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DOE OSTI2022

Preliminary Design of Reactor Cavity Cooling System for a Horizontal Compact HTGR

The Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed MIGHTR in 3 years and support its commercialization as a safe and low-cost HTGR. Argonne National Laboratory (Argonne) is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. This report documents the preliminary design study of the RCCS for the HC-HTGR. It includes the establishment of the design requirements, a high-level design study by initial scoping calculations, and preliminary performance calculations of the HC-HTGR RCCS design. Design requirements for the HC-HTGR RCCS have been established to guide preliminary design activities and scoping performance calculations. Initial scoping calculations including estimation of the water inventory, estimation of HVAC thermal capability, and a parametric study on loop dimensions by standalone RCCS analysis. Based on scoping calculation results, a set of baseline dimensions of the HC-HTGR RCCS was derived. A water panel modeling approach was investigated to explore various potential design options for the water panel under consideration for the HC-HTGR RCCS using RELAP5-3D. A test case study was performed to assess the prediction capability of two modeling approaches. The results were compared with CFD simulations conducted in constant RPV temperature and heat flux boundary conditions. It confirms the capability of the RELAP5-3D modeling approach to include all important heat transfer mechanisms expected in the HC-HTGR RCCS operation conditions. Then, a reference RELAP5-3D model for the 1/8 th of a compartment of the preliminary design of the HC-HTGR RCCS was developed. A preliminary performance analysis was conducted to evaluate single-phase natural circulation performance with different top tank temperature values and panel conduction performance in various operation conditions. From single-phase natural circulation performance analysis, the system operation mode was investigated in normal operating and limiting design conditions. It showed operation mode in a subcooled state with a proper top tank water cooling system. Parasitic heat loss by both internal air flow and RCCS was estimated, showing it satisfies maintaining below target maximum heat loss of the HC-HTGR RCCS. From the panel conduction performance analysis, two candidate materials for the riser tube such as carbon steel and stainless steel were compared in the thermal performance of HC-HTGR RCCS. From a single water panel test compared with CFD simulation results, it was confirmed that the current capability of the RELAP5-3D modeling approach for the water panel predicts the thermal conduction of two different materials of the water panel. Then, system-level thermal performance analysis was performed for 1/8 th of the compartment of the preliminary HC-HTGR RCCS design. It was first observed that the current preliminary HC-HTGR RCCS design had minimal impact on the overall thermal performance of the water panel by changing pipe material from carbon steel to stainless steel. From Argonne’s effort on the ongoing water-based NSTF testing program, several considerations other than the thermal performance point of view were addressed to be considered in selecting pipe materials.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Analytic error analysis of cross section interpolation methods in nodal diffusion codes - II: Numerical results

This paper is the second part of a two-part paper that documents the numerical results for the partial derivatives model presented in part I. In this paper, we derive the error bounds for the analytical point-wise error expression and verify our bounds with numerical experiments. The point-wise error expressions make available, and bound, the sources that contribute to the total error of the interpolated cross section in terms of the Lagrange interpolation errors and the model form error. MPACT is used to generate two-group homogenized cross sections for Westinghouse's AP1000 Region 4 lattice to evaluate the accuracy of the bounds. Error bounds calculated over a grid are compared to numerical data for uni-variate and multi-variate interpolation. The point-wise error bounds of a typical case matrix - two branches in each state variable - are displayed for bi-variate interpolation in the state variables: moderator density, fuel temperature, and boron concentration. The error bounds are shown to be highly accurate compared to numerical results, and in accordance with the underlying physics. We then discuss and show how the sources of error contribute to the total error, and consider the improvement of each error source. Finally, we mention future work such as propagating our cross section error bounds through a reactivity calculation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Completion document for MRT 8824 - Upgrade NIF’s gaseous radiochemistry diagnostics (RAGS) to support weapons science

