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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 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

fuel cycle cost analysis of Th/U MOX fuel in a block-type HTR - 18619

With a rapid development of nuclear energy in these years, the supply of nuclear fuel is becoming more and more important. Thorium regains attention because of its abundant reserves and excellent physical properties. As a next generation reactor, the high temperature reactor (HTR) is a promising choice for thorium utilization because of its deeper burnup, harder spectrum and better inherent safety. Therefore, a 600 MWth 108-column modular high temperature reactor (MHR) core model with 20%UO{sub 2}-80%ThO{sub 2}; MOX fuel is adapted. A traditional two-step calculation scheme, lattice calculation then core calculation, is used to get the neutronic performance of equilibrium cycle including uranium enrichment and mass of fuel. Using these data, the fuel cycle cost of HTR can be calculated by the levelized lifetime cost method. In order to take into account the variation of unit price, a well-established, standard statistical analysis technique is used by Monte Carlo simulation, (thus, the probability distribution of fuel cycle cost can be obtained. Furthermore, sensitivity calculations are made to analyze the impact on the fuel cycle cost of variations in the unit prices for each fuel cycle component. The result shows that the most influential parameters are unit price of natural uranium, enrichment and discount rate. Finally a comparison of fuel cycle cost between 100%UO{sub 2} and 20%UO{sub 2}-80%ThO{sub 2} MOX fuel is made. The latter is 11% cheaper than the former. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A Better Method to Calculate Fuel Burnup in Pebble Bed Reactors Using Machine Learning

Burnup measurement is an important step in material control and accountancy (MC&A) at nuclear reactors, and may be done by examining gamma spectra of fuel samples. Traditional approaches rely on known correlations to specific photopeaks (e.g. 137 Cs) and operate via a standard linear regression method. However, the quality of these regression methods is limited even in the best case, and is significantly poorer at short fuel cool-down times, due to the elevated radiation background by short life-time isotopes, and self-shielding effect of the fuel. For practical operation of pebble bed reactors (PBRs), quick measurements (in minutes) and short cooling times (in hours) are required from a safety and security perspective. We investigated the efficacy and performance of machine learning (ML) methods to predict the burnup of the pebble fuel from full gamma spectra (rather than specific discrete photopeaks) and found a full-spectrum ML approach to far outperform baseline regression predictions in all measurement and cooling conditions - including in operational-like measurement conditions. We also performed model and data ablation experiments to determine the relative performance impact of our ML methods' capacity to model data nonlinearities and the inherent additional information in full spectra. Applying our ML methods, we found a number of surprising results, including improved accuracy at shorter fuel cooling times (the opposite of the norm), remarkable robustness to spectrum compression (via rebinning), and competitive burnup predictions even when using background signal only (i.e. explicitly omitting known isotope photopeaks).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

Comparing Ecological Risk between D and D Facilities and Operating Facilities at the Hanford Site - 19531

