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Development for Improvement Methodology of Radiation Shielding Evaluation Efficiency about PWR SNF Interim Storage Facility
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 Title & Authors
Development for Improvement Methodology of Radiation Shielding Evaluation Efficiency about PWR SNF Interim Storage Facility
Kim, Taeman; Seo, Myungwhan; Cho, Chunhyung; Cha, Gilyong; Kim, Soonyoung;
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 Abstract
For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.
 Keywords
Spent nuclear fuel;Dry interim storage facility;Shielding analysis;MCNP;External dose estimate;
 Language
Korean
 Cited by
 References
1.
Korea Radioactive Waste Agency. Technology development for implementation of spent nuclear fuel transportation & storage system. KORAD/TR-2014-01. 2014.

2.
U.S. Nuclear Regulatory Commission. Criteria for radioactive materials in effluents and direct radiation from an ISFSI or MRS. 10CFR72.104. 2014.

3.
Gauld IC. ORIGEN-ARP: Automatic rapid processing for spent fuel depletion, decay, and source term analysis(Ver.6). Oak Ridge National Laboratory. ORNL/TM-2005/39. 2005.

4.
Sweezy JE. MCNP-A General Monte Carlo NParticle transport code version 5. Los Alamos National Laboratory. LA-UR-03-1987. 2003.

5.
International Commission on Raidological Protection. Conversion coefficients for use in radiological protection against external radiation. ICRP Publication 74 Vol.26(3-4). 1996.

6.
Cha GY, Kim SY, Noh KY, Lee WG, Kim TM, Baeg CY. A spent fuel storage cask modeling for improving the calculation efficiency about off-site dose Monte Carlo. Korean Association for Radiation Protection 2012 Autumn Meeting. 2012 November 29-30. Jeju, Korea.

7.
Noh KY, Kim SY, Cha GY, Kim ES, Yim HK, Kim TM, Baeg CY. A study on the 2 step calculation method for reducing computational time in directly radiation dose at interim storage facilities. Korean Association for Radiation Protection 2014 Spring Meeting. 2014 April 24-25. Busan, Korea.