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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Characterization Report for the MP-1 Experiment Fabrication Campaign (Rev.1)

This report covers the results of MP 1 fuel characterization analysis to provide a detailed understanding of the as-fabricated fuel that would be irradiated in the MP 1 (Mini-Plate 1) experiment. The work was performed at Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL), and Los Alamos National Laboratory (LANL). Characterization of the as-fabricated fuel was performed in accordance with Characterization Plan for the Fabrication of U 10Mo for the MP 1 Experiment, hereinafter the MP 1 Characterization Plan (MAQP 2016). Under the MP 1 Characterization Plan, fuel foils at different points in the foil fabrication process (e.g., hot rolled annealed, cold rolled annealed) were studied to understand the effect of various thermomechanical processes on the fuel microstructure. Samples were sent to each of the three organizations so that testing and analysis could be performed independently using similar equipment and a standardized set of measurement and analysis procedures. The MP 1 experiment will provide an opportunity to understand the effects of processing conditions on the final fuel microstructure, to compare independently obtained results, and achieve a three-way validation. PNNL characterized 28 uranium–10 wt% molybdenum (U 10Mo) samples. Six pieces/specimens from each sample/foil were sectioned, in accordance with the MP 1 Characterization Plan. Similarly, INL and LANL also examined 24 and 17 U 10Mo specimens, respectively. These total 69 samples consist of six types of MP 1 characterization foils fabricated by Babcock and Wilcox Technologies (BWXT) from ingots produced by the Y-12 National Security Complex: (a.) 0.047 in.-thick hot-rolled and annealed samples with and without Zr layers; (b.) 0.02 in.-thick cold-rolled and annealed samples with and without Zr layers; and (c.) 0.01 in.-thick cold-rolled and annealed samples with and without Zr layers. Microstructure, chemical composition, carbide morphology, U 10Mo foil thickness, Zr thickness, mechanical properties (microhardness), and density were evaluated in both longitudinal and transverse directions for foils of the three different thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

PNNL's Intermediate Characterization Summary for the MP-1 Experiment

Characterization of the as-fabricated fuel is being performed at Pacific Northwest National Laboratory (PNNL) in accordance with Characterization Plan for the Fabrication of U-10Mo for the MP-1 Experiment, hereinafter the MP-1 Characterization Plan. Similar characterization work is also being performed at two other organizations, INL and Los Alamos National Laboratory, to provide a detailed understanding of the as-fabricated fuel that would be irradiated in the MP-1 experiment. Under the MP 1 Characterization Plan, fuel foils with different fabrication parameters (such as rolling condition, rolling thickness reduction, co-rolling with Zr layers) are being studied. Similar samples from master foils were sent to each of the three organizations, so that testing and analysis can be performed independently using similar equipment and a standardized set of measurement and analysis procedures. A final consolidated report will be prepared based on this work and it will summarize all of the information obtained from different labs. The MP-1 experiment will provide an opportunity to understand the effects of processing conditions on the final fuel microstructure, to compare results obtained independently, and achieve a three-way validation. Currently, PNNL characterized 28 U-10Mo samples, by sectioning six pieces/specimens from each sample/foil, in accordance with the MP-1 Characterization Plan. These 28 samples consist of six types of MP-1 characterization foils from BWXT: 0.047 in. thick hot-rolled samples with and without Zr layers; 0.02 in. thick cold-rolled and annealed samples with and without Zr layers, and 0.01 in. thick cold-rolled and annealed samples with and without Zr layers. During the first year, PNNL completed sectioning of two subsamples per sample/foil and performed the necessary characterization on all 28 U-10Mo samples. This report covers the results of the first year’s analyses of PNNL’s MP-1 fuel characterization samples. Microstructure, chemical composition, carbide morphology, U-10Mo foil thickness, Zr thickness, mechanical properties (microhardness), and density were evaluated in both longitudinal and transverse directions for foils of the three different thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Enabling the Broader Use of MOOSE for Nuclear Energy and Other Simulation

