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SCWR transient safety analysis code SCAC-CSR1000
Progress in Nuclear Energy 83 (2015) 318e325
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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
SCWR transient safety analysis code SCAC-CSR1000 Liang Liu a, b, Tao Zhou a, b, *, Yu Li a, b, Juan Chen a, b, Muhammad Zeeshan Ali a, b, Zejun Xiao c a
North China Electric Power University Nuclear Thermal Safety and Standardization Research Institute, Beijing 102206, China Beijing Passive Technology Key Laboratory, Beijing 102206, China c China Nuclear Power Institute Nuclear Reactor System Design Technology Laboratory, Chengdu 610041, China b
a r t i c l e i n f o
a b s t r a c t
Article history: Received 30 October 2014 Received in revised form 31 March 2015 Accepted 3 April 2015 Available online 16 May 2015
CSR1000 was selected as the research object. A code named SCAC-CSR1000 has been developed based on the SCAC code. The reliability of the code was verified by comparing the result of SCAC-CSR1000 and SCTRAN. Then the safety analysis was carried out. Five events were selected, that are partial loss of reactor coolant flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure and loss of feed water heating. By the numerical analyses, it was found that the MCST does not exceed 1260 C, meets the design safety requirements. The 2nd MCST are higher than 1st MCST. The isolation of main steam line has the least safety margin. © 2015 Elsevier Ltd. All rights reserved.
Keywords: CSR1000 Transient Safety SCAC
1. Introduction Supercritical water-cooled reactor (SCWR) (Cheng and Liu, 2008) has been considered as one of the most promising Generation-IV reactors. The two advantages are the compactness of the plant system due to the high specific enthalpy of supercritical water and the simplicity of the plant system without the recirculation system and dryers of BWRs and steam generators of PWRs. Various aspects of SCWRs are being carried out around the world. For pressure-vessel type SCWR, The feasibility assessments of the Super Light Water Reactor (Super LWR) and the Super Fast Reactor (Super FR) (Oka et al., 2010) were proposed by the University of Tokyo since 1990s. In NERI project, the US SCWR has been put forward since 2001 (MacDonald et al., 2004). Research and development of SCWR are being done in Europe as the High Performance Light Water Reactor (HPLWR) with funding by the European Union (Schulenberg et al., 2011). SCWR research in Korea has been mainly promoted by the Korea Atomic Energy Research Institute (KAERI) and Korea Electric Power Research Institute (KEPRI) (Bae et al., 2007). A pressure-tube type CANDU SCWR has been developed by AECL and various universities in CANADA (Yu et al., 2009). In China, a new thermal spectrum SCWR concept named CSR1000 is
* Corresponding author. http://dx.doi.org/10.1016/j.pnucene.2015.04.003 0149-1970/© 2015 Elsevier Ltd. All rights reserved.
proposed by Nuclear Power Institute of China (NPIC) (Xiao et al., 2013). It is an essential process to validate the feasibility of the safety performance of SCWR. In the past researches, many researchers has done the safety analysis work by various methods. The SPRATDOWN, SPRAT-DOWN-DP and SCRELA reflood code (Ishiwatari et al., 2005) were made by University of Tokyo aiming to evaluate the safety characteristic of Super LWR. It can do the safety analysis on transients and accidents. The safety analysis on US SCWR had been evaluated by the RELAP5-3D/ATHENA computer program. The results showed that US SCWR meet the safety thermal limits (MacDonald et al., 2004). For the CANDU SCWR, a code named SUBCHAN was made to analysis the sub channel characteristic (Yu et al., 2009). So far, there are many other analysis code or program are made, such as RELPA5, CATHARE, APROS, ATHLET 2.1 and SCTRAN. Using different codes or programs to analysis the same reactor type can make the reliability more confident. As for CSR1000, the Preliminary safety evaluation has been made (Wu et al., 2014). The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable. This paper introduces a new analysis code named SCACCSR1000. It is a code for the safety analysis. It is developed from the SCAC code (Chen, 2013), which is made by FORTRAN language. Using the Watts correlation and the pressure drop balance method
L. Liu et al. / Progress in Nuclear Energy 83 (2015) 318e325
Mom
Nomenclature N
b l b C S t z
r G u g f D P h A l Q''
Total fission power total effective delayed neutron fraction Neutron decay constant of the delayed group Effective share of the delayed neutron group The fission power of delayed neutron group The external neutron source term Time(s) Position(m) Density(kg/m3) Mass flux(kg/s) Flow rate(m/s) Acceleration of gravity(m/s2) Friction coefficient Equivalent diameter(m) Pressure(MPa) Specific enthalpy(J/kg) Circulation area(m2) Node height(m) Heat release line power density of fuel rods(W/m)
319
Momentum(kg/m2,s)
Superscript K Time step Subscripts i The number of the neutron group W Water rod f Fuel rod DC Downcomer path round Round path WR Water rod path Abbreviation SCWR Supercritical water-cooled reactor CR Control rod MCST Maximum cladding surface temperature RCP Reactor coolant pump PWR Pressurized water reactor BWR Boiling water reactor
can make the result more conservative and scientific. Safety characteristics of CSR1000 under abnormal transients and accidents in supercritical pressure conditions are proposed along with the sensitive analyses results. These results are meaningful for improving the performance of CSR1000, especially the increasing value and changes over time of MCST. 2. Research object 2.1. The overall parameters CSR1000 (Wu et al., 2014) is a thermal spectrum reactor, which is cooled and moderated by light water. The thermal efficiency is 43.5%. In order to achieve this thermal efficiency, the core inlet temperature is 280 C, outlet temperature is 500 C. Table 1 shows the specific parameters. The Figs. 1 and 2 show the fuel component. From Table 1 and Figs. 1e2, we can see that the total number of fuel assemblies is 177. They are divided into first and second pass assemblies in accordance with different neutron energy spectrum and different coolant flow. Wherein the number of first pass assemblies are 57. The number of second pass assemblies are 120. There are 244 fuel rods and 4 water rod in each assembly. It should be noted that the fuel assembly use cruciform control rod design, evenly distributed in the reactor pressure vessel.
Fig. 1. Component distribution.
