DUPIC Fuel Compatibility Assessment - International Nuclear ...

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KAERI/RR-1999/99

A Study on Direct Use of Spent PWR Fuel inCANDU Reactors

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S U M M A R Y

I. Project Title

DUPIC Fuel Compatibility Assessment

II. Objective and Importance of Project

A. Objective

The purpose of this study is to assess the compatibility of DUPIC (Direct Use of Spent PWR

Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology

being developed to utilize the spent PWR fuel in CANDU reactors. The objective of this project

has been set as follows;

- Final Objectives

• Assessment of DUPIC fuel compatibility with CANDU reactors

• Economic analysis of DUPIC fuel cycle

• Technical feasibility analysis for practical use of DUPIC fuel

- Objectives of Phase I (1997.7.21-2000.3.31)

• Feasibility analysis on applicability of the current core design method

• Feasibility analysis on operation of the DUPIC fuel core

• Compatibility analysis on individual reactor system

• Sensitivity analysis on the fuel composition

• Economic analysis on DUPIC fuel cycle

- Objectives of Phase II (2000.4.1-2002.3.31)

• Reactor safety analysis and licensing possibility

• Development of DUPIC fuel core design method

• Technology development for providing optimum DUPIC fuel material

• Technical feasibility analysis for practical use of DUPIC fuel

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KAERI/RR-1999/99

B. Importance of Research and Development

As the nuclear power generation continues, the accumulation of spent fuel and its disposal become

an urgent problem and, therefore, each country is developing its own technology for the back-end

fuel cycle. The DUPIC technology has been developed to convert the spent PWR fuel into the

CANDU fuel, which can resolve the accumulation of spent PWR fuel and reduce the spent

CANDU fuel. In order to prove the feasibility of the DUPIC technology, the compatibility analysis

of DUPIC fuel with current CANDU reactors and the economic analysis have been performed

by Korea Atomic Energy Research Institute (KAERI) during Phase I period (1997.7.21 -

2000.3.31).

In order to prove the compatibility of the DUPIC fuel, the reference DUPIC fuel composition

was determined, the performance of the reactor analysis code was assessed, the compatibility

of individual reactor system was analyzed, and the uncertainty due to the fuel composition was

estimated. These studies have demonstrated that the current DUPIC fuel model is compatible

with the current CANDU 6 reactor. This result will be used as a basis for Phase II study, which

focuses on the feasibility of practical use of the DUPIC fuel and the reactor safety analysis.

Once the safety analysis is completed, the DUPIC fuel compatibility analysis will be accomplished.

Based on this, the possibility of licensing and key issues for the practical use of DUPIC fuel

can be discussed.

- Technical Aspect of Research and Development

The reactor physics analysis on the DUPIC fuel core utilizes the existing analysis method but

also requires new methods that can quantitatively estimate the sensitivity of the core performance

parameters to the fuel composition and power distribution. Therefore, it is possible to improve

the current CANDU core analysis technology and acquire a leading technology in CANDU fuel

development by establishing the DUPIC fuel core analysis method.

- Economic/Industrial Aspect of Research and Development

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KAERI/RR-1999/99

Until now, the CANDU core design and analysis are partly dependent on the foreign technology.

During the Phase I period, it was possible to develope and localize a part of such technology.

As far as the compatibility analysis is concerned, it is more important to have the technology

that can be used to determine the compatibility of the DUPIC fuel than the compatibility of

the DUPIC fuel itself. It is expected that the localization of technology will contribute to the

import substitute in the future.

- Social/Cultural Aspect of Research and Development

There are diverse technical questions on the feasibility of loading DUPIC fuel in a CANDU

reactor, because the CANDU reactor was originally designed for natural uranium fuel. In order

to resolve those technical uncertainties and to provide a rationale to develope safeguardable

DUPIC technology, the compatibility analysis of the DUPIC fuel should be performed.

HI. Scope and Contents of Project

A) Feasibility analysis on applicability of the current core design method

• Design parameter

- Nuclear fuel and reactor core design data

- Physics and thermal-hydraulic design requirements

• Physics, thermal-hydraulic and safety analysis

- Analysis model and input parameters for computer codes

• Computer codes for operation

- Benchmark calculation using reactor physics measurement results

B) Feasibility analysis on operation of the DUPIC fuel core

• Neutronic characteristics analysis - Fuel bundle and channel power distribution

• Thermal-hydraulic characteristics analysis - Thermal-hydraulic parameters

• Operation characteristics analysis - 600-FPD refueling simulation

• Safety of key performance parameters - Assessment of ten reactivity coefficients

C) Compatibility analysis on individual reactor system

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• Reactor control system - Reactivity and power controllability

• Reactor shutdown system - Shutdown capability

• Fuel transportation system - Criticality and radiation level

• Fuel storage system - Criticality and cooling capacity

• Fuel handling system - Radiation level

• Fuel loading system - Loading path and fueling machine capacity

• Reactor structural material - Radiation effect on welding and joint area

D) Sensitivity analysis on the fuel composition

• DUPIC fuel composition analysis - Analysis on 3600 spent PWR fuel assemblies

• Sensitivity study on DUPIC fuel composition

- Uncertainty estimation of core performance parameters

E) Economic analysis on DUPIC fuel cycle

• Fuel fabrication cost - Preliminary conceptual design and cost estimation

• Fuel cycle unit cost - Fuel handling and disposal cost estimation

• Fuel cycle cost - Calculation of DUPIC and direct disposal fuel cycle cost

IV. Results and Proposal for Applications

A. Results of Research and Development

1) Feasibility analysis on applicability of the current core design method

The major differences between DUPIC and standard CANDU fuel are the fuel bundle

configuration and fuel material composition. The DUPIC fuel bundle adopts 43- element model

which has been developed for natural and slightly enriched uranium (SEU), while the standard

fuel has 37 fuel elements. For the fuel composition, the center rod of DUPIC fuel bundle contains

burnable poison material, while the standard fuel has no poison material. The DUPIC core model

is the same as the standard CANDU 6 reactor model. However, the fuel management method

of the DUPIC core is a 2-bundle shift refueling scheme, while the standard core uses an 8-bundle

shift scheme. For the analysis of the DUPIC fuel system, a transport code WIMS-AECL is

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used for lattice calculations and SHETAN code is used for the reactivity device analysis. Other

computer codes for the core and safety analysis are the same as those used for the standard

core analysis.

The benchmark calculation of the lattice code WIMS-AECL has been performed for the DUPIC

fuel using MCNP code. For the criticality calculation, the eigenvalue (koo) error was within

0.73% 5" k. In general, the error increases as the fuel burnup increases. The void reactivity

estimated by WIMS-AECL matches that of MCNP within 5%. However, the fuel temperature

(Doppler) coefficient has a relatively large error of 80% at the discharge state. The results of

benchmark calculations have shown that the WIMS-AECL is in general acceptable for DUPIC

physics design and analysis. However, for the slow transient that includes the fuel temperature

chang, there is a slight loss of accuracy due to the imperfectness of the temperature data.

The validation calculation of WIMS/SHETAN/RFSP code system, which is used for the DUPIC

core analysis, was performed using Wolsong 2 physics measurement data. The validation

calculation includes the calibration of boron reactivity, adjuster rod worth, shutoff rod worth,

mechanical control absorber worth, temperature coefficients of the coolant and moderator, and

power distribution. The results have shown that the average error of zone controller unit (ZCU)

worth is less 3%, which enhances an error of 0.2% 5" k for criticality. In general, the error of

reactivity device worth is less than the permissible error of 15%. The flux scan error is about

10% which is also less than permissible root-mean-square (RMS) error of 15%. The coolant

temperature coefficient was consistent with the measurement result. The results of validation

calculations have confirmed that the current core analysis system is acceptable for the feasibility

study of the DUPIC fuel compatibility analysis.

2) Feasibility analysis on operation of the DUPIC fuel core

At the initial burnup state, the reactivity coefficients of DUPIC fuel are different from those

of natural uranium fuel For natural uranium, as the fuel burnup increases, the reactivity

coefficients such as the fuel and moderator temperature coefficients increase rapidly because

Pu content increases significantly. On the other hand, the reactivity coefficients of the DUPIC

fuel are not favorable as compared with natural uranium fuel at the initial burnup stage. However,

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the change of the reactivity coefficient is small because Pu content changes little as the fuel

burnup increases. Therefore, the overall behavior of the DUPIC fuel is better than that of natural

uranium fuel at the equilibrium burnup state.

To assess the performance of the operating reactor, the probability to exceed the administrative

limits was estimated based on the result of 600-FPD refueling simulation. For convenience, the

administrative limit was set as 95% of the operation limit. The administrative limits for the

channel and bundle powers are 6935 kW and 888 kW, respectively. The channel power peaking

factor of 1.10 and ZCU level of 0.2-0.8 were chosen as the administrative limits. For the DUPIC

fuel core, the administrative limits of the channel and bundle powers were not exceeded, whereas

the natural uranium core has the probability of 0.3%, which shows there is no significant difference

in the reactor power characteristics between two cores. The probability to exceed the limit of

channel power peaking factor was 0.17% for DUPIC core whereas the natural uranium core

was found to never exceed this limit. For ZCU level, the probability was 0.12% and 0.15%

for the DUPIC and natural uranium cores, respectively. This revealed that both cores are almost

the same in the viewpoint of the operational performance.

The thermal-hydraulic properties were assessed for the time-average cores. The total coolant

flow rate for DUPIC core is 8124 kg/s, which is 2% less than that of natural uranium core.

The maximum channel flow is 26.7 kg/s for the DUPIC core, which is also 2% less than that

of natural uranium core. Therefore the coolant flow rate of DUPIC core meets the design

requirement. The maximum channel pressure drop is 718 kPa at the flow of 23.9 kg/s, which

also meets the design requirement. The minimum critical channel power for DUPIC core is

5148 kW, which is a 3.5% improvement over that of natural uranium core. Therefore it can

be concluded that the thermal-hydraulic performance of DUPIC core is no worse than that of

natural uranium core. The critical channel power ratio is 1.502 and 1.440 for the DUPIC and

natural uranium core, respectively, which satisfy the design requirement of 1.12. The exit quality

of the hottest channel is 3.9% and 3.7% for the DUPIC and natural uranium core, respectively,

which also meet the design requirement of 4.0%. Based on these results, the thermal-hydraulic

characteristics of the DUPIC core are not significantly different from those of natural uranium

core.

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3) Compatibility analysis on individual reactor system

For ZCU system, the spatial control capability after refixeling was assessed. The Xe transient

analysis has shown that the power oscillation is controlled for both the DUPIC and natural uranium

cores. For the adjuster system, the functional requirements were confirmed for the power flattening

capability, Xe reactivity compensation during the restart after a shutdown, shim operation during

the malfunctioning of the refueling machine, and reactor power level change. A distinctive feature

of the DUPIC core from the standard core is that the time necessary to reach the full power

is delayed, which is partially due to Xe feedback effect of the DUPIC core. However, the time

delay does not cause any adverse effect on the operation of DUPIC fuel core. For regional

overpower protection (ROP) trip set points, a comparative study has been performed for 232

design base cases. As a result, the ROP trip set point of DUPIC core was found to be 123%,

which is acceptable as compared with 122% for natural uranium core.

For shutdown system 1, the dynamic reactivity is 72.5 mk for DUPIC core, which is large

enough keep the reactor subcriticality. The insertion characteristics of the shutdown rods are

similar to those of natural uranium core. For the assessment of shutdown rod performance, the

reactor inlet header (RIH) 20% loss of coolant accident (LOCA) was analyzed. Though the power

pulse for DUPIC core rises a little earlier as compared with natural uranium core, the overall

integrated thermal energy deposition in the fuel during first 3 sec is almost the same for both

cores. The possibility of fuel disrupture was assessed based on the accumulated thermal energy

during the power pulse. The margin to the fuel rupture criterion, 840 J/g, for DUPIC core decreases

by 3.9% compared with natural uranium core.

For the fuel transportation technology, a transportation cask was conceptually studied, which

contains two spent fuel baskets being used for the dry storage facility at Wolsong nuclear plant.

The transportation capacity is 120 fuel bundles. The overall dimension of the cask is 53.3 cm

in radius, 15 cm in lead shield for photon shielding, and 10 cm in polyethylene for neutron

shielding. The radiation level on the surface and 2 m away from the surface of the cask satisfies

the design requirements of 200 mrem/hr and 10 mrem/hr, when the spent DUPIC fuel cooled

for 10 years is used as the source term. The weight of the cask is —17 ton including the spent

fuel, which can be handled by the head crain of the plant (the maximum capacity of 30 ton).

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The criticality of the transportation cask has no problem.

In the storage pool, the criticality of DUPIC fuel is 0.923 when piled up in the conventional

way. Therefore the subcriticality requirement (0.95) is satisfied. The storage capacity corresponds

to the amount of spent fuel from 12 years operation. In this case the cooling capacity of the

storage pool needs to be increased by 1 MW, considering the decay heat from the spent DUPIC

fuel. "When the cooling system is down, the time for the storage pool to reach 49°C is 10 hours

and, therefore, the cooling system should be restored in that time.

The annual dose of the fresh DUPIC fuel is 6.47 Sv/y for the whole body and, therefore, the

DUPIC fuel should be handled and transported by remote operation. When the DUPIC fuel

is transferred to the plant, they should be located at the loading system by either manual or

remote handling. First of all, the fresh fuel needs to be moved from the storage area to the

loading station (designed for dry storage facility). Once the transportation cask is opened in

the loading station, the basket containing the fuel is taken out and moved to the tilt station.

The fuel basket is moved horizontally to the fuel tray using the tilt station. Therefore, if the

DUPIC fuel transportation cask is compatible with the loading station and the fuel basket is

compatible with the tilt station, there would be no technical problem in handling the fresh DUPIC

fuel.

The compatibility of the fueling machine needs to be considered in two aspects: the endurance

time and thermal capacity. In case of DUPIC fuel, when four channels are refueled everyday

with two fuel bundles loaded in a channel, the endurance time could be reduced by a half.

The heavy water inside the magazine of the refueling machine should meet the normal operating

condition (maximum magazine temperature < 57°C) and design requirement (maximum magazine

temperature < 149°C). This requirement is satisfied if the fueling machine takes fuel bundles

necessary to load two channels at a time. Therefore there is no need to modify the fueling

machine cooling system.

The radiation shielding analysis has shown that the radiation dose outside the primary shield

of the DUPIC fuel core is less than those of natural uranium core except the end shield. Even

for the end shield, the design requirement is satisfied for th DUPIC fuel core. The heat flux

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and temperature requirements for the thermal shield are also met for the DUPIC fuel core.

Radiation damage rate of the critical reactor component was assessed by calculating displacement

per atom (DPA). The analyses have been performed for roll joint, calandria welding area, and

edges/corners of calandria tank assuming 30 years operation, and the results have shown that

the lifetime of the critical component of the DUPIC core could be reduced by 30% compared

with that of natural uranium core.

4) Sensitivity analysis on the fuel composition

Once three inter-assembly mixing operations are taken for 3600 spent PWR fuel assemblies,

there will be 450 distinctive fuel compositions. The distribution of these fuel compositions was

computed with 95% confidence level. In principle, there is no variation in U235 and Pu239

content if the fissile isotopic content is adjusted to the target value. On the other hand, the

standard deviation of fissile content is 2-3% when the reactivity is adjusted to the target value.

The burnup-dependency of nuclear characteristics was assessed through the depletion calculation.

For the fissile content adjustment method, the dispersion of the initial reactivity is relatively

large but it gradually decreases as the fuel burns. However, the trend is reversed for the reactivity

control method. Therefore it can be said that the fissile content adjustment method is more

effective in reducing the DUPIC fuel composition heterogeneity effect than the reactivity

adjustment method.

The uncertainty of the core performance parameter due to the variation of DUPIC fuel composition

has been estimated using a computer code GENOVA which was developed in the course of

this project. The uncertainty of the maximum channel power was obtained based on 600-FPD

refueling simulation. The uncertainty is 1.3% and 7.0% for the fissile content adjustment and

reactivity control method, respectively. To confirm the uncertainty level of the core performance

parameter obtained by the deterministic method and to realistically simulate the composition

heterogeneity on the core performance parameter, a refueling simulation was performed using

30 different DUPIC fuel types. As a result, the uncertainty of the maximum channel power

was confirmed to be less than 1%. Therefore, it can be concluded that the DUPIC fuel composition

heterogeneity does not to impose any serious effect on the reactor operation if the fuel composition

is adjusted.

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KAERI/RR-1999/99

5) Economic analysis on DUPIC fuel cycle

In order to estimate DUPIC fuel fabrication cost, a conceptual design of the DUPIC fuel

fabrication facility was performed for the fabrication capacity of 400 MTHE/yr. The levelized

fuel fabrication cost was estimated to be 558 $/kgU. The DUPIC fuel fabrication cost is much

higher than the current light water reactor (LWR) fuel fabrication cost of 275 $/kgU and heavy

water reactor (HWR) fuel fabrication cost of 65 $/kgU, which is due to the fuel fabrication

facility construction cost that requires remote hot cell processes. The effect of using fresh uranium

to control the fuel composition was assessed too. For example, when the slightly enriched and

depleted uranium is added by 6.5% and 10.8%, respectively, the DUPIC fuel fabrication cost

619 $/kgU.

The cost of DUPIC fuel transportation and storage was indirectly estimated based on that of

LWR spent fuel (230 $/kgHM) quoted from OECD/NEA(1993) publication, uisng the decay

heat ratio. As a result, the transportation and storage cost of the DUPIC fuel was found to

be 170 $/kgHM. The DUPIC fuel handling cost was estimated too, which includes the design

modifications to load the DUPIC fuel following the reverse path in the plant. Considering both

the hardware and software modifications, the DUPIC fuel handing cost was estimated to 4500

million Won (reference year 2000), which is then converted to the levelized cost of 5.13 $/kgHM.

The DUPIC fuel disposal facility model utilizes Canadian room-and-pillar design concept. In

order to estimate the appropriate disposal cost, the domestic reactor type strategy was established

at first. Then, the disposal cost was estimated by interpolating the cost data based on the domestic

electricity generation capacity. The levelized cost was calculated by discounting the cost for

the operation period. The disposal costs for the spent LWR1, HWR and DUPIC fuel are 403,

118, and 220 $/kgHM. Considering that the current spent LWR and HWR fuel disposal costs

are 610 and 73 $/kgHM in OECD/NEA publication, the disposal costs estimated in this study

are significantly conservative to be used for the DUPIC fuel cycle cost analysis.

The DUPIC fuel cycle cost was estimated using one-batch model and compared with the direct

disposal fuel cycle. For the estimation of fuel cycle cost, both the deterministic and Monte Carlo

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methods were used. The deterministic analysis has shown that the DUPIC fuel cycle cost is

5.25-5.43 mills/kWh depending on the composition adjustment method while the direct disposal

fuel cycle cost is 5.19-5.24 mills/kWh. However the Monte Carlo simulation has shown that

the standard deviation of the fuel cycle cost is 0.34-0.38 mills/kWh. Therefore, it is believed

that the DUPIC fuel cycle is comparable to the once-through fuel cycle.

B. Proposals for Application

The DUPIC core analysis technology can be used for assessing the compatibility of the advanced

HWR fuel to be developed in the future with the standard CANDU core and for designing

a new HWR core. By developing DUPIC core and compatibility analysis technology, the design

and analysis technology of the standard CANDU core can also be improved. If the accuracy

of the CANDU core design and analysis method is confirmed through the experimental

verification, it is possible to achieve the technology independence for the CANDU core design

and analysis.

The ROP analysis and computer codes developed (fuel management program and perturbation

code) during Phase I of DUPIC project are expected to replace the technology imported for

the HWR core design and analysis. Also the advanced design and analysis method developed

during the course of DUPIC project can be used continuously to the development of advanced

CANDU fuels and could be exported to the developing countries in the future.

Because DUPIC core design and analysis technology includes new analysis methods as well

as the existing methodology, it can be used for the analyses of fuel loading behavior and other

related characteristics, which is necessary to demonstrate the performance of the advanced fuel

being developed. It can also be used for the following research activities;

• Cobalt production in a HWR

• Higher actinide transmutation under high thermal flux

• Thorium fuel cycle.

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C O N T E N T

PREFACE 1

SUMMARY (KOREAN) 3

SUMMARY (ENGLISH) , 13

CONTENT 25

TABLE CONTENT 51

FIGURE CONTENT 59

CHAPTER 1. INTRODUCTION - 69

1.1 PURPOSE OF PROJECT 72

1.2 OBJECTIVES AND SCOPE 73

1.3 CURRENT STATUS OF TECHNOLOGY DEVELOPMENT 16

1.3.1 Foreign Technology 76

1.3.2 Domestic Technology • 77

1.3.3 Review of Technology Development Cases 78

1.4 DETAILED TECHNOLOGY ITEMS 80

1.4.1 Foreign Technology 80

1.4.1.1 Design/analysis code validation 80

1.4.1.2 Reactor design/analysis method 80

1.4.1.3 Reactor system compatibility assessment 80

1.4.1.4 Reactor safety analysis 81

1.4.1.5 Fuel handling technology 81

1.4.1.6 Fuel cycle economics analysis 81

1.4.2 Domestic Technology 81

1.4.2.1 Design/analysis code validation 81

1.4.2.2 Reactor design/analysis method 82

1.4.2.3 Reactor system compatibility assessment 82

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1.4.2.4 Reactor safety analysis 82

1.4.2.5 Fuel handling technology 83


1.5 REFERENCES 84

CHAPTER 2. VALIDATION OF NUCLEAR DESIGN METHOD 87

2.1 CANDU CORE ANALYSIS METHODOLOGY 90

2.1.1 Core Analysis Procedure 90

2.1.1.1 Lattice calculation 90

2.1.1.2 Representation of reactivity devices 90

2.1.1.3 Core calculation 91

2.1.1.4 Kinetics calculation 91

2.1.1.5 Reactor stability and control 91

2.1.1.6 Fuel management strategy • 92

2.1.2 Nuclear Design Data 92

2.1.3 Computer Codes • 92

2.1.3.1 Lattice code 92

2.1.3.2 Core analysis code 93

2.2 LATTICE CODE VALIDATION 107

2.2.1 Calculation Model 107

2.2.2 Natural Uranium Fuel 108

2.2.2.1 Burnup reactivity 108

2.2.2.2 Coolant void reactivity • 108

2.2.2.3 Fuel temperature coefficient 109

2.2.2.4 Primary heat transport system reactivity I l l

2.2.2.5 Relative pin power ratio 112

2.2.2.6 Reaction rates 112

2.2.3 DUPIC Fuel 113

2.2.3.1 Burnup reactivity 113

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KAERI/RR-1999/99

2.2.3.2 Coolant void reactivity 113

2.2.3.3 Fuel temperature coefficient 114

2.2.3.4 Primary heat transport system reactivity 115

2.2.3.5 Relative pin power ratio 115

2.2.3.6 Reaction Rates 115

2.2.4 Summary 115

2.3 VALIDATION OF CORE ANALYSIS CODE 146

2.3.1 Calculation Procedure 146

2.3.1.1 Lattice model 146

2.3.1.2 Reactivity device model , 147

2.3.1.3 Core analysis model • 151

2.3.2 Natural Uranium Fresh Core 152

2.3.2.1 Criticality measurement 152

2.3.2.2 Reactivity device worth 152

2.3.2.3 Reactivity coefficient measurement 154

2.3.2.4 Flux distribution 154

2.3.3 Natural Uranium Equilibrium Core 155

2.3.3.1 Time-average core • 155

2.3.3.2 Refueling simulation 157

2.3.3.3 Summary 158

2.3.4 DUPIC Fuel Equilibrium Core 158

2.3.4.1 DUPIC core model 158

2.3.4.2 Comparison of results 159

2.4 SUMMARY AND CONCLUSION 196

2.5 REFERENCE 197

CHAPTER 3. REACTOR PHYSICS DESIGN AND ANALYSIS 201

3.1. PHYSICS DESIGN REQUIREMENTS OF A CANDU REACTOR 204

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KAERI/RR-1999/99

3.1.1 Power Controllability 204

3.1.2 Compliance to Design Limits during Normal and Transient Conditions 205

3.1.3 Reactivity Controllability 205

3.1.4 Shutdown System 206

3.1.5 On-line Flux Mapping 206

3.1.6 Regional Overpower Protection 206

3.2. REACTOR PHYSICS ANALYSIS METHOD 207

3.2.1 DUPIC Fuel Cross-Section Generation 207

3.2.1.1 Base cross-section 207

3.2.1.2 Incremental cross-section 208

3.2.1.3 Xenon cross-section 208

3.2.2 DUPIC Fuel Core Calculation 209

3.2.2.1 Time-average core model 209

3.2.2.2 Instantaneous core model 210

3.2.2.3 Refueling simulation model 211

3.3. REFERENCE DUPIC FUEL COMPOSITION 217

3.3.1 Fissile Content Adjustment Option 217

3.3.1.1 Reference DUPIC fuel composition 218

3.3.1.2 Multiple spent PWR fuel mixing 218

3.3.1.3 Fuel cycle cost 219

3.3.2 Reactivity Control Option by SEU/DU 219

3.3.2.1 Spent PWR fuel mixing 219

3.3.2.2 Reactivity control by SEU/DU 220

3.3.2.3 Optimum target reactivity 221

3.3.3 Reactivity Control Option by Natural Uranium 221

3.3.3.1 Utilization of high reactivity fuel 221

3.3.3.2 Utilization of linear reactivity fuel 222

3.3.3.3 Optimum target reactivity 222

3.3.4 Characteristics of DUPIC Fuel 223

3.3.4.1 Fissile content 224

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KAERI/RR-1999/99

3.3.4.2 Plutonium content 224

3.3.4.3 Fission products 224

3.3.5 Summary 225

3.4. DUPIC FUEL LATTICE PROPERTY 238

3.4.1 Comparison of Lattice Property 238

3.4.1.1 Lattice parameters 238

3.4.1.2 Relative pin power distribution 238

3.4.1.3 Delayed neutrons 238

3.4.2 Temperature Reactivity Effect 239

3.4.2.1 Moderator temperature reactivity effect 240

3.4.2.2 Coolant temperature reactivity effect 240

3.4.2.3 Fuel temperature reactivity effect 241

3.4.2.4 Reactivity change from full power to zero power 241

3.4.3 Void Reactivity Effect 242

3.4.3.1 Void reactivity 242

3.4.3.2 Void reactivity versus degree of voiding 242

3.4.3.3 Void reactivity versus fuel irradiation 242

3.4.3.4 Effects of absorbers on void reactivity 243

3.4.3.5 Detailed core simulation for coolant voiding 243

3.4.4 Power Coefficient 243

3.4.4.1 DUPIC fuel core • 244

3.4.4.2 Natural uranium core 244

3.4.5 Miscellaneous Reactivity Perturbations 245

3.4.5.1 Moderator purity reactivity effect 245

3.4.5.2 Coolant purity reactivity effect 245

3.5. COMPATIBILITY OF REACTIVITY DEVICES 266

3.5.1 Zone Controller Unit 266

3.5.1.1 Static reactivity worth 266

3.5.1.2 Adequacy of zone control system in suppressing spatial oscillation 267

3.5.1.3 Effect of draining zone controllers 268

3.5.2 Adjuster Rod System 269

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KAERI/RR-1999/99

3.5.2.1 Static reactivity worths 269

3.5.2.2 Adjuster bank reactivity insertion characteristics 269

3.5.2.3 Startup after a short shutdown 270

3.5.2.4 Startup after a poison-out 271

3.5.2.5 Reactivity shim operation 271

3.5.2.6 Power reduction (stepback) 272

3.5.3 Mechanical Control Absorber 273

3.5.4 Shut-Down System 273

3.5.4.1 Static reactivity of shut-off rods 274

3.5.4.2 Performance of shut-off rod system 275

3.5.4.3 Performance of liquid poison injection system 276

3.5.5 Xenon Transient 277

3.5.5.1 Shutdown from various power levels 278

3.5.5.2 Transients after startup 278

3.5.5.3 Power stepbacks from full power 278

3.5.5.4 30-minute xenon load 279

3.5.6 Xenon Spatial Oscillation 279

3.5.6.1 Xenon oscillation 279

3.5.6.2 Instability analysis 281

3.5.7 Summary 283

3.6. REGIONAL OVERPOWER PROTECTION SYSTEM 328

3.6.1 ROP Analysis Model • 329

3.6.1.1 Trip coverage equation 329

3.6.1.2 ROP detector 329

3.6.2 ROP Calculation Procedure 330

3.6.2.1 Flux shape and channel powers 330

3.6.2.2 Thermal-hydraulic analysis 330

3.6.2.3 Detector response 331

3.6.2.4 Ripple calculation 331

3.6.2.5 Trip setpoint calculation 331

3.6.3 Result of ROP Trip Setpoint Calculation 332

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KAERI/RR-1999/99

3.6.3.1 ROP trip setpoint 332

3.6.3.2 Single-detector failure 333

3.6.3.3 REFORM calculation 333

3.7. DUPIC FUEL CORE CHARACTERISTICS 337

3.7.1 Reference Core Simulation 337

3.7.1.1 Time-Average core 337

3.7.1.2 Instantaneous core • 338

3.7.1.3 Refueling simulation 339

3.7.2 Deterministic Analysis of Fuel Composition Heterogeneity 340

3.7.2.1 Sensitivity method for CANDU core analysis 340

3.7.2.2 Perturbation method 344

3.7.2.3 Uncertainty estimation 347

3.7.3 Statistical Analysis of Fuel Composition Heterogeneity 349

3.7.3.1 Heterogeneous core 349

3.7.3.2 Equilibrium core 352

3.7.4 Summary 353

3.8. SUMMARY AND CONCLUSION 394

3.9 REFERENCE 396

CHAPTER 4. RADIATION PHYSICS ANALYSIS 399

4.1. COMPUTER CODES AND LIBRARIES 403

4.1.1 Shielding Analysis Codes 403

4.1.2 Assessment of Shielding Analysis Code 405

4.1.2.1 Natural uranium core calculation 405

4.1.2.2 DUPIC core calculation 409

4.2. PRIMARY SHIELD ANALYSIS OF CANDU REACTOR 420

4.2.1 CANDU Primary Shield System 420

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KAEM/RR-1999/99

4.2.1.1 End shield 420

4.2.1.2 Side shield 421

4.2.1.3 Top shield 421

4.2.1.4 Bottom shield 421

4.2.2 Primary Shield Design Criteria 421

4.2.3 Primary Shield Analysis for DUPIC Fuel 422

4.2.3.1 Source term calculation 422

4.2.3.2 End shield calculation 423

4.2.3.3 Side shield calculation 424

4.2.3.4 Top shield calculation 425

4.2.3.5 Bottom shield calculation 426

4.2.4 Radiation Heat Generation 426

4.2.4.1 Axial shield heat deposition 426

4.2.4.2 Radial shield heat deposition 427

4.2.5 Summary 428

4.3. RADIATION EFFECT ON REACTOR HARDWARE 451

4.3.1 Radiation Damage Analysis 451

4.3.1.1 Fuel channel system 452

4.3.1.2 Calandria shell system 454

4.3.2 Thermal Shield Analysis 455

4.3.2.1 Side thermal shield 455

4.3.2.2 End thermal shield 457

4.4 DUPIC FUEL HANDLING, TRANSPORTATION AND STORAGE 467

4.4.1 Dose Rate of DUPIC Fuel Bundle 468

4.4.1.1 Calculation model 468

4.4.1.2 Fresh DUPIC fuel 469

4.4.1.3 Spent DUPIC fuel 471

4.4.2 Transportation Cask for DUPIC Fuel 471

4.4.2.1 Transportation cask model for DUPIC fuel 471

4.4.2.2 Radiation source 473

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KAERI/RR-1999/99

4.4.2.3 Shielding analysis of transportation cask 475

4.4.3 Criticality Calculation for Storage Bay 478

4.4.3.1 Stack model 478

4.4.3.2 Calculation method 480

4.4.3.3 Results and discussion 481

4.5. SUMMARY AND CONCLUSION 508

4.6 REFERENCE 510

CHAPTER 5. THERMAL-HYDRAULICS ANALYSIS 515

5.1 THERMAL-HYDRAULIC ANALYSIS MODEL 518

5.1.1 Thermal-hydraulic Design Requirements for CANDU 6 518

5.1.2 Thermal-hydraulic Model 518

5.1.2.1 Pressure drop model 519

5.1.2.2 Heat transfer model 520

5.1.2.3 Critical channel power calculation 521

5.1.3 Input Parameters and Operating Conditions 522

5.1.3.1 Feeders 522

5.1.3.2 Fuel channel and fuel bundle 523

5.1.3.3 Header-to-header boundary condition 523

5.2 RESULTS OF THERMAL-HYDRAULIC ANALYSIS 529

5.2.1 Results and Discussion 529

5.2.1.1 Channel power distribution 529

5.2.1.2 Channel flow rates 529

5.2.1.3 Critical channel power 530

5.2.1.4 Critical power ratio 530

5.2.1.5 Channel Exit Quality 531

5.2.2 Uncertainty of Critical Channel Power Prediction 531

5.2.2.1 Source of uncertainty in critical channel power calculation 531

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KAERMRR-1999/99

5.2.2.2 Radial correction factor and associated uncertainty 532

5.3 VALIDITY OF MODELS AND CODES 555

5.3.1 NUCIRC Critical Heat Flux Model 555

5.3.2 ASSERT Code Validation 556

5.4 CONCLUSION 565

5.5 REFERENCE 566

CHAPTER 6. FUEL CYCLE ECONOMICS ANALYSIS 569

6.1 DUPIC FUEL FABRICATION COST 572

6.1.1 DUPIC Facility Design Requirements 573

6.1.1.1 Facility performance requirements 573

6.1.1.2 DUPIC fuel processing requirements 576

6.1.2 Conceptual Design of DUPIC Process System 577

6.1.2.1 Process mass balance 577

6.1.2.2 Conceptual design of fuel process system 579

6.1.2.3 Facility description 581

6.1.3 Fabrication Cost Estimation 582

6.1.3.1 Cost evaluation data 582

6.1.3.2 Fuel fabrication cost 584

6.1.4 Conclusion and recommendations 587

6.2. DUPIC FUEL HANDLING COST 608

6.2.1 Current Fuel Loading and Unloading Path 609

6.2.2 DUPIC Fuel Loading and Unloading Method 609

6.2.2.1 Utilization of current loading path 610

6.2.2.2 Utilization of current unloading path (reverse path) 612

6.2.3 Fueling Machine and Spent Fuel Storage Bay 613

6.2.3.1 Fueling machine cooling capacity 615

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KAERI/RR-1999/99

6.2.3.2 Storage bay requirements and capacity 618

6.2.3.3 Storage bay cooling capacity 620

6.2.4 DUPIC Fuel Handling Cost in Wolsong Nuclear Power Plant 625

6.2.4.1 Compatibility of dry storage facility and DUPIC fuel handling equipment ••• 625

6.2.4.2 Pushing ram and dryer 626

6.2.4.3 Spent fuel storage and reception bay cooling system 626

6.2.4.4 Gamma detector for refueling operation 626

6.2.4.5 Fuel handling program change 627

6.2.4.6 Design documentation change 627

6.2.5 Design Modification Items and Cost 628

6.2.5.1 Fuel handling cost 628

6.2.5.2 Fuel handling unit cost 628

6.2.6 Summary and Conclusion 629

6.3. SPENT FUEL DISPOSAL COST 668

6.3.1 Basic Design Concept of Disposal Facility 669

6.3.1.1 Current status of HLW disposal technology 669

6.3.1.2 Disposal facility model • 670

6.3.2. Estimation of Scaled Disposal Cost of Spent Fuel 672

6.3.2.1 Reference disposal cost model 672

6.3.2.2 Disposal container model 674

6.3.2.3 Disposal vault layout 675

6.3.2.4 Cost of spent fuel disposal 677

6.3.3 Levelized Unit Cost of Spent Fuel Disposal 678

6.3.3.1 Analysis of electricity generation size in Korea 678

6.3.3.2 Levelized unit disposal cost 679

6.3.4 Conclusion 681

6.4 DUPIC FUEL CYCLE COST 699

6.4.1 Reference DUPIC Fuel Model 700

6.4.1.1 Fissile content adjustment (Option 1) 701

6.4.1.2 Reactivity control by SEU/DU (Option 2) 701

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KAERI/RR-1999/99

6.4.1.3 Isotopic composition control by partial mixing (Option 3) 701

6.4.2 Unit Cost of DUPIC Fuel Cycle Composition 702

6.4.2.1 DUPIC fuel fabrication cost 702

6.4.2.2 DUPIC fuel handling cost 702

6.4.2.3 Interim storage cost •• 703

6.4.2.4 Spent fuel disposal cost 703

6.4.3 Fuel Cycle Cost Analysis 704

6.4.3.1 Fuel cycle cost calculation method 704

6.4.3.2 Reference plant and fuel cycle model 706

6.4.3.3 Fuel cycle cost calculation 707

6.4.4 Summary and Recommendations 710

6.5 REFERENCE 739

CHAPTER 7. ACHIEVEMENT OF OBJECTIVE AND FUTURE WORKS 743

7.1. ACHIEVEMENT OF OBJECTIVE 746

7.1.1 Major Achievement of Detailed Research Objective 746

7.1.2 Summary and Discussion of Detailed Research Objective 747

7.1.2.1 Feasibility analysis on applicability of current core design method 747

7.1.2.2 Feasibility analysis on operation of DUPIC fuel core 747

7.1.2.3 Compatibility analysis on individual reactor system 747

7.1.2.4 Sensitivity analysis on fuel composition 748

7.1.2.5 Economic analysis on DUPIC fuel cycle 748

7.1.3 Research Products 749

7.1.3.1 Summary table • 749

7.1.3.2 Foreign SCI journal publication 749

7.1.3.3 Foreign conference publication 750

7.1.3.4 Domestic conference publication 752

7.1.3.5 Technical report 753

7.1.3.6 Computer program 755

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KAERI/RR-1999/99

7.2 FUTURE WORKS • 756

7.2.1 Application of Research Product 756

7.2.1.1 Economic/Industrial aspect of research and development 756

7.2.1.2 Social/Cultural aspect of research and development 756

7.2.1.3 Technical aspect of research and development 756

7.2.2 Future Works • 757

7.2.2.1 Validation of nuclear design method 757

7.2.2.2 Reactor physics design and analysis 758

7.2.2.3 Radiation physics analysis 760

7.2.2.4 Thermal-hydraulic analysis 761


7.2.3 Research and Development Strategy for Phase-II 763

7.2.3.1 Final objective 763

7.2.3.2 Annual research objective and content 764

7.2.3.3 Research and development strategy 765

7.2.3.4 Research and development organization 766

- 37 -

KAERI/RR-1999/99

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3.5.5.4 30£ 4 ^ -f *> 279

3.5.6 4 f e g-?> ^l^i- , • 279

3.5.6.1 *fiir *!-§• 279

3.5.6.2 # * > ^ ^ *H^ 281

3.5.7 A ^ 283

3.6 ^H f 2 f # ^ JLSL 7)1^ *H^ 328

3.6.1 ROP 711A> ^ ^ - g - • 329

3.6.1.1 ^ °}^+\ 329

3.6.1.2 ROP ?j)^-7l 329

3.6.2 ROP 7j] > ^x\ 330

3.6.2.1 ^ * H T "S ^ *1I^ %^ 330

3.6.2.2 ^ ! ^ ^ &q 330

3.6.2.3 4^-7] -g-^- 331

3.6.2.4 Ripple Tjl l •' 331

3.6.2.5 HtJ ^ ^ ^ 1 7j]^ 331

- 43 -

KAERI/RR-1999/99

3.6.3 S^J -g^* l >n-t> ^ 332

3.6.3.1 M.% -£^*1 332

3.6.3.2 I7fl 711^71 *&J£ 333

3.6.3.3 REFORM 7 ^ ^ 2 } 333

3.7 DUPIC i - £ ^ ^ H 337

3.7.1 7]& ±M * H 337

3.7.1.1 ^ l ^ - ^ ^ r i i ^ 337

3.7.1.2 ^ # ic^J 338

3.7.1.3 Al V ^ ^ - * 1 | ^ S 4 339

3.7.2 : g ^ € - 3 yo^°fl ^ t > «15:^S. «fl 4 340

3.7.2.1 CANDU J c ^ * H # - S l t t *&&£. ^^4 tio^^- 340

3.7.2.2 ^ ^ Hf 344

3.7.2.3 - i-^S. TflAl; 347

3.7.3 ^-7)1^ UO <H1 ^?> Hl^gJE «H^ 349

3.7.3.1 tif^l ii-y 349

3.7.3.2 *m j-Ell ic^J 352

3.7.4 A«? 353

3.8 -S.<^ ^ ^ ^ r 394

3.9 Q3. €•*! 396

4 ^ . DUPIC J n ^ «ol-A>>a #&1 «l|^ 399

4.1 <&-£3^. 51 « J 4 S 403

4.1.1 2}sfl*lH 2 H 403

4.1.2 ^ H * H SH5] ^7> 405

4.1.2.1 ^ - f e j - f f Jn-y 7}R! 405

4.1.2.2 DUPIC Jt-y ^^> 409

4.2 CANDU ^ } S . ^ 1*1- 1 * H 420

- 44 -

KAERI/RR-1999/99

4.2.1 CANDU 1*} ^ l 7 i l # 420

4.2.1.1 ##*M 420

4.2.1.2 <§^I*M 421

4.2.1.3 ^ * M 421

4.2.1.4 *Ff*M • 421

4.2.2 1*> xMXW ^ 4 *)ltt*l • 421

4.2.3 DUPIC ^SH&S.oW c M \x\ ^ 1 ^ 1 1 - i%q 422

4.2.3.1 ^ ^ 7fl# 422

4.2.3.2 ^ > * } s f l X k ] ; 423

4.2.3.3 4^W^1^> 424

4.2.3.4 ^^WR!: 425

4.2.3.5 Sff^MTiRV 426

4.2.4 <i ^ - ^ fltm 426

4.2.4.1 ^«oVtS *&*¥$ $}# • • 426

4.2.4.2 «>3 W ^ ^ ^ ^]^> 427

4.2.5 &.<$ 428

4.3 #x\S. ^ S # ^ l tH*> ^ H l <%> 451

4.3.1 aouHd ^ J - *1N 451

4.3.1.1 ^<&£. *H^ l-f- 452

4.3.1.2 2HKE:TIM ^^|] 454

4.3.2 <ixM Sfl^ 455

4.3.2.1 ^ <i }3fl 455

4.3.2.2 f^> «i^H 457

4.4 «?<£S 41^-, ^ ^ ^ ^ *M • 467

4.4.1 DUPUC ^<&M. c>^5} ^ i ^ l - 468

4.4.1.1 ^1^> s.ig 468

4.4.1.2 DUPIC *!^<*IS 469

4.4.1.3 DUPIC ^HH^ ^(&g. 471

4.4.2 DUPIC ^<£5. £ ^ § 7 l 471

4.4.2.1 DUPIC ^<gSt:>^ 1 ^ § 7 ] JE.'i 471

- 45 -

KAERI/RR-1999/99

4.4.2.2 uo

uRl-& • 473

4.4.2.3 ^£-§-71 ^j-spB^ 475

4.4.3 CANDU «|&& t\^ ^%3&% ^ t > ti^HS. Jg7} 478

4.4.3.1 * | # a.«i 478

4.4.3.2 7fl*> J2.«i 480

4.4.3.3 ^2]- ^ SL$] 481

4.5 &.<% *i ^-g- 508

4.6 ^JL^rt i 510

4 5 # . DUPIC «|«iS. <g^M 6<f^^ ^ r ^ 515

5.1 < g ^ ^ sfl^j j u g 518

5.1.1 CANDU-6 <i^ ^7^| A 3 518

5.1.2 < i ^ ^ J3.»i 518

5.1.2.1 ^ ^ - * > S-^i 519

5.1.2.2 ^ ^ ^ S.<i 520

5.1.2.3 ^ ^ l ^ l l ^ l - ^ 7fl4>i«i 521

5.1.3 ti^^>S H ^ ^ ^ 3 522

5.1.3.1 i^7] 522

5.1.3.2 ^«SS. ^B^ 9i ^ ^ ^ - 523

5.1.3.3 ^lt^^> ^ 7 f l ^ ^ 523

5.2 < i ^ *H ^^4 529

5.2.1 ^3} ^J ^ ^ 529

5.2.1.1 fl #^ ^:S 529

5.2.1.2 ^ - f r ^ 529

5.2.1.3 ^ ^ H ^ 1 -^ 530

5.2.1.4 ^7fl#^Hl 530

5.2.1.5 ^H^#9- &S. 531

5.2.2 ^ T p f l ^ - t ^ 1 - % ^ ^ 531

- 46 -

KAERI/RR-1999/99

5.2.2.1 1 - S H I ^ Q°A 531

5.2.2.2 «>^nj-^ ^ 9i 1 - % ^ 532

5.3 S . ^2} SH.SI -frJBL^ 555

5.3.1 NUCIRC 2 H S ] <£7i|<i^r S . ^ 555

5.3.2 ASSERT 2 E £ ^ • 556

5.4 ^ § 565

5.5 #^*i 566

*il 6 # . DUPIC ^# .§ .^7 ] 3 * 8 ^ ^Sj 569

6.1 DUPIC ^ < £ S a|jS«l-g- 572

6.1.1 DUPIC *]<£ -iTfl^L^ 573

6.1.1.1 A]>y ^ A ^ 573

6.1.1.2 DUPIC ^<&g. %•% SL& 576

6.1.2 DUPIC «}<£S ^§-^ 7fl [ -g7(1 • 577

6.1.2.1 ^g #^ ^# 577

6.1.2.2 ^ ^ S . ^ 1 ^ ^ - ^ ! 7H^ -gTfl 579

6.1.2.3 A ^ 7flJSL 581

6.1.3 ^ 1 S Hj-g- ^ 7 } • 582

6.1.3.1 H|-§- 3g7]- >S 582

6.1.3.2 ^<&£. ^JiHl-g- 584

6.1.4 2 [ ^ ^J ^ J S l ' W 587

6.2 DUPIC ^ « ^ S ^l^-Hj-g- 608

6.2.1 7 ]^ - ^ ^ S . ^<& *£ «1# 7&S. 609

6.2.2 DUPIC « ? g £ ^ 6 J ^ Hjj^. «t-o> 6 0 9

6.2.2.1 7 ] ^ ^ - ^ $3. %-§- »^^> 610

6.2.2.2 7 | § «ol-# ?gS %-§- aoK> ( ^ ^ ^ Ho^) 612

6.2.3 ^<&g. J3J&7] vjl Af-g-^ m&g. ^ H V S ^-^4 613

- 47 -

KAERI/RR-1999/99

6.2.3.1 DUPIC ^<^^.J2.%7] ^4-§-eoV^r^ 615

6.2.3.2 4 3 - W & 5 . *\%&. J2L# 5l a^-fHg 618

6.2.3.3 ^ H ^ f ^ S ^#2. zHHl? r*l 620

6.2.4 DUPIC *?«i&Sl 4}^ &*}£- *J-§- «]-§- ^ 625

6.2.4.1 ?l*l*|# ^Hjif DUPIC ^<&g. ^ J ^ H l S j J S L ^ 625

6.2.4.2 *HH=. ^ £ 5 . 1 M #*J 7l-^(ram) J ^3:7] .' 626

6.2.4.3 4 - g ^ «f^lS ^ ^ 3 : 51 ^ r ^ 4^1-f- 626

6.2.4.4 DUPIC ^ ^ S . 3>B 4^°11 ^ r ^ l ^ ^ ^ 4 ^ # 7 ] 626

6.2.4.5 ^ ^ S . ^ l ^ - ^ 1 ^ ^f lolSSH^ ^ ^ 627

6.2.4.6 -i l -H ^ 3 627

6.2.5 ^ U | ^ %KE. ^ ^ § al-g- 628

6.2.5.1 ^<$.£. 4-1^-al-S- 628

6.2.5.2 ^ ^ S . 4 1 ^ - ^Jflttl-ig- 628

6.2.6 ^ 5 | ^ - 629

6.3 Aj~§-^ ^<&3. *|£-ti]-g- 668

6.3.1 ^ ^ ) ^ ^ 1 7 l ^ ^ ^ l 7B^ 669

6.3.1.1 J L ^ l l 7 l # ^^71^51 ^ ^ § 669

6.3.1.2 * ^ l i i JS.^ 670

6.3.2 Aj-g-^ ^ ^ S 5 ] *|-gHl-g- 3g7V 672

6.3.2.1 7|^- ^-S-al-g- S -^ 672

6.3.2.2 ^^-§-7] SM 674

6.3.2.3 x\& #S tifl^l 675

6.3.2.4 4 - § - ^ «*^1S *|£Hl-g- 677

6.3.3 4 - § - ^ ^ ^ ^ ^ ^ r ^ Jg^Sf ^>^1 «!-§- 678

6.3.3.1 ^-4 ^ ^ 3 . Sfl*} • 678

6.3.3.2 ^ § 2 f T£4\ ^^r«l-g- 679

6.3.4 ^-g- 681

6.4 DUPIC ^ < £ ^ 7 ] al-g- 699

6.4.1 7 ] ^ *$<&S. 3.*i 700

6.4.1.1 ^^:<i^ # ^ i Q$<£ 1) 701

- 48 -

KAERI/RR-1999/99

6.4.1.2 SEU/DUofl £J«> **•§-.£ 2& (yoK> 2) 701

6.4.1.3 - f £ ^o\] SR> - § ^ 4 : 2& (*£*> 3) 701

6.4.2 DUPIC ^<$.£. ^ 7 ] - f & l ^>7f 702

6.4.2.1 DUPIC ^ £ 5 . 4^ H l -§" 702

6.4.2.2 DUPIC «JgS. 4l^-«)-§- 702

6.4.2.3 -§•#*!# 91 T^«l-§- • 703

6.4.2.4 A>^f ^<&3. *!•§• B]-§- 703

6.4.3 n<&£. ^7l«]-g- ^ 4 704

6.4.3.1 « } £ l ^ 7 l Hj-g- 7?1 > ^ 704

6.4.3.2 7 ] ^ t&Ufc 91 «J«1S. ^ 7 ] J2.ig 706

6.4.3.3 ^ ^ 5 . ^ 7 1 Hl-g- TiKl 707

6.4.4 ^ . ^ 91 ^JJL A> J- 710

6.5 ^ 3 . £ r ^ 739

7 # . «!9-7l|«i ^ -S ^ ^ S . «1 *^^ «a^- 743

7.1 ^9-711 ^ " S Tg^S. 746

7.1.1 M]^- 919- •^•S'i ^ . S . ^ ^ 746

7.1.2 Afl-f <*Kz- ^ - ^ ^ A ^ 91 ^ ^ . 747

7.1.2.1 71$ i ^ ^ i l T f l ^ S } *)-§- B > ^ ^7> 747

7.1.2.2 DUPIC ^<££. ^^f^ ^gr^ B>^-^ ^ 747

7.1.2.3 Qx\S. 4 ^ l * ^ - ^ 6ovUA^ ^7 f 747

7.1.2.4 ^<&g. # ^ 3L^ ^ l ^ S . ^r^4 748

7.1.2.5 DUPIC ^ < £ S >g^j|^ ^ - ^ 748

7.1.3 <£^- ^ 4 # 749

7.1.3.1 #^S • 749

7.1.3.2 ^j-S] SCI 7114 ^ € r • 749

7.1.3.3 ^r$\ & *Hr 750

7.1.3.4 ^-t(I ^ S fe^ 752

7.1.3.5 7l^(*l2g-^H4) M.3L*\ 753

7.1.3.6 ££12^4 755

- 49 -

KAERI/RR-1999/99

7.2 *£%-

7.2.1

7.2.1.1

7.2.1.2

7.2.1.3

7.2.2

7.2.2.1

7.2.2.2

7.2.2.3

7.2.2.4

7.2.2.5

7.2.3 2 ^

7.2.3.1

7.2.3.2

7.2.3.3

7.2.3.4

756

756

756

756

• 756

757

757

# ] H^ 758

1-51 * H ^ 760

761

761

763

763

764

765

766

- 50 -

KAERI/RR-1999/99

Table 2.1-1 Design Data of Nominal Core 96

Table 2.1-2 Design Data of Fuel Channel • 97

Table 2.1-3 Design Data of Fuel 98

Table 2.1-4 Description of Computer Codes used in CANDU Core Physics

Analysis 99

Table 2.2-1 Comparison of km for Natural Uranium Fuel Lattice 117

Table 2.2-2 Comparison of Void Reactivity for Natural Uranium Fuel 118

Table 2.2-3 Comparison of Fuel Temperature Coefficient for Natural Uranium

Fuel 119

Table 2.2-4 Relative Changes in Four Factors for Natural Uranium Fuel

at Fresh State 120


at Equilibrium State 121


at Discharge State 122

Table 2.2-7 Relative Pin Power for Natural Uranium Fuel 123

Table 2.2-8 Reaction Rate Ratio for Natural Uranium Fuel 124

Table 2.2-9 Comparison of km for DUPIC Fuel Lattice 125

Table 2.2-10 Comparison of Void Reactivity for DUPIC Fuel 126

Table 2.2-11 Comparison of Fuel Temperature Coefficient for DUPIC Fuel 127

Table 2.2-12 Relative Pin Power for DUPIC Fuel 128

Table 2.2-13 Reaction Rate Ratio for DUPIC Fuel 129

Table 2.3-1 Lattice Parameters for Natural Uranium Initial Core 161

Table 2.3-2 Lattice Parameters for Irradiated Natural Uranium Fuel 162

Table 2.3-3 Incremental Cross-sections for Initial Core 163

Table 2.3-4 Incremental Cross-sections for Equilibrium Core 164

Table 2.3-5 Reactivity Change with Boron Concentration in Moderator 165

Table 2.3-6 Comparison of ZCU Reactivity Worth 166

Table 2.3-7 Calibration of Zone Controller 167

Table 2.3-8 Comparison of Average Zone Level Worth 168

- 51 -

KAERI/RR-1999/99

Table 2.3-9 Reactivity Worth of Individual Adjuster Rod 169

Table 2.3-10 Reactivity Worth of adjuster Bank 170

Table 2.3-11 Reactivity Worth of Individual Mechanical Control Absorber 171

Table 2.3-12 Reactivity Worth of Mechanical Control Absorber Bank 172

Table 2.3-13 Reactivity Worth of Individual Shutoff Rod 173

Table 2.3-14 Reactivity Change due to Heat Transport System Temperature 174

Table 2.3-15 Reactivity Change due to Moderator Temperature 175

Table 2.3-16 Comparison of Critical Core Performance Parameters 176

Table 2.3-17 Comparison of Fixed Burnup Core Performance Parameters 177

Table 2.3-18 Comparison of Zone Controller Unit Worth 178

Table 2.3-19 Comparison of Adjuster Rod Worth 179

Table 2.3-20 Comparison of Mechanical Control Absorber Worth 180

Table 2.3-21 Comparison of Shutoff Rod Worth 181

Table 2.3-22 Comparison of 600-FPD Refueling Simulation 182

Table 3.3-1 Composition Variation for Fissile Content Adjustment Option 226

Table 3.3-2 Summary of Fissile Content Adjustment Option 227

Table 3.3-3 Unit Cost of Fuel Cycle Components 228

Table 3.3-4 Summary of Reactivity Control by SEU/DU 229

Table 3.3-5 Summary of Utilization of Linear Reactivity Fuel 230

Table 3.3-6 Comparison of k<» and Isotopic Composition 231

Table 3.3-7 Comparison of koo Variation 232

Table 3.3-8 Comparison of Thermal Absorption Cross-Section 233

Table 3.3-9 Comparison of Neutron Production Cross-section (X100) 234

Table 3.4-1 Comparison of Design Parameters for DUPIC and Natural Uranium

Fuel 246

Table 3.4-2 Lattice Parameters for DUPIC Fuel 247

Table 3.4-3 Lattice Parameters for Natural Uranium Fuel 248

Table 3.4-4 Kinetic Parameters of DUPIC Fuel 249

Table 3.4-5 Kinetic Parameters of Natural Uranium Fuel 250

Table 3.4-6 Comparison of void Reactivity 251

Table 3.4-7 Reactivity Feedback (mk) due to Power Level Change 252

- 52 -

KAERI/RR-1999/99

Table 3.5-1 Reactivity Worth and Power Tilt vs. ZCU Level for DUPIC Core •• 285

Table 3.5-2 Reactivity Worth and Power Tilt vs. ZCU Level for Natural Uranium

Core 286

Table 3.5-3 Comparison of Form Factor vs. ZCU Level 287

Table 3.5-4 Power Perturbation Coefficients in DUPIC Core 288

Table 3.5-5 Thermal Flux Perturbation Coefficients in DUPIC Core 289

Table 3.5-6 Adjuster Band Reactivity Insertion Characteristics for DUPIC Core 290

Table 3.5-7 Adjuster Bank Reactivity Insertion Characteristics for Natural Uranium

Core 291

Table 3.5-8 Simulation of Startup after Short Shutdown for DUPIC Core 292

Table 3.5-9 Simulation of Startup after Short Shutdown for Natural Uranium Core •• 293

Table 3.5-10 Simulation of Startup after Poison-out Shutdown for DUPIC Core 294

Table 3.5-11 Simulation of Startup after Poison-out Shutdown for Natural Uranium

Core 295

Table 3.5-12 Simulation of Adjuster Shim Operation for DUPIC Core 296

Table 3.5-13 Simulation of Adjuster Shim Operation for Natural Uranium Core 297

Table 3.5-14 Simulation of Stepback to 60% Full Power for DUPIC Core 298

Table 3.5-15 Simulation of Stepback to 60% Full Power for Natural Uranium Core 299

Table 3.5-16 Comparison of SOR Static Reactivity Worth 300

Table 3.5-17 Comparison of SOR Insertion Characteristics 301

Table 3.5-18 Damping Factors for Xenon Oscillation 302

Table 3.5-19 Damping Factors of DUPIC Fuel Core for Different Power Levels 303

Table 3.5-20 Damping Factors of DUPIC Fuel Core fir Various Refueling Schemes •• 304

Table 3.6-1 Estimated ROP Errors and Uncertainties for DUPIC Core

(90% Confidence) 334

Table 3.6-2 Confidence for DUPIC Fuel Core with ROP Setpoint of 125%

(25 worst cases) 335

Table 3.6-3 Setpoints for Single Detecter Failure 336

Table 3.7-1 Characteristics of DUPIC Core vs. Refueling Scheme 355

Table 3.7-2 Summary of 30 Instantaneous Calculations 356

Table 3.7-3 Comparison of Refueling Simulation for 600-FPD 357

Table 3.7-4 Comparison of Probability to Exceed Administrative Limits 358

- 53 -

KAERI/RR-1999/99

Table 3.7-5 Constrained Sensitivity to Thermal Absorption Cross Section 359

Table 3.7-6 Constrained Sensitivity to Neutron Production Cross Section 360

Table 3.7-7 Comparison of Sensitivity to Thermal Absorption Cross Section 361

Table 3.7-8 Comparison of Sensitivity to Neutron Production Cross Section 362

Table 3.7-9 Sensitivity Coefficient to Thermal Absorption Cross Section for Selected

Burnup 363

Table 3.7-10 Sensitivity Coefficient to Neutron Production 364

Table 3.7-11 Uncertainty of Lattice Parameters for DUPIC Fuel Option 1 365



Table 3.7-14 Uncertainty of Performance Parameters for DUPIC Fuel Option 1 368



Table 3.7-17 Sensitivity of Clustering Group 371

Table 3.7-18 Uncertainty due to Group-average Fuel Type 372

Table 3.7-19 Comparison of Performance Parameters by Refueling Simulation 373

Table 4.1-1 Atomic Densities of Materials Used in CANDU Primary Shield

Calculation 411

Table 4.1-2 Reference Number of Meshes and Dimensions Used in End Shield

ANISN Calculation 412

Table 4.1-3 Comparison of Dose Rate through End Shield between ANISN and

MCNP-4B • 413

Table 4.1-4 Comparison of Dose Rate through End Shield for DUPIC Fuel Core .... 414

Table 4.2-1 Summary of CANDU Primary Shield Thickness and Design Criteria 430

Table 4.2-2 Comparison of Dose Rates through Primary Shields 431

Table 4.2-3 Number of Meshes and Dimensions for side shield Calculation 432

Table 4.2-4 Number of Meshes and Dimensions for Top Shield Calculation 433

Table 4.2-5 Number of Meshes and Dimensions for Bottom Shield Calculation 434

Table 4.2-6 Total Heating in Two End Shield Components during Reactor

Operation 435

Table 4.2-7 Total Heating in Side Shield Components during Reactor Operation 436

- 54 -

KAERI/RR-1999/99

Table 4.3-1 DPA at Innermost Groove of Pressure Tube to End-Fitting Rolled Joint

during 30 Years Reactor Operation 459

Table 4.3-2 DPA at Weld between Heavy Steel Plates used to Construct Calandria

Side Tube Sheets during 30 Years Reactor Operation 460

Table 4.3-3 DPA at Corner of Calandria Sub-shells and Annular Plates during

30-Years Reactor Operation 461

Table 4.4-1 Percentage of Volatile and Semi-Volatile Fission Products Removed 485

Table 4.4-2 Actinide Activity and Annual Doses from Airborne Contamination

from Fresh DUPIC Fuel 486

Table 4.4-3 Annual Gamma Dose Rates from Fresh DUPIC Bundle 487

Table 4.4-4 Neutron Sources from Spent Natural Uranium Fuel According to Cooling

Time (Unit: Meuirons/sec.MTHM) 488

Table 4.4-5 Neutron Sources from Nominal Spent DUPIC Fuel According to

Cooling Time (Unit: Meutrons/secMTHM) 489

Table 4.4-6 Neutron Sources from Over-Burned Spent DUPIC Fuel According to

Cooling Time (Unit: Meutrons/sec.MTHM) 490

Table 4.4-7 Total Gamma Source Spectrum for Conventional Spent Natural Uranium

Fuel (Unit: Meutrons/sec.MTHM) 491

Table 4.4-8 Total Gamma Source Spectrum for Nominal Spent DUPIC Fuel

(Unit: Meutrons/secMTHM) 492

Table 4.4-9 Total Gamma Source Spectrum for Over-burned Spent DUPIC Fuel

(Unit: Meutrons/sec.MTHM) 493

Table 4.4-10 Dose Rates through Cask Axial Shield 494

Table 4.4-11 Dose Rates through Cask Radial Shield 495

Table 4.4-12 Dose Rates through Cask Radial Shield Depending on Gamma Shield

Thickness 496

Table 4.4-13 keff of One Core-Load of Fuel Bundles in Contact with Each Other .... 497

Table 4.4-14 keff of One Core-Load of Fuel Bundles in 0.4 mm Separation

(Moderator-to-Volume Ratio = 1.0811) - 498

Table 4.4-15 keSf for Fuel Bundles Infinitely Stacked Criss-Crossed 499

Table 4.4-16 keff for Fuel Bundles in Single Transport Module 500

Table 4.4-17 keff for Fuel Bundles in Infinite Transport Module 501

- 55 -

KAERI/RR-1999/99

Table 5.2-1 Result of Sensitivity Analysis 534

Table 5.2-2 Selected Fuel Channels for Radial Correction Factor Calculation 535

Table 5.2-3 Radial Correction Factor 536

Table 6.1-1 Characteristics of DUPIC Fuel Bundle 588

Table 6.1-2 Nominal Level of Recycle Stream for DUPIC Process 589

Table 6.1-3 Material Flow in Main Process Building 590

Table 6.1-4 Estimated DUPIC Direct Capital Cost - 591

Table 6.1-5 Estimated Annual DUPIC Labor Cost 592

Table 6.1-6 Estimated Annual DUPIC Non-Labor Cost 593

Table 6.1-7 Inputs for Life Cycle and Unit Cost Estimation 594

Table 6.1-8 Life Cycle Cost and Unit Cost Estimation of DUPIC Fuel Fabrication

. (Discount 5 % Capacity 400 MT, Contingency 25%) 595

Table 6.1-9 Estimated Costs for DUPIC Fuel Fabrication Plant of 400 MTHE/yr

Capacity 596

Table 6.1-10 Sensitivity Analysis on Cost Parameters 597

Table 6.1-11 Sensitivity Analysis for Adding Natural Uranium 598

Table 6.1-12 Sensitivity Analysis for Adding Slightly Enriched Uranium 599

Table 6.2-1 Comparison of DUPIC Fuel Loading Path (Front Loading) 631

Table 6.2-2 Comparison of DUPIC Fuel Loading Path (Reverse Loading) 632

Table 6.2-3 Time History for Defueling of 8 Bundles per Channel 633

Table 6.2-4 Time History for Defueling of 4 Bundles in 2 Channels

(2 Bundles per Channel) 634

Table 6.2-5 Parameters for Calculation of D2O Temperature in Fueling Machine 635

Table 6.2-6 Parameters for Calculation of D2O Temperature in Storage Bay 636

Table 6.2-7 Storage Bay Temperature and Heat Load due to Spent Fuel Decay

Heat 637

Table 6.2-8 Heat Load in Storage Bay due to Instantaneous Discharge of

Full and Half Core 638

Table 6.2-9 Storage Bay Temperature and Decay Heat of Spent Fuel due to

Core Discharge 639

- 56 -

KAERI/RR-1999/99

Table 6.2-10 Time to Reach 49°C due to Malfunction of Storage Bay Cooling

System 640

Table 6.2-11 Fatigue Usage Factor for Fueling Machine 641

Table 6.2-12 Capital Cost for DUPIC Fuel Handling 642

Table 6.2-13 Unit Cost and Economic Parameters of DUPIC Fuel Handling 643

Table 6.2-14 Life Cycle and Unit Cost of DUPIC Fuel Handling 644

Table 6.3-1 Cost Estimates for Packaging and Geological Disposal of Spent Fuel •••• 682

Table 6.3-2 Comparison of Disposal Containers 683

Table 6.3-3 Summary of Repository Data for 5000 TWh Electricity Production 684

Table 6.3-4 Summary of Repository Operation Data 685

Table 6.3-5 Breakdown of Disposal Costs (1991 U$ million) 686

Table 6.3-6 Nuclear System Scenario up to 2030 687

Table 6.3-7 Results of Material Flow and Electricity Generation for Fuel Cycle

Options • 688

Table 6.3-8 Cost Break-Down for Disposal Facility (1991 U$ million) 689

Table 6.3-9 Discounted Disposal Costs for CANDU-NU Spent Fuel 690

Table 6.3-10 Discounted Disposal Costs for CANDU-DUPIC Spent Fuel 691

Table 6.3-11 Discounted Disposal Costs for PWR Spent Fuel 692

Table 6.3-12 Disposal Unit Costs for Three Different Spent Fuels 693

Table 6.4-1 Characteristics of Reference DUPIC Fuel 712

Table 6.4-2 Input Values for Fuel Cycle Components 713

Table 6.4-3 Distribution Parameters of Input Values for Uncertainty analysis 714

Table 6.4-4 Characteristics of Reference Reactors and Fuels for Once-through

and DUPIC Fuel Cycles 715

Table 6.4-5 Material Flow of Once-through Fuel Cycle based on One-Batch

Equilibrium Model 716

Table 6.4-6 Material Flow of once -through Fuel Cycle Based in One CANDU

Reactor 717

Table 6.4-7 Levelized Costs (mills/kWh) of Once-through and DUPIC Fuel Cycle

for Option 1 (Deterministic Method) 718



- 57 -

KAERI/RR-1999/99



Table 6.4-10 Summary of Levelized Fuel Cycle Costs by Deterministic

Method (mills/kWh) 721

Table 6.4-11 Results of Monte Carlo Simulation for Uncertainty Analysis of Fuel

Cycle Cost (Statistical Parameters and Percentile) 722

Table 6.4-12 Summary of Environmental Benefit of DUPIC Fuel Cycle 723

- 58 -

KAERI/RR-1999/99

Figure 2.1-1 A Simple Chart of Physics Analysis for a CANDU Reactor 100

Figure 2.1-2 Configuration of a DUPIC Fuel Lattice 101

Figure 2.1-3 Face View of Reactor Showing Fuel Channels and Calandria Shell 102

Figure 2.1-4 Plan View of Reactor Showing Layout of Reactivity Devices 103

Figure 2.1-5 Face View of Reactor Showing Zone Controllers and Adjusters 104

Figure 2.1-6 SHETAN Model of a Lattice Cell 105

Figure 2.1-7 Time-Average 8-Bundle Shift in a 12-Bundle Channel 106

Figure 2.2-1 Variation of kw for Natural Uranium Fuel Lattice 130

Figure 2.2-2 Void Reactivity Change for Natural Uranium Fuel at Fresh State 131

Figure 2.2-3 Void Reactivity Change for Natural Uranium Fuel at Equilibrium State 132

Figure2.2-4 Void Reactivity Change for Natural Uranium Fuel at Discharge State ••• 133

Figure 2.2-5 Temperature Reactivity Change for Natural Uranium Fuel

at Fresh State 134


at Equilibrium State 135


at Discharge state 136

Figure 2.2-8 Reactivity Change versus System Temperature Following a Reactor

Shutdown 137

Figure 2.2-9 Variation of km for DUPIC Fuel Lattice 138

Figure 2.2-10 Void Reactivity Change for DUPIC Fuel at Fresh State 139

Figure 2.2-11 Void Reactivity Change for DUPIC Fuel at Equilibrium State 140

Figure 2.2-12 Void Reactivity Change for DUPIC Fuel at Discharge State 141

Figure 2.2-13 Temperature Reactivity Change for DUPIC Fuel at Fresh State 142

Figure2.2-14 Temperature Reactivity Change for DUPIC Fuel at Equilibrium state •••- 143

Figure 2.2-15 Temperature Reactivity Change for DUPIC Fuel at Discharge State 144

Figure 2.2-16 Reactivity Change versus System Temperature Following a Reactor

Shutdown 145

Figure 2.3-1 SHETAN Model for Fuel Channel 183

Figure 2.3-2 SHETAN Model for Reactivity Device 184

- 59 -

KAERI/RR-1999/99

Figure 2.3-3 WIMS-AECL Slab Model for Structural Materials 185

Figure 2.3-4 Typical RFSP Nodal Model for XY Plane 186

Figure 2.3-5 Typical RFSP Nodal Model for XZ Plane 187

Figure 2.3-6 Calibration of Zone Controller 188

Figure 2.3-7 Heat Transport System Temperature Effect 189

Figure 2.3-8 Moderator Temperature Effect 190

Figure 2.3-9 Horizontal Flux Scan 191

Figure 2.3-10 Vertical Flux Scan 192

Figure 2.3-11 Comparison of Channel Power for Equilibrium Natural Uranium Core

(Critical Core) 193

Figure 2.3-12 Comparison of Channel Power for Equilibrium Natural Uranium Core

(Fixed Burnup) 194

Figure 2.3-13 Comparison of Bundle Power Distribution for Equilibrium DUPIC Core 195

Figure 3.2-1 DUPIC Fuel Lattice Model 212

Figure 3.2-2 SHETAN Model for Fuel Channel 213

Figure 3.2-3 SHETAN Model for Reactivity Device 214

Figure 3.2-4 Front View of CANDU-6 Core 215

Figure 3.2-5 Plan View of Reactivity Device Layout 216

Figure 3.3-1 Distribution of k«, for Fissile Content Adjustment Option 235

Figure 3.3-2 Distribution of kM for Spent PWR Fuel 236

Figure 3.3-3 Distribution of k» for Reactivity Control Option 237

Figure 3.4-1 Variation of £„ and keff with Burnup (PPV+WIMS) 253

Figure 3.4-2 Variation of Relative Element Linear Power with Burnup 254

Figure 3.4-3 Reactivity Change due to Moderator Temperature (WIMS) 255

Figure 3.4-4 Reactivity Change due to Coolant Temperature (WIMS) 256

Figure 3.4-5 Reactivity Change due to Fuel Temperature (WIMS) 257

Figure 3.4-6 Reactivity Change due to System Temperature Following a Reactor

Shutdown (WIMS) 258

Figure 3.4-7 Reactivity Increase due to Complete and Partial Voiding of

Coolant (WIMS) 259

Figure 3.4-8 Variation of Coolant Void Reactivity with Fuel Burnup (WIMS) - 260

- 60 -

KAERI/RR-1999/99

Figure 3.4-9 Dependence of Coolant Void Reactivity on Amount of Boron in

Moderator and Coolant Purity for DUPIC Fuel 261

Figure 3.4-10 Dependence of Coolant Void Reactivity on Amount of Boron in Moderator

and Coolant Purity for Natural Uranium Fuel (WIMS) 262

Figure3.4-ll Comparison of Power Coefficients 263

Figure 3.4-12 Reactivity Change due to Moderator D2O Purity (WIMS) 264

Figure 3.4-13 Reactivity Change due to Coolant D2O Purity (WIMS) 265

Figure 3.5-1 Comparison of ZCU Static Reactivity Worth 305

Figure 3.5-2 Power Tilts after Refueling Transient 306

Figure 3.5-3 Comparison of Xenon Load at 30-min after Shutdown 307

Figure 3.5-4 Comparison of ADJ Bank Insertion Characteristics 308

Figure 3.5-5 Xenon Buildup after Shutdown 309

Figure 3.5-6 Static Reactivity Worth of MCA 310

Figure 3.5-7 Comparison of Static Reactivity Worth Insertion Characteristics 311

Figure 3.5-8 Reactor Power for 20% RIH Break LOCA Shutdown by SDS1 312

Figure 3.5-9 Dynamic Reactivity for 20% RIH Break LOCA Shutdown by SDS1 313

Figure 3.5-10 Reactor Power for 100% RIH LOCA Shutdown by SDS2 314

Figure 3.5-11 Dynamic Reactivity for 100% RIH LOCA Shutdown by SDS2 315

Figure 3.5-12 Xenon Load after Reactor Shutdown 316

Figure 3.5-13 Xenon Load after Reactor Startup 317

Figure 3.5-14 Xenon Load after Power Setback from Full Power 318

Figure 3.5-15 Comparison of Top-to-Bottom Tilt 319

Figure 3.5-16 Comparison of Side-to-Side Tilt 320

Figure 3.5-17 Comparison of Front-to-Back Tilt 321

Figure 3.5-18 Comparison of Top-to-Bottom Oscillation with Different Power Levels

for DUPIC Core 322

Figure 3.5-19 Axial Power Shape of Central Channel for Various Refueling Schemes 323

Figure 3.5-20 Comparison of Front-to-Back Tilt for Various Refueling Schemes of

DUPIC Core 324

Figure 3.5-21 ZCU Controllability of Top-to-Bottom Oscillation of DUPIC Core 325

Figure 3.5-22 ZCU Controllability of Side-to-Side Oscillation of DUPIC Core 326

Figure 3.5-23 ZCU Controllability of Front-to-Back Oscillation of DUPIC Core 327

- 61 -

KAERI/RR-1999/99

Figure 3.7-1 Comparison of Axial Power of Channel L-3 374

Figure 3.7-2 Comparison of Horizontal Channel Power for Row M 375

Figure 3.7-3 Comparison of Vertical Channel Power for Column 11 376

Figure 3.7-4 Axial Power Shape for 2-Bundle Shift Core 377

Figure 3.7-5 Channel Power Map of Reference DUPIC Core 378

Figure 3.7-6 Maximum Channel Power for 600-FPD Simulation 379

Figure 3.7-7 Maximum Bundle Power for 600-FPD Simulation 380

Figure 3.7-8 Zone Controller Level for 600-FPD Simulation 381

Figure 3.7-9 Channel Power Peaking Factor 600-FPD Simulation 382

Figure 3.7-10 Flow Diagram of Sensitivity Calculation 383

Figure3.7-ll Age Distribution of Instantaneous Core 384

Figure 3.7-12 Distribution km for 30 Fuel Types (Option 1) 385

Figure 3.7-13 B5U Content Distribution for 30 Fuel Types (Option 2) 386

Figure 3.7-14 235U Content Distribution for 30 Fuel Types (Option 3) 387

Figure 3.7-15 Channel Power Uncertainty due to Group-Average Fuel Property

(Option 1) 388

Figure 3.7-16 Bundle Power (Position 6) Uncertainty due to Group-Average Fuel

Property (Option 1) 389

Figure 3.7-17 Channel Power Peaking Factor Uncertainty due to Group-Average

Fuel Property (Option 1) 390

Figure 3.7-18 Heterogeneity Effect on MCP during 600-FPD Simulation 391

Figure 3.7-19 Heterogeneity Effect on MBP during 600-FPD Simulation 392

Figure 3.7-20 Heterogeneity Effect on CPPF during 600-FPD Simulation 393

Figure 4.1-1 One-dimensional Model for the End Shield System 415

Figure 4.1-2 Core Channel Map for CANDU-6 Reactor 416

Figure 4.1-3 Comparison of Total Dose Rate for End Shield 417

Figure 4.1-4 Comparison of Heat Deposition Rate through End Shield for Natural

Uranium Core 418

Figure 4.1-5 Comparison of Heat Deposition Rate through End Shield

for DUPIC Core 419

Figure 4.2-1 CANDU Primary Shield System 437

- 62 -

KAERI/RR-1999/99

Figure 4.2-2 Fission Neutron Spectrum for both Natural Uranium and DUPIC Fuel •• 438

Figure 4.2-3 Coordinates Used for Source Term Generation 439

Figure 4.2-4 Axial Power Distribution for End Shield Calculation

(Average over Channels L-ll , L-12, M-ll and M-12) 440

Figure 4.2-5 Comparison of Dose Rates through End Shield 441

Figure 4.2-6 Bundle Power Distribution on Core Side

(Average over Channels L-l, L-22, M-l and M-22) 442

Figure 4.2-7 Radial Power Distribution for Side Shield Calculation 443

Figure 4.2-8 Comparison of Dose Rates through Side Shield • 444

Figure 4.2-9 Radial Power Distribution for Top Shield Calculation 445

Figure 4.2-10 Comparison of Dose Rates through Top Shield 446

Figure 4.2-11 Radial Power Distribution for Bottom Shield Calculation 447

Figure 4.2-12 Comparison of Dose Rates through Bottom Shield 448

Figure 4.2-13 Comparison of Heat Deposition Rates through End Shield during

Full Power Operation 449

Figure 4.2-14 Comparison of Heat Deposition Rates through Side Shield during

Full Power Operation 450

Figure 4.3-1 Fuel Channel System 462

Figure 4.3-2 Calandria Shell 463

Figure 4.3-3 Configuration of Natural Uranium Fuel Lattice 464

Figure 4.3-4 Configuration of DUPIC Fuel Lattice 465

Figure 4.3-5 End Thermal Shielding 466

Figure 4.4-1 a-n and Fission Neutrons from Fuel Bundle after 10-Year Decay 502

Figure 4.4-2 Spent DUPIC Fuel Storage Basket 503

Figure 4.4-3 Spontaneous Fission Spectrum of 2i2Cf 504

Figure 4.4-4 keff of Infinite Hexagonal Lattice of 37-Element Standard Natural Uranium

Fuel Bundle at Discharge Burnup State 505

Figure 4.4-5 keff of Infinite Hexagonal Lattice of 43-Element DUPIC Fuel Bundle

at Fresh Burnup State 506

Figure 4.4-6 kef/ of Infinite Hexagonal Lattice of 43-Element DUPIC Fuel Bundle

at Discharge Burnup State 507

- 63 -

KAERI/RR-1999/99

Figure 5.1-1 Slave Channel Analysis Model in NUCIRC Code 525

Figure 5.1-2 CCP Calculation Scheme in NUCIRC 526

Figure 5.1-3 Geometry of Inlet Feeder to Channel N19 527

Figure 5.1-4 Feeder Geometry Input of NUCIRC 528

Figure 5.2-1 Channel Power Distribution of DUPIC Core (kW) 537

Figure 5.2-2 Channel power Distribution of Standard Core (kW) 538

Figure 5.2-3 Axial Power Distribution of DUPIC and Standard Fuel for Channel

Ll l at 100% F.P. Normal Operating Condition 539

Figure 5.2-4 Channel Flow Rate of DUPIC Core (kg/s) 540

Figure 5.2-5 Channel Flow of Standard Core (kg/s) 541

Figure 5.2-6 Critical Channel Power in DUPIC Core 542

Figure 5.2-7 Critical Channel Power in Standard Core 543

Figure 5.2-8 Critical Power Ratio in DUPIC Core 544

Figure 5.2-9 Critical Power Ratio in Standard Core 545

Figure 5.2-10 Channel Exit Quality of DUPIC Core 546

Figure 5.2-11 Channel Exit Quality of standard Core 547

Figure5.2-12 Enthalpy of DUPIC fuel in Channel Ll l under CHF Condition 548

Figure5.2-13 Enthalpy of Standard Fuel in Channel Ll l under CHF Condition 549

Figure5.2-14 Void Distribution of DUPIC Fuel in Channel L l l under CHF

Condition 550

Figure 5.2-15 Void Distribution of Standard Fuel in Channel Ll l under CHF

Condition 551

Figure5.2-16 CHFR of DUPIC Fuel in Channel L l l under CHF Condition 552

Figure5.2-17 CHFR of Standard Fuel in Channel Ll l under CHF Condition 553

Figure5.2-18 Axial Distribution of CHFR in Channel Ll l under CHF Condition 554

Figure 5.3-1 Heat Flux Versus Quality at Location of BT, Freon-114 Annulus Data,

Dl=0.563 in., D2=0.875 in. and 12 ft Heated Length 558

Figure 5.3-2 Critical Quality Versus Boiling Length, Freon-114 Annulus Data 559

Figure 5.3-3 Critical Quality Versus Boiling Length Data, 12.6 mm Round Tube

3.66 m Heated Length 560

Figure 5.3-4 Subchannel and Rod Numbering in ASSERT Validation for Standard

Fuel Bundle Simulation 561

- 64 -

KAERI/RR-1999/99

Figure 5.3-5 Pressure Drop and Void Profiles Simulated by ASSERT Code 562

Figure 5.3-6 Measured and Computed Fuel Rod Surface Temperature in Different

Subchannels 563

Figure 5.3-7 Measured and Computed CHF for Standard Fuel Bundle Experiments ••• 564

Figure 6.1-1 Schematic Process for DUPIC Facility Cost Evaluation 600

Figure 6.1-2 Configuration of DUPIC Fuel Bundle 601

Figure 6.1-3 Accounting Methodology in DUPIC 602

Figure 6.1-4 Pictorial Illustration of DUPIC Process 603

Figure 6.1-5 DUPIC Process Mass Balance Schematic 604

Figure 6.1-6 DUPIC Facility Area Plot 605

Figure 6.1-7 Main Process Building Floor Plot 606

Figure 6,1-8 Sensitivity of Cost Parameters 607

Figure 6.2-1 CANDU-6 Refueling Sequence 645

Figure 6.2-2 Current CANDU-6 Fuel Transfer Path 646

Figure 6.2-3 CANDU-6 Fuel Handling System 647

Figure 6.2-4 Spent Fuel Discharge Elevator 648

Figure 6.2-5 Spent Fuel Transfer Equipment 649

Figure 6.2-6 Spent Fuel Storage Tray 650

Figure 6.2-7 Spent Fuel Shielded Basket Drying and Welding Station 651

Figure 6.2-8 Short-Term Spent DUPIC Fuel Decay Heat per Bundle 652

Figure 6.2-9 Long-Term Spent DUPIC Fuel Decay Heat per Bundle 653

Figure 6.2-10 Magazine Temperature from 4-Bundle Shift (2 Bundles per Channel)

Refueling 654

Figure 6.2-11 Magazine Temperature from 8-Bundle Shift (2 Bundles per Channel)

Refueling • 655

Figure 6.2-12 Magazine Temperature from 8-Bundle Shift Refueling per Channel 656

Figure 6.2-13 Storage Bay Temperature from Spent Fuel Spent Fuel Decay Heat 657

Figure 6.2-14 Storage Bay Heat Load from Spent Fuel Decay Heat 658

Figure 6.2-15 Storage Bay Temperature from Full Core Dump after 31-Years

of Reactor Operation 659

Figure 6.2-16 Storage Bay Heat Load form Full Core Dump after 31-Years

- 65 -

KAERI/RR-1999/99


Figure 6.2-17 Storage Bay Temperature from Full Core Dump after 12-Years


Figure 6.2-18 Storage Bay Heat Load from Full core Dump after 12-Years


Figure 6.2-19 Storage Bay Temperature from Half Core Dump after 31-Years


Figure 6.2-20 Storage Bay Heat Load from Half Core Dump after 31-Years


Figure 6.2-21 Storage Bay Temperature from Half Core Dump after 12-Years


Figure 6.2-22 Storage Bay Heat Load from Half Core Dump after 12-Years


Figure 6.2-23 DUPIC Fuel Transfer Path 667

Figure 6.3-1 Procedure of HLW Disposal Cost Estimation 694

Figure 6.3-2 Spent Fuel Disposal Facility Perspective 695

Figure 6.3-3 Waste Emplacement Geometry for an Underground Facility 696

Figure 6.3-4 Installed Capacity of Nuclear Power Plants 697

Figure 6.3-5 Disposal Unit Costs for Spent CANDU-NU, CANDU-DUPIC and

PWR Fuels 698

Figure 6.4-1 Procedure of Cost Analysis of DUPIC Fuel Cycle 724

Figure 6.4-2 Triangular Distribution Function of Natural Uranium

(Minimum=15, Mode=19.2 and Maximum=35$/kg) 725

Figure 6.4-3 Components and Time Frame of Once-through and DUPIC Fuel Cycles 726

Figure 6.4-4 Histogram of Fuel Cycle Cost for DUPIC Fuel Option 1 727



Figure 6.4-7 Comparison of Probabilistic Density Function of Fuel Cycle Cost for

Option 1 730


Option 2 731


- 66 -

KAERI/RR-1999/99

Option 3 732

Figure 6.4-10 Sensitivity of Fuel Cycle Component for Option 1 733



Figure 6.4-13 Natural Uranium Saving and Spent Fuel Reduction for DUPIC Fuel

Option 1 (based on annual requirement of one CANDU reactor) 736





- 67 -

KAERI/RR-1999/99

KAERI/RR-1999/99

1. M

DUPIC

(SEU),M

AIROX

y 1990 1

DUPIC ^ ^ ^ ^ 7 ] 71^-

2000V1 3 ^ 31 <g

A ^ ^ ^ S . 1997V1 7 ^ 21

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1.1

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. n sm DUPIC4^5171

DUPIC

- 72 -

KAERI/RR-1999/99

1.2

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71$

, DUPIC

DUPIC

. DUPIC

, ZL # DUPIC ^ DUPIC

DUPIC 71

DUPIC

DUPIC D U P I C

DUPIC

lfe DUPIC

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- 73 -

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-&

(1997.7.21-

1998.3.31)

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DUPIC

, Poison

7]

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DUPIC

•DUPIC

• DUPIC

DUPIC

(1998.4.1-

1999.3.31)

DUPIC

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- 74 -

KAERl/RR-1999/99

(1998.4.1-

1999.3.31)

(1999.4.1-

2000.3.31)

DUPIC i^QS.

