Preliminary Safety Analysis for the IRIS Reactor

Proceedings of ICONE-10 10th International Conference on Nuclear Engineering

April 14–18, 2002, Arlington, VA

Proceedings of ICONE1010th International Conference on Nuclear Engineering

Arlington, VA, April 14-18, 2002

ICONE10-22398

ICONE-22398

PRELIMINARY SAFETY ANALYSIS OF THE IRIS REACTOR

M. E. Ricotti Polytechnic of Milan

Dept. Nuclear Engineering Via Ponzio 34/3, 20133 Milano, Italy

[email protected]

A. Cammi Polytechnic of Milan

Dept. Nuclear Engineering Via Ponzio 34/3, 20133 Milano, Italy [email protected]

A. Cioncolini Polytechnic of Milan


[email protected]

A. Cipollaro University of Pisa – DCMNP

Via Diotisalvi, 1, 56126 Pisa, Italy [email protected]

F. Oriolo University of Pisa – DCMNP

Via Diotisalvi, 1, 56126 Pisa, Italy [email protected]

C. Lombardi Polytechnic of Milan


[email protected]

L. E. Conway Westinghouse Electric Company

1344 Beulah Road Pittsburgh, PA 15235, USA Phone: +1-412-256-1089

[email protected]

A. C. Barroso Comissao Nacional de Energia Nuclear

– CNEN, Brazil Rua General Severiano 90

22294-900 Rio de Janeiro, Brazil [email protected]

ABSTRACT A deterministic analysis of the IRIS safety features has

been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, a preliminary set of accident transients for the whole primary and safety systems was investigated. Since the project was in a conceptual phase, the reported analyses must be considered preliminary. In fact, neither the reactor components, nor the safety systems and the reactor signal logics were completely defined at that time.

Three “conventional” design basis accidents have been preliminary evaluated: a Loss Of primary Flow Accident, a Loss Of Coolant Accident and a Loss Of Feed Water accident.

The results show the effectiveness of the safety systems also in LOCA conditions; the core remains covered for the required grace period. This provides the basis to move forward to the preliminary design.

INTRODUCTION IRIS[1,2] is a small modular pressurized water reactor, which

is designed to reach a higher safety level through plant layout simplification, intrinsically safe process and passive safety features.

IRIS main characteristic is the integral reactor concept, being all the major reactor coolant system components, i.e. coolant pumps, steam generators and pressurizer, located inside the pressure vessel.

The project is developed by a large international consortium, led by Westinghouse, joining the efforts of industries, research centers, universities and utilities.

The IRIS safety[3,4] is based on a safety by design approach, attempting to first eliminate the possibility of accident sequences from occurring, and secondly to reduce the severity of their consequences and/or the probability of occurrence. The integral reactor vessel configuration, is an ideal layout for implementing this approach: lack of large loop

Keywords: IRIS, Integral reactor, Accident analyses, Safety

systems, RELAP, LOFA, LOCA, Generation IV

1 Copyright © 2002 by ASME

piping, substantial percentage of natural circulation, large water inventory, plus a high pressure containment.

THE "SAFETY BY DESIGN” APPROACH The main safety objective of IRIS is to increase the degree

of inherent safety of the system, by eliminating or mitigating the initiating events and accidents via a new approach to the design of LWR.

Examples of this strategy, identified as ‘Safety by Design’ (the progressive application of inherent safety principles), are: i) the adoption of an integral pool reactor (to eliminate large vessel penetrations and thus eliminating the possibility of a Large LOCA), ii) the elimination of low level penetrations in the same RPV (thus lessening the consequences of small LOCAs), iii) the adoption of Once-Trough Steam Generators with the primary fluid flowing on the outside of the tube bundle (eliminating the concern for MSGTR), and iv) a reactor design with a high degree of Natural Circulation, up to 100% (to eliminate LOFAs). Table 1 shows some examples of the IRIS Safety by design approach. The IRIS concept has been heavily influenced by this approach, in which a high degree of safety is not achieved by designing active or passive systems to cope with accidents, but rather by dealing directly with them through the main design.

