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Evaluation of accident tolerant cladding materials in a severe accident of the BNPP
Annals of Nuclear Energy 129 (2019) 214–223
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Evaluation of accident tolerant cladding materials in a severe accident of the BNPP H. Ebrahimgol, M. Aghaie, A. Zolfaghari, Gh. Alahyarizadeh Engineering Department, Shahid Beheshti University, G.C, P.O. Box: 1983963113, Tehran, Iran
a r t i c l e
i n f o
Article history: Received 11 April 2018 Received in revised form 2 November 2018 Accepted 3 February 2019
Keywords: Accident tolerant fuels BNPP SBO accident SiC FeCrAl
a b s t r a c t Accident Tolerant Fuels (ATFs) are fuels or fuel clads with improved features in comparison with standard UO2-Zircaloy in commercial LWRs. The ATFs could tolerate in vessel loss of coolant accidents. IronChromium-Aluminum (FeCrAl) alloy and Silicon carbide (SiC) are being introduced as enhanced cladding candidates to mitigate accident consequences. These cladding materials have slower oxidation kinetics and high resistance to core degradation which reduce hydrogen accumulation in severe accidents. This paper deals with the evaluation of ATFs in hydrogen mitigation during Station Blackout Accident (SBO). In this analysis, a numerical model based on MELCOR for Bushehr Nuclear Power Plant (BNPP) in SBO is prepared and replacement of ATF cladding materials (FeCrAl and SiC) is studied. The thermal hydraulic response of the BNPP and core heat-up with clad oxidations are evaluated and the hydrogen generation due to the pernicious interaction of hot steam and cladding materials for FeCrAl, SiC and zircaloy-4 are calculated. The results show the total hydrogen accumulation in the SBO of the BNPP are reduced 68% and 47% for FeCrAl, SiC, respectively. The accumulated energy in the core are reduced 88% and 81% for FeCrAl and SiC, respectively and the debris generation and lower head failure are improved, significantly. The results illustrate the lower hydrogen generation and oxidation kinetics in case of using the ATF materials and the performances are demonstrated. Ó 2019 Published by Elsevier Ltd.
1. Introduction The Fukushima Daiichi accident renewed attention to the pernicious interaction between zirconium alloys and hot steam during severe accidents. Finding applicable solutions to reduce accident consequences such as hydrogen production and core degradation due to loss of coolant in core have been evaluated by researchers at nuclear laboratories and universities considerably (Terrani et al., 2013; Gauntt et al., 2005a). Past SiC investigation has focused on the high temperature gas reactor tri-structural-isotropic (TRISO) fuel and fusion reactor structural components (Yueh and Terrani, 2014), but nowadays interest of using fully ceramic SiC for LWR cladding is significantly increasing (Bragg, 2012). Recently many projects at nuclear laboratories such as Idaho National Laboratory with US Nuclear Regulatory Commission (NRC) have been focused on ATFs development in LWRs (Carmack and Goldner, 2014; Merrill et al., 2017). One of the main proposed solutions to mitigate the consequences of the severe accident is ATF materials for reducing the rates of heat and hydrogen generation due to cladding oxidation with hot steam
E-mail address: [email protected] (M. Aghaie) https://doi.org/10.1016/j.anucene.2019.02.008 0306-4549/Ó 2019 Published by Elsevier Ltd.
(Terrani, 2014). During 2012, several candidate cladding materials (such as 310 SS, FeCrAl, and SiC) were demonstrated to have significantly lower oxidation kinetics and hydrogen generation (Pint et al., 2013). The crucial approaches for development of ATFs are improvement of fuel properties, enhancement of cladding properties to maintain core coolability and retain fission products and improvement of reaction kinetics with steam to minimize enthalpy and hydrogen generation (Farmer et al., 2014). The SiC concept is expected to gain more benefits than zirconium alloy, smaller neutron absorption cross section, lower chemical interaction, ability to endure higher fuel burn-ups and higher temperatures, exceptional inherent radiation resistance, lack of progressive irradiation growth, and low induced activation/low decay heat (Katoh et al., 2012). FeCrAl alloy, have the ability to enhance oxidation resistance by creating a protective oxide layer; for instance, the aluminum found in FeCrAl undergoes a reaction with oxygen to produce an oxide layer (Al2O3) capable of oxidation kinetics reduction in hot steam environments (Terrani et al., 2013). Implementation of new cladding material in MELCOR was tried at Idaho National Laboratory (INL) in 2012. The INL applied these materials to simulate the Three Mile Island Unit 2 (TMI-2) lossof-coolant accident (Merrill, 2012). The TMI-2 accident was SBLOCA in a two-loop Babcock and Wilcox (B&W) pressurized
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water reactor (PWR) (Gauntt et al., 2002). In 2016 an effort was initiated for replacing zircaloy with SiC and ferritic stainless-steel alloy (FeCrAl alloy) (Merrill et al., 2017). Recently some experimental studies related to the performance of proposed ATF concepts in steam environments at high temperature and pressure have been performed (Pint et al., 2013; Cheng et al., 2012). As a result of these studies, important kinetic parameters have become available for ATF materials. Based on recent experimental researches, ATFs performance modeled for a range of severe accidents was analyzed at Fuel Cycle Research & Development Advanced Fuels Campaign (Robb et al., 2016). To develop investigation of the accident tolerant reactor concept, the Georgia Institute of Technology was asked to conduct an integrated research project separately (Lindley et al., 2016). Other research activities in the university of Illinois and the university of Manchester are being conducted to discuss two pathways in order to enhance the modification of monolithic zircaloy cladding as the foundation of ATF cladding (Yueh and Terrani, 2014). After decades of using zirconium cladding components in LWRs reactors, the transition to using ATF materials represents a revolutionary prototype shift. Due to the impact associated with any such transition, associated challenges will need to be evaluated via predictive fuel performance analysis. In this paper, two candidates of ATF cladding materials (FeCrAl and SiC) as appropriate substitutes of zircaloy-4 during station blackout in BNPP have been studied. Using the MELCOR code, the core materials such as fuel clad and required number of thermophysical characteristics are replaced and a model for BNPP during the SBO is prepared. For accurate evaluation of the materials performances, the updated characteristics such as oxidation kinetics are prepared. The core heat up and thermal hydraulic assessment of the core is evaluated against Final Safety Analysis Report (FSAR, 2007) of Bushehr nuclear power plant. The effect of the ATF materials in thermal hydraulic response of the plant is evaluated and in core meltdown, the hydrogen generation, energy accumulation in core, debris and lower head creep rupture are calculated. The purpose of this paper is to demonstrate the impact of the advanced cladding materials on the fuel performance on the BNPP (WWER1000 reactor) under the beyond design basis accidents (BDBAs). 2. The BNPP description and model nodalization The plant of interest to analyze is a four-loop pressurized water reactor WWER1000-V446 (Located in Bushehr, Iran) with 3000 MW thermal power generation. The main distinguishing features of the WWER compared to other PWRs are; horizontal steam generators, hexagonal fuel assemblies, no bottom penetrations in the pressure vessel and high-capacity pressurizers providing a large reactor coolant inventory.
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The BNPP reactor core consists of 163 fuel assemblies in hexagonal design and different fuel enrichment. A part of fuel assemblies contains burnable absorber rods (BAR) bundles or absorbing rods of control and protection system (CPS), which are moved by the CPS drives. The fuel bundle has a shape of hexahedron in section with maximum dimension of 235.1 mm between the faces. In the bundle cross section there were placed 311 fuel rods, 18 guiding channels for positioning CPS or BAR, central channel and channel for in-core instrumentation detector (ICID). Each loop has cold leg and hot leg pipes connected to reactor vessel nozzles, a main coolant pump to circulate the reactor coolant system with nominal capacity of 21,200 m3/h and 0.624 MPa head and steam generators (SG) type PGV-1000 M(B) is intended to produce dry saturated steam of 6.27 MPa pressure, in the form of one vessel with horizontal arrangement of tubes constituting the heat exchanging surface. One pressurizer (PRZ) is connected to the undetachable part of the single hot leg of main circulation loop by connecting pipeline to control of loops pressure. The relief system connected to PRZ is intended to remove steam, steamwater and steam-gas mixture from the pressurizer during operation of the emergency gas removal system and/or excessive pressure protection system. Secondary loop is designed as SG source and sink with its safety relief valves. They have been disconnected at the beginning of the SBO accident as described in Table 1. The reactor pressure vessel (RPV) is modeled in 5 control volumes: down-commer, bypass, lower head, core, upper plenum. The reactor core is separated in 5 rings and 19 axial levels. The BNPP nodalization model is shown in Fig. 1. 3. Simulation of the SBO accident in the BNPP In this section, the SBO accident is simulated and results are compared with FSAR (2007). The BNPP’s final safety analysis report (FSAR) includes the SBO accident until core heat up. During the station blackout, due to loss of both the normal A.C. power supply and the emergency diesel generators, core heat sink will be lost. The sequences of events in this accident are reported in Table 1 (FSAR, 2007). In this work, simulation of the SBO accident revealed the following occurrences. At the beginning of the accident, reactor control and protection system (RCP) is tripped and scram occurs. The reactor power decreases due to onset of control rods motion. Fig. 2 depicts the reactor core power during scram. It is clear that the reactor power decreases to decay heat level significantly at the beginning of the insertion and follows the decay heat generation in next. The reactor coolant pump sets switch off and all blow-down makeup systems of the primary system fail. The pressurizer power supply system is disconnected and BRU-K valves designed to dump steam from the steam generator into the turbine condenser close.