This milestone highlights the successful upgrade and deployment of the Radiochemical Analysis of Gaseous Samples (RAGS) system, which delivers high-quality measurements (<20% uncertainty) of activated gaseous species from NIF implosions. These measurements are critical for supporting Stockpile Stewardship Program (SSP) relevant platforms, including LANL’s Double Shell and LLNL’s Pushered Single Shell campaigns. The RAGS diagnostic technique enables analysis of short-range mixing in implosions using high-Z shells, which are otherwise inaccessible to conventional x-ray diagnostic methods. By facilitating the investigation of high-Z material mixing into fusion burn, this system provides essential data for quantifying and interpreting results in high-energy density (HED) experiments. This report details the physics motivation, diagnostic fundamentals, planned and enacted upgrade work, and the quantification of uncertainty for the upgraded system.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2019

FGR 13 Dose Conversion Factor Files

The purpose of this document is to discuss the construction of two MELCOR Accident Consequence Code System (MACCS) dose conversion factor (DCF) files in some detail, an older file created in 2007 named FGR13DCF.inp and a newer file created in 2018 called FGR13GyEquiv_RevAinp. Very briefly, the difference between the two files is that the older file follows the standard conventions of assigning a radiation weighting factor of 20 for alpha radiation for all tissues and organs; whereas, the newer file complies with the Federal Guidance Report (FGR) 13 health effects modeling and uses modified radiation weighting factors (referred to as relative biological effectiveness (RBE) factors) for alpha radiation of 10 for breast and of 1 for red bone marrow. During an intermediate period, a file called FGR13GyEquiv.inp was created and used for the State-of-the-Art Reactor Consequence Analysis (SOARCA) calculations. This file was not released to the MACCS user community, but it is also discussed briefly in this document.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2016

NDMAS implementation and data qualification for accident tolerant fuel experiments

In order to meet quality and regulatory requirements associated with the U.S. Department of Energy Accident Tolerant Fuels (ATF) experiments, all data and documentation associated with ATF component fabrication, characterization, traceability, testing and post irradiation examination are being managed using the Nuclear Data Management and Analysis System (NDMAS). NDMAS melds commercial off-the-shelf software to provide the full range of needed functionality. The core component of NDMAS is the SAS Enterprise Business Intelligence software, which provides data manipulation, statistical and graphical capabilities, and delivery of live content to web pages. Content delivery to both Idaho National Laboratory researchers and external users via a web interface is managed using SharePoint, which is access-controlled. Additionally, scripts, macros, and commercial wrappers may be used or written by the NDMAS team to control interfaces between SAS and the structured query language server, where much of the data reside. ATF experiment data and documentation stored in the NDMAS database may be utilized to support Nuclear Regulatory Commission licensing of ATF concepts for commercial Light Water Reactor demonstration. Therefore, Data Qualification Plans are being developed to screen and tag all document and data input for ATR water loop testing, Transient Reactor Testing (TREAT), and post irradiation examination. A project-wide document library has been created in the ATF NDMAS SharePoint Site for general unrestricted document and information access. Restricted libraries have also been created to allow individual customer access to restricted documents. Data from ATR water loop instrumented lead experimentation, TREAT, and post-irradiation examination will be available to program participants on the SharePoint site as they are collected. This offers a unique capability for team collaboration while these tests are being conducted. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Mechanical Properties and Deformation Behavior of Additively Manufactured 316L Stainless Steel (FY2020)

