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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Air and radon pathways screenings methodologies for the next revision of the E-area PA

The strategic plan for the next E-Area Low-Level Waste Facility Performance Assessment includes recommended changes to the screening criteria used to reduce the number of radioisotopes that are to be considered in the air and radon pathways incorporated into the GoldSim® atmospheric release model (ARM). For the air pathway, a revised screening methodology was developed based on refinement of previous E-Area PA screening approaches and consideration of the strategic plan recommendations. The revised methodology has three sequential screening steps for each radioisotope: (1) volatility test using the Periodic Table of the Elements, (2) stability test based on half-life, and (3) stability test based on volatility as measured by the Henry’s Law constant for the assumed dominant gaseous species or vapor pressure in the case of tritiated water. Of the 1252 radioisotopes listed in the International Commission on Radiological Protection Publication 107, only the 10 that satisfied all three steps of the revised screening methodology will be included in the ARM. They are: Ar-37, Ar-39, Ar-42, C-14, H-3, Hg-194, Hg-203, Kr-81, Kr-85, and Xe-127. For the radon pathway, a revised screening methodology was developed that also has three sequential steps: (1) identify all decay chains that terminate at Rn-222, (2) screen out parents that decay through U-238 because of its 4.5-billion-year primordial half-life, and (3) eliminate remaining parents whose half-life is shorter than one day. Of the 86 possible decay chains leading to Rn-222, six decay chains consist of 15 unique radioisotopes that will be incorporated into the ARM. The 15 radioisotopes are: U-238, Th-234, Pa-234m, Pu-238, U-234, Th-230, Ra-226, Cf-246, Cm-242, Am-242m, Am-242, Np-238, Np-234, Pa-230, and Rn-222.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Characterization of Aluminum Metal Powders for Plutonium-238 Program

As part of radioisotope thermoelectric generator (RTG) production for power sources in deep space exploration, Oak Ridge National Laboratory (ORNL) produces plutonium-238 ( 238 Pu). To produce 238 Pu, neptunium-237 ( 237 Np) targets are fabricated at ORNL and subsequently irradiated at the High Flux Isotope Reactor (HFIR) at ORNL and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Specifically, current flowsheets utilize neptunium dioxide ( 237 NpO 2 ) targets. Aluminum (Al) powder is blended with 237 NpO 2 target material to improve thermal conductivity during irradiation. The current composition of 237 NpO 2 /Al pellets (i.e., cermets, or ceramic–metallics) is 20% 237 NpO 2 , 70% Al, and 10% void space. Al powders utilized by the 238 Pu program are high-fired under vacuum before blending into targets. The primary objective of this work is to measure the physical, chemical, and thermal properties of Al before and after the high-firing process. Properties of interest include the morphology, surface texture, particle size, crystal phase(s), and thermal conductivity of the material. These properties are measured with existing materials characterization equipment at ORNL, including powder x-ray diffraction (pXRD), scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS), laser flash analysis (LFA), and thermogravimetric analysis (TGA). A secondary objective of this work is to compare the Al powders before and after the high-firing procedure to determine the effects of high-firing on chemical, physical, and thermal properties and to determine the efficacy of the high-firing process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Neptunium mononitride as a target material for Pu-238 production

Deep space exploration requires specialized sources for both thermal and power applications. Radioactive decay heat of plutonium-238 (238Pu) provides these sources in the form of radioisotope thermoelectric generators (RTGs). The 238 Pu is produced via neutron capture reaction involving neptunium-237 ( 237 Np) target material. Continual optimization of 237 Np target materials and evaluation of potential alternative targets for production of 238 Pu RTGs are advantageous for meeting ongoing space power system resource requirements. Current production of 238 Pu for RTGs for the United States space program utilizes neptunium dioxide ( 237 NpO 2 ) targets; however, the use of neptunium mononitride ( 237 NpN) presents an opportunity to increase the mass of 237 Np per target compared to the dioxide form, as well as increase the thermal conductivity of the target. To assess the viability of a 237 NpN target material, the material chemistry must be thoroughly evaluated, including synthesis methods and dissolution and reprocessing schemes. This review presents a summary of synthesis pathways for 237 NpN based on published literature on actinide mononitrides. Specific literature on 237 NpN is limited, necessitating evaluation of other actinide systems to gather parallels. This suggests a need for additional experimental studies on 237 NpN. A particular limitation in the existing literature is a lack of information on the differences in material characteristics, such as morphology, particle size, and trace chemical impurities, as a function of synthesis method. These parameters may affect subsequent reactor performance or dissolution of irradiated targets. The evaluation of existing literature is presented with a focus on the efficacy of 237 NpN targets for 238 Pu production.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Analysis of neptunium oxides produced through modified direct denitration

