SCWR Core Design – General (SC07) - ICTP

2291-8

Joint ICTP-IAEA Course on Science and Technology of Supercritical Water Cooled Reactors

Laurence LEUNG

27 June - 1 July, 2011

Atomic Energy of Canada Ltd. (AECL) Stn. 87, Chalk River Laboratories

Chalk River K0J 1J0 Ontario CANADA

SCWR CORE DESIGN � GENERAL

SCWR Core Design – General(SC07)

SCWR Core Design – General(SC07)

Laurence LeunggManager, Advanced Concepts and Collaboration

Joint ICTP IAEA Course on Science andJoint ICTP-IAEA Course on Science andTechnology of SCWRs, Trieste, Italy June 27 – July 1, 2011

UNRESTRICTED / ILLIMITÉ

Objectives

• To present an overview of current SCWR core designs

• To highlight core-design criteria for SCWR

• To introduce the importance of coupling between neutronics and thermal-hydraulics in SCWR core design

• To demonstrate applications of analytical tools (subchannel and CFD codes) in support of SCWR core(subchannel and CFD codes) in support of SCWR coredesign

SCWR Core Design – General (SC07) Joint ICTP-IAEA Course on Sci. & Tech. of SCWRs, Trieste, Italy, June 27 – July 1, 2011UNRESTRICTED / ILLIMITÉ 2

Introduction

• Overview of current SCWR core designs–Similarities and differences

• Design Goals–Generation IV International

Forum (GIF)• Design Requirements

– International standards and local guides

• Core design criteria–Limits for normal operationsLimits for normal operations

and abnormal scenarios• Reactor design process

–Five categories

ht tp ://www.g en-4.org/Techn olo gy/syste ms/scwr. htm

SCWR Core Design – General (SC07) Joint ICTP-IAEA Course on Sci. & Tech. of SCWRs, Trieste, Italy, June 27 – July 1, 2011

Five categories

UNRESTRICTED / ILLIMITÉ 3

Overview of SCWR Design Status

• Global development of SCWR design concepts was introduced in Lecture S04

Korea and United States are no longer pursuing their design concepts– Korea and United States are no longer pursuing their design concepts• Canada completed their pre-conceptual thermal-spectrum core

design– Just started the next phase to complete the conceptual designp p p g

• China is working on the conceptual mixed-spectrum core design– A thermal-spectrum pressure-tube concept is also being pursued

separatelyEU (HPLWR) d J (JSCWR) l d h i l• EU (HPLWR) and Japan (JSCWR) completed their conceptualthermal-spectrum core designs– Awaiting for their government approval of the next phase

• Japan is also working on the conceptual fast spectrum core• Japan is also working on the conceptual fast-spectrum coredesign (Super Fast Reactor)

• Russia completed their conceptual fast-spectrum core design– Unclear about the next phase


Unclear about the next phase


Current SCWR Thermal Core Designs

T. Schulenberg et al., 2009

K Yamada et al 2011

M. Yetisir et al., 2011

K. Yamada et al., 2011


General Similarities and Differences in Thermal Core Designs

• SimilaritiesDirect cycle

• Differences–Reactor type–Direct cycle

–Pressure at 25 MPa–Light-water coolant

Reactor type– Pressure tube vs pressure

vessel–Reactor powerLight water coolant

–Modified stainless-steel cladding

Reactor power–Moderator

– Light water vs heavy waterI l d l l–Vertical core

–Batch fuelling (3 cycles)–Inlet and outlet coolant

temperatures–Fuel designg

– Uranium vs Pu-Thorium–Flow pass

– Single vs 2 or 3 passes


– Single vs 2 or 3 passes


Current SCWR Mixed and Fast Core Designs

closure headtop head

T N k t k t l 2010inlet nozzle

outlet nozzle

protective pipes blockprotective tube unit T. Nakatsuka et al., 2010

Vessel

FACS nozzle

buffle

shaftcore barrel

S. Ryzhov et al., 2011X. Cheng et al., 2007


General Similarities and Differences in Mixed and Fast Core Designs

• SimilaritiesDirect cycle

• DifferencesReactor power–Direct cycle

–Pressure vessel type–Pressure at ~25 MPa

–Reactor power–Inlet and outlet coolant

temperaturesPressure at 25 MPa–Light-water coolant–Modified stainless-steel

–Fuel design (seed and blanket configuration)

