skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident

Abstract

The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWtmore » initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)« less

Authors:
; ; ; ; ; ; ;  [1];  [2];  [3]
  1. Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, Nuclear Science and Technology Development Center, 101 Section 2, Kuang Fu Rd., HsinChu, Taiwan (China)
  2. Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C., 1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan (China)
  3. Department of Mechanical Engineering, Chung yuan Christian University, 200, Chungpei Rd., Chungli District, Taoyuan City, Taiwan (China)
Publication Date:
OSTI Identifier:
22992035
Resource Type:
Journal Article
Journal Name:
Transactions of the American Nuclear Society
Additional Journal Information:
Journal Volume: 114; Journal Issue: 1; Conference: Annual Meeting of the American Nuclear Society, New Orleans, LA (United States), 12-16 Jun 2016; Other Information: Country of input: France; 8 refs.; Available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 United States; Journal ID: ISSN 0003-018X
Country of Publication:
United States
Language:
English
Subject:
97 MATHEMATICAL METHODS AND COMPUTING; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; CLADDING; CONCRETES; FUEL STORAGE POOLS; FUKUSHIMA DAIICHI NUCLEAR POWER STATION; LBLOCA; NUCLEAR POWER PLANTS; PWR TYPE REACTORS; REACTOR COOLING SYSTEMS; RELIEF VALVES; SAFETY ANALYSIS; SEVERE ACCIDENTS; SIMULATION; SPENT FUELS; STATION BLACKOUT; STEAM GENERATORS

Citation Formats

Wang, Jong-Rong, Tseng, Yung-Shin, Wang, Ting-Yi, Chiang, Yu, Hsu, Wen-Sheng, Chen, Hsiung-Chih, Chen, Shao-Wen, Shih, Chunkuan, Lin, Hao-Tzu, and Teng, Jyh-Tong. The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident. United States: N. p., 2016. Web.
Wang, Jong-Rong, Tseng, Yung-Shin, Wang, Ting-Yi, Chiang, Yu, Hsu, Wen-Sheng, Chen, Hsiung-Chih, Chen, Shao-Wen, Shih, Chunkuan, Lin, Hao-Tzu, & Teng, Jyh-Tong. The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident. United States.
Wang, Jong-Rong, Tseng, Yung-Shin, Wang, Ting-Yi, Chiang, Yu, Hsu, Wen-Sheng, Chen, Hsiung-Chih, Chen, Shao-Wen, Shih, Chunkuan, Lin, Hao-Tzu, and Teng, Jyh-Tong. 2016. "The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident". United States.
@article{osti_22992035,
title = {The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident},
author = {Wang, Jong-Rong and Tseng, Yung-Shin and Wang, Ting-Yi and Chiang, Yu and Hsu, Wen-Sheng and Chen, Hsiung-Chih and Chen, Shao-Wen and Shih, Chunkuan and Lin, Hao-Tzu and Teng, Jyh-Tong},
abstractNote = {The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)},
doi = {},
url = {https://www.osti.gov/biblio/22992035}, journal = {Transactions of the American Nuclear Society},
issn = {0003-018X},
number = 1,
volume = 114,
place = {United States},
year = {Wed Jun 15 00:00:00 EDT 2016},
month = {Wed Jun 15 00:00:00 EDT 2016}
}