Production of neptunium-237 ( 237 Np) target materials for plutonium-238 ( 238 Pu) radioisotope thermoelectric generators (RTGs) for deep space exploration requires advanced chemistry and engineering development. Currently, the domestic Pu-238 Supply Program at Oak Ridge National Laboratory produces neptunium dioxide (NpO 2 ) for target material using a modified direct denitration (MDD) flowsheet. Although the chemistry, reaction mechanisms, and product characteristics of MDD are well understood for uranium, corresponding studies of the neptunium system are still needed to continue optimization of target material properties, production equipment design, and production flowsheets. Here, the objective of this work is to characterize crystalline phases, morphology, surface texture, and particle size of NpO 2 produced via MDD reactions. Solid-phase characterization techniques, including powder X-ray diffraction (pXRD) and scanning electron microscopy with energy-dispersive spectroscopy (SEM-EDS), were employed to achieve this objective. Subsequent data processing using the Morphological Analysis for Material Attribution (MAMA) software was performed to analyze particle morphology and size. Broadly, the powders were found to contain a mixture of NpO 2 and Np 2 O 5 after denitration with a variety of morphologies. After high-firing, the product was found to be NpO 2 with a typical polycrystalline oxide morphology and a grain size ranging from 0.72 to 0.94 µm. These analyses provide knowledge on the reaction pathway for a non-traditional NpO 2 synthesis method and offer additional unique insight into production-scale environments for transuranic materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Predictive Contaminant Transport Simulation with the P2R Model for the Composite Analysis Base Case

In order to meet the requirements of U.S. Department of Energy (DOE) in DOE O 435.1 Chg. 1, Radioactive Waste Management, a composite analysis (CA) must be completed for the Central Plateau at the Hanford Site. The updated Hanford Site CA requires estimates of fate and transport of radionuclides in the groundwater from multiple sources within and downgradient of the Central Plateau. Numerical simulations using the MT3D-MST computer code were conducted to evaluate the fate and transport of existing distributions of 16 radionuclides as identified in CP-62184, Hanford Site Composite Analysis: Radionuclide Selection for Groundwater Pathway Evaluation: tritium (H-3), carbon-14 (C-14), chloride- 36 (Cl-36), strontium-90 (Sr-90), technetium-99 (Tc-99), rhenium-187 (Re-187), radium-226 (Ra-226), thorium-230 (Th-230), uranium-232 (U-232), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), neptunium-237 (Np-237), and uranium-238 (U- 238). This document details the application of the P2R Model version 8.3 to predict the fate and transport of contaminants in groundwater on the Central Plateau for the 10,052 year simulation to support the CA. The simulation of fate and transport of contaminants reported in this environmental calculation file (ECF) will support dose predictions as part of the updated Hanford Site CA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Comparison of Advanced Processing of Irradiated UO{sub 2} Fuels and NpO{sub 2} Targets