2.2. Flow scheme Different from the conventional SCWR design, the flow scheme of CSR1000 adopts a flow pattern of inside to the outside. The coolant flow pattern is shown in Fig. 3. From Fig. 3, we can see that 76.7% of the coolant is directed to the top dome, which is then divided into three parts, including 35.9% Table 1 The parameters of CSR1000. Core
CSR1000
Coolant inlet/outlet temperatures ( C) Primary coolant line/main steam line Fuel rod diameter/pitch/cladding thickness (mm) The wall thickness of the moderator channel (mm)
280/500 2/2 9.5/10.5/0.57 0.80
Fig. 2. CSR100 Fuel component.
320
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Fuel rod number/water rod number Fuel assemblies/first pass assemblies/second pass assemblies Core pressure/main steam pressure (MPa) Heat/electric power (MW)/Thermal Efficiency (%) Main coolant flow rate (kg/s) Neutron spectrum Cladding material Average power density (MW/m3) Active core height (m)
flowing downward through the fuel channel of the first pass, 10.8% flowing downward through the water rods of the first pass and 30% flowing downward through the water rods of the second pass. The coolant from the downcomer (23.3% of the total), the fuel channel of the first pass and the water rods mixed together in the lower plenum, and then flows upwards through the fuel channel of second pass. This specific flow scheme is favorable to improving the neutron utilization, reducing the hot channel factor and increasing the coolant outlet temperature. 2.3. Control system The control system of CSR1000 use the same design arrangement of Super LWR (Oka et al., 2010). The actuation conditions of the safety system is summarized by Xi'an Jiaotong University (Wu et al., 2014). The control system is shown in Fig. 4. From Fig. 4, we can see that there are three control systems. They are pressure control system, power control system and steam temperature control system. The turbine control valves and the turbine bypass valve formed the pressure control system. The CRs formed the power control system. The feedwater pumps formed the steam temperature system. 3. Calculation model
224/4 177/57/120 25/24.5 2300/1000/43.5 1190 Thermal Stain steel (310S) 60 4.2
6 X dNðtÞ ðrðtÞ bÞ ¼ NðtÞ þ li Ci ðtÞ þ S dt L i¼1
(1)
dCi ðtÞ b ¼ li Ci ðtÞ þ i NðtÞ dt L
(2)
ð i ¼ 1e6Þ
where，N(t) is the total fission power, b s the total effective delayed neutron fraction, L is the time of each prompt neutron generation, li is the neutron decay constant for the delayed group i, bi is the effective share of the delayed neutron group ii, Ci(t) is the fission power of delayed neutron group i, S is the external neutron source term. 3.2. Governing equations Governing equations of mass, momentum and energy conservations are shown below.
vrðz; tÞ vGðz; tÞ þ ¼0 vt vz
(3)
vðruÞ v ru2 2f vP þ ¼ rg cos q þ ru2 H vt Dh vz vz
(4)
00 00 vfrðz; tÞhðz; tÞg vGðz; tÞhðz; tÞ 1 þ ¼ lf Q ðz; tÞ lf Qw ðz; tÞ vt vz Aw
3.1. Point reactor kinetics model Using reactivity r and prompt neutron generation timeL, we can also write the point reactor kinetics equations as follows.
Fig. 3. Coolant flow pattern.
(5)
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Table 3 Relief valve function. Relief valve function
Relief valve (20% capacity 8units)
Open(MPa)
Close Open(MPa)
Number
26.2 26.4 26.6 26.8
25.2 25.4 25.6 25.8
1 1 3 3
2010). The flow rate ratio between the first meshes of the two flow paths is iteratively changed to satisfy the momentum conservation.
Momkround Momk1 DPtotWR DPtotDC round ¼ Dt lround
Where, l is the length (m), Mom is the momentum per unit volume (kg/m2,s), Dt is the time step spacing (s), superscript K is the time step, subscript DC is the downcomer path, subscript round is the round path, subscriptWR is the water rod path.
Fig. 4. Control system.
vfrðz; tÞhðz; tÞg vGðz; tÞhðz; tÞ 1 00 þ ¼ lw Qw ðz; tÞ vt vz Aw
(7)
4. Code description
(6)
where, t is the time(s), z is the position(m), r is the density (kg/ m3), G is the mass flux (kg/s), u is the flow rate (m/s), g is the acceleration of gravity (m/s2), f is the friction coefficient, fi ¼ 0:0791Re0:25 , Dh is the Equivalent diameter(m), P is the presi sure (MPa), h is the specific enthalpy (J/kg), Aw is the circulation area 00 (m2), lf is the Node height(m), Q'', Qw are the heat release line power density of fuel rods and moderator (W/m). Note that, Equation (5) is for the coolant energy conservation equation. Equation (6) is for the moderator energy conservation equation. The governing equations are discretized using the upwind difference scheme and the explicit scheme. The boundary conditions are the feedwater flow rate, the feedwater temperature, and the turbine inlet flow rate.
4.1. Node division Each flow channel of CSR1000 is simplified to an average channel and a hot channel. Wherein, the power distribution based on the heat factor of the channel is set to 1.26. Core node division is shown in Fig. 5. From Fig. 5, we can see that the average thermal path channel and the hottest channel are divided into 40 axial nodes in order to improve the accuracy of the calculation process. Wherein, all the coolant channels have heat exchange with moderator channel. Axial power distribution in the core is calculated by neutron kinetics, and finally get the normalized axial power distribution. 4.2. Calculation flow
3.3. Heat transfer correlation There are upward and downward flow regions in the CSR1000 flow scheme design. The watts correlation (Watts and Chou, 1982) has considered the difference between the upward and downward flow. Table 2 shows the supercritical water heat transfer correlation and scope. By using Watts correlation, the MCST is higher than which is calculated by OKA-Koshizuka correlation (Oka et al., 2010), Bishop correlation (Bishop et al., 1964) or DittuseBoelter correlation (Yu & Zhu, 2002). Using watts correlation can get the more conservative result. So the SCAC-CSR1000 code use this correlation. 3.4. Pressure drop balance model At the normal operating conditions, the total pressure drop is calculated by adding the four pressure drops, that are friction drop, acceleration drop, buoyancy drop and the orifice drop. The 4.2 m core height total pressure drop of each path is 0.019 MPa (Oka et al.,
The SCAC-CSR1000 code is developed from SCAC code. On the basis of the original code SCAC, the new code adds the safety control modules and the pressure drop balance module. The IAPWS-IF95 water property package (Wagner et al., 2002) is also added into the new code to expand the calculation range of water property. The density feedback and the Doppler feedback are added into the new code. The density coefficient is a function of the coolant density. The Doppler coefficient is a function of the fuel temperature (Oka et al., 2010). Both of them are used for calculation in point reactor kinetic equation. The calculation flow is shown in Fig. 6. From Fig. 6, we can see that the code reads the initial parameters firstly. The results of initial parameters are got by steady-state calculation. For the difference comparing to the widely used code such as RELAP5 and CATHARE, the redistribution of mass flow rate in each channel is achieved by pressure drop balance model. This model can be easily adjusted to every different situation by changing the condition settings. This can be a more flexible solution for the flow redistribution problem. The solving method for
Table 2 Supercritical water heat transfer correlations and scope. Correlations name
Pressure/MPa
Mass flow density/kg m2 s1
Heat flux/MW m2
DittuseBoelter Bishop Swenson Yamagata Xufeng et al. Watts
Subcritical conditions 22.8e27.6 22.8e41.4 22.6e29.4 23e30 25
e 680e3600 542e2150 310e8130 600e1200 160e1060
e 0.31e3.46 0.2e1.8 0.116e0.93 0.1e0.6 0.175e0.44
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Fig. 5. Core node division.
