Safety evaluation of the inherent and passive safety features of the smart design

Safety evaluation of the inherent and passivesafety features of the smart design

Kyoo Hwan Bae *, Hee Cheol Kim, Moon Hee Chang, SukKu Sim

Korea Atomic Energy Research Institute (KAERI), Power Reactor Technology Development Team,

150, Dukjin-dong, Yusong-gu, Taejon, South Korea 305-353

Received 18 February 2000; received in revised form 15 May 2000; accepted 16 May 2000

Abstract

SMART (system-integrated modular advanced reactor) is a 330 MWt advanced integralPWR, which is under development at KAERI for seawater desalination and electricity gen-eration. The conceptual design of the SMART desalination plant produces 40,000 m3/day of

potable water and generates about 90 MW of electricity, which are assessed as su�cient for apopulation of about 100,000. The SMART enhances safety by adopting the inherent safetydesign features such as the elimination of large break loss of coolant accidents, substantially largenegative moderator temperature coe�cients, etc. In addition, the safety goals of the SMART are

achieved through the adoption of passive engineered safety systems such as an emergency corecooling system, passive residual heat removal system, safeguard vessel, and reactor and con-tainment overpressure protection systems. This paper describes the design concept of the

major safety systems of the SMART and presents the results of the safety analyses using aMARS/SMR code for the major limiting accidents including transient behaviors due to desa-lination system disturbances. The analysis results employing conservative initial/boundary

conditions and assumptions show that the safety systems of the SMART conceptual designadequately remove the core decay heat and mitigate the consequences of the limiting accidents,and thus secure the plant to a safe condition. # 2000 Elsevier Science Ltd. All rights reserved.

Keywords: Integral reactor; SMART; Desalination plant; Design basis accidents; Safety analysis

1. Introduction

Recently, small and medium sized integral type advanced reactors for the diverseutilization of nuclear energy are getting much attention from the international

Annals of Nuclear Energy 28 (2001) 333±349

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* Corresponding author. Tel.: +82-42-868-8687; fax: +82-42-868-8990.

E-mail address: [email protected] (K.H. Bae).

nuclear community (IAEA, 1996, 1997). Large capacity power reactors are noteconomically viable for non-electric applications. These small and medium sizednuclear cogeneration reactors thus diversify the peaceful uses of nuclear energy inthe areas of seawater desalination, district heating, process heat generation, and shippropulsion.To develop an economically viable and safer advanced reactor both for seawater

desalination and small-scale electricity generation, the 330 MWt SMART develop-ment program was launched at KAERI in 1996 (Chang and Kim, 1997). The con-ceptual design of the SMART and its application system for seawater desalinationwas completed in March, 1999 (Chang et al., 1999) and the basic design is currentlyunderway. The SMART nuclear desalination plant aims to produce 40,000 m3/dayof potable water using approximately 10% of the total energy produced and gen-erate about 90 MW of electricity using the remaining energy.Since the SMART desalination plant is designed to operate with a di�erent

environment and objectives from the large capacity loop type commercial reactors,protection of the product water from possible contamination by radioactive mate-rials and minimization of the radiation release to the environment during accidentsare the principal safety aspects to be considered in the design. Therefore, it isrequired to simplify the design for operation and maintenance through a compactand modular design of the systems and components. Also, it is required to enhancethe safety and reliability by adopting inherent safety design characteristics andadvanced passive design features.Di�erent from the loop type commercial reactors, the SMART NSSS adopts the

design concept of containing most of the primary circuit components, such as thecore, four glandless canned motor main coolant pumps (MCPs), twelve helicallycoiled once-through steam generator (SG) cassettes, and the N2 gas pressurizer(PZR) in a single leak-tight reactor pressure vessel (RPV). Also, the design char-acteristics of the large, self-controlled gas PZR and the substantially large negativemoderator temperature coe�cient (MTC) to realize the soluble boron free core areinherent safety features. Due to these design characteristics, the SMART can fun-damentally eliminate the possibility of large break loss of coolant accidents(LBLOCAs), improve the natural circulation capability, and better accommodateand thus enhance resistance to a wide range of transients and accidents.In addition to these inherent safety design features, the safety goals of the

