Relevant thermalhydraulic aspects of new generation LWRs

Nuclear Engineering and Design 145 (1993) 241-259 North-Holland

241

Relevant thermalhydraulic aspects of new generation LWRs F. D ' A u r i a a, M. M o d r o b, F. O r i o l o a a n d K. T a s a k a c "Dept. of Mech. and Nuclear Constructions, Unic'. of Pisa, Via Diotisah,i 2, 56100 Pisa, Italy h INEL, P.O. Box 1625, Idaho Falls, 119 83415-2404, USA c Dept. of Nuclear Eng., Unit. of Nagoya, Furo-cho, Chikusa-Ku, Nagoya-464, Japan

Received 1 April 1993

The present paper deals with the evaluation of thermalhydraulic aspects retained of importance for the assessment of safety of the new generation nuclear plants. Following a survey of the reactor concepts proposed for the future, the attention will be focused toward SBWR, AP-600 and PIUS whose characteristics, under many respects, bound the features introduced in the largest part of the new reactors. Expected relevant phenomena typical of the mentioned plants will be discussed in the paper: on this basis a critical overview of the experimental activities planned or in progress is presented and a judgement about the suitability of available computer codes is formulated. Conclusions are drawn in relation to the assessment of the new design proposals from a thermalhydraulic point of view.

I. Introduction

The Light W a t e r Reactors ( L W R ) are considered primary nuclear reactors for power generation in the first half of 21st century. Nevertheless, the safety of the systems must be continuously improved to cope with the increase in the capacity of the electric power generated by the nuclear source in developing countries as well as in developed countries and to preserve the environment on the earth at the same time. In this context, the public demand for improved safety and the utilities request for increased economic profit, also looking at investment cost, led the vendors to the proposal of advanced reactors. In the domain of LWR, two main classes can be distinguished: - evolutionary reactors (e.g. Advanced B W R , System 80 + PWR, etc); - innovative reactors (e.g. Simplified B W R , Advanced PWR-600, PIUS, etc.). In the former class, the systems and the functions of the present generation reactors have been fully optimized on the basis of the operational experience and the current technology. A substantial reduction of the core melt frequency together with a reduction of the investment cost obtained through the simplification of the whole system, 0029-5493/93/$06.00

constitute the strategic objective of the second class of reactors. Current technology shall be used in the design that will considerably shorten construction schedule. Extensive use of passive systems, especially for residual heat removal purposes, is made in this category of reactors. The core melt frequency reduction can be estimated in the range 10-100 and significant improvement of economic performance either looking at the kilowatthour cost or at the investment, are foreseen in both classes of reactors. From a thermalhydraulic point of view nominal and off-normal behaviour of the advanced reactors, with main regard to the above mentioned second class, may involve the occurrence of p h e n o m e n a that are either of low importance or not present at all in the current generation LWR. Following a review of the main peculiarities of the considered reactors, the main purpose of the present paper is to identify new thermalhydraulic p h e n o m e n a that are expected to play a role in the operation and in the safety of advanced reactors. On this basis, considering an overview of the experimental activities in progress or planned and the experience gained in the application of available system codes to the prediction of advanced reactors transient scenarios, the need for

© 1993 - E l s e v i e r S c i e n c e P u b l i s h e r s B.V. A l l rights r e s e r v e d

20

19

10 11 12 13 14 15 16 17 18

9

1 2 3 4 5 6 7 8

Thermal power Numer of external loops Vessel height Vessel diameter Reactor pressure Core inlet flow Core inlet temperature Core f l o w / thermal power Core inlet p o w e r / thermal power Number of fuel rods Core active height Rod diameter Linear power-average Linear power-maximum Volumetric power Primary fluid volume Primary fluid mass Thermal p o w e r / fluid mass Number of tubes per steam generator Number of primary pumps

-

-

kW/kg

m mm kW/m kW/m MW/m 3 m3 kg. 103

-

k g / s MWt

MWt m m MPa kg/s K

2 P WR W

4.87

4 PWR KWU

4.73

4.99

2785 3765 3 4 12.3 12.3 3.98 5 15.5 15.8 13200 18000 559

3 PWR FRA

2

4674-V

16.7

182 112

4

3382-V

347

3

274

4

6.78 7.03 28435 50952 47100 57900 3.66 3.66 3.66 3.9 9.5 9.5 9.5 9.5 18.1 18.3 16.2 16.7

4.55

1986 3423 2 4 11.2 3.69 15.5 16.0 8550 16700 560

1 PWR W

3.70

3014 2 21 5.6 7.1 11160

6 BWR-6 GE

-H 4

2

-

15.7

192

10.5 50856 38688 3.53 3.8 9.1 12.3 16.7 20.5 45.0

6.27

3000 4 10.9 4.1 15.7 18800 595

5 VVER USSR

10

-

50.5

54064

3.67

3926 0 21.0 7.1 7.3 14440

7 ABWR GE

Table 1 Relevant characteristics of present and new generation reactors (evolutionary and innovative types)

4

56876 3.81 9.7 17.5

3800 2 15.3 4.62

8 PWR-80+ CENP

-V 4

45

9.1

12.4 78683

7.45

3300 4 11.7 5.4 15.7 24600 595

9 VVER USSR

4

6307-V

7.13

6.8 38280 3.66 9.5 13.4 32.2 78.8 340 272

4.73

1940 2 11.5 4.0 15.5 9185 549

12 SBWR GE

2

6.5

9 13000 533

2000 4 45

13 PlUS ABB

0

-

41.9

4

0.7

72.0 3300 3000

5.53 9.0 43920 67308 2.74 2.5 12.3 16.6 11.9

3.83

1825 2000 2 0 15.7 24.6 6.0 7.1 7677 551

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F. D'Auria et al. / Relevant therrnalhydraulic aspects new experiments is evaluated and a judgement is given concerning the suitability of available computer models.