This Final Scientific and Technical Report summarizes work performed under the Phase IIA SBIR project “Enabling the Broader Use of MOOSE for Nuclear Energy and Other Simulation” (DE-SC0020906) from August 2023 through August 2025. The objective of the Phase IIA effort was to mature and harden capabilities developed during Phase II, with the goal of enabling practical interoperability between Coreform’s isogeometric analysis (IGA) technologies and the Multiphysics Object-Oriented Simulation Environment (MOOSE), while improving robustness, performance, and scalability for complex, nuclear-relevant geometries. Over the course of Phase IIA, the project established and validated an extraction-based interoperability pathway between Coreform tools and MOOSE. A combined mesh and matrix format was defined collaboratively with MOOSE developers and integrated into the solver, enabling standard MOOSE workflows to operate on data exported from Coreform’s IGA and Flex Representation Method (FRM) pipelines. Early demonstrations validated architectural compatibility using linear solid mechanics problems, while later efforts focused on benchmark testing and external use. By the end of the project period, engineers at BWXT were able to independently set up and execute a simulation using the Coreform–MOOSE workflow and provide direct feedback that informed further refinement. In parallel, substantial effort was devoted to improving the robustness of trimmed U-spline construction for complex CAD geometries. A growing test suite of nuclear-relevant models was compiled through collaboration with multiple stakeholders and used to drive extensive bug fixing and reliability improvements. These efforts resulted in improved robustness and performance, including the addition of fallback capabilities that enhance reliability when the underlying commercial CAD kernel fails. Performance-oriented work progressed later in the project, with the development and demonstration of methods to decompose complex geometries into structured subregions and updated data representations to support more efficient solver processing. Additionally, extensive enhancements to threadsafe parallel data structures and trimming operations established a foundation for scalable processing of large assemblies. Collaboration with Sandia National Laboratories on the SGM geometric modeling kernel advanced to a functioning interface test case, positioning the workflow for future kernel integration. Overall, the Phase IIA effort successfully transitioned the project from architectural proof-of-concept to externally exercised, solver-integrated capability, while clarifying remaining technical challenges related to standardization, performance optimization, and kernel integration.

42 ENGINEERING↗
NASA NTRS2020

Simulation of Hydrogen Flow in a Nuclear-Thermal Propulsion Reactor Subsystem: Conjugate Heat Transfer in Surrogate Fuel Elements and Flow Through the Dome Inlet to the Fuel Elements

NASA is performing research related to a Nuclear-Thermal Propulsion (NTP) engine as a step toward future manned missions to Mars. BWX Technologies (BWXT), in conjunction with NASA, is developing the reactor subsystem for that NTP engine design. As a subset of the reactor subsystem, the fuel elements have continued to evolve in design, as well. A set of experiments within the Nuclear Thermal Reactor Element Environmental Simulator at NASA Marshall has been developed to compliment that design process, including a series of surrogate fuel elements. Before the experiments of each surrogate can be completed, they must be deemed safe using structural analysis, which itself requires thermal and conjugate heat transfer analyses of the articles. Therefore, conjugate heat transfer simulations of the hydrogen flow through the test article fuel element surrogates were developed to provide estimates of heating, as well as to provide assessments of the fluid flow through the elements. The set of simulations involved not only the separate test conditions, but also a sensitivity study to estimate the effect of physical parameter uncertainty on the selected quantities of interest. Additionally, the fluid flow through the fore section inner plenum dome to the entrances of the fuel elements, another section of the reactor subsystem still under active development, was simulated and analyzed. The simulations show the extent of non-uniformity in the mass flow distribution across the dome outlets. Furthermore, additional analyses estimate the effects of required flow blockages on the potential development of flow dynamics. These analyses were used in the design cycle to make improvements to the reactor subsystem design.

Kalen E. Braman↗
DOE OSTI2018

USHPRR FUEL FABRICATION PILLAR: FABRICATION STATUS, PROCESS OPTIMIZATIONS, AND FUTURE PLANS

The Fuel Fabrication (FF) Pillar, a project within the U.S. High Performance Research Reactor Conversion program of the National Nuclear Security Administration’s Office of Material Management and Minimization, is tasked with the scale-up and commercialization of high-density monolithic U-Mo fuel for the conversion of appropriate research reactors to use of low-enriched fuel. The FF Pillar has made significant steps to demonstrate and optimize the baseline co-rolling process using commercial-scale equipment at both the Y-12 National Security Complex (Y-12) and BWX Technologies (BWXT). These demonstrations include the fabrication of the next irradiation experiment, Mini-Plate 1 (MP-1), and casting optimizations at Y-12. The FF Pillar uses a detailed process flow diagram to identify potential gaps in processing knowledge or demonstration, which helps direct the strategic research agenda of the FF Pillar. This paper describes the significant progress made toward understanding the fuel characteristics, and models developed to make informed decisions, increase process yield, and decrease lifecycle waste and costs.

