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Title: Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Abstract

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm 2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleatemore » boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).« less

Authors:
 [1];  [1];  [1];  [2];  [2];  [2];  [2]
  1. Argonne National Lab. (ANL), Argonne, IL (United States)
  2. Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation (NA-20)
OSTI Identifier:
1330567
Report Number(s):
ANL/GTRI/TM-14/8 Rev. 1
130671; TRN: US1700450
DOE Contract Number:
AC02-06CH11357
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; SLIGHTLY ENRICHED URANIUM; BERYLLIUM; CRITICAL HEAT FLUX; BR-2 REACTOR; THERMAL HYDRAULICS; NUCLEATE BOILING; STEADY-STATE CONDITIONS; REACTOR CORES; INSTABILITY

Citation Formats

Licht, J. R., Bergeron, A., Dionne, B., Van den Branden, G., Kalcheva, S, Sikik, E, and Koonen, E. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU. United States: N. p., 2016. Web. doi:10.2172/1330567.
Licht, J. R., Bergeron, A., Dionne, B., Van den Branden, G., Kalcheva, S, Sikik, E, & Koonen, E. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU. United States. doi:10.2172/1330567.
Licht, J. R., Bergeron, A., Dionne, B., Van den Branden, G., Kalcheva, S, Sikik, E, and Koonen, E. 2016. "Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU". United States. doi:10.2172/1330567. https://www.osti.gov/servlets/purl/1330567.
@article{osti_1330567,
title = {Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU},
author = {Licht, J. R. and Bergeron, A. and Dionne, B. and Van den Branden, G. and Kalcheva, S and Sikik, E and Koonen, E},
abstractNote = {BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).},
doi = {10.2172/1330567},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

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  • BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux ofmore » 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.« less
  • The NATCON code provides a capability for the analysis of the thermal-hydraulics of plate-type research reactors cooled by natural convection. The code includes steady-state estimates of what are often considered safety margins. These estimates may include power peaking factors, hot channel factor components, and laminar flow entrance effects. A listing of the code and a description of the input are included, along with some sample results. The current version of the code has only properties data for light water. 4 refs., 3 figs., 11 tabs.
  • This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity.more » The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.« less
  • The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less