Many countries are faced with a large remediation task resulting from World War II, the Cold War, and subsequent industrial development. In the U.S. the Department of Energy (DOE) faces the largest hazardous waste remediation tasks, with a commitment to protect human health and the environment during and following remediation. DoE's biggest challenge is remediation of the many remaining cleanup areas at the Hanford Site, Washington. Some Hanford sites have extensive ecological resources and ecosystems that require protection or restoration. We briefly 1) describe a rapid evaluation method for ecological resources on units slated for cleanup that allows comparison among the different areas at the Hanford Site, 2) examine some D4 (deactivation, decommissioning, decontamination and demolition) facilities and Operating facilities that require remediation, 3) describe potential ecological harm to resources on these sites, and 4) compare how risks from potential harm vary at the Hanford Site. The ecological evaluation includes three phases: 1) general steps (identify resource categories, identify cleanup areas), 2) ecological descriptions (describe resources on the site to be cleaned up), and 3) ecological ratings (rate the value of the resources on a scale that can be applied across remediation sites). After completion of these steps for the sites requiring cleanup, a comparison of potential risks can be made among types of facilities (e.g. D4 facilities, legacy waste sites, operating facilities). D4 facilities are those slated for deactivation, decommissioning, decontamination and demolition, while operating facilities at Hanford include waste treatment plants, sludge basins, waste trenches, Central Waste Complex, storage facilities, and disposal facilities, among others required to support the D4 and other remediation activities, before they too are remediated. Currently the potential harm to ecological (and eco-cultural) resources on the Hanford Site for remediation at D4 facilities is very low, but increases to very high (for reactors) during remediation when there is physical disruption, increased traffic and personnel, and possible increased contamination. Following remediation, the potential harm to ecological resources is low, and the value may increase due to restoration of native vegetation on sites that were largely industrial prior to remediation. In contrast, the potential harm to ecological resources at Operating Facilities was higher currently than for D4 facilities, but was low to medium following cleanup. The higher harm (risk) at D4 facilities partly derives from the location of the reactors near the Columbia River, a resource extremely important to Tribal Nations and to other peoples in the Northwestern United States, as well as for the iconic and threatened salmon populations. Methods described in this paper provide Congress, DOE managers, regulators, Tribes, and the general public with assurance that ecological and eco-cultural resources and the environment are being protected during and following remediation. The analysis of specific sites provides managers, regulators and the public with information to make cost-effective decisions about remediation that continues to enhance ecological value while protecting human health. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Upgrading Fast Pyrolysis Liquids

The liquid formed by the fast pyrolysis (FP) of biomass is commonly termed bio-oil. Biomass is converted to bio-oil through FP at 400-550°C typically at 1 atm. During FP, the biomass is held at temperature for a residence time on the order of 1-2 seconds. FP is beneficial from the standpoint that a large portion of the total mass and carbon contained in the biomass is converted and retained in the liquid bio-oil fraction. However, bio-oil is acidic, viscous, and thermally unstable. Bio-oil must be upgraded in order to be considered as a liquid transportation fuel or separated and further treated to produce value-added chemicals. Bio-oil is generally regarded as difficult to upgrade due to its thermal instability and chemical complexity which is derived in part from the incorporation of highly reactive species such as aldehydes and terminal olefins as well as the presence of sugars and anhydrosugar species which undergo secondary polymerization reactions when heated. Upgrading strategies considered in this chapter include methods to stabilize and deoxygenate bio-oil through low temperature hydroprocessing to stabilize bio-oil followed by higher temperature “deep hydrotreating” to remove the majority of oxygen present; converting bio-oil catalytically (often with zeolites such as HZSM-5) prior to condensation with catalysts inside the FP reactor or in a separate fixed or fluidized bed after the FP reactor; catalytic cracking with acid catalysts such as zeolites after condensation; or through physical or chemical separations to improve bio-oil stability, extract valuable chemicals, or produce H¬2. Bio-oil upgrading by co-processing with refinery streams such as vacuum gas oil in a fluid catalytic cracking (FCC) unit in a refinery has also been investigated. Numerous excellent reviews, perspectives and other chapters have been published in the preceding 7 years.[1-10] Reviews specific to individual sections within this chapter are at the beginning of the individual sections. This chapter is intended as an update to the original article authored by Prof. Anthony V. Bridgwater, Thermochemical Processing of Biomass: Conversion into Fuels, Chemicals and Power, First Edition, Edited by Robert C. Brown, © 2011 John Wiley & Sons, Ltd. We gratefully acknowledge the original contribution by Prof. Bridgwater and have kept much of his original discussions intact with updated references. The scope of this updated chapter includes references published since about 2010. In the interest of space many of the original references included in the original version were not included. Thus, Prof. Bridgwater’s original chapter should be viewed as a companion to this updated chapter to provide a holistic set of references to the numerous bio-oil upgrading processes discussed herein.

