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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

ILAW Container Cost Reduction and Optimization Study

This study, performed on behalf of Washington River Protection Solutions (WRPS), examines the potential for making cost savings for the Immobilized Low Active Waste (ILAW) containers over the mission life of the Waste Treatment and Immobilization Plant (WTP). it builds on earlier study assessments 24590-LAW-RPT-M-01-001 Rev 0 ILAW Product Container Specifications Optimization Study and DOE/ORP-2013-02 Rev 0 Decision Analysis of Low-Active Waste Container Finishing Handling System Alternatives. The WTP design and installed equipment has evolved over an extended period ~20 years. Consequently, some of the existing requirements relating the ILAW containers may no longer be applicable under current mission plans where the glass filled containers are placed into the Integrated Disposal Facility (IDF) rather than stacked in an engineered purpose-built ventilated store as originally envisioned. The study team reviewed available WTP project reference documents to establish the functions the container must perform and assign requirements against each of these functions. With the assumptions that some of the existing WTP contract restrictions would not necessarily apply to the subsequent facility operations contract the container requirements were reviewed to identify those which were ‘hard constraints‘, imposed by regulatory requirements. Initially each of these hard constraints were reviewed, evaluated and documented to see if they precluded the use of an alternate less expensive material of construction, carbon steel, to the existing 304 L stainless steel in the current container specification. That review did not generate any fatal flaws for a change in material of construction to carbon steel allowing the study team to examine the impacts of material cost savings, potential changes to the design and requirements and impacts on interfaces of the containers with the existing WTP process equipment from container receipt, transfers, glass filling, cooling and export. In addition, a change in container target fill level was evaluated and qualitative, and quantitative where feasible, impacts to the container fabrication process assessed. Lastly, additional follow on development or testing activities to support the use of carbon steel were assessed and a recommended path forward with a ROM cost and timescale developed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Requesting an Exemption from Standard Compliance: EPAct State and Alternative Fuel Provider Fleet Program Guidance Document

The U.S. Department of Energy established the Alternative Fuel Transportation Program (Program) and associated regulatory requirements pursuant to the Energy Policy Act of 1992. The Program, otherwise known as the State and Alternative Fuel Provider Fleet Program, requires covered state government and alternative fuel provider fleets operating under Standard Compliance to acquire alternative fuel vehicles (AFVs) as a specific percentage of their annual non-excluded light-duty vehicle acquisitions. The opportunity for covered fleets operating under Standard Compliance to request exemptions from their AFV-acquisition requirements serves as administrative relief in the unlikely event a fleet is unable to satisfy its requirements through the normally available compliance alternatives. These alternatives include the acquisition of light-duty AFVs, the acquisition of other, creditable vehicles (e.g., gasoline-fueled hybrid electric vehicles), making certain investments, the purchase of biodiesel for use in medium- or heavy-duty vehicles to the maximum extent allowed, and purchasing or trading for banked AFV credits. This document addresses requests for exemptions from the AFV-acquisition requirements to help covered fleets better understand: How to file a request for an exemption, information and documentation DOE needs to process an exemption request, and important policies relevant for filing exemption requests.

ADVANCED PROPULSION SYSTEMS,ENERGY PLANNING, POLIC↗
NASA NTRS2011

A Human Factors Approach to Bridging Systems and Introducing New Technologies

The application of human factors in aviation has grown to cover a wide range of disciplines and methods capable of assessing human-systems integration at many levels. For example, at the individual level, pilot workload may be studied while at the team level, coordinated workload distribution may be the focal point. At the organizational level, the way in which individuals and teams are supported by training and standards, policies and procedures may introduce additional, relevant topics. A consideration of human factors at each level contributes to our understanding of successes and failures in pilot performance, but this system focused on the flight deck alone -- is only one part of the airspace system. In the FAA's NextGen plan to overhaul the National Airspace System (NAS), new capabilities will enhance flightdeck systems (pilots), flight operations centers (dispatchers) and air traffic control systems (controllers and air traffic managers). At a minimum, the current roles and responsibilities of these three systems are likely to change. Since increased automation will be central to many of the enhancements, the role of automation is also likely to change. Using NextGen examples, a human factors approach for bridging complex airspace systems will be the main focus of this presentation. It is still crucial to consider the human factors within each system, but the successful implementation of new technologies in the NAS requires an understanding of the collaborations that occur when these systems intersect. This human factors approach to studying collaborative systems begins with detailed task descriptions within each system to establish a baseline of the current operations. The collaborative content and context are delineated through the review of regulatory and advisory materials, letters of agreement, policies, procedures and documented practices. Field observations and interviews also help to fill out the picture. Key collaborative functions across systems are identified and placed on a phase-of-flight timeline including information requirements, decision authority and use of automation, as well as level of frequency and criticality.

