POSCA: A computer code for fission product plateout and circulating coolant activities within the primary circuit of a high temperature gas-cooled reactor
- 1. Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
Description
Numerical prediction of fission product plateout and circulating coolant activities under normal operating conditions is crucial in the design of a high temperature gas-cooled reactor (HTGR). The results are used for the maintenance and repair of the components as well as the safety analysis regarding early source terms under loss of coolant accident scenarios. In this work, a new computer code named POSCA (Plate-Out Surface and Circulating Activities) was developed based on a one-dimensional model to evaluate fission product plateout and circulating coolant activities within the primary circuit of a HTGR. The verification and validation of study for the POSCA code was done using available analytical results and two in-pile experiments (i.e., OGL-1 and VAMPYR-1). The results of the POSCA calculations show that POSCA is able to simulate plateout and circulating coolant activities in a HTGR with fast computation and reasonable accuracy
Additional details
Publishing Information
- Journal Title
- Nuclear Engineering and Technology
- Journal Volume
- 52
- Journal Issue
- 9
- Series
- 22 refs, 18 figs, 7 tabs
- Journal Page Range
- p. 1974-1982
- ISSN
- 1738-5733
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 52113446
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- COMPUTER CODES; COOLANTS; FISSION PRODUCTS; HTGR TYPE REACTORS; LOSS OF COOLANT; MAINTENANCE; ONE-DIMENSIONAL CALCULATIONS; P CODES; PLATES; PRIMARY COOLANT CIRCUITS; REPAIR; SOURCE TERMS; SURFACES; VALIDATION; VERIFICATION
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; COOLING SYSTEMS; ENERGY SYSTEMS; GAS COOLED REACTORS; GRAPHITE MODERATED REACTORS; ISOTOPES; MATERIALS; RADIOACTIVE MATERIALS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; TESTING