SMART are enhanced through highly reliable passive safety systems such as thepassive residual heat removal system (PRHRS), passive emergency core coolingsystem (ECCS), safeguard vessel, and reactor and containment overpressure pro-tection systems (Seo et al., 1996; Chang et al., 1998).The PRHRS cools down the NSSS by natural circulation for emergency situations

where normal steam extraction or feed water supply are unavailable, and providesthe core decay heat removal capability for 72 h without any operator action. TheECCS provides coolant to make-up the primary coolant inventory loss in case of asmall break loss of coolant accident (SBLOCA) by accumulators pressurized withN2 gas. The safeguard vessel completely encloses the RPV assembly and has thecapacity to entrap all the primary coolant leakage from the RPV. The reactor and

334 K.H. Bae et al. / Annals of Nuclear Energy 28 (2001) 333±349

containment overpressure protections (ROPS and COPS) are achieved by the safetyvalves and passive energy removal by the heat conduction mechanism through thecontainment wall and water jacket around the safeguard vessel.Considering these advanced design characteristics for the SMART, it is required

to perform systematic safety analyses in the conceptual design stage to secure thereliability and safety margin of the SMART. For the assessment of the SMARTsafety, major design basis accidents (DBAs) such as a steam line break, feedwaterline break, total loss of ¯ow, and small break loss of coolant accidents are analyzedemploying conservative initial and boundary conditions. These accidents arebelieved to envelope SMART DBAs. The following sections review the major pas-sive engineered safety systems of the SMART conceptual design and present thesafety analysis results for these limiting accidents using a MARS/SMR code which isa modi®ed version of MARS (Jeong et al., 1999) for the SMART performance andsafety analyses.

2. Safety systems of the SMART

Fig. 1 shows a schematic drawing of the safety systems of the SMART conceptualdesign. The SMART safety systems along with the multiple safety barrier designsuch as fuel, RPV, safeguard vessel, and the double containment prevent coredamage and minimize the radiation release to the environment during accidents.The major primary circuit components are housed in the RPV. The core and PZR

are located at the lower and upper part of the RPV, respectively. Four MCPs andtwelve SG cassettes are symmetrically arranged along the annular region betweenthe core support barrel and the RPV wall. After removing the heat in the core, thereactor coolant ¯ows upward through the upper core region, and enters the suctionheader of the MCPs. Passing via the MCPs, coolant is distributed to the shell side ofthe SG cassettes and transfers heat to the secondary coolant. The secondary sidefeedwater enters to the bottom of the SG, ¯ows upward inside the helically coiledtube to remove the heat from the shell side primary coolant and exits the SG assuperheated steam.The safeguard vessel is a steel-made, leak-tight pressure vessel housing the RPV,

pressurizer gas cylinders, ECCS tanks, and the associated valves and pipelines. Theprimary function of the safeguard vessel is to con®ne any radioactive release from theprimary circuit within the vessel under DBAs related to the loss of integrity of theprimary system. The containment is made of a steel structure with a concrete buildingaccommodating the safeguard vessel, makeup system pipelines, and PRHRS. Itsfunction is to con®ne the release of radioactive products inside the containmentunder the postulated beyond design basis accidents related to the loss of integrity ofthe safeguard vessel.The safety systems, such as the reactor shutdown system, ECCS, safeguard vessel,

PRHRS, ROPS, and COPS are depicted in this ®gure. These safety systems aredesigned to meet the redundancy and independency design requirements to ensurehigh reliability and safety. The safety systems designed to operate passively and

K.H. Bae et al. / Annals of Nuclear Energy 28 (2001) 333±349 335

credited for the safety analysis are described below and their major design para-meters are listed in Table 1.