DUPIC ^<&.g.

- DUPIC ^g^l

• *%<&£. M

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- DUPIC ^<?1

•«»-*— jzt ^i O

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8"*1 7 1 ^ Jg7|-

• ^ 1 ' ^ . ' ^ (^1^.5.

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431,

- 75 -

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1.3 ^ l f l * &%

1.3.1

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Pu239/241

1.2-1.5

^ . e ^ 235^ ^HO>O] o.7

wt% ILK} 3.7] n^6\)

7 ] # 3 . AIROX

AIROX

MOX ^ ^ S . A f ^ H1J2. ^ 7 } 1 - ^*5t> ^ ^ 1 - AIROX

^ CANDU 7||3O> *%<&$. ^7]S.x\ ^ ^ ^ f s f e (SEU),

(RU), ^ ^ « f ^ S (MOX) ^ - l 4

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NRC ^*)1*W AECLo]

3.71]

- 76 -

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CANDU

22 «f 84

-fettle ^ 50%<Hl

20%

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^r WIMS

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- 77 -

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DUPIC ^<£^ 4i^Sofl c||t> OT^ ^ 7 f e DUPIC

1.3.3

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SEU

^ ZL

- 78 -

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n.

- 79 -

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1.4 4 l } ^ j £ ^ ^

1.4.1

1.4.1.1 ^Tfl/SJH 2 H

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^ SHETAN) ^

^ ^ S j 5 i c > . ZLB|T4 SEU iJ

n ^ ^ ^ ^ ^ . *H<>1| ^f§-5l^I ^ f e WIMS

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DUPIC J n ^ ^ ^ ^ j g 7 ^ <*m6\]x\ ^*$51 aj- ^u]-. ZL^X} AECL^

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KAERI/RR-1999/99

1.4.1.4 # ^ ^ }

43.3. ^ ^ < > 1 ^^5|<H 51^ . DUPIC

1.4.1.5

7 l ^ & o)n)

. ZLB]V|- DUPIC « | 1 ^ | ^ V 4 ^ ^ ^

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1.4.1.6

It:}

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1.4.2.1

o

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MCNP#

1.4.2.2

DUPIC

AECLO]

DUPIC

1.4.2.3

DUPIC

|, DUPIC

4 T^^ofl cfl*H

1.4.2.4

DUPIC AECLo]

- 82 -

1.4.2.5

1.4.2.6

KAERI/RR-1999/99

DUPIC AECLo] $1

DUPIC

- 83 -

KAERI/RR-1999/99

1.5 ^ J L ^ r ^ i

1. P.S.W. CHAN and A.R. DASTUR, "Fuelling Schemes for the Conversion of Existing

CANDU's from Natural to Enriched Fuel Cycles", Proc. Topical Meeting on Advances in

Fuel Management, Pinehurst, USA, 1986.

2. P.G. BOCZAR, H.G. BLUNDELL and MT. VAN DYK, "Fuel Management Simulations

for a Part-Core Loading of Slightly Enriched Uranium in a CANDU-600", AECL-9530, Atomic

Energy of Canada Limited, 1987.

3. M.H. YOUNIS and P.G. BOCZAR, "Equilibrium Fuel-Management Simulations for 1.2%

SEU in a CANDU 6", AECL-9986, Atomic Energy of Canada Limited, 1989.

4. M.H. YOUNIS and P.G. BOCZAR, "Axial Shuffling Fuel-Management Schemes for 1.2%

SEU in CANDU", AECL-10055, Atomic Energy of Canada Limited, 1989.

5. P.G. BOCZAR, IJ. HASTINGS and A. CELLI, "Recycling in CANDU of Uranium and/or

Plutonium from Spent LWR Fuel", AECL-10018, Atomic Energy of Canada Limited, 1989.

6. R.E. GREEN, P.G. BOCZAR and I.J. HASTINGS, "Advanced Fuel Cycles for CANDU

Reactors", AECL-9755, Atomic Energy of Canada Limited, 1988.

7. S.N. JAHSHAN and T.J. MCGEEHAN, "An Evaluation of the Deployment of AIROX-

Recycled Fuel in Pressurized Water Reactors", Nucl. Tech. Vol.106, 1994.

8. H. FEINROTH, "An Overview of the AIROX Process and Its Potential for Nuclear Fuel

Cycle", Proc. Int. Conf. and Tech. Exhibition on Future Nuclear System: Emerging Fuel Cycles

and Waste Disposal Options, GLOBAL'93, Seattle, USA, 1993.

9. L.L. PEREZ TUMINI et al., "Study of a Tandem Fuel Cycle Between a Brazilian PWR

(ANGRA-I) and an Argentinean CANDU (EMBALSE)", Annals of Nuclear Energy Vol.22,

pp. 1-10, 1995.

- 84 -

KAERI/RR-1999/99

10. D.F. TORGERSON, P.G. BOCZAR and A.R. DASTUR, "CANDU Fuel Cycle Flexibility",

9th Pacific Basin Nuclear Conference, Sydney, May 1994.

11. G. HORHOIANU, D.R. MOSCALU, G. OLTEANU and D.V. IONESCU, "Development

of SEU-43 Fuel Bundle for CANDU Type Reactors", Annals of Nuclear Energy, Vol.25,

pp.1363-1372, 1998.

12. I. PATRULESCU, M. CONSTANTIN, E. RADES and V. BALACEANU, "Development

of CANDU Reactor Physics Calculation System Based on WIMS Code", Annals of Nuclear

Energy, Vol.24, pp.1105-1125, 1997.

13. J.B. SLATERr and C.S. RIM, "AECL-KAERI Joint Research Program TANDEM Fuel Cycles

Phase I Program - 1983", CRNL-2772, Atomic Energy of Canada Limited, 1984.

14. ^ ^ 4,

, KAERI/RR-1123/92, June 1992.

15. H.B. CHOI, B.W. RHEE and H.S. PARK, "Physics Study on Direct Use of Spent Pressurized

Water Reactor Fuel in CANDU (DUPIC)", Nucl. Sci. Eng: 126, pp.80-93, 1997.

16. a l % ^ }, % ^

, KAERI/RR-1696/96, July 1997.

17. H.B. CHOI, J.W. CHOI and M.S. YANG, "Composition Adjustment on Direct Use of Spent

Pressurized Water Reactor Fuel in CANDU", Nucl. Sci. Eng: 131, pp.62-77, 1999.

18. C.J. JEONG and H.B. CHOI, "Compatibility Analysis on Existing Reactivity Devices in

CANDU 6 Reactors for DUPIC Fuel Cycle", Nucl. Sci. Eng: 134, pp.1-16, 2000.

- 85 -

KAERI/RR-1999/99

K DUPIC

- 87 -

KAERI/RR-1999/99

2. DUPIC

DUPIC

DUPIC

CANDU

. DUPIC

1.45 wt%# 7)

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POWDERPUFS-V (PPV) [Ref. 4], MULTICELL [Ref. 5],

H-i- Af§-$Vfe ^>^^1, DUPIC ^ ^ 5 . *H^^fe ^

[Ref. 7], 3 ^ } ^ ^ - ^ a . S ^ l SHETAN [Ref. 8], ~Le]Jl

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[Ref.

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PPV/MULTICELL/RFSP

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^ DUPIC

CANDU

g: WIMS/SHETAN/

- 89 -

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2.1 CANDU

CANDU

CANDU

2.1.1

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.1-2fe CANDU-6 DUPIC

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- 91 -

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WIMS

- 92 -

KAERI/RR-1999/99

4

£: SHETAN fl}

.el^l ^^^B <8<3£| ^ r ^ ^ r * WIMS-AECL

$ # } SHETAN

SHETAN S ^ f e fl^ £

1/8 ^<$.S. t\*£ SMio] ^^g-g-^) tflsfl n^J 2.1-6^1

2.1.3.2

2ZL1- ^ > 3 H t i RFSP

. RFSP S ^ #

1 r"2

= -f- J Ha)dm (2.1-1)

- 93 -

KAERI/RR-1999/99

L, Aa> = 0)2

. A 0)1= 4

(2.1-2)

ZL 0)2 7} ^ ^ 1

_1 —

coo

(2-1-4)

- 94 -

KAERI/RR-1999/99

5}

Silt:}.

RFSP

f(i,i)*(o>2(k) - (2.1-5)

±3. RFSP S ^

- 95 -

Table 2.1-1

Design Data of Nominal Core

KAERI/RR-1999/99

Number of fuel channels

Lattice pitch

Inner radius of main calandria

Inner radius of subcalandria

Length of calandria notch

Length of fuel channel (12 fuel bundles)

Extrapolated* length of fuel channel

Extrapolated reactor radius

Reactor core radius+

Reflector thickness

380

28.575 cm (square)

379.7 cm

337.8 cm

96.52 cm

594.4 cm

606.0 cm

384.7 cm

314.3 cm

65.6 cm

Moderator/reflector volumetric average temperature

Moderator/reflector D2O purity

69 °C"

99.85 wt%+

Number of adjusters

Number of light water zone control units

Number of mechanical control absorbers (cadmium)

Number of shutoff rods (cadmium)

Number of liquid poison injection nozzles

Number of vertical flux detector assemblies

Number of horizontal flux detector assemblies

21

6

4

28

6

26

7

Total fission power

Total reactor power

Total electrical power

2158.5 MW

2061.4 MW(th)

713 MW(e)"'

**

***

Extrapolated boundaries are used for the purpose of core diffusion calculations.

This is given by K X (core radius)2 = 380 x (pitch)2

Temperature at the moderator outlet.

Nominal design value, operating purity might be higher to improve fuel burnup.

Gross nominal

- 96 -

KAERI/RR-1999/99

Table 2.1-2

Design Data of Fuel Channel

Number of fuel channels 380

Length of fuel channel (12 bundles) 594.4 cm

Pressure tube (Zr-2.5% Nb) inside diameter 10.3378 cm

Average pressure tube wall thickness 0.4343 cm

Calandria tube (Zr-2) inside diameter 12.8956 cm

Average calandria tube wall thickness 0.1397 cm

Coolant temperature averaged over channel 288 °C+

Coolant D2O Purity 99.10 wt%+

Average-to-maximum channel power in core (Radial form factor) 0.821*

Average-to-maximum bundle power in core (Overall form factor) 0.559*

Average-to-maximum bundle power in a channel,

averaged for all channels (Axial form factor) 0.672*

* Assuming bi-directional eight-bundle shift fueling

+ Nominal design value

- 97 -

KAERI/RR-1999/99

Table 2.1-3

Design Data of Fuel

Bundle design

Element (sheath) outside diameter

Average sheath wall thickness

Pellet outside diameter

Stack length

43-element cluster

1.350 cm (large), 1.150 cm (small)

0.039 cm (large), 0.036 cm (small)

1.2665 cm (large), 1.0725 cm (small)

48.2 cm

Fuel material

Pellet density23SU content239Pu content

Fissile content

Dysprosium in center rod

(U-Pu-X)O2, spent PWR fuel

10.4 g/cm3

1.0 wt%

0.45 wt%

1.488 wt%

4.64 wt%

Weight per bundle (kg)

Uranium

Plutonium

Actinides

(U-Pu-X)

(U-Pu-X)O2

17.686

0.131

17.844

18.372

20.837

- 98 -

KAERI/RR-1999/99

Table 2.1-4

Description of Computer Codes used in CANDU Core Physics Analysis

Type Name Description

Lattice codes POWDERPUFS-V Basically a one-group (Westcott) treatment; uses

(one-dimensional) semi-empirical expressions derived from experiment

data

WIMS-AECL Multi-group transport code used to provide lattice cell

data benchmark purposes

Supercell codes MULTICELL

SHETAN

3-D two-group diffusion calculation using the

supercell approach

3-D multi-group transport code

Core design code RFSP Flux calculation module based on CHEBY; calculates

time-average and instantaneous fluxes, power and

burnup distributions, simulates different kinds of

reactor operation (in particular, refueling under

various rules) by taking time steps based on

previously calculated fluxes.

Kinetic code CERBERUS 3-D two group kinetics code based on the Improved

Quasi-static Method

Also exists as module CERBERUS in RFSP

Spatial control RFSP Diffusion calculation based on CHEBY; calculates

xenon transients and spatial control; automatically

searches the time when the change in xenon balances

a defined reactivity insertion.

- 99 -

KAERI/RR-1999/99

REACTOR DESIGN REQUIREMENT:

-FUELCHANNEL- POWER LIMITATIONS

KINETIC STUDIES:

- CERBERUS

LATTICE CODES:

- POWDERPUFS- WIMS-AECL

CORE DESIGNCONTROL DEVICESSHUTDOWN SYSTEMS:

-RFSP

FUEL MANAGEMENTSTUDIES:

-RFSP

1I]

REACTIVITY DEVICESIMUATIONS;

- MULTICELL-SHETAN

^ ,J

REACTOR CONTROLTRANSIENPERTURBS

-RFSP- CHEBXEIV

TS DUE TOLTIONS:

1AX

Fig. 2.1-1 A Simple Chart of Physics Analysis for a CANDU Reactor

- 100 -

KAERI/RR-1999/99

Fuel Elements

D2O Primary Coolan

Pressure Tube

Gas Annulus

Calandria Tube

Moderator

Fig. 2.1-2 Configuration of a DUPIC Fuel Lattice

- 101 -

KAERI/RR-1999/99

CHANNEL COLUMNDESIGNATIONS

CHANNEL ROWDESIGNATIONS

20 21

uV

w

\ \\

22

337.8cm

379.7cm

Fig. 2.1-3 Face View of Reactor Showing Fuel Channels and Calandria Shell

- 102 -

KAERI/RR-1999/99

N

CHANNEL ROW DESIGNATION T

1 2 3 4 5 6 7 8 9 10 11

SIGNA'379.7 cm

337.8 cm

2 13 14 15 16 17 18 19 20 21 22i i i i

FUEL STRING INSIDE OF TUBE SHEET\

© ADJUSTER RODS(21) AND ROD NUMBER

0 ZONE CONTROL RODS(6)

l2 MECHANICAL CONTROL ABSORBERSC-t)

Fig. 2.1-4 Plan View of Reactor Showing Layout of Reactivity Devices

- 103 -

KAERI/RR-1999/99

LIQUID ZONECONTROLLERS

GRADED ADJUSTERS

L.P.=LATTICE PITCH

Fig. 2.1-5 Face View of Reactor Showing Zone Controllers and Adjusters

- 104 -

KAERI/RR-1999/99

\

x Stainless SteelTube-and-RodAdjuster

\Fuel Channel

Fig. 2.1-6 SHETAN Model of a Lattice Cell

- 105 -

KAERI/RR-1999/99

POSITION 1 2 3 4 5 6 7 8 9 10 II 12

FLUX (J) • 0. ^) ^J, ^ ) , i$ . 0- 0K 0 ^ i n ^ ^

t = 0

t = T

1 = 0

O>,=

0

C02 =

0

0 ) , =

0

C04=

0

" 5 =

0

" 6 =

0

0 ) , =

0 0

0 ) 9 =

q>,T cp,T

C012 =

<P4T

c o , = 0) , -'2

(D,T <t>+T

0 ) 7 =

<t>7T

£0, &),„= w,,=

0), CO2 =

o" 6 co ,=

CP2T

co 1 2 =

AVERAGE DISCHARGE =1/8 ( O 5

IRRADIIAT1ONT/8 { ( D , + 0 2 +

Fig. 2.1-7 Time-Average 8-Bundle Shift in a 12-Bvindle Channel

- 106 -

KAERI/RR-1999/99

2.2

WIMS-AECL

[15-19H . DUPIC

Mosteller^ 20

MCNPt-

CANDU

. WIMS-AECL S H f e

DUPIC

J£#5|(doppler)

MCNP

*H ENDF/B-VI release

}B.e |B |^ TRX-1, 2, BAPL-1, 2,

i f BAPL

# Wfe 2 STD(Standard Deviations)

KENO

ENDF60-2J- ENDF50 Bl-o]a.

KENO

}-. TRX

MCNP «>-§-

MCNP BW^-eMfe

2.2.1

CANDU

-t>#>JL MCNP 7i

VI e fo l^^Bl^ - Afg-*H

4 1362} 7 S

>. MCNP 7)1 A

^- WIMS-AECL^

WIMS-AECL 2{x} ENDF/B-Vi]-

0.5

- 107 -

te «J4(specular reflection)

40007fl£| ^

KAERI/RR-1999/99

. JE.^- MCNP

1000 % ^

2.2.2

CANDU

71S

o, 3980, 7228

2.2.2.1

>. ENDF/B-V

r MCNP#

^ o | WIMS-AECL

WIMS-AECL

0.42% dk

o]t\. ZLS]v} ENDF/B-VI B>oia.

2.2.2.2

7]5. ^^rfe

. WIMS-AECL#

3. 0.0001

MCNP 7Jl^>^- 0.807859^ 0.0001

CANDU

0.807859, 0.7, 0.5, 0.3, 0.1

7}*]

- 108 -

KAERI/RR-1999/99

(0.807859 g/c

av («*)== 1000 xf-r-1

L "nominominal "perturb

ZL^J 2.2-2, 2.2-3, ZLZ]3L 2.2-4<Hl £^|*fSdt^-^, S 2.2-2

0.0001 g/cm3AM l ^Sf^ ^ 4 4 71S ^^M- 4 <

WIMS-AECLofl

MCNPif H]3.*||# icfl 5%

2.2.23

CANDU

7]

(2-2"2)

A,21- A2fe 4 4

960.16°K

MCNPif

- 109 -

KAERI/RR-1999/99

WIMS-AECL (ENDF/B-V) Tfl-tftteJ ^tfl M _£*fe 0.68 X 1O*JA/K°|;2, ^tfl

^ ~77%o]t:}. ENDF/B-VI

WIMS-AECL-^

1.51*10*

MCNPSJ

2STD l ) R > |

293.16, 473.16, 673.16, 873.16, 960.16, 1073.16, 1273.16 H.?]3. 1473.16°K«H]

T x 1 (2.2-3)"960.16 A ( A J

ZL^ 2.2-5O1H 2.2-7^

1 / f e 1STD ^ 68%^ ^1S]S.# ^ ^ c } . ENDF/B-V

WIMS-AECL^ «>-§-£ A ^ MCNPl- o|-g-«> ^ ^ 3 f «]3 .*H # 4 2STD

(95.5%^ - i l^£)5] ^ ^ o f l ^ <y*l*M, ENDF/B-VI

oflA-] 1STD ^^1

WIMS-AECL

(2.2-4)

- 110 -

/ =

5L7\

2.2.2.4

cell

KAERI/RR-1999/99

(2.2-6)

(2.2-7)

(2.2-8)

o)t:f. 2.2-6^ 4

o]

4- CZP

PHTS

56O.16°K<H1

1 96O.16°K<>11

2.2-8^)

7il<i

- Ill -

KAERI/RR-1999/99

3.71

ENDF/B-V e M - * M s l # °]-8"t> WIMS-AECL^ PHTS

MCNP TllAVl^if H]J2.*J; ttfl 1STD3} ^ $ J

2STDJit:l- QQ 3.7\] ^cfl ^ 7 } % T : > . ENDF/B-VI B M ^ -

WIMS-AECL^} PHTS «>-§-,£ 7 } ] ^ ^ 3:7) ^EUoJ-Hfe- I S T D L H H z±

2STD o]* fa 2f^ 3§ 7>^r:f. O]BJ^> ^ ^ - ^ ZL^| 2.2-6<Hl

fl ENDF/B-VI-1- 4-§-*ffe WIMS-AECL

2.2.2.5

(2-2-9)

z i

2.2-7^1

K ENDF/B-VSf -VI

-g-MCNP

ENDF/B-V

Milgramol26

WIMS-AECLS

2.0%

# 4 , ENDF/B-VI

WIMS-AECL T

I CANDU

2.2.2.6

h§-

- 112 -

KAERI7RR-1999/99

, a28, P2 8 , c* )#

£ a|j3L*H a 2.2-8&J1

WIMS-AECL «>-§--§•

. ENDF/B-VI B H J ^ -

ENDF/B-V B H -

2.2.3 DUPIC

CANDU

DUPIC

DUPIC

CANDU

. H5lu> DUPIC

0, 7419 ^ 14825

2.2.3.1

DUPIC

MCNP

fi= MCNP

] WIMS-AECL^

ENDF/B-Vif ENDF/B-VI

*Hfe 4 4 0.73 0.05%

WIMS-AECL

i|ofl ^ i r} . MCNP

ENDF/B-V e}o)a.

ENDF/B-VI e W

o ] # S 2.2-9^1

WIMS-AECL^

2.2.3.2

2.2-10

0.0001

- 113 -

KAERI/RR-1999/99

g/cm3*}*!

^ MCNP

H ] 3 . ^ ttfl WIMS-AECL

2STD

> WIMS-AECL 7

. 7 1 5

2.2-

MCNP

ENDF/B-Vi} -VI a H ^ -

4 4 3.4

DUPIC

2.2.3.3

DUPIC ^

$- <&£[ rfe MCNP

WIMS-AECL ^l^^l - MCNP

fe fi 2.2-1- ENDF/B-V e}ol«.

1STD

]. ENDF/B-VI

(50%)o]t;>. DUPIC ^ ^ S - S WIMS-AECL5.

^ ^ i S . €• A*>7} sac>. nelvl- MCNP ^]

ufl, WIMS-AECLol ofl^.*}^ ^ - s . ^ ] ^ ^ MCNP

, 239Pu ^

DUPIC

, DUPIC

DUPIC

MCNP

fe MCNP

MCNP

^ 2STD

WIMS-AECL

WIMS-AECLS.

. ZL ^2}fe ^ ^ 2.2-13 J j ^ 2.2-15^1

1STD (68% ^.S\S.

ENDF-V s>o| H.

; ENDF/B-VI 3 H

- 114 -

KAERI/RR-1999/99

2.2.3.4 £} £ | | K

PHTS^ «hg-S.^3H- ZL J 2.2-16«^

H«R ^ * 3ter ^ - f 5 f e (transuranic)

H f * h } ) ^ > o|-j | .s, DUPIC ^ ^ ^ . ^ PHTS

}. ENDF/B-v eH-^-

PHTS «>-§-£ 7ll>iKgr MCNP

A-] ^ *i*l*}fe a J ^ ENDF/B-VI e}oia . s le ]^ . o|-g-«> WIMS-

AECL3| PHTS «}-§-£ Til^^r ZL^ 2.2-8^

2.2.3.5

fffl # 3. -S-M DUPIC2.2-12ofl

oB ^ ^ 5 - ^- # ^ ^ ^^H *Hfe ENDF/B-v f -vi tiHH.2.6 91

2.2.3.6

DUPIC ^ ^ ^ ^ « K g ^ i - WIMS-AECL5.

2.2-13oH A^*>5ic.>. WIMS-AECL2]- MCNP Afo]^

2.2.4

LH*1 WIMS-AECL-i-

^ H ^ ^ ^ S . ^ - ^^ [ -T-5 | -H- ^ DUPIC

7 ] ^ & ^ ^ 1 ^ : MCNPl- A]-g-*|-^[i:l-. ^<?l-?-e}^ ^<aS.^l tW<*l, WIMS-

- 115 -

KAERI/RR-1999/99

AECL

fe ENDF/B-V

3.$^ DUPIC

MCNP

ENDF/B-VI

CANDU

MCNPAj-

WIMS-AECL 7

DUPIC ^^S<HI

WIMS-AECL611

5% <>

MCNP 0.73%8k

f. ZL5^u>, WIMS-AECL#

WIMS-AECL BMH.

DUPIC

- 116 -

KAERI/RR-1999/99

Comparison ofTable 2.2-1for Natural Uranium Fuel Lattice

Bumup

(MWd/T)

0

3980

7228

khot

1.11776±0.00047a

1.11678 (-0.00098)

1.11870 (0.00094)

1.04389 ±0.00043

1.04803 (0.00414)

1.04559 (0.00170)

0.98907 ±0.00037

0.99331 (0.00424)

0.99015 (0.00108)

Kvoid

1.13846 ±0.00047

1.13820 (-0.00026)

1.13978 (0.00132)

1.06025 ±0.00042

1.06382 (0.00357)

1.06116 (0.00091)

1.00285 ±0.00039

1.00744 (0.00459)

1.00410 (0.00125)

kcold

1.12869±0.00046

1.12990 (0.00121)

1.13057 (0.00188)

1.04744 ±0.00042

1.05146 (0.00402)

1.04776 (0.00032)

0.98979 ±0.00038

0.99318 (0.00339)

0.98886( -0.00093)

Code

MCNP

WIMS(V)

WIMS(VI)

MCNP

WIMS(V)

WIMS(VI)

MCNP

WIMS(V)

WIMS(VI)aone standard deviation

( ) difference from MCNP

WIMS(V): WIMS-AECL calculation with ENDF/B-V library

WIMS(VI): WIMS-AECL calculation with ENDF/B-VI library

- 117 -

KAERI/RR-1999/99

Table 2.2-2

Comparison of Void Reactivity8 for Natural Uranium Fuel

Burnup

(MWd/T)

0

3980

7228

MCNP

16.267 ±0.587"

14.782 ±0.573

13.893±0.538

WIMS-AECL

ENDF/B-V

16.851 (0.584)

14.163 (-0.619)

14.120 (0.227)

WIMS-AECL

ENDF/B-VI

16.532 (0.266)

14.033 (-0.749)

14.031 (0.139)

Computed as (.i/kM-i/kv<M)^im , (mk)

computed as V oL* (1000//&,,)2 + aL,* (1000/*L


- 118 -

KAERI/RR-1999/99

Table 2.2-3Comparison of Fuel Temperature Coefficient*1 for Natural Uranium Fuel

Burnup

(MWd/T)

0

3980

7228

MCNP

-11.345 ±0.934"

2.261 ±0.868

4.659 ±0.802

WIMS-AECL

ENDF/B-V

-13.665 (-2.320)

0.507 (-1.754)

4.815 (0.156)

WIMS-AECL

ENDF/B-VI

-12.164 (-0.819)

2.027 (-0.234)

6.238 (1.579)

"computed as (*n73.I6-*96(>.16)/(1473.16 K-960.16K)xl06 ,

bcomputed as VUO6/(1473.16 K-960. i6K)} 2 x(4 n i 6 + (&,.„.)


- 119 -

KAERI/RR-1999/99

Table 2.2-4

Relative Changes in Four Factors for Natural Uranium Fuel at Fresh State

FuelTemp.(K)

4°r

Aal<e>

API<P>

4f/<f>

293.16

10.397

3.314

-0.636

7.601

1.393

473.16

7.095

2.309

-0.415

5.117

0.936

673.16

3.594

1.220

-0.203

2.484

0.511

873.16

1.041

0.366

-0.055

0.705

0.149

960.16

0.000

0.000

0.000

0.000

0.000

1073.16

-1.325

-0.496

0.074

-0.873

-0.192

1273.16

-3.542

-1.352

0.240

-2.311

-0.532

1473.16

-5.656

-2.177

0.378

-3.642

-0.862

< > denotes average value of reference (960.16 K) and selected temperature

* ApT is slightly different from the sum of four factor changes. This is due to a slightly

different meaning of the milli-k (mk) used in ApT and four-factor changes.

- 120 -

KAERI/ER-1999/99

Table 2.2-5

Relative Changes in Four Factors for Natural Uranium Fuel at Equilibrium State

FuelTemp.(K)

ApT

Aq/<v>

Aef<e>

Apf<p>

Afl<f>

293.16

3.113

-3.869

0.138

8.371

-1.371

473.16

1.554

-2.991

0.157

5.661

-1.201

673.16

0.319

-1.819

0.129

2.798

-0.782

873.16

0.064

-0.528

0.037

0.776

-0.227

960.16

0.000

0.000

0.000

0.000

0.000

1073.16

-0.046

0.663

-0.046

-0.971

0.283

1273.16

0.018

1.868

-0.120

-2.551

0.815

1473.16

0.237

3.117

-0.212

-4.036

1.358




- 121 -

KAERI/RR-1999/99

Table 2.2-6

Relative Changes in Four Factors for Natural Uranium Fuel at Discharge State

FuelTemp.(K)

ApT

As/<s>

Ap/<p>

4fKf>

293.16

-0.132

-6.808

0.404

8.634

-2.373

473.16

-0.842

-5.117

0.358

5.853

-1.935

673.16

-1.075

-3.023

0.239

2.907

-1.209

873.16

-0.345

-0.880

0.064

0.819

-0.357

960.16

0.000

0.000

0.000

0.000

0.000

1073.16

0.446

1.096

-0.092

-1.016

0.437

1273.16

1.397

3.069

-0.239

-2.682

1.231

1473.16

2.497

5.060

-0.395

-4.241

2.046




- 122 -

KAERI/RR-1999/99

Table 2.2-7Relative Pin Power for Natural Uranium Fuel

Burnup

(MWd/T)

0

3980

7228

Ring 1

0.2109

±0.175%

(1.336%)

(0.981%)0.2123

±0.194%

(1.764%)

(1.098%)0.2155

±0.191%

(1.921%)

(1.214%)

Ring 2

0.2218

±0.108%

(0.569%)

(0.339%)0.2228

±0.108%

(1.481%)

(0.489%)0.2254

±0.106%

(0.963%)

(0.480%)

Ring 3

0.2520

±0.081%

(0.005%)

(0.009%)0.2517

±0.081%

(-0.172%)

(-0.059%)0.2510

±0.096%

(-0.042%)

(-0.026%)

Ring 4

0.3153

±0.072%

(-1.297%)

(-0.902%)0.3132

±0.072%

(-2.110%)

(-1.044%)0.3081

±0.078%

(-2.014%)

(-1.179%)

Code

MCNP

WIMS(V)

WIMS(VI)MCNP

WIMS(V)

WIMS(VI)MCNP

WIMS(V)

WIMS(VI)

WIMS(V): WIMS-AECL calculation with

WIMS(VI): WIMS-AECL calculation with

ENDF/B-V library.

ENDF/B-VI library.

- 123 -

KAERI/RR-1999/99

Table 2.2-8Reaction Rate Ratio for Natural Uranium Fuel

289

Sr

*286r

C

Ring 1

0.4664±0.629%(-1.200%)(-3.359%)

0.0354±0.432%(-0.483%)(-0.722%)

0.0712±0.338%(-2.924%)(-3.153%)

0.9826±0.298%(-0.228%)(-0.997%)

Ring 2

0.4453±0.288%(-0.606%)(-2.878%)

0.0337±0.237%

(0.151%)(-0.120%)

0.0667±0.172%(-0.135%)(-0.348%)

0.9681±0.149%(-0.075%)(-0.841%)

Ring 3

0.4057±0.206%(-2.259%)(-4.716%)

0.0299±0.237%

(0.508%)(-0.001%)

0.0556±0.142%

(0.205%)(-0.215%)

0.9401±0.122%(-0.588%)(-1.332%)

Ring 4

0.3831±0.171%(-1.774%)(-4.686%)

0.0245±0.135%

(2.040%)(0.930%)0.0385

±0.099%(0.805%)(0.049%)0.9216

±0.085%(-0.527%)(-1.302%)

Average

0.4028±0.118%(-1.705%)(-4.330%)

0.0280±0.111%

(1.055%)(0.346%)0.0495

±0.075%(0.236%)

(-0.258%)0.9368

±0.063%(-0.463%)(-1.226%)

Code

MCNP

WIMS(V)WIMS(VI)MCNP

WIMS(V)WIMS(VI)MCNP

WIMS(V)WIMS(VI)MCNP

WIMS(V)WIMS(VI)

WIMS(V): WIMS-AECL calculation with ENDF/B-V library.

WIMS(VI): WIMS-AECL calculation with ENDF/B-VI library.

- 124 -

KAERI/RR-1999/99

Table 2.2-9Comparison of k^ for DUPIC Fuel Lattice

Burnup

(MWd/T)

0

7419

14825

khot

1.15099 ±0.00047a

1.15222 (0.00123)

1.15049 (-0.00050)

1.03880 ±0.00041

1.04397 (0.00517)

1.03879 (-0.00001)

0.91889 ±0.00033

0.92617 (0.00728)

0.91802 (-0.00010)

1.16371 ±0.00048

1.16505 (0.00134)

1.16330 (-0.00041)

1.05243 ±0.00041

1.05727 (0.00484)

1.05213 (-0.00030)

0.93097 ±0.00035

0.93832 (0.00735)

0.93103 (0.00006)

kcold

1.15865±0.00046

1.15972 (0.00107)

1.15622 (-0.00243)

1.04296 ±0.00040

1.04779 (0.00483)

1.04173 (-0.00123)

0.91980 ±0.00033

0.92564 (0.00584)

0.91761 (-0.00219)

Code

MCNP

WIMS(V)

WIMS(VI)

MCNP

WIMS(V)

WIMS(VI)

MCNP

WIMS(V)

WIMS(VI)aone standard deviation


WIMS(V): WIMS-AECL calculation with ENDF/B-V library

WIMS(VI): WIMS-AECL calculation with ENDF/B-VI library

- 125 -

KAER17RR-1999/99

Table 2.2-10

Comparison of Void Reactivity11 for DUPIC Fuel

Burnup

(MWd/T)

0

7419

14825

MCNP

9.497±0.580b

12.467 ±0.552

14.121 ±0.523

WIMS-AECL

ENDF/B-V

9.558 (0.061)

12.050 (-0.418)

13.981 (-0.140)

WIMS-AECL

ENDF/B-VI

9.571 (0.075)

12.206 (-0.262)

14.309 (0.188)

"computed as

Computed as

, (mk)

L x (iooo/*L)2 + o


- 126 -

KAERI/RR-1999/99

Table 2.2-11Comparison of Fuel Temperature Coefficient8 for DUPIC Fuel

Burnup

(MWd/T)

0

7419

14825

MCNP

-5.497 ±1.219°

-1.891 ±1.096

4.522 ±0.997

WIMS-AECL

ENDF/B-V

-5.517 (-0.019)

-1.715 (0.175)

3.840 (-0.682)

WIMS-AECL

ENDF/B-VI

-4.581 (-0.916)

-0.936 (0.955)

4.425 (-0.097)

"computed as (*I4J3.,6-*960.,6)/(i473.l6 K-960.i6K)xio6

Computed as ^{IO6/(1473.16 K-960.16 K)}2x (a{mM + <&,.„,)


- 127 -

KAERI/RR-1999/99

Table 2.2-12Relative Pin Power for DUPIC Fuel

Burnup

(MWd/T)

0

7419

14825

Ring 1

0.1143

±0.263%

(2.101%)

(1.317%)0.1547

±0.243%

(1.717%)

(0.939%)0.1938

±0.194%

(1.451%)

(0.712%)

Ring 2

0.1998

±0.127%

(2.638%)

(2.008%)0.2226

±0.117%

(2.002%)

(1.456%)0.2338

±0.112%

(1.657%)

(1.180%)

Ring 3

0.2727

±0.089%

(0.874%)

(0.805%)0.2724

±0.090%

(0.412%)

(0.430%)0.2626

±0.093%

(0.098%)

(0.183%)

Ring 4

0.4132

±0.072%

(-2.434%)

(-1.867%)0.3503

±0.072%

(-2.351%)

(-1.674%)0.3099

±0.077%

(-2.240%)

(-1.491%)

Code

MCNP

WIMS(V)

WIMS(VI)MCNP

WIMS(V)

WIMS(VI)MCNP

WIMS(V)

WIMS(VI)



- 128 -

KAERI/RR-1999/99

Table 2.2-13Reaction Rate Ratio for DUPIC Fuel

P28

Ring 1

2.7260±0.698%(-6.427%)(-7.211%)

0.1737±0.526%(-3.131%)(-1.714%)

0.2413±0.410%

(0.068%)(0.703%)1.5367

±0.538%(-3.996%)(-5.005%)

Ring 2

1.5002±0.295%(-3.280%)(-4.619%)

0.0985±0.250%(-2.520%)(-1.455%)

0.1447±0.198%(-2.238%)(-1.845%)

1.0842±0.213%(-1.617%)(-2.668%)

Ring 3

1.1219±0.233%(-3.469%)(-5.443%)

0.0727±0.180%(-1.257%)(-0.757%)

0.1041±0.142%(-1.269%)(-1.302%)

0.9341±0.150%(-1.659%)(-2.839%)

Ring 4

0.8870±0.184%(-3.566%)(-6.077%)

0.0494±0.140%

(2.375%)(1.951%)0.0631

±0.100%(0.306%)

(-0.200%)0.8391

±0.108%(-1.745%)(-2.928%)

Average

1.1435±0.097%(-3.664%)(-5.566%)

0.0706±0.079%(-0.612%)(-0.172%)

0.0982±0.063%(-1.011%)(-1.000%)

0.9409±0.062%(-1.799%)(-2.940%)

Code

MCNP

WIMS(V)WIMS(VI)MCNP

WIMS(V)WIMS(VI)MCNP

WIMS(V)WIMS(VI)MCNP

WIMS(V)WIMS(VI)



- 129 -

KAERI/RR-1999/99

1.00-

0.98-

- - - - - WIMS(ENDF5)- ° - WIMS(ENDF6)—*— MCNP-4B

1000 2000 3000 4000 5000 6000 7000 8000

Bumup (MWD/T)

Fig. 2.2-1 Variation of k^ for Natural Uranium Fuel Lattice

- 130 -

KAERI/RR-1999/99

2 -

-• WIMS-AECLx—MCNP-4B

0.0 0.1 0.2 0.3 0.4 0.5 0.6

Coolant Density (g/cc)

0.8

Fig. 2.2-2 Void Reactivity Change for Natural Uranium Fuel at Fresh State

- 131 -

KAERI/RR-1999/99

16

14-J

£ 12-cCD

$

I

§O

10-

8-

6 -

4 -

2 -

•

I.

I.

I ,

^ \

• 1 • 1 • 1 • 1 • 1

• 1 • J 1

-----WIMS-AECL—*— MCNP-4B •

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8Coolant Density (g/cc)

Fig. 2.2-3 Void Reactivity Change for Natural Uranium Fuel at Equilibrium State

- 132 -

KAERI/RR-1999/99

3?

14

12-

10-

f 8oroa>a:

IJ5ooO

6 -

•£ 4 -

2 -

-—- WIMS-AECL-*— MCNP-4B

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8Coolant Density (g/cc)

Fig. 2.2-4 Void Reactivity Change for Natural Uranium Fuel at Discharge State

- 133

KAERI/RR-1999/99

1 0 -

— - WlMS(ENDF5)- - • - WIMS{ENDF6)— MCNP-4B

400 600 800 1000

Fuel Temperature (K)

1200 1400

Fig. 2.2-5 Temperature Reactivity Change for Natural Uranium Fuel at Fresh State

- 134 -

KAERI/RR-1999/99

encco

6I

3.0-

2.5-

1.5-

f 1.0-

0.0

WIMS(ENDF5)WIMS(ENDF6)MCNP-4B

-0.5-

-1.0400 600 800 1000


1200 1400

Fig. 2.2-6 Temperature Reactivity Change for Natural Uranium Fuel at Equilibrium State

- 135 -

KAERI/RR-1999/99

3.5-r-

3.0-

2.5-

2.0-

1.5-

1.0-

0.5-

0.0 - •

-0.5-

-1.0-

-1.5-

-2.0 - •

I1o

----- WIMS{ENDF5)--O- WIMS(ENDF6)- * — MCNP-4B

400 600 800 1000


1200 1400

Fig. 2.2-7 Temperature Reactivity Change for Natural Uranium Fuel at Discharge State

- 136 -

KAERI/RR-1999/99

CO

O>,

sa:

1 2 -

10-

8 -

6 -

4 -

2 -

0 -

- 2 -

-4 -

- 6 -

•

1

1 *

— " ^ - C l

*

•

^¥—

•

'

i *

• | i j i

X—

Fresh Stc

: _.

Equilibrii

•

- W!MS(ENDF5)- WIMS(ENDF6)- MCNP-4B

ate

m Fuel

.....

1

300 400 500 600 700

Temperature (K)

800 900 1000

Fig. 2.2-8 Reactivity Change versus System Temperature Following a Reactor Shutdown

- 137 -

KAERI/RR-1999/99

o•cCOu

oQ.

3

I

1.15-<

1.10-

1.05-

1.00-

0.95-

0.90- 1 • I • I . 1 . 1

1 ' 1

- - . -WIMS(ENDF5)- D - WIMS(ENDF6)

- x — MCNP-4B-

-

2000 4000 6000 8000 10000 12000 14000 16000

Bumup (MVMDAT)

Fig. 2.2-9 Variation of £M for DUPIC Fuel Lattice

- 138 -

KAERI/RR-1999/99

—- WIMS-AECL— MCNP-4B

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8


Fig. 2.2-10 Void Reactivity Change for DUPIC Fuel at Fresh State

- 139 -

KAERI7RR-1999/99

%Qi

C03szO

'>t5TO

a:•o'o>

ola

oo

1 3 -

12 -J

11 -

1 0 -

9 -

8 -

7 -

6 -

5 -

4 -

3 -

2 -

1 -

0 -

1 I ' ) ' 1

: ^

• I ' I ' 1

• 1 ' 1

• i • i

i • i • i

-

- • - -WIMS-AECL—«— MCNP-4B

_

•

_

--

•

-

•

-

-

• i * i • i

0.0 0.1 0.2 0.3 0.4 0.5 0.6

Coolant [Density (g/cc)

0.7 0.8

Fig. 2.2-11 Void Reactivity Change for DUPIC Fuel at Equilibrium State

- 140 -

KAERI/RR-1999/99

- - - W I M S - A E C L—*— MCNP-4B

0.3 0.4 0.5 0.6 0.7 0.8


Fig. 2.2-12 Void Reactivity Change for DUPIC Fuel at Discharge State

- 141 -

KAERI/RR-1999/99

—• - WIMS(ENDF5)- ° - WIMS(ENDF6)-*— MCNP-4B

400 600 800 1000


1200 1400

Fig. 2.2-13 Temperature Reactivity Change for DUPIC Fuel at Fresh State

- 142 -

KAERI/RR-1999/99

-—WIMS(ENDF5)-°-WMS(ENDF6)—*— MCNP-4B

-1 -

400 600 800 1000


1200 1400

Fig. 2.2-14 Temperature Reactivity Change for DUPIC Fuel at Equilibrium State

- 143 -

KAERI/RR-1999/99

3.5

3.0-

2.5--

? 0 -

- . -WIMS(ENDF5)- ° - VUMS(ENDF6)—*— MCNP-4B

-1.0-

-1.5-

-2.0

*•—

400 600 800 1000


1200 1400

Fig. 2.2-15 Temperature Reactivity Change for DUPIC Fuel at Discharge State

- 144 -

KAERI/RR-1999/99

3 -

2 -

t I"cCO

6>

0 -

-1 -

- 2 -

- 3 -

- 4 -

WIMS(ENDF5)WIMS(ENDF6)

— MCNP-4B

300 400 500 600 700

Temperature (K)

800 900 1000

Fig. 2.2-16 Reactivity Change versus System Temperature Following a Reactor Shutdown

- 145 -

KAERI/RR-1999/99

2.3

CANDU RFSP 3-S. RFSP 1 H

RFSP

SHETAN

MCNP S ^

MULTICELL# ^*fl ^flg^^x;}.. DUPIC ^ ^

^\^^S\ ^§-^r^^$] ^ A > ^ - 4 4 WIMS-AECLi)-

. WIMS-AECL ^ SHETAN2J- #&Q RFSP 3 £ * | Bj-

Phase-B i i # 5 l >y^ ^5f

87}# fl«H WIMS-AECL4

RFSP s s s

i DUPIC

CANDU

2.3.1

RFSP S E

2.3.1.1

CANDU-6

4 308.12, 308.12, 308.12 ^ 302.16°K.

4 960.16, 561.16, 561.16 ^ 342.16°K.

^-Tilfe 19.1 kg.

| ^ 48.2 cm.

49.53 cm tfofl

4

4

- 146 -

KAERI/RR-1999/99

89 ofiM*! -L## ol-g-^uK ^ # 7 1 1 ^ #n Benoist

"Bl W # ^f§-^K ^^x£ 1 5 . ^^>^l 3*M<H^=r WIMCORE [Ref. 33]

2.3.1.2

CANDU £U)-5.oM ^-%-S. *\}o]%x\ (ADJ, ZCU, MCA, HB}3. SOR)ij-

(guide tube, liquid poison injection tube, tension spring, locator, 7]Bf)i) ^

7}.

2.3-1 g 2.3-2AJ-

WIMS-AECL# *]-§-«]

>. SHETAN

. SHETAN

i, ^ ^ 8 } ^ # > f e ? ^ # ^ ^ ] ^ ^ - ^ SS304 (p =

7.9 g/cm3)^. -7-^Sj<H SI^-^ ^ 1 ^ ^ : 3.81 cm<>lt:}.

- 147 -

KAERI/RR-1999/99

71

Zircaloy-2 ( p = 6.55 g/cm3)S-

r 4.579/4.572 cmo]^ , <

4 4 6.38/6.53 ^ 0.939/1.042

^ Zircaloy-2S. ^>#«^5l-*.^ v f l i l ^ ^ ^ . 4 4 2.54 cm ^ 2.79 cm<>lt;K

7fl 1-<H ^o.v> H 3.7]7}

7? ttfl^-ofl SHETAN S^^lA-Jfe t f l l ^ A S ^ ^ # 2.79 c m £

^ ^ ^ ^ ^ S ^ 6.5232

^- 4 4 5.715/5.842 cm©]*}.

CANDU

WIMS-AECL#

- 148 -

KAERI/RR-1999/99

Inconel 0=f l* l iS^) , Zircaloy

SHETANofl

3-2]

r WIMS-AECL

SHETAN^r ^ ] ^ 3 ^ 1 ^ 7 # WIMS-AECL^]^

Inconel

SHETAN

cfl*>

fe 6.26

0.9525 g| 1.5621

4 4 1.5875 gj 1.56845

2.4, 2.4 ZLH]3. 2.6

SHETAN

-^: Zircaloy

. o]

itfl,

- 149 -

KAERI/RR-1999/99

71

^ \A 108.111224.195 cm3ol^ %• eo>^ 2.6 kgolcK

3.7}

7} 3.3. -^07> «^^cf. 4=-i-;g ^«y^- ^1^1 >*1 a l ^ ^ - ^ l ^ J ^ 13.97x18.7325x

12.7 cm ° H ^-^fe 1-27 cmo]!} 7lE} ^l^l^S] ^l-^^: 9.525x16.8295x12.7 cm

1^ 0.9525 cm

SHETAN#

0.635 cmoln^ ^ . 3§2>A] -O1^ ^ > ^ ^ - 7.62

2730.06 cm3

M 1^ Phase-B

2.3-3011

2.3-4^1

- 150 -

KAERI/RR-1999/99

2.3.1.3

RFSP 2 H

fe 4 4 XY 1LP

CANDU

°l n l 44x36x225.

"L^ 2.3-4 g 2.3-5<Hl

. X

. Y-

2.3-46)1

X-Y-Z

(1-7), (8-14), 1151:2. (15-21)5. /fl 7f£|

. 4

- 151 -

KAERI/RR-1999/99

2.3.2

W1MS-AECL 91 SHETAN4 g j ^ RFSP 3 E S ] ^ ^ . o . ^*H, Phase-B

}. Phase-Bfe ^

2.3.2.1

WIMS/RFSP f

^ . ^ , Phase-B

16.94%©lt:l-.

1)if ?J4f45l £J£fe 4 4 99.63 ^ 99.84 wt%o)t:f.

4 ^ - 4 ^ ^S.fe 4 4 308.12 9J 302.16°Ko|cl-.

55% ^<a^t:>.

9.0 ±0.5

l 3 rak ^ u f 0.997ol^JL,

^ ^ l f %51 -^^}# 4 ^ 8.5

if ^

2.3.2.2

7}.

RFSP 3.B.-

. ZCU£) #-§v£7fe

-£-3. RFSP SH<>1|

152 -

KAERI/RR-1999/99

7.7495 mk°lt:)-.

0, 10, 20, 40, 55, 60, 80, ZLB]JL 100%<H ufl ZCU

^ *HH>M SM-S.7H z%*Y PPV/RFSP^ ^

2.3-64 i-tB}i|5it|-. PPV^l^^ 71^-^S. WIMS-AECLJf V]^ 4

^ ^ 0.058 ppmoH, o | ^ -0.45

= Az3 + £22 + Cz+ D (2.3-1)

^ 2.3-6^1

0.01 K ]$^^\\ §

3.0%

RFSP

^-§-£7} i ^ f e 20% o|*fo]t^, RMS (Root Mean Square)

9.2%o]n]-. S ^ ^ - ^ - ^ 7^1^^4fe S 2.3-104 ^ M ^ l ^ K WIMS/RFSP

PPV/RFSP

14%

2.3-124

2.3-134

- 153 -

KAERI/RR-1999/99

RMS J2.*fe 4 4 11.16.8% RMS J2.*fe 4 4

25.54} 12.

WIMS/RFSP

7}

2.3.2.3

^ 8.5

ppmojt:]-. 35dfl^

fe 99.64

2.3-14 *£ ZL^ 2.3-

fe 8.5 69-f-B] 35°C7W

-c- 99.84

2.3-155} a^J 2.3-8*11

WIMS/RFSP# fe WIMS/RFSP

2.3.2.4

Phase-B

154 -

KAERI/RR-1999/99

Case 1:

Case 2: ^r 50% ^ ^ S ] a , JS

Case 3: JS.^

Case 4:

Case 5:

1, 2, 3, H <>]

8.5

91

6.6%o]t:f.

fe 34.469%^.

^ RFSP SS .u |{^ INTREP

VFD #19i} HFD #1

2.3-9^ 2.3-10

RMS ^^fe 4

2.3.3

RFSP

CANDU

RFSP#

2.3.3.1

PPV/RFSP1-

- 155 -

KAERI/RR-1999/99

50%o] u>.

7}. ti^H i

PPV/RFSP , WIMS/RFSP

WIMS/RFSP

}. ofs}# ^cfl

RMS

0.5%

4 4 4.95} 1.

^Ai 3711 l-fB

. WIMS/RFSP

. S 2.3-16^

WIMS/RFSP

WIMS/RFSP ^ 1 ^ ^

. WIMS/RFSPA]- PPV/RFSP

RMS Hfe 4 4

PPV/RFSP

0.03%

2.5%o]%3L,

2.3-17ofl

WIMS/RFSP

PPV/RFSPi}

2.3-18, 2.3-19, 2.3-20 D.

- 156 -

KAERI/RR-1999/99

2.3-2lofl

5% o]i4|S UB»gtc>. n}BM WIMS/RFSP

. 71^- PPV/RFSP SJEJ>]

4 4 4% 51 6

2.3.3.2

, 600-FPD >. DUPIC

IFPD #q±

(CPPF) - CPPF^ ROPT

•Elimination -

•Ordering -

• Selection - -f «h§-S.

- 157 -

KAERI/RR-1999/99

PPV £ fl e ^ } 4 M H r# 3 3 ^ ^ # 45- ^ ^ ^ 1 S 2.3-22<Hl AW&uf. 600-

FPD ^*> *§^ i ajcfl af l^t-s^ PPV £ WIMS 7]& >H*HH AA 6849 ^6853 kWojcf. ^-^ -f-<Hl cJ|*H *J4 ^>#1-^^: 4 4 855 «J 852

0.01 o j^oju} . ©]if ^ - ^ ^ 2 f ^ WIMS/RFSP

^ 1)71171- 2\o]5L $.<&^Z}^ n<&3.o)) t%*± CANDU

2.3.3.3

WIMS/RFSP

, WIMS/RFSP S . ^ 7 l ^ ^ CANDU

2.3.4 DUPIC

lfe. PPVAf- RFSP

^-SlM- DUPIC

f MCNP

2.3.4.1 DUPIC

DUPIC Jn'gS] ^H-fe ^ I ^ S ^ -f-W-oIcf. nfBfA] DUPIC «|

-. DUPIC

- 158 -

KAERI/RR-1999/99

RFSP 2 H S 4h&*}&t:]-. 4

CANDU

>. ZLB|x+ MCNP a.Ho

, RFSP

1/4 ^ i i f g ]7964 ^J 7206 MWd/T

4 3 7 ^

LAT» ^V§-*>Saa, 7]& ^^H-3,7} ^«> JjLg-i- <g^*}7l 41

*H U(Universe)l- A]^-tr}^o.^, 7 ] ^ 4 4 ^ S » ^fl# - f"^# <&7]*}7] 41*11

FILL ^ ^ > # AVg-*>Sit:f. <>]e|*i

2.3.4.2

t-gr DUPIC «J«1S in^J^ ^ti||7Jl^^- # ^ ^ - S # ^7] 41 *H

K MCNP Tj]^:^ i^7]# 100,000 <y^Aj- ^ 120^711-

t l ^ I # ^•«ll>II41fe 1.03825±0.00024

4 -S^fe ^cN 0.12%5koli:}. ZLe]i+

ZL^ 2.3-13OH £ A ] ^ ^ ^ ) ^ 3}tj) A^f7} 7.2%S>|

^ f ^ # ^ ^ r RFSP 3 = .

- 159 -

KAERI/RR-1999/99

MCNP

A12 ^ L22

^ RFSP MCNP

3%!-

o|

- 160 -

KAERI/RR-1999/99

Table 2.3-1

Lattice Parameters for Natural Uranium Initial Core

Natural

Uranium

Depleted

Uranium

Reflector

-£trl

2.4711E-01

2.4712E-01

2.5686E-01

-£ti2

3.7684E-01

3.7662E-01

4.0369E-01

1.6590E-03

1.6224E-03

Za2

4.0828E-03

3.7528E-03

3.8712E-04

v In

5.1039E-03

4.0697E-03

8.9448E-03

8.9660E-03

9.9859E-03

H

0.27282

0.21732

- 161 -

KAERI/RR-1999/99

Table 2.3-2

Lattice Parameters for Irradiated Natural Uranium Fuel

Burnup

(MWd/T).0

3.2

157.0

791.1

1582.4

2372.5

3161.8

3950.4

4738.8

5526.9

6314.8

7102.7

7890.5

8678.3

9466.3

10254.2

11042.3

11830.4

12618.7

13407.0

14195.5

14984.0

15772.5

16561.2

17349.9

18138.6

18927.4

19716.3

20505.1

21294.1

22083.0

22871.9

23660.9

24449.8

-£trl

2.3905E-01

2.3905E-01

2.3904E-01

2.3905E-01

2.3907E-01

2.3908E-01

2.3910E-01

2.3911E-01

2.3912E-01

2.3913E-01

2.3914E-01

2.3915E-01

2.3915E-01

2.3916E-01

2.3916E-01

2.3917E-01

2.3917E-01

2.3918E-01

2.3918E-01

2.3918Er01

2.3919E-01

2.3919E-01

2.3919E-01

2.3919E-01

2.3919E-01

2.3920E-01

2.3920E-01

2.3920E-01

2.3919E-01

2.3920E-01

2.3920E-01

2.3921E-01

2.3919E-01

2.3919E-01

Eta

3.6147E-01

3.6151E-01

3.6165E-01

3.6192E-01

3.6217E-01

3.6233E-01

3.6244E-01

3.6252E-01

3.6259E-01

3.6263E-01

3.6267E-01

3.6270E-01

3.6271E-01

3.6274E-01

3.6274E-01

3.6277E-01

3.6277E-01

3.6278E-01

3.6280E-01

3.6280E-01

3.6283E-01

3.6284E-01

3.6284E-01

3.6285E-01

3.6287E-01

3.6288E-01

3.6289E-01

3.6290E-01

3.6290E-01

3.6291E-01

3.6293E-01

3.6294E-01

3.6295E-01

3.6295E-01

1.6506E-03

1.6506E-03

1.6504E-03

1.6639E-03

1.6932E-03

1.7230E-03

1.7496E-03

1.7725E-03

1.7925E-03

1.8103E-03

1.8263E-03

1.8410E-03

1.8546E-03

1.8674E-03

1.8795E-03

1.8910E-03

1.9019E-03

1.9124E-03

1.9225E-03

1.9323E-03

1.9417E-03

1.9508E-03

1.9597E-03

1.9682E-03

1.9765E-03

1.9845E-03

1.9923E-03

1.9997E-03

2.0070E-03

2.0142E-03

2.0211E-03

2.0277E-03

2.0342E-03

2.0405E-03

3.5245E-03

3.5359E-03

3.6015E-03

3.7130E-03

3.8057E-03

3.8653E-03

3.9058E-03

3.9341E-03

3.9543E-03

3.9689E-03

3.9798E-03

3.9883E-03

3.9954E-03

4.0014E-03

4.0068E-03

4.0120E-03

4.0172E-03

4.0224E-03

4.0279E-03

4.0337E-03

4.0398E-03

4.0461 E-03

4.0527E-03

4.0594E-03

4.0662E-03

4.0731 E-03

4.0800E-03

4.0868E-03

4.0936E-03

4.1003E-03

4.1069E-03

4.1133E-03

4.1196E-03

4.1257E-03

v Ea

4.6937E-03

4.6811E-03

4.6105E-03

4.7831 E-03

4.9171E-03

4.9782E-03

4.9970E-03

4.9889E-03

4.9639E-03

4.9279E-03

4.8854E-03

4.8396E-03

4.7933E-03

4.7469E-03

4.7019E-03

4.6591E-03

4.6192E-03

4.5823E-03

4.5487E-03

4.5185E-03

4.4915E-03

4.4676E-03

4.4464E-03

4.4277E-03

4.4114E-03

4.3971 E-03

4.3845E-03

4.3734E-03

4.3637E-03

4.3551E-03

4.3474E-03

4.3405E-03

4.3343E-03

4.3286E-03

ER

8.6383E-03

8.6383E-03

8.6381E-03

8.6270E-03

8.6023E-03

8.5771E-03

8.5546E-03

8.5352E-03

8.5184E-03

8.5035E-03

8.4902E-03

8.4780E-03

8.4668E-03

8.4565E-03

8.4467E-03

8.4375E-03

8.4289E-03

8.4206E-03

8.4127E-03

8.4052E-03

8.3980E-03

8.391 OE-03

8.3843E-03

8.3779E-03

8.3717E-03

8.3657E-03

8.3598E-03

8.3543E-03

8.3489E-03

8.3437E-03

8.3386E-03

8.3337E-03

8.3289E-03

8.3243E-03

H

.25094

.25025

.24582

.25086

.25401

.25427

.25290

.25056

.24766

.24443

.24107

.23771

.23445

.23130

.22833

.22556

.22302

.22069

.21858

.21670

.21503

.21355

.21225

.21110

.21010

.20921

.20844

.20777

.20717

.20665

.20618

.20577

.20540

.20506

- 162 -

KAERI/RR-1999/99

Table 23-3

Incremental Cross-sections for Initial Core

D-TYPEC-OUTERC-INNERB-TYPE

A-OUTERA-INNER

ADJGT

SORGT

VFDGT

LPIGN

SOR/MCA

ZCRTS

ADJTS

SORTS

MODIN

ZCRBL

ADJBL

MCABL

VFDBL

ADJSB

ADJSC

ZCRNUT

ADJNUT

SORNUT

ZCAIR1

ZCAIR2

ZCAIR3

ZCH2O1

ZCH2O2

ZCH2O3

ZCRD2O

8.3098E-047.7719E-042.0964E-032.1142E-031.0179E-031.1413E-03

-4.9889E-05

-1.1012E-04

-5.4896E-05

-7.9364E-05

1.6947E-03

-1.5860E-03

-8.2475E-03

-1.3340E-03

2.4255E-02

-1.6699E-03

-3.1814E-03

-3.2742E-03

-5.8271E-03

-2.8551E-05

7.1436E-05

1.6506E-02

-4.4693E-03

2.1655E-02

-1.3629E-02

-1.7417E-02

-1.0267E-02

2.2657E-02

2.6746E-02

1.8978E-02

-9.3430E-05

1.6139E-031.4923E-033.4226E-033.4752E-03

1.9013E-032.1312E-03

-2.4244E-04

-9.7927E-04

-1.1796E-04

-6.7058E-04

8.3499E-03

-3.1247E-03

-3.7007E-03

-9.4947E-04

2.4659E-02

1.8206E-03

1.7908E-03

1.6076E-03

2.0840E-03

-1.2376E-03

5.7876E-05

4.1125E-03

-1.5992E-03

9.5742E-03

-1.9477E-02

-3.0468E-02

-8.2894E-03

1.6531E-01

1.8309E-01

1.4718E-01

-7.4697E-04

1.8248E-051.7026E-054.4053E-054.4412E-052.2182E-052.4940E-05

3.4485E-06

8.5162E-06

2.1302E-06

7.0960E-06

2.0496E-04

4.6328E-04

9.3310E-05

5.0620E-04

7.5433E-04

5.0613E-05

3.2130E-06

8.1900E-06

-7.4404E-05

1.3742E-05

5.0345E-06

6.1480E-04

7.3030E-06

7.1071E-04

-2.3138E-05

-5.6720E-05

6.4995E-06

1.8565E-04

2.2126E-04

1.5393E-04

7.0564E-06

4.7338E-044.3746E-04

9.8971E-041.0061E-035.5575E-046.2287E-04

8.4201E-06

3.7670E-05

4.9602E-06

2.4120E-05

5.8185E-03

8.4548E-04

-2.0700E-05

1.5804E-03

7.8126E-03

7.5609E-04

7.1506E-04

7.1558E-04

7.6774E-04

2.8540E-05

5.1210E-05

2.1703E-03

1.4874E-04

3.3836E-03

1.3014E-04

4.7100E-05

2.1748E-04

1.2491E-03

1.3785E-03

L1145E-03

2.7270E-05

1.2843E-041.1906E-042.7026E-042.7399E-04

1.5150E-041.6915E-04

4.6394E-06

1.8703E-05

4.1504E-06

1.1315E-05

1.5334E-03

0.0000E+00

0.0000E+00

0.0000E+O0

0.0000E+00

O.OOOOE+00

O.OOOOE+00

O.0000E+00

O.OOOOE+00

1.0059E-05

1.4384E-05

O.OOOOE+00

O.OOOOE+00

O.OOOOE+00

5.9792E-05

5.9323E-05

6.4975E-05

1.9735E-04

2.0770E-04

1.8215E-04

1.3063E-05

-4.6426E-06-1.1363E-05-8.0904E-05-8.1714E-05-1.3826E-05-1.0418E-05

-1.2143E-05

-3.4526E-05

-3.3304E-06

-2.0867E-05

-1.9035E-04

2.4219E-02

2.4284E-02

2.4016E-02

5.0044E-03

2.4860E-02

1.6329E-02

2.5370E-02

1.0074E-02

-1.0242E-05

5.9344E-06

2.4161E-02

2.0741E-02

2.4532E-02

-4.1472E-04

-6.9732E-04

-1.7105E-04

1.4150E-03

1.7131E-03

1.1541E-03

-2.5337E-05

6.4811E-036.0086E-031.3639E-021.3827E-02

7.6456E-038.5363E-03

2.3413E-04

9.4390E-04

2.0942E-04

5.7101E-04

7.7386E-02

O.OOOOE+00

O.OOOOE+00

O.OOOOE+00

O.OOOOE+00

O.OOOOE+00

O.OOOOE+00

O.OOOOE+00

O.OOOOE+00

5.0762E-04

7.2590E-04

O.OOOOE+00

O.OOOOE+00

O.OOOOE+00

3.0175E-03

2.9938E-03

3.2790E-03

9.9597E-03

1.0482E-02

9.1926E-03

6.5923E-04

- 163 -

KAERI/RR-1999/99

Table 2.3-4

Incremental Cross-sections for Equilibrium Core

A-INNERA-OUTERB-TYPE

C-IINNERC-OUTERD-TYPE

ZCH201ZCAIR1ZCH202ZCAIR2ZCH203ZCAIR3

SOR/MCA

ADJGTSORGTVFDGTLPIGN

ZCRD20ADJSBADJSC

ADJTSZCRTSSORTS

ADJUNTZCRNUTSORNUT

ADJBLZCRBLMCABLVFDBLMODIN

1.0814E-031.7147E-031.8710E-031.5957E-038.8084E-041.6676E-03

1.8277E-02-1.2996E-022.1351E-02

-1.5813E-021.5446E-02

-0.0431E-02

1.4875E-03

-9.8050E-062.3484E-051.5706E-05

-3.4273E-061.3411E-058.4937E-05

-1.2577E-06

1.0652E-021.6708E-021.7026E-02

1.5238E-025.498E-024.0863E-02

1.6947E-021.8005E-021.6409E-021.4348E-024.5594E-02

7.3037E-046.8393E-041.0200E-039.5502E-044.6411E-046.6051E-04

1.6592E-01-1.9172E-02

1.8334E-01-1.9347E-02

1.4819E-01-8.7079E-03

2.7164E-02

-5.7408E-04-9.7942E-04-3.9753E-04-7.4506E-04-7.2628E-04-8.6811E-04

1.6630E-05

3.3O39E-O34.0286E-036.2321E-03

5.5299E-031.1604E-021.7194E-02

8.8995E-038.9312E-038.7059E-039.1403E-033.2451E-02

AS*

2.2379E-058.1587E-054.2268E-053.5353E-054.0298E-058.0176E-05

1.1068E-03-2.6164E-042.2993E-03

-1.6589E-051.0773E-039.4225E-04

3.5110E-04

6.2969E-069.8608E-063.8903E-067.7015E-067.5123E-061.0151E-052.9779E-07

5.9065E-049.8182E-041.0313E-03

4.8169E-041.0883E-031.1910E-03

4.8672E-045.2269E-044.7931E-044.0750E-041.2949E-03

5.8670E-045.2827E-048.9088E-047.9733E-043.5166E-045.0680E-04

1.1822E-031.4154E-041.3042E-036.6180E-051.1568E-032.2022E-04

5.5623E-03

1.8510E-053.6580E-051.1391E-052.2940E-052.4111E-052.8850E-051.2100E-05

1.9332E-032.7858E-033.4687E-03

2.0989E-034.0915E-035.2820E-03

2.6380E-032.6347E-032.6724E-032.6724E-039.4873E-03

AvEn

1.1846E-041.1354E-041.8061E-041.6081E-047.4229E-051.0914E-04

7.8823E-055.3540E-057.9021E-055.9665E-057.5835E-055.0427E-05

1.0705E-03

7.3439E-061.8118E-054.4042E-067.7914E-069.1461E-061.1058E-054.6296E-06

0000E+00OOOOE+000000E+00

0000E+000000E+000000E+00

0000E+000000E+000000E+000000E+000000E+00

-7.5577E-058.9456E-04

-6.1804E-05-6.6685E-05

3.5375E-048.9811E-04

1.4371E-03-3.3652E-04

1.7317E-03-6.0985E-041.1635E-03-8.6142E-05

8.5676E-04

-1.5253E-05-2.2230E-05

1.0040E-05-1.2175E-05-1.2172E-05-1.3395E-05

2.3925E-05

-1.4028E-03-1.6888E-03-1.8813E-03

-3.7996E-03-1.6955E-03-1.3851E-03

-7.1686E-03-9.6484E-04-6.1141E-04-1.2776E-02-2.0651E-02

AH

5.9784E-035.7301E-039.1144E-038.1152E-033.7460E-035.5080E-03

3.9778E-032.7019E-033.9878E-033.0110E-033.8271E-032.5448E-03

5.4022E-02

3.7259E-049.1434E-042.2227E-043.9318E-044.6158E-045.5805E-042.3365E-04

0000E+000000E+00OOOOE+00

0000E+00OOOOE+000000E+00

0000E+000000E+000000E+00OOOOE+000000E+00

- 164 -

KAERI/RR-1999/99

Table 2.3-5

Reactivity Change with Boron Concentration in Moderator

Boron Concentration (ppm)

