Application of MCS Code for Spent Fuel Cask Analysis
- 1. Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)
Description
The spent fuel located inside the cooling pool must be transported out of the site to enable the decommission process. Thus, there is a demand for spent fuel casks capable of keeping spent fuels as well as personnel safe during handling and transportation of the spent nuclear fuel. In this study, criticality of various spent fuel casks, namely KN12, KN18, GBC32, KORAD21 are analyzed together with the flux calculation for KORAD21, using the Monte Carlo code named MCS developed in UNIST. The calculated results are also compared with results from MCNP6 and KENO VI code. In this study, the application of MCS code for spent fuel cask analysis has been presented. The multiplication factor obtained from MCS shown good agreement with results from other Monte Carlo code.
Additional details
Identifiers
- URL
- https://www.kns.org;
Publishing Information
- Publisher
- KNS
- Imprint Place
- Daejeon (Korea, Republic of)
- Imprint Title
- Proceedings of the KNS 2018 Fall Meeting
- Imprint Pagination
- vp.
- Journal Page Range
- [3 p.]
Conference
- Title
- 2018 Fall Meeting of the KNS
- Dates
- 24-26 Oct 2018
- Place
- Yeosu (Korea, Republic of)
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 50060242
- Subject category
- S42: ENGINEERING;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- ERRORS; FUEL ASSEMBLIES; KORI-1 REACTOR; M CODES; MONTE CARLO METHOD; REDUCTION; SHIELDING; SPENT FUEL CASKS; TRANSPORT; USES
- Descriptors DEC
- CALCULATION METHODS; CASKS; CHEMICAL REACTIONS; COMPUTER CODES; CONTAINERS; ENRICHED URANIUM REACTORS; POWER REACTORS; PWR TYPE REACTORS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 7 refs, 5 figs, 2 tabs