Evaluation of MSNDs for Fast-Neutron Detection and the TREAT Hodoscope

Journal Article · Fri Jul 01 00:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:23042763

Ghosh, Priyarshini; Fu, Wenkai; Fronk, Ryan; McGregor, Douglas S.; Roberts, Jeremy A. ^[1]

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1. Department of Mechanical and Nuclear Engineering Kansas State University (United States)

The need for next-generation, accident-tolerant fuels has generated significant interest in the restart of domestic fuels testing at the Transient Reactor Test (TREAT) facility. A key feature of the TREAT reactor core is an open, horizontal channel that allows line-of-sight viewing of test specimens using various instrumentation during steady-state and transient operation. This instrumentation includes the hodoscope, which consists of over 300 steel-collimated channels coupled to Hornyak buttons (and other detectors) for detection of fast neutrons emitted from fission of the test specimen. Posterior analysis can be used to reconstruct fuel motion and other properties from the collimated channel signals. A large-scale effort is underway to develop next generation instrumentation for transient testing at TREAT (and, potentially, other facilities). Work at KSU specifically aims to provide improved detection systems for hodoscope applications. It was initially proposed to adapt KSU-developed Microstructured Semiconductor Neutron Detectors (MSNDs) for use as fast-neutron detectors. The current generation of MSNDs is designed for thermal-neutron detection based on {sup 6}Li and can achieve intrinsic efficiencies nearing 30% owing to the large (n, t) cross section of {sup 6}Li and clever optimization of the geometry. In order to evaluate MSNDs for fast-neutron detection, two approaches were considered: (1) use of a hydrogenous material to thermalize neutrons to exploit the intrinsic efficiencies of current MSNDs, and (2) substitution of a fast-sensitive reactant for {sup 6}Li. The former approach is attractive because it makes use of proven, off-the-shelf technology and leads to reasonable time resolution and efficiency. Unfortunately, use of thermalization leads to prohibitive degradation of spatial resolution. The second approach will require further experimental efforts, but scoping analyses suggest {sup 235}U- or {sup 237}Th-loaded MSNDs may achieve sufficient efficiencies. The rest of this summary examines these issues in more detail. (authors)

• OSTI ID: 23042763
• Journal Information: Transactions of the American Nuclear Society, Vol. 115; Conference: 2016 ANS Winter Meeting and Nuclear Technology Expo, Las Vegas, NV (United States), 6-10 Nov 2016; Other Information: Country of input: France; 8 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US); ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ACCIDENT-TOLERANT NUCLEAR FUELS
CROSS SECTIONS
FAST NEUTRONS
FISSION
HODOSCOPES
LITHIUM 6
MICROSTRUCTURE
NEUTRON DETECTION
NEUTRON DETECTORS
REACTOR CORES
SEMICONDUCTOR MATERIALS
SPATIAL RESOLUTION
STEELS
TESTING
THERMAL NEUTRONS
THORIUM 237
TIME RESOLUTION
TRANSIENTS
TREAT REACTOR
URANIUM 235