Preliminary safety evaluation for CSR1000 with passive safety system

Annals of Nuclear Energy 65 (2014) 390–401

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Annals of Nuclear Energy

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Preliminary safety evaluation for CSR1000 with passive safety system

0306-4549/$ - see front matter � 2013 Elsevier Ltd. All rights reserved.http://dx.doi.org/10.1016/j.anucene.2013.11.031

⇑ Corresponding author. Tel./fax: +86 029 82663769.E-mail address: junligou@mail.xjtu.edu.cn (J. Gou).

Pan Wu a, Junli Gou a,⇑, Jianqiang Shan a, Bo Zhang a, Xiang Li b

a School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Chinab Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu 610041, China

a r t i c l e i n f o

Article history:Received 26 July 2013Received in revised form 15 October 2013Accepted 17 November 2013Available online 15 December 2013

Keywords:SuperCritical Water-cooled Reactor (SCWR)Passive safety systemSafety analysisCSR1000

a b s t r a c t

This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor(CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core designapplied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of thecore at normal operation condition. Each fuel assembly is made up of four sub-assemblies with down-ward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the largewater inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 toincrease the safety reliability at abnormal conditions. In this paper, accidents of ‘‘pump seizure’’, ‘‘lossof coolant flow accidents (LOFA)’’, ‘‘core depressurization’’, as well as some typical transients are analysedwith code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate thatthe maximum cladding surface temperatures (MCST), which is the most important safety criterion, of theboth passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivityanalyses of the delay time of RCPs trip in ‘‘loss of offsite power’’ and the delay time of RMT actuation in‘‘loss of coolant flowrate’’ were also included in this paper. The analyses have shown that the core designof CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequencesof the selected abnormalities.

� 2013 Elsevier Ltd. All rights reserved.

1. Introduction

SuperCritical Water-cooled Reactor (SCWR) is a promising gen-eration IV reactor concept for its considerable advantages in sys-tem simplicity, high thermal efficiency and enhanced safety.Many countries and organizations have proposed their specificSCWR concepts and devoted themselves to improving their designsin the past decade. Several concepts of the pressure-vessel typeSCWR were proposed by the University of Tokyo since the 1990s.The feasibility assessments of the Super Light Water Reactor (SuperLWR) and the Super Fast Reactor (Super FR) are ongoing jointly bythe universities, research institutes and industries in Japan (Okaet al., 2010). A NERI project aiming to develop a reference US SCWRhas been conducted since 2001 (MacDonald et al., 2004). In Europe,the High Performance Light Water Reactor (HPLWR) is still underdevelopment (Schulenberg et al., 2011). SCWR-R is a pressure ves-sel concept developed by Republic of Korea that utilizes a solidZrH2 moderator. SCWR researches in Korea include a feasibilitystudy, a core conceptual study, experiments for supercritical heattransfer, and an investigation of the corrosion and radiation effectson candidate materials (Bae et al., 2007). A pressure-tube typeCANDU SCWR has been developed by AECL and various universi-

ties in CANADA (Leung, 2010). In China, a new thermal spectrumSCWR concept named CSR1000 is proposed and its research anddesign are carried out by Nuclear Power Institute of China (NPIC)(Xiao et al., 2013).

Safety analysis is an essential procedure to evaluate the feasibil-ity of a new SCWR concept. In the past researches, a comprehen-sive evaluation of the Super LWR was carried out by theUniversity of Tokyo with SPRAT-DOWN, SPRAT-DOWN-DP andSCRELA reflood module. Various transients and accidents includingLOFA and LOCA, were chosen for the safety analysis of the SuperLWR and a good performance was achieved (Ishiwatari et al.,2007). The response of the US SCWR concept during LOFA andLOCA had been evaluated using the RELAP5-3D computer program.The results showed that the proposed system for US SCWR allowedthe core to meet cladding thermal limits (MacDonald et al., 2004).The analyses of the incidents addressing heat storage capacity, corecooling in case of loss-of-flow, core heat capacity, RIAs and ATWSwere carried out for HPLWR. Various safety analysis codes wereintroduced into the analyses to increase the confidence in the re-sults, such as RELPA5, CATHARE, APROS, KIKO3D-ATHLET, andTRAB-3D/SMABRE. There were still open issues to be discussedfor the safety analysis of HPLWR (Andreani et al., 2012). The LOCAanalysis of SCWR-M with a passive safety system was carried outby Shanghai Jiaotong University. The passive safety system con-taining the isolation cooling system, accumulator injection system,