2. Design features of considered reactors In order to evaluate the thermathydraulic performance of new reactors especially looking at the possible different accident scenarios compared to the present generation reactors, two broad investigation areas can be considered: (1) comparison of the characteristic values assumed by important nominal conditions; (2) evaluation of hardware characteristics and of the intervention modalities of engineered safety features that affect the evolution of the off-normal transients. Relevant characteristics (investigation area Nr. 1) of present generation and next generation reactor designs are summarized in Table 1 *. The considered thermalhydraulic parameters (e.g. items 6 to 9, 13, etc.) give an idea of the differences among the various concepts and of the differences expectable during transient conditions. In the vertical columns, reactors 1 to 6 are examples of the present generation, reactors 7 to 9 are part of the evolutionary class and the remaining ones belong to the previously defined innovative category. It should be noted that the selection of the reactors in the table has been done considering the availability of data to the authors and the attempt to differentiate the various classes. Review of the data shows that the majority of reported parameters are essentially the same for the current generation reactors and the evolutionary reactors. However in the case of the innovative reactors some values are quite different, such as the core linear power (parameter 13 in Table 1), the core inlet power (obtained by multiplying flowrate and enthalpy) over thermal power (9), the thermal power density (18). These parameters have values quite smaller for the innovative reactors than for the present generation

* An important class of future LWR has not been included in Table l: the high converter reactors (both PWR and BWR). These are characterized by very tight lattice, by a ratio moderator over fuel volume four times lower than current reactors and by a fuel burn up of the order of 100 GWd/ton. The design appears in a earlier stage with respect to the others, constituting the main reason for not including these reactors in the table.

243

reactors: the new values apparently shift the nominal conditions in a direction toward the increase of safety margins. Evaluation of the engineered safety systems shows that these are still very similar for the current generation reactors and the innovative ones. However, the extensive use of passive systems in innovative reactors (also resulting from the desire to avoid the intervention of operators for extended periods - tens of hours) led to the need of founding the reactor safety upon the availability of large heat sinks inside the containment. As a consequence, almost all the transients are forced, by automatic means, into low pressure long term cooling processes where the interaction between the primary circuit and the containment system becomes important. For this reason and because the passive safety systems are highly interconnected to the primary circuit with feedback capabilities of the primary system on their performance, new accident scenarios are foreseeable that must be carefully studied. In conclusion, only innovative reactors need extensive R & D studies before their commercialization. A detailed investigation of the various thermalhydraulic situations requiring specific research plans in each of the innovative reactors is beyond the purpose of the paper. So only three reactors will be considered in some details: SBWR, AP-600 and PIUS. For completness a brief description of the main features of these reactors is given below. 2.1. S B W R features A sketch of the S B W R is given in Fig. 1 that includes the primary system (up to the main isolation valves), the containment and most of the safety features of the plant. The S B W R is the result of an extensive simplification of the existing B W R plant, the major simplification being the elimination of any pump to recirculate the coolant inside the vessel [1]. The use of natural circulation results in a simple system with less piping and less reactor vessel penetrations below top of the core. Simplifications have been introduced at several levels in the plant including the conventional part and the control room. The containment and the systems for the handling of operational transients and accidents, are described with some detail hereafter. The containment is based on the principle of pressure suppression through the condensation of primary steam or two phase mixture in a devoted pool in the same way as current generation reactors. The core

F. D'Auria et al. / Releuant thermalhydraulic aspects

244 IC

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--GDCCS

POOL

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ACC C R D POOL -

Fig. 1. Sketch of SBWR.

position in the vessel is at lower elevation than in present BWR increasing the margin to core uncovery. The basic feature of the SBWR plant is the complete integration of the containment and of the safety systems. In the majority of the foreseeable accident conditions special depressurization valves (DPV) allow the equalizations of containment and vessel pressure. This makes possible the injection of liquid into the vessel by gravity from a special pool situated in the

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upper part of the containment (Gravity Driven Cooling System, GDCS). Additionally, in case of reactor isolation transients, the steam produced in the core is driven (whatever is the presure) into a heat exchanger that transfers the thermal power to another pool on the top of the containment. The steam condensed in this heat exchanger flows back by gravity into the reactor vessel. This system is called Isolation Condenser (IC). Additional condensers are used to remove energy from the containment drywell (Passive Containment Cooling Systems, PCCS). These condensers are located in similar pools as the IC on top of the containment. PCCS is equipped with special system that allows purging of the noncondensable gases from the condensers into the suppression pool space. The condensate from the PCCS flows by gravity into the GDCS pools. The PCCS as well as the IC and GDCS do not require any active components (e.g. pumps). The use of these systems allows also for a relatively small containment. 2.2. AP-600 features

A sketch of AP-600 is given in Fig. 2 that includes the primary system, the containment and most of the safety features of the plant. Even in the case of AP-600, the essential technical concept underlying the design is simplification [2], [3] that also allows a modular construction approach. In

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245

F. D'Auria et al. / Releeant thermalhydraulic aspects

order to evaluate the transient thermalhydraulic behaviour, three areas can be distinguished that are different from the corresponding ones in current PWR: primary circuit, containment and engineered safety features. Apart from those outlined in Table 1, the differences in primary circuit concern the elimination of the loop seals; two canned rotor pumps are directly connected with the outlet plena of each steam generator and deliver the coolant to devoted cold legs (4 cold legs are part of the primary loop). Basically the containment has the same features as the current PWR. It is a full pressure steel vessel containment but, additionally it is equipped with special Passive Containment Cooling Systems (PCCS) that is sufficient to remove long term decay heat by cooling the external walls of the steel containment vessel. The cooling is provided by thin film evaporation of water passively injected on the containment walls and by natural circulation of air flow up along the containment walls. The passive emergency core cooling system consists of two Core Make Up Tanks (CMT), two large accumulators and an In-containment Refueling Water Storage Tank (IRWST). The ECC water is fed through two safety injection lines directly into the downcomer of the reactor vessel. In case of a non-LOCA accident, core cooling is provided via natural circulation Passive Residual Heat Removal (PRHR) system with a heat exchanger submerged in the IRWST. The CMT are located above the reactor coolant loop; pressure balancing lines, that open in case of accident, make possible the gravity injection of the borated liquid from the tanks to the vessel. An Automatic Depressurization System (ADS) is provided with valves connected to the pressurizer and the hot leg, that allows reduction of the RCS pressure to the containment pressure and therefore draining of the IRWST water into the RCS and establishing of long term cooling. This also makes easier the gravity driven reflood in case of core uncovely. As indicated, the coolant inventory control and the removal of decay heat during accidents in AP-600 will be achieved by using low head gravity draining or natural circulation. The low motivation forces and the multiple parallel circulation paths may lead to safety system performance disturbances due to phenomena interactions and equipment malfunction perturbances. 2.3. P l U S features

A sketch of PIUS is given in Fig. 3 from which the main features of the designs can be recognized. The

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concept of PIUS is "more" innovative if compared with the two previous designs [4], [5]. The primary coolant circulation system is immersed into a pool of cold borated water confined by a larger concrete vessel. The primary loop is connected to the borated pool at a lower and a upper elevation through so-called density locks. Within this density locks an interface between the hot primary coolant and the cold pool water is maintained via operation of the reactor coolant pumps. When the pressure drops of the flowing coolant inside the steel vessel equal the gravity head of the cold liquid in the concrete vessel, there is no flow from the large borated pool to the primary circuit. However, if a pressure balance disturbance should occur due to void formation in the core or reactor coolant pump speed drop, cold borated water from the pool will penetrate the primary system through the lower density lock leading to reactor shutdown. A natural circulation and core cooling will be established with water flowing through the lower density lock into the core, then rising through the core and the riser and exiting the primary circuit at the top of the riser through the upper density lock into the pool. Special natural circulation loops are active to keep cooled the borated water pool and are effective to remove the decay heat.