Kanki, Barbara G.↗
DOE OSTI2026

Reassessing Double-Ended Guillotine Break Requirements: Evidence-Based Analysis of Regulatory Assumptions After Five Decades of Nuclear Operation

After five decades of nuclear power operation encompassing more than 20,000 reactor-years across 35 countries and 647 reactors, zero double-ended guillotine breaks (DEGBs) have been documented in commercial reactor coolant systems—despite DEGB being the fundamental design-basis assumption driving Emergency Core Cooling System (ECCS) sizing, structural protection requirements, and containment design specifications. This report examines the basis for DEGB requirements in nuclear power plant design. The DEGB postulate assumes the instantaneous, complete circumferential severance of the largest diameter pipes in reactor coolant systems, driving major design requirements under 10 Code of Federal Regulations 50.46, General Design Criterion 4 and containment design specifications. The United States (4,880 reactor-years) and France (2,505 reactor-years) contribute the largest operational datasets. Probabilistic assessments estimate direct DEGB occurrence probabilities with extremely low event frequencies, far below the 10-5/reactor-year thresholds typically used to define non-credible events in nuclear-safety analyses; i.e., events with probability this low fall into beyond-design-basis events. Current material-science knowledge demonstrates that the ductile steel materials used in nuclear piping systems exhibit stable crack-growth behavior fundamentally incompatible with instantaneous severance. International regulatory experience, particularly Germany’s comprehensive break-preclusion implementation, and successful leak-before-break (LBB) applications in almost all of U.S. pressurized water reactor units validate that alternatives can maintain safety performance while reducing economic burden. Current DEGB protection systems impose estimated lifetime costs of hundreds of millions of dollars per unit, over the life of a plant across the nuclear industry (including ongoing costs), representing substantial resource allocation toward scenarios with extremely low probability. Although this report acknowledges uncertainties regarding long-term aging effects, potential synergistic degradation mechanisms, and site-specific seismic considerations that warrant continued evaluation as regulatory policy evolves, there remains no documented evidence that a DEGB has occurred as a consequence of the conditions or mechanisms described in this report. This report acknowledges the Nuclear Regulatory Commission’s (NRC’s) recent efforts—outlined in the draft Interim Staff Guidance (ISG) NRC-DSS-ISG-2025-XX (“Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Draft Interim Staff Guidance”)—to reduce overly conservative requirements for large-break loss of coolant accidents through technical justifications and exemptions. However, extensive operating experience and validated methodologies—such as LBB and in-service inspection programs—demonstrate that the probability of a DEGB in reactor coolant-loop piping is extremely low, even under seismic conditions. The authors and reviewers of this report recommend that DEGB be removed as a design-basis event through formal rulemaking, rather than case-by-case exemptions, to better reflect credible failure modes, align with current data, and align with modern, risk-informed safety analysis.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Proposal to use the DOE-STD-3009-2014 Methodology to Prepare the Documented Safety Analysis (DSA) for Research Reactor Nuclear Facilities

Sandia National Laboratories (SNL) personnel operate a low power research reactor (the Annular Core Research Reactor, or ACRR), and a zero-power critical experiment assembly referred to as CX. In accordance with 10 CFR 830, Subpart B, Appendix A, the acceptable methodology for developing a Documented Safety Analysis (DSA) for DOE nuclear reactors is the Nuclear Regulatory Commission’s (NRC’s) Regulatory Guide 1.70 (RG 1.70). RG 1.70 does not address certain areas required by 10 CFR 830 and expected by DOE (e.g., full facility hazard analysis).Thus, the current DSAs for SNL’s reactor nuclear facilities are based on RG 1.70, but also of necessity supplemented by DOE-STD-3009-94 methods. SNL personnel, in consultation with the National Nuclear Security Administration (NNSA) Sandia Field Office (SFO), have concluded that an alternate methodology is preferred to RG 1.70. The details of the proposal, and the reasons motivating its development, are discussed in the order described below. The proposed alternate methodology will be applicable to the ACRR and the CX (i.e., it will be applicable to nuclear facilities in which a reactor and/or a critical assembly will be operated).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

DECOVALEX-2019 (Task E Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. This document is the final report of Task E which was proposed and coordinated by Andra, the National Radioactive Waste Management Agency in France, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. The purpose of Task E of the DECOVALEX-2019 project is to investigate upscaling THM modelling from small-scale experiments (some cubic meters) to full-scale experiments (some ten cubic meters) and finally to the scale of the waste repository (cubic kilometers). To achieve this aim, the data of two in-situ heating experiments performed by Andra (the French National Radioactive Waste Management Agency) in the Meuse/Haute-Marne Underground Research Laboratory (MHM URL) have formed the basis for the understanding of the THM behavior of the COx at different scales. The first experiment provided the reference values of the THM parameters by means of a calibration exercise and they were used for a blind prediction and an interpretative analysis of the second one.

