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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

SAM Two-Phase Flow Model Development and Applications for Operational Transients in Advanced Reactors

As advanced nuclear technologies continue to develop, the need for the flexible operation and generation of these advanced reactors becomes necessary to maximize economic potential. As large-scale experiments are not always feasible, modeling and simulations of advanced reactors play a crucial role in design optimization and analysis. The SAM (System Analysis Module) code developed at Argonne National Laboratory is a state-of-the-art system-level thermal-hydraulic code aimed at simulating advanced reactor systems. Recent code developments have implemented two-phase flow modeling using the homogeneous equilibrium model, and a new steam generator component has been developed to utilize the two-phase flow implementation. In addition to verification tests, a load-following simulation was performed to model a realistic load-following transient in a proposed integrated system consisting of a conceptual advanced reactor known as the Advanced Burner Test Reactor (ABTR) and thermal energy storage (TES) tanks. The integrated system model uses two large TES tanks designed for sodium and a model helical coil steam generator to simulate the operational load-following transient. The flow rates of the feedwater and secondary loops are regulated to meet a prescribed steam generator load consistent with the electricity demand over a 24-h period. In conclusion, the results found the ABTR system was able to maintain stable reactor conditions and primary- and secondary-side characteristics over the course of the load-following transient.

Advanced Burner Test Reactor (ABTR)↗
DOE OSTI2020

Oak Ridge National Laboratory Technical Input for the Nuclear Regulatory Commission Review of the 2017 Edition of ASME Section III, Division 5, ‘High Temperature Reactors’

To assist the Nuclear Regulatory Commission in its decision making on endorsement of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code Section III, Division 5 (2017 Edition) for development of advanced non-light water reactors, the following Division 5 portions were reviewed: Article HBB-2000 Material; Article HCB-2000 Material; Article HGB-2000 Material; Mandatory Appendix HBB-I-14 Tables and Figures; and, Nonmandatory Appendix HBB-U Guidelines for Restricted Material Specifications to Improve Performance in Certain Service Applications. In addition to the 2017 Edition, the same parts of the 2019 Edition have also been reviewed as indicated in various sections of the report. This review was conducted by a collaboration of national laboratory and private sector participants with significant industrial experience, including some heavy lifting and deep diving from Clarus Consulting, LLC., all intended to achieve an objective, independent, and practical perspective. The report provides recommendations, descriptions of the evaluation methods, and the source references for the data used. To build confidence required for endorsement of the Code, this review was conducted as a verification and validation of the above Code contents. The objective of verification is to ensure that the Code is free of error – direct or implied; contains the information needed for its use, including proper coverage of the Code-specified materials for the intended application, and completeness and adequacy of references to other portions of the Code. The objective of validation is to authenticate that the Code tabulations and graphs represent design inputs consistent with what are determined using rules and methods specified by the Code. The authentication process used data that were assembled and/or generated independent of Code development, while the methods of analysis followed Code-specified methods where appropriate. The designated portions for this review cover the five alloys codified for high temperature reactor applications in Division 5, i.e. 316 SS, 304 SS, 800H, 2¼Cr-1Mo, and 9Cr-1Mo-V, regarding their general requirements, permitted specifications and design stress intensity values for pressure-retaining applications, deterioration in service, fatigue acceptance test, permissible weld materials, tensile and yield strength, expected minimum stress-to-rupture values (including for Alloy 718), weld stress rupture factors, permissible materials for bolting use, and restricted specifications in certain service applications. Additionally, stress intensity values for bolting materials including 316 SS, 304 SS and alloy 718 were reviewed. Analysis and discussion are also provided on contents outside of these designated Code portions where it was deemed relevant and necessary to develop a technically sound understanding of issues relating to the designated portions. Due to unavailability of sufficient test data on welds during the review period, the weld stress rupture factors in Tables HBB-I-10.14A to E, which cover a total of ten tables for the five alloys welded with twenty-eight different weld metals (some with similar properties), have been deferred to a future review effort. The review identified mainly two types of issues. The first type includes instances where the Code is found factually incomplete or incorrect, such as obsolete materials specifications listings, missing tabulation of stresses for bolting. Changes to the Code are recommended in these cases. The second type of issue includes instances where the Code tabulations and graphs are found to be less conservative than the review analysis results. In these cases, recommendations are made for further review and consideration where the difference in conservatism exceeds 10%, which is our threshold for questioning technical adequacy, meriting a risk assessment by the Nuclear Regulatory Commission and/or reactor designers. It is noted that this effort has been executed using all available data and established methods of analysis, including methods and criteria specified and used by the Code. As such, the findings that are presented in quantitative detail, in a format for convenient comparison with the Code, and with identification of where further review is recommended, should provide a sound technical basis for decisions about quantifying the implications of the reduced design margins and technical adequacy/inadequacy to form a basis for conditioning specific Code tabulation values on endorsement. Recommendations for specific changes to the Code, however, entail design conservatism considerations beyond the scope of this review effort, and are not made in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