Albrecht, Karl O.↗
DOE OSTI2023

Fuel Salt Qualification and Associated Fuel Cycle Opportunities & Issues

NRC staff have defined fuel qualification as a process which provides high confidence that the physical and chemical behavior of fuel is sufficiently understood so that it can be adequately modeled for both normal and accident conditions, reflecting the role of the fuel design in the overall safety of the facility. Uncertainties are defined so that calculated fission product releases include the appropriate margins to ensure conservative calculation of radiological dose consequences. The Advisory Committee on Reactor Safeguards (ACRS) has accepted the methodology described in NUREG/CR-7299 as an acceptable method for fuel qualification for molten salt reactors. The fuel salt qualification process described in NUREG/CR-7299 is based upon understanding the role that the fuel salt plays in achieving the fundamental safety functions, which in-turn is based on the fuel salt thermophysical and thermochemical properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Large Cask, Small Building, AREVA's Safe MLLW Disposal Solution Despite Facility Limitations - 16295

In order to begin decommissioning of a 1950's research reactor, Class C Mixed Low Level Waste (MLLW) had to be removed from the facility and disposed of. The AREVA TN RAM cask, designed and licensed for shipping Type B waste packages, was chosen as the most appropriate shipping solution. Due to the building layout, crane capacity and pool size, the standard TN RAM cask loading process could not be utilized. Unique solutions and approaches had to be developed to complete the project. AREVA designed a smaller liner insert that would house the four MLLW components in a single waste package to solve the problems related to crane capacity and physical pool dimensions. AREVA also designed and fabricated a shield bell to provide radiation shielding while moving the liner insert from the reactor pool to the cask that was staged adjacent to the reactor pool. The AREVA team worked with the disposal facility to cover requirements for disposal of Class B MLLW. This included designing spacers filled with inert material to fulfill the requirement for maximum 10% void space in the waste package. This disposal method for Class C waste met the requirements of macro-encapsulation. Therefore, no additional treatment action was required when disposing of Class C MLLW. AREVA's solution provides a model for disposing MLLW requiring macro-encapsulation in a single waste package. This solution also provides a model for how to adapt large shipping casks and liners to smaller aging commercial, DOE or research facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Reactor physics benchmark of Westinghouse PWR core design suite for high burnup/high enrichment fuel - Part 2: 2D and 3D core simulations

Westinghouse has performed a comprehensive set of code-to-code benchmark comparisons to corroborate application of its state-of-the-art NEXUS-based code suite to high burnup and high enrichment (HB/HE) fueled PWR core design and reload analysis. The benchmark consisted of a progression in complexity from modeling single 2D pin cells, through 2D lattices and finally to whole core 2D and then 3D scenarios. The accompanying paper (Part 1) presented the results for a variety of pin cell and lattices representative of high energy cores and established the ability of the MPACT code to accurately match the predictions of the Monte-Carlo code SERPENT2 such that the former may be used as a surrogate reference code in spatially complex whole core modeling applications where using Monte-Carlo methods is somewhat impractical. This paper focuses on comparisons at the whole core level using a representative high power density 3-loop Westinghouse NSSS core operating for 24 month cycles employing high enriched fuel with use of combined burnable absorbers to hold down excess reactivity. The evolution of the core reactivity and power distribution characteristics throughout the cycle are calculated using the Westinghouse NEXUS-based code suite and compared to the results using MPACT. The results of this whole core benchmark confirm the conclusions that the Westinghouse code suite accurately models HB/HE fuel designs, with very good prediction of core reactivity and power distribution throughout the cycle. The industry can be confident that the NEXUS code system can be deployed to support PWR reload safety and operational analyses for cores featuring HB/HE fuel and 24-month cycle operation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Response of thermal and fast-ion transport to beam ion population, rotation and T e / T i in the DIII-D steady state hybrid scenario

