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Roberts, Jeremy A.

Publications and source records attributed to Roberts, Jeremy A..

17 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Extension of the discrete generalized multigroup method using SPH factors

The discrete generalized multigroup (DGM) method provides a way to treat the energy dependence of neutron transport similarly to the standard multigroup approximation. However, DGM uses an orthogonal basis to retain the energy dependence in higher-order terms. Using correction factors similar to traditional Superhomogénéisation (SPH) factors, the DGM method may be extended to produce cross sections that are homogenized over both space and energy. Additionally, since some fine-group energy dependence is retained, the resulting homogenized cross sections are more problem-independent than cross sections homogenized by SPH factors alone. In particular, a 44-group set of cross sections is collapsed to approximately a 1% error in the pincell fission densities for a test problem using three DOF per coarse energy group.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

A high-efficiency, low-Ĉerenkov Micro-Layered Fast-Neutron Detector for the $\mathrm{TREAT}$ hodoscope

Here this paper presents an enabling technology for high-efficiency fast-neutron detection for the Transient Reactor Test (TREAT) Facility hodoscope. The hodoscope is currently outfitted with 25.4-mm long Hornyak buttons for fast-neutron detection. However, they require improvements, as they suffer from low detection efficiency, Ĉerenkov radiation contaminating the signal, and non-linearity of detector response at large transients. To address these issues, a layered ZnS:Ag device has been constructed. The layered design provides various advantages: it incorporates greater scintillation volume, Ĉerenkov background is significantly reduced by eliminating the need for additional light guides, and the detection process is simplified by achieving gamma-ray rejection using only pulse-height discrimination. The result is a greater number of fast neutron counts with reduced background noise contaminating the signal. On being irradiated with a 252Cf source (whose neutron energy spectrum closely mimics a reactor fission spectrum), a 25.4-mm long layered device was found to exhibit an intrinsic fast-neutron detection efficiency of 8.1% for neutrons with an average energy of 2 MeV, while a 43-mm long device allows for a greater number of fast neutron to be detected, increasing the efficiency to 9.1%. Further lengthening of the device up to 20 cm has been numerically predicted to exhibit even higher efficiencies. The detectors can be mass-produced easily and inexpensively.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

A Geant4 evaluation of the Hornyak button and two candidate detectors for the TREAT hodoscope

The performance of traditional Hornyak buttons and two proposed variants for fast-neutron hodoscope applications was evaluated using Geant4. The Hornyak button is a ZnS(Ag)-based device previously deployed at the Idaho National Laboratory's TRansient REActor Test Facility (better known as TREAT) for monitoring fast neutrons emitted during pulsing of fissile fuel samples. Past use of these devices relied on pulse-shape discrimination to reduce the significant levels of background Cherenkov radiation. Proposed are two simple designs that reduce the overall light guide mass (here, polymethyl methacrylate or PMMA), employ silicon photomultipliers (SiPMs), and can be operated using pulse-height discrimination alone to eliminate background noise to acceptable levels. Geant4 was first used to model a traditional Hornyak button, and for assumed, hodoscope-like conditions, an intrinsic efficiency of 0.35% for mono-directional fission neutrons was predicted. The predicted efficiency is in reasonably good agreement with experimental data from the literature and, hence, served to validate the physics models and approximations employed. Geant4 models were then developed to optimize the materials and geometries of two alternatives to the Hornyak button, one based on a homogeneous mixture of ZnS(Ag) and PMMA, and one based on alternating layers of ZnS(Ag) and PMMA oriented perpendicular to the incident neutron beam. For the same radiation environment, optimized, 5-cm long (along the beam path) devices of the homogeneous and layered designs were predicted to have efficiencies of approximately 1.3% and 3.3%, respectively. For longer devices, i.e., lengths larger than 25 cm, these efficiencies were shown to peak at approximately 2.2% and 5.9%, respectively. Furthermore, both designs were shown to discriminate Cherenkov noise intrinsically by using an appropriate pulse-height discriminator level, i.e., pulse-shape discrimination is not needed for these devices.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Preliminary Analysis of Damage to MPFDs Caused by Reactor Pulses