P. Wu et al. / Annals of Nuclear Energy 65 (2014) 390–401 391

gravity driven cooling system, and automatic depressurization sys-tem, was able to keep the SCWR-M core at safety condition duringloss of coolant accident (Liu et al., 2013).

This paper concentrates on the safety analyses of CSR1000.Transients and accidents of different types were simulated withSCTRAN to make a preliminary assessment of the feasibility ofCSR1000. The sensitivity analyses of some crucial parameters werealso presented to provide essential references for the futureimprovements.

2. CSR1000 and its passive safety systems

2.1. Descriptions of the CSR 1000

CSR1000 is a pressure-vessel SCWR concept, which is developedby the Nuclear Power Institute of China (NPIC). The core is cooledand moderated by light water. The rated thermal and electric pow-ers of the CSR1000 are 2300 MW and 1000 MW respectively. In or-der to achieve a high thermal efficiency, the coolant temperaturesof the core inlet and outlet are set to be 280 �C and 500 �C. Thetwo-pass core design adopted by the CSR1000 increases the coreheating length and decreases the temperature difference in the ax-ial direction effectively. The coolant temperature distribution inthe core of CSR1000 is illustrated in Fig. 1. From the figure wecan see that the temperature difference along each flow path issmall and it’s good for the axial power flattening.

The two-pass core of CSR1000 is made up of 177 fuel assem-blies. The first-pass core is comprised of 57 assemblies located inthe center part, while the remaining 120 assemblies make up thesecond-pass. The arrangement of the assemblies in the core isshown in Fig. 2(a) and the flow scheme of the two-pass core isillustrated in Fig. 2(b). 76.7% Of the coolant is directed to the topdome, which is then divided into three parts, including 35.9% flow-ing downward through the fuel channel of the first pass, 10.8%flowing downward through the water rods of the first pass and30% flowing downward through the water rods of the second pass.The coolant from the downcomer (23.3% of the total), the fuelchannel of the first pass and the water rods mixed together inthe lower plenum, and then flows upwards through the fuel chan-nel of second pass. This specific flow scheme is favorable toimproving the neutron utilization, reducing the hot channel factorand increasing the coolant outlet temperature. Fig. 3 shows the

Fig. 1. Coolant temperature

cross-section of the fuel assembly which contains four sub-assem-blies with downward water rods. For each sub-assembly, the fuelbundle of 9 by 9 fuel rods configuration with an square waterrod in the center is surrounded by an assembly wall. The waterin the water rods serves as the moderator and coolant at the sametime. The water inventory in the water rods is favorable to pre-venting the core heat-up when the abnormal incidents take place.Besides, the cruciform control rod is adopted (Xia et al., 2013).

The main parameters of the CSR1000 are listed in Table 1.

2.2. The passive safety system design

As passive safety systems depend on only natural forces, such asgravity, natural circulation, and compressed gas-simple physicalprinciples, no pumps or other rotating machinery are requiredand the failure possibility of the safety system is reduced. Passivesafety systems have been widely utilized in many advanced nucle-ar reactor, such as ESBWR (Cheung et al., 2005), and AP1000(Schulz, 2006), to provide significant improvements in safety, reli-ability, and plant costs. In order to deal with different design basisaccidents of CSR1000, an innovative passive safety system has beenproposed for CSR1000 by Xi’an Jiaotong University and NPIC. Thepassive system of CSR1000 is shown in Fig. 4. Its mainly made upof the high pressure reactor make-up tank (RMT), the isolation con-denser system (ICS), the automatic depressurization system (ADS),the gravity driven core cooling system (GDCS), the passive contain-ment cooling system (PCCS) and the direct vessel injection (DVI).

