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At least 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
NASA NTRS2022

Satellites Quantify the Spatial Extent of Cyanobacterial Blooms Across the United States at Multiple Scales

Previous studies indicate that cyanobacterial harmful algal bloom (cyanoHAB) frequency, extent, and magnitude have increased globally over the past few decades. However, little quantitative capability is available to assess these metrics of cyanoHABs across broad geographic scales and at regular intervals. Here, the spatial extent was quantified from a cyanobacteria algorithm applied to two European Space Agency satellite platforms—the MEdium Resolution Imaging Spectrometer (MERIS) onboard Envisat and the Ocean and Land Colour Instrument (OLCI) onboard Sentinel-3. CyanoHAB spatial extent was defined for each geographic area as the percentage of valid satellite pixels that exhibited cyanobacteria above the detection limit of the satellite sensor. This study quantified cyanoHAB spatial extent for over 2,000 large lakes and reservoirs across the contiguous United States (CONUS) during two time periods: 2008–2011 via MERIS and 2017–2020 via OLCI when cloud-, ice-, and snow-free imagery was available. Approximately 56% of resolvable lakes were glaciated, 13% were headwater, isolated, or terminal lakes, and the rest were primarily drainage lakes. Results were summarized at national-, regional-, state-, and lake-scales, where regions were defined as nine climate regions which represent climatically consistent states. As measured by satellite, changes in national cyanoHAB extent did have a strong increase of 6.9% from 2017 to 2020 (|Kendall’s tau (τ)| = 0.56; gamma (γ) = 2.87 years), but had negligible change (|τ| = 0.03) from 2008 to 2011. Two of the nine regions had moderate (0.3 ≤ |τ| < 0.5) increases in spatial extent from 2017 to 2020, and eight of nine regions had negligible (|τ| < 0.2) change from 2008 to 2011. Twelve states had a strong or moderate increase from 2017 to 2020 (|τ| ≥ 0.3), while only one state had a moderate increase and two states had a moderate decrease from 2008 to 2011. A decrease, or no change, in cyanoHAB spatial extent did not indicate a lack of issues related to cyanoHABs. Sensitivity results of randomly omitted daily CONUS scenes confirm that even with reduced data availability during a short four-year temporal assessment, the direction and strength of the changes in spatial extent remained consistent. We present the first set of national maps of lake cyanoHAB spatial extent across CONUS and demonstrate an approach for quantifying past and future changes at multiple spatial scales. Results presented here provide water quality managers information regarding current cyanoHAB spatial extent and quantify rates of change.

Harmful Algal Blooms↗
DOE OSTI2011

SOURCE OF MICROBUNCHING AT BNL NSLS SOURCE DEVELOPMENT LABORATORY

We report experimental studies of the origins of electron beam microbunching instability at BNL Source Development Laboratory (SDL). We eliminated laser-induced microbunching by utilizing an ultra-short photocathode laser. The measurements of the resulting electron beam led us to conclude that, at SDL, microbunching arising from shot noise is not amplified to any significant level. Our results demonstrated that the only source of microbunching instability at SDL is the longitudinal modulation of the photocathode laser pulse. Our work shows that assuring a longitudinally smoothed photocathode laser pulse allows mitigating microbunching instability at a typical FEL injector with a moderate microbunching gain. In this paper we investigated the source of microbunching instability at the SDL. To distinguish microbunching induced by shot noise from that arising from the longitudinal modulation of the photocathode laser, we studied the beam created by a very short laser pulse, thus eliminating the possibility of laser-induced microbunching. While the measured energy spectra of compressed beam did reveal severe longitudinal fragmentation, an analysis of the beam dynamics proved this to be due to self-fields acting on a beam with an initially smooth longitudinal profile, and not due to microbunching instability. Such fragmentation only was possible with the very short bunch chosen for these studies, and is absent in routine SDL operations. Our experiment shows that in the absence of the initial laser-induced beam modulation, microbunching instability at the SDL is not observed, and must be well below the levels that would limit the FEL performance. This result agrees with assumption of previous SDL studies that (when present under different machine conditions) microbunching instability at the SDL was laser-induced. Microbunching instability gain at the SDL is moderate. This is mainly because the SDL utilizes a single stage bunch compressor as well as due to the small compression ratio. Since the design of the SDL injector is typical of the majority of FEL injectors, our experiment proves that one possible way to control microbunching instability in such machines (that by design have a moderate microbunching gain) is to maintain a sufficiently smooth longitudinal profile of the photo-cathode laser. We note that the general principles for designing a machine with a moderate microbunching instability gain are presented in [12]. In conclusion, our experiment demonstrates that microbunching instability can be eliminated from a typical FEL injector with single stage bunch compressor (and operating without a laser heater) as long as the photocathode laser is longitudinally smooth. For machines with multi-stage bunch compressors, our results offer an important benchmark to establish a minimal laser heater power for instability-free operation.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Codisposal Waste Package Loading Options for DOE SNF and HLW

