AN ASSESSMENT OF THE RADIATION DOSE RATE DUE TO AN OCCURRENCE OF THE DEFECT ON THE SPENT NUCLEAR FUEL ROD

  • Lee, Sang-Hun (Department of Energy & Environment System Engineering, Dongguk University) ;
  • Moon, Joo-Hyun (Department of Energy & Environment System Engineering, Dongguk University)
  • Published : 2009.09.30

Abstract

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

References

  1. Oak Ridge National Laboratory, SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, ORNL/TM-2005/39, Version5.1, Vols. I-III, 2006
  2. Gauld IC et al., ORIGEN-ARP: Automatic Rapid Processing for Spent Fuel Depletion, Decay, And SourceTerm Analysis, ORNL/TM-2005/39, Radiation Safety Information Computational Center at Oak Ridge National Laboratory, 2006
  3. Rhoades WA and CHILDS RL. The TORT Three-Dimensional Discrete Ordinates Neutron/PhotonTransport Code, ORNL-6268, Oak Ridge NationalLaboratory, 1987
  4. White JE et al. BUGLE-96: Coupled 47 Neutron, 20Gamma-Ray Group Cross Section Library Derived from ENDF/B-IV for LWR Shielding and Pressure Vessel Dosimetry Applications, Radiation Safety Information Computational Center Data Library Collection DLC-185,Oak Ridge National Laboratory, 1996
  5. Rhoades WA. The GIP Program for Preparation of Group-Organized Cross Section Libraries, Oak Ridge National Laboratory, April 1978
  6. International Commission on Radiological Protection.Conversion Coefficients for use in Radiological Protection against External Radiation, ICRP Publication 74, London UK;Elservier, 1996
  7. Engle Jr. WW. A User's Manual for ANISN: A One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering, K-1693, Oak Ridge Gaseous Diffusion Plant, 1967
  8. Oak Ridge National Laboratory. ANISN: A One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering, CCC-254, 1994