58 GEOSCIENCES↗
DOE OSTI2022

Ageing Management for Extended Long-Term Dry Storage of Spent Nuclear Fuel and Transportation

The principal objectives of the Coordinated Research Project (CRP) and CRADA were to (1) investigate how ageing effects leading to degradation of materials used in the spent fuel dry storage systems could be managed by ageing management programs (AMPs) and (2) using existing AMPs as a basis, establish guidance on how to develop, generate, and maintain AMPs for dry storage systems of spent nuclear fuel (SNF) that can be accomplished in various ways. The PI (Dr. Liu) is an internationally recognized expert on ageing management for license renewal of nuclear power plants and independent spent fuel dry storage installations. Dr. Liu was invited by IAEA to participate in this CRP and served as Chair of the Working Group, supporting the IAEA CRP Lead, with other members from Argentina, the Czech Republic, France, Germany, Hungary, Japan, Pakistan, Spain, Switzerland, United Kingdom, and the United States of America. Argonne’s scope of work in the CRADA for the CRP included (1) ageing management guidance documents developed by Argonne for DOE and used by the Nuclear Regulatory Commission (NRC) and industry and (2) research leveraged from the DOE Office of Nuclear Energy on the mechanical properties of high-burnup fuel cladding and the ARG-US remote monitoring systems technology developed for the DOE Packaging Certification Program, Office of Packaging and Transportation, Office of Environmental Management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1999

Interagency Depainting Study Status

This document discusses coating removal in general. Sections of this presentation include: regulatory background, the initial parameters of the study, the current parameters of the study, stages in each sequence, preparation of the test specimen, material evaluation testing, chemical stripping, CO2 blasting, Flashjet coating removal, plastic media blasting, sodium bicarbonate wet stripping, water stripping, and wheat starch stripping.

Clark-Ingram, Marceia↗
DOE OSTI2025

A Microreactor Program Plan for The Department of Energy

The DOE Microreactor Program was established in FY 2019 to support research and development (R&D) of technologies related to the development, demonstration, and deployment of low-power, transportable reactors to provide power and heat for decentralized generation in civilian, industrial, and defense energy sectors. The program conducts both fundamental and applied R&D to de-risk technology performance and manufacturability readiness of microreactors. R&D projects and work packages are selected to support concept-neutral technology maturation. The intent is to ensure those concepts can be licensed and deployed by commercial entities to meet specific use case requirements. At the same time, the program will also support R&D specific to certain reactor technology groups (e.g., heat pipe reactors and gas-cooled reactors) to ensure relevancy and address the technology needs of commercial developers. The program will ensure coordination of work and activities across participating laboratories and universities, establish, and manage stakeholder interactions, and support program meetings. These stakeholders include, but are not limited to, industry developers, the U.S. Nuclear Regulatory Commission, the Department of Energy, policymakers, and end users. This document provides an overview of the overall Microreactor Program, including its vision, key technical objectives, and scope of the current and proposed R&D portfolio. It covers a 5-year rolling currently from Fiscal year 2025 through Fiscal year 2030. This document will be revised at least biennially to reflect changing priorities.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

Assessment of Adequacy and Completeness of Standards Endorsed by Regulatory Guides to Molten Salt Reactors

The benefits of using voluntary consensus standards and industry guidance documents and reports are evident in the design and licensing of the current generation of nuclear power plants. The use of voluntary consensus standards and industry guidance and reports facilitates the design and licensing of advanced reactors to improve the effectiveness and efficiency of the licensing and regulation of non-light-water reactor (non-LWR) technologies. Most of the regulations, guidance, and standards applicable to nuclear power plants were developed for water-cooled plants, so they may not adequately address factors such as the coolants, materials, temperatures, operations, testing, and maintenance proposed for advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Updates to Nek5000: RANS wall functions, NRC support, and documentation