The thermal and fast-ion transport properties of DIII-D steady-state hybrid discharges with normalized beta β N ≳ 3 are studied at low injected torque and an increased electron to ion temperature ratio T e / T i . Linear stability analysis performed with the TGLF turbulent code indicates that a high-k mode is usually dominant at smaller radii, whereas a low-k mode is usually dominant at larger radii in these plasmas. A reduction in the net injected torque from 8.6 to 4.3 N-m leads to reduced E × B shear and hence, an enhanced turbulence that was observed on the Doppler backscattering diagnostic and was also computed with TGLF. As T e / T i in the core was increased from 0.57 to 0.66 by adding electron cyclotron current drive (ECCD) to these plasmas, higher levels of transport are observed with increased high-k modes indicated by TGLF. The fast-ion transport level varied over an order of magnitude in these discharges depending on whether Alfvén eigenmodes, fishbones, or no instabilities were observed. Hybrid plasmas with fishbones have decreased fast-ion transport, compared to plasmas with Alfvén eigenmodes, since they are resonant with a smaller portion of phase space and their resonance is farther from the wall. This reduction in fast-ion transport with ECCD mitigates the increase in turbulent transport, resulting in higher performance than expected during strong electron heating. Similarly, the lowest fast-ion transport was observed in the low torque plasma, which also led to better than expected performance at this torque value. Finally, the thermal and fast-ion transport changes observed as the torque/rotation and T e / T i are varied indicate possible methods for transferring this scenario to a reactor.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Review of capture cross sections relevant for intentional nuclear forensics

The NA-22 Intentional Forensics Venture is developing a system for tagging nuclear fuel using various methods of information encoding. One of the main methods under development is the insertion of isotopically enriched tracers into the fuel. In order to aid in the understanding of the neutronic performance of these taggants, we assess the quality of the nuclear data underpinning simulations, which are driven by the neutron-capture cross sections. We present these cross sections of naturally occurring isotopes of the elements provided in the neutron sublibrary of the planned ENDF/B-VIII.1 Feb. 2023 library release. We make this assessment using a rubric designed for this effort, which quantifies orthogonal features related to the overall quality. The quality metric highlights 6 aspects: experimental data, resonance evaluations, integral metrics, covariances, fission products, and documentation. We focus on energy ranges relevant for reactor applications. We also discuss additional sources for new, high-quality cross-section data that may be utilized on the time scale of the venture, including existing global data, new experiments, and computational methods. Finally, overall outlook is presented with conclusions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Development of spectroscopy apparatus for molten salt reactor safeguards and material characterization

A molten salt reactor (MSR) produces nuclear fission energy using mixture of fissile and non-fissile salts. Salt in the reactor functions as the nuclear material and the coolant allowing for a safer design. International nuclear safeguards is the monitoring of nuclear reactors, nuclear fuel, and irradiated material to prevent the proliferation of nuclear weapon grade material in hostile foreign nations. The mixture of salt creates a proliferation risk as it generates a conglomeration of fuel, daughter nuclei, and corrosive material. A method to monitor and measure the materials present in the reactor is essential to the scalable deployment of MSRs. The objective of this project was to design a spectroscopy apparatus for aqueous salt so it can be measured and collected consistently in the aerosol phase.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

Accelerate Nuclear Research and Development by Reducing Time and Cost Spend in the Pre-conceptual Design Phase of Advanced Reactor Experiments

