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International Journal of Nuclear EnergyScience and Technology

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IJNEST addresses original research, ideas and developments in all areas ofnuclear energy science and technology. Its scope embraces fundamentalsolid state physics, nuclear fuel reserves, fuel cycles and cost, materials,processing, system and component technology, design, optimisation, directconversion of nuclear energy sources, environmental control, reactorphysics, heat transfer, fluid dynamics, structural analysis, fuel management,future developments, nuclear fuel, safety, nuclear aerosol, neutron physics,computer technology, risk assessment, reactor thermal hydraulics,packaging and transportation.

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30/01/19 11.10International Journal of Nuclear Energy Science and Technology (IJN…blishers - linking academia, business and industry through research

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Topics covered includeReactor physics

Reactor research

Alternative reactor technologies

Radiation shielding

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Nuclear materials, neutron radiation effects in materials

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Simulating nuclear safety

17 October, 2018

Commercial operation of the CHASNUPP-1 996 megawatt intermediate type

pressurised water reactor began in May 2000 in Pakistan. It is a conventional two-loop

PWR and is run by the Pakistan Atomic Energy Commission. Now, scientists Khurram

Mehboob and Mohammad Aljohani of the Department of Nuclear Engineering at King

Abdul Aziz University in Saudi Arabia have carried out simulations of the activity of the

unit using MATLAB to probe the risks associated with a putative coolant leak that

might see radioactivity entering the environment. The team reports details of their

study in the International Journal of Nuclear Energy Science and Technology [...]

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01/09/18 05.40International Journal of Nuclear Energy Science and Technology (IJNEST) - Inderscience Publishers

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International Journal of Nuclear Energy Science andTechnology

This journal also publishes Open Access articles

Editor in Chief: Dr. Arturo Delfín LoyaISSN online: 1741-637XISSN print: 1741-63614 issues per yearSubscription price IJNEST addresses original research, ideas and developments in all areas ofnuclear energy science and technology. Its scope embraces fundamental solidstate physics, nuclear fuel reserves, fuel cycles and cost, materials,processing, system and component technology, design, optimisation, directconversion of nuclear energy sources, environmental control, reactor physics,heat transfer, fluid dynamics, structural analysis, fuel management, futuredevelopments, nuclear fuel, safety, nuclear aerosol, neutron physics, computertechnology, risk assessment, reactor thermal hydraulics, packaging andtransportation.

Honorary EditorMaïsseu, André, WONUC, France

Editor in ChiefDelfín Loya, Arturo, Instituto Nacional de Investigaciones Nucleares, Mexico(a_delfin_l hotmail.com)

Regional Editor Latin AmericaBarroso, Antonio Carlos De Oliveira, Instituto de Pesquisas Energéticas e Nucleares, Brazil

Honorary Advisory BoardChidambaram, Rajagopala, Bhabha Atomic Research Centre (BARC), India

Scientific and Editorial Committee ChairSzatmáry, Zoltán, Budapest University of Technology and Economics, Hungary

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Bojić, Milorad, University of Kragujevac, SerbiaCirimello, Roberto, Argentine Atomic Energy Commission, ArgentinaCoccoz, Guillermina D.H., CAC – CNEA, ArgentinaCuttler, Jerry M., , CanadaDe Alemeida Carvalho, Everton, WANO, UKEspinosa-Paredes, Gilberto, Universidad Autónoma Metropolitana-Iztapalapa, MexicoFoos, Jacques, CNAM, FranceFujii, Yasuhiko, Tokyo Institute of Technology, JapanGagarinski, Andrei, Kurchatov Institute, Russian FederationGhitescu, Petre, "Politechnica" University of Bucharest, RomaniaGomeno, Maria J., University of Zaragoza, SpainGreyvenstein, Gideon P., North-West University, South AfricaKinoshita, Chiken, Kyushu University, JapanKlein, Andrew C., Oregon State University, USAKondo, Shunsuke, University of Tokyo, JapanKruychkov, Eduard F., Moscow Engineering Physics Institute, Russian FederationLandsberger, Sheldon, University of Texas at Austin, USALatek, Stanislaw, National Atomic Energy Agency, PolandLeonidou, D. J., National Technical University of Athens, GreeceLung, Michel, , FranceMavko, Borut, University of Ljubljana, SloveniaMerle-Lucotte, Elsa, LPSC/ENSPG, FranceMinguez, Emilio, Universidad Politecnica de Madrid, SpainMonchaud, Serge, Technical University of Sofia, BulgariaNečas, Vladimír, Slovak University of Technology, SlovakiaNghiep, Tran Dai, Vietnam Atomic Energy Commission, VietnamPázsit, Imre, Chalmers University of Technology, SwedenPór, Gábor, Budapest University of Technology and Economics, HungaryPalamidessi, Hugo, Asociacion de Professionales de la Comision Nacional de Energia Atomica,ArgentinaPanella, Bruno, Politecnico di Torino, ItalyPonomarev-Stepnoi, Nikolai N., Academy of Sciences, Russian FederationSandquist, Gary M., University of Utah, USATaboada, Horacio, CNEA, ArgentinaTaleb, Safia, Université Djillali Liabès, AlgeriaTorgerson, David, Atomic Energy of Canada Ltd. (AECL), CanadaVanmol, Chris, CCMECL, BelgiumVapirev, Emil, University of Sofia and Technical University of Sofia, BulgariaVargas, Gustavo Alonso, National Institute for Nuclear Research, MexicoWilliams, R. A., University of Leeds, UKWisnubroto, Djarot S., National Nuclear Energy Agency - BATAN, IndonesiaYoshida, H., The Nagoya University Museum, JapanYu, Suyuan, Tsinghua University, ChinaZou'bi, Moneef R, IAS, Jordan