2.1. Passive residual heat removal system (PRHRS)

The PRHRS passively removes the core decay heat by natural circulation foremergency situations when normal steam extraction or feed water supply is una-vailable. The PRHRS is designed with four independent trains and the operation ofany two trains will be su�cient to remove the decay heat. Each train has a heatexchanger submerged in an emergency cooldown tank (ECT), a compensating tank(CT), a check valve, remote control valves, and pipelines. The ECT is located highenough above the RPV to remove the core decay heat by natural circulation whenthe secondary system loses the capability for heat removal. The CT, pressurized with5 MPa N2 gas, makes up the initial inventory loss in the PRHRS pipe lines and SGregion, and ®lls up the voids formed during the cooldown process. A check valveinstalled on the pipeline between the CT and the heat exchanger is used to protectthe reverse ¯ow to the heat exchanger that prevents the natural circulation ¯ow.

Fig. 1. Schematic diagram of the SMART engineered safety features.


After a reactor trip of the initiating event, the main steam and feedwater isolationvalves are closed and the PRHRS cut-o� valves are opened to connect to the SG.The cold liquid in the feedwater piping enters the lower part of the SG, ¯ows upthrough a helical, once-through steam generator removing heat from the primarysystem, becomes superheated steam, exits to the steam line, condenses in the heatexchanger by the pool boiling heat transfer in the ECT, and returns to the SG. Twotrains have su�cient capability to remove the core decay heat for 72 h without anyoperator actions during the postulated design basis accidents.The SECS consists of four independent trains and the operation of any two trains

will be su�cient to remove the decay heat.

2.2. Emergency core cooling system (ECCS)

Since the SMART design inherently eliminates the potential for LBLOCA byeliminating a large size pipe penetration to the RPV, the ECCS is provided to pro-tect the core damage by the make-up of the primary coolant inventory during aSBLOCA. As summarized in Table 1, the ECCS consists of two independent trains

Table 1

Design parameters for the passive safety systems

Parameter Value

PRHRS

No. of trains 4

Heat removal/train 4.6 MW

CT total volume 2 m3

CT N2 gas volume 0.3 m3

CT pressure 5.0 MPa

ECT water volume 50.0 m3

ECCS

No. of trains 2

N2 gas pressure 10.0 MPa

Total volume/tank 5.0 m3

Water volume/tank 3.0 m3

Connected systems Makeup system

ROPS

No. of trains 2

POSRV open pressure 17.51 MPa

POSRV close pressure 13.86 MPa

COPS

Water jacket water volume 35 m3

Water jacket water temperature 50�CSafeguard vessel net free volume 400 m3

Safeguard vessel design pressure/temperature 3 MPa/250�CContainment net free volume 3000 m3

Containment design pressure/temperature 0.3 MPa/120�C


and the operation of one train will be su�cient to provide its function. Each trainincludes a 5 m3 cylindrical water tank pressurized with nitrogen gas at 10 MPa,isolation and check valves, and a pipe 20 mm in diameter connected to the RPVannular cover. When a SBLOCA occurs and the primary system depressurizes belowthe ECCS tank pressure, the coolant in the tank is injected into the upper annularcavity of the pressurizer. Each train of the ECCS piping shares the RPV penetrationpiping with the make-up system as well as the emergency boron injection system.

2.3. Reactor and containment over-pressure protection system

The function of the ROPS is to reduce the RPV pressure under the over-pressur-izing DBAs. The system consists of two parallel trains that are connected to thePZR through a single pipeline. These two trains are also connected to a singlepipeline connected to the internal shielding tank. Each train is equipped with aPOSRV. When the reactor pressure rises up to 17.51 MPa, the POSRV is openedand steam or the two-phase mixture is discharged into the internal shielding tankthrough the sparging device and then condensed.The safeguard vessel and containment over-pressure protection are accomplished

in a passive manner. The steam released by the loss of integrity of the primary sys-tem is condensed by conduction through a steel structure and the water jacketaround the safeguard vessel. The heat accumulated in the containment is removedby the conduction mechanism through the steel structure itself and through theemergency cooldown tanks installed inside the containment. A rupture disc is alsoprovided in the containment to protect the steel structure from over-pressure duringthe postulated beyond DBAs.