The design process of every new concept, such as advanced nuclear reactors or associated experiments, starts with the pre-conceptual design phase. In this phase, the viability of a wide range of design options needs to be assessed quickly, to understand the operating envelope and its feasibility. A variety of physics models (thermal-hydraulics, neutronics, mechanical design, etc.) has to be considered at this very first design stage and optimum component sizes and materials (e.g. heat exchangers, piping, turbomachinery, coolant type, etc.) have to be chosen for a given set of boundary conditions (e.g. heat source, heat sink, flow rate, etc.). Detailed solutions such as provided by high fidelity methods like computational fluid dynamics (CFD), Monte Carlo methods, etc. and even lower fidelity tools such as system or subchannel codes, etc. are usually not used during the pre-conceptual design due to the relatively long time needed to create input models, the computational time to obtain a solution and the lack of flexibility to quickly investigate different combinations of components, individual component sizes and material properties. High fidelity tools are usually only employed in the conceptual design and later phases once a base concept has been identified during the pre-conceptual design stage. The current practice during the pre-conceptual design stage is that analysts collect the needed equations, material properties, closure laws, etc. and create ad-hoc solutions form scratch for every new problem. There clearly is a lack of a flexible scoping tool that can be used during pre-conceptional design before higher fidelity tools (as described above) come into play. To reduce user errors in ad-hoc solutions and increase fidelity and efficiency, this project aims to investigate and develop a user-friendly scoping tool to address the thermal-hydraulic designing needs during preconceptual experiment design, i.e. Thermal-hydraulic Research Universal Scoping Tool (TRUST). The success of TRUST will provide the nuclear engineers with an easy-to-use and affordable calculator for early reactor system design and optimization.

42 ENGINEERING↗
DOE OSTI2026

Ceramic Composite Experimental Testing Status

Over recent years, ceramic matrix materials such as SiC–SiC and C–C have been gaining interest for use in fusion reactors, light water reactors (LWRs), and high-temperature reactors (HTRs). These materials are good candidates to operate in very high temperature and moderate to high radiation environments. The evaluation of composite materials, in general, is challenging because of variations in precursor materials, variations in the fabrication process across fabricators, and the wide range of potential fiber architectures, to name a few. However, the need to evaluate neutron-irradiated properties adds another layer of complexity, which includes cost, timeline, and specimen size limitations (often associated with irradiation testing). A qualification methodology for the use of ceramic composites is provided in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III-5-HHB. The methodology is supported by ASTM International (ASTM) guides, which provide a pathway to accomplish this effort. Part of the qualification strategy is for the designer to collect material property data on environmental conditions representative of its design envelope. These data include irradiation effects. This report presents an experimental study and test campaign developed to partially address this gap by providing initial mechanical and physical property data required for design. A variety of different materials using different manufacturing techniques are considered as part of this campaign. The test plan suggests performing a screening or partial irradiation study to assist the designer during the material selection process. The designer can then perform a more comprehensive qualification study if the material performance is promising. This work focuses on the status of the specimen preparations (machining of samples), the current test methods and failure analysis as well as the preparation of irradiation vehicles for the irradiation campaign. The irradiation will be performed at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR) and at Idaho National Laboratory (INL) in the Advanced Test Reactor (ATR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Polarized fusion and potential in situ tests of fuel polarization survival in a tokamak plasma

