R&D DC Electrical Safety [Slides]
Objectives: Recognize DC and pulsed electrical hazards; Improve risk assessment skills; Concepts to improve worker safety; Engineered vs. administrative controls; tools of the trade.
42 ENGINEERING↗
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Results for “safety management”
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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High Fidelity Simulations of Air-Cooled Reactor Cavity Cooling System
Nuclear energy is increasingly acknowledged as pivotal in the global shift towards cleaner energy solutions. Advanced nuclear technologies, including High Temperature Gas-cooled Reactors (HTGRs), stand out as appealing options among Generation IV reactors due to their high temperature heat output and potential for cogeneration. HTGR designs incorporate passive safety systems, such as the Reactor Cavity Cooling System (RCCS), which utilize natural principles to manage heat dissipation from the reactor pressure vessel (RPV) during accidents or routine shutdowns. Regulatory bodies require thorough validation of safety systems like the RCCS to ensure they meet specified standards. Consequently, there is a pressing need within the industry for advanced simulation tools capable of assessing these systems’ performance accurately. There is a knowladge gap in the literature concerning high-fidelity data for the RCCS, which motivates the focus of this study. This research focuses on a specific RCCS designed for the Modular High-Temperature Gas Reactor developed by General Atomics (GA- MHTGR). Experimental studies on a scaled version of the air-cooled RCCS of GA-MHTGR were conducted by the University of Wisconsin-Madison. This work contributes to a broader initiative aimed at establishing a numerical benchmark based on the UW-Madison experiments. As first step we performed high fidelity simulations of the experimental facility setup, to analyze flow physics in such systems and validate NekRS and the MOOSE heat transfer and radiation modules.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
The DOE/NNSA Nuclear Criticality Safety Program and the Purpose of the Technical Program Review [Slides]
This presentation provides key background information regarding the Nuclear Criticality Safety Program (NCSP), including the program's mission, vision, organization, and technical program element activities. Additionally, it details the agenda of the 2022 NCSP Technical Program Review (TPR) and establishes guidelines.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Brief Overview of the NCSP [Slides]
The mission of the NCSP is to provide sustainable expert leadership, direction, and the technical infrastructure necessary to develop, maintain, and disseminate the essential technical tools, training, and data required to support safe, efficient fissionable material operations within the Department of Energy. The vision of the NCSP is a continually improving, adaptable, and transparent program that communicates and collaborates globally to incorporate technology, practices, and programs to be responsive to the essential technical needs of those responsible for developing, implementing, and maintaining nuclear criticality safety.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Subcritical Assembly at ORNL [Slides]
This presentation discusses the US Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP), specifically regarding the training and education courses. Also discussed is the AGN-201M reactor and fuel disk. Additionally, the ORSA conceptual design and design changes are presented. In conclusion, the ORSA design has moved from conceptual to implementable. ORSA hands-on experiments will remain subcritical under normal and credible abnormal conditions. The ORSA location will have minimal security and safety basis impact and is feasible for a diversity of students. The implementation of ORSA will improve upon NCSP T&E’s vision to remain adaptable and responsive.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Integrating Immersive Visualization in Molten-Salt Reactor Waste Management for Experimental Design and Planning
Molten-salt reactors (MSRs) represent a promising solution for next-generation nuclear energy, offering advantages in safety, fuel efficiency, and waste minimization. However, their liquid-fueled design presents unique challenges for spent fuel management, making post-shutdown waste characterization essential for developing effective strategies. Despite this need, there is a notable absence of visualization platforms specifically tailored to the unique characteristics and analytical requirements of MSR waste management. Existing tools in the nuclear industry are primarily designed for reactor operations or generic data exploration and lack both integration with MSR-specific multiphysics frameworks and the ability to simultaneously visualize time-dependent thermal fields, chemical composition evolution, and radiation distribution patterns. To address these limitations, this paper presents an immersive virtual reality (VR) visualization platform that processes and displays high-fidelity multiphysics simulation output from the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework in real-time, using Unity. The platform visualizes MSR waste characteristics such as nuclide decay, salt cooling, and corrosion by using Exodus II output data and running on a VR headset. It includes a user-friendly interface with features such as visibility toggling, cross-sectional slicing, and time-series animation for exploring simulation data. These capabilities support experimental design, stakeholder engagement, and public communication by making complex reactor behavior more accessible and understandable. By enhancing spatial reasoning and reducing cognitive load, this immersive environment fosters more effective communication and decision-making in MSR waste management.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclear Safety [Vol. 33, No. 4, October-December 1992]
Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 499 Technical Note: Nuclear Regulatory Commission’s Examination Question Bank System, C. L Jensen and N. L. Skinner; 506 Report on the Specialists’ Meeting on Passive and Active Safety Features of Liquid-Metal Fast Breeder Reactors Organized by the International Atomic Energy Agency at Oarai Engineering Centre of Power Reactor and Nuclear Development Corporation, Japan, November 5-7,1991, S. R. Paranjpe; ACCIDENT ANALYSIS: 514 A Review of Hydrogen Production During Melt/Water Interaction in LWRs, D. F. Fletcher, B. D. Turland, and S. P. A. Lawrence; 535 A Review of the U 3 0 8 -Al Reaction as a Potential Heat Source in Research and Test Reactor Accidents, J. L. Snelgrove; CONTROL AND INSTRUMENTATION: 543 Multilevel Interfaces for Power Plant Control Rooms II: A Preliminary Design Space, K. J. Vicente; DESIGN FEATURES: 549 Twenty-Second DOE/NRC Nuclear Air Cleaning and Treatment Conference, R. R. Bellamy, D. W. Moeller, and M. W. First; 562 Book Review: Safety Assessment of Unit 5 (WWER-440/W-213) of the Greifswald Nuclear Power Station, M. H. Fontana; WASTE AND SPENT FUEL MANAGEMENT: 576 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 586 Analysis of Corrective Actions Applied to Nuclear Power Plant Operations, A. E. Cross; 593 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 596 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; 609 Non-Commercial Power Operating Experience; 614 World Nuclear Performance, Compiled by M. D. Muhlheim; RECENT DEVELOPMENTS: 623 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 635 Reports, Standards, and Safety Guides, D. S. Queener; 639 Proposed Rule Changes as of June 30, 1992; ANNOUNCEMENTS: 513 Fourth International Conference on Nuclear Fuel Reprocessing and Waste Management RECOD '94; 585 International Topical Meeting on Advanced Reactors Safety; 622 International Workshop on Implementation of ALARA at Nuclear Power Plants; 648 Short Course on Nuclear Criticality Safety; 648 8th Topical Meeting on Problems in Reactor Physics MEPhi-VOLGA-93; 643 The Authors; 646 Reviewers of Nuclear Safety, Vol. 33