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types, such as light- and heavy-water-moderated reactors, graphite-moderated reactors, and breeder reactors, with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties resulting in a large diversity of reactor and fuel designs. Because of the wide variety and condition of SNF, a robust canister termed the DOE Standardized SNF Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard SNF Canister had four variations consisting of a 10- or 15-foot length and an 18- or 24-inch diameter. For ultimate disposal in the Yucca Mountain Repository these canisters were to be grouped with 24-inch diameter HLW canisters in a codisposal waste package. The 18-inch DOE Standardized SNF Canister could be placed in the middle of five HLW canisters. The 24-inch DOE Standardized SNF Canister would take the place of one the five HLW canister on the outer ring in the codisposal waste package. No DOE Standardized SNF Canisters have been loaded, and many fuel types can be loaded in multiple canisters. Depending on the selected diameter of the DOE Standardized SNF Canister for a particular fuel, the number of potential codisposal waste packages could vary. In addition, different types of DOE SNF may be able to be combined in a single canister. This could significantly change the number of DOE SNF canisters that are loaded and result in a decrease of the number of total potential codisposal waste packages. Ultimately, if too many HLW canisters need disposition, then a reduction in the number of DOE Standardized SNF Canisters will change the final total of codisposal waste packages very little. Likewise, if the ratio of DOE Standardized SNF Canisters to HLW canister produced is too high, then DOE Standardized SNF Canisters will not have enough HLW canisters to surround them in the codisposal waste package. This paper examines the results of differing loading strategies for a few DOE SNFs. It also compares the ranges of DOE Standardized SNF Canisters that may be produced to the ranges of HLW canisters that may be produced in order to determine potential scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Yttrium Hydride Post-Irradiation Examination Plan

Micro nuclear reactors are expected to use metal hydride moderators to ensure thier compact design for the flexible operational purposes. Department of Energy’s Microreactor program therefore aims to provide infrasture and qualification bed for the development of micro nuclear reactors. Because micro nuclear reactors must be compact to offer a flexible operational purposes, metal hydrides are expected to be used as neutron moderator materials due to their high hydrogen number density. Los Alamos National Laboratory (LANL) directed by the DOE to provide a qualified fabrication techniques of metals hydride, especially the yttrium hydride (YHx) which provides advantages over other metal hydride candiates. Qualification of YHx involves fabrication, property, and performance pillars, which must be met for the deployment of microreactors. In the support of the qualification process metal hydride, LANL has developed fabrication techniques based in direct hydriding and powder compacting. For both manufacturing techniques, fresh material properties of metal hydrides have been reported in the literature. To establish the performance limits and the qualification parameters, an irradiation campaign has been initated at Idaho National Laboratory (INL). The irradation campaign aims to investigate the critical irradiated YHx properties as a function of neutron fluence and temperaute. The first phase of the irradiation campaign includes the effect of the operating temperature under neutron irradiation which provides initial data on the metal hydrides and informs the manufacturing processes. So that, drop-in irradations in Advanced Test Reactor (ATR) has been designed. This document depicts the post-irradation examination (PIE) plan of the ATR irradiated YHx. Objective of the PIE plan is to serve as a qualification basis for YHx as a neutron moderator material by understanding the impacts irradiation has on the physical and thermophysical properties; to establish a link between fabrication and the irradiated YHx properties; to institute a PIE strategy for other solid-state neutron moderator candidates, and; to inform micro nuclear reactor fuel performance and safety codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Electrochemically modulated molten salt reactor

An electrochemically modulated molten salt reactor (EMMSR) that contains a vessel and a power source. The vessel houses a fuel salt, at least a portion of a neutron moderator, and at least a portion of an insulator. The fuel salt includes enough dissolved fissile isotopes to cause continued self-sustaining fission reactions during the operation of the EMMSR. The neutron moderator is configured to slow down fast neutrons produced by the dissolved fissile isotopes. The insulator is configured to electrically isolate the neutron moderator from the vessel. The power source has a positive potential and a negative potential. The positive potential is received by the neutron moderator and the negative potential is received by the vessel.

