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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.

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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Evaluation of BISON's Transient Fission Gas Release Model on High-Burnup UO 2 MiniFuel

BISON's fission gas release (FGR) model has been evaluated and tested over several hypothetical thermal transient cases at high burnup and validated against FGR experimental data from an annealing test at high burnup. Diffusion-controlled FGR is accounted for in the model by including several physical mechanisms such as the diffusion of fission gas to grain faces, grain boundary sweeping, grain growth, and intergranular bubble growth. Under temperature variations, FGR is increased due to the development of micro-cracks at grain faces. This existing transient capability, however, was developed based on transient fuel behavior during power ramps and the associated micro-cracking process. Nevertheless, a fraction of the FGR during temperature transients such as LOCA can ensue from different mechanisms (in particular, fuel fragmentation) and present an important dependency on local burnup and high burnup structure formation. It follows that the performance and necessary developments of the existing BISON fission gas model for the analysis of LOCA-type transients need to be investigated through a systematic assessment study. Such a study is performed in this work. The model shows an overall good qualitative agreement with experimental observations on FGR during annealing tests with different temperature conditions. It also accounts well for microstructural effects on FGR. When quantitatively compared to FGR data from previously irradiated 103 MWd/kgU UO 2 discs under thermal annealing, the model shows a less satisfactory agreement with the experimental data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

The iron oxidation state of Ryugu samples

The Hayabusa2 mission sampled Ryugu, an asteroid that did not suffer extensive thermal metamorphism, and returned rocks to the Earth with no significant air exposure. It therefore offers a unique opportunity to study the redox state of carbonaceous Cb-type asteroids and evaluate the overall redox state of the most primitive rocks of the solar system. An analytical framework was developed to investigate the iron mineralogy and valence state in extraterrestrial material at the micron scale by combining x-ray diffraction, conventional Mössbauer (MS), and nuclear forward scattering (NFS) spectroscopies. An array of standard minerals was analyzed and cross-calibrated between MS and NFS. Then, MS and NFS spectra on three Ryugu grains were collected at the bulk and the micron scales. In Ryugu samples, iron is essentially accommodated in magnetite, clay minerals (serpentine–smectite), and sulfides. Only a single set of Mössbauer parameters was necessary to account for the entire variability observed in MS and NFS spectra, at all spatial scales investigated. These parameters therefore make up a fully consistent iron mineralogical model for the Ryugu samples. As far as MS and NFS spectroscopies are concerned, Ryugu grains are overall similar to each other and share most of their mineralogical features with CI-type chondrites. In detail however, no ferrihydrite is found in Ryugu particles even at the very sensitive scale of Mössbauer spectroscopy. The typical Fe 3+ /Fe tot of clay minerals is much lower than typical redox ratios measured in CI chondrites (Fe 3+ /Fe tot = 85%–90%). Furthermore, magnetite from Ryugu is stoichiometric with no significant maghemite component, whereas up to 12% of maghemite was previously identified in the Orgueil's so-called magnetite. These differences suggest that most CI meteorites suffered terrestrial alteration and that the preterrestrial composition of these carbon-rich samples was less oxidized than previously measured. However, it is not clear yet whether or not the parent bodies of CI chondrites were as reduced as Ryugu. Finally, the high spatial resolution of NFS allows to disentangle the redox state and the crystal chemistry of iron accommodated in serpentine and smectite. The most likely polytype of serpentine is lizardite, containing <35% of Fe 3+ , a fraction of which being tetrahedrally coordinated. Smectite is more oxidized (Fe 3+ /Fe tot > 65%) and mainly contains octahedral ferric iron. In conclusion, this finding implies that these clays formed from highly alkaline fluids and the spatial variability highlighted here may suggest a temporal evolution or a spatial variability of the nature of this fluid.