The Transformational Challenge Reactor (TCR) program plans to build most of the TCR core components through additive manufacturing (AM) processes. These processes include laser powder bed fusion (LPBF) for the metallic (316L) components and the newly developed combined process of binderjet printing and chemical vapor infiltration (CVI) for the SiC fuel matrix. Mechanical testing and characterization tasks have been carried out since the beginning of the TCR program to (1) build a property database for the AM materials that will be used in TCR core and (2) to assess the materials’ performance in TCR-relevant conditions. This document reports the outcome of the testing and characterization efforts for the fiscal year with a focus on the mechanical performance data of AM 316L stainless steel (SS). Baseline tensile testing over a wide temperature range of room temperature–600 °C was completed for the AM 316L alloy in as-built, stress-relieved, and solution-annealed conditions. The as-built 316L showed the highest strength, and the alloy after the post-build treatments showed reduced strengths in the low-strain range. However, the strength differences among the AM materials became insignificant in the later part of deformation. Furthermore, regardless of post-build processing, the AM 316L SS showed higher strength and comparable ductility when compared with wrought 316L SS. Thermal creep testing and microstructural evolution during creep deformation were also performed under selected conditions. It was found that the AM 316L steel showed the best creep resistance in the stress-relieved condition. In-situ tensile tests were performed using scanning electron microscopy and 1-ID beamline at the Advanced Photon Source to elucidate the deformation and fracture behavior of AM 316L and the evolution of crystalline stress, dislocations, and pore distribution. Using these in-situ testing data, an in-depth analysis of the roles of microstructural features in deformation and fracture processes is presented herein. The final section of the document introduces ongoing and future activities for materials testing and characterization, including irradiation effect studies and ball punch testing on AM 316L and AM IN718 alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds (Annual Status Report FY 2022)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-EP-0645). The estimates are compared with the performance objectives defined in U.S. Department of Energy requirements (DOE O 435.1 and its companion documents DOE M 435.1-1 and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 is expected. Within the active burial grounds, low-level and mixed low-level waste currently may be disposed only in two lined trenches in the 218-W-5 Burial Ground (Trenches 31 and 34) until they are either filled or a decision is made to close these trenches. Some mixed low-level waste is also disposed at the Environmental Restoration Disposal Facility in the 200 West Area (which is covered under a separate PA). During this (fiscal year 2022) reporting period (October 1, 2021, through September 30, 2022), waste was disposed to the 200 West Area LLBGs. Continued groundwater monitoring of the 200 West Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis assumptions of a postclosure facility that will not be degraded by human activity. The LLBGs are in an area identified for waste management and containment of residual contamination (DOE/EIS-0391). The current closure plan for the LLBGs (DOE/RL-2000-70) estimates that the 200 West LLBGs will be closed in the 2050 timeframe. The Disposal Authorization Statement, other technical basis documents, and the radioactive waste management basis are of continued adequacy to meet the performance objectives of DOE O 435.1. Overall, there are no substantive changes to primary PA assumptions and no changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the Disposal Authorization Statement is maintained. A new PA to evaluate the long-term impacts of three disposal trenches that are currently active within the 200 East and 200 West Areas (Trench 94 in the 200 East Area and Trenches 31 and 34 in the 200 West Area) was initiated in fiscal year 2019 and completed in fiscal year 2022. Corrective actions addressing 3 key issues and 31 secondary issues identified during the review process were developed and submitted to the Low-Level Waste Disposal Facility Federal Review Group Co-Chairs for review and approval. This PA provides additional technical basis for the continued adequacy of the existing Operating Disposal Authorization Statement.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Optimizing the Performance of Sparse-Matrix Vector Products on Next-Generation Processors

Matrix-vector products are ubiquitous in high-performance scientific applications and have a growing set of occurrences in advanced data analysis activities. Achieving high performance for these kernels is therefore paramount, in part, because these operations can consume vast amounts of application execution time. In this report we document the development of several sparse-matrix vector product kernel implementations using a variety of programming models and approaches. Each kernel is run on a broad set of matrices selected to demonstrate the wide variety of matrix structure and sparsity that is possible with a single, generic kernel. For benchmarking and performance analysis, we utilize leading computing architectures for the NNSA/ASC program including Intel's Knights Landing processor and IBM's POWER8.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Code Coverage Status of the ARC Code DASSH-F