As a key part of the U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation program (NEAMS), the next generation of tools for nuclear reactor design are being developed. A primary metric for success of this program is the adoption of those tools by industry stakeholders. To help meet this metric for Nek5000, development efforts this year have focused on the expansion of the available Reynolds-averaged Navier-Stokes (RANS) turbulence models to include wall function models, continued support of the U.S. Nuclear Regulatory Commission in simulating a hydrogen mitigation benchmark experiment, and expansion of the Nek5000 documentation to enhance the code’s usability. The implementation of wall functions is intended to significantly decrease time to-solution for a wide range of problems and it has been identified as an important feature by stakeholders across the nuclear community. The initial implementation of standard and pressure corrected wall functions has been tested in Nek5000 on the classical problems of channel flow and a backward facing step. It has also been tested on a 2D molten salt fast reactor core and a T-junction. The recommended formulation of choosing y + = 30 as a boundary value combined with a Neumann-Neumann formulation of the boundary condition for k and τ produces the most consistent results compared to a wall resolved approach. The collaboration with the NRC is a continuing exercise that has been ongoing for multiple years. This year we have concluded investigations into appropriate inlet conditions, showing that fully developed turbulent conditions are adequate for the full PANDA domain. We have also concluded simulations for an unobstructed jet and are now refocusing to the original obstructed jet case. The expansion of the code documentation has also been an ongoing effort over the past few years to address feedback from the various stakeholders. The documentation now more fully covers many of the available features and closes identified SQA gaps. Finally we report on three training sessions that were offered over the past year in an effort to expand the Nek5000 user base.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Molten Salt Reactor Technical and Safety Considerations Outside of Guidance Documents

This document provides information on distinctive characteristics of liquid salt–fueled molten salt reactors (MSRs) for US Nuclear Regulatory Commission (NRC) consideration as it seeks to achieve effective and efficient advanced reactor mission readiness. The NRC has requested that Oak Ridge National Laboratory provide advice regarding technical and safety considerations that might be advantageous to address via policy, rulemaking, or license conditions (i.e., beyond the scope of guidance documents) for accommodating MSRs. It is recognized that MSRs could be regulated based on existing rules, with exceptions, to reflect their distinctive characteristics and technologies. However, directly applying the existing regulatory processes that evolved with the light water reactor fleet to MSRs could be unduly burdensome to the point that their application would inhibit deployment of MSRs in the United States.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Results for the Fourth Quarter Calendar Year 2018 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2018 (CY18) sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) Limits and Targets. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter CY18 sampling of Tank 50 were requested by SRR personnel via a Task Technical Request (TTR) and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is part of Deliverable 2 relating to Task 1 from the SRR request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the TTQAP for the Tank 50 Saltstone task.

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

Update of Receptor Locations for INL National Emissions Standard for Hazardous Air Pollutants (NESHAP) Assessments

This report documents the selection of updated public receptor locations for use in conducting regulatory assessments of radioactive air emissions for Idaho National Laboratory (INL) facilities. The basis for this receptor location analysis is found in 40 CFR 61 Subpart H, “National Emissions Standards for Emissions of Radionuclide Other than Radon from Department of Energy Facilities,” where the requirements for identifying National Emission Standards for Hazardous Air Pollutants (NESHAP) receptor locations are outlined. Receptor locations (residences, schools, businesses, or farm operations) were identified using high-resolution satellite imagery. This assessment identified 31 unique receptor locations for all INL Site facilities, and 30 receptor locations for INL in-town facilities that have the potential to emit radioactive materials. Identification of receptors for each in-town facility is new to this report.

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

Survey of Aging and Monitoring Concerns for Cables and Splices Due to Cable Repair and Replacement