Table 1 Sequence of system and device actuation (FSAR, 2007). Time, s (FSAR)
Time, s (MELCOR)
Event
Interlocks, set point for actuation or other cause
0.0
0.0
Loss of all A.C. off-site and on-site power supply sources (power unit blackout)
0.6 1.4 1.7 7.0 2800. 7000. 10000.
0.6 1.4 – 7.0 3200. 7100. 10000.
Trip of all RCP sets Trip of the main and auxiliary feed water systems of the secondary side Trip of makeup-blowdown system of the primary system BRU-K disconnection Disconnection of PRZ system power supply Closing the turbine generator stop valves Scram signal generation The onset of control rod motion BRU-A opening SG drainage Onset of the core heat-up End of calculation
Turbine emergency protection action NPP blackout Emergency protection action Reaching SG pressure of 7,15 MPa
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Fig. 1. Nodalization scheme of the BNPP.
Fig. 2. Reactor power.
Fig. 3. SG secondary pressure.
Immediately, the main and auxiliary feedwater systems of the secondary sides are closed in order to protect the turbine. As shown in Fig. 3, the heat transfer between the primary and secondary sides causes vapor generation inside SG shell and the secondary side pressure increases. The secondary side pressure is maintained between 6.27 MPa and 7.15 MPa by steam dump valves. As Fig. 4 presents, the SGs water inventory dry out completely in the absence of secondary side feedwater. Discharge through the dump devices decrease SG water inventory until 3150 s and finally, SG will lose its heat removal ability. SGs drying results in a slight primary side pressure increasing. Once SGs dry out, the primary pressure increases drastically up to it reaches the PRZ safety relief valves setpoint. Coolant discharge from safety relief valves results loss of the primary coolant, and the core is further heated-up. Without acceptance of management measures, damage occurs to the core and vessel at high pressure. The pressure variation in primary side of the BNPP during SBO accident is shown in Fig. 5. In first seconds, due to the trip of all RCP sets and reactor scram, the primary pressure reduces sharply.
Fig. 4. SG water level.
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Fig. 7. Coolant flowrate at the reactor inlet. Fig. 5. Coolant pressure in Pressurizer.
In following, loss of coolant inventory in shell side of SGs results pressure increasing, slightly. By SGs drainage and coolant heat up, primary pressure and PRZ water level increase. While the primary pressure reaches to 18.1 MPa, PRZ safety relief valves act to remove excessive energy by coolant discharge to relief tank and PRZ water level decreases. This valve is closed once the primary pressure drops to 17.2 MPa. The water level in the PRZ is reported in Fig. 6. Coolant evaporation in primary side and coolant discharge via relief valves lead to water level decrement as depicted in Fig. 6. Figs. 5 and 6 present the PRZ relief valve actions during the SBO accident. The relief valve setpoints (17.2&18.1 MPa) make the pressure oscillations as depicted in Fig. 5. Fig. 7 presents the coolant flow rate in primary side during the SBO. Coolant flowrate with loss of all A.C. powers decreases significantly at the first seconds and continues slightly until 5000 s as a result of buoyancy driven natural circulation. Finally, the coolant circulation stops completely during 200 s. Fig. 8 shows water volume in reactor pressure vessel. Once coolant circulation fully stops, coolant heat up occurs and in-vessel water evaporation begins. The pressure vessel coolant dries until 7000 s. Once the coolant flow in primary side stops (5200 s), the temperature at the reactor outlet increases and the RPV water inventory decreases. The RPV drainage accelerates and temperature rises as presented in Fig. 9. In SBO accident, the fuel cladding temperature increases slightly due vapor generation in the RPV. The trend of vapor tem-
Fig. 8. Reactor water volume.
Fig. 9. Coolant temperature at reactor outlet.
Fig. 6. PRZ level.
perature at reactor outlet is similar to cladding temperature and it is increasing. Temperature increment of fuel cladding continues with reactor vessel drainage until the fuel rods fail. As Fig. 10 depicts, once core heat removal fails the cladding temperature increases slightly. As soon as RPV dries out, the core is heated up and the cladding temperature increases significantly until the fuel rod fails at 9200 s. The accumulated heat in core is due to decay heat and cladding oxidation.
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Fig. 12. Density. Fig. 10. Maximum temperature of fuel rod claddings.
4. Requirements for modification of cladding materials This investigation is performed by MELCOR code. This code is developed to analyze light water reactors under severe accidents which allow us to alter cladding material and consider the behavior of this alteration under off-normal conditions. The default cladding material in the MELCOR code is zircaloy. In this work, the material properties specified for zircaloy are modified to survey different cladding materials. The following explicate how the material properties in MELCOR are modified to depict the advanced materials and the modeling limitations encountered. To define new cladding materials to the MELCOR, the candidate cladding materials including SiC and FeCrAl are overwritten. The new clads are assumed to have the same geometry and dimensions as zircaloy clad in the BNPP. The modifications include core oxidation kinetics and material property routines consist of thermal conductivity (Fig. 11), density (Fig. 12), specific heat (Fig. 13), enthalpy (Fig. 14), latent heat of fusion (Table 2) and emissivity. Thermophysical properties of ATF materials will be discussed in next section.
Fig. 13. Specific heat.
4.1. Thermophysical properties of ATF materials 4.1.1. The FeCrAl and oxidized FeCrAl thermophysical properties The FeCrAl thermal conductivity increases along with temperature increment and for temperatures higher than 1700 K remains constant (Wang et al., 2017). The thermal conductivity of oxidized FeCrAl is assumed to be 20 W/mK as displayed in Fig. 11. The oxidized FeCrAl alloy consists of Fe3O4, Al2O3 and Cr2O4 are formed
Fig. 14. Enthalpy.
Table 2 ATFs latent heat of fusion (Wang et al., 2017; Richet et al., 1982).
Fig. 11. Thermal conductivity.
Latent heat of fusion
Value (kJ/kg)
FeCrAl FeCrAl Oxide SiC Silica (SiO2)
260 580 360 930
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after core uncovery. The FeCrAl density descends from 8000 kg/m3 to 7000 kg/m3 between 300 K and 1500 K and will be remained 7000 kg/m3 all over the ranges. The density of oxidized FeCrAl is 5000 kg/m3 as shown in Fig. 12. The Specific heat of the FeCrAl alloy increases almost linearly with temperature increment up to 1600 K then remains constant. The specific heat of oxidized FeCrAl is estimated to be 500 J/kg. K as illustrated in Fig. 13 (Wang et al., 2017). The enthalpy of FeCrAl and oxidized FeCrAl increases linearly except in temperatures between 1500 K and 2000 K as displayed in Fig. 14. The latent heat of fusion of FeCrAl is estimated to be 260 kJ/K and the latent heat of fusion of oxidized FeCrAl is set to 580 kJ/kg. The melting point of FeCrAl alloy is 1600 K and the melting point of oxidized FeCrAl is around 1800 K, as shown in Table 3 (Wang et al., 2017). The emissivity of oxidized FeCrAl is set to 0.70 (Robb et al., 2016). 4.1.2. The SiC and SiC oxide thermophysical properties The SiC thermal conductivity decreases considerably below temperature 1300 K then reduces mildly up to 5000 K. For SiO2, this value increases before reaching 1400 K, then remains constant, as presented in Fig. 11. The density of SiC and SiO2 are assumed to be 2900 kg/m3 and 2000 kg/m3, respectively as shown in Fig. 12. The specific heat of SiC increases below temperature 2500 K then decreases slightly. For temperature below 1200 K, the trend of specific heat of SiO2 is similar to SiC. For temperatures above 1500 K due to lack of experimental data, the linear extrapolated values to 2000 K are considered as presented in Fig. 13 (Merrill,2012). The melting temperature of SiC and SiO2 are 2900 K and 1873 K, respectively. The enthalpy progression of SiC and SiO2 along with the increase of temperature. The significant increase in enthalpy of SiC is around 2900 K where the SiC phase change takes place. It is around 1873 K for SiO2 shown in Fig. 14. The latent heat of fusion of SiC and SiO2 are assumed to be 360 (kJ/kg) and 930 (kJ/kg), respectively (Table 2). The SiC cladding inner surface emissivity is set to 0.75 (Merrill, 2012).