Table 1 Examples of the IRIS “Safety by Design” Philosophy

Effect on Different Design Choices

Design Characteristic

Safety Implication

Related Accident Disposition

Integral configuration

No external loop piping

Large LOCAs Eliminated

High degree of natural circulation

LOFAs

(e.g., pump seizure)

Either eliminated (full natural circulation) or made acceptable (partial natural circulation)

Tall vessel with elevated steam generators

Can accommodate internal control rod drives

Reactivity insertion due to control rod ejection

Eliminated

Long life core No partial refueling

Refueling accidents

Reduced probability

Large water inventory inside vessel

Slows transient evolution

Helps to keep core covered

Reduced size, higher pressure containment

Inside the vessel heat removal

Reduced driving force through primary opening

Small-medium LOCAs

Core remains covered with no safety injection

A limited number of simple passive safety system was anticipated to complete the IRIS safety strategy, but the key features of the Safety by design philosophy is to eliminate or reduce the number of safety systems (and thus increase the safety and reduce the cost of the plant) by designing the system to prevent the accidents, instead of coping with them.

The possibility of implementing this philosophy in IRIS descends mainly from feature i), i.e. the choice of an integral layout (see Figure 1), a radically new approach with respect to current commercial water reactors design.

Figure 1 IRIS Integral Primary System

SAFETY SYSTEMS The primary safety strategy in IRIS relies on two main

items. A small, high pressure, pressure suppression containment is adopted to limit the initial blow down following a small-to-medium Loss Of Coolant Accident (LOCA), with the suppression pools used to limit the initial pressure peak within acceptable limits. Secondly, decay heat is removed from the containment by a Passive Emergency Heat Removal System (PEHRS) so that the loss of water inventory by the RPV is limited by condensing the produced steam within the vessel, and not in the containment.

The PEHRS (see Figure 2) is made up of four trains, each connected to the feed and steam line of a pair of Steam Generators. IRIS has eight modular, helical coil SGs, paired


together to four steam and feed-lines. A passive system is connected to each steam/feed line to guarantee redundancy. Following an accident the Main Feed/Steam Isolation Valves (MFIV/MSIV) are closed after an S-signal (low pressure in the RPV or low level) as the safety valves on the PEHRS loop are opened. Each steam and feed water line will have redundant isolations in order to minimize the likelihood of containment bypass scenarios following a Steam Generator Tube Rupture.

Figure 2 P&I Diagram of the Passive Emergency Heat Removal System

The steam for the SGs is thus no longer directed to the

turbine and secondary loop, but rather deviated to a U-tube heat exchanger (this system is yet under definition) located in the refueling tank. The steam is condensed in this pool heat exchanger and then the liquid flows through a line connected to the feed line and back to the reactor. A natural circulation loop is thus established and decay heat can be removed from the reactor. The whole system is designed (as are the secondary system steam and feed-lines up to the isolation valves) for full design pressure (155 bar, 2250 psia).

Various level and temperature sensors are added for monitoring of the PEHRS. According to the IAEA classification, this is a class IV passive system due to the presence of valves that need to be actuated to start the system.

This system is very simple from a design and conceptual point of view, but it still requires careful study and examinations from a thermal hydraulic perspective, in particular to (1) verify the start-up behavior of the system and the possible effect of plants conditions on start-up and (2) to verify the correct performance of the system under different conditions during the evolution of accident sequences that require the intervention of the PEHRS.

The PEHRS is complemented by a water tank which is added to the line taking the liquid back to the feed line. Its purpose is to replenish the circuit with a sufficient amount of water and to prevent the progressive drying up of the system following small water leakages through the isolation valves.

Two important features of the IRIS design (the elimination of large piping penetrations in the vessel thanks to the integral layout, and the elimination of low level penetrations in the vessel) act in synergy with this strategy to actually eliminate the need for short term water makeup to the RPV, guaranteeing a long grace period following a LOCA. Henceforth the LOCA coping strategy in IRIS is an example of the application of the Safety by design concept.