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International Journal of Nuclear EnergyScience and Technology

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Pages Title and authors

111-126

Estimation of radioactivity released from CHASNUPP-1 nuclearpower plant during loss of coolant accidentKhurram Mehboob; Mohammad S. AljohaniDOI: 10.1504/IJNEST.2018.10015404

127-138

The enhancement of energy gain in a p6Li inertial fusion reactorby laser-driven protonsJavad Bahmani; Baharak Eslami; Farhad Mohammad JafariDOI: 10.1504/IJNEST.2018.10015405

139-160

Uncertainty calculation in small break LOCA in the emergencycore cooling system connected to the hot leg of Angra 2 nuclearpower plantEduardo Madeira Borges; Gaianê Sabundjian; Francesco D'Auria;Alessandro PetruzziDOI: 10.1504/IJNEST.2018.10015406

161-171

Validating COMSOL multiphysics for VVER-1000 whole-core-steady-state via AER benchmark problemNed Xoubi; Abdelfattah Y. SolimanDOI: 10.1504/IJNEST.2018.10015407

172-195

An R-package for water and steam properties for scientific andgeneral useB.D. Baptista Filho; E.L.L. Cabral; A.C.O. BarrosoDOI: 10.1504/IJNEST.2018.10015408

196-212

Prediction of peak cladding temperature in a three-looppressurised water reactor with accident-tolerant fuel duringloss-of-coolant accidentAlexander AgungDOI: 10.1504/IJNEST.2018.10015409

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01/09/18 05.38Article: "Prediction of peak cladding temperature in a three-loop pre… of simulations show that during LB-LOCA with partially functioning

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Int. J. of Nuclear Energy Science and Technology » 2018 Vol.12, No.2 Title: Prediction of peak cladding temperature in a three-loop pressurised water reactor with accident-tolerant fuel during loss-of-coolant accident Author: Alexander Agung Address: Department of Nuclear Engineering and Engineering Physics, Faculty of Engineering, UniversitasGadjah Mada, Yogyakarta 55281, Indonesia Abstract: Safety analysis of a PWR fuelled with ATF (Accident-Tolerant Fuel) has been performed at LB-LOCA condition. The ATF being used is uranium silicide (U3Si2) and FCMF (Fully CeramicMicroencapsulated Fuel) with silicon carbide (SiC) and FeCrAl alloy as a cladding material. The objective ofthis research is to obtain dynamic characteristics of ATF-fuelled PWR at LB-LOCA condition. RELAP5-3Dsystem code was used to model the reactor and simulate the transient. A safe shutdown of the reactor wasassumed after a depressurisation following a double-ended guillotine breach in the main pipe. The results ofsimulations show that during LB-LOCA with partially functioning ECCS, the transient PCTs were far below themaximum allowable limit. The use of ATF could decrease the maximum transient PCT. It is shown that U3Si2fuel with FeCrAl cladding has the minimum PCT transient and the shortest quench time to steady statecondition after transient initiation. Keywords: nuclear safety; peak cladding temperature; accident-tolerant fuel; loss-of-coolant accident;RELAP5-3D. DOI: 10.1504/IJNEST.2018.10015409 Int. J. of Nuclear Energy Science and Technology, 2018 Vol.12, No.2, pp.196 - 212 Date of acceptance: 18 Jun 2018Available online: 10 Aug 2018

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196 Int. J. Nuclear Energy Science and Technology, Vol. 12, No. 2, 2018

Copyright © 2018 Inderscience Enterprises Ltd.

