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At least 469 records · Page 26

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DOE OSTI2013

The AN neutron transport by nodal diffusion

The two group diffusion model combined to a nodal approach in space is the preferred scheme for the industrial simulation of nuclear water reactors. The main selling point is the speed of computation, allowing a large number of parametric studies. Anyway, the drawbacks of the underlying diffusion equation may arise with highly heterogeneous interfaces, often encountered in modern UO{sub 2} and MO{sub x} fuel loading patterns, and boron less controlled systems. This paper aims at showing how the simplified AN transport model, equivalent to the well known SPN, can be implemented in standard diffusion codes with minor modifications. Some numerical results are illustrated. (authors)

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Preparation of Waste Fingerprints for the Miscellaneous Beta Gamma Waste Feeds to the Box Encapsulation Plant at Sellafield - 16080

There are a number of legacy pond and silo storage facilities on Sellafield Site that contain a broad range of ILW arising from the early Windscale Pile and Magnox Fuel Reprocessing Programmes. The waste comprises of solid Miscellaneous Beta Gamma Waste (MBGW) from operations (including filters, containerised waste, thermocouple wire, support struts, pumps, soft wastes) and associated sludges that have arisen from the corrosion of the fuel, Magnox Swarf cladding and the MBGW. It is intended to retrieve and package this waste to meet the requirements of interim storage and final disposal in the Box Encapsulation Plant (BEP) at Sellafield. As part of this process each package must be assigned a realistic and justifiable radionuclide and material inventory. These inventories can either be assigned by radiometric measurements including the associated errors or developed by fundamental data underpinning these assessments and a 'chain of custody' approach of knowing where the waste has been and how it has evolved. The subject of this paper is concerned with the development of the waste fingerprints for these legacy waste streams. As the wastes have arisen over a ∼70 year period there have been an assortment of different working practices and developments in equipment design that have led to significant variations in the waste condition and its composition. During storage the waste items have become mixed, corroded and sludge deposits have arisen. Because of this, and that early consignment records are either incomplete or missing, the task of assigning material and radionuclide fingerprints has been very challenging. In addition, to comply with regulatory requirements the inventory assignment has to be sufficiently robust to meet the processing envelope of the encapsulation process i.e. to allow controls to be made on the amounts of reactive materials (e.g. aluminium, Magnox and uranium) that can be processed within each product and allow identification of excluded items from being consigned from the donor plants. Also, as the waste comprises actinides, fission and activation products it was noted that whilst the UK standard inventory code FISPIN10 predicts the actinides and fission products well, there was a significant discrepancy in the prediction of activation products when compared with those predicted by the specialised fusion code FISPACT. It should be noted that the use of FISPACT to predict actinides and fission products is not recommended. To address these issues, - position statements were produced for each of the waste streams that would be consigned to BEP; - gaps within the data were assessed and where possible inactive trial work was identified to address it; - material and radionuclide fingerprint justification documents were developed from the position statements based upon available data, the experimental work and justified assumptions; - where FISPIN runs were required, these were run with a modified activation data library based on the European Activation File 2010 (EAF2010) to model the activation products. This paper summarises the BEP process, discusses examples of how the material and radionuclide waste stream fingerprints have been justified by revisiting the sample data, original consignment records, engineering drawings and recent remotely operated vehicle (ROV) pond surveys and making reasoned assumptions based on operational practices and non-active experimental work. Incorporation of the EAF2010 library within the FISPIN code has offered significant benefits to BEP and indeed other plants processing mixed fission and activated wastes as it allows comparable predictions of activation products compared with that made by FISPACT and for the plant to maintain one data processing stream rather than two if it had to utilise both modelling codes. This work will underpin the inventory submission in support of the Interim Letter of Compliance (ILoC) to Radioactive Waste Management Limited (RWM). The process of examining the original records, plant operation schedules and engineering drawings has enabled removal of some of the pessimisms that have been made in previous assessments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Aging Management Considerations for Dry Cask Storage Canisters - 18472