The Safety by design strategy is then completed by the definition of a limited number of passive safety systems to achieve a very high degree of safety and very low CDFs (the target for IRIS CDFs is <10-8). It must be pointed out that the deterministic analyses such as those discussed in this paper, are iterated with the Probabilistic Safety Analysis (PSA) carried out in parallel5. This approach is intended to provide valuable insights to the project in the early design stage, thus implementing a “joint deterministic & probabilistic informed design strategy”. Significant information coming from a simplified PSA analysis already led to modifications of the PEHRS. As a further example, an alternative solution for the water make-up tank in the PEHRS is to increase the size of the descending pipe line, in order to store the same amount of water. A PSA response on the two options will be taken into account for the decision.

PRELIMINARY ACCIDENT ANALYSIS The preliminary deterministic analysis of the IRIS safety

features has been carried out by means of the best-estimate code RELAP5mod3.2.

First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, a preliminary set of accident transients for the whole primary and safety systems was investigated. Since the project was in a conceptual phase, the reported analyses must be considered preliminary. In fact, neither the reactor components, nor the safety systems and the reactor signal logics were completely defined.

The following transients were simulated: 1. Loss Of Primary Flow Accident (LOFA); 2. Loss Of Coolant Accident on the primary side (LOCA); 3. Loss Of Feed Water on the secondary side (LOFW).

The plant configuration adopted for the initial RELAP model was a prior version to the current one previously shown in Figure 1. This version (Figure 3) had canned motor pumps located in the cold area to improve their performance. Later adoption is of the internal spool pumps[6] allowed eliminating the vessel penetrations close to the top of the core. The RELAP nodalization performed for the interim configuration is depicted in Figure 4. In order to facilitate further IRIS RELAP model development and maintenance, the information about each component is contained in a set of database worksheets. Thus, all the input is easily accessible and can be updated when new data become available.


https://www.researchgate.net/publication/236396076_The_IRIS_Spool-Type_Reactor_Coolant_Pump?el=1_x_8&enrichId=rgreq-788db3ff-42be-4932-aff3-b880273f696f&enrichSource=Y292ZXJQYWdlOzIzNjU1Mzg0OTtBUzoxOTg3MTA3MDYzNDgwMzRAMTQyNDM4NzczMjM5Ng==

Figure 3 IRIS integral RPV (earlier version with canned pumps)

The primary and secondary coolant systems are divided

into nine “main” or “general” zones: the lower plenum, the core, the riser, the downcomer, the intact heat transfer loop made up by a couple of Steam Generator modules and connected PEHRS, the eventually broken heat transfer loop, the pressurizer, the boundaries for the PEHRS and the control components.

A detailed description of the modular, helical-coil Steam Generator is presented in ref. 7.

Loss of primary Flow Accident (LOFA) A complete loss of forced flow accident may result from a

simultaneous loss of electrical power to the reactor coolant pumps.

As the pumps start to coast-down, the immediate effect is a rapid increase in the coolant temperature. The pump underspeed setpoint is almost instantaneously reached and the reactor is tripped; then natural circulation flow is established.

Figure 4 RELAP5/MOD3.2 Nodalization of IRIS – Primary System

The sequence of the main imposed trips is shown in Table

2. In the analysis there is no trip related to low pump speed and the scram is conservatively actuated by the over-temperature trip when the core outlet coolant temperature is higher than 610K, corresponding to a core temperature increase of 118% over nominal conditions.

The aim of this analysis is to show that: - The coastdown rate of the reactor coolant pumps is

sufficient to prevent reaching the minimum departure from nucleate boiling ratio;

- The natural circulation on the primary side is not hindered for any reason during the transient and it is large enough to ensure a stable safe plant shutdown with the pumps stopped and the control rods fully inserted;

- The natural circulation of the Passive Emergency Heat Removal System (PEHRS) operates properly and is able to remove the decay heat.