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. The TREAT reactor is designed to produce large, transient fluxes that replicate the effects of a wide variety of accident conditions on an irradiated fuel sample. Peak powers generated during short pulses approach 20 GW with a neutron flux of approximately 10{sup 17} n cm{sup -2} s{sup -1}. As part of the restart effort, new vehicle designs are being explored for sample irradiation, including the Static Environment Rodlet Transient Test Apparatus (SERTTA) and variants. As part of the SERTTA designs, micro-pocket fission detectors (MPFDs) are being explored for localized flux measurements before, during, and after transient operations. These detectors are also being considered as part of future, core-wide instrumentation for the TREAT reactor. However, the high fluxes associated with the pulses in the TREAT core present certain thermodynamic concerns given their intensity and the amount of energy they will release within the MPFDs within such a short amount of time. Assuming each fission that occurs in a MPFD results in a fission fragment depositing 85 MeV of energy in the substrate on average and each MPFD has 0.5 μg of natural uranium deposited on it, a 100 ms long pulse in the TREAT reactor at 1017 n cm{sup -2} s{sup -1} flux will produce approximately 4.6 x 10{sup 6} fissions in an MPFD. This pulse deposits 6.3 x 10{sup -5} J in the substrate over a 100 ms duration. That amount of energy is sufficient to melt roughly 33% of the Pt under the UO{sub 2} deposition or a circular spot with a diameter of approximately 860 μm. Of course, the above estimate makes a number of simplifying assumptions and ignores heat transfer, but does serve to illustrate the point that this rapid heating effect could be detrimental to the integrity and life of these in-core MPFDs. To study how these layers are affected by large transient neutron fluxes, five samples were prepared, pre-analyzed, and then subjected to a pulse in the central thimble of the Kansas State University TRIGA Mark II reactor to a peak power of 1.58 GW. After being allowed to decay to safer radiation levels, the sample substrates and coatings were inspected using scanning electron microscopy (SEM) to identify any changes to their surface morphology. Indeed, certain changes were found and are discussed in greater detail in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Reconstructing Solar Particle Event Spectra from Absorbed Dose Measurements

The space radiation environment is unlike anything encountered on Earth. It is dominated by heavy charged particles consisting of highly charged nuclei with atomic numbers ranging from 1 to 94 with a broad spectrum of kinetic energies. The three primary sources of radiation in space are trapped belts, galactic cosmic rays (GCRs), and solar particle events (SPEs). The trapped belts are only of concern near bodies with significant magnetic fields and GCR flux spectra are relatively well-predicted using models that account for solar modulation of the local interstellar spectrum. SPEs, which consist primarily of protons, are difficult to predict in terms of occurrence, duration, intensity, and energy spectrum. SPE intensity and energy spectrum are critical for understanding radiation exposure-related health effects for astronauts in space. The National Aeronautics and Space Administration (NASA) relies on data from the Geostationary Operational Environmental Satellite series to characterize intensities and energy spectra of SPEs near Earth. However, these data are not applicable for spacecraft in other regions of the solar system. To understand the radiation environment for astronauts embarking on travel beyond Earth orbit, NASA is developing the Hybrid Electronic Radiation Assessor (HERA), a distributed dosimetry system that employs a solid-state silicon sensor coupled with the Timepix readout chip. The pixelated nature of the detector permits detailed analyses of charged particle tracks which can then be used to characterize the fluence spectrum at the location of the detector. Despite the unprecedented level of detail regarding the radiation environment provided by these detectors, knowledge of the fluence spectrum at all locations within the vehicle, specifically inside of astronauts, is desired to fully characterize the associated risk of harmful health effects. Radiation transport of the external radiation environment, considering the shielding provided by the vehicle, other objects, and even astronauts, is performed to obtain this information. Thus, there is clearly a need to define the external radiation environment using measurements performed within the vehicle. The number of detectors that may be flown as a part of HERA is limited by logistics, including data, power, and mass restrictions. The current configuration for crewed flight consists of six detectors at various locations with in the Orion Multi-Purpose Crew Vehicle (MPCV). It follows that an external proton fluence spectrum can be characterized with at most six energy bins with traditional matrix inversion techniques. The purpose of the present study is to determine whether it is possible to reconstruct SPE spectra with greater resolution from absorbed dose measurements and knowledge of the response of individual detectors using linear programming.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Manual reactor power oscillations were observed in real-time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A study of reactivity biases and their dependence on spatial discretization in depleted TRIGA fuel

The impact of spatial discretization on reactivity biases in TRIGA fuel models was analyzed here. In particular, unit-cell analyses in 2-D and 3-D were performed using Serpent to understand how spatial discretization affects the accuracy with which the effects of material evolution and temperature feedback are resolved. For temperature-dependent cases, a simple, single-channel model was employed. Analysis of 2-D models showed that essentially no radial discretization resolution is needed to eliminate biases (i.e., to reduce biases to the level of stochastic uncertainties) due to material evolution but that more than eight, equal-area, radial regions are needed to resolve temperature-feedback effects. Analysis of 3-D models showed that at least seven, equal-volume, axial regions may be required to resolve material evolution with temperature feedback leading to insignificant additional bias. Because of memory constraints, a full-core model with radially- and axially-resolved fuel elements may be impractical, especially for production-level analyses. Consequently, an “effective Doppler temperature” was determined empirically as a function of the radially-averaged temperature and may be used for future, full-core analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