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Overview and Status Update of the Nuclear Criticality Safety Training and Pipeline Program [Slides]
This presentation is on the overview and status update of the Nuclear Criticality Safety Training and Pipeline Program. This lecture includes an introduction to the program, a program overview, and a status update at the end. Steps to complete moving forward are also provided.
96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
Demolition Notifications Environmental Protection and Compliance Waste Management Programs [Slides]
The laboratory creates a variety of wastes, each with its own health and safety concerns, regulatory requirements, and disposal pathways. However, the basic process, from preplanning to waste disposal, is always essentially the same.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Hand Calculations for Nuclear Criticality Safety – Primer Revision [Slides]
This presentation discusses progress on the revision of the Hand Calculations for Nuclear Criticality Safety Primer. The new Primer should be approved for unlimited release by the end of the summer 2022. ORNL and LLNL are working to incorporate the website complement onto the NCSP website.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Mobile Hot Cell Digital Twin: End-of-life Management of Disused High Activity Radioactive Sources
Sealed radioactive sources are used in various settings such as nuclear facilities, universities, hospitals, and industry. When these sources reach the end of their lifespan, they become waste and are recaptured and stored long-term. A digital twin can be used in the design of a solution for recapturing spent sources, offering benefits in deployment time, operator safety, and cost reduction. We propose a digital twin framework for managing the end-of-life process of disused high activity radioactive sources and have created a prototype implementation to demonstrate its feasibility.
61 RADIATION PROTECTION AND DOSIMETRY↗
Modeling interconnections of safety and financial performance of nuclear power plants, part 3: Spatiotemporal probabilistic physics-of-failure analysis and its connection to safety and financial performance
Here, this paper is a byproduct of a line of research by the authors to analyze interrelationships of safety and financial performance of nuclear power plants (NPPs). The result of this line of research is summarized in three parts: Part 1 covers a categorical review of relevant literature and the theoretical bases that support the methodological developments in Part 2. Part 2 introduces an Integrated Enterprise Risk Management (I-ERM) methodological framework to quantify the interconnections of safety and financial performance with a focus on operation and maintenance (O&M) of NPPs. Part 2 has also demonstrated the applicability and values of the I-ERM methodology through an NPP case study. This paper is Part 3, where detailed development and implementation of one of the I-ERM modules, i.e., probabilistic physics-of-failure (PPoF) analysis, and its connection with safety and financial performance is reported. In this article, the physical failure modeling for hardware components is advanced by incorporating finite element analysis (FEA) into PPoF analysis and coupling the FEA-based PPoF with the maintenance performance through a renewal process model. This article covers two scientific contributions: (i) first-of-its-kind incorporation of FEA into the PPoF model of thermal fatigue for NPP components; and (ii) advancing the interface between the PPoF analysis and the renewal process model in order to deal with spatiotemporal FEA outputs and to efficiently estimate the physical transition rates even when the PPoF outputs are dominated by success data. Through the incorporation of FEA, the resolution of the PPoF analysis is enhanced as spatiotemporal conditions such as stress and temperature can be considered explicitly instead of relying on simplified assumptions or analytical models with reduced spatiotemporal dimensions. To demonstrate an application of the FEA-based PPoF analysis and its coupling with maintenance through the renewal process model, a case study is conducted using excess letdown elbow piping in the chemical and volume control system of a Pressurized Water Reactor.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Decay heat analysis for advanced reactor spent fuel transportation and storage applications
Accurate characterization of nuclide inventories and decay heat in spent nuclear fuel is critical for ensuring its safe handling, storage, transportation, and disposal. Although extensive research has been conducted on light-water reactor fuel, advanced reactors present unique challenges due to their diverse core configurations, fuel characteristics, neutron energy spectra, and burnup levels. Building upon previous efforts that developed representative reactor core models for various advanced reactor types and fuels, this study evaluates reactor-specific decay heat characteristics. The results highlight significant variations across advanced reactor types as well as across reactor designs within the same reactor type, and they provide comparison to typical commercial light-water reactor fuel. For example, thermal-spectrum reactor fuels were observed to have an approximately 100-fold decrease in decay heat over the first decade of cooling, whereas the reduction was 10-fold for fast-spectrum reactor fuels. Mass-specific decay heat at discharge can differ by three orders of magnitude among the fast and thermal reactor systems considered. Overall, for the analyzed advanced reactor fuel, fewer than 17 nuclides account for over 99% of total decay heat at 0.5 years, and that number drops to fewer than 7 nuclides at 100 years of cooling. By quantifying reactor-specific decay heat trends and nuclide contributions, this work provides a technical basis to support the development of spent fuel management strategies for advanced reactor fuels as well as safety evaluations for storage, transportation, and long-term waste disposal.
