Papers by Robert Youngblood
Journal of Nuclear Engineering and Radiation Science
Dr. Romney Duffey is an internationally recognized multi-disciplinary scientist, consultant, mana... more Dr. Romney Duffey is an internationally recognized multi-disciplinary scientist, consultant, manager, speaker, author, and poet. Born on June 26, 1942, and educated in England, Dr. Duffey has over 50 years of unique experience in the UK, USA, and Canada on nuclear technology development, risk assessment, industrial safety, nuclear-system design, and accident analysis. As an applied physicist, his career has included a wide span of senior power-industry and government positions as researcher, executive advisor, senior manager, published author, lecturer, and consultant. Dr. Duffey is globally known as a developer of new concepts and designs with innovation advantages and market potential, for contributions to risk management and reliability applications, and to the enhancement of our understanding of the physical world. In addition to working in the USA, Canada, and the UK his international industrial, laboratory, and technical connections are worldwide.
This information was prepared as an account of work sponsored by an agency of the U.S. Government... more This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product, or process disclosed herein, or represents that its use by such a third party would not infringe on privately owned rights. The views expressed herein are not necessarily those of the U.S. Department of Energy.
This paper presents methods and findings of a systems interaction study of Indian Point 3. The st... more This paper presents methods and findings of a systems interaction study of Indian Point 3. The study was carried out in support of the resolution of Unresolved Safety Issue A-17 on Systems Interactions. Fault tree methods were employed. Among the study's findings is a single active failure in the low pressure injection function; this discovery led to a plant modification. In addition to providing support to the staff in resolving USI A-17, the project discovered an important new class of failure modes which led the utility to implement a hardware modification. The scope of the project is indicated, key features of the method are highlighted findings are discussed, and comments are offered on the usefulness of this type of, principal study. 9 refs., 1 fig., 1 tab.
Journal of the Royal Statistical Society: Series B (Statistical Methodology), 2010
We address the problem of how to conduct a minimally informative, nonparametric Bayesian analysis... more We address the problem of how to conduct a minimally informative, nonparametric Bayesian analysis. The central question is how to devise a model so that the posterior distribution satisfies a few basic properties. The concept of "local mass" provides the key to the development of the limiting Dirichlet process (or limdir) model. This model is then used to provide an engine for inference in the compound decision problem and for multiple comparisons inference in a one-way analysis of variance setting. Our analysis in this setting may be viewed as a limit of the analyses developed by Escobar (1994) and by Gopalan and Berry (1998). Computations for the analysis are described, and the predictive performance of the limdir model is compared to that of mixture of Dirichlet processes models.
Evaluation of core damage sequences initiated by loss of reactor coolant pump seal cooling
This report is concerned with core damage accident sequences initiated by loss of component cooli... more This report is concerned with core damage accident sequences initiated by loss of component cooling water, leading to loss of reactor coolant pump seal cooling, subsequent primary coolant leakage, and failure to make up the coolant loss. Three plants are considered: Indian Point Unit 3, Midland Unit 2, and Calvert Cliffs Unit 1. It is shown that design differences in
NASA System Safety Handbook, Volume 1: System Safety Framework and Concepts for Implementation
Since its founding, NASA has been dedicated to the advancement of aeronautics and space science. ... more Since its founding, NASA has been dedicated to the advancement of aeronautics and space science. The NASA scientific and technical information (STI) program plays a key part in helping NASA maintain this important role. The NASA STI program operates under the auspices of the Agency Chief Information Officer. It collects, organizes, provides for archiving, and disseminates NASA's STI. The NASA STI program provides access to the NASA Aeronautics and Space Database and its public interface, the NASA Technical Report Server, thus providing one of the largest collections of aeronautical and space science STI in the world. Results are published in both non-NASA channels and by NASA in the NASA STI Report Series, which includes the following report types: TECHNICAL PUBLICATION. Reports of completed research or a major significant phase of research that present the results of NASA Programs and include extensive data or theoretical analysis. Includes compilations of significant scientific and technical data and information deemed to be of continuing reference value. NASA counterpart of peer-reviewed formal professional papers but has less stringent limitations on manuscript length and extent of graphic presentations. TECHNICAL MEMORANDUM. Scientific and technical findings that are preliminary or of specialized interest, e.g., quick release reports, working papers, and bibliographies that contain minimal annotation. Does not contain extensive analysis. CONTRACTOR REPORT. Scientific and technical findings by NASA-sponsored contractors and grantees. CONFERENCE PUBLICATION. Collected papers from scientific and technical conferences, symposia, seminars, or other meetings sponsored or co-sponsored by NASA. SPECIAL PUBLICATION. Scientific, technical, or historical information from NASA programs, projects, and missions, often concerned with subjects having substantial public interest. TECHNICAL TRANSLATION. English-language translations of foreign scientific and technical material pertinent to NASA's mission. Specialized services also include creating custom thesauri, building customized databases, and organizing and publishing research results.