F. D'Auria et al. / Rele~)ant thermalhydraulic aspects

246

Table 2 R e l e v a n t t h e r m a l h y d r a u l i c p h e n o m e n a identified for t h e c u r r e n t g e n e r a t i o n r e a c t o r s 0 BASIC PHENOMENA

• E v a p o r a t i o n d u e to D e p r e s s u r i z a t i o n • E v a p o r a t i o n d u e to H e a t I n p u t • C o n d e n s a t i o n d u e to P r e s s u r i z a t i o n - C o n d e n s a t i o n d u e to H e a t R e m o v a l • I n t e r f a c i a l Friction Vertical Flow • I n t e r f a c i a l Friction H o r i z o n t a l Flow • Wall to Fluid Friction • P r e s s u r e D r o p s at G e o m e t r i c D i s c o n t i n u i t i e s •Pressure wave propagation -Breaks L / D < 1 • Valves •Pipes

1 CRITICAL FLOW

2 PHASE SEPARATION/VERTICAL

FLOW

WITH AND WITHOUT MIXTURE LEVEL

-P i p e s / P l e n a •C o r e •O o w n c o m e r

3 STRATIFICATION HORIZ. FLOW

•Pipes

4 PHASE SEPARATION AT BRANCHES

•B r a n c h e s

5 ENTRAINMENT/DEENTRAINMENT

•C o r e •U p p e r P l e n u m •O o w n c o m e r -SG-Tube SG-Mixing C h a m b e r (PWR) • H o t leg with L O C I ( P W R ) •

6 LIQUID-VAPOUR CONDENSATION

MIXING WITH

-Core •D o w n c o m e r •U p p e r P l e n u m •L o w e r P l e n u m SG-Mixing C h a m b e r (PWR) •

• E C C I in H o t a n d Cold L e g ( P W R ) 7 CONDENSATION

IN STRATIFIED

CONDITIONS

•P r e s s u r i z e r ( P W R ) S G - P r i m a r y Side ( P W R ) S G - S e c o n d a r y Side ( P W R ) • H o r i z o n t a l Pipes •

•

8 SPRAY EFFECTS

•C o r e ( B W R ) •P r e s s u r i z e r ( P W R ) • O T S G S e c o n d a r y Side ( P W R )

9CCF/CCFL

•U p p e r T i e Plate • C h a n n e l I n l e t Orifices ( B W R ) • H o t a n d Cold L e g •S G T u b e ( P W R ) •D o w n c o m e r •S u r g e l i n e ( P W R )

10 G L O B A L M U L T I D I M E N S I O N A L F L U I D TEMPERATURE, VOID AND FLOW DISTRIBUTION

•U p p e r P l e n u m •C o r e •D o w n c o m e r • S G - S e c o n d a r y Side

F. D'Auria et al. / Relevant thermalhydraulic aspects

247

Table 2 (continued) 11 HEAT TRANSFER NATURAL FORCED CONVECTION SUBCOOLED/SATURATED/NUCLEATE BOILING DNB/DRYOUT POST CHF HT RADIATION CONDENSATION

•Core, SG, Structures •Core, SG, Structures •Core, SG, Structures •Core, SG, Structures •Core •SG, Structures

12 QUENCH FRONT PROPAGATION/REWET

•Fuel Rods •Channel Walls and Water Rods (BWR)

13 LOWER PLENUM FLASHING 14 GUIDE TUBE FLASHING (BWR) 15 ONE AND TWO PHASE IMPELLER-PUMP BEHAVIOUR 16 ONE AND TWO PHASE JET-PUMP BEHAVIOUR (BWR) 17 SEPARATOR BEHAVIOUR 18 STEAM DRYER BEHAVIOUR 19 ACCUMULATOR BEHAVIOUR (PWR) 20 LOOP SEAL FILLING AND CLEARANCE (PWR) 21 ECC BYPASS/DC PENETRATION 22 PARALLEL CHANNEL EFFECTS INSTABILITIES (BWR) 23 BORON MIXING AND TRANSPORT 24 NONCONDENSIBLE GAS EFFECT (PWR) 25 LOWER PLENUM CCF

It should be noted that the safety of P I U S does not depend on specific engineered safety systems but is ensured by features typical of the design itself.

3. Expected relevant phenomena Thermalhydraulic p h e n o m e n a relevant to the evolutionary type nuclear plant can be considered the same as those valid for the current generation reactors. A suitable review of applicable p h e n o m e n a can be found in CSNI reports 132 [6] and 161 [7] and in N R C report 1230 [8]. For completness the list is reported in Table 2. Limited specific research activity in this area appears necessary, if one excludes new domains like Accident M a n a g e m e n t and special topics like instability in boiling channels where the interest is common to the present generation reactors. In the case of innovative reactors the foreseeable relevant thermalhydraulic p h e n o m e n a can be grouped into two categories [9]: (a) p h e n o m e n a that are relevant also to the present generation reactors; (b) new kinds of p h e n o m e n a a n d / o r scenarios. For the category (a) the same considerations apply as for the evolutionary reactors and the p h e n o m e n a of concern are therefore well documented in refs. [6]-[8]. However, it has to be noted that significance of various

p h e n o m e n a may be different for the innovative reactors. Nevertheless, it is believed that the data base, understanding and modelling capabilities acquired for the current reactors are adequate for p h e n o m e n a in category (a) in the sense defined in refs. [6-8]. P h e n o m e n a of the category (b) can be subdivided into three classes: (bl) p h e n o m e n a related to the containment processes and interactions with the reactor coolant system, (b2) low pressure phenomena, (b3) p h e n o m e n a related specifically to new components, systems or reactor configurations. In current generation reactors the thermalhydraulic behaviour of the containment system and of the primary system are studied separately. This is not any more possible in most of the new design concepts; suitable tools must be developed to predict the performance of the integrated system. A peculiarity common to almost all the innovative reactors is the presence of devices that depressurize the primary loop essentially to allow the exploitation of large amounts of liquid at atmospheric pressure and to minimize the risk of high pressure core melt: in this case the p h e n o m e n a may be similar (or the same) as those reported for present generation reactors (Table 2) but the range of parameters and the safety relevance can be much different. Finally, the presence of new systems or components and some geometric peculiarities of innovative reactors