There is one RMT for each of the two trains. The RMT is de-ployed between the DVI line and the steam line. The RMTs are con-nected to the steam line upstream of the main steam isolationvalves via a pressure balance line so that the pressure in each tankis maintained equal to the pressure in the main steam line. The iso-lation valves are placed between the bottom of RMT and the DVIline to control the coolant flowrate. The RMTs are placed outsidethe containment to provide sufficient high pressure coolant injec-tion during different accidents. The tanks are filled with supercrit-ical water of 25 MPa and 50 �C. When LOFA and LOCA occur, theisolation valve will be opened automatically by the actuation sig-nal. The low temperature coolant in the tank will flow into the coreby gravity.

Two trains of ICS are proposed to passively remove the decayheat through natural circulation for the long-term cooling after

distribution in the core.

Fig. 2. Fuel assembly distribution and flow scheme in the core of CSR1000. (a) Fuel assembly distribution. (b) Flow scheme in the core.

Fig. 3. Cross-section of the fuel assembly.

Table 1The main parameters of the CSR1000.

Core pressure (MPa) 25Thermal/electrical power (MW) 2300/1000Efficiency (%) 43.5Inlet/outlet temperatures (�C) 280/500Neutron spectrum ThermalQuantity of fuel assemblies 177Coolant flow scheme Two-passMain coolant flowrate (kg/s) 1190Average power density (MW/m3) 60Active core height (m) 4.2 mCladding material Stain steel (310S)Maximum cladding surface temperature (�C) 650Fuel doppler feedback coefficient ($/K) �3.54e�3Moderator density feedback coefficient ($/(kg/m3)) 2.046e�2

392 P. Wu et al. / Annals of Nuclear Energy 65 (2014) 390–401

LOFA and abnormal transients without depressurization. Undernormal operating conditions, the heat exchanger of ICS is filledwith water, being isolated by a check valve at the inlet and an iso-lation valve at the outlet. After a system trip, the isolation valvesare opened and ICS will start to work automatically.

The systems of GDCS and PCCS refer from ESBWR. The GDCScontaining four GDCS pools is designed for the low pressureinjection in LOCA. The PCCS is deployed outside the first steelcontainment and provide a natural circulation to remove theheat in the containment passively. The PCCS pool is the ultimateheat sink.

The ADS is made up of 8 safety relieve valves (SRVs) and 8depressurization valves (DPVs). The SRVs are designed for theoverpressure protection and the high temperature coolant will bereleased into the suppression pool. The DPVs are provided for therapid depressurization of the vessel and the coolant will be in-duced to the drywell of the containment.

The initial geometric parameters without considering the con-figuration in the containment are shown in Table 2. With referringto the previous investigations on the safety analysis of SCWRs andin consideration of the passive safety features of the CSR1000, theactuation conditions of the passive safety systems are preliminarilydetermined as given in Table 3 (Wu et al., 2013).

3. Methodology

Referring to code RETRAN-02, SCTRAN is developed by Xi’anJiaotong University. SCTRAN is a one-dimensional safety analysiscode for SCWRs and applies the fission decay heat equation andpoint neutron kinetics equation with six groups of delayed neutronto calculate the core power. Its ability to simulate the transientsand accidents of SCWR has been verified by comparing with APROScode and RELAP5-3D code, respectively. The results imply that theSCTRAN code is capable of carrying out safety analysis for SCWRs

Fig. 4. Passive safety system of CSR1000.

Table 2Main parameters of the passive safety system.

Parameters Name Value

RMT Quantity of RMTs 2RMT diameter (m) 4.3Initial water level (m) 6.2Elevation from the core (m) 10

ICS Quantity of heat transfer tubes 1200Heat transfer tube diameter (mm) 12Temperature of hot trap (�C) 50Heat transfer tube length (m) 5

GDCS Quantity of GDCS pools 4Elevation from the core (m) 4.572GDCS diameter (m) 6.88

ADS (8DPV) DPV area (m2) 0.09348

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(Wu et al., 2013). In this paper, the code SCTRAN will be applied tomake a preliminary safety evaluation of CSR1000.