Dry cask storage systems (DCSSs) are used to store used nuclear fuel (UNF) at independent spent fuel storage installations (ISFSIs) prior to final geologic storage or reprocessing. As storage durations increase beyond the twenty-year period commonly implemented for licensing the storage sites, it is important to consider aging related degradation issues for DCSS components. A large portion of the DCSS designs used throughout the world employ a ventilated concrete overpack with a welded stainless steel canister that confines the UNF. This paper explores key issues related to aging management for these canister components, including: (1) canister environment, (2) inspection limitations, and (3) aging management program guidance and requirements. While the austenitic stainless steel used for the canisters are generally resistant to most forms of environmental attack, it is possible for conditions to arise that lead to pitting and even stress corrosion cracking of the canister systems. A large number of studies related to canister degradation have been performed by industry groups and government researchers, both in the US and worldwide. These advances in the collective knowledge provide insight into risk factors for DCSS. Such information is beneficial to improve ISFSI siting decisions and ensure successful relicensing of existing ISFSIs. This paper highlights relevant research and also discusses applicability to degradation of other equipment and containers used to process and isolate UNF and other nuclear waste streams from the public and the environment. Inspection of the canister surface is an important tool to investigate and evaluate the potential for degradation. Existing DCSS designs typically do not offer ready access to the canister surface in order to perform the inspection. Further complicating inspection efforts are the surface conditions present on the sheltered canisters after extended storage times, including atmospheric debris such as dust and pollen. Given ongoing development efforts, inspection capabilities and associated limitations should be assessed at multiple points in time over the course of an aging management program to incorporate new advances in the field. This paper explores the limitations of currently available capabilities in inspection and delivery, and provides insight into the gaps remaining to be filled to permit an effective inspection program. Recommendations and requirements related to programs for aging management of canisters have been published or are in progress from a variety of groups, including industry groups, regulatory authorities, and Codes and Standards. An effective aging management program should consider the technical bases of these various inputs and maintain compliance with requirements. Given multiple sources, decisions will be required when considering conflicting guidance, and this paper will include discussion of the relative merits of the various approaches. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

WIPP Safety Significant Confinement Ventilation System (SSCVS) Quality Assurance Program - 20237

Construction and major equipment fabrication are underway for a new ventilation exhaust system at the Waste Isolation Pilot Plant (WIPP). The new system will provide continuous HEPA filtration of the exhaust from the WIPP underground while meeting the ventilation demands for expanded waste emplacement and mining operations over the next 40 years. The SSCVS project has proceeded with the awarding of CD-3A for Long Lead Procurements followed by CD-2/3 approval for the construction of the facilities. The SSCVS project team made the decision to increase the quality oversight throughout the complete project per the Project Quality Assurance Plan (PQAP) such that it can ensure the final products meet applicable NQA-1 requirements and general quality requirements, and that the project will proceed to operations without significant delays due to post installation quality concerns. The PQAP was established prior to approval of the project at the CD-2/3 milestone. The SSCVS project has 2 major long lead procurements which commenced with the awarding of CD-3A. These procurements include the HEPA filter housing assemblies and the Exhaust Fans. There are 22 HEPA filter assemblies for the SSCVS, which are designated Safety Significant for pressure boundary confinement, per the approved PDSA. The exhaust fans also have Safety Significant components to maintain the pressure confinement boundary. NWP made the decision to install resident Quality Inspectors at the fabrication facilities to ensure the documentation for each fabricated unit meets all the QA requirements as defined in the respective contract documents. For the construction of the SSCVS Salt Reduction Building (SRB) and New Filter Building (NFB), dedicated, on-site quality inspectors have been utilized to ensure that all work being performed will comply with quality requirements and the Special Inspections as defined in IBC-2015. At each stage of inspection, hold points are placed into the work packages to ensure NWP QA personnel are present to inspect and verify the installations are in accordance with the design specifications, codes and standards. With this level of oversight, NWP anticipates that the level of quality related issues, as seen on multiple other DOE complex site projects, will be avoided and that the commissioning, start-up and hot operations will progress in a timely fashion, as all documentation will be in place and approved prior to initiation of these activities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Increasing Industry Involvement: A Solution from a Student's Perspective

Many engineering employers are finding that graduates are not sufficiently equipped to begin contributing productively when they exit a 4-year institution and enter the workforce. The Intern Professional Development Center (IPDC) is a proposed business model that exposes engineering undergraduates to industry-relevant knowledge, while providing them with working experience during their degree. Using a tri-stage approach, the IPDC engages with students early in their academic careers, provides continuous training and development opportunities, and finally connects the students with potential employers. The two topics most commonly under-represented in engineering programs are: - Familiarity with drafting software - Knowledge of relevant codes and standards. The Intern Professional Development Center (IPDC): A tri-stage solution for recruiting, training, and encouraging the workforce of tomorrow. An IPDC allows students to be active members of industry during their undergraduate careers, leading to: - Improved local talent retention; - Improved graduation and academic performance statistics; - Near-guarantee for post-graduation employment. In March of 2019, the first IPDC prototype began operating in the Tri-Cities, Washington. Since then, 3 undergraduate students have entered the IPDC at Stage 2, beginning with the standard training and professional development. Of those 3 students, 2 have already moved on to Stage 3, with successful ongoing internships with local employers. At a performance review for one of the IPDC's Stage 3 students after 90 days of employment, the direct manager said: 'How soon can we get another one?'.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

