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in creator/contributor: "Roth, P. A."

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Assessment of RELAP5/MOD3 Version 7 based on the BETHSY Test 6. 2 TC

This document provides a discussion of the BETHSY test 6.2 TC which was conducted to investigate thermal hydraulic phenomena during a 5% cold leg SBLOCA and to provide high quality data for advanced thermal-hydraulic code assessment. BETHSY test 6.2 TC was analyzed using RELAP5/MOD3 version 7o.

Date: January 1, 1992

Creator: Choi, C. J.; Roth, P. A. & Schultz, R. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 … more

Date: July 1, 1992

Creator: Roth, P. A.; Schultz, R. R. & Choi, C. J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Assessment of RELAP5/MOD3 Version 7 based on the BETHSY Test 6.2 TC

Date: August 1, 1992

Creator: Choi, C. J.; Roth, P. A. & Schultz, R. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d`Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d`Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 … more

Date: July 1, 1992

Creator: Roth, P. A.; Schultz, R. R. & Choi, C. J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, tha… more

Date: November 1, 2009

Creator: Chang, G.S.; Roth, P. A. & Lillo, M. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Applications of the 3-D Deterministic Transport Attila{reg_sign} for Core Safety Analysis

An LDRD (Laboratory Directed Research and Development) project is ongoing at the Idaho National Engineering and Environmental Laboratory (INEEL) for applying the three-dimensional multi-group deterministic neutron transport code (Attila{reg_sign}) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the model development, capabilities of Attila, generation of the cross-section libraries, and comparisons to an ATR MCNP model and future.

Date: October 6, 2004

Creator: Lucas, D.S.; Gougar, D.; Roth, P.A.; Wareing, T.; Failla, G.; McGhee, J. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department