Wight, Jared M.↗
DOE OSTI2017

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2018, the activities for the INL ART TDO fuel development are focused on final characterization to support the AGR-5/6/7 experiments and technical support to the Generation-IV International Forum (GIF) pre-irradiation leach-burn-leach (LBL) round robin effort with China and South Korea.

+↗
DOE OSTI2022

Initial Observations from AGR 5/6/7 Capsule 1

The fourth and final irradiation experiment in the Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) fuel development and qualification program is designated as AGR-5/6/7. Data collected from the fabrication, irradiation, and post-irradiation examination (PIE) of this tristructural isotropic (TRISO) fuel are intended to serve as the primary data set for the qualification of this fuel for use in high-temperature gas-cooled reactors (INL 2021, Collin 2018b). However, data collected from the three preceding irradiations (i.e., AGR-1, AGR-2, and AGR-3/4) may also be used to supplement data collected from AGR-5/6/7. All components of the AGR-5/6/7 fuel (i.e., UCO kernels, TRISO coatings, and fuel compacts) were produced on an engineering scale at BWXT (Lynchburg, Virginia USA) according to the fuel specification (Marshall 2017). This fuel was irradiated in the northeast flux trap (NEFT) at the Advanced Test Reactor (ATR) at Idaho national Laboratory (INL) from February 16, 2018 to July 22, 2020 (Pham et al. 2021). Measurements in the fission product monitoring system (FPMS) indicated unexpected and significant numbers of failures of TRISO particles in Capsule 1 near the end of the sixth irradiation cycle (ATR Cycle 166A). In the fourth cycle (ATR Cycle 164B) and beyond, the sweep gas flow became very low (presumably from degradation of the capsule gas outlet line via an unidentified mechanism), and the program deliberately isolated Capsule 1 from gas flow periodically. In later cycles, attempts to reestablish any kind of flow in Capsule 1 were unsuccessful. With little or no flow through Capsule 1, FPMS measurements and enumerations of failed particles in Capsule 1 were difficult or impossible as was the ability to control the helium/neon gas mixture used for temperature control. Gas flows and fission gas activity in the effluent gas from the other AGR-5/6/7 capsules were also impacted by the Capsule 1 gas flow issues and the large increase in fission gas released from the Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

U.S. High Temperature Materials Highlights

U.S. GIF VHTR work is continuing on graphite qualification, Alloy 617 regulatory issues beyond the Code space, Alloy 800H weldments, and ASME Codes and Standards R&D is still considering both pebble bed and prismatic and steam generator and heat exchanger U.S. DOE Advanced Reactor Demonstration Program (ARDP) Two U.S.-based teams were selected to demonstrate advanced nuclear reactors in the United States that can be operational by 2027 One of the teams is X-energy (Rockville, MD) which will demonstrate a modular gas-cooled reactor design (Xe-100) with four 80 MWe, TRISO fuel, pebble bed reactors A number of U.S.-based teams were selected to design and develop safe and affordable reactor technologies that can be licensed and deployed over the next 10 to 14 years (Risk Reduction) One of the teams is BWXT Advanced Technologies, LLC which will develop a commercially viable transportable microreactor with the design focused on using TRISO fuel particles and silicon carbide (SiC) matrix A number of U.S.-based teams were selected to assist the progression of advanced reactor designs in their earliest phases (Advanced Reactor Concepts-20) One of the teams is Massachusetts Institute of Technology which will mature the Modular Integrated Gas-Cooled High Temperature Reactor (MIGHTR) concept with a horizontal compact design from a pre-conceptual stage to a conceptual stage to support commercialization

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Direct Reduction of Waste through Refining of DOE Metal Assets - 13632