6

7

8

9

10

Average

Excessive Reactivity (mk)

19.432

11.635

3.869

-3.867

-11.566

Boron Coefficient (mk/ppm)

7.797

7.766

7.736

7.699

7.750*

* 8.310 by PPV/RFSP

- 165 -

KAERI/RR-1999/99

Table 2.3-6

Comparison of ZCU Reactivity Worth

Average Zone Level

(%)

0

10

20

40

55

60

80

100

Reactivity Worth (mk)

PPV/RFSP

0

0.733

1.475

2.962

4.029

4.359

5.524

6.405

RFSP/RFSP

0.784

1.566

3.119

4.229

4.561

5.742

6.608

Difference (%)

-

6.96

6.17

5.30

4.96

4.63

3.95

3.17

- 166 -

KAERI/RR-1999/99

Table 2.3-7

Calibration of Zone Controller

Batch

1

2

3

4

5

6

7

8

9

10

11

Total

Boron

(ppm)

0.058

0.057

0.058

0.058

0.058

0.058

0.058

0.058

0.058

0.058

0.057

0.635

Reactivity

(ink)

0.449

0.442

0.451

0.449

0.449

0.445

0.447

0.449

0.453

0.445

0.444

4.924

Measured

ZCU Level

Change(%)

10.1

8.59

7.18

6.54

6.64

6.15

6.14

5.98

5.96

6.37

6.12

77.87

WIMS/RFSP

ZCU Level

Change(%)

9.560

8.090

7.430

6.980

6.450

6.050

5.900

5.860

6.000

6.000

5.630

73.950

ZCU Worth

(mk/%AVZL)

0.0470

0.0547

0.0607

0.0644

0.0697

0.0736

0.0757

0.0767

0.0755

0.0742

0.0799

- 167 -

KAERI/RR-1999/99

Table 2.3-8

Comparison of Average Zone Level Worth

AVZL(%)

20 ~ 60

20 ~ 80

Measured

0.07166 mk/%AVZL

0.06769 mk/%AVZL

WIMS/RFSP

0.07368 mk/%AVZL

0.06938 mk/%AVZL

Difference(%)

2.818

2.496

- 168 -

KAERI/RR-1999/99

Table 2.3-9

Reactivity Worth of Individual Adjuster Rod

Adjuster

Withdrawn

1

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

Total

Measurement

0.215

0.551

0.696

0.381

0.703

0.553

0.215

0.247

0.674

0.904

0.518

0.911

0.723

0.284

0.216

0.520

0.700

0.370

0.709

0.572

0.219

10.881

WIMS/RFSP

0.199

0.519

0.642

0.353

0.644

0.513

0.189

0.229

0.660

0.850

0.490

0.847

0.657

0.225

0.194

0.514

0.644

0.350

0.644

0.510

0.187

10.06

Difference(%)

-7.41

-5.84

-7.79

-7.32

-8.45

-7.29

-12.06

-7.41

-2.07

-5.93

-5.35

-7.04

-9.07

-20.87

-10.19

-1.06

-7.93

-5.43

-9.19

-10.85

-14.57

-7.54

PPV/RFSP

0.232

0.584

0.726

0.374

0.727

0.577

0.235

0.265

0.726

0.94

0.501

0.943

0.717

0.267

0.232

0.577

0.731

0.374

0.727

0.577

0.232

11.264

Difference(%)

7.95

5.96

4.27

-1.81

3.35

4.28

9.34

7.15

7.73

4.03

-3.22

3.50

-0.76

-6.12

7.41

11.07

4.51

1.06

2.51

0.87

5.99

3.52

- 169 -

KAERI/RR-1999/99

Table 2.3-10

Reactivity Worth of Adjuster Bank

Adjuster Bank

Withdrawn

1

2

3

4

5

6

7

Total

No. rod

1,7,11,15,21

2,6,18

4,16,20

8,9,13,14

3,14

5,17

10,12

Measurement

1.36

1.53

1.51

2.33

1.77

1.79

3.37

13.66

WIMS/RFSP

1.236

1.399

1.387

2.021

1.500

1.524

2.703

11.77

Difference

(%)

-9.12

-8.56

-8.15

-13.26

-15.25

-14.86

-19.79

-13.84

PPV/RFSP

1.38

1.53

1.52

2.27

1.69

1.71

3.02

13.12

Difference

(%)

1.47

0.00

0.66

-2.58

-4.52

-4.47

-10.39

-3.95

- 170 -

KAERI/RR-1999/99

Table 2.3-11

Reactivity Worth of Individual Mechanical Control Absorber

MCA rod

Inserted

1

2

3

4

Total

Measurement

1.885

1.944

1.876

2.009

7.713

WIMS/RFSP

2.070

2.068

2.097

2.092

8.327

Difference(%)

9.84

6.38

11.80

4.12

7.96

PPV/RFSP

2.080

2.065

2.075

2.065

8.285

Difference(%)

10.37

6.22

10.63

2.78

7.41

- 171 -

KAERI/RR-1999/99

Table 2.3-12

Reactivity Worth of Mechanical Control Absorber Bank

MCA Bank

Inserted

1 (MCA#1, #4)

2 (MCA#2, #3)

Total

Measurement

4.85

4.73

9.58

WIMS/RFSP

5.60

5.60

11.20

Difference(%)

15.42

18.39

16.89

PPV/RFSP

5.437

5.436

10.873

Difference(%)

12.10

14.93

13.50

- 172 -

KAERI/RR-1999/99

Table 2.3-13

Reactivity Worth of Individual Shutoff Rod

SOR inserted

1

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

22

23

24

25

26

27

28

Total

Measurement

1.292

1.601

1.598

1.310

0.913

1.891

1.957

0.980

1.313

2.266

2.398

2.321

1.395

1.314

1.421

1.573

2.210

2.363

2.334

1.382

0.906

1.846

1.946

1.008

1.264

1.593

1.630

1.351

45.378

WIMS/RFSP

1.328

1.645

1.64

1.323

1.043

2.206

2.213

1.042

1.533

2.456

2.563

2.448

1.527

1.601

1.596

1.549

2.498

2.612

2.494

1.549

1.095

2.317

2.312

1.1

1.43

1.795

1.788

1.427

50.13

Difference(%)

2.75

2.78

2.60

0.96

14.24

16.63

13.10

6.41

16.76

8.39

6.88

5.46

9.43

21.81

12.29

-1.52

13.03

10.52

6.88

12.05

20.89

25.48

18.82

9.09

13.14

12.71

9.64

5.59

10.47

PPV/RPSP

1.288

1.635

1.634

1.284

0.964

2.173

2.157

0.954

1.495

2.497

2.567

2.483

1.479

1.476

1.463

1.497

2.493

2.567

2.484

1.481

0.957

2.161

2.161

0.952

1.29

1.635

1.632

1.283

48.142

Difference(%)

-0.34

2.15

2.23

-2.02

5.59

14.89

10.24

-2.57

13.87

10.20

7.05

6.97

5.99

12.30

2.94

-4.82

12.80

8.62

6.45

7.13

5.66

17.04

11.06

-5.59

2.06

2.66

0.074

-5.06

6.09

- 173 -

KAERI/RR-1999/99

Table 2.3-14

Reactivity Change due to Heat Transport System Temperature

Coolant Temperature

(°C)

35.25

50.06

64.87

79.92

96.85

110.84

125.53

140.82

155.26

174.14

187.21

199.78

215.04

230.07

245.09

259.84

Measured (mk)

-

-0.890

-0.738

-0.733

-0.738

-0.674

-0.725

-0.597

-0.537

-0.576

-0.487

-0.339

-0.408

-0.307

-0.254

-0.200

WIMS (mk)

0

-0.809

-0.761

-0.739

-0.818

-0.624

-0.627

-0.623

-0.537

-0.636

-0.448

-0.393

-0.433

-0.358

-0.289

-0.219

Difference (%)

-

-9.09

3.12

0.82

10.84

-7.41

-13.52

4.38

-0.02

10.42

-8.09

15.89

6.25

16.45

13.71

9.41

- 174 -

KAERI/RR-1999/99

Table 2.3-15

Reactivity Change due to Moderator Temperature

Moderator Temperature

CO)

69

65

60.06

54.45

50.25

45.52

40.42

34.99

Measured (mk)

-

-0.104

-0.172

-0.188

-0.158

-0.189

-0.225

-0.274

WIMS/RFSP (mk)

-

-0.057

-0.081

-0.097

-0.085

-0.105

-0.117

-0.146

Difference (%)

-

-45.02

-52.93

-48.38

-46.09

-44.55

-48.08

-46.64

- 175 -

KAERI/RR-1999/99

Table 2.3-16

Comparison of Critical Core Performance Parameters

keff

Maximum channel power (kW)

Maximum bundle power (kW)

Radial peaking factor

Discharge burnup (MWhr/Bundle)

Inner Core

Outer Core

Whole Core

Reference

(PPV/RFSP)

1.00000

6702

832

0.809

3623

3128

3301

WIMS/RFSP

1.00000

6734

827

0.806

3536

3189

3314

Difference(%)

0.0

0.5

-0.6

-0.4

-2.4

2.0

0.4

- 176 -

KAERI/RR-1999/99

Table 2.3-17

Comparison of Fixed Burnup Core Performance Parameters

keff



Radial peaking factor

Discharge burnup (MWhr/bundle)

Inner core

Outer core

Whole core

Reference

(PPV/RFSP)

1.00000

6702

832

0.809

3623

3128

3301

WIMS/RFSP

0.99997

6747

830

0.804

3622

3129

3298

Difference(%)

-0.003

0.7

-0.2

-0.6

-

-

-

- 177 -

KAERI/RR-1999/99

Table 2.3-18

Comparison of Zone Controller Unit Worth

zculevel

0102030405060708090

100

PPV/RFSP

keff

1.00371

1.00295

1.00221

1.00147

1.00073

1.00000

0.99930

0.99867

0.998070.997550.99710

Excess reactivity(mk)

-3.696-2.941-2.205-1.468-0.7290.0000.6961.3351.9312.4532.909

WIMS/RFSP

keff

1.003871.003081.00231.001531.000751.000020.999280.998600.997980.997440.99698

Excess reactivity(mk)

-3.855-3.071-2.295-1.528-0.749-0.0200.7181.4022.0212.5673.033

Difference

(%)

4.34.44.14.12.7-

3.05.04.74.64.3

- 178 -

KAERI/RR-1999/99

Table 2.3-19

Comparison of Adjuster Rod Worth

123456789101112131415161718192021All

SUM

PPV/RFSP

keff

1.000241.000781.001091.000661.001081.000771.000241.000281.000861.001181.000741.001181.000861.000281.000241.000771.001081.000661.001071.000761.000241.01705

Rod worth(ink)0.2400.7791.0890.6601.0790.7690.2400.2800.8591.1790.7391.1790.8590.2800.2400.7691.0790.6601.0690.7590.24016.76415.047

WIMS/RFSP

keff

1.000211.000741.001081.00061.001071.000741.000211.000241.000831.001191.000691.001181.000811.000251.000211.000741.001081.000601.001091.000731.000211.01666

Rod worth(mk)0.2100.7391.0790.6001.0690.7390.2100.2400.8291.1890.6901.1790.8090.2500.2100.7391.0790.6001.0890.7290.21016.38714.488

Difference

(%)

-12.50-5.12-0.92-9.09-0.92-3.89-12.50-14.28-3.490.85-6.750.00-5.81-10.71-12.50-3.890.00-9.091.87-3.94-12.50-2.25-3.72

179 -

KAERI/RR-1999/99

Table 2.3-20

Comparison of Mechanical Control Absorber Worth

1

2

3

4

All

SUM

PPV/RFSP

keff

0.99750

0.99754

0.99750

0.99754

0.98809

Rod worth (mk)

2.503

2.466

2.504

2.470

12.059

9.944

WIMS/RFSP

keff

0.99752

0.99757

0.99753

0.99757

0.98802

Rod worth (mk)

2.485

2.441

2.478

2.433

12.126

9.837

Difference

(%)

-0.72

-1.02

-1.04

-1.51

0.56

-1.07

- 180 -

KAERI/RR-1999/99

Table 2.3-21

Comparison of Shutoff Rod Worth

.12345678910111213141516171819202122232425262728AllSUM

PPV/RFSP

keff

0.9987500.9984580.9984660.9987720.9989650.9976030.9976180.9989870.9983590.9969300.9966790.9969590.9983950.9982980.9983320.9983630.9969220.9966760.9969550.9984000.9989740.9976060.9976200.9989880.9987570.9984650.9984680.9987730.929665

Rod worth (ink)

1.2521.5441.5361.2301.0362.4032.3881.0141.6443.0793.3323.0501.6081.7051.6711.6403.0883.3353.0541.6031.0272.4002.3861.0131.2451.5371.5341.22975.65653.551

WIMS/RFSP

keff

0.9987590.9984800.9984830.9987750.9989480.9976540.9976850.9989760.9983450.9970260.9968010.9970560.9983900.9982750.9983200.9983500.9970260.9968160.9970590.9983910.9989550.9976560.9976850.9989750.9987550.9984800.9984870.9987750.931262

Rod worth

(mk)1.2431.5221.5191.2271.0532.3522.3201.0251.6582.9833.2092.9531.6131.7281.6831.6532.9833.1942.9501.6121.0462.3502.3201.0261.2471.5221.5151.22773.81252.730

Difference (%)

-0.72-1.43-1.11-0.241.64-2.13-2.821.090.85-3.14-3.69-3.200.311.350.720.80-3.39-4.23-3.430.561.85-2.09-2.741.290.16-0.98-1.24-0.16-2.44-1.59

- 181 -

KAERI/RR-1999/99

Table 2.3-22

Comparison of 600-FPD Refueling Simulation

Peak channel power (kW)

Peak bundle power (kW)

Channel power peaking factor

Zone controller level

Inner core discharge burnup (MWhr/kgU)

Outer core discharge burnup (MWhr/kgU)

Inner core refueling Rate (channels/FPD)

Outer core refueling Rate (channels/FPD)

Whole core refueling Rate (channels/FPH

Reference(PPV/RFSP)

6849

855

1.055

0.50

188

161

0.69

1.30

1.99

WIMS/RFSP

6853

852

1.063

0.50

182

163

0.71

1.28

1.99

Difference(%)

0.06

-0.35

0.76

0.0

-3.19

1.23

2.90

-1.54

0.0

- 182 -

KAERI/RR-1999/99

24.765cm,

/ / /

>// /i

14.2875cm

r /

/

S/\ / /// / /

// /

/ // /

/ // /

71

/ // // // //// /

W

///

V

//

/

//

f

////

/£•••

J*

/

/

/

V

/

/

/

V

//.—V

/

/

)

/

V

Fig. 2.3-1 SHETAN Model for Fuel Channel

- 183 -

KAERI/RR-1999/99

24.765c

14.2875cm

/ / / /

/ 7 / /

/_ /_

7

/ y

/ ..-••

Liir

///M

/

V

/

/

/

/

Fig. 2.3-2 SHETAN Model for Reactivity Device

- 184 -

KAERI/RR-1999/99

1 2

GT

3

TS

4

CN

5

CR

6

BL

7

Source

Vacuum

Calandria

SHETAN-modelled Region

No.

1

2

3

4

5

6

7

Coordinate (cm)

14.2875

28.595

51.459

54.015

60.295

79.375

82.2325

Material

FuelD2OGuide tubeD2OTension springCoupling rodGuide tubeCoupling rodCoupling nutGuide tubeD2OCoupling rodD2OCoupling rodLocatorBracketD2O

Stainless steel

Fig. 2.3-3 WIMS-AECL Slab Model for Structural Materials

- 185 -

KAERI/RR-1999/99

A

B

C

D

F.

F

G

H

T

K

L

M

N

0

P

Q

R

S

T

U

V

W

...

1 2 3 4

|

5 e

1/8

2/9

...

7

;

9

444-i

:

10 i i 12 13

_ m I.. Li_LLJ_|

3/10

4/11

5/12

14 15 16

TTTTTTT

17 18

_ ! _

7/14

19

i

20 21

! i

22

-•

l

9

10

12

13

14

15

17

18

19

20

22

23

24

25

27

28

29

30

31

32

33

34

35

1 2 3 4 5 8 9 11 13 15 17 19 21

10 12 14 16 18 20

2 3 2 5 2 7 2 9 31 3 3 3 5 37

2 2 2 4 2 6 2 8 3 0 3 2 3 J 3 6

39 40 41 42 *

Fig. 2.3-4 Typical RFSP Nodal Model for XY Plane

- 186 -

KAERI/RR-1999/99

-*1 2 3 4

LZC

1/2

ADJ

1

ADJ

8

ADJ

IS

Lzca9

7 S

ADJ

2

ADJ

9

ADJ

16

c 10

ADJ

3

ADJ

10

ADJ

17

11 12

3/4/S

ALIJ

4

ADJ

11

ADJ

IS

LZC1

0/11

m

3

AD.

5

AD

12

AD;

11

4 IS 16

ADJ

6

ADJ13

ADJ

20

17 IS

LZC

57

ADJ

7

ADJ

14

ADJ

?!

LZC1

VI4

19 20 21 22

" 1

2

3

5

6

7

8

9

10

11

12

« 24

44

1 2 3 4 5 6 7 9 10 12 14 16 IS 20 22 24 26 28 30 32 34 36 37 38

11 13 15 17 19 21 23 2L 2 29 31 33 35

Fig. 2.3-5 Typical RFSP Nodal Model for XZ Plane

- 187 -

KAERI/RR-1999/99

-WIMS/RFSP

Measurement

4 5 6 7Boron Batch

Figure 2.3-6 Calibration of Zone Controller

- 188 -

KAERI7RR-1999/99

I1/1

5a

2 -

- 2 -

- 4 -

- 6 -

- 8 -

-10

- WXMS/RFSP

M«asurment

50 100 150 200

Temperature <°C)

2 5 0 300

Figure 2.3-7 Heat Transport System Temperature Effect

- 189 -

KAERI/RR-1999/99

0.5-

£ 0 0

"•§ -0.5 -

3a« - i o -u

a-1.5-

-2.0-

WIMSflRFSP

• Measurment

••

30 35 40 45 50 55 60 65Temperature (°C)

70

Figure 2.3-8 Moderator Temperature Effect

- 190 -

KAERI/RR-1999/99

1.0-

o».

O.4.111 '

MS 'A

\

\

500 600 TOO 000 BOO 1000 1100 I20O

l P«itk>n<cm>

f'j

If'/ • '

JrS.hf \

• \

.1r rrrTrr,500 600 700 600

Hurizonlal Po»hion(™>

CASE 1 CASE 2

I ..:

CASE 3 CASE 4

Horfaunt.l PoiiUon(cm)

CASE 5

Figure 2.3-9 Horizontal Flux Scan

- 191 -

KAERI/RR-1999/99

>TOO SOO 900 1000 1100 I20O 1300

V r t k a l Po*Hk>n<cm>

CASE 1 CASE 2

0 6 .

O.2

/

WMSitnr

900 1000 1100 1200 1300 1400

V*rtle»l potttfa>n<cm)

1.0-

as-

0.0-

o.a /

* \

\

wmurtr V7DD tOO SCO 1100 1300 1300

Vert ical PMhfe*Kcm>

CASE 3 CASE 4

TOO aOO 900 1000 JIOO I ICO 1300 \AC

Vrr tkat Po*itlun(cm>

CASE 5

Figure 2.3-10 Vertical Flux Scan

- 192 -

KAERI/RR-1999/99

8 9 10 11 12 13 14 15 16 17 18 19 20 21 22

A

B

r.D

F.

F

G

H

J

K

1,

M

N

O

P

0R

S

T

U

V

W

3.4

2.7

2.9 2.2

2.7 1.6

2.4 1.3

2.4 1.2

2.5

2.3

1.2

14

1.4

16

MaximumAverage ]

3.7

3.4

2.8

1.8

1.2

.7

.3

.3

.3

.4

.9

.8

.4

3.6

3 ?

2.5

1.7

.9

.4

1

0

-.3

-5

-6

- 4

-.3

-.5

-1.1

3 7

3.2

7 4

1.7

.9

.2

.0

6

5

.1

-8

-1 3

-14

-15

-1.8

-1.7

-7 3

4.93%.54%

3.9

3 4

2.4

18

1.4

.8

.2

.5

9

10

9

.5

-4

-14

-15

-19

-2.4

-2.3

-7.3

-3 3

3.4

7 5

1.7

14

1.5

1.7

1.1

.6

4

7

0

-.2

-4

-5

- 6

-16

-2.5

-2.8

-7..9

-3?

2.9

7 0

1.3

1 1

1.4

1.8

1.2

.7

3

0

-.4

-5

-4

- 6

-16

-Z7

-3 1

-3 5

3.1

2.6

.9

8

1.5

1.5

1.0

.5

1

- 4

-.6

-7

-8

- 9

- . 5

-2.9

-3 4

-3 6

-3 7

-4.9

3.3

2.2

1 1

.5

1.3

1.4

.9

.4

-1

- 5

- 7

-.9

-9

-9

-10

-17

-3.2

-3 8

-40

-40

-4.3

3.3

2.1

8

4

1.8

2.2

1.6

.6

-8

-1 1

-1.1

-7

-3

-13

-3.4

-41

-44

-4?

-4.4

3.2

2.1

8

.2

4

1.8

2.2

1.6

.6

- 8

-1 1

-1.1

-7

-3

-13

-3.5

-41

-44

-4?

-4.4

3.2

2.1

1 1

.5

4

1.3

1.3

.9

.4

-1

- 5

- 8

-.9

-9

-1 0

-1 1

-17

-3.2

-3 8

-41

-4 1

-4.4

3.0

2.5

1 5

.8

7

1.4

1.4

.9

.5

0

- 3

- 5

-.7

-8

-9

-10

-16

-3.0

-3 5

-3 7

-3 7

-4.9

2.8

19

1.2

10

1.3

1.7

1.1

.5

1

-1

-4

-.5

-6

-6

- 7

-17

-2.8

-3?

-3 4

-3 5

3.3

7 4

1.5

17

1.4

1.6

1.0

.5

0

-•>

-.3

-5

-6

- 7

-17

-2.7

-3 0

-3 0

-3 3

3.7

3 3

2.3

16

1.2

.6

.0

.3

7

8

7

.3

-6

-1 6

-17

-2.0

-2.5

-7 5

-7 5

-3 5

3 5

3.0

7 ?

1.5

.7

.0

-.2

4

-.2

-10

-1 6

- 1 6

-1.8

-2.1

-19

-7 4

3.3

7 9

2.2

1.4

.6

.1

- 3

- 5

-.6

-8

-8

- 6

-.6

-.7

-13

3 4

3.1

2.5

1.5

.8

4

n-.1

0

i

6

.5

.2

3.0

2.3

1.8 2.4

1.2 2.3

.9 2 0

8 1 9

.8

1.0

1 1

1.3

2.1

19

Fig. 2.3-11 Comparison of Channel Power for Equilibrium Natural Uranium Core (Critical Core)

- 193 -

KAERI/RR-1999/99

7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22

A

R

C

D

E

F

G

H

J

K

L

M

N

0

P

Q

R

S

T

U

V

w

5.0

48

4.5

44

45

4?

5.5

4.7

4.1

35

3.1

30

30

31

3.2

3.3

5.9

55

4.7

3.6

2.9

?3

2.0

1 8

1 7

17

1.9

2.4

2.4

2.0

5.7

5.2

44

3.4

2.3

1.7

1 ?

1.0

8

6

5

.5

.8

1.0

.9

.4

5.9

5.2

4.2

33

2.1

1.1

.7

.9

7

3

-5

-.8

-.6

-6

-.7

-.4

L?

61

5.5

4.3

3.3

76

1.5

.4

.0

1

.1

-1

-4

-1 1

-1.5

-1.4

-1.4

-1.5

-1.3

-1.1

1-2.0

57

4.5

3.4

2.6

7?

1.5

.4

-.6

-1 1

-1.5

-1 7

-18

-18

-1.6

-1.3

-1.5

-2.0

-2.0

-1.8

-2.0

51

3.9

2.8

2.1

17

1.1

-.1

-1.1

-19

-2.3

-7 6

-?fi

-7.5

-2.2

-1.8

-2.0

-2.4

-2.4

-2.3

-2.3

5.3

47

3.3

2.3

1.5

1?

.3

-.8

-1.7

-7 4

-2.9

-3?

-3 3

-3?

-2.9

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-2.8

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4?

2.8

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-2.1

-7 9

-3.5

-3 8

-3 8

-3 7

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-3.4

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5.5

47

2.5

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11

.6

-.6

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-3?

-3.9

-43

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-3 7

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-3.8

-3.7

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5.5

4?

2.5

1.4

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10

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-.7

-2.1

-3?

-4.0

-43

-4?

-3 7

-2.9

-2.4

-2.7

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-3.8

-3.7

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5.5

4?

2.8

1.7

1.0

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-.1

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-3.6

-3 9

-3 9

-3 7

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-3.0

-3.0

-3.5

-3.5

-3.3

-3.0

-3.2

5.3

46

3.3

2.2

1.4

1 1

.2

-.9

-1.8

-?6

-3.0

-3 4

-3 4

-3 3

-3.0

-2.7

-2.6

-3.0

-3.0

-2.9

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-3.8

50

3.8

2.7

2.0

16

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-7 8

-7 7

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55

4.4

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2.5

71

1.3

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-.8

-1 4

-1.7

-1 9

-7 0

-7 0

-1.8

-1.5

-1.7

-2.2

-2.1

-1.9

-2.1

60

5.3

4.1

3.1

2.4

1.3

.2

-.2

-4

-.1

-3

-7

-14

-1.8

-1.6

-1.6

-1.7

-1.4

-1.3

-2.2

5.7

5.0

3.9

3.0

1.8

.8

.4

6

.6

4

0

-7

-1.1

-.9

-.9

-1.0

-.6

-1.1

5.5

4.9

4.1

3.1

2.0

1.3

9

.6

4

3

7

.2

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.7

.6

.1

5.5

5.1

4.3

3.2

2.5

1 9

1.6

1 4

1 3

13

1.5

2.0

2.0

1.7

5.0

4.3

3.7

3 1

2.7

?S

">5

2.6

2.7

2.9

4.4

43

4.0

3R

39

3.7

Maximum 6.10%Average 2.45%

Fig. 2.3-12 Comparison of Channel Power for Equilibrium Natural Uranium Core (Fixed Burnup)

- 194 -

KAERI/RR-1999/99

A

B

C

u

b

r

\3

14ri

j

V

1

122700.562814.07356

3560.0428A AA4280.04890.41488

-0.205250.405280.575450.39544

-0.185610.385630.365850.37580

-0.855940.37592

-0.346050.37601

-0.666050.386050.0

132660.552743.01343A AQ3481.46420A AAv.*r*r

4210.244830.41481

-0.415270.39522

-0.955390.38538

-0.195580.385590.185750.375750.05920.36588

-0.686000.36596

-0.676030.37601

-0.33

142590.562704.25328A CfiKJ.jyJ

3341.83405n A S4050.04620.424660.875100.405100.05280.395290.195510.38550

-0.185690.37566

-0.535830.37579

-0.695870.375880.175920.375920.0

15

310A ciU.Jl

3172.26381A A£.

3820.264440.424440.04950.40491

-0.815340.39525

-1.695410.38537

-0.745600.37554

-1.075700.37567

-0.535800.37575

-0.865780.385800.35

16

2750.542843.27344n Aftu.^o3491.454140.44413

-0.244720.41463

-1.705140.40502

-1.185250.39518

-1.335400.38537

-0.565580.38550

-1.435660.37559

-1.245650.38564

-0.18

17

2350.592527.23302A CO3102.653710.473730.544300.43426

-0.934780.41470

-1.675080.40502

-1.185350.39525

-1.875360.38529

-1.315420.38539

-0.555520.38544

-1.45

18

261U.JO

2755.363150.503273.813790.463800.264260.434270.234650.41464

-0.225000.40491

-1.805160.39510

-1.165310.39522

-1.695350.39528

-1.31

19

2710.552834.433150.503273.813700.463741.084110.444130.494430.434440.234720.41466

-1.274880.41481

-1.434950.41488

-1.41

20MCNP%STDRFSP%DIFF

2610.562765.753090.513182.913460.483521.733870.46382

-1.294080.44406

-0.494250.43421

-0.944330.44429

-0.92

21

DIFF(%)=

22

(R-MVMX100

T

InnerCore

2810.542986.053140.513181.273320.493350.903490.483490.03570.48356

-0.28

2580.522704.652680.552742.242770.562811.44

Fig. 2.3-13 Comparison of Bundle Power Distribution for Equilibrium DUPIC Core

- 195 -

KAERI/RR-1999/99

2.4

DUPIC

MCNP DUPIC

0.73%

5%

10"6(5k/K),

2STD

A U (1.75x

MCNP5]

irfl, WIMS-AECLofl MCNP

WIMS/RFSP

RMS

RMS

0.3%5k

- 1 2 %

, PPV/RFSP

}, WIMS/RFSP

71

DUPIC ^

MCNP S«i^& CANDU

ZL ^ 3 f i - WIMS/RFSP

4 0.12% 5 k

] ^ . MCNP

MCNPif WIMS/RFSP

WIMS/RFSP 7j-i>ol

DUPIC

MCNP

fe DUPIC

DUPIC

- 196 -

KAERI/RR-1999/99

1. H. KEIL, P.G. BOCZAR, and H.S. PARK, "Options for the Direct Use of Spent PWR Fuel

in CANDU (DUPIC)", Proceedings of Third International Conference of CANDU Fuel, Chalk

River, Canada, 1992.

2. J.S. LEE et al., "Reaserch and Development Program of KAERI for DUPIC (Direct Use

of Spent PWR Fuel in CANDU Reactors)", Proceedings of International Conference and

Technology Exhibition on Future Nuclear System: Emerging Fuel Cycles and Waste Disposal

Options, GLOBAL'93, Seattle, USA, 1993.

3. H.B. CHOI, B.W. RHEE, and H.S. PARK, "Physics Study on Direct Use of Spent PWR

Fuel in CANDU (DUPIC)", Nucl. Sci. Eng.: 126, pp.80-93, May 1997.

4. E.S.Y. TIN and P.C. LOKEN, "POWDERPUFS-V Physics Manual", TDAI-31 Part 1, Atomic


5. A.R. DASTUR et al., "MULTICELL User's Manual", TDAI-208, Atomic Energy of Canada

Limited, 1979.

6. D.A. JENKINS and B. ROUBEN, "Reactor Fuelling Simulation Program - RFSP: User's

Manual for Microcomputer Version", TTR-321, Atomic Energy of Canada Limited, 1993.

7. J.V. DONNELLY, "WIMS-CRNL: A User's Manual for the Chalk River Version of WIMS",

AECL-8955, Atomic Energy of Canada Limited, 1986.

8. H. CHOW and M.H.M. ROSHD, "SHETAN - A Three-Dimensional Integral Transport Code

for Reactor Analysis", AECL-6787, Atomic Energy of Canada Limited , 1980.

9. J.F. BRIESMEISTER, ed., "MCNP- A General Monte Carlo N-Particle Transport Code,

Version 4B," LA-12625-M, Los Alamos National Laboratory, 1997.

- 197 -

KAERI/RR-1999/99

10. M.A. SHAD, "RFSP Reactor Physics Model for Point Lepreau Generating Station Unit 1",

TTR-386 Rev.l, Atomic Energy of Canada Limited, 1994.

11. B. ROUBEN and A.R. DASTUR, "CERBERUS User's Manual (Revision 1)", TDAI-177,

Atomic Energy of Canada Limited, 1986.

12. CM. BAILEY, P. AKHTAR, P.E. TREMBLAY, O.A TROJAN and J.H. CARSWELL,

Jr., "The SORGHUM Code: Program Description and User's Guide", TDAI-133, Atomic


13. B. ROUBEN et al., "CHEBXEMAX User's Manual", TDAI-187, Atomic Energy of Canada

Limited, 1980.

14. M.H.M. ROSHD and H.C. CHOW, "The Analysis of Flux Peaking at Nuclear Fuel Bundle

Ends Using PEAKAN", AECL-6174, Atomic Energy of Canada Limited, 1978.

15. D.S. CRAIG, "Testing ENDF/B-V Data for Thermal Reactors", AECL-7690, Atomic Energy

of Canada Limited, 1984.

16. J.V. DONNELLY, "Progress in the Development of WIMS at Chalk River", AECL-8807,


17. J.V. DONNELLY, "Validation of WIIMS with ENDF/B-V Data for Pin-cell Lattices",


18. J.V. DONNELLY, "Description of the Resonance Treatment in WIMS-AECL", AECL-10550,


19. J.V. DONNELLY, "Acceptance Tests for a New WIMS-AECL ENDF/B-V Nuclear Data

Library", FFC-RCP-010 Rev.0, Atomic Energy of Canada Limited, 1997.

20. R.D. MOSTELLAR and L.D. EISENHART, "Benchmark Calculation for the Doppler

- 198 -

KAERI/RR-1999/99

Coefficient of Reactivity," Nucl Sci. Eng., 107, 265 (1991)

21. F. RAHNEMA et al., "Boiling Water Reactor Benchmark Calculations", Nuclear Technology

184, 117, 1997.

22. "Cross-Section Evaluation Working Group Benchmark Specifications," BNL-19302

(ENDF-202), Brookhaven National Laboratory, 1974.

23. L.M. PETRIE and N.F. LANDERS, "KENO V.a An Improved Monte Carlo Criticality

Program with Supergrouping," Section Fl l , NUREC/CR-0200, Vol. 2, U.S. Nuclear

Regulatory Commission, 1984.

24. G.H. ROH and H.B. CHOI, "Assessment of Neutron Transport Codes for Application to

CANDU Fuel Lattice Analysis", KAERI/TR-1377/99, Korea Atomic Energy Research

Institute, 1999.

25. G.H. ROH, H.B. CHOI, and J.W. PARK, "Sensitivity Analysis on Various Parameters for

Lattice Analysis of DUPIC Fuel with WIMS-AECL Code", Proceedings of the Korean Nuclear

Society Autumn Meeting, Taegu, Korea, Oct. 1997.

26. M.S. MILGRAM, "A Comparison of the Reactor Physics Predictions for CANFLEX Fuel

Using the Codes MCNP, WIMS-AECL and LATRAP", Topical Meetings on Advances in

Reactor Physics, Charleston, 1992.

27. D.A. JENKINS, "Comparison of RFSP with HBAL and with Phase rB' Experiments at Point

Lepreau", Appendix D of Addendum to TDAI-440 Part I, Atomic Energy of Canada Limited,

1991.

28. M.A. SHAD and A.C. MAO, "Comparison of History-Based Core Tracking and Powermap

Simulations with Conventional Production Runs for Point Lepreau", TTR-486, Atomic Energy


- 199 -

KAERI/RR-1999/99

29. H.C. CHOW, "Post-Simulation of 1992 Point Lepreau Restart Physics Tests", TTR-480,


30. D.A. JENKINS, "Simulation of 1992 SDS1 Trip Test at Point Lepreau", TTR-532, Atomic


31. H.C. CHOW and D.A. JENKINS, "RFSP Simulation of Point Lepreau Derating Events",

TTR-544, Atomic Energy of Canada Limited, 1994.

32. E.V. CARRUTHERS and J.V. DONNELLY, "Validation of WIMS-AECL against Actual

Operating History: Core Following of Point Lepreau Nuclear Generating Station with

POWDERPUFS-V and WIMS-AECL", 5th International Conference on Simulation Methods

in Nuclear Engineering, Montreal, Canada, 1996.

33. J. GRIFFITHS, "WIMCORE: An Interface between WIMS-AECL and the codes FMDP,

RFSP, HQSIMEX, SHETAN and GETRANS", Internal Memorandum RC-906, Chalk River

Laboratories, 1993.

34. D. JENKINS et al., "AMAD for Physics Simulations", Addendum to TDAI-440 Part I, Atomic


35. H. CHOW and C. NEWMAN, "Post-Simulation of the Point Lepreau 1992 Startup Physics

Tests", 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal,

1994.

36. H.B. CHOI, "A fast-running fuel management program for a CANDU reactor", Annals of

Nuclear Energy, Vol.27, pp.1-10, 1999.

- 200 -

KAERI/RR-1999/99

. DUPIC

- 201 -

KAERI7RR-1999/99

3. DUPIC m°*^ ^

^ | fl DUPIC

CANDU

. U DUPIC

DUPIC

CANDU

^ DUPIC «|«iJ5. ^ 7 1 ^ ^-^^r, DUPIC

4 , CANDU $J*f.3AiH && ^ $Xr}. o ] # ^«B CANDU

DUPIC ^ ^ ^ ^ o ^ j g « .^^- ^ * J ^ ^ o . ^ 5 o | ^ 7]^ . CANDU

DUPIC

DUPIC ^ ^ S 5 ] Efig-^^- i i l -e l ^7)| ^ ^ J | - 7> g- 2f^o] 61-5-^, °lfe DUPIC

CANDU Qx}M.$\ ^ - ^ # ^ ^ * W ^ ^

] 4 ^ . DUPIC

fg 7]^. DUPIC

l.45wt%olt:|-.3

CANDU ^

fe DUPIC ^ ^ S i f 7 ] ^ CANDU 6

il 3 .2^<H1A^ CANDU

7]^- DUPIC ^ ^ S S ^ ^ ^ 5| ^^f - ^ - i - 3.3^ ^ 3.4^<^1 4 4 7l

. DUPIC ^

- 203 -

KAERI/RR-1999/99

3.1

CANDU

CANDU

^ 27H1-

CANDU

CANDU

3.1.1

CANDU

51

- 204 -

KAERI/RR-1999/99

4^0.14

5% - 95%

3.1.2

4 4 800 kW 91 6500 4 4 935 kW

7300 kWojt:]-.

CANDU 9S4

fonn factor7|-

9J 9J shim

3.1.3

n]

91

- 205 -

KAERI/RR-1999/99

16.6

.^- 14.9

3.1.4

4 ^^1 ^ 1 ^ ^ #*}MM %*}M*i 5X3- *}*Anm ^ m ^r $ls-

>. CANDU

7 1 1 ^ 3 ^ ^ - o U i , ^ 2 3 ^

o|-g-«>c:f. 4 3*1 Tfl-fvg: ij4i ^§-7f^ 7)-^ ^^(Minimum Acceptable

Performance Specifications; MAPS)*H^

3.1.5 On-line i f ^ * H r Mapping

mapping 7«^&

of 7fl ofl Af§-^nf. On-line #3*Hf-^3E ^>Sfe 3 ^ 5l ^ 1 ^ S. S^ofl cfl

*1I 4^-€^>. -£3*Hr mapping

3.1.6 ^-Jf- 3 ^ # ^ J i ^ Tfl-f-

7 1-f- (Regional Overpower Protection; ROP)^- in^fl

(SDS1 : 10.8%/s, SDS2: 17.3%/s)o]r:>.

- 206 -

KAERI/RR-1999/99

3.2

CANDU POWDERPUFS-V(PPV) [Ref. 5]

^ RFSP [Ref. 7]

fe DUPIC

^ RFSP

MULTICELL [Ref. 6]

. DUPIC

31-SrM WIMS-AECL [Ref. 8] 3.

SHETAN [Ref. 9] 3 £ <

c>. DUPIC

3.2.1 DUPIC

3.2-1^ ^ o ] DUPIC

ENDF/B-V

437|

K WIMS-AECL 89-$

3.2.1.1

WIMS-AECL

. RFSP S ^

RFSP

(3.2-1)

- 207 -

KAERI/RR-1999/99

3.2.1.2

CANDU ^l

S. 719-71-S^<a SHETAN 3 H S Tll^^c}. <§•£ ^^^^Sr ^ ^ - ^ ^©1 #

719-7}

^r ZL^ 3.2-2 iJ 3.2-3ofl

3.2.1.3

RFSP T i l t H ^ 135Xe

61

(3.2-3)

(3.2-4)

= £(r)(x(r)-xref) (3.2-5)

- 208 -

KAERI/RR-1999/99

. RFSP S H

, 7 l f e 135Te

WIMS-AECL 3 .H

3.2.2 DUPIC

3.2.2.1

CANDU 4 l > f e fla ^ ^

i i ^ ^ ^^^"Ell 71 -^ - £71 $J*H RFSP S ^

J t

r 26co

o>2fe- 4 4

CANDU 6 ^ 4

3.2-4^} ^-o) DUPIC

(3.2-6)

o*

(3.2-7)

. SDS-21- ^ l ^ t > tiJ:-§-£ 7)^- ^|*1 J g ^ S . # H ^ 3.2-5<Hl

- 209 -

KAERI/RR-1999/99

3.2.2.2

H,

(Maximum Channel Power; MCP>§- o j [ ^ ^ afl,

( P m a x )

RFSP

o>(i,j.k) = ^ ( k ) + f(i,j) x (a,2(k) - (3.2-8)

91

4 * B ^91

si7]

- 210 -

KAERI/RR-1999/99

to,

3.2.2.3

± RFSP

: 4

}Hi o|

CANDU |-.10 o| a

- 211 -

KAERI/RR-1999/99

Moderator

Fuel RodCoolantPressure TubeAir GapCalandria Tube

Fig. 3.2-1 DUPIC Fuel Lattice Model

- 212 -

KAERI/RR-1999/99

y_24.765CI

* / / / / /

V / \/ / /

Fig. 3.2-2 SHETAN Model for Fuel Channel

- 213 -

KAERI/RR-1999/99

*A / / / /// y / / /

24.765v / / / /

V / A / /</ / /

V / /

14.2875cm

/ / ^

LWCic.

Fig. 3.2-3 SHETAN Model for Reactivity Device

- 214 -

KAERI/RR-1999/99

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22

A

B

C

D

E

F

G

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J

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L

M

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P

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Fig. 3.2-4 Front View of CANDU-6 Core

- 215 -

KAERI/RR-1999/99

>-» x-direction

Back Side

••

o•••••

•

•

•

•

•

•

•

•

o••••D

a

a

••

a

a

•

oa•a•a

•••

o o• •

o

Front Side

• Adjuster Rod (21) • Shutoff Rods (28)

O Zone Controller Units (14) • Mechanical Control Absorber (4)

Fig. 3.2-5 Plan View of Reactivity Device Layout

- 216 -

KAERI/RR-1999/99

3.3 7)^ DUPIC m& %

DUPIC

ofl, DUPIC t*}<£3.^ J2.-5.

o] ^Sf^>u>. o]e|^> ^ ^ ^ a l ^ ^ A ^ D U P I C

CANDU

DUPIC

- DUPIC

s-a-- DUPIC «|«iS. ^ 7 l H l # ^ ^ S Hl^^: ^ ^ S ^71H1J2X1- ^711

3.5wt% ^ 1 ^ ^ -f-B}w (Slightly enriched uranium; SEU), 0.25wt%

SEU 9J DU

SEU 91

36007fl*l

2/tf-i- ^^* f7 l ^ * H CASMO3 [Ref. 11]

91 0RIGEN2 [Ref. 12] S ^ # 4-§-*}^3., DUPIC ^ 9 l S 5 l ^ ^ - ^ WIMS-

AECL S J ^ #

- 217 -

KAERI/RR-1999/99

H f e iB % |-fe 239Pu239Pu 2.$ ^^m £\&Sk*W-c\.

SEU tfl DU£f a l l - a i ^ ^ ^ S . ^ ^-^«> 235U

3.3.1.1 7 l ^ DUPIC

DUPIC ^ « 1 S . ^ S ^ ^ : 1.0wt% 235U ^ 0.45wt%

:^*H-H *J 90%^ A^g-^ ^ ^ S . ^ ^ ^ . 7 } 7]

DUPIC ^ « ^ S 5 | ^ ^ # W * } ^ * I DUPIC 3 ) ^ 1 5 . ^fl7}^-l ^ sac>. o|n||,SEU ^J DU^ 6O> . 4 4 ^^| | A^J . « .^o | 7>8 ni

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, CANDU

) 4 ] DUPIC

(5.46 mills/kWh)14

3.3.1.2

can]..

2 3 9Pu

~7 el 3 3_1— ZF ^ ^ } - ^ ^ ^ ^ a > - g - i ^5j -#- T-J-EMTL o l r t ti]-M- J?.— DUPIC

3.3-l<H| a]J2.*>Sit:>. ^ ^ ^ ^ ^ ^ 7 }

- 218 -

KAERI/RR-1999/99

DUPIC

3.3.1.3

3.3-2&fl

DUPIC

7]-g- DUPIC «J<*I5.£1 ^ l^Hlfe 558 $/kgHMS^15 DUPIC

DUPIC

3.3-3^1

DUPIC ^

^ H ^ 0.03

3.3.2

DUPIC ^<^.g.S>) «>-§-S.7}

2 3 5 u

DUPIC

3.3.2.1

«K§-3£.

3.3-2^

- 219 -

KAERI/RR-1999/99

CANDU #

5 *> 90%

(>1.25811) 7}

ZL J 3.3-3^ -f Bfeo] ^ 7 ^ 7 ] * H $1}"^ -^^ ^ofl rc}$

f } \ > | U ^ ^ f ^ J l ! £ ^ ^ f l ^ h g ] 1 . 1 6 6 ±

0 . 0 0 1 4 f l 1 1 ^ ^ i }

3.3.2.2 SEU/DUofl

SEUi} DU «K§-S. A>o]6l| S i A ^ , DUPIC

SEU1- ^ ^ t l c } . ^ ^ - ^ ^ SEU^ ° o ^ DUPIC

^ 1 ^ ±0.001

DUPIC

SEU/DU# A|~§-*> «>-§-£ ^ ^ 4 ^ ^ a 3.3-4611 v|-E}i-jj$at:}. o|

SEU

SEU/DU

- DUPIC *<|<5i£. «a^H7} ^ 7 } * H ^ ^ 5 . 7 1 Hi 7} ^ ^ = 5 . a ] ^ ^ - 7} 2} a]

CANDU

- 220 -

KAERI/RR-1999/99

3.3.2.3

SEU/DU»

37W

1.19

DU

DUPIC

43.1.17

, SEU/

^ 1.185.

3.

100%

3.3.3.1

DUPIC

. 35]

^r DUPIC

«>-§-£# 1.16 g 1.21 A>o]6JI ^ 7 1

25%

- 221 -

KAERI/RR-1999/99

B k

(2%

7}

= 1.21)51

).

3.3.3.2 <>)-§-

°l-8-i-ol

3.3-2)<Hl $ a ^

SEU

88%

3.3.3.3

- 222 -

KAERI/RR-1999/99

CPPF# 1.064

1.197}

3.3.4 DUPIC

DUPIC

239Pu

3.3-7~3.3-9<Hl

3.c}. a

SEU/DU S i

DUPIC

DUPIC n^S.^ ^ «J-S-^^ 1-14 4 1.46 wt% g| 0.57 wt%o]u}. ojAJ.

37}x)

- 223 -

KAERI/RR-1999/99

3.3.4.1

DUPIC i ^ H H f e 2-

3.3.4.2

DUPIC

9%

7] $M DUPIC

Doppler

3000 MWd/t

3.3.4.3

DUPIC

oilDUPIC

DUPIC

7}

- 224 -

KAERI/RR-1999/99

3.3.5 £.<$

-M^Hr DUPIC

96o/o7}

^r DUPIC ^ £

CANDUAf ^ ^ ^J^>Sofl>fc| ^J-ô] SjuK SEU/

g-^-°l 3.4%7}

7} t<[£<£^ # ^

y,i^-H SôU^fe A ^ ^ T -fete^ ^ & o?-8ô1 *l=fl 88%ol

SEU/DUcHl

- 225 -

KAERI/RR-1999/99

Table 3.3-1

Composition Variation for Fissile Content Adjustment Option

Actinides

Fission

Products

2 3 5 u

239Pu241Pu240Pu

241Am

155Gd149Sm143Nd15ISmI03Rh

Number of Mixing Operation

0

0.0*

0.0

20.3

8.9

11.4

21.2

4.3

7.6

6.7

9.0

1

0.0

0.0

14.7

8.8

10.7

18.7

1.7

8.7

6.4

10.1

2

0.0

0.0

9.8

5.6

6.6

12.1

0.9

5.1

4.1

6.1

3

0.0

0.0

6.7

4.1

5.1

8.3

0.7

3.0

2.6

3.9

* Two standard deviations in percent

- 226 -

KAERI/RR-1999/99

Table 3.3-2

Summary of Fissile Content Adjustment Option

Number of assemblies mixed

Spent PWR fuel utilization (%)

DUPIC fuel composition (wt%)

Spent PWR fuel

SEU feed

DU feed

Manufacturing process (400 t/yr capacity)

Annual SEU feed (MTU)

Annual SEU cost (M$) - O&M cost

Manufacturing Cost ($/kgU)

Fuel cycle cost (mills/kWh)

First Mixing

2

90

82.5

7.8

9.7

31.2

29.3

631

5.49

Second Mixing

4

90

82.6

7.3

10.1

29.2

27.4

627

5.48

Third Mixing

8

96

82.7

6.5

10.8

26.0

24.4

619

5.46

- 227 -

KAERI/RR-1999/99

Table 3.3-3

Unit Cost of Fuel Cycle Components

Component

Uranium(U3O8) ($/lb)

- PWR

- CANDU

Conversion ($/kgU)

- PWR

- CANDU

Enrichment ($/SWU)

Fabrication ($/kgU)

- PWR

- CANDU

- DUPIC

Transportation ($/kgHM)

- DUPIC

Transportation &

Storage ($/kgHM)

- PWR

- CANDU

- DUPIC

Disposal ($/kgHM)

- PWR

- CANDU

- DUPIC

Loss Rate

(%)

0.5

0.5

1

1

1

Lead/Lag time

(months)

-24

-17

-18

-13

-12

-6

-10

-10

120

120

120

120

360

360

360

Unit Cost

19.2

19.2

8

8

110

275

65

558

50

230

48a

170b

610

73

316C

a transportation cost [Ref. 14] + storage cost [Ref. 16]b transportation & storage cost for DUPIC = transportation and storage cost for CANDU

x 4.5c disposal cost for DUPIC = disposal cost for CANDU X 4.33

- 228 -

KAERI/RR-1999/99

Table 3.3-4

Summary of Reactivity Control by SEU/DU

Average ka>

Poisoned koo

Spent PWR fuel utilization

Fuel composition (wt%)

Spent PWR fuel

SEU feed

DU feed


Manufacturing process (400 T/year)

Annual SEU feed (MTU)

Annual SEU cost (M$) - O&M

Manufacturing cost ($/kgU)

Discharge burnup (MWd/T)


Target Reactivity ( koo )

1.16

1.16022

1.12749

100.0

90.6

0.0

9.4

4.22

0.0

0.0

558

11954

5.53

1.17

1.17053

1.13736

100.0

92.7

0.9

6.4

4.85

3.6

3.4

566

13395

5.43

1.18

1.18021

1.14650

100.0

96.6

2.3

1.1

5.15

9.4

8.8

580

14523

5.36

1.19

1.19012

1.15589

100.0

95.9

4.1

0.0

5.27

16.3

15.3

596

15562

5.32

1.20

1.20005

1.16506

100.0

93.9

6.1

0.0

5.45

24.4

23.0

615

16570

5.28

1.21

1.21008

1.17435

100.0

91.8

8.2

0.0

5.63

32.7

30.8

635

17590

5.24

- 229 -

KAERI/RR-1999/99

Table 3.3-5

Summary of Utilization of Linear Reactivity Fuel

Average kco

Poisoned koo

Spent PWR fuel utilization

DUPIC Fuel composition (wt%)

Spent PWR fuel

Natural uranium feed


Manufacturing process (400 T/year)

Annual NU feed (MTU)

Annual NU cost (k$) - O&M

Manufacturing cost ($/kgU)

Discharge burnup (MWd/T)


Target Reactivity ( koo )

1.16

1.16018

1.12813

87.8

99.9

0.1

4.72

0.4

20

558

12522

5.45

1.17

1.16995

1.13712

77.4

99.8

0.2

4.84

0.8

40

558

13274

5.40

1.18

1.18006

1.14640

66.2

99.2

0.8

4.93

3.2

160

558

14006

5.35

1.19

1.18998

1.15572

54.7

99.9

0.1

5.10

0.4

20

558

14866

5.30

1.20

1.19984

1.16494

43.3

99.6

0.4

5.27

1.6

80

558

15700

5.25

1.21

1.20999

1.17444

32.4

99.7

0.3

5.49

1.2

60

558

16627

5.19

- 230 -

KAERI/RR-1999/99

Table 3.3-6

Comparison of koo and Isotopic Composition

Isotope

Koo

Actinides

Fission

Products

235U

B 9Pu

241Pu

240Pu

241Am

155Gd

I49Sm

M3Nd

151Sm

103Rh

Fissile Content

Adjustment

1.15623

10000.0

4500.0

423.7

1758.9

617.1

8.8

3.3

608.6

10.3

318.6

± 1.26%*

±0.0%

±0.0%

±6.7%

±4.1%

±5.1%

±8.3%

±0.7%

±3.0%

±2.6%

±3.9%

Reactivity Control

by SEU/DU

1.14614

9663.8

5325.4

500.7

2075.4

727.7

10.4

3.9

720.0

12.2

376.4

±0.20%

±2.9%

±1.7%

±7.0%

±2.9%

±4.1%

±5.6%

± 1.6%

±4.3%

±3.6%

±4.1%

Reactivity

by Natural

1.12817

8660.3

5473.0

520.8

2153.7

758.8

10.7

4.0

742.4

12.5

389.3

Control

Uranium

±0.25%

±3.3%

±3.6%

± 10.4%

±5.0%

±6.5%

±7.9%

±2.1%

±6.6%

±4.9%

±6.4%

* Average ± two standard deviations

- 231 -

KAERI/RR-1999/99

Table 3.3-7

Comparison of koo Variation

Burnup (MWD/T)

.0

3.3

164.8

824.9

1650.1

2474.9

3299.5

4124.2

4948.9

5773.4

6598.0

7422.6

8247.4

9072.1

9896.7

10721.4

11546.1

12370.8

13195.7

14020.5

14845.4

Equilibrium

Fissile Content

Adjustment

1.15619

1.15204

1.14695

1.15393

1.14305

1.12997

1.11634

1.10247

1.08849

1.07448

1.06047

1.04649

1.03258

1.01878

1.00511

.99160

.97829

.96520

.95235

.93980

.92755

1.04616

±.01453*

±.01434

±.01191

±.00960

±.00894

±.00851

±.00813

±.00777

±.00742

±.00708

±.00676

±.00643

±.00613

±.00584

±.00555

±.00528

±.00502

±.00477

±.00453

±.00431

±.00410

± .00642

Reactivity Control

by SEU/DU

1.14614

1.14219

1.13945

1.14904

1.13858

1.12554

1.11187

1.09796

1.08396

1.06992

1.05592

1.04198

1.02814

1.01440

1.00084

.98745

.97426

.96132

.94862

.93623

.92414

1.04568

±.00231

±.00244

±.00279

±.00456

±.00530

±.00577

±.00620

±.00659

±.00696

±.00731

± .00762

±.00790

±.00815

±.00835

±.00850

± .00861

±.00868

±.00870

±.00866

±.00858

± .00845

±.00783

Reactivity

by Natural

1.12817

1.12431

1.12408

1.13392

1.12295

1.10940

1.09525

1.08091

1.06654

1.05220

1.03796

1.02384

1.00988

.99612

.98258

.96929

.95627

.94355

1.04596

Control

Uranium

±.00286

±.00302

±.00299

±.00553

±.00653

± .00714

±.00765

±.00811

±.00852

±.00890

±.00923

± .00951

±.00973

±.00990

±.01001

±.01006

±.01005

± .00998

±.00904

* Two standard deviations

- 232 -

KAERI7RR-1999/99

Table 3.3-8

Comparison of Thermal Absorption Cross-Section

1

Burnup (MWD/T)

.0

3.3

164.8

824.9

1650.1

2474.9

3299.5

4124.2

4948.9

5773.4

6598.0

7422.6

8247.4

9072.1

9896.7

10721.4

11546.1

12370.8

13195.7

14020.5

14845.4

Equilibrium

Fissile Content

Adjustment

.53647

.53712

.53599

.53015

.52591

.52185

.51783

.51382

.50985

.50590

.50199

.49813

.49431

.49056

.48687

.48325

.47972

.47628

.47294

.46970

.46658

.49804

± .00259*

±.00256

±.00226

±.00204

±.00207

±.00211

±.00216

±.00220

±.00224

±.00228

±.00231

±.00233

± .00235

± .00237

±.00238

±.00239

±.00239

±.00238

±.00238

±.00236

±.00234

± .00233

Reactivity Control

by SEU/DU

.54905

.54965

.54816

.54184

.53735

.53311

.52890

.52473

.52059

.51648

.51242

.50840

.50443

.50053

.49669

.49292

.48923

.48563

.48212

.47871

.47541

.50947

±.00679

±.00676

±.00682

±.00675

±.00681

±.00670

±.00698

±.00705

±.00711

±.00716

±.00720

±.00723

±.00724

±.00724

±.00723

±.00720

±.00716

±.00710

±.00702

±.00693

±.00682

± .00722

Reactivity

by Natural

.54778

.54839

.54639

.53972

.53502

.53058

.52620

.52187

.51759

.51335

.50918

.50507

.50103

.49706

.49319

.48940

.48571

.48213

.51152

Control

Uranium

±.01133

±.01129

±.01146

±.01132

±.01136

±.01143

± .01149

±.01153

±.01155

±.01156

±.01155

±.01152

±.01147

±.01140

±.01131

±.01121

±.01107

±.01092

±.01156


- 233 -

KAERI/RR-1999/99

Table 3.3-9

Comparison of Neutron Production Cross-section (XlOO)

Burnup (MWD/T)

.0

3.3

164.8

824.9

1650.1

2474.9

3299.5

4124.2

4948.9

5773.4

6598.0

7422.6

8247.4

9072.1

9896.7

10721.4

11546.1

12370.8

13195.7

14020.5

14845.4

Equilibrium

Fissile Content

Adjustment

.78249

.78063

.77551

.77167

.75830

.74389

.72930

.71472

.70025

.68595

.67183

.65793

.64427

.63089

.61780

.60503

.59260

.58054

.56886

.55759

.54674

.65761

±.00507*

±.00501

± .00399

± .00295

± .00255

±.00226

±.00203

±.00184

±.00167

± .00155

±.00145

±.00139

±.00136

±.00135

±.00136

±.00138

±.00141

±.00143

±.00146

±.00148

±.00149

±.00139

Reactivity Control

by SEU/DU

.79785

.79599

.79187

.78922

.77555

.76062

.74547

.73035

.71536

.70054

.68594

.67159

.65751

.64373

.63027

.61713

.60435

.59195

.57993

.56833

.55715

.67540

±.01396

±.01396

±.01415

±.01529

±.01569

±.01590

±.01606

±.01619

±.01627

±.01632

±.01632

±.01629

±.01622

±.01611

±.01595

±.01576

±.01552

±.01524

±.01492

±.01455

±.01415

±.01630

Reactivity

by Natural

.78362

.78180

.77873

.77588

.76168

.74627

.73069

.71520

.69992

.68488

.67013

.65569

.64160

.62787

.61453

.60157

.58904

.57694

.67842

Control

Uranium

±.02199

±.02198

±.02215

±.02373

±.02413

±.02427

±.02430

± .02428

±.02420

± .02407

±.02388

±.02365

±.02336

±.02304

±.02264

±.02220

±.02171

±.02116

±.02399


- 234 -

KAERI/RR-1999/99

300

1.12 1.14 1.16 1.18 1.20

300

250-

g 200-

8 ISO-

'S 100-

62 50-

Second Mixing

" . 1

1.12 1.14 1.16 1.18 1.20

XIE<s>

fd2

250-

200-

150-

100-

50-

0-

Third Mixing

•B

•

• • • •1.10 1.12 1.14 1.16 1.18 1.20

k-inf

Fig. 3.3-1 Distribution of k«> for Fissile Content Adjustment Option

- 235 -

KAERI/RR-1999/99

0 500 1000 1500 2000 2500 3000 3500

Assembly Number

Fig. 3.3-2 Distribution of k«> for Spent PWR Fuel

- 236 -

era

oo

g2,

No. of Assembly No. of Assembly No. of Assembly

I

H

x'(D

3 N

5

KAERI/RR-1999/99

3.4 DUPIC

WIMS-AECL

, «Kg-s.

DUPIC

DUPIC 37-g-

3.4.1

3.4.1.1

DUPIC

fe DUPIC 3.4-24J-

25-

. DUPIC

], DUPIC

3.4.1.2

-LH 3.4-2^ DUPIC

>. DUPIC

, DUPIC

3.4.1.3

CANDU

CANDU

- 238 -

KAERI/RR-1999/99

f ^ ^ } fi 3.4-4

DUPIC ^<^.3.6\) tfl# ^lU -§"8*1- $ # 2 f -§-3j

fe- S 3A-5O]}

3.4.2

WIMS-AECL

(Fresh Clean Fuel)

2 - 3

K DUPIC |

(Equilibrium Fuel)

3.4.2.1

- 239 -

KAERI/RR-1999/99

(resonance

escape probability)*!] <3*<M" u]*M, <>}$.

f «>-§- #^S>] ^ S f ^ < i ^ ^ ^ f 4 ^ - ^ 61^]- (thermal

reproduction factor)^] r)*\

HQ 3.4-3^

(poisoned) ^ ^ f l L f S 4 ^ i « } £ S # J i e } ^ tc|f,

3.4.2.2

fe oj

. DUPIC

-L5]vf, DUPIC

^ > } H ^ > ^ ^ f e I & ) fe <^ 140°C

- 240 -

KAERI/RR-1999/99

3.4.2.3 JL:zf

6\]

fet;)-. DUPIC 37-g-

MX-}

3.4.2.4

ZL^ 3.4-6 «>-§-J£*l

shutdown)^

shutdown)^

DUPIC

^fe 4f 12

- 241 -

KAERI/RR-1999/99

3.4.3

3.4.3.1 713E

JL

. DUPIC

Hl«H 7.3 mk7} 4 ^ :

37-g- m<&3.°\) «1«J 2.1 mk

9.6 mkti^l, <>lfe 37-g-

DUPIC ^

3.4.3.2

3.4-7^

O] ^ s f s ^ - B j , 7]

^ - 7 f e DUPIC ^l^S.ol] T=H«H Jicf

DUPIC ^<£S7l- ^ 4 ^ r 71 a «

Ufl^ofl, 37-g- ^ ^ -

3.4.3.3

DUPIC

71

- 242 -

KAERI/RR-1999/99

>. ZLQ 3.4-8

3.4.3.4

7)S

D2O H2O

713E

3.4-10<HJ

3.4.3.5

71S

. Helvf, CANDU-6 44

RFSP

71S 6.51

fl 0.89 mk

DUPIC

3.4.4

- 243 -

KAERI/RR-1999/99

7}

3.4.4.1 DUPIC

100%

3.4-11611 100% £ # e

, DUPIC

«1*H

DUPIC 71SE4I-S 7l^-7]

DUPIC

^ S . , DUPIC

3.4.4.2

CANDU-6 , WIMS-AECL

fe 3.71. 100% , 0.31

-0.45

WIMS-

- 244 -

KAERI/RR-1999/99

, WIMS-AECL e>o}tt.

3.4.5

^ 7|E}- «!-§-£ J L ^ # # 7 1 ^ * ^ © . ^ , o]Bj$> jL3ffe WIMS-AECL

3.4.5.1

fe 99.85 wt% (99.833 at%)o]nf ^ t^^) 0.15 wt%fe

) 3.7}

98.0-99.8 at%

17.5 mk/at%, ^g^Jn^ ^ ^ S ^ A - | f e 20.9 mk/at%3.

3.4.5.2

^^1 47] ofi -cHl

ofl H]s|| £ ^ * 1 4^1-. 3 . ^ 3.4-13^

-. 99.2-99.8 at% ^^6flA-| ^4*fl ^ £ T J ^ ^ . e l S ^ ^ - c - 0.36 mk/at%,

fe 0.46 mk/at%S

- 245 -

KAERI/RR-1999/99

Table 3.4-1

Comparison of Design Parameters for DUPIC and Natural Uranium Fuel

Parameters

Bundle Design

Element outside diameter (cm)

Average sheath wall thickness (cm)

Pellet outside diameter (cm)

Stack length (cm)

Fuel Material

Pellet density (g/cm3)235U content (wt%)239Pu content (wr%)

Fissile content (wt%)

Dy in center rod

Weight per Bundle (kg)

(U-Pu-X)

U

(U-Pu-X)O2

UO2

DUPIC

43-element

1.350 (large), 1.150 (small)

0.039 (large), 0.036 (small)

1.267 (large), 1.073 (small)

48.2

(U-Pu-X)O2, Spent PWR fuel

10.4

1.0

0.45

1.488

4.64

18.372

20.837

Natural Uranium

37-element

1.308

0.040

1.220

48.2

UO2

10.6

0.711

0.711

19.100

21.782

- 246 -

KAERI/RR-1999/99

Table 3.4-2

Lattice Parameters for DUPIC Fuel

Burnup

(MWd/T)

.02.5

123.71238.72477.03715.24953.36191.47429.68667.99906.1

11144.512382.813621.314859.916098.517337.318576.419815.620435.221055.021674.922294.722914.623534.624154.624774.625394.726014.926635.127255.427875.728496.029116.5

I*

2.3857E-012.3857E-012.3857E-012.3857E-012.3858E-012.3859E-012.3859E-012.3859E-012.3860E-012.3860E-012.3861E-012.3861E-012.3861E-012.3862E-012.3861E-012.3861E-012.3862E-012.3861E-012.3861E-012.3861E-012.3860E-012.3860E-012.3860E-012.3859E-012.3859E-012.3859E-012.3858E-012.3858E-012.3858E-012.3858E-012.3857E-012.3857E-012.3857E-012.3857E-01

Eta

3.6489E-013.6491E-013.6493E-013.6476E-013.6465E-013.6455E-013.6443E-013.6433E-013.6422E-013.6412E-013.6401E-013.6392E-013.6383E-013.6373E-013.6365E-013.6357E-013.6349E-013.6341E-013.6336E-013.6331E-013.6329E-013.6326E-013.6324E-013.6322E-0I3.6318E-013.6315E-013.6314E-013.6311E-013.6310E-013.6308E-013.6306E-013.6303E-013.6302E-013.6301E-01



Xn


IK

8.2904E-038.2904E-038.2905E-038.2886E-038.2860E-038.2834E-038.2809E-038.2785E-038.2760E-038.2737E-038.2714E-038.2692E-038.2670E-038.2649E-038.2628E-038.2606E-038.2585E-038.2564E-038.2543E-038.2533E-038.2522E-038.2512E-038.2501 E-038.2491E-038.2480E-038.2470E-038.2460E-038.2449E-038.2439E-038.2428E-038.2417E-038.2407E-038.2396E-038.2386E-03

H

.39320

.39242

.38964

.38446

.37284

.36100

.34926

.33771

.32642

.31543

.30478

.29454

.28473

.27539

.26657

.25829

.25058

.24345

.23692

.23387

.23097

.22822

.22561

.22313

.22080

.21859

.21652

.21457

.21273

.21101

.20940

.20789

.20648

.20516

- 247 -

KAERI/RR-1999/99

Table 3.4-3

Lattice Parameters for Natural Uranium Fuel

Burnup

(MWd/T)

.03.2

157.0791.1

1582.42372.53161.83950.44738.85526.96314.87102.77890.58678.39466.3

10254.211042.311830.412618.713407.014195.514984.015772.516561.217349.918138.618927.419716.320505.121294.122083.022871.923660.924449.8

, . ,


r-


Xal


3.5245E-033.5359E-033.6015E-033.7130E-033.8057E-033.8653E-033.9058E-033.9341 E-033.9543E-033.9689E-033.9798E-033.9883E-033.9954E-034.0014E-034.0068E-034.0120E-034.0172E-034.0224E-034.0279E-034.0337E-034.0398E-034.0461 E-034.0527E-034.0594E-034.0662E-034.0731 E-034.0800E-034.0868E-034.0936E-034.1003E-034.1069E-034.1133E-034.1196E-034.1257E-03

la

4.6937E-034.6811E-034.6105E-034.7831E-034.9171E-034.9782E-034.9970E-034.9889E-034.9639E-034.9279E-034.8854E-034.8396E-034.7933E-034.7469E-034.7019E-034.6591E-034.6192E-034.5823E-034.5487E-034.5185E-034.49I5E-034.4676E-034.4464E-034.4277E-034.4114E-034.3971E-034.3845E-034.3734E-034.3637E-034.3551E-034.3474E-034.3405E-034.3343E-034.3286E-03

SR

8.6383E-038.6383E-038.6381E-038.6270E-038.6023E-038.5771 E-038.5546E-038.5352E-038.5184E-038.5035E-038.4902E-038.4780E-038.4668E-038.4565E-038.4467E-038.4375E-038.4289E-038.4206E-038.4127E-038.4052E-038.3980E-038.3910E-038.3843E-038.3779E-038.3717E-038.3657E-038.3598E-038.3543E-038.3489E-038.3437E-038.3386E-038.3337E-038.3289E-038.3243E-03

H

.25094

.25025

.24582

.25086

.25401

.25427

.25290

.25056

.24766

.24443

.24107

.23771

.23445

.23130

.22833

.22556

.22302

.22069

.21858

.21670

.21503

.21355

.21225

.21110

.21010

.20921

.20844

.20777

.20717

.20665

.20618

.20577

.20540

.20506

- 248 -

KAERI/RR-1999/99

Table 3.4-4

Kinetic Parameters of DUPIC Fuel

Group

1

2

3

4

5

6

lP

A

Vl

V2

Fresh Condition*

(0 MWd/T)

B.000284

.001121

.000983

.002235

.000735

.000180

.005537

A

.000595

.031482

.121806

.317582

1.394136

3.759629

0.000513 sec

0.000453 sec

9.54 X106 cm/sec

2.88 X105 cm/sec

Equilibrium Condition

(7442 MWd/T)

0.000263

.001072

.000930

.002106

.000714

.000172

.005257

A

.000563

.031344

.123257

.320546

1.404731

3.729208

0.000553 sec

0.000539 sec

9.55 XI06 cm/sec

2.87 X105 cm/sec

* No boron in moderator

- 249 -

KAERI/RR-1999/99

Table 3.4-5

Kinetic Parameters of Natural Uranium Fuel

Group

1

2

3

4

5

6

lP

A

V]

V2

Fresh Condition*

(0 MWD/T)

B.000380

.001496

.001356

.003192

.001022

.000233

.007680

A

.000726

.031731

.117089

.312620

1.401893

3.910627

0.000826 sec

0.000757 sec

9.17X106 cm/sec

2.85 x 10s cm/sec

Equilibrium Condition

(3977 MWD/T)

0.000282

.001093

.000964

.002179

.000717

.000183

.005417

A

.000591

.031527

.122425

.317652

1.384846

3.758042

0.000731 sec

0.000712 sec

9.26 X106 cm/sec

2.85 x l 05 cm/sec

* No boron in moderator

- 250 -

KAERI/RR-1999/99

Table 3.4-6

Comparison of Void Reactivity

Fresh Clean Fuel

Equilibrium Fuel

DUPIC

9.284 mk

11.779 mk

Natural Uranium

16.584 mk

13.879 mk

- 251 -

KAERI/RR-1999/99

Table 3.4-7

Reactivity Feedback (mk) due to Power Level Change

Power

Level

(%)

130

120

110

100

90

80

70

60

50

DUPIC Fuel Core

TfUel

-0.29

-0.19

-0.10

0.0

0.10

0.19

0.27

0.37

0.45

Icool

1.11

0.57

0.20

0.0

-0.10

-0.18

-0.25

-0.34

-0.40

Pcool

-1.57

-1.16

-0.65

0.0

0.76

1.90

3.36

5.29

7.91

Natural Uranium Core

Tfuel

-0.11

-0.09

-0.04

0.0

0.06

0.12

0.17

0.25

0.31

Pcool

1.19

0.62

0.20

0.0

-0.11

-0.20

-0.27

-0.38

-0.45

Xenon

-0.94

-0.71

-0.39

0.0

0.45

1.13

2.01

3.35

5.17

- 252 -

KAERI/RR-1999/99

- Natural Uranium•DUPIC

0 2000 4000 6000 8000 10000 12000 14000

Bumup (MWD/T)

Fig. 3.4-1 Variation of *„, and kelf with Burnup (PPV+WIMS)

- 253 -

1.3

co

but

istr

Q

(U

oX

ive

at

a:

1.2 -

1.1 -

1.0 -

0.9 -

0.8 -

0.7 -

0.6 -

0.5 -

0.4 -

KAERL/RR-1999/99

OUTER

INTERMEDIATE

SCENTER

- Natural Uranium•DUPIC

2000 4000 6000 8000 10000 12000 14000

Burnup (MWD/T)

Fig. 3.4-2 Variation of Relative Element Linear Power with Burnup

- 254 -

KAERI/RR-1999/99

I

Ias

DC

10 20 40 50 60

Moderator Temperature (Xi)

70 80 90

Fig. 3.4-3 Reactivity Change due to Moderator Temperature (WIMS)

- 255 -

KAERI/RR-1999/99

-2 -

CD

-4 -O

USas

s. -6

- 8 -

Equilibrium - Natural Uranium- DUPIC

40 80 120 160 200 240

Coolant Temperature ( t ! )

280 320

Fig. 3.4-4 Reactivity Change due to Coolant Temperature (WIMS)

- 256 -

KAERI/RR-1999/99

1 0 -

8 -

6 -

t- 4 _a>O)

1 2 -o

ts8. -2-

- 4 -

- 6 -

i | • | * | i

\ ^

Fresh r-^N.

Equilibriurn ^ ^ . . . ' " - - ^~\^^

-

DUPIC

-

-

•

^ " \ ^ ^ -

200 400 600 800

Fuel Temperature CC)

1000 1200

Fig. 3.4-5 Reactivity Change due to Fuel Temperature (WIMS)

- 257 -

KAERI/RR-1999/99

100 200 300 400

Temperature (t;)

500 600 700

Fig. 3.4-6 Reactivity Change due to System Temperature Following a Reactor

Shutdown (WIMS)

- 258 -

KAERI/RR-1999/99

- Natural Uranium-DUPIC

0.000 0.125 0.250 0.375 0.500 0.625


0.750 0.875

Fig. 3.4-7 Reactivity Increase due to Complete and Partial Voiding of Coolant

(WIMS)

- 259 -

KAERI/RR-1999/99

0 2000 4000 6000 8000 10000 12000 14000

Burnup (MWD/U)

Fig. 3.4-8 Variation of Coolant Void Reactivity with Fuel Burnup (WIMS)

- 260 -

KAERI/RR-1999/99

24-

2 2 -

2 0 -

-eas

o

i>

Rea

ctiv

void

Coo

lant

18

16

14

12

1 0 -

- Coolant Purity = 99.00 a/o- Coolant Purity = 97.23 a/o

Equilibrium Fuel

0.00 1.25 2.50 3.75 5.00 6.25 7.50

Moderator Boron Concentration (ppm)

8.75 10.00

Fig. 3.4-9 Dependence of Coolant Void Reactivity on Amount of Boron in Moderator

and Coolant Purity for DUPIC Fuel

- 261 -

KAERI/RR-1999/99

24

1 0 -

- Coolant Purity = 99.00 a/o- Coolant Purity = 97.23 a/o

0.00 1.25 2.50 3.75 5.00 6.25 7.50

Moderator Boron Concentration (ppm)

8.75 10.00

Fig. 3.4-10 Dependence of Coolant Void Reactivity on Amount of Boron in

Moderator and Coolant Purity for Natural Uranium Fuel (WIMS)

- 262 -

KAERI/RR-1999/99

t

1.0015

1.0010

1.0005

1.0000

WIMS-AECL/RFSP

50 60 70 80 90 100

Reactor Power

Natural Uranium >

110 120 130

Fig. 3.4-11 Comparison of Power Coefficients

- 263 -

KAERI/RR-1999/99

10

-10-

C03

o

101

8.

-30-

-50-

-60

Fresh

--"7Equilibrium

\Fresh


98.00 98.25 98.50 98.75 99.00 99.25 99.50

Moderator Purity (atom percent)

99.75 100.00

Fig. 3.4-12 Reactivity Change due to Moderator D2O Purity (WIMS)

- 264 -

KAERI/RR-1999/99

(U

cto

O

8.

