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This work summarizes the radiation transport–based design for a new D2O-moderated ex-core irradiation facility in the Washington State University (WSU) TRIGA reactor for optimization of 135Xe sources used for calibration and quality control testing of Xe gas detection equipment in support of the Comprehensive Test Ban Treaty (CTBT). Three-dimensional (3-D) particle transport analysis characterizing the WSU reactor core using MCNP6.2 (3-D Monte Carlo) and PENTRAN (3-D deterministic parallel SN) form the basis for the computational optimization. Excellent agreement between MCNP6.2 and PENTRAN predictions is observed. A fundamental fuel bundle depletion analysis is applied to enable a more accurate prediction of neutron flux and neutron spectrum distribution, which drives production rates of 135Xe and 133Xe. The results of various model simulations were used to inform recommendations for the final irradiation chamber design, which has been optimized for safe placement in the reactor tank prior to startup and will allow for insertion and rotation of xenon "bean" samples using existing WSU irradiation equipment, while remaining within operational parameters. The irradiation chamber is expected to produce samples that will remain viable for use in CTBT standards applications for durations 70% to 80% longer than samples produced using current procedures. Thus, this design is expected to improve CTBT-related calibrations and performance testing and to support the continued stability of the CTBT monitoring network. [ABSTRACT FROM AUTHOR]

Here we perform the matching of neutron counts in two detector gasses through capture reactions and radiation transport–optimized moderating materials. One of our detectors uses helium-3 (3He) gas and has been widely used as a neutron detection material in proportional detector tube designs. This study examines boron trifluoride (BF3) as a potential gas for neutron detection in place of 3He based on a previously studied "spectrally matched" design derived from deterministic adjoint analyses that closely mimic the spectral response of 3He. The integrated spectral response of each tube, i.e., the count rate, is calculated and measured at various distances from an isotropic neutron source where similar "total sources" are achieved in either detection system. Our results show the integrated spectral response of a dual BF3 tube detector was within 10% of a single 3He tube when exposed to a similar source. GEANT4 Monte Carlo simulations were used to calculate the total source for each detector and showed count rates within 5% of those produced by MCNP, providing a strong confidence in its behavior in the thermal energy regime. We provide results in this study to partially validate the replacement based on the spectrally matched design, which will lead to further validation through the utilization of multiple neutron spectra via simulated and experimental studies. [ABSTRACT FROM AUTHOR]

We present a systematic computational dose rate evaluation for a packaged 1.8-Ci 241AmBe source using both Monte Carlo and deterministic approaches, with some experimental measurements for correlation. The 241AmBe source is stored in an extended 55-gal-drum container. Computational dose rate analysis is performed using MCNP6 (Monte Carlo) and PENTRAN (SN) on the Center for High Performance Computing system at the University of Utah. Limited information is available regarding internal drum shielding construction, and a reverse engineering approach is presented here to estimate the dose rate and compare with measured experimental values. Our analysis shows that a deterministic three-dimensional quadrature (SN) and anisotropic scattering (PN) order of S20P2 is sufficient for dose rate calculations of the 241AmBe source with polyethylene surrounding the source as shielding material. Higher quadrature orders, i.e., at least S70 for neutrons and S40 for photons, are needed in the presence of air due to severe streaming effects, and this is dependent upon the distance between the source and measurement locations. With air surrounding the 241AmBe source, the Monte Carlo method is considered to be better for neutron dose calculations while the SN method is considered better for photon dose calculations. Good agreement from both computational verification and experimental validation are observed for the dose "hot spot" in the extended 55-gal drum. The differences noted between the MCNP6/PENTRAN calculations are within 6% for the neutron dose rate and 30% for the photon dose rate. It is observed that more than 95% of the dose is attributed to neutrons. Detailed studies including a literature data validation, PENTRAN SN convergence study, buildup factor analysis, and dose rates with different shielding materials are presented in the narrative. [ABSTRACT FROM AUTHOR]

AbstractHere we present a new method of irradiating 132Xe capsules with neutrons to produce 133mXe gas standards that are used for radiation detector calibration at radioxenon measurement laboratories in support of the Comprehensive Test Ban Treaty (CTBT). This method is designed to maximize the production of 133mXe compared to 133Xe, both of which are competing products from the 132Xe(n, g) reaction. The 133mXe is produced at a much higher fraction for high-energy neutron absorptions in 132Xe (~50% for fast neutrons versus ~11% for thermal neutrons).We performed “spectral tuning” of the Washington State University (WSU) TRIGA reactor neutron spectrum inside the 132Xe ampules to maximize the number of fast neutrons and minimize the number of thermal neutrons available for 132Xe absorption. Spectral tuning analysis, done with Monte Carlo simulations, provided valuable insights into a future final design for a 132Xe irradiation capsule. With no spectral tuning, the fractional yield of 133mXe in the WSU reactor was ~11.7%. By surrounding the 132Xe capsule with a 0.5-cm-thick layer of tungsten and a 2.83-cm layer of europium (III) oxide and placing it in the reactor’s cadmium rotator tube next to the fuel elements, the fractional yield of 133mXe can be increased to 24.6%, a 111% increase in yield. Thus, by improving the fractional yield of 133mXe through spectral tuning, the CTBT will have better quality gas standards to use for radioxenon detector calibration to assist in the CTBT’s mission. [ABSTRACT FROM AUTHOR]

Here we perform the matching of neutron counts in two detector gasses through capture reactions and radiation transport–optimized moderating materials. One of our detectors uses helium-3 (3He) gas and has been widely used as a neutron detection material in proportional detector tube designs. This study examines boron trifluoride (BF3) as a potential gas for neutron detection in place of 3He based on a previously studied "spectrally matched" design derived from deterministic adjoint analyses that closely mimic the spectral response of 3He. The integrated spectral response of each tube, i.e., the count rate, is calculated and measured at various distances from an isotropic neutron source where similar "total sources" are achieved in either detection system. Our results show the integrated spectral response of a dual BF3 tube detector was within 10% of a single 3He tube when exposed to a similar source. GEANT4 Monte Carlo simulations were used to calculate the total source for each detector and showed count rates within 5% of those produced by MCNP, providing a strong confidence in its behavior in the thermal energy regime. We provide results in this study to partially validate the replacement based on the spectrally matched design, which will lead to further validation through the utilization of multiple neutron spectra via simulated and experimental studies. [ABSTRACT FROM AUTHOR]