CVMR{sup R} presents a technology for refining nickel from the enrichment barrier materials of the DOE that is proven through 100 years of use by the metals industry. CVMR{sup R} applies modern controls, instrumentation for process and monitoring of the system, and innovative production methods to produce a wide spectrum of products that generate new technology applications and improvements to our society and economy. CVMR{sup R} will receive barrier materials as a secure operation and size reduce the metal to a shred that is fed to a carbonylation reactor where nickel is reacted with carbon monoxide and generate nickel carbonyl. The carbonyl will be filtered and decomposed with heat to form a variety of products that include high value nano powders, coated substrates, net shapes and pure nickel. The residue from the reactor will retain radionuclides from enrichment activities. The carbon monoxide will only react and extract nickel under the operating conditions to leave volumetric contamination in the unreacted residue. A demonstration plant was designed and built by CVMR{sup R} and operated by BWXT, to demonstrate the systems capabilities to DOE in 2006. A pilot plant operation precedes the detailed design of the nickel refinery and provides essential data for design, safe work practices, waste characterizations and system kinetics and confirms the project feasibility. CVMR{sup R} produces nickel products that are cleaner than the nickel in U.S. commerce and used by industry today. The CVMR{sup R} process and systems for nickel refining is well suited for DOE materials and will provide value through environmental stewardship, recovery of high value assets, and support of the DOE environmental remediation programs as the refined nickel generates additional long term benefits to local communities. (authors)

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Liquid Waste Cell Decommissioning at the West Valley Demonstration Project - 16454

CH2M HILL BWXT West Valley, LLC (CHBWV) began decontamination and deactivation activities in the Liquid Waste Cell in the Main Plant Process Building (MPPB) at the West Valley Demonstration Project (WVDP) in 2013. The work is being done under CHBWV's Phase 1 Facility Disposition Contract with the U.S. Department of Energy, and is a prerequisite for demolition of the 3,716 square meter (40,000 square foot) MPPB, which is planned to begin in 2018. At the start of deactivation activities, the Liquid Waste Cell contained nine large contaminated vessels and approximately 1,220 meters (4,000 linear feet) of piping. Due to high levels of contamination and high radiation dose rates, frequent or prolonged manned entries into the cell were undesirable. CHBWV's strategy for preparing the cell for building demolition was based on conducting dose reduction and contamination stabilization activities using remotely deployed long-reach tools where possible, supported by manned entries where situations warrant or conditions were favorable. This paper discusses the conditions, challenges, decontamination approach, specialized tooling and expected end result of the Liquid Waste Cell deactivation project. (authors)

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

A Scaling-based Applicability Assessment of the CAER Integrated Systems Test Facility

This paper presents a scaling-based applicability assessment of the Integrated Systems Test (IST) facility located at the Center of Advanced Engineering Research in Forest, Virginia. The assessment draws upon both traditional and contemporary scaling methods, seeking an optimal representation of light water reactor (LWR) and small modular reactor (SMR) characteristics during design-basis events. Based on published LWR loss-of-coolant-accident phenomena identification and ranking tables compiled with a similar assessment completed by BWXT mPower, Inc., for the mPower{sup TM} SMR, the phenomena considered in the IST design include break flow, decay heat, loop thermal-hydraulics, and primary-to-secondary heat transfer. Among these, the single- and two-phase natural circulation and system depressurization, by association to loop thermal-hydraulics, were the principal challenges. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Overcoming Technical Challenges During Demolition of the West Valley Vitrification Facility - 19387

In February 2019, CH2M Hill BWXT West Valley, LLC will complete demolition and waste disposal of the first vitrification facility in the history of the United States. This project marked the largest and most complex radiological demolition project undertaken at the U.S. Department of Energy-led West Valley Demonstration Project (WVDP) to date. Provisions put in place during demolition monitored for and mitigated contamination migration during demolition. The work was carried out in a well-planned manner to address specific areas where higher levels of radiological activity were known to exist and where difficult and oversized equipment removal or sensitive items required special handling. Inaccessible contamination on ceiling mounted equipment, controlled removal of two massive steel shield doors, intact removal of lead glass shield windows and a confined demolition boundary area were some of the limiting factors that required special consideration during demolition. This paper contains a discussion of planning and implementation of special handling techniques utilized during some of the more difficult evolutions encountered during open-air demolition of the Vitrification Facility. The work techniques, provisions to mitigate inclement weather and lessons learned have been incorporated into the planning for follow-on radiological demolition work at the WVDP. (authors)