58 GEOSCIENCES↗
DOE OSTI2026

r -process Nucleosynthesis from Hyperaccreting Neutron Stars in Common Envelopes

We investigate nuclear reactions and feedback in hyperaccreting neutron star environments, considering accretion rates in the range 0.3–3 × 10 4 M ⊙ yr −1 , typical of short-period compact-object binaries in common envelopes. Our models account for weak reactions, neutrino energy loss, nuclear energy release, pair production, degenerate equations of state, and general relativistic hydrodynamics. Depending on the accretion rates, these systems can develop both proton- and neutron-rich atmospheres with strong convective instabilities linking the neutrino sphere to the outgoing accretion shock inside the radiation trapping zone. Convection drives nucleons through multiple heating and cooling cycles, with photodisintegration dominating during the heating phase and heavy element synthesis during the cooling phase, ejecting material with abundances that depend on the accretion rate and depth of the final decompression trajectory. The turbulent nature of convective currents is conducive to creating a wide range of nuclear products through a variety of effects, including nuclear statistical equilibrium freeze-out and the r-, p-, and γ-processes. We also observe a novel multistep process in reheated trajectories, consisting of proton-capture and photodissociation reactions operating on r-process seeds, producing overall neutron-deficient isotopes. A significant amount of infalling gas experiences high entropy and short (millisecond) freeze-out timescales capable of making r-process elements with high overabundances through a disequilibrium effect between neutrons and α-particles that does not require an excess of neutrons.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2022

Calculating Radiation Damage (DPA) from Transmutation Products

This is a poster for an INL poster session. Accurate models for radiation damage are crucial for predicting material performance in radiation environments. The uncertainty of state-of-the-art radiation damage models is large, contributing to excessive safety margins. A major source of this uncertainty is neglecting the effect that transmutation products have on radiation damage. Transmutation products are new nuclides formed by neutron activation during irradiation; they can contribute to radiation damage by additional neutron capture or decay events. Ignoring the contribution of transmutation products leads to a significant underprediction of the radiation damage (e.g., >10% error in 316 stainless steel). This underprediction is accounted for in part by adding larger safety margins to designs. Currently, the state of the art explicitly accounts for only a single transmutation product, namely nickel-59, during the radiation damage calculation. All other transmutation products are assumed to not contribute to the radiation damage, because there is currently no established method to systematically track all or a selection of radiation damage contributions of transmutation products during activation. In the case of nickel-59, the current method is to apply a precalculated correlation that cannot be used for any other nuclide and is largely dependent on all nuclear engineers being experts in this niche topic. This project proposed to methodically find other transmutation products that cause significant radiation damage, and then to develop a general framework for systematically tracking the radiation damage from these nuclides. This was accomplished by combining the radiation damage calculation into the transmutation calculation already performed for irradiated structural materials. The key idea of our framework is to introduce radiation-damage "pseudo-nuclides" to the list of nuclides used in the transmutation analysis. This allows radiation damage to be tracked alongside the creation and destruction of transmutation products. The main deliverable of this project is a general framework for computing radiation damage while the damaged material undergoes transmutation; this capability allows a significantly more accurate estimation of radiation damage, and in turn reduce required safety margins thereby reducing the cost to construct reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

First-Scatter Source Generation in GADRAS

A new scatter calculation algorithm has been implemented in the Gamma Detector Response and Analysis Software (GADRAS) package that accounts for spectral effects of scattering materials not in the line of sight of the detector and the source. Previously, GADRAS would only apply scattering effects due to materials that fall between the source and detector. This new routine will allow better modeling of various scenarios including gamma imagers, collimated detectors, or traditional gamma detectors where scattering materials are present.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Advanced Fuel Cycle Cost Basis Report: Supporting Documents 2 Production Based Costing