The SCTRAN model of CSR1000 system is briefly illustrated inFig. 5. Each pass of the two-pass core was simulated by an averagefuel channel and a hot channel. The hot channel factors of bothpasses were set 1.26, which were achieved by neutron calculation.The water rods and the fuel channels were separately simulated by10 nodes in the axial direction. The four upright black solids stand-ing next to the fuel channels in the figure were presented to sim-ulate the fuel rods in the core. The 40 small black solidsrepresented the water rod wall to model the heat transfer betweenthe coolant channel and the corresponding water rods. The corepower distribution in the axial direction, the doppler and densityfeedback coefficients were obtained from the neutron calculation.A time dependent junction with temperature of 280 �C and flow-rate of 1180 kg/s, as well as a time dependent volume with tem-perature of 500 �C and pressure of 25 MPa, were separately set asthe boundary conditions of the main coolant line and the main

steam line in each loop at normal condition. The reactor coolingsystem was modeled as two loops while only one loop was shownin the figure. Except the ICS located in the loop 1, the safety deviceswere equipped to the both loops. The control system of CSR1000was not involved in this paper. Focusing on the thermal–hydraulicperformance inside the reactor pressure vessel (RPV), this paperdid not take the function of PCCS into consideration.

4. Event selection and safety principle

In this paper, some typical events ranging from operational con-ditions to accidents were selected to make an comprehensivesafety assessment of CSR1000. The events to be analysed couldbe classified into three categories by their causes:

(a) decrease in core coolant flowrate, i.e., ‘‘partial loss of reactorcoolant flow’’, ‘‘loss of offsite power’’, ‘‘main coolant controlsystem failure’’;

(b) abnormality in reactor pressure, i.e., ‘‘isolation of mainsteam line’’;

(c) abnormality in reactivity, i.e., ‘‘loss of feedwater heating’’,‘‘uncontrolled CR withdrawal’’.

Furthermore, analyses of typical accidents including ‘‘pump sei-zure’’, ‘‘loss of coolant flowrate accident (LOFA)’’ and ‘‘core depres-surization’’ were performed. The LOCA analysis will be carried outin the future.

In the evaluation, the maximum cladding surface temperaturewas regarded as the most important criterion. As the stainless steelis selected to be the cladding material of the CSR1000, the claddingtemperature criterion of the transients and accidents were set to be850 �C and 1260 �C, respectively. Besides, the reactor pressureshould be kept below 30.3 MPa to maintain the pressure boundaryintegrity. In the following analysis, the flowrate ratio of the first-pass and the second-pass is obtained by being divided by the main

Table 3Actuation conditions of the passive safety system.

Actuation conditions

Reactor Pressure low (Level 1): 24 MPaMass flowrate low (Level 1): 90%

Scram Pressure high (Level 1): 26 MPaPower high: 120%

MSIV Pressure low (Level 2): 23.5 MPaMass flowrate low: 6%

RMT Pressure low (Level 2): 23.5 MPaMass flowrate low (Level 1): 90%

SRV (10% of the rated flowrate) Relief valve Safety valveOpen number Close Open number (MPa)(MPa) (MPa)26.2 25.2 1 27.0 226.4 25.4 1 27.2 326.6 25.6 3 27.4 326.8 25.8 3

ADS Pressure low (Level 2): 23.5 MPaMain coolant flowrate low: 6%

ICS RMT water level low: <50%GDCS Pressure difference between GDCS pool and RPV

Fig. 5. SCTRAN model of CSR1000 system.

394 P. Wu et al. / Annals of Nuclear Energy 65 (2014) 390–401

coolant flowrate (1190 kg/s). The MCST variations of both passesare obtained from MCST values of the corresponding hot channels.