Prediction of peak cladding temperature in a three-loop pressurised water reactor with accident-tolerant fuel during loss-of-coolant accident

Alexander Agung Department of Nuclear Engineering and Engineering Physics, Faculty of Engineering, Universitas Gadjah Mada, Yogyakarta 55281, Indonesia Email: [email protected]

Abstract: Safety analysis of a PWR fuelled with ATF (Accident-Tolerant Fuel) has been performed at LB-LOCA condition. The ATF being used is uranium silicide (U3Si2) and FCMF (Fully Ceramic Microencapsulated Fuel) with silicon carbide (SiC) and FeCrAl alloy as a cladding material. The objective of this research is to obtain dynamic characteristics of ATF-fuelled PWR at LB-LOCA condition. RELAP5-3D system code was used to model the reactor and simulate the transient. A safe shutdown of the reactor was assumed after a depressurisation following a double-ended guillotine breach in the main pipe. The results of simulations show that during LB-LOCA with partially functioning ECCS, the transient PCTs were far below the maximum allowable limit. The use of ATF could decrease the maximum transient PCT. It is shown that U3Si2 fuel with FeCrAl cladding has the minimum PCT transient and the shortest quench time to steady state condition after transient initiation.

Keywords: nuclear safety; peak cladding temperature; accident-tolerant fuel; loss-of-coolant accident; RELAP5-3D.

Reference to this paper should be made as follows: Agung, A. (2018) ‘Prediction of peak cladding temperature in a three-loop pressurised water reactor with accident-tolerant fuel during loss-of-coolant accident’, Int. J. Nuclear Energy Science and Technology, Vol. 12, No. 2, pp.196–212.

Biographical notes: Alexander Agung is currently an Assistant Professor at the Department of Nuclear Engineering and Engineering Physics, Universitas Gadjah Mada, in Yogyakarta, Indonesia. From 2009 to 2014 he worked at Chalmers University of Technology, Sweden, as a research scientist. He received PhD degree in Nuclear Engineering from Delft University of Technology, The Netherlands, MSc degree in Energy Engineering from Royal Institute of Technology, Sweden and BEng degree in Nuclear Engineering from Universitas Gadjah Mada, Indonesia. His research interests are nuclear reactor physics, computational thermal-hydraulics, nuclear safety analysis, advanced reactor development and in-core fuel management.

Prediction of peak cladding temperature in a three-loop pressurised water reactor 197

1 Introduction

It has been proven that nuclear reactors are a reliable, sustainable and cost-effective source of electricity production. Worldwide operating nuclear power plants are mostly light water cooled reactor (LWR), 70% of which are of pressurised water reactor (PWR) type. In a typical 1000 MWe-class PWR, the average power density is around 104.5 MWth/m3 (Todreas and Kazimi, 2012), or approximately 10 times larger than that of fossil nuclear power plants. This feature is advantageous as the size of reactor core is smaller and the volume of nuclear fuel that is used is less than that of fossil power plants. On the other hand, larger power density creates substantial thermal stress. When the nuclear reactor is shutdown, 7% of the power is generated by the decay of radioactive fission product. The amount decreases to around 1% and 0.2% at 4 hours and 10 days after shutdown, respectively (ANSI/ANS, 2014).

When a loss-of-coolant accident (LOCA) occurs, the pressure of coolant system decreases drastically and core degradation might occur as decay heat generation continues although the reactor has been shutdown. When the cladding temperature exceeds 800°C, an exothermic reaction occurs between Zircaloy metal and water according to the following reaction (Baker and Just, 1962):

� � � � � � � �2 2 22 2 583.6 kJ/(mole Zr)Zr s H O g ZrO s H g� o � � (1)

The produced hydrogen can then be oxidised:

� � � � � �2 2 21 241.8 kJ/(mole Zr)2

H g O g H O g� o � (2)

Core degradation might propagate to a fuel melt, creating a corium which later could interact with structural concrete (Todreas and Kazimi, 2012). Metal oxidation in the corium takes place according to the following reaction:

