Direct measurement of the oxygen vacancies produced in calcium tungstate by fast reactor neutrons.
Oxygen vacancies caused by fast neutrons in calcium tungstate single crystals measured by electron spin resonance
Kikuchi, C.↗
SEARCH · Engineering Papers
Results for “fast neutron”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Fast Hyperspectral Neutron Tomography
Hyperspectral neutron computed tomography is a tomographic imaging technique in which thousands of wavelength-specific neutron radiographs are measured for each tomographic view. In conventional hyperspectral reconstruction, data from each neutron wavelength bin are reconstructed separately, which is extremely time-consuming. These reconstructions often suffer from poor quality due to low signal-to-noise ratios. Consequently, material decomposition based on these reconstructions tends to produce inaccurate estimates of the material spectra and erroneous volumetric material separation. In this paper, we present two novel algorithms for processing hyperspectral neutron data: fast hyperspectral reconstruction and fast material decomposition. Both algorithms rely on a subspace decomposition procedure that transforms hyperspectral views into low-dimensional projection views within an intermediate subspace, where tomographic reconstruction is performed. The use of subspace decomposition dramatically reduces reconstruction time while reducing both noise and reconstruction artifacts. We apply our algorithms to both simulated and measured neutron data and demonstrate that they reduce computation and improve the quality of the results relative to conventional methods.
Chowdhury, Mohammad Samin Nur [Purdue University]↗
Crystal growth, scintillation properties & fast neutron-gamma discrimination of cubic halide perovskite CsCaCl 3 :(Eu 2+ , Tl + )
The vast variety of nuclear security applications require radiation detection materials tailored to their operational needs. A scintillator’s properties are strongly influenced by the choice of luminescent dopant, which facilitates customization to different applications. In this work, transparent Ø12 mm single crystals of undoped CsCaCl 3 , CsCaCl 3 :1% Eu, CsCaCl 3 :1% Tl, and CsCaCl 3 :1% Eu, 1% Tl were grown via the Vertical Bridgman method. Their scintillation properties and fast neutron-gamma discrimination capabilities were investigated. Undoped CsCaCl 3 had a light yield of 2,500 ph/MeV, which is the highest reported to date for this CVL material. The incorporation of Eu 2+ or Tl + into CsCaCl 3 as luminescence centers resulted in significantly higher light yields of ∼16,000 ph/MeV and energy resolutions of ∼8% at 662 keV. Compared to the single dopant counterparts, CsCaCl 3 :Eu, Tl had significantly suppressed afterglow; however, this came at the cost of reduced light yield. Among the materials tested, only CsCaCl 3 :Tl showed effective fast neutron and gamma discrimination capabilities, achieving a Figure of Merit of 3.2 between gamma-rays and fast neutron captures that produce protons and 1.6 between gamma-rays and fast neutron captures that produce alpha particles.
crystal growth↗
Methods for determining burnup in enriched uranium-235 fuel irradiated by fast reactor neutrons
Discussing monitors for determining burnup rate of enriched uranium 235 fuel in fast neutron environment
Richardson, P. J.↗
MECHANISTIC MULTIPHYSICS MODELING OF CLADDING RUPTURE IN NUCLEAR FUEL RODS DURING LOSS-OF-COOLANT ACCIDENT CONDITIONS
The Loss of Coolant Accident (LOCA) is a design basis accident that is included as part of the safety analysis of nuclear power plants. As the nuclear industry desires to increase the cycle length and discharge burnup of existing nuclear power plants they must demonstrate safe operation during a LOCA on high burnup fuel. During a LOCA transient on high burnup fuel rods, the rods may undergo a process known as fuel fragmentation, relocation, and dispersal (FFRD). To permit dispersal, the cladding encapsulating the fuel must rupture with an opening size large enough to allow the fragmented fuel particles to release. Current licensing tools used by industry and the United States Nuclear Regulatory Commission are limited in geometric fidelity and materials that can be analyzed. These simulation tools generally employ a quasi-two-dimensional (1.5D or Layered1D) or 2D-RZ axisymmetric geometric representations exclusively. While a valid approach under some instances, there are times when important physics have an asymmetric behavior in the fuel rod. Examples include fuel fragmentation, thermal-hydraulic boundary conditions, and cladding rupture, all of which are important for LOCA analysis. As industry pursues burnup extensions it must be demonstrated that fuel dispersal can be mitigated or eliminated. To do this, an understanding of the rupture opening after cladding failure is required. This work presents the development of a model for predicting the size and location of the rupture opening in failed fuel rods during LOCA conditions using advanced modeling and simulation tools. In order to supply the rupture model with appropriate boundary conditions, improvements to fuel fragmentation, axial relocation and oxidation modeling were required. First, the eXtended Finite Element Method (XFEM) is used to mechanistically predict the number of fuel fragments that form due to material strength randomization, criteria for strength randomization, mesh density, power ramping rates and irradiation effects. These predictions with associated uncertainty were compared to empirical correlations developed for UO2 verifying that they can be used with increased confidence in subsequent axial relocation analyses. Secondly, a new first-of-its-kind Layered2D computational