Table 2 Sequence of Trips for the LOFA

Event # Event Description Set point Values

1 Loss of all 4 pumps (initiating event) coastdown start

0.0 sec

2 Scram signal setpoint Core ∆T > 118% nominal core ∆T

3 Intact and broken loop feed water isolation

valves closure

Scram time + 5.0 sec (*)

4 Main Steam Isolation Valve closure

FWIV closure + 5.0 sec (*)

5 Emergency Heat Removal Globe Valve opening

MSIV closure + 5.0 sec (*)

* Assumed delay for valve sequence is an educated guess. As noted below, the main assumptions for this analysis

concern the delay times for isolation of the feed and steam lines and actuation of the Passive Emergency Heat Removal System. Moreover, the reference calculation considers the case that all four loops of the PEHRS are used, but it is also possible to simulate the same transient assuming that, for example, one out of four PEHRS subsystems is not operational. No remarkable differences were found since the heat transfer surface of the in-pool condenser was large enough to reject the decay heat.

The analysis of this accident includes 1000 s of steady-state calculation and six hours of transient calculation after the event initiation. This long period of time is due to the phenomena occurring in the SG secondary sides as discussed below.

The initiating event is the simultaneous loss of the electrical power supply to all reactor coolant pumps. Upon the loss of power, the RCPs coastdown, the primary system pressure losses decrease and the primary coolant temperature increases. The reactor power (Figure 5) drops following the reactor trips and the PEHRS is actuated to remove the decay heat.

With no RCPs operating, the cooling of the primary system is ensured by the natural circulation capability of the IRIS configuration. This corresponds to a mass flow rate about 10% of nominal full flow at the beginning of the transient and around 7% average for the entire transient (see Figure 6). The PEHRS pool needs to be modeled in more detail to accurately reproduce all the physical phenomena that occur in the pool and natural circulation flow path. However, indications are that the capability of the PEHRS to guarantee a safe shutdown of the reactor and remove the decay power is sufficient to avoid any DNB or core damage. Natural circulation is achieved in both the RCS and the PEHRS loops and at the beginning of the transient the extracted power is very high due to the cool water coming from the descending leg of the PEHRS and the large thermal difference between the hot primary system and the cold PEHRS pool heat sink. The pool temperature starts to increase

and reaches saturation after about 3000 s, and then water starts to evaporate and after 24 hours the pool is almost empty.

Thus, this transient can be divided in two phases: the first stage with efficient heat transfer by the pool exchangers, resulting in the primary system pressure (Figure 7) and coolant temperature decrease. The second phase, from about 12000 seconds to the end, with an increase in both primary system pressure and coolant temperature due to the progressive evaporation of the heat sink and consequent heat transfer degradation.

Loss Of Feed Water accident (LOFW) For the RELAP simulation of this accident two loops were

modeled, one detailed, single PEHRS loop (Loop 1) and one lumped loop, simulating the remaining 3 PEHRS loops (Loop 2).

The single loop is made of two Steam Generator modules and one in-pool Heat Exchanger. The lumped loop represents the symmetric behavior of the other SG modules and passive safety systems. Thus, Loop 1 is equivalent to 1 PEHRS circuit out of 4 and 2 SGs modules out of 8, while Loop2 is equivalent to 3 PEHRS circuits out of 4, and 6 SGs modules out of 8.

The RELAP nodalization for the RPV has been maintained as depicted in Figure 4. Loop 1 has 25 components, subdivided in 190 volumes, while Loop 2 is made up by 26 components, 240 volumes.

The resulting RELAP nodalization is shown in Figure 8. Other main assumptions taken into account for the set up of the accident simulation are: - all PEHRS (4 circuits) are activated; - the reactor is tripped; - no heat losses on piping are simulated.

The accident sequence is summarized in Table 3.