3. Safety assessment of the SMART

The safety of the SMART conceptual design adopting the inherent safety designfeatures and the passive engineered safety systems is assessed for the major limitingsafety related DBAs, such as a steam line break, feedwater line break, total loss of¯ow, and small break loss of coolant accidents.As discussed above, the passive engineered safety systems such as PRHRS, ECCS,

ROPS, and COPS play important roles to mitigate the consequences of these DBAsand thus ensure SMART safety requirements. The PRHRS is an ultimate passivedecay heat removal system for all DBAs. The SMART safety goal of a 72 h graceperiod without any operator action should be achieved by the PRHRS capability toremove decay heat for 72 h without core damage. The ECCS is a dedicated passivesafety system to mitigate the consequences of a postulated SBLOCA by reactorinventory make-up ensuring complete core coverage with the coolant. The ROPSand COPS are used to reduce the pressure and temperature of the reactor, safeguardvessel, and containment, and thus maintain their integrity throughout the accident.The ROPS is actuated to mitigate the consequences of overpressure accidents, suchas main feedwater line break (MFLB) and total loss of ¯ow (TLOF) accidents, and


the COPS is actuated for the main steam line break (MSLB) accidents and SBLO-CAs.

3.1. Analysis model

The computer code used for the analysis is MARS/SMR, which is a best-estimatethermal-hydraulic system analysis code based on a two-¯uid model for two-phase¯ows. This code is a developmental version of MARS (Lee et al., 1999) for theSMART safety and performance analyses. A number of SMART speci®c modelsre¯ecting the SMART's design characteristics, such as a helical tube SG, gas pres-surizer, and critical ¯ow with non-condensable gas, have been addressed in the code(Bae et al., 1999).The MARS code has been developed at KAERI by consolidating and restructur-

ing the RELAP5/MOD3.2.1.2 (INEL, 1998) and COBRA-TF (Thurgood et al.,1983), which has the capability of analyzing the one-dimensional and/or three-dimensional thermal-hydraulic system and the fuel responses of light water reactortransients. In the analysis of the accidents presented in this paper, a one-dimensionalanalysis method is used. Other codes employed for the analysis are MATRA (Yooand Hwang, 1998) and CONTEMPT4/MOD5/PCCS (Hwang et al., 1998) for thecalculations of the detailed DNBR of the hot sub-channel and the temperature andpressure of the containment, respectively. For the DNBR calculation of the tran-sient, the AECL-86 CHF Look-up table is used.Fig. 2 shows the SMART MARS/SMR nodalization for the safety analysis. To

well predict the physical phenomena expected during the transient, the system ismodeled in detail using 324 volumes and 346 junctions. The SMART core, com-

Fig. 2. SMART MARS/SMR nodalization.


posed of 57 industry proven KOFAs (Korean Optimized Fuel Assemblies) adoptingthe low-enriched 17�17 UO2 fuel rod array (Chang et al., 1999), is modeled as oneaverage channel having 56 fuel assemblies (P200), one hot channel (P210) for a hotassembly, and one core bypass region (P190). The fuel rods are modeled as 10 axialnodes and 7 radial nodes. Four MCPs (280, 281, 290, 291) are modeled separatelyand the PZR component is divided by the upper annular cavity (P300), intermediatecavity (P310), end cavity (P320), and the pipes (P305, P315) connecting them.Twelve SG cassettes are modeled as four sections and each section is divided by 17axial nodes to properly predict the heat transfer phenomena. Four trains of thePRHRS connecting the SG sections are modeled independently. Also, all of the heatstructures of the system are modeled.

3.2. Initial/boundary conditions and assumptions

Conservative initial and boundary conditions, as well as conservative assumptionsare employed to evaluate the safety envelope of the SMART conceptual design.Table 2 shows the major parameters used for the accident analyses.The initial core power and feedwater ¯ow rate are assumed to be 103% of the

nominal values considering the measurement uncertainty. The pressures of the PZRand SG steam are 15 and 3 MPa, respectively, and 10% of the SG tube plugging isconsidered as a design requirement. For the conservative results, the moderatordensity and doppler reactivity values are selected as the least or most negative onesdepending on the transient characteristics of the initiating accidents. The minimumshutdown rod worth with the most reactive rod stuck out is credited and a total roddrop time of 8 s is used. A conservative ANS-73 decay heat curve is used with a 1.2multiplication factor. The failure of one PRHRS train is considered as a single fail-ure assumption.