Jiang, Junhua↗
DOE OSTI2015

Prompt Gamma Ray Analysis of Soil Samples

Neutron moderation effects were measured in bulk soil samples through prompt gamma ray measurements from water and benzene contaminated soil samples using 14 MeV neutron inelastic scattering. The prompt gamma rays were measured using a cylindrical 76 mm x 76 mm (diameter x height) LaBr{sub 3}:Ce detector. Since neutron moderation effects strongly depend upon hydrogen concentration of the sample, for comparison purposes, moderation effects were studied from samples containing different hydrogen concentrations. The soil samples with different hydrogen concentration were prepared by mixing soil with water as well as benzene in different weight proportions. Then, the effects of increasing water and benzene concentrations on the yields of hydrogen, carbon and silicon prompt gamma rays were measured. Moderation effects are more pronounced in soil samples mixed with water as compared to those from soil samples mixed with benzene. This is due to the fact that benzene contaminated soil samples have about 30% less hydrogen concentration by weight than the water contaminated soil samples. Results of the study will be presented. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Minimum critical values for various fissile media with very low content of plutonium mixed in different kinds of matrix - 14340

For materials slightly contaminated by fissile nuclides (such as radioactive sludge), it could be useful to take into account the low fissile density in order to define the nuclear criticality safety limits (mass, volume,...) when this density is lower than the one of the optimally moderated fissile-water mixture (e.g. 30 g/dm{sup 3} for {sup 239}Pu). In such a case, if the material is modelled by the usual fissile-water mixture only, the fissile medium still contains a void fraction when the optimal moderation ratio is reached. Then, the question is to know whether modeling the other elements of the material can impact the nuclear criticality safety limits, in particular when they are good scattering elements. This paper aims at studying the impact of replacing the void fraction in plutonium fissile media, with various apparent densities, by different kinds of matrix containing scattering elements. Minimum critical values (mass, volume, diameter and thickness) have been determined for plutonium media, composed of 100 % {sup 239}Pu, with densities of 2, 5.8 and 7.3 g/dm{sup 3} in different kinds of matrix containing scattering elements (graphite, carbon, alumina, silica, magnesia, fluorite and calcium oxide). Different contents of the aforementioned scattering elements in the fissile media were studied and additional moderation was ensured by water. The results highlight that minimum critical values are lower when considering a fissile medium containing graphite, carbon, silica, alumina and magnesia with additional water moderation, compared to plutonium-water mixture only with the same low fissile apparent density. However, there is still a benefit of crediting density (critical values for a low fissile density are still higher than critical values without a density control). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Atalante research facility new mass limits in a unit containing plutonium and minor actinides - 14402

ATALANTE, located in Marcoule, is one of the main Nuclear Facilities of the French Alternative Energies and Atomic Energy Commission (CEA). This facility is mainly dedicated to research and development for nuclear energy related to the back end of the fuel cycle: spent fuel and final waste management. It is a large research complex housing chemistry and analytical laboratories and shielded cells (240 glove boxes and 11 shielded cells), equipped for basic studies through to demonstration experiments. Within the facility, one unit can contain plutonium and curium at the same time. In this case, nuclear criticality safety is achieved by exercising control over mass. A subcritical mass limit is defined for plutonium (280 g for {sup 239}Pu) and an exemption mass limit (3 g for {sup 245}Cm) is used for the actinide. In order to increase the maximum mass (in particular {sup 245}Cm) which can be handled in the unit, new mass control limits were studied, based on control over the sum of the mass ratio to the upper subcritical mass, noted m{sub lim}, for each nuclide (350 g for {sup 239}Pu; 34.8 g or 23 g for {sup 245}Cm) as shown in equation (1) below. m({sup 239}Pu)/ m{sub lim}({sup 239}Pu)+ m({sup 245}Cm)/m{sub lim}({sup 245}Cm)≤1 (1) To validate this criticality safety control method, a study was conducted to demonstrate that, for different masses of {sup 239}Pu and {sup 245}Cm meeting equation (1), the calculated effective multiplication factor is always lower than the effective multiplication factor corresponding to a sphere of 350 g of pure {sup 239}Pu, moderated, with 20 cm water reflection. The two pseudo fissile media are: - metallic plutonium (100% {sup 239}Pu, density 19.86), moderated with water, - metallic curium (100% {sup 245}Cm, density 13.566), moderated with water. The calculation model is a sphere of {sup 245}Cm metal positioned in a sphere of {sup 239}Pu metal, with optimum moderation for each fissile material, and a full water reflection (20 cm). The results show that, under certain conditions, the effective multiplication factor of the mixture can be higher than the effective multiplication factor of each pure fissile material when using the upper subcritical mass of 34.8 g for {sup 245}Cm. When using 23 g of {sup 245}Cm, the k{sub eff} of the mixture is always lower than the k{sub eff} of pure {sup 239}Pu. These calculations enable the safety documentation to be changed by introducing a new process operational procedure, based on equation (1). This change in procedure allows the operator to work with a larger quantity of curium in the unit. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Microstructural characterization of the oxidation of nuclear graphite under chronic and accident conditions via XCT and synchrotron XCT