RAPID

Parallel computer code for the simulator for dynamics of power systems which has the capability to initiate the system and create different faults for the dynamic analysis. The code is based on time-parallel method (Parareal) with Adaptive Method Reduction (AMR). The coarse solvers for the Parareal algorithm include several Semi Analytical Solution methods. Also, Integrated simulation of coupled transmission and distribution systems can be studied.

Simunovic, Srdjan [Oak Ridge National Lab. (ORNL),↗
DOE OSTI2026

ROSE

Developed at Lawrence Livermore National Laboratory (LLNL), ROSE is an open source compiler infrastructure to build source-to-source program transformation and analysis tools for large-scale C (C89 to C23), C++ (C++98 to C++23), UPC, Fortran (Fortran4, 66, 77, 95, 2003), OpenMP, Java, Python, and Binary applications. ROSE users range from experienced compiler researchers to library and tool developers who may have minimal compiler experience. ROSE is particularly well suited for building custom tools for static analysis, program optimization, arbitrary program transformation, domain-specific optimizations, complex loop optimizations, performance analysis, and cyber-security. ROSE is: A library (and set of associated tools) to quickly and easily apply compiler techniques to one's code in order to improve application performance and developer productivity. A research and development compiler infrastructure for for writing custom source-to-source translators to perform source code transformations, analysis, and optimizations. Is

Pinnow, NathanT [Lawrence Livermore National Labor↗
DOE OSTI2016

Variations in Medicare Reimbursement in Radiation Oncology: An Analysis of the Medicare Provider Utilization and Payment Data Set

Purpose: The purposes of this study were to summarize recently published data on Medicare reimbursement to individual radiation oncologists and to identify the causes of variation in Medicare reimbursement in radiation oncology. Methods and Materials: The Medicare Provider Utilization and Payment Data: Physician and Other Supplier Public Use File (POSPUF), which details nearly all services provided by radiation oncologists in 2012, was used for this study. The data were filtered and analyzed by physician and by billing code. Statistical analysis was performed to identify differences in reimbursements based on sex, rurality, billing of technical services, or location in a certificate of need (CON) state. Results: There were 4135 radiation oncologists who received a total of $1,499,625,803 in payments from Medicare in 2012. Seventy-five percent of radiation oncologists were male. The median reimbursement was $146,453. The code with the highest total reimbursement was 77418 (radiation treatment delivery intensity modulated radiation therapy [IMRT]). The most commonly billed evaluation and management (E/M) code for new visits was 99205 (49%). The most commonly billed E/M code for established visits was 99213 (54%). Forty percent of providers billed none of their new office visits using 99205 (the highest E/M billing code), whereas 34% of providers billed all of their new office visits using 99205. For the 1510 radiation oncologists (37%) who billed technical services, median Medicare reimbursement was $606,008, compared with $93,921 for all other radiation oncologists (P<.001). On multivariate analysis, technical services billing (P<.001), male sex (P<.001), and rural location (P=.007) were predictive of higher Medicare reimbursement. Conclusions: The billing of technical services, with their high capital and labor overhead requirements, limits any comparison in reimbursement between individual radiation oncologists or between radiation oncologists and other specialists. Male sex and rural practice location are independent predictors of higher total Medicare reimbursements.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2019

