skip to main content

Title: Neutronics Analyses of the Minimum Original HEU TREAT Core

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to validate the MCNP model of the TREAT reactor with the well-documented measurements which were taken during the start-up and early operation of TREAT. Furthermore, the effect of carbon graphitization was also addressed. The graphitization level was assumed to be 100% (ANL/GTRI/TM-13/4). For this purpose, a set of experiments was chosen to validate the TREAT MCNP model, involving the approach to criticality procedure, in-core neutron flux measurements with foils, and isothermal temperature coefficient and temperature distribution measurements. The results of this study extended the knowledge base for the TREAT MCNP calculations and established the credibility of the MCNP model to be used in the core conversion feasibility analysis.
Authors:
 [1] ;  [1] ;  [1] ;  [1]
  1. Argonne National Lab. (ANL), Argonne, IL (United States)
Publication Date:
OSTI Identifier:
1224961
Report Number(s):
ANL/GTRI/TM--14/13
121427; TRN: US1500840
DOE Contract Number:
AC02-06CH11357
Resource Type:
Technical Report
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)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; HIGHLY ENRICHED URANIUM; TREAT REACTOR; CARBON; NEUTRON FLUX; TEMPERATURE COEFFICIENT; GRAPHITIZATION; FEASIBILITY STUDIES; KNOWLEDGE BASE; TEMPERATURE DISTRIBUTION; CRITICALITY; REACTOR CORES; M CODES; REACTOR KINETICS MCNP