� � � � � � � �2 2 22 2 598.2 kJ/(mole Zr)Zr l H O g ZrO s H g� o � � (3)

� � � � � � � �2 2 729.5 kJ/(mole Zr)Zr l CO g ZrO s C s� o � � (4)

Hydrogen production from a molten core and corium interaction with concrete depend strongly on the condition of an accident. The produced hydrogen may accumulate to an extremely high concentration. If the concentration of hydrogen gas and other conditions meet at a particular limit, a deflagration process may happen. In a specific condition, depending on the percentage of water steam, air and hydrogen, deflagration process might occur very rapidly and causes shockwave compression. This process is called detonation and may cause severe static and dynamic load to the containment (Pershagen, 1989). Such hydrogen detonation has occurred during Fukushima Daiichi accident in 2011.

As a result of a large tsunami following an earthquake off the Pacific coast of Tohoku in Japan, backup (emergency) generators that supplied electricity to the plant were immersed by sea water. This event caused a total station blackout (SBO) condition, meaning that decay heat removal process by cooling water could not take place. The next subsequential events led to a melting core, produced hydrogen massively and caused detonation (IAEA, 2015; Ishikawa, 2014).

198 A. Agung

Learning from Fukushima accident, intensive efforts to improve the level of nuclear reactor safety have been sought. From the viewpoint of nuclear fuel design, an exploratory investigation has been conducted to replace monolithic uranium dioxide (UO2) fuel and Zircaloy cladding with other material which can withstand accident conditions, including the design basis accident (DBA) and severe accident (SA). The replacement fuel in this context is called Accident-Tolerant Fuel (ATF). Three strategies exist in the development of ATF (Zinkle et al., 2014):

1 Modifying Zr metal alloy cladding in order to withstand oxidation, for example by using coating,

2 Replacing Zr metal alloy cladding with alternative oxidation-resistant cladding material,

3 Replacing monolithic ceramic oxide fuel with other alternative forms of fuels.

Adding coating material to the Zircaloy cladding requires the smallest change from manufacturing perspective. Some candidates of coating material that have been proposed include Cr (Brachet et al., 2014; Kim et al., 2015), ZrO2, TiAlN, Ti2AlC, Ti3AlC2 (Stewart, 2015) and SiC (Usui et al., 2015). Material researches performed within the last decade also lead to several material systems which resist oxidation at high temperature. Such materials, for example, are advanced steel FeCrAl (Tang et al., 2015; Yamamoto et al., 2015), refractory metal alloy like Mo (Nelson et al., 2014; Cheng et al., 2015) and SiC composite (Belgacem et al., 2014; Deck et al., 2015).

Some considerations to use alternative types of fuel have been put into practice, for example, by using TRISO particles within SiC matrix to form fuel pellet. Such type of fuel is called fully ceramic microencapsulated fuel (FCMF) (Terrani et al., 2012; Powers et al., 2013; Snead et al., 2014). The consideration to use TRISO as LWR fuel is based on its excellent track record as the fuel of high-temperature gas-cooled reactor (HTGR), whereby the retention to radioactive release from the fuel kernel and its thermal conductivity are much better than the monolithic UO2. Other alternatives are to use higher fissile density fuel to improve the economic value of ATF and to reduce the amount of accumulated enthalpy in the fuel. The chosen fuel for this case is nitrides (UN) and silicides which have large thermal conductivity (Harp et al., 2015; Jaques et al., 2015; Ortega et al., 2016).

Development of ATF from fuel design and manufacturing point of view has been in the third phase based on the technology readiness level (TRL) assessment, which reaches the verification of new fuel concept, including fabrication and sample irradiation test (Kurata, 2016; Carmack, 2014). In order to use ATF in a commercially operated LWR to replace the monolithic UO2 fuel with Zircaloy cladding, the ATF should have reached the ninth phase in the TRL assessment. Henceworth, numerous researches and developments should be performed before the ATF technology becomes mature for commercial operation.