-0.5-

-1.0-

-1.597.6 98.0 98.4 98.8 99.2

Coolant Purity (atom percent)

99.6

Fig. 3.4-13 Reactivity Change due to Coolant D2O Purity (WIMS)

- 265 -

KAERI/RR-1999/99

3.5 719"

CANDU-6 ^^}S.<^lfe 3^~g- (Adjuster rods: ADS), ^ *] ^ ^ * ] (Zone controUer

unit: ZCU), £L*|-5. ^{*] 7)|J§- (Shutdown system: SDS), - I K ^ r (Mechanical control absorber:

^£ 47}*]

. (H^l 3.2-5

3.5.1

ZCU

(bulk control) (spatial control)

3.5.1.1

(AZL)

Sum of volumes of water in all compartmentsTotal volume of all compartments 1UU

100%

- 266 -

KAERI/RR-1999/99

4 5.76 mk 9| 3.22 inkS.

91 3.65 mkS

37-g-

100% 9| 50%<^ 4 4 6.50 mk

JE 3.5-12} & 3.5-2^- ^ 4 Hhg-5.7]- 7fl*H o]-g-S}

DUPIC n<&£. 9J

DUPIC 9J 3.5-2^ n}

top-to-bottom tilt(%) == (Pi + p3 + ps) ~ (P2 + Ps + PT)-h Jf 2 X ( 3 5 . 2 )

side-to-side tilt(%) = 100 (3.5-3)

3.5-3<H]

i = 1. 2, - ,

form factor^

DUPIC 9J

H]jZ«fl

JE-^-

3.5.1.2

RFSP

form factor7l-

(Spatial Oscillation)

600-FPDitJ

. DUPIC 9J

- 267 -

KAERI/RR-1999/99

J-14, S-3,

J-14

L-97>

S-3 Hi L-9

& DUPIC

, DUPIC

3.5.1.3

^AS.

14711

(mk), 4

- 268 -

KAERI/RR-1999/99

3.5-45+ fi 3.5-5< ) 7 ) 1 ^ # 3 ^ Tfl^ *£ # ^ * H r $& 7fl<M- 4 4 M"H|*B DUPIC

DUPIC

nl*>c}. ^ , DUPIC ^

7}

3.5.2

. o) %6\)x\±; DUPIC

shim

3.5.2.1

RFSP 3 H 5 ) A]

DUPIC 51 ^^-^-efe ^"€ i=--y^] tfl^H 4 4 10.2 mkif 16.6 mk^

DUPIC 91 ^ ^ 1 f f e 8 ^ ^ n flH fl<x) ^ 30^ ^ ]^# afi^ ^^- -f-^fe 6.8

3.4 mk W

DUPIC

3.5.2.2

- 269 -

KAERI/RR-1999/99

$ 50%5L

DUPIC «£ ££-T-BBr #*l ii-fcH cfl*H 4 4 S 3.5-6

3.5-7ofl

383 cm ^ 1 \ \ }

171 cm o H ^ I ^*11>^. S^> ^1^> ^2f, DUPIC

. ZL^xln>, DUPIC J ^ ^ i | ^^-^1 S^-g-51 ^o) b |

3.5.2.3

^ ^ ^ 50%«

^ i DUPIC ^ ^<a-fefe «?^1S ^ i i^^l cfl*H 4 4 S 3.5-83.5-9«Hl A^*>Sii:>. fi 3.5-8^ ^

^: 4

DUPIC

DUPIC ^ ^ ^ ^190^o] x]<&5\o\ 3 9 5 ^ ]

, DUPIC J^ ^ f

- 270 -

KAERI/RR-1999/99

DUPIC

$fe 90%

3.5.2.4 Poison-out

°) £4

DUPIC

50%

ufl

36.4*1 #

DUPIC

H ^ 3.5-5^1

DUPIC 3.5-102f

3.5-1H] DUPIC

3.5.2.5 «>-§-£ shim

7 ] ^ , CANDU

CANDU-6

mk/FPDo]i;|-. DUPIC

3.5-13^]

. DUPIC

4 4 0.40 mk/FPDAf 0.48

3.5-12^]

| ^ shim 47H5] 7J[SJ

- 271 -

KAERI/RR-1999/99

Shim

&JL 94%S - f r * ] ^ - . Shim

1.11 mkS

2.8

shim

(FPD)

4.0<H

, DUPIC

31*1

6> ZLelU, shim

3.5.2.6 # ^ ^»J (Stepback)

60%

DUPIC

3.5-145} a 3.5-15^] 4 4

DUPIC 84

70%

DUPIC

DUPIC

- 272 -

KAER1/RR-1999/99

3.5.3

CANDU-6

(MCA)©]

471151

DUPIC 8.36

2.99 mk7>

3.0

3.5-6^1

EL7\]

3.5.4

V. CANDU-6

- 273 -

KAERI/RR-1999/99

o}Jf- ^7K&el breakup assessment)^

A}JL(1OSS of coolant accident: LOCA)A| ^ n i z^*\ 3 ^ c|^-<^ ^

>. LOCAA] ^ 7 ] f e power pulse %•<& o|AfSf

71^^1*1 840

k W V t W ^^peaking)]

840 j / r (3-5-5)

20%

100%

3.5.4.1 ^g^j-g- ^8^ «hg-J£

£4

71 S3

3.5-16cHl M-B>LflSdc>. DUPIC

72.5 mk^Ic-il,

.7>7> 40.5

, 287fl ut 267H3] ^ ^ 1 ^ - ^ cfltl ^z\ «K§-£7fe 4 4 87 mki} 56 mk

£4

- 274 -

KAERI/RR-1999/99

'6 o

3.5-172]- ZL% 3.5-7^1

3.5.4.2

DUPIC

(reactor inlet header: R1H) 20%

*> s a * . ^ ,4% crept

0.00528S

28.3% ]<H1 HlSfl DUPIC

^ DUPIC

10.6% 335.

ZL1] 3.5-8^ 3.5-9^ 20%

power pulse7f DUPIC high log rate S

high log rate S H

- 275 -

KAERI/RR-1999/99

61 *> power pulsed ^ &£ 7\$ *}Q]7\ &-b 5 ^ S

DUPIC ii^<HH i-M*M itfl ofl ^ ^ ) n } * ! ^ DUPIC

RFSP 3.B.*\ CERBERUS 3L#ofl £|Sj} 7 ) 1 ^ u>^ 5]cfl 3j-g-^°1H*l-c- DUPIC

5} 3*^- 3257.1 kW • solJL, 7 l $ i ^ £ j -f- 3274.3 kW • &o\x%. oMy ^ 2.7} # ^ o | DUPIC5] ^g-f 745.8 kWo]jL 7]^% ^ - f 789.4

ig. o]A>S}~f e ] ^ / # ^ . S ^ 5 j -7117> DUPIC ^ - f 20.8 kg

21.7kgo]^, c } ^ ufl «>3J*y-*o* peaking factor7> DUPIC j

1.13H71 itjl-g-oll, ^ 2 f ^ ^ . S DUPIC I fuel breakup

*1*1, DUPIC i i ^ ^ n>xlol 7lî^ôfl HlSfl 3t>^7]ô] 840

3.9% #

3.5.4.3

SDS25] ^ ^ ^ 7 ] - l - $}*t DUPIC

^ o ] SDS2^ ^•^3g7]- 7lÂ>jLol ^^>S. Q^t (RIH) «J|^ 100%

37.8%^«^] Hl«H DUPIC

c>. SDSl^ ^-f^ l WlH nj-xtol 3 . o]-^-^ RIH 100% ^ - f blowdown rate7>

tc|-B> power pulse7> RIH 20% ^ - f 6\] ti^jj ^.*] ^-7>*>i^, o]oj| rc>B|-

£ RIH

ZLQ 3.5-103]- 3.5-11^- -T- iôf l tH*M ^^>S. ^ ^ - ^ nfl^ 100%

power pulse7l- DUPIC ±M°) ^ ^ ^7>*}<^ high log rate

< 0.95^

S high log rate S U ^ . S . ^^l£|o1 ^ 0.953:^

2.39# Ji&irt-L °M ijtll # ^ ^ ^ *}$, *1*1 power pulsed ^

- 276 -

KAERI/RR-1999/99

, 7]Bf

o}

DUPICS]

. ^ CERBERUS

2276.2 kW

850.7

peaking factor^ ^ } o | S <&%%, DUPIC5|

^ ^ # n l ^ , ^ 3 ] - ^ A 5 . DUPIC

840 J/g^- 7 l ^ A 5 . 5.8%

DUPIC

Sj*1|

2138.8

713.8

^-71]

DUPIC

24.4%,

3.5.5

( I 3 5Xe, I0SRh, I 13Cd, I 4 9Sm, I 5 1Sm

29

40%!-

7] 135Xe

0o]

4-f

70%#,

135 80 mkS

- 277 -

KAERI/RR-1999/99

35A) # 28

. CANDU-6

3.5.5.1

3.5-12^ 4

, DUPIC

20, 40, 60, 80, 100%

37-g-

37-g-

ioo%

37%

3.5.5.2

DUPIC

20, 40, 60, 80, 100%

3.5-13^

100%

3%

irfl

3.5.5.3

ZL^ 3.5-14 100% 80, 60, 40, 20, 0%S

- 278 -

KAERI/RR-1999/99

37-g-

37-g-

DUPIC

37% ^ ^

3.5.5.4 3 0 ^

CANDU-6

DUPIC

7mkc| ^

DUPIC

& 6.8

10.2

3.5.6

<§»oK§- Stfe ^ ^ ^ .

37],

51*11

DUPIC CANDU-6

3.5.6.1

- 279 -

KAERI/RR-1999/99

5f S.B. (higher harmonic modes)

CANDU

*\] 7}?]

*iSL?\7)

50%

5, 6, 7, 12, 13, 14, 19, 20,

(3.5-2)^-

front-to-back tilt (%) =P- —

x 100 (3.5-6)

45.4%^

3]

3.5-15ofl

170

44.5%^ 120

] - ^ ^ , DUPIC i - ^

37%

170 u}E}^r:}.

6l*H DUPIC 9J Sfe

. DUPIC

, DUPIC

- 280 -

KAERI/RR-1999/99

3.5.6.2

(damping factor) (threshold power)

fe DUPIC Hla.*]-7]

4

7>. ^-^- t[^f (Damping Factor)

(3.5-7)

(phase)

(3.5-8)

±= q (3.5-7)

- 281 -

KAERI/RR-1999/99

45% _ L 7 6 7

X10"2

37% ^J 79% . ^ , DUPIC

DUPIC

a 3.5-i8oH^

(Threshold Power)

# - 4 #

DUPIC

20% ^

10%

3.5-19

DUPIC

DUPIC

DUPIC

. DUPIC

cf

4

Uj-i}

- 282 -

KAERI/RR-1999/99

form factor# «]3.*H

j£ 3.5-20<Hl

fonn factor^

CANDU-6

>€ DUPIC

4 4 ^.H 3.5-21, 3.5-22

3.5.7

DUPIC

CANDU-6

! •§•

^- DUPIC

01 cf.

- DUPIC

poison-out 4

shim

60%

DUPIC

- 283 -

KAERI/RR-1999/99

, DUPIC i i ^ ^ ^lfe -f-^7} 37^-

. DUPIC

DUPIC ^ ^ ] j # ^

DUPIC fl

- 284 -

KAERI/RR-1999/99

Table 3.5-1

Reactivity Worth and Power Tilt vs. ZCU Level for DUPIC Core

• Average

Zone

Controller

Level

(%)

25

50

75

100

Reactivity

Worth

(mk)

1.626

3.222

4.657

5.761

Number of tfeO filled LP* divided by total number

of LP in ZCU compartment

Centre

Rod

Upper

1.875/7.5

3.750/7.5

5.625/7.5

7.500/7.5

Centre

Rod

Middle

1.75/7

3.50/7

5.25/7

7.00/7

Centre

Rod

Lower

1.75/7

3.50/7

5.25/7

7.00/7

Side

Rod

Upper

2.125/8.5

4.250/8.5

6.375/8.5

8.500/8.5

Side

Rod

Lower

2/8

4/8

6/8

8/8

Power Tilt

(top-to-

bottom)

1.00442

1.01121

0.96301

0.90049

*LP = Lattice Pitch

- 285 -

KAERI/RR-1999/99

Table 3.5-2

Reactivity Worth and Power Tilt vs. ZCU Level for Natural Uranium Core

Average

Zone

Controller

Level

(%)

25

50

75

100

Reactivity

Worth

(ink)

1.836

3.647

5.245

6.499

Number of H2O filled LP* divided by total number

of LP in ZCU compartment

Centre

Rod

Upper

1.875/7.5

3.750/7.5

5.625/7.5

7.500/7.5

Centre

Rod

Middle

1.75/7

3.50/7

5.25/7

7.00/7

Centre

Rod

Lower

1.75/7

3.50/7

5.25/7

7.00/7

Side

Rod

Upper

2.125/8.5

4.250/8.5

6.375/8.5

8.500/8.5

Side

Rod

Lower

2/8

4/8

6/8

8/8

Power Tilt

(top-to-

bottom)

1.00244

1.01566

0.98008

0.92736

*LP = Lattice Pitch

- 286 -

KAERI/RR-1999/99

Table 3.5-3

Comparison of Form Factor vs. ZCU Level

Average Zone

Controller

Level (%)

0

25

50

75

100

DUPIC Core

Radial Form

Factor

0.8173

0.8342

0.8385

0.8343

0.8006

Overall Form

Factor

0.5955

0.6004

0.6000

0.6039

0.5990

37-element NU Core

Radial Form

Factor

0.8022

0.8190

0.8242

0.8251

0.7996

Overall Form

Factor

0.5673

0.5702

0.5638

0.5491

0.5252

- 287 -

KAERI/RR-1999/99

Table 3.5-4

Power Perturbation Coefficients in DUPIC Core

*

1

2

3

4

5

6

7

8

9

10

11

12

13

14

Regional power % change per 1 mk perturbation of ZCU

1

41.06

7.20

9.09

2.38

-6.37

-7.15

-11.41

3.90

-3.81

-3.22

-7.28

-11.39

-11.53

-13.78

2

12.80

45.60

-5.67

3.25

9.04

-10.20

-6.25

-1.82

5.19

-10.72

-6.85

-2.95

-13.78

-10.60

3

9.66

-6.79

36.41

4.63

-9.15

9.04

-7.11

-3.99

-11.23

2.66

-7.24

-13.69

-3.84

-11.16

4

5.58

4.22

4.48

22.13

4.19

5.63

4.17

-6.94

-7.24

-7.37

-6.32

-7.53

-7.22

-7.19

5

-4.74

14.38

-8.45

6.47

38.47

-4.36

15.12

-9.74

-1.96

-12.88

-6.57

3.56

-10.38

-1.95

6

-7.20

-11.11

8.81

2.55

-6.18

41.22

7.28

-11.11

-13.80

-3.04

-7.11

-11.29

4.61

-3.57

7

-10.53

-6.23

-5.91

3.35

9.53

12.76

45.74

-13.06

-10.62

-10.78

-6.84

-2.98

-2.01

5.16

8

4.48

-3.73

-3.12

-6.90

-10.79

-11.37

-14.63

40.31

6.83

9.18

2.71

-6.64

-7.34

-10.80

9

-2.00

5.23

-10.26

-6.63

-3.09

-13.48

-11.12

12.45

44.31

-6.20

3.51

9.97

-10.75

-6.13

10

-3.86

-11.59

2.37

-6.91

-13.01

-3.96

-11.89

8.83

-6.93

37.83

4.86

-9.87

9.78

-6.68

11

-7.21

-7.49

-7.11

-6.07

-7.29

-7.13

-7.58

5.52

3.87

4.64

22.66

4.29

5.58

3.95

12

-10.28

-2.07

-12.22

-6.28

3.14

-10.07

-1.92

-4.27

14.44

-9.01

6.72

40.27

-4.90

13.91

13

-11.47

-14.22

-3.18

-6.95

-10.85

4.20

-3.92

-7.04

-11.01

9.68

2.69

-6.82

41.66

6.83

14

-13.72

-10.98

-10.32

-6.55

-2.95

-1.81

5.57

-10.03

-6.17

-6.02

3.56

9.59

12.95

44.64

* Perturbed zone number

- 288 -

KAERI/RR-1999/99

Table 3.5-5

Thermal Flux Perturbation Coefficients in DUPIC Core

*

1

2

3

4

5

6

7

8

9

10

11

12

13

14

Detector flux level % change per 1 mk perturbation of ZCU

1

69.17

8.61

11.43

6.01

-5.51

-5.75

-10.19

-1.24

-6.31

-5,94

-9.07

-12.37

-12.32

-14.24

2

15.54

69.17

-4.23

7.12

10.86

-8.94

-4.9

-4.8

-0.58

-11.62

-8.77

-5.97

-14.23

-11.62

3

11.02

-6.11

63.15

7.59

-8.47

10.47

-6.38

-6.45

-12.24

-1.84

-9.17

-14.51

-6.36

-12.15

4

7.17

3.69

9.33

39.96

4.12

7.15

3.67

-8.57

-9.42

-8.51

-8.84

-9.92

-8.86

-9.33

5

-3.72

15.75

-7.14

10.1

64.09

-3.37

16.43

-10.84

-5.04

-13.59

-8.69

-1.36

-11.53

-5

6

-5.73

-9.93

11.29

6.3

-4.92

69.91

8.69

-11.88

-14.24

-5.83

-8.96

-12.27

-0.83

-6.12

7

-9.29

-4.95

-4.47

7.21

11.23

15.51

68.42

-13.49

-11.6

-11.67

-8.74

-5.91

-4.99

-0.47

8

-0.93

-6.44

-5.82

-8.65

-11.77

-12.19

-15.17

68.6

8.12

11.75

6.61

-5.38

-5.9

-9.7

9

-4.98

-0.56

-11.12

-8.41

-5.89

-13.98

-12.21

15.17

66.57

-4.71

7.53

11.74

-9.51

-4.93

10

-6.38

-12.65

-1.97

-8.79

-13.87

-6.43

-12.99

10.24

-6.23

65.3

7.97

-9.22

11.13

-6.08

11

-8.86

-9.74

-8.22

-8.53

-9.6

-8.79

-9.89

7.01

3.33

9.68

40.82

4.17

7.18

3.36

12

-11.45

-5.24

-12.91

-8.31

-1.59

-11.24

-5.14

-3.33

15.64

-7.65

10.51

66.93

-3.9

15.18

13

-12.29

-14.72

-5.79

-8.68

-11.82

-1.04

-6.54

-5.68

-9.92

12.11

6.5

-5.56

70.18

8.12

14

-14.21

-12.04

-11.18

-8.38

-5.86

-4.84

-0.39

-8.82

-4.92

-4.49

7.63

11.46

15.74

67.75

* Perturbed zone number

- 289 -

KAERI/RR-1999/99

Table 3.5-6

Adjuster Bank Reactivity Insertion Characteristics for DUPIC Core

Configuration

No AdjusterZone Level 50%

Bank 7 being inserted

Bank 7 fully insertedBank 6 being inserted

Banks 7, 6 fully insertedBank 5 being inserted

Banks 7, 6, 5 fully insertedBank 4 being inserted

Banks 1, 6, 5, 4 fully insertedBank 3 being inserted

Banks 7, 6, 5, 4, 3 fully insertedBank 2 being inserted

Banks 7, 6, 5, 4, 3, 2 fully insertedBank 1 being inserted

Insertion*(cm)

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15171.45

171.4557.15

-57.15-171.45

k-eff

1.011041

1.0106811.0100831.0093421.008968

1.0086561.0081781.0076031.007319

1.0069881.0065211.0059741.005709

1.0055341.0051011.0045701.004314

1.0040511.0036201.0031741.002983

1.0027071.0022681.0018481.001654

1.0015071.0011861.0008611.000718

Bank Worth(mk)

0.350.941.662.03

0.310.781.341.62

0.330.791.331.59

0.170.601.131.38

0.260.691.131.32

0.270.711.131.32

0.150.470.790.93

* Distance of the bottom of the adjusters from the horizontal mid-plane

- 290 -

KAERI/RR-1999/99

Table 3.5-7

Adjuster Bank Reactivity Insertion Characteristics for Natural Uranium Core

Configuration

No AdjusterZone Level 50%

Bank 7 being inserted

Bank 7 fully insertedBank 6 being inserted

Banks 7, 6 fully insertedBanks 5 being inserted

Banks 7, 6, 5 fully insertedBank 4 being inserted

Banks 7, 6, 5, 4 fully insertedBank 3 being inserted

Banks 7, 6, 5, 4, 3 fully insertedBank 2 being inserted

Banks 7, 6, 5, 4, 3, 2 fully insertedBank 1 being inserted

Insertion*(cm)

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

171.4557.15

-57.15-171.45

k-eff

1.01815

1.017681.016691.015311.01457

1.014191.013461.012521.01202

1.011621.010901.009991.00951

1.009271.008611.007631.00707

1.006721.006071.005311.00492

1.004551.003911.003151.00278

1.002571.002081.001491.00118

Bank Worth(mk)

-

0.451.402.743.46

0.381.082.002.48

0.391.101.992.46

0.230.881.852.40

0.340.981.732.12

0.371.001.752.12

0.210.701.291.60

* Distance of the bottom of the adjusters from the horizontal mid-plane

- 291 -

KAERI/RR-1999/99

Table 3.5-8

Simulation of Startup after Short Shutdown for DUPIC Core

Power

Level

(%)

0.1

56

65

65

68

68

76

76

87

87

91

91

95

95

100

100

100

Adjuster

Banks

All BK out

BK 1-6 out

BK 1-5 out

BK 1-4 out

BK 1-3 out

BK 1-2 out

BK 1 out

All BK in

k-effective

1.00072

1.00072

1.00070

1.00073

1.00071

1.00072

1.00072

1.00072

1.00073

1.00071

1.00073

1.00072

1.00072

1.00072

1.00072

1.00081

1.00076

Xenon Time

Step

(min)

44.5

67.4

0.0

91.4

0.0

73.1

0.0

53.1

0.0

34.1

0.0

34.7

0.0

41.4

0.0

60.0

30.0

Average

Zone

Level (%)

50.0

70.0

19.9

70.0

34.8

66.2

35.0

66.1

40.1

67.2

43.6

67.1

44.8

67.3

51.9

78.2

90.0

Maximum

Channel

Power

(kW)

8.0

4781

4909

5241

5196

5297

5670

5673

6214

6247

6282

6252

6409

6315

6519

6485

6807

Maximum

Bundle

Power

(kW)

0.1

627

581

639

631

649

660

673

709

710

735

726

730

722

751

742

765

- 292 -

KAERI/RR-1999/99

Table 3.5-9

Simulation of Startup after Short Shutdown for Natural Uranium Core

Power

Level

0.1

56

65

65

ON

O

N00

00

76

76

87

87

91

91

95

95

100

100

100

100

Adjuster

Banks

All BK out

BK 1-6 out

BK 1-5 out

BK 1-4 out

BK 1-3 out

BK 1-2 out

BK 1 out

All BK in

k-effective

1.00118

1.00118

1.00117

1.00119

1.00115

1.00120

1.00116

1.00117

1.00120

1.00120

1.00118

1.00118

1.00118

1.00119

1.00120

1.00119

1.00119

1.00119

Xenon Time

Step

(min)

36.1

8.2

0.0

32.6

0.0

30.2

0.0

28.8

0.0

26.0

0.0

30.2

0.0

47.6

0.0

90.0

90.0

90.0

Average

Zone

Level (%)

50.0

62.9

9.0

69.4

24.0

66.3

23.7

65.1

26.2

66.2

32.8

66.7

34.2

66.8

43.1

75.3

78.9

76.5

Maximum

Channel

Power

(kW)

0.9

4845

5176

5546

5371

5625

6130

6082

6534

6643

6567

6559

6633

6498

6679

6538

6549

6563

Maximum

Bundle

Power

(kW)

0.1

698

663

764

713

783

796

836

853

896

856

873

825

823

822

829

831

821

- 293 -

KAERI/RR-1999/99

Table 3.5-10

Simulation of Startup after Poison-out Shutdown for DUPIC Core

Power

Level

0.1

0.1

0.1

0.1

56

65

65

68

68

76

76

87

87

91

91

95

95

Adjuster

Banks

All BK in

All BK out

BK 7 in

BK 6,7 in

BK 5-7 in

BK 4-7 in

BK 3-7 in

BK 2-7 in

k-effective

0.95316

0.96611

0.98726

1.00072

1.00072

1.00072

1.00073

1.00072

1.00072

1.00071

1.00071

1.00071

1.00071

1.00072

1.00072

1.00073

1.00072

Xenon Time

Step

(min)

540.0

540.0

540.0

125.4

4.2

0.0

7.7

0.0

6.4

0.0

6.1

0.0

5.0

0.0

5.0

0.0

5.2

Average

Zone

Level (%)

50.0

50.0

50.0

50.0

69.9

34.7

69.9

42.0

68.4

42.9

68.3

47.2

68.0

47.7

68.2

47.8

69.4

Maximum

Channel

Power

(kW)

6

6

6

4

4296

4623

4755

4778

4865

5274

5302

5835

5883

5990

5976

6222

6159

Maximum

Bundle

Power

(kW)

0.7

0.7

0.7

0.5

529

521

545

548

562

581

588

653

651

683

673

693

690

- 294 -

KAERI/RR-1999/99

Table 3.5-11

Simulation of Startup after Poison-out Shutdown for Natural Uranium Core

Power

Level

0.1

0.1

0.1

0.1

56

65

65

ON

O

N00

00

16

16

87

87

91

91

95

95

Adjuster

Banks

All BK in

All BK out

BK 7 in

BK 6,7 in

BK 5-7 in

BK 4-7 in

BK 3-7 in

BK 2-7 in

k-effective

0.91359

0.93595

0.96968

1.00119

1.00119

1.00118

1.00119

1.00118

1.00119

1.00119

1.00117

1.00118

1.00118

1.00117

1.00117

1.00118

1.00117

Xenon Time

Step

(min)

600.0

600.0

600.0

382.5

1.9

0.0

5.6

0.0

4.5

0.0

4.4

0.0

4.1

0.0

4.0

0.0

4.3

Average

Zone

Level (%)

50.0

50.0

50.0

50.0

69.6

22.9

70.8

33.9

68.8

34.5

67.8

37.9

68.1

40.2

67.9

40.1

68.1

Maximum

Channel

Power

(kW)

6

6

6

8

4637

4848

5076

5019

5160

5541

5599

6090

6184

6183

6223

6367

6342

Maximum

Bundle

Power

(kW)

0.7

0.7

0.7

1

689

589

643

620

651

679

710

713

751

748

760

732

745

- 295 -

KAERI/RR-1999/99

Table 3.5-12

Simulation of Adjuster Shim Operation for DUPIC Core

Power

Level

(%)

100

94

94

94

87

87

87

82

82

82

79

79

79

68

68

68

61

61

52

52

Adjuster

Banks

All BK in

BK 1 out

BK 1,2 out

BK 1-3 out

BK 1-4 out

BK 1-5 out

BK 1-6 out

All BK out

k-effective

1.00263

1.00264

1.00265

1.00376

1.00376

1.00376

1.00536

1.00535

1.00535

1.00692

1.00692

1.00693

1.00834

1.00833

1.00833

1.01054

1.01053

1.01261

1.01262

1.01525

Xenon Time

Step

(hrs)

steady state

0.0

4.0

steady state

0.0

4.0

steady state

0.0

4.0

steady state

0.0

4.0

steady state

0.0

4.0

steady state

0.0

steady state

0.0

steady state

Average

Zone

Level (%)

20.1

34.2

20.8

20.2

40.3

24.1

20.0

40.7

29.2

20.2

40.3

34.9

20.8

45.6

14.0

20.0

46.7

19.6

55.4

20.1

Maximum

Channel

Power

(kW)

6495

6224

6198

6180

5840

5830

5794

5653

5649

5606

5645

5669

5614

4961

4895

4854

4551

4443

4048

3924

Maximum

Bundle

Power

(kW)

754

700

697

699

683

684

679

644

646

642

634

641

631

577

576

561

520

504

492

467

- 296 -

KAERI/RR-1999/99

Table 3.5-13

Simulation of Adjuster Shim Operation for Natural Uranium Core

Power

Level

(%)

100

94

94

94

87

87

87

82

82

82

79

79

79

68

68

68

61

61

52

52

Adjuster

Banks

All BKin

BK 1 out

BK 1,2 out

BK 1-3 out

BK 1-4 out

BK 1-5 out

BK 1-6 out

All RK" r>ntA l l J3IS. OUI

k-effective

1.00328

1.00327

1.00328

1.00498

1.00497

1.00497

1.00721

1.00723

1.00721

1.00950

1.00952

1.00951

1.01199

1.01202

1.01204

1.01484

1.01482

1.01760

1.01762

1.02108

Xenon Time

Step

(hrs)

steady state

0.0

4.0

steady state

0.0

4.0

steady state

0.0

4.0

steady state

0.0

4.0

steady state

0.0

4.0

steady state

0.0

steady state

0.0

steady state

Average

Zone

Level (%)

21.0

42.4

29.3

20.8

49.5

33.9

21.2

50.3

40.1

20.5

53.4

50.3

18.9

54.8

20.3

17.7

57.6

15.9

76.3

20.8

Maximum

Channel

Power

(kW)

6594

6386

6377

6334

6018

6027

5979

5907

5919

5822

5980

6037

5919

5277

5205

5135

4874

4719

4506

4309

Maximum

Bundle

Power

(kW)

787

766

761

753

756

763

746

735

744

708

793

814

769

702

688

677

651

619

662

619

- 297 -

KAERI/RR-1999/99

Table 3.5-14

Simulation of Stepback to 60% Full Power for DUPIC Core

Adjuster

Bank

Position

All BK in

BK 1 out

BK 1,2 out

BK 1-3 out

BK 1-4 out

BK 1-3 out

BK 1,2 out

BK 1 out

All BK in

k-effective

1.00072

1.00072

1.00072

1.00071

1.00071

1.00072

1.00072

1.00072

1.00072

1.00073

1.00072

1.00073

1.00072

1.00072

1.00072

1.00073

Xenon Time

Step

(min)

22.0

0.0

12.3

0.0

21.3

0.0

27.9

0.0

600.0

0.0

79.3

0.0

227.9

0.0

316.9

0.0

Average Zone

Level

(%)

20.0

34.0

20.0

40.0

19.6

39.8

19.7

39.9

67.0

42.8

67.7

46.1

68.1

47.8

69.6

54.9

Maximum

Channel

Power

(kW)

3884

3963

3953

4016

3987

4122

4103

4295

4364

4196

4179

4044

3997

3979

3924

3847

Maximum

Bundle

Power

(kW)

451

446

452

468

467

469

470

483

506

All

All

All

458

449

446

454

- 298 -

KAERI/RR-1999/99

Table 3.5-15

Simulation of Stepback to 60% Full Power for Natural Uranium Core

Adjuster Bank

Position

All BKin

BK 1 out

BK 1,2 out

BK 1-3 out

BK 1-4 out

BK 1-3 out

BK 1,2 out

BK 1 out

All BK in

k-effective

1.00118

1.00117

1.00118

1.00118

1.00118

1.00118

1.00121

1.00119

1.00118

1.00118

1.00118

1.00117

1.00118

1.00119

1.00118

1.00117

Xenon Time

Step

(min)

12.9

0.0

11.0

0.0

19.1

0.0

29.6

0.0

360.0

0.0

216.0

0.0

325.8

0.0

523.2

0.0

Average Zone

Level

(%)

20.1

40.9

19.6

47.9

19.3

47.8

19.0

51.4

71.9

32.3

64.8

33.6

68.9

38.8

68.5

46.6

Maximum

Channel

Power

(kW)

3942

4064

4031

4136

4122

4312

4303

4572

4706

4429

4399

4184

4129

4059

4036

3926

Maximum

Bundle

Power

(kW)

467

484

475

516

512

533

522

601

650

567

566

527

521

485

493

475

- 299 -

KAERI/RR-1999/99

Table 3.5-16

Comparison of SOR Static Reactivity Worth

Case

Reference

(All ADJs in)

28 SORs

Inserted

26 SORs

Inserted

DUPIC Core

keff

1.000720

0.933067

0.961706

Radial

Form

Factor

0.8392

0.6593

0.2369

Overall

Form

Factor

0.6004

0.3426

0.0758

Worth

(mk)

-72.5

-40.5

Natural Uranium Core

keff

1.001014

0.920887

0.947409

Radial

Form

Factor

0.8219

0.6614

0.2729

Overall

Form

Factor

0.5579

0.5055

0.0966

Worth

(mk)

-86.9

-56.5

- 300 -

KAERI/RR-1999/99

Table 3.5-17

Comparison of SOR Insertion Characteristics

Position of

SOR

Outside

Reactor

+ 9 LP *

+ 5 LP

+ 1 LP

- 5 LP

- 7 LP

- 9 LP

DUPIC Core

keff

1.000720

1.000446

0.998254

0.994759

0.976688

0.956873

0.933079

Radial

Form

Factor

0.839

0.724

0.552

0.410

0.228

0.205

0.659

Overall

Form

Factor

0.600

0.533

0.410

0.294

0.150

0.138

0.342

Worth

(mk)

-0.3

-2.5

-6.0

-24.6

-45.8

-72.4

37-Element Nu Core

keff

1.001014

0.999670

0.996988

0.992792

0.971202

0.948245

0.920877

Radial

Form

Factor

0.822

0.726

0.554

0.412

0.227

0.208

0.661

Overall

Form

Factor

0.558

0.482

0.366

0.267

0.142

0.134

0.505

Worth

(mk)

-1.3

-4.0

-8.3

-30.7

-55.6

-86.9

* +9/-9 LP means that the bottom of the shutoff rod is 10 lattice pitches

above/below the centre line

- 301 -

KAERI/RR-1999/99

Table 3.5-18

Damping Factors for Xenon Oscillation

Oscillation Type

Top-to-Bottom

Side-to-Side

Front-to-Back

DUPIC Core

-1.767X10"2

-1.594X10"2

-2.517 X10"2

Natural Uranium

-3.244 X10"2

-2.635 X10"2

-1.198X10-'

- 302 -

KAERI/RR-1999/99

Table 3.5-19

Damping Factors of DUPIC Fuel Core for Different Power Levels

Power Level (%)

100

80

60

40

20

10

Damping Factor* (h/1)

-1.767X10"2

-2.336 X10"2

-3.085 X10"2

-4.391 X 10"2

-1.212 X10'2

-

* Top-to-bottom oscillation

- 303 -

KAERI/RR-1999/99

Table 3.5-20

Damping Factors of DUPIC Fuel Core for Various Refueling Schemes

Fueling Scheme

2-Bundle Shift

4-Bundle Shift

6-Bundle Shift*

8-Bundle Shift*

Axial Form Facror

0.716

0.695

0.557

0.540

Damping Factor (hr"1)

-1.767X10"2

-4.121 X10"2

-

-

* Actually 6- and 8-bundle shift refueling schemes are not practicable for DUPIC

core.

- 304 -

KAERI/RR-1999/99

— 5

s

DUPIC

3 7 - E l e m e n t NU

20 40 60 80

A v e r a g e Z o n e C o n t r o l l e r L e v e l (%)

1 00

Fig. 3.5-1 Comparison of ZCU Static Reactivity Worth

- 305 -

KAERI/RR-1999/99

6 8Time (hr)

DUPIC Core

Top-To-Bottom

Side-To-Side

Front-To-Back

Natural Uranium-

Core

10 12 14

Fig. 3.5-2 Power Tilts after Refueling Transient

- 306 -

KAERI/RR-1999/99

18

16

14

3T£ 12

co

Ic

10

3 7 - E l e m e n t N a t u r a l U r a n i u m F u e l C o r e ,I n c r e a s e o f X e n o n L o a d 30 M i n u t e sa f t e r s h u t d o w n = 1 3 . 5 1 7 m k

D U P I C c o r e ,I n c r e a s e o f X e n o n L o a d 30 M i n u t e sa f t e r s h u t d o w n = 6 . 8 0 3 m k

9 12 15 18 . 21 24 27

Tim e A f t e r S h u t d o w n ( m i n )

30 33 36

Fig. 3.5-3 Comparison of Xenon Load at 30-min after Shutdown

- 307 -

KAERI/RR-1999/99

2 3.0

o 2.0

5 1.5m 10

o.s

—•—37-BementNU

Bank #7 Being Inserted , / #

/ • • • * " "

25

20

1.5

1.0

Q5

0.0

Bank#3 Being Inserted

Bank#4,5,6,7Areadylnserted

.-••. - •

400 300 200 100 -10D -200 -300

3.0

2.S

2.0

1.5

1.0

0.5

Bank #6 Being InsettedBank #7 Already Inserted

400 300 200 100 0 -100 -200 -300400 303 200 100 -100 -200 -300

Bank 05 Being InsertedBank # 6.7 Already Inserted

Distance Relative To Centre Line (cm) 30O 20O 100 0 -10O -200 -300

Distance Relative To Centre Line (cm)

100 -100 -200 -300

Fig. 3.5-4 Comparison of ADJ Bank Insertion Characteristics

- 308 -

KAERI/RR-1999/99

co(1)X

-20

U -40-

Natural Uranium

0 10 20 30 40 50 60 70

-80-

-100-

-120

Time after Shutdown (hrs)

Fig. 3.5-5 Xenon Buildup after Shutdown

- 309 -

-2

E

•5

CO

&

oICO

-8

-10

-12

KAERI/RR-1999/99

DUPICFuel

• Natural Uranium

400 300 200 100 0 -100 -200 -300 -40

Insertion Depth (cm)

Fig. 3.5-6 Static Reactivity Worth of MCA

- 310 -

KAERI/RR-1999/99

u •

-10

-20

mk)

j-ao-os

t>S-50UL

:ic

%*>

-70

-80

-9D

DJRC

37-BenBrtNU

Cfertne

-X••X

\ \1 \\\\\\\V \

\ \

\ \

' \\ \

jne \ x

10 5 0 -5 -1

Position a SORFrom Centre Line (Lattice Pftch)

Fig. 3.5-7 Comparison of Static Reactivity Worth Insertion Characteristics

- 311 -

KAERI/RR-1999/99

N 1.5 -

0.0 0.5 1.0 1.5 2.0

Time (seconds)

3.0 3.5

Fig. 3.5-8 Reactor Power for 20% RIH Break LOCA Shutdown by SDS1

- 312 -

KAERI/RR-1999/99

-1000.0 1.0 1.5 2.0

Time (seconds)

2.5 3.0 3.5

Fig. 3.5-9 Dynamic Reactivity for 20% RIH Break LOCA Shutdown by SDS1

- 313 -

KAERI/RR-1999/99

2.5-

•2 0 -

a. 1.5-

rmal

ized

d

2 0.5-

0.0-

A• fM

./I/ 1• / \7 L

SDS2 RIH 100% BreakNAT-U Trip at 0.3390 s

Inj Starts at 0.9228 s

• DUPIC Trip at 0.3420 sInj Starts at 0.92S8 s

y

•" * • • • • • •

1 | 1 | < | > | • 1 ' 1 ' 1

-0.5 0.0 0.5 1.0 1.5 2.0 2.5

Time After Break (sec)

3.0 3.5

Fig. 3.5-10 Reactor Power for 100% RIH LOCA Shutdown by SDS2

- 314 -

KAERI/RR-1999/99

20

0-

-20-

CO

a:o£CO

-60-

-80-

-100

*********

SDS2RIH100% Break

NAT-U Trip at 0.3390 s

Inj Starts at 0.9228 s

DUPIC Trip at 0.3420 s

Inj Starts at 0.92SS s

-0.5 0.0 0.5 1.0 1.5 2.0 2.5

Time After Break (sec)

3.0 3.5

Fig. 3.5-11 Dynamic Reactivity for 100% RJH LOCA Shutdown by SDS2

- 315 -

KAERI/RR-1999/99

•oCDO

coc0)X

Power Level

Before Shutdown

(DUPIC Core)

100%

80%60%40%

20%

37-Element Natural Uranium Core

Shutdown from 100% Power

10 20 30 40 50 60

Time after Reactor Shutdown (hr)

Fig. 3.5-12 Xenon Load after Reactor Shutdown

- 316 -

KAERI/RR-1999/99

37-Element Natural Uranium

fe\\ Core, Startup to 100% Power

-300 20 30 40 50

Time after Startup (hr)

100%

80%

60%

40%

20%

Fig. 3.5-13 Xenon Load after Reactor Startup

- 317 -

KAERI/RR-1999/99

0

-20-1

Equilibrium Poisoning at Full

/Power = 28.989 mk

•oroo_icoc<DX

- 80%

- 60%

• 40%

- 20%

- 0%

37-Element Natural Uranium Core, _

Reduce Power Level to 0%

10 20 30 40 50 60

Time after Power Reduction (hr)

70

Fig. 3.5-14 Xenon Load after Power Setback from Full Power

- 318 -

KAERI/RR-1999/99

oP-.

50 100 150Time (hr)

200 250

Fig. 3.5-15 Comparison of Top-to-Bottom Tilt

- 319 -

KAERI/RR-1999/99

50-

0

DUPIC

Natural uranium

50 100 150Time (hr)

200 250

Fig. 3.5-16 Comparison of Side-to-Side Tilt

- 320 -

KAERI/RR-1999/99

oOH

0 50 100 150Time (hr)

200 250

Fig. 3.5-17 Comparison of Front-to-Back Tilt

- 321 -

KAERI/RR-1999/99

50

ilt

(%)

oOH

40-

30-

20-

10.

o.-10.

-20.

-30-

-40

-5050 100 150

Time (hr)

200 250

Fig. 3.5-18 Comparison of Top-to-Bottom Oscillation with Different Power Levels

for DUPIC Core

- 322 -

KAERI/RR-1999/99

1100

O

CO

1000-

900'

800'

700'

600'

500^

400.

300^

200^

100.

o

Fueling direction

0

_. - - V i,

2-bundle shift

4-bundle shift

6-bundle shift

8-bundle shift

4 6 8 10

Axial bundle position

12

Fig. 3.5-19 Axial Power Shape of Central Channel for Various Refueling Schemes

- 323 -

KAERI/RR-1999/99

CD

oOH

50

40-

30-

20-

1.0.

0.

-10-

-20-

-30-

-40

-50

A

\ .1XI

0

• * * •w

50

2-bundle shift

4-bundle shift

• 6-bundIe shift

8-bundle shift

100 150

Time (hr)

200 250

Fig. 3.5-20 Comparison of Front-to-Back Tilt for Various Refueling Schemesof DUPIC Core

- 324 -

KAERI/RR-1999/99

Uncontrolled

ZCU controlled

40 60

Time (hr)

80 100

Fig. 3.5-21 ZCU Controllability of Top-to-Bottom Oscillation of DUPIC Core

- 325 -

KAERI/RR-1999/99

50

40-

30^

20.

-10.

-20^

-3<

-40

-500 20

Uncontrolled

• ZCU controlled

40 60

Time (hr)

80 100

Fig. 3.5-22 ZCU Controllability of Side-to-Side Oscillation of DUPIC Core

- 326 -

KAERI/RR-1999/99

50

40-

30-

20-

10.

0

-10-

-20-

-30.

-40.

-500

• Uncontrolled

• ZCU controlled

20 40 60Time (hr)

80 100

Fig. 3.5-23 ZCU Controllability of Front-to-Back Oscillation of DUPIC Core

- 327 -

KAERI/RR-1999/99

3.6

CANDU

3. \

ROP) H ^

Jfl*fl

. ROP

K CANDU

it^l(Regional Overpower Protection;

27fl7r

ROP

. CANDU ROP

CANDU

ROP

DUPIC ^: ROP

fg. DUPIC

-fe channel-random J£-fe- common-random

- 43-g-

DUPIC . 2327fl -gTfl 7]

- 328 -

KAERI/RR-1999/99

3.6.1 ROP

3.6.1.1 M.

ripple

5]7] . ROP

(TSP)

TSP

Do

I"CPRL

jofl

jp7> ^

100% #^^1^-1 TJl^-7] 3L7]

k, ripple qofl cflt> 71) -71 j p 5 |

k, ripple q«Hl *%# 711^7] J5]

ripple q g # ^

k, ripple qo||

-. <±, o]

711

(3.6-1)

(3.6-2)

3.6.1.2 ROP

4 ROP Tfl^f^ 20~507H^ icvjj 711^-715. , 4

fe ' t ^ ^ S . , SDS2 ROP

, SDSI ROP

- 329 -

KAERI/RR-1999/99

271151

3.6.2 ROP

ROP A^L-%: *1*B RFSP [Ref. 7]

^ NUCIRC [Ref. 22] 3H7J- A ^ ^ u f . i i l - s |

ROVER-F [Ref. 23]

3.6.2.1

ROP 7ll# ^TllcH) Af^-sj 23271)^

. o] 7«

3.6.2.2

DUPIC^ S , DUPIC

91 orifice7f ^

Bfl*l£M 517] nB^-^1, ^B^ -fr^ ^:S7f DUPIC

DUPIC

AECLS] Xc-Lb ^ ^ ) A ] # Af-g-^f^uf. ©1 ^ 1 ^ ^ : 37-^-

- 330 -

KAERI/RR-1999/99

37-g-

3.6.2.3 711 -7]

(lead cabled

3.6.2.4 Ripple

^ rippM

ROP 7}]471TT rippleo]

. NUCIRC

fe INTREP7 3 J = . #

ROVER-F

CANDU

ripple"^ ^g^t>t:f. ROP

100%

rippled 600 FPD

3.6.2.5

(Detector-random error):

- 331 -

KAERI/RR-1999/99

(Channel-random error):

icff

(Common-random error):

o] J2.*).i=,

(Systematic error):

3.6-H1

(root-sum-square)

3.6.3 S *

I7l|5J

DUPIC

M: tcfii]- reform

>. off

3.6.3.1

2327H ^7^1 71

98%

DUPIC

25711

J}, DUPIC

< > 1 ^ ^ < g, CANDU-6 Q*

3.6-2^ ^ B

- 332 -

KAERI/RR-1999/99

3.6.3.2 1711 711^71 ^ £*J

. DUPIC

n SDS1

SDS2 7^1^-7]6flA^ E^} ^ ^ ^ 7 1 - ijcfl 11%

3.6.3.3 REFORM

Reform |>L> ; 7 l § # ^ ^ S # H H 1 ^ ROP

reform 7^*>.^*fe ^3.^-ti 23<Hl ^ W

DUPIC i^^<Hl t iN reform .7^]A1:# ^ * ^ M ROP

9ic}. Reform 7 ^ ^3f, S^j ^ ^ * ] f e 125.7%S

- 333 -

KAERI/RR-1999/99

Table 3.6-1

Estimated ROP Errors and Uncertainties for DUPIC Core (90% Confidence)

Source of Errors

1. Detector-Related ErrorsTrip SetpointBuffer AmplifierDynamic Compensation

2. Flux-Shape ErrorsSimulation ErrorChange due to BoilingLead-Cable ContributionsOff-Nominal Core

3. CCP ErrorsCHF Correlation ErrorsIncomplete InstrumentationNUCIRC Pressure Loss TermUncertainty In HTS Boundary ConditionsChange in Ref. HTS Boundary

ConditionsHTS variationsChannel Age CorrectionCCP ChangeNormal Operating Flux TiltDifferent Fuel TypeAllowance for PT Creep

4. Calibration ErrorsCP/CPPF CalculationThermal Power CalculationCPPF Drift ErrorCalibration Drift Error

Total

Estimated Magnitude (%)

DetectorRandom

± 0.18± 0.10± 0.70

+ 1.94

± 0.20

± 1.60

± 2.60

ChannelRandom

± 1.09

+ 0.89

± 0.16± 0.11

+ 0.93

+ 1.97

CommonRandom

+ 0.14± 0.10

± 1.08

± 0.10± 0.80

± 1.70

± 0.83± 2.32

± 0.15

± 1.57± 1.70± 0.80

+ 4.18

BiasError

~0

+ 0.10- 0.20- 0.20

0.66-0.20-2.20

3.18

-0.20

-LOO

0.20

+0.14

- 334 -

KAERI/RR-1999/99

Table 3.6-2

Trip Confidence for DUPIC Fuel Core with ROP Setpoint of 125% (25 worst cases)

49

42

112

44

39

108

46

51

114

123

115

110

126

121

129

122

120

38

50

128

130

53

60

45

127

Required

Case

D12C80

D05C80

MCAN1H

D07C80

D02C80

MCAC1H

D09C80

D14C80

MCAN2H

ZTSESF

MCAN2F

MCAC2H

ZTT045

ZT1ABT

ZTT315

ZTSFSE

ZT1ATB

D01C80

D13C80

ZTT225

ZT2A01

D02N50

D09N50

DO8C8O

ZTT135

ROP Setpoint

CPRL

1.287278

1.287262

1.172954

1.287740

1.293773

1.290789

1.285568

1.296551

1.102006

1.255475

1.148860

1.168160

1.265897

1.271510

1.258256

1.257559

1.283709

1.315854

1.318809

1.259769

1.313649

1.337646

1.331552

1.321406

1.267906

for 98% Trip

SDS1

.9984

.9966

.9993

.9982

.9996

.9993

.9968

.9965

.9951

.9976

.9954

.9972

.9992

.9957

.9959

.9974

.9997

.9974

.9977

.9968

.9990

.9995

.9988

.9974

.9990

Confidence =

SDS2

.9749

.9751

.9785

.9859

.9900

.9908

.9918

.9919

.9944

.9952

.9997

.9955

.9957

.9993

.9986

.9962

.9965

.9967

.9968

.9978

.9969

.9972

.9972

.9986

.9975

1.2336

Limiting

Detectors

9E

4D

11D

9F

2E

2E

6D

10E

11D

ID

2E

2E

ID

10F

IF

HE

10F

5F

9F

HE

ID

2E

6D

5E

7F

8H

7J

7H

4H

2G

8J

3H

4G

7H

2G

8J

7H

8J

8H

6H

1G

8H

2H

4H

6G

2G

3J

3H

6G

7G

- 335 -

KAERI/RR-1999/99

Table 3.6-3

Setpoints for Single Detector Failure

SDS1 Detector

Failed Detector

ID2D3D4D5D6D7D8D9D10D11D12D

IE2E3E4E5E6E7E8E9E10E

HE

IF2F3F4F5F6F7F8F9F10F11F

RequiredSetpoint

1.23361.23361.23361.23361.23361.23361.23361.23361.23361.23361.22541.2336

1.21591.23361.23361.23361.23361.23361.23361.23361.23361.23361.2336

1.23361.23361.23361.23361.23361.23361.23361.23361.23361.23361.2336

SDS2 Detector

Failed Detector

1G2G3G4G5G6G7G8G

1H2H3H4H5H6H7H8H

U2J3J4J5J6J7J8J

RequiredSetpoint

1.23361.22261.23341.23321.23361.23361.21001.2202

1.23321.23121.21651.21111.23231.23361.12041.2002

1.23331.23261.22151.16891.23361.23361.19471.1960

- 336 -

KAERI/RR-1999/99

3.7 DUPIC l ^ ^ ^

DUPIC 3J*}3/g£ ^ £ 3 . 3 : ^ # 4*l*k2. 7\& CANDU-6

, DUPIC ^«iS7]- # S 3 CANDU-6

CANDU-6 # # ^ 1 t> £ #

DUPIC «|&S.7]- ^ " ^ ^ CANDU-6 SJ

DUPIC

$7}*}%^}.

3.7.1 7 ] ^

, DUPIC

^ t - y 71)A>

91 £# ^ ^ ^I^Afife fl 91

3.7.1.1

CANDU 4l^Sofl-Hfe- 4 ^ ^ ^r^^l 4

K DUPIC

7H

- 337 -

KAERI/RR-1999/99

DUPIC^ 3.7- ZL^ 3.7-2

44

(fueling zone)ofl

6621 kW

765 M-3, M-5, M-ll

MTHM)

ZL^ 3.7-5^1

371.4 5J 350.7

358.3 MWh/kgHM (= 14929.3 MWd/

(dwell time>gr

D U P I C u - A l ^

4 4 1.6%if 7.5%

3.6%

3.7.1.2

& RFSP 3^$] lt

S c | ^ . D U P I C

>. 30711^

- 338 -

KAERI/RR-1999/99

3J-26\]M JiL-fe ti^ ^ o ] , 6-T-m g 8-cH*

7300 kWif 935

^ DUPIC

W, DUPIC

3.7.1.3 A]^> ^ ^ Û^^. 3.*} (Time-Dependent Refueling Simulation)

600 # # ^ (FPD) J§-<>}°) ^ H v ^ S^Vl- ^-n 7\& DUPIC

CANDU ^ > S . ^ 5 | ^ ^ A4

., DUPIC «?«1S i ^ ^ ^ ^g^- MCP Rj MBP

^ o . ^ , Jji i Â]^ ^§5- CPPFfe

xtfl, DUPIC i i ^ ^ : *ff-ofl 47fl£l

*£ DUPIC iiôfl^fe 27B, &<&$-Bfe Jtôfl^fe 87flo}7]DUPIC i

# ^ ^ r 7024 kWolnf, ^ § 5 MCP^ 6844 kW<>|i;K 5E$t,

^r 827 kWojnf, JS.A> 7}]^ %-<&£l $^:&-& 804 kWo]r:>.

ripple^ <£-§•

^ D U P I C

3]cl| rippled l.lOo]^ Jg^^l fe 1.065o]t V- ZL^ 3.7-9^1 7 ^

MCP ^ MBP-fe- <di^\ 7} *]]?>*] Jit:]- ^£. ^ 5 . ^

j- %^£\ <*•]%} ^-^ 7]^-g- A J ' S J J ^ ] 4|$]-o^> ^ I^7f îltl^l-^l 95%-§-

(administrative limit).5L -^^]*}^t:}. Û1^ nj ~^}^ i

- 339 -

KAERI/RR-1999/99

4 4 6935 kW J- 888 kW°lt:}. ^ ^ # ^*H, CPPF *

4 4 1.103J- 0.2/0.83. ^smsUrt -

600 FPDS} zfl#*l £ 4

DUPIC ^ \ ^f , 1 ^ ^ W

DUPIC i n ^ ^ ^ ^ CPPF7}

^^f^: % # ^ DUPICDUPIC i-ys} Afl # ^ o l 71^^1^-f-Bl cj

%-i-S DUPIC i t^î ^-foll c-1 3L7fl v^yjc}. it(-sH, CPPFzcu ^ H S a ^ l ^ $a^ DUPIC

3.7.2 ^^g^r^ «0ôfl 31 «>

7]-g- DUPIC Jn^Sj A^-^ ^ ^ r (performance parameter) #^g

}. CANDU i ^ } Sfl^ 3 .ô] RFSP S H -

^V8-*W r-fi- i ^ ^ A ^^ ^ ^ r (key core

performance parameter) # ^ # % ^ £ # 7 ) 1 ^ } & I ; } . Afl 7}xl

4 ^ r J 71$ DUPIC

cfl V ^ ^ ig^- ^}S. (composition variation data)#

3.7.2.1 CANDU

CANDU

71 $\n ZCU

- 340 -

KAERI/RR-1999/99

zcu H ^ $ }

7}.

CANDU Q} H 1 ^

Z C U

RS] ^4f nl^J-s. (constrained

sensitivity) §fe (a,s)<Hl cU«> i?^| ^ - ^ ^ ^ S - (unconstrained sensitivity) «J ZCU

zcu

RFSP S ^ # -f *H ZCU

<H ZCU

ZCU

(3.7-2)

(3.7-3)

(3.7-4)

147fl ZCU

- 341 -

ZCU

KAERI/RR-1999/99

zcu

(3.7-5)

(3-7-6)

(3.7-7)

ZCU

ZCU ZCU

-^P . -Cf l r ,^ - ) , ; = 1 , 2 , - , 1 4 (3.7-8)

(3.7-9)

zcu

ZCU

- 342 -

KAERI/RR-1999/99

-I- -I- QPi z" — —' ' > ° zu*~> a —

(3.7-10)

SA4 •$ z l I C^"u §^2 1 1 Q^H ft 14

i, " I TOjjOj T ' "T OZ|]O,,

ef. ZCU

RFSP ZCU

(user-specified limit)

H (3.7-5)) £ 71^

ZCU

(3.7-10)) S # (constraint)^

Q =Wl (SP2[Xl

p")2sp

a")

(3.7-n)

Lagrange

(3.7-11)31

o l # 0o]

(stationary condition)^ *,.ofl t:D*>

- 343 -

KAERI/RR-1999/99

a1NxN = bx

= b2(3.7-12)

n p

wkS2.S2.- (3.7-13)

N (3.7-14)

ZCU fe ^ (3.7-4)3} ^ (3.7-12)#

*, S*,.,

z c u

, 147fl

3.7.2.2

ZCU

. RFSP 3 H f UJ-^-^-S. *> GPT

MCP, MBP,

7f.

- 344 -

KAERI/RR-1999/99

Sa=~a <KFM

ZCU

CANDU ^ J ^ S f e 147H5] ^ I ^ A S i - H ^ H , 4

1"&>7) ^

(linear functional)^.

o]5>

- 345 -

(3.7-15)

, t o > i - Pi<8H,<f>0>c- ^.{SM- A0SF)</>0) (3.7-17)

~~ (3J"18)

KAERI/RR-1999/99

zcu

tf.

ZL^ 3.7-lOofl

RFSP

TIME-AVER g ^<&3.

GENOVA [Ref.

I PERTXS«

(DUPIC

SIMULATE

CANDU

, RFSP 3 =

CALCON SJE:

DUPIC CANDU

MCP,

MBP,

3.7-5 ^

^ RFSP

Sat}. ZIB] vf, MCP ^ MBP7}

MBP7f

3.7-7

- 346 -

KAERI/RR-1999/99

2. ZCU

100, 200, 300, 400, 500, 600 FPD -§• ^

}. JE 3.7-9 9J 3.7-lOoiH ^ f e

MCP, MBP, CPPF.S}

K MCP

3.7.2.3

DUPIC

DUPIC

Sit:}.

V = SJfS'

3.7-9

MCP g| MBP5]

t nH,3.7))

91

(3.7-19)

- 347 -

KAERI/RR-1999/99

7\.

DUPIC ^<&g. 3V$ «15-^S.# %4>*]7]7] # « M 4 1 ^ 4

DUPIC

*> DUPIC

S 3.7-11, 3.7-12, 3.7-

95% >*d^SiI S ^ S * } (2

600 ^ ^ H } ^ JDUPIC n&

M fi 3.7-14, 3.7-15,

f. DUPIC «}^5. »^9i # 1 ^ tfl*H, MCP, MBP, CPPF

1-3, 2.5, 1.2%3. ^

U-235

-& <$ 1.0%

DUPIC tq&S. ^91 #2o\] tfl*> MCP, MBP, CPPFS] # % ^ S . f e 6.6, 11.3, 7.5%o]

n?, DUPIC ^ ^ S . ^*> #3o1] c ^ 1-%-ySfe- ^^> A o ^ ^ # % ^ £ ( S 3.7-13

10.3, 15.0, 9.4%S &7}#-c}. M] 7}*} DUPIC

- 348 -

KAERI/RR-1999/99

3.7.3

3.7}

(agglomerative hierarchical

clustering, AHC [Ref. 29])o]

3.7.3.1

. o] S . 1 ^ A ^ ZL^ 3.7-

(clustering group)

7}.

, 10, 15, 20, 25, 30<>j ^.ei

, RFSP I H f ^ * I S^Hi ^r 5ii^ ^ ^ S %EH^ M r:fl 7 j | ^ ^ 4

357f|

- 349 -

KAERI/RR-1999/99

(strategy)^

- 4^ ^.^g (neutronic property)^-

50% ^ ) s L ZCU g ^ ^

^ ^ S . ^ B U ^ 7 | |^7> 107fl

DUPIC ^<&£- *&*±o\] cR> MCPS MBP, CPPFfe

DUPIC

107H

307}*] *gEU<SJ ^ ^ S # H l ^ ^ S SB^^I 4-§-^7 |5 . ^^*}^lt:f. 4 4 ^ DUPIC

ZL^ 3.7-12, 3.7-13, 3.7-14^

- 350 -

KAERI/RR-1999/99

3.7-176)1 MCP,

MBP, CPPF#

^ MCP#

MCP, MBP,

, DUPIC

average)

4 307fl)

, 4 (group-

, 4

, 44^

^ 1007M £ 4

(group-average) ^ ^ S . ^ - ^ # ^f-§-t! ±M 2.*}

i%, c>«J- # ^ , CPPF51 *W-§. ^A>*>SiJl, DUPIC

ZL^ 3.7-15, 3.7-16, 3.7-17ofl ^KHSd^K OL *}°}^ 95%

3.7-18«Hl 4 7\x\ DUPIC

>. DUPIC ^ ^ 5 . ^ - ^

fe 1.20, 0.89, 1.20%£ 1+E^o.t^, DUPIC

DUPIC ^ 9 l S *&<& #1*)

44

#2

- 351 -

KAERI/RR-1999/99

3.7.3.2 ^ ^

600 ^ # ^ < a (FPD)

-§•<& *Hif- ^ £ S . ^Hv^i HS3.13 (auto-refueling program)

7}

7\. f%<£S. *Vi%& £ 4

£4£4 1^-^ MCP, MBP^r H ^ 3.7-18

fe ] y # <t ^ ol^cf. CPPF^b ZL^ 3.7-20

(0.34%) #7>*fe ^ ^ S .

-i-^-^S.7} S % ^ ^ , MCP, MBP,

o.5O, 0.69, o.8O%5. ofl-a

U-235

DUPIC «J«iS ^ ^ #2

DUPIC ^ ^ S ^91 #13} 7\$\ %-<£*\^\. DUPIC n<&£. ^91 #2*]} cH*M, * ^

^ ^<^S 2 ^ A S 1*1 #%^£7V S^-5]^, MCP, MBP, CPPF5] - i^-^sfe

4 4 0.58, 1.34, 0.73%olc}. DUPIC 1%<&3. *£9± #3<Hl cfl«> MCP, MBP, CPPFS]

1-^-ys.fe- 5 ^ 3 : ^ ^ 5 - ^ ^ - ^ S ^ ^ l - % ^ £ # S # * M 4 4 0.47, 1.45,

0.52%o]c>. a]^- -L ^ } 7 } DUPIC

(reactor regulating system, RRS)^|

- 352 -

KAERI/RR-1999/99

CANDU ^ f 3 # ^ <M § £nfl, ZCU <^-fe

ZCU

^XI ^ S l ^r Sa K ^ S H , CANDU

3.7.4

DUPIC ^ ^ ^ 5 . #*L€ CANDU-6)^.0^xi 7 ] ^ D U P I C

35L4 X l - i ^ ^^*l -55l t :> . TjfXV ^ 3 4 , £]cfl ^ ^ ^ t : ] - ^

CPPF ^ g .

4 7M DUPIC

^ r<>l] tfl*i ^ ^ ^ i ^ y j ^ ^ S <g*oKi-

#0)7)

, MCP, MBP, CPPF^ 1 - ^ S T T 4 4

1.3, 2.5, 1.2%S. Tfl^&c)-. ^ ^ ^ ^ U^ofl^fe aJ:-S-U^r4 cfl*i

- 353 -

KAERI/RR-1999/99

Sfe 307M

DUPIC n<&3. *y- >oU tH^> w ] ^ ^ H

S. S-^-i- ^ f e ^>6a ^BD «a^- 91 307W

>Si^. ^ 1 ^ 1 4 , MCP, MBP, CPPFofl

0.6, 1.5, 0.8% J i t} -Ti) v}Ef^t:}. rc}eM, a l ^ ^ s

r]- tS->c|E>3t, MCP 9J MBP^ ^r^i ^|*>^1*1 7300 kW ^J 935 kWJicl- ^^1 c|

4 1 } K f e 1-R-fe DUPICCANDU ]

, DUPIC

- 354 -

KAERI/RR-1999/99

Table 3.7-1

Characteristics of DUPIC Core vs. Refueling Scheme

Peak Channel

Power (kW)

Peak Bundle

Power (kW)

Form Factor

(Avg./Max.)

Discharge Burnup

(MWhr/Bundle)

Refueling Rate

(Channels/Day)

Inner

Outer

Inner

Outer

Radial

Axial

Whole

Inner

Outer

Whole

Inner

Outer

Whole

2-Bundle

Shift

6621

6621

766

748

0.82

0.72

0.59

6630

6260

6395

1.48

2.57

4.06

4-Bundle

Shift

6627

6627

797

784

0.82

0.70

0.57

6853

6250

6372

0.74

1.29

2.03

6-Bundle

Shift

6671

6644

880

900

0.81

0.62

0.50

6656

6251

6328

0.51

0.85

1.36

8-Bundle

Shift

6632

6632

820

833

0.82

0.66

0.54

6421

6175

6266

0.38

0.65

1.03

8-Bundle

(Nat. U)

6729

6732

821

827

0.81

0.68

0.55

3537

3188

3313

0.70

1.25

1.95

- 355 -

KAERI/RR-1999/99

Table 3.7-2

Summary of 30 Instantaneous Calculations

Peak Channel

Power (kW)

Peak Bundle

Power (kW)

Channel Power

Peaking Factor

Radial Form Factor

Max.

Avg.

Min.

Max.

Avg.

Min.

Max.

Avg.

Min.

Max.

Avg.

Min.

2-Bundle

Shift

6699

6629

6580

783

771

763

1.129

1.110

1.088

0.824

0.818

0.810

4-Bundle

Shift

7232

7044

6911

906

875

858

1.148

1.122

1.104

0.785

0.770

0.750

6-Bundle

Shift

8157

7766

7447

1147

1078

1029

1.277

1.207

1.169

0.728

0.699

0.665

8-Bundle

Shift

9054

8374

7769

1200

1107

1024

1.425

1.303

1.234

0.698

0.649

0.599

8-Bundle

(Nat. U)

7135

6875

6704

894

872

840

1.127

1.093

1.080

0.809

0.789

0.760

- 356 -

KAERI/RR-1999/99

Table 3.7-3

Comparison of Refueling Simulation for 600-FPD




Refueling rate (Channels/day)

DUPIC reference

6844

804

1.063

4.05

Natural Uranium

6853

852

1.063

1.99

- 357 -

KAER17RR-1999/99

Table 3.7-4

Comparison of Probability to Exceed Administrative Limits

Administrative limits

Channel power

Bundle power

CPPF

ZCU level

6935 kW

888 kW

1.10

0.2 / 0.8

DUPIC reference

0.0*

0.0

0.17

0.12

Natural Uranium

0.33

0.0

0.0

0.15

* Percent

- 358 -

KAERI/RR-1999/99

Table 3.7-5

Constrained Sensitivity to Thermal Absorption Cross Section

Response

(Location)

MCP (0-17)

MBP (O-IO- 3)*

CPPF (M- 4)

Sensitivity

method

-0.054

-0.876

-0.148

Direct

calculation

-0.054

-0.812

-0.147

Error

(%)

0.0

7.9

0.7

* The MBP bundle is located at third bundle position out of 12 bundles in channel O-l 0.

- 359 -

KAERI/RR-1999/99

Table 3.7-6

Constrained Sensitivity to Neutron Production Cross Section

Response

MCP

MBP

CPPF

Sensitivity

method

0.027

0.664

0.119

Direct

calculation

0.028

0.729

0.121

Error

(%)

-3.6

-8.9

-1.7

- 360 -

KAERI/RR-1999/99

Table 3.7-7

Comparison of Sensitivity to Thermal Absorption Cross Section

Response

MCP

MBP

CPPF

Constrained

sensitivity

-0.054

-0.876

-0.148

Unconstrained

sensitivity

-0.186

-0.994

-0.260

Ratio

0.29

0.88

0.57

- 361 -

KAERI/RR-1999/99

Table 3.7-8

Comparison of Sensitivity to Neutron Production Cross Section

Response

MCP

MBP

CPPF

Constrained

sensitivity

0.027

0.664

0.119

Unconstrained

sensitivity

0.160

0.792

0.236

Ratio

0.17

0.84

0.50

- 362 -

KAERI/RR-1999/99

Table 3.7-9

Sensitivity Coefficient to Thermal Absorption Cross Section for Selected Burnup

FPD

100

200

300

400

500

600

Average

MCP (Channel)

-0.103 (P- 8)

-0.110 (Q-14)

-0.085 (Q-13)

-0.075 (S-11)

-0.107 (P- 8)

-0.106 (Q-14)

-0.091

MBP (Bundle)

-0.901 (O- 9-10)

-0.877 (L-14-10)

-0.797 (Q-13- 9)

-0.893 (O-l0- 3)

-0.843 (P- 8- 9)

-0.841 (N-15- 3)

-0.861

CPPF (Channel)

-0.102 (O- 9)

-0.100 (K-15)

-0.134 (N-20)

-0.178 (M-20)

-0.192 (K-20)

-0.089 (M-13)

-0.135

- 363 -

KAERI/RR-1999/99

Table 3.7-10

Sensitivity Coefficient to Neutron Production Cross Section for Selected Burnup

FPD

100

200

300

400

500

600

Average

MCP

0.077

0.083

0.058

0.047

0.080

0.079

0.064

MBP

0.692

0.664

0.568

0.683

0.615

0.632

0.645

CPPF

0.076

0.075

0.104

0.149

0.164

0.064

0.107

- 364 -

KAERI/RR-1999/99

Table 3.7-11

Uncertainty* of Lattice Parameters for DUPIC Fuel Option 1

Bumup(MWd/T)

0.0

3.3

164.8

824.9

1650.1

2474.9

3299.5

4124.2

4948.9

5773.4

6598.0

• 7422.6

8247.4

9072.1

9896.7

10721.4

11546.1

12370.8

13195.7

14020.5

14845.4

15670.5

16495.6

17320.8

18146.2

18971.5

0.00000679

0.00000766

0.00000979

0.00000958

0.00000673

0.00000668

0.00001056

0.00000707

0.00000758

0.00000717

0.00001531

0.00001247

0.00000779

0.00000707

0.00001583

0.00001369

0.00001208

0.00000917

0.00000908

0.00001060

0.00000995

0.00000943

0.00001166

0.00001305

0.00001011

0.00000999

0.00005119

0.00004763

0.00004485

0.00004046

0.00004237

0.00004353

0.00004267

0.00004396

0.00004383

0.00004466

0.00004217

0.00004487

0.00004406

0.00004256

0.00004193

0.00004235

0.00004314

0.00004196

0.00004102

0.00004785

0.00003953

0.00003879

0.00003849

0.00003674

0.00004051

0.0*004272

0.00258670

0.00256098

0.00225999

0.00203942

0.00206698

0.00211444

0.00216032

0.00220290

0.00224135

0.00227593

0.00230615

0.00233231

0.00235306

0.00236937

0.00238076

0.00238751

0.00238882

0.00238465

0.00237459

0.00235898

0.00233739

0.00230972

0.00227683

0.00223871

0.00219591

0.00214768

0.00143958

0.00143941

0.00143043

0.00140157

0.00136725

0.00133466

0.00130354

0.00127345

0.00124468

0.00121713

0.00119061

0.00116020

0.00113863

0.00111608

0.00109230

0.00106889

0.00104627

0.00102429

0.00100164

0.00097578

0.00095317

0.00093589

0.00091705

0.00089646

0.00087577

0.00085556

2A

0.00200876

0.00200848

0.00199701

0.00195996

0.00191730

0.00187642

0.00183713

0.00179890

0.00176195

0.00172622

0.00169142

0.00164974

0.00162175

0.00159172

0.00155937

0.00152723

0.00149609

0.00146562

0.00143374

0.00139684

0.00136433

0.00133939

0.00131185

0.00128158

0.00125139

0.00122119

0.00507154

0.00501301

0.00398896

0.00294586

0.00254516

0.00226390

0.00203147

0.00183583

0.00167440

0.00154729

0.00145395

0.00139263

0.00136020

0.00135200

0.00136049

0.00138055

0.00140657

0.00143463

0.00145920

0.00147791

0.00148859

0.00149025

0.00148225

0.00146386

0.00143399

0.00139392

H

0.00259615

0.00256538

0.00202940

0.00147867

0.00127667

0.00113412

0.00101334

0.00090880

0.00082025

0.00074846

0.00069415

0.00065748

0.00063769

0.00063249

0.00063812

0.00065143

0.00066906

0.00068853

0.00070661

0.00072181

0.00073286

0.00073903

0.00073992

0.00073509

0.00072399

0.00070726

* Numbers are two standard deviations in percent

- 365 -

KAERI/RR-1999/99

Table 3.7-12


Bumup

(MWd/T)

0.0

3.3

164.8

824.9

1649.9

2474.6

3299.1

4123.6

4948.1

5772.7

6597.2

7421.8

8246.3

9070.8

9895.3

10719.9

11544.6

12369.3

13194.2

14018.9

14843.8

15668.9

16493.9

17319.1

18144.3

18969.7

2*

0.00003806

0.00003884

0.00003777

0.00003751

0.00003685

0.00003755

0.00003876

0.00003861

0.00003699

0.00003696

0.00003756

0.00003772

0.00003794

0.00003776

0.00003779

0.00003735

0.00003922

0.00003809

0.00003831

0.00003818

0.00003742

0.00003774

0.00003761

0.00003896

0.00003841

0.00003720

0.00013875

0.00013862

0.00014493

0.00013812

0.00014079

0.00014225

0.00014003

0.00014191

0.00014191

0.00014231

0.00014542

0.00014445

0.00014249

0.00013998

0.00014176

0.00014030

0.00013911

0.00013652

0.00013521

0.00013824

0.00013048

0.00012721

0.00012300

0.00012701

0.00012182

0.00011301

0.00679272

0.00676275

0.00682074

0.00675141

0.00681108

0.00689571

0.00697653

0.00704864

0.00711061

0.00716177

0.00720072

0.00722843

0.00724242

0.00724202

0.00722789

0.00719916

0.00715618

0.00709660

0.00702133

0.00692952

0.00682127

0.00669885

0.00656021

0.00640784

0.00624191

0.00606305

2m

0.00182791

0.00182760

0.00182083

0.00179715

0.00176798

0.00173947

0.00171169

0.00168479

0.00165785

0.00163047

0.00160306

0.00157545

0.00154541

0.00151356

0.00149460

0.00147279

0.00144763

0.00142225

0.00139693

0.00137157

0.00134694

0.00132157

0.00129582

0.00126923

0.00124302

0.00121424

2A

0.00283908

0.00283861

0.00283000

0.00279524

0.00275282

0.00271083

0.00266889

0.00262767

0.00258599

0.00254340

0.00249964

0.00245552

0.00240661

0.00235331

0.00232140

0.00228448

0.00224144

0.00219790

0.00215337

0.00210887

0.00206471

0.00201942

0.00197327

0.00192561

0.00187849

0.00182718

v2&

0.01395604

0.01396206

0.01415304

0.01528809

0.01568692

0.01590409

0.01606448

0.01618558

0.01626885

0.01631574

0.01632412

0.01629386

0.01622167

0.01610667

0.01595098

0.01575748

0.01552080

0.01523948

0.01491684

0.01455290

0.01415196

0.01371606

0.01324431

0.01274177

0.01221178

0.01165914

H

0.00714449

0.00714751

0.00724964

0.00785575

0.00808488

0.00821811

0.00832062

0.00840166

0.00846192

0.00850210

0.00852090

0.00851841

0.00849274

0.00844354

0.00837188

0.00827895

0.00816226

0.00802105

0.00785706

0.00767040

0.00746328

0.00723693

0.00699089

0.00672805

0.00645009

0.00615968


- 366 -

KAERI/RR-1999/99

Table 3.7-13


Bumup(MWd/T)

0.0

3.3

164.7

824.7

1649.6

2474.2

3298.7

4123.1

4947.4

5771.8

6596.2

7420.7

8245.1

9069.6

9894.1

10718.7

11543.3

12368.0

13192.8

14017.7

14842.6

15667.6

16492.8

17318.1

18143.5

18969.0

0.00001058

0.00001094

0.00001028

0.00001064

0.00001122

0.00001039

0.00001112

0.00001242

0.00001527

0.00001766

0.00001269

0.00001305

0.00001353

0.00001777

0.00001419

0.00001401

0.00001563

0.00001314

0.00001230

0.00001372

0.00001352

0.00001340

0.00001267

0.00001290

0.00001325

0.00001480

0.00022133

0.00022053

0.00022419

0.00022212

0.00022102

0.00022111

0.00022107

0.00022063

0.00021897

0.00021711

0.00021619

0.00021429

0.00021345

0.00021072

0.00020806

0.00020401

0.00020248

0.00019702

0.00019356

0.00018865

0.00018369

0.00017786

0.00017525

0.00016615

0.00016346

0.00015481

0.01132726

0.01128695

0.01145883

0.01132418

0.01136224

0.01142915

0.01148567

0.01152658

0.01155094

0.01155800

0.01154771

0.01151860

0.01147060

0.01140246

0.01131459

0.01120533

0.01107398

0.01091996

0.01074326

0.01054490

0.01032456

0.01008321

0.00982277

0.00954473

0.00925041

0.00894294

0.00282375

0.00282373

0.00281403

0.00277740

0.00273161

0.00268642

0.00264171

0.00259705

0.00255269

0.00250782

0.00246091

0.00241622

0.00237515

0.00236265

0.00229674

0.00225380

0.00221088

0.00216839

0.00212577

0.00208243

0.00203825

0.00199432

0.00195071

0.00190729

0.00186505

0.00182376

2*

0.00440943

0.00440941

0.00439729

0.00434295

0.00427539

0.00420812

0.00414062

0.00407239

0.00400391

0.00393399

0.00385977

0.00378835

0.00372236

0.00367771

0.00359461

0.00352361

0.00345230

0.00338116

0.00330913

0.00323581

0.00316091

0.00308610

0.00301153

0.00293713

0.00286447

0.00279338

v2&

0.02199285

0.02198165

0.02214853

0.02372677

0.02413316

0.02426507

0.02430310

0.02427881

0.02419921

0.02406655

0.02388249

0.02364641

0.02336024

0.02303976

0.02263770

0.02219852

0.02170649

0.02116256

0.02057013

0.01993284

0.01925030

0.01852584

0.01776616

0.01697475

0.01615573

0.01531550

H

0.01070746

0.01070272

0.01079170

0.01163009

0.01189261

0.01201544

0.01208955

0.01213024

0.01214040

0.01212068

0.01207156

0.01199249

0.01188431

0.01175417

0.01158010

0.01138267

0.01115465

0.01089666

0.01061049

0.01029822

0.00995985

0.00959731

0.00921427

0.00881278

0.00839521

0.00796511


- 367 -

KAERI/RR-1999/99

Table 3.7-14

Uncertainty of Performance Parameter for DUPIC Fuel Option 1 (%)

FPD

1

100

200

300

400

500

600

Average

MCP

2.3

1.0

1.0

1.3

1.8

1.0

1.1

1.3

MBP

2.3

2.2

2.0

3.0

2.2

2.6

2.8

2.5

CPPF

1.6

0.7

0.7

1.8

1.6

1.4

0.3

1.2

- 368 -

KAERI/RR-1999/99

Table 3.7-15


FPD

1

100

200

300

400

500

600

Average

MCP

7.3

7.5

5.9

6.7

8.3

5.2

5.4-

6.6

MBP

9.5

11.9

9.8

10.7

11.6

15.4

10.1

11.3

CPPF

9.5

14.3

2.9

8.4

8.3

5.2

4.0

7.5

- 369 -

KAERI/RR-1999/99

Table 3.7-16


FPD

1

100

200

300

400

500

600

Average

MCP

11.6

9.3

7.9

8.0

9.4

13.8

11.9

10.3

MBP

14.7

13.2

14.5

13.1

14.5

13.4

21.6

15.0

CPPF

13.9

3.8

9.2

12.4

3.4

10.5

12.5

9.4

- 370 -

KAER1/RR-1999/99

Table 3.7-17

Sensitivity* of Clustering Group

Performance

Parameter

Maximum Channel

Power (kW)

Maximum Bundle

Power (kW)

Channel Power

Peaking Factor

Clustering Group

1

10

15

20

25

30

1

10

15

20

25

30

1

10

15

20

25

30

DUPIC Fuel Option

Option 1

6885

6843

6849

6844

6843

6849

796

805

804

805

805

805

1.051

1.055

1.054

1.056

1.054

1.054

Option 2

6831

6835

6833

6835

6831

6834

796

810

810

809

809

810

1.052

1.052

1.052

1.052

1.052

1.052

.. Option 3

6811

6818

6806

6810

6822

6813

809

823

819

817

822

818

1.043

1.043

1.044

1.045

1.044

1.043

* With Spatial and Bulk Control

- 371 -

KAERI/RR-1999/99

Table 3.7-18

Uncertainty due to Group-average Fuel Type

Performance Parameter

Channel power

Bundle power


Maximum

Average

Maximum

Average

Maximum

Average

DUPIC Fuel Option

Option 1

1.20*

0.49

0.89

0.57

1.20

0.46

Option 2

1.18

0.40

1.49

0.59

1.17

0.37

Option 3

0.77

0.34

1.11

0.57

0.76

0.33

* Percent

- 372 -

KAERI/RR-1999/99

Table 3.7-19

Comparison of Performance Parameters by Refueling Simulation

DUPIC

Fuel

Option 1

Option 2

Option 3

Performance Parameter










DUPIC Core Model

Single Fuel

Type

6844

804

1.0625

6831

800

1.0627

6746

794

1.0620

30 Fuel Types

6843

805

1.0661

6843

806

1.0665

6755

801

1.0640

Difference

(%)

0.01

0.12

0.34

0.18

0.75

0.36

0.13

0.88

0.19

- 373 -

KAERI/RR-1999/99

2-Bundle Shift4-Bundle Shift6-Bundle Shift8-Bundle Shift

Fig. 3.7-1 Comparison of Axial Power of Channel L-3

- 374 -

KAERI/RR-1999/99

7000

6500

6000

*• 5500

o% 5000c

I

° 4500

4000

3500


6 8 10 12 14 16 18 20 22

Column Position

Fig. 3.7-2 Comparison of Horizontal Channel Power for Row M

- 375 -

KAERI/RR-1999/99

7000

6500

6000

§ 5500

i_

| 5000Q."«3§ 4500CO

6

4000

3500

3000


10 12 14

Row Position

16 18 20 22

Fig. 3.7-3 Comparison of Vertical Channel Power for Column 11

- 376 -

KAERI/RR-1999/99

- Channel L-3Channel L-11

5 6 7 8

Bundle Position

10 11 12

Fig. 3.7-4 Axial Power Shape for 2-Bundle Shift Core

- 377 -

KAERI/RR-1999/99

10 11 12 13 14 15 16 17 18 19 • 20 21 22

A

B

C

D

E

F

G

H

J

K

L

M

N

O

P

QR

S

T

U

V

W

3415

3608

3753

3760

3622

3459

3694

4127

4414

4686

4844

4874

4758

4536

4254

3818

3360

4001

4483

4941

5287

5557

5725

5781

5696

5492

5135

4644

4111

3431

3455

4071

4693

5171

5559

5849

6082

6268

6357

6320

6167

5836

5380

4815

4159

J3506

|3319

4042

4701

5213

5527

5791

5971

6115

6361

6491

6525

6480

6157

5777

5311

4766

4144

3395

2996

3836

4645

5241

5657

5866

6026

6045

6127

6343

6475

6547

6570

6440

6179

5842

5411

4840

4002

3091

3485

4377

5147

5658

5973

6064

6152

6180

6242

6358

6463

6536

6584

6567

6444

6322

6024

5457

4632

3662

3932

4779

5485

5906

6125

6160

6215

6222

6261

6332

6412

6490

6563

6621

6589

6579

6400

5902

5119

4183

3283

4158

5026

5668

5993

6025

6105

6165

6193

6220

6263

6329

6410

6492

6550

6559

6549

6582

6162

5433

4462

[3485

3350

4302

5129

5713

5984

5979

6067

6135

6156

6155

6165

6219

6317

6436

6519

6548

6554

6621

6249

5573

4649

3591

3418

4330

5082

5606

5862

5887

6051

6131

6130

6072

6019

6060

6190

6401

6523

6556

6510

6498

6140

5518

4705

3687

3417

4330

5081

5605

5860

5886

6050

6129

6129

6070

6018

6059

6188

6400

6522

6555

6509

6496

6139

5517

4705

3687

3348

4300

5126

5710

5980

5975

6063

6130

6151

6151

6161

6215

6313

6432

6515

6544

6550

6617

6246

5571

4647

3590

3280

4154

5021

5662

5987

6018

6097

6156

6185

6211

6255

6321

6403

6486

6544

6553

6543

6577

6158

5429

4460

3483

3927

4772

5478

5897

6114

6149

6203

6210

6249

6321

6402

6480

6555

6613

6581

Us726394

5896

5115

4180

1

3480

4370

5137

5647

5959

6049

6136

6164

6227

6344

6450

6525

6574

6558

6435

6314

6017

5452

4627

3658

2990

3827

4635

5228

5642

5848

6007

6025

6108

6325

6460

6534

6557

6428

6168

5832

5403

4833

3997

3088

3310

4031

4686

5195

5508

5770

5948

6093

6341

6472

6510

6464

6143

5764

5301

4757

4137

3389

3443

4055

4674

5150

5536

5824

6057

6244

6335

6300

6149

5819

5365

4803

4149

3499

|

3344

3981

4461

4917

5262

5530

5700

5757

5673

5472

5116

4628

4097

3421

3671

4103

4389

4660

4818

4849

4735

4516

4234

3801

3390

3582

3728

3736

3600

3438

Fig 3.7-5 Channel Power Map of Reference DUPIC Core

- 378 -

KAERI/RR-1999/99

7500

7000 -

to

6 6500 -

6000300

Full Power Day

500 600

Fig. 3.7-6 Maximum Channel Power for 600-FPD Simulation

- 379 -

KAERI/RR-1999/99

950 -

900 -

Q.

m

s0-

850 -

800

750100 200 300 400

Full Power Day

500 600

Fig. 3.7-7 Maximum Bundle Power for 600-FPD Simulation

- 380 -

KAERI/RR-1999/99

1.0

0.8

5> 0.6

iij2

8 0.4

0.2

0.0

! ;'

^ ^ ^ W ^ ^ ^

- Average Level- Highest/Lowest Level

100 200 300 400

Full Power Day

500 600

Fig. 3.7-8 Zone Controller Level for 600-FPD Simulation

- 381 -

KAERI/RR-1999/99

1.15

oo2.O)c(0a>a.

oa.