The line power density is a key influence factor for coolant temperature. And it is also an important parameter in Neutronics/ Thermal-hydraulics coupling. The calculated power result is compared with the design power data for the power verification. The power share is the proportion of the sum of each node power in the total power. The calculated result is shown in Fig. 7. As can be seen from Fig. 6, the deviations of line power density, first pass power share and the second pass power share are all less than 0.2%. The calculated result is closed to the design data. The code power calculation is reliable.
linearly reduced to 50%. The transient calculation result is shown in Fig. 10. As can be seen from Fig. 10, the inlet flow rate decreased to 90% 1 s after the pump tripped. Then the reactor start to shut down after 0.5 s delay. The CRs were inserted into the core rapidly. Then the reactor power decreased rapidly. The two passes coolant flow rate are all decreased. Wherein the flow rate of first pass decreased rapidly, while that of the second pass decreased slower. The two passes flow rate respectively reach the minimum value at 5.2 s and 14 s. The 1st pass flowrate eventually maintains above 50%. It is because the 1st pass coolant channel has the lower orifice coefficient. And the pressure drop balance model balances the flowrate in each channel. The decreasing of the coolant flow rate leaded to the increasing of the MCST. But after the two passes MCSTs reached the maximum value at 8 s, the MCSTs decreased after that. It is because the heat production from fuel rod is decreased, but the coolant removed more heat. Because of the change of the flow and the density of the water, the core pressure decreased obviously in the first 5 s. And then it kept stable at 24.2 MPa.
5.2. Coolant temperature verification
6.2. Isolation of main steam line
The coolant temperature is the key parameter for the characteristic of the core. The calculation time is set to 300 s. The result of stead-state coolant temperature is got. Then the result is compared with the same reactor analysis code SCTRAN (Wu et al., 2014). The results are shown in Figs. 8 and 9. As can be seen from Figs. 8 and 9, the result of SCACdCSR1000 is similar to that of SCTRAN. The two codes can meet the inlet temperature at 280 C and the outlet temperature at 500 C. While the highest coolant temperatures of first pass are between 400 and 420 C. The deviation is within 5%. The SCAC-CSR1000 code and the SCTRAN code use different heat transfer correlations and the solving methods. Because the Watt's correlation gets the lower heat transfer correlation, the heat transferred to the coolant become less so the coolant temperature become lower.
The turbine valve started to close at the 0.6 s. The valve opening rapidly reduced to 5% in 1 s. The turbine valve totally closed at the 3 s. The turbine bypass valve opens at the pressure greater than 26.2 MPa. The behavior of the relief valve is shown in Table 3. In the process of turbine valve closing, the pressure of the core increased. So it is required that the turbine bypass relief valve open to decrease the core pressure. The electric power cannot be supplied to the coolant pump after the turbine valve has been closed. After losing external power, the coolant pump can still work 10 s. The coolant pump idle time is 5 s. The transient calculation result is shown in Fig. 11. As can be seen from Fig. 11, the core pressure had a significant increase after the turbine valve started to close. At the 3.5 s, the core pressure exceeded 26 MPa triggered the shutdown signal. The CRs were inserted into the core rapidly. Then the reactor power decreased rapidly. When the pressure exceeded 26.2 MPa, the turbine bypass relief valve opened. With the constant opening and closing of the pressure relief valve, the core pressure continuously changed between 25 MPa and 26.8 MPa. The two passes coolant flow appeared regular pulse changes. Also that was caused by constant opening and closing of the pressure relief valve. Between
governing equation is developed so that it can be used for backflow calculation. In neutron kinetics calculation module, the reactor power and reactivity coefficient are got by six groups of delayed neutrons point reactor model calculation. 5. Code verification 5.1. Power verification
6. The results 6.1. Partial loss of flow At the 0 s time, one of the two coolant pump tripped. The coolant pump idle time is 5 s. In 5 s the primary coolant flow
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2.0
Start
1.8
Line power density First pass power share Second pass power share
1.6
Reads the initial parameters
Deviation(%)
Solving Governing equations
1.4 1.2 1.0 0.8 0.6 0.4
Convergence judgment
0.2
0.18259
0.13589
0.0547
0.0
(1)
(2)
(3)
Fig. 7. Deviation between calculation and design.
Next node calculation
Judgment on pressure drop balance of each channel
The hottest channel calculation
Neutron kinetics calculation
Results output Fig. 8. SCAC-CSR1000 coolant temperature.
End Fig. 6. Calculation flow.
15 and 20 s, the first pass coolant flow occurred a backflow phenomenon. This is because the coolant in the coolant pump is not provided. The coolant flow was redistributed. The second pass MCST appeared downward trend after it had a significant increasing in the beginning 17 s.
6.3. Uncontrolled CR withdrawal The maximum reactivity worth of a CR cluster depends on the loading pattern of the fuel assemblies and the CR pattern. Any limitation of the loading pattern or any interlock of the CR pattern, like the rod worth minimizer of BWRs, is not considered conservatively. The highest reactivity worth of a CR luster under all the considerable loading patterns and CR patterns is estimated as 1.05%
Fig. 9. SCTRAN coolant temperature.