0

100

200

300

400

500

600

700

800

900

1000

1100

-10 0 10 20 30 40 50 60 70 80 90 100

Time (s)

Pow

er (M

W)

Figure 5 LOFA: Reactor Power


Time (s) % (

NC

mas

s fl

ow

rat

e/n

om

inal

mas

s fl

ow

rat

e)

Figure 6 LOFA: Percentage of Natural Circulation

in Primary System

Figure 7 LOFA: Primary System Pressure

Figure 8 RELAP Nodalization for the Single PEHRS Loop Table 3 Sequence of Trips for the LOFW Transient

Event # Event description Set point

values 1 steady state at full power 0 s up to

1000-s

2 (instantaneous) feed water closure

1000+s

3 (instantaneous) scram 1000 +10 s

4 instantaneous isolation valves closure (feedwater + steam lines)

1000 + 15 s

5 instantaneous PEHRS isolation valve opening

1000 + 20 s

6 end of simulation 6000 s

The analysis of the LOFW accident starts with a period of

1500 s to reach steady state conditions prior to the initiating event and 5000 s of transient calculation after the initiating event.

The accidents starts with a sudden stop of the feed water flow, which is a conservative assumption because no flow inertia is taken into account. A 10 s delay in the scram activation is assumed. A subsequent 5 s delay is assumed in the isolation action for the steam and feed lines. A further delay of 5s occurs before the PEHRS trains enter operation. The global system behavior leads to a heavily dumped oscillation of the primary fluid temperature early in the transient (Figure 9), always well under the nominal values, with the primary flow rate maintaining its value because of the primary pumps. The PEHRS shows the persistence of a significant ∆T across the steam generator in the secondary side throughout the transients. This is consistent with a stable natural circulation flow rate and determines the removal of the thermal power from the RPV, which is rejected to the pool (Figure 10). The power rejected is well above the decay power produced by the core, and the pool water starts boiling well after 1 hour in the transient.

The large amount of heat extracted by the passive trains depressurizes dramatically the primary system, without intervention of the ADS. The PEHRS shows the characteristic sliding pressure behavior (Figure 11), following the temperature history of the primary fluid.

The Passive System fulfills its mission, maintaining the rejected power well above the decay power. The flow rate in the steam generator modules coupled in parallel to the PEHRS, shows an oscillating behavior that deserves to be investigated, in order to identify whether this is due to simulation effects (e.g., numerical stability) or thermal-hydraulic instabilities.

Future activities will include investigation of the coupled thermal-hydraulic behavior of parallel tubes in a single steam generator module and of coupled steam generator modules in parallel. Parametric analyses for the PEHRS dimensioning will be performed according to the results of the PSA analyses.


100

150

200

250

300

350

0 1000 2000 3000 4000 5000 6000Time (s)

Pri

mar

y T

emp

erat

ure

s (K

)

tempf 110010000 tempf 110200000

Steady-state

Figure 9 LOFW: Primary Fluid Temperatures – core inlet

(blue), core outlet (red).

-200

0

200

400

600

800

1000

1200

0 1000 2000 3000 4000 5000 6000

Time (s)

Po

wer

(M

W)

Total SG heat exchange (MW)Total Pool heat exchange (MW)

Total core power (MW)

Steady-state

Figure 10 LOFW: Thermal Powers – Decay (red), Extracted

(blue), Rejected (green).

0.E+0

1.E+6

2.E+6

3.E+6

4.E+6

5.E+6

6.E+6

7.E+6

8.E+6

9.E+6

1.E+7

0 1000 2000 3000 4000 5000 6000

Time (s)

Pre

ssu

re (

Pa)

p 657010000

Steady-state

Figure 11 LOFW: Secondary Pressure Inside PEHRS Circuit

Loss Of Coolant Accident (LOCA) The unique IRIS approach to deal with small-to-medium LOCAs via the thermohydraulic coupling of the vessel and

containment was discussed in ref. 3. Even though RELAP does not model the containment and therefore cannot accurately predict the IRIS response to a LOCA, a RELAP preliminary analysis of a LOCA was performed for the following reasons: - To provide a comparison point with other reactors analyses,

since RELAP is the “classical” code used for LOCA analyses;

- To confirm the results under a set of conservative assumptions (i.e., neglecting the transient increase in the containment back pressure);

- To investigate separately the effect of the IRIS integral vessel configuration (internal heat removal, large water inventory) and to assess its effectiveness even in the absence of the coupling to the containment.

