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Peters, Curtis D.

Publications and source records attributed to Peters, Curtis D..

14 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

A neutron fluence map of the Los Alamos National Laboratory Godiva IV critical assembly

A neutron fluence map and a total ionizing dose map of the Los Alamos National Laboratory Godiva IV fast burst critical assembly was generated using passive reactor dosimetry, comprised of sulfur pellets and thermoluminescent dosimeters. Godiva IV is an unmoderated, fast burst, critical assembly constructed of approximately 65 kg of highly enriched uranium fuel alloyed with 1.5 % molybdenum for strength. The mapping was performed during a single 75.6 ºC temperature rise burst operation, with the top and sides of the cylindrical Godiva-IV Top Hat covered in passive dosimetry. Dosimetry was placed in a symmetric pattern around the Top Hat, with higher concentrations near the control rods and burst rod. A specific portion of the lower quadrant of the burst rod was mapped to confirm a testing region where the neutron fluence varied by no more than ± 5%. The results will be used to assess the neutron, gamma, and total ionizing dose environment in three-dimensional space around the assembly for higher fidelity experiment placement, active dosimetry positioning, and radiation field characterization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Radiation characterization summary of the NETL beam port 1/5 free-field environment at the 128-inch core centerline adjacent location

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the University of Texas at Austin Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor for the beam port (BP) 1/5 free-field environment at the 128-inch location adjacent to the core centerline has been accomplished. NETL is being explored as an auxiliary neutron test facility for the Sandia National Laboratories radiation effects sciences research and development campaigns. The NETL reactor is a TRIGA Mark-II pulse and steady-state, above-ground pool-type reactor. NETL is intended as a university research reactor typically used to perform irradiation experiments for students and customers, radioisotope production, as well as a training reactor. Initial criticality of the NETL TRIGA reactor was achieved on March 12, 1992, making it one of the newest test reactor facilities in the US. The neutron energy spectra, uncertainties, and covariance matrices are presented as well as a neutron fluence map of the experiment area of the cavity. For an unmoderated condition, the neutron fluence at the center of BP 1/5, at the adjacent core axial centerline, is about 8.2×10 12 n/cm 2 per MJ of reactor energy. About 67% of the neutron fluence is below 1 keV and 22% above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is roughly 1.6×10 12 n/cm 2 per MJ of reactor energy.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Radiation characterization summary for the WSMR fast burst reactor environment at the 6-inch location

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields for the White Sands Missile Range (WSMR) Fast Burst Reactor, also known as molybdenum-alloy Godiva (Molly-G) has been assessed at the 6-inch irradiation location. The neutron energy spectra, uncertainties, and common radiation metrics are presented. Code-dependent recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. The Molly-G core was designed and configured similarly to Godiva II, as an unreflected, unmoderated, cylindrical annulus of uranium-molybdenum-alloy fuel with molybdenum loading of 10%. At the 6-inch position, the axial fluence maximum is about 2.4×10 13 n/cm 2 per MJ of reactor energy; about 0.1% of the neutron fluence is below 1 keV and 96% is above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is estimated at 2.2×10 13 n/cm 2 per MJ of reactor energy. The prompt gamma-ray dose is roughly 2.5E+03 rad(Si) per MJ and the delayed gamma-ray dose is about 1.3E+03 rad(Si) per MJ.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Radiation Characterization Summary: NETL Beam Port 1/5 Free-Field Environment at the 128-inch Core Centerline Adjacent (NETL-FF-BP1/5-128-cca)

This document presents the facility-recommended characterization of the neutron, prompt gamma ray, and delayed gamma ray radiation fields in the University of Texas at Austin Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor for the beam port 1/5 free-field environment at the 128-inch location adjacent to the core centerline. The designation for this environment is NETL-FF-BP1/5-128-cca. The neutron, prompt gamma ray, and delayed gamma ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma ray fluence profiles within the experiment area of the cavity. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Radiation Characterization Summary: WSMR Fast Burst Reactor Environment at the 6-Inch and 24-Inch Irradiation Locations

This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields for the White Sands Missile Range (WSMR) Fast Burst Reactor, also known as molybdenum-alloy Godiva (Molly-G), at the 6-inch and the 24-inch irradiation locations. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented. Code dependent recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Transient Thermal Analysis of Calorimeters Used in Characterization of the ACRR Radiation Environments

Silicon calorimeters have been used for active radiation dosimetry in the central cavity of the Annular Core Research Reactor (ACRR) for over a decade. Recently, there has been interest in using other materials for calorimetry to accurately measure the prompt gamma-ray energy deposition in the mixed neutron and gamma-ray environment. The calorimeters used in the ACRR use a thermocouple (TC) to measure the change in temperature of specific materials in the radiation environment. The temperature change is related to the instantaneous dose received by the material in a pulse-transient operation. SOLIDWORKS Simulation and ANSYS Mechanical were used to model the calorimeter and analyze the thermal behavior under pulse-transient conditions. This report compares the results from modeling to experimental results for selected calorimeter materials and radiation environments. These materials include bismuth, tin, zirconium, and silicon. Calorimeters assembled with each material were irradiated in the ACRR central cavity in the free- field, LB44, CdPoly, and PLG radiation environments. The neutronics code Monte-Carlo N- Particle (MCNP) was used to calculate the neutron and gamma-ray response of the calorimeter materials at the experimental locations in the central cavity. Different response tallies were used and found to give different results for the gamma-ray energy deposition. It was determined that performing the neutron/gamma-ray/electron transport in MCNP using the *F8 electron tally gave the overall best agreement with the experimental results. The *F8 tally, however, is much more computationally intensive than the neutron/gamma-ray transport calculations. Also, this report contains parametric analyses that examine the ways to improve the current design of the calorimeters. One finding from the parametric analysis was that the TC should be placed closer to the outer radius of the disks to obtain a measurement closer to the maximum temperature of the disk. Also, the parametric analysis showed that the most dominant mechanism of heat loss in the calorimeters is conduction through the alumina posts. In future designs, the conduction should be minimized to reduce the effect of heat loss on the measurements.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