1.10 -

•53 1.05 |-

1.00100 200 300

Full Power Day

400 600

Fig. 3.7-9 Channel Power Peaking Factor for 600-FPD Simulation

- 382 -

RFSP code

TIME-AVER:Time-average calculation

INSTANTAN:Instantaneous calculation

SIMULATE:Refueling simulation

GENOVA code

ADJOINT:Adjoint calculation

PERTXS:Unconstrained sensitivity

Processing code

CALCON:Constrained sensitivityUncertainty estimation

KAERI/RR-1999/99

Covariance data

Fig. 3.7-10 Flow Diagram of Sensitivity Calculation

- 383 -

KAERI/RR-1999/99

10 11

|.S74 .855 .095

12 13 14 15 16

,797 .321,5551

17 18 19 - 20 21 22

BCDEF

.947 .187

.350 .587

.508j.666 .908.068 .311.463 .705.868 .110

197 .716 .961597 .121 .361000 .518 .755397 .918 .161

.242 .479

.637 .879

.042 .279

.437 .676

.208 .726

.608 .132Oil .529.408 .929

800 .324 .558 .840 .079

.500

.900

.300

.687,811 .334 .5681.850 .100

.971 .253

.371 .647

.766 .053

.171 .447

GH

.297

.692

.205

.605

.008

.405

.724 .968

.129 .368.250 .487

.645 .887.526 .763 .050 .287

.926 .168

.332 .566.445

237 .753 .997,634 .158 .395,040 .553

.276 .516

.674 .916.224 .740.621 .142

.795.684 .434 .955 .195

.837 .358 .595

.076 .318

.474 .713

.876 .118

.026 .540

.421 .940

.984 .263

.382 .658

.776 .063

.503

.904

.305

.218 .734

.616 .137

979 .258,376 .653

.021 .534 .771 .058 .300

.497

.897

.182 .458

.582 .863.700.105

.416 .934 .176 .453,576 .858

.695.016.413.816.097 .847 .087 .824 .345 .818 .340 .100

MNOPQ

.495

.895

.295

.221

.618

.024

.418

.821

.737 .982

.140 .379

.537 .774

.937 .179

.261

.655

.061

.455 .697

.500

.900

.303

.226 .742 .987

.624 .145 .384

.029 .542 .779

.424 .942 .184

.266 .505

.661 .905

.066 .308

.461 .703

.234 .750

.632 .155

.037 .550

.432 .953

.995 .274

.392 .671

.792 .074

.192 .471

.513

.913

.316

.711

.203 .721.126

.966 .247

.366 .642.524 .761 .047

.403 .924 .166 .442

.603

.005

.342 .579 .861 .103826 .347 .584 .866 .108 834 .355 .592 .874 .116 .805 .329 .563 .845

.484

.884

.284

.682

.084

.216

.613

.018

RSTUV

271 .511.211 .729J974 .255 .492 ,200 .718 .963 |.245

611 .134 .374013 .532 .768,411 .932 .174.813 .337 .571

.650 .892

.055 .292

.450 .700

.853 .092

,482.882.282.679

803 .326 .560 .842 .082

,600 .124.003 .521.400 .921

.363 .640

.758 .045

.163 .440

W ^•958 .240 .476 .213 .732 .976 I

Fig. 3.7-11 Age Distribution of Instantaneous Core

- 384 -

KAERI/RR-1999/99

DUPIC Fuel Option 1

1.13 1.18

Fig. 3.7-12 Distribution k^ for 30 Fuel Types (Option 1)

- 385 -

KAERI/RR-1999/99

oO

<5XI

E

3 U

40

30

20

10

g

* J * 1 "

DUPIC Fuel Option 2

n

n

r

flmll0.00020 0.00021

i • i •

fin

(11(1 [ f l n0.00022 0.00023 0.00024 0.00

U235 Number Density

Fig. 3.7-13 235U Content Distribution for 30 Fuel Types (Option 2)

- 386 -

KAERI/RR-1999/99

DUPIC Fuel Option 3

0.00018 0.00019 0.00020 0.00021

U235 Number Density

0.00022

235'Fig. 3.7-14 U Content Distribution for 30 Fuel Types (Option 3)

- 387 -

KAERI/RR-1999/99

7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22

.77

.83 .871.10 .891.20 .92.81 .77.65 .69.61 .63.57 .63.60 .60.62 .64.81 .691.18 .78.78 .67.54 .58.44 .48.39 .46

37) .2634 .3039 .2737 .30

26 .2223 .2222 .1514 .13

59 .4856 .5261 .4965 .5587 .58

18 .4216 .3120 .1522 .2028 .2835 .4541 .8030 .64

.42 .25

.38 .30

.43 .29

.41 .33

.46 .32

.43 .37

.43 .3336] .31

.51 .47

.56 .51

.57 .56

24 .3021 .2716 .2619 .2042 .2634 .22.18 .21.19 .20.27 .26.46 .30.78 .40.68 .29

.38

.33

.33

.28

.30

.2630.27.29.24

.45

.44

.39

.40

.36

.40

.38

.42

.35

.30

.60

.50

.49

.44

.47

.47

.57

.62

.52

.37

.70

.62

.54

.53

.52

.60

.751.12.69.49

.31 .36

.28 .35

.85

.79

.76

.68

.67

.62

.64

.62

.66

.66

.65

.56

.5346

.72

.70

.66

.65

.62

.64

.63

.65

.61

.59

.68

.65

.65

.62

.63

.61

Fig. 3.7-15 Channel Power Uncertainty due to Group-Average Fuel Property (Option 1)

- 388 -

KAERI/RR-1999/99

9 10 11 12 13 14 15 16 17 18 19 20 21 22

ARCDEFGHJKLMNOPQRSTuVW

.84

.89

.88

.88

.85

.85

.80

.84

.85

.87

.87

.86

.86

.82

.84

.80

|.69.73.77.82.83.93.87.88.82.81.84.78.79.75

1I

r.64.69.72.79.78.83.83.85.82.80.81.79.75.74.74R01L

11

.52.60

.63

.67

.68

.74

.76

.78

.82

.83

.80

.76

.76

.70

.70

.787477

—]

47.52.54.58.59.64.68.73.80.77.75.75.70.71.65.67.69737673

AT>.48.49.54.54.60.69.69.71.73.69.67.65.64.62.62.6766r>73

139

.43

.48

.53

.53|

.55

.58

.61

.61

.61

.61

.58

.56

.55

.55

.59|

.59647074

1

.3239.41.47.49.55.52.54.53.50.51.49.49.46.48.46.49.57616669.72

.3535

.3842.47.51.52.48.46.41.43.39.41.37.39.39.44.53586469.69

.2730.36.39.45.47.48.46.37.36.29.31.31.31.30.35.41.50596365.68

.3030.36.40.44.47.44.46.40.32.33.27.28.26.30.31.37.46556364.66

.2938

.38

.42

.44

.47

.46

.44

.38

.35

.34

.31

.29

.28

.28

.30

.36

.465?6165.67

35|43

.45

.46

.4750|.52.47.45.38.37.35.31.31.34.36.39|.4554556167|

43.49.46.49.53.51.55.51.48.47.41.38.36.33.44.44.48515560

5?

.53

.52

.56

.55

.56

.57

.53

.52

.52

.46

.46

.44

.41

.43

.47

.46535758

56.57.57.60.60.59.61.57.56.56.51.50.50.44.43.47.46535759

~M\.63.62.62.62.62.62.60.59.57.56.53.51.48.49.525559|

.67|

.67

.64

.65

.66

.65

.64

.62

.58

.57

.59

.54

.57

.55

.53

55f

.67

.70

.67

.67

.67

.67

.66

.65

.61

.59

.57

.59

.55

.52

.69

.68

.69

.71

.68

.67

.64

.63

.62

.60

.67

.69

.70

.65

.64

.65

Fig. 3.7-16 Bundle Power (Position 6) Uncertainty due to

Group-Average Fuel Property (Option 1)

- 389 -

KAERI/RR-1999/99

ABCDEFGHJKLMNOPQRSTUVw

8 9 10 11 12 13 14 15 16 17 18 19 20 21 22

.16 .24.35 .29 .23 .19 .21 .19 .25 .31 .45 .55

.58 .48 .29 .28 .21 .25 .24 .33 .37 .44 .69 .92

.91 .52 .36| .27 .27 .23 .28 .29 ,37| .51 .69 1.08.62 .511 .50 .36 .32 .23 .23 .25 .25 .39 .51 .591 .72 .73

.66 .54 .54

.64 .61.72 .72 .58.78 .69 .65.71 .73 .61.76 .62 .66

.70 .96

.49 .37 .32 .24 .17 .18 .26 .32 .44 .55.66 .67 .65.56 .46 .40 .27 .21 .15 .20 .19 .30 .46 .43 .48.57 .45 .38 .25 .21 .42 .43 .24 .26 .46 .41 .55.55 .52 .33 .30 .19 .32 .36 .23 .26 .34 .44 .46.62 .53 .40 .30 .15 .15 .21 .21 .27 .42 .49 .55.68 .52 .46 .30 .22 .22 .20 .22 .30 .40 .61 .76.89 .55 .42 .29 .35 .28 .30 .28 .32 .47 .63 1.14

.71 1.03 .85.69 .57

.58 .48 .32 .31 .43 .48 .33 .34 .39 .49

.43 .37 .39 .40 .75 .75 .40 .28 .28 .33

.59 .56

.58 .58 .64

.52 .63 .58

.65 .66 .59

.84 .63 .611.20 .80

.69 .74 .67

.44 .53.56 .42 .37| .28 .30 .67 .67 .30 .27| .21 .36 .37

.47 .42

Fig. 3.7-17 Channel Power Peaking Factor Uncertainty due to

Group-Average Fuel Property (Option 1)

- 390 -

KAERI/RR-1999/99

7500

7000

License limit

6500

6000

• Homogeneous Case• Heterogeneous Case

100 200 300 400

Full Power Day

500 600

Fig. 3.7-18 Heterogeneity Effect on MCP during 600-FPD Simulation

- 391 -

KAERI/RR-1999/99

950 -

900

I850

• a

800

750

License limit


100 200 300 400

Full Power Day

500 600

Fig. 3.7-19 Heterogeneity Effect on MBP during 600-FPD Simulation

- 392 -

KAERI/RR-1999/99

1.15

1.00


100 200 300 400

Full Power Day

500 600

Fig. 3.7-20 Heterogeneity Effect on CPPF during 600-FPD Simulation

- 393 -

KAERI/RR-1999/99

3.8

DUPIC

5.

CANDU

ROP H

.. SEU/DU S ^

DUPIC

. Doppler

. DUPIC

DUPIC

71 51*11

I, ROP S

5 .

DUPIC

7>

DUPIC , DUPIC

- 394 -

KAERI/RR-1999/99

DUPIC

CANDU-6 ^ > ? ^ 4 ^ i ^ ^] f } DUPIC

- 395 -

KAERI/RR-1999/99

3.9 ^H^Urrti

1. J.S. LEE et al., "Reaserch and Development Program of KAERI for DUPIC (Direct Use

of Spent PWR Fuel in CANDU Reactors)," Proc. Int. Conf. and Technology Exhibition on

Future Nuclear System: Emerging Fuel Cycles and Waste Disposal Options, GLOBALf93,

Seattle, USA, 1993.

2. H.B. CHOI, B.W. RHEE, and U.S. PARK, "Physics Study on Direct Use of Spent PWR

Fuel in CANDU (DUPIC)," Nucl. Sci. Eng.: 126, pp.80-93, May 1997.

3. H.B. CHOI, J.W. CHOI, and M.S. YANG, "Composition Adjustment on Direct Use of Spent

Pressurized Water Reactor Fuel in CANDU," Nucl. Sci. Eng.: 131, pp.62-77, Jan. 1999.

4. "Design Manual: CANDU 6 Generating Station Physics Design Manual," 86-03310-DM-000

Rev. 1, Atomic Energy of Canada Limited, 1995.

5. E.S.Y. TIN and P.C. LOKEN, "POWDERPUFS-V Physics Manual," TDAI-31 Part 1, Atomic


6. A.R. DASTUR et al., "MULTICELL User's Manual," TDAI-208, Atomic Energy of Canada

Limited, 1979.

7. D.A. JENKINS and B. ROUBEN, "Reactor Fuelling Simulation Program - RFSP: User's

Manual for Microcomputer Version," TTR-321, Atomic Energy of Canada Limited, 1993.

8. J.V. DONNELLY, "WIMS-CRNL: A User's Manual for the Chalk River Version of WIMS,"


9. H. CHOW and M.H.M. ROSHD, "SHETAN - A Three-Dimensional Integral Transport Code

for Reactor Analysis," AECL-6787, Atomic Energy of Canada Limited , 1980.

10. H.B. CHOI, "A fast-running fuel management program for a CANDU reactor," Annals of

- 396 -

KAERI/RR-1999/99

Nuclear Energy, 27, pp.1-10, 1999.

11. M. EDEMUS and B.H. FORSSEN, "CASMO-3 A Fuel Assembly Burnup Program User's

Manual Version 4.4," STUDSVIK/NFA-89/3, Studsvik of America, Inc., 1989.

12. A.G. CROFF, "A User's Manual for the ORIGEN2 Computer Code," ORNL/TM-7175, Oak

Ridge National Laboratory, 1980.

13. H.B. CHOI and G.H. ROH, "A Sensitivity Study on DUPIC Fuel Composition,"

KAERI/TR-942/97, Korea Atomic Energy Research Institute, 1998.

14. The Economics of the Nuclear Fuel Cycle, Organization for Economic Cooperation and

Development, Nuclear Energy Agency, 1993.

15. J.W. CHOI, W.I. KO, J.S. LEE, M.S. YANG, and H.S. PARK, "Cost Assessment of a

Commercial Scale DUPIC Fuel Fabrication," Proc. the 6th Int. Conf. on Radioactive Waste

Management and Environmental Remediation, Singapore, Oct. 12 - 16 1997.

16. H.S. PARK et al., "The Construction of an Interim Spent Fuel Storage Facility," KAERI-

NEMAC/PR-35/94, Korea Atomic Energy Research Institute, 1994.

17. P.J. PERSIANI et al., "Fuel Reprocessing Data Validation Using the Isotope Correlation

Technique," Proc. Int. Conf. on the Physics of Reactors: Operation, Design and Computation,

Marseille, France, 1990.

18. N. ENSSLIM et al., "Analysis of Initial In-Plant Active Neutron Multiplicity Measurements,"

LA-UR-93-2631, Los Alamos National Laboratory, 1993.

19. Y.D. HARKER et al., "Precise Measurement of Fuel Content of Irradiated and Nonirradiated

Materials," 25th Annual Meeting of Institute of Nuclear Materials Management, 1984.

20. W.R. CORCORAN et al., "Damping of xenon oscillation in the Main Yankee reactor," Proc.

- 397 -

KAERI/RR-1999/99

American Nuclear Society 6th biannual conference on reactor operating experience, Myrtle

Beach, South Carolina, 1973.

21. F.A.R LARATTA, G.K.J. GOMES and CM. BAILEY, "Design and Assessment of the

Replacement ROPT Systems for Wolsong-1," TTR-289, Part 1(W1), Atomic Energy of

Canada Limited, 1995.

22. M.R. SOULARD, "NUCIRC Code Validation," TTR-301, Atomic Energy of Canada Limited,

1991.

23. J. PITRE, "ROVER-F Manual," TTR-605, Rev.l, Atomic Energy of Canada Limited, 1999.

24. E. GREENSPAN, "Sensitivity Functions for Uncertainty Analysis," J. LEWINS and M.

BECKER, Eds., Advances in Nuclear Science and Technology, Vol.14, Plenum Press, New

York, 1982.

25. J.J. DUDERSTADT and L.J. HAMILTON, Nuclear Reactor Analysis, John Wiley & Sons,

Inc., New York, 1976.

26. D.H. KIM, J.K. KIM, and H.B. CHOI, "A Generalized Perturbation Theory Program for

CANDU Core Analysis," Annals of Nuclear Energy, 2000.

27. C.R. WEISBIN, "Sensitivity and Uncertainty Analysis," J. LEWINS and M. BECKER, Eds.,

Advances in Nuclear Science and Technology, Vol.14, Plenum Press, New York, 1982.

28. G. WILLERMOZ, G. BRUNA, R. CASTELLI, and P. BETHOUX, "Studies on the Sensitivity

of PWR Core Parameters to Fuel Manufacturing Uncertainties via Statistical and Perturbation

Methods," 1994 Topical Meeting on Advances in Reactor Physics, Knoxville, 1994.

29. G.N. LANCE and W.T. WILLIAMS, "A General Theory of Classification Sorting Strategies

1. Hierarchical System," The Computer Journal, 9, pp.373-380, 1966.

- 398 -

KAERI/RR-1999/99

. DUPIC

- 399 -

KAERI/RR-1999/99

4. DUPIC

^ 7 ] (DUPIC)

CANDU ^^•S.^-M ° ^

.1'2 ZLBfth DUPIC

$17} nfl-grofl, DUPIC

^ ^ ] 1 ^ ] CANDU

71 4^r<Hl CANDU J&^>S.ofl DUPIC

CANDU

CANDU

D U P I C CANDU

o]

o]

7]

7} CANDU

ZLBiP.^, DUPIC

- ^ S . CANDU ^[^}

(discrete ordinates) ^ ^ ^ f Point Kernel

^ 7]^ 4^1 *VAnj DUPIC 4.1

fe, 7}

CANDU

- 401 -

KAERI/RR-1999/99

4.3*1 ofl/fe CANDU

- 402 -

DUPIC «¥<d.S.7} ^ -^£1 CANDU

1, DUPIC

KAERI/RR-1999/99

4.1 3L& ^ ^ }

^ Boltzmann

Boltzmann *!•§• * H * K r 3 f e

*B TP- 7>*1 yov^, -2T 4 ^ : ^ ^ (Discrete Ordinates Method)

Carlo)

CANDU5} ^ ^ 7f«y- # ^

EXTERMINATOR [Ref. 9] i f MAC-RAD [Ref. 10]# 4

ANISN [Ref. 11], DOT4.2 [Ref. 12]5f QAD-CG [Ref. 13]

4.1.1

fe ^ 4 CANDU

H«J ANISN SJE^r

^ H Boltzmann ^ £ ^ ^ ^ ^ ^^^)1 >«>f§-^>. ANISN

i« r jS^o ] *fliS*f7fl 7)1^>5]^ ^ # ^ f - ^ ^ ^ l (Deep Penetration

Problem)# *fl^*>7l JH*H ^V§-^>. CANDU

CANDU ^ ^ S . ^H^H^^lfe ^ l ^ A S . DLC-37D B M - M S]I4oflA-| AJA>^ 3 8

5- n1-^^! * f4S .# Af§-t>t>. DLC-37D e ro|_ti5]ej^ Experimental Power Reactor

(EPR) 2} q-fr^S. 2*&4\

ENDF/B-IV# ^ 7 ] 5 . 7H^*fSElc>.1S ^ . ^^K DLC-37D e>ola.

, 2. &$£$}5L $X^ ^^*i l ^11^1 a l ^ ^ ^ ^ - f Sfe Sife DUPIC

CANDU &

- 403 -

KAERI/RR-1999/99

BUGLE96 s>o| tL^e]! -

(LWR)

^ 51 o.n}, ENDF/B-VI [Ref.

3 f e *¥*KS°Jn}. BUGLE96

(Infinite dilute) ^ 9 3

(Self-shielded) 4downcomer

47$ ^ j - 20$

(BWR)

10571)

MCNP-4B 2 = f e

SJE-olc]-.18 MCNP-4B

cross-section)

>. NCNP-4B

MCPLIB1-,

(pointwise

«g^t;K MCNP-4B#

fe RMCCS $• ENDF60#

fe TMCCS1- -Si

DORT 3.^*1

4

GIP

, ^ i

^ ^ ^ * r S # ^^bcf. ANISN

fe GIP 3.^$] activity tabled]

ANSI/ANS-6.1.1-1977 [Ref.

(flux-to-dose conversion factor)# tH

BUGLE96., 4

- 404 -

KAERI/RR-1999/99

4.1.2 M q %\

, CANDU

ANISN 7f*M- ^ * f c 2 . ^ E U ^ S

3.717}

-^r ^B>H.el<>f ^ -a .AlS (calandria side tube

sheet), ^^rvfl^j B>4i7ot- ^- (carbon steel balls in light water) 9 |

# ^ . A ) ^ (refueling machine side tube sheet)^. ^Qr.}. 0} ^M.

^ 1600 cm^| ^ ? > ^ 1 ^ 137 cm

4.1.2.1

, CANDU ^ f ^ r

V l 2 0 CANDU

, CANDU Q

(time-average core)

7716 3} 6847

7228 MWdMTUoju]-.21

7]-. ANISN

CANDU ^ 1 ]

| l , BUGLE96 ^ ^ ^ } S # >)-§-*l 1^>^ ANISN

4-8-^ ^ - ^ 1 - ^ ^ ^^f^ n ^ £ ^ S 4.2-lofl

^r WIMS-AECL 3.^. ^ J i ^ ] ^ # ^ g ^ ^ ^ i -#Efl (3500

- 405 -

KAERI/RR-1999/99

3.

BUGLE96 * ^ £ 3 * r ^ l #^$a^l &^r £ 7r*l ^ # <*, l33Cs,150Dy, 166Er, 155Gd, 165Ho, 127I, 83Kr, I39La, 143Nd, 105Pd, I47Pm, 103Rh, 147Sm, 99Tc £131Xe g-ofl cU*1Hfe BUGLE96

<H, BUGLE96

. ^ -^m, CANDU ^

(L-ll, L-12, M-ll ^ M-12H

^ . \ c ^ # d o H ^ ^ ^5\Q i ^ ^ H5L±r 7.797X

10" fissions/cm3sec ^ ^ 723.125 kWth/bundle

<g*ot . n]^lx;f. d t ^ ^ x f (coarse mesh)

(spatial discretization),

(angular discretization), y ] ^ ^ ^ 4-ITEJ-^S. (order of anisotropy treatment),

(quadrature set) ^-o]] nj-ef x}o}7\ ^ ^ ^l^K CANDU #^>*fs|]

ANISN 7114H ^f§-5]^ * a ^ § ^ ^^g^-i- Al^^7l ^1*H, Bfl^^ ^ ,

§ ANISN

- 406 -

KAERI/RR-1999/99

\. MCNP-4B

ANISN

MCNP-4B 3 £

71

MCNP-4B nfl,

MCNP-4B

MCNP-4B

( v )

, MCNP-4B 3.

N0NU7f 5a

NONU

NONU

-i- #0)7]

1 7 ] ^ (variance reduction technique)^ A]~g-*}Sit:>.23 -f-^*> ^1^CH]A]^ A«

#t:-| n^^ T l l ^ l ^ i . 6>B||*V7] ^ * H , 4 ^ 1 ^ " ^ (particle splitting)

Russian roulette 3f ^ - ^ 7fl^|^ ^ ^ [ ^ ^ (population control method)#

71

10.0%

s. (PWT)l-MCNP-4B

*H # 20,000,0007])^

MCNP-4B

- 407 -

KAERI/RR-1999/99

- * S , ANISN

0.05%<Hl

MCNP-4B3.HS} S # Tally Tally (F4)# o]

Z) = f DFi<f>(E)dE (4.1-1)

.-fe ANSI/ANS-6.1.1-1977<HlA-|

Tally 9>£.

Tally ^}^6l F 6

MCNP-4B

5Nd, ' V

i 124 «?<tofl nflt> *?*fc§.# S.^}3. $lJL, ENDF/B-VI

fe ENDF60 ^ ^ > ^ ^ 4 ^ ^ # ^ S . 4-§-^>^t:|-. "15] 14, ENDF60

^ ^ : ^ ^ ^ , 83Kr, !01Ru, 103Rh, 10SPd, 108Pd, 113Cd, 13IXe, I43Nd,

, 148Nd, 149Sm, 151Sm, l52Sm, 154En, 155Eu, ^ 242mAm^ ENDF/B-V

fe RMCCS

^8: Thinning ENDF/B-VI, ENDF60#

Release 2 #

t a } ^ 139La, 160Dy, 161Dy, I62Dy, 163Dy, 164Dy, 166Er, 167Er, I31I, 144Nd, 146Nd, I50Nd,

238Np, 148mPm, 15!Pm, 148Sm, 153Sm,

Fission Product)^

!33Xe

TMCCS

. ( P s e u d o

-fe- MCPLIB02

>. ANISN ^ MCNP

], ANISN ^ MCNP-4B

- 408 -

KAERI/RR-1999/99

$L5L, BUGLE96#

ANISN 7 ] ] ^ -§^*h ^ U # -&*&#£• MCNP-4B T l l ^ ^ H]J2.*H 4

4 31.9%, 7.09% g 13.4% ^7)1 uj-Bf^r]-. #<g MCNP-4B Tfl^KgafSj X ^ S * M -

-f, #*} 91 ^ ^ i^ l -^ l cB*]: 7>^- ^ ^ r *>o|fe 4 4 1.2% g 2.7%o]

, CANDU € ^ S ^ l ^ ^ l ^ l f - 4 ><Hl BUGLE96

ANISN 3.^.%

e>. ANISN ^ MCNP

MCNP

ANISN

MCNP-4B ^ 2 } i f H]J2.^; ttfl, ANISN^ ^ 2 f e ^^>He]o> â.A]S<Hl^fe ^ 8%

CANDU ^J^>5.^ 1*> ^3Efl^lî) « i ^ ^ # ^lôfl BUGLE96

ANISN 3 - = #

4.1.2.2 DUPIC i^-y 7j|-il

DUPIC ^ ^ [ ^ ^ S^^gr WIMS-AECL S H 5 ] ^ ^ 7 ) ] ^ ^ ^^^J-Efl (7400

MWd/MTU)<HM ^-^f^cl-. ^ ^ « i >i3a}

ANISN2} MCNP

DUPIC ^ ^ S 7 } ^ ^ ^ CANDU

^ 4.437X10" fissions/cm3sec ^ f e 553.5 kWth/bundle<>l cf.

MCNP-4B 3 H S . ^-*> ^ ^ l - ^ a 2.4<H] Hl^j-ûj-. BUGLE96

ANISN S^<^1 *|?> # ^ 4 , %*} 9J ^ ^lBoll-^4 4 35.4%, 10.9% *i 13.4% ^711 ^}E>^t:}. MCNP-4B

4.0% if- 2.5%

- 409 -

KAERI/RR-1999/99

MCNP-4B

10% ^ ^ | I ^

, BUGLE96

MCNP-4BSf

- 410 -

KAERI/RR-1999/99

Table 4.1-1

Atomic Densities of Materials Used in CANDU Primary Shield Calculation

No.

1

2

3

4

5

6

Region ID

Stainless Steel304L

( p =7.9g/cm3)

Carbon SteelBall/H2O

(60/40 region)

Ordinary Concrete(,o=3.36g/cm3)

Moderator(p=1.09g/cm3)

Water

Air

Element

CSiCr

HC0

HC0

Mg

HD

H

0

Atomic Density(atoms/b-cm)

1.387E-041.271E-031.734E-02

2.674E-027.794E-041.337E-02

9.583E-031.143E-024.531E-026.018E-03

1.839E-46.599E-2

6.639E-2

5.018E-5

Element

MnFeNi

SiMnFe

AlSiCaFe

O

O

Atomic Density(atoms/b-cm)

1.732E-035.812E-028.107E-03

2.525E-045.010E-045.002E-02

1.534E-041.783E-037.498E-031.112E-04

3.309E-2

3.346E-2

- 411 -

KAERI/RR-1999/99

Table 4.1-2

Reference Number of Meshes and Dimensions Used in End Shield ANISN

Calculation

Region

1

2

3

4

5

6

Total

Region ID

Core

Calandria Side TubeSheet

Carbon Steel Ball &Water

Fuelling Machine SideTube Sheet

Concrete Wall (I)

Concrete Wall (II)

Number ofMeshes

90

10

90

15

20

50

275

Thickness(cm)

297.18

5.08

78.74

7.62

30.48

82.52

Radial Distance fromCore Center (cm)

297.18

302.26

381.00

388.62

419.10

525.62

- 412 -

KAERI/RR-1999/99

Table 4.1-3

Comparison of Dose Rate through End Shield between ANISN and MCNP-4B

Code

ANISN

MCNP-4B

Cross SectionLibrary

BUGLE96

ENDF60(ENDF/B-VI)

Dose Rates (/iSv/hr)

neutron

1.2624

1.8543(1 ±0.0889)

gamma

55.7034

59.8872(1 ±0.0583)

total

56.9658

65.7415(1± 0.1063)

- 413 -

KAERI/RR-1999/99

Table 4.1-4

Comparison of Dose Rate through End Shield for DUPIC Fuel Core

Code

ANISN

MCNP-4B

Cross SectionLibrary

BUGLE96

ENDF60(ENDF/B-V1)

Dose Rates (^Sv/hr)

neutron

1.6026

2.4811(1 ±0.0914)

gamma

66.4504

73.8065(1 ±0.0602)

total

66.0530

76.2826(1 + 0.1094)

- 414 -

KAERI/RR-1999/99

Calandria SideTube Sheet

Fuelling MachineTube Sheet

Core

297.18cm

CarbonSteelBalls

/Water

78.74cm

Concrete

106.68cm

5.08cm 7.62cm

Fig. 4.1-1 One-dimensional Model for the End Shield System (Not Scaled)

- 415 -

KAER1/RR-1999/99

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22

A

B

C

D

E

F

G

H

J

K

L

M

N

O

P

Q

R

S

T

U

V

W /

//

//

f

Inner/Outer Core Boundary

Fig. 4.1-2 Core Channel Map for CANDU-6 Reactor

- 416 -

Dose Rate Distribution

i

no3ari;

Io

CD

nustar

T I

33oore

Cen

CD

I

8"

KAERI/RR-1999/99

10"°"

300 310 320 330 340 350 360

Distance From Core Center (cm)

370 380

Fig. 4.1-4 Comparison of Heat Deposition Rate through End Shield

for Natural Uranium Core

- 418 -

do'

ynIICD

a•SoenSiOS

sm

cOno3

Heat Deposition Rate (mW/cm;

ga

T1

Ioo3

> S1 oCO Z

DO

Fueling Machine Side Tube Sheet..i i i i i

KAERI/RR-1999/99

4.2 CANDU

Mli)

ii)

iii)

iv) ZL

61 cf. CANDU

2*}

4.2.1 CANDU

f. ZI^} 4.2-

^-7} ^71

CANDU

^ t : } . CANDU DUPIC

DUPIC

al7l nfl^ofl, DUPIC

4.2.1.1

5.07 cm .a.A]^, 78.74 cm

^ . ^ ) ^ 91.44 cm o]u>.

i ^ 304LS ^>#<H %

- 420 -

KAERI/RR-1999/99

4.2.1.2

160 cm

122 cm H , 7.6 cm

4.2.1.3

>3. 532 cm

(reactivity mechanism deck)

122 cm -^

:>. CANDU

4.2.1.4

122 cm

220 cm 122 cm

44

4.2.2

&JL,

30 mSv

Atomic Energy Control Act^]

602.(2)

50

-. CANDU4.2-lofl

- 421 -

KAERI/RR-1999/99

4.2.3 DUPIC

, DUPIC

. DUPIC

CANDU

CANDU

CANDU €

DUPIC

ANISN

4.2.3.1

DUPIC

^ ^ 1 RFSP [Ref. 26]#

ANISN

xQcxl03[W/kW]x3.1xlQ10 [fissions/W • sec]

»

= 5.797 x 1011 [ fissions/cm3 • sec].

(MW)

(kW),

h^ (MW)

(cm) °]3.

(cm)

^ 4

- 422 -

KAERI/RR-1999/99

SL ^ S ^ M Tfl^^cf. ^ - f e B ? ^ DUPICWIMS-AECL 7ff-&©3.-fB| <£$X3., ^ «|<2Sofl

t>. ^ - f s H ? ^ DUPIC ^<£3.ofl tR> ^ ^ ti<M-5fe #A3*HNr 4 42.66632]- 2.7094 o]v\. ^ < £

4.2.3.2

> # ^ - S ^ ANISN 2.H

4 ^ ^ 5 ] 3.71-fe- 4.1.2^^

]-. 4 7H£| 7>^ ^ ^^(^1 cfl*> ^-^*o> # ^ ^ S ^ H ^ 4.2-4ofl

^I ^ ^ l ^ ^ § ^ ^ ^ i ^ ^ € ^ ^ B f e ±^3}- DUPIC

4 4 5.797X10" and 4.437 xlO11 fissions/cm3 sec o|t:f.

, DUPIC

^ ^ l lH, DUPIC

, DUPIC Jt-yil ^-f, ^*J - t^^r 4

DUPIC

DUPIC in^]*] #^^]-3S|l^]l- ^-*> # ^ ^ 1 " ^ : 3.i& 4.2-561)

^ : S 4.2-26H ^SdL^>. S 4.2-2^1^

(cosine-shape) ±3. a o v # ^ ^ *^ R$ ^ [ ^ ^ i ^ (disc

- 423 -

KAERI/RR-1999/99

source) o]z\ 7\ii\ (flux) £ t ^ s f £0}

(314 cm)

l (1600 cm)

= 0.0193 ol

4.2-2^ n]

(water region) o] ^ 7 ] rcfl ofl

4.2.3.3

4.2-3ofl

^ £ ^ S 2.iofl

DUPIC

DUPIC

DUPIC 7

DUPIC

A

DUPIC

^ H ^ 4.2-82} JE 4.2-2ofl

length correction factor) -if

factor)#

(finite

(non-uniform source correction

- 424 -

KAERI/RR-1999/99

? B l DUPIC

(0.9958)^- £cK € ^4.2-2^

25/iSv/hr# ^ S i J l , o]o]] tf|Sj]Af^ CANDU

CANDU ^4x]-S<H| ^ ^ ^ DUPIC

4.2.3.4

ANISN

fe o]# ^ ^ ^ f e ^ ^ S-^A^ ^*<Hxl ^ ^ ^ w^s . ^6\7]7\

DUPIC

DUPIC Jn^Sj ^ ^ W * # ^ ^ 3 E f e ZL^ 4.2-9ofl £A]^f^t;].. o]

ZLQ 4.2-10c>fl

fe 0.9534olJL, ^ ^ - f e

DUPIC ii^ofl cBtl H ] ^ ^ ^ ^ i ^ ^ l ^ f e 4 4 0.6805 } 0.7644o]r:f.

- 425 -

KAERI/RR-1999/99

4.2.3.5 *F

44 9

4.2-llofl

DUPIC

0.9944olt:f. *J<£-f5^ ii^^J- DUPIC

fe 4 4 0.8653^} 0.8637 °]&i:}. 5 . 3.2^1 A-] j t

DUPIC

4.2.4

] c||tb ^ * f # (heat load) # ^<g«fee(| ^A*fcK 4.2.1^^]^ <&&# Hf f

1, DUPIC ^ ^ S 7 f ^ -^s l CANDU

6l-g"*H DUPIC

4.2.4.1

4. 4 7H 5]^4 *1| *1 4 ^ o #^

DUPIC ii^oil cflsH 4 4

- 426 -

KAERI/RR-1999/99

5.797X10" £} 4.437X10" fissions/cm3

, DUPIC i ^ M cfl > %• <g-§Ml- ^S-fe ^ - ^ H i ? iiôKI «1*B ^ 17%

(form factor)# # # * M 4

^ f DUPIC iiôfl cHtt Êfl^l^fe 4 4 0.711 2f 0.862

. ANISN 3 . H S m

Total Nuclear Energy Deposition (MW)

= [Neutron Heating (mW/cm2) + Gamma Heating (mW/cm2)]

x [Area of two end shields (2 x 380 channels x 28.5752 cm2/channels)]

X [Radial form factor at core end (0.711 for Natural Uranium and 0.862 for DUPIC)]

x[10"9 MW/mW].

44.2-6^1 ô]x\ 6iu}. 0} S S - J f e ] , DUPIC

11%, 29% 9J 31%

(2158.5 MWth) 3} H]J2.*>^, DUPIC it^©fl &<>H ^^>^N1^1] 4

4.2.4.2

ANISN 3 . ^ ^ ^^*}$Elr:K ©]

- 427 -

KAERI/RR-1999/99

., DUPIC

# i ^ ^ ? fe ^ ! 35%

± DUPIC i i ^^ l tBt> ^BH^^fe 4 4 0.635 ^ 0.794 o|u|-.

ANISN 3.B.S. n*ys± 2r&*W 4 * ^ 4

Total Nuclear Energy Deposition (MW)

= [Neutron Heating (mW/cm) + Gamma Heating (mW/cm)]

x[Core Length (594.36cm)]

X [Axial form factor at core end (0.635 for Natural Uranium and 0.794 for DUPIC)]

x [10'9 MW/mW].

, DUPIC i i^£) ^ -^^M 7i] <Hl ^ - ^ 5 ) ^ #<>llM*fe ^«?l

. ^ 1 ^ 4^}. H5iS.S, o] ^2}S.-?-Bl, ^^|j 7}%-^-9l CANDU

DUPIC «}^S7i- J- SlSEl-i- ^-f, ^ ^ - f e f e ii-y^]- a]ja*H

4.2.5

2]- DUPIC

DUPIC

, DUPIC

- 428 -

- 429 -

KAERI/RR-1999/99

, DUPIC

CANDU &*}£.*]} DUPIC

4.2-62f 4.2-7

, DUPIC

KAERI/RR-1999/99

Table 4.2-1

Summary of CANDU Primary Shield Thickness and Design Criteria

Shield System

End Shield

Side Shield

Top Shield

Bottom Shield

Composition and Thickness

- the calandria side tube sheet 5.08cm thick- the carbon steel balls and light water region 78.74cm thick- the fueling machine side tube sheet 7.62cm thick- the concrete containment wall 137cm thick

- vault water 122cm thick- vault concrete 122cm thick- air gap 7.6cm thick- reactor building cross-wall concrete 160cm thick

- the vault water 532cm thick- the reactivity mechanism deck 122cm thick

- the vault water 220cm thick- the vault concrete 122cm thick- the concrete ceiling 122cm thick above room Rl-012

Design Criteria(In Operation)

< 6//Sv/hr

< 25 ju Sv/hr

< 250// Sv/hr

< 25//Sv/hr

- 430 -

KAERI/RR-1999/99

Table 4.2-2

Comparison of Dose Rates through Primary Shields

(Unit : ^Sv/hr)

End Shield

Side Shield

Top Shield

Bottom Shield

Core Type

Natural Uranium

DUPIC

Natural Uranium

DUPIC

Natural Uranium

DUPIC

Natural Uranium

DUPIC

Neutron

1.2624

1.6026

4.183E-9

4.076E-9

2.523 8E-20

2.3731E-20

2.128E-12

2.2719E-12

Gamma

55.7034

66.4504

79.1951

57.0177

26.8640

18.6702

49.9925

39.3647

Total

56.9658

68.0530

79.1951

57.0177

26.8640

18.6702

49.9925

39.3647

Calibrated*

1.0801

1.3134

69.0740

51.8177

18.2810

13.6050

43.4485

34.1528

*The attenuation factor is considered for the end shield which the finite length

correction factor and the non-uniform source correction factor are considered

for the side, the top the bottom shield system.

- 431 -

KAERI/RR-1999/99

Table 4.2-3

Number of Meshes and Dimensions for side shield Calculation

Region

1

2

3

4

5

6

7

Total

Region ID

Core

Reflector

Calandria Wall

Vault Water

Liner

Concrete I

Concrete II

Number ofMeshes

110

36

5

80

5

40

80

356

Thickness(cm)

314.30

65.46

2.858

121.92

0.635

121.92

160.02


314.30

379.76

382.62

504.54

505.17

627.09

787.11

- 432 -

KAERI/RR-1999/99

Table 4.2-4

Number of Meshes and Dimensions for Top Shield Calculation

Region

1

2

3

4

5

6

7

8

9

10

11

12

13

Total

Region ID

Core

Reflector

Calandria Wall

Vault Water(l)

Vault Water(2)

Vault Water(3)

Vault Water(4)

Air Gap

Steel Lower Deck Plate

Ordinary Concrete

Steel Upper Deck Plate

Air Gap

Steel Tread Plates

Number ofMeshes

110

35

5

160

200

200

140

2

5

27

5

2

10

901

Thickness(cm)

314.30

65.46

2.86

121.92

152.54

152.54

104.81

30.48

2.54

71.12

2.54

35.56

10.16


314.30

379.76

382.62

5.4.54

657.08

809.62

914.43

949.91

947.45

1018.57

1021.11

1056.67

1066.83

- 433 -

KAERI/RR-1999/99

Table 4.2-5

Number of Meshes and Dimensions for Bottom Shield Calculation

Region

1

2

3

4

5

6

7

8

Total

Region ID

Core

Reflector

Calandria Wall

Vault Water

Vault Water

Liner

Concrete I

Concrete II

Number ofMeshes

110

36

5

80

65

3

40

40

379

Thickness(cm)

314.30

65.46

2.858

121.92

97.79

0.635

121.92

121.92


314.30

379.76

382.62

504.54

602.33

602.97

748.89

846.81

- 434 -

KAERI/RR-1999/99

Table 4.2-6

Total Heating in Two End Shield Components during Reactor Operation

Component

Calandria SideTube Sheet

Carbon Steel Ball/Water Region

Fuelling MachineTube Sheet

Natural Uranium

Heat Rates (mW/cm2)

Neutron

1.74E+1

1.07E+2

2.35E-5

Gamma

2.76E+3

1.96E+3

9.06E-3

Total

2.79E+3

2.07E+3

9.08E-3

Total(MW)

1.23E+0

9.14E-1

4.01E-6

DUPIC

Heat Rates (mW/cm2)

Neutron

3.46E+1

1.36E+2

2.98E-5

Gamma

2.57E+3

2.25E+3

1.09E-2

Total

2.60E+3

2.38E+3

1.09E-2

Total(MW)

1.39E+0

1.28E+0

5.85E-6

- 435 -

KAERI/RR-1999/99

Table 4.2-7

Total Heating in Side Shield Components during Reactor Operation

Component

Reflector

Calandria Shell

Vault Water

Steel Liner

Concrete

Natural Uranium

Heat Rates (mW/cm2)

Neutron

1.71E+6

4.19E+2

2.42E+2

1.80E-6

1.01E-4

Gamma

1.03E+7

2.83E+6

1.48E+6

5.32E+3

3.12E+4

Total

1.20E+7

2.83E+6

1.48E+6

5.32E+3

3.12E+4

Total(MW)

4.52E+0

1.07E+0

5.60E-1

2.01E-3

1.18E-2

DUPIC

Heat Rates (mW/cm2)

Neutron

1.67E+6

3.04E+2

2.24E+2

1.75E-6

9.86E-5

Gamma

8.50E+6

2.06E+6

1.08E+6

3.85E+3

2.25E+4

Total

1.06E+7

2.06E+6

1.08E+6

3.85E+3

2.25E+4

Total(MW)

4.80E+0

9.71E-1

5.09E-1

1.82E-3

1.06E-2

- 436 -

KAERI/RR-1999/99

Bottom Shield i

InaccessibleArea DuringOperation

Accessible Area Dunng Operation

\ D;O Reflector(65.48cm)

Fig. 4.2-1 CANDU Primary Shield System (Not Scaled, Unit : cm)

- 437 -

0.40

KAERI/RR-1999/99

10- 10" 10'Energy, (MeV)

Fig. 4.2-2 Fission Neutron Spectrum for both Natural Uranium and DUPIC Fuel

- 438 -

KAERI/RR-1999/99

otoCOCM

to

VSide ShieldSource Term

Top ShieldSource Term

End ShieldSource Term

4

297.18 cm

^ • Z

Bottom ShieldSource Term

Fig. 4.2-3 Coordinates Used for Source Term Generation

- 439 -

KAERI/RR-1999/99

800

700-

600-

1.3?-ac

m

500-

400-

300-

200-

1003 4

Plane Number

Fig. 4.2-4 Axial Power Distribution for End Shield Calculation

(Average over Channels L-ll, L-12, M-ll and M-12)

- 440 -

Dose Rate (uSv/hr)o o o o o o o o o o o o o

(TO

4^

o

II

Ien

O

f

I(0

I

(Jl

013a.

Q>ter

oarb

oCO

Ball

V\

Calandria Side Tube Sheet

Fueling Machine

Co,

QCO

SO

I5'3a>2-sall

Ij/

/

/

f

•a:l'n'f

'/!'/'/ill

11IS

, . i , ,

/

3 / y3 Z#\/f

/

/

X / S i d e Tube Sheeta I

I

,j j j j j ,

DcO

/

y

zc3c33

I

imij

/

-

-

-

_

_

KAERI/RR-1999/99

500-

400 -

5 300oQ.

I 200-m

100-

-

•

-

1 1 1

i I

Natural UraniumDUPIC

1

1

-

_

-

1 2 3 4 5 6 7 8

Plane Number

9 10 11 12

Fig. 4.2-6 Bundle Power Distribution on Core Side

(Average over Channels L-l, L-22, M-l and M-22)

- 442 -

KAERI/RR-1999/99

800-

750 -

7 0 0 -

=-• 650 -

ower

« 6 0 ° -c

m 5 5 0 -

500-

4 5 0 -

1 1 1 1 1 1 1 1 1 1

i__^_r-r~J -

; ;-

-- •

DUPIC

-

-

11 10 9 8 7 6 5 4 3 2 1

Channel Column Number

Fig. 4.2-7 Radial Power Distribution for Side Shield Calculation

- 443 -

KAERI/RR-1999/99

350 400 450 500 550 600 650 700 750

Distance from Core Center (cm)

Fig. 4.2-8 Comparison of Dose Rates through Side Shield

- 444 -

KAERI/RR-1999/99

800-

750-

700-

i 650-

| 600 -a>

§ 550-1m

500-

4 5 0 -

400

_ l I I I I I

-Natural Uranium-DUPIC

K Ji I

H G F E D

Channel Row Number

Fig. 4.2-9 Radial Power Distribution for Top Shield Calculation

- 445 -

KAERI/RR-1999/99

350 400 450 500 550 600 650 700 750 800 850 900 950 1000 1050


Fig. 4.2-10 Comparison of Dose Rates through Top Shield

- 446 -

KAERI/RR-1999/99

850

800-

750-

700-

650-

o0- 600-

I 550-f500-

450-

400

- Natural Uranium• DUPIC

i i i i i r iM N O P Q R S T

Channel Row Number

U V W

Fig. 4.2-11 Radial Power Distribution for Bottom Shield Calculation

- 447 -

KAERI/RR-1999/99

10'1

10°

I ' I ' 1

• Natural Uranium• DUPIC

350 400 450 500 550 600 650 700 750 800


Fig. 4.2-12 Comparison of Dose Rates through Bottom Shield

- 448 -

Heat Deposition Rate (mW/cm3)

ostaneeF

rom

Co

3o(DZJ

&

D

300

CO

o ~

S3-O

8-o

340

w8"

•COO) -o

S3-o

8 -o

/'

Y,,,\

S?8-o32?eel

Balland

Watei

/

/

i

Calandria Side Tube Sheet

/'

//

/ '

/

//

//

/

/

DCTJO

Fueling Machine Side Tube Sheet

i i i , ,

/*

-

zc5LC

ss3C3

*

KAERI/RR-1999/99


350 400 450 500 550 600 650 700 750


Fig. 4.2-14 Comparison of Heat Deposition Rates through Side Shield

during Full Power Operation

- 450 -

KAERI/RR-1999/99

4.3 &} i i l fl> ^Ki §%*

CANDU Qx}S.*\] DUPIC

$1*11

(pressure tube)2} # ^ o ] ^ - (end fitting joint)^ 7}Q * } ^ •§• (groove),

elo} -f-^^1 (subshells)^l- % ^ ^ (annular plates) -M-o|Sj

CANDU ^J^Sofl DUPIC

7| £|3rH DPA (Displacement per Atom; *3:

CANDU ^ H ^

\. o] <&*Wl^ ^ ^ H B l o > ^ * | } (mainshell)

1 1-&1 Pfi^^1 (embedment ring)# ^ . ^ ^ f e

(steel slab shieldHr.}. o] Jf ^ l a f l ^ ^ - 3 . s l S ^ ^ £ 7 > 65.6OC-S.t:l-

K ^-S]J5LS., CANDU 414S&11 DUPIC

DPA

ANISNi)-

4.3.1

- 451 -

KAERI/RR-1999/99

(electronic excitation)* «*lJ>zl ^ $i°M,

(cascade)fe «J:3:(recoil)

1 ^ ^ , DPA

DPA = (4.3-1)

0.8©l

31

NJOY97.95 [Ref.

Cr, Fe

^ 25

40

DPA

ENDF/B-VI

ife

H ^ 4.3-2

4.3.1.1

5.08 cm Jf-n*) *±&2=LT!\O}

78.74 cm - ^

^ 6.8

7.62 cm

>. 4

28.58 cm

sat}, -L

- 452 -

KAERI/RR-1999/99

ield plug

assemblies)^.

^ 4.3-KHl

3 8 0 7 ^

4

MCNP-4B

28.575

^ 3 | - DUPIC

4.3-33|- 4.

WIMS-AECL

.8200, .9185 ^J 1.126254JL, DUPIC

.7869,

.5399, .7792, .9387 ^

4.2.3.1

4.3-l3f 4.3-2ofl 4 4

30% DUPIC

nfl,23%

- 453 -

KAERI/RR-1999/99

4.3.1.2 ^ e > ^

CANDU (shell) 7} $X

304LS , 2.86 cm

337.82

379.73

plate)^

(sub-shell)4 ^ - ^ ^ 1 (main-shell)^. 9"

(annular

1.9

CANDU ^^>S.fe 3 8 0 7 ^ , 4(on-power)

MCNP

1/8

. MCNP-4B

(repeated structure) #*}$. vj|5]

71 g :

4DUPIC

o)

- 454 -

KAERI/RR-1999/99

4 ^<$.£. -S-^ # « j £ WIMS-AECL

3.*\zM*\2\ DPA

S 4.3-3 ] ^M$X^\. ^ - f sHi r i i ^ 4 nlJl*H, DUPIC

DUPIC ^ ^ ^ # ^ ^ fll^

DPA #7>eo># ^ . ^ ^ ttfl, DUPIC «J«lS7l- CANDU

4.3.2

65.6°CS.

- ^ 20 mW/cm2

4.3.2.1

- 455 -

KAERI/RR-1999/99

%• <i%Kg- ANISN 3 H

ZL J 4.2-14^

exp (4.3-2)

H H 4.2-14o)|

ft ^ ^ r 0.0582 Af 0.0578

DUPIC

DUPIC

. ^^-, ^ ^ i - f e f e ^ DUPIC

9.9524 mW/cm2 i{- 7.2552 mW/cm2o|

20

s^ 44

^ 2O.o°c^

54°C ol:n, # S 49"

3.7] 4^<>1| (

* f e A B] (x

627.09 cm),

- 456 -

-PL

KAERI/RR-1999/99

]• (4.3-3)

(57.51), ?2fe

(20.0TC), L ^ : •g-S. 3 . B | H ^ ^ J (281.94 cm), *fe

(17.31 mW/cm°C), ZLe l i <?ofe # ^

. (4.3-4)

DUPIC « | ^ S i l<Hl tflSfl 4 4 23.42 cm^ 19.25 cm^tf. o) nfl, ^ -S

JH^ ^ H ^ S ^ r ^ ^ - f Bfe^- DUPIC ^ ^ 5 . i^^^l tflsfl 4 4 55.51

53.82°C^14. aeiJE^., ^^-fefe^l- DUPIC

7} t

4.3.2.2

fe 122 cm M | I ^ ^ M^ $-ISli-K # ^ ^ ^ ^ J f *l*)*g (support ring)^

S . f e 122 cmJB.i;> ^ ^

(end-wallH 1 ^

<H1A-| i f e H}ir 4 ° 1 , 15 cm ^ | 5 ] ^l^Bo1 *rsfl»]; (support

ring shielding slab)i|- 4 7^^ ^ 1 $ ^}^]^V (curtain shielding slab; ^-V\] 5 cm)©] _SL

- 457 -

KAERI/RR-1999/99

DORT#

BUGLE96 J ^

44

DUPIC

3500 MWd/MTU, DUPIC: 7400 MWd/MTU)<Hl>H WIMS-AECL S H f o)-g- r}<*}

4 ^ * g ^ ^ < i ^35. ^ 5 ^ ZL^ 4.2-45f 4.2-7^1

DUPIC

^ ^ ^ ^ ^ 1 ^ : € ^ f f e f DUPIC0.1011 mW/cm33} 0.1215 mW/cm3^]c>. 4.4.1^<HH -S-^l

^ DUPIC «)<^^ i^^<^] cB*H - g ^ - t 1.0 mW/cm3^l AoV-§-^fe ^ ^ ^ r 4 4

17.1842 mW/cm2^ 17.3109 mW/cm2<>lv\. tc}sH, ^ Aofl-H^ < i ^ ^ : 4 4 1-7376

mW/cm2 5J 2.1030 mW/cm2o|t:>. ^ ^ - f Bffe- Jt^l-^ <i^-^l «1*H DUPIC

20 2

(65.

^ DUPIC

- 458 -

KAERI/RR-1999/99

Table 4.3-1

DPA at Innermost Groove of Pressure Tube to End-Fitting Rolled Joint

during 30 Years Reactor Operation

Natural uranium fuel

DUPIC fuel

DPA

0.40238 (1 ±0.0069)

0.52460 (1 ±0.0147)

Comparison oftolerable time

-

I 23.3%

- 459 -

KAERI/RR-1999/99

Table 4.3-2

DPA at Weld between Heavy Steel Plates used to Construct Calandria Side Tube

Sheets during 30 Years Reactor Operation


DUPIC fuel

DPA

1.16235 (1 ±0.0047)

1.50077 (1 ±0.0119)


-

| 22.5%

- 460 -

KAERI/RR-1999/99

Table 4.3-3

DPA at Corner of Calandria Sub-shells and Annular Plates

during 30 Years Reactor Operation


DUPIC fuel

DPA

0.05298 (1 ±0.0711)

0.05807 (1 ±0.0664)


-

i 8.8%

- 461 -

KAERI/RR-1999/99

Annuls Gas DPA Calculating PoinTLattice Tube *- Calandria side Tube sheet

.-Sleeve Insert Annulus Gas

Coolant End Fitting

Coolant

lugFuelAdaptor Fuel Bundle

Pressure Tube

Fig. 4.3-1 Fuel Channel System

- 462 -

KAERI/RR-1999/99

. r •. _

. • CALANOFUA VAULT ENO WALL : •

SUPPORT SHELL

V

o

REACTIVITYMECHANISMTHIMBLE

ENOSHIELDSHELl

WELDS OUTSIDE OF CALANOHIATUBESHEET ARE NOT INCLUDEDIN THE CALANORIA VESSEL BOUNDARY

SLEEVE INSERT -

SUPPORT•S PLATE

ns=2,, CALAN0R1A ANNULAR p-.ATE V

DPA CalculatingPoint

LAI I ICE TUBE

^FUELLING MACHINE SIDE TUBESHEET

END SHIELD

CALANDRIACALANORIA SIOETUBESHEET

.PRESSURE TUBE

I L JCALANORUTUBE

LEGEND:CALANORIA VESSEL PRESSURE BOUWJARYBALANCE OF ENDSHIELD. ISTRUCTJRALSUPPORT AND SHIELDING ATTACHMENT)

Fig. 4.3-2 Calandria Shell

- 463 -

KAERI/RR-1999/99

Fuel Elements

D2O Primary Coolant

Pressure Tube

Gas Annulus

Xalandria Tube

Moderator

Fig. 4.3-3 Configuration of Natural Uranium Fuel Lattice

- 464 -

KAERI/RR-1999/99

Fuel Elements

D2O Primary Coolant

Pressure Tube

Gas Annulus

Calandria Tube

Moderator

Fig. 4.3-4 Configuration of DUPIC Fuel Lattice

- 465 -

KAERI/RR-1999/99

LEAD AND STAINLESSSTEEL WOOL RADIATIONSHIELDING

RING SHIELDINGSLAB

HEAVYWATERCALANORIA

Fig. 4.3-5 End Thermal Shielding

- 466 -

KAERI/RR-1999/99

4.4

DUPIC

DUPIC

CANDU

DUPIC

^ S i d W M S j DUPIC

DUPIC ^<&g. ^ - ^ 91 ^HHf DUPIC

^ DUPIC

fe DUPIC

DUPIC

ZL

DUPIC

. DUPIC

^ CANDU

DUPIC

. ZLSm DUPIC

DUPIC

DUPIC

}, DUPIC

DUPIC

el*];

^ DUPIC

DUPIC

91fe CANDU DUPIC

- 467 -

KAERI/RR-1999/99

4.4.3^ <Hl4^r CANDU

4.4.1 DUPUC

DUPIC « | ^ S S £ 3 1 3 ^ CANDU-6 Q6a%^S- ^ ^ W . £ W<HM*r DUPIC

DUPIC ajel&Sl ^ W ^l^S^l cfl«j{ ^AU*| - a ^ ^ c } . ^ ^^-fe 43^-

37-g- ^^^.u]-y i# 5:^:^5. -HgSl&i:}. 43-g-

, 37^-

4.4.1.1

35

WOBI #^*ll ^17>33 4-§-5|5d^K ^^r^- 4-g-^f *|^S-S.-f-B| CANDU

4.4-

DUPIC

, DUPIC

C A N D U 4 » ^ f l ^ ^ f ] 1 ^ 4 ^

DUPIC

WOBI Tfl-ihg: CANDU ^^>Sofl ^ ^ ^ DUPIC

- 468 -

KAERI/RR-1999/99

fe 15 MWd/kgHMS}

4.4.1.2 DUPIC

fe WOBI

i f ^ o | 220.4 cm*|

gr 3.5

3 0.4025 cm

, 35 MWd/kg ^ 1 ^ A f s j ^ t : } . PWR

I v f o } ^ ^ . s 4.4-2*1]

0.28 , 37 «J«iS-g- 0.28 mgo]u}-.

10%-b

]-. CANDU-6 ^ ^ ] ^ 1 ^

1 tio^ ^ ^ f S ^ i - i - :g7l7fg- (RBVS; SI

7 1 ^ »£ R-109if R-110# ^ -SM 680 m3/hS., service port room^i

R-115# ^f-*l| 340 595 m3/h

4.4-2^]

(committed effective dose equivalent) £• H50

4.4-2ofl 1, o|

- 469 -

KAERI/RR-1999/99

R-201^ J E ^ J L ^ ^ : ^ 1 ^ ^ ^ . 5 - 5 . <H]^"5lo1- ^ 7 1 ^ A - ) DUPIC

., DUPIC

f. R B V S # ^

j ^ 3 5 . S a ^ . - < H 7 H f e ^ ^ ^ r ^ ^ ^ l R-2102} 7 ^ ] ^ ^ ] ^ (stairwells)

R-102 # j j l l f ^ ^ l

}. DUPIC

DUPIC

- DUPIC m&

, QAD-CGGP [Ref. 36]

, WOBI

5g-f, 30 cm (1 foot) I M * ! ^ ^ - ^ 5.5 m (ig feet)

5.5 m <i<H^l ^-^1 ^ l ^ ^ d ^ l - ^ : 43 cm (17")^| -g-SBlH ^ o ] i 4 12.7 cm

22

>. DUPIC

4.4-16]] Ji&t]-. ^ ] # *M ^$a# ^4", DUPIC AJ

3.2 ^Sv/h O|JL, ^ - f s f e ^ f § ^ -^QS. t\^S>\ &<%*} ^ ^ 1 - ^ 1.32 //Sv/h

- 470 -

KAERI/RR-1999/99

^ : DUPIC &

DUPIC

4.4.1.3 DUPIC

c M & 1.3

DUPIC A}^~f «}<>iS^?-B^ - § ^ ^ -*1%*#^: 163.6

fe^>^, DUPIC ^ f g - ^ ^ < ^ ^ ^ ^ % * # ^ ^ ^ ^ - S f e ^ - C f Afl71 oH

( a ^ 4.4-1

4.4.2 DUPIC

l ^ f ^ g 1 1 i g H DUPIC

K DUPIC ^«^S.7> ° > # < H 1 ^ ^ >

\dz #<>}&£}. a e j u } , DUPIC

ofl CANDU ^ #

DUPIC

(IAEA)^ Safety Series

No.6 [Ref. 37]if n ]^ - ^*H^-*fl£]*l5] (USNRC)^ 10CFR71 [Ref.

I ^ § ^ 2 mSv/hr <>1*H3. 2 m

^ : 0.1 mSv/hr 6}*f^ ^»,

4.4.2.1 DUPIC

, 4 7B 1 i%<&3. v}»^ 7

- 471 -

KAERI/RR-1999/99

^ Sa-fe KSC-4i+ KSC-77V

S H DUPIC

, o]

3.0^ CANDU

DUPIC

DUPIC ^ < ^ 5 . ^ ;g-f, JL

•> ^1 °flM^l (0.01-0.575 MeV)

(0.85-2.75MeV) ^ f > i ^ ^ ^ ^

(> 3.5 MeV) ^ } ^ i ^ e .

^.^^ <$ 40-50% VJ-^-'^,

DUPIC

CANDU >

CANDU -?

7.06 mo]

KSC-4#

^ <$ 2.2 molt;},

CANDU

DUPIC

CANDU

4.85 41-42

DUPIC

ZL ^4S.A-1, DUPIC

DUPIC

DUPIC

DUPIC

DUPIC 120

^ S 2 7fi5]

$1

15 cm

- 472 -

KAERI/RR-1999/99

10 cm

4.4.2.2

DUPIC

0RIGEN2

DUPIC 35000 MWd/tHM5]

L, DUPIC fe CANDU

DUPIC

Afl7]7]-

K 0RIGEN2

10

15000

DUPIC

MWd/tHM^l

MWd/tHM£)

4000 MWd/tHM

DUPIC

(over-burned) DUPIC

fe CANDU 3J*fc§.oM 7500

CANDU ^J^fSofl-H 19000

DUPIC

^ DUPIC

0, 10, 20, 30, 40 £ 50

7K

238U

- 473 -

KAERI/RR-1999/99

<£ m^<£T$ 2 - 3 7 ^242Cm, 244Cm, 238Pu5f ^ ^ . ^ - e f e ^ 4 : ^ ( f f , n)

ORIGEN2

ZL ^3}fe 7 l ^ ^ *i£-feHr *h%-^ « )^5 . , DUPIC

DUPIC ^ ^ § ^ «}^Sof] cH*|| 4 4 S 4.4-4, 4.4-5 ^ 4.4-6^| ^<>f^ 5ir:>. o]

ORIGEN2

Watt

/ ( £ ) = Cexp(-^-)sinh\T6£5 (4.4-1)

, a$ b^ 252Cf<Hl t|(*l| 4 4 1.025 MeV # 2.926

o] Q& Los Alamos National Laboratory Group T-2£] Madland<H]

n ^ 4.4-3^ 252Cf ^ y i ^ ^ < i ^ ^ # i ^ ^ T ^ , o| ^

^r 51^ BUGLE ^ # £ 4 ^}S^5 | 47$

- 474 -

KAERI/RR-1999/99

^ # S ] i r ^ 7 } 90%

ORIGEN2 4^A^.-f-B

, DUPIC A>^f ^ ^ ^ uj JL^ i t DUPIC

4.4-7, 4.4-8

4.4.2.3

DUPIC

-o| 2 mSv/hr

-§-71 ^^<H1^ 2 m 1<H^1 ^ |^]<H1A^ -a^l-ol 0.1 mSv/hr o|8fo|c

ANISN 7^1^1# ^ t g ^ & c } , ^ssJI l -^^: ^ > ^ l f ^I«> 15 cm

^ M « ^ * M ^ H ^ ^ SJmo]] 10 cm

611\. t t l-sH, ANISN ) ^ : ^

^ 4 4 15, 15 92 I O ^ S . *}&JL, 4 ^^<H1^^ ^^f^ 3 7 1 ^ -c}. *M 78

44

(7200 MWd/tHM), DUPIC ^ < £ 5 . ti^#^^J£ (15000

MWd/tHM) ^ :2.<*[dt DUPIC «J« iS tioV#^4:S (19000 MWd/tHM)ofl

- 475 -

KAERI/RR-1999/99

BUGLE96 ?OHx}g. > ^ ^ # ° l -§ -*H ANISN Tfl^M- -M§*}$ic} . oj-g- * $ £ . § . t:}-

WIMS-AECL

, BUGLE96ofl

, 133Cs, I60Dy, 166Er, 155Gd, 165Ho, 127I, 83Kr,

139La, 143Nd, 105Pd, 147Pm, 103Rh, 147Sm, 99Tc, 131Xe ^*\] V-M BUGLE96

BUGLE96 «?^>S. ^ ^ ^ ^ o]-§-*}<H ^ 1 ^ ] ^ ^ - ^^ t r>7] ^ ^ M , 0RIGEN2 71]

BUGLE965]

K 0RIGEN2 7 ^ 1 ^ ^ . ^ ^ ^

BUGLE963J 475: ^^3- ^9^*>5i t :> . ^ A ^ f ^ ^ * | ^-f, ORIGEN2

205"

, ANISN

, DUPIC

4i DUPIC 1 fll 1^ ^

4.4-10^1 L-j

. (Ref. 28 ^ B

(4.4-2)

- 476 -

KAERI/RR-1999/99

0.4298

(53.5 cm) ojuj.. o] xcfl,

jL^ 2 m <i<H*} *|;g ( L = 225

. o]

( L = 25 cm) 2} -£•

Hr 4

DUPIC

CANDU

. 3 . ^ , DUPIC

2851

CANDU

4.4-11 ofl

(finite length correction factor)#

0.6133o]t:>.