The RELAP model adopted a detailed nodalization for the

PEHRS and a simplified nodalization for the RPV. For the RPV, half of the volumes were adopted with respect to the model reported in Figure 4. The circuit scheme for the passive safety system to be simulated is the same shown in Figure 8. The nodalization includes single units for the PEHRS circuit, the Steam Generator and the in-pool Heat Exchanger, equivalent to 4 PEHRS trains, 8 SGs and 4 Heat Exchangers, respectively, for a total amount of 220 volumes. This is due to the assumption in the accident sequence that all the PEHRS trains are activated and functioning in the same way, without asymmetry in their behavior.

Other assumptions are: - Reactor scram is activated; - No piping heat losses are considered; - Complete pump trip occurs; - RPV simplified modeling is adopted; - No containment simulation is performed, the presence of

the containment is described by a steady value of 2 bar for the environment backpressure;

- The break size is 3 ½”, occurring at elevation +20 m from RPV bottom.

The accident sequence is reported in Table 4. The analysis of the LOCA includes 1500 s of steady state

calculation prior to the initiating event and 8500 s of transient calculation after the initiating event. Both the reactor primary coolant and the feedwater pumps are assumed to trip upon occurrence of the break opening in the RPV. A 10 s delay is assumed before the scram activation and the feed and steam line isolation. A further delay of 5 s is supposed to occur before the PEHRS activation.

The integral mass discharged through the break and the fluid remaining in the RPV are shown in Figure 12. The water mass still stored in the RPV at the end of the transient is sufficient to avoid fuel damage. The assumption of 2 bar, steady environment back-pressure in the containment is very conservative: detailed analysis on the containment shows that the pressure remains largely over this value for the first 10 hours in the transient.


Table 4 LOCA Simulated Reference

Event # Event Description Set point

Values Additional

Information 1 steady state at full power 0 s up to

1500-s

2 Reactor coolant pumps and feed water closure;

break opening

1500+s

3 (instantaneous) scram, feed and steam lines

isolation

1500 + 10 s

4 PEHRS isolation valve opening

1500 + 15 s

5 end of simulation 10000 s

The containment

backpressure is maintained at a steady

value (2 bar)

The pressures in the RPV is shown in Figure 13. The

passive loop, starting from the nominal pressure condition of the secondary steam and feed lines, adapts very rapidly to the primary pressure in a characteristic “sliding pressure” behavior. This is driven by the heat exchange capacity of the passive loop, which is determined by the natural circulation flow rate, the primary temperature and the pool temperature.

The time profiles for the inlet-outlet temperatures for the core and for the pool are reported in Figures 14 and 15. The primary and secondary temperatures are consistent with the systems pressure: they show that a critical heating of the core does not occur. The pool is still far from boiling.

The thermal power generated from the core (decay heat), extracted from the RPV by the PEHRS and rejected to the pool are depicted in Figure 16. The passive system is capable of rejecting a thermal power well in excess of the decay value. An oscillating behavior for the primary system is expected, always with a net negative (i.e., rejected) power exchanged with the passive loop. Even the PEHRS behaves normally in an oscillating way, with the same feature of a net negative power rejected to the pool.

Future analyses will include: - Investigation of a small bypass between the riser and the

steam generator at mid elevation to evaluate the effectiveness of an alternative flow path for the fluid, when the liquid level falls under the riser top end;

- Consideration of the effect of the heat structures and their thermal inertia on the PEHRS behavior;

- A parametric investigation of break size and break elevation, to validate a previous investigation carried out by means of simplified models. The RELAP analysis will also be coupled with the

containment investigation results, in order to correctly simulate the environment backpressure.