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shutdown signal was triggered because the reactor power rate exceed 120%. The CRs were inserted into the core rapidly. Then the reactor power decreased rapidly. The first pass coolant flow showed a phenomenon of substantial increase after the shutdown signal was triggered. There are three reason cause it. The first reason is that the pressure drop and the coolant density increase. The second reason is that the first pass coolant channel has a small orifice correlation. The third reason is the pump power increase when the pressure is dropped. The inlet flow rate had a sharply drop after a short rising. The two passes MCST and the pressure had a deep decline. According to the relationship of pressure drop between two passes, the first pass coolant pressure had greater drop than that of the second pass. 6.4. Reactor coolant pump seizure
Fig. 10. Partial loss of flow.
At the 0 s, coolant flow from one of the tow RCPs is assumed to suddenly stop. The primary coolant flow rate reduced to 50% instantly. The results is shown in Fig. 13. As can be seen from Fig. 13, different from the partial loss of flow transient, the inlet flowrate shows a different result. It is because the inlet flowrate is decreased to 50% at the 0 s. And after the shutdown signal was triggered, the only running reactor coolant pump losses the power supply at the 15 s. The two passes MCST increased faster than that of the partial loss of flow accident. In the first 6 s, the two passes MCST increased because of the instantaneous reduce of coolant flow rate. The shutdown signal was triggered at 0 s. The core power and pressure decrease rapidly. The inlet flow rate, first pass coolant flow rate and the second pass coolant flow rate were all showed a slowly rising trend. It was caused by slowly pressure drop. The first pass coolant flow rate showed a backflow phenomenon after the 39 s. 6.5. Loss of feedwater heating
Fig. 11. Isolation of main steam line.
dk/k at the normal operating condition. The calculation result is shown in Fig. 12. As can be seen from Fig. 12, the reactivity increased with the uncontrolled CR withdrawal started from the 0 s. The reactor power increased. But the power increasing rate is small due to the reactivity feedbacks from the water density and fuel temperature. The cladding temperature increased by only 17 C because the main coolant flow rate was increased by the control system so as to keep the main steam temperature. At the 80 s time, the
Fig. 12. Uncontrolled CR withdrawal.
Loss of one stage of the feedwater heating will cause a 35 C drop of the feedwater temperature. In the safety analysis, it is conservatively assumed as 55 Cas is done in the safety analysis of Super LWR. The calculation result is shown in Fig. 14. As can be seen from Fig. 14, there is a short decrease of the two passes coolant flow rate in the first 5 s. It is because the density of the two passes coolant increased. During that time, the two passes MCSTs has a rapidly increase. After that, in order to keep 100% of the power, the control system withdrawn the CRs. Meanwhile, in order to maintain the main steam temperature at 500 C, the control system increased the primary coolant flow. The MCST decreased and the reactor power increased during the flow rate
Fig. 13. Reactor coolant pump seizure.
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7. Conclusion
Fig. 14. Loss of feedwater heating.
The SCAC-CSR1000 code is developed from SCAC code (Chen et al., 2013). The steady-state results of power and coolant temperature are got. The results are compared with the same reactor type analysis code SCTRAN. The reliability of the code is verified. After that the five events are selected for the safety analysis by using the new code. The CSR1000 reactor is able to ensure the MCSTs are below the safety criterion of 1260 C. The second pass MCST is higher than that of the first pass. The isolation of main steam line has the least safety margin. This study introduces a new analysis code named SCACCSR1000 and clarifies the characteristics of the CSR1000 during events at supercritical pressure condition. The results of the safety analysis and the sensitivity analysis, especially the increase in the cladding temperature and its duration, are useful for improving the CSR1000 performance. Further safety analysis is necessary to clarify the characteristics of the CSR1000 at all the abnormal transients and accidents. Acknowledgments The research is supported by the Research Fund of the China Key Laboratory of Nuclear Reactor System Design Technology (2014BJ0041), the Research Fund of the China Key Laboratory of nuclear thermal-hydraulic technology (2013B40), the Central Universities special funds basic research and business projects (2014BJ0086) (2014BJ0087). References
Fig. 15. Sensitivity analysis.
increasing time. After the main steam temperature reached 500 C, both the coolant flow rate and the power are decreased. When the core inlet flow rate was reduced to 90%, the shutdown signal was triggered. The CRs were inserted into the core rapidly. Then the reactor power decreased rapidly. While the two passes MCST dropped. 6.6. Sensitivity analysis Oxidation behavior of stainless steel and Ni-alloy is expected to be milder than that of Zircaloy. In this study, the criterion of cladding temperature is set to 1260 C for stainless steel cladding, taken from the criterion for LOCA of U.S.-PWR with stainless steal cladding. The MCSTs of five events are summarized. The order of the events are partial loss of flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure, loss of feedwater heating. The result is shown Fig. 15. As can be seen from Fig. 15, the MCSTs of all the events did not exceed the 1260 C safety criterion. The second pass MCST is higher than that of the first pass. In accordance with the safety margin, the descending order is uncontrolled CR withdrawal, loss of feedwater heating, partial loss of flow, reactor coolant pump seizure, isolation of main steam line.
