Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)
Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm2 and temporary heat flux limit of 600 W/cm2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.
- Research Organization:
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation
- DOE Contract Number:
- AC02-06CH11357
- OSTI ID:
- 1327832
- Report Number(s):
- ANL/RTR/TM-16/3; 130471; TRN: US1700327
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
SLIGHTLY ENRICHED URANIUM
HIGHLY ENRICHED URANIUM
BR-2 REACTOR
THERMAL HYDRAULICS
CONVERSION
NUCLEAR FUELS
PRESSURE RANGE MEGA PA
REACTOR CORES
SECONDARY COOLANT CIRCUITS
LOSSES
COOLANTS
LOSS OF COOLANT
SAFETY ANALYSIS
COMPUTERIZED SIMULATION
TEMPERATURE DEPENDENCE
STEADY-STATE CONDITIONS
FLOW RATE
HEAT FLUX