Advanced processing methods for UO{sub 2}-based used nuclear fuels have been developed and demonstrated in several experimental programs at the Oak Ridge National Laboratory's Radiochemical Engineering Development Center (REDC). The Consolidated Fuel Recycle Program (1979-1986) and the Advanced Fuel Cycle Initiative (2002- 2008) were sponsored by the U.S. Department of Energy Office of Nuclear Energy (DOE/NE). Currently, DOE/NE is administering a new program for the National Aeronautics and Space Administration to produce {sup 238}Pu by irradiating NpO{sub 2}, and the irradiated targets are to be chemically processed using the same equipment and shielded facilities in the REDC that were used previously to process irradiated UO{sub 2}-based fuels. Flowsheets for processing irradiated NpO{sub 2} targets are now being developed and demonstrated. There are both similarities and differences in the processing of the two types of irradiated fuels/targets, and a comparison of these is useful for application to new types of feed material. A comparison of similarities and differences for processing UO{sub 2}- and NpO{sub 2}-based feed materials has been useful in planning for advanced processing methods and new requirements. Use of existing multistage solvent extraction equipment and shielded processing facilities at the ORNL REDC has enabled development, demonstration, and application of new and improved separations methods. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 ( 238 Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 ( 237 Np) dioxide (NpO 2 )/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs that were necessary due to the limiting fabrication and irradiation data available. The calculated parameters in the final production target model are the most accurate and comprehensive, while still conservative. Over 210 permutations in irradiation time and position were evaluated, and are supported by the most recent inputs and highest fidelity methodology. The results of these analyses show that the models presented in this report provide a robust and reliable basis for previous, current and future experiment safety analyses. In addition, they reveal an evolving knowledge of the steady-state behavior of the NpO 2 /Al pellets under irradiation for a variety of target encapsulations and potential conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Opportunities for the Multi Recycling of Used MOX Fuel in the US - 12122

Over the last 50 years the US has accumulated an inventory of used nuclear fuel (UNF) in the region of 64,000 metric tons in 2010, and adds an additional 2,200 metric tons each year from the current fleet of 104 Light Water Reactors. This paper considers a fuel cycle option that would be available for a future pilot U.S. recycling plant that could take advantage of the unique opportunities offered by the age and size of the large U.S. UNF inventory. For the purpose of this scenario, recycling of UNF must use the available reactor infrastructure, currently LWR's, and the main product of recycling is considered to be plutonium (Pu), recycled into MOX fuel for use in these reactors. Use of MOX fuels must provide the service (burn-up) expected by the reactor operator, with the required level of safety. To do so, the fissile material concentration (Pu-239, Pu-241) in the MOX must be high enough to maintain criticality, while, in current recycle facilities, the Pu-238 content has to be kept low enough to prevent excessive heat load, neutron emission, and neutron capture during recycle operations. In most countries, used MOX fuel (MOX UNF) is typically stored after one irradiation in an LWR, pending the development of the GEN IV reactors, since it is considered difficult to directly reuse the recycled MOX fuel in LWRs due to the degraded Pu fissile isotopic composition. In the US, it is possible to blend MOX UNF with LEUOx UNF from the large inventory, using the oldest UNF first. Blending at the ratio of about one MOX UNF assembly with 15 LEUOx UNF assemblies, would achieve a fissile plutonium concentration sufficient for reirradiation in new MOX fuel. The Pu-238 yield in the new fuel will be sufficiently low to meet current fuel fabrication standards. Therefore, it should be possible in the context of the US, for discharged MOX fuel to be recycled back into LWR's, using only technologies already industrially deployed worldwide. Building on that possibility, two scenarios are assessed where current US inventory is treated; Pu recycled in LWR MOX fuels, and used MOX fuels themselves are treated in a continuous partitioning-transmutation mode (case 2a) or until the whole current UNF inventory (64,000 MT in 2010) has been treated followed by disposal of the MOX UNF to a geologic repository (case 2b). In the recycling scenario, two cases (2a and 2b) are considered. Benefits achieved are compared with the once through scenario (case 1) where UNF in the current US inventory are disposed directly to a geologic repository. For each scenario, the heat load and radioactivity of the high activity wastes disposed to a geologic repository are calculated and the savings in natural resources quantified, and compared with the once-through fuel cycle. Assuming an initial pilot recycling facility with a capacity of 800 metric tons a year of heavy metal begins operation in 2030, ∼8 metric tons per year of Pu is recovered from the LEUOx UNF inventory, and is used to produce fresh MOX fuels. At a later time, additional treatment and recycling capacities are assumed to begin operation, to accommodate blending and recycling of used MOX Pu, up to 2,400 MT/yr treatment capacity to enable processing UNF slightly faster than the rate of generation. Results of this scenario analysis study show the flexibility of the recycling scenarios so that Pu is managed in a way that avoids accumulating used MOX fuels. If at some future date, the decision is made to dispose of the MOX UNF to a geologic repository (case 2b), the scenario is neutral to final repository heat load in comparison to the direct disposal of all UNF (case 1), while diminishing use of natural uranium, enrichment, UNF accumulation, and the volume of HLW. Further recycling of Pu at the end of the scenario (case 2a) would exhibit further benefits. As expected, Pu-241 and Am-241 are the source of long term HLW heat load and Am-241 and Np-237 are the source of long term radiotoxicity. When advanced technology is available, introduction of minor actinide recycling, in addition to Pu recycling, by the end of this scenario, or sooner, would have a major impact on final repository heat load and long term radiotoxicity of the HLW. This scenario is also compatible with a gradual introduction of a small number of FR's for Pu management. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