The purpose of this section is to outline a method of cost analysis whereby a significantly better representation of “should achieve” costs may be attained for NOAK systems. Based on economists’ notions of producer theory, and grounded in the cost analysts’ and project managers’ tool called the Work Breakdown Structure (WBS), this section describes best practices in cost estimation. It then illustrates how the Code of Accounts (COA) structure, developed by the Economic Modeling Working Group (EMWG) of the Generation IV International Forum in “Cost Estimating Guidelines for Generation IV Nuclear Energy Systems [EMWG 2007] (hereafter “Gen IV Guidelines Document”), can be used to differentiate “should achieve” versus “did experience” costs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Quantifying the effect of hydride microstructure on zirconium alloys embrittlement using image analysis

Since hydride microstructure can significantly influence Zr alloy nuclear fuel cladding’s ductility, a new metric has been developed to quantify hydride microstructure in 2D micrographs and relate it to crack propagation. As cladding failure usually results from a hoop stress, this new metric, called the Radial Hydride Continuous Path (RHCP), is based on quantifying the continuity of brittle hydride particles along the radial direction of the cladding tube. Compared to previous metrics, this approach more closely relates to the propensity of a crack to propagate radially through the cladding tube thickness. The RHCP takes into account hydride length, orientation, and connectivity to choose the optimal path for crack propagation through the cladding thickness. The RHCP can therefore be more closely linked to hydride embrittlement of the Zr alloy material, thus creating a relationship between material structure, properties, and performance. The new definition, along with previously proposed metrics such as the Radial Hydride Fraction (RHF), the Hydride Continuity Coefficient (HCC), and the Radial Hydride Continuity Factor (RHCF), have been implemented and automated in MATLAB. These codes are made available with this publication as supplementary materials. These metrics were verified by comparing their predictions of hydride morphology against expected values in simple cases, and the implementation of the new metric was validated by comparing its predictions with manual measurements of hydride microstructure performed on ImageJ. The RHCP was also validated against experimental measurements of fracture behavior and it was shown to correlate with cladding failure better than previous metrics. The information provided by these metrics will help accurately assess cladding integrity during operation, transportation, and storage.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Optical modeling of a radioisotope atmospheric chamber

Abstract Volatile radioisotopes represent a substantial health risk when released into the environment. To better understand the environmental fate of radioisotopes, the authors constructed a cylindrical steel-walled chamber to simulate the atmospheric processing of volatile radioactive gases. Optical modeling was performed for the properties of simulated sunlight in the chamber to better characterize atmospheric reaction studies. Optical simulations were performed using two wall materials (steel and thin-film silica) and validated against experimental measurements. This optical analysis methodology can be used to improve the fidelity of atmospheric models by accounting for optical inhomogeneities enabling a firmer grasp of radioisotopes’ environmental fate. Graphical abstract

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

NA-22 Write for Consortium Efforts

We are asked to look into the improvements for the measurement of Pu of a variety of material types (MT): MT41 (Pu-242, 20 – 60% enriched), MT42 (Pu-242 > 60% enrichment), and MT-82 (Np-237). Many of these are legacy items that were packaged years ago (Fig 1&2), and when the items are re-packaged for long-term storage, they must be re-measured using either NDA (nondestructive analysis methods or radiochemical methods. It is of interest to improve the NDA and radiochemistry to reduce the un-certainty in the measurements, as in the case of larger items (250 g to 2 kg items) the uncertainty in measurements makes it difficult to certify the mass of the material down to the 0.5 g accountability requirement per DOE orders by only doing NDA measurement. It would be of interest to look into methods to improve the detection of these materials, which would not only benefit the weapon manufacturing community, but also has the ability to significantly improve the non-proliferation community.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Comparison of Temperature-Dependent Cross Section Treatments Within the Shift Monte Carlo Radiation Transport Code

Multiple calculation methods have been developed to account for Doppler broadening of cross sections in Monte Carlo radiation transport simulations. This work compares the accuracy and performance of four different broadening methods, including an on-the-fly multipole treatment, applied to three different geometries of increasing complexity, including a nuclear reactor assembly with a realistic temperature profile. The comparisons reveal how the combination of physics implementations in the Shift Monte Carlo code result in slightly different temperature-dependent behavior, and different levels of performance, for a variety of materials and situations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Spent fuel nondestructive assay integrated characterization from active neutron, passive neutron, and passive gamma