Layer assembly vs separatecladding–Vertical core

MOX f l

– Layer assembly vs separateassemblies

–Flow pass–MOX fuel – Single vs 2 passes


Design Goals

• GIF– Enhanced safety (strengthen public acceptance)y ( g p p )– Improved economic (reduce capital and operating costs)– Enhanced sustainability (reduce waste and improve resource

utilization)utilization)– Increased proliferation resistance (strengthen public confidence)

• Competitiveness to other power generation sources and systems– Base load compared to peak load– Supply stability (e.g., oil price fluctuation)pp y y ( g , p )


Enhanced Safety is the Top Priority


Economic


Sustainability

• Uranium fuel is used in all power reactors

Global Uranium Reserves

– Natural– Enriched

• Largest uranium reserves

50

60 PlannedUnder Construction

g(over 100,000 tons)– Australia, Brazil, Canada,

China, Jordan, Kazakhstan,

30

40

50

33

of N

ucle

ar R

eact

orsNamibia, Niger, Russia, South

Africa, Ukraine, United States, Uzbekistan

R id th i l

0

10

20

1 1 2

23

1 14

1 2 1

116

2 2 2 12 2

1

20

2

12

2

14

6

62

9

Num

ber • Rapid growth in nuclear

plant construction– 61 units under construction


– 111 units planned/ordered

Design Requirements

• Functional Requirements• Performance Requirements

• Layout and Modulization• Interfacing Systems

• Safety Requirements• Applicable Codes,

Standards and Classification

• Decontamination and Decommissioning

• Materials and Chemistry• Environmental Conditions• Overpressure Protection• Inspection and Testing

y• Loads, Loads Combinations

and Service Limits• Human Factors and other Inspection and Testing

• Reliability and Maintainability

Design Requirements and Constraints

• International Standards• IAEA Safety Standards Series (Safety of Nuclear

Power Plant: Design) Requirements, NS-R-1• Local Standards/Guides


Local Standards/Guides• Applicable to installation at individual country


Five Categories in Design Process

• Nuclear Design– Neutron economy consideration

• Materials Selection– Minimize corrosion, chemical

– Fuel cycle, core configuration, fuel geometry, fuel-to-moderator ratio

• Thermalhydraulics Design– Efficient/effective heat removal from

reaction, and corrosion product transport

– Enhanced mechanical properties (e.g., high strength) Ch i t t l t tEfficient/effective heat removal from

fuel and core– Fuel rod, fuel assembly, core

configuration • Economics

– Chemistry control strategy– Structural, cladding, fuel, coolant,

moderator, insulator (key to SCWR)• Safety and Control

– Minimize energy costs and enhance plant reliability

– Capital costs (materials, fabrication and construction requirements)

– Facilitate steady-power operation– Limit severity and/or mitigate

consequences of potential accidents– Minimize radiation exposures for

t d l bliq )

– Net electrical output (minimize in-plane energy consumption)

– Operating and maintenance costs (system reliability, waste handling, f l b f l l t f l

operators and general public– Monitoring and protective systems,

shut-down system, containment– Control rods, trips, poisons, fuel,

core configuration


fuel, burnup, fuel cycle, spent-fuelstorage)

– Decommissioning

core configuration


Core Design Criteria Introduction

• All five categories in the design process are linked to the core design/configurationg g