MCNP6 simulated performance of Micro-Pocket Fission Detectors (MPFDs) in the Transient REActor Test (TREAT) Facility

Micro-Pocket Fission Detectors (MPFDs) are under development for in-core neutron flux measurements at the Transient REActor Test facility (TREAT) and in other experiments at Idaho National Laboratory (INL). The sensitivity of MPFDs to the energy dependent neutron flux at TREAT has been determined for 0.0300-μm thick active material coatings of 242 Pu, 232 Th, natural uranium, and 93% enriched 235 U. Self-shielding effects in the active material of the MPFD was also confirmed to be negligible. Lastly, fission fragment energy deposition was found to be in conformance with previously reported results.

36 MATERIALS SCIENCE↗
DOE OSTI2016

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

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)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Impact of Spatial Discretization on Reactivity Biases in Depleted TRIGA Fuel

Ongoing work at Kansas State University (KSU) and the University of Wisconsin involves the determination of fuel compositions in TRIGA reactors. Modeling of depletion requires a number of approximations to be made, which often includes selection of an appropriate spatial discretization for tracking material evolution in time. Incorporation of temperature feedback also requires such spatial zoning. Any approximation made leads to computational biases, i.e., differences between computed and accurate values of some parameter of interest. The objective of this study is to understand the potential magnitude of biases introduced by the selected spatial discretization. Specifically, biases in the reactivity of a square-pitch unit cell representative of a typical TRIGA fuel element were studied as a function of burnup for different radial and axial discretizations. Temperature effects were also studied by introducing a fixed, radial temperature profile. The results indicate a very fine radial discretization (> 10 radial regions) may be needed to eliminate biases (of statistical significance) from temperature feedback effects, while an axial discretization of {approx_equal}7 regions may be sufficient to eliminate biases due to evolution of the fuel composition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Initial Application of the Reactivity Method to Analyze KSU TRIGA Fuel

The Kansas State University (KSU) TRIGA Mk II reactor was refueled over 40 years ago with fuel having stainless steel cladding. Much of that fuel had significant burnup, and little work has been done since then to quantify the core composition. The present work aims to provide data in support of efforts to understand this composition. Currently available data consist of 26 critical configurations, including the location of fuel elements, the position of control rods, and various temperatures that have been documented in operational log books. Additional measurements will be made using nondestructive techniques to quantify the burnup of fuel elements irradiated in the KSU reactor. Gamma spectroscopy is a well-established method for determining the burnup of spent fuel, and plans are in place at KSU to perform such measurements. Another option explored here is use of the reactivity method, which relates reactivity worths to burnup using linear or other correlations. Reactivity measurements of fuel elements with wide ranges of burnup were conducted to help identify the present compositions of those fuel elements and to verify the historical inventories of those fuel elements. Significant biases were observed between measured and modeled reactivities for several elements in the KSU reactor. It is expected future measurement of burnups using gamma spectroscopy will help to reduce these biases and produce a best-estimate model of the core composition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Updating a PWR Simulator in Python

A nuclear reactor is a complicated system that is comprised of many systems and components. In order to fully model a reactor, several approximations and correlations must be used for a real-time simulation. In 1984 as part of a doctoral dissertation at Massachusetts Institute of Technology (MIT), a FORTRAN 77 program was developed to simulate a four-loop, Westinghouse pressurized water reactor (PWR). The program was later called PRISM and extended to include a comprehensive set of features accessible in a DOS-based user interface. This present work seeks to update the original PRISM program for cross platform deployment using the Python language. The goal of this present work is to update and improve a teaching tool for both students and reactor operators. Originally, PRISM was created as part of a PhD project in FORTRAN 77, but the code no longer works readily on modern operating systems. Python was chosen to rewrite the program because of the ease of use of the Python modules including Kivy as well as the portability of the program to various operating systems. This project aims to develop desktop and mobile versions of the software, allowing the user flexibility in choice of platform. This simulator is scheduled to be deployed in time for a live demonstration at the 2016 Annual ANS meeting. The project aims to produce a reactor simulator that will showcase the dynamics of a PWR system in real-time under various conditions. Some potential scenarios include startup conditions, scram operations, or failure conditions such as a break in a coolant leg. The user will be able to induce various failure scenarios and the program will simulate how the reactor will behave. The simulation will include a suite of monitoring functions that will allow automatic systems to attempt to adapt to the changing conditions, e.g., control rod movement. This program will provide an educational resource for anyone learning about nuclear reactor kinetics or dynamics of a reactor system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