Bae, Y.Y., Jang, J., Kim, H.Y., Yoon, H.Y., Kang, H.O., Bae, K.M., 2007. Research activities on a supercritical pressure water reactor in Korea. Nucl. Eng. Technol. 39, 273e286. Bishop, A.A., Sandberg, R.O., Tong, L.s., 1964. Forced Convection Heat Transfer to Water at Near-Critical Temperatures and Supercritical Pressures[R]. Report WCAP-2056, Part IV. Westinghouse Electric Corp, Pittsburgh, USA. Chen, J., 2013. Coupled Neutronics/Thermal-hydraulics Simulation and Safety Analysis of Supercritical Light Water Reactor. Doctoral thesis. North China Electric Power University. Chen, J., Zhou, T., et al., 2013. Influence analysis of coupled neutronics and thermalhydraulics on characteristics of supercritical water-cooled reactor system. At. Energy Sci. Technol. 47 (5), 804e810. Cheng, X., Liu, X.J., 2008. Research status and prospect of super critical water-cooled reactor. At. Energy Sci. Technol. 42, 168e172. Ishiwatari, Y., Oka, Y., Koshizuka, S., Yamaji, A., Liu, J., 2005. Safety of super LWR, (I) safety system design. J. Nucl. Sci. Technol. 42, 927e934. MacDonald, P., Buongiorno, J., Davis, C., Witt, R., 2004. Feasibility Study of Supercritical Water Cooled Reactors for Electric Power Production (Final Report), INEEL/EXT-04e02530. Oka, Y., Koshizuka, S., Ishiwatari, Y., Yamaji, A., 2010. Super Light Water Reactors and Super Fast Reactors. Springer, New York. czy, C., Laurien, E., Schulenberg, T., Starflinger, J., Marsault, P., Bittermann, D., Mara Nijeholt, J.A., Anglart, H., Andreani, M., Ruzickova, M., Toivonen, A., Lycklama a 2011. European supercritical water cooled reactor. Nucl. Eng. Des. 241, 3505e3513. Wagner, et al., 2002. The IAPWS formulation 1995 for the thermodynamic properties of ordinary water substance for general and scientific use. J. Phys. Chem. Ref. Data 31 (2), 387e535. Watts, M.J., Chou, C.T., 1982. Mixed convection heat transfer to supercritical pressure water[C]. In: Proceedings of the 7th International Heat Transfer Conference, Munchen, Germany, 3, pp. 495e500. Wu, P., Gou, J., Shan, J., et al., 2014. Preliminary safety evaluation for CSR1000 with passive safety system. Ann. Nucl. Energy 65, 390e401. Xiao, Z., Li, X., Huang, Y.-P., Tang, R., Luo, Q., Zang, F.-G., Li, Q., Li, P.-Z., Yi, W., 2013. Overview of research and development (Phase I) on key technologies for supercritical water-cooled reactor. Nucl. Power Eng 1e4. Yu, P.A., Zhu, J.Z., 2002. Nuclear Reactor Thermal Analysis. Shanghai Jiaotong University Press. Yu, J., Liu, H., Jia, B., 2009. Sub-channel analysis of CANDUeSCWR and review of heat-transfer correlations. Prog. Nucl. Energ. 51, 246e252.
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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
SCWR transient safety analysis code SCAC-CSR1000 Liang Liu a, b, Tao Zhou a, b, *, Yu Li a, b, Juan Chen a, b, Muhammad Zeeshan Ali a, b, Zejun Xiao c a
North China Electric Power University Nuclear Thermal Safety and Standardization Research Institute, Beijing 102206, China Beijing Passive Technology Key Laboratory, Beijing 102206, China c China Nuclear Power Institute Nuclear Reactor System Design Technology Laboratory, Chengdu 610041, China b
a r t i c l e i n f o
a b s t r a c t
Article history: Received 30 October 2014 Received in revised form 31 March 2015 Accepted 3 April 2015 Available online 16 May 2015
CSR1000 was selected as the research object. A code named SCAC-CSR1000 has been developed based on the SCAC code. The reliability of the code was verified by comparing the result of SCAC-CSR1000 and SCTRAN. Then the safety analysis was carried out. Five events were selected, that are partial loss of reactor coolant flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure and loss of feed water heating. By the numerical analyses, it was found that the MCST does not exceed 1260 C, meets the design safety requirements. The 2nd MCST are higher than 1st MCST. The isolation of main steam line has the least safety margin. © 2015 Elsevier Ltd. All rights reserved.
Keywords: CSR1000 Transient Safety SCAC
1. Introduction Supercritical water-cooled reactor (SCWR) (Cheng and Liu, 2008) has been considered as one of the most promising Generation-IV reactors. The two advantages are the compactness of the plant system due to the high specific enthalpy of supercritical water and the simplicity of the plant system without the recirculation system and dryers of BWRs and steam generators of PWRs. Various aspects of SCWRs are being carried out around the world. For pressure-vessel type SCWR, The feasibility assessments of the Super Light Water Reactor (Super LWR) and the Super Fast Reactor (Super FR) (Oka et al., 2010) were proposed by the University of Tokyo since 1990s. In NERI project, the US SCWR has been put forward since 2001 (MacDonald et al., 2004). Research and development of SCWR are being done in Europe as the High Performance Light Water Reactor (HPLWR) with funding by the European Union (Schulenberg et al., 2011). SCWR research in Korea has been mainly promoted by the Korea Atomic Energy Research Institute (KAERI) and Korea Electric Power Research Institute (KEPRI) (Bae et al., 2007). A pressure-tube type CANDU SCWR has been developed by AECL and various universities in CANADA (Yu et al., 2009). In China, a new thermal spectrum SCWR concept named CSR1000 is
* Corresponding author. http://dx.doi.org/10.1016/j.pnucene.2015.04.003 0149-1970/© 2015 Elsevier Ltd. All rights reserved.
proposed by Nuclear Power Institute of China (NPIC) (Xiao et al., 2013). It is an essential process to validate the feasibility of the safety performance of SCWR. In the past researches, many researchers has done the safety analysis work by various methods. The SPRATDOWN, SPRAT-DOWN-DP and SCRELA reflood code (Ishiwatari et al., 2005) were made by University of Tokyo aiming to evaluate the safety characteristic of Super LWR. It can do the safety analysis on transients and accidents. The safety analysis on US SCWR had been evaluated by the RELAP5-3D/ATHENA computer program. The results showed that US SCWR meet the safety thermal limits (MacDonald et al., 2004). For the CANDU SCWR, a code named SUBCHAN was made to analysis the sub channel characteristic (Yu et al., 2009). So far, there are many other analysis code or program are made, such as RELPA5, CATHARE, APROS, ATHLET 2.1 and SCTRAN. Using different codes or programs to analysis the same reactor type can make the reliability more confident. As for CSR1000, the Preliminary safety evaluation has been made (Wu et al., 2014). The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable. This paper introduces a new analysis code named SCACCSR1000. It is a code for the safety analysis. It is developed from the SCAC code (Chen, 2013), which is made by FORTRAN language. Using the Watts correlation and the pressure drop balance method
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Mom
Nomenclature N
b l b C S t z
r G u g f D P h A l Q''
Total fission power total effective delayed neutron fraction Neutron decay constant of the delayed group Effective share of the delayed neutron group The fission power of delayed neutron group The external neutron source term Time(s) Position(m) Density(kg/m3) Mass flux(kg/s) Flow rate(m/s) Acceleration of gravity(m/s2) Friction coefficient Equivalent diameter(m) Pressure(MPa) Specific enthalpy(J/kg) Circulation area(m2) Node height(m) Heat release line power density of fuel rods(W/m)
319
Momentum(kg/m2,s)
Superscript K Time step Subscripts i The number of the neutron group W Water rod f Fuel rod DC Downcomer path round Round path WR Water rod path Abbreviation SCWR Supercritical water-cooled reactor CR Control rod MCST Maximum cladding surface temperature RCP Reactor coolant pump PWR Pressurized water reactor BWR Boiling water reactor
can make the result more conservative and scientific. Safety characteristics of CSR1000 under abnormal transients and accidents in supercritical pressure conditions are proposed along with the sensitive analyses results. These results are meaningful for improving the performance of CSR1000, especially the increasing value and changes over time of MCST. 2. Research object 2.1. The overall parameters CSR1000 (Wu et al., 2014) is a thermal spectrum reactor, which is cooled and moderated by light water. The thermal efficiency is 43.5%. In order to achieve this thermal efficiency, the core inlet temperature is 280 C, outlet temperature is 500 C. Table 1 shows the specific parameters. The Figs. 1 and 2 show the fuel component. From Table 1 and Figs. 1e2, we can see that the total number of fuel assemblies is 177. They are divided into first and second pass assemblies in accordance with different neutron energy spectrum and different coolant flow. Wherein the number of first pass assemblies are 57. The number of second pass assemblies are 120. There are 244 fuel rods and 4 water rod in each assembly. It should be noted that the fuel assembly use cruciform control rod design, evenly distributed in the reactor pressure vessel.