Spent nuclear fuel comprises a wide range of irradiated isotopic material compositions, and characterization through nondestructive measurements is beneficial in verifying declared parameters before the fuel is placed in storage, final disposal, and/or reprocessed. This paper discusses results from three nondestructive assay instruments, including passive gamma, passive neutron, and active neutron methods, that measured fifty spent fuel assemblies at the Clab interim storage facility in Sweden. Here, integrated analysis of the measurements from the three different instruments allowed parametric assessments of cooling time, burnup, neutron multiplication, fissile mass, initial enrichment, and decay heat of each individual fuel assembly. Passive gamma measurements were found to be the most beneficial in predicting cooling time, passive neutron for determining burnup, active neutron in estimating initial enrichment, and both passive and total neutron for multiplication correlations. Fissile mass was best estimated using any combination of any two of the instruments such that corrections for isotopic changes in the fuel could be accounted for with the first set of measurements and multiplication of the assembly in the second. In conclusion, the nondestructive assay technologies demonstrated through this effort enhance the characterization of spent nuclear fuel assemblies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

GIF Supercritical Water Cooled Reactor: Proliferation Resistance and Physical Protection White Paper

This report is part of a series of six white papers, prepared jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) and the six System Steering Committees (SSCs) and provisional System Steering Committees (pSSCs). This publication is an update to a similar series published in 2011 presenting the status of Proliferation Resistance & Physical Protection (PR&PP) characteristics for each of the six systems selected by the Generation IV International Forum (GIF) for further research and development, namely: the Sodium-cooled fast Reactor (SFR), the Very high temperature reactor (VHTR), the gas-cooled fast reactor (GFR), the Molten salt reactor (MSR) and the Supercritical water–cooled reactor (SCWR). This white paper represents the status of Proliferation Resistance and Physical Protection (PR&PP) characteristics for the Supercritical Water-cooled Fast reactor (SCFR) reference designs selected by the Generation IV International Forum (GIF) SCWR System Steering Committee (SSC). The intent is to generate preliminary information about the PR&PP features of the SCWR reactor technology and to provide insights for optimizing their PR&PP performance for the benefit of SCWR system designers. It updates the SCWR analysis published in the 2011 report “Proliferation Resistance and Physical Protection of the Six Generation IV Nuclear Energy Systems”, prepared Jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) and the System Steering Committees and provisional System Steering Committees of the Generation IV International Forum, taking into account the evolution of both the systems, the GIF R&D activities, and an increased understanding of the PR&PP features. The white paper, prepared jointly by the GIF PRPPWG and the GIF SCWR SSC, follows the highlevel paradigm of the GIF PR&PP Evaluation Methodology to investigate the PR&PP features of the eight proposed GIF SCWR designs. Two small modular reactors, the Canadian SSR and the Canada/China/Europe ECC-SMART are also mentioned. An overview of the fuel cycles for the GIF designs are provided. For PR, the document analyses and discusses the proliferation resistance aspects in terms of robustness against State-based threats associated with diversion of materials, misuse of facilities, breakout scenarios, and production in clandestine facilities. Similarly, for PP, the document discusses the robustness against theft of material and sabotage by non-State actors. The document follows a common template adopted by all the white papers in the updated series.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Considerations for Defining G-Values for Aluminum-Clad Spent Nuclear Fuel