5. Safety analyses of CSR1000

5.1. Transient analysis

5.1.1. Partial loss of main coolant flowThe partial loss of main coolant flow was caused by the loss of

the feedwater pump in the second loop. The main coolant flowratein this loop was supposed to decrease linearly to zero in 5 s (actu-ally depends on the pump inertia), while the feedwater pump ofthe other loop maintained the original state. ‘‘The low coolantflowrate (90%)’’ signal triggered the reactor scram signal andRMT actuation at 1 s. After 0.5 s delay, the control rods started todrop and the core power decreased. Fig. 6 shows the systemparameter variations. The RMT valves were opened after 4.0 s de-lay. The RMT in the second loop quickly drained water into the coreby gravity. However, in the first loop, as the main coolant line andthe steam line were connected through the RMT and the feedwaterpump was still running, part of the feedwater was injected to the

steam line, which decrease the coolant flowrate in the first loop.The total coolant flowrate flowing through the core maintained ahigh level in the first 5 s of the incident and restrained the MCSTincrease of both passes. The mismatch of core flow and core powerincreased the MCST of the second pass by 50 �C. The core pressurekept below the initial values.

5.1.2. Loss of offsite powerIn normal conditions of CSR1000, two turbine-driven RCPs

were provided to circulate the coolant system. Besides, two mo-tor-driven RCPs were supplied as backups at startup and shut-down conditions. The loss of offsite power did not affect theoperation of turbine-driven RCP. In this incident, the ‘‘loss of off-site power’’ signal triggered the reactor scram directly. With thesupply of steam from moisture separator reheater, the turbine-driven RCPs could keep supplying coolant regardless of the lossof the offsite power. At about 10 s, the steam supplied for theRCPs became less and the RCPs started to run out. Then the‘‘low coolant flowrate (90%)’’ signal triggered the RMT valves tobe opened. With the aids of RMT, the main coolant flowratewas maintained to be about 50% of the original level. During this

Fig. 6. Partial loss of main coolant flow.

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transient, the MCSTs of both passes stayed below the initial val-ues all the time, as shown in Fig. 7.

In this transient, the delay time of RCPs trip was decided by thesteam supply from the deaerator. The delay time of RCPs trip wasan important parameter because it determined when the maincoolant flowrate started to decrease. Here a sensitivity analysis ofthe delay time of RCP trip was performed. Three conditions weresimulated, whose delay time were 3, 5 and 10 s separately. Thevariations of main coolant flowrate and MCST of the second-passare illustrated in Fig. 8. The solid and dotted lines of different colorsrepresent different conditions. From Fig. 8 we can see that when-ever the main coolant flowrate started to decrease, the timelyintervention of RMT could compensate the coolant inventory inthe core. The MCST of second pass stayed below the initial valuesin all the three conditions due to main coolant flowrate recovery.We can imagine that in the limit case that the delay time equals0 s, the consequence will be similar to that of ‘‘loss of coolant flow-rate accident’’.

Fig. 7. Loss of of

5.1.3. Main coolant control system failureThis is a typical flowrate increasing incident. The failure of

main coolant control system caused the main coolant flowrateincreasing to 138% of its original value in 5 s. The simulation re-sult is shown in Fig. 9. The increasing coolant flowrate meant thatmore coolant was injected into the core. The core was bettercooled while the cladding temperature and the coolant tempera-ture decreased. Due to the fuel doppler and moderator feedbackprinciple, a positive reactivity feedback was produced. The corepower started to increase because of the positive reactivity. Itcan be observed from the figure that the growth rate of the maincoolant flowrate was larger than that of the power. Thus theMCSTs and the coolant temperature decreased, which further in-creased the core power. When the power increased to 120% of itsoriginal value, the reactor scram signal was actuated and thepower started to decrease after 0.5 s delay. When the reactorcoolant pump stopped, the RMTs would be launched and a largeamount of cold water was injected to the core, like what

fsite power.

Fig. 8. Sensitivity analysis of RCP delay time in loss of offsite power.

396 P. Wu et al. / Annals of Nuclear Energy 65 (2014) 390–401

happened in the above incidents. The core was effectively cooledand the MCSTs of the both passes stayed below the initial valuesin the process.