The FeCrAl oxidation is calculated using standard parabolic kinetics, with appropriate rate constant expressions similar to zircaloy. The square of the thickness of the oxidized FeCrAl scale formed x is proportional to exposure time t as Eq. (1) (Merrill et al., 2017).
x2 ¼ k p t
4.2.1. The FeCrAl oxidation kinetics Ferritic stainless-steel alloys, such as FeCrAl, are gaining much consideration and attraction for ATF cladding material because of their ability to make protective oxide layer limiting further corrosion. For instance, the aluminum in FeCrAl undergoes a reaction with oxygen to form an oxide layer (Al2O3) capable of reducing oxidation kinetics in high temperature steam environment and thus making FeCrAl an excellent ATF candidate (George et al., 2015). The high neutron absorbing cross section nature of stainless steel alloys still exists, but the high strength and slower oxidation of iron-based alloys allows for the reduction of cladding thickness. Consequently, assuming the clad outer diameter remains constant the fuel pellet diameter can be increased (George et al., 2015). This FeCrAl alloy contains 69% Fe, 21.6% Cr, 4.9% Al, (weight percent) plus other minor constituents.
Table 3 ATFs melting point (Wang et al., 2017; Merrill et al., 2017). Melting point
Value (K)
FeCrAl FeCrAl Oxide SiC Silica (SiO2)
1600 1800 2900 1873
ð1Þ
where the kp (µm2/h) is the parabolic rate constant. This rate constant can be derived by the equation as Eq. (2).
kp¼ k0 expðEa =RT Þ
ð2Þ
where k0 is a pre-exponential constant; R is the universal gas constant (J/mol-K); Ea is the activation energy (J/mol). For this research, the constants for FeCrAl and air interaction are k0 = 6.6998 (kg2FeCrAl/m4s) and Ea = 294 (kJ/mol-K). These constants for the FeCrAl and steam interactions are k0 = 0.5213 (kg2-FeCrAl/m4s) and Ea = 260 (kJ/mol-K). The values could be acquired by digitizing of APMT curves from Pint’s data (Pint et al., 2013). After substituting these values into Eq. (2) yields the FeCrAl oxidation kinetics for air and steam (Eqs. (3) and (4), respectively). Where T is temperature in Kelvin and k(T) is oxidation rate (kg2 /m4s).
kðT Þ ¼ 6:6998 expð35360=T Þ
ð3Þ
kðT Þ ¼ 0:5213 expð31271=T Þ
ð4Þ
4.2.2. The SiC oxidation kinetics Oxidation of SiC in air and steam is acquired by parabolic reaction kinetics. They can be modeled in straightforward manner. Oxidation of SiC in hot air follows the chemical Eq. (5) (Merrill et al., 2017).
SiC ðsÞ þ 1:5O2 ðg Þ ! SiO2 ðsÞ þ COðg Þ
ð5Þ
Based on experimental data, the activation energy is determined to be 190 kJ/mol and the pre-exponential constant to be k0 (for air) of 6.478E-5 (kg2-SiC/m4s). After substituting these parameters into Eq. (5) and simplifying, SiC oxidation kinetics with hot air is modeled by Eq. (6).
kðT Þ ¼ 6:478E 5 exp ð22852=T Þ
4.2. Oxidation kinetics for ATF materials
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ð6Þ
Reaction between SiC and hot steam is more complicated than SiC–O2 reaction. As described above, this reaction consists of two chemical processes (Eqs. (7) and (8)). First, SiC reacts with steam and SiO2 is formed. It undergoes a chemical reaction with steam and Si(OH4) is produced.
SiC þ H2 O ! SiO2 þ H2
ð7Þ
SiO2 þ 2H2 O ! SiðOHÞ4
ð8Þ
The values for this relationship proposed by Opila (2003), which are k0 (for steam) of 6.915E-9 (kg2-FeCrAl/m4s) and the Ea (for steam) of 35 (kJ/mol-K) (Merrill et al., 2017). Substituting these values into Eq. (2) gives SiC oxidation kinetics with hot steam for directly modification in MELCOR as Eq. (9). Where T is temperature in Kelvin and k(T) is oxidation rate constant (kg2/m4s).
kðTÞ ¼ 6:915E 9expð4209=TÞ
ð9Þ
4.3. Oxidation heating of ATF materials The oxidation of zirconium (Zr) with water vapor and air follow the chemical Eqs. (10) and (11).
Zr ðsÞ þ 2H2 Oðg Þ ! ZrO2 ðsÞ þ 2H2 ðg Þ þ Q ox
ð10Þ
Zr ðsÞ þ O2 ðg Þ ! ZrO2 ðsÞ þ Q ox
ð11Þ
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The heat of oxidation (Qox) for zircaloy-steam reaction (Eq. (10)) at 298.15 K is set to nominal value 5.797E6 (J/kg-Zr) and it is estimated to be 1.2065E7 (J/kg-Zr) for zircaloy-O2 reaction (Gauntt et al., 2005b). This type of reaction is exothermic which means energy releases from oxidation take place. The heat of FeCrAl oxidation with steam and oxygen are not exactly reported in the literature at this simulation time, thus based on approaches that were used by SNL-NM MELCOR developers to estimate FeCrAl oxidation heating, the heat of reaction was assumed equivalent heat of stainless steel oxidation, thus the released energy is estimated to be 7.808E6 (J/kg-FeCrAl) for FeCrAl-O2 reaction and 1.278E6 (J/kgFeCrAl) for FeCrAl-H2O reaction. This determination is to moleweight the heat of reaction of the constituent elements (Fe, Cr, and Al) to calculate a heat of reaction for this alloy. The heat of SiC oxidation at high pressure and temperature condition (accident condition) estimated based on the heat of formation of reactants and products of oxidation reaction at 298.15 K (Merrill et al., 2017). Based on standard molar enthalpy or heat of formation (DH0f) of reactants and products of SiC-O2 reaction at 298.15 K given by Weast and Astle (1980), the released energy for this reaction is 2.4E7(J/kg-SiC). The generated oxidation heat of the SiC-H2O reaction, based on heat of formation data given by (Merrill et al., 2013) is 4.15E6(J/kg-SiC), and for the next oxidation reaction (Eq. (8)), the released heat is 1.02E6(J/kg-SiO2) based on heat of formation data presented by Allendorf et al. (1995). These results represent all SiC oxidation energies are exothermic at 298.15 K. However, oxidation heat in MELCOR is hardcoded but because of the greater magnitude of zircaloy oxidation heat in MELCOR than FeCrAl oxidation heat and SiC oxidation heat as mentioned previously, this simulation could meet plant safety margins.
Fig. 16. PRZ level.
water level is higher than zircalloy-4 and by utilizing of FeCrAl, PRZ water level is going down faster than zircalloy-4. Using of SiC and FeCrAl alloy as cladding material, the SG water inventory in secondary side evaporates in same manner, approximately. Fig. 17 shows the SG water level in SBO accident for cladding materials. Better heat transfer of these materials in comparison of zircalloy-4 leads secondary side pressure increases faster as shown in Fig. 18. As shown in Fig. 19, the RPV water inventory dries a little faster due to have higher thermo-physical
5. Evaluation of the ATF materials in SBO accident In this section the performances of the ATF cladding materials are compared with zircaloy-4 in SBO accident. As a consequence of reactor scram in the SBO accident primary pressure drops significantly. Using of SiC and FeCrAl, fall in PRZ pressure is a little less than zircaloy-4 due to have better thermo-physical properties, such as higher thermal conductivity. Fig. 15 compares the ATF performances against zircaloy-4 in pressure variation of the SBO. As shown in Fig. 16, utilizing of ATF materials, water injection to hotleg is a little less than zircalloy-4, therefore PRZ water inventory is approximately same as zircalloy-4 until 5200 s. After safety valves opening due to reaching pressure 18.1 MPa in the PRZ, using of SiC, ejected water from PRZ is less than zircalloy-4, thus PRZ
Fig. 17. SG water level.
Fig. 15. Coolant pressure in Pressurizer.
Fig. 18. SG secondary pressure.
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Fig. 19. Reactor water volume.
Fig. 20. Coolant temperature at reactor outlet.
properties including thermal conductivity for ATF materials. Fig. 20 presents the coolant temperature in reactor outlet during the SBO accident. For SiC and FeCrAl alloy, temperature at reactor outlet increases because of excellent heat transfer characteristics of these materials especially in specific heats. Generally, based on this calculation a good agreement between zircaloy-4 and FeCrAl and SiC performances during accident progression as shown in Figs. 15–20 is demonstrated.
Fig. 21. Maximum temperature of fuel rod claddings.
Fig. 22. Cumulative in-vessel energy.
As seen in Fig. 23, once the pressure vessel lower head fails, debris mass ejects to the cavity in a high-pressure condition. In the case of zircaloy-4, lower head fails at 11,600 s, while lower head of FeCrAl fails at 13,200 s. However, the decay heat plays the main role in the failure of core materials, but the oxidation heat is the main cause of this delay in the RPV failure. The total debris mass is 148.1 tons for zircaloy-4 and for FeCrAl, it is 146.6 tons, almost 1.5 tons less than zircaloy-4.
