by Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, National Technical Information Service, [distributor in Washington, DC, Springfield, VA .
Written in English
|Statement||prepared by P. Brodie, P.C. Hall|
|Series||International agreement report -- NUREG/IA-0065|
|Contributions||Hall, P. C, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research|
|The Physical Object|
|Pagination||v, 19 p.|
|Number of Pages||19|
Analysis of semiscale test S-LH-2 using RELAP5/MOD2 (OCoLC) Online version: Brodie, P. Analysis of semiscale test S-LH-2 using RELAP5/MOD2 (OCoLC) Material Type: Government publication, National government publication, Internet resource: Document Type: Book, Internet Resource: All Authors / Contributors. Analysis of Semiscale Test S-LH-2 using RELAP5/MOD2 P. Brodie and P.C. Hall, Analytical Investigation Section The RELAP5/MOD2 code is being used by National Power Nuclear Technology-Division for calculating Small Break Loss of Coolant Accidents (SBLOCA) and pressurised transient sequences for the Sizewell 'B' PWR. Analysis of semiscale test S-LH-2 using RELAP5/MOD2 Technical Report Brodie, P. ; Hall, P.C. The RELAP5/MOD2 code is being used by National Power Nuclear Technology Division for calculating Small Break Loss of Coolant Accidents (SBLOCA) and pressurized transient sequences for the Sizewell " B. Topics: 11M - Material degradation, corrosion, fracture mechanics, 10P - Nuclear reactor technology, 13L - Safety engineering, Analysis of semiscale test S-LH-2 using RELAP5/MOD2 [ Nuclear power coolant faults diagnostics].
An Analysis of Semiscale Mod–2C S–FS–1 Steam Line Break Test Using RELAP5/MOD2: NUREG/IA Analysis of Semiscale Test S–LH–1 Using RELAP5/MOD2: NUREG/IA Analysis of Semiscale Test S–LH–2 Using RELAP5/MOD2: NUREG/IA RELAP5/MOD2 Analysis of LOFT Experiment L9–4: NUREG/IA The Semiscale Mod-2A facility is modeled, as a base case, by using single core channel model. The base case calculation is executed, the result is compared with the experiment data and code predictability on the important thermal-hydraulic phenomena is discussed. Arne et al. () presented the assessment of RELAP5/MOD2 computer code against the Natural Circulation Test Data from Yong-Gwang Unit 2. The results of the RELAP5/MOD2 computer code simulation for the Natural Circulation Test in Yong-Gwang Unit 2 were analyzed and compared with the plant operation by: model (use for estimation of ability), Likert scale (measures human attitude) are the examples of such scales in Psychometrics used widely in the social science & educational research [3,4,5]. Likert scale was devised in order to measure ‘attitude’ in a scientifically accepted and validated manner in [6,7]. An attitude can be defined.
- 3 - SEP 1 of the Semiscale S-LH-1 experiment6 and (4) a break spectrum a~alysis of the~ RESAR-3S plant using both TRAC-PFl/MODl and RELAP5/ previous results have been previously transmitted to your staff. Comprehensive analysis with RELAP5/MOD2 is performed to predict the transient thermal-hydraulic responses of the experiment. Test S-IB-3 is a %, communicative cold leg break LOCA experiment using Semiscale Mod-2A facility in , for the principal objective to provide reference data for comparison of Semiscale test results to LOBI facility B-R1M test results. The present report describes a RELAP5/MOD2 analysis of the small LOCA test S-LH-2 which was performed on the Semiscale Mod-2C Facility. S-LH-2 simulated a . An integral loop test and MARS code analysis for a DVI line break LOCA in the APR The transient thermal–hydraulic phenomena of a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in pressurized water reactor, APR, were investigated.