Table 2

Initial/boundary conditions and assumptions

Parameter Value

Core power 339.9

(103% of normal), MW

Feedwater ¯ow 157.075

(103%of nominal), kg/s

Primary coolant ¯ow rate, kg/s 1544

PZR pressure, MPa 15

SG pressure, MPa 3

SG tube plugging, % 10

Reactor trip total delay time, s 1.65

Shutdown rod worth, %�r ÿ8.9Control rod drop time, s 8.0

Decay heat curve 120% of ANS-73

Single failure 1 PRHRS train failure


3.3. Analysis results

3.3.1. Main steam line break (MSLB)Due to the thermal coupling between the SMART NSSS and the desalination

system, any transient of the desalination system can directly impact the reactorsafety. The increased steam ¯ow events due to the excess load on the desalinationsystem or steam line pipe breaks can cause the increase of heat removal by the sec-ondary system. The MSLB is a limiting accident for the decrease in heat removal bythe secondary system, and may occur as a result of thermal stress or cracking in themain steam line. A double-ended guillotine type break in the 160 mm steam pipesystem is assumed in the accident analysis. The rupture of a main steam line causesan uncontrolled steam blowdown and excessive heat removal from the rupturedsteam generator cassette, which results in a rapid cooldown of the primary system.The core inlet coolant temperature decrease combined with the large negative MTCcauses a core power increase, which results in a reactor trip on a high power signal.Simultaneous with the reactor trip, the turbine trips and the MCPs start to coastdown. Consequently, the main feedwater and steam isolation valves are closed, andthe PRHRS is connected to the intact SGs. In this steam line break accident onlytwo trains of the PRHRS are available for the system cooldown.For this accident, the hot channel DNBR is a major parameter of concern. Fig. 3

shows the hot channel DNBR behavior during the initial period of the transient withthe major related parameters of core pressure, core power, and core coolant ¯owrate. The primary pressure decreases with the initial cooldown of the primary cool-ant, whereas the core power increases by the positive reactivity insertion due to theincreased core coolant density, and reaches the high power trip setpoint of 115% at

Fig. 3. Transient behaviours of the major parameters (MSLB).


8 s into the transient. When the reactor, MCP, and turbine trips occur at about 10 s,the core power and the core inlet ¯ow decrease and the core pressure increases,respectively. The core pressure then decreases as the core power rapidly drops by thesu�cient insertion of the shutdown rods, and the PRHRS continuously removes thecore decay heat.The DNBR decreases as the core power increases, and then decreases more when

theMCPs start to coast down. The minimumDNBR of 1.334 is reached at 13 s, whichis higher than the SMART speci®ed acceptable fuel design limit (SAFDL) of 1.3.When the core power decreases to the decay heat level, the DNBR abruptly rises.Fig. 4 shows the long-term system pressures and primary system coolant tem-

peratures. As shown in this ®gure, the primary and secondary pressures are wellbelow the safety criteria of 110% of the design pressure, 18.7 MPa. Also, there is nopossibility of return-to-power during the accident due to a small water inventory ofthe once-through steam generator. Thus, for the MSLB accident at full poweroperation, the decay heat generated in the reactor core is well removed by two trainsof the PRHRS natural circulation.

3.3.2. Main feedwater line break (MFLB)A sudden stop of the steam ¯ow to the desalination system due to an unexpected

desalination system shutdown, turbine trip, and the loss of feedwater ¯ow to thesecondary system can cause a decrease in heat removal by the secondary system. TheMFLB is a limiting accident for the decrease in heat removal by the secondary sys-tem, and may occur as a result of thermal stress or cracking in the main feedwaterpipe system.

Fig. 4. Long-term system responses (MSLB).