Abstract The use of spin-polarized fusion fuels would provide a significant boost towards the ignition of a burning plasma. The cross section for D + T → α + n, would be increased by 1.5 if the fuels were injected with parallel polarization. Furthermore, our simulations demonstrate additional non-linear power gains in large-scale machines such as ITER, due to increased alpha heating. Such benefits require the survival of spin polarizations for periods comparable to the particle confinement time. During the 1980s, calculations predicted that polarizations could survive a plasma environment, although concerns persisted regarding the cumulative impacts of wall recycling. In that era, technical challenges prevented direct tests and left the large scale fueling of a power reactor beyond reach. Over the last decades, this situation has changed dramatically. Detailed simulations of ITER have predicted negligible wall recycling in a high-power reactor, and recent advances in laser-driven sources project the capability of producing large quantities of ∼100% polarized D and T. The remaining crucial step is an in-situ demonstration of polarization survival in a plasma. For this, we outline a measurement strategy using the isospin-mirror reaction, D + 3 He → α + p. Polarized 3 He avoids the complexities of handling tritium, while encompassing the same spin-physics. We evaluate two methods of delivering deuterium, using dynamically polarized Lithium-Deuteride (with vector polarization P V D of 70%) or frozen-spin Hydrogen-Deuteride (with P V D of 40%), together with a method of injecting optically-pumped 3 He (with 65% polarization). Pellets of these materials all have long polarization decay times (∼6 min for LiD at 2 K, ∼2 months for HD at 2 K, and ∼3 d for 3 He at 77 K), all far greater than a plasma shot in a research tokamak such as DIII-D (∼20 s). Both species can be propelled from a single cryogenic injection gun. We review plasma requirements and strategies for detecting polarization survival. Polarization alters both fusion yields and the angular distribution of fusion products, and each of these provides a potential signal. In this paper we simulate a selection of shots with similar characteristics in a future high-T ion H plasma, and find ratios of yields from shots with fuel spins parallel and antiparallel reaching 1.3 (HD + 3 He) to 1.6 (LiD + 3 He) over a wide range of poloidal angles. (A companion paper finds sensitivity to fusion product angular distributions as reflected in the pitch angles of protons and alphas reaching the plasma facing wall.)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Decayheatml

This code is designed to predict and analyze the decay heat generated in molten salt reactors (MSRs) using a hybrid approach that combines machine learning and segmented polynomial fitting. The accurate prediction of decay heat is essential for reactor safety and the optimization of spent fuel storage. The code operates through several key components: 1) Data Architecture: It incorporates a modular data architecture that handles various MSR-specific operational parameters such as power density, humidity content, and air ingress. These parameters are sampled using Sobol sequences to ensure comprehensive coverage of operational uncertainties. 2) Machine Learning Framework: The code employs a diverse set of machine learning models, including polynomial regression, decision trees, random forests, gradient boosting, support vector regression, k-nearest neighbors, multi-layer perceptrons, and symbolic regression. These models are trained to predict decay heat over a wide temporal range, from immediate shutdown up to 10,000 years. 3) Region-Optimized Training: The temporal domain is divided into multiple regions, each modeled separately to capture distinct decay heat characteristics across different time scales. This approach significantly improves the accuracy and interpretability of predictions. 4) Segmented Polynomial Interpretation (SPI): The SPI method translates machine learning predictions into piecewise polynomial equations. These equations are physically interpretable and can be directly integrated into existing engineering workflows and safety analyses. 5) Front-End Interfaces: The code includes both a Jupyter notebook interface for research development and a Streamlit web application for operational deployment. These interfaces allow users to interactively explore decay heat predictions, adjust operational parameters, and visualize results in real-time. 6) Applications: The framework supports various applications, including safety system validation and spent fuel container optimization. It enables real-time evaluation of worst-case decay heat scenarios, informing the design of passive safety systems and optimizing container designs for long-term storage. Overall, this code provides a robust, accurate, and user-friendly tool for predicting decay heat in MSRs, enhancing reactor safety, and optimizing spent fuel management.

Retamales, Mauricio Eduardo Tano [Idaho National L↗
DOE OSTI2022

Demonstration of photoreactor platform for on-sun unassisted photoelectrochemical hydrogen generation with tandem III–V photoelectrodes

Photoelectrochemical (PEC) water splitting is a direct solar-driven technology that converts solar energy to storable chemical energy in the form of hydrogen. In addition to semiconductor and catalyst development for improved photoelectrodes, designing reactors for testing and operation of PEC systems under inherently dynamic outdoor solar illumination is necessary to further the commercial viability of PEC technology. Herein, we present a versatile photoreactor system mounted on a solar tracker for outdoor PEC testing and demonstrate unassisted PEC water splitting under real world on-sun conditions. GaInP 2 /GaAs tandem absorber photoelectrodes with a MoS 2 catalyst were fabricated and exhibit >8% solar-to-hydrogen efficiency. On-sun efficiency and stability of the photoelectrodes were characterized on both sunny and partly cloudy days with continuous monitoring of insolation and weather conditions. Furthermore, the versatile photoreactor and outdoor PEC testing capabilities and methods presented here can accelerate the development of other solar fuel generating systems and technologies.