DOE OSTI2017

Characterizing the Benefits of Seismic Isolation of Nuclear Structures in Terms of Reduced Risk and Cost

Seismic isolation is a mature technology that has been implemented in many civil structures, including buildings, bridges, liquid natural gas tanks, and off shore oil platforms, both in the United States and other countries, to mitigate the damaging effects of earthquakes. Seismic isolation has also been used in nuclear structures in France and South Africa, but not yet in the United States: neither in the Department of Energy facilities nor in commercial nuclear power plants (NPPs). This is primarily due to a lack of guidelines, codes and standards for the analysis, design and construction specific to seismically isolated nuclear structures. However, seismic isolation of nuclear structures has seen increased research interest in the recent years and the recently published national consensus standard, America Society of Civil Engineers (ASCE) Standard 4-16 'Seismic analysis of safety related nuclear structures', now incorporates language and commentary (Chapter 12) for seismically isolating surface or near-surface-mounted nuclear facilities, including NPPs. Seismic isolation substantially reduces horizontal seismic loads (demands) on structures, systems, and components. Reduction in demand results in four potential benefits: (1) economic: reduction in capital cost, (2) increased safety: reduction in the mean annual frequency of unacceptable performance, (3) insurance: protection against increases in the known seismic hazard after construction by minimizing the effort to re-qualify and re-certify structures, systems and components, and (4) recertification: the opportunity to certify an existing NPP design for a region of higher seismic hazard. Item (2) above, an increase in safety (reduction of seismic risk), was explored in Huang et al., wherein it was demonstrated that the implementation of seismic isolation reduced seismic risk in nuclear power plant structures. Studies that assess item (1) have not been performed although the use of isolation may lead to large reductions in the capital cost of safety-related nuclear facilities. Funding provided by Nuclear Safety Research and Development in the Department of Energy (DOE) allowed the authors to develop a framework for assessing the economic benefits and reductions in seismic risk afforded by the use of seismic (base) isolation. The framework includes probabilistic risk assessment and estimation of overnight capital costs (OCC's) of a sample generic DOE nuclear facility (GDNF). The project is documented in detail in Bolisetti et al. and Yu et al., and key results are presented here. The GDNF considered in this study is based on a facility constructed at the Idaho National Laboratory. The GDNF is a two-story reinforced concrete building with an embedded basemat and is founded on soft rock at the boundary of Site Class B and Site Class C (B/C boundary) as characterized by ASCE/SEI Standard 7. The GDNF is assumed to handle radiological materials, and its structure, systems and components (SSC's) are designed to effectively confine these materials in the event of an internal failure. The GDNF is populated with components that are generic to safety-related nuclear structures. A single system, which is intended to confine a loss of materials-at-risk (MAR) to the interior of the structure, is chosen for this study. Accordingly, a hypothetical event tree and fault tree are assumed for risk assessment. The isolation system for the GDNF is designed per Chapter 12 of ASCE/SEI 4-16 but is not optimized. The isolation system consists of 38 lead rubber bearings installed in one horizontal plane below the basemat. Details of the GDNF, including the SSC's, event tree and fault tree, and isolation system, are presented in the project report and Yu et al.. This study includes seismic risk assessment and cost estimation for four cases: (1) the GDNF is located at a site of low to moderate seismic hazard and is constructed on a conventional (non-isolated) foundation; (2) the GDNF is located at the same site as case 1, but is seismically isolated; (3) the GDNF is located at a site of high seismic hazard and is constructed on a conventional foundation; and (4) the GDNF is located at the same site of high seismic hazard as case 3, but is seismically isolated. In this study, the DOE site at Idaho National Laboratory (INL) is chosen as the low to moderate seismic hazard site for cases 1 and 2, and the Los Alamos National Laboratory (LANL) is chosen as the high seismic hazard site for cases 3 and 4.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Implementation of a generalized coherent elastic scattering formulation for thermal neutron scattering analysis - 14487