Fig. 1. Component distribution.
2.2. Flow scheme Different from the conventional SCWR design, the flow scheme of CSR1000 adopts a flow pattern of inside to the outside. The coolant flow pattern is shown in Fig. 3. From Fig. 3, we can see that 76.7% of the coolant is directed to the top dome, which is then divided into three parts, including 35.9% Table 1 The parameters of CSR1000. Core
CSR1000
Coolant inlet/outlet temperatures ( C) Primary coolant line/main steam line Fuel rod diameter/pitch/cladding thickness (mm) The wall thickness of the moderator channel (mm)
280/500 2/2 9.5/10.5/0.57 0.80
Fig. 2. CSR100 Fuel component.
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Fuel rod number/water rod number Fuel assemblies/first pass assemblies/second pass assemblies Core pressure/main steam pressure (MPa) Heat/electric power (MW)/Thermal Efficiency (%) Main coolant flow rate (kg/s) Neutron spectrum Cladding material Average power density (MW/m3) Active core height (m)
flowing downward through the fuel channel of the first pass, 10.8% flowing downward through the water rods of the first pass and 30% flowing downward through the water rods of the second pass. The coolant from the downcomer (23.3% of the total), the fuel channel of the first pass and the water rods mixed together in the lower plenum, and then flows upwards through the fuel channel of second pass. This specific flow scheme is favorable to improving the neutron utilization, reducing the hot channel factor and increasing the coolant outlet temperature. 2.3. Control system The control system of CSR1000 use the same design arrangement of Super LWR (Oka et al., 2010). The actuation conditions of the safety system is summarized by Xi'an Jiaotong University (Wu et al., 2014). The control system is shown in Fig. 4. From Fig. 4, we can see that there are three control systems. They are pressure control system, power control system and steam temperature control system. The turbine control valves and the turbine bypass valve formed the pressure control system. The CRs formed the power control system. The feedwater pumps formed the steam temperature system. 3. Calculation model
224/4 177/57/120 25/24.5 2300/1000/43.5 1190 Thermal Stain steel (310S) 60 4.2
6 X dNðtÞ ðrðtÞ bÞ ¼ NðtÞ þ li Ci ðtÞ þ S dt L i¼1
(1)
dCi ðtÞ b ¼ li Ci ðtÞ þ i NðtÞ dt L
(2)
ð i ¼ 1e6Þ
where，N(t) is the total fission power, b s the total effective delayed neutron fraction, L is the time of each prompt neutron generation, li is the neutron decay constant for the delayed group i, bi is the effective share of the delayed neutron group ii, Ci(t) is the fission power of delayed neutron group i, S is the external neutron source term. 3.2. Governing equations Governing equations of mass, momentum and energy conservations are shown below.
vrðz; tÞ vGðz; tÞ þ ¼0 vt vz
(3)
vðruÞ v ru2 2f vP þ ¼ rg cos q þ ru2 H vt Dh vz vz
(4)
00 00 vfrðz; tÞhðz; tÞg vGðz; tÞhðz; tÞ 1 þ ¼ lf Q ðz; tÞ lf Qw ðz; tÞ vt vz Aw
3.1. Point reactor kinetics model Using reactivity r and prompt neutron generation timeL, we can also write the point reactor kinetics equations as follows.
Fig. 3. Coolant flow pattern.
(5)
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Table 3 Relief valve function. Relief valve function
Relief valve (20% capacity 8units)
Open(MPa)
Close Open(MPa)
Number
26.2 26.4 26.6 26.8
25.2 25.4 25.6 25.8
1 1 3 3
2010). The flow rate ratio between the first meshes of the two flow paths is iteratively changed to satisfy the momentum conservation.
Momkround Momk1 DPtotWR DPtotDC round ¼ Dt lround
Where, l is the length (m), Mom is the momentum per unit volume (kg/m2,s), Dt is the time step spacing (s), superscript K is the time step, subscript DC is the downcomer path, subscript round is the round path, subscriptWR is the water rod path.
Fig. 4. Control system.
vfrðz; tÞhðz; tÞg vGðz; tÞhðz; tÞ 1 00 þ ¼ lw Qw ðz; tÞ vt vz Aw
(7)
4. Code description
(6)
where, t is the time(s), z is the position(m), r is the density (kg/ m3), G is the mass flux (kg/s), u is the flow rate (m/s), g is the acceleration of gravity (m/s2), f is the friction coefficient, fi ¼ 0:0791Re0:25 , Dh is the Equivalent diameter(m), P is the presi sure (MPa), h is the specific enthalpy (J/kg), Aw is the circulation area 00 (m2), lf is the Node height(m), Q'', Qw are the heat release line power density of fuel rods and moderator (W/m). Note that, Equation (5) is for the coolant energy conservation equation. Equation (6) is for the moderator energy conservation equation. The governing equations are discretized using the upwind difference scheme and the explicit scheme. The boundary conditions are the feedwater flow rate, the feedwater temperature, and the turbine inlet flow rate.