Sealed-canister dry storage of aluminum-clad spent nuclear fuel (ASNF) generated by research reactors is an alternative to current storage and disposition pathways as directed by the U.S. Department of Energy. The major challenge faced for this storage approach is radiolytic H 2 generation, including from the aluminum (oxy)hydroxide layers on the surface of ASNF. Experimental and modeling activities have been carried out to characterize the radiolytic yield as part of a DOE-sponsored research program to develop the technical basis for ASNF dry storage. The G-value is a commonly way to report results of radiolysis testing and is defined as the radiolytic yield of a species (e.g. molecular hydrogen) per unit radiation energy deposited into the material system. An independent technical review of the ASNF dry storage technical basis performed by Pacific Northwest National Laboratory raised questions about differences in G-value definitions used for experiments on ASNF surrogates consisting of aluminum samples with adherent (oxy)hydroxides compared to G-values reported in prior literature and how the magnitudes compared between different studies. Material systems resembling ASNF pose complications for measuring/defining G-values to predict the evolution of H 2 in a sealed canister, including i) accounting for radiolytic yields potentially arising from multiple sources, i.e., residual free (vapor), physisorbed, and chemisorbed/chemically bound waters; ii) deciding what portions of the multi-material system to include in the absorbed energy (radiation dose) calculation, considering possible energy exchange between materials as well as measurement limitations, and iii) capturing variations in G-value associated with non-linear yield vs. dose curves and/or dependence on the cover gas. This report summarizes previous literature information on radiolytic H 2 generation and associated G-values from mixed-material systems (generally oxides in contact with water or organic compounds) and from (oxy)hydroxides/hydrates to compare with the definitions and values for ASNF surrogate samples containing adherent aluminum (oxy)hydroxides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Application of BISON to UO 2 MiniFuel fission gas release analysis

There has been a recent push to accelerate fuel qualification by developing revolutionary capabilities to reduce irradiation periods, and thereby, reduce the time required to qualify a new fuel system. One such capability is the MiniFuel irradiation capsule designed to miniaturize fuel samples and irradiate “mini” fuel samples under isothermal temperature conditions. MiniFuel allows steady-state irradiations to decouple the traditionally coupled fission rate (i.e., power) and temperature parameters to understand and generate microstructures observed in fuel operated in a commercial reactor. Furthermore, this process offers the possibility to gather in situ data as well as postirradiation or transient data such as thermal conductivity, specific heat, fission gas diffusion and release, etc. However, accelerating fuel qualification is not solely reliant on generating large amounts of data but also on developing an informed test matrix designed to rapidly generate impactful data. Additionally, this process is reliant on fuel performance codes, such as BISON, to evaluate MiniFuel irradiations using existing material models. This process pinpoints model/data gaps, identifies desired irradiation conditions, and subsequently supports model validation and development. This work describes the use of BISON to perform a number of sensitivity studies designed to understand conditions that lead to fission gas release (FGR) under steady-state isothermal irradiation conditions and temperature transient conditions. The model is applied to a UO 2 MiniFuel example and shows an overall good qualitative agreement with experimental FGR annealing tests under different temperature conditions. It also accounts well for microstructural effects on FGR. When quantitatively compared with FGR data from previously irradiated 103 MWd/kgU UO 2 discs under thermal annealing, the model shows a less satisfactory agreement with the experimental data. Finally, a UO 2 MiniFuel test matrix is proposed to help to extend the model's operational range and validate the new FGR model capabilities to higher burnups and transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Evaluation of Heat Exchanger Technology for Integrated Energy System: Methodology and Preliminary Application

This report proposes a novel heat exchanger evaluation method, based on integrated QFD-AHP technique, to support strategic optimal selection of heat exchanger (HX) technology for integrated energy systems. The term “optimal selection” is defined as selecting a product (i.e., HX) that best meets various aspects of end user demand, including both technical and economic requirements. Compared to previous HX evaluation methods, the proposed approach allows decision makers to explicitly take into account the practical demand of customers (i.e., HX end users) which may vary depending on the HX application target. In addition, the proposed evaluation method can be used not only for high-level decisions to select HX type, but also for low-level decisions such as the selection of the key components including HX material or heat transfer fluid. The focus of this study is to describe the new HX evaluation method and to illustrate the overall HX evaluation and selection processes using the three nuclear heat application cases that represent the thermal integration between advanced nuclear reactors and industrial processes. As a preliminary study, three different types of HX, i.e., shell-and-tube heat exchanger, printed circuit heat exchanger, and heat pipe heat exchanger, were compared.

