A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly

Citation
Ws. Kim et al., A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly, ANN NUC ENG, 29(3), 2002, pp. 303-321
Citations number
21
Language
INGLESE
art.tipo
Article
Categorie Soggetti
Nuclear Emgineering
Journal title
ANNALS OF NUCLEAR ENERGY
ISSN journal
0306-4549 → ACNP
Volume
29
Issue
3
Year of publication
2002
Pages
303 - 321
Database
ISI
SICI code
0306-4549(200202)29:3<303:ASACMF>2.0.ZU;2-E
Abstract
In sodium cooled liquid metal reactors design limits are imposed on the max imum temperatures of the cladding and fuel pins. Thus an accurate predictio n of the core coolant/fuel temperature distribution is essential to LMR cor e thermal hydraulic design. The detailed subchannel thermal hydraulic analy sis code MATRA-LMR is being developed for LMFBR core design and analysis ba sed on COBRA-IV-I and MATRA. The major modifications and improvements imple mented in MATRA-LMR are as follows: sodium property calculation subprogram, sodium coolant heat transfer correlations, and most recent pressure drop c orrelations. To assess the development status of this code, benchmark calcu lations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR were comp ared to the measurements and to the SABRE4 and SLTHEN code calculation resu lts, respectively. Finally, the major technical results of the conceptual d esign for the KALIMER U-10%Zr binary alloy fueled core have been compared w ith the calculations of the MATRA-LMR, SABRE4 and SLTHEN codes. (C) 2001 Pu blished by Elsevier Science Ltd.