solar fuels↗
DOE OSTI2020

DOE-NE LWRS Integrated Program Plan - Physical Security Pathway

Domestic nuclear power is facing increased financial pressures from a variety of areas and there is pressure on these utilities to reduce their cost of operation. Currently, about 20%-30% of all on-site personnel are related to physical security. The LWRS Program recognized that R&D related to physical security could play a role in providing nuclear utilities technical and staffing efficiency options to meet their physical security commitments, but utilities often lack the technical basis or the ability to create the technical basis to realize or implement these efficiencies; towards this end, the LWRS Program created the Physical Security Pathway in September 2019. The pathway performs R&D to develop methods, tools, and technologies to optimize and modernize a nuclear power facility’s security posture. The pathway will: (1) conduct research on risk-informed techniques for physical security that account for a dynamic adversary; (2) apply advanced modeling and simulation tools to better inform physical-security scenarios and reduce uncertainties in force-on-force modeling; (3) assess benefits from proposed enhancements and novel mitigation strategies and explore changes to best practices, guides, or regulation to enable modernization; and (4) enhance and provide the technical basis for stakeholders to employ new security methods, tools, and technologies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Treatment of Nonuniform Temperature Distribution by Subgroup Method

The subgroup method divides an energy group into several subgroups according to the magnitude of the cross section. The subgroup fixed source equation is formulated on the energy range of a subgroup, the ratio of which to the energy range of an energy group is defined as subgroup probability. Typically the physical probability table or the mathematical probability table is applied. However, both of the tables are not guaranteed to share the same set of subgroup probabilities for a resonant nuclide at different temperatures. This means that the subgroup fixed source equation cannot be formulated and the subgroup method will not work when the temperature distribution is nonuniform. To overcome this defect, the subgroup method based on partial cross section fit scheme (PXSFS) and the simplified partial cross section fit scheme (SPXSFS) are proposed. These methods fit the cross sections at different temperatures as partial cross sections to share a same set of subgroup probabilities. The new methods is compared to the conventional subgroup method (CSM) based on physical probability table and the preexisting methods: the correlation model (CM), the subgroup level adjustment scheme (SLAS) and the number density adjustment scheme (NDAS). The numerical results show that the new methods can better predict the spatial dependent reaction rate than preexisting methods. Within these two methods, the simplified scheme consumes less time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Ab initio generation of thermal neutron scattering law for uranium dioxide

Neutronic analysis of moderated nuclear fission systems, e.g. nuclear reactors, relies upon an accurate prediction of the thermal neutron spectrum. Thermal neutrons have wavelengths on the order of the interatomic spacing between atoms in matter. Consequently, the scattering cross-sections of these neutrons are affected by the scattering law of the material (i.e. S(α,β), where α and β are dimensionless momentum and energy transfer variables), which describes the atomic structure and thermodynamic motion of the atomic nuclei. Uranium-dioxide (UO{sub 2}) is the primary fuel source for nuclear power reactors moderated by light water. In such systems (e.g. nuclear power reactors, spent fuel pools, etc.) the nuclear fuel comprises a large fraction of the system volume, approximately 1/3 in light-water reactors. Therefore, the scattering law of the fuel material may impact the thermal neutron spectrum and criticality of these systems. In this work previously established ab initio lattice dynamics (AILD) methods were utilized to calculate the phonon density of states (DOS), which acts as a fundamental input to the scattering law. Subsequently, the S(α,β) of UO{sub 2} was generated for publication as an ENDF thermal scattering library (TSL), which will be available for use in standard neutronics computer applications. (authors)

07 ISOTOPE AND RADIATION SOURCES↗