42 ENGINEERING↗
DOE OSTI2021

Material Accountancy Performance Indicator Toolkit

The MAPIT - Material Accountancy Performance Indicator Toolkit includes statistical tests and calculations based on open literature that calculate safeguards metrics. The tool can be used with user supplied datasets or with the several included datasets which are generated from models that are maintained at Sandia National Laboratories. The models are not included with the software. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525. SAND2021-0467 O

Shoman, Nathan↗
DOE OSTI2022

SERPENT 2 and TRIPOLI-4 neutronics benchmarking of the Jules Horowitz reactor

Jules Horowitz Reactor (JHR) is a future center for nuclear material testing in Europe, currently under construction at CEA Cadarache in southern France. It has been modelled extensively with CEA's HORUS3D/N neutron calculation package, that includes a Monte-Carlo neutron transport code TRIPOLI-4. In order to support the future computational analyses on the reactor core physics, VTT's Monte Carlo neutronics code SERPENT 2 has been used to create an alternative model of the JHR, taking advantage of the pre-existing Constructive Solid Geometry (CSG) structures on which the TRIPOLI-4 model was built. This work describes a benchmark, in which two JHR core configurations were analyzed using both of these models. The first configuration is a fresh fuel core at the beginning of life, in which aluminum based dummy-devices are installed in the reactor test positions. The second configuration represents the end-of-cycle equilibrium core with a heterogeneous burnup distribution, in which CALIPSO devices containing cladding material samples are installed at the in-core experimental positions, while the reflector positions are occupied by DEN-REP devices containing UO{sub 2} fuel. Neutron flux spectra and heat deposition are the parameters of interest in this study, due to their high importance in creating the desired experimental irradiation environment. The impact of delayed gamma energy deposition will be studied in detail, since the two models account for it with different methods. This work aims to establish a SERPENT 2 model of JHR, understanding its differences and agreements with TRIPOLI-4, and discussing its potential future applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

GIF Gas Cooled Fast Reactor: Proliferation Resistance and Physical Protection White Paper

This white paper represents the status of Proliferation Resistance and Physical Protection (PR&PP) characteristics for the Gas-cooled Fast reactor (GFR) reference designs selected by the Generation IV International Forum (GIF) GFR System Steering Committee (SSC). The intent is to generate preliminary information about the PR&PP features of the GFR reactor technology and to provide insights for optimizing their PR&PP performance for the benefit of GFR system designers. It updates the GFR analysis published in the 2011 report “Proliferation Resistance and Physical Protection of the Six Generation IV Nuclear Energy Systems”, prepared Jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) and the System Steering Committees and provisional System Steering Committees of the Generation IV International Forum, taking into account the evolution of both the systems, the GIF R&D activities, and an increased understanding of the PR&PP features. The white paper, prepared jointly by the GIF PRPPWG and the GIF GFR SSC, follows the high-level paradigm of the GIF PR&PP Evaluation Methodology to investigate the PR&PP features of the GIF GFR 2400 MWth reference design. The ALLEGRO reactor is also described. The EM2 and HEN MHR reactor are mentioned. An overview of fuel cycle for the GFR reference design and for the ALLEGRO reactor are provided. For PR, the document analyses and discusses the proliferation resistance aspects in terms of robustness against State-based threats associated with diversion of materials, misuse of facilities, breakout scenarios, and production in clandestine facilities. Similarly, for PP, the document discusses the robustness against theft of material and sabotage by non-State actors. The document follows a common template adopted by all the white papers in the updated series.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