235-F GoldSim Fate and Transport Model: Uncertainty Quantification

Building 235-F was configured with two missions in mind: Actinide Billet Line (ABL) and the fabrication of Pu-238 oxide for space program applications. ABL produced Np-237 billets for use in SRS reactors, whereas the design process, fabrication, and examination of Pu-238 oxide powder occurred within the following areas, respectively: Plutonium Experimental Facility (PEF), Plutonium Fuel Form (PuFF), and Old Metallography Lab (OML). By 1990 production ceased and by 2006 de-inventory occurred; however, assays have shown significant holdup remains within ABL and PuFF. As a result, 235-F is a Category 2 nuclear facility, with plans to undergo deactivation and decommission (D and D) via In-Situ Disposal (ISD). The purpose of this project is to ensure United States Environmental Protection Agency (USEPA) groundwater radiation maximum contaminant level (MCL) and dosage standards are met during the D and D of 235-F by quantifying uncertainty through probabilistic modeling and evaluation of various ISD alternatives. GoldSim is a dynamic modeling software package with a graphical, object-oriented interface capable of capturing the influence of complex system input variability on probabilistic system outcomes. A GoldSim stochastic fate and transport model for 235-F was developed and matched with a PORFLOW deterministic model to simulate probabilistic release and flow of radionuclides from ABL and PuFF into the vadose zone, the Upper Three Runs (UTR) Aquifer, and UTR Creek. The GoldSim model was used to probabilistically evaluate four ISD scenarios against USEPA groundwater radiation MCLs and dosage standards. The deterministic 235-F GoldSim fate and transport model continues to be refined to match the results of the PORFLOW deterministic model to ensure the model accurately represents radionuclide movement through the groundwater system. The stochastic variables that are utilized within the GoldSim model are founded on the most current data; a conservative perspective is taken where needed. Alignment with the PORFLOW deterministic model, coupled with input stochastic variability, allows the probabilistic 235-F GoldSim model to capture the conservative breadth of possible outcomes for radionuclide fate within this particular system. This ensures that the USEPA MCLs and dosage limits hold even in the worst case scenarios.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Thermal decomposition of neptunyl ammonium nitrate: mechanistic insights and structural characterization of the Np 2 O 5 intermediate phase

Neptunium (Np) possesses a rich and unique chemistry that often diverges from other actinide elements yet remains relatively underexplored compared with the other light actinides. A resurgence of interest in Np has been spurred by the application of 237 Np for plutonium-238 ( 238 Pu) production for use in radioisotope thermoelectric generators (RTGs), necessitating evaluation of Np chemical reactions and materials. The work presented here studied the thermal decomposition of neptunyl ammonium nitrate (NH 4 Np VI O 2 (NO 3 ) 3 ) for synthesis of neptunium dioxide (NpO 2 ), which is the target material used for production of 238 Pu. Additionally, structural characterization of the intermediate solid Np pentoxide (Np 2 O 5 ) was performed. Advanced solid-state characterization techniques, including simultaneous thermal analysis (STA), powder X-ray diffraction (pXRD), Raman spectroscopy, and density functional theory (DFT) modeling have been combined to study the reaction pathways. Analysis revealed that NH 4 Np VI O 2 (NO 3 ) 3 thermally decomposes to a proposed neptunyl nitrate intermediate, followed by Np 2 O 5 and finally NpO 2 , all within the temperature range of 150 °C–600 °C. Further characterization of the pentoxide intermediate provided the first Raman spectra of pure-phase Np 2 O 5 and associated DFT modeling confirmed Raman peak assignments for this phase. These findings provide mechanistic information to advance production of the critical radioisotope 238 Pu and advance the state of knowledge on Np materials chemistry using modern characterization techniques.