From a nuclear reactor safety point of view, the TRL phase 3 poses many challenges to answer before the ATF is proven safe for commercial operation at normal, transient, DBA and beyond DBA (BDBA) conditions. Safety analysis should be performed to understand the influence of using ATF in the performance of reactor at accident situations. A double-ended guillotine large break loss-of-coolant accident (LB-LOCA) is chosen as a limiting DBA condition. The acceptance criterion for such accident scenario is that the peak cladding temperature (PCT) should not exceed 1470 K and no melting in


the centreline of fuel (TF < 3003 K) (US NRC, 10 CFR 50.46). For the purpose of this safety analysis, a thermal-hydraulic code, RELAP5-3D, is used as plant system simulator (Relap5-3D Development Team, 2015).

This article describes the activity to investigate the influence of using ATF as alternative fuel and cladding materials to safety-related issues. The accident was chosen as it is a limiting condition and it could give an insight into the capability of the emergency safety systems. The investigation was originally part of safety assessment performed by request of safety authority and later it was extended by the author by implementing the ATF to the model. The structure of this article is as follows. The description of the nuclear power plant unit is presented, followed by a description of the accident event. Modelling and simulation schemes are later described and the results of the simulations are discussed in the next section. Conclusions and possible future works are mentioned in the final section.

2 Description of the plant

The safety analysis described in this paper was performed using a typical 3-loop Westinghouse PWR. The fuel pins consist of monolithic UO2 fuel pellet and Zircaloy-4 as cladding. The technical data of the reactor is shown in Table 1 and flow diagram of the plant is given in Figure 1 (US-NRC, 1993). Table 1 Technical operation data of the reactor

Parameter Unit Value Nuclear steam supply system (NSSS) power MW 3160 Reactor power MW 3151 Reactor coolant flow kg/h 50.5 u 106

Reactor coolant pressure bar 155.1 Core outlet temperature K 598.35 Core inlet temperature K 556.45 Steam pressure Bar 60.61 Steam flow kg/s 1679.8 Steam temperature K 549.35 Feedwater temperature K 483.65

The primary cycle or the reactor cooling system (RCS) consists of a reactor, steam generators and reactor coolant pumps. Thermal output of PWR is determined by the size of the reactor and the number of loops in the primary system. The secondary system or the steam cycle starts from the shell side of the steam generator where feedwater is boiled using the heat transferred from U-tubes containing hot reactor-coolant.

Steam produced in the steam generator is directed to the high-pressure side of the main turbines through steam line isolation valves. Steam is then directed to the steam separator/reheater where excess moisture is reduced and steam is reheated using higher enthalpy steam from the main steam line. Reheated dry steam is then directed to low-pressure turbine and is discharged to the main condenser. Complementary steam line is also provided to bypass steam turbine and discharge steam directly to the main condenser at specified plant conditions.

200 A. Agung

Figure 1 Flow diagram of a typical PWR (US NRC Technical Training Center, 1993)

Steam is condensed in the condenser by redirecting the steam through pipes containing circulation water and is then collected in the hotwell section. The pump sucks condensate in the hotwell and redirects it to the suction part of feedwater pump through the tube side of feedwater heaters.

Feedwater heaters are used to increase plant efficiency. Main feedwater pump discharges water to the steam generator through water level controller valves and boiling takes place to produce steam. This completes the steam cycle and the process is restarted.

Accumulators contain a reserve of borated water maintained at pressure by an overblanket of nitrogen. They are connected to each cold leg of the main reactor coolant system (RCS). This component injects borated water into the RCS if loss-of-coolant accident (LOCA) occurs. When RCS pressure decreases below the level set for the accumulator, nitrogen will force borated water into the RCS to provide water for core cooling and boron to keep the reactor shutdown.

The residual heat removal system (RHRS) located inside the containment building has to serve two functions. The first function relates to a normal operational condition whereby the RHR is used to remove decay heat during reactor shutdown by pumping hot water in the RCS from the hot leg through the steam generator and back to the RCS through the cold leg. The second function operates during accident by pumping cold borated water from the refuelling water storage tank (RWST) to the RCS after a loss-of-coolant accident.

Safety injection system is part of the emergency system and is located inside the containment building. It also provides borated water for injection from the RWST to the RCS at LOCA. This system has smaller capacity but at higher pressure than RHRS.


The chemical and volume control system (CVCS) keeps the purity of the RCS. Several parts of the CVCS also serve to provide high-pressure borated water to the RCS during an emergency situation.

During LOCA, hot coolant water from RCS outflows to the containment and flashes into the steam in order that the containment pressure increases. The containment spray pumps will remove water from RWST to the containment through ring-shaped spray pipe inside the containment. Sprayed cold water will condense steam and return the containment pressure to its design limit. This is to prevent any damage to the containment building as well as to prevent any possibilities of releasing contaminated water spilled from the RCS.