The Argonne Reactor Code (ARC) software system supports users in their fast reactor design goals by providing neutronic, thermal-hydraulic, and structural analysis capabilities. DASSH-F serves as a steady state thermal hydraulic capability within the ARC system and replaces the SE2-ANL software that preceded it. This document identifies the set of test problems used to assess the code coverage for DASSH-F. The goal is to document what parts of the existing DASSH-F code are touched by the set of test problems and which are not. Because the verification work remains to be done on DASSH-F, one can assume that most of these issues will be resolved as part of that work. The code coverage analysis of DASSH-F was performed with the Code Coverage Tool of the Intel Fortran compiler which requires modifications to the compilation of DASSH-F. The code coverage tables are given for each submodule of DASSH-F. Because DASSH-F links to modules in DIF3D, some details on coverage changes to the DIF3D linked files is provided. As will be seen, most of the uncovered parts/files can be ignored because they are either for error message and debugging output or obviously not needed by DASSH-F today. Seven features of the DASSH-F code were identified to not be covered by the existing testing suite and thus additional verification test problems are suggested to fully cover these sections.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

PVAnalytics: A Python Package for Automated Processing of Solar Time Series Data

Multiple publicly available software packages exist that analyze solar time series data, including RdTools and Solar Data Tools, among others. Several of these packages contain their own unique quality assurance (QA) and feature recognition algorithms. The python PVAnalytics package was developed to offer an internally consistent source for these analysis tools, making it easier for the end user to deploy these routines on his or her solar data. The PVAnalytics package currently contains routines for outlier detection, inverter clipping detection, irradiance and temperature checks, orientation checks, and data shift detection, among other functions. These functions have been aggregated from various sources including Solar Forecast Arbiter, RdTools, and the QA process developed by NREL's PV Fleets Initiative. We are continuously adding new functionality to the package, including documentation, examples and algorithms. By bundling QA functionality into a single software package, we hope to make PVAnalytics a comprehensive software library to support analysis of solar metadata and time series data.

data cleaning↗
DOE OSTI2020

Distribution of Infiltration in the 216-U-10 and 216-B-3 Pond Systems 1944-1997

A modeling system to simulate fate and transport of contaminant plumes in groundwater beneath the Central Plateau to support the Cumulative Impact Evaluation (CIE) and Composite Analysis (CA) is being developed. The CIE will evaluate the cumulative effect of multiple Central Plateau radiological and chemical sources on groundwater quality (DOE/RL-2018-69, Cumulative Impact Evaluation Technical Approach Document ). This will enable cleanup decisions to be evaluated in the context of all Central Plateau waste sites and existing groundwater conditions. The CA evaluates the cumulative effect of all sources of radioactive materials on radiological dose to future members of the public (DOE/RL-2018-60, Annual Status Report (FY 2018): Composite Analysis of Low Level Waste Disposal in the Central Plateau of the Hanford Site ). Fate and transport modeling will be used to forecast future groundwater concentrations for both the CIE and CA. The CIE/CA subsurface modeling system is divided into source (inventory), vadose zone, and saturated zone facets. Source inventory is a necessary input to the vadose zone facet, and releases from the vadose zone to groundwater are inputs to the saturated zone facet. The vadose zone facet simulates historical and future releases from the vadose zone to groundwater for both liquid waste and solid waste disposal sites. Effluent release volumes and radionuclide inventory estimates for liquid waste sites are contained in ECFHANFORD- 17-0079, Hanford Soil Inventory Model (SIM-v2) Calculated Radionuclide Inventory of Direct Liquid Discharges to Soil in the Hanford Site’s 200 Areas , herein referred to as SIM-v2. The liquid waste sites in SIM-v2 include the 216-B-3 Pond near 200 East Area and the 216-U-10 Pond in 200 West Area. The 216-B-3 Pond system consisted of a main pond and 3 expansion lobes that were operational during overlapping time periods. It also included influent ditches, but these are not listed as liquid waste sources in SIM-v2 and are not included in the CIE/CA modeling. The total effluent volume released to the pond system was measured, but the partitioning of this volume to the various expansion lobes was not determined and is not accounted for in SIM-v2. Regarding the 216-U-10 Pond, it received water from several influent ditches. The pond and the influent ditches are listed as liquid waste sites in SIM-v2 and are included as sources in the CIE/CA modeling. Because water in the ditches flowed into the pond, only a portion of the effluent volumes and contaminant inventories assigned to the ditches in SIM-v2 infiltrated from the ditches. Thus, partitioning of infiltration and contaminant inventories between the ditches and the pond is not fully accounted for in SIM-v2. Further, some of the water in the 216-U-10 Pond entered overflow ditches and the amount of this overflow was not recorded and is not accounted for in SIM-v2. Thus, the purpose of this Environmental Calculation File (ECF) is to partition the infiltration of wastewater and contaminant inventory between the main pond and each expansion lobe of the 216-B-3 Pond system, and to partition the infiltration of wastewater and contaminant inventory between the ditches and pond of the 216-U-10 Pond system.