4.1. Node division Each flow channel of CSR1000 is simplified to an average channel and a hot channel. Wherein, the power distribution based on the heat factor of the channel is set to 1.26. Core node division is shown in Fig. 5. From Fig. 5, we can see that the average thermal path channel and the hottest channel are divided into 40 axial nodes in order to improve the accuracy of the calculation process. Wherein, all the coolant channels have heat exchange with moderator channel. Axial power distribution in the core is calculated by neutron kinetics, and finally get the normalized axial power distribution. 4.2. Calculation flow
3.3. Heat transfer correlation There are upward and downward flow regions in the CSR1000 flow scheme design. The watts correlation (Watts and Chou, 1982) has considered the difference between the upward and downward flow. Table 2 shows the supercritical water heat transfer correlation and scope. By using Watts correlation, the MCST is higher than which is calculated by OKA-Koshizuka correlation (Oka et al., 2010), Bishop correlation (Bishop et al., 1964) or DittuseBoelter correlation (Yu & Zhu, 2002). Using watts correlation can get the more conservative result. So the SCAC-CSR1000 code use this correlation. 3.4. Pressure drop balance model At the normal operating conditions, the total pressure drop is calculated by adding the four pressure drops, that are friction drop, acceleration drop, buoyancy drop and the orifice drop. The 4.2 m core height total pressure drop of each path is 0.019 MPa (Oka et al.,
The SCAC-CSR1000 code is developed from SCAC code. On the basis of the original code SCAC, the new code adds the safety control modules and the pressure drop balance module. The IAPWS-IF95 water property package (Wagner et al., 2002) is also added into the new code to expand the calculation range of water property. The density feedback and the Doppler feedback are added into the new code. The density coefficient is a function of the coolant density. The Doppler coefficient is a function of the fuel temperature (Oka et al., 2010). Both of them are used for calculation in point reactor kinetic equation. The calculation flow is shown in Fig. 6. From Fig. 6, we can see that the code reads the initial parameters firstly. The results of initial parameters are got by steady-state calculation. For the difference comparing to the widely used code such as RELAP5 and CATHARE, the redistribution of mass flow rate in each channel is achieved by pressure drop balance model. This model can be easily adjusted to every different situation by changing the condition settings. This can be a more flexible solution for the flow redistribution problem. The solving method for
Table 2 Supercritical water heat transfer correlations and scope. Correlations name
Pressure/MPa
Mass flow density/kg m2 s1
Heat flux/MW m2
DittuseBoelter Bishop Swenson Yamagata Xufeng et al. Watts
Subcritical conditions 22.8e27.6 22.8e41.4 22.6e29.4 23e30 25
e 680e3600 542e2150 310e8130 600e1200 160e1060
e 0.31e3.46 0.2e1.8 0.116e0.93 0.1e0.6 0.175e0.44
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Fig. 5. Core node division.
The line power density is a key influence factor for coolant temperature. And it is also an important parameter in Neutronics/ Thermal-hydraulics coupling. The calculated power result is compared with the design power data for the power verification. The power share is the proportion of the sum of each node power in the total power. The calculated result is shown in Fig. 7. As can be seen from Fig. 6, the deviations of line power density, first pass power share and the second pass power share are all less than 0.2%. The calculated result is closed to the design data. The code power calculation is reliable.
linearly reduced to 50%. The transient calculation result is shown in Fig. 10. As can be seen from Fig. 10, the inlet flow rate decreased to 90% 1 s after the pump tripped. Then the reactor start to shut down after 0.5 s delay. The CRs were inserted into the core rapidly. Then the reactor power decreased rapidly. The two passes coolant flow rate are all decreased. Wherein the flow rate of first pass decreased rapidly, while that of the second pass decreased slower. The two passes flow rate respectively reach the minimum value at 5.2 s and 14 s. The 1st pass flowrate eventually maintains above 50%. It is because the 1st pass coolant channel has the lower orifice coefficient. And the pressure drop balance model balances the flowrate in each channel. The decreasing of the coolant flow rate leaded to the increasing of the MCST. But after the two passes MCSTs reached the maximum value at 8 s, the MCSTs decreased after that. It is because the heat production from fuel rod is decreased, but the coolant removed more heat. Because of the change of the flow and the density of the water, the core pressure decreased obviously in the first 5 s. And then it kept stable at 24.2 MPa.
5.2. Coolant temperature verification
6.2. Isolation of main steam line
The coolant temperature is the key parameter for the characteristic of the core. The calculation time is set to 300 s. The result of stead-state coolant temperature is got. Then the result is compared with the same reactor analysis code SCTRAN (Wu et al., 2014). The results are shown in Figs. 8 and 9. As can be seen from Figs. 8 and 9, the result of SCACdCSR1000 is similar to that of SCTRAN. The two codes can meet the inlet temperature at 280 C and the outlet temperature at 500 C. While the highest coolant temperatures of first pass are between 400 and 420 C. The deviation is within 5%. The SCAC-CSR1000 code and the SCTRAN code use different heat transfer correlations and the solving methods. Because the Watt's correlation gets the lower heat transfer correlation, the heat transferred to the coolant become less so the coolant temperature become lower.
The turbine valve started to close at the 0.6 s. The valve opening rapidly reduced to 5% in 1 s. The turbine valve totally closed at the 3 s. The turbine bypass valve opens at the pressure greater than 26.2 MPa. The behavior of the relief valve is shown in Table 3. In the process of turbine valve closing, the pressure of the core increased. So it is required that the turbine bypass relief valve open to decrease the core pressure. The electric power cannot be supplied to the coolant pump after the turbine valve has been closed. After losing external power, the coolant pump can still work 10 s. The coolant pump idle time is 5 s. The transient calculation result is shown in Fig. 11. As can be seen from Fig. 11, the core pressure had a significant increase after the turbine valve started to close. At the 3.5 s, the core pressure exceeded 26 MPa triggered the shutdown signal. The CRs were inserted into the core rapidly. Then the reactor power decreased rapidly. When the pressure exceeded 26.2 MPa, the turbine bypass relief valve opened. With the constant opening and closing of the pressure relief valve, the core pressure continuously changed between 25 MPa and 26.8 MPa. The two passes coolant flow appeared regular pulse changes. Also that was caused by constant opening and closing of the pressure relief valve. Between
governing equation is developed so that it can be used for backflow calculation. In neutron kinetics calculation module, the reactor power and reactivity coefficient are got by six groups of delayed neutrons point reactor model calculation. 5. Code verification 5.1. Power verification
6. The results 6.1. Partial loss of flow At the 0 s time, one of the two coolant pump tripped. The coolant pump idle time is 5 s. In 5 s the primary coolant flow
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2.0
Start
1.8
Line power density First pass power share Second pass power share
1.6
Reads the initial parameters
Deviation(%)
Solving Governing equations
1.4 1.2 1.0 0.8 0.6 0.4
Convergence judgment
0.2
0.18259
0.13589
0.0547
0.0
(1)
(2)
(3)
Fig. 7. Deviation between calculation and design.
