Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask

2018 ◽  
Vol 114 ◽  
pp. 495-509 ◽  
Author(s):  
Jaerim Jang ◽  
Wonkyeong Kim ◽  
Sanggeol Jeong ◽  
Eun Jeong ◽  
Jinsu Park ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
And Storage ◽  
Criticality Safety

scholarly journals Criticality safety analysis of spent fuel pool for TRIGA Puspati Reactor using MCNP5

2019 ◽  
Vol 555 ◽  
pp. 012001
Author(s):  
Nurlaila Syamsul Bahri ◽  
Jasman Zainal ◽  
Khaidzir Hamzah ◽  
Muneer Aziz Mohammad Saleh ◽  
Mohamad Hairie Rabir
Keyword(s):  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Criticality Safety

scholarly journals Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Claubia Pereira ◽  
Jéssica P. Achilles ◽  
Fabiano Cardoso ◽  
Victor F. Castro ◽  
Maria Auxiliadora F. Veloso
Keyword(s):  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Operation Conditions ◽  
Transport Code ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Criticality Safety

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized 

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2013 ◽  
Vol 479-480 ◽  
pp. 543-547
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Wan Yun Li ◽  
Shao Wen Chen ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Injection ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Study Case

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

scholarly journals Thermal Safety Analysis of Spent Fuel Pool in Marine Nuclear Power Platform

2018 ◽  
Vol 189 (4) ◽  
pp. 042003
Author(s):  
Wang Jue ◽  
Hu Chen ◽  
Zhu Meng ◽  
Yin Yu-hao
Keyword(s):  
Nuclear Power ◽  
Safety Analysis ◽  
Thermal Safety ◽  
Spent Fuel ◽  
Spent Fuel Pool

A study of burnup credit in criticality safety analysis for PBR spent fuel pebbles

2019 ◽  
Vol 132 ◽  
pp. 347-356 ◽  
Author(s):  
Shang-Chien Wu ◽  
Der-Sheng Chao ◽  
Jenq-Horng Liang
Keyword(s):  
Safety Analysis ◽  
Spent Fuel ◽  
Criticality Safety

Criticality Safety Assessment of a TRIGA Reactor Spent-Fuel Pool Under Accident Conditions

1997 ◽  
Vol 117 (2) ◽  
pp. 248-254 ◽  
Author(s):  
Bogdan Glumac ◽  
Matjaž Ravnik ◽  
Marjan Logar
Keyword(s):  
Safety Assessment ◽  
Spent Fuel ◽  
Triga Reactor ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
Criticality Safety

Application of TRACE and CFD in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2011 ◽  
Vol 145 ◽  
pp. 78-82 ◽  
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Yung Shin Tseng ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Energy Research ◽  
Important Work ◽  
Spent Fuel Pool

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. After Fukushima NPP event, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of the spent fuel pool for Chinshan NPP which also assumed the cooling system of the spent fuel pool failed. The geometry of the Chinshan NPP spent fuel pool is 12.17 m × 7.87 m × 11.61 m and the initial condition is 60 ¢J / 1.013 × 105 Pa. In general, the NPP safety analysis is performed by the thermal hydraulic codes. The advanced thermal hydraulic code named TRACE for the NPP safety analysis is developing by U.S. NRC. Therefore, the safety analysis of the spent fuel pool for Chinshan NPP is performed by TRACE. Besides, this safety analysis is also performed by CFD. The analysis result of TRACE and CFD are similar. The results show that the uncovered of the fuels occur in 2.7 days and the metal-water reaction of the fuels occur in 3.5 days after the cooling system failed.

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