5.1. Evaluation of FeCrAl performance This section contains the performance of the FeCrAl against zircaloy-4 in the SBO accident. Fig. 21 presents the maximum cladding temperature at the center region of the core. Till 5200 s of the accident, FeCrAl cladding temperature is same as zircaloy-4 due to the core cooling. Once coolant flow stops, so RPV water level decreases and the core begins to be uncovered. Thereafter fuel rods start to be heated up. The much lower oxidation kinetics of FeCrAl compared to zircaloy-4, retards the progress of fuel degradation about 2770 s compared with zircaloy-4. As shown in Fig. 22, the cumulative oxidation heat in the core for zircaloy-4 is 53.44GJ, however for FeCrAl alloy, it is 6.35 GJ. It is almost 88% less than zircaloy-4 oxidation. The much lower oxidation heating of FeCrAL leads to fuel failure with 2770 s delay in comparison of zircaloy-4. Also, a delay (1600 s) in lower head failure is observed. This value is the upper limit of oxidation heat, because of oxidation heating is hardcoded in MELCOR as described previously.
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Fig. 23. Debris mass through vessel.
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mances are evaluated. Based on the calculation, compared to conventional UO2-Zircaloy fuel rod, implementation of the FeCrAl alloy as cladding material results in more than 47.09GJ (88%) decrement in cumulative heat generated by oxidation reaction. It leads to fuel rods failure with 2770 s delay and retards lower head failure about 1600 s. Other benefit gained by using this advanced cladding material is decrement of debris mass about 1.5 tons. As the most important benefit of this alteration is more than 285 kg (68%) decrement in hydrogen generation. In the case of replacement of SiC with zircaloy-4, the following advantages achieved. The total in-vessel cumulative heat of oxidation is reduced about 43.18GJ (81%). The total debris mass ejected through lower head is 2.1 tons less than zircaloy-4 and the total in-vessel hydrogen production decreases more than 197 kg (47%). These results prove using FeCrAl and SiC, improve performance and safety characteristics in Bushehr power plant during severe accidents. Fig. 24. Total hydrogen mass.
Acknowledgments An important phenomenon happens during severe accidents is the hydrogen generation from the exothermal reaction between oxidation of core components (Gharari et al., 2018). The hydrogen accumulation in containment maybe leads to explosion and leakage of radioactive material to the atmosphere. Fig. 24 implies total in-vessel hydrogen production of zircaloy-4 is 419 kg, however, in the case of FeCrAl, it is 134 kg which means 68% decrement.
The authors are gratefully indebted to Shahid Beheshti University for partial support of this work.
5.2. Evaluation of SiC performance
References
In this section the performance of the SiC against the zircaloy-4 is evaluated. As seen in Fig. 21, after core dry out, the heat transfer between fuel rods and steam is considerably low and temperature of clad increases significantly until clad melts. The SiC clad failure occurs at 12,300 s. It is about 3100 s over zircaloy-4 due to having the higher melting point and much lower magnitude of oxidation rate, mainly. As shown in Fig. 22, total cumulative oxidation heat of SiC is 10.26 GJ, which means 81% decrement in energy in case of using SiC that leads to delay in fuel degradation and failure of RPV lower plenum in comparison of zircaloy-4. Once failure of the support plate occurs, the core relocation starts. As seen in Fig. 23, after the relocation of molten core materials to lower plenum, the RPV fails due to creep rupture of the lower head at 13,700 s in case of SiC clad. The lower plenum of SiC could withstand more than 2100 s in comparison of zircaloy4. Total debris mass ejected through vessel of SiC is 146 tons and 2.1 tons less than debris mass of zircaloy-4. After the core uncovered, the steam-clad reaction, accompanied by hydrogen pick up. As Fig. 24 illustrates, however, the hydrogen generation begins almost at the same time for both zircaloy-4 and SiC, due to having much less oxidation rate of SiC, total in-vessel generated hydrogen of SiC is 222 kg. The result indicates, using SiC concludes 47% decrement in hydrogen production.
Allendorf, M.D., Melius, C.F., Ho, P., Zarchariah, M.R., 1995. Theoretical study of the thermochemistry of molecules in the Si-O-H system. J. Phys. Chem. 99 (41), 15285–15293. Bragg-Sitton, S.M., 2012. Light Water Reactor Sustainability Program Advanced LWR Nuclear Fuel Cladding System Development Technical Program Plan. Technical Report INL/MIS-12-25696. Idaho National Laboratory. Carmack, J., Goldner, F., 2014. Forward for special JNM issue on accident tolerant fuels for LWRs. J. Nucl. Mater. 1, 373. Cheng, T., Keiser, J.R., Brady, M.P., Terrani, K.A., Pint, B.A.J., 2012. Oxidation of fuel cladding candidate materials in steam environments at high temperature and pressure. J. Nucl. Mater. 427, 396–400. Farmer, M.T., Leibowitz, L., Terrani, K.A., Robb, K.R., 2014. Scoping assessments of ATF impact on late-stage accident progression including molten core–concrete interaction. J. Nucl. Mater. 448, 534–540. Final Safety Assessment Report (FSAR) for BNPP, 2007. Russia Federal Agency on Nuclear Energy (RFANE). Accident Analysis, Book 4, Moscow. Gharari, R., Kazeminejad, H., Mataji-Kojouri, N., Hedayat, A., 2018. A review on hydrogen generation, explosion, and mitigation during severe accidents in light water nuclear reactors. Int. J. Hydrogen Energy 43 (4), 25. Gauntt, R.O., Cash, J.E., Cole, R.K., Erickson, C.M., Humphries, L.L., Rodriguez, S.B., Young, M.F., 2005a. MELCOR Computer Code Manuals, Vol. 1: Primer and Users’ Guide. Gauntt, R.O., Cash, J.E., Cole, R.K., Erickson, C.M., Humphries, L.L., Rodriguez, S.B., Young, M.F., 2005b. MELCOR Computer Code Manuals, Vol. 2: Reference manuals. Gauntt, R.O., Ross, K., Wagner, K., 2002. Simulation of TMI-2 Phase2 With an Enhanced 2-Dimensional In-Vessel Natural Circulation Model. In: 10th International Conference on Nuclear Engineering, Arlington, Virginia, USA, Vol. 3. American Society of Mechanical Engineers. pp. 487–494. George, N.M. et al., 2015. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors. Annals Nucl. Energy 75, 703–712. Katoh, Y. et al., 2012. Radiation effects in SiC for nuclear structural applications. Curr. Opinion Solid State Mater. Sci. 16 (3), 143–152. Lindley, B.A., Kotlyar, D., Parks, G.T., Lillington, J.N., Petrovic, Bojan, 2016. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes. EPJ Nuclear Sci. Technol. 2, 14. Merrill, B.J., Bragg-Sitton, S.M., Humrickhouse, P.W., 2017. Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials. Nucl. Eng. Design 315, 170–178. Merrill, B.J., Bragg-Sitton, S.M., 2013. Advanced Cladding Modeling Work to Assess Cladding Performance Under Accident Conditions. INL/EXT-13-30206. Merrill, B.J., 2012. MELCOR SiC Modifications for Severe Accident Analysis Applications, Idaho National Laboratory Report, INL/LTD-12-27961. Opila, E.J., 2003. Oxidation and volatilization of silica formers in water vapor. J. Am. Ceram. Soc. 86 (8), 1238–1248. Pint, B.A., Terrani, K.A., Brady, M.P., Cheng, T., Keiser, J.R., 2013. High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments. J. Nucl. Mater. 440, 420–427.
6. Conclusion Today for mitigation of accident consequences, it is needed to evaluate the accident tolerant clads over the traditional zirconium alloy cladding employed in the nuclear industry. In this work, it is focused mainly on FeCrAl and SiC as optimized cladding materials in comparison with zircaloy-4 during station blackout accident in the BNPP. An important advantage of these materials is slower oxidation kinetics conducts to produce much less hydrogen than zircaloy-4. In this research, the thermophysical properties and oxidation kinetics for these ATFs replaced with zircaloy-4 and perfor-
Appendix A. Supplementary material Supplementary data to this article can be found online at https://doi.org/10.1016/j.anucene.2019.02.008.
H. Ebrahimgol et al. / Annals of Nuclear Energy 129 (2019) 214–223 Richet, P., Bottinga, Y., Denielou, L., Petitet, J.P., Tequi, C., 1982. Thermodynamic properties of quartz, cristobalite and amorphous SiO2: drop calorimetry measurements between 1000 and 1800 K and a review from 0 to 2000 K. Geochimica et Cosmochimica Acta 46 (12), 2639–2658. Robb, K.R., McMurray, J.W., Terrani, K.A., 2016. Severe accident analysis of BWR core fueled with UO2/FeCrAl with update materials and melt properties from experiments. Fuel Cycle Res. Develop. Adv. Fuels Campaign. Terrani, K.A., Pint, B.A., Parish, C.M., Silva, C.M., Snead, L.L., Katoh, Y., 2014. Silicon carbide oxidation in steam up to 2 MPa. J. Am. Ceram. Soc. 97 (8), 2331–2352.
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Terrani, K.A., Zinkle, S.J., Snead, L.L., 2013. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding. J. Nucl. Mater. 448, 420–435. Wang, J., Mccabe, M., Wu, L., Dong, X., Wang, X., Haskin, T.C., Corradini, M.L., 2017. Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out. Nucl. Eng. Design 313, 458–469. Weast, R.C., Astle, M.J., 1980. Handbook of Chemistry and Phyics. CRC Press, p. B234. Yueh, K., Terrani, K.A., 2014. Silicon carbide composite for light water reactor fuel assembly applications. J. Nucl. Mater. 448 (1–3), 380–388.