248

F. D'Auria et a L / Relet~ant thermalhydraulic aspects

Table 3 Relevant thermalhydraulic p h e n o m e n a of interest in the innovative reactors Phenomena occurring due the interaction between primary system and containment 1. Behaviour of large pools of liquid: thermal stratification - n a t u r a l / f o r c e d convection - s t e a m condensation (e.g. chugging, etc.) heat and mass transfer at the upper interface (e.g. vaporization) liquid draining from small openings (steam and gas transport) 2. Tracking of non-condensibles (essentially, H2, N2, air): -effect on mixture-to-wall heat transfer coefficient -mixing with liquid phase -mixing with steam phase -stratification in large volumes at very low velocities 3. Condensation on the containment structures: -coupling with conduction in larger structures 4. Behaviour of containment emergency systems (PCCS, external air cooling, etc.): - interaction with primary cooling loops 5. Thermofluid-dynamics and pressure drops in various geometrical configurations: -3-D flow paths around open doors, connection of big pipes with pools, etc. - g a s / l i q u i d phase separation at low Re and in laminar flow -local pressure drops -

-

-

Phenomena occurring at atmospheric pressure 6. Natural circulation:

among parallel circulation loops inside and outside the vessel - Influence of non-condensables - i n t e r a c t i o n

7. Steam liquid interaction: direct condensation -pressure waves due to condensation -

8. Gravity driven reflood: heat transfer coefficients - pressure rise due to vaporization -consideration of a closed loop -

9. Liquid temperature stratification: - lower plenum of vessel - downcomer of vessel horizontal/vertical piping -

Phenomena originated by the presence of new components and systems or special reactor configurations 10. Behaviour of density locks: -stability of the single interface (temperature and density distribution) between two density locks 11. Behaviour of check valves: - o p e n i n g / c l o s u r e dynamics partial/total failure 12. Critical and supercritical flow in discharge pipes: -shock waves - supercritical flow in long pipes -behaviour of multiple critical sections 13. Behaviour of Isolation Condenser 14. Stratification of boron: between chemical and thermohydraulic problems -time delay for the boron to become effective in the core - i n t e r a c t i o n

-

- i n t e r a c t i o n

F. D'Auria et al. / Relevant thermalhydraulic aspects require the evaluation of additional scenarios and phenomena. A list of identified phenomena belonging to subclasses bl, b2 and b3 is given in Table 3; comments related to few of these are reported below.

Behaviour of large pools of liquid (item 1 in Table 3) Large pools may have a very wide spectrum of geometric configurations. Heat transfer in one very limited zone in terms of volume (e.g. by condensing steam or by isolation condenser) does not imply homogeneous or nearly homogeneous temperature in the pool. Three-dimensional convection flows develop affecting the heat transfer process. Liquid drain from relatively small openings may cause rotation of the fluid and entrainment of the gas phase (vortex formation).

Tracking of non-condensables (item 2 in Table 3) The non-condensable gases play much more important role in safety evaluation of the innovative reactors than for the current generation reactors, particularly because of the coupling of the containment with primary system. Flow of the non-condensables within the containment and potential stratification and separation of steam constitute an important factor to consider since it may affect heat transfer within the containment and out of the containment Also, within the primary system the nitrogen transport from accumulators may affect the coolant distribution and pressure response to the degree that gravity driven safety injection may be impacted. Non-condensable gases are also an important factor in efficiency of isolation condensers and the condensing heat exchangers.

Thermofluiddynamics and pressure drops in various geometrical configurations (item 5 in Table 3) Owing to the lack of pumping power the circulation among the various zones of the system constituted by primary circuit and containment depends upon relatively small driving forces. The presence of obstacles like bends, valves, etc., that have no relevance when pumps are running, can be important for the evolution of the phenomena. Various mechanisms of phase separation at very small Reynold number may interfere with the establishment of flowrates.

Natural circulation (item 6 in Table 3) Natural circulation is a complex phenomenon depending upon some other phenomena mentioned in Table 3. The interaction between multiple parallel flow paths may be critical especially in long lasting transients.

249

Gravity driven reflood (item 8 in Table 3) Reflood has been widely considered in safety studies related to the present generation reactors. Additional aspects of interest in the innovative reactors are the presence of feedback between the velocity of the quench front, the pressure rise due to vaporization (at the quench front) and the condensation of steam possibly in the same tank supplying liquid for reflood.

Behaviour of density locks (item 10 in Table 3) The stability of the interface of the density locks, especially when two density locks are present, appears a critical aspect; possible long term variations of static head in the pool (e.g. due to stratification or heating) may change the interface position in each density lock and the stability characteristics.

Behaviour of check valves (item 11 in Table 3) Check valves connect the primary circuit with very large volumes through large pipes. Conditions in the piping with check valves may arise that cause rapid condensation on one side resulting in slam close of the valve. The small driving forces may not reopen the valves again. The failure to open may prevent the coolant to flow into the primary circuit; the failure to close may cause fast draining of primary circuit; opening and closure cycles cause critical oscillations in the flow rates.

4. Overview of experimental activities in progress

Experimental activities carried out, in progress or planned for the three considered reactor types have been reviewed. Few results available from the literature are summarized in the following three sections. Only preliminary conclusions can be drawn about the consistency between the researches objectives and the relevant phenomena listed in Table 3.

4.1. SBWR experimental activities Examples of experimental activities related to SBWR are documented in refs. [10] to [15].