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

KENO-VI Primer: Performing Calculations using SCALE’s Criticality Safety Analysis Sequence (CSAS6) with Fulcrum

The SCALE code system developed at Oak Ridge National Laboratory is widely used and accepted around the world for criticality safety analysis. The well-known KENO-VI three-dimensional Monte Carlo criticality computer code is one of the primary criticality safety analysis tools in SCALE. The KENO-VI primer is designed to help a new user understand and use the SCALE/KENO-VI Monte Carlo code for nuclear criticality safety analysis. It assumes that the user has a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with SCALE/KENO-VI in particular. The primer is designed to teach by example, with each example illustrating two or three features of SCALE/KENO-VI that are useful in criticality analysis. The primer is based on SCALE 6.2 and 6.3, which includes the Fulcrum graphical user interface. Each example uses Fulcrum to provide the framework for preparing input data and viewing output results. Starting with a Quickstart section, the primer gives an overview of the basic requirements for SCALE/KENO-VI input and allows the user to quickly run a simple criticality problem with SCALE/KENO-VI. Each following section begins with a list of basic objectives identifying the goal of the section and the individual SCALE/KENO-VI features covered in detail in the section’s sample problems. Upon completion of the primer, a new user should be comfortable using Fulcrum to set up criticality problems in SCALE/KENO-VI. The primer provides a starting point for the criticality safety analyst who uses SCALE/KENO-VI. Complete descriptions are provided in the SCALE/KENO-VI manual. Although the primer is self-contained, it is intended as a companion volume to the SCALE/KENO-VI training and documentation. The SCALE manual and training schedule are available at https://scale.ornl.gov. The primer provides specific examples of using SCALE/KENO-VI for criticality analysis; the SCALE/KENO-VI manual provides information on the use of SCALE/KENO-VI and all its modules. The primer also contains an appendix with sample input files. In addition, this primer, its errata, and sample inputs are also available at https://code.ornl.gov/scale/primers/kenovi/.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Performance Assessment for the Environmental Restoration Disposal Facility (Annual Status Report FY 2021)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA) (CP-60089), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and that the determination must consider the results of data collection and analysis from research, field studies, and monitoring as well as the need to update any Radioactive Waste Management Basis (RWMB) documents. Beginning in 1996, the Environmental Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2021, extending from October 1, 2020, through September 30, 2021), approximately 9.14E+04 metric tons (1.01E+05 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2021, approximately 16.9 million metric tons (18.9 U.S. tons) of waste has been disposed of at ERDF, which equates to consumption of approximately 88.7% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed. As a condition of the DAS, disposal operation within ERDF must be in accordance with the waste acceptance criteria (ERDF-00011) that provide specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE M 435-1.1. The ERDF waste acceptance criteria stipulate that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. There have been no changes to the physical configuration of ERDF or to the waste forms (source, physical form, etc.). No new unreviewed disposal question screenings or evaluations have been generated in this reporting period. Therefore, there are no noted impacts to the PA, CA, DAS, or RWMB resulting from the evaluations and screenings. Sum-of-fractions analysis shows that the disposed inventory meets both the concentration and inventory threshold requirements. A sum-of-fractions value is computed for ERDF sensitive radionuclides contributing to the all pathways and the air pathway inventory limits. Computed values were 7.63E-03 and 1.49E-03, respectively. The disposed waste inventory remained well under the PA imposed limits, as shown in Table 4 and Table 5 in the main text of this report. Required monitoring was satisfactorily completed during the fiscal year reporting period. Compliance with performance objectives were met as each of the reported values were well below the established limit. Overall, there are no substantive changes to primary PA assumptions or changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the DAS is maintained.