The purpose of this report is to survey aging and monitoring concerns for electrical cable splices in nuclear power plants (NPPs) in long term operation. As portions of existing electrical cable runs in nuclear are replaced over time due to localized events, the total number of splices in NPPs is expected to increase. Relative to cables, the body of knowledge regarding aging of splices and splices in combination with aging cables in nuclear service environments in long-term operations is low. A few reports have considered the aging of cable system components other than cables (Jacobus 1990; Nelson 1998; Villaran and Lofaro 2002), but the nuclear industry has two decades of operating experience since these were published to further enlighten this issue. Herein we discuss electrical cables and splices commonly found in U.S. nuclear power plants, their qualification in safety-related application, and methods for monitoring their health condition. Common environmental stresses that can give rise to cable and splice failure are discussed. The Nuclear Regulatory Commission (NRC) Licensee Event Reports (LER) database was used to identify documented issues of cable and splice failure. The trend in the resultant data over time is considered to see if failures are increasing as plants age. Observations and conclusions of this work include: 1. Cables and splices are highly reliable components. Occurrence rates for events of interest were low and nearly constant over the last 20 years. 2. Common-cause failure for evaluated cable events of interest was observed to primarily be associated with loose connections, which may manifest associated with workmanship issues, thermal cycling, and/or vibration. 3. Replacement of cables is more common than repair, leading to an increase in proportion of new generation cables in the plant over time. 4. Splices on degraded cables have been observed to be problematic. Due to aging NPP infrastructure, including electrical cables, it is expected that such issues will continue to increase. 5. Condition monitoring approaches, while shown to be fruitful for cables, have been shown to be insensitive to degradation of splice sleeves, which are critical to the continued performance of splices. Additional condition monitoring (CM) work is needed to evaluate methods which are sensitive to the degradation of splice components. 6. Extended Material Degradation Assessment (EMDA) knowledge gaps for electrical cables (Bernstein et al. 2014) have not been investigated for splices but may represent similar concerns such as for the accelerated aging process historically used in environmental qualification.

42 ENGINEERING↗
DOE OSTI2023

United States–Brazil Joint Study: A Preliminary Assessment of Opportunities and Challenges for Small Modular Reactors in Brazil

This report supplies the technical information to enable a broader understanding of different technology choices for nuclear reactors in Brazil. The assessment characteristics of this process include reactor technology, economic costs for deployment, and regulatory analyses of small modular reactors, advanced reactors, and microreactors. The analysis documented in this report provides economic and technological information to assist Brazil, whose official indicative scenarios show an increase in operating nuclear power from 2 Gigawatt in 2022 to 8–10 Gigawatt by 2050, in policy making and long-term energy planning. To protect company proprietary information, the information in this report was limited to publicly available information and therefore can only inform preliminary reactor technology assessments. This report does not recommend or endorse any specific reactor developer or technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Monitoring Plan for the Idaho National Laboratory Remote Handled Low Level Waste Disposal Facility

This monitoring plan for Idaho National Laboratory’s Remote-Handled Low Level Waste Disposal Facility was developed to meet the requirements for monitoring low-level waste disposal facilities according to the U.S. Department of Energy (DOE) Order 435.1, “Radioactive Waste Management,” and the guidance provided in the associated technical standard “Disposal Authorization Statement and Tank Closure Documentation” (DOE-STD-5002-2017). The purpose of this monitoring plan is to document a monitoring strategy that includes (1) compliance monitoring activities to demonstrate compliance with regulatory standards/limits and (2) performance monitoring to build confidence the facility is performing as demonstrated in the facility performance assessment (PA) (DOE-ID 2018a), composite analysis (CA) (DOE ID 2012), and CA addendum (DOE-ID 2018b). The de minimus impact to the aquifer predicted by the PA suggests that aquifer compliance monitoring should be augmented with performance monitoring of the drainage course materials and sedimentary interbeds in the vadose zone beneath the facility to provide a more effective means of identifying performance deviations. The monitoring approach delineated in this document was informed by the systems evaluation of natural and engineered facility features presented in the PA, an assessment of aquifer baseline conditions (INL 2017d), the dose analysis conducted in support of the PA and CA, and monitoring data collected during the first four years of facility operations (baseline monitoring phase) (INL 2023b). This plan provides monitoring locations, sampling frequencies, and sampling methods; recommendations for data evaluation; and a description of the monitoring plan implementation. Collected data will be used to demonstrate facility compliance and to identify conditions that are not consistent with the key assumptions made by the PA and CA.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2023

Assessment of Sodium Fast Reactor Specific Consensus Standards and Recommendations for Future Regulatory Development for Standards Activities

The U.S. has designed, constructed, and operated several sodium fast reactors (SFRs). To support these efforts and to document best practices, activities associated with the development of standards and consensus standards were carried out, however, the vast majority of those SFR standards are currently in a withdrawn or inactive status. Consensus standards are useful during design and licensing of reactors, as is evident in the current nuclear power generation fleet. Consensus standards provide certainty and predictability between stakeholders during licensing and other regulatory interactions. These standards offer guidance, reflect expert positions, and establish effective practices for a wide range of topics. Consensus standards are developed by committees with representation from a broad cross-section of industry, organizations, professional societies, and government agencies. However, most existing consensus standards applicable to nuclear power plants were developed specifically for light-water reactor (LWR) designs and may not adequately address specific advanced non-LWR design topics such as those related to SFRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