Advanced reactors↗
Preliminary Evaluation of Joint Electricity-Hydrogen Concept of Operations
An initial thermal power dispatch (TPD) concept of operations was evaluated that couples a nuclear power plant to a nearby hydrogen production plant. GSE Systems’ generic pressurized water reactor full-scope simulator was modified with a TPD model comprised of a thermal power extraction and delivery system. A prototype human-system interface (HSI) was developed to interact with the TPD model and allow participants to execute the basic operating scenarios for normal operations. Four retired operators performed the evaluation, and due to COVID-19 travel restrictions, the original in-person experimental design was restructured to support a remote participator evaluation using a web meeting platform. Data from operator feedback, observations from the research team, and quantitative survey responses revealed that the initial TPD concept of operations is feasible. The operators were comfortable with the engineered system and HSI and could manage it without adverse impacts to reactor power, plant safety, or equipment. Findings are discussed in terms of both the TPD system design and HSI performance.
human factors↗
HyRAM+ (Hydrogen Plus Other Alternative Fuels Risk Assessment Models) v.6.1
SAND2025-11565O HyRAM+ (Hydrogen Plus Other Alternative Fuels Risk Assessment Models) is a tool for conducting quantitative risk assessment (QRA) in transportation systems. HyRAM+ contains validated, simplified release behavior models, engineering models, and generic data relevant to hydrogen installations. HyRAM+’s platform integrates models and data to conduct QRA on user-defined hydrogen or other alternative fuel systems. The software will enable the international safety research community to add validated models to the HyRAM+ platform for use in QRAs. HyRAM (hydrogen-only) versions 1.0 to 3.1 were developed by Sandia for the Department of Energy (DOE) Hydrogen and Fuel Cell Technologies Office. The following agencies contributed to the development of HyRAM+ version 4.0 regarding the addition of methane (natural gas) and propane models: the DOE Vehicle Technologies Office and the Department of Transportation Pipeline and Hazardous Material Safety Administration. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.
Groth, Katrina [Sandia National Lab. (SNL-NM), Alb↗
Radiation-Generating Device Safety Self-Study [Slides]
Self-study slides are included for the following modules: RGD Safety Course Presentation, Radiation Detection, Biological Effects, RGDs, Radiation Protection Principles, Production of X-Rays, Protective Measures, and Responsibilities for X-Ray Safety.
96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
Sample Cybersecurity Clauses for EV Charging Infrastructure Procurements
This is the final version of PNNL-34373, with sponsor updates. The proposed sample cybersecurity clauses for EVCI procurements are designed to assist in managing the risk of cyberattacks that may degrade the safety, security, and reliability of EVCI. The sample clauses are intended to be tailored and incorporated into procurement specifications for equipment and services related to the National EV Infrastructure Formula Program deployments. Widespread adoption of the sample cybersecurity procurement language will integrate cybersecurity throughout the life cycle of the infrastructure.
33 ADVANCED PROPULSION SYSTEMS↗
Case Study of the City of Reno: Resilience Valuation and Planning for Solar and Storage on Critical Infrastructure
This case study report documents the City of Reno's efforts to quantify the value of resilience and the entrepreneurial ways through which the City approached the questions of what role solar and storage could play in their resilience efforts and how to value and pay for solar-plus-storage systems. It documents how and why the City of Reno came to the specific conclusion to integrate solar and storage into their Public Safety Center retrofit project and apply for Federal Emergency Management Agency (FEMA) Building Resilient Infrastructure and Communities (BRIC) funding. This report provides an overview for other local governments to draw inspiration from when considering how to prioritize community values and pursue multi-stakeholder dialogue around resilient energy solutions. The City of Reno's experience can serve as a starting point for other communities interested in valuing the resilience benefits of solar and storage.
14 SOLAR ENERGY↗