The scattering interactions of thermal neutrons in matter are classified into inelastic and elastic components. In crystalline materials, coherent elastic scattering represents a dominant form of interaction. Traditionally, the treatment of this interaction in standard computer codes (e.g., the LEAPR module of the NJOY code system) was performed within the cubic approximation and restricted the analyst to a few materials that were hardwired into the code. In general, these restrictions can hinder the production of ENDF style thermal neutron scattering libraries for new materials outside the current portfolio of treated materials. In this work, a general formulation based on the exact theory of coherent elastic scattering was developed. This formulation is capable of processing the crystalline structure of any material. The cubic approximation is relaxed by accounting for the partial forces from asymmetric crystal structures in the calculation. The atom sites approximation is relaxed by utilizing a site-wise Debye-Waller factor. Specifically, the Debye-Waller factor is no longer regarded as a uniform constant across the crystal but treated as a matrix tensor that uses the integral of the partial phonon density of states. Therefore, the calculation of the coherent elastic scattering cross section is executed by applying the matrix-form of the Debye-Waller factor on each reciprocal space point and summing them up. Furthermore, the developed algorithm supports compounds with any number of elements and of any ratio to each other. This general formulation was successfully implemented and contributed to the production of ENDF style thermal neutron scattering libraries for silicon dioxide (α-SiO{sub 2}) and silicon carbide (3C-SiC). The needed input for the analysis was generated using density functional theory (DFT) atomistic simulations and dynamical matrix calculations for both materials. (authors)

42 ENGINEERING↗

DOE OSTI2024

Leveraging NREL's ResStock & ComStock Dataset to Evaluate Building Stock Electrification: Preprint

Residential and commercial buildings accounted for 40% of U.S. energy consumption in 2022 and represent a significant opportunity for decarbonization through energy efficiency and electrification, and for grid planning. Building stock energy modeling is a powerful tool that can evaluate what-if scenarios as utilities, municipalities, policymakers, building owners and others work towards equitable building decarbonization and climate goals. This presentation will highlight several high-impact use cases of the National Renewable Energy Laboratory (NREL)'s highly granular, bottom-up building stock energy modeling tools, ResStock and ComStock. These use cases cover a wide range of project scale, from neighborhood electrification analysis and municipality long-term energy planning, to state energy code development and national policy evaluation. This presentation will showcase specific real-world applications for which ResStock and ComStock have been utilized across the country, including California codes and standards cost-effectiveness analysis, New York City affordable housing electrification cost gap analysis, and California targeted electrification and gas decommissioning analysis. For each use case, this presentation will illustrate how ResStock and ComStock played a crucial role in accurately characterizing regional building stocks, providing discrete and aggregated end-use load shapes, and calculating lifecycle consumption, emissions, and costs for a variety of building electrification strategies and scenarios. Finally, this presentation will demonstrate how the data provided by ResStock and ComStock can help unlock significant outcomes for these use cases, including but not limited to, customer bill impact, incentive and program design, and energy equity analyses.

building stock modeling↗

DOE OSTI2025

Graphite for Advanced Nuclear Reactors: Deployment Readiness Review

Historically, graphite has been used in numerous reactor technologies, including research/test reactors and, commercially, in advanced gas reactors. These non-metallic materials can play a key role as internal core structures, reflectors, or neutron moderators, making them important for deployment of certain advanced reactor technologies. This report explores the industry readiness for graphite material deployment, documenting Codes and Standards applicable to their design, qualification, and manufacturing. Discussions also examine the manufacturing processes, aging/degradation, inspection techniques, and disposal options. The report covers the current status, documents gaps, and proposes some conclusions about approaches to managing some of the current gaps.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Battery Safety Factsheet

A brief overview of battery safety topics presented as a factsheet for a community audience. Battery safety topics include: battery management systems, thermal runaway, what to do in a battery fire emergency situation, battery storage considerations, commissioning, and battery codes and standards.

Battery Energy Storage↗

DOE OSTI2023

SPC-70692 Rev 0 MARVEL B4C Powder Specification

A. Idaho National Laboratory (INL) is developing a microreactor to produce electrical power utilizing a small nuclear core and Stirling engines under the Microreactor Application Research Validation and Evaluation (MARVEL) program. This procurement specification defines the requirements for procurement of the B4C Powder. B. Any conflict between this specification and referenced Codes and Standards, or any supplementary specifications in the procurement documents requires written clarification from the Contractor prior to proceeding with any work. Any deviation from the procurement documents requires approval by the Contractor with the change request process.