Graphite is a candidate material to moderate fast neutrons and for structural components in the US next-generation graphite-moderated reactors. A graphite core is conceived as a large formation of interconnected bricks that primarily serves as the moderator of fast neutrons, holds instrumentation, fuel elements, control rods, and is a receptacle for molten salts for Molten Salt Reactors (MSRs) designs. During the operation of a graphite-moderated power plant, graphite components might be subjected to chronic oxidation during normal operating conditions or aggressive oxidation as a result of accidental ingress of air that reacts vigorously with the graphite core. Using synchrotron, x-ray computed tomography (XCT), this research systematically characterized microstructural changes that accompanied these two oxidation scenarios. Chronic oxidation was studied by characterizing IG-110, PCEA, and NBG-18 specimens that were gradually oxidized in air at a low temperature (520°C). The accidental ingress of air into the graphite reactor was simulated by subjecting the grades mentioned above to high-temperature oxidation at approximately 750°C. This research is the first, in situ, systematic characterization of nuclear graphite microstructural evolution that can be associated with the two possible oxidation scenarios and provide insight into related repercussions. The results show that the microstructure and pore connectivity influence the rate of oxidation and evolution of the microstructure under the two oxidation regimes. These results are crucial to understanding which grades are more resilient to each type of oxidation and documenting the damage created in the graphite.

Arregui Mena, Jose'↗
DOE OSTI2024

Sum-of-Fractions Method

Sum-of-fractions is a method intended to make sure a subcritical margin for aqueous solutions and slurries of fissionable isotopes exists. The method indicates that a system is subcritical if the sum of the ratios of the mass of each isotope (in a mixture) to its individual minimum subcritical mass limit is less than or equal to one. Historically, the basis of the sum-of-fractions has been derived from allowances given in the American National Standards Institute (ANSI)/ American Nuclear Society (ANS)-8.15-1981. However, the allowance was removed in ANSI/ANS-8.15-2014 due to a lack of technical basis. A methodology was developed to assess the validity of using the sum-of-fractions for water- or polyethylene-moderated systems for the following nuclides: 232U, 233U, 234U, 235U, 237Np, 236Pu, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 242mAm, 243Am, 242Cm, 243Cm, 244Cm, 245Cm, 246Cm, 247Cm, 249Cf, and 251Cf. The methodology uses available benchmark data for mixtures of 233U, 235U, and 239Pu to establish the calculational margin, and a mass limit reduction to establish the margin of subcriticality. Water- or polyethylene-moderated and -reflected mixtures containing the nuclides are evaluated with the code system, SCALE 6.2.4. Including the calculational margin, subcritical mass limits for each nuclide were computed for optimally water- or polyethylene-moderated and fully reflected systems. These masses were used to create nuclide mixtures in which the sum of the mass to subcritical mass limit ratios is one. The various nuclide mixtures were modeled over a range of moderation and demonstrate the keff does not exceed the calculational margin. For additional assurance of subcriticality, a significant mass reduction is applied to each computed minimum critical mass of the nuclides without adequate benchmark data consistent with the method in ANSI/ANS-8.15-2014.

criticality safety, Actinide↗
DOE OSTI2012

Radiation Effects in Graphite

The requirements for a solid moderator are reviewed and the reasons that graphite has become the solid moderator of choice discussed. The manufacture and properties of some currently available near-isotropic and isotropic grades are described. The major features of a graphite moderated reactors are briefly outlined. Displacement damage and the induced structural and dimensional changes in graphite are described. Recent characterization work on nano-carbons and oriented pyrolytic graphites that have shed new light on graphite defect structures are reviewed, and the effect of irradiation temperature on the defect structures is highlighted. Changes in the physical properties of nuclear graphite caused by neutron irradiation are reported. Finally, the importance of irradiation induced creep is presented, along with current models and their deficiencies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Elastic moduli of high-density, sintered monoliths of yttrium dihydride