71 §

, DUPIC DUPIC

> ^ DUPIC

4.4-12©«

tl (15 cm

ANISN 711

18 cm

Jit:> ^ 20%

^ 3 cmf-

30

^ CANDU

(p =11.34 g/cm3), g

^ 4 4 ^ 2.5, 0.3, 12.7 £ 1.0 §-©13.,

(p=0.92 g/cm3)cHl

^ ^ 1 7

- 477 -

KAERI/RR-1999/99

20

3 cm

4.4.3 CANDU

DUPIC

CANDU

4.4.3.1

DUPIC

MCNP-4B

43-g- DUPIC

, DUPIC

CANDU

CANDU

37-g-

S.1!^. «.*]- 6 4 4 4 (infinite hexagonal lattice),

(one core-load of the fuel bundles), ^ 4 ^ - ^ - £ #

bundles stacked criss-crossed against each layer) ^

(fuel bundles in the transport module) ^]^\. S.-&

4 7>4 CANDU «|«iS. t r i h ^-, ^ t - ^ ^ S Ao^

43-g- DUPIC

(infinite

7}. -*> 64 4 4

CANDU

H keff , 44

- 478 -

KAERI/RR-1999/99

S. Jfnjofl cH^ ^ ^ 1 1 -f-sj^l H]^ . 37-g- S ^ CANDU ^ ^ S '^}^3\- 43-g-

DUPIC t ^ ^ S c f ^ ] t^m 4 4 0.732f 0.78o]cK «^^] T j ] ^ ^<&

30

31 Cftij:, ^ o ] 31

B]«H Itn} ^^^^1 S ^ ^ ^ i 7}^H15] 10 cm

Jf*> 6 4

51 ?>^ol o.8

CANDU «J«!S c>y^Jf 43-g- DUPIC

o) sxg£. JZ.Q 64

2.4

- 479 -

KAERI/RR-1999/99

2.4

4X1251 fe 48711-S] ^ E f l t N ^ ^

CANDU

10 cm

a e l J L 1.2 g/cm3^

0.5 g/crn3*] ^

4 7M

^ 0.5, 1.0

4.4.3.2

CANDU

H»J MCNP-4B#

KCODE

. 4 ^

. MCNP 3 E

(inactive) VA 4-8-3 (active)

, MCNP 5LEl-b

(estimator)^ ^ ^

CANDU

37-g- (0 MWd/t)

(7200 MWd/t) ^

#i (14800

- 480 -

KAERI/RR-1999/99

MWd/t) ^EflSl 43-g- DUPIC

^ WIMS-AECL

r- 4 CANDU

MCNP

MCNP

DUPIC

ENDF/B-VI Release

NJOY

97.62OJJL, NJOY

DUPIC

MCNP-4B fl f

Afg-*>5jlc>. NJOY 3 H 5 ]

(fractional tolerance)^ 0.001#

300

WIMS-AECL

<I ^ A ^ # (pseudo fission product; PFP)S. T4E}\5C>. WIMS-AECL

MCNP-4B S-H#£ l *H<&5. 2L*i<>\} mi ^ ^ -fr*l*l-7l ^ * M , MCNP

S. ^>^^«H1 cH*> PFP 4 ^ - f e RMCCS *%*}£. ^>^^<H] S % ^ jg^ . ^ ^ -

(average fission product) * r S # °l-§-*H ^^I^f^^f. o]ttfl, MCNP *

4.4.3.3

CANDU

MCNP-4B

, 4 TflAKg;

- 481 -

KAERI/RR-1999/99

7}. ^ - « 6 4

471

500

1000 7H^ # ^ * h 50

MCNP

^ 4.4-4, 4.4-5

64

^K 471

ke//

37-g-

, 43^- DUPIC

H)7f 1.081 l«y

14 mk7f

43-g- DUPIC

, DUPIC

6 4

^ 3000

MCNP

100 7fl51

CANDU

4.4- CANDU

0-8

i t 4.4-14oj| § ^-ffoil 0.8

- 482 -

CANDU

o]

O)

KAERI/RR-1999/99

Jio]*] -, DUPIC

o]

o]v.\. MCNP

1000

2000

4 4 0.1 mm if 0.24 mm

100 7 ^ a ] % ^

2fe S 4.4-15011

CANDU

CANDU

3000

71

. MCNP

100 , 500

}fe 4 4 S 4.4-162f 4.4-

DUPIC J

0.5, 1.0 ^ 0.5 g/cm39l 37-g-

43-g- DUPIC

0.5 g/cm

ZLBll4, DUPIC 1.0 0.5

- 483 -

KAERI/RR-1999/99

1.0 g/cm3

^ ^ ^ g l Htt ^ ^ ^ 1 CANDU

CANDU ^ | |

- 484 -

1.2 g/cm3*]- ^ o ] 1.0 g/cm3

KAERI/RR-1999/99

Table 4.4-1

Percentage of Volatile and Semi-Volatile Fission Products Removed

Element

I

Br

Kr

Xe

Cs

Rb

% Removed

99

99

99

99

90

90

- 485 -

KAERI/RR-1999/99

Table 4.4-2

Actinides Activity and Annual Doses from Airborne Contamination

from Fresh DUPIC Fuel

Isotopes

U232U234U235U236U238

Np237Pu238Pu239Pu240Pu241Pu242Am241

Am242mAm243Cm242Cm243Cm244Cm245Cm246Cm247

Total

After 10-year decay

Inventrory

(g*g)

5.26E-071.20E-029.28E+004.42E+009.39E+024.02E-011.29E-015.79E+002.41E+008.51E-015.16E-015.58E-018.45E-041.09E-012.21E-062.87E-042.26E-021.11E-031.41E-042.29E-06

9.64E+02

Activity(Curies)

1.16E-057.47E-052.01E-052.86E-043.16E-042.83E-042.21E+003.59E-015.48E-018.80E+012.04E-031.92E+008.86E-038.82E-032.19E-027.31E-031.48E-021.83E+001.90E-044.32E-05

9.49E+01

SurfaceContami-

nation(g/m2)

1.15E-132.63E-092.03E-069.68E-072.06E-048.79E-082.83E-081.27E-065.29E-071.86E-071.13E-071.22E-071.85E-102.40E-084.83E-136.29E-114.96E-092.42E-103.08E-115.01E-13

2.11E-04

AirborneActivity(Ci/m3)

3.86E-152.49E-146.67E-159.51E-141.05E-139.42E-147.35E-101.19E-101.82E-102.93E-086.79E-136.37E-102.95E-122.93E-127.27E-122.43E-124.94E-126.10E-106.32E-141.44E-14

3.16E-08

H50Effective

WholeBody Dose

(Sv)

1.44E-051.90E-054.74E-066.93E-057.09E-052.01E-041.57E+002.76E-014.21E-011.20E+001.69E-031.47E+00O.OOE+006.77E-036.33E-044.10E-036.14E-031.52E+001.57E-04O.OOE+00

6.47E+00

D50LungDose(Sv)

123E-041.59E-043.91E-055.74E-045.97E-04O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00

1.49E-03

D50 ...BoneDose(Sv)

3.91E-064.28E-061.07E-061.55E-051.59E-051.84E-052.61E+014.88E+007.45E+002.18E+012.77E-022.60E+01O.OOE+001.20E-015.81E-036.91E-021.05E-012.60E+012.69E-03O.OOE+00

1.13E+02

D50LLINTDose(Sv)

O.OOE+002.16E-085.69E-097.94E-088.21E-08O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00

1.89E-07

- 486 -

KAERI/RR-1999/99

Table 4.4-3

Annual Gamma Dose Rates from Fresh DUPIC Bundle

Distance

Contact

1 foot away

18 feet away

Dose Rate

(mSv/a)

2.96EX104

2.70EX103

9.61

- 487 -

KAERI/RR-1999/99

Table 4.4-4

Neutron Sources from Spent Natural Uranium Fuel According to Cooling Time

(Unit: Neutrons/sec.MTHM)

( a , n) Neutron Source

Cooling Time

PU238PU239PU240AM241CM242CM244

SUB-TOTAL

0 YR

5.116E+041.225E+051.805E+056.772E+032.063E+061.425E+04

2.439E+06

10 YR

5.574E+041.265E+051.8O3E+O52.711E+051.055E+029.729E+03

6.445E+05

20YR

5.151E+041.264E+051.801E+054.302E+051.004E+026.635E+03

7.960E+05

30 YR

4.760E+041.264E+051.800E+055.243E+059.595E+014.525E+03

8.838E+05

40 YR

4.399E+041.264E+051.798E+055.784E+059.167E+013.086E+03

9.326E+05

50 YR

4.065E+041.263E+051.796E+056.077E+058.758E+012.105E+03

9.573E+05

Spontaneous Fission Neutron Source

Cooling Time

U238PU238PU240PU242CM242CM244CM246

SUB-TOTAL

TOTAL

0 YR

1.250E+048.343E+039.516E+058.956E+041.001E+071.716E+061.895E+03

1.279E+07

1.523E+07

10 YR

1.250E+049.090E+039.506E+058.957E+045.119E+021.171E+061.892E+03

2.236E+06

2.881E+06

20 YR

1.250E+048.400E+039.496E+058.956E+044.873E+027.989E+051.889E+03

1.862E+06

2.658E+06

30 YR

1.250E+047.763E+039.486E+058.956E+044.656E+025.448E+051.886E+03

1.606E+06

2.490E+06

40 YR

1.250E+047.174E+039.477E+058.956E+044.448E+023.716E+051.884E+03

1.431E+06

2.364E+06

50 YR

1.250E+046.630E+039.467E+058.956E+044.250E+022.534E+051.881E+03

1.311E+06

2.268E+06

- 488 -

KAERI/RR-1999/99

Table 4.4-5

Neutron Sources from Nominal Spent DUPIC Fuel According to Cooling Time


{a, n) Neutron Source

Cooling Time

PU238PU239PU240AM241AM243CM242CM243CM244

SUB-TOTAL

0 YR

2.947E+071.577E+054.502E+054.252E+064.830E+044.916E+092.273E+068.574E+06

4.961E+09

10 YR

4.378E+071.606E+054.547E+055.124E+064.828E+047.542E+041.782E+065.848E+06

5.728E+07

20YR

4.046E+071.608E+054.576E+055.623E+064.823E+047.120E+041.398E+063.988E+06

5.221E+07

30 YR

3.739E+071.610E+054.595E+055.892E+064.819E+046.802E+041.096E+062.720E+06

4.784E+07

40 YR

3.455E+071.611E+054.606E+056.020E+064.814E+046.499E+048.593E+051.855E+06

4.403E+07

50 YR

3.193E+071.612E+054.612E+056.061E+064.810E+046.209E+046.738E+051.265E+06

4.067E+07


Cooling Time

PU238PU240PU242CM242CM244CM246

SUB-TOTAL

TOTAL

0 YR

4.806E+062.373E+062.506E+062.385E+101.032E+096.836E+06

2.490E+10

2.986E+10

10 YR

7.139E+062.397E+062.507E+063.660E+057.041E+086.826E+06

7.235E+08

7.808E+08

20 YR

6.598E+062.412E+062.507E+063.455E+054.802E+086.816E+06

4.989E+08

5.511E+08

30 YR

6.097E+062.422E+062.507E+063.301E+053.275E+086.807E+06

3.457E+08

3.935E+08

40 YR

5.635E+062.428E+062.507E+063.153E+052.233E+086.797E+06

2.411E+08

2.851E+08

50 YR

5.207E+062.431E+062.508E+063.013E+051.523E+086.787E+06

1.696E+08

2.103E+08

- 489 -

KAERI/RR-1999/99

Table 4.4-6

Neutron Sources from Over-Burned Spent DUPIC Fuel According to Cooling Time


{a, n) Neutron Source

Cooling Time

PU238PU239PU240AM241AM243CM242CM243CM244

SUB-TOTAL

0 YR

2.979E+071.612E+054.921E+052.302E+065.958E+043.651E+091.999E+061.336E+07

3.699E+09

10 YR

3.981E+071.643E+054.994E+053.215E+065.956E+044.141E+041.567E+069.116E+06

5.448E+07

20YR

3.679E+071.645E+055.042E+053.751E+065.951E+043.893E+041.229E+066.217E+06

4.876E+07

30 YR

3.399E+071.647E+055.073E+054.054E+065.945E+043.719E+049.637E+054.240E+06

4.403E+07

40 YR

3.141E+071.648E+055.093E+054.214E+065.939E+043.553E+047.557E+052.891E+06

4.005E+07

50 YR

2.903E+071.648E+055.104E+054.285E+065.934E+043.395E+045.925E+051.972E+06

3.666E+07


Cooling Time

PU238PU240PU242CM242CM244CM246

SUB-TOTAL

TOTAL

0 YR

4.858E+062.594E+062.860E+061.772E+101.609E+091.293E+07

1.935E+10

2.305E+10

10 YR

6.492E+062.633E+062.860E+062.009E+051.098E+091.292E+07

1.123E+09

1.177E+09

20 YR

5.999E+062.658E+062.860E+061.889E+057.485E+081.290E+07

7.732E+08

8.220E+08

30 YR

5.544E+062.674E+062.860E+061.805E+055.105E+081.288E+07

5.347E+08

5.787E+08

40 YR

5.123E+062.685E+062.860E+061.724E+053.481E+081.286E+07

3.719E+08

4.120E+08

50 YR

4.734E+062.691E+062.860E+061.647E+052.374E+081.284E+07

2.608E+08

2.975E+08

- 490 -

KAERI/RR-1999/99

Table 4.4-7

Total Gamma Source Spectrum for Conventional Spent Natural Uranium Fuel

(Unit: Photons/sec.MTHM)

AverageEnergy(MeV)

1.00E-022.50E-023.75E-025.75E-028.50E-021.25E-012.25E-013.75E-015.75E-018.50E-011.25E+001.75E+002.25E+002.75E+003.50E+005.00E+007.00E+009.50E+00

TOTAL

0 YR

2.481E+184.235E+173.744E+173.590E+177.028E+176.172E+177.862E+174.277E+176.561E+176.899E+173.785E+171.441E+177.106E+163.064E+161.521E+166.494E+155.914E+131.315E+1O

8.162E+18

10 YR

4.319E+149.520E+131.102E+148.505E+134.973E+134.109E+134.136E+132.103E+137.510E+143.641E+132.240E+134.713E+113.061E+101.691E+092.188E+081.127E+051.291E+041.480E+03

1.686E+15

Cooling Time

20 YR

3.268E+146.678E+138.069E+136.705E+133.631E+132.729E+133.035E+131.295E+135.644E+I47.721E+128.217E+122.150E+113.118E+073.679E+064.504E+059.656E+041.104E+041.263E+03

1.229E+15

30 YR

2.579E+145.208E+136.247E+135.506E+132.815E+I31.983E+132.355E+139.908E+124.464E+143.807E+123.508E+121.134E+117.754E+061J79E+062.015E+058.550E+049.751E+031.115E+03

9.628E+14

40 YR


7.613E+14

50 YR


6.043E+14

- 491 -

KAERI/RR-1999/99

Table 4.4-8

Total Gamma Source Spectrum for Nominal Spent DUPIC Fuel

(Unit: Photons/sec.MTHM)

AverageEnergy(MeV)

1.00E-022.50E-023.75E-025.75E-028.50E-021.25E-012.25E-013.75E-015.75E-018.5OE-O11.25E+001.75E+002.25E+002.75E+003.50E+005.00E+007.00E+009.50E+00

TOTAL

0 YR


7.584E+18

10 YR


4.487E+15

Cooling

20 YR


2.866E+15

I Time

30 YR


2.241E+15

40 YR


1.805E+15

50 YR


1.476E+15

- 492 -

KAERI/RR-1999/99

Table 4.4-9

Total Gamma Source Spectrum for Over-burned Spent DUPIC Fuel

(Unit: Photons/secMTHM)

AverageEnergy(MeV)

1.00E-022.50E-023.75E-025.75E-028.50E-021.25E-012.25E-013.75E-015.75E-018.50E-011.25E+001.75E+002.25E+002.75E+003.50E+005.00E+007.00E+009.50E+00

TOTAL

0 YR

2.298E+184.219E+173.691E+173.587E+175.865E+175.257E+177.098E+174.265E+176.716E+176.861E+173.738E+171.454E+177.004E+162.902E+161.353E+165.313E+155.091E+131.155E+1O

7.690E+18

10 YR


5.303E+15

Cooling Time

20 YR


3.371E+15

30 YR


2.622E+15

40 YR


2.094E+15

50 YR


1.693E+15

- 493 -

KAERI/RR-1999/99

Table 4.4-10

Dose Rates through Cask Axial Shield

Neutron

Gamma

Total

CoolingTime

0 YR10 YR20 YR30 YR40 YR50 YR



Cask surface(Criterion: 2.0 mSv/hr)

NAT37

3.040E-045.750E-055.306E-054.969E-054.718E-054.528E-05

6.465E+032.761E-021.029E-024.772E-032.482E-031.463E-03

6.465E+032.766E-021.034E-024.822E-032.529E-031.508E-03

DUP15

7.062E-011.847E-021.303E-029.308E-036.744E-034.973E-03

7.977E+032.969E-011.491E-018.650E-025.383E-023.561E-02

7.977E+033.154E-011.621E-019.581E-026.057E-024.059E-02

DUP19

5.452E-012.785E-021.944E-021.369E-029.744E-037.036E-03

7.737E+033.608E-011.878E-011.123E-017.142E-024.768E-02

7.737E+033.887E-012.073E-011.260E-018.116E-025.471E-02

2 m apart from surface(Criterion: 0.1 mSv/hr)

NAT37

9.017E-061.706E-061.574E-061.474E-061.399E-061.343E-06

2.029E+028.617E-043.211E-041.489E-047.739E-054.560E-05

2.029E+028.634E-043.227E-041.503E-047.879E-054.694E-05

DUP15

2.093E-025.472E-043.863E-042.758E-041.999E-041.474E-04

2.504E+029.229E-034.625E-032.680E-031.666E-031.101E-03

2.504E+029.776E-035.012E-032.956E-031.865E-031.248E-03

DUP19

1.616E-028.253E-045.761E-044.056E-042.887E-042.085E-04

2.428E+021.120E-025.821E-033.477E-032.208E-031.473E-03

2.428E+021.203E-026.397E-033.883E-032.497E-031.681E-03

Note) NAT37 : Conventional spent natural uranium fuel (7200 MWd/tHM)

DUP15 : Nominal spent DUPIC fuel (15000 MWd/tHM)

DUP19 : Over-burned spent DUPIC fuel (19000 MWd/tHM)

- 494 -

KAERI/RR-1999/99

Table 4.4-11

Dose Rates through Cask Radial Shield

Neutron

Gamma

Total

CoolingTime





NAT37

2.948E-045.576E-055.145E-054.819E-054.575E-054.391E-05

6.780E+032.871E-021.070E-024.942E-032.584E-031.522E-03

6.780E+032.876E-021.075E-024.990E-032.630E-031.566E-03

DUP15

6.726E-011.758E-021.241E-028.864E-036.422E-034.736E-03

8.365E+033.054E-011.519E-018.777E-025.442E-023.584E-02

8.366E+033.230E-011.643E-019.663E-026.084E-024.058E-02

DUP19

5.191E-012.652E-021.851E-021.304E-029.278E-036.700E-03

8.114E+033.695E-011.907E-011.137E-017.201E-024.787E-02

8.114E+033.960E-012.092E-011.267E-018.128E-025.457E-02


NAT37

6.540E-051.237E-051.141E-051.069E-051.0I5E-059.741E-06

1.800E+037.504E-032.786E-031.281E-036.636E-043.860E-04

1.800E+037.516E-032.797E-031.291E-036.737E-043.957E-04

DUP15

1.492E-013.901E-032.754E-031.967E-031.425E-031.051E-03

2.221E+037.492E-023.624E-022.045E-021.242E-028.034E-03

2.221E+037.882E-023.900E-022.242E-021.384E-029.085E-03

DUP19

1.152E-015.884E-034.108E-032.892E-032.058E-031.486E-03

2.154E+038.915E-024.465E-022.605E-021.621E-021.062E-02

2.154E+039.504E-024.876E-022.894E-021.826E-021.211E-02

Note) NAT37 : Conventional spent natural uranium fuel (7200 MWd/tHM)

DUP15 : Nominal spent DUPIC fuel (15000 MWd/tHM)

DUP19 : Over-burned spent DUPIC fuel (19000 MWd/tHM)

- 495 -

KAERI/RR-1999/99

Table 4.4-12

Dose Rates through Cask Radial Shield Depending on Gamma Shield Thickness

Gammashield

thickness

16 cm17 cm18 cm20 cm


Neutron

1.681E-021.606E-021.538E-021.410E-02

Gamma

2.000E-011.461E-011.176E-019.257E-02

Total

2.168E-011.621E-011.330E-011.067E-01


Neutron

3.760E-033.625E-033.500E-033.261E-03

Gamma

4.795E-023.416E-022.696E-022.088E-02

Total

5.171E-023.778E-023.046E-022.414E-02

- 496 -

KAERI/RR-1999/99

Table 4.4-13

of One Core-Load of Fuel Bundles in Contact with Each Other

Natural uraniumdischarged fuel bundle

DUPIC freshfuel bundle

DUPIC dischargedfuel bundle

H2O reflector(10 cm thickness)

0.78704 ±0.00059

0.88253 ±0.00072

0.69738 ±0.00056

Against concretecorner

0.78578 ±0.00062

0.88370 ±0.00063

0.69824 ±0.00060

- 497 -

KAERI/RR-1999/99

Table 4.4-14

keff of One Core-Load of Fuel Bundles in 0.4 mm Separation

(Moderator-to-Volume Ratio = 1.0811)




H2O reflector(10 cm thickness)

0.80515 ±0.00062

0.93746 ±0.00065

0.73370 ±0.00059

Against concretecorner

0.80557 ±0.00058

0.93942 ±0.00066

0.73450 ±0.00063

- 498 -

KAERI/RR-1999/99

Table 4.4-15

keff for Fuel Bundles Infinitely Stacked Criss-Crossed




Bundle-to-bundle and layer-to-layerseparation

1.0 mm

0.80651 ±0.00053

0.93897 ±0.00057

0.73592 ±0.00048

2.4 mm

0.80268 ±0.00051

0.94879 ±0.00054

0.74019 ±0.00050

- 499 -

KAERI/RR-1999/99

Table 4.4-16

keff for Fuel Bundles in Single Transport Module




Light water density

0.5 g/cm3

0.63624 ±0.00060

0.75900 ±0.00066

0.60487 ±0.00055

1.0 g/cm3

0.62829 ±0.00055

0.79165 ±0.00067

0.59191 ±0.00059

1.2 g/cm3

0.60169 ±0.00056

0.76919±0.00071

0.58257 ±0.00054

- 500 -

KAERI/RR-1999/99

Table 4.4-17

keff for Fuel Bundles in Infinite Transport Module




Light water density

0.5 g/cm3

0.77181 ±0.00058

0.92368 ±0.00067

0.72080 ±0.00058

1.0 g/cm3

0.66095 ±0.00057

0.83406 ±0.00067

0.63583 ±0.00054

1.2 g/cm3

0.62245 ±0.00055

0.79673 ±0.00065

0.60388 ±0.00055

- 501 -

KAERI/RR-1999/99

1.00E+08

1.00E+07

1.00E+06

1.00E+05

- —— . — - — . - —

-—. - «

• —

mm—

.

— —

— —

Spent PWR

Spent DUPIC

-— " Nat. Uranium

— • —

t

100 1000

Days

10000

Fig. 4.4-1 a-n and Fission Neutrons from Fuel Bundle after 10-Year Decay

- 502 -

KAERI/RR-1999/99

Seal weld

56 cm

Base plate

Lift shaftCover

Cylindrical wall

Steel grid Fuel bundle

Diameter 107 cm

Fig. 4.4-2 Spent DUPIC Fuel Storage Basket

- 503 -

KAERI/RR-1999/99

>

0.30

0.25-

0.20-

0 . 1 5 -

0.10-

0.05-

0.00

10"° 10-5 10 10'3 10'2

Emergy (MeV)

10° 101

252yFig. 4.4-3 Spontaneous Fission Spectrum of Cf

- 504 -

KAERI/RR-1999/99

1.00

•PHp=0.5g/cm

1.0 1.2 1.4 1.6

Moderator-to-Fuel-Volume Ratio

1.8 2.0

Fig. 4.4-4 keff of Infinite Hexagonal Lattice of 37-Element Standard Natural Uranium

Fuel Bundle at Discharge Burnup State

- 505 -

KAERI/RR-1999/99

1.00-

2 °-95

0.65-

0.600.8

—*~~ PKO = ° - 5 9'cm

PHp=1.2g/cm3

1.0 1.2 1.4 1.6


1.8 2.0

Fig. 4.4-5 kgff of Infinite Hexagonal Lattice of 43-Element DUPIC Fuel Bundle

at Fresh Burnup State

- 506 -

KAERI/RR-1999/99

1.00

0.600.8 1.0 1.2 1.4 1.6


1.8 2.0

Fig. 4.4-6 £e// of Infinite Hexagonal Lattice of 43-Element DUPIC Fuel Bundle

at Discharge Burnup State

- 507 -

KAERI/RR-1999/99

4.5

CANDU

DUPIC

^ DUPIC

DUPIC

71-

DUPIC

CANDU DUPIC

Vl DUPIC

DPA

30%

30

D P A ^ j - ^ . b> 1 0o / o

DUPIC CANDU

DUPIC

DUPIC

DUPIC

}, DUPIC

41 *M, CANDU

- 508 -

KAERI/RR-1999/99

, DUPIC & DUPIC

ANISN 3 E f

CANDU

^ CANDU

, DUPIC

MCNP-4B

CANDU

(0.95)

-, DUPIC

14 mk

CANDU

- 509 -

KAERI/RR-1999/99

4.6 REFERENCES

1. H.B. CHOI et al., "Physics Study on Direct Use of Spent PWR Fuel in CANDU (DUPIC),"

Nucl. Sci. Eng., 126, 80, 1997.

2. J.S. LEE et al., "Research and Development Program of KAERI for DUPIC (Direct Use

of Spent PWR Fuel in CANDU Reactors)," Int. Conf. and Tech. Exhibition on Future Nuclear

System: Emerging Fuel Cycles and Waste Disposal Options, GLOBAL'93, Seattle, USA,

1993.

3. H.B. CHOI et al., "Comparison of Refueling Schemes for DUPIC Core," 4th International

Conference on CANDU Fuel, Pembroke, Canada, Oct. 1-4, 1995.

4. K.Y. KIM et al., "Shielding Design Manual, Part 1 - Reactor Building," 86-03320-DM-001,

Rev. 1, Korea Atomic Energy Research Institute and Atomic Energy of Canada Limited,

1995.

5. K.Y. KIM et al., "Radiation Heating Report," 86-03320-AR-004, Rev. 2, Korea Atomic

Energy Research Institute and Atomic Energy of Canada Limited, 1995.

6. K.Y. KIM et al., "Improvement of Top Shield Analysis Technology for CANDU 6 Reactor,"

KAERLTR-738/96, Korea Atomic Energy Research Institute, 1996.

7. J.J. DUDERSTADT and W.R MARTINE, Transport Theory, Chapter 1, John Wiley and

Sons, New York, 1979.

8. N.M. SCHAEFFER, ed., Reactor Shielding for Nuclear Engineers, Chapters 4 and 7, USAEC,

1973.

9. T.B. FOWLER, "EXTERMINATOR-2: A Fortran-IV Code for Solving Multigroup Diffusion

Equations in Two Dimensions," ORNL-4078, Oak Ridge National Laboratory, 1967.

- 510 -

KAERI/RR-1999/99

10. U. CANALI et al., "MAC-RAD, A Reactor Shielding Code," Euratom, EUR 2152.e, European

Atomic Energy Community, 1964.

11. W.W. ENGLE Jr., "A Users Manual for ANISN - A One Dimensional Discrete Ordinates

Transport Code for Neutron and Gamma Ray Shielding Calculation," RSIC-CCC-307, Oak

Ridge National Laboratory, 1979.

12. W.A. RHOADES et al., "DOT IV - Two-Dimensional Discrete Ordinates Radiation Transport

Code System, Version 4.2," RSIC-CCC-320, Oak Ridge National Laboratory, 1979.

13. Bechtel Power Corporation, "QAD-CG, Combinatorial Geometry Version of QAD-P5A, A

Point Kernel Code for Neutron and Gamma Ray Shielding Calculations," RSIC-CCC-307,

Oak Ridge National Laboratory, 1979.

14. W.A. RHOADES and R.L. CHILDS, "The DORT Two-Dimensional Discrete Ordinates

Transport Code," Nucl. Sci. Eng., 99, 1, 1988.

15. W.E. FORD, "Coupled 100 Neutrons-21 Gamma Ray Group, Pg Cross Section Library for

EPR," ORNL/TM-5249, Oak Ridge National Laboratory, 1976.

16. "ENDF/B-IV Summary Documentation," BNL-NCS-17541 (ENDF-201) 2nd Edition,

Brookhaven National Laboratory, 1975.

17. "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived

from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications,"

RSIC-DLC-185, Oak Ridge National Laboratory, July 1996.

18. P.F. ROSE ed., "ENDF/B-VI Summary Documentation," BNL-NCS-17541 (ENDF-201), 4th

Edition, Brookhaven National Laboratory, 1991.

19. J.F. BRIESMEISTER ed., "MCNP - A General Monte Carlo N-Particle Transport Code,

Version 4B," LA-12625-M, Los Alamos National Laboratory, 1997.

- 511 -

KAERI/RR-1999/99

20. ANS Working Group 6.1.1, "Neutron and Gamma-Ray Flux-to-dose Rate Factors," ANSI/

ANS-6.1.1-1977, American Nuclear Society, 1977.

21. "Design Manual : CANDU 6 Generating Station Physics Design Manual," 86-03310-D000,

Rev. 1, Atomic Energy of Canada Limited, 1995.

22. KEPCO, "Final Safety Analysis Report : Wolsong NPP Units No. 2/3/4 Chapter 4," 1995.

23. J.V. DONNELLY, "WIMS-CRNL, A User's Manual for the Chalk River Version of WIMS,"


24. T.E. BOOTH, "A Sample Problem for Variance Reduction in MCNP," LA- 10363-MS, Los

Alamos National Laboratory, 1985.

25. R.C. LITTLE, "High-Temperature MCNP Cross Sections," X-6-IR-87-505, Los Alamos

National Laboratory, 1987.

26. KEPCO, "Final Safety Analysis Report : Wolsong NPP Units No. 2/3/4 Chapter 12," 1995.

27. D.A. JENKINS and B. ROUBEN, "Reactor Fuelling Simulation Program - RFSP : User's

Manual for Microcomputer Version," TTR-321, Atomic Energy of Canada Limited, 1993.

28. G.H. ROH and H.B. CHOI, "Assessment of CANDU Primary Shield System for DUPIC

Fuel," KAERI/TR-1056/98, Korea Atomic Energy Research Institute, 1998.

29. J.D. KIM et. al., "Calculation and Comparative Analysis of Fe-56 DPA Cross Sections

Processed from Evaluated Nuclear Data Libraries," Proceedings of the Korean Nuclear Society

Autumn Meeting, Seoul, Korea, 1998.

30. L.R. GREENWOOD and R.K. SMITHER, "Displacement Damage Calculations with ENDF/

B-V," Proceeding of the Advisory Group Meeting on Nuclear Data Radiation Damage

- 512 -

KAERI/RR-1999/99

Assessment and Reactor Safety Aspect, IAEA, Vienna, Austria, 1981.

31. R.E. MACFARLANE and D.W. MUIR, "The NJOY Data Processing System Version 91,"

LA-12740-M, Los Alamos National Laboratory, 1994.

32. M.M. EL-WAKIL, Nuclear Heat Transport, Chapter 6, International Textbook Company,

1971.

33.1.C. GAULD, "WOBI-An Integrated 2-Dimensional WIMS-AECL and ORIGENS Burmip

Analysis Code System", RC-1808, Atomic Energy of Canada Limited, 1997.

34.1.E. OLDAKER, "Core Fuel for a CANDU 6 Reactor - 37-Element Bundles", AECL Technical

Specification TS-XX-37000-6, 1989.

35. J.R. JOHNSON and D.W. DUNFORD, "Dose Conversion Factors for Intakes of Selected

Radionuclides by Infants and Adults", AECL-7919, Atomic Energy of Canada Limited, 1983.

36. K.A. LITWIN, I.C. GAULD and G.R. PENNER, "Improvements to the Point Kernel Code

QAD-CGGP: A Code Validation and User's Manual", AECL Report RC-1214, Atomic Energy


37. IAEA, "IAEA Safety Series No.6, Transport Regulations," 1985.

38. National Archives and Records Administration, "Packaging and Transportation of Radiative

Materials, Code of Federal Regulations, Title 10, Part 71," USNRC, 1992.

39. H.Y. GANG et. al., "KSC-4 Shipping Cask Safety Analysis Report," KAERI/TR-137/89,

Korea Atomic Energy Research Institute, 1989.

40. J.H. YOON and J.R. CHOI, "Shielding Analysis for KSC-7 Spent Fuel Shipping Cask,"

KAERI-NEMAC/TR-03/93, Atomic Energy Research Institute, 1993.

- 513 -

KAERI/RR-1999/99

41. KEPCO, "Preliminary Safety Analysis Report for Wolsong Unit 3 & 4," Chapter 9, 1992.

42. K.S. CHUN et. al., "Safety Analysis for the Storage Capacity Expansion of the Existing

Spent Fuel Bay at KNU#3," KAERI/RR-596/87, Korea Atomic Energy Research Institute,

1987.

43. E.A. ENGHOLM, "Shielding Aspects of LWR Spent-Fuel Shipping Cask," Proceeding of

Third International Symposium on Packaging and Transportation of Radioactive Materials,

CONF-710801, 1971.

44. K.T. TSANG et. al., "Storage of Natural-Uranium Fuel Bundles in Light Water : Reactivity

Estimates," Fifth International Conference on Simulation Methods in Nuclear Engineering,

Montreal, Canada, 1996.

- 514 -

KAERI/RR-1999/99

DUPIC n&

- 515 -

KAERI/RR-1999/99

5. DUPIC

DUPIC s^ ! ) iM o

^ * M NUCIRC 3.S.1- 43-*>&i;}.. NUCIRC S H f e

NUCIRC ^

DUPIC

CANDU ^^}S.<Hl DUPIC ^J ^

CANDU # \ (&$ ))H ^

fe DUPIC ^^5.0^) cfl*H

NUCIRC s ^ . # <>l-8-*M DUPIC

. H ^ 4 DUPIC 91

DUPIC «foiSofl 3j-g-*> ^2fo]cf. ixj-eH DUPIC

$]-5}o] ASSERT-PV

DUPIC 91

DUPIC «?«iS.7l- ^ g - ^ ^ S i } H ^ ^ ^ ^ - ^ ^ 1 ^ 3.7)} t\M

- 517 -

KAERI/RR-1999/99

5.1 < i ^ iH i

^r CANDU-6

5.1.1 CANDU-6

DUPIC « J ^ S # ^ ^ « > CANDU

^ CANDU-6

1) ^cfl^-g- ^ ^ ^ ^ K r ^H^-H-^l 23.9 kg/s^ HH 718 kPa

2) ^tflpi-g- ^ ^ - H - 3 o ^ 27.4

3)

5.1.2

NUCIRC S ^ ^ f e

O.lpsi

- 518 -

KAERI/RR-1999/99

5.1.2.1

A P ^ A P W/ actual flow \2 / reference density \r f r i c ref I reference flow j \ l d i t )

W/ actual flow \ / reference density \ref I reference flow j \ actual density )

2 (Af

- 519 -

(5-1-1)

(5.1-2)

Proh = ^1^>€ P f c - A P f c - A P ^ - APof (5.1-3)

AP i e f =AP W c +AP a c c (5.1-4)

5.1.2.2

KAERI/RR-1999/99

— AP elevation (5-1-7)

A P = AP -m+ APorifice+ APventuri+ APorifice+ APventuri+ APnozzle (5.1-9)

7]

NUCIRC

Hi -H i + 1 =1707xP c x[ f (Xi ) - f (X i + 1) (5.1-10)

Pc, x 9i f fe #^ ,

- 520 -

KAERI/RR-1999/99

44

= 100.0 h s a t ) (5.1-11)

(5-1-12)

— « 2 v /JP ^ / actual flow \2 .-^CM><^PrefX( r e f e r e n c e f l o w ) >

reference density \actualdensity )

4M = (1- ar0

(5.1-14)

fe 44

5.1.2.3

NUCIRC

n.% 5.1-20]] . 4

- 521 -

= A4PAsGA6

! = 0.8589, A2 = 0.21967

4 = 0.18396, A5 = 0.62492,

= -0.59870,

6 = -0.44954

5.1.3

5.1.3.1

= 1.0

= 0.06

= 1.0

= 1.0

- 522 -

KAERI/RR-1999/99

o]n) fe Xc-Lb

(5.1-16)

Xc, P, G, Lb,

X c - L b CANDU

(5.1-16W

KAER17RR-1999/99

15 S.

0.13

15 S.

= 0.0003 in

5.1.3.2

= 53.05

190,000 lbm/hr

= 12.3 psi

= 46.71

190,000 lbm/hr

= 0.06 ft2

12

19.5 ft

0.037865 ft2

0.2499 ft

0.00002 in

12

19.5 ft

0.03889 ft2

0.2459 ft

0.00002 in

5.1.3.3

511.8 °F

590.0 "F

- 523 -

= 194.7 psi

5] 7]

KAERI/RR-1999/99

100%

ttfl

- 524 -

KAERI/RR-1999/99

Inlet Feeder

AP

Fuel Channel

Outlet Feeder

Inlet End Fitting Outlet End Fitting

Fig. 5.1-1 Slave Channel Analysis Model in NUCIRC Code

- 525 -

KAERI/RR-1999/99

PathB

P'roh=Prih-(APfa+AP&+AP_,+APJ

Path A is used for the initial evaluation

And Path B is used thereafter.

Fig. 5.1-2 CCP Calculation Scheme in NUCIRC

- 526 -

KAERI/RR-1999/99

2'-0

0.7715'

4.2.1a

Fig. 5.1-3 Geometry of Inlet Feeder to Channel N19

- 527 -

KAERI/RR-1999/99

Inlet HEDGrayloc GEN

GEN

GEN

GEN

GEN

GEN

GEN

50 GEN

GEN55 GEN

GEN

GEN

75 GEN

80 GEN

INLET ENDj

reeder

5

7

9

11

13

14

15

16

1719

20

21

23

24

-0.468

-0.1979

-3.9479

-1.4063

0.0

0.0

0.0

0.0

0.00.0

0.0

0.00.0

0.0

N19,0.293

0.0

0.0

1.4063

4.3021

4.0

4.0

4.0

2.00.0

0.0

0.0

8.7978

SOUTH,0.0

0.5417

0.0

0.0

0.0

0.0

0.0

0.0

0.0-2.3125

-4.2292

-5.02.3689

0.2598

INLET,0.25

0.25

1.25

1.25

1.25

1.25

0.7715

11

CASE12^ -

2

2

2

2

2-*—

K—

33

3

33

3•

1

, HOT, FULL

Bend/Elbow

—Straight Pipe Only

—Existence of Reducer

Material IndexI t ^^ fM k ^ t £4 f^^r M B1 few T^^ ^ B 1 fet^Xi

ivauius oi K^ urvature7 f*n«m*fl inciteMa ^U"UI UUIdlC

V fn-nrriin»tp\. \^r\J Ul UlllalC

X Co-ordinate

Fig. 5.1-4 Feeder Geometry Input of NUCIRC

- 528 -

KAERI/RR-1999/99

5.2 < i ^ n^ ^

5.2.1 ^3} nj fe^

CANDU ^^^<^1 && ^«&S ^ DUPIC *|*iS4l W * H 100% #

5.2.1.1

: DUPIC

DUPIC ^ ^ S ^ | ^ o ^ # ^ ^

5.2-3< l SKHi^r «>Af ^ o | Lll

Lll l

, DUPIC

DUPIC t^*} «>^»y- #^^:Sfe S ^ Jc^sf -^-4*1 ^ ^ 5 . <g^ 61 cf.#^^r DUPIC

, DUPIC

5.2.1.2

NUCIRC

^-eo># 7j]A>iy- ^ ait:]-. H ^ 5.2-4

DUPIC ^ f r ^ ^

- 529 -

KAERI/RR-1999/99

DUPIC ^r 4 4 8214 kg/s^- 8300

kg/s o)c}

-§-^1*1 27.4 kg/s#

23.9 kg/s^ -fr^S a

DUPIC

Stic}.. DUPIC

718 kPa

8689

NUCIRC S S .

5.2.1.3

DUPIC ZL^ 5.2-621- 5.2-

l, DUPIC^f

DUPIC

DUPIC

£• 5.2.2

5.2.1.4

DUPIC

1.5025. H07

. 006 ^fl^^l . ©] ^ J f f e

- 530 -

KAERI/RR-1999/99

1.12

5.2.1.5

4

DUPIC 5.2- DUPIC

5.2-104

. NUCIRC 3J=ir

5.2.2

5.2.2.1

/ s 5.2-I [Ref.

NUCIRC a.

DUPIC

- 531 -

KAERI/RR-1999/99

a)

b)

c)

NUCIRC

JL7] ^-c}. ZLBlJL b) 91 c) % ^ NUCIRC 3 S 5 f ^ ^

, DUPIC

ASSERT-PV 3 H [Ref. 12]

B, DUPIC2]-

ZLQ 5.2-12-fB]

5.2.2.2 «fcg

ASSERT-PV

DUPIC

DUPICNUCIRC vDUPIC A ^ R

- 532 -

KAERI/RR-1999/99

DUPIC

DUPIC

ASSERT-PV

ASSERT

5.2-25}

5.2-3011

1.0

S 5.2-3^

DUPIC

. o] DUPIC

CCP _ A CCP , A FRDUPIC — A NU + A DUPIC

EAEUPIC = 0.93%

- 533 -

KAERI/RR-1999/99

Table 5.2-1

Result of Sensitivity Analysis [Ref. 7]

Parameter

Reference CCP

IF Entrance K

Inlet Feeder K

Inlet Feeder (f.L)

Inlet Feeder (ID)

Orifice (K)

IEF DeltaP

Fuel Bundles K

Fuel Bundle CFL/D

OEF DeltaP

Outlet Feeder (K)

Outlet Feeder (f.L)

Outlet Feeder (ID)

Two-Phase Multiplier (Fuel)

Two-PhaseMultiplier (OEF)

Two-Phase Multiplier (OF)

CHF Correlation

Inlet Temperature

H A Peader-to-Header

C » & A P H-to-Hombined T

Cin&AP H-to-Hombined T

Outlet Pressure

CCP (MW)

9.03

9.03

9.024

9.007

9.021

9.030

9.022

9.011

9.009

8.993

9.002

8.997

9.010

8.975

9.011

8.975

8.889

8.871

8.898

8.852

8.852

8.968

SENSITIVITY

0.

-0.0006 MW/%

-0.0046 MW/%

-0.0218 MW/%

-0.000 MW/%

-0.0008 MW/%

-0.0098 MW/%

-0.0108 MW/%

-0.0037 MW/%

-0.0028 MW/%

-0.0065 MW/%

-0.0496 MW/%

-0.0158 MW/%

-0.0038 MW/%

-0.0110 MW/%

-0.0325 MW/%

-0.0797 MW/%

-0.0033 MW/oC

-0.0623 MW/KPa

-0.0111 MW/kPa

-0.0006 MW/kPa

- 534 -

KAERI/RR-1999/99

Table 5.2-2

Selected Fuel Channels for Radial Correction Factor Calculation

Channel Flow (kg/s)

DUPIC

24.23 (H07)

26.73 (LOS)

25.20 (P08)

11.63 (A14)

24.33 (N04)

Standard

24.94 (O06)

27.20 (L05)

26.80 (M05)

11.71 (A09)

24.63 (N04)

Fuel Channel Conditions

Minimum CPR channel

Maximum channel flow

Maximum channel power

Maximum channel exit quality

Maximum fuel element temperature*

- 535 -

KAERI/RR-1999/99

Table 5.2-3

Radial Correction Factor

Fuel Channel

A09

A14

H07

L05

M05

N04

O06

P08

Channel Flow (kg/s)

11.71

11.63

24.23

26.97

26.80

24.48

24.94

25.20

Critical Channel Power (kw)

CCPNU

6950

6920

11220

12200

12140

11310

11500

11310

CCPDUPIC

6790

6750

11150

12190

12120

11250

11410

11250

Mean Value

Standard Deviation

Radial

Correction

Factor (FR)

0.9770

0.9754

0.9938

0.9992

0.9984

0.9947

0.9922

0.9947

0.9907

0.009236

(0.93%)

- 536 -

KAERI/RR-1999/99

ABCDEFGHJKLMNOP0RsTUVW


1

31993412

3506

3535

33593220

123384

427650775552

58695814

6001

5974

6001

58335808

5729

58705944

6082

5992

5969

5870

5603

4987

4285

3308

2

3534

3911

4217

4420

4596

4559

4477

4209

3965

3501

133281

42845071

5708

59275965

5945

6041

595159775870

59435914

60466004

6053

5939

6037

5644

5082

4193

3245

3

3207

3867

4288

4762

5024

5310

5388

5468

5309

5150

4749

4323

3772

3159

143232

40955008

55975983

5933

6039

5994

6049595960255966

6063

6018

6100

5993

6004

5945

5623

4907

4059

3110

4

3321

3966

4524

5024

5326

5636

5768

5968

5954

5948

5718

5450

4959

4471

3813

3223

15

3894

47065455

58146072

6013

61106000

606060076105

6052

61396087

6091

59775854

5340

4671

3764

5

3202

3962

4562

5102

5340

5631

5725

5885

6008

6149

6070

6059

5684

5373

48794408

37833103

16

3405

43375054

5606

5840

5968

59666017

596160916067

6155

6078

6106

5896

5815

5467

4996

41893320

6

2888

3771

4533

5162

5505

5749

5824

5879

5850

6070

6076

61606074

5999

5678

5404

4944

4452

36352802

17

2928

3749

458751185564

5690

5881

5809591059936136

6080

61315924

5723

5336

4976

4393

3651

2760

7

3425

4285

5094

5544

5897

5902

6036

5953

6036

6027

6146

6092

6158

6047

3973

5769

5536

4963

42433305

18

3248

39364614

5054

5393

55695777

581560626073

6121

5982

5729

5304

4911

4348

3798

3054

8

3847

4741

5394

5868

6007

6080

6042

6072

5993

6084

6038

6131

6075

6166

6031

6049

5807

540546403812

19

33663934

4570

49695374

5570581758936002

5871

57625377

4994

4410

3831

3171

93193

4124

4953

5647

5919

5997

5971

6064

5981

6033

5956

6042

5996

6095

6034

6069

5945

6014

55784964

4033

3149

20

3241

3825

4324

4706

50655244

54295397

5346

50794780

4265

3794

3111

103306

4239

5115

5648

5985

5899

6012

5972

6021

59085942

5874

5988

5978

6079

5987

6010

5981

57055040

4237

3223

21

3488

3941

41674451

45354588

44154231

3907

3521

113351

4315

5027

5606

5809

5878

5934

6043

5932

5903

5739

5799

5801

6017

6011

60635904

5932

55505036

4248

3339

22

321933663524

3483

3372

3168

Fig. 5.2-1 Channel Power Distribution of DUPIC Core (kW)

- 537 -

KAERI/RR-1999/99

ABCDEFGHJKLMNOP

QRSTUV

w

ABCDEFGHJKLMNOPQRSTUVW

1

332535303657364034763269

123389435451265656590359536078608959935865575057325808597560696030588356975240454636892641

2

3617

40644365462947644763461143484002

3516

133316431951725762601059906032604159975932588158775919598460075963587858025335459236352549

3

33043962

4464492152585510565056695538528348724332

37383024

14321241625061571360156034606760716035599959845992601960486045.5979587857725262446934642396

4

3412

40694700

5179555558276035618762356147593655415022

439836902955

15

39084801572259356135616861636089606560796105613161596174606459135622503941993199

5

32804054

47265241

5543

578859486083629763836358622558295373

4835422235352701

16

34534384517156815988608461216076608361496203622762276163596057015298465637812761

6

295238484682

52885700

5890

60206040610462816368638163186074573053044789414232572266

17

29343823464952485653583959655984604762246313632862686028568952694760411932402255

7

34704407520057156027

6126

616661226130619562486269626762015995

57315324467837972772

18

325340184682518854845723588060146228631762956165577553264796419035112684

8

39224820554559626165

6201

61986124610061146139616461906204609159385644505742133209

19

3375402246425112548257505956610861586074586954794968435536562930

9321941725075573060346055

6091

609560606025600860176042607060675999

589757885276448034712400

20

32543901439448435176542555665586546052134807427636912990

10332043255181577260216003

6045605560125947589658915933599760205975589058125344459836402525

21

3543398442834543467846814534427639363455

113391435651295659590759576083

609459975870575557385813598060736034

588757005243454936902642

22

324334453572356034033201

Fig. 5.2-2 Channel Power Distribution of Standard Core (kW)

- 538 -

KAERI/RR-1999/99

0.20

0.15

CD

O

g> 0.10

CD

on

0.05

0.004 5 6 7 8 9

Bundle Position

10 11 12

Fig. 5.2-3 Axial Power Distribution of DUPIC and Standard Fuel for Channel Lll

at 100% F.P. Normal Operating Condition

- 539 -

KAERI/RR-1999/99

ABCDEFGHJ

KLMN0P

QRSTUV

w

ABCDEFGHJKLMN0P

QRSTUVW

1

12.7213.5914.0913.9713.9612.94

1213.1517.7821.8524.6325.9924.7924.4124.2924.0524.1923.9024.3023.6623.4723.9524.4024.5325.8823.7321.4116.2412.78

2

14.2516.3916.2517.2920.4920.3618.2517.1116.1714.07

1313.2217.6821.6424.3925.9325.3825.3424.8125.2624.9124.9424.6724.9124.6524.9925.0224.8825.8623.8221.3516.4212.56

3

13.4016.0117.3719.8521.2821.8322.4522.1621.5420.4819.9118.9216.2812.34

1411.6317.9320.3623.9125.3725.0025.0425.3425.2725.3625.1026.3125.1525.2125.0125.1524.8625.0822.8520.1315.3811.91

4

12.9415.5818.6020.9423.1024.0524.5225.0824.9724.3323.4722.5723.5019.0416.0413.25

15

16.9219.4623.2224.1725.1525.1925.3825.8725.7526.0725.6425.8225.2525.4124.8925.1523.7422.1219.3515.23

5

13.0915.8018.7621.4523.0824.1225.5126.0826.7326.3126.3624.8224.2724.1022.4118.2615.2111.94

16

13.3616.8420.2922.6424.0824.1024.4525.3325.5925.4925.4625.1725.0124.3823.9423.5622.1719.4916.6112.75

6

11.8014.1217.2220.6022.8924.1224.7125.3425.4625.6925.8325.2624.6724.5923.4823.0720.4917.8613.8910.79

17

11.7614.1517.1320.7122.7424.2624.5525.4825.3225.8925.6825.4724.5024.7923.3623.2020.4217.9613.8610.84

7

13.3316.9220.1822.8223.9224.2924.2325.5025.4025.6525.2625.3524.7824.5623.7323.6922.0319.5616.5312.78

18

13.0215.8618.6521.5522.9624.2525.4426.1526.6526.5026.2125.0324.1624.1522.3918.3415.1812.00

8

16.9819.3823.3624.0125.3424.9925.5925.6825.9425.8725.8225.6225.4325.2025.0524.9723.8721.9919.4015.17

19

12.8815.6418.5021.0722.9924.1924.4025.2524.8324.5323.3422.7223.4719.1216.0113.31

911.6717.8920.5023.7925.5624.8125.2525.1325.4625.1625.2826.1825.3425.0125.2024.9525.0124.9222.9520.0215.4311.87

20

13.3716.0817.3119.9621.2021.9722.3522.3321.4420.6319.8518.9816.2612.38

1013.1917.7621.5424.5425.7725.5725.1525.0225.0725.0924.7524.8524.7124.8524.7825.1924.7125.9823.6921.4316.3412.59

21

14.3116.3416.3217.2320.5220.3418.3617.0716.2414.05

1113.2017.7121.9524.5526.0624.6224.6024.0724.2424.0124.0924.1323.8523.2724.1624.2024.6925.7923.8421.3316.3012.75

22

12.6913.6714.0614.0413.9412.98

Fig. 5.2-4 Channel Flow Rate of DUPIC Core (kg/s)

- 540 -

KAERI/RR-1999/99

ABCDEFGHJKLMNOP

0RSTUVW

ABCDEFGHJKLMN0P

QRSTUV

w

1

12.6513.5114.0313.9813.9412.98

1213.2417.9222.2625.1326.6825.2625.0624.8024.8524.8224.7524.9524.4924.0924.7925.0725.4526.6024.5221.9316.9913.09

2

14.2616.3616.2917.2520.6920.5018.3317.1416.3214.16

1313.2717.8921.9325.0226.5826.1925.9825.6825.9525.8025.6525.5725.6425.5925.7625.9925.7026.7524.6021.9917.0312.95

3

13.4116.0517.3519.9621.3421.9922.5022.3521.5920.6920.1019.1716.4512.53

1411.7218.0820.7024.3526.2025.5825.8726.0126.1626.0726.0026.9926.0925.9525.9725.9425.8325.9423.7520.7616.0412.25

4

12.9215.5818.6021.0823.2824.3924.7525.4925.1624.6923.6622.9923.9019.4016.2913.50

15

17.0919.6123.6924.6325.9225.6926.2026.5326.6126.7426.5226.5026.0926.0725.7826.0024.7722.8519.8615.63

5

13.0715.8118.8021.7023.3024.5325.9826.6627.2026.8026.6825.3224.7024.5922.7818.6915.5512.32

16

13.3817.0020.4423.1424.4824.6424.8826.0826.1826.2526.0225.9225.5125.1024.5224.4422.9320.2917.1713.19

6

11.8014.1517.2720.8123.1124.5725.0725.8725.8026.1726.1325.6824.9425.2624.0023.6921.0318.4614.4411.10

17

11.8214.1817.3120.8923.2024.6725.1825.9625.9026.2926.2525.8025.0525.3524.0623.7121.0518.4614.4511.10

7

13.3616.9720.3723.0624.3924.5524.7726.0026.0926.1725.9425.8325.4325.0224.4524.3922.9020.2717.1613.18

18

13.1115.8818.8721.7923.4024.6426.0526.7227.3126.9426.8125.4624.7924.6122.7918.7015.5612.33

8

17.0819.5823.6524.5725.8525.6126.1226.4826.5726.7026.4626.4426.0326.0225.7325.9824.7522.8419.8615.63

19

12.9715.6618.7021.2323.4024.5324.9025.6525.3424.8623.8123.1023.9119.4316.3013.50

911.7118.0820.6724.3126.1625.5425.8225.9626.1226.0325.9626.9826.0525.9125.9325.9225.8125.9423.7420.7616.0412.25

20

13.4516.1317.4620.1021.4822.1522.6722.5221.7520.8220.1719.2116.4812.55

1013.2617.8921.9225.0026.5726.1725.9525.6525.9325.7825.6325 .5625.6225.5725.7425.9725.6926.7424.6021.9917.0312.95

21

14.3616.4816.3817.3620.7720.6018.4517.2416.3914.22

1113.2417.9222.2625.1326.6825.2525.0524.7924.8424.8124.7424.9524.4824.0824.7925.0625.4426.6024.5221.9316.9913.09

22

12.7513.6514.1314.0614.0113.04

Fig. 5.2-5 Channel Flow of Standard Core (kg/s)

- 541 -

KAERI/RR-1999/99

ABCDEFGHJKLMN0PQRSTUVW

ABCDEFGHJKLMNOP

QRSTUVW

1

5702

5929

6322

6333

6206

6045

12

5909

7408

8355

8909

9292

9123

9251

9144

9215

9164

9038

9142

9113

9070

9167

9193

9217

9375

8948

8355

7025

6012

2

6286

6882

7121

7346

7854

7794

7476

7181

7005

6354

135862

7399

8295

8956

9140

9207

9172

9179

9295

9302

9241

9275

92479280

9247

9329

9223

9464

8949

8373

7015

5765

3

6095

6771

7270

7856

82428416

8577

8583

8416

8188

8135

7595

6971

5791

14

5543

7342

7941

8842

9119

9118

9191

9175

9355

9297

9398

9477

9362

9309

9316

9252

9258

9227

8884

8077

6780

5614

4

5845

6657

7538

8017

8660

88939072

9277

9151

9085

8861

8667

8685

7797

68505976

15

7091

7790

8762

8913

9182

9145

9222

9353

9388

9478

9475

9473

9374

9328

9264

9215

8998

8579

7984

6675

5

5789

6703

7624

8316

8650

8909

92029312

9463

9432

9376

9195

8956

8954

8550

76276668

5579

16

5944

7321

8006

8650

8955

9034

9038

9309

9294

94619410

9348

9263

9205

9051

8964

8664

7992

7087

5850

6

5436

6322

7369

8183

8674

8953

9011

92479233

9358

94369324

9187

91608970

8788

8165745761765154

17

5444

6295

7390

8163

8692

8931

9029

9228

9254

9334

94619300

9210

9142

8988

8770

8178

7444

6201

5148

7

5970

7302

8025

8637

8980

9010

9064

9284

9323

94279441

9318

9291

91809072

8942

8673

797371065833

18

5794

6678

7642

8301

8669

8888

9218

9297

9485

9411

9397

9174

8973

8934

8562

7615

6691

5577

8

7073

7816

8749

8938

9159

9173

9194

9385

9352

9512

9436

9511

9343

9356

92399236

8975

858979636693

19

5852

6632

7558

8005

8670

8874

90919255

9171

90658874

8660

8700

7778

6872

5973

95534

7365

7924

8858

9098

9145

9162

9208

9319

9339

93609514

9324

9345

9284

9279

9233

92478867809267575624

20

6104

6747

7290

7835

8257

8402

8596

8568

84298174

8152

7581

6992

5786

105873

7382

8310

8936

9160

9182

9202

9145

93359265

9278

92389284

9243

92809300

92479446

8967835970335754

21

6278

6905

7098

7367

7836

7813

7458

7203

6985

6365

115898

7426

8346

8923

9274

9149

9220

9180

9177

9201

9005

91819077

9108

9133

9222

9194

93928931836770036020

22

5714

5913

6345

6312

62306041

Fig. 5.2-6 Critical Channel Power in DUPIC Core

- 542 -

KAERI/RR-1999/99

ABCDEFGHJKLMN0PQRSTUVW

ABCDEFGHJKLMNOP

QRSTUV

w

1

54635684

6071

6062

5972

5772

125669

7188

8175

8793

9190

9019

9100

9016

90569048

8887

9026

89588935

8996

9064

9069

9282

8801

8182

6805

5830

2

6034

6637

6866

7113

7645

7599

7235

6957

6734

6101

135648

7153

8129

8803

9040

9082

9065

9031

9191

91589129

9125

9137

9125

9129

9184

9099

9344

8805

8201

6814

5609

3

5829

6518

7035

7635

8054

8233

8404

8392

8233

7976

7925

7351

6730

5528

145295

7150

7741

8700

8982

9005

9050

9066

9213

91949251

9385

9214

9195

9169

9132

9109

9098

8710

7893

6582

5474

4

5581

6422

7310

7823

8494

8733

8930

9123

9011

8915

8697

8485

85207544

6635

5723

15

6830

7608

8572

8778

9043

9029

9083

9255

92569382

93359370

9227

9207

9113

9090

8830

8402

7791

6466

5

5538

6478

7398

8133

8489

8769

9103

9206

9368

9304

9268

9045

8831

8759

8337

7344

64665351

16

5736

7051

7826

8464

8824

8875

8916

9182

9194

9326

9302

9202

9139

9048

8904

8801

8480

7791

6861

5664

6

5110

6092

7133

7995

8512

8816

8902

9136

9131

9226

9323

9177

9058

9023

88398614

7978

72175989

4966

17

5200

6091

7133

7994

8516

8815

8903

9135

9135

92269322

9178

9057

9024

8837

8617

7977

7216

5989

4967

7

5736

7051

7828

8463

8825

8873

8917

9179

9192

9327

9298

9203

9138

9050

89028804

8476

7792

68605664

18

5539

6477

7401

8134

8492

8768

9104

92079369

9305

9267

9045

8829

8764

8336

7344

6465

5352

8

6831

7606

8570

8779

9041

9030

9083

9256

9256

9382

9336

9370

9228

9 206

9114

9088

8830

8401

7791

6466

19

5582

6419

7316

7821

8495

8736

8931

91239011

8916

8697

8486

8519

7548

6633

5724

95295

7150

7741

8700

8982

9006

9048

9066

92139194

9252

9384

92159194

9170

9131

9110

9097

87107889

6583

5474

20

5830

6517

7037

7639

8055

82328408

83968234

7976

7925

7357

6727

5533

105648

7153

8130

8803

9041

9082

9067

9030

9191

9157

9131

9128

9136

9126

9127

9186

90989344

88058202

6813

5609

21

6033

6641

6866

71157647

7603

7238

6956

6739

6092

115668

7188

8175

8793

9188

9021

9099

9017

9056

9048

8887

9025

8958

8933

89979062

9072

9281

88058182

6805

5830

22

5469

56876077

6067

5973

5775

Fig. 5.2-7 Critical Channel Power in Standard Core

- 543 -

KAERI/RR-1999/99

A

BC

DEF

G

HJKLMNOP

QRS

T

U

V

w

ABCDEFGHJKLMNOPQR

sTUVW

1

1.783 11.737 11.804 11.792 11.847 11.877 1

1

1

121.7451.7321.6461.6041.5831.5691.5411.5311.5361.5711.5571.5961.5531.5261.5071.5341.5441.5961.5961.6761.6391.819

2

.779

.760

.688

.662

.709

.709

.670

.705

.767

.799

131.7871.7271.6361.5691.5421.5431.5431.5191.5621.5571.5751.5611.5641.5351.5401.5411.5531.5681.5861.6481.6731.777

3

1.9001.7511.6951.6501.6401.5851.5921.5701.5841.5901.7131.7411.8311.833

14.714.794.585.580

1.5241.5371.5221.5301.5461.5611.5601.5881.5441.5471.5271.5441.5421.5521.5801.6471.6711.806

4

1.7601.6781.6661.5961.6261.5781.5731.5541.5371.5271.5501.5901.7361.7271.7971.853

15

1.8051.6411.591.533.511.521.509.559.549.577.552

1.5661.5271.5321.5211.5421.5371.6061.7091.774

5

1.8081.6911.6711.6301.6201.5821.6071.5831.5751.5341.5451.5171.5761.6521.7371.7301.7631.799

16

1.7281.6721.5701.5431.5331.5141.5151.5471.5591.5531.5511.5191.5241.5071.5351.5411.5841.6001.6911.762

6

1.8811.6761.6261.5851.5761.5571.5471.5731.5781.5421.5531.5131.5121.5271.5801.6261.6521.6751.6981.839

17

1.8591.6791.6111.5951.5631.5701.4351.5891.5661.5571.5421.5291.5021.5431.5711.6441.6431.6951.6981.864

7

1.726 11.688 11.561 11.558 11.523 11.527 11.502 11.560 11.545 11.565 ]1.536 ]1.529 11.508 ]1.518 11.5191.5501.5671.6071.6741.765

18

1.7851.696 11.656 11.643 11.607 ]1.595 11.596 11.598 11.5651.550 11.5351.5341.5671.6701.7281.7511.7621.826

8]

.822 1

.633 ]

.607 1

.523 1

.524 1

.508 1

.522

.545 1

.561

.564

.563

.551

.538

.517

.5321.5271.5451.5891.7151.756

19

.738

.685 1

.654 1

.611 1

.613 1

.593 1

.563 1

.571 1

.528 1

.544

.540

.6101.7261.7481.7951.883

9.733.787.600.569.537.525

1.534.518.558.548.572.574

1.5551.5331.5381.5291.5531.5371.5901.6301.6761.787

20

.883

.764

.686 1

.665 1

.630 1

.602 1

.583 1

.587 1

.577 1

.609 1

.706 1

.761 1

.827

.859

10.777.741.624.583.530.557.531

1.5311.550L.569L.5621.5721.5511.5461.5261.5531.5381.5791.5721.6581.6601.786

21

.801

.752

.703

.656

.728

.703

.690

.702

.789

.792

11

1.7601.7201.6601.5921.5961.5571.5541.5181.5471.5591.5691.5821.5651.5131.5181.5211.5571.5831.6091.6611.6501.804

22

1.7751.7571.8011.8131.8471.905

Fig. 5.2-8 Critical Power Ratio in DUPIC Core

- 544 -

KAERI/RR-1999/99

ABCDEFGHJKLMN0P0RSTUVW


1

1.6501.6171.6661.6721.7241.772

12.679.658.602.561

1.563.521.503.486

1.517.549.552

1.581.548.501

1.4881.5081.5481.6361.6871.8071.852'216

2

1.675 11.640 11.578 11.542 11.611 11.602 11.575 11.606 11.688 11.724 1

11

13.709.663.578.534.510.522.509.501.539.550.559.559.550.531.525

1.5461.5541.617.656.793.881

2.209

3

.771

.653

.581

.559

.538

.500

.493

.486

.492

.515

.633

.687

.791

.835

141.6561.7241.5351.5291.4991.4991.4981.5001.5321.5381.5521.5731.5371.5261.5221.5331.5561.5831.6611.7731.9072.293

4

11.643 11.584 11.562 11.516 11.535 11.504 11.485 11.480 11.451 11.456 11.471 11.537 11.688 11.706 11.806 11.944 1

15

1.7371.5751.5441.4851.4801.4701.4801.5261.5321.5491.5351.5341.5041.4981.5081.5431.5771.6741.8622.029 :

5

1.694 1.604 1.571.558 1.538 1.521.537.520.494.463.463.459 ].521.621.714.745.836.990

16

1.6521.6001.5051.4961.4791.4651.4631.5171.5171.5221.5051.4841.4741.4741.5001.5501.6071.6791.8222.058 :

6

.768

.589

.529

.517

.500

.503

.485

.518

.502

.475

.470

.4441.440.492

1.549.631

1.6731.7481.8452.201 '

17

1.780.599

1.5401.529.512

1.516.499

1.5331.516.488.483.456.451

1.5031.5591.6421.6831.759

7

1.6451.5911.4971.4861.4701.4541.4521.5051.5051.5111.4941.4741.4641.4651.4911.5421.5981.6721.8152.050 :

18

.708

.619 11.586 11.574 1.555 I.538 1.554 1.537 ].510 ].479 1.478 1.473

1.5351.6361.7281.7601.848

1.855 2.0032.211

8

1.7311.5691.5371.4781.4721.4621.4711.5171.5231.5411.5271.5261.4971.4901.5021.5371.5711.6671.8562.023

19

.661

.602 1

.582 1

.536 1

.556 1

.525 1

.505 1

.500 1

.469 1

.474 1

.488 1

.555 1

.705 1

.723 1

.822 1

.961

91.6521.7201.5311.5241.4951.4931.4921.4931.5261.5321.5461.5661.5311.5201.5171.5281.5511.5781.6571.7681.9032.289

20

.799

.677

.608

.583

.562

.523

.516

.508

.513

.536

.655

.711

.813

.856

10.707.660.575.531

1.5071.5191.5061.4981.5351.5461.5551.5561.5461.5281.5221.5431.5511.6141.6541.7911.8782.207

21

1.7091.6731.6101.5721.6411.6311.6031.6331.7181.754

111.678.657.601.560.562

1.5201.5021.4851.5161.548.550

1.5791.5471.5001.4871.5071.5471.6351.6861.8061.8512.215

22

1.6921.6571.7071.7101.7621.812

Fig. 5.2-9 Critical Power Ratio in Standard Core

- 545 -

KAERI/RR-1999/99

ABCDEFGHJKLMNOP0RSTUVW

ABCDEFGHJKLMNOPQRSTUVW

1

1.541.401.481.81.781.58

122.10.83.06-.57-.52.161.091.061.35.68.71.221.191.611.60.95.77-.56.20-.032.432.30

2

1.39.59

2.351.95-.78-.84.981.101.111.72

131.38.94.22.13-.38.12.07.75.20.53.16.60.31.92.49.65.39.02.26.391.752.10

3

.81

.801.24.55.28.81.601.141.131.51.64-.13.28

2.05

143.91-.171.02.16.19.32.58.19.49.13.58-.54.59.40.79.34.61.251.10.86

2.502.41

4

2.021.79.89.47-.14.09.20.48.43.91.85.70

-1.60.49.45.91

15

.251.07.50.62.62.39.50-.11.13-.18.40.12.76.46.86.271.02.64.791.18

5

1.061.51.91.44-.11.03-.70-.64-.65.03-.25.86.06-.55-.96.761.362.21

16

2.142.481.531.18.771.14.81.34-.05.52.45.87.751.341.021.051.091.881.512.27

6

1.192.922.591.49.63.43.16-.08-.28.21.20.811.00.85.78.10.701.292.272.22

17

1.542.742.961.22.96.12.46-.40.00-.17.50.391.33.451.03-.23.89.91

2.411.83

7

2.302.141.80.821.09.761.22.02.33.20.85.531.181.001.46.791.471.671.882.13

18

1.441.321.24.17.16-.28-.47-.92-.42-.32.01.45.30-.83-.82.411.471.78

8

-.031.30.19.93.28.75.14.25-.21.19.06.51.41.87.54.64.76.99.611.52

19

2.411.551.18.15.11-.23.46.12.69.501.11.32

-1.45.16.57.51

93.56.00.67.42-.14.67.22.56.14.51.22-.20.24.80.44.73.30.60.851.192.302.75

20

1.07.511.48.23.51.45.84.731.351.09.82-.44.421.65

101.58.66.47-.17-.10-.21.42.37.57.17.54.24.70.55.90.31.76-.24.57.16

2.061.91

21

1.05.79

2.012.16-1.07-.69.601.25.731.88

111.821.07-.20-.32-.77.49.721.46.961.06.34.58.80

2.041.181.35.43-.30-.07.23

2.172.55

22

1.721.021.621.43.881.18

Fig. 5.2-10 Channel Exit Quality of DUPIC Core

- 546 -

KAERI/RR-1999/99

ABCDEFGHJKLMNOP0RSTUV

w

ABCDEFGHJKLMNOP0RSTUVW

1

2.482.222.452.441.501.83

122.001.06-.06-.54-.84.31.891.09.74.36-.07-.20.371.25.92.61-.14

-1.57-1.61-2.15-1.35-2.46

2

1.86 :1.39 13.073.01-.26-.121.511.751.141.71

131.54.91.38-.11-.52-.31-.06.16-.09-.20-.26-.22-.15.07-.01-.28-.35-1.29-1.34

3

.41

.24> . l l.11.151.441.54.741.971.861.00-.30-.05.87

143.60.01.93.26-.23.27.12.00-.13-.19-.14-.86-.16.01-.03-.23-.45-.88-.85

-2.01 -1.46-1.67 -1.39-3.01 -3.05

4

2.642.341.67.95.55.51.95.931.231.321.48.71

-1.62-.11-.43-1.06

15

.171.34.40.71,31.56.13-.24-.40-.37-.16-.05.26.30.16-.48-.49

-1.01-1.67-2.32

5

1.571.981.60.78.45.27-.26-.37-.04.45.451.06.26-.87-1.37-.47-.36-1.08

16

2.412.541.881.071.001.131.03.05-.02.21.52.59.891.01.82.03-.11-.23

-1.06-1.90

6

1.643.333.261.811.17.59.56.10.32.57.941.231.64.63.59-.69-.35-.68-.59

-2.35

17

1.483.143.051.58.93.35.31-.14.07.31.69.981.38.41.40-.83-.48-.79-.69

-2.44

7

2.542.682.051.251.191.351.26.25.18.40.72.781.081.19.99.16.01-.13-.98

-1.83

18

1.351.721.34.51.17-.02-.52-.63-.31.16.17.77.02

-1.05-1.53-.62-.50-1.20

8

.251.45.51.84.45.72.29-.10-.26-.23-.02.08.39.43.29-.38-.40-.94-1.60-2.26

19

2.332.001.31.57.20.15.58.56.85.951.13.41

-1.83-.32-.62-1.22

93.67.061.01.34-.15.36.22.11-.03-.09-.04-.78-.06.11.06-.14-.37-.82-.80

-1.42-1.35-3.01

20

1.04.841.66.68.71.991.091.301.541.47.66-.58-.32.61

101.57.94.43-.06-.48-.26.00.22-.03-.14-.20-.16-.09.13.05-.23-.30

-1.26-1.31-1.99-1.65-2.98

21

1.31.87

2.542.47-.67-.521.041.30.731.28

112.021.07-.05-.52-.83.33.911.12.77.38-.05-.18.391.27.94.63-.12

-1.56-1.60-2.14-1.34-2.45

22

1.791.531.821.85.971.32

Fig. 5.2-11 Channel Exit Quality of Standard Core

- 547 -

KAERI/RR-1999/99

Fig. 5.2-12 Enthalpy of DUPIC Fuel in Channel Lll under CHF Condition

- 548 -

KAERI/RR-1999/99

Fig. 5.2-13 Enthalpy of Standard Fuel in Channel Lll under CHF Condition

- 549 -

KAERI/RR-1999/99

Fig. 5.2-14 Void Distribution of DUPIC Fuel in Channel Lll under CHF Condition

- 550 -

KAERI/RR-1999/99

Fig. 5.2-15 Void Distribution of Standard Fuel in Channel Lll under CHF Condition

- 551 -

KAERI/RR-1999/99

Fig. 5.2-16 CHFR of DUPIC Fuel in Channel Lll under CHF Condition

- 552 -

KAERI/RR-1999/99

Fig. 5.2-17 CHFR of Standard Fuel in Channel Lll under CHF Condition

- 553 -

1

KAERI/RR-1999/99

- Standard (CHF at 11.12MW)

- DUPIC (CHF at 11.06MW)

i

300i

400 500 600

Axial Position (cm)

Fig. 5.2-18 Axial Distribution of CHFR in Channel Lll under CHF Condition

- 554 -

KAERI/RR-1999/99

5.3

5.3.1 NUCIRC

5.3-1 [Ref.

U+A

(5.3-1)

(5.3-2)

a ^ 5.3-2 [Ref.

H ^ 5.3-3

[Ref. 7M

L, JE

- 555 -

a(p,G)Lb

b(p,G,DH)+Lb

I

(5.3-5)1-

(n/m)Lbr [(GAhfg)/(mPH)

KAERI/RR-1999/99

(5.3-6)51 ^ i f e ^ (5.1-16H ^o]^ l NUCIRC

^- CANDU

5.3.2 ASSERT 3 . ^

7 = n(G,p)-m(G,p)xc (5.3-4)

River) ^ ^ ^ ^ Ul ^ ^ I f e CANDU ^

^ ^ ^ ^ ^ r CANDU

fe 2L£SL$. 13.9 Mpa^ ^ ^ ^ 17.0 ] ^

- 556 -

KAERI/RR-1999/99

, 6

I*}, ZL

. ©1

5.3-7cHl [Ref. 17]

ASSERT

ASSERT SL^

, ASSERT S H f e

CANDU

ofl^-*>jL

5.3-4^

. ^ . ^ 5.3-6ofl j£

. ASSERT 3 . ^ o]

- 557 -

< IS

8

KAERI/RR-1999/99

o

o

o

SYMBOL

VoAD••

AXIAL HEAT FLUX SHAPE

12 11 UNIFORM

6 ft UNIFORM

6 ft COSINE

12 ft COSINE

12 tt OUTLET PEAK

12 U INLET PEAK

A

A

V

V

V

••

•• a

a

0.18 0.22< X (i) >. LOCAL QUALITY

Fig. 5.3-1 Heat Flux Versus Quality at Location of BT, Freon-114 Annulus Data,

Dl=0.563 in., D2=0.875 in. and 12 ft Heated Length [Ref. 15]

- 558 -

KAERI/RR-1999/99

t 0.20

O

BEST FIT LINE

SYMBOL

Vo•A•

TD

FREON ANNULUS DATA3/106 - 0.540 lb/h-»t2


12 ll UNIFORM

6 ft UNIFORM

12 It COSINE

6 It COSINE

12 M OUTLET PEAK

12 fi INLET PEAK

6 It HALF COSINE

(SHIRALKAFf. 19721

CRITICAL BOILING LENGTH, LB Ht>

Fig. 5.3-2 Critical Quality Versus Boiling Length, Freon-114 Annulus Data [Ref. 14]

- 559 -

KAERI/RR-1999/99

50

X , 0

SYMBOL

o•V

A

MAXIMUM/MINIMUMHEAT FLUX RATIO

1

1.91

2.99

4.7


UNIFORM

EXPONENTIAL DECREASE

EXPONENTIAL DECREASE

SYMMETRICAL CHOPPED COSINE

2.0 2.5

CRITICAL BOILING LENGTH (ml

Fig. 5.3-3 Critical Quality versus Boiling Length Data, 12.6 mm Round Tube 3.66 m

Heated Length [Ref. 16]

- 560 -

KAERI/RR-1999/99

Fig. 5.3-4 Subchannel and Rod Numbering in ASSERT Validation for Standard Fuel

Bundle Simulation [Ref. 17]

- 561 -

KAERI/RR-1999/99

BUNDLE AVERAGE- - SUBCHANNEL 1

SUBCHANNEL 10

0.0 3.0

AXIAL POSITION m6.0

Fig. 5.3-5 Pressure Drop and Void Profiles Simulated by ASSERT Code [Ref. 17]

- 562 -

KAERI/RR-1999/99

300

760

1 '

EXPT •_ ASSERT —

1 /

1

/ROD

1

2

12000 4000

POWER kw

340

300 -

2602000 4000

POWER kw

260

340

2000 4000

POWER kw

300

260

1

1

1 1

<**ROO

|

18

—

2000 2000

POWER kw

Fig. 5.3-6 Measured and Computed Fuel Rod Surface Temperature in Different

Subchannels [Ref. 17]

- 563 -

KAERI/RR-1999/99

1200

E

oLLJ

Q

a.300 -

300 600 900 1200

EXPERIMENTAL HEAT FLUX KW m"2

Fig. 5.3-7 Measured and Computed CHF for Standard Fuel Bundle Experiments

[Ref. 17]

- 564 -

KAERI/RR-1999/99

5.4

DUPIC

DUPIC

. NUCIRGSJE^J

, DUPIC

DUPIC

DUPIC

DUPIC

7)S.

, NUCIRC 3 H

^r 0.9907

0.93%

DUPIC 9| 3E

^ , DUPIC

DUPIC

DUPIC

- 565 -

KAERI/RR-1999/99

1. H. CHOI, B.W. RHEE and H. PARK, "Physics Study on Direct Use of Spent Pressurized

Water Reactor Fuel in CANDU (DUPIC)," Nucl. Sci. Eng., Vol. 126, 1997.

2. C.J. JEONG, J.W. PARK, and J. PITRE, "Preliminary ROP Assessment for CANDU-6 with

DUPIC Fuel," Korea Nuclear Society Conference, 1999.