The component cooling water (CCW) is used as a cooling medium for potentially radioactive components, such as heat exchangers and pump oil coolant. This system is a closed loop, cooled by service water system which receives water from river or lake near which the plant is located.

3 Description of accident event

LOCA in a PWR can be categorised according to the break size of the flow area. There are three categories, i.e., large break LOCA (LB-LOCA) with break size of at least 250 mm, medium break LOCA (MB-LOCA) with break size of 80 to 250 mm and small break LOCA (SB-LOCA) with break size less than 80 mm. To replace the loss of coolant from the system, one or more emergency core cooling systems (ECCS) is used, i.e., high-pressure safety injection (HPSI), accumulator and low-pressure safety injection (LPSI). HPSI and LPSI are actuated by signals indicating safety system injection, while accumulator will operate when the reactor pressure decreases below 4 MPa. When the injection phase is over, water recirculation is performed to remove long-term decay heat.

LB-LOCA starts by a postulated guillotine break in a cold-leg pipe of the primary coolant loop. The sequence of events can be categorised into four phases (Pershagen, 1989):

1. Blowdown phase, whereby rapid depressurisation and intensive break flow occur during 20–40 seconds,

2. Refill phase, when the break flow stops and supply water start to refill the reactor vessel. During this period, the reactor core contains steam and cooling process deteriorates in such a way that cladding temperature rises rapidly,

3. Reflood phase starts when the water level reaches the bottom of the reactor core. During this period, a maximum cladding temperature is reached around 1 to 2 minutes after the breach, and

4. Long-term cooling phase starts when the cladding temperature decreases to its normal level. Long-term cooling continues as long as necessary to ensure access of the core for fuel removal and maintenance activity can be started.

The breach of the pipe triggers a reactor scram and safety injection based on signals indicating low pressure in the pressuriser or high pressure in the containment. For the first 10–25 seconds, the pressure is low enough to trigger the accumulator to inject water. The LPSI system starts pumping water into the reactor after 20–30 seconds. The accumulator tank will deplete after around 50–100 seconds. The LPSI system continues

202 A. Agung

to supply water until the borated water storage tank is empty, which is predicted to occur within 20 minutes. The reactor operators should then redirect the LPSI to circulate water from the containment pool through the heat exchanging device in the RHRS.

During the blowdown phase, rapid depressurisation occurs until saturated pressure where water starts to boil is reached and break flow becomes limited. Blowdown phase is terminated for around 15 seconds when the pressures in the primary system and containment reach an equilibrium of around 0.4–0.5 MPa and the flow terminates. Before this condition is reached, the accumulator is actuated. During the blowdown phase, some of the injected water is prevented from reaching the core by reverse flow in the downcomer. This event is called an ECCS bypass. Some of the injected water then spill out through the breach.

The reactor vessel is then filled with water, and the core is initially flooded by water from the accumulator and then from the LPSI. During the refill and reflood phases, no bypass flow occurs, but water gets resistance from steam inside the core that has to be forced out of the core for the water level to increase. This phenomenon is called steam blockage and its occurrence becomes severe if the location of the breach is between the main coolant pump and steam generator, as the flow resistance reaches its maximum.

Critical heat flux can be achieved quite rapidly during blowdown phase. When water starts to boil, fuel rods are cooled down by a mixed flow of water and steam, and the cladding temperature reaches its maximum. When the core is exposed, the cooling process becomes worse until the fuel rods are wetted during the reflood phase and the cladding temperature reaches its maximum for the second time.

4 Acceptance criteria for the ECCS

The acceptance criteria for the ECCS in LWR state that the ECCS has to be designed and constructed in such a way that the following limitations shall be fulfilled during or after a postulated LOCA (US NRC 10 CFR 50.46):

1 Peak cladding temperature is less than 1204°C (2200°F).

2 The maximum cladding oxidation cannot exceed 17% of the total thickness of the cladding.

3 The maximum amount of hydrogen produced from water or steam reaction to cladding cannot exceed 1% of the amount of hydrogen produced if all cladding reacts chemically with water or steam.