Yttrium dihydride has been considered as potential nuclear reactor moderator for micro-reactors operating in the thermal energy spectrum due to their ability to retain hydrogen to high temperatures while having a relatively low impact on neutron economy. Despite these advantages, it is difficult to prepare yttrium dihydride in high-density, near net-shape monoliths for moderator applications due to volume expansion upon hydriding and the pyrophoric nature of the material. In this work, high-density monoliths of yttrium dihydride were prepared by direct hydriding and using powder metallurgical methods. Resonant ultrasound spectroscopy was used to measure the elastic moduli of the sintered monoliths, while nanoindentation was used to determine the Young’s modulus and hardness of the directly-hydrided yttrium and the sintered monoliths. Single crystal elastic constants and related properties of yttrium dihydride were also determined using density functional theory (DFT) for comparison. We report mechanical properties of materials produced by both methods were observed to be consistent with the values from literature for yttrium dihydride. Due to the novelty of producing yttrium dihydride by powder metallurgy, this result indicated that the monoliths produced by powder metallurgy were of high quality. To that end, powder metallurgy is believed to be a viable method for large-scale production of yttrium dihydride monoliths for nuclear reactor moderator applications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Thermophysical properties of high-density, sintered monoliths of yttrium dihydride in the range 373–773 K

Yttrium dihydride is a promising nuclear reactor moderator for high-temperature, thermal micro-reactors due to its ability to retain hydrogen to high temperatures while having a relatively low impact on neutron economy. However, it is difficult to fabricate yttrium dihydride in high-density, near net-shape monoliths for moderator applications due to challenges associated with the hydrogen absorption process. Here, high-density monoliths of yttrium dihydride were prepared by powder metallurgical methods. The lattice strain and thermal expansion coefficient were measured using neutron diffraction of powders of directly-hydrided yttrium, while the molar heat capacity and thermal diffusivity of sintered yttrium dihydride were measured using differential scanning calorimetry (DSC) and laser flash analysis (LFA), respectively. The molar heat capacity of yttrium dihydride was also calculated using density functional theory (DFT) for comparison. These parameters were used to then calculate the thermal conductivity and resistivity of yttrium dihydride as a function of temperature. The thermophysical properties of materials produced by both methods were observed to be consistent with the values from literature for yttrium dihydride. Due to the novelty of producing yttrium dihydride by powder metallurgical methods, this result indicated that the sintered monoliths were of high quality and that powder metallurgy is a viable method for large-scale production of yttrium dihydride monoliths for nuclear reactor moderator applications.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Characterization of the microstructure of yttrium hydride under proton irradiation

High moderation per unit volume solid moderator materials like yttrium hydride (YH x ) are necessary for compact nuclear microreactors. However, the phase stability and hydrogen transport processes of YH x under high-temperature irradiation are largely unknown. Proton irradiation was conducted on YH x at 300 °C and 580 °C to 0.2 dpa using 1 MeV or 2 MeV protons in a high-vacuum environment. The hydrogen concentration was determined before and after irradiation using elastic recoil detection analysis, and microstructural evolution was examined via post-irradiation scanning transmission electron microscopy and Raman spectroscopy. Dislocation loops and cavities were observed in all conditions; their distribution was correlated with the bombarding proton energy and ion irradiation temperature. This work revealed that hydrogen retention is proportional to the formation of traps for hydrogen gas atoms and identified pathways for hydrogen release. The relative contributions of bulk or fast diffusion paths, such as grain boundaries, delamination boundaries, and stacking faults are discussed; the primary mechanisms of hydrogen loss are likely based on diffusion, ruling out artefacts of the experimental design. In conclusion, the study suggests proton irradiation may be a strong surrogate to study hydrogen transport in hydride moderator materials under irradiation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