Toshiba activities The main purpose of the facility constructed and operated by Toshiba is the real time simulation of long term transient occurring in the SBWR [10], [11]. The test rig (sketch in Fig. 4) has a volume scaling ratio of 1/400, a height scaling ratio of 1/1, operating pressure of around 0.3 MPa. These tests were designed to evaluate the condensation heat transfer in the PCCS condensers in presence

250

F. D'Auria et al.

®

/

Relecant thermalhydraulic aspects

FLOW R A T E

? Fig. 4. Flowsheet of IC-Toshiba test facility.

result, it was found that IC had sufficient heat removal capacity even in presence of nitrogen and that the degradation in heat transfer coefficient for forced condensation was much milder than the values foreseen for stagnant condensation. A wide data base has been measured in the above facility in relation to the overall system response in case of an accident. The process is as follows in case of a steam line break: the steam coming from the vessel directly reaches the drywell where it mixes with nitrogen. The steam-nitrogen mixture reaches the IC where steam is condensed (going back to the vessel) and nitrogen causes pressure increase up to clear special vent lines connecting IC with the suppression pool. At this time nitrogen vents to the suppression pool chamber increasing again the heat removal capability of the IC. Suppression pool and drywell are also connected by direct nitrogen vent lines and by vacuum breakers. Vent submergence of nitrogen vent lines is such to maintain the flow from drywell to IC to suppression pool up to limit values of nitrogen concentration; vacuum breakers opening cause direct flow from suppression pool to drywell. Experimental data related to the operation of the whole system essentially confirm the adequacy of the nitrogen venting mechanism [11] and suggest that there is the possibility to optimize the system response by varying the relative submergences inside the suppression pool of the lines connecting this zone with IC and with the drywell. Acticities at P S I a n d S I E T

of non-condensables. Also these tests provided data on non-condensable purging. The PCCS condensers are connected to the containment drywell and during steam release to the drywell (due to steam line break or action of the automatic depressurization system) steam and containment nitrogen flow into the condensers. Presence of noncondensable gas reduces significantly efficiency of the condensers. The design calls for a purging system that removes the nitrogen from the condenser to the suppression pool. The degradation of the heat transfer coefficient as a function of nitrogen concentration is shown in Fig. 5. Local and overall values of heat transfer coefficients are considered: the differences between the two, increase when nitrogen concentration increases and is connected with the larger area needed for the condensation. In the same figure, the prediction obtained by the Sparrow model [10] is also reported; in this connection it should be noted that the Sparrow derivation does not account for system effects. So the "conservatism" in the prediction should be expected. As a main

Construction of a large facility (1/24) to evaluate the SBWR containment behavior is planned at PSI [12]. This facility will provide for scaled suppression pool, drywell, and IC and PCCS condensers pools. Interactions between individual compartments of the

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\~ "" -.

0.2 0

o

0.02

~

0.04

/ ~"1 |

o •

TEST (LOCAL] TEST (OVERALL] Sparrow (Convective) Sparrow (Stagnant)

0.06

0.08

0.1

0.12

PNZ / P tot

Fig. 5. Heat transfer degradation coefficient measured in Toshiba facility as a function of nitrogen partial pressure.

251

F. D'Auria et al. / Rele~'ant thermalhydraulic aspects

containment and phenomena such as steam noncondensable stratification will be studied additionally to the PCCS performance. Also aerosol pool scrubbing phenomena will be investigated. Two isolation condensers are now in the final stage of construction at SIET [13], one at low pressure for testing the passive containment cooling and the other one at high pressure for measuring the heat removal capability from the primary circuit. The available power is up to 20 MW and design pressure is 10 MPa for the facility simulating primary circuit behaviour.

;..-.

Use of PIPER-ONE facility Piper-one simulates a BWR-6 with volume and height scaling ratios of 1/2200 and 1/1 respectively and is installed at DCMN of University of Pisa. The electrical power supplied to the core rods is, roughly 25% of the ideally scaled value and is sufficient to simulate long lasting transients with scram. The facility has been modified two times with the introduction of the Gravity Driven Cooling System GDCS [14] and the IC [15], respectively. In both cases the vessel configuration remained the same simulating the BWR-6; so the experimental data can be used for the evaluation of the qualitative performance of the two added systems (and obviously for code validation purposes). Three experiments were performed in the GDCS configuration (Fig. 6) and demonstrated a huge influence of the boundary conditions and of pressure drops upon the involved phenomena: in test PO-SD-6A (break and GDCS pool connected to the atmosphere) the pressure rise at the quench front prevented core reflood (Fig. 7); in test PO-SD-6C (break discharging into GDCS pool), GDCS was effective in quenching the core (Fig. 8). In the IC configuration [15] a pipe was added at the top of the main vessel carrying the steam into a heat exchanger placed inside a pool at atmospheric pressure. The bottom line of the heat exchanger primary side was connected with the lower plenum of the facility. Tests have been performed by varying the core power (up to 15% of the decay value) and system pressure (up to 5 MPa). The evaluation of experimental data is in progress. As a preliminary result, strong temperature stratification was observed inside the IC pool, where the heat exchanger was purposely put in the upper part. The pool was boiling at the top with temperatures as low as 293 K in the bottom.

4.2. AP-600 experimental actiuities

j ZERO

LP B O T T O M

LEVEL

T Fig. 6. Sketch of PIPER-ONE (GDCS configuration).

800 T

(K) 700

., Le

600 / 5OO

., L~W, ~

!

W (KW)

400

300 ~ -100

/

Heating p o w e r

~ 100 TIME

200

100

0 300

(s)

Fig. 7. Rod surface temperature trends and core power during

Westinghouse conducts a comprehensive testing program that covers all the new passive safety features.

PIPER-ONE GDCS experiment (pool at atmospheric pressure).

252

F. D 'Auria et al. / Releuant thermalhydraulic aspects

700

100

T [KI

Level G "

/

600

F

600

0

400

-J

~

'-

KW] 100

/~H~st'n9 p°wer~

20 0

.

~ 0

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1 ~ 0.05

I 0.1 FIow-kg/sec

._---~3oo F I 0.15

o

-100

0

100

200

300

TIME [s)

Fig. 8. Rod surface temperature trends and core power during PIPER-ONE GDCS experiment (pool connected with break line).