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

Open-air Demolition: Validation of EPA Approved Alternative Method for Emission Estimation - 19397

Emissions from open-air demolition of US Department of Energy (DOE) facilities are regulated by the US Environmental Protection Agency (EPA) under 40 CFR 61 subpart H. This regulation requires that facilities estimate emissions using an EPA approved method from activities such as open-air demolition prior to commencing the activity. Should those emissions exceed certain thresholds, the facility may be required to obtain an Approval to Construct or Modify the facility and/or installed monitoring methods meeting methods acceptable to EPA. There are no EPA approved methods that are specific to open-air demolition methods. For this project the DOE and CH2M Hill BWXT elected to prepare an alternative method specific to open-air demolition. EPA conditionally approved the method but required validation of the method during the demolition of the Vitrification Facility at the West Valley Demonstration Project (WVDP) before use for open-air demolition at other WVDP structures. The validation method included detailed source characterization, close-in ambient samplers, and a comparison of measured results to predicted emissions using AERMOD (American Meteorological Society (AMS) and U.S. Environmental Protection Agency (EPA) Regulatory Model). The methods evaluated include hydraulic hammers, mechanical shears, load out of waste and handling of the debris piles. The validation method involved the collection and analysis of ambient air samples from air samplers located at strategic locations within the projected plume path of the demolition activities. The analytical results from the air samplers were then compared to ambient air concentrations predicted at the sampler locations with the EPA model AERMOD. The input to the AERMOD software is a source term developed weekly using the alternative method and material at risk disturbed during each week. To account for the building wake effects of the surrounding structures, virtual point sources were developed in the AERMOD model and placed at the locations of the demolition, debris piles, debris handling and load out areas. It was found that the factors used for hydraulic hammering were extremely conservative and that emissions from the debris pile were not only the primary emission source, but that the moisture content of the debris pile and the wind speed were critical factors that must be considered in future open-air demolition projects. Based on the results of the validation process, adjustments were made to the alternative method. The final validated alternative method provides an approach for estimating airborne emissions from open-air demolition methods and techniques. The method will be used in lieu of 40 CFR 61 Appendix D to estimate emissions from future open-air demolition projects at the WVDP. (authors)

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

Panel Session 120B: DOE Portsmouth Paducah Project Office (PPPO) Update

This panel focused on the Portsmouth/Paducah Project Office (PPPO) of the US DOE Office of Environmental Management, Lexington, Kentucky, the US DOE Featured Site for Thursday. It showcased recent progress, challenges and planned efforts in radioactive waste management, waste disposition, environmental remediation and D and D at the Portsmouth (Ohio) and Paducah (KY) Gaseous Diffusion Plant Sites, as well as the Depleted Uranium Hexafluoride (DUF6) Conversion Project with facilities at both sites Presentations proposed include: 1 - Portsmouth/Paducah Project Office (PPPO) (Joel Bradburne, US DOE Portsmouth/Paducah Project Office (USA)); 2 - Portsmouth (Jeff Bettinger, US DOE Portsmouth/Paducah Project Office (USA)); 3 - Paducah (Jennifer Woodard, US DOE Portsmouth/Paducah Project Office (USA)); 4 - Depleted Uranium Hexafluoride Conversion Project (Reinhard Knerr, US DOE Portsmouth/Paducah Project Office (USA)); 5 - Portsmouth Deactivation and Demolition Project (Bob Smith, Fluor-BWXT Portsmouth (USA)); 6 - Paducah Deactivation and Remediation Project (Jeff Bradford, Four Rivers Nuclear Partnership - FRNP (USA)); 7 - Seven Lines-One Project (Zack Smith, SNC Lavalin (USA))

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

Removing the Cloak of Secrecy from Legacy Cleanup at Los Alamos national Laboratory - 20395

