• Title, Summary, Keyword: Shipping Cask

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Neutron Dose Rate Analysis of PWR Spent Fuel Transport Cask Using Monte Carlo Method

  • Do, Mahnsuck;Kim, Jong-Kyung;Yoon, Jeong-Hyoun
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.847-852
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    • 1995
  • A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.

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A Study on the Side Drop Impact of a Nuclear Spent Fuel Shipping Cask (사용후 핵연료 수송용기의 수평낙하충격에 관한 연구)

  • Chung, Sung-Hwan;Lee, Young-Shin
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.21 no.3
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    • pp.457-469
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    • 1997
  • A nuclear spent fuel shipping cask is required by IAEA and domestic regulations to withstand a 9m free drop condition. In this paper, the structural analysis under the 9m side drop condition was performed to understand the dynamic impact behavior and to evaluate the safety of the cask for 7 PWR nuclear spent fuel assemblies. The analysis result was compared with the measured value of the 9m side drop test for the 1/3 scaled-down model and the accuracy of the 3D analysis was confirmed. Analysis in accordance with the diameter of impact limiters for the proto-type cask were performed. Through the analysis, the impact behaviors due to the side drop and the effects dependent on the diameter of impact limiters were grasped. Maximum stress intensities on each part of the cask were respectively calculated by using the stress evaluation program and the structural safety of the cask was finally evaluated in accordance with the regulations.

Radiation Shield Analysis for Spent Fuel Shipping Cask (핵연료 수송용기의 방사선 차폐해석)

  • Cho, Kun-Woo;Kim, Hee-Won;Kwon, Seog-Kun;Kwak, Eun-Ho;Moon, Philip-S.
    • Journal of Radiation Protection and Research
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    • v.10 no.2
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    • pp.148-154
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    • 1985
  • Radiation shield design for a shipping cask, KSC-1, was evaluated to verify that the cask can be used in the transportation of a spent fuel assembly discharged from KNU 5 & 6. Radiation source term of the spent fuel assembly was calculated with the computer program ORIGEN-79, QAD-CG, ANISN-KA and DOT 3.5 codes Were used in the shielding calculations and the nuclear cross section data needed was extracted from the DLC-23/CASK library. It is concluded that KSC-1 shipping cask satisfies the requirements specified in the relevant regulations under normal conditions of transport and under accident conditions in transport.

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Design Study of A Spent Fuel Shipping Cask for Korea Nuclear Unit-1 (고리 1호기의 기사용 핵연료 집합체 수송용기 설계에 관한 연구)

  • Moo Han Kim;Chang Sun Kang
    • Nuclear Engineering and Technology
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    • v.14 no.4
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    • pp.196-203
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    • 1982
  • To transport the spent fuel assemblies of Korea Nuclear Unit 1, which is a Westinghouse type two loop pressurized water reactor, it has been found that steel is the most appropriate material for the design of a shipping cask in comparison with lead and depleted uranium. The proposed shipping cask will transport nine fuel assemblies at the same time and is well within the weight limit of transportation by unrestricted rail car. The cask requires 33cm thick steel shield and 27cm thick water region to satisfy the 3 feet apart dose rate limit set forth in 10 CFR 71, and 1.27cm thick steel boron fuel basket to hold the fuel elements inside the cask and control the effective multiplication factor. As a safety analysis, the fuel cladding and centerline temperatures were calculated under the accident condition of complete loss of water coolant, and it was found that the temperature was much lower than the limit of the melting point. k$_{eff}$ was calculated with fresh fuel assemblies, which was found to be well lower than 0.95. For shielding computation, the multipurpose Monte Carlo code MORSE-CG and one dimensional discrete ordinates transport code ANISN were used, and the Monte Carlo codes KENO and MORSE-CG were used for criticality calculation. The radiation source terms were calculated using ORIGEN-79.9.