Only limited information about this program was published in the open literature. References [16] and [17] provide some overview of planned and performed AP600 testing activities. 4' ACCUMULATOR

I

350 KW HEATER

G I

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111 II,

I I~ TUSES II =

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--©

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Fig. 10. Westinghouse PRHR heat exchanger test: heat transfer versus PRHR tube flow (per tube basis; saturated tank conditions). A prototypic representation of a complete passive residual heat removal (PRHR) loop (Fig. 9) has been in operation by Westinghouse [16]. A series of tests at different temperatures in single phase flow demonstrated the suitability of the component in transferring to the IRWST adequate amount of thermal power. Typical data of heat removed per tube as a function of tube flow rate are shown in Fig. 10; the parameter on the right is the entrance temperature of primary fluid. As expected, higher primary fluid temperature and higher mass flow rates result in higher heat transfer. Another area of interest in the research deals with the problem of stratification and natural convection inside the pool. Roughly, 40 K have been found as the maximum temperature difference inside the pool with the largest part of the fluid volume at boiling conditions. Installation of a special baffle inside the pool and the proper position of the heat exchanger led to the optimization of the pool heat sink capacity. Plans for modifying the Spes facility available at SIET (PWR simulator with 1/427 volume and power scaling factor) have been defined into detail [17]. The new facility (called Spes-2) will simulate AP-600, with a volume scaling ratio given by 1/395. Now, design calculations have been performed to compute the predicted scenarios in AP-600 and in Spes-2. This led to a modification of volume scaling ratio of the core makeup tanks to better reproduce the predicted plant behaviour. 4.3. P l U S experimental actit,ities

TUBES

Fig. 9. Sketch of Westinghouse PRHR test facility.

The PIUS concept apparently contains more innovation with respect to the previous reactors; so more

253

F. D'Auria et al. / Relevant thermalhydraulic aspects

presence of hot water affects the pressure balance causing a net inflow into the core from the cold water tank through the lower density lock. The inflow of cold (borated) water is intermittent until the pump controller restores the interface in the lower density lock. A similar loss of feedwater transient was considered in the Japanese facility [20] assuming a different control of the primary circulation pumps. Relevant results are shown in Figs. 15 and 16. In this case the hot water inlet in primary circuit is balanced by the pump velocity increase up to the limit velocity (50 Hz) when pump trip occurs (after 2000 s in Fig. 15). This causes stable flow of cold water into the main loop through the lower density lock. The experimental research by Pind and Fredell [21] was focused on the evaluation of the transport processes in a single density lock. Mechanisms identified

researches can be expected. Example of experiment related to PIUS can be found in refs. [18] to [21]. Two facilities have been constructed in Sweden and in Japan and are shown in Figs. 11 and 12, respectively. These have essentially the same objectives of demonstrating the possibility of the reactor concept and to validate computer models. The reactor simulator in the Atle loop has the same height as the reactor in the real plant; the overall volume scale is 1/308. The Japanese simulator [19] is a low pressure scaled loop and comprises the various zones of the plant. Several transient scenarios have been measured in both test rigs. Typical transient data from Atle are shown in Figs. 13 and 14. Following a loss of heat sink, hot water arrives in the downcomer of the facility causing temperature increase in the core region (Fig. 13). The

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MAKEUP WATER I

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254

F. D'Auria et al. / Releeant thermalhydraulic aspects

COMPUTED

Primary

2

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were the boron mixing and diffusion at molecular level and the instability of the interface caused by natural convection flow in the cold pool. The first problem makes necessary a boron purification loop in the PIUS; the second problem confirmed the occurrence of "sloshing" oscillations, requiring full scale testing.

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Fig. 16. Japanese rig: system response in case of loss of feedwater.

F. D'Auria et al. / Relevant thermalhydraulic aspects

5. Suitability of computer codes in applications to innovative reactors

K [ ] Base Case 1200

A large number of calculations have been performed with LWR system codes (Relap5 Trac, Cathare, Thyde) and with specially developed codes (e.g. Rigel, Trip, Fumo, etc.). Examples of analyses documented in the literature can be found in refs. [22] to [27] related to the first group of codes and in refs. [28] to [30] related to the second group. Additional code applications can also be found in the previously listed references.

255

•

Sensitivity

1000

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5.1. AP-600 calculations

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1.2

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Fig. 18. AP-600 interfacing L O C A simulation: predicted trends

System codes are widely used at University of Pisa (e.g. [31],[32]); R e l a p 5 / m o d 2 code deficiences have been found in the application to transient analysis in both SBWR and AP-600 plants. Relevant results from the application to large break and interfacing LOCA (ILOCA) analyses related to the last reactor are given below [27]. It should be noted that an extensive analysis could reduce the discrepancies between the various sensitivity calculations that are presented below: a substantial help in this direction could be obtained with the release of new code versions (e.g. R e l a p 5 / M o d 3 ) and, especially following the analysis of the experimental data in the new integral facilities simulating AP-600 (e.g. SPES-2 and ROSA-V).

16

./I-I~

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A

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Position of sparger in the IRWST The reference ILOCA calculation (Fig. 17) was performed assuming that the sparger of the depressuration system in the IRWST pool is located near the top of the volume occupied by the liquid. In this case only partial condensation of the steam discharged from the primary system is observed; the remaining part of the steam contributes to the pressurization of the pool volume, allowing the drainage of the liquid from the pool towards the core. A second calculation was carried out assuming that the sparger is located in the bottom of the pool. In this case, a nearly complete steam condensation prevents the pool liquid from discharging and leads to extended core dryout (Fig. 18).

,:0L /

MPa Break Opens

of rod surface temperature (comparison between reference calculation and sensitivity study results).

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of primary system pressure and core level (reference calculation).

%.2 ; TIME

(s]

Fig. 19. AP-600 interfacing L O C A simulation: predicted trends of rod surface temperature (comparison between reference calculation and sensitivity study results).

F. D',4uria et al. / Rele~,ant thermalhydraulic aspects

256

5.2. Specific deficiencies of codes

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Fig. 20. AP-600 large break LOCA simulation: predicted trends of liquid level in the CMTs (comparison between reference calculation and sensitivity study).

Effect of liquid temperature in the pool Again with reference to the ILOCA in the AP-600 plant, while the reference calculation was done assuming an initial temperature of the liquid in the Core Make up Tanks equal to 373 K (to prevent the stop of the calculation owing to numerical problems), a second calculation was run considering a lower (323 K) liquid temperature in the Core Make up Tanks. In this case the condensation in the connection zone between the CMTs lines and the vessel creates a steam upward flow in the lines that prevents liquid drainage and causes dryout early in the transient (Fig. 19).

Effect of nodalization of the CMT discharge lines The line connecting the CMT to the vessel presents a complex layout in the vertical plane. Different approaches in modelling the line result in quite different predictions of CMT discharge. The CMT levels in the reference case for LBLOCA, are compared with those obtained in the sensitivity study in Fig. 20 *. In the reference case the fairly coarse nodalization of the discharge lines modifies the pressure drop situation in the plant causing flow from the cold leg to one of the CMT, that does not experience any level decrease despite the liquid delivery to the vessel. The nearly constant level situation in one tank prevents the activation of the depressurization system. The more detailed nodalization used in the sensitivity calculation allows the complete discharge of both CMT's. * The steps observable in the curves for the reference case occur when the level crosses the volume boundary and constitute a nodalization effect.