NEUP Final Report: Analysis of Thermal Transients for a Sodium-CO 2 Heat Exchanger

The design of heat exchangers for use in a supercritical carbon dioxide (sCO 2 ) Brayton cycle power converter must provide for acceptable performance for duty cycle events encompassing anticipated transients and postulated accidents. Argonne National Laboratory participated in the NEUP project, “Thermal Hydraulic and Structural Testing and Modeling of Compact Diffusion-Bonded Heat Exchangers for Supercritical CO 2 Brayton Cycle” with the task of conducting a comprehensive analysis of thermal transients for sCO 2 cycle heat exchangers, with emphasis on the sodium-to-CO 2 heat addition heat exchanger. A range of transients, from normal operation to severe accidents, were simulated with the coupled PDC and SAS4A/SASSYS-1 system level dynamic analysis computer codes. For each transient, the calculated change in the heat exchanger wall temperature is determined as a measure of the thermal loading. In addition, an analysis to find an equivalent helium-CO 2 heat exchanger was carried out to facilitate testing with helium rather than sodium on the hot side.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

ADAPTIVE SURROGATES WITHIN THE RAVEN FRAMEWORK FOR DYNAMIC PROBABILISTIC RISK ASSESSMENT ANALYSIS

The RAVEN code has been under development at the Idaho National Laboratory since 2012. Its main goal is to create a multi-purpose platform for the deploying of all the capabilities needed for Probabilistic Risk Assessment (PRA), Uncertainty Quantification (UQ), data mining analysis and optimization studies. RAVEN has demonstrated a good level of maturity in terms of state-of-art and advanced analysis methodologies. In the past year the RAVEN code has been released as open-source project and is available to download (raven.inl.gov) free of charge. The main subject of this paper is to present a new system within the RAVEN framework to drastically accelerate PRA and UQ analysis using acceleration schemes based on surrogate modeling. This system is based on the heavy usage of surrogate modeling and artificial intelligence algorithms, able to emulate the response of a physical system in closed domains (domains where the surrogate models have been constructed from). The aim of this paper is to show to the community how the employment of surrogate modeling techniques can be a viable solution for the current and future challenges in terms of PRA, UQ and Safety Margins characterizations. In addition, in this manuscript it is presented the construction of a RAVEN new entity named “HybridModel” that is aimed to switch between a high-fidelity code and its linked surrogate, accelerating the PRA and UQ processes without largely reducing the accuracy of the results of the analyses. All these new capabilities will be shown with an application example about a Station Black Out scenario for a prototypical Pressurized Water Reactor (modeled with RELAP5-3D).

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Evaluation of effective-stress-function algorithm for nuclear fuel simulation

In a pressurized water reactor (PWR), the mechanical integrity of nuclear fuel is the most critical issue as it is an important barrier for fission products released into the environment. The integrity of zirconium cladding that surrounds uranium oxide can be threatened during off-normal operation owing to a pellet-cladding mechanical interaction (PCMI). To analyze the fuel and cladding behavior during off-operation, the fuel performance code should calculate an inelastic analysis in two - or three-dimensional calculations. In this paper, the effective stress function (ESF) algorithm based on a two-dimensional FE module has been implemented to simulate the inelastic behavior of the cladding with stability and accuracy. The ESF algorithm solves the governing equations of the inelastic constitutive behavior by calculating the zero of the appropriate effective-stress-function. To verify the accuracy of the ESF algorithm for an inelastic analysis, a code-to-code benchmark was performed using the commercial FE code, ANSYS 13.0. To demonstrate the stability and convergence of the implemented algorithm, the number of iterations in the ESF algorithm was compared with that in a sequential algorithm in the case of an inelastic problem. Consequently, the evaluation results demonstrate that the implemented ESF algorithm improves the efficiency of the computation without a loss of accuracy for an inelastic analysis. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Applicability of the ARCON96 Code to the Radiological Consequence Analysis of Releases from the Radioassay and Non-Destructive Testing (RANT) Facility