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Criticality Analysis of KSC-4 Spent Fuel Shipping Cask (KSC-4 수송용기의 핵임계도 분석)

  • Choi, B.I.;Shin, H.S.;Park, C.M.;Ro, S.G.
    • Journal of Radiation Protection and Research
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    • v.14 no.1
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    • pp.56-65
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    • 1989
  • The nuclear criticality of the KSC-4 shipping cask which can load four assemblies of PWR spent fuel was analyzed using KENO-IV computer code and 19-group nuclear cross section set generated from 218-group neutron cross section library(DLC-43/CSRL) using AMPX system. In accordance with 10CFR71, the analysis was performed for fresh fuel assemblies, instead of the spent fuels, under both norml transportation and hypothetical accident conditions. The calculated maximum multiplication factors(Keff) of the KSC-4 cask were 0.85289 and 0.94185 for the normal transportation and hypothetical accident conditions, respectively. The highest Keff of the KSC-4 cask is within the subcritical limit prescribed in l0CFR71 and accordingly the KSC-4 cask is safely designed in terms of nulear criticality.

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Shielding Design of Shipping Cask for 4 PWR Spent Fuel Assemblies (PWR집합체 4개 장전용 수송용기의 차폐설계)

  • Kang, Hee-Yung;Yoon, Jung-Hyoun;Seo, Ki-Seog;Ro, Seung-Gy;Park, Byung-Il
    • Nuclear Engineering and Technology
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    • v.20 no.1
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    • pp.65-70
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    • 1988
  • A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.

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Criticality Analyses of Spent Fuel Shipping Cask (핵연료(核燃料) 수송용기(輸送容器)에 대(對)한 핵림계분석(核臨界分析))

  • Min, Duck-Kee;Ro, Seung-Gy;Kwack, Eun-Ho
    • Journal of Radiation Protection and Research
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    • v.9 no.2
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    • pp.97-102
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    • 1984
  • Criticality analyses of the KSC-1(Korean Shipping Cask-1) spent fuel shipping cask have been performed with the help of KENO-IV Monte Carlo computer code and 19-group CSLIB 19 cross section set which was generated from AMPX modular system. The analyses followed a benchmark calculation which has been made regard to the B & W CX-10 criticality facility in order to validate the Monte Carlo code cross section set described above. The KSC-1 shipping cask seems to be safe in the criticality point of view for the transport of one PWR spent fuel assembly under the normal conditions as well as the hypothetical accident conditions.

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Analysis of the criticality of the shipping cask(KSC-7) (KSC-7 사용후핵연료 수송용기 핵임계해석)

  • Yoon, Jung-Hyun;Choi, Jong-Rak;Kwak, Eun-Ho;Lee, Heung-Young;Chung, Sung-Whan
    • Journal of Radiation Protection and Research
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    • v.18 no.2
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    • pp.47-59
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    • 1993
  • The criticality of the shipping cask(KSC-7) for transportion of 7PWR spent fuel assemblies has been calculated and analysised on the basis of neutron transport theory. For criticality analysis, effects of the rod pitches, the fixed neutron absorbers(borated sus+boral) were considered. The effective multiplication factor has been calculated by KENO-Va, Mote Carlo method computer code, with the HANSEN-ROACH 16 group cross section set, which was made for personal computer system. The criticality for the KSC-7 cask was calculated in terms of the fresh fuel which was conservative for the aspects of nuclear critility. From the results of criticality analysis, the calculated Keff is proved to be lower than subcritical limit during normal transportation and under hypothetical accident condition. The maximum calculated criticalities of the KSC-7 were lower the safety criticality limit 1.0 recommended by US 10CFR71 both under normal and hypothetical accident condition. Also, to verify the KSC-7 criticality calculation results by using KENO-Va, it was carried out benchmark calculation with experimental data of B & W(Bobcock and Wilcox) company. From the 3s series of calculation of the KSC-7 cask and benchmark calculation, the cask was safely designed in nuclear criticality, respectively.

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