Deficiencies and capabilities of system codes are widely discussed by the international community. Findings relevant to the new generation reactors are discussed hereafter; these are the direct outcome of the experience acquired in the use of Relap5/mod2, but can be extended to other advanced system codes. In connection with the use of codes in predicting transient scenarios in the innovative reactors, essentially, two phases in the event time sequences can be distinguished: (a) primary system presure greater than about 0.5 MPa; (b) subsequent period including a tight interaction between the primary loop and the containment system and the long term behaviour of the passive safety systems. The physical situations foreseeable in the first period are characterized by parameter ranges for which a very wide data base (experiments and code calculations) already exists. However, this data base does not cover all of the new design features. Some of those will operate in ranges for which models were not developed or properly assessed. Example of such are the CMTs in the AP-600 design. The large thermal gradient in the CMTs causes very high condensation rate. One can expect that during CMT draining a layer of saturated liquid will form over the subcooled water significantly reducing the condensation. Models for such thermal stratification do not exist in present codes. However, codes capabilities and limits can be retained the same as applicable for the present generation reactors and will not be discussed furtherly in this paper. On the other hand, most of the phenomena foreseeable in the phase (b) should be considered outside the qualification boundary of the codes. Some of them are also outside the validity limits of the correlations and of the numerical structure of the codes. For a systematic evaluation of the codes limits and capabilities, all the phenomena listed in Table 3 should be considered in the assessment process. Such results are not available for the time being, and only few generic aspects are emphasized below.

List of deficiencies" 1.

At pressure close to the atmospheric value, very large oscillations may occur in the physical quantities. In some cases this is the results of the applied numerical scheme and of discontinuities of the functions simulating the water properties with main concern to the derivatives. As a consequence of this, very small time steps must be used (this is not

F. D'Auria et al. / Relevant thermalhydraulic aspects

pratical for long lasting transients) and frequent interruption of the calculation may occur. 2. The simulation of the fluid behaviour downstream of a critical section (supercritical flows) is not allowed in any geometric situation. Steam superheating may be important in the condensation process inside pools. 3. The transition between critical flow model and the ordinary differential equation model to calculate flowrate leads to oscillations in the calculation. 4. The possible occurrence of multiple critical sections in a complex piping appears to be outside the prediction capabilities. 5. The stratification of temperatures in pools or tanks is not well calculated by the present system codes: even the natural convection circulations establishing when a heat source and sink are present are not calculated. This may result in wrong prediction of gravity head, and errors in mass flowrates and heat exchange coefficients. 6. The natural circulation occurring in several parallel loops (the common element to almost all of these being the reactor core) largely depends upon local loss coefficients in complex three-dimensional geometries. These have much more importance when pumping power is lacking and must be supplied as input by the user. Integral system data will be required to asses the capability of the codes to predict the natural circulation in complex systems. 7. The capability to track non-condensable gases in the whole spectrum of foreseeable conditions of temperature, velocity and gas fraction does not appear adequate. Especially the steam-gas separation process is not considered. 8. The evaluation of both direct (e.g. at ECC port, inside large pools, etc.) and indirect condensation (e.g. inside IC tubes) does not appear adequate, particulary in presence of non-condensable gases. 9. The zero dimensional neutronic kinetics is not suitable for simulating 3D behaviour of large cores especially of BWR type (this important limitation also applies for current reactors). 10. The codes numerical solution scheme should be improved to handle transients lasting several hours. As a consequence, especially of items 5, 6 and 8 the codes are not able to simulate the integrate behaviour of primary system and containment.

6. Conclusions

An overview has been given in this paper of relevant thermalhydraulic aspects applicable to the new

257

generation reactors. The thermalhydraulics of evolutionary reactors do not imply the occurrence of new accident scenarios compared to current generation reactors. The same conclusion does not apply to the innovative reactors. SBWR, AP-600 and PlUS have been specifically considered summarizing some results of experimental studies and of code applications. The main outcomes are the classification of a series of phenomena important for the evaluation of transient performance of the mentioned reactors and the identification of specific code limits. Relevant differences between scenarios in current and innovative type of reactors can be addebted to two facts: - evolution of the largest part of scenarios at low pressure (near the atmospheric value) in the innovative reactors; - tight interaction between primary system and containment and the passive safety systems also implying the occurrence of several parallel natural circulation loops each one including (possibly) pools where direct steam condensation takes place and the transport of large amounts of non-condensible gas. The comparison between the foreseeable new plant phenomena and the objectives of the documented experimental researches, demonstrate that some critical issues have been considered, but the spectrum of potentially interesting thermalhydraulic aspects is far larger than the number of phenomena taken into account. Furthermore the available data base must be considered still preliminary and not fully exploited. New researches can be planned on this basis. The application of currently available system codes to off-normal conditions typical of the innovative reactors allowed to distinguish two periods that are separated by a pressure boundary set at about 0.5 MPa: when the primary system pressure is above that value, the code suitability and applicability is essentially the same as in current generation codes; at pressure below that value, the occurrence of new phenomena and intrinsic code limitations (of low interest for current reactors) prevent, in the general case, the possibility of a reliable simulation of plant scenarios. A list of 10 specific areas in the codes that need improvement has been produced. Finally innovative reactors contain technological features that are common to the present generation reactors. From a thermalhydraulic point of view, PIUS is characterized by the largest innovation: the safety is not dependent upon added external circuits or components but is intrinsic to the reactor concept. As such it requires larger investigation to prove its suitability.