With the establishment in 2018 of the Los Alamos Legacy Cleanup Contract, the Department of Energy's (DOE) Environmental Management Los Alamos Field Office (EM-LA) and its contractor Newport News Nuclear BWXT Los Alamos (N3B) embarked on an expansive program to engage a broad range of local and regional stakeholders in the decision-making process for cleanup of legacy wastes (pre-1999) at the Los Alamos National Laboratory (LANL). From the inception of LANL during the Manhattan Project until the issuance of the new cleanup contract, the site was managed by a single management and operations (M and O) contactor whose main focus was national security. Segmenting from laboratory operations a separate cleanup contract managed by DoE's Office of Environmental Management (EM) brought new opportunities for the cleanup of legacy waste. That change likewise has opened new opportunities for engaging the community on legacy cleanup decisions. The new legacy cleanup contract - an unclassified contract - emphasizes public outreach, marking a cultural shift in how stakeholders in Los Alamos and throughout northern New Mexico are engaged. Without the constraints of classified work, EM-LA and N3B have increased the level of transparency with the public regarding cleanup activities. EM-LA and N3B have crafted an approach that best meets the values, priorities and needs of the neighboring communities by adapting the model successfully employed at Rocky Flats. Through enhanced stakeholder involvement, EM-LA will implement decisions that have considered broad input and deliver a safe and efficient cleanup of the site. EM-LA and N3B began the expanded stakeholder outreach program in January 2019, leading up to four Environmental Management Cleanup Forums held in the summer of 2019. EM-LA and N3B are now in the process of initiating a series of meetings with various stakeholder groups to determine their 'values' as they relate to cleanup using a novel approach that elicits and records values. The entire process is documented, ensuring that should questions arise about how, when and why decisions were made, they can be easily addressed. In addition to other public meetings to present cleanup scenarios and solicit active public involvement in cleanup decisions, the EM-LA and N3B approach involves discussions with stakeholders who want to explore specific technical issues and cleanup discussions in depth. EM-LA and N3B are also evaluating and implementing a variety of communications tools to help ensure the engagement plan sets the gold standard for transparency in legacy cleanup and stakeholder involvement. (authors)

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

Automated Data Review of Analytical Laboratory Results at Los Alamos National Laboratory - 20299

Newport News Nuclear BWXT-Los Alamos, LLC (N3B) collects samples in support of the U.S. Department of Energy's (DOE) Office of Environmental Management (EM) Los Alamos Legacy Cleanup Contract (LLCC). N3B receives and reviews over 1.6 million sample data points annually in support of various ongoing environmental monitoring and remediation projects of the LLCC. N3B must demonstrate and document that reported external analytical laboratory data produced for the LLCC are of sufficient quality to fulfill their intended purpose and to support defensible decision making as described in EPA QA/G4 Guidance for the Data Quality Objectives Process 1994. In 2018, N3B assumed management of the LLCC along with the Environmental Information Management (EIM) database that contains all historical and current environmental data associated with the LLCC. The entire EIM database is shared between N3B, Triad National Security, LLC (Triad), and New Mexico Environment Department (NMED). These three parties jointly manage the database, its configuration, and changes / updates. All environmental data that are entered into EIM are updated and available, on a daily basis, in the linked public database Intellus New Mexico (Intellus). The quality and defensibility of the environmental data generated from sampling activities is a key component of an effective remediation process. Providing quality data is accomplished through a data assessment process that includes examination, verification, and validation. Examination is the assessment of completeness of the deliverables, identification of any reporting errors, and determining the usability of the data based on the laboratory's evaluation of its data as described in the case narrative received with the data. Verification consists of an evaluation of the Electronic Data Deliverables (EDD) data report to determine the extent to which the external analytical laboratories met method and contract-specific quality control and reporting requirements. Validation consists of determining the data quality and the extent to which the external analytical laboratories accurately and completely reported all sample and quality control results and satisfied all contract requirements. EIM contains an automatic Data Validation Module which performs automated data review (DVM ADR). DVM ADR is a tool to assist in the validation process. When DVM ADR is used in conjunction with manual examination of sample data packages, the combination of the two will meet and exceed the requirements of verification. N3B recognized an opportunity for process improvement, focusing on DVM ADR configuration and enhancements in EIM. Testing EIM's configuration provided proof of the DVM ADR's capabilities and flexibility to accurately perform routine data checks based on analytical methods and regulatory requirements. In addition, the DVM ADR module was improved through enhancements for all analytes, particularly upgrades for radiochemistry data. Extensive testing of the DVM ADR module occurred using EDDs from actual laboratory analyses on the EIM testing site. During this process, N3B manipulated EDD information to verify that the actual outcomes matched the expected outcomes. The results of this testing were shared with the database architects, and configuration improvements were identified to address these results. During this process, N3B identified that the radiochemical DVM ADR capabilities were underutilized, and so enhanced the DVM ADR functionality with respect to radioanalytical assessment. N3B environmental data uploads to Intellus on a daily basis from EIM, once the analytical data undergoes examination and verification. As such, it is important to have a high level of confidence in the quality and defensibility of the data. The process of manual examination, along with the DVM ADR, in conjunction with full validation of a percentage the data specified through the Data Quality Objectives greatly increases efficiency of data review and confidence level of the quality of the data, and gives the project managers, governmental offices, and the public expedited access to high-quality data. (authors)