framework was developed that provides the ability to apply azimuthally varying boundary conditions while providing discrete layers to track fuel movement during the LOCA. An existing fuel axial relocation model developed for Layered1D was extended to work within the Layered2D framework. A large sensitivity study was performed on the initial version of the model to identify modeling parameters of particular importance, with the emissivity used for radiation after blowdown being the primary source of uncertainty. Then, a simplistic approach to incorporate mechanical degradation of the cladding due to oxidation was also developed to investigate the impact of reduced cladding thickness on predictions of the time to failure of cladding tubes. It was found that the cladding will typically rupture prior to a reduction in thickness that is sufficient to impact the rupture behavior. A model was then developed for predicting cladding rupture that transfers the cladding surface temperatures, rod internal and external pressures, fast neutron fluence, and fast neutron flux from a more detailed Layered1D, Layered2D, or 2D-RZ analysis to a 3D cladding only analysis. Comparisons of the rupture model to a few experiments indicate reasonable predictions. The rupture model was then applied to two accident tolerant fuel concepts (FeCrAl and Cr-coated Zircaloy) where it predicted that both ATF concepts would have smaller rupture openings and delayed rupture times than the standard Zircaloy-4 cladding material under identical loading conditions.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of HELIOS/BIPR/PARCS/MCNP6 Computation Route for WWER RPV Neutron Fluence Analysis and Validation Against Ex-Vessel Detector Measurement Data
Here, a computational route was developed for precise calculation of fast neutron fluence on a WWER-type reactor pressure vessel (RPV). The method is based on the transfer of neutronics data from HELIOS-2 lattice calculations and nodal diffusion neutronics data (power, density, and temperature) from BIPR7.1 and PARCS 3.36/PATHS core calculations into a three-dimensional (pinwise axially distributed) fixed neutron source for modeling of transport of fast neutrons from the reactor core to the outer surface of the RPV using MCNP6.2. Validation of the proposed computational method was carried out based on comparative analysis of MCNP6.2-predicted and neutron dosimetry–measured reaction rates [ 54 Fe(n,p) 54 Мn, 93 Nb(n,n') 93 mNb, and 58 Ni(n,р) 58 Со] on the outer surface of the Armenian Nuclear Power Plant (ANPP) Unit 2 RPV. Validation revealed that the MCNP6.2-predicted fast neutron fluence results are very sensitive to the ENDF-B neutron data. Particularly, MCNP6.2 with ENDF/B-VIII.0 significantly underpredicts (20% to 30%) fast neutron fluence while using ENDF/B-VII.1 data overpredicts it. Adding revised beta-released evaluations of 54 Fe, 56 Fe, 57 Fe, and 16 O from the International Nuclear Data Evaluation Network (INDEN) to ENDF/B-VIII.0 allows one to obtain reasonable agreement with measurement results for all types of measured reaction rates.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Evaluation of a positron-emission-tomography-based SiPM readout for compact segmented neutron imagers
Gamma-ray emission from special nuclear material (SNM) is relatively easy to shield from detection using modest amounts of high-Z material. In contrast, fast-neutrons are much more penetrating and can escape relatively thick high-Z shielding without losing significant energy. Furthermore, fast neutrons provide a clear and unambiguous signature of the presence of SNM with few competing natural background sources. The challenge of detecting fast neutrons is twofold. First, the neutron flux from SNM are only a fraction of the corresponding gamma-ray flux. Second, fast neutrons can be difficult to differentiate from gamma rays. The ability to discriminate gamma rays from neutrons, combined with a neutron imaging capability, can in some situations enable the localization of SNM neutron sources distinct from background. With the developments of pulse-shape-sensitive plastic scintillators that offer excellent gamma-ray/neutron discrimination, and arrays of silicon photomultipliers combined with highly scalable and fast positron-emission-tomography (PET) multi-channel readout systems, field-deployable neutron imagers suitable for SNM detection might now be within reach. In this paper, we present a characterization of the performance of a recently available commercial PET-scanner readout, including its sensitivity to pulse-shape differences between fast neutrons and gamma rays, energy and timing resolution. Here, we find that, while the pulse-shape discrimination is achievable with stilbene, further improvement of the readout is required to achieve it with the PSD-capable available plastic scintillators.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Open cycle gas core nuclear rockets
The open cycle gas core engine is a nuclear propulsion device. Propulsion is provided by hot hydrogen which is heated directly by thermal radiation from the nuclear fuel. Critical mass is sustained in the uranium plasma in the center. It has typically 30 to 50 kg of fuel. It is a thermal reactor in the sense that fissions are caused by absorption of thermal neutrons. The fast neutrons go out to an external moderator/reflector material and, by collision, slow down to thermal energy levels, and then come back in and cause fission. The hydrogen propellant is stored in a tank. The advantage of the concept is very high specific impulse because you can take the plasma to any temperature desired by increasing the fission level by withdrawing or turning control rods or control drums.
