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Title: Validation of the MULCH-II code for thermal-hydraulic safety analysis of the MIT research reactor conversion to LEU.

Abstract

An in-house thermal hydraulics code was developed for the steady-state and loss of primary flow analysis of the MIT Research Reactor (MITR). This code is designated as MULti-CHannel-II or MULCH-II. The MULCH-II code is being used for the MITR LEU conversion design study. Features of the MULCH-II code include a multi-channel analysis, the capability to model the transition from forced to natural convection during a loss of primary flow transient, and the ability to calculate safety limits and limiting safety system settings for licensing applications. This paper describes the validation of the code against PLTEMP/ANL 3.0 for steady-state analysis, and against RELAP5-3D for loss of primary coolant transient analysis. Coolant temperature measurements obtained from loss of primary flow transients as part of the MITR-II startup testing were also used for validating this code. The agreement between MULCH-II and the other computer codes is satisfactory.

Authors:
; ; ; ; ;
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
971169
Report Number(s):
ANL/NE/CP-60723
TRN: US1001050
DOE Contract Number:  
DE-AC02-06CH11357
Resource Type:
Conference
Resource Relation:
Conference: 2007 RERTR International Meeting; Sep. 23, 2007 - Sep. 27, 2007; Prague, Czech Republic
Country of Publication:
United States
Language:
ENGLISH
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; COMPUTER CODES; COOLANTS; DESIGN; LICENSING; NATURAL CONVECTION; RESEARCH REACTORS; SAFETY; SAFETY ANALYSIS; TEMPERATURE MEASUREMENT; TESTING; THERMAL HYDRAULICS; TRANSIENTS; VALIDATION

Citation Formats

Ko, Y. C., Hu, L. W., Olson, A. P., Dunn, F. E., Nuclear Engineering Division, and MIT. Validation of the MULCH-II code for thermal-hydraulic safety analysis of the MIT research reactor conversion to LEU.. United States: N. p., 2007. Web.
Ko, Y. C., Hu, L. W., Olson, A. P., Dunn, F. E., Nuclear Engineering Division, & MIT. Validation of the MULCH-II code for thermal-hydraulic safety analysis of the MIT research reactor conversion to LEU.. United States.
Ko, Y. C., Hu, L. W., Olson, A. P., Dunn, F. E., Nuclear Engineering Division, and MIT. Mon . "Validation of the MULCH-II code for thermal-hydraulic safety analysis of the MIT research reactor conversion to LEU.". United States. doi:.
@article{osti_971169,
title = {Validation of the MULCH-II code for thermal-hydraulic safety analysis of the MIT research reactor conversion to LEU.},
author = {Ko, Y. C. and Hu, L. W. and Olson, A. P. and Dunn, F. E. and Nuclear Engineering Division and MIT},
abstractNote = {An in-house thermal hydraulics code was developed for the steady-state and loss of primary flow analysis of the MIT Research Reactor (MITR). This code is designated as MULti-CHannel-II or MULCH-II. The MULCH-II code is being used for the MITR LEU conversion design study. Features of the MULCH-II code include a multi-channel analysis, the capability to model the transition from forced to natural convection during a loss of primary flow transient, and the ability to calculate safety limits and limiting safety system settings for licensing applications. This paper describes the validation of the code against PLTEMP/ANL 3.0 for steady-state analysis, and against RELAP5-3D for loss of primary coolant transient analysis. Coolant temperature measurements obtained from loss of primary flow transients as part of the MITR-II startup testing were also used for validating this code. The agreement between MULCH-II and the other computer codes is satisfactory.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 2007},
month = {Mon Jan 01 00:00:00 EST 2007}
}

Conference:
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