• No SCWR cores in operation– No defined standards/guidelines

N ti i– No operating experience

• Evolved from existing design criteria for PWRs– Maximum cladding surface temperature at nominal conditionsg p– Maximum fuel centerline temperature for nominal fuel rod dimension

at 100% Power– Minimum departure from nucleate boiling ratio (DNBR) at nominalMinimum departure from nucleate boiling ratio (DNBR) at nominal

operating conditions– Irrelevant to SCWR since there is no phase change


SCWR Cladding Temperature Criteria

• Minimum Nominal Cladding Temperature Ratio– Ratio of maximum calculated design cladding temperature and the

defined cladding limit temperature– Define the maximum allowable cladding temperature at normal

operating conditionsE ti l f f th f l d– Ensure optimal performance of the fuel rod

– Limit corrosion rates to acceptable values

• Minimum Safety Cladding Temperature RatioR i b h i l l d i l ddi– Ratio between the maximum calculated transient claddingtemperature and the defined cladding safety limit temperature

– Define the maximum allowable cladding safety limit temperature at transient conditions in safety analysestransient conditions in safety analyses– Based on 95% probability and 95% confidence for the fuel cladding

integrity– Higher than normal operation cladding temperature for short periods of


g p g p ptime


SCWR Fuel Temperature Criteria

• Maximum Fuel Temperature for Nominal Conditions and Transient Analysesy– Fuel centreline temperature criteria

– Prevent fuel melting– Fuel and composition dependentFuel and composition dependent

– More crucial for SCWR– High operating temperature


Deteriorated Heat Transfer Phenomena

• DNB-like phenomena with sharp cladding temperature rises at the pseudo-critical

Schmidt (1959)

5-mm ID tube, G = 700 kg m-2 s-1

rises at the pseudo criticalpoint– Observed at low flows and high

heat fluxes in tubes and annuliU l h th it b– Unclear whether it can beencountered in bundles– Spacing devices enhance

turbulence suppressing this phenomenaphenomena

• Ongoing debate on its importance– Corresponding peak cladding temperatures could be lower than those

at high bulk-fluid temperatures– The phenomena is ignored in some preliminary safety analyses

– Concerns:– Its occurrence at axial peak power location– Non-uniform occurrences around fuel rod leading to bowing

SCWR Core Design – General (SC07) Joint ICTP-IAEA Course on Sci. & Tech. of SCWRs, Trieste, Italy, June 27 – July 1, 2011UNRESTRICTED / ILLIMITÉ

Non uniform occurrences around fuel rod leading to bowing– Material corrosion

18

Common SCWR Design Challenges: Materials

SS316 and SS316L25 MPa, 270-550oC, >500 h

• In-core components (internals and fuel cladding):

No single alloy with sufficient– No single alloy with sufficientinformation to confirm its performance

– Zirconium-based alloys are not viable material

D. Guzonas, 2011

viable material– Need thermal and corrosion-

resistant barrier

Out of core components• Out-of-core components– Based on material characteristics for fossil-fire power plants– Different acceptance requirements on corrosion for nuclear power

plantsp• Demonstrate performance in key areas:

– Corrosion and stress corrosion cracking; strength, embrittlement and creep resistance; and dimensional and microstructural stabilityQ tif i di ti ff t


– Quantify irradiation effect

Common SCWR Design Challenges: Chemistry

• Changes in chemical properties due to marked change in SCW density through the critical point

• SCWR In-core radiolysis is markedly different from those of conventional water-cooled reactors– Extrapolation of the behavior is inappropriate

• Identification of an appropriate water

p pp p– Strong impact on corrosion and stress corrosion cracking

D. Guzonas, 2011

appropriate waterchemistry to minimize– Corrosion rates– Stress corrosion crackingStress corrosion cracking– Deposition of deposits on fuel

cladding and turbine blades• Establish a chemistry-


control strategy

Common SCWR Design Challenges: Thermalhydraulics

• Cladding temperature limits have been adopted as the design criteria

Traditional CHF criteria is no longer appropriate due to the lack of– Traditional CHF criteria is no longer appropriate due to the lack ofphase change