3. M.F. LIGHTSTONE, "NUCIRC-MOD 1.505 Users Manual," TTR-516, Atomic Energy Canada

Limited, 1993.

4. M. SOULARD, "NUPREP505 Users Manual," AECL Draft, 1994.

5. G.D. HARVEL, "NUCIRC: Part I and II," NUCIRC Training Material at KEPCO, 1999.

6. G.H. RHO, H. CHOI and J.W. PARK, "Sensitivity Analysis on Various Parameters for Lattice

Analysis of DUPIC Fuel with WIMS-AECL Code," Proceedings of the Korean Nuclear Society

Autumn Meeting, Taegu, Korea, 1997.

7. L.C. CHOO, "Critical Channel Power Analysis: Wolsong NPP," AECL 86-03500-AR-021,

1994.

8. G.D. HARVEL, "Wolsong 3, 4, PHT System Flow Verification Procedure," AECL Technical

Document 86-33100-610-001, 1998.

9. G.D. HARVEL, 1998, "Wolsong 3 PHT Flow Verification: Disposition for 100.0% F.P. with

Fuel Commissioning Flows," AECL Memo 86-33100-640-003, 1998.

10. J.W. PARK, "A Subchannel Analysis of the DUPIC Fuel Bundle in CANDU Reactor," Annals

of Nuclear Energy, Vol. 26, No. 1, 1998.

11. J.W. PARK and G.M. CHAE, "ASSERT-PV Simulation of Two-Phase Flow in Horizontal

- 566 -

KAERMRR-1999/99

and Vertical Channels," Canadian Nuclear Society '99, 1999.

12. E.K. ZARIFFEH, G.M. WADDINGTON, N. HAMMOUDA, L.N. CARLUCCI, V.C.

FRISIMA, J.C. KITELEY, D.S. ROWE, and P. PFEIFFER, "ASSERT-PV V2R8 Users

Manual," FFC-FCT-133, COG-97-460, Atomic Energy Canada Limited, 1998.

13. CM. BAILEY and G.K.J. GOMES, "ROPT Error Analysis for Wolsong-1," TTR-289 Part

3, Atomic Energy Canada Limited, 1995.

14. R.T. LAHEY, Jr. and F.J. MOODY, "The Thermal-Hydraulics of A Boiling Water Reactor,"

American Nuclear Society Monograph, 1993.

15. A. TAPUCU, A. TEYSSEDOU, P. TYE and N. TROCHE, "The Effect of Turbulent Mixing

Models on the Prediction of Subchannel Codes," Nuclear Engineeing & Design, Vol. 149,

1994.

16. R.K.F. KEEYS, J.C. RALPH and D.N. ROBERTS, "Post Burnout Heat Transfer in High

Pressure Steam-Water Mixtures in a Tube with Cosine Heat Flux Distribution," AERE-R6411,

AERE, UK, 1971.

17. M.B. CARVER, J.C. KITELEY, R.Q.N. ZHOU, S.V. JUNOP and D.S. ROWE, "Validation

of the ASSERT Subchannel Code: Prediction of Critical Heat Flux in Standard and Nonstandard

CANDU Bundle Geometries," Nuclear Technology, Vol. 112, 1995.

- 567 -

KAERI/RR-1999/99

- 568 -

KAERI/RR-1999/99

6 S-. DUPIC

- 569 -

KAERI/RR-1999/99

6. DUPIC

CANDU Qx}S.°\} *}&¥• 3 ^S«J <££.-§ *J3 *1l4-g-*fe DUPIC

(U.

-f-Sfl DUPIC

c}. D. # J L # ^ DUPIC ^ < a ^ l - 4 2 t ^r Stl

OREOX [Ref. 2]

OREOX 7}^o\]

^.iH, DUPIC ^ ^ S ^ 7 ] ^ ^ S f 7 } ^ ^ # ^SB DUPIC

DUPIC ^ ^ S ^ ^ ^ <a^"7} ^tgSJjL Stic}.3-4

OECD/NEA(1993) [Ref. 5]7}

W , DUPIC

uj-g-o] -^^*}7|| Qv\. ^ «14 DUPIC

DUPICDUPIC

H l § & 4 ^ } } ^ DUPIC

DUPIC ^ ^ ^ ^ 4

6.2 ^loflA-|^ DUPIC

o.i^, DUPIC

- 571 -

KAERI/RR-1999/99

6.1 DUPIC %& ]

DUPIC ^<£S.^7l =gaJM3£: DUPIC q& ^ ^ ]

^fl^^S.A-1 DUPIC

DUPIC

^ DUPIC ^ ^ ] ) § 4 p g f l f g i ^ ^ r ^ fl^i]M ^ l £ ^ 1 ^ ] ] # ^ H^ l DUPIC

DUPIC ^ l ^ ^ l ^1^1 7||^^7^1 ^ r ^ ^ r 1992^1 Idaho National Engineering Laboratory

AIROX ^.3-A^l 7 5-7l$>of ^*J*|-^l^r. AIROX

DUPIC

Oak Ridge National Laboratory (ORNL)^]

DUPIC ^ < ^ S . ^ H l - g - g : ^ 400 MTHE ^ f S ^ ^ ) ^ ^ ) ) fl^

7]

^ 4 JJL

aela DUPIC

fe ZL 6.1-

- 572 -

KAERI/RR-1999/99

6.1.1 DUPIC #

D U P I C M ) ^ ^ ) g 4 g } § ^ $}}] fl|, ^ f ]

l ! 4 #4 S H & ^ } j | ] AIROX

D U P I C 4

71^011 cH«U A I R O X

6.1.1.1

CANDU

DUPIC

>: DUPIC A|>£ -§-5o^ v£ Af^-^ ^ ^ S . ^ < ^ 5 . 400 MTHEo]t;>

2. ^ u | oj-g-^-: X\^i$] ^ - S # J L ^ * M , ^ «

3.4. -gn] ^ ^ ] : DUPIC

CANDU S f ^ S S . ^ ^ * f e S.& ^ a l # ^ J l ^l^f. CANDU

, DUPIC A l ^ S

. DUPIC A ] ^ ^ - DUPIC ^ ^ 5 . ^>^61] ^^.«> S . ^ -fi- 1 Al>^ . DUPIC A ] ^ ^ -

- 573 -

KAERI/RR-1999/99

2.

3. 7«^ iJ |<H (I&C): A]A^^

1. 4-g-^ ^ ^ S ^-^-r DUPIC X\^6]}

^ t > lOVi o | # u|2]-^ S f 17X17

2. DUPIC ^ < £ S ^^1-i-: ^ 1 ^ ^ ij-f- ^ ^ # ^ CANDU ^J^S-g- 43-g-

^ > ^ ° 1 ^ . DUPIC ^ ^ S . t:>«^ -gTd 4 6 o ^ ^3L^ 4 4 S 6.1-12}

3. ^ ^ S S ^ : 7 |§ ^^r«i^ # ^ ^ ^ r 1-60

7HJ*>t*. 1.60

S. 35 MWdTigUofl

4. n^:<i

- 574 -

KAERI/RR-1999/99

> > | > 100 MTHE)

6. DUPIC ^ ^ 5 . * i # : ^HV^g^ ^ 1 ^ ^ ^ : *l4i?> DUPIC

# ^-§-*}S^- ^ ^ 5 ] ^ o > t>t>. (50MTHE)

7. 3^)711- * | # : DUPIC A

7].

f. DUPIC

: DUPIC ) ^ f ^ ] ^

ANS 8 ^ | e ) Z : ^

> * U , ^ 4 : 27}*]$)

A]>b^£ ANS 8.3-21

2. 4 ^

3. ^ H | *}afl: ^ ^ 4 ^ o f l ^ 5 i aoVA}^ § ^ 7 } 0.5 mrem/hr

DUPIC

(NRTA) A ] ^

3.Q 6.1-3^]

DUPIC # ^ 4 3 # l ( )^ , D U P I C

- 575 -

KAERI/RR-1999/99

6.1.1.2 DUPIC

DUPIC ^<$.M. :gjg,o. A ] ^ S J tifl^l-t ^ ^ 3 1 ^ -£j&.# W c f . DUPIC

OREOX ^-^^r DUPIC ^ ^ 5 . 4 ^ ^ ^ -S-^^f 3-711-

7}. DUPIC

1.

0.07 MT/MTHES.

0.29 MT/MTHES.

3.71 sj

4.

. DUPIC

£n> -

- 576 -

KAERI/RR-1999/99

3.

>. DUPIC

2.

3.

CANDU

. DUPIC

^r S 6.1-2ofl

6.1.2 DUPIC

6.1.2.1

DUPIC

- ( ^ , 400 MTHE/yr

DUPIC 400 MTHE/yro]t:|-.

1 o> 30%5] A } ^ ^ 7 l ^ * W o j ^ ^ . o)

- 570 MTHE/yr (^- 400/0.7=571)^1

. # 600

- 577 -

KAERI/RR-1999/99

^ £3.7] <>]

-g-ô] 4 4 150 MTHE/yr l 47H5| ^ ^ ? > ^ g Bj- T

4 4 27fl}. 0} 27113] J B ^ ^ T ^-^SLS. 4 4

sat;]-. 4 ?mds] 2711 s>oi^

batch 3.71

>^. _ 3 5 0 kgHE %£.°M (#7> ^ ^ - S U235 2

3.7]«] «1*I 200 kgHE# batch 3.71 S

200kgHE# batch 3.7] S.

.^ &C\ ^ ^ « > batch

3.71-i- ^L3l F<H, batch 3 7 ] f # D| ^-^*f<^ 160kgHE (^, batch?> 3.7]

20% <H^-s s]-§-)Bl- 7H8^>i;K o] batch

3.7]

80 kgHE£] batch

batch 3.711- ^ - ^ ^ f e o ] ^ - ^ ^ ^ ^ ^ - %ôU]^7] ^Â-]x:].. (^, ^

batch

, 80 kgHE)^] ^ ^ ^ r « 7700

t^ (^, 10.4 g/cm3 ^^^11 ig£), ^]^^r

cm3) ^ £ 7 } f 7 m c } . # ^ JL#5loM J1^<>1 (S. 6.1-3

, 10.4 g/cm3 ^^^11 ig£), ^]^^r ^^ ^ ^ r TjÎ ô]] 32% (-10,200

460 kgHEolnf. O|B|^> batch

^ batch 7f ^ o ] £ <^^ 7fl (^, 460/80,

- 578 -

KAERI/RR-1999/99

6.1.2.2

DUPIC A ] ^ ^ 471)51

DUPIC

9i

4.7 g/cm3

7]).

, -40013 , :5

(~500°C,

600°C,

batch

- 579 -

KAERI/RR-1999/99

ZL

ofl

n>.

, DUPIC

43-g- f s . ^ , DUPIC

- 580 -

KAERI/RR-1999/99

6.1.2.3 X\Jg 7fl.fi.

DUPIC - ^7 , ] 7 |

DUPIC

fe ZL^] 6.1-6<H1 ^Bj-uf 61c}. o) A ] ^ ^ c f l ^ 0.4 km

30

5 mrem/yr

23

7H

ufl- 9.75 m,

85 m, ^ o ] 7 > 20 mojt}-. o|

829

- 581 -

KAERI/RR-1999/99

DUPIC

batchJL

DUPIC $ 4 ^ #^ ^ ^ §* ^ltl ^4^ l DUPIC

4 W S J ^tô]i4 ô>^ ^<H1^ 1.6m ^ S ^ ^-ô] ^tlSlojoTT-f. ^-fc. 1.6 m£j $°]^ 7}$. -f- •%•% Bf T 4

6.1.3

DUPIC ^ ^ 5 . 4 3 - «l-§^- 6 . 1 . 1 ^ ^ ^ ^ ^ t l AôH tcl-Bf DUPIC

6.1.3.1 Hj-g- 3g7|-

71 ^ ^

DUPIC l g

^ - ^ ^ INEL

Rockwell International^! 5l*l| ^ 7 ^ H f S l ^ M 4-§"tb «!-§•

- 582 -

KAERI/RR-1999/99

(Richardson's Construction Estimating Standards)^ n | o | ^ ^|^H]-§-^f^. (Mean's

Facilities Cost Data)7} ©l-g-SjSfcr- <yjf #a|ofl t f l s j j ^ ORNLÂ-] *flA|# H]-g-

: 2020-2059 (40id)

l^>: 2015-2019

7]-g- i d S : 1995

: 5%

^ S 6.1-4ofl

3007|| ^5L£|

530

: ^>fl (14%), ^ÎM^^^]- ^ ^ ^B] (10%), Î«l7f (20%), ^ #

(3%), 4^-^f ^-^ (2%), "y^H1- 91 ^ ^ ^ ^ (6%), 1 ^ ^ 3*1- (20%),

-& 413

25%Z}JL 7f^^rf^, 400 MTHE/yr ^-2. A } ^ * | ^ ^ | ^[û]-g-^- 1179

- 583 -

KAERI/RR-1999/99

1 *]• 90000 $/yr, <a*]i-H 50000 $/yr, 7]&*\ 50000 $/yr, ^ ^ ^ V 3700

$/yr, ZLBJJL Af?-^J 25000 $/yr#

10% ^ £ . 5 . 7}^«:t;>. ^ ^ r « O U^ ^ 7 H - ^ ^-e]5f n]%- ^7\M: 10

m3, -S-B]S|. * > * 1 ^ ^ 7]-i- 41 m3, <!& ^ < £ S ^-2L^ 65 m3, 7)B|- s j | 7 l# 764

>. 400 MTHE/yr<Hl cflsfl Jg7>S ^^|] <£# ^r^/-fr^»]-§-& 140

6.1.3.2

( 6 J - 2 )

N P B ^ ^ ^ 7 > 5 } o|^.(net present benefit)©)^, Q . ^

- 584 -

KAERI/RR-1999/99

to, 7 1 ^ Aj>g -§-50>ofl power factor ^ ^ . *?-§-*H

7}. 7 1 ^ J£i£

DUPIC *]-gSj *}el -§-3oNr 400 MTHE/yroJc}. o ^ CANDU Q*}S. 77\6\]

r l# S.^. LCC-fe S 6.1-8< 1

2015^^1 2059\l Tft!:6!! *1^5l ^i^» ^

! LCGb 7115

^ NPV# 5L#*KH #±3£\

^ 1188 M$o]r:}. LCC5| NPV7} ^ ^ 5 ] ^ LUCS. 7\]*h?} <$• Stic}

5.^5] LUCfe 558 $/kgHE^ ^1^5)^^}. 7]-g- DUPIC

6.1-9011

DUPIC ^711 2}J§oH *Xo)*\, DUPIC

l 150% ^^1<^1 cjjsfl ^7}t}^3. ZL %3}^ S. 6.1-103} ~L

6.1-

- 585 -

KAERI/RR-1999/99

500 MTHE5. ^ S W * 1 £ ^I^Hl-g^: 4 4 17% «£ 28%

}HS. DUPIC

DUPIC ^ ^ S ^^^gr DUPIC

4MM DUPIC

50 $/kgU5. 7>^*>n^, o\%-£: OECD/NEA

SEU (3.5 wt%) 7 } 3 £ 940

50 $/kgU, 8 $/kgU, ZLE] L 110 $/SWUB>JL

(SEU) ^-^r ^ ^ - f e f e <^# 7ff-i>*ic>. ZL t[£-, SEU

- SEUu>

6.1-11311- 6.1-12ofl ^ ^ - f e f e j } SEU<1 ^* i n]4s . ^ ^ ^2 f# 4 4. a 6.1-1 H I t}E}v* y}^- 4°L DUPIC «j«is 4 3

4 ^ : ^ 7 f # 7 ^ ^ } . ^91-fSfe# 50%

583 S/kgVS. A^r 4.5% ^ * } &Dfl-f

- 586 -

6.1.4

KAERI/RR-1999/99

10% M^ Hl-S-°1 5 5 8 ^ 652 $/kgUS 17%

DUPIC

DUPIC

7)fif

DUPIC

o] ^ ^ . # ^ . * H 400 MTHE/yr -g-

S.5. (558 $/kgHM) ^Tll

-. o]

K DUPIC

- 587 -

KAERI/RR-1999/99

Table 6.1-1

Characteristics of DUPIC Fuel Bundle

Physical geometry

Bundle diameter

Number of large pins

Number of small pins

Length

Heavy metal weight per bundle

Bundle weight

Number of pellets in a large pin

Number of pellets in a small pin

Pellet density

Pellet surface finish

102.5 mm max

8

35

49.53 mm

17.64 kg

23.6 kg

30

36

10.4 (0.15) g/cm3

0.8-1.6 micron RA

- 588 -

KAERI/RR-1999/99

Table 6.1-2

Nominal Level of Recycle Stream for DUPIC Process

Reference fuel material flow

Fuel material loss

- In fuel decladding process step

- In dust form (e.g., trapped in HEPA filters and

non repairable equipment)

Recycle streams

- Rejected pellets before sintering

- Rejected pellets after sintering

- Rejected pellets after grinding/finishing

- Net rejected fuel pin after welding

- Initial rejected fuel pin

- Repairable fuel pins

- Net rejected fuel bundle after welding

- Initial rejected fuel bundle

- Repairable fuel bundle

Total recycled fuel material to oxidation/reduction process

1 MTHE of processed

spent PWR fuel

1 wt%

Negligible

0.5 wt%

5.0 wt%

5.4 wt%

1.0 wt%

3.0 wt%

2.0 wt%

0.1 wt%

1.0 wt%

0.9 wt%

12.0 wt%

- 589 -

KAERI/RR-1999/99

Table 6.1-3

Material Flow in Main Process Building

Net DUPIC facility throughput

Design throughput at 70% plant availability

Number of parallel process lines

Design throughput per process line

Daily process rate (i.e., 150/365)

PWR spent fuel disassembly rate (0.41/0.44)

PWR fuel rod decladding rate (264)

Fuel oxidation/reduction process rate (w/o recycle stream)

Fuel oxidation/reduction process rate with recycle stream

Larger pellet production rate (w/o recycle stream)

Larger pellet production rate with recycle stream

Smaller pellet production rate (w/o recycle stream)

Smaller pellet production rate with recycle stream

Larger fuel pin production rate (w/o recycle stream)

Larger fuel pin production rate with recycle stream

Smaller fuel pin production rate (w/o recycle stream)

Smaller fuel pin production rate with recycle stream

CANDU DUPIC bundle production rate

400 MTHE/year

600 MTHE/year

4

150 MTHE/year

0.41 MTHE/day/line

1 fuel assembly/day/line

264 fuel rods/day/line

410 kgHE/day/line

460 kgHE/day/line

4,830 pellets/day/line

5,410 pellets/day/line (+12%)

28,980 pellets/day/line

32,460 pellets/day/line (+12%)

161 pins/day/line

163 pins/day/line (+1.1%)

805 pins/day/line

814 pins/day/line (+1.1%)

23 bundles/day/line

- 590 -

KAERI/RR-1999/99

Table 6.1-4

Estimated DUPIC Direct Capital Cost

Element

Site Preparation

Process System

Main Process BuildingHealth Physics Facility

Safeguards and SecurityUtilitiesFire Department

Simulation and TrainingAdministration Facilities

Specialty Gases BuildingWarehouseOff-Site FacilitiesTotal Direct Cost

Estimate ($)

11,406,000

297,682,000

185,172,000

6,640,0007,343,000

10,971,0001,550,000

360,0001,360,0001,534,000

525,0005,476,000

530,019,000

- 591 -

KAERI/RR-1999/99

Table 6.1-5

Estimated Annual DUPIC Labor Cost

Department/Division

General manager staff (total)Staff

Administration (total)Manager/StaffLegalHuman resourcesProcurementComptrollerComputer/Information sciencePublic relations

Safety and health (total)Manager/StaffIndustrial safetyRadiation safetyMedicalEmergency preparednessAnalytical laboratoryData processing/Records

Safeguards and security (total)Manager/StaffNuclear material safeguards & accountabilitySecurity

Environmental & waste management (total)Manager/StaffEnvironmental/Waste management & complianceWaste Operations

Engineering and quality assurance (total)Manager/StaffNuclear safety engineeringProcess mechanical engineeringElectrical/Instrumentation engineeringConstruction engineeringQuality engineering

Compliance/Standards (total)Manager/Staff

Operations (total)Manager/StaffProcess/Utility operationsOperational maintenance

DUPIC Staff Total

Staff

556525141310156

934

20176

20206623194054216368641718197

211515113126734

493

Labor($)

255 •

115199547550399765315

1891015813275937915215

152952375

1158021675

165865915965365859

611

49833001615

Total($)255

2890

4359

1479

2592

4134

611

5413

21733

- 592 -

KAERI/RR-1999/99

Table 6.1-6

Estimated Annual DUPIC Non-Labor Cost

Category

Materials

Equipment

replacementUtilities

RadwasteDisposalNon-Labor

Total

Gross($)

20,235

21,200

9,150

67,800

118,385

Specific Item

Dysprosium pins

DUPIC fuel assembly componentsProcess gasesAnalytical suppliesMaterials for waste treatment

Kr-85 cylindersVitrified waste canistersGreater-than-Class-C containersLow level waste containersLiquid argon/NitrogenFilters (HEPA, charcoal, liquid)

Health physics contamination suppliesPersonnel protective equipment

Electricity

Fuel oilTransportation fuel and lubricants

Spare partsChemicalsMiscellaneousJanitorial suppliesMiscellaneous (e.g., Office supplies)

Cost($)

800

18,0007550

200150150400607520075

21,200

6,000

25075

2,500755012575

67,800

Cost Basis

1 pin per bundle/21,500 bundles per year

21,500 CANDU bundles.per yearScale up from AIROX 200MTHEScale up from AIROX 200MTHE

264 Cylinders per year41.2 MT glass/24 cubic meters per year60 cubic meters per year756 cubic meters per yearScale up from AIROX 200MTHEScale up from AIROX 200MTHEScale up from AIROX 200MTHEScale up from AIROX 200MTHE

1/10 of total cost for process equipment

46 million kWh per year

Scale up from AIROX 200MTHEScale up from AIROX 200MTHE1/10 of total cost for utility equipmentScale up from AIROX 200MTHEScale up from AIROX 200MTHEScale up from AIROX 200MTHEScale up from AIROX 200MTHE

Scale up from AIROX disposal cost

- 593 -

KAERI/RR-1999/99

Table 6.1-7

Inputs for Life Cycle and Unit Cost Estimation

Content

Capital Cost

Operation & Maintenance

Cost (annual basis)

Decommissioning Cost

(annual basis)

Sub-content

Direct cost

Site preparation

Process systems

Main processing building

Site support facilities

Indirect cost

Contingency

TotalStaff

Utilities

MaterialsEquipment replacementRadwaste disposal

TotalDecommissioning cost

Cost(k$)

530,019

11,406

297,682

185,172

35,759

413,416

235,859

1,179,29421,733

9,150

20,23521,20067,800

140,1188,282

- 594 -

KAERI/RR-1999/99

Table 6.1-8

Life Cycle Cost and Unit Cost Estimation for DUPIC Fuel Fabrication

(Discount 5%, Capacity 400 MT, Contingency 25%)

1

Year

201520162017201820192020202120222023202420252026202720282029203020312032203320342035203620372038203920402041204220432044204520462047204820492050205120522053205420552056205720582059Total

Cost (k$)

Capital

117929235859235859353788235859

1179294

Net Present Values (Levelized Unit Cost

Operation&

Maintenance

1401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401181401185604720

Decontamination &

decommission.

8282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282828282

331280

;k$) = 1187607($/kg) = 558

Total

1179292358592358593537882358591484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484001484007115294

NPV

4444684660806281151837313243823417363974937856360533433632701311442966128249269032562224402232402213421080200761912018209173421651615730149811426813588129411232511738111791064710140965791978759834279457566720668636536

1187607

Production(MTHE)

40040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040040016000

DiscountedProduction(MTHE)

118113107102979388848076736966636057545249474542403837353332302927262524232120191918

2128

- 595 -

KAERI/RR-1999/99

Table 6.1-9

Estimated Costs for DUPIC Fuel Fabrication Plant of 400 MTHE/yr Capacity

(Reference Case)

Item

Capital Life Cycle Cost Direct Costs

Indirect Costs

Contingency

Operation and Maintenance Staff

Costs (annual basis) Utilities

Materials

Equipment Replacement

Process Waste Disposal

Decontamination and Decommissioning Life Cycle

Cost40-years Life Cycle Cost (M$) in Net Present Value

Levelized Unit Cost ($/kgHE)

NPV (M$)

179

140

80

117

48

106

112

362

44

1,188

Fraction of

Levelized Unit Cost

33.6 %

62.7 %

3.7 %

558

- 596 -

KAERI/RR-1999/99

Table 6.1-10

Sensitivity Analysis on Cost Parameters

Items

Discount Rate (%)

Contingency (%)

Production (MTHE)

Sensitivity Variable

2.503.755.006.257.5012.5018.7525.0031.2537.50200300400500600

Unit Cost ($/kg HE)

494524558596637537548558568579686606558535501

- 597 -

KAERI/RR-1999/99

Table 6.1-11

Sensitivity Analysis for Adding Natural Uranium

Natural uranium

fraction (wt%)5101520253035404550

Annual natural

uranium feed (MTU)20406080100120140160180200

Annual natural

uranium cost (k$)10002000300040005000600070008000900010000

Fabrication cost

($/kgHM)561563566568571573576578581583

- 598 -

KAERI/RR-1999/99

Table 6.1-12

Sensitivity Analysis for Adding Slightly Enriched Uranium

SEU fraction

(wt%)12345678910

Annual SEU feed(MTU)481216202428323640

Annual SEU cost

(k$)376075201128015040188002256026320300803384037600

Fabrication cost

($/kgHM)567577586596605614624633643652

- 599 -

Anchor Facility

Material/Labor Cost

Levelized UnitCost Model

Facility Performance &Process Requirements

Facility Design

Process FlowEquipmentHot Cell LayoutBuildingAuxiliary SystemsOthers

Evaluation of Capital Costs

and Operating Costs

Life Cycle Cost& Unit Cost Analysis

KAERI/RR-1999/99

Reference Process Model

Other Reference

Financial Factors- discount rate- contingency

Construction Scenario

Refinement Calculation of DUPIC FuelFabrication Cost and Its Sensitivity Analysis

Fig. 6.1-1 Schematic Process for DUPIC Facility Cost Evaluation

- 600 -

KAERI/RR-1999/99

1 Zircaloy Bearing Pads

2 Zircaloy Fuel Sheath

3 Zircaloy End Support Plate

4 Fuel Pellets

Fig. 6.1-2 Configuration of DUPIC Fuel Bundle

- 601 -

KAERI/RR-1999/99

INPUT- NDA (Spent PWR assembly)- Shipper/Receiver difference

Item Handling

Assembly/Rod - No material-uncounted-for (MUF)

MBA-1

FabricationProcess

NDA

Bulk Handling- MUF- Waste (hulls, scrap, etc.): NDA- Physical inventory test at each key

measurement point by weight- Operator/IAEA sharing accounting

data (NDA, DA, weight)

MBA-2 -

DUPIC Bundles

Item Handling

OUTPUT (NDA)MBA-3 -1

Fig. 6.1-3 Accounting Methodology in DUPIC

- 602 -

KAERI/RR-1999/99

Spent PWR Fuel

Decladding

Oxidation/Reduction

PWR Rods

Cut to size

Structural waste

;iVolatile waste

Volatile/Semi-volatilewaste

Pelletization

SinteringV ^ ^ ^ A

/ k

Parts

! «£

DUPIC rod DUPIC Bundle

Fig. 6.1-4 Pictorial Illustration of DUPIC Process

- 603 -

KAERI/RR-1999/99

PWR Spent fuel

(1)

StructuralHardware

Disassembly

Solid waste

Fuel fragmentsand off-gas

(7)

Off-gas waste

(8)

Oxidation/ReductionProcess

Volatile/Semi-volatile wastes

10.9wt%offuel material

Powder

Pellet Forming/Sintering

(9)

Volatile/Semi-volatile wastes

(10)

INew non-fuel

component

(4) Pellets

(6)Fuel Pin/Bundle

Fabrication

l.lwt%offuel material

New DUPIC fuelBundles

(5)

Fig. 6.1-5 DUPIC Process Mass Balance Schematic

- 604 -

KAERI/RR-1999/99

Isolation Zone Secondary alarm statioi

Central alarm station — P j

/ IWarehouse Fire

House

Hypothetical SiteBound

AdministrationBuilding

EOC/VisitorsCenter •

— — — • " • " ^ ™**» • ^ ^ ^ • • ^ • • ^ • • • • • • • • • • • • • • • _ ^ ^ ^ ^ ^ ^ ^ w w « v ^ « V i ^ • ^ • • •» ^ . ^ ^ • • _J

Fig. 6.1-6 DUPIC Facility Area Plot

- 605 -

KAERI/RR-1999/99

Cold shop

Mock-upwork area

Rail carair lock

Waste productstorage

PWRspent fuelstoragevault

Elevator/stairs

Material.—.transterlhatch

#1 Rail cartruck bay

Rail carair lock

#2 Rail cartruck bay

Office New DUPICfuel storage

Cold truck bayNou-fuel component

storageElevator/

stairs

Process monitoring and operating access gallery (around canyon)

# 1 Process line

ft 2 Process line# 1 Process cell (canyon)

Process monitoring and operating access gallery

# 3 Process line

# 4 Process line2 Process celt (canyon)

New DUPICfuel transferhatch | ~ j

#1 Canyoncontrol room

Computerandsupportequipment

Securityoffice

Process materialanalytical lab.

Elevatorstairs

Healthphysics

laboratoryand office

DUPICbundle

inspection

Stairs

T9m

64m

Fig. 6.1-7 Main Process Building Floor Plot

- 606 -

KAERI/RR-1999/99

800

750

700 i

9 650

|? 600

r 550CO

O 500

| 450

400

350

300

discount rate i

contingency !

. production rate !

* • • - .

05 075 1 J O O 125

Sensitivity Factors (value/reference value)

150

Fig. 6.1-8 Sensitivity of Cost Parameters

- 607 -

KAERI/RR-1999/99

6.2 DUPIC ^

DUPIC (Direct Use of PWR fuel In CANDU Reactors)

^ DUPIC

3. 41^-^ ^ € 7l7]£) ^ ^ ^ ? l t H ° l ^ r - r £ 1 ^ DUPIC

DUPIC

J ° S DUPIC ^ ^ 5 . ^ 5]% 3L7l^ 7 | $ - § ^ 5 . ^ ^ « ^ ^ ^ f ^n>. HB^tf- DUPIC

-b o |o | 71] ^ t t l » | ^ o | o]s.<>|^o> sVcf. ^ DUPIC

DUPIC 9 r }\

A]

, 2, 3

DUPIC

DUPIC

DUPIC ^<&.3.*) ^ ^ r ^ - ^ ^ S . 6l^> 4 ^ - 2 ^ 1 *M-§-3oM- ^^^f^ ^ T f l ^ ^ A]

- 608 -

KAERI/RR-1999/99

6.2.1

(discharge

(reception

n i g 6.2-24)-

A]

^-§1)

6.2-3^ - L ^ 6.2-4±r

6.2-5

. ZL^ 6.2-6^

6.2-7^

IAEA -g-

-§-

6.2.2 DUPIC

DUPIC

- 609 -

KAERtfRR-1999/99

1'12 -r-

. DUPIC

DUPIC

DUPIC t\

A.

B.

DUPIC

DUPIC

DUPIC

6.2.2.1

DUPIC

7)

-LB]J1 DUPIC

7}. a«g 6.2-1

DUPIC

DUPIC

}. o]

DUPIC

- 610 -

KAERI/RR-1999/99

DUPIC

o| ^ A } ^ ^ ^ 5 . a .^7] S ]H^ 3.717}

A-2) (ZL^ 6.2-1

o] «OVA1^ ^^--g-71 vfl^ A j g S ^-^^-ôil ^ ^ ^ ^ # ^ * 1 * H , DUPIC

s ^ 1 ^ - A] iH8*Rf ^.<g# j q ^ s # o ] ^ ôic}.. o) ^*jofl^ ^>î ^ ^

DUPIC «¥&5.fe 4^-8-71 ^5 ] 5J- iS # 7 ^ ^ 7B-g^t:>. ZLBlJl DUPIC

. o)

DUPIC

- 611 -

KAERI/RR-1999/99

6.2.2.2 (ZL^J 6.2-1

DUPIC

DUPIC

DUPIC

DUPIC

DUPIC

DUPIC

Qv.}. n

(ZL^ 6.2-1

DUPIC

*> DUPIC

- 612 -

KAERI/RR-1999/99

B-2) ( a H 6.2-1

DUPIC ^ # ^ t i l 1 o^, DUPIC

^ DUPIC

n e l i DUPIC

717151

ZLelJL ^1 DUPIC ^ ^ ^ . - ^ ^ I ^ S . ^ ^ - i - ^*B ^ : ^ ^ # 3 ^-§-71

^ ^ 4 J r K 1^^ f 6.2-2^44^1 ^ s ^ i cflt> Hia

6.2.3

DUPIC } d l ^ g f e g f f # ^

$ ^ ^ - f efe ^ ^ s . # a ^ ^ # ^ ^ -t! DUPIC

. DUPIC n £ ) ^ ^ ^ > i ^ ^

DUPIC «J«i§.5l ^g-f ^ 5 . ^ 6.2.2

- 613 -

KAERI/RR-1999/99

DUPIC i #

DUPIC

3.^7)7} 87i^ ^ ^ U H M - ^-g-^ ^ sa>nj ^ l ] - . DUPIC

1-77H

AI D U P I C

., 4 «I^S*H^6H -f- ^ ^r^-^- ^ ^ 5 . 3.^:71 S. Al~g c DUPIC77fl ^ ^ ^ ^ f e f J2.^*>3., H ^ 47B *fl^^

X\

# ^ ^ 4 . ZL5|uf DUPICins. #^1-7)) 5]<H ^^) i - l l ^^^^ 1/23. # ^ ^ c K

n}7fl(closure plug)^ iij^is)(snout plug)^ vfl -«g«> <%X\ 1/23.

^r Slfe 3 ^ ^1 DUPIC «^oiS.5f 4

DUPIC ^91^-7} $^£)^] * l ^ yo^°14. nfsH A<Bf4

*\ DUPIC ^ ^ S ^ I ^ f e 2, 4, 6, 8 tHHi} . ^.5]^- °H1

- 614 -

KAERI/RR-1999/99

JE. 6.

2, 4, 6, 8

S.oflA-1 A] DUPIC

7H

DUPIC

6.2-3

DUPIC

6.2.3.1 DUPIC

7H ^

^r

Tc- PdcdQ0UtT+ h~ - Tamb) = (C dMd+ (6.2-1)

Q

Md

Mm

Qtn

Q.01U

dT

dz

T

Tc

1 amb

Pd

Cd

il71 £

D2O

D2O

D2O

5o* (kg)

%* (kg)

(m3/sec)

(irtVsec,

(°C/sec)

(=

(kg/m3)

(J/kg°C=W.sec/kg°C)

- 615 -

KAERI/RR-1999/99

cm : i$7\*l til«i (J/kg°C)

Am : n W ^ l S]Jf #2\ (m2)

(W/m2oC)

3) yJS; % ^

C=

Q(r)o)r.\.

A + q - B • T= C~^ (6.2-2)

c^7\M A, B ^ C ^

^+gU+ggr0)e"cr

r = ^ (6.2-3)

- 616 -

KAERI/RR-1999/99

(6.2-4)6\]*\ T><r,-=ro)=To ^ <?',< r ,= r0) = ?

, H ^ 6.2-82} 6.2-9^

- a 6.2-4^

DUPIC

DUPIC ^ ^ ^ §

ZL^] 6.2-10, 6.2-11, ZLZ]3. 6 .2-12^

. H o]

=L% 6.2-10^:

fe S 6.2-451

54.2°C

(130°F) ^ W

51.6TC (125'F)

6.2-4

57.8°C (136°F)

57.2°C (135°F) K ^ . ^ 6.2-12^

3. 70 °C (158T)

- 617 -

KAERI/RR-1999/99

0°C (140°F)

5.5. ^<^.3.S!) £2^7}

baU screw ^ - g - ^ 1 ] ^ ^ 7 1 ] ^ ^ ^ £ 7 } 150°C (30

0°F), ^ 7 ^ ^ ^ - 9J BflolS H e f o l H . ^ ^ T f l ^ ^ ^;S.7> 65°C (150'F), ZLZ]3L 7}

21°C (70°F)<>li:K

65 °C

DUPIC } £ b H^ ^ flI ^ » 11

« ? i ^ * } ^ ^ ^ l l ^ ^ * ^ ^ - f a f l S , ^-U 6.2-

65 °C (150-F)

7 1 ^ O)AOVO1

6.2.3.2

- 618 -

KAERI/RR-1999/99

(6.2-3)^

2)

3)

4)

80% # ^ £ #

38°C (100T)

80% # ^ ^ ^ i

49 °C (120T)

38°C (100T)

^ ^4 -g-

49°C (120°F)

38°C (100T)

2) ^ J ^ > ^ # 80%

80%#

49°C (120T)

3) A]

4 MW5] <g

^r 0.2 MWo|c>.

- 619 -

KAERI/RR-1999/99

^f ls 4DUPIC

7]7} <

27fl

DUPIC

2711*1

DUPIC

43,0087flJif

43008/8=5376

98X24X16=37,6325.

^ safe

71

. 6\

112X24X16=

DUPIC

DUPIC

6.2.3.3

- 620 -

KAERI/RR-1999/99

(6.2-4)# tf ^§^ ^ $& £ %} £ # 3 |§

-§•

nrCYT — nrQT-\- n, ,— k ,,A «— ^ ^ - — h * A / T— T C\^o\v-* c f-^x^-i i *dfuel /cwalir^-wall ^i^ flwater**-$nrf\ -* -1 amb/

** (6.2-5)

^ B | * M ^T (6.2-4)iif -

" == PcQTc-\ ~J~ Tenv^ hwater™-surfTami,

B= PCQ+ "^^ Wa Tenv-

C= CwaterMwater+ C/ue!M/uei

- 621 -

KAERI/RR-1999/99

Q

T

Tc

T

* env

1 amb

dx

P

Cwaler

Cfuel

hsurf

%* (m3/sec)

(= TCOTS,°C)

(°C)

20 m

(2TCS.

(20 m)

(J/kg°C=W.sec/kg°C)

(J/kg°C)

(0.061 W/m°C)

= 1.32(zJT/L)025)

20 m

ZLBli L ^ 12 m>§- A

^ S 6.2-6ofl

1.32 W/m2 °C

DUPIC

-8

A

- 622 -

KAERI/RR-1999/99

80% ^ 100%S.

^ 80%

100%

31.6°C

6.2-7011

JL3|*}3. <i*f# ^71512 MW

2.5MWS,

6.2-8

6.2-8^: tc]-H.

- 623 -

KAERI/RR-1999/99

46.2-16^

ZL^ 6.2-174 6.2-18^ 12

ZL^| 6.2-194 6.2-20^

. -L*U 6.2-214 6.2-22^

6.2-9

. 80% 71

o] ^ ^ ^ .

^-§-71^0]

36.

. ttj-BM o)

MW, Hl^^-BH 4

sell on

- ^ ^ > J I (80% 3.4 MWS

^ 4 4 ^^r^ 2.5 MWS

3.25

4.2 MW5.

^ 6-7 MW<H] o l ^

61

4

5 MWS

- 624 -

KAERI/RR-1999/99

S 6.2-1

JOS. ^oHNNH, -g-^«i# 7l^j$.S xH'-S ^ 4 - ^ ^ ^ : 2 MW^Al 2.5

4 Mw^]^ 5

6.2.4 DUPIC «|*[jg.Sl ^ ^ ^^>S. ^-g- H]-§-

DUPIC

AH 4-§-f « ? ^ ^ 3E^.# -f-*H «?-^3. ^1 DUPIC

^ DUPIC f ^ j

ZL^J 6.2-23

DUPIC ^ & ^ ^ § 1 ^ % ^ $

J £2,7]

6.2.4.1 ^ i ^ ^ H 1 - ^ « 1 ^ DUPIC

DUPIC

- 625 -

KAERI/RR-1999/99

DUPIC

DUPIC

6.2.4.2

7)

D U P I C

6.2.4.3

DUPIC 71

6.2-8 CHIA-1 6.2-12ofl 50°C

. 6.2.3

6.2.4.4 DUPIC

DUPIC

i fDUPIC

. o] ^B}^

- 626 -

KAERI/RR-1999/99

6.2.4.5

M f *1 DUPIC

-§•

o)

6.2.4.6

W . Jt 6.2-ll^r ^ ^

2, 3 ^ 45171^ ^ ^ ^ 3.^71 ( ^ ^ 13L71 - ^^ ) ^ ^ ^ 7 f l - f -#^ 30^ 7 ] ^ ^

2]S.X]^(fatigue usage factor)o]c>. SL6\) uj-E]-^ ^ ^ 6J-^ ^7j]^-#eH] cfl

magazine rotor, B-ram, latch-ram,

C-ram, ^-^^ 3tf, #<$ 7]*\ £ 4 # ^ « - ^

^ l DUPIC

, DUPIC

-S-i-

- 627 -

6.2.5.1

KAERI/RR-1999/99

6.2.5

6 . 2 . 4 ^ ^ s f i - H } % ^ S »i&^ i - H I ^ DUPIC

CANDU-6

S . * > ^ 3BE 6.2-125}

7}7] ^

2000^1

3,750,000 U$7>

6.2.5.2

(NPV)

(6-2"7)

(6-2"8)

- 628 -

KAERI/RR-1999/99

(6-2-9)

5%S.

4 DUPIC ^

1 4 I S ^ tl| 5]= 15,000

DUPIC «|elS.6fl tfl^^n}. 4|^.al-§-^- >t>^*fSir;>. S 6.2-132]- 6.2-14

DUPIC «J« iS^ 41^-Hl-g-^ t B S ^ ^ . 5 . 5.12 $/kgHMo]r]-.

6.2.6

DUPIC

DUPIC

DUPIC ^ < £ 5 . # QxtS-O)} ^ - ^ ^ ^ $XT\) ^ C > . DUPIC

DUPIC

- 629 -

KAERI/RR-1999/99

DUPIC ^ ^ 5 . JS-tajofl tflsfl*] 5.0-5.3 $/kgHMS.

DUPIC

- 630 -

KAERI/RR-1999/99

Table 6.2-1

Comparison of DUPIC Fuel Loading Path (Front Loading)

MI -§-

3.2)tt*7|

2 4 ^ 1

2. ^<g.3. ^

4*S# *

1 Mo^ g 7?

2 'S|64 JJ..2.

3 4 J-8-71

2. DUPIC ^

3 . DUPIC *!§

•8-<>l*l-s.

^ 2. A-1

[7] 21 S # ol-g-

g;7l 21SM1 4 ^ * 1

l-f1 U54 Si ^^f #*11-11 §"71 ^ S f ^ l

^7l S l^ .^^ T 1 - ^ ^

4 ^ - § - 7 l i-H<H] l

*7 l ^*W«1 Hi*i<Hl DUPIC ^ ^ 5 . # ^-g-^

wtt r"rf r 3 ^ l ^ L " ? f l ' ^ ^ TT "ft")-"?!

5. o l * (^5

4 tf*7l i1.4 -§-71

/ig3^1 4<

2. 4^--g-7li

4. ^^"-§-7li

1-4^i

3.^^s^a.'

1.4 -712.4^-8-71

1 -f 4 :?

S. A)

i-fl<Hl DUPIC t^<^

a# *f7i*m ^-11 *7 l ^gS]-^-*

-11 tl<&£. ^ ^

4^-8-71 UHIA-

Lfl 4 ^ A\ 4

ufl^l M ^

vfl-f- * 7 l ^ 5 f

Mis *H*

- o l *.3. 7fl*y-

-a*i

^ «

^8-71

a] Hl-g-

- 631 -

KAERI/RR-1999/99

Table 6.2-2

Comparison of DUPIC Fuel Loading Path (Reverse Loading)

B)

S.B-1 B-2

Ml -8- DUPIC DUPIC

5.DUPIC DUPIC

DUPIC

2. uovo v # ^ S A f DUPIC

^ 7 1

4-S-)

6.2-1

- 632 -

KAERI/RR-1999/99

Table 6.2-3

Time History for Defiieling of 8 Bundles per Channel

1

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

4 <3 Ml *

NfFP to Reactor face

*Mn}7fl Efl7]*H *H\ S S T } 1 * ^ e ] 7 l ^ g ]3 i*H 27fl «?^SC>^ fflTl^^S. o)^.

^ «14 27B ajgscl-i}- 4 7 1 ^ ^ 5 . oi^.

-¥• ^ H 27fl ^ ^ s c m 4T1^<H) ^ H ^ S4 tt*B 27H «?«is4^: 47i?i-5.s ol-i-4 ^^B 27fl ^<gSt;}«J: nH>| ofl ^$#5.Ml ^ ^ 27B <^S.t:l-tiJ: DU71^1^.5. o ] ^

Ml ^^H 2711 ^ < ^ S t : m n | 7 l ^ ^ ^ ^ ^ S

^} lo>7Jl RJ W 4 7 f ) # ^JL *B^^-H ^rS]^<aS-2.%7l Af-g-^^S 5 ^ ol^-gr}^ .4J-

^7115] «^0iSt:>^ I S f # * H yoV#3:^- O]^•¥•711 «|^sn>^- s i t #*>^ iM-ss o]^.

0

207

445

479

624

649

751

820

916

943

1883

2532

2744

2987

3197

3395

120

139

576

596

207

238

34

145

25

102

69

96

27

940

649

212

243

210

198

- 633 -

KAERI/RR-1999/99

Table 6.2-4

Time History for Defueling of 4 Bundles in 2 Channels (2 Bundles per Channel)

I

2

3

4

5

6

7

8

9

10

11

12

13

14

15

NFP to Reactor face

^ 4 ^ 27H « ) ^ s . t ; ^ nil7]£.£3. ©l-i-

^«i4 *m 27H ^ ^ s t m ^Tj oi] ^ H ^ S -

-¥• ^*H *M 2711 «!|«iSr:>^ n | 7 i £ ^ . S . o)-§.

¥ ^ f l ^fl1^ 27B *J&g.t*1* nflTl cH} ^ J g - ^ S

A>3 n>7H ^J #^>nl-7ll- ^JL AB^ofl-H ^ e l

0

207

445

479

1419

1558

2134

2730

2937

3082

3107

4047

4696

4908

5151

43 #

120

139

576

596

207

238

34

940

139

576

596

207

145

25

940

649

212

243

- 634 -

KAER1/RR-1999/99

Table 6.2-5

Parameters for Calculation of D2O Temperature in Fueling Machine

*1171*! ^ f " r ^ (Md)

fflTi^ ifl^- *>Sfl<*] ^ (Mm)

* M # -fr^/-B-# -fV£ (Q=Qin=Qou.)

^ «]<! (crf)

"flTi*! « 1 ^ (CjB)

^ ^ wlf1 (^)

^-°a^^r ^ £ (Tc)

A>^-^<a^. #<a ?a pfl7> l ^ £ (To)

ft3210

6159

0.00068 (=0.68ysec)

4232 at 40°C heavy water

4188 at 40 °C light water

460

1100 at 40 °C

307-317 (=35-45 °C)

307-317 (=35-45°C)

# ^

kg

kg

m3/sec

J/kg°C

J/kg°C

Kg/m3

°K

°K

- 635 -

KAERI/RR-1999/99

Table 6.2-6

Parameters for Calculation of D2O Temperature in Storage Bay

2047 m2,040,859 kg

682 m679,954 kg

(Q) 0.152 (=152 Vsec) m /sec

(Q) 0.0758 (=75.8 Vsec) m /sec

4188 at 30 °C light water J/kg°C

997 at 30 °C kg/m3

0.061 W/m°C

1.32 W/m2°C

719.27 m235.8 m523.15 m105.14 m460 J/kg°C

(Tc) 299 (=27 °C) °K

(To) 322 (=50°C) °K

- 636 -

KAERI/RR-1999/99

Table 6.2-7

Storage Bay Temperature and Heat Load due to Spent Fuel Decay Heat

Al#°l^(yr)1

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

22

23

24

25

26

27

28

29

30

31

80% 3 # £*I*1 -s .JE (C)

28.8

29.2

29.5

29.6

29.7

29.8

29.9

30.0

30.1

30.2

30.3

30.4

30.5

30.5

30.6

30.7

30.8

30.8

30.9

31.0

31.0

31.1

31.2

31.2

31.3

31.3

31.4

31.5

31.5

31.6

31.6

^HM*! (kW)1163

1456

1605

1697

1773

1837

1898

1956

2013

2066

2120

2170

2221

2269

2317

2363

2410

2454

2499

2541

2584

2625

2667

2706

2746

2784

2823

2860

2897

2933

2969

100% # £#

^#2: £ £ (°C)29.2

29.8

30.1

30.3

30.4

30.6

30.7

30.8

30.9

31.0

31.1

31.2

31.3

31.4

31.5

31.6

31.7

31.8

31.9

32.0

32.0

32.1

32.2

32.3

32.4

32.4

32.5

32.6

32.7

32.7

32.8

# :5-^ <! (kW)

1447

1814

2000

2116

2209

2289

2367

2439

2510

2577

2643

2706

2770

2830

2890

2948

3006

3062

3117

3171

3224

3275

3327

3377

3426

3474

3522

3568

3615

3660

3705

- 637 -

KAERI/RR-1999/99

Table 6.2-8

Heat Load in Storage Bay due to Instantaneous Discharge of Full and Half Core

flflt- $•^ 4*12: (day)

0

5

10

15

20

25

30

35

40

45

60

75

90

105

^ ^ (Full Core)

130

6.72

5.03

4.22

3.72

3.36

3.08

2.85

2.66

2.51

2.2

2.0

1.86

1.8

1- <l*l-ir (MW)

^ i c 9 (Half Core)

64.8

3.36

2.52

2.11

1.86

1.68

1.54

1.43

1.33

1.25

1.1

1.0

0.93

0.9

- 638 -

KAERI/RR-1999/99

Table 6.2-9

Storage Bay Temperature and Decay Heat of Spent Fuel due to Core Discharge

3 U d *

^ V^]^1^Ji|^

313.-&S-

31*

37

6630

4200

38

7150

3250

80% ^

5°C

(kW)

(kW)

3°C

(kW)

(kW)

36.2 °C

5830

3400

37.

6350

2500

(kW)

(kW)

o°c(kW)

(kW)

100% ^

31W

38.6°C

7350 (kW)

4800 (kW)

39.4 °C

7900 (kW)

3850 (kW)

37.0 °C

6350

3800

37.

6900

2850

(kW)

(kW)

8°C

(kW)

(kW)

- 639 -

KAERI/RR-1999/99

Table 6.2-10

Time to Reach 49 °C due to Malfunction of Storage Bay Cooling System

Time (hour)

After 31 year

80% power

7

100% power

5.3

After 12 year

80% power

10.3

100% power

7.9

- 640 -

KAERI/RR-1999/99

Table 6.2-11

Fatigue Usage Factor for Fueling

Component

Snout Assembly

Snout EmergencyLock Assembly

Magazine

SeparatorAssembly

Gland Plate

CoolantConnector

Ram HousingAssembly

Ball Screw SealAssembly

Gearbox, MainShaft and TapeDrive

Center SupportCenter Support Seal Holder RingLock RingScrew GearClamping BarrelWedge SegmentCenter Support BoltEmergency Lock-CoverCapscrews Holding the EmergencyLock-CoverLock Assembly Mounting CapscrewsMain HousingEnd Cover FlangeRam EndBracketsTechlok ClampGearbox HousingEnd CoverClamp StudCylinder BlockCylinder HeadCap ScrewPistonsGland PlateGland Plate BoltsFlange & HubBoltsMagazine Housing ExtensionRam HousingRear Forging10" Techlok Clamp10" Techlok Clamp StudHousingAssembly BoltsRetainerSeal SleeveRetaining NutGearbox and Tape DriveMain ShaftBolts

Snout Assembly Cavity Outlet Fitting

Machine

Fatigue UsageFactor

0.9910.00.1

0.090.60.30.60.06

0.4

0.40.90.10.30.150.40.000.860.840.060.060.210.060.021

0.40.370.870.5

0.70.050.30.2

0.5

0.70.72.30.040.140.060.7

0.34

ReplacementInterval

N/A3 yearN/AN/AN/AN/AN/AN/A

N/A

N/AN/AN/AN/AN/AN/AN/AN/AN/A

N/A

N/AN/AN/AN/AN/AN/AN/A

N/A

N/AN/AN/A

N/AN/A

N/AN/A

13 yearsN/AN/AN/A

N/AN/A

- 641 -

KAERI/RR-1999/99

Table 6.2-12

Capital Cost for DUPIC Fuel Handling

Content

CapitalCost

per Plant

Sub-content

Spent Fuel Storage Bay

Loading Equipment

Spent Fuel Port Pusher

Spent Fuel Port Blow

Dryer

Gamma Ray Detector

Fuel Loading System

Control Program

Modification

Design Documentation

Cooling Capacity Increase

of Spent Fuel Pool

Storage (Increase of Heat

Exchanger Capacity)

Total

Hard Ware

(l,000won)

1,000,000

500,000

100,000

100,000

NA

NA

1,000,000

2,700,000

Man-hour

2,000

3,000

1,000

1,000

4,000

6,000

1,000

18,000

Total Cost

(l,000Won)

1,200,000

800,000

200,000

200,000

400,000

600,000

1,100,000

4,500,000

Total Cost

(k$)

1,000.0

666.7

166.7

166.7

333.3

500.0

916.7

3,750.0

* Assumption

-100,000 Won per man-hour

-l,200Won/$

- 642 -

KAERI/RR-1999/99

Table 6.2-13

Unit Cost and Economic Parameters of DUPIC Fuel Handling

Content

Capital Cost per Plant

Burnup of DUPIC Fuel

(MWd/MTU)Annual Feed of

DUPIC fuel (MTHM)

Economic Parameters

Sub-content

Spent Fuel Storage Bay Loading Equipment

Sent Fuel Port Pusher

Spent Fuel Port Blow Dryer

Gamma Ray Detector

Fuel Loading System Control Program

Modification

Design Documentation

Cooling Capacity Increase of Spent Fuel

Pool Storage

Total

14,900

47.592

discount rate(%)

basis year

Levelized Unit Cost ($/kg)

Cost (k$)

1,000.0

666.7

166.7

166.7

333.3

500.0

916.7

3,750.0

5.00

2000

5.12571

*Exchange rate assumption : 1200 Won/US$

- 643 -

KAERI/RR-1999/99

Table 6.2-14

Life Cycle Cost and Unit Cost of DUPIC Fuel Handling

Year

2020202120222023202420252026202720282029203020312032203320342035203620372038203920402041204220432044204520462047204820492050Total

Capital Cost

(US$)

3,750,000

Net Present

Value (US$)

1,413,336

1,413,335.56

Annual Feed of

DUPIC Fuel (kg)

47,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,59247,592

Net Present Value

of Feed (kg)

17,08316,26915,49514,75714,05413,38512,74712,14011,56211,01210,4879,9889,5129,0598,6288,2177,8267,4537,0986,7606,4386,1325,8405,5625,2975,0454,8044,5764,3584,150

275,734.49

Levelized Unit Cost ($/kg) = 5.12571

- 644 -

KAERI/RR-1999/99

Hot Cell (Root A21

SPENT FUEL DISCHARGE B A Y - J ~L

Storage Boy Transfer Station (Root B2)

SPENT FUEL STORAGE BAY

Transfer Station (Root B l l

DEFECTED FUEL BAY

CANNEDDEFECTED FUEL

STORAGE TRAYS

Fig. 6.2-1 CANDU-6 Refueling Sequence

- 645 -

KAERI/RR-1999/99

Fig. 6.2-2 Current CANDU-6 Fuel Transfer Path

- 646 -

KAERI/RR-1999/99

JL REACTOR2. FUELUNO MACHINE BRIDGES3. FUEUJNO MACHINE CARRIAGES4. FUELUNO MACHINE HEADS5. FUELLING MACHINE CATENARIES6. SHIELDING DOORS7. CATENARY SYSTEMSS. FUEUJNS MACHINE MAINTENANCE LOCK TRACK9. FUELUNS MACHINE MAINTENANCE LOCK

10. NEW FUEL TRANSFER MECHANISMS1 1 . SPENT FUEL DISCHARGE BAY12. SPENT FUEL DISCHARGE PORTS13 . CALIBRATION PORTS14. ANCILLARY PORTS15. REHEARSAL FACILITY16. SPENT FUEL LADLE DRIVES17. DEFECTED FUEL CANNING EQUIPMENTS18. SPENT FUEL TRANSFER CONVEYOR19. SPENT FUEL RECEPTION BAY2 a RECEPTION BAY EQUIPMENT2 1 . DEFECTED FUEL STORAGE BAY22. DEFECTED FUEL STORAGE BAY EQUIPMENT23. SPENT FUEL STORAGE BAY24. STORAGE BAY MANBRIDGE26. STORAGE TRAY SUPPORTS26. SEMIAUTOMATIC SPENT FUEL HANDUNG SYSTEM27. NEW FUEL STORAGE ROOM23. FUEUING MACHINE HEAD TRANSPORT CART29. EQUIPMENT AIRLOCK30. FUEL HANDLING SYSTEM CONTROL CONSOLE

Fig. 6.2-3 CANDU-6 Fuel Handling System

- 647 -

KAERI/RR-1999/99

1 END FITTING2 BALL VALVES3 ELEVAVNG LADLE HOISTS4 ELEVAVNG LADLE DRIVE (IN NEW FUEL ROOM)5 ELEVAVNG LADLE6 MAIN ELEVATOR RAILS7 GUIDE RAILS8 FUEL POSITIONING ASSEMBLIES9 LOWER RAIL SUPPORT

10 AUXILIARIES11 SPRAY HEADERS12 REMOVABLE PLATFORMS13 FUEL TRANSFER EQUIPMENT14 DEFECTED FUEL CANNING EQUIPMENT15 SAFEGUARD MONITORS

Fig. 6.2-4 Spent Fuel Discharge Elevator

- 648 -

KAERl/RR-1999/99

SPENT FUEL DISCHARGE EQUIPMENTTRANSFER RACK DETECTIONDISCHARGE BAY CONVEYORTRANSFER CANAL CONVEYORTRANSFER CARTCONVEYOR DRIVES

T. RECEPTION BAY8. TRANSFER RACK9. TRANSFER RACK HANDUNO TOOL

10. RACK HANDUNd TOOL STORAGE BRACKET11. 2 RON RECEPTION BAY CRANE12. SINGLE RACK STAND-OFF13. EMPTY RACKS ON TRIPLE RACK STANDOFF14. STORAGE TRAY STAND15. PARTIALLY FILLED TRAY ON STANDI S . BUNDLE LIFTING TOOL17. FUEL STORAGE TRAYS18. STORAGE TRAY CONVEYORI T . CONVEYOR DRIVE20. STORAGE TRAY UFHNQ TOOL2 1 . SPENT FUEL STORAGE BAY22. EMPTY STORAGE IRAYS23. DEFECTED FUEL TRANSFER EQUIPMENT24. DEFECTED FUEL STORAGE TRAYS25. DEFECTED FUEL BAY ISOLATION VALVE28. ISOLATION VALVE DRIVE

Fig. 6.2-5 Spent Fuel Transfer Equipment

- 649 -

KAERI/RR-1999/99

MAIN STORAGE BAY FLOOR -

DETAIL "A1

CAP SEAL

Fig. 6.2-6 Spent Fuel Storage Tray

- 650 -

KAERI/RR-1999/99

FUEL TRANSFERFLASK

SHIELDED BASKETDRYING AND

WELDING STATION

BASKET ATDRYING POSITION

WELDING TORCH

TV CAMERA

WELDING TORCHLOADING SHAFT

SPENT FUEL BAY IRRADIATEDFUEL BAY WALL

T" FUEL TILT TABLE

UNDERWATERWORK TABLE

Fig. 6.2-7 Spent Fuel Shielded Basket Drying and Welding Station

- 651 -

KAERI/RR-1999/99

Fig. 6.2-8 Short-Term Spent DUPIC Fuel Decay Heat per Bundle

- 652 -

KAERI/RR-1999/99

X

10,000

1,000

100

10

1

—»*•• ••

H s1

\

» -

**<

h

0.0001 0.001 0.01 0.1 1

Time (Year)

10 100

Fig. 6.2-9 Long-Term Spent DUPIC Fuel Decay Heat per Bundle

- 653 -

KAERI/RR-1999/99

<L>

I

80

75

70

65

60

55

50

45

40

4 bundle shift (2 bundle per channel) refueling

I

:

;I

*

r

r i i i

* < . /

i i i i

^

0 10 20

L_

30 40 50

Time (min)

60 70 80

Fig. 6.2-10 Magazine Temperature from 4-Bundle Shift (2 Bundles per Channel) Refueling

- 654 -

KAERI/RR-1999/99

p

n.

I

80

75

70

65

60

55

50

45

40

8 bundle shift (2 bundle per channel) refueling

•

rfwmJ

u

1

1

N

20 40 60 80 100Time (min)

120 140 160 180

Fig. 6.2-11 Magazine Temperature from 8-Bundle Shift (2 Bundles per Channel) Refueling

- 655 -

KAERI/RR-1999/99

80

75

704)

H

pei

Tem

65

6055

50

45

40

8 bundle shift refueling per channel

i /-

//

/

, , ,\

10 20 30Time (min)

40 50

Fig. 6.2-12 Magazine Temperature from 8-Bundle Shift Refueling per Channel

- 656 -

KAERI/RR-1999/99

- 80% reactor power operation • 100% reactor power operation

Fig. 6.2-13 Storage Bay Temperature from Spent Fuel Decay Heat

- 657 -

KAERI/RR-1999/99

. 80% reactor power operation . 100% reactor powe operationr

4000

3500

: 3000

, | j 2500

! 1 2000

; S? 1500to

• a IOOO

: 500

i

i—""

( = •

10 15 20

Time (year)

25 30 35 :

Fig. 6.2-14 Storage Bay Heat Load from Spent Fuel Decay Heat

- 658 -

KAERI/RR-1999/99

80% reactor power op. 100% reactor power op.

25

10 20 30

Time (day)

Fig. 6.2-15 Storage Bay Temperature from Full Core Dump after 31-Years of Reactor

Operation

- 659 -

KAERI/RR-1999/99

I -9000 |

7son

•^ 6000

| | 4500 t

j 1* 3000 (

j 1500

1 oi (

- 80% reactor power op.

D 10

—*— 100%

RF~ ~. —. *

20 30

Time (day)

reactor power op.

40 50

Fig. 6.2-16 Storage Bay Heat Load from Full Core Dump after 31-Years of ReactorOperation

- 660 -

KAERWRR-1999/99

• 80% reactor power op. 100% reactor power op.

40

8OS

g

35

30

ca

25

10 20 30 40 50

Time (day)

Fig. 6.2-17 Storage Bay Temperature from Full Core Dump after 12-Years of Reactor

Operation

- 661 -

KAERI/RR-1999/99

• 80% reactor power op.

9000 r

7500

6000

4500

3000

1500

0

10


Ar.Vd xzzszzz

20 30

Time (day)

40 50

Fig. 6.2-18 Storage Bay Heat Load from Full Core Dump after 12-Years of ReactorOperation

- 662 -

KAERI/RR-1999/99

100% reactor power op.80% reactor power op.

10 15Time (day)

20 25

Fig. 6.2-19 Storage Bay Temperature from Half Core Dump after 31-Years of Reactor

Operation

- 663 -

KAERI/RR-1999/99

Hea

t (K

Dec

! t

9000

7500

6000

4500

3000 '

1500

0

— 80% reactor power op.

I 1

3 5 10

—

15

Time (day)

100% reactor

• - * • * t. t

>->-« t i

20

power op. j

|

•

25 30

Fig. 6.2-20 Storage Bay Heat Load from Half Core Dump after 31-Years of ReactorOperation

- 664 -

40

£ 353

4 - *

8.

30

25

KAERI/RR-1999/99

• 80% reactor power op. 100% reactor power op.

N-*-*-T-^V . . — _ •

* ^ a A *•» » ^ i •

10 15 20

Time (day)

25 30

Fig. 6.2-21 Storage Bay Temperature from Half Core Dump after 12-Years of Reactor

Operation

- 665 -

KAERI/RR-1999/99


(KW

)H

eati

;cay

Q

9000

7500

6000

4500

3000

1500

0

100% reactor power op.

Ar \....

•+-+-* ft a-.

10 15 20

Time (day)

25 30

i

Fig. 6.2-22 Storage Bay Heat Load from Half Core Dump after 12-Years of ReactorOperation

- 666 -

KAERI/RR-1999/99

Fig. 6.2-23 DUPIC Fuel Transfer Path

- 667 -

KAERI/RR-1999/99

6.3 ] § f %& ^ $ l

CANDU ^

7H^*H 1990V1 5^1 DUPIC

- DUPIC ^ £ 5 . ^ 7 ] ^ ^ 3 . 4-§-^ *|&g.l- CANDU

CANDU

(once-through cycle 2.^- direct disposal option) H] J2.tr}-©}

DUPIC ^ ^ 5 . ^ 7 ] fl ^ ^ ^ ]

^•*1| DUPIC 7 ]^o

DUPIC ^ ^ S ^ 7 ] ^ ,S.i;> 7^1^^ ^ ^ # ^ ^ ^ $ 1 4 . DUPIC

, DUPIC ^ ^ S ^ 7 ) «]-§-

DUPIC. ZL*tl 6.3-l gr *l^:aI-§- ^7} 2HJ# i i^^u]- . ^*1|3, Af§-

9i DUPIC

AECLo] i l j l ^ "The Disposal of Canada's Nuclear Fuel Waste: Engineering for a

Disposal Facility" [Ref. 16]^- £

- 668 -

KAERI/RR-1999/99

6.3.1

6.3.1.1

- 669 -

KAERI/RR-1999/99

#t*1|(buflfer), nf^fl(backfill), ZL

^71

6.3.1.2

DUPIC

room-and-pillar

r 7H^o|c>. room-and-pillarfe ^1^« | -^ nH^Lflofl *>i4 ^ ^ - n. o|

^ roora-and-pillar

- 670 -

KAERI/RR-1999/99

7}.

CANDU fe ~10,100,0007fl CANDU

-191,000 Mg

cf. a

4

57H

1000 m

fe zj-

- 671 -

KAERI/RR-1999/99

6.3.2.

OECD/NEA

[Ref. 18]fe

"The Cost of High Level Waste Disposal in Geological Repositories"

H]

CANDU

fe 6.3.2.2^1 7}

4 6.3.2.3^2}- 6.3.2.4^<>\]

6.3.2.1

6.3-

6.3-

JE. 6.3-

- 672 -

KAERI/RR-1999/99

71-

71-g- \dH-i- i^m^cf. o| Hj-g-^ 1991 L*

S.€: a}-§-^ ^^1 ^ 1 ^

NEA A}^-^-^ i | * H ^*BStK ^* | |S ^ 7 f e 1991 id

M 7

1991 id 7^

OECD/NEA

6.3-15]

- 673 -

KAERI/RR-1999/99

3.71

, CANDU ^QS-S!] ^ # « a ^ £ f e nfl-f- ^7]

6.3.2.2

DUPIC 4 - § - ^ ^<£^.£| ^ ^ a | ^ - o | , ^^-Si!} CANDU

' Slit:}.

A

D U P I C

7}.

CANDU ^«g^-e fe DUPIC ZLZ\JL ^^S. *}&%- ^ « ^ S . ^ 37>

}. ol^«> «Jo4S5j y J # ^ ^ £ ^ ^ ^ - ^ - e f e DUPIC, l ie]

JZ. ^ g ^ ^ ^^S.<H1 tU*H 4 4 8, 56, ZLelJL 35 MWdTcgHEojt:}. DUPIC

S^l HM-^Êfe ^ ^ r S i } CANDU ^4^-<Hl^^ ^ ^ S . Q±S. Î

DUPIC ^ ^ S ^ l 7f^«> ilcfl ^#«g4iSfe CANDU ^^>ôfl>H 21 MWd/kgHES.

6.3-lofl ^<H^1 « ^ e o ^ ^ ^ 1 1 - J 1 ^ * H , ^ ^ ^ 5000

- 674 -

KAERI/RR-1999/99

2000 TWh^- 10900 TWh# til-g-^M- 4\Q *£ • *}th$i-£..S. A]~g-

CANDU #*}3.c% tfl*fl>*| «iJL

1 £ 5330 TWhofl cfl*> DUPIC

13320 MTHE (=5330X10756000/24/0.3)0)4.

CANDU ^ ^ - f e f e A]~g-^ ^ ^ ^ 6 f l c|*> 7 ] ^ ^^r-g-7]^ 7 2 7 ^

(1362.7 kgU)# ^-§-*fe ^M¥ -S-7]olcl-. 10\i^> ^ 4 ^ 4 - § - ^ ^ ^ 5 . 727H^: -337W oju]-. ^ ^^ .of l^^ . DUPIC

DUPIC

-8-7]if -8-*}*M-. OECD/NEA ^ ^ - ^ ] 1 8 ^ * } ^ , ^7 l 2000 TWh#

-7840 MgHMo]3.5 o|^^h ^ 4 ^ ^ ^ ^ ^ -5300

^ ^ € ^ > . 4 5 W , 4 -§"71^ Af-g-^ « | o iS -1,480 kgHM

, o] -g-7]^ 40Td?I: U I 4 € A]-§-^ t ^ ^ ^ S *U^^[ 4 , -1050

6.3.2.3

DUPIC

^<$.£. ^ ^ ^ ^r£^lt>^r OECD/NEA ^J IA^] 1 8 ^7]*}^ 85°C

- 675 -

KAERI/RR-1999/99

7}. CANDU

*J 5 m, ^ o ] 8

^ ^ y ^ : 1000 m

5000 TWhS]

. CANDU DUPIC

DUPIC 4 S ^

>. DUPIC 4 g ^ ^ ^ ^ ^ ^

flll ^ } ^ ^ r ^ ^ r 1000 m

10 m 1& fll^^ S l H -733

7)

85°C o

71 ^ * H , -8-71^- 6 m ^ A S MH I JL ^^r^J- 3<>lfe 500 m l

5000 TWh

-68^014.

- 676 -

KAERI/RR-1999/99

S 6.3-3ofl

c}. 7 ^ ] DX, DY HelJL HX ("L^ 6.3-3)

}. 7] el DX, DY, T£

7]e]

t fecf. S 6.3-3^ ^ 7 1 ^ e o > 5000

71 el DX, DY, a

£ S 6.3-4ofl ^A]SlSd4. DUPIC

CANDU * U 1 - ^ | &7] ^^^S] ^°1^K ^ " i " # ^ , 5330 TWh

CANDU ^VS<HJ-H^ ^71-^^^= 2000

3330 TWh£| ^olc>. ^[7l ^ ^ 1 % ^ 13330 TWhif 29070

CANDU *J*fc3.oflA-1^1 ^ 7 l ^ ^ ^ ^ S r 4 4 5000Jf 10900

CANDU

zj-

3^Stic}.

^1*}7l fl*B, 3E 6.3-4O11

H «l.fi-# ^ ^ # 4 ^ t > ^ 1 ^ 7 l # ^ H v ^ 4 # T4B}H«C|-. S 6.3-3

6.3-4OJ1

6.3.2.4

OECD/NEAfe

OECD/NEAfe

- 677 -

KAERI/RR-1999/99

CANDU *i<£-f e f e DUPIC

, OECD/NEA<HI

. 3. 6.3-

-8-71

6.3.3

CANDU ^i^-f-efe, DUPIC

6.3.3.1

6.3.3.1 ^-tfl

12716

MWe)7} ^ r ^ ^«Hl $a^, 57l(#-§-3o> 5000 MWe)7> ^ ^ #o1] SdcK 1999^

45484 MWeofl ^Qv.}. o ) ^ ^ 28%7} ^

7\.

2030 ^M Q

-.21 2016^

- 678 -

KAERI/RR-1999/99

ZL^ 6.3-4^ £ § Qx}^ iHldfcSj ^ 7]Q ^-<£ %• -§-*£ £ 3 f # J i o ^ t : } . 40

7] £| 3-^SSj- 1971 £j CANDU ^^j-^7} $X^*\], < > 1 ^ DUPIC

f. 2016^

^ 0.711

^ 0.25 wt%]

DUPIC

7] * ^ . a E NUFCAP[Ref. 26]# Aj-g-ô^ sg7>*>Sit:K o| SJE.^ 1996^

6.3.3.2

CANDU *I<2-feB=f, DUPIC %6.3.3.1 aloM - ^ - S ^ ti^3ov ^ - J S . ^ S 6.3-5^1

JE.-E- ^ ^ 5 . ^ 7 ] »J-6>ofl cfl*foj ^ #ê o >^ 12411

-, # w i ? i % ^ ^ ^ S 9289 TWhi f CANDU ^1^>S 3219 TWh^.

. CANDU î^-?-efe DUPIC

- 679 -

KAERI/RR-1999/99

4 4 77, 168, 9| 270 $/kgHE o|t>.

6.3-8^ &£, ^^g ^ ^ H l * 6 . 3 . 3 . 1 ^ ^ ^ 7

^>^ti|-§-^r CANDU € ^ - T - ^ } ^ - O | 1.59 M$/TWh,

DUPIC ^ " £ 5 . ^ 0.53 M$/TWh, H5]^L ^^S. ^ ^ S ^ 1.37 M$/TWho] v\.

LUC ^ ^ ^

- 680 -

L U C = N P B ( 6 3 " 2 )

( 6-3 '3 )

KAERI/RR-1999/99

(2020-2046)

10\d (2010-2019)

Si ^ ^ l t h A ^ ^ o ] ^ £ 5 } Jf 2Td (2047-2048)

: 2020

5%

S. 6.3-9, 6.3-10 Rj 6.3-11^- ^-g- JL#3f g|A-| ^*> 7>^^-S . 71]^:^ CANDU

-, DUPIC ZLels. ?g^S. A]~§-^ ^^S.^1 tnt> a ^ l - g s j.3-12fe 37W Al-g-^ ^ ^ ^ EHofl tflsfl ^^^7lHl-g-^ NPV

g CANDU

118 $/kgHM, DUPIC «(q4S 220 $/kgHM, ZLB]3.

403$/kgHM ©jr;]-.