In this safety analysis, a double-ended guillotine type break of 150 mm ID feed-water line piping between the feedwater isolation valve and the inlet of the steamgenerator is assumed. A rupture of the feedwater line piping causes a two phase mix-ture in the secondary system pipe line to discharge through the break, resulting in arapid decrease of the secondary system pressure and an increase in the primary pres-sure due to the reduced heat removal by the secondary system. The reactor trip occurson the high primary pressure trip signal followed by the instantaneous turbine andMCP trips. Upon a reactor trip, the main steam and feedwater isolation valves areclosed and the PRHRS isolation valves are opened to connect the PRHRS to pas-sively remove the core decay heat. Only two trains of the intact PRHRS are creditedto be operable assuming a single failure of one train of the PRHRS.Fig. 5 shows the major parameters a�ecting the hot channel DNBR. The initial

core pressure increase results in an increase of the moderator density, which in turncauses an increase in the core power by the positive reactivity. When the reactor andMCP trips occur at about 6 s, the core power and the core inlet coolant ¯ow ratedecrease rapidly. As shown in this ®gure, the DNBR decreases with the initial corepower increase. A minimum DNBR of 1.441 is reached at around 6 s. However, thisvalue is higher than the value of 1.3, which is a preliminary SAFDL of the SMARTconceptual design.Fig. 6 shows the primary and secondary pressures, and the primary coolant tem-

peratures during this accident. The primary pressure and temperature increaserapidly after the accident due to the reduction in heat removal by the secondarysystem. The opening of the PZR safety valves at 6 s alleviates the primary pressureincrease. The primary pressure reaches the maximum value of 18.37 MPa at around8 s into the transient. The primary pressure and the coolant temperature at the SG

Fig. 5. Transient behaviours of the major parameters (MFLB).


inlet decrease after the reactor trip. Initially, after the PRHRS is connected thecoolant in the PRHRS piping and heat exchanger piping of the emergency cooldowntank interrupts the natural circulation. The primary coolant temperature at the exitof the SG thus increases slightly even after the reactor trip. The system pressure andtemperature decrease monotonically after the natural circulation ¯ow is well estab-lished in the PRHRS loops, which assures the capability of the two trains of thePRHRS in mitigating the feedwater line pipe break accident.The results show that the peak primary and secondary pressures are well below

the safety limit of 110% of the design pressure, 18.7 MPa, with two trains of thePRHRS. Also, fuel damage due to the DNB is not anticipated during the transient.

3.3.3. Total loss of ¯ow (TLOF)The TLOF is a typical accident of decrease in the reactor coolant ¯ow rate and is

caused by a complete loss of power supply to all MCPs in operation. This accidentresults in a complete loss of forced circulation of primary coolant ¯ow and thusproduces the largest degradation in the DNBR margin than any other partial forcedreactor coolant ¯ow accident. The only credible mechanism for a loss of powersupply to all four MCPs is a loss of o�site power.Upon the loss of o�site power, the turbine trip, the feedwater ¯ow termination,

and the coast down of all MCPs occur concurrently. The feedwater ¯ow rate reachesthe low-¯ow trip setpoint (10% of the rated ¯ow rate) at 0.3 s into the transient, butan additional delay time of 1.15 s is assumed for the actual reactor trip. With thereactor trip, the main feedwater and steam isolation valves are closed and threetrains of the PRHRS are connected, accounting for a single failure assumption ofthe PRHRS.

Fig. 6. Long-term system responses (MFLB).


As the MCPs coast down, the coolant ¯ow through the core decreases, whichcauses an increase in the core average coolant temperature, which in turn causes adecrease in the core power by the negative MTC. The core power drops rapidly tothe decay heat level by the insertion of shutdown rods. The primary pressureincreases due to the reduction of heat removal by the turbine trip and decreaseswhen the PZR safety valves open and the core power decreases rapidly by the su�-cient insertion of shutdown rods at around 6 s.For this accident, the major parameter of concern is the minimum hot channel