Establish technical justification to demonstrate that the horizontal diffusion parameter (σy) and vertical diffusion parameter (σz) that are embedded in the Nuclear Regulatory Commission (NRC) Atmospheric Relative CONcentrations (ARCON96) code are representative of the meteorological conditions at LANL. Compare the ARCON96 atmospheric diffusion factor ( X /Q) to the MACCS2-calculated X /Q for the determination of radiological consequences of ground releases from the Radioassay and Non-Destructive Testing (RANT) facility on the key receptors, the colocated worker (CW) and Maximally-Effected Offsite Individual (MEOI).

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Draft genome of the most devastating insect pest of coffee worldwide: the coffee berry borer, Hypothenemus hampei

The coffee berry borer, Hypothenemus hampei, is the most economically important insect pest of coffee worldwide. We present an analysis of the draft genome of the coffee berry borer, the third genome for a Coleopteran species. The genome size is ca. 163 Mb with 19,222 predicted protein-coding genes. Analysis was focused on genes involved in primary digestion as well as gene families involved in detoxification of plant defense molecules and insecticides, such as carboxylesterases, cytochrome P450, gluthathione S-transferases, ATP-binding cassette transporters, and a gene that confers resistance to the insecticide dieldrin. A broad range of enzymes capable of degrading complex polysaccharides were identified. We also evaluated the pathogen defense system and found homologs to antimicrobial genes reported in the Drosophila genome. Ten cases of horizontal gene transfer were identified with evidence for expression, integration into the H. hampei genome, and phylogenetic evidence that the sequences are more closely related to bacterial rather than eukaryotic genes. We find the draft genome analysis broadly expands our knowledge on the biology of a devastating tropical insect pest and suggests new pest management strategies.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2015

Preliminary critical experiments validation of a lattice physic code LATC - 14415

Core and System Integrated Engine for design and analysis (COSINE) is an integrated nuclear engineering code package, which is being developed by State Nuclear Power Software Development Center (SNPSDC) in China. A lattice physics code named LATC is included in COSINE.LATC is a 2D, steady-state code for modeling PWR fuel. A WIMS-D 69 group micro library based on ENDF/B-VI is used and there are two transportation methods in LATC, such as two steps method (collision probability method for cell homogenization, discrete coordinate method for assembly homogenization) and one step method (MOC). The UO{sub 2} fuel assembly, which may contain guide tubes, burnable poisons such as Integral Fuel Burnable Absorbers (IFBA) / Wet Annular Burnable Absorbers (WABA)/ Pyrexes or control rods etc, can be calculated in LATC. LATC provides the required cross-section data for the core physics simulator analysis code in COSINE. In order to evaluate the quality of the LATC, a strategy of validation is researched and applied to the LATC. Firstly, a validation requirement analysis is performed, and the corresponding critical experiments are collected, and then a critical experimental matrix considering the variety of fuel enrichments, fuel pitch and water to uranium ratio is generated. Finally, a comparison of results between LATC using collision probability and discrete coordinate method and critical experiments are made to make sure that the code works well. According to the validation strategy discussed above, the preliminary validation is carried out and critical experiments are calculated according to the critical experimental matrix and results from LATC code are compared with reference results. The result shows that the calculated results agree well with critical experiments, and do not indicate any significant trends with respect to the various conditions considered. (authors)

42 ENGINEERING↗
DOE OSTI2016

Charged particle tracking through electrostatic wire meshes using the finite element method

Wire meshes are used across many disciplines to accelerate and focus charged particles, however, analytical solutions are non-exact and few codes exist which simulate the exact fields around a mesh with physical sizes. A tracking code based in Matlab-Simulink using field maps generated using finite element software has been developed which tracks electrons or ions through electrostatic wire meshes. The fields around such a geometry are presented as an analytical expression using several basic assumptions, however, it is apparent that computational calculations are required to obtain realistic values of electric potential and fields, particularly when multiple wire meshes are deployed. The tracking code is flexible in that any quantitatively describable particle distribution can be used for both electrons and ions as well as other benefits such as ease of export to other programs for analysis. The code is made freely available and physical examples are highlighted where this code could be beneficial for different applications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Software Verification Toolkit (SVT): Survey on Available Software Verification Tools and Future Direction