Ragsdale, Robert↗
Differential neutron energy spectra measured on spacecraft in low Earth orbit
Two methods for measuring neutrons in the range from thermal energies to dozens of MeV were used. In the first method, alpha-particles emitted from the 6Li(n,alpha)T reaction are detected with the help of plastic nuclear track detectors, yielding results on thermal and resonance neutrons. Also, fission foils are used to detect fast neutrons. In the second method, fast neutrons are recorded by nuclear photographic emulsions (NPE). The results of measurements on board various satellites are presented. The neutron flux density does not appear to correlate clearly with orbital parameters. Up to 50% of neutrons are due to albedo neutrons from the atmosphere while the fluxes inside the satellites are 15-20% higher than those on the outside. Estimates show that the neutron contribution to the total equivalent radiation dose reaches 20-30%.
Cosmos 1887 Project↗
Atmospheric neutrons
Contributions to fast neutron measurements in the atmosphere are outlined. The results of a calculation to determine the production, distribution and final disappearance of atmospheric neutrons over the entire spectrum are presented. An attempt is made to answer questions that relate to processes such as neutron escape from the atmosphere and C-14 production. In addition, since variations of secondary neutrons can be related to variations in the primary radiation, comment on the modulation of both radiation components is made.
Korff, S. A.↗
Measurement of muon-induced high-energy neutrons from rock in an underground Gd-doped water detector
In this paper, we present a measurement of the rate of correlated neutron captures in the WATCHBOY detector, deployed at a depth of approximately 390 meters water equivalent (m.w.e.) in the Kimballton Underground Research Facility (KURF). WATCHBOY consists of a cylindrical 2 ton water target doped with 0.1% gadolinium, surrounded by a 40 ton undoped water hermetic shield. We present a comparison of our results with the expected rate of correlated neutron captures arising from high-energy neutrons incident on the outside of the WATCHBOY shield, predicted by a hybrid FLUKA/GEANT4-based simulation. The incident neutron energy distribution used in the simulation was measured by a fast neutron spectrometer, the 1.8-ton Multiplicity and Recoil Spectrometer (MARS) detector, at the same depth. We nd that the measured detection rate of two correlated neutrons is consistent with that predicted by simulation. The result lends additional confidence in the detection technique used by MARS, and therefore in the MARS spectra as measured at three different depths. Confirmation of the fast neutron flux and spectrum is important as it helps validate the scaling models used to predict the fast neutron fluxes at different overburdens.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
A fast neutron-gamma detector for space research.
Test results on anticoincidence system for charged particle shield in fast neutron-gamma detector for space research applications
Lockwood, J. A.↗
Accelerated Fuel Qualification of Fast Modular Reactor Fuel in a Thermal Reactor: Modeling and Simulation Paired with Irradiation Testing
The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO 2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. Finally, this approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.
Advanced test reactor (ATR)↗
Seasonal CO2 Observations on North and South of Mars as Seen by HEND (Mars Odyssey) and MOLA (MGS)
The first year of neutron mapping measurements from the Mars Odyssey spacecraft are presented based on observations from the High Energy Neutron Detector (HEND). The HEND instrument is a part of GRS suite responsible for registration of epithermal and fast neutrons originating in the Mars subsurface layer. The gamma ray and neutron spectrometers measure the scattering of fast neutrons from the Martian surface, which is caused by bombardment of primary cosmic rays and is strongly sensitive to the presence of hydrogen atoms. Even several percent subsurface hydrogen significantly depresses the flux of epithermal and fast neutrons. The recent Mars Odyssey observations detected a considerable amount of hydrogen, almost certainly corresponding to water ice, in the shallow near surface of the southern and northern hemispheres of Mars.