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

Enhanced electrical penetration components meet the high safety standards required for China's HTR-PM Shidaowan and generation mPower small modular reactors - 300

Enhanced glass-to-metal feedthroughs will be used to carry key instrumentation and control (I and C) signals in China's high-temperature, gas-cooled pebble-bed modular reactor (HTR-PM) at Shidaowan, and in Small Modular Reactor (SMR) prototyping. Chinergy Co., Ltd and Jiamusi Electric Machine Co., Ltd will use the feedthroughs, also called electrical penetration assemblies (EPAs), at the Shidaowan twin-reactor high temperature reactor (HTR). These feedthroughs are manufactured using specialized processes that make them far more robust than counterparts made with organic or ceramic materials. Improved EPAs manufactured by SCHOTT are the only feedthrough solution that can be used in the primary loop of an HTR because organic seals cannot withstand the high temperatures and pressures of this environment. Glass-to-metal seals (GTMS) are made using inorganic, non-aging materials which enable a virtually unlimited lifespan in high-temperature, high-radiation and high-pressure applications. This I and C technology package is being enhanced by the development and testing of new mineral-insulated (MI) cable connectors and EPAs made of crystallized glass-ceramic combined with superalloys. The EPAs have a maintenance-free lifespan of over 60 years, preventing any outage time related to these components and subsequently reducing costs associated with shutdowns. They are far more robust than organic seals and provide increased safety at a lower lifetime cost. Resistance against high pressure and high temperature are especially ideal qualities that make EPAs a valuable choice for environments with high standards for fire protection, such as nuclear plants, subsea oil drilling, or any other underwater application. Due to their significant offering of safety-relevant qualities, the EPAs have also been chosen by BWXT for implementation in its mPower SMR. Small modular reactor requirements are complex and very challenging but are met by the new EPA components, which will help form a strong I and C safety chain. Unlike ordinary compression glass-to-metal seals, GTMS with crystallized glass-ceramic have the ability to withstand temperatures of over 350 deg. C / 662 deg. F and operating pressures of 3500 bar. Furthermore, their lifetime is qualified for up to 60 years, including pressure and radiation requirements in accordance with the latest IEC/IEEE standards. Higher-specification EPAs and MI cable connectors have been designed into severe accident I and C instruments, including advanced hydrogen and oxygen sensors that exceed the challenging requirements of both HTR and SMR applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

RELAP5-3D Solutions to Exercise 1 of the OECD-NEA HTTF Benchmark

Discussion of RELAP5-3D solutions to Exercise 1 of the OECD-NEA High Temperature Test Facility (HTTF) Benchmark. Results from ongoing benchmark activities supported by the Advanced Reactor Technologies - Gas-Cooled Reactor Program with use of the High Performance Computing Center at Idaho National Laboratory. Also discussion of Ultra Safe Nuclear Corporation and BWXT plans for the decade, and issues of modern thermal hydraulic tools for prismatic HTGR modeling. Open discussion of verification and validation and ART-GCR campaign to spearhead the development of an HTGR thermal hydraulics benchmark based on the HTTF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