Lawson, Kathryn M. [Oak Ridge National Laboratory ↗
DOE OSTI2012

Proliferation resistant fuel for pebble bed modular reactors

We show that it is possible to denature the Plutonium produced in Pebble Bed Modular Reactors (PBMR) by doping the nuclear fuel with either 3050 ppm of {sup 237}Np or 2100 ppm of Am vector. A correct choice of these isotopes concentration yields denatured Plutonium with isotopic ratio {sup 238}Pu/Pu {>=} 6%, for the entire fuel burnup cycle. The penalty for introducing these isotopes into the nuclear fuel is a subsequent shortening of the fuel burnup cycle, with respect to a non-doped reference fuel, by 41.2 Full Power Days (FPDs) and 19.9 FPDs, respectively, which correspond to 4070 MWd/ton and 1965 MWd/ton reduction in fuel discharge burnup. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Prompt and Delayed Inelastic Scattering Reactions from Fission Neutron PGAA - First Results of FaNGaS

The new instrument Fast Neutron Gamma Spectroscopy (FaNGaS) has been installed at the SR10 beam line of the FRM II Research Reactor in Garching and tested successfully. Complimentary to cold neutron PGAA, with FaNGaS inelastic scattering reactions induced by fission neutrons can be studied. Gamma lines from (n,n'γ) reactions up to now have been rarely studied and no adequate compilation of the emitted gamma energies exist. In developing nondestructive analytical techniques using neutron generator based PGAA such data are badly needed for quantification of heavy metals and actinides in e.g. nuclear waste or safeguards samples. A number of elements and relevant actinides have been irradiated in the fast neutron beam SR10 at the FRM II reactor in Garching, Germany. A heavily shielded 50% eff. HPGe detector perpendicular to the beam is looking at the samples exposed to 2.3 E8 cm{sup -2}s{sup -1} fission neutrons. Prompt gamma spectra have been taken and evaluated using the available data in scattered sources. Additional gamma lines have been detected and are being compiled to create a data base for (n,n') reactions. Particular emphasis is given on actinides including {sup 238}U, {sup 232}Th, {sup 237}Np, {sup 242}Pu and {sup 241}Am. Some examples will be given and first results will be discussed in this contribution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Comparison of neutron diffusion and Monte Carlo models for a fission wave

Many groups have used neutron diffusion simulations to study fission wave phenomena in natural or depleted uranium. However, few studies of fission wave phenomena have been published that use Monte Carlo simulations to confirm the results of diffusion models for this type of system. In the present work we show the results of a criticality and burnup simulation of a traveling wave reactor using MCNPX 2.7.0. The characteristics of the fission wave in this simulation are compared with those from a simple one-dimensional, one-group neutron diffusion model. The diffusion simulations produce a wave speed of 5.9 cm/yr versus 5.3 cm/yr for the Monte Carlo simulations. The axial flux profile in the Monte Carlo simulation is similar in shape to the diffusion results, but with different peak values, and the two profiles have an R2 value of 0.93. The {sup 238}U, {sup 239}Np and {sup 239}Pu burnup profiles from the diffusion simulation show good agreement with the Monte Carlo simulations, R values of 0.98, 0.93 and 0.97 respectively are observed. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

