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Title: Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SNmore » is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.« less
Authors:
 [1] ;  [1]
  1. Argonne National Lab. (ANL), Argonne, IL (United States)
Publication Date:
OSTI Identifier:
1227442
Report Number(s):
ANL/NE--15/36
123161; TRN: US1500915
DOE Contract Number:
AC02-06CH11357
Resource Type:
Technical Report
Research Org:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; TREAT REACTOR; NUCLEAR FUELS; VALIDATION; COMPUTERIZED SIMULATION; NEUTRON TRANSPORT THEORY; MONTE CARLO METHOD; VERIFICATION; COMPARATIVE EVALUATIONS; CROSS SECTIONS; GEOMETRY; TRANSIENTS; P CODES; REACTOR START-UP; MESH GENERATION