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Evaluation of accident tolerant cladding materials in a severe accident of the BNPP H. Ebrahimgol, M. Aghaie, A. Zolfaghari, Gh. Alahyarizadeh Engineering Department, Shahid Beheshti University, G.C, P.O. Box: 1983963113, Tehran, Iran
a r t i c l e
i n f o
Article history: Received 11 April 2018 Received in revised form 2 November 2018 Accepted 3 February 2019
Keywords: Accident tolerant fuels BNPP SBO accident SiC FeCrAl
a b s t r a c t Accident Tolerant Fuels (ATFs) are fuels or fuel clads with improved features in comparison with standard UO2-Zircaloy in commercial LWRs. The ATFs could tolerate in vessel loss of coolant accidents. IronChromium-Aluminum (FeCrAl) alloy and Silicon carbide (SiC) are being introduced as enhanced cladding candidates to mitigate accident consequences. These cladding materials have slower oxidation kinetics and high resistance to core degradation which reduce hydrogen accumulation in severe accidents. This paper deals with the evaluation of ATFs in hydrogen mitigation during Station Blackout Accident (SBO). In this analysis, a numerical model based on MELCOR for Bushehr Nuclear Power Plant (BNPP) in SBO is prepared and replacement of ATF cladding materials (FeCrAl and SiC) is studied. The thermal hydraulic response of the BNPP and core heat-up with clad oxidations are evaluated and the hydrogen generation due to the pernicious interaction of hot steam and cladding materials for FeCrAl, SiC and zircaloy-4 are calculated. The results show the total hydrogen accumulation in the SBO of the BNPP are reduced 68% and 47% for FeCrAl, SiC, respectively. The accumulated energy in the core are reduced 88% and 81% for FeCrAl and SiC, respectively and the debris generation and lower head failure are improved, significantly. The results illustrate the lower hydrogen generation and oxidation kinetics in case of using the ATF materials and the performances are demonstrated. Ó 2019 Published by Elsevier Ltd.
1. Introduction The Fukushima Daiichi accident renewed attention to the pernicious interaction between zirconium alloys and hot steam during severe accidents. Finding applicable solutions to reduce accident consequences such as hydrogen production and core degradation due to loss of coolant in core have been evaluated by researchers at nuclear laboratories and universities considerably (Terrani et al., 2013; Gauntt et al., 2005a). Past SiC investigation has focused on the high temperature gas reactor tri-structural-isotropic (TRISO) fuel and fusion reactor structural components (Yueh and Terrani, 2014), but nowadays interest of using fully ceramic SiC for LWR cladding is significantly increasing (Bragg, 2012). Recently many projects at nuclear laboratories such as Idaho National Laboratory with US Nuclear Regulatory Commission (NRC) have been focused on ATFs development in LWRs (Carmack and Goldner, 2014; Merrill et al., 2017). One of the main proposed solutions to mitigate the consequences of the severe accident is ATF materials for reducing the rates of heat and hydrogen generation due to cladding oxidation with hot steam
E-mail address: [email protected] (M. Aghaie) https://doi.org/10.1016/j.anucene.2019.02.008 0306-4549/Ó 2019 Published by Elsevier Ltd.
(Terrani, 2014). During 2012, several candidate cladding materials (such as 310 SS, FeCrAl, and SiC) were demonstrated to have significantly lower oxidation kinetics and hydrogen generation (Pint et al., 2013). The crucial approaches for development of ATFs are improvement of fuel properties, enhancement of cladding properties to maintain core coolability and retain fission products and improvement of reaction kinetics with steam to minimize enthalpy and hydrogen generation (Farmer et al., 2014). The SiC concept is expected to gain more benefits than zirconium alloy, smaller neutron absorption cross section, lower chemical interaction, ability to endure higher fuel burn-ups and higher temperatures, exceptional inherent radiation resistance, lack of progressive irradiation growth, and low induced activation/low decay heat (Katoh et al., 2012). FeCrAl alloy, have the ability to enhance oxidation resistance by creating a protective oxide layer; for instance, the aluminum found in FeCrAl undergoes a reaction with oxygen to produce an oxide layer (Al2O3) capable of oxidation kinetics reduction in hot steam environments (Terrani et al., 2013). Implementation of new cladding material in MELCOR was tried at Idaho National Laboratory (INL) in 2012. The INL applied these materials to simulate the Three Mile Island Unit 2 (TMI-2) lossof-coolant accident (Merrill, 2012). The TMI-2 accident was SBLOCA in a two-loop Babcock and Wilcox (B&W) pressurized
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water reactor (PWR) (Gauntt et al., 2002). In 2016 an effort was initiated for replacing zircaloy with SiC and ferritic stainless-steel alloy (FeCrAl alloy) (Merrill et al., 2017). Recently some experimental studies related to the performance of proposed ATF concepts in steam environments at high temperature and pressure have been performed (Pint et al., 2013; Cheng et al., 2012). As a result of these studies, important kinetic parameters have become available for ATF materials. Based on recent experimental researches, ATFs performance modeled for a range of severe accidents was analyzed at Fuel Cycle Research & Development Advanced Fuels Campaign (Robb et al., 2016). To develop investigation of the accident tolerant reactor concept, the Georgia Institute of Technology was asked to conduct an integrated research project separately (Lindley et al., 2016). Other research activities in the university of Illinois and the university of Manchester are being conducted to discuss two pathways in order to enhance the modification of monolithic zircaloy cladding as the foundation of ATF cladding (Yueh and Terrani, 2014). After decades of using zirconium cladding components in LWRs reactors, the transition to using ATF materials represents a revolutionary prototype shift. Due to the impact associated with any such transition, associated challenges will need to be evaluated via predictive fuel performance analysis. In this paper, two candidates of ATF cladding materials (FeCrAl and SiC) as appropriate substitutes of zircaloy-4 during station blackout in BNPP have been studied. Using the MELCOR code, the core materials such as fuel clad and required number of thermophysical characteristics are replaced and a model for BNPP during the SBO is prepared. For accurate evaluation of the materials performances, the updated characteristics such as oxidation kinetics are prepared. The core heat up and thermal hydraulic assessment of the core is evaluated against Final Safety Analysis Report (FSAR, 2007) of Bushehr nuclear power plant. The effect of the ATF materials in thermal hydraulic response of the plant is evaluated and in core meltdown, the hydrogen generation, energy accumulation in core, debris and lower head creep rupture are calculated. The purpose of this paper is to demonstrate the impact of the advanced cladding materials on the fuel performance on the BNPP (WWER1000 reactor) under the beyond design basis accidents (BDBAs). 2. The BNPP description and model nodalization The plant of interest to analyze is a four-loop pressurized water reactor WWER1000-V446 (Located in Bushehr, Iran) with 3000 MW thermal power generation. The main distinguishing features of the WWER compared to other PWRs are; horizontal steam generators, hexagonal fuel assemblies, no bottom penetrations in the pressure vessel and high-capacity pressurizers providing a large reactor coolant inventory.
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The BNPP reactor core consists of 163 fuel assemblies in hexagonal design and different fuel enrichment. A part of fuel assemblies contains burnable absorber rods (BAR) bundles or absorbing rods of control and protection system (CPS), which are moved by the CPS drives. The fuel bundle has a shape of hexahedron in section with maximum dimension of 235.1 mm between the faces. In the bundle cross section there were placed 311 fuel rods, 18 guiding channels for positioning CPS or BAR, central channel and channel for in-core instrumentation detector (ICID). Each loop has cold leg and hot leg pipes connected to reactor vessel nozzles, a main coolant pump to circulate the reactor coolant system with nominal capacity of 21,200 m3/h and 0.624 MPa head and steam generators (SG) type PGV-1000 M(B) is intended to produce dry saturated steam of 6.27 MPa pressure, in the form of one vessel with horizontal arrangement of tubes constituting the heat exchanging surface. One pressurizer (PRZ) is connected to the undetachable part of the single hot leg of main circulation loop by connecting pipeline to control of loops pressure. The relief system connected to PRZ is intended to remove steam, steamwater and steam-gas mixture from the pressurizer during operation of the emergency gas removal system and/or excessive pressure protection system. Secondary loop is designed as SG source and sink with its safety relief valves. They have been disconnected at the beginning of the SBO accident as described in Table 1. The reactor pressure vessel (RPV) is modeled in 5 control volumes: down-commer, bypass, lower head, core, upper plenum. The reactor core is separated in 5 rings and 19 axial levels. The BNPP nodalization model is shown in Fig. 1. 3. Simulation of the SBO accident in the BNPP In this section, the SBO accident is simulated and results are compared with FSAR (2007). The BNPP’s final safety analysis report (FSAR) includes the SBO accident until core heat up. During the station blackout, due to loss of both the normal A.C. power supply and the emergency diesel generators, core heat sink will be lost. The sequences of events in this accident are reported in Table 1 (FSAR, 2007). In this work, simulation of the SBO accident revealed the following occurrences. At the beginning of the accident, reactor control and protection system (RCP) is tripped and scram occurs. The reactor power decreases due to onset of control rods motion. Fig. 2 depicts the reactor core power during scram. It is clear that the reactor power decreases to decay heat level significantly at the beginning of the insertion and follows the decay heat generation in next. The reactor coolant pump sets switch off and all blow-down makeup systems of the primary system fail. The pressurizer power supply system is disconnected and BRU-K valves designed to dump steam from the steam generator into the turbine condenser close.