K West Basin Sand Filter Backwash Sample Analysis

A sand filter is used to help maintain water clarity at the K West Basin where highly radioactive sludge is stored. Eventually that sand filter will require disposal. The radionuclide content of the solids trapped in the sand filter will affect the selection of the sand filter disposal pathway. The Pacific Northwest National Laboratory (PNNL) was contracted by the K Basin Operations & Plateau Remediation Project (operations contractor CH2M Hill) to analyze the radionuclide content of the solids collected from the backwash of the K West Basin sand filter. The radionuclide composition in the sand filter backwash solids will be used by CH2M Hill to determine if the sand filter media and retained sludge solids will be designated as transuranic waste for disposal purposes or can be processed through less expensive means. On October 19, 2015, K Basin Operations & Plateau Remediation Project staff backwashed the sand filter into the North Load-Out Pit (NLOP) and immediately collected sample slurry from a sampling tube positioned 24 in. above the NLOP floor. The 764 g sand filter backwash slurry sample, KW-105 SFBW-001, was submitted to PNNL for analysis on October 20, 2015. Solids from the slurry sample were consolidated into two samples (i.e., a primary and a duplicate sample) by centrifuging and measured for mass (0.82 g combined – wet centrifuged solids basis) and volume (0.80 mL combined). The solids were a dark brown/orange color, consistent with iron oxide/hydroxide. The solids were dried; the combined dry solids mass was 0.1113 g, corresponding to 0.0146 weight percent (wt%) solids in the original submitted sample slurry. The solids were acid-digested using nitric and hydrochloric acids. Insoluble solids developed upon dilution with 0.5 M HNO 3 , corresponding to an average 6.5 wt% of the initial dry solids content. The acid digestate and insoluble solids were analyzed separately by gamma spectrometry. Nominally, 7.7% of the 60 Co was present in the insoluble solids; less than 1% of other gamma-emitters (i.e., 137 Cs, 154/155 Eu, and 241 Am) were present in the insoluble solids. Aliquots of the acid digestate were analyzed directly using gamma energy analysis (GEA) and after separations for 238 Pu, 239+240 Pu, 237 Np, and 241 Am radioisotopes using alpha energy analysis (AEA). The 90 Sr was measured by liquid scintillation counting (LSC) on the Sr-separated fraction. The plutonium isotopic distribution of the acid digestate was analyzed following Pu separations by thermal ionization mass spectrometry (TIMS). A table summarizes the results for the primary and duplicate samples. The 239+240 Pu concentration (µCi/g dry) relative to 90 Sr and to 137 Cs concentrations (µCi/g dry) was examined. The K West Basin sludge has a 239+240 Pu/ 90 Sr ranging from 0.1 to 1.2 and the 239+240 Pu/ 137 Cs ratio ranging from 0.10 to 0.47. In contrast, the sand filter backwash solids 239+240 Pu/ 90 Sr ratio was 10.6 and the 239+240 Pu/ 137 Cs ratio was 2.0. The ratio differences indicate a relative enhancement of the Pu concentration in the sand filter solids relative to the 137 Cs and 90 Sr sludge concentrations currently in the K West Basin. A dose-to-curie radioisotope evaluation of the sand filter waste form may need to consider this dissimilarity.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Sintering effects on NpO 2 grain size and morphology: The role of precursor Np phase

Neptunium dioxide (NpO 2 ) is a key phase in nuclear material processing as a target material for the production of plutonium-238 ( 238 Pu) and has been historically synthesized via the calcination of a Np oxalate precursor. Alternative synthesis methods for NpO 2 are now more prevalent, necessitating their study and comparison with the more common oxalate route. The purpose of this work was to investigate the microstructural properties of NpO 2 synthesized from a nitrate-based Np precursor phase via the assessment of NpO 2 particle size and morphology as a function of calcination temperature and time. Scanning electron microscopy (SEM) was used to probe the primary grain size and morphology of NpO 2 after calcination at temperatures ranging from 700 to 1100 °C and hold times ranging from 1 to 10 h. Post-image analysis using ImageJ software enabled the quantification of mean particle diameter. This analysis indicated that particle diameter increases with both increasing calcination temperature and hold time. Primary particles were shown to be clumped in irregular patterns into the overall rough, blocky aggregates, but this macroscopic morphology was not affected by calcination time or temperature. Although trends in primary grain size of NpO 2 were consistent with available literature from other Np precursor phases, the macroscopic morphology of the NpO 2 aggregates was quite different than reported for other precursors. Through comparison with historical literature on Np oxalate, this work emphasizes the importance of Np precursor on the physical properties of NpO 2 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