6.3.4

DUPIC

CANDU ^ ^ - f e f e o l 1.59 M$/TWh, DUPIC « J « 1 S ^ 0.53 M$/TWh, IL

1.37 M$/TWh<>lt:f. DUPIC

^ - ^ , CANDU

CANDU

DUPIC ZLB13. %^S. Aj-g^f ^^S«H1 cfl^M 4 4 H8, 220, IielJL 403

DUPIC

fe DUPIC

- 681 -

KAERI/RR-1999/99

Table 6.3-1

Cost Estimates for Packaging and Geological Disposal of Spent Fuel [Ref. 18]

Spent Fuel (tU)

Corresponding Electricity generation(TWh)

Volume of waste (mj)

Packaging

Characteristicsof repository

Estimated cost

Normalizedcost

Inclusion of packing cost

Container (thickness)

Depth (m)

Host rock

Volume of excavatedrock (Mm3)

Operating period (year)

Sealing material

In national currency unit(base year)

In billion of U$ of July1991

Cost per unit electricitygeneration (M$/TWh)

Cost per unit weight ofwaste (k$/tU)

Canada

191000

10900

99000

Yes

Titanium (6.3 mm)

1,000

Crystalline rock

7.2

41

Bentonite/sand

9,500 M C$ (1990)

8.7

0.80

46

Sweden

7840

2000

12900

Yes

Copper (10 cm)

500

Crystalline rock

0.8

27

Bentonite/sand

20.2 b SKr (1990)

3.2

1.6

410

- 682 -

KAERI/RR-1999/99

Table 6.3-2

Comparison of Disposal Containers

Overall length (mm)

Overall diameter (mm)

Thickness (mm)

Capacity (no. of fuel assemblies)

Spent Fuel Type

CANDU-NU

2246

645

6.3

72

CANDU-DUPIC

2246

645

6.3

72

. PWR

4500

800

100

4

- 683 -

KAERI/RR-1999/99

Table 6.3-3

Summary of Repository Data for 5000 TWh Electricity Production

Cooling time of spent fuel

Container capacity (max. no. of assemblies)

Actual no. of fuel assemblies per container

Actual amount of fuel per container (kgHM)

Initial container heat output (W)

No. of containers across the room width

Borehole spacing across the room width (m)

Pitch distance along the room length (m)

Center-to-center room spacing (m)

Room width (m)

Room length (m)

Max. container outer-surface temperature (C)

CANDU

-NU

10

72

72

1363

337

3

2.1

3.1

30

8

230

89

CANDU

-DUPIC

50

72

60

1095

733

1

-

10

16

4

230

89

PWR

40

4

4

-1480

-1050

1

-

6

25

4

250

<85

- 684 -

KAERI/RR-1999/99

Table 6.3-4

Summary of Repository Operation Data

Spent fuel repository

(TWh)

Amount of spent fuel

(Mg HE)

Number of containers

Disposal rate

(containers/year)

Years of operation

Sub-surface plan area

(km2)

CANDU-NU

(Ti container)

2000

33330

24460

3471

8

0.8

5000

83330

61150

3471

18

2.1

10900

181660

133300

3471

39

4.2

CANDU-DUPIC

(Ti container)

5330

13220

12100

3471

4

2.0

13330

33060

30200

3471

9

5.1

29070

72070

65800

3471

19

10.2

PWR

(Cu container)

2000

7840

5300

196

27

1.4

5000

19610

13250

196

68

3.5

10900

42750

28890

196

147

7.6

- 685 -

KAERI/RR-1999/99

Table 6.3-5

Breakdown of Disposal Costs (1991 U$ million)

Spent fuel

repository (TWh)

Construction

OperationDirect

Indirect

Decommissioning

CANDU-NU

(Ti container)

2000

1380

1265

345

920

5000

1610

2990

920

1265

10900

2070

6900

2070

1725

CANDU-DUPIC

(Ti container)

5330

1380

1380

230

920

13330

1725

3450

460

1265

29070

2300

7475

1035

1725

PWR

(Cu container)

2000

1955

1265

345

690

5000

2415

3105

805

1380

10900

2875

6670

1725

3450

- 686 -

KAERI/RR-1999/99

Table 6.3-6

Nuclear System Scenario up to 2030

Year

19781979198019811982198319841985198619871988198919901991199219931994199519961997199819992000200120022003200420052006200720082009201020112012201320142015201620172018201920202021202220232024202520262027202820292030

Nuclear Power Plant

Name

Kortfl

Kori#2/Wolsong#l

Kori#3Kori#4Yongkwang#2

Yongkwani>#2Uljin#lUljin#2

Yongkwang#3Yongkwang#4

Wolsung#2Uljin#3/Wolsong#3Uljin#4/Wolsong#4

Yongkwang#S,#6

Uljin#5Uljin#6

NPP"#1,#2NPP#3NPP#4

NPP#5/KNGR°#INPP#6/KNGR#2

KNGR#3KNGR#4

CANDU#1,#2KNGRS5,#6/CANDU#3

CANDU#4,#5KNGR#7,#8

KNGR#9/CANDU#6,#7CANDU#8,#9

KNGR#10/CANDU#10KNGR#U/CANDU#11

KNGR#12CANDUS12

KNGR#13,#14CANDIK13

KNGR#15,S16KNGR*17/CANDU#I4KNGR#23/CANDU#I5

Generating Capacity (MWe)New

PWR587

650

9501,900

950950950

1,0001,000

1,0001,000

2,000

1,0001,000

2,0001,0001,0002,3002,300

1,3001,300

2,600

2,6001,300

1,3001,3001,300

2,600

2,6001,3001,300

PHWR

679

700700700

1,400700

1,400

1,4001,400

700700

1,400

700

700700

DecommissionPWR

587

650

9501,900

950950950

1,0001,000

1,0001,000

PHWR

679

700700700

Installed Capacity (MWe)

PWR587587587587587

1,2371,2372,1874,0875,0375,9876,9376,9376,9376,9376,9376,9377,9378,9378,9379,937

10,93710,93710,93712,93712,93713,93714,93714,93716,93717,35018,35020,65022,95022,95023,60024,90023,95022,05023,70022,75024,40025,70025,70027,00028,30029,60028,60030,20030,20031,80032,10033,400

PHWR

679679679679679679679679679679679679679679

1,3792,0792,7792,7792,7792,7792,7792,7792,7792,7792,7792,7792,7792,7792,7792,7792,1002,1002,1003,5004,2005,6005,6007,0008,4009,1009,8009,800

11,20011,20011,20010,50010,50011,200

Total587587587587587

1,9161,9162,8664,7665,7166.6667,6167,6167,6167,6167,6167.6168,6169,616

10,31612,01613,71913,71613,71615,71615,71616,71617,71617,71619,71620,12921,12923,42925,72925,72925,70027,00026,05025,55027,90028,35030,00032,70034,10036,10038,10039,40039,80041,40041,40042,30042,60044,600

Power Generation (MWyr)

PWT205364393329 '434636888

1,1272,5823,7133,8724,6825,4625,8115,8275,9315,5506,3507,1507.1507,9508,7508,7508.750

10,35010,35011,15011,95011,95013,55013,88014,68016,52018,36018,36018.88019,92019,16017.64018.96018.20019,52020,56020,56021,60022.64023.68022,88024,16024.16025,44025,68026,720

PHWR0000

0316455638543631536618584618581672577577577

1,1721,7672,3622,3622,3622,3622,3622,3622,3622,3622,3622,3622,3622,3622,3622,3622,3621,7851,7851,7852,9753,5704,7604.7605,9507,1407,7358,3308,3309,5209,5209,5208,9258,9259,520

Total205364393329434952

1,3431,7663,1254,3444,4085,3006,0456,4296,4086,6036,1276,9277,7278,3229,717

11,11211,11211,11212,71212,71213,51214,31214,31215,91216,24217,04218,88220,72220,72220,66521,70520,94521,61522,53022,96024,28026,51027,70029,33530,97032,01032.40033,68033,68034,36534,60536,240

*The nuclear system form the year 2016 is based on the following assumptions:-Electricity capacity reserve ratio is 20% from the year 2016.-Average Increase rate of maximum electricity demand is 2%/year-Nuclear share of electricity capacity is 37% up to the year 2020. 40% up to the year 2030.-Plant load factor is 80% for CANDU.-Plant life-time is 30 year for all types.*NPP means the type of Korean Standard Nuclear Power Plant (Uljin#3,#4)°KNGR means the Korean Next Generation Reactor being developed

- 687 -

KAERI/RR-1999/99

Table 6.3-7

Results of Material Flow and Electricity Generation for Fuel Cycle Options

Items

DUPIC

Cycle

Direct

Disposal

PWR Interim Storage (ton)

CANDU/DUPIC Interim Storage (ton)

DUPIC Facility (ton)

Disposal Capacity

(ton)

Cumulated Electricity

Generation (TWh)

PWR

CANDU

DUPIC

PWR

CANDU

Total

PWR Interim Storage (ton)

CANDU Interim Storage (ton)

Disposal Capacity

(ton)

Cumulated Electricity

Generation (TWh)

PWR

CANDU

PWR

CANDU

Total

Total treated amount

23,230

31,700

21,188

7,930

10,512

21,188

9,289

3,129

12,411

29,118

42,094

29,118

42,094

9,289

3,129

12,411

- 688 -

KAERI/RR-1999/99

Table 6.3-8

Cost Break-Down for Disposal Facility (1991 U$ million)

Spent fuel repository

(TWh)Spent fuel repository

(MTHM)

Construction

Operation

Direct

Indirect

Total

Decommissioning

Total Cost

CANDU-NU

(Ti container)

3129

53671

1467

1914

561

2475

1050

4992

CANDU-DUPIC

(Ti container)

12411

36861

1685

3212

434

3646

1225

6556

PWR

(Cu container)

9289

36861

2749

5697

1474

7171

2885

12805

- 689 -

KAERI/RR-1999/99

Table 6.3-9

Discounted Disposal Costs for CANDU-NU Spent Fuel

Year

201020112012201320142015201620172018201920202021202220232024202520262027202820292030203120322033203420352036203720382039204020412042204320442045204620472048Total

Cost (k$)

Capital

146700.00146700.00146700.00146700.00146700.00146700.00146700.00146700.00146700.00146700.00

1467000.00

O&M

90296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.30

2438000.00

Decom.

525000.00525000.001050000.00

Total

146700.00146700.00146700.00146700.00146700.00146700.00146700.00146700.00146700.00146700.0090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.3090296.30525000.00525000.004955000.00

Total NPV

58054.17155289.68752656.84550149.37647761.31145486.96243320.91741258.01639293.34837422.23721937.15320892.52719897.64418950.13818047.75017188.33316369.84115590.32514847.92914140.88413467.50912826.19912215.42811633.74111079.75310552.14610049.6639571.1079115.3408681.2768267.8827874.1747499.2137142.1086802.0076478.1026169.62134163.20132536.382874680.25

Production(MTHM)

1987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.811987.81

53671.00

DiscountedProduction(MTHM)

482.93459.94438.03417.18397.31378.39360.37343.21326.87311.30296.48282.36268.91256.11243.91232.30221.24210.70200.67191.11182.01173.34165.09157.23149.74142.61135.82

7425.17

Net Present Values (k$) = 874680.2


- 690 -

KAERI/RR-1999/99

Table 6.3-10

Discounted Disposal Costs for CANDU-DUPIC Spent Fuel

Year

201020112012201320142015201620172018201920202021202220232024202520262027202820292030203120322033203420352036203720382039204020412042204320442045204620472048Total

Cost (k$)

Capital

168500.00168500.00168500.00168500.00168500.00168500.00168500.00168500.00168500.00168500.00

1685000.00

O&M

135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04

3646000.00

Decom.

612500.00612500.001225000.00

Total

168500.00168500.00168500.00168500.00168500.00168500.00168500.00168500.00168500.00168500.00135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04135037.04612500.00612500.006556000.00

Total NPV

66681.17263505.87860481.78857601.70354858.76552246.44349758.51747389.06445132.44242983.27832806.75131244.52529756.69128339.70526990.19625704.94824480.90323315.14622204.90121147.52420140.49919181.42818268.02717398.12116569.63915780.60815029.15114313.47713631.88312982.74612364.52011775.73311214.98410680.93710172.3219687.9259226.59539857.06837959.1121122865.11

Production(MTHM)

1365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.22

36861.00

DiscountedProduction(MTHM)

331.68315.88300.84286.51272.87259.88247.50235.72224.49213.80203.62193.92184.69175.89167.52159.54151.94144.71137.82131.26125.01119.05113.38107.98102.8497.9493.28

5099.58



- 691 -

KAERI/RR-1999/99

Table 6.3-11

Discounted Disposal Costs for PWR Spent Fuel

Year

201020112012201320142015201620172018201920202021202220232024202520262027202820292030203120322033203420352036203720382039204020412042204320442045204620472048Total

Cost (k$)

Capital

274900.00274900.00274900.00274900.00274900.00274900.00274900.00274900.00274900.00274900.00

2749000.00

O&M

265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56

7170000.00

Decom.

1442500.001442500.002885000.00

Total

274900.00274900.00274900.00274900.00274900.00274900.00274900.00274900.00274900.00274900.00265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.56265555.561442500.001442500.0012804000.00

Total N P V

108787.265103606.91998673.25693974.53089499.55285237.66981178.73277313.07873631.50370125.24164515.74561443.56758517.68355731.12653077.26350549.77548142.64245850.13643666.79641587.42539607.07137721.02035924.78134214.07732584.83531033.17729555.40628148.00626807.62525531.07124315.30623157.43422054.69921004.47520004.26219051.67818144.45693867.46289397.5832057234.33

Production

(MTHM)

1365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.221365.22

36861.00

Discounted

Production

(MTHM)

331.68315.88300.84286.51272.87259.88247.50235.72224.49213.80203.62193.92184.69175.89167.52159.54151.94144.71137.82131.26125.01119.05113.38107.98102.8497.9493.28

5099.58



- 692 -

KAERI/RR-1999/99

Table 6.3-12

Disposal Unit Costs for Three Different Spent Fuels

Construction cost total (M$)

Annual operation and maintenance(M$)

Decommissioning total (M$)

Life cycle cost in net presentvalue (M$)

Waste production in net presentvalue (MTHM)

Levelized unit cost ($/kgHM)

CANDU-NU(Ti container)

1467

90

1050

875

7425

118

CANDU-DUPIC(Ti container)

1685

135

1225

1123

5100

220

PWR(Cu container)

2749

266

2885

2057

5100

403

- 693 -

KAERI/RR-1999/99

Literature Survey on HLW

Disposal Facility

Reference Selection for HLW

Disposal Cost Estimation

Analysis of Disposal Capacity

in Korea

Future Nuclear Grid Analysis

ICumulated Electricity

Generation

ISpent Fuel Arising

Unit Cost Estimation of HLW

Disposal for Spent CANDU

and DUPIC fuels

Fig. 6.3-1 Procedure of HLW Disposal Cost Estimation

- 694 -

KAERI/RR-1999/99

EIS 4-1.03

Fig. 6.3-2 Spent Fuel Disposal Facility Perspective [Ref. 16]

- 695 -

KAERI/RR-1999/99

TitaniumContainer

Fig. 6.3-3 Waste Emplacement Geometry for an Underground Facility [Ref. 20]

- 696 -

KAERI/RR-1999/99

50,000

45,000

40,000

a, 35,000

| 30,000

.-£• 25,000

g. 20,000

O 15,000

10,000

5,000

0

/

J

total of

• total

40

of

• i

Units\'i , -

• iI Jn^

IDPWR"\ IBPHWR

A

1978 1988 1998 2008 20182028 2038

Year2048 2058 2068 2078

Fig. 6.3-4 Installed Capacity of Nuclear Power Plants

- 697 -

KAERI/RR-1999/99

PWR (9,289 TWh, 270$/kgHE)

CANDU-DUPIC(12,41 lTWh, 168$/kgHE)

CANDU-NU(3,129 TWh, 77$/kgHE)

0 5000 10000 15000 20000 25000 30000Electricity Generation (TWh)

Fig. 6.3-5 Disposal Unit Costs for Spent CANDU-NU, CANDU-DUPIC and PWRFuels [Ref. 20]

- 698 -

KAERI/RR-1999/99

6.4 DUPIC ^ I M ] §

CANDU &*}$. 47)7}

PWR ^<as<Hlfe €*}3.ofl 4-g- 7|-^t> a ^ - S # * H £<2 -f e>0.7 wt%)^ 2Bl| O]^1- S ^ 5 ] ^ 5Ll7l 4^:^] °11- ^7Fg-*H CANDU ^ ^ }

CANDU ^^}SÎ ^^.*> ^ g - f s t e S . ^ A ^711 5]fe ^ ^ ^ ^ ^ )

. DUPIC r o|^«> 7 ] ^ 7||^§H1 PWR f CANDU

DUPIC « } ^ 5 . ^7}S] B>^Jg ^ - ^ ^ ^ DUPIC ^ 7 | ^

(safeguardability)<>l

^ ^ ^ fSÎ DUPIC

DUPIC

ZL ^ - 9 } D U P I C ^ ^ > # ] ^ 1 H ^ l 1 ? > ? J ^

l ^ ^ l % & ] ^ . D U P I C

|, DUPIC

DUPIC « ? ^ ^ ^ CANDU Î -Sof l -H^ ^ « l - § - , ^

} 4 DUPIC

DUPIC ^ ^ ] fl^l}

|{ al-g- ^7j-7f 9}x}r\ AECL ^

1 ^ CANDU ^ * } ^ H ^ 1 DUPIC

DUPIC ^ ^îfe 3§7>e> ^ ^ a l ^ f . jEta 7 ] ^ i=>4l«]-§-#

- 699 -

KAERI/RR-1999/99

DUPIC

6.4.1 7 ^ ^<&3.

DUPIC «)o4S^7]6fl^ 4 - g - ^ PWR ^<&£.±r 3*1 ^ 7 > ^ # 7^^ CANDU

DUPIC ^ 9 1 S . # ?>#ôl : ?>i;>. o] #AT

DUPIC ^ ^ ^ o l l f e S

%&&$ l ^ ] % ^ l ^c>. o | ^ AÂ>^ DUPIC

P W R ^ £ ] ^ J ^ 1 fH

DUPIC

SEU

DU# *Q[2&%«Qi*\ %7}tfe 7ô\t\. ojel*]; 7 f l ^ # 7 ^ 2 ^ ^ DUPIC

CANDU

I: SEU v± DU#j l O i-~ -r rrf27

2: SEU 5J DU#

3: A > ^ f PWR ! i # ? & ^ ] 1 ^ ^ ^

S 6.4-H

- 700 -

KAERI/RR-1999/99

6.4.1.1

PWR

blending

r DUPIC

PWR

wt% DU# °l-g-*HS U235 9J Pu239

%±3. ^}]dt\. o] ^ - f ofl 7 ] § DUPIC

0.45 wt%# Tg$^ ^ Sa^.^ 4

SEU

^ ^ 3.5 wt% SEU |- 0.25

DUPIC ^ ^ S . # «># ^ $X^

S ^ ^ U235 1.0 wt% % Pu239

96%#

6.4.1.2 2)

DUPIC

. DUPIC

PWR

PWR 7HJ-

>?Kg.£(targpt reactivity)!-

^ . DUPIC

^ . f e 100%

DU 1.1 wt%S.

35]

SEU ^-^r DU

PWR

S E U 2.3 wt%

6.4.1.3 3)

PWR DUPIC

- 701 -

DUPIC

1.57 wt°/o©lt}.

DUPIC

KAERI/RR-1999/99

6.4.2 DUPIC

OECD/NEA

6.4-2 1 t^Lfl Sit}. DUPIC DUPIC

6.4.2.1 DUPIC

DUPIC

DUPIC

DUPIC

5%,

558

4 4 619

40()S.

SEU ^ DU

580 $/kgHES.

6.4.2.2 DUPIC

DUPIC CANDU

- 702 -

KAERI/RR-1999/99

3* g ^ l S d l f DUPIC

C A N D U | i l }]

^7} ^-H] til-g-i)- - i ^ H l - g ^ JS.J^ JL3|*> CANDU

3,750,000 $ (2000\1

DUPIC ^<?1S.^-

5.0 $/kgHM< ]A-| 5.3

6.4.2.3

CANDU 9J DUPIC ^i^S. ^MQ til-g-^l OECD/NEA<M|

CANDU ^ ^ 5 . ^ ^ H ' - W l - S ^ " ^7|-*f31, ^ i^ 1 # £ * } # wl-8-ofl 7}

DUPIC

^ CANDU

g- 35

ZLBl L CANDU Al^g-^ « } ^ S ^^Hj -g .^ . OECD/NEA 2LJL*H \}S\ $X^ 13

$/kgHM#

DUPIC ^ ^ ^ . 5 ] -g-2|<i%^ Af-g-^ CANDU

conditioning plant7>

DUPIC ^^5.^1 ^ ^ H v «l-g^ CANDU

170

6.4.2.4 4-8-^*1 g £

PWR !%<$.£., 4 - g - ^ CANDU « } ^ S ^ 4 - g - ^ DUPIC

] ^}u}t:f AECL611 16

- 703 -

KAERI/RR-1999/99

PWR «?«1S, 4-S-^1 CAM)U ^ ^ 5 . ^J A f § ^ D u p i c

403 $/kgHE, 118 $/kgHE £ 220 32

6.4.3

one-batch ^ ^ J S . ^ ^ ^ ^ 5 . ^ 7 ] H]-g-

M f 1 DUPIC ^«^5.^7] ^ - ^ # JL^*><H 7 ] ^ PWR

CANDU ^ 4 ^ # ^d^*>31 7 1 ^ ^J^S51 < i ^ # , o

6.4.3.1 ^<?!S.^7l H]-g-

^ ^ : OECD/NEAoflA-1

71

F.{t)

to : ?U

- 704 -

KAERI/RR-1999/99

L :

T, : ^ 7 ] #5) 3qcfl?)t

T2 :

«]-§- (Levelized fuel cycle cost, LFCQ

LFCC = (6.4-2)

uf.

7}7\

(non-parametric distribution)<>] A]~§.*>sit;>. o |

=

mode

- 705 -

KAERI/RR-1999/99

SD=

( 6 4 . 5 )

(6.4-6)

a 6.4-3^ £

s>BWN ft (

^ OECD/NEA

$/kgU, ZLeU

mode ft)o] ^

. ^ - ^ 6.4-2

130 $/kgU ^ ^

80 $/kgU, mode ^ HO

6.4.3.2 7] 71

DUPIC

MOX

^71

o)n) DUPIC

PWR

PWR

DUPIC

CANDU

DUPIC

DUPIC

CANDU & DUPIC

lead time^f lag ti

leadAag ti

. DUPIC

AR (At

AR

- 706 -

KAERI/RR-1999/99

^ ^§§Î 950 MWe^- PWRJf 713

CANDU ^ * > S # ?}%*}&&. o) #

6.4.3.3 ^<£S^7 l Hj-g-

S. 6.4-Hfe fcafl ^ - ^ ^ ^ S . ^ 5 g € 71$ DUPIC

CANDU-6 ^>Sof l> | ^ DUPIC

fe. DUPIC

14,900 MWd/MTHM, ^^1 27} 14,500 MWd/MTHM,

15,400

6.4-5<>flfe 71$ PWRJf CANDU - ^ ^ ^ . ^ A ^ # ^ ^ 1 # # ^Bf^fl ôlcl-. PWR

]$J»S. tj-^KHSi ».nf, CANDU^ ^ - f 17H

a. 6.4-6^ DUPIC «?«!S^7loflAl CANDU

I7fl i c ^ #5O>^. 7lôjg. *}#£. xt))^ ^ ^ ^ ^ 7 1 ^ -g -^ # ^ *

. DUPIC 7l$*K!jS.n}-i;} ^4iS7f ^?V*| C>s.7] 4 ^ ^ PWR

# : 0.711 wt%.

Tail assay: 0.25 wt%.

(Loss factors)

(conversion) : 0.5%

PWR ^ DUPIC *%<$.£. *\]£. : 1%

- 707 -

KAERI/RR-1999/99

- CANDU : 0.5%

S. 6.4-

CANDU

ratio)

DUPIC

£>t;r.

PWR

(PWR-to-CANDU reactor

6.4-7, 6.4-8

, DUPIC

4 4 >

DUPIC

DUPIC ^ CANDU

^ PWR

DUPIC

, 2

o]

o l ^ DUPIC

CANDU

mills/kWh,

DUPIC ^ « ^ ^ . *#«# HIA-}*] DUPIC

- 5.19 mills^Whl- }LO]JL ^-C}. DUPIC

^131 al-i-g- <£ r 5 1 ^ . tiJ:^<Hl, DUPIC

DUPIC ^7 lHJ -§ -£ 5.25 mffls/kWh,

5.42

,£- 5.24

6.4-lOg: S 6.4-7, 6.4-8

0.24 mills/kWh,

. DUPIC

0.1 mills/kWh,

0.01 mills/kWhl-

^ S . ^ DUPIC

DUPIC # l

SEU

71

- 708 -

KAERI/RR-1999/99

Latin Hypercube

tfl*1H, DUPIC

mills/kWh, HB|JL

mills/kWhS.

5.39 mills/kWh,

5.30 mills/kWh,

0.09

DUPIC

# 10,000*1

5.55 mills/kWh,

5.26 mills/kWh,

] ^ 0.29

0.35

0.37

0.34 mills/kWh, Zi

0.37 mills^WhS

\f DUPIC

S 6.4-

percentile

75th percentile ^ - ^ 95th

percentile O]

6.4-4, 6.4-5 £ 6.4-6^r 4

6.4-7, 6.4-8 9i 6.4-9^

positively skewed *fc2.20%

positive skewness $17]

6.4-10, 6.4-11 g 6.4-12^ ^

^ r ^ ^ r Spearman Rank Correlation ao ^ ° | 3 4

3.nDUPIC fe CANDU ^ ^ - f c ^

, DUPIC

DUPIC

- 709 -

KAERI/RR-1999/99

6.4-15<Hl

o]

39$2}- DUPIC

fe ^ 6.4-13, 6.4-14 £

3.5 wt% ^ - f f i f e 0>

^ - f fife(U3O8)^ 978$

, DUPIC ^71*1 ^ - f 3.5 wt% -^Hr-sH? <$

SEU % DU7> ^ j t * > L | CANDU

4 4 20% 51 23%*j

2 51 3*11 cH«> DUPIC

%SLS. 4 4 4^1"^} ( S 6.4-12

DUPIC

15,000

DUPIC

DUPIC

65%

P W R

CANDU

DUPIC

6.4.4

~b DUPIC

MOX

n] DUPIC

l ^ 7 1 (thermal recycling)^ «J

DUPIC

- 710 -

KAERI/RR-1999/99

DUPIC t!}<£5. *\}2. ^ - ^ # ^ 1 SEU ^ DU#

, DUPIC

PWR

DUPIC

o> ^ ^ O.

- DUPIC

- 711 -

# 3 DUPIC

3f DUPIC

23% ^§£ ^ ^ % ^ ^1-5.^, Afg-^ ^ ^ ^ . i t ^ ^ S *f 65%

KAERI/RR-1999/99

Table 6.4-1

Characteristics of Reference DUPIC Fuel

Fissile content

(wt%)

Fuel composition

(%)

23>U239Pu24!Pu

Spent PWR fuel

SEU (3.5 wt%)

DU (0.25 wt%)

NU (0.71 wt%)

Spent PWR fuel utilization (%)

Discharge.burnup in CANDU (MWd/T)

Annual Fuel requirement (MTU)

DUPIC Fuel Model

Option 1

1.00

0.45

0.04

82.7

6.5

10.8

0.0

96

14900

47.6

Option 2

0.97

0.53

0.05

96.6

2.3

1.1

0.0

100

14500

48.9

Option 3

0.98

0.54

0.05

100.0

0.0

0.0

0.0

80

15400

46.0

- 712 -

KAERI/RR-1999/99

Table 6.4-2

Input Values for Fuel Cycle Components

Component

Uranium ($/lbU3O8)

- PWR

- CANDUConversion ($/kgU)

- PWR

- CANDU

Enrichment ($/SWU)

Modification of CANDU

Reactor for DUPIC ($/kgHM)

Fabrication ($/kgHM)

- PWR

- CANDU

- DUPIC


- DUPICTransportation/Storage ($/kgHM)

- PWR

- CANDU

- DUPICDisposal ($/kgHM)

- PWR

- CANDU

- DUPIC

Loss Rate

(%)

0.5

0.5

1

1

1

Lead/Lag

(months)

-24

-17

-18

-13

-12

-12

-6

-10

-10

120

120

120

120

360

360

360

Unit Cost

19.2

19.2

8

8

110

Option 1: 4.183

Option 2: 4.007

Option 3: 4.287

275

65

Option 1: 619

Option 2: 580

Option 3: 558

50

230

48

170

403

118

220

- 713 -

KAERI/RR-1999/99

Table 6.4-3

Distribution Parameters of Input Values for Uncertainty analysis

Component

Uranium ($/lbU3O8)

- PWR

- CANDU

Conversion ($/kgU)

- PWR

- CANDU

Enrichment ($/SWU)

Modification of CANDU

Reactor for DUPIC

($/kgHM)

Fabrication ($/kgHM)

- PWR

- CANDU

- DUPIC


- DUPIC

Transportation/Storage ($/kgHM)

- PWR

- CANDU

- DUPIC

Disposal ($/kgHM)

- PWR

- CANDU

- DUPIC

Distribution

Triangular

Triangular

Triangular

Triangular

Triangular

Triangular

Triangular

Triangular

Minimum

15

15

6

6

80

Option 1: 3.76

Option 2: 3.61

Option 3: 3.85

200

47

Option 1: 450

Option 2: 422

Option 3: 406

13

60

13

44

92

27

50

Most likely

value

(Mode)

19.2

19.2

8

8

110

4.183

4.007

4.287

275

65

619

580

406

50

230

48

170

403

118

220

Maximum

35

35

11

11

130

7.11

6.81

7.27

350

83

788

738

710

63

290

61

214

443

130

242

- 714 -

KAERI/RR-1999/99

Table 6.4-4

Characteristics of Reference Reactors and Fuels for Once-through and DUPIC Fuel Cycles

Item

Reactor

- Electric power (MWe)

- Thermal power (MWth)

- Specific power (MW/tonU)

- Load factor

- Cycle length (Full Power Day)

- No. of fuel assemblies or bundles per core

- No. of batches for PWR

- Loading per core (MTU)

- Annual fuel requirement (MTU)

Fuel

- Initial enrichment

- No. of fuel rods per assembly

- Discharge burnup (MWd/kgHM)

- Reference cooling time for refabrication of

spent PWR fuel into DUPIC fuel (year)

Characteristic Value

PWR

950

2,775

40.2

0.8

290

157

3

69.1

23.15

3.5%

264

35

10

CANDU

713

2,159

25.5

0.9

4,560

80.3

94.5

Nat. U

37

7.5

CANDU

-DUPIC

713

2,159

25.5

0.9

4,560

80.3

Table 6.4-1

Table 6.4-1

43

Table 6.4-1

- 715 -

KAERI/RR-1999/99

Table 6.4-5

Material Flow of Once-through Fuel Cycle based on One-Batch Equilibrium Model

Uranium purchase (lb U3O8)

Conversion (MTU)

Enrichment (TSWU)

Fabrication (MTU)

Reactor condition

Electricity power (MWe)

Thermal efficiency (%)

Specific power (MW/MTU)

Burnup (MWd/MTU)

Transportation (MTHM)

Interim storage (MTHM)

Disposal (MTHM)

PWR

428461

168.80

111.90

23.26

950

34.23

40.2

35000

23.03

23.03

23.03

CANDU

223748

86.06

85.63

713

33.03

25.5

7900

84.79

84.79

84.79

MTU: Metric Ton Uranium

TSWU: Ton Separative Work Unit

MTHM: Metric Ton Heavy Metal

- 716 -

KAERI/RR-1999/99

Table 6.4-6

Material Flow of Once-through Fuel Cycle based on One CANDU Reactor

Uranium purchase (lb UjOg)

Conversion (MTU)

Enrichment (TSWU)

Fabrication (MTU)

PWR core (MTU)


Fabrication (MTHM)

Spent PWR fuel (%)

SEU (%)

DU (%)

Natural U (%)Transportation (MTHM)CANDU reactor

Electric power (MWe)

Thermal efficiency (%)

Specific power (MW/MTU)

Burnup (MWd/MTU)


Interim storage (MTHM)

Disposal (MTHM)

Equilibrium core ratio

DUPIC Fuel Option

Option 1

1,317,551

560.78

344.11

71.53

70.82

70.82

85.63

82.7

6.5

10.8

0.084.79

713

33.03

25.5

14900

84.79

84.79

84.79

1.688

Option 2

1,539,002

591.96

401.95

83.55

82.27

82.2785.63

96.6

2.3

1.1

0.084.79

713

33.03

25.5

14500

84.79

84.79

84.79

2.037

Option 3

1,593,170

612.79

416.10

86.49

85.63

85.63

85.63

100

84.79

713

33.03

25.5

15400

84.79

84.79

84.79

1.989

MTU: Metric Ton Uranium

TSWU: Ton Separative Work Unit

MTHM: Metric Ton Heavy Metal

SF: Spent PWR Fuel

NU: Natural Uranium

SEU: Slight Enriched Uranium (3.5 wt%)

DU: Depleted Uranium

Equilibrium core ratio = CANDU annual requirement/ PWR annual Requirement

- 717 -

KAERI/RR-1999/99

Table 6.4-7

Levelized Costs (mills/kWh) of Once-through and DUPIC Fuel Cycle for Option 1

(Deterministic Method)

Components

P

W

R

C

A

N

D

U

Uranium (U3O8)

Conversion

Enrichment

Fabrication

Transportation

Trans. & Storage

Disposal

Uranium (U3O8)

Conversion

Fabrication

Transportation

Plant Modification*

Transportation & Storage

Disposal

Total

Once-through

PWR

1.170

0.183

1.667

0.846

-

0.363

0.239

CANDU

-

-

-

-

-

-

-

0.191

0.030

0.241

-

-

0.162

0.092

5.185

DUPIC

1.170

0.183

1.667

0.846

0.079

-

-

-

-

1.133

0.076

0.016

0.173

0.085

5.427

*Fuel handling equipment for DUPIC fuel loading and heat exchanger for pool storage capacity

increase etc. are included in the cost.

- 718 -

KAERI/RR-1999/99

Table 6.4-8



Components

P

WR

C

AN

D

U

Uranium (U3O8)

Conversion

Enrichment

Fabrication

Transportation

Trans. & Storage

Disposal

Uranium (U3O8)

Conversion

FabricationTransportation

Plant Modification


Disposal

Total

Once-through

PWR

1.211

0.189

1.726

0.875-

0.375

0.248

CANDU

-

-

--

-

-

-

0.165

0.026

0.208-

0.140

0.080

5.244

.DUPIC

1.2110.189

1.726

0.875

0.082-

-

-

-

0.941

0.078

0.013

0.154

0.075

5.345

- 719 -

KAERI/RR-1999/99

Table 6.4-9



Components

P

W

R

C

A

N

D

U

Uranium (U3O8)

Conversion

Enrichment

Fabrication

Transportation

Trans. & Storage

Disposal

Uranium (U3O8)

Conversion

Fabrication

Transportation

Plant Modification


Disposal

Total

Once-through

PWR

1.206

0.189

1.719

0.872

-

0.374

0.247

CANDU

-

-

-

-

-

-

-

0.168

0.027

0.212

-

-

0.143

0.081

5.237

DUPIC

1.206

0.189

1.719

0.872

0.081-

-

-

-

0.871

0.0780.014

0.1470.072

5.249

- 720 -

KAERI/RR-1999/99

Table 6.4-10

Summaiy of Levelized Fuel Cycle Costs by Deterministic Method (mills/kWh)

Direct fuel cycle

(Once-through)

DUPIC fuel cycle

DUPIC Fuel Option

Option 1

5.185

5.427

Option 2

5.244

5.345

Option 3

5.24

5.25

- 721 -

KAERI/RR-1999/99

Table 6.4-11

Results of Monte Carlo Simulation for Uncertainty Analysis of Fuel Cycle Cost

(Statistical Parameters and Percentile)

Items

Minimum (mills/kWh)

Maximum (mills/kWh)

Mean (mills/kWh)

Std. Dev. (mills/kWh)

Variance

Skewness

Kurtosis

Mode (mills/kWh)

Percentile

(mills/kWh)

5%

10%

15%

20%

25%

30%

35%

40%

45%

50%

55%

60%

65%

70%

75%

80%

85%

90%

95%

Option 1

DUPIC

4.3198

6.9141

5.5514

0.3467

0.1202

0.1724

2.7371

5.0690

5.0067

5.1119

5.1917

5.2522

5.3014

5.3518

5.4008

5.4460

5.4896

5.5381

5.5786

5.6263

5.6765

5.7306

5.7844

5.8567

5.9290

6.0195

6.1480

Direct

4.2069

6.5050

5.2603

0.3719

0.1383

0.2140

2.6887

4.9032

4.6777

4.7920

4.8722

4.9385

4.9940

5.0474

5.0960

5.1444

5.1897

5.2353

5.2844

5.3362

5.3909

5.4493

5.5138

5.5881

5.6728

5.7658

5.9018

Option 2

DUPIC

4.3561

6.7810

5.4804

0.3535

0.1250

0.2099

2.7292

4.8000

4.9327

5.0362

5.1051

5.1672

5.2253

5.2757

5.3244

5.3673

5.4160

5.4638

5.5139

5.5598

5.6095

5.6625

5.7165

5.7847

5.8624

5.9581

6.0954

Direct

4.1628

6.5760

5.3235

0.3780

0.1429

0.2276

2.6632

4.8681

4.7344

4.8485

4.9268

4.9921

5.0491

5.1014

5.1512

5.2017

5.2526

5.3004

5.3508

5.4024

5.4583

5.5167

5.5780

5.6495

5.7422

5.8427

5.9793

Option 3

DUPIG

4.2203

6.6311

5.3882

0.3446

0.1187

0.2076

2.7164

4.9146

4.8578

4.9558

5.0243

5.0848

5.1404

5.1904

5.2342

5.2790

5.3245

5.3724

5.4176

5.4613

5.5082

5.5615

5.6229

5.6841

5.7624

5.8586

5.9842

Direct

4.2287

6.5708

5.3042

0.3737

0.1396

0.2508

2.6791

4.5244

4.7271

4.8390

4.9138

4.9768

5.0330

5.0894

5.1369

5.1826

5.2277

5.2790

5.3290

5.3782

5.4298

5.4877

5.5564

5.6359

5.7246

5.8227

5.9482

- 722 -

KAERI/RR-1999/99

Table 6.4-12

Summary of Environmental Benefit of DUPIC Fuel Cycle

Natural Uranium Saving

Rate (%)

Disposal Waste (HLW)

Reduction Rate (%)

DUPIC Fuel Option

Option 1

19.7

64.5

Option 2

20.4

65.5

Option 3

22.7

67.2

- 723 -

KAERI/RR-1999/99

references

unit costs datafor components

engineeringevaluation

model selectionfor fuel cycles

selection of referencenuclear power plant

material flow analyses

levelized unit cost

sensitivity anduncertainty analysis

• electrical power• capacity factor• power efficiencj

product enrichmenttails contentloss factor

discount ratelead/lag time

Fig. 6.4-1 Procedure of Cost Analysis of DUPIC Fuel Cycle

- 724 -

KAERI/RR-1999/99

!

15 19.2$/kgU3O8

35

Fig. 6.4-2 Triangular Distribution Function of Natural Uranium

(Minimum = 1 5 , Mode = 19.2 and Maximum = 35 $/kg)

- 725 -

KAERI/RR-1999/99

Uraniu

>

m Ore

(-24)*

f

Conversion

>

Enricl

>

(-18)

f

unent

(-17)

f

Conversion

(-12)

f

PWR FuelFabrication

>(-6)

Uranium Ore

\(-24)

f

Conversion

>

(-18)

f

Enrichment

(-13)>

(

CANDU FuelFabrication

t

(^ PWR J)

SpenA R S

SpenAFR!

>

tFueltorage

(+120)

t

tFuelStorage

(-12)

f

PWR FuelFabrication

(-6)

f

(-10) /• Nf PWR J)

f ;

CCANDU J

(+360)

f

Spent FuelDisposal

t

Spent FuelAR Storage

>(+120)

f

Spent FuelAFR Storage

•

Once-through


(+120)

(+360)t

Spent FuelDisposal

Fuel Cycle

* Lead/Lag Time(months)

DUPIC FuelFabrication

->(" CANDU J )

i


>(+120)

i

Spent FuelAFR Storage

(+360)

i

Spent FuelDisposal

DUPIC Fuel Cycle

Fig. 6.4-3 Components and Time Frame of Once-through and DUPIC Fuel Cycle

- 726 -

KAERI/RR-1999/99

mills/kwh

Lebelized Unit Cost(mills/kwh) for DUPIC option in Option 1

0.14

rn ill s/kwh

Lebelized Unit Cost(mills/kwh) for Once-through option in Option 1

Fig. 6.4-4 Histogram of Fuel Cycle Cost for DUPIC Fuel Option 1

- 727 -

KAERI/RR-1999/99

0.12

tora

ooo*

oc0)3a£

0.10

0.08

0.06

0.04

0.00

0.10

0.08

g 0.06CD33cra>

0.02

mills/kwh


£ tP «? «? A # O s T? ^ t? 6 A ^ <S <? ? § &t . - ^ • fc- t . - !»• v <o- <o- <r>- <r>- < r <v «>• *>• %•• «>• 6 ' e>- fe-

mills/kwh



- 728 -

KAERI/RR-1999/99

0.16

0.14

mills/kwh


mills/kwh



- 729 -

KAERI/RR-1999/99

0.16

0.12

~ o.ioooo

0.08

O0.06

0.02

DUPIC Option

minimum : 4.3198 mills/kwh

maximum : 6.9141 mills/kwh

mean :5.5514 mills/kwh

std deviation : 0.3467 mills/kwh

variance : 0.1202

skewness : 0.1724

kurtosis : 2.7371

mode : 5.069 mills/kwh

Option 1

\\\

^ V

\ \

\ \

\ \

\ \

Once-through Option



mean : 5.2603 mills/kwh

std deviation : 0.3719 mills/kwhvariance : 0.1383

skewness : 0.214O

kurtosis : 2.6887


4 .5 5.5

mills/kwh

7.5 8.5

Fig. 6.4-7 Comparison of Probabilistic Density Function of Fuel Cycle Cost for Option I

- 730 -

KAERI/RR-1999/99

0.14

0.12

ooo

oca2

0.08

0.06

0.04

0.02

0.00

Option 2

DUPIC Optionminimum : 4.3561 mills/kwhmaximum : 6.7810 mills/kwhmean : 5.4804 mills/kwhstd deviation : 0.3535 mills/kwhvariance : 0.1250skewness : 0.2099kurtosis : 2.7292mode : 4.800 mills/kwh

\

•a '

\

\

Once-through Option



mean : 5.3235 mills/kwh

std deviation : 0.3780 mills/kwh

variance : 0.1429

skewness : 0.2276

kurtosis : 2.6632modee : 4.8681 mills/kwh

\ \

3.5 4.5 5.5

mills/kwh

6.5

Fig. 6.4-8 Comparison of Probabilistic Density Function of Fuel Cycle Cost for Option 2

- 731 -

KAERI/RR-1999/99

0.16

0.14

CD

0.02

0.00

CO

oooo

CD

0.

0.

0.

0

12

10

08

06

DUPIC Option



mean : 5.3862 mills/kwhstd deviation : 0.3446 mills/kwh

variance : 0.1187

skewness : 0.2O76

kurtosis : 2.7164mode : 4.9146 mills/kwh

Option 3

i

\

\

\ \

\ \

U

\ \

Once-through Option



mean : 5.3042 mills/kwhstd deviation : 0.3737 mills/kwhvariance ". 0.1396skewness : O.25O8kurtosis : 2.6791


5.S

mills/kwh

6.5

Fig. 6.4-9 Comparison of Probabilistic Density Function of Fuel Cycle Cost for Option 3

- 732 -

KAERI/RR-1999/99

Plant Modification

DUPIC Disposal

PWR Conversion

DUPIC Transport

DUPIC Interim storage

PWR Fab.

DUPIC Fab. Cosl

Enrichment

Uranium(U3O8)

0.00 0.30 0.40 0.S0

Rank Correlation Coefficient

DUPIC Fuel Cycle

CANDU Conversion

CANDU Disposal

0.20 0.30 0.40 0.50 0.60

Rank Correlation Coefticient

Once-through Cycle

0.70 0.80

Fig. 6.4-10 Sensitivity of Fuel Cycle Component for Option 1

- 733 -

KAERI/RR-1999/99

Plant Modification

DUPIC Disposal

PWR Conversion


0.00 0.10 0.20 0.30 0.40 0.50 0.60

Rank. Correlation Coefficient

DUPIC Fuel Cycle

CANDU Conversion

CANDU Disposal

CANDU Fab.

PWR Conversion

CANDU Interim Stor.

PWR Disposal

PWR Interim Storaoe

PWR Fab.

Enrichment

Uranium(U3O8)

0.30 0.40 0.50 0.60

Rank Correlation Coelficient

Once-through Cycle

o.7o o.ao


- 734 -

KAERI/RR-1999/99

Plant Modification

DUPIC Disposal

PWR Conversion


DUPIC Transport

PWR Fab.

DUPIC Fab. Cost

Enrichment

Uranium (U3O8) mmmmmmmmmmmmm0.30 0.40 0.50


DUPIC Fue! Cycle

CANDU Conversion

CANDU Disposal

PWR Conversion

CANDU Fab.

CANDU Interim Stor.

PWR Disposal

PWR interim Storage

PWR Fab.

Enrichment

Uranium (U3O8)

m t

:•

§) !;

i0.10 0.20 0.30 0.40 0.50 0.60


Once-through Cycle

0.70 0.80


- 735 -

KAERI/RR-1999/99

Uranium(U3O8)(728.31 klb)


PWR(39.36 MTU)

CANDU(94.55 MTU)

Disposal(39.36 MTHM)


(Once-Through Fuel Cycle)

Uranium(U3OS)(728.31 klb)

Uranium(U3O8)(57.24 k Ib)

PWR(39.36 MTU)

K SEU:6.5% .DU:10.8%

r

CANDU(47.59 MTU)


Equilibrium core ratio = 1.70Uranium saving rate of DUPIC fuel cycle : 19.7%Disposal reduction rate of DUPIC fuel cycle : 64.5%

(DUPIC Fuel Cycle)

Fig. 6.4-13 Natural Uranium Saving and Spent Fuel Reduction for DUPIC Fuel Option 1

(based on annual requirement of one CANDU reactor)

- 736 -

KAERI/RR-1999/99

Uranium(U3O8)(874.19klb)

Uraniutn(U3O8)(249.51 klb)

PWR(47.24 MTU)

CANDU(94.55 MTU)





Uranium(U3O8) 1 ^SEU(20.81 k lb) | DU:1

PWR(47.24 MTU)

2.3% w. 1 %

r

CANDU(48.90 MTU)


Equilibrium core ratio = 2.04Uranium saving rate of DUPIC fuel cycle : 20.04%Disposal reduction rate of DUPIC fuel cycle : 65.5%

(DUPIC Fuel Cycle)



- 737 -

KAERI/RR-1999/99

Uranium(U3O8)(852.07 k Ib)


W

PWR(46.05 MTU)

CANDU(94.55 MTU)




Uranium(U3O8)(852.07 k tb)

Uranium(U3O8)(Oklb)

^ S E UDU:

PWR(46.05 MTU)

0% w

r

CANDU(46.05 MTU)


Equilibrium core ratio = 1.99Uranium saving rate of DUPIC fuel cycle: 22.7%Disposal reduction rate of DUPIC fuel cycle : 67.2%

(DUPIC Fuel Cycle)



- 738 -

KAERI/RR-1999/99

6.5 ^ J L ^ r t S

1. J.S. LEE et al., "Research and Development Program of KAERI for DUPIC (Direct Use

of Spent PWR Fuel in CANDU Reactors)," Proceedings of International Conference and

Technology Exhibition on Future Nuclear System: Emerging Fuel Cycles and Waste Disposal

Options, GLOBAL'93, Seattle, 1993.

2. MS. YANG et al. "Conceptual Study on the DUPIC Fuel Manufacturing Technology,"

Proceedings of International Conference and Technology Exhibition on Future Nuclear System:

Emerging Fuel Cycles and Waste Disposal Options, GLOBAL'93, Seattle, 1993.

3. H.B. CHOI, B.W. RHEE, H.S. PARK, "Physics Study on Direct Use of Spent PWR Fuel

in CANDU (DUPIC)," Nucl. Sci. Eng.: 126, pp.80-93, 1997.

4. C.J. JEONG and H.B. CHOI, "Compatibility Analysis on Reactivity Devices for Direct Use

of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)," Nucl. Sci. Eng.:

134, pp. 1-16, 2000.

5. The Economics of the Nuclear Fuel Cycle, Organization for Economic Cooperation and

Development/Nuclear Energy Agency, 1993.

6. H.S. PARK, "A Study on the Direct Use of Spent PWR Fuel in CANDU (Phase 1 Feasibility

Study)," KAERI/RR-1244/92, Korea Atomic Energy Research Institute, pp.151-181, 1993.

7. T.R. THOMAS, "AIROX Nuclear Recycling and Waste Management," Proceedings of

International Conference and Technology Exhibition on Future Nuclear System: Emerging

Fuel Cycles and Waste Disposal Options, GLOBAL'93, Seattle, 1993.

8. Y.G. LEE, H.R. CHA, J.S. HONG, H.O. MENLOVE and P.M. RINARD, "Development

of DUPIC Safeguards Neutron Counter," Proceedings of the Institute of Nuclear Materials

Management, 38th Annual Meeting, Pheonix, 1997.

- 739 -

KAERI/RR-1999/99

9. J.T. THOMAS, "Nuclear Safety Guide, TID-7016 (Rev. 2)," NUREG/CR-0095, Oak Ridge

National Laboratory, 1978.

10. H.B. CHOI, J.W. CHOI, M.S. YANG, "Composition Adjustment on Direct Use of Spent

Pressurized Water Reactor Fuel in CANDU," Nucl. Sci. Eng.: 131, pp.62-77, 1999.

11. R.K. NAKAGAWA, " New DUPIC Fuel Loading System Feasibility Study," DUPIC-

AC-RT-01, Atomic Energy of Canada Limited, May 1992.

12. R. BOYD and D.J. KOIVISTO, "An Assessment of Fuel Handling Systems for DUPIC

Fuel in CANDU 6," DUPIC-AE-030, Atomic Energy of Canada Limited, August 1998.

13. AECL document, "Design Manual - Fuelling Machine D2O Control System," 86-35230-

DM-001, Rev.l, Atomic Energy of Canada Limited, 1999.

14. AECL document, "Design Requirement - Spent Fuel Bay Cooling and Purification System,"

86-34410-DR-001, Rev.2, Atomic Energy of Canada Limited, 1997

15. J.Y LEE, A. BALDOR, "Design Manual: Spent Fuel Storage Wolsong NPP 234," 86-35360-

DM-001, Rev.l, Atomic Energy of Canada Limited, 1996.

16. G.R. SIMMONS and P. BAUMGARTNER, "The Disposal of Canadas Nuclear Fuel Waste:

Engineering for a Disposal Facility," AECL-10715, Atomic Energy of Canada Limited, 1994.

17. C.S. KANG et al., "Survey and Analysis of the Domestic Technology Level for the Concept

Development of High Level Waste Disposal," KAERI/CM-231/98, Korea Atomic Energy

Research Institute, 1998.

18. The Cost of High Level Waste Disposal in Geological Repositories An Analysis of Factors

Affecting Cost Estimate, Organization for Economic Cooperation and Development/Nuclear

Energy Agency, 1993.

- 740 -

KAERI/RR-1999/99

19. R.J. ELLIS, "Sensitivity of the DUPIC CANDU-Stage Burnup to PWR Model

Representations," DUPIC-AE-004, Atomic Energy of Canada Limited, 1996.

20. Y. ATES and P. BAUMGARTNER, "Preliminary Disposal Cost Comparison of Spent PWR,

CANDU-NU and CANDU-DUPIC Fuels," RC-1879, DUPIC-AE-025, Atomic Energy of

Canada Limited, 1997.

21. The Fourth Long-term Supply and Demand Plan of Nuclear Energy (1998-2015), MOCIE

Notice No. 1998-93, Korean Ministry of Commerce, Industry and Energy, 1998.

22. Y.K. YOON, "A Study on the Formulation of Long-term Nuclear Energy Policy Direction

for Korea," Korean Nuclear Society, Korea Ministry of Science and Technology, 1994.

23. S.H. CHANG, "A Study on the Establishment of Comprehensive Promotion Plan for

Utilization, Research and Development of Nuclear Energy," Korean Nuclear Society, Korea

Ministry of Science and Technology, 1994.

24. "Radioactive Waste Management Plan in Korea," the 249th AEC, Korean Atomic Energy

Commission, 1998.

25. K.Y. LEE, Y.D. WHANG, J. LEE and Y.I. KIM, "Establishment of the Phase II NSSS

Development Plan for the Next Generation Reactor," KAERI/TR-592/95, Korea Atomic

Energy Research Institute, 1995.

26. K.D. SONG, M.K. LEE, K.H. MOON, S.S. KIM and B.R. LEE, "Economic Evaluation

of Nuclear Fuel Cycle in Korea," KAERI/RR-1687/96, Korea Atomic Energy Research

Institute, 1996.

27. H. CHOI, W.I. KO, and M.S. YANG, "Reactivity Control Method for Direct Use of Spent

Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)," Nitcl. Sci. Eng., June, 2000.

28. W.I. KO, J.W. CHOI, H.B. CHOI, J.S. LEE, and J.H. WHANG, "Development of the

- 741 -

KAERI/RR-1999/99

Combination Method for Minimizing Composition Variability of DUPIC Fuel Feedstock,"

Proceedings of Korean Nuclear Society Autumn Meeting, Seoul (1998).

29. J.W. CHOI, W.I. KO, J.S. LEE, M.S. YANG, and H.S. PARK, "Cost Assessment of a

Commercial Scale DUPIC Fuel Fabrication," Proceedings of the 6th International Conference

on Radioactive Waste Management and Environmental Remediation, Singapore (1997).

30. B.G. NA and I. NAMGUNG, "Assessment of Wolsong NPP Fuel Handling System for DUPIC

Fuel," KAERI/CM-347/99, Korea Atomic Energy Research Institute (2000).

31. H.S. PARK, S.W. PARK and W.I. KO, "The Construction of an Interim Spent Fuel Storage

Facility," KAERI-NEMAC/PR-35/94, Korea Atomic Energy Research Institute (1994).

32. W.I. KO, H. CHOI, and M.S. YANG, "Cost Evaluation for Disposal of Spent DUPIC Fuel,"

KAERI/TR-1440/99, Korea Atomic Energy Research Institute (1999).

33. W.I. KO, J.W. CHOI, C.H. KANG, J.S. LEE, and K.J. LEE, "Nuclear Fuel Cycle Cost

Analysis Using a Probabilistic Simulation Technique," Annals of Nuclear Energy, Vol. 25,

pp.771-789 (1998).

34. W.I. KO, J.W. CHOI, J.S. LEE, H.S. PARK and K.J. LEE, "Uncertainty Analysis in DUPIC

Fuel Cycle Cost Using A Probabilistic Simulation Method," Nuclear Technology, Vol. 127,

pp. 123-140 (1999).

35. D. VOSE, Quantitative Risk Analysis: A Guide Monte Carlo Simulation Modeling, pp.165-168,

John Wiley and Sons, Inc., New York (1996).

36. M. EVANS, N. HASTINGS and B. PEACOCK, Statistical Distribution (2nd Edition),

pp.31-37, John Wiley and Sons, Inc., New York (1993).

- 742 -

KAERI/RR-1999/99

- 743 -

KAERI/RR-1999/99

fe DUPIC «j<££.;M 314s. ^ ^ ^ ^ 91 DUPIC *|«ig.sj ;§

DUPIC 3 #1

DUPIC

. DUPIC

DUPIC } ^

>. DUPIC

DUPIC

7H

DUPIC

fe DUPIC

- 745 -

KAERI/RR-1999/99

7.1

7.1.1

AI]J=L <^^2. ^ . ^

DUPIC ^ ^ 1 5 . TI^I-S-

DUPIC ^ < g S ^^11^

• DUPIC

. M ( 4 ^

* ' > T " S ' B |

• DUPIC

• DUPIC

• SV-S-S:

•ROP *

7 ] ^ . ^ . ^ 7

7|.g. i->y <:

71^- (ZCU,

H 71^ ^

- c* yfi - f| 5

^ 7fl^ ^ 7 j

«1 ^«1 711^

^ 7fl^ ^ ^

- W 7B «

3 ^ (MCNP gj ^ ^ Jbc#5l - y ^ )

• (AUTOREFUEL)

. £ S . Z ^ 7H^ (WIMTAB)

ADJ, MCA, SOR) ^ ^ s§7l-

HI 1 7i |^ ^7^1 ( ^ ^-^. Jf-S)

l^t (CLUSTER)

t (GENOVA)

1 91 ul-§-^7l- (558$/kg)

! i i ^^"Wl-§- 3§7V (220$/kg)

- 746 -

KAERI/RR-1999/99

7.1.2 fl S ^ ^ ^ 1

7.1.2.1 7)^- i ^ *4X] ^ S ] ^-g-

DUPIC

^ DUPIC

WIMS

7.1.2.2 DUPIC

7.1.2.3

^r DUPIC

DUPIC ii^ofl-HS. ^>b*W ^°1- DUPIC

3 . Sa^l-. ZL5ii4 DUPIC Jn^6U^

Bt> Sfl^ ly-oVo] ^.^-sjt:!-. DUPIC

^ , DUPIC

- 747 -

o} ^ ^

37)7}

KAERI/RR-1999/99

7.12A

DUPIC

DUPIC

7.1.2.5 DUPIC

OECD/NEA

DUPIC

DUPIC

fe DUPIC

DUPIC

3.B.S.

- 748 -

KAERI/RR-1999/99

7.1.3

7.1.3.1

i*rvi£

(1997)

3

7

5

(1998)

8

5

**\

2

8

(1999)

8

2

7

10 19

14

**\

9

23

7.1.3.2 SCI

1. D.H. Kim, J.K. Kim, and H.B. Choi, "A Generalized Perturbation Theory Program for CANDU

Core Analysis", Annals of Nuclear Energy, June 2000.

2. C.J. Jeong and H.B. Choi, "Instability of Xenon Spatial Oscillation in a CANDU-6 Reactor

with DUPIC Fuel", Annals of Nuclear Energy, May 2000.

3. H.B. Choi, W.I. Ko and M.S. Yang, "Reactivity Control Method for Direct Use of Pressurized

Water Reactor Fuel in CANDU Reactors (DUPIC)", Nucl. Sci. Eng., June 2000.

4. Hyung-Seok Lee, Won Sik Yang, Man Gyun Na and Hangbok Choi, A Pin Power

Reconstruction Method for CANDU Reactor Cores Based on Coarse-Mesh Finite Difference

Calculations, Nuclear Technology: 130, pp. 1-8, April 2000.

5. C.J. Jeong and H.B. Choi, "Compatibility Analysis on Existing Reactivity Devices in CANDU

6 Reactors for DUPIC Fuel Cycle", Nucl. Sci. Eng.: 134, pp.265-280, March 2000.

6. Chang Joon Jeong and Hangbok Choi, "Xenon transient analysis for direct use of spent

pressurized water reactor fuel in CANDU reactors (DUPIC)", Annals of Nuclear Energy, Vol.27,

pp.269-278, Dec. 1999.

7. Hangbok Choi, "A fast-running fuel management program for a CANDU reactor", Annals

of Nuclear Energy, Vol.27, pp.1-10, Sept. 1999.

- 749 -

KAERI/RR-1999/99

8. H.B. Choi, J. W. Choi, M.S. Yang, "Composition Adjustment on Direct Use of Spent Pressurized

Water Reactor Fuel in CANDU", Nitcl. Sci. Eng.: 131, pp.62-77, Jan. 1999.

9. Jee-Won Park, "A subchannel analysis of DUPIC fuel bundles for the CANDU reactor", Annals

of Nuclear Energy, Vol. 26, pp.29-46, Jan. 1999.

7.1.3.3

1. Dongwhan Park, Hangbok Choi and Changjoon Jeong, "Assessment of CANDU Physics

Analysis Tools Using Measurement Data of Wolsong Nuclear Power Plant 2, The 2000 ANS

International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation

into the Next Millennium, PHYSOR2000, Pittsburgh, May 7-11, 2000.

2. D.H. Kim, J.K. Kim, and H.B. Choi, "A Sensitivity Method for CANDU Core Analysis,

The 2000 ANS International Topical Meeting on Advances in Reactor Physics and Mathematics

and Computation into the Next Millennium, PHYSOR2000, Pittsburgh, May 7-11, 2000.

3. Hangbok Choi and Jee-Won Park, "Power Coefficient Calculation of a CANDU Reactor",

Sixth International Conference on CANDU Fuel, Niagara Falls, Sept. 26-30, 1999.

4. C.J. Jeong and H.B. Choi, "Analysis of Xenon Spatial Oscillation in a CANDU-6 Reactor

with DUPIC Fuel", Sixth International Conference on CANDU Fuel, Niagara Falls, Sept. 26-30,

1999.

5. Gyuhong Roh and Hangbok Choi, "Benchmark Calculations for CANDU Fuel Bundles", Proc.

Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear

Applications, Madrid, Sept. 26-31, 1999.

6. Hangbok Choi and Jee-Won Park, "Comparison of Reactivity Feedback due to Power Level

Change of a CANDU Reactor", Proc. Mathematics and Computation, Reactor Physics and

Environmental Analysis in Nuclear Applications, Madrid, Sept. 26-31, 1999.

7. Hangbok Choi, Won Ik Ko, Myung S. Yang, "Comparison of DUPIC Fuel Composition

Heterogeneity Control Methods", Proc. International Conference on Future Nuclear System:

GLOBALr99, Jackson Hole, Aug. 29 - Sept. 3, 1999.

8. Chang-Joon Jeong and Hangbok Choi, "Xenon Characteristics of Spent PWR Fuel in a

CANDU-6 Reactor", Tran. Am. Nucl. Soc, Vol.80, Boston, June 6-10, 1999.

9. C.J. Jeong and H.B. Choi, "Assessment of Reactivity Devices for a CANDU 6 Reactor with

spent PWR Fuel", Tran. Am Nucl. Soc, Vol.80, Boston, June 6-10, 1999.

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10. J.-W. Park, K.M. Chae, and H. Choi, "ASSERT-PV Simulations of Two-Phase Flow in

Horizontal and Vertical Subchannels", 20th Annual Conference of the Canadian Nuclear Society,

Montreal, May 30 - June 2, 1999.

11. Hangbok Choi, Jongwon Choi, Won II Ko, Jae Sol Lee, "Preliminary Study on Sensitivity

of DUPIC Fuel Cycle Cost to Fuel Composition", Int. Conf. on the Physics of Nuclear Science

and Technology, Long Island, New York, Oct. 5-8, 1998.

12. G.H. Roh and H.B. Choi, "Assessment of DUPIC Physics Calculation by MCNP", Int. Conf.

on the Physics of Nuclear Science and Technology, Long Island, New York, Oct. 5-8, 1998.

13. Jee-Won Park, Hangbok Choi, "The Effect of Channel Flow Reduction on Thermal

Performance of DUPIC Fuel Bundle Strings in the CANDU Reactor", Int. Conf. on the Physics

of Nuclear Science and Technology, Long Island, New York, Oct. 5-8, 1998.

14. Gyuhong Roh and Hangbok Choi, "One-Dimensional Analysis of CANDU Primary Shield

for DUPIC Fuel", Int. Conf. on the Physics of Nuclear Science and Technology, Long Island,

New York, Oct. 5-8, 1998.

15. Do Heon Kim, Jong Kyung Kim, Hangbok Choi, Gyuhong Roh, Won Sik Yang, "Development

of a Generalized Perturbation Program for a CANDU Reactor", Int. Conf. on the Physics of

Nuclear Science and Technology, Long Island, New York, Oct. 5-8, 1998.

16. Jee-Won Park and Hangbok Choi, "An Analysis of Void Fraction Propagation in Two-Phase

Flow Through Multiple Flow Channels", Proc. of 33rd Intersociety Energy Conversion Eng.

Conf., IECEC98, Colorado Springs, Colorado, Aug. 2-6, 1998.

17. Hangbok Choi and Jee-Won Park, "Nuclear Characteristics of Spent Pressurized Water Reactor

Fuel in a Heavy Water Reactor", Proc. of the 6th Int. Conf. on Nucl. Eng., ICONE-6, San

Diego, California, May 10-14, 1998.

18. Jee-Won Park and Hangbok Choi, "Thermal Behaviors of Proposed DUPIC Fuel Bundle

in the CANDU Reactor", Proc. of the 6th Int. Conf. on Nucl. Eng., ICONE-6, San Diego,

California, May 10-14, 1998.

19. Pritam D. Krishnani, Hangbok Choi, Chang J. Jeong, "Method of Calculating the Effect

of Heterogeneous Fuel Composition on a Fuel Bundle Power Ramp", Tran. Am. Nucl. Soc,

Vol.77, Albuquerque, Nov. 16-20, 1997.

20. H.B. Choi, G.H. Roh, J.W. Park, "CANDU Core Analysis with Spent PWR Fuel of Fixed

235U and 239Pu Content", Tran Am. Nucl Soc, Vol.77, Albuquerque, Nov. 16-20, 1997.

21. Hangbok Choi, Gyu H. Roh, Ho H. Lee, "Sensitivity Study on DUPIC Fuel Composition",

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Int. Conf. on Future Nuclear Systems: GLOBAL'97, Yokohama, Japan, Oct. 5-10, 1997.

22. Hangbok Choi, Gyu H. Roh, Chang J. Jeong, Bo W. Rhee, Jong W. Choi, C.R. Boss,

"Preliminary Assessment on Compatibility of DUPIC Fuel with CANDU-6", Int. Conf. on Future

Nuclear Systems: GLOBAL'97, Yokohama, Japan, Oct. 5-10, 1997.

23. Jee-Won Park and Hangbok Choi, "Assessment of Average Thermal-Hydraulic Governing

Equations Used in PWR/PHWR System Design and Safety Analysis", Eighth Int. Topical Meeting

on Nuclear Reactor Thermal-Hydraulics: NURETH-8, Kyoto, Japan, Sept. 30 - Oct. 4, 1997.

7.1.3.4

1. Dongwhan Park and Hangbok Choi, Assessment of WIMS-AECL using Physics Measurement

of Wolsong-2 Reactor, Proc. of the Korean Nuclear Society Autumn Meeting, Seoul, Korea,

Oct. 29-30, 1999.

2. Do Heon Kim, Jong Kyung Kim and Hangbok Choi, GPT Estimation of Unconstrained

Sensitivity Coefficients for CANDU Core Analysis, Proc. of the Korean Nuclear Society Autumn

Meeting, Seoul, Korea, Oct. 29-30, 1999.

3. Hyug-Seok Lee, Won Sik Yang, Man Gyun Na, and Hangbok Choi, A Pin Power

Reconstruction Method for CANDU Reactor Cores, Proc. of the Korean Nuclear Society Autumn

Meeting, Seoul, Korea, Oct. 29-30, 1999.

4. G.H. Roh and H.B. Choi, "Reactivity Coefficient Calculation of CANDU Fuel Lattices by

MCNP", Proc. of the Korean Nuclear Society Spring Meeting, Pohang, Korea, May 28-29, 1999.

5. H.B. Choi and W.I. Ko, "Reactivity Control Option for DUPIC Fuel by Natural Uranium",

Proc. of the Korean Nuclear Society Spring Meeting, Pohang, Korea, May 28-29, 1999.

6. Hangbok Choi, "Power Coefficient Calculation of a CANDU Reactor", Proc. of the Korean

Nuclear Society Spring Meeting, Pohang, Korea, May 28-29, 1999.

7. Chang-Joon Jeong and Hangbok Choi, "Damping Analysis of Xenon Oscillation in CANDU-6

Reactor with DUPIC Fuel", Proc. of the Korean Nuclear Society Spring Meeting, Pohang, Korea,

May 28-29, 1999.

8. Chang-Joon Jeong, Jee-Won Park and J. Pitre, "Preliminary ROP Assessment for CANDU-6

with DUPIC Fuel", Proc. of the Korean Nuclear Society Spring Meeting, Pohang, Korea, May

28-29, 1999.

9. C.J. Jeong and Hangbok Choi, "Xenon Load Analysis for CANDU-6 with DUPIC Fuel",

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Proc. of the Korean Nuclear Society Autumn Meeting, Seoul, Korea, Oct. 30-31 1998.

10. C.J. Jeong and Hangbok Choi, "Assessment of Reactivity Devices for CANDU-6 with DUPIC

Fuel", Proc. of the Korean Nuclear Society Autumn Meeting, Seoul, Korea, Oct. 30-31 1998.

11. Do H. Kim, Jong K. Kim, Hangbok Choi, Gyuhong Roh, Won Sik Yang, "Estimation of

CANDU Reactor Zone Controller Level by Generalized Perturbation Method", Proc. of the Korean

Nuclear Society Autumn Meeting, Seoul, Korea, Oct. 30-31 1998.

12. Do H. Kim, Jong K. Kim, Hangbok Choi, Gyuhong Roh, Won Sik Yang, "A Generalized

Perturbation Program for CANDU Reactor", Proc. of the Korean Nuclear Society Spring Meeting,

Suwon, Korea, May 29, 1998.

13. Hangbok Choi, Gyuhong Roh, "A Sensitivity Study on Neutronics Performances of DUPIC

Fuel", Proc. of the Korean Nuclear Society Spring Meeting, Suwon, Korea, May 29, 1998.

14. Jee-Won Park, Hangbok Choi, Bo W. Rhee, "Enthalpy and Void Distributions in Subchannels

of PHWR Fuel Bundles", Proc. of the Korean Nuclear Society Spring Meeting, Suwon, Korea,

May 29, 1998.

15. Gyuhong Roh, Hangbok Choi, "Benchmark Calculation of CANDU End Shielding System",

Proc. of the Korean Nuclear Society Spring Meeting, Suwon, Korea, May 29, 1998.

16. Hangbok Choi, "MCNP Basic Features", * H ^ * H M 1 * 1 - S l 1998*d *\] 165] ^7fl«]-

^ U S ] , tfl*l, April 1998.

17. Jee-Won Park, Gyu-hong Roh, Hangbok Choi, "Inconsistency in the Average Hydraulic

Models Used in Nuclear Reactor Design and Safety Analysis", Proc. of the Korean Nuclear


18. Gyuhong Roh, Hangbok Choi, Jee-Won Park, "Sensitivity Analysis on Various Parameters

for Lattice Analysis of DUPIC Fuel with WIMS-AECL Code", Proc. of the Korean Nuclear


19. Do H. Kim, Jong K. Kim, Hangbok Choi, Gyuhong Roh, In H. Jeong, "Fission Product

Inventory Calculation with CASMO-ORIGEN-Coupled Program", Proc. of the Korean Nuclear

Society Autumn Meeting, Taegu, Korea, Oct 1997.

7.1.3.5

1. Won II Ko, Hangbok Choi and Myung Seung Yang, Cost Evaluation for Disposal of Spent

DUPIC Fuel, KAERI/TR-1440/99, Dec. 1999.

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2. Gyuhong Roh and Hangbok Choi, "Assessment of Neutron Transport Codes for Application

to CANDU Fuel Lattice Calculation", KAERI/TR-1377/99, August 1999.

3. Won HKo, Hangbok Choi and Myung S. Yang, "Cost Evaluation of a Commercial-Scale

DUPIC Fuel Fabrication Facility (Part III) - Appendix", KAERI/TR-1374/99, August 1999.

4. Won II Ko, Hangbok Choi and Myung S. Yang, "Cost Evaluation of a Commercial-Scale

DUPIC Fuel Fabrication Facility (Part II) - Preliminary Conceptual Design", KAERI/TR-1373/99,

August 1999.

5. Won II Ko, Hangbok Choi and Myung S. Yang, "Cost Evaluation of a Commercial-Scale

DUPIC Fuel Fabrication Facility (Part I) - Summary", KAERI/TR-1372/99, August 1999.

6. Hangbok Choi, "Composition Heterogeneity Analysis for DUPIC Fuel (I) - Statistical Analysis",

KAERI/TR-1371/99, August 1999.

7. Hangbok Choi and Won II Ko, "Comparison of DUPIC Fuel Composition Heterogeneity

Control Method", KAERI/TR-1370/99, August 1999.

8. Hangbok Choi, Gyuhong Ron, and Do H. Kim, "DUPIC Fuel Cycle Economics Assessment

(II)", KAERI/AR-537/99, April 1999.

9. Hangbok Choi, Gyuhong Ron, and Do H. Kim, "DUPIC Fuel Cycle Economics Assessment

(I)", KAERI/AR-535/99, April 1999.

10. Hangbok Choi, "Power Coefficient Calculation of a CANDU Reactor", KAERI/TR-1246/99,

March 1999.

11. G.H. Roh and H.B. Choi, "MCNP Tutorials and Samples", KAERI/TR-1219/99, Feb. 1999.

12. Chang-Joon Jeong and Hangbok Choi, "Assessment of CANDU-6 Reactivity Devices for

DUPIC Fuel", KAERI/TR-1160/98, November 1998.

13. Chang-Joon Jeong and Hangbok Choi, "Xenon Load Analysis for CANDU 6 with DUPIC

Fuel", KAERI/TR-1132/98, September 1998.

14. Jee-Won Park and Hangbok Choi, "A Development of Two-Fluid Multifield Model for Low

Quality Boiling Transition Simulations", KAERI/TR-1127/98, September 1998.

15. Hangbok Choi, "A Method to Calculate the Effect of Heterogeneous Composition on Bundle

Power", KAERLTR-1126/98, September 1998.

16. Hangbok Choi, "Nuclear Data Uncertainty Analysis on a Minor Actinide Burner for

Transmuting Spent Fuel", KAERI/TR-1112/98, August 1998.

17. Hangbok Choi, "Neutronics Design Study on a Minor Actinide Burner for Transmuting Spent

Fuel", KAERI/TR-1111/98, August 1998.

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18. Gyuhong Roh and Hangbok Choi, "Assessment of CANDU Primary Shield System for DUPIC

Fuel", KAERI/TR-1056/98, May 1998.

19. Hangbok Choi, Gyuhong Roh, "A Sensitivity Study on DUPIC Fuel Composition",

KAERI/TR-942/97, Jan. 1998.

20. Jee-Won Park, "An Assessment of Thermal Behavior of the DUPIC Fuel Bundle by

Subchannel Analysis", KAERI/TR-938/97, Dec. 1997.

21. Jee-Won Park, Gyuhong Roh, "A Review of Critical Heat Flux Prediction Technique and

Its Application in CANDU Reactor", KAERI/AR-473/97, Sept. 1997.

7.1.3.6 HS-2.

1. ^S - t i , £ } % ^ , CALCON, 99-01-12-6348,

1999.12.31.

2. ^S- tL ^ % ^ , GENOVA, 99-01-12-6349,

1999.12.31.

3. a ) % ^ , WIMTAB, 98-01-12-4011, tH^^SS.ZL 'gJ iL^ lS ' ] , 1998.8.1.

4. 3 ^ - 4 , NINEM, 98-01-12-4010, S H ^ ^ B J H S U ' S J i ^ , 1998.8.1.

5. ^ 1 % ^ , CONEX, 98-01-12-4009, *H-£-*fBlHSll3i.&3LSl, 1998.8.1.

6. ^1 % ^ , VOPT, 98-01-12-4008, *>^ |#BiHS^H3.eJLS| , 1998.8.1.

7. Jfc^gP, ^ % ^ , ANSGEN, 98-01-12-4007, S H ^ ^ S S H ' g J i S L S I , 1998.8.1.

8. i ] % ^ , FIXCOM, 97-01-12-6355, ^^ -Sf lHHSH^JLSLi l , 1997.12.24.

9. i | ^ 4 , WIMPRO, 97-01-12-6354, t>^-^-^-B]HSll^AaLSI, 1997.12.24.

10. 5 ] % ^ , Jn-3^F, WIMKIN, 97-01-12-5226,

1997.11.7.

11. ^ % ^ , ii^rHF, CLUSTER, 97-01-12-5225,

1997.11.7.

12. 5 ] % ^ , AUTOREFUEL, 97-01-12-4685, ^ ^ - ^ ^ ^ = ^ 1 1 ^ ^ 1 3 : 5 1 , 1997.10.16.

13. o l^^- , ^ ^ S - , Q*^, RSTORE, 97-01-12-4684, * > ^ - ^ ^ B l S S . - L ^ J E L ^ ,

1997.10.16.

14. 3 1 ^ 4 , ic^~^, COCUP, 97-01-12-4683, SH*g-^s tS -L 'g . e jL3 | , 1997.10.16.

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- 759 -

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- 760 -

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- 762 -

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BIBLIOGRAPHIC INFORMATION SHEETPerforming Org.

Report No.Sponsoring Org.

Report No.Standard Report No. INIS Subject Code

KAERI/RR-1999/99

Title/Subtitle | A Study on Direct Use of Spent PWR Fuel in CANDU Reactors

- DUPIC Fuel Compatibility Assessment

Project Managerand Department

Hangbok Choi (Nuclear Fuel Design Technology Development Team)

Researcher andDepartment

G.H. Rho, J.W. Park, C.J. Jeong, B.W. Rhee, S.S. Kim, W.I. Ko,W.K. Kim, J.W. Choi, J.S. Lee, K.H. Byun, J.J. Park, H.S. Kim, Y.O. Lee,K.J. Ahn, H.M. Lee, K.W. Moon, K.K. Bae, D.H. Kim, K.M. Chae,J.K. Kim, D.W. Park

PublicationPlace

Taejon Publisher KAERIPublication

Date2000. 3.

Page 766 p. 111. & Tab. Yes ( V ), No ( ) Size 26 Cm.

Note Nuclear Research and Development Program of MOST

ClassifiedOpen ( ), Restricted ( ),

Class Document, Internal Use Only ( V )Report Type Research Report

Sponsoring Org. Contract No.

Abstract (15-20 Lines)

The purpose of this study is to assess the compatibility of DUPIC (Direct Use of Spent PWR Fuelin CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology beingdeveloped to utilize the spent PWR fuel in CANDU reactors. The Phase I study of this projectincludes the feasibility analysis on applicability of the current core design method, the feasibilityanalysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system,the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle.The results of the validation calculations have confirmed that the current core analysis system isacceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of coresimulations have shown that both natural uranium and DUPIC fuel cores are almost the same from_the viewpoint of the operational performance. For individual reactor system including reactivitydevices, the functional requirements of each system are satisfied in general. However, because ofthe pronounced power flattening in the DUPIC core, the radiation damage on the critical componentsincreases, which should be investigated more in the future. The DUPIC fuel composition heterogeneitydoes not to impose any serious effect on the reactor operation if the fuel composition is adjusted.The economics analysis has been performed through conceptual design studies on the DUPIC fuelfabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuelcycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costsof the fuel cycle components. The results of Phase I study have shown that it is feasible to usethe DUPIC fuel in CANDU reactors without major changes in hardware. However further studiesare required to confirm the safety of the reactor under accident condition.

Subject Keywords

(About 10 words)

CANDU, DUPIC, Compatibility, Economics, Heterogeneity,

Uncertainty, Validation, Thermal-hydraulics, Radiation, Refueling

DUPIC Fuel Compatibility Assessment - International Nuclear ...

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