Table 1 Sequence of system and device actuation (FSAR, 2007). Time, s (FSAR)
Time, s (MELCOR)
Event
Interlocks, set point for actuation or other cause
0.0
0.0
Loss of all A.C. off-site and on-site power supply sources (power unit blackout)
0.6 1.4 1.7 7.0 2800. 7000. 10000.
0.6 1.4 – 7.0 3200. 7100. 10000.
Trip of all RCP sets Trip of the main and auxiliary feed water systems of the secondary side Trip of makeup-blowdown system of the primary system BRU-K disconnection Disconnection of PRZ system power supply Closing the turbine generator stop valves Scram signal generation The onset of control rod motion BRU-A opening SG drainage Onset of the core heat-up End of calculation
Turbine emergency protection action NPP blackout Emergency protection action Reaching SG pressure of 7,15 MPa
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Fig. 1. Nodalization scheme of the BNPP.
Fig. 2. Reactor power.
Fig. 3. SG secondary pressure.
Immediately, the main and auxiliary feedwater systems of the secondary sides are closed in order to protect the turbine. As shown in Fig. 3, the heat transfer between the primary and secondary sides causes vapor generation inside SG shell and the secondary side pressure increases. The secondary side pressure is maintained between 6.27 MPa and 7.15 MPa by steam dump valves. As Fig. 4 presents, the SGs water inventory dry out completely in the absence of secondary side feedwater. Discharge through the dump devices decrease SG water inventory until 3150 s and finally, SG will lose its heat removal ability. SGs drying results in a slight primary side pressure increasing. Once SGs dry out, the primary pressure increases drastically up to it reaches the PRZ safety relief valves setpoint. Coolant discharge from safety relief valves results loss of the primary coolant, and the core is further heated-up. Without acceptance of management measures, damage occurs to the core and vessel at high pressure. The pressure variation in primary side of the BNPP during SBO accident is shown in Fig. 5. In first seconds, due to the trip of all RCP sets and reactor scram, the primary pressure reduces sharply.
Fig. 4. SG water level.
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Fig. 7. Coolant flowrate at the reactor inlet. Fig. 5. Coolant pressure in Pressurizer.
In following, loss of coolant inventory in shell side of SGs results pressure increasing, slightly. By SGs drainage and coolant heat up, primary pressure and PRZ water level increase. While the primary pressure reaches to 18.1 MPa, PRZ safety relief valves act to remove excessive energy by coolant discharge to relief tank and PRZ water level decreases. This valve is closed once the primary pressure drops to 17.2 MPa. The water level in the PRZ is reported in Fig. 6. Coolant evaporation in primary side and coolant discharge via relief valves lead to water level decrement as depicted in Fig. 6. Figs. 5 and 6 present the PRZ relief valve actions during the SBO accident. The relief valve setpoints (17.2&18.1 MPa) make the pressure oscillations as depicted in Fig. 5. Fig. 7 presents the coolant flow rate in primary side during the SBO. Coolant flowrate with loss of all A.C. powers decreases significantly at the first seconds and continues slightly until 5000 s as a result of buoyancy driven natural circulation. Finally, the coolant circulation stops completely during 200 s. Fig. 8 shows water volume in reactor pressure vessel. Once coolant circulation fully stops, coolant heat up occurs and in-vessel water evaporation begins. The pressure vessel coolant dries until 7000 s. Once the coolant flow in primary side stops (5200 s), the temperature at the reactor outlet increases and the RPV water inventory decreases. The RPV drainage accelerates and temperature rises as presented in Fig. 9. In SBO accident, the fuel cladding temperature increases slightly due vapor generation in the RPV. The trend of vapor tem-
Fig. 8. Reactor water volume.
Fig. 9. Coolant temperature at reactor outlet.
Fig. 6. PRZ level.
perature at reactor outlet is similar to cladding temperature and it is increasing. Temperature increment of fuel cladding continues with reactor vessel drainage until the fuel rods fail. As Fig. 10 depicts, once core heat removal fails the cladding temperature increases slightly. As soon as RPV dries out, the core is heated up and the cladding temperature increases significantly until the fuel rod fails at 9200 s. The accumulated heat in core is due to decay heat and cladding oxidation.
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Fig. 12. Density. Fig. 10. Maximum temperature of fuel rod claddings.
4. Requirements for modification of cladding materials This investigation is performed by MELCOR code. This code is developed to analyze light water reactors under severe accidents which allow us to alter cladding material and consider the behavior of this alteration under off-normal conditions. The default cladding material in the MELCOR code is zircaloy. In this work, the material properties specified for zircaloy are modified to survey different cladding materials. The following explicate how the material properties in MELCOR are modified to depict the advanced materials and the modeling limitations encountered. To define new cladding materials to the MELCOR, the candidate cladding materials including SiC and FeCrAl are overwritten. The new clads are assumed to have the same geometry and dimensions as zircaloy clad in the BNPP. The modifications include core oxidation kinetics and material property routines consist of thermal conductivity (Fig. 11), density (Fig. 12), specific heat (Fig. 13), enthalpy (Fig. 14), latent heat of fusion (Table 2) and emissivity. Thermophysical properties of ATF materials will be discussed in next section.
Fig. 13. Specific heat.
4.1. Thermophysical properties of ATF materials 4.1.1. The FeCrAl and oxidized FeCrAl thermophysical properties The FeCrAl thermal conductivity increases along with temperature increment and for temperatures higher than 1700 K remains constant (Wang et al., 2017). The thermal conductivity of oxidized FeCrAl is assumed to be 20 W/mK as displayed in Fig. 11. The oxidized FeCrAl alloy consists of Fe3O4, Al2O3 and Cr2O4 are formed
Fig. 14. Enthalpy.
Table 2 ATFs latent heat of fusion (Wang et al., 2017; Richet et al., 1982).
Fig. 11. Thermal conductivity.
Latent heat of fusion
Value (kJ/kg)
FeCrAl FeCrAl Oxide SiC Silica (SiO2)
260 580 360 930
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after core uncovery. The FeCrAl density descends from 8000 kg/m3 to 7000 kg/m3 between 300 K and 1500 K and will be remained 7000 kg/m3 all over the ranges. The density of oxidized FeCrAl is 5000 kg/m3 as shown in Fig. 12. The Specific heat of the FeCrAl alloy increases almost linearly with temperature increment up to 1600 K then remains constant. The specific heat of oxidized FeCrAl is estimated to be 500 J/kg. K as illustrated in Fig. 13 (Wang et al., 2017). The enthalpy of FeCrAl and oxidized FeCrAl increases linearly except in temperatures between 1500 K and 2000 K as displayed in Fig. 14. The latent heat of fusion of FeCrAl is estimated to be 260 kJ/K and the latent heat of fusion of oxidized FeCrAl is set to 580 kJ/kg. The melting point of FeCrAl alloy is 1600 K and the melting point of oxidized FeCrAl is around 1800 K, as shown in Table 3 (Wang et al., 2017). The emissivity of oxidized FeCrAl is set to 0.70 (Robb et al., 2016). 4.1.2. The SiC and SiC oxide thermophysical properties The SiC thermal conductivity decreases considerably below temperature 1300 K then reduces mildly up to 5000 K. For SiO2, this value increases before reaching 1400 K, then remains constant, as presented in Fig. 11. The density of SiC and SiO2 are assumed to be 2900 kg/m3 and 2000 kg/m3, respectively as shown in Fig. 12. The specific heat of SiC increases below temperature 2500 K then decreases slightly. For temperature below 1200 K, the trend of specific heat of SiO2 is similar to SiC. For temperatures above 1500 K due to lack of experimental data, the linear extrapolated values to 2000 K are considered as presented in Fig. 13 (Merrill,2012). The melting temperature of SiC and SiO2 are 2900 K and 1873 K, respectively. The enthalpy progression of SiC and SiO2 along with the increase of temperature. The significant increase in enthalpy of SiC is around 2900 K where the SiC phase change takes place. It is around 1873 K for SiO2 shown in Fig. 14. The latent heat of fusion of SiC and SiO2 are assumed to be 360 (kJ/kg) and 930 (kJ/kg), respectively (Table 2). The SiC cladding inner surface emissivity is set to 0.75 (Merrill, 2012).
The FeCrAl oxidation is calculated using standard parabolic kinetics, with appropriate rate constant expressions similar to zircaloy. The square of the thickness of the oxidized FeCrAl scale formed x is proportional to exposure time t as Eq. (1) (Merrill et al., 2017).
x2 ¼ k p t
4.2.1. The FeCrAl oxidation kinetics Ferritic stainless-steel alloys, such as FeCrAl, are gaining much consideration and attraction for ATF cladding material because of their ability to make protective oxide layer limiting further corrosion. For instance, the aluminum in FeCrAl undergoes a reaction with oxygen to form an oxide layer (Al2O3) capable of reducing oxidation kinetics in high temperature steam environment and thus making FeCrAl an excellent ATF candidate (George et al., 2015). The high neutron absorbing cross section nature of stainless steel alloys still exists, but the high strength and slower oxidation of iron-based alloys allows for the reduction of cladding thickness. Consequently, assuming the clad outer diameter remains constant the fuel pellet diameter can be increased (George et al., 2015). This FeCrAl alloy contains 69% Fe, 21.6% Cr, 4.9% Al, (weight percent) plus other minor constituents.