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

Performance Assessment for the Environmental Restoration Disposal Facility (Annual Status Report FY 2022)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA) (CP-60089), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and that the determination must consider the results of data collection and analysis from research, field studies, and monitoring as well as the need to update any Radioactive Waste Management Basis (RWMB) documents. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2022, extending from October 1, 2021, through September 30, 2022), approximately 8.52E+04 metric tons (9.39E+04 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2022, approximately 17.0 million metric tons (18.7 million U.S. tons) of waste has been disposed at ERDF, which equates to consumption of approximately 89.1% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed. As a condition of the DAS, disposal operations within ERDF must be in accordance with the waste acceptance criteria (ERDF-00011) that provide specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE M 435-1.1. The ERDF waste acceptance criteria stipulate that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. There have been no changes to the physical configuration of ERDF or to the waste forms (source, physical form, etc.). No new unreviewed disposal question screenings or evaluations have been generated in this reporting period. Therefore, there are no noted impacts to the PA, CA, DAS, or RWMB resulting from the evaluations and screenings. Sum-of-fractions analysis shows that the disposed inventory meets both the concentration and inventory threshold requirements. A sum-of-fractions value is computed for ERDF sensitive radionuclides contributing to the groundwater pathways and the air pathway inventory limits. Computed values were 5.50E-04 and 3.30E-03, respectively. The disposed waste inventory remained well under the PA imposed limits, as shown in Table 4 and Table 5 in the main text of this report. Required monitoring was satisfactorily completed during the fiscal year reporting period. Compliance with performance objectives were met as each of the reported values were well below the established limit. Overall, there are no substantive changes to primary PA assumptions or changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the DAS is maintained.

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

TEM Specimen Manufacturing to Support Analysis of IASCC Mechanisms in Irradiated Austenitic Steels – Light-Water– Relevant Materials

The present report describes the manufacturing of samples for advanced microstructure studies of mechanisms of irradiation-assisted stress corrosion cracking (IASCC) in the materials of light-water reactor (LWR) internal components. The samples were manufactured via FIB (focused ion beam milling) from 4-point-bend specimens subjected to oxidation and mechanical loading in a simulated LWR environment (primary water, PW). The corrosion tests were conducted at the University of Michigan (UM), and the tested specimens were transported to ORNL to perform FIB work. Finally, the manufactured FIB specimens were shipped back to the UM for microstructure characterization. Section 1 in the report provides a brief introduction; Section 2 describes materials, their in-service and irradiation history, and the geometry of specimens for IASCC testing. Section 3 documents typical work steps (specimen mapping) and reports the number of manufactured FIB objects. Section 4 briefly describes the preliminary results of microstructure analysis performed at the University of Michigan and highlights the key findings. The summary and conclusions section discusses future possible future steps and generalizes the obtained results, their unique character, and their importance for the nuclear industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Advances in Metallic Fuel Database Development and Data Qualification