• Accurate prediction of supercritical water (SCW) heat transfer is essential to establish the power output and safety margin– Need experimental data for relevant bundle geometry at conditions of

interest• Most experimental data on SCW heat transfer were obtained

with tubeswith tubes– Applicable to the fossil-plant boiler but not directly to SCWR geometry

and conditions• No experimental data on SCW heat transfer in relevant bundle p

geometry– Only a few bundle-subassembly data– No information on separate effects


Common SCWR Design Challenges: Safety

• Establishment of design-basis accidents and potential initiators for severe accidents

• Transient experimental data on supercritical heat transfer– Pressure transient through the pseudo-critical point

• Experimental SCW data on critical flow– Designs of safety/relief valve and depressurization system– Support of large break loss-of-coolant accident analyses

• Susceptibility to dynamic oscillationsp y y– Large variation of coolant density in the axial direction – Strong coupling of the neutronic and thermal-hydraulic behavior– Need experimental data and analytical model to predict the onset of

instability• Applicability of safety analysis code

– Validations were performed for subcritical applications only


– Need integral test data at supercritical conditions


Nuclear and Thermalhydraulics Designs

• Core coolant density variations at normal operationoperation– Small in LWRs and HWRs– Large in SCWRs

• Large density variation in• Large density variation inthe core affecting– Moderation and absorption– Neutron spectrumNeutron spectrum– Fission power distribution– In turn, impacting the coolant

and moderator density profiles• Coupled neutronics and

thermalhydraulics analysis is required in design

l l ti J Q Shan et al 2010


calculations

UNRESTRICTED / ILLIMITÉ

J.Q. Shan et al., 2010

23

Neutronics and Thermalhydraulics Coupling

• Coupling analyses are performed using two suites of analytical toolsets– Neutronics codes (such as MCNP) to establish radial and axial

distributions of power– Thermalhydraulics codes (subchannel code) to determine fuel

temperature and coolant temperature (and in turn the coolant density)• An iterative approach is applied in the calculations

R di l d A i l Di t ib ti f P

Neutronics Code ThermalhydraulicsC d

Radial and Axial Distributions of Power

Neutronics Code Code

Fuel Temperature

SCWR Core Design – General (SC07) Joint ICTP-IAEA Course on Sci. & Tech. of SCWRs, Trieste, Italy, June 27 – July 1, 2011UNRESTRICTED / ILLIMITÉ

Coolant TemperatureCoolant Density

24

Illustration of Fuel Channel Density Variations

• Established core power profiles at BOC conditions using the RFSP codeusing the RFSP code– Coolant flowing uni-

directionally inside the fuel channel 0 6

0.7

0.8

/cc)

E ilb i– Large moderator pool• Coolant properties are

defined at 5 set positions l th f l h l 0.3

0.4

0.5

0.6

nt D

ensi

ty (g

/ EquilbriumVoidedFlooded

along the fuel channel• Consider Three Cases:

– Equilibrium Coolant Density ( d T t ) Di t ib ti 0

0.1

0.2

0.3

Coo

lan

(and Temperature) Distribution– Uniform Coolant Density, Inlet

Conditions– Uniform Coolant Density

0 100 200 300 400 500

Distance From Inlet (cm)


Uniform Coolant Density,Outlet Conditions


Impact of Density Variation on Axial Power Profiles in Fuel Bundle

• Channel power changes determined using WIMS with

Heavy Water Moderator

Liner Pressure TubeInsulator

determined using WIMS witha fixed total neutron flux

• Voided channel– Similar to channel outlet 70

Fuel Spacer GridLight Water Coolant

Similar to channel outletconditions

– Overall power decrease• Flooded channel 50

60

HE)

– Similar to channel inlet conditions– Overall power increase

• Equilibrium channel 20

30

40

Pow

er (W

/g H

Equilibrium

– Transition from flooded channel to void channel case

– Power peak close to the upstream end

0

10

20

0 100 200 300 400 500

VoidedFlooded


upstream end 0 100 200 300 400 500Distance From Inlet (cm)