5.1.4. Main steam line valve closureIn order to observe the safety system response to the abnormal-

ities addressing the core pressure, the MSIV closure transient wassimulated. The valves of both main steam lines were supposed tobe closed in 5 s. The main coolant flowrate was assumed to keepconstant in 10 s. The closure of MSIV pressurized the core, whichfurther led to the power scram (high pressure signal). The corepressure reached the threshold value of SRVs and the relief valveswere opened to release coolant to avoid overpressure. Thus thecore pressure oscillated between 25.02 and 26.72 MPa due to theSRV opening and closing. The flow oscillation amplitude of thesecond pass was more violent than that of the first pass becauseof buoyancy effect and drag force in the downward channel. TheMCSTs of both passes had a slight increase of 10 �C, as shown inFig. 10.

Fig. 9. Main coolant con

5.1.5. Loss of feedwater heatingThe loss of steam supply to the last high-pressure heater caused

the feedwater temperature decreasing from 280 �C to 273.6 �C. Theexcursion of feedwater temperature was set as 40 �C to achieve aconservative result. The result is shown in Fig. 11. The temperaturedecrease of feedwater reduced the coolant temperature and in-creased the coolant density simultaneously. The reactor scramwas actuated after the coolant temperature feedback and dopplerfeedback made the core power rise to 120% of the rated value.The mismatch of the coolant flow and the core power resulted inthe cladding temperature increase. The MCST of the second-passhad a peak temperature of 767 �C.

5.1.6. Uncontrolled CR withdrawalThe maximum reactivity value of a CR cluster in CSR1000 was

not determined yet. The previous literature about the analysis of‘‘uncontrolled CR withdrawal’’ was referred (Ishiwatari et al.,2005a,b). In this simulation, a CR cluster whose reactivity valuewas 2.0$ was regarded to be drew out of the core at a uniform

trol system failure.

Fig. 10. Main steam line valve closure.

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speed of 114 cm/min. The core power increased gradually with theCR withdrawal. Thus the MCST of the both passes rose together.When the power reached a certain level, the reactor scram signalwas actuated and the power started to decrease after 0.5 s delay.The main coolant flowrate could still maintain for a while andthe MCST of the both passes started to decrease after reaching apeak value of 750 �C, as shown in Fig. 12. After the coolant pumpstopped, the coolant from the RMTs would make up the core flowand cool the core effectively.

5.2. Accident analysis

5.2.1. Pump seizureThe consequence of ‘‘pump seizure’’ is shown in Fig. 13. In this

accident the feedwater pump in the second loop got stuck, whichresulted in a sudden loss of the coolant flowrate in 0.1 s. The scramsignal was quickly detected because of the ‘‘the low feedwaterflowrate (90%)’’. After 4 s delay, the valves of RMT in both loopswere opened. Like the process of ‘‘partial loss of coolant’’, the

Fig. 11. Loss of feed

opening of RMT was helpful for the water injection in the secondloop while it was disadvantage for the first loop. However, the totalcoolant was still sufficient to cool down the core after the MCSTsexperienced an increase of 100 �C. This accident was more severethan the ‘‘partial loss of coolant flow’’ because the coolant in oneloop lost instantly.

5.2.2. Total loss of coolant flow (LOFA)The loss of coolant flow accident (LOFA) was mainly mitigated

by the RMT that provided low temperature coolant by gravity inthe initial stage and the ICS that cooled the coolant from the coreby natural circulation. LOFA was initiated by the both RCPs trip.The coolant flowrate was assumed to ramp linearly to zero in 5 s.The ‘‘low feedwater flowrate (90%)’’ signal triggered the scram sys-tem and the RMT actuation. The control rods started to insert intothe core at 1.0 s and the RMT valve was opened at 5.0 s. FromFig. 14 we can see that, in the first 5 s, the mismatch of powerand main coolant flow led to the cladding temperature increase.With the injection of cold water from RMT, coolant flowrate of

water heating.

Fig. 12. Uncontrolled CR withdrawal.

Fig. 13. Pump seizure.

398 P. Wu et al. / Annals of Nuclear Energy 65 (2014) 390–401

both passes recovered. The MCSTs of both passes started to de-crease after reaching a temperature peak of 815 �C. The oscillationsof system pressure and core flowrate were resulted from the openand closure of SRV.