The Fuels Irradiation and Physics Database (FIPD [1]) is a comprehensive repository of data and documents related to Uranium-Zirconium based metallic fuel test pins. This database stores operational conditions of these pins, calculated using a suite of Argonne National Laboratory analysis codes developed during the Integral Fast Reactor (IFR) program. Key calculated data include axial distributions of power, temperature, fluence, burnup, and isotopic densities. Additionally, the FIPD holds post-irradiation examination (PIE) data such as fission gas release, gas chemistry measurements, and axial distributions derived from profilometry, gamma scanning, and neutron radiography. Complementing these data is an extensive archive of documents related to various pins and experiments. These include raw PIE records, design details, safety analyses, and operational reports. More detail about FIPD can be found in ref. [2]. The database development is an ongoing effort covering metallic fuel experiments from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The recent improvements to the database and the data QA status are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

SAFETY ANALYSIS REPORT FOR PACKAGING SHIELDING & NUCLEAR CRITICALITY SAFETY COURSES DEVELOPED AND CONDUCTED BY OAK RIDGE NATIONAL LABORATORY

The US Department of Energy (DOE) Packaging Certification Program, Office of Packaging and Transportation, is offering Safety Analysis Report for Packaging (SARP) Shielding and Nuclear Criticality Safety (NCS) courses for SARP generalists and analysts. The SARP generalist course is designed for project managers, supervisors, NCS/shielding subject matter experts (SME), and SMEs in non-NCS/shielding technical areas (e.g., structural, thermal, package design) who need to improve their understanding of how NCS/shielding analyses fit into the broader spectrum of SARP documentation. The generalist course provides an overview of the regulations and guidelines for the criticality and shielding analyses included in a SARP, and it illustrates how the NCS/shielding chapters integrate with the other parts of the SARP. After course material is presented, students in the generalist course will review an actual SARP document to examine key elements of the shielding and criticality analyses. The analyst course provides detailed instructions on the radioactive material package shielding analyses and NCS evaluation fundamentals needed by analysts/practitioners, including safety analysts and/or technical reviewers, to prepare and/or review technical analyses for SARP documentation. The analyst course also provides an overview of regulations and guidelines, in addition to detailed in-class exercises associated with package shielding and NCS analyses. Analysis teams will be faced with staged SARP examples in which several important decision processes in the generation of a SARP will be demonstrated and discussed. Both courses are offered at Oak Ridge National Laboratory, and for interested students, the analyst course is offered for one hour of graduate credit (course ME-690) from the University of Nevada at Reno. An overview of the course will be provided to ensure that the NCS community is aware of the course and its intent to support the shielding and NCS aspects of packaging and transportation analyses for SARP activities.

Bowen, Douglas G.↗
DOE OSTI2016

FY 2016 Status Report: Documentation of All CIRFT Data including Hydride Reorientation Tests (Draft M2)

The first portion of this report provides a detailed description of fiscal year (FY) 2015 test result corrections and analysis updates based on FY 2016 updates to the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) program methodology, which is used to evaluate the vibration integrity of spent nuclear fuel (SNF) under normal conditions of transport (NCT). The CIRFT consists of a U-frame test setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages connecting to a universal testing machine. The curvature SNF rod bending is obtained through a three-point deflection measurement method. Three linear variable differential transformers (LVDTs) are clamped to the side connecting plates of the U-frame and used to capture deformation of the rod. The second portion of this report provides the latest CIRFT data, including data for the hydride reorientation test. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNF samples into a single zone. A detailed examination revealed that, at the same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), LMK, and mixed uranium-plutonium oxide (MOX). Just looking at the strain, LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The third portion of this report provides finite element analysis (FEA) dynamic deformation simulation of SNF assemblies . In a horizontal layout under NCT, the fuel assembly’s skeleton, which is formed by guide tubes and spacer grids, is the primary load bearing apparatus carrying and transferring vibration loads within an SNF assembly. These vibration loads include interaction forces between the SNF assembly and the canister basket walls. Therefore, the integrity of the guide tubes and spacer grids critically affects the vibration intensity of the fuel assembly during transport and must be considered when developing the multipurpose purpose canister (MPC) design for safe SNF transport.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