High temperature elastic properties of sub-stoichiometric yttrium dihydrides

Yttrium hydrides are considered as candidate materials for neutron moderation applied in microreactors (akin transportable tiny nuclear reactors) owing to their superior thermal stability and hydrogen retention. The evolution of elastic properties of these materials at elevated temperatures, needed for predicting the thermomechanical response and performance of the moderator during in-service reactor conditions, however, is lacking. Here, we report the Young’s and shear elastic moduli of three stoichiometries of bulk yttrium hydride (YH x , x = 1.61, 1.82, and 1.84) from room temperature to 1000°C. In situ temperature-dependent measurements of the longitudinal and shear wave velocities were performed using a laser ultrasonic technique while heating the sample in a vacuum-pumped heating stage. The elastic moduli increased linearly with increasing hydrogen content and decreased by ~10% during heating from room temperature to 1000°C in the three YH x compositions. The linear relationship between the elastic moduli and the hydrogen content in yttrium hydride was verified by atomistic calculations based on density functional theory (DFT). The absence of abrupt changes in the temperature-dependent measurements of elastic modulus of the YH x samples suggested negligible loss of hydrogen at elevated temperatures. Excellent agreement was found between the measured and calculated dependence of the elastic moduli on the stoichiometry, thereby providing a new approach for investigating the effects of fabrication-induced parameters (such as porosity) on the elastic moduli. Furthermore, this study demonstrates the utility of the combined approach involving DFT-based atomistic calculations and measurements of the elastic moduli for the informative development of metal hydrides and can be used as a metric for novel moderator materials investigations for emerging microreactors and beyond.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2010

PRESSURE FLUCTUATION BEHAVIOR IN THE CRYOGENIC HYDROGEN SYSTEM CAUSED BY A 100 kW PROTON BEAM INJECTION

Supercritical hydrogen (1.5 MPa and around 20 K) has been selected as a moderator material for the intense spallation neutron source (JSNS) in J-PARC. The cryogenic hydrogen system provides the supercritical hydrogen for the moderators and removes the nuclear heating at the moderators, which is estimated to be 3.8 kW for a proton beam power of 1 MW. The pressure control system was designed to mitigate pressure fluctuation caused by suddenly turning a proton beam on and off, which is composed of a heater as an active controller for thermal compensation and an accumulator as a passive volume controller. A 109 kW proton beam was injected to the JSNS in December 2007. The pressure fluctuation behaviors were studied for the 109 kW proton beam operation. As soon as the proton beam was injected, the accumulator spontaneously started to constrict. The heater control succeeded in maintaining a constant heat load applied to the cryogenic hydrogen system. The pressure control system successfully reduced the pressure fluctuation within 6 kPa. It was confirmed that the pressure control system was effective.

43 PARTICLE ACCELERATORS↗
DOE OSTI2017

Neutron-beam-shaping assembly for boron neutron-capture therapy

A neutron-beam-shaping assembly consisting of a moderator, a reflector, and an absorber is used to form a therapeutic neutron beam for the boron neutron-capture therapy of malignant tumors at accelerator neutron sources. A new structure of the moderator and reflector is proposed in the present article, and the results of a numerical simulation of the neutron spectrum and of the absorbed dose in a modified Snyder head phantom are presented. The application of a composite moderator and of a composite reflector and the implementation of neutron production at the proton energy of 2.3MeV are shown to permit obtaining a high-quality therapeutic neutron beam.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965 have been evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study "power plants for the production of electrical power in space vehicles." The third configuration in this series of experiments was a beryllium-reflected assembly of stainless-steel-clad, highly enriched uranium (HEU)-O 2 fuel mockup of a potassium-cooled space power reactor. Reactivity measurements cadmium ratio spectral measurements and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The fuel tube effect and neutron moderating and absorbing material worth measurements are the focus of this paper. Additionally, a measurement of the worth of potassium filling the core region was performed but has not yet been evaluated Pellets of 93.15 wt.% enriched uranium dioxide (UO 2 ) were stacked in 30.48 cm tall stainless steel fuel tubes (0.3 cm tall end caps). Each fuel tube had 26 pellets with a total mass of 295.8 g UO 2 per tube. 253 tubes were arranged in 1.506-cm triangular lattice. An additional 7-tube cluster critical configuration was also measured but not used for any physics measurements. The core was surrounded on all side by a beryllium reflector. The fuel effect worths were measured by removing fuel tubes at various radius. An accident scenario was also simulated by moving outward twenty fuel rods from the periphery of the core so they were touching the core tank. The change in the system reactivity when the fuel tube(s) were removed/moved compared with the base configuration was the worth of the fuel tubes or accident scenario. The worth of neutron absorbing and moderating materials was measured by inserting material rods into the core at regular intervals or placing lids at the top of the core tank. Stainless steel 347, tungsten, niobium, polyethylene, graphite, boron carbide, aluminum and cadmium rods and/or lid worths were all measured. The change in the system reactivity when a material was inserted into the core is the worth of the material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