DNBR. Fig. 7 shows the hot channel DNBR behavior drawn with the core pressure,core power, and core coolant ¯ow rate. The DNBR decreases with a decrease in thecore ¯ow rate and an increase in the coolant temperature and reaches a minimumvalue of 1.53 at 6.5 s when the core power rapidly decreases.Fig. 8 shows the long-term behavior of the system pressures and temperatures. A

turbine trip results in the secondary system steam ¯ow termination, which causes aheat-up of both the primary and secondary systems. The primary pressure (PZRTop) and temperature (SG Inlet) increase initially due to the reduced heat removalby the secondary system and decrease after the core power reduces rapidly by thesu�cient insertion of the shutdown rods into the core. With the reactor trip, themain feedwater and steam isolation valves are closed and three trains of the PRHRSare connected, accounting for a single failure assumption of the PRHRS. After thePRHRS is connected, the secondary system pressure increases for a period until thenatural circulation ¯ow is well established in the PRHRS loops.Afterward, the primary and secondary system ¯ows are safely transferred to the

natural circulation modes to continuously remove the core decay heat. After 150 sinto the transient, the natural circulation ¯ow in the PRHRS loops reaches 5.7% of

Fig. 7. Transient behaviours of major parameters (TLOF).


the initial feedwater ¯ow, and the primary and secondary coolant temperaturesdecrease at a rate of 90�C/s. The capacity of the three-train PRHRS is thus proved tobe su�cient for the removal of the core decay heat and the primary and secondarypressures are well below the safety limit of 110% of the design pressure, 18.7 MPa.

3.3.4. Small break loss of coolant accident (SBLOCA)Since the SMART has inherently eliminated the LBLOCA, only a SBLOCA is

postulated and thus su�cient time to mitigate the possibility of core uncovery isassured.The instantaneous guillotine rupture of the pipeline connecting the pressurizer end

cavity and the N2 gas cylinders is considered as the largest SBLOCA for theSMART. In this accident, the primary coolant is released via a pipe 20 mm in dia-meter into the safeguard vessel. The rupture of the gas cylinder pipe causes the N2

gas in the pressurizer end cavity to initially discharge through the break and thus theprimary system pressure rapidly decreases, as shown in Fig. 8. When the primarysystem pressure reaches down to the low-pressure trip setpoint of 12 MPa, thereactor is tripped at about 16 s into the accident. Simultaneously with the reactortrip, the turbine trips and the MCPs also trip with the assumption of a loss of o�sitepower. The PRHRS passively removes the core decay heat after the main feedwaterand steam isolation valves are closed by the reactor trip signal. Flashing due to therapid decrease in primary pressure, as well as the boiling in the core, steam and theN2 gas mixture discharge to the break resulting in a slower primary pressuredecrease as shown in Fig. 9.The ECCS actuates at around 80 s into the transient when the primary system

pressure decreases below 10 MPa, which compensates for the loss of primary

Fig. 8. Long-term system responses (TLOF).


coolant inventory. The continuous steam discharge through the break causes adecrease in the primary system inventory and an increase in the safeguard vesselpressure. When the primary system pressure decreases to the safeguard vessel pres-sure that is calculated using the CONTEMPT4/MOD5/PCCS code at around9300 s, the break discharge ¯ow ceases and no more system inventory loss occurs.

Fig. 9. System pressures and break ¯ow rate (SBLOCA).

Fig. 10. RPV water level and inventory (SBLOCA).


Fig. 10 shows the primary system water inventory as well as the reactor vessel col-lapsed water level. The actuation of the ECCS during a SBLOCA keeps the coolantlevel well above the top of the core. Also, the primary coolant temperature as well asthe hot spot cladding temperature do not exceed the initial value, as shown in Fig.11.Therefore, the capability of the ECCS with three trains of the PRHRS is proven to

adequately remove the core decay heat and mitigate the consequences of the acci-dent, and thus secure the plant to a safe condition even for the case of a postulatedSBLOCA.