Impact of Water Rods on Density Profiles

• INEEL performed coupled neutronic and thermal-neutronic and thermal-hydraulics analysis

• SC Water flows down inside t d ( d t )water rods (as moderator)

and up along the fuel rods (as coolant) of the fuel

blassembly• Water temperature

increases continuously and water density decreases– The decrease is steeper along

the fuel rods than the water d


rodsP. MacDonald et al., 2005


Impact of Density Variation on Axial Power Profiles in Fuel Assembly

• INEEL applied MCNP, ORIGEN2, and MOCUP

d i t icodes in neutroniccalculations

• Reference case with 5% if i huniform enrichment

–Power peak at the downstream end of the core

– Influence of water-roddensity variations

• Power peak reduction i th i h tusing three enrichment

zones–Flatten the axial power

filP. MacDonald et al., 2005


profile


Thermalhydraulics Tools

• Density distributions were based on cross-sectional averaged calculationsg

• Local-conditions variations are important due to flow and enthalpy imbalances within bundles

I t l id d d i f l bl– Internal, side, and corner rods in a fuel assembly– Rods in various rings of a fuel bundle

• Detailed calculations are required using analytical toolsq g y– Subchannel codes

– Essential for fuel design and optimizationComputational fluid dynamics tools– Computational-fluid dynamics tools– Support fuel design (e.g., spacers) and subchannel-code development– Design support of other components (e.g., outlet nozzles)


Subchannel Codes

• Local flow conditions and wall temperature predictions

S t th fl i th f l b dlReactor Core

CANDU Reactor Core Showing Fuel Channels

– Separate the flow area in the fuel bundleor assembly into interconnected subchannels

– Each subchannel is treated as a tube-equivalent flow channel

Grid Spacers

End Plate

Calandria Tube

– Local flow conditions inside each subchannel are calculated using equations of conservation and boundary conditions

– Constitutive equations are introduced to t bli h h t t f t iti i t

Pressure Tube

Fuel ChannelShowing Fuel Bundles

Subchanneli (k)

establish heat transfer, transition points,mixing between subchannels, void drift, buoyancy, etc.

• Subchannel flow modelling:– Sub-critical conditions Subchannel

Subchannelj (k)

Axial NodeJGap (k)

Axial NodeJ - 1

Sub critical conditions– Homogeneous flow– Drift-flux model– Two fluids

– Super-critical conditionsNo phase change Distance, l

Control Volumes

j (k)

Angle, Fi (k)


– No phase change– Pseudo-liquid and pseudo-vapor flows

Fuel BundleShowing Subchannels

,Centroid-to-Centroid

Specifications


Subchannel Analysis for SCWR Fuel Assemblies

• Mainly for normal operating conditions– Relatively simpler than analyses for current fleets of reactorsy p y

– No phase change– Provide detailed distributions of local flow conditions and cladding

temperature in the fuel assembly and bundletemperature in the fuel assembly and bundle– Support fuel design and optimization

• Issues– Lack of relevant data to establish constitutive equations (e.g., mixing)– Reflecting mainly the parametric trends of the implemented

correlations– Observed effect in tubes may not be present in bundles

– Lack of prediction methods for separate effects (e.g., spacer, gap size, etc.)


– Lack of relevant bundle data for validation


Illustrations of Subchannel-Code Predictions of Cladding Temperatures


P. MacDonald et al., 2005.


Fuel Bundle Design Optimization Using Subchannel Code

Exit PressureInlet Temperature

25 MPa350oC

844.80751.78 760.10670.09

Temperature (oC)Cladding(Peak)

Coolant(Outlet)

Temperature (oC)Cladding(Peak)

Coolant(Outlet)