Writing software is difficult. However, writing complex, well tested and designed, and functionally correct software is incredibly difficult. An entire field of study is devoted to the validation and verification of software to address this problem, and in this paper we analyze the landscape of currently available third party software. We have divided our analyses into three separate subsections with regards to software validation: formal methods, static analysis, and test generation. Formal verification is the most complex method in which to validate software correctness, but also the most thorough as it truly validates the mathematical validity of the source code. Static analysis generally is relegated to abstract syntax tree traversal techniques to find errors related to faulty software such as memory leaks or stack overflow issues. Automatic test generation is similar in implementation to static analysis, but pushes a bit further in verifying the boundedness of function inputs and outputs with regards to annotated or parsed criteria. The crux of this report is to analyze and describe the software tools that implement these techniques to validate and verify software. Pros and cons related to installation, utilization, and capabilities of the frameworks are described, and reproducible examples are provided with a focus on usability. The initial survey concluded that the most interesting tools of note are Z3, Isabelle/HOL, and TLA+ with regards to formal verification; and Infer, Frama-C, and SonarQube with regards to static analysis. With these tools in mind, a final conjecture is provided that describes future avenues of utilizing these tools for developing a verification framework to assist in validating existing software at Sandia National Laboratories.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Numerical Analysis of Novel Plate Type Heat Exchanger with Oval-Twisted Channels

The novel heat exchanger (HX) designs implementing geometrical and surface modification combinations are expected to perform better than traditional HX technologies. Applied enhancement techniques seek to achieve (1) higher overall heat transfer performance, (2) increased compactness, and (3) simplified or comparable manufacturability. One innovative enhancement technique is to generate swirling flow vortices induced by channel or tube twisting. The turbulence generated by the twisted cross-section greatly enhances the heat transfer rate with minimal increases in pressure drop. Plate-type HXs and Printed Circuit Heat Exchangers (PCHXs) are compact designs that achieve high heat transfer rates per unit volume by utilizing several small channels, which maximizes the heat transfer surface area between the hot and cold fluids. Currently, advanced manufacturing technologies enable the design and fabrication of compact-type units with complicated channel geometries to achieve the highest performance and meet the compactness criteria of innovative HX technology. The proposed HX design concept combines the compactness of plate-type HXs and twisted channels, which provide additional turbulence and flow swirl enhancement. The plate-type oval-twisted HX (PTOTHX) is a crossflow configuration, with 16 short channels on one side (for hot fluid) and 8 long channels on the perpendicular side (for cold fluid). The inlet plenums have flow guide vanes to redirect flow and produce uniformity across the various flow paths. The compact size and purportedly improved heat transfer performance of the PTOTHX investigated herein prove its viability in various applications. Some notable potential nuclear applications of the PTOTHX include reactor core, spent fuel cooling, and residual heat dissipation. To establish a reference case, circular channels (PTCHX) are also considered in the present study for comparison with (PTOTHX). This paper aims to outline the numerical analysis procedures for determining the viability of the PTOTHX by comparing its heat transfer performance with the PTCHX units. The computational study used STAR-CCM+, a commercial computational fluid dynamics (CFD) code. Sensitivity analysis and model selection studies are conducted to determine the appropriate mesh density and turbulence model to provide the reported results. Numerical analyses comparing the Nusselt number (Nu) of the PTOTHX design with a comparable HX unit, including a circular cross-section and no twisting (PTCHX), show an overall heat transfer performance increase of 29-55% for balanced flow and 29-59% for imbalanced flow. Oval-cross-sectional twisted channels induce swirling flow vortices, enhancing the working fluid's convective heat transfer capabilities.