Next node calculation
Judgment on pressure drop balance of each channel
The hottest channel calculation
Neutron kinetics calculation
Results output Fig. 8. SCAC-CSR1000 coolant temperature.
End Fig. 6. Calculation flow.
15 and 20 s, the first pass coolant flow occurred a backflow phenomenon. This is because the coolant in the coolant pump is not provided. The coolant flow was redistributed. The second pass MCST appeared downward trend after it had a significant increasing in the beginning 17 s.
6.3. Uncontrolled CR withdrawal The maximum reactivity worth of a CR cluster depends on the loading pattern of the fuel assemblies and the CR pattern. Any limitation of the loading pattern or any interlock of the CR pattern, like the rod worth minimizer of BWRs, is not considered conservatively. The highest reactivity worth of a CR luster under all the considerable loading patterns and CR patterns is estimated as 1.05%
Fig. 9. SCTRAN coolant temperature.
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shutdown signal was triggered because the reactor power rate exceed 120%. The CRs were inserted into the core rapidly. Then the reactor power decreased rapidly. The first pass coolant flow showed a phenomenon of substantial increase after the shutdown signal was triggered. There are three reason cause it. The first reason is that the pressure drop and the coolant density increase. The second reason is that the first pass coolant channel has a small orifice correlation. The third reason is the pump power increase when the pressure is dropped. The inlet flow rate had a sharply drop after a short rising. The two passes MCST and the pressure had a deep decline. According to the relationship of pressure drop between two passes, the first pass coolant pressure had greater drop than that of the second pass. 6.4. Reactor coolant pump seizure
Fig. 10. Partial loss of flow.
At the 0 s, coolant flow from one of the tow RCPs is assumed to suddenly stop. The primary coolant flow rate reduced to 50% instantly. The results is shown in Fig. 13. As can be seen from Fig. 13, different from the partial loss of flow transient, the inlet flowrate shows a different result. It is because the inlet flowrate is decreased to 50% at the 0 s. And after the shutdown signal was triggered, the only running reactor coolant pump losses the power supply at the 15 s. The two passes MCST increased faster than that of the partial loss of flow accident. In the first 6 s, the two passes MCST increased because of the instantaneous reduce of coolant flow rate. The shutdown signal was triggered at 0 s. The core power and pressure decrease rapidly. The inlet flow rate, first pass coolant flow rate and the second pass coolant flow rate were all showed a slowly rising trend. It was caused by slowly pressure drop. The first pass coolant flow rate showed a backflow phenomenon after the 39 s. 6.5. Loss of feedwater heating
Fig. 11. Isolation of main steam line.
dk/k at the normal operating condition. The calculation result is shown in Fig. 12. As can be seen from Fig. 12, the reactivity increased with the uncontrolled CR withdrawal started from the 0 s. The reactor power increased. But the power increasing rate is small due to the reactivity feedbacks from the water density and fuel temperature. The cladding temperature increased by only 17 C because the main coolant flow rate was increased by the control system so as to keep the main steam temperature. At the 80 s time, the
Fig. 12. Uncontrolled CR withdrawal.
Loss of one stage of the feedwater heating will cause a 35 C drop of the feedwater temperature. In the safety analysis, it is conservatively assumed as 55 Cas is done in the safety analysis of Super LWR. The calculation result is shown in Fig. 14. As can be seen from Fig. 14, there is a short decrease of the two passes coolant flow rate in the first 5 s. It is because the density of the two passes coolant increased. During that time, the two passes MCSTs has a rapidly increase. After that, in order to keep 100% of the power, the control system withdrawn the CRs. Meanwhile, in order to maintain the main steam temperature at 500 C, the control system increased the primary coolant flow. The MCST decreased and the reactor power increased during the flow rate
Fig. 13. Reactor coolant pump seizure.
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7. Conclusion
Fig. 14. Loss of feedwater heating.
The SCAC-CSR1000 code is developed from SCAC code (Chen et al., 2013). The steady-state results of power and coolant temperature are got. The results are compared with the same reactor type analysis code SCTRAN. The reliability of the code is verified. After that the five events are selected for the safety analysis by using the new code. The CSR1000 reactor is able to ensure the MCSTs are below the safety criterion of 1260 C. The second pass MCST is higher than that of the first pass. The isolation of main steam line has the least safety margin. This study introduces a new analysis code named SCACCSR1000 and clarifies the characteristics of the CSR1000 during events at supercritical pressure condition. The results of the safety analysis and the sensitivity analysis, especially the increase in the cladding temperature and its duration, are useful for improving the CSR1000 performance. Further safety analysis is necessary to clarify the characteristics of the CSR1000 at all the abnormal transients and accidents. Acknowledgments The research is supported by the Research Fund of the China Key Laboratory of Nuclear Reactor System Design Technology (2014BJ0041), the Research Fund of the China Key Laboratory of nuclear thermal-hydraulic technology (2013B40), the Central Universities special funds basic research and business projects (2014BJ0086) (2014BJ0087). References
Fig. 15. Sensitivity analysis.
increasing time. After the main steam temperature reached 500 C, both the coolant flow rate and the power are decreased. When the core inlet flow rate was reduced to 90%, the shutdown signal was triggered. The CRs were inserted into the core rapidly. Then the reactor power decreased rapidly. While the two passes MCST dropped. 6.6. Sensitivity analysis Oxidation behavior of stainless steel and Ni-alloy is expected to be milder than that of Zircaloy. In this study, the criterion of cladding temperature is set to 1260 C for stainless steel cladding, taken from the criterion for LOCA of U.S.-PWR with stainless steal cladding. The MCSTs of five events are summarized. The order of the events are partial loss of flow, isolation of main steam line, uncontrolled CR withdrawal, reactor coolant pump seizure, loss of feedwater heating. The result is shown Fig. 15. As can be seen from Fig. 15, the MCSTs of all the events did not exceed the 1260 C safety criterion. The second pass MCST is higher than that of the first pass. In accordance with the safety margin, the descending order is uncontrolled CR withdrawal, loss of feedwater heating, partial loss of flow, reactor coolant pump seizure, isolation of main steam line.
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