0.0E+0

0.5E+5

1.0E+5

1.5E+5

2.0E+5

2.5E+5

3.0E+5

3.5E+5

4.0E+5

0 2000

4000 6000

8000 10000

Time [s]

Mas

s [k

g]

Steady - state

Figure 12 LOCA: RPV Stored (blue) and Discharged

(red) Mass

0.0E+0

4.0E+6

8.0E+6

1.2E+7

2.0E+7

1.6E+7

0 2000

4000 6000

8000 10000

Time [s]

Pre

ssu

re [

Pa]

Steady - state

Figure 13 LOCA: RPV Pressure

Figure 14 LOCA: Core Inlet (blue) and Core Outlet (red) Temperatures

350

400

450

500

550

600

650

0 2000

4000 6000

8000 1000

0 Time [s]

Tem

per

atu

re [

K]

Steady - state


Figure 15 LOCA: Pool Temperature

Figure 16 LOCA: Power exchanged by Core (red), Steam

Generators (blue) and in-pool Condenser (green)

CONCLUSIONS Notwithstanding the preliminary level of the analysis, it can

be stated that the inherent safety feature of the IRIS design and the Passive Safety System (PEHRS) fulfill their mission.

The higher degree of natural circulation for the primary flow achievable by the integral reactor, in respect to that obtainable by current loop-type PWRs, downgrades the severity of the accidents.

The IRIS design is not susceptible to damage during the postulated accidents analyzed here, since IRIS safety by design provides means to remove core decay heat to the environment, maintain primary system water inventory for core cooling and provide necessary containment cooling. These functions occur with no need for AC-power or other support systems.

Further development in the design of the passive system will refer mainly to the in-pool condenser component and to the optimization of the heat transfer capabilities, in order to reach a more uniform behavior in the heat extraction from the RPV and rejection to the heat sink.

250

270

290

310

330

350

370

0 2000 4000 6000 8000 10000

Time [s]

Tem

per

atu

re [

K]

Steady-state

-200

0

200

400

600

800

1000

0 2000 4000 6000 8000 10000

Time [s]

Po

wer

[M

W]

Steady-state

Detailed dynamic analyses for the SG modules and the in-pool condenser will be performed, to avoid two-phase or single phase flow instabilities in the whole range of operation of the PEHRS. ACKNOWLEDGEMENTS

The financial support of the IRIS program (NERI Grant DE-FG03-99SF21901) by the US DOE is gratefully acknowledged. Contributions by the IRIS consortium members were invaluable to the progress of the project.

REFERENCES

[1] Carelli, M.D., et al, “IRIS Reactor Development,” Proceedings of the 9th International Conference on Nuclear Engineering, Nice, France, April 8-12, 2001.

[2] Carelli, M. D., L. E. Conway, B. Petrovic, D. V.

Paramonov, M. Galvin, N. E. Todreas, C. V. Lombardi, F. Maldari, M. E. Ricotti and L. Cinotti, “IRIS Reactor Conceptual Design,” Intl. Conf. On the Back-End of the Fuel Cycle (Global 2001), Paris, France, September 9-13, 2001.

[3] Conway, L. E., et al, “Simplified Safety and Containment

Systems for the IRIS Reactor,” Proceedings of the 9th International Conference on Nuclear Engineering, Nice, France, April 8-12, 2001.

[4] Carelli, M. D., et al, “Safety by Design: A New Approach

to Accident Management in the IRIS Reactor,” IAEA International Seminar on Status and Prospects for Small and Medium sized Reactors, Cairo, Egypt, May 27-31, 2001.

[5] Mizuno, Y., K. Ogura, H. Ninokata, L. Conway,

“Preliminary Probabilistic Safety Assessment of the IRIS Plant”, ICONE-10 Conf., paper #22095.

[6] Kujawski, J. M., D. M. Kitch, L. E. Conway, “The IRIS

Spool-Type Reactor Coolant Pump,” ICONE-10 Conf., paper #22572.

[7] L. Cinotti, M. Bruzzone, N. Meda, G. Corsini, C.

Lombardi, M. Ricotti, L. E. Conway, “Steam Generator of the International Reactor Innovative and Secure”, ICONE-10 Conf., paper #22570.


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Preliminary Safety Analysis for the IRIS Reactor

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