4 Any changes in the core geometry should keep the configuration to be coolable.

5 Core temperature after injection phase should be kept low. Decay heat should be removed for long period of time.

5 Modelling

5.1 Plant model

The generic three-loop PWR was modelled using RELAP5-3D (Relap5-3D Development Team, 2015). The components in the primary coolant loop were almost modelled individually. Components in the secondary coolant loop, however, are not modelled


entirely. Components beyond turbine control valves and turbine bypass valves in the steam line, as well as components before feedwater control valves in the feedwater line, are not modelled.

Figure 2 Nodalisation of the primary side

Inside the reactor pressure vessel, the 157 fuel assemblies in the core region were individually modelled for the hydrodynamics and the heat structures. Those fuel assemblies were axially nodalised into eight levels. The core and the downcomer were split into three parallel channels to retain the structure of the three-loop primary coolant side. The core was also complemented by three bypass channels per loop to represent the baffle-barrel space, the open guide thimble and the flow path at the core-periphery, respectively.

204 A. Agung

A fuel assembly was modelled with a heated channel, in which the heat source for the fuel rod is provided by the neutron point kinetics available in the RELAP5-3D. A fuel element has nine radial mesh points.

The current nodalisation of the generic PWR also includes several major parts, such as three cold legs and three hot legs, residual heat removal systems, main circulating pumps, steam generators with vertical heat exchanger tubes, a pressuriser model with spray systems and electrical heaters, feedwater systems, simplified turbine models, as well as steam dumping lines. Figure 2 shows the nodalisation of the primary side, while Figure 3 shows the nodalisation of the secondary side.

Figure 3 Nodalisation of the secondary side


The control and protection systems were modelled accordingly. The level control system maintained the level in the steam generators. The input signals were fed into the controller and the output signals determined the stem position of the feedwater control valves in the model.

The pressure and level of the pressuriser were controlled by their control components. The spray system was actuated in case of excessive primary pressure and at a lower pressure the heaters were actuated accordingly.

To account for a double-ended gouillotine break, three valves were modelled, i.e. one isolation valve (component #288) and two break valves (component #950 and #960). Those valves were located in the cold leg of loop number 2 between the ECCS injection point and the reactor vessel inlet nozzle. The two break valves were connected to a single volume, representing a containment.

5.2 Thermophysical model

Within the scope of this study, three types of material were used for the fuel, i.e. the monolithic UO2, uranium silicide (U3Si2) and FCMF (Fully-Ceramic Micro-encapsulated Fuel) and for the cladding, i.e. Zircaloy-4, silicon carbide (SiC) and FeCrAl alloy. The thermophysical data for UO2 and Zircaloy-4 has been provided by default in RELAP5-3D. Thermophysical data for uranium silicide (White et al., 2015), FCMF (Lee and Cho, 2015), SiC (Snead et al., 2007) and FeCrAl alloy (Ott et al., 2014) were put into the corresponding RELAP5-3D input deck.

5.3 Simulation setup

Within current LOCA analysis, Ransom-Trapp critical flow was activated in the model. The time stepping was chosen to be automatic between semi-implicit scheme and nearly-implicit scheme to handle the Courant limit with the minimum and maximum time steps being 10–7 and 0.025 s, respectively, as LOCA analysis prone to have demanding fast transient calculation.

The simulations were split into two steps, i.e. the steady-state part and the transient part. The steady state part was performed to obtain a steady condition for all parameters in the plant. This step was calculated for 300 seconds. During this steady state part, the isolation valve (#288) was open, while the two break valves (#950 and #960) were closed. A transient calculation was subsequently performed by restarting the simulation based on the steady-state parameters obtained in the previous step. This transient step was calculated for 350 seconds.

The transient was started by closing the isolation valve and opening the break valves, leading the working fluid to flow into the containment. The reactor was scrammed and the subsequent power generation was coming from the decay heat, being tabulated based on ANSI/ANS 5.1-2014. A worst single failure scenario was assumed for the ECCS, i.e. a failure of one out of the three HPSI systems and one out of the two LPSI systems was not functioning. The initiation of the injection system was triggered based on predefined setpoints of low-low signals of the pressuriser. The secondary system was isolated from the beginning of the transient to account a conservative assumption.

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6 Results and discussions

This study simulates the LB-LOCA for a set of combinations between fuel materials (UO2, U3Si2 and FCMF) and cladding materials (Zircaloy-4, SiC and FeCrAl alloy). For each combination, the simulation results are presented, i.e. the fuel centreline temperature and the inner cladding temperature, as shown in Figure 4 to Figure 12. Both parameters are considered at eight axial nodalisation positions, i.e. #1 is at the bottom and #8 is at the top.