258

F. D'Auria et al. / Relevant thermalhydraulic aspects

References [1] A.S. Rao, C.D. Sawyer, R.L. Huang, F.M. Paradiso, H.E. Townsend, Simplified BWR performance and safety, Int. Workshop on the Safety of Nuclear Installations of the Next Generation and Beyond, Chicago (US) Aug. 28-31, 1989. [2] H.I. Bruschi, T.S. Anderson, The Westinghouse AP-600: the leading technology for proven safety and simplicity, IAEA Technical Committee Meet. on Progress in Development and Design Aspects of Advanced Water-Cooled Reactors, Rome (I) Sept. 9-12, 1991. [3] R.M. Kemper, C.M. Verter, Loss of Coolant accident performance of the Westinghouse 600 MWe Advanced Pressurized Water Reactor, J. Nuclear Technology 91 (July 1990). [4] C. Pind, The Secure Heating Reactor, J. Nuclear Technology 79 (Nov. 1987). [5] C. Sundqvist, L. Nilsson, T. Pedersen, PIUS, Aspects on Containment. Philosophy and Design, IAEA Technical Committee Meet. on Progress in Development and Design Aspects of Advanced Water-Cooled Reactors, Rome (1) Sept. 9-12, 1991. [6] CSNI Group of Experts, CSNI code validation matrix of Thermalhydraulic Codes for LWR LOCA and Transients, CSNI Report 132, Paris (F) (March 1987). [7] CSNI Group of Experts, Thermohydraulic of Emergency Core Cooling in Light Water Reactors, CSNI Report 161, Paris (F) (Oct. 1989). [8] US NRC, Compendium of ECCS Research for Realistic LOCA Analysis, USNRC Report NUREG 1230, Washington (USA) (Dec. 1987). [9] F. D'Auria: Overview of requisites for thermalhydraulic codes in view of applications to the new generation reactors, Communication to CSNI THSB Task Group Meeting, Paris (F), June 25-27, 1991. [10] H. Nagasaka, K. Yamada, M. Katoh, S. Yokobori, Heat removal test of Isolation Condenser applied as a passive containment cooling system, The 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE), Tokyo (J), Nov. 4-7, 1991. [I1] S. Yokobori, H. Nagasaka, T. Tobimatsu, System response test of Isolation Condenser applied as a passive containment cooling system, The 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE), Tokyo (J), Nov. 4-7, 1991. [12] P. Coddington, S. Guntay, H. Dreier, L. Fisher, M. Huggenberger, H. Smith, L. Duff, C. Varadi, G. Yadigaroglu, ALPHA: the long term passive heat removal program at PSI, Switzerland, Int. Top. Meet. on Nuclear Reactor Thermalhydraulics (NURETH-5), Salt Lake City (USA), Sept. 20-24, 1992. [13] S. Botti, F. Mazzacurati, C. Medich, O. Vescovi, ISOCON: an experimental test facility to qualify the SBWR Isolation condenser and Passive Containment cooling system, Int. Top. Meet. on Nuclear Reactor Thermalhy-

[14]

[15]

[16]

[17]

[18]

[19]

[20]

[21]

[22]

[23]

[24]

[25]

draulics (NURETH-5), Salt Lake City (USA), Sept. 2024, 1992. R. Bovalini, F. D'Auria, M. Mazzini, Experiment of core coolability by a gravity driven system performed in PIPER-ONE apparatus, 7th Proc. of Nuclear Thermalhydraulics, 1991 ANS Winter Meet. San Francisco (USA), Nov. 10-14, 1991. R. Bovalini, F. D'Auria, M. Mazzini, P. Vigni, Isolation condenser performance in PIPER-ONE apparatus, Submitted at 1992 European Two-Phase Flow Group Meet.Stockolm (S), May 1992. M.M. Corletti, L.E. Hochreiter, Advanced light water reactor passive residual heat removal heat exchanger test, The 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE), Tokyo (J), Nov. 4-7, 1991. C. Medich, M. Rigamonti, O. Vescovi, Spes-2, an integral test facility for AP-600 Advanced PWR Safety Research, 7th Proc. of Nuclear Thermalhydraulics, 1991 ANS Winter Meet. San Francisco (USA), Nov. 10-14, 1991. D. Babala, U. Bredolt, J. Kemppainen, A study of the dynamics of Secure Reactors: comparison of experiments and computations, Nucl. Engrg. Des. 122 (1990). T. Watanabe, Y. Asahi, M. Fujii, Y. Anoda, K. Tasaka, Y. Kukita, Transient analysis of loss of feedwatyer at PlUS experimental apparatus, The 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE), Tokyo (J), Nov. 4-7, 1991. K. Tasaka, M. Tamaki, M. Fujii, Y. Amada, H. Murata, Y. Kukita, S. lmai, H. Kohketsu, S. Fukuchi, Small scale thermalhydraulic experiment of a PlUS type reactor, The 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE), Tokyo (J), Nov. 4-7, 1991. C. Pind, J. Fredell, Summary of theoretical analysis and experimental verification of PlUS density lock development Program, IAEA Technical Committee Meet. on Progress in Development and Design Aspects of Advanced Water-Cooled Reactors, Rome (1) Sept. 9-12, 1991. H. Oikawa, K. Arai, H. Nagasaka, Optimisation study on SBWR isolation condenser heat removal performance, The 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE), Tokyo (J), Nov. 4-7, 1991. K.M. Vierow, H.E. Townsend, J.R. Fitch, J.G.M. Andersen, M. Alamgir, V.E. Schrock, BWR passive containment cooling system by condensation driven natural circulation, The 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE), Tokyo (J), Nov. 4-7, 1991. A. Cheung, T. Schulz, D. Holderbaum, E. Carlin, C. Verter, AP-600 core make up tanks elimination, a design feability study, 7th Proc. of Nuclear Thermalhydraulics, 1991 ANS Winter Meet. San Francisco (USA), Nov. 111-14, 1991. K. Okabe, N. Umezawa, K. Sugizaki, T. Matsuoka, LOCA analysis for new medium size simplified PWR, 7th Proc. of Nuclear Thermalhydraulics, 1991 ANS Winter Meet. San Francisco (USA), Nov. 10-14, 1991.

F. D'Auria et aL / Relevant thermalhydraulic aspects

[26] B. Barbucci, C. Bertani, C. Carbone, G. Del Tin, F. Donatini, G. Sobrero, Assessment of PlUS primary system using analytical and experimental models, IAEA Technical Committee Meet. on Progress in Development and Design Aspects of Advanced Water-Cooled Reactors, Rome (I) Sept. 9-12, 1991. [27] P. Andreuccetti, P. Barbucci, F. Donatini, F. D'Auria, G.M. Galassi, F. Oriolo, Capabilities of Relap5 in simulating SBWR and AP-600 thermalhydraulic behaviour, IAEA Technical Committee Meet. on Progress in Development and Design Aspects of Advanced Water-Cooled Reactors, Rome (I) Sept. 9-12, 1991. [28] D. Abdollahian, C.B. Jonhson, J. Yedidia, Analysis of Simplified Boiling Water Reactor ECCS performance, 7th Proc. of Nuclear Thermalhydraulics, 1991 ANS Winter Meet. San Francisco (USA), Nov. 10-14, 1991. [29] E. Brega, C. Lombardi, M. Ricotti, A. Rilli, Simulation of operational and accident transients of the PIUS reactor

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