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연구자들이 자신의 연구와 전문성을 널리 알리고, 새로운 협력의 기회를 만들 수 있는 네트워킹 공간이에요.
논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2008.1
- 수록면
- 319 - 326 (8page)
이용수
초록· 키워드
오류제보하기
To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the
Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell
padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an
easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials,
polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon
shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code
system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask
with the two candidates were estimated within the limit (2 mSv/hr).
목차
등록된 정보가 없습니다.
참고문헌 (10)
참고문헌 신청
D.B. Pelowitz / 2005 / MCNPX User’s Manual / LA-CP-05- 0369, Los Alamos National Laboratory
E.E. Lewis / 1984 / Computational Methods of Neutron Transport / John Wiley & Sons
I.G. Gauld / 2004 / ORIGEN-ARP: Automatic Rapid Processing for Spent Fuel Depletion, Decay, and Source Term Analysis
J. Spanier / 1969 / Monte Carlo Principles and Neutron Transport Problems / Addison-Wesley Publishing Company
J.J. Laidler / 1997 / Development of Pyroprocessing Technology / Prog. in Nucl. Energy 31 : 131 ~ 140
E.E. Lewis / 1984 / Computational Methods of Neutron Transport / John Wiley & Sons
I.G. Gauld / 2004 / ORIGEN-ARP: Automatic Rapid Processing for Spent Fuel Depletion, Decay, and Source Term Analysis
J. Spanier / 1969 / Monte Carlo Principles and Neutron Transport Problems / Addison-Wesley Publishing Company
J.J. Laidler / 1997 / Development of Pyroprocessing Technology / Prog. in Nucl. Energy 31 : 131 ~ 140
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