25 ENERGY STORAGE↗
DOE OSTI2016

limiting rods selection in large break LOCA analysis with the RELAP5-3D code

The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10 CFR 50.46(c), the US NRC put forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently more fuel rods need to be analyzed under LOCA conditions to maintain the safety margin, compared to the current practice in which only one hot rod needs to be analyzed. New analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of the limiting rods under LOCA conditions. In this work, the best-estimate plus uncertainty analysis capability for large break LOCA with the new cladding embrittlement criteria using the RELAP5-3D code is established and demonstrated with a reduced set of uncertainty parameters. With the new analysis method, the limiting transient case and the limiting rods can be easily identified in response to the proposed new rulemaking. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Quantitative Analysis of Color Differences within High Contrast, Low Power Reversible Electrophoretic Displays

Here, quantitative color analysis is performed on videos of high contrast, low power reversible electrophoretic deposition (EPD)-based displays operated under different applied voltages. This analysis is coded in an open-source software, relies on a color differentiation metric, ΔE * 00 , derived from digital video, and provides an intuitive relationship between the operating conditions of the devices and their performance. Time-dependent ΔE * 00 color analysis reveals color relaxation behavior, recoverability for different voltage sequences, and operating conditions that can lead to optimal performance.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Conjugate Heat Transfer Analysis in Pressurized Helium Gas Cooling Channels

Los Alamos National Laboratory (LANL) is working with NorthStar Medical Radioisotopes (NMR) on their efforts to produce Mo-99 from the irradiation of Mo-100 targets. The NMR target consists of an Inconel window that allows the electron beam to penetrate and irradiate a stack of Mo-100 disks. The irradiation process generates large amounts of heat and is cooled using pressurized helium gas flowing at 400 g/s. LANL provides both numerical analysis support and experimental support. Computational fluid dynamics (CFD) and conjugate heat transfer (CHT) analysis are performed on the NMR target to analyze the helium flow and volumetric heating through the NMR system. In this iteration of numerical analysis, we update the Mo-100 disk thickness to 0.74 mm and the gap between the Inconel window and the first disk to 0.69 mm. Estimates of the heat flux from the electron beams from Monte Carlo N-Particle code (MCNP) analysis were combined with CFD to perform the CHT analysis. The pressure drop through the targets, force applied on the first disk, beam window and target disk temperatures are some of the quantities estimated using the simulations. The pressure drop through the target was estimated as 0.198 MPa, while the force due to venturi effects across the first disk was estimated to be 100 N. The results show maximum window temperatures over 600 °C and peak disk temperatures over 1600 °C. Past work has shown that beyond 650-700 °C, the Inconel window becomes susceptible to yield and rupture. The increased temperature in these results reduces the margin of error possible on beam spot size. The disk temperatures showed a sinusoidal distribution with the 6 th – 8 th disk from each window exhibiting highest disk temperatures. The peak temperature of the stainless steel housing was close to 316 °C and that of the laminations and spacers was approximately 582 °C.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Risk-Informed Mechanistic Source Term Calculation (Final CRADA Report)

As part of an FY18 U.S. Department of Energy (DOE) Gateway for Accelerated Innovation in Nuclear (GAIN) award, a CRADA agreement was established between Argonne National Laboratory (Argonne) and Oklo. The CRADA focused on the following three tasks to assist Oklo with the regulatory licensing of their innovative nuclear reactor design: • Task 1 – Coupling and Demonstration Analysis using The Nek5000, Diablo, and OpenMC Codes: Establishing a coupling between the neutronic, structural, and thermohydraulic analysis codes and utilizing the code package for a high-fidelity demonstration analysis of radial core expansion due to temperature changes and its impact on core reactivity. • Task 2 – Mechanistic Source Term Analysis: Performing a realistic evaluation of the potential release of radionuclides to the environment during postulated transient event sequences. • Task 3 – Data in Support of Metal Fuel Qualification: An assessment of past postirradiation examination data of metal fuel utilized in the Argonne EBR-II reactor to assist in the potential qualification of the fuel typer for regulatory licensing. The details regarding each analysis and the associated conclusions were provided to Oklo in three separate project reports. Due to the utilization of proprietary and export controlled information, the current summary report was created, which provides an uncontrolled overview of each of the three tasks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