Table 3 ATFs melting point (Wang et al., 2017; Merrill et al., 2017). Melting point
Value (K)
FeCrAl FeCrAl Oxide SiC Silica (SiO2)
1600 1800 2900 1873
ð1Þ
where the kp (µm2/h) is the parabolic rate constant. This rate constant can be derived by the equation as Eq. (2).
kp¼ k0 expðEa =RT Þ
ð2Þ
where k0 is a pre-exponential constant; R is the universal gas constant (J/mol-K); Ea is the activation energy (J/mol). For this research, the constants for FeCrAl and air interaction are k0 = 6.6998 (kg2FeCrAl/m4s) and Ea = 294 (kJ/mol-K). These constants for the FeCrAl and steam interactions are k0 = 0.5213 (kg2-FeCrAl/m4s) and Ea = 260 (kJ/mol-K). The values could be acquired by digitizing of APMT curves from Pint’s data (Pint et al., 2013). After substituting these values into Eq. (2) yields the FeCrAl oxidation kinetics for air and steam (Eqs. (3) and (4), respectively). Where T is temperature in Kelvin and k(T) is oxidation rate (kg2 /m4s).
kðT Þ ¼ 6:6998 expð35360=T Þ
ð3Þ
kðT Þ ¼ 0:5213 expð31271=T Þ
ð4Þ
4.2.2. The SiC oxidation kinetics Oxidation of SiC in air and steam is acquired by parabolic reaction kinetics. They can be modeled in straightforward manner. Oxidation of SiC in hot air follows the chemical Eq. (5) (Merrill et al., 2017).
SiC ðsÞ þ 1:5O2 ðg Þ ! SiO2 ðsÞ þ COðg Þ
ð5Þ
Based on experimental data, the activation energy is determined to be 190 kJ/mol and the pre-exponential constant to be k0 (for air) of 6.478E-5 (kg2-SiC/m4s). After substituting these parameters into Eq. (5) and simplifying, SiC oxidation kinetics with hot air is modeled by Eq. (6).
kðT Þ ¼ 6:478E 5 exp ð22852=T Þ
4.2. Oxidation kinetics for ATF materials
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ð6Þ
Reaction between SiC and hot steam is more complicated than SiC–O2 reaction. As described above, this reaction consists of two chemical processes (Eqs. (7) and (8)). First, SiC reacts with steam and SiO2 is formed. It undergoes a chemical reaction with steam and Si(OH4) is produced.
SiC þ H2 O ! SiO2 þ H2
ð7Þ
SiO2 þ 2H2 O ! SiðOHÞ4
ð8Þ
The values for this relationship proposed by Opila (2003), which are k0 (for steam) of 6.915E-9 (kg2-FeCrAl/m4s) and the Ea (for steam) of 35 (kJ/mol-K) (Merrill et al., 2017). Substituting these values into Eq. (2) gives SiC oxidation kinetics with hot steam for directly modification in MELCOR as Eq. (9). Where T is temperature in Kelvin and k(T) is oxidation rate constant (kg2/m4s).
kðTÞ ¼ 6:915E 9expð4209=TÞ
ð9Þ
4.3. Oxidation heating of ATF materials The oxidation of zirconium (Zr) with water vapor and air follow the chemical Eqs. (10) and (11).
Zr ðsÞ þ 2H2 Oðg Þ ! ZrO2 ðsÞ þ 2H2 ðg Þ þ Q ox
ð10Þ
Zr ðsÞ þ O2 ðg Þ ! ZrO2 ðsÞ þ Q ox
ð11Þ
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The heat of oxidation (Qox) for zircaloy-steam reaction (Eq. (10)) at 298.15 K is set to nominal value 5.797E6 (J/kg-Zr) and it is estimated to be 1.2065E7 (J/kg-Zr) for zircaloy-O2 reaction (Gauntt et al., 2005b). This type of reaction is exothermic which means energy releases from oxidation take place. The heat of FeCrAl oxidation with steam and oxygen are not exactly reported in the literature at this simulation time, thus based on approaches that were used by SNL-NM MELCOR developers to estimate FeCrAl oxidation heating, the heat of reaction was assumed equivalent heat of stainless steel oxidation, thus the released energy is estimated to be 7.808E6 (J/kg-FeCrAl) for FeCrAl-O2 reaction and 1.278E6 (J/kgFeCrAl) for FeCrAl-H2O reaction. This determination is to moleweight the heat of reaction of the constituent elements (Fe, Cr, and Al) to calculate a heat of reaction for this alloy. The heat of SiC oxidation at high pressure and temperature condition (accident condition) estimated based on the heat of formation of reactants and products of oxidation reaction at 298.15 K (Merrill et al., 2017). Based on standard molar enthalpy or heat of formation (DH0f) of reactants and products of SiC-O2 reaction at 298.15 K given by Weast and Astle (1980), the released energy for this reaction is 2.4E7(J/kg-SiC). The generated oxidation heat of the SiC-H2O reaction, based on heat of formation data given by (Merrill et al., 2013) is 4.15E6(J/kg-SiC), and for the next oxidation reaction (Eq. (8)), the released heat is 1.02E6(J/kg-SiO2) based on heat of formation data presented by Allendorf et al. (1995). These results represent all SiC oxidation energies are exothermic at 298.15 K. However, oxidation heat in MELCOR is hardcoded but because of the greater magnitude of zircaloy oxidation heat in MELCOR than FeCrAl oxidation heat and SiC oxidation heat as mentioned previously, this simulation could meet plant safety margins.
Fig. 16. PRZ level.
water level is higher than zircalloy-4 and by utilizing of FeCrAl, PRZ water level is going down faster than zircalloy-4. Using of SiC and FeCrAl alloy as cladding material, the SG water inventory in secondary side evaporates in same manner, approximately. Fig. 17 shows the SG water level in SBO accident for cladding materials. Better heat transfer of these materials in comparison of zircalloy-4 leads secondary side pressure increases faster as shown in Fig. 18. As shown in Fig. 19, the RPV water inventory dries a little faster due to have higher thermo-physical
5. Evaluation of the ATF materials in SBO accident In this section the performances of the ATF cladding materials are compared with zircaloy-4 in SBO accident. As a consequence of reactor scram in the SBO accident primary pressure drops significantly. Using of SiC and FeCrAl, fall in PRZ pressure is a little less than zircaloy-4 due to have better thermo-physical properties, such as higher thermal conductivity. Fig. 15 compares the ATF performances against zircaloy-4 in pressure variation of the SBO. As shown in Fig. 16, utilizing of ATF materials, water injection to hotleg is a little less than zircalloy-4, therefore PRZ water inventory is approximately same as zircalloy-4 until 5200 s. After safety valves opening due to reaching pressure 18.1 MPa in the PRZ, using of SiC, ejected water from PRZ is less than zircalloy-4, thus PRZ
Fig. 17. SG water level.
Fig. 15. Coolant pressure in Pressurizer.
Fig. 18. SG secondary pressure.
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Fig. 19. Reactor water volume.
Fig. 20. Coolant temperature at reactor outlet.
properties including thermal conductivity for ATF materials. Fig. 20 presents the coolant temperature in reactor outlet during the SBO accident. For SiC and FeCrAl alloy, temperature at reactor outlet increases because of excellent heat transfer characteristics of these materials especially in specific heats. Generally, based on this calculation a good agreement between zircaloy-4 and FeCrAl and SiC performances during accident progression as shown in Figs. 15–20 is demonstrated.
Fig. 21. Maximum temperature of fuel rod claddings.
Fig. 22. Cumulative in-vessel energy.
As seen in Fig. 23, once the pressure vessel lower head fails, debris mass ejects to the cavity in a high-pressure condition. In the case of zircaloy-4, lower head fails at 11,600 s, while lower head of FeCrAl fails at 13,200 s. However, the decay heat plays the main role in the failure of core materials, but the oxidation heat is the main cause of this delay in the RPV failure. The total debris mass is 148.1 tons for zircaloy-4 and for FeCrAl, it is 146.6 tons, almost 1.5 tons less than zircaloy-4.
5.1. Evaluation of FeCrAl performance This section contains the performance of the FeCrAl against zircaloy-4 in the SBO accident. Fig. 21 presents the maximum cladding temperature at the center region of the core. Till 5200 s of the accident, FeCrAl cladding temperature is same as zircaloy-4 due to the core cooling. Once coolant flow stops, so RPV water level decreases and the core begins to be uncovered. Thereafter fuel rods start to be heated up. The much lower oxidation kinetics of FeCrAl compared to zircaloy-4, retards the progress of fuel degradation about 2770 s compared with zircaloy-4. As shown in Fig. 22, the cumulative oxidation heat in the core for zircaloy-4 is 53.44GJ, however for FeCrAl alloy, it is 6.35 GJ. It is almost 88% less than zircaloy-4 oxidation. The much lower oxidation heating of FeCrAL leads to fuel failure with 2770 s delay in comparison of zircaloy-4. Also, a delay (1600 s) in lower head failure is observed. This value is the upper limit of oxidation heat, because of oxidation heating is hardcoded in MELCOR as described previously.