Inlet TemperatureOutlet TemperatureMass VelocityAverage Heat Flux

350 C625oC

806 kg/(m 2·s)0.788 MW/m2

775.88

793.11

810.34

827.57

665.54

687.10

708.66

730.22

736.66

742.52

748.38

754.24

635.03

643.80

652.56

661.32

724.19

741.42

758.65

600.87

622.43

643.99

713 22

719.08

724.94

730.80

599 98

608.74

617.51

626.27

672.50

689.73

706.96

536.20

557.75

579.31

701.50

707.36

713.22

582.46

591.22

599.98

Optimized BundleReference Bundle


Optimized BundleReference Bundle


Needs of Computational Fluid Dynamics Analyses

• Issues related to subchannel code applications prompt the need of further details in thermal-hydraulics calculationsy

• In addition, supercritical fluid heat transfer depends strongly on the near-wall characteristics

T b l t t d t b d l ti f ll– Turbulence structure due to buoyancy and acceleration of near-wallfluid– Deteriorated heat-transfer phenomena

• Subchannel code is based on cross-sectional averaged conditions within a subchannel– Cannot capture such detailsCannot capture such details

• Applications of computational fluid dynamic tools to support fuel design and optimization and supplement i f ti t b h l l


information to subchannel analyses


CFD Analysis for SCWR Fuel

• Computational fluid dynamics (CFD) tools apply numerical methods and algorithms to evaluate fluid flow and heat transfer in open or closed channels

• Turbulence modelling– Model the length and time scales associated with turbulent flow– Examples: Direct numerical simulation, Large eddy simulation, Reynolds-

averaged Navier–Stokes equations etcaveraged Navier Stokes equations, etc.• Mainly for normal operating conditions

– Relatively simpler than analyses for current fleets of reactors– No phase change

CFD for two phase flow is still under development– CFD for two-phase flow is still under development– Provide detailed distributions of flow conditions and cladding temperature inside

a subchannel• Issues

R i id bl ti li iti it li ti t i l– Require considerable computing resources limiting its applications to simplegeometry or small sections of fuel assembly at this point

– Lack of turbulence measurements to establish appropriate turbulence models– Lack of detailed measurements of flow and temperature of relevant fixtures or