4. Summary and conclusions

The performance of the passive engineered safety systems such as the ECCS,PRHRS, and reactor and containment overpressure protection systems of theSMART conceptual design are evaluated. For the assessment of the SMART safety,the major limiting accidents such as a steam line break, feedwater line break, totalloss of ¯ow, and small break loss of coolant accidents are analyzed including tran-sients due to desalination system disturbances using the MARS/SMR, MATRA,and CONTEMPT4/MOD5/PCCS codes.The results of the safety analyses using the conservative initial/boundary condi-

tions and assumptions show that the feedwater line break, steam line break, andSBLOCA give the maximum peak RCS pressure, minimum DNBR, and the mini-mum collapsed core level, respectively. The peak RCS pressure, SAFDL on mini-mum DNBR, and the minimum collapsed core level for these accidents are within

Fig. 11. Primary coolant and cladding surface temperatures (SBLOCA).


the design limits of 110% of the design pressure, 1.30, and no core uncovery,respectively.The primary and secondary system ¯ows are safely transferred to the natural cir-

culation modes after the reactor trip and the core decay heat is su�ciently removedby the PRHRS for all of the accidents. For the feedwater line break and total loss of¯ow accidents, the reactor overpressure protection system and the PRHRS maintainthe system pressure and DNBR within the preliminary safety limits of the SMARTconceptual design. The actuation of the ECCS with PRHRS during the SBLOCAkeeps the coolant and cladding temperatures well below the design limit, and thecoolant inventory is maintained well above the top of the core. Therefore, the safetysystems of the SMART conceptual design are evaluated to adequately mitigate theconsequences of these limiting accidents and thus secure the plant to a safe condi-tion.As the SMART design evolves, detailed passive safety system designs will be

®nalized, and the performance of these systems will be assessed in detail using thevalidated safety analysis methodology and computer codes.

Acknowledgements

This study has been performed under a contract with the Korean Ministry ofScience and Technology.

References

Bae, K.H., Chung, Y.J., Lee, G.H., Sim, S.K., Song, J.H., 1999. Safety Evaluation for the SMART

Conceptual Design. Technical Report, KAERI/TR-1396/99, KAERI, Taejon, Korea.

Chang, M.H., Kim, S.H., 1996. Korea's R & D program and activities for nuclear desalination. Pro-

ceedings of Symposium on Desalination of Seawater with Nuclear Energy, Taejon, Korea.

Chang, M.H, et al., 1998. Design characteristics of SMART for nuclear desalination. Proceedings of 9th

International Conference on Emerging Nuclear Energy Systems, Tel-Aviv, Israel.

Chang, M.M. et al., 1999. Development of Advanced Reactor Technology. Research Report, KAERI/

RR-1896/98, KAERI, Taejon, Korea.

Hwang, YD., Kim, YI., Bae, YY., Chang, MM., 1998. Development of the Containment Transient

Analysis Code for the Passive Reactor. Technical Report, KAERI/TR-1058/98, KAERI, Taejon, Korea.

IAEA, 1996. Design and Development Status of Small and Medium Reactor Systems 1995. IAEA-TEC-

DOC-881.

IAEA, 1997. Proceedings of Symposium on Desalination of Seawater with Nuclear Energy, Taejon,

Korea, 16±29 May.

INEL, 1998. RELAP5 Code Manual. NUREG/CR-5535, USNRC.

Jeong, J.J., Ha, K.S., Chung, B.D., Lee, W.J., 1999. Development of a multi-dimensional thermal-

hydraulic system code. MARS 1.3.1. Annals of Nuclear Energy 26, 1611±1642.

Seo, J.K., Yoon, J.H., Kim, J.P., Lee, D.J., Chang, M.H., 1996. Advanced Integral Reactor (SMART) for

Nuclear Desalination. IAEA-SM-347/40.

Thurgood, M.J. et al., 1983. COBRA/TRAC: A Thermal-hydraulics Code for Transient Analysis of

Nuclear Reactor Vessels and Primary Coolant Systems. NUREC/CR-3046, USNRC.

Yoo, Y.J., Hwang, D.H., 1998. Development of a Subchannel Analysis Code MATRA (Version Alpha).

Technical Report, KAERI/TR-1033/98, KAERI, Taejon, Korea.


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Safety evaluation of the inherent and passive safety features of the smart design