Nuclear characterization of a general-purpose instrumentation and materials testing location in TREAT

Abstract

The Transient Reactor Test facility (TREAT) was constructed in the late 1950s, provided thousands of transient irradiations before being placed in standby in 1994, and resumed operations in 2017 in order to reclaim its crucial role in nuclear-heated safety research. The latter half of TREAT’s historic operation was best known for integral-scale testing of fuel specimens under postulated reactor plant accident conditions, while TREAT’s earlier history included extensive simpler phenomena identification tests that elucidated fundamental behaviors and paved the way for these integral-scale tests. Advances in modern computational capabilities and a resurgence of interest in novel reactor technology have created an opportunity for emphasizing modernized science-based and separate effects test capabilities once again at TREAT. An innovative approach to this type of testing has been developed to leverage minor radioactivity built-in during brief TREAT irradiations in low activation hardware to facilitate handling for materials and instrumentation testing. This capability, termed the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation vehicle system, will be used for inaugural fueled experiments in TREAT’s modern era as well as novel approaches to study materials undergoing neutron irradiation and instrumentation development and qualifications. This paper describes a comprehensive nuclear characterization, obtained via computational modeling, for the Materials and Instrumentation Modular Irradiation Capability (MIMIC) module in the Broad Use Specimen Transient Experiment Rig (BUSTER) test position of the MARCH system. Though results are directly applicable to MIMIC and BUSTER, they also provide general quantification of the nuclear performance of the reactor and potential test materials, crucial for evaluating potential experiment design and response in TREAT. The neutron and photon flux environments were calculated via MCNP with ENDF/B-VII.1 nuclear data. Wire heating rates and atomic displacement calculations were also performed; sample calculations using historic TREAT operational data were provided to demonstrate example conditions. These calculations were performed to provide steady-state baseline reference values typical in both half- and full-slotted TREAT core configurations, enabling design scoping analysis prior to development of more specific design testing needs and transient testing experimentation. Due to the limited availability of historic data to validate current calculations, validation experiments are planned once the BUSTER test vehicle has been constructed.