geometry for validation


geometry for validation


Illustrations of CFD Modeling of Turbulent Flow near a Rod

0.03

0.040.05

0.06

0.06

0.060.070.07

0.070.2

u'/U

F rame 001 20 Apr 2010 fluent6.3.26

0.0150.02 0.020.025 0.025

0.0 0.0 3 0.0350.035 0.040.04

0.045

0.0450 .045

0.05

0.05 0 .05

0.05

0.060.2

v' /U


0. 01

0.02

00

0.02

0.0

3

0 .03

0.04

0.04

0.04

0 .04

0.04

0.04

0

0.04

0 .05

0. 05

0 .05

0.05

0.05

0.05

0 .05

0.06

Y 0.1

0.15

0.080.070.060.050.040.030.020.01

0.010.015

0.02

0 .025

0.03

0.035

0.03

5

0.035

0.04

0. 04

0.04

0.04

00

0 .0 4

0.04

5

0

0 .04

5

0.0

45

0.045

0.04

5

0.05

0.05

0.05

0.05

0.05

0.05

0.0

55

0.055

0.0 5

5

5 5

0.05

5

0.055

0.0 55

0.0

6

0 .06

0.06

Y 0.1

0.15

0.060.0550.050.0450.040.0350.030.0250.020.0150.01

’/U ’/U’/U ’/U0.01

0 01

02

0 02 0.02

0 .0 2

0.03

0 03

0.03

0.04

0 0 4

0.0 4

0.05

0.05 0.0 50.06

0 .06

0.06

X0 0 .05 0.1 0. 15 0.2

0

0.05

0.01

0 0101

0. 015

0 0150 015

0.0 2

0 02

0.025

0 025

00

3

0.03

0.03

0035

0.035

0.03 5

04

0.04

0 .04

0 .04 5

0.0450.04

5

0.05

0.050 .0

55

0.06

0 .06

0 .06

0.06

X0 0 .05 0.1 0. 15 0.2

0

0.05

0 0750 075 0 080 08 0 085 0 0850 090 2


0 003 0 004 0 0050 00 5 0 0060 007 0 0070 00 80 009 0 009 0 010 2


u’/Ub v’/Ubu’/Ub v’/Ub

0.03

0. 04

0.045

0.045

0.050.05

0.05

0. 055

0 .055

5

0 .06

0.0 6

0.06

0.06

0.065

0.06

5

0.065

0.0

7

0.07

0 .0 7

0.07

0.0 7

0.075

0.07

5

0.075

0.0750.075

0.07

5

0 8

0.0 8

0.08

0.080 .08

0.0

8

0.08

0.085

0.085

0 .085

0.085

85

0.08

5

0.085 0.0850.090.09

0.09

0.09

0.0

99

0.09

Y 0.1

0.15

0.2

w'/U

0.090.0850.080.0750.070.0650.060.0550.050.0450 04 02

0.003

0.003

0.00

3

0 .0040.

004

0.004

0 .004

0.0

04

0 .005

0 .00

5

0.005

0 .0050.00 5

0.00

5

0.006

0 .00

6

0.006

0.006

0.00

60.

006

0.0060.007

0.007

0.00

7

0.00

7

0.007 0.0070.0 08

0.0080.00 8

0 .00 8

0.00

8

0

0.00

8

0.0090 .009 0.009

0.009

0.00

9

0.0

090.

009

0.01

0 .0 10.011Y 0.1

0.15

0.2

k/U2

0.0140.0130.0120.0110.010.0090.0080.0070.0060.0050 004

0 .02

0.025

0.0 25

0. 03

0 .0 3

5

0.035

0 035

0

0.04

0 04

0 .0 45

0 0450 045

0.0 5

0.0 5

0 0 50 05

0.05 5

0.05

5

0.055

0 055

0.0

6

0.0 6

0.0 6

0 060 06

0 .0 6

5

0.0 6

5

0.0

65

0 .06

5

0 065

007

0.07

0.07

0.07

0 0 7

0.07 5

0.07 5

0 .07

5

0075

0 075

0.075

0.0 7

5

0 .0 8

0.08

0.080

08

0 08

0.0 80 .0 8

0.0

8 5

0 .0850 .0 85

0.0 8

0.0850.0850 085

0.085

0 .0 9

0 .09

0 .09

0

0.09

0.09

0 09 0 090.09

X0 0 .05 0.1 0. 15 0.2

0

0.05

0.040.0350.030.0250.02

0.00

1

0.001

0 001

0.00

0.002

0 0020 003 0 0030.003

0.00

3

0.004

0004

0 004 0 004

0.0 0

4

0.0 05

0 .005

000

5

0 005

0 .0 05

0.0 0

5

0. 006

0.006

0.006

0 0060. 006

0.007

0. 00 7

0.00

7

0.0070 007

0.007

0.0 0

8

0. 00 8

.008

0.00

8

0.00 8

0.

0.009

0 .00 9X

0 0 .05 0.1 0. 15 0.20

0.05

0.0040.0030.0020.001

w’/Ub k/Ub2

w’/Ubk/Ub

2


Illustrations of CFD Modeling of SCW Heat Transfer of a Wire-Wrap Spacer

WiWire wrap spacer

Fuel rod

J.A. Lycklama et al., 2011.


Conclusions

• Several SCWR concepts are being pursued and have been completed at different stages

• SCWR reactor design goals match closely to those established by GIF

• Core design criteria are mainly related to cladding and fuelCore design criteria are mainly related to cladding and fueltemperature

• Strong challenges remain in key design areasC l d t i d th lh d li l i i• Coupled neutronic and thermalhydraulics analysis isnecessary in view of the large variation in fluid density in the core impacting power distribution and stability

• Subchannel and CFD codes are important tools to assist in SCWR core and fuel designs– Further developments are still necessary


p y


References (1)

• IAEA, “Description of SCWR Design Concepts”, Chapter 2 of Heat Transfer Behaviour and Thermohydraulics Codes Testing for SCWRs, IAEA Technical Document, (in draft)IAEA “S f t f N l P Pl t D i ” IAEA S f t St d d S i• IAEA, “Safety of Nuclear Power Plant: Design”, IAEA Safety Standard Series,Requirements, IAEA-NS-R-1, 2000.