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Fig. 23. Debris mass through vessel.
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mances are evaluated. Based on the calculation, compared to conventional UO2-Zircaloy fuel rod, implementation of the FeCrAl alloy as cladding material results in more than 47.09GJ (88%) decrement in cumulative heat generated by oxidation reaction. It leads to fuel rods failure with 2770 s delay and retards lower head failure about 1600 s. Other benefit gained by using this advanced cladding material is decrement of debris mass about 1.5 tons. As the most important benefit of this alteration is more than 285 kg (68%) decrement in hydrogen generation. In the case of replacement of SiC with zircaloy-4, the following advantages achieved. The total in-vessel cumulative heat of oxidation is reduced about 43.18GJ (81%). The total debris mass ejected through lower head is 2.1 tons less than zircaloy-4 and the total in-vessel hydrogen production decreases more than 197 kg (47%). These results prove using FeCrAl and SiC, improve performance and safety characteristics in Bushehr power plant during severe accidents. Fig. 24. Total hydrogen mass.
Acknowledgments An important phenomenon happens during severe accidents is the hydrogen generation from the exothermal reaction between oxidation of core components (Gharari et al., 2018). The hydrogen accumulation in containment maybe leads to explosion and leakage of radioactive material to the atmosphere. Fig. 24 implies total in-vessel hydrogen production of zircaloy-4 is 419 kg, however, in the case of FeCrAl, it is 134 kg which means 68% decrement.
The authors are gratefully indebted to Shahid Beheshti University for partial support of this work.
5.2. Evaluation of SiC performance
References
In this section the performance of the SiC against the zircaloy-4 is evaluated. As seen in Fig. 21, after core dry out, the heat transfer between fuel rods and steam is considerably low and temperature of clad increases significantly until clad melts. The SiC clad failure occurs at 12,300 s. It is about 3100 s over zircaloy-4 due to having the higher melting point and much lower magnitude of oxidation rate, mainly. As shown in Fig. 22, total cumulative oxidation heat of SiC is 10.26 GJ, which means 81% decrement in energy in case of using SiC that leads to delay in fuel degradation and failure of RPV lower plenum in comparison of zircaloy-4. Once failure of the support plate occurs, the core relocation starts. As seen in Fig. 23, after the relocation of molten core materials to lower plenum, the RPV fails due to creep rupture of the lower head at 13,700 s in case of SiC clad. The lower plenum of SiC could withstand more than 2100 s in comparison of zircaloy4. Total debris mass ejected through vessel of SiC is 146 tons and 2.1 tons less than debris mass of zircaloy-4. After the core uncovered, the steam-clad reaction, accompanied by hydrogen pick up. As Fig. 24 illustrates, however, the hydrogen generation begins almost at the same time for both zircaloy-4 and SiC, due to having much less oxidation rate of SiC, total in-vessel generated hydrogen of SiC is 222 kg. The result indicates, using SiC concludes 47% decrement in hydrogen production.
Allendorf, M.D., Melius, C.F., Ho, P., Zarchariah, M.R., 1995. Theoretical study of the thermochemistry of molecules in the Si-O-H system. J. Phys. Chem. 99 (41), 15285–15293. Bragg-Sitton, S.M., 2012. Light Water Reactor Sustainability Program Advanced LWR Nuclear Fuel Cladding System Development Technical Program Plan. Technical Report INL/MIS-12-25696. Idaho National Laboratory. Carmack, J., Goldner, F., 2014. Forward for special JNM issue on accident tolerant fuels for LWRs. J. Nucl. Mater. 1, 373. Cheng, T., Keiser, J.R., Brady, M.P., Terrani, K.A., Pint, B.A.J., 2012. Oxidation of fuel cladding candidate materials in steam environments at high temperature and pressure. J. Nucl. Mater. 427, 396–400. Farmer, M.T., Leibowitz, L., Terrani, K.A., Robb, K.R., 2014. Scoping assessments of ATF impact on late-stage accident progression including molten core–concrete interaction. J. Nucl. Mater. 448, 534–540. Final Safety Assessment Report (FSAR) for BNPP, 2007. Russia Federal Agency on Nuclear Energy (RFANE). Accident Analysis, Book 4, Moscow. Gharari, R., Kazeminejad, H., Mataji-Kojouri, N., Hedayat, A., 2018. A review on hydrogen generation, explosion, and mitigation during severe accidents in light water nuclear reactors. Int. J. Hydrogen Energy 43 (4), 25. Gauntt, R.O., Cash, J.E., Cole, R.K., Erickson, C.M., Humphries, L.L., Rodriguez, S.B., Young, M.F., 2005a. MELCOR Computer Code Manuals, Vol. 1: Primer and Users’ Guide. Gauntt, R.O., Cash, J.E., Cole, R.K., Erickson, C.M., Humphries, L.L., Rodriguez, S.B., Young, M.F., 2005b. MELCOR Computer Code Manuals, Vol. 2: Reference manuals. Gauntt, R.O., Ross, K., Wagner, K., 2002. Simulation of TMI-2 Phase2 With an Enhanced 2-Dimensional In-Vessel Natural Circulation Model. In: 10th International Conference on Nuclear Engineering, Arlington, Virginia, USA, Vol. 3. American Society of Mechanical Engineers. pp. 487–494. George, N.M. et al., 2015. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors. Annals Nucl. Energy 75, 703–712. Katoh, Y. et al., 2012. Radiation effects in SiC for nuclear structural applications. Curr. Opinion Solid State Mater. Sci. 16 (3), 143–152. Lindley, B.A., Kotlyar, D., Parks, G.T., Lillington, J.N., Petrovic, Bojan, 2016. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes. EPJ Nuclear Sci. Technol. 2, 14. Merrill, B.J., Bragg-Sitton, S.M., Humrickhouse, P.W., 2017. Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials. Nucl. Eng. Design 315, 170–178. Merrill, B.J., Bragg-Sitton, S.M., 2013. Advanced Cladding Modeling Work to Assess Cladding Performance Under Accident Conditions. INL/EXT-13-30206. Merrill, B.J., 2012. MELCOR SiC Modifications for Severe Accident Analysis Applications, Idaho National Laboratory Report, INL/LTD-12-27961. Opila, E.J., 2003. Oxidation and volatilization of silica formers in water vapor. J. Am. Ceram. Soc. 86 (8), 1238–1248. Pint, B.A., Terrani, K.A., Brady, M.P., Cheng, T., Keiser, J.R., 2013. High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments. J. Nucl. Mater. 440, 420–427.
6. Conclusion Today for mitigation of accident consequences, it is needed to evaluate the accident tolerant clads over the traditional zirconium alloy cladding employed in the nuclear industry. In this work, it is focused mainly on FeCrAl and SiC as optimized cladding materials in comparison with zircaloy-4 during station blackout accident in the BNPP. An important advantage of these materials is slower oxidation kinetics conducts to produce much less hydrogen than zircaloy-4. In this research, the thermophysical properties and oxidation kinetics for these ATFs replaced with zircaloy-4 and perfor-
Appendix A. Supplementary material Supplementary data to this article can be found online at https://doi.org/10.1016/j.anucene.2019.02.008.
H. Ebrahimgol et al. / Annals of Nuclear Energy 129 (2019) 214–223 Richet, P., Bottinga, Y., Denielou, L., Petitet, J.P., Tequi, C., 1982. Thermodynamic properties of quartz, cristobalite and amorphous SiO2: drop calorimetry measurements between 1000 and 1800 K and a review from 0 to 2000 K. Geochimica et Cosmochimica Acta 46 (12), 2639–2658. Robb, K.R., McMurray, J.W., Terrani, K.A., 2016. Severe accident analysis of BWR core fueled with UO2/FeCrAl with update materials and melt properties from experiments. Fuel Cycle Res. Develop. Adv. Fuels Campaign. Terrani, K.A., Pint, B.A., Parish, C.M., Silva, C.M., Snead, L.L., Katoh, Y., 2014. Silicon carbide oxidation in steam up to 2 MPa. J. Am. Ceram. Soc. 97 (8), 2331–2352.
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Terrani, K.A., Zinkle, S.J., Snead, L.L., 2013. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding. J. Nucl. Mater. 448, 420–435. Wang, J., Mccabe, M., Wu, L., Dong, X., Wang, X., Haskin, T.C., Corradini, M.L., 2017. Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out. Nucl. Eng. Design 313, 458–469. Weast, R.C., Astle, M.J., 1980. Handbook of Chemistry and Phyics. CRC Press, p. B234. Yueh, K., Terrani, K.A., 2014. Silicon carbide composite for light water reactor fuel assembly applications. J. Nucl. Mater. 448 (1–3), 380–388.