With the draining out of RMT, the ICS valve was opened toprovide long-term cooling for LOFA. Fig. 15(a) shows clearly thatICS was started at 60 s and the power removed by ICS was largerthan the decay heat of the core at about 630 s, which meant thatthe ICS was capable of providing the long-term decay heat re-moval. Fig. 15(b) shows the coolant flowrate evolutions of bothpasses with time. When the stable natural circulation was estab-lished, the downward flow in the first pass was reversed becauseof the big flow resistance in the flow channel. The MCST of thefirst pass had a slight increase in this period. However, it didnot change the overall trend of MCST variations. With the startupof the RMT and the ICS in different stages, the MCSTs of bothpasses decreased and kept to be a low level, as shown inFig. 15(c). In the simulation of ICS, the temperature of the ICSpool was set constant.

From the above analysis, we can realize the importance of RMTin the abnormalities about decreasing in main coolant flow. How-ever, the opening of RMT valve at inappropriate time will makethings worse because the feedwater from the main coolant linemay flow directly to the main steam line through the RMT, likethe situations of the second-loop in ‘‘partial loss of main coolantflow’’ and ‘‘pump seizure’’. Here a sensitivity analysis about theRMT delay time was conducted for the total loss of flow accident,which was most severe in the abnormalities addressing the maincoolant decrease. Four different conditions whose RMT delay timeswere 0, 2, 4 and 6 s, were simulated. The solid and dotted lines ofdifferent colors in Fig. 16 represents the variations of the second-pass MCST and the main coolant flowrate in different conditionsseparately. The main coolant flow decreased most quickly whenRMT delay time equaled zero, which resulted in the biggest MCSTincrease. When delay time equaled 6 s, the core encountered ashort period with no coolant injection from RMTs for about 1 s,which was bad for the core cooling. A very good performancewas achieved when delay time equaled 2 s or 4 s. 2 s or 4 s After

Fig. 14. Variations of core parameters in the first 40 s of LOFA.

Fig. 15. Thermal–hydraulic response in long-term LOFA. (a) The ICS parameters. (b) The core flowrate variations. (c) The MCST evolutions of both passes.

P. Wu et al. / Annals of Nuclear Energy 65 (2014) 390–401 399

the reactor scram, the main coolant flowrate dropped to a low le-vel. The water in RMT can flow to the core through the feedwaterline by gravity. Therefore the delay time of RMT should be decidedby the main coolant flowrate variation in accidents, which de-pended on the coolant pump inertia. On the premise that theRCP run-down time is 5 s, we regarded 4 s as the best delay timeof RMT actuation.

5.2.3. Core depressurizationThis incident was simulated to observe the reactor behavior

during the core depressurization. The ADS valves were assumedto be opened at 0 s. The coolant flowed through the ADS valvesat a critical velocity. The power scram signal was actuated laterbecause of the low coolant flowrate in the core. The RMT valveswere opened at 4 s and the main coolant injection stopped at 5 s

Fig. 16. Sensitivity analysis of RMT actuation delay time in LOFA.

Fig. 17. Core depressurization. (a) System parameter variation with time. (b) The void fraction of the lower and upper plenum and the GDCS flowrate. (c) The MCST variationof both passes in 2000 s.

400 P. Wu et al. / Annals of Nuclear Energy 65 (2014) 390–401

because of the pump run-down. At the initial stage of the blow-down phase, coolant in the pressure vessel was led to the steamlines through the two-pass core with the intervention of the ADS.Additionally, the coolant temperature decreased because of thecore depressurization (isenthalpic expansion). Thus the MCSTs de-

creased quickly. In this period, the coolant from the upper plenummainly flowed downward through the water rods of the first-passand second-pass other than the first-pass core because there was abig flow resistance in the coolant channel caused by the core heat-ing. That is the reason why the MCST of the first pass started to

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increase after a slight decrease, as shown in Fig. 17(a). In this pro-cess, all the coolant gathering in the down plenum had to be direc-ted to the steam line through the second pass core and the MCST ofsecond-pass kept decreasing before 30 s. With the blowing out ofthe coolant, the coolant inventory in the core became less andthe ADS flowrate decreased with the core pressure dropping. TheMCSTs of both passes started to increase at about 30 s.