• M. Yetisir et al., “Conceptual Mechanical Design for A Pressure-Tube Type Supercritical Water-Cooled Reactor”, Proc. 5th International Symposium on Supercritical Water-cooled Reactors Vancouver Canada March 13-17 2011Supercritical Water cooled Reactors, Vancouver, Canada, March 13 17, 2011.

• T. Schulenberg et al., “European Supercritical Water Cooled Reactor”, FISA 2009, Prague, Czech Republic, June 22-24, 2009.

• K. Yamada et al., “Overview of The Japanese SCWR Concept Developed under The GIF Collaboration”, Proc. 5th International Symposium on Supercritical Water-, y p pcooled Reactors, Vancouver, Canada, March 13-17, 2011.

• X. Cheng et al., “A Mixed Core for Supercritical Water-cooled Reactors”, Nuclear Engineering and Technology, 40(2), pp. 117-126, 2007.

• T. Nakatsuka et al., “Current Status of Research and Development of Supercritical C (S )Water-Cooled Fast Reactor (Super Fast Reactor) in Japan, Proc. IAEA Technical

Meeting on Heat Transfer, Thermal-Hydraulics and System Design for Super-Critical Water-Cooled Reactors, Pisa, Italy, July 5-8, 2010.

• S. Ryzhov et al., “Concept of A Single-Circuit RP with Vessel Type Supercritical Water-Cooled Reactor” Proc 5th International Symposium on Supercritical Water-


Water-Cooled Reactor , Proc. 5th International Symposium on Supercritical Water-cooled Reactors, Vancouver, Canada, March 13-17, 2011.


References (2)

• U.S. DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, “A Technology Roadmap for Generation IV Nuclear System”, GIF-002-00, December 2002.R A K i f “N l E i i Th d T h l f C i l N l• R.A Knief, “Nuclear Engineering: Theory and Technology of Commercial NuclearPower”, Hemisphere Publishing Corporation, Taylor & Francis, Washington DC 20005-3521, ISBN 1-56032-089-3, 1992.

• P. MacDonald et al., “Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production – 2nd QuarterlyReactors for Actinide Burning and Electric Power Production 2nd QuarterlyReport”, Idaho National Engineering and Environmental Laboratory, Report INEEL/EXT-02-00759, March 2002.

• P. MacDonald et al., “Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production – Final Report, 12th Q t l R t” Id h N ti l E i i d E i t l L b tQuarterly Report”, Idaho National Engineering and Environmental Laboratory,Report INEEL/EXT-04-02530, January 2005.

• D. Guzonas, “Materials and Chemistry for a Supercritical Water-Cooled Reactor –Progress and Challenges”, Proc. Int. Symposium of SCWR, Vancouver, BC, Canada Mar 13-17 2011Canada, Mar. 13-17, 2011.

• Shan, J.Q. et al., “Coupled Neutronics/thermalhydraulics Analysis of CANDU-SCWR Fuel Channel”, Annals of Nuclear Energy, Vol. 37, pp. 58-65, 2010.

• J.A. Lycklama et al., “Development of A Heat Transfer Correlation For The HPLWR Fuel Assembly by Means of CFD Analyses”, Proc. Int. Symposium of


HPLWR Fuel Assembly by Means of CFD Analyses , Proc. Int. Symposium ofSCWR, Vancouver, BC, Canada, Mar. 13-17, 2011.


SCWR Core Design – General (SC07) - ICTP

Documents

Transcript of SCWR Core Design – General (SC07) - ICTP