Litvak, M. L.↗
The Fuel Motion Monitoring System at TREAT - Current Status and Future Plans
An important component of the United States Nuclear Fuel Safety Transient Testing Program, the Fuel Motion Monitoring System (FMMS) at Idaho National Laboratory's Transient Reactor Test Facility (TREAT) is fast-neutron hodoscope capable of imaging the location, movement, and relocation of nuclear fuel experiments under simulated transient accident conditions. The FMMS was refurbished in parallel with the TREAT restart program starting in 2014, restoring 96-channels of fast-neutron detection. Since returning to operation in 2017 the FMMS has supported many fuel safety experiments supporting accident tolerant fuel development, light-water reactor safety, space thermal nuclear propulsion fuel development, and advanced reactor research and development. In Phase 2 of the FMMS restoration. work is now under way to expand the FMMS' field-of-view by adding an additional 96 channels of fast neutron detectors to the system's hodoscope, along with an expanded data acquisition system and associated transient timing electronics. An overview of the FMSS system and its fast-neutron detectors will be presented along with examples of current FMMS imaging performance and associated information.
46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
4-D Model of CO2 Deposition at North and South of Mars from HEND/Odyssey and MOLA/MGS
The first 1.5 year of neutron mapping measurements onboard Mars Odyssey spacecraft are presented based on High Energy Neutron Detector (HEND) observations. HEND instrument is a part of GRS suite responsible for registration of epithermal and fast neutrons originating in Mars subsurface layer. The scattering of fast neutrons in Mars surface caused by primary cosmic rays is strongly sensitive to presence of hydrogen atoms. Even several percents of subsurface water significantly depress epithermal and fast neutron flux. It turns orbit neutron spectroscopy into one of most efficient methods for finding distribution of subsurface water. The Mars Odyssey observations revealed huge water- ice regions above 60N and 60S latitudes. It was founded that distribution of subsurface water has layered structure at these regions. It is thought that more than 50% wt water ice covered by relatively dry layer with different thickness.
M L Litvak↗
HFIR SiC-SiC Composite Clad Tube Bowing Test: Pre-Irradiation Characterization
This report describes the pre-irradiation characterization of the SiC-SiC composite clad tube bowing experiment that will be irradiated in the High Flux Isotope Reactor (HFIR). There are concerns that SiCSiC fuel cladding in light water reactors could undergo bowing because of the non-uniform fast neutron flux profiles. This experiment is combined with the SiC-SiC channel box experiment, which has similar concerns regarding control blade movements in boiling water reactors. The HFIR experiment aims to validate the predicted deformation and stresses in the composite clad specimen after the exposure of fast neutron flux gradients. Significant radial fast neutron flux gradients that exist in the permanent reflector of HFIR were thoroughly characterized by using detailed 3D neutronic calculations. The 3D displacement damage dose rate profile and the resulting volumetric swelling in SiC were used as inputs to structural analyses to determine the predicted deformation that will affect the specimens. The tube specimens were characterized by using traditional dimensional inspection and surface profilometry to provide detailed information regarding the pre-irradiation condition as the first step to evaluate the radiation-induced deformation. Furthermore, fine engraving markers were inscribed along all outer surfaces of the specimen and mapped by using a digital microscope and a multidimensional stage. This allows marker spacings to be accurately measured and compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. The experiment successfully completed two HFIR cycle irradiations, cycles 492 and 493, which finished on July 25, 2021.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
New constraint on the Np 237 ( n , γ ) Np 238 integral cross section using the Godiva-IV critical assembly
Accurate knowledge of the 237 Np(n, γ) 238 Np cross section at fast neutron energies is important for applied nuclear science. The presently available experimental data has large disagreements in the fast neutron region. Perform a model-independent measurement of the 237 Np(n, γ) 238 Np integral cross section using a well characterized fast neutron source and compare the result with previous measurements and current nuclear data evaluations. Provide an integral measurement that can be used as a benchmark for current evaluations. Multiple samples of 237 Np were irradiated in the Godiva-IV critical assembly. Following the irradiation, the samples placed in a γ-ray counting setup and the γ-rays emitted from the decay of 238 Np were measured over a time period of approximately 7 days. Multiple γ-ray decay branches of 238 Np were observed. The observed activity of 238 Np was used to calculate the amount of 238 Np produced during the irradiation via the 237 Np(n, γ) 238 Np reaction and an integral cross section of 342(11) mb was measured for the Godiva-IV neutron spectrum. Further, the 238 Np half-life has been measured with a result of 50.31(5) hours. The 237 Np(n, γ) 238 Np integral cross section measured in this work is in agreement with overlapping 1σ error bands to ENDF/B-VIII.0. However, the measured value is 3σ away from the calculated integral cross section using JENDL-5. This measurement offers a reliable benchmark for future 237 Np(n, γ) 238 Np cross section evaluations.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