When the pressure dropped to be lower than the containmentpressure, which was set 0.2 MPa in this paper, the GDCS poolwould take in charge of providing low temperature coolant forthe core. In Fig. 17(b), the GDCS valves were opened at about100 s, which means the start of the first reflooding phase. Coolantof low temperature flowed into the DVI lines and cooled the lowerplenum. The lower plenum were fully cooled until 230 s. In thisperiod, the MCST of both passes kept rising. After peaking to850 �C and 760 �C separately, the MCSTs of both passes started todecrease due to the coolant from GDCS pool entering the coresection.

Once the cold coolant entered the fuel channel, it began to cooldown the fuel rod and the MCSTs of both passes started to de-crease. Meanwhile, the coolant was heated to steam by the coredecay heat, which led to the pressurization of the RPV. The increaseof the RPV pressure decreased GDCS flowrate gradually and coolantinjection from GDCS was stopped at about 400 s, as shown inFig. 17(b). When the RPV pressure became larger than that of thecontainment, the coolant in the pressure vessel drained into thecontainment again through the ADS valves. The flowrate in bothpasses recovered and the MCSTs kept decreasing. The RPV depres-surization started again due to the blowing out of the coolant intothe containment. When the RPV pressure was low enough, thecoolant injection into the RPV from the GDCS pool (second refillingphase) started. The variations of the MCSTs were similar to those inthe first refilling phase. With the decrease of the core power andthe starts of the second reflooding phase, the MCSTs of both passesbegan to decrease after reaching the second peak and finally keptat a low level of around 150 �C, as shown in Fig. 17(c).

6. Conclusion

The concept development of the Chinese Supercritical watercooled Reactor (CSR1000), which is a pressure-vessel type, thermalspectrum SCWR of 1000 MWe, is being conducted by NuclearPower Institute of China (NPIC). The analyses of 6 transients and3 accidents of CSR1000 were carried out with the system codeSCTRAN. In all the incidents addressing the core flowrate decrease,the RMT actuation were triggered and the low temperature coolantfrom the RMT was injected into the core by gravity. The core underthis type of incidents could be cooled down effectively by the RMTin the prophase. In the ‘‘main steam line valve closure’’, the openand closure of SRVs protected the core from overpressure. In the‘‘loss of feedwater heating’’ and ‘‘uncontrolled CR withdrawal’’analysis, the power scram design stopped the core from over-power. Additionally, the ICS design did a god job in removing thedecay heat of the core in LOFA. In the ‘‘core depressurization’’,the ADS intervention increased the core flowrate and helped thecore depressurize to a low level. It was quite helpful for the corecooling in the process of depressurization. The intervention ofGDCS provide a necessary low pressure injection for the systemand finally cooled down the core.

In the above analyses, we can see that the design basis incidentsare mitigated effectively. Different passive safety designs proposedin this paper, i.e., RMT, ICS, SRV, DPV, and GDCS work quite welland satisfy the original requirements. The biggest MSCTs in tran-sient and accident analyses are separately 767 �C and 850 �C,which is far below the corresponding temperature criterions. Thecore pressure in all the incidents are kept below 30.3 MPa. Conclu-sion can be achieved that the passive safety systems greatly miti-gate the consequences of these incidents and the inherent safetyof CSR1000 is enhanced. The results will be a useful reference forthe future development of CSR1000.

The simulations of other transients and the sensitivity analyseswill be carried out to optimize the core design as well as the pas-sive safety system of CSR1000 in the future. The analyses of loss ofcoolant accident (LOCA) will also be the next topic. Additionally,the code SCTRAN would be coupled with some neutron calculationcodes to improve the calculation accuracy in the reactivityincidents.

Acknowledgments

The research is supported by the Research Fund for the DoctoralProgram of Higher Education of China (20120201110043) and Nu-clear Power Institute of China (NPIC).

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