Figure 4 Centreline fuel temperature (left) and inner cladding temperature (right) of UO2 fuel and Zircaloy-4 cladding

Figure 5 Centreline fuel temperature (left) and inner cladding temperature (right) of UO2 fuel and SiC cladding

It is shown that the highest fuel and cladding temperature of all material combinations during the transient occurs at node 6. This is consistent with the fact that the heat generation source has the shape of a slightly upward-skewed sinusoid. The highest fuel


temperature during the transient can be found for the combination of UO2 fuel and Zircalloy-4 cladding. Furthermore, the maximum temperature of UO2 fuel exceeds 1000°C regardless of the type of cladding being used. The use of U3Si2 (Figures 7–9) and FCMF (Figures 10–12) reduces the temperature significantly, although, for Zircalloy-4 cladded fuel, the temperatures are still close to 1000°C. It is shown in Figure 12 that the combination of FCMF and FeCrAl alloy cladding gives the smallest increase in the maximum fuel temperature.

Figure 6 Centreline fuel temperature (left) and inner cladding temperature (right) of UO2 fuel and FeCrAl alloy cladding

Figure 7 Centreline fuel temperature (left) and inner cladding temperature (right) of U3Si2 fuel and Zircaloy-4 cladding

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Figure 8 Centreline fuel temperature (left) and inner cladding temperature (right) of U3Si2 fuel and SiC cladding

Figure 9 Centreline fuel temperature (left) and inner cladding temperature (right) of U3Si2 fuel and FeCrAl alloy cladding

Figure 10 Centreline fuel temperature (left) and inner cladding temperature (right) of fully ceramic microencapsulated fuel and Zircaloy-4 cladding


Figure 11 Centreline fuel temperature (left) and inner cladding temperature (right) of fully ceramic microencapsulated fuel and SiC cladding

Figure 12 Centreline fuel temperature (left) and inner cladding temperature (right) of fully ceramic microencapsulated fuel and FeCrAl alloy cladding

Figure 13 shows the PCT as a function of time for the duration of the transient being simulated. Simulation results from all combinations of fuel and cladding materials show that the PCT is lower than the maximum permissible value. This shows that the safety system (accumulator, HPSI and LPSI) is well-functioning to prevent cladding material from damaging.

The combination of UO2 and Zircaloy-4 cladding gives the highest PCT as compared to other material combinations. UO2 fuel also produces high PCT although it is combined with SiC and FeCrAl alloy as cladding. This is consistent with the result of maximum fuel temperature. The decrease in cladding temperature to around 400 K takes the longest time for UO2 fuel. The longest quench time being 375 seconds for UO2-Zr fuel.

The smallest increase in PCT is shown by U3Si2 fuel with FeCrAl cladding, in which the maximum transient PCT is 900 K. The U3Si2 fuel with SiC cladding also shows a similar maximum value of transient PCT, but the quench time to reach steady condition takes place 3 seconds longer. For U3Si2-Zr material, it is shown that the silicide fuel is

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able to reduce the maximum PCT about 150 K, but the quench time to reach steady state is 50 seconds longer than that of the SiC or FeCrAl-cladded fuel. This shows the thermal inertia of Zr is relatively higher.

In general, the performance of FCMF is in between that of UO2 and U3Si2 fuels. This is valid for the maximum PCT as well as for the quench time to reach steady state.

Figure 13 Peak cladding temperature during the duration of transient

7 Conclusions

From the simulation of LB-LOCA performed for different combinations of fuel and cladding materials, several conclusions can be drawn as follows. During LB-LOCA with a well-functioning ECCS, the transient PCT is below the maximum permissible level. UO2 fuel with Zr cladding shows the highest PCT compared to other material combinations being considered in this study. The use of accident tolerant fuel (ATF) is able to reduce the maximum PCT. The U3Si2 fuel with FeCrAl cladding has the smallest PCT with a maximum value of 637°C and the shortest quench time to reach steady state, i.e. around 215 seconds after initiation of the accident.

The model does not consider a full spectrum of malfunction of safety injection systems. However, a worst single failure scenario was assumed to account malfunction in ECCS, i.e. one out of three HPSI systems and one out of two LPSI systems are not functioning. The accumulators, however, are still functioning. For this reason, it is proven that no safety violation occurred even with the regular fuel. The use of ATF, in this case, gives a more relaxed safety margin.


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