NASA Risk Management Handbook
The purpose of this handbook is to provide guidance for implementing the Risk Management (RM) req... more The purpose of this handbook is to provide guidance for implementing the Risk Management (RM) requirements of NASA Procedural Requirements (NPR) document NPR 8000.4A, Agency Risk Management Procedural Requirements [1], with a specific focus on programs and projects, and applying to each level of the NASA organizational hierarchy as requirements flow down. This handbook supports RM application within the NASA systems engineering process, and is a complement to the guidance contained in NASA/SP-2007-6105, NASA Systems Engineering Handbook [2]. Specifically, this handbook provides guidance that is applicable to the common technical processes of Technical Risk Management and Decision Analysis established by NPR 7123.1A, NASA Systems Engineering Process and Requirements [3]. These processes are part of the \Systems Engineering Engine. (Figure 1) that is used to drive the development of the system and associated work products to satisfy stakeholder expectations in all mission execution do...
The project manager and the authors express their gratitude to NASA Office of Safety and Mission ... more The project manager and the authors express their gratitude to NASA Office of Safety and Mission Assurance (OSMA) management for their support and encouragement in developing this document, the second and final volume of the NASA System Safety Handbook. Building upon the work that resulted in the first volume of this handbook, the development effort leading to this document was conducted in stages, and was supported through reviews and discussions by the NASA System Safety Steering Group (S3G) and by the additional contributors listed below (in alphabetical order).
Nuclear Technology, 2020
The safety goals adopted by the U.S. Nuclear Regulatory Commission (NRC) consist of two qualitati... more The safety goals adopted by the U.S. Nuclear Regulatory Commission (NRC) consist of two qualitative safety goals backed up by two quantitative health objectives (QHOs). The QHOs establish risk limits for severe accidents in terms of their radiological consequences to affected individuals, in particular, the average individual health risks of early fatality and latent cancers from radiation exposure of members of the public living in the vicinity of a nuclear power plant. This paper is devoted to a reexamination of the coverage of the current safety goals as they constrain (or fail to constrain) the total (radiological and nonradiological) risk posed by nuclear power plant operation. Specifically, we suggest the need to address societal consequences. By societal consequences, we mean measures of consequences that reflect the number of people affected and the offsite effects both radiological and nonradiological, not just the individual risks. Recent Level 3 probabilistic risk assessments suggest that given a high likelihood of evacuation of the close-in population before any release occurs the current QHOs are satisfied by large margins, and the experience of an actual severe accident at Fukushima showed that actual human health effects from released radiation were not the dominant consequences, as there were no early fatalities and no measurable increases expected in cancer rates above the baseline rates in the Japanese population. Hence, regardless of accident probability, Fukushima-type accidents with evacuation would satisfy the NRC's health-related safety goals. However, there were very significant societal costs in that large numbers of people were relocated for long periods and there was substantial property damage and community disruption along with the costs of recovery and decontamination. We argue that, in addition to the risks addressed in the current safety goals, societal risk should also be considered. This paper discusses specific possibilities for a goal and an associated quantitative objective.
Progress in Nuclear Energy, 2020
Computational Fluid Dynamics (CFD) is one of the modeling approaches essential to identifying the... more Computational Fluid Dynamics (CFD) is one of the modeling approaches essential to identifying the parameters that affect Containment Thermal Hydraulics (CTH) phenomena. While the CFD approach can capture the multidimensional behavior of CTH phenomena, its computational cost is high when modeling complex accident scenarios. To mitigate this expense, we propose reliance on coarse-grid CFD (CG-CFD). Coarsening the computational grid increases the grid-induced error thus requiring a novel approach that will produce a surrogate model predicting the distribution of the CG-CFD local error and correcting the fluid-flow variables. Given sufficiently fine-mesh simulations, a surrogate model can be trained to predict the CG-CFD local errors as a function of the coarse-grid local flow features. The surrogate model is constructed using Machine Learning (ML) regression algorithms. Two of the widely used ML regression algorithms were tested: Artificial Neural Network (ANN) and Random Forest (RF). The proposed CG-CFD method is illustrated with a three-dimensional turbulent flow inside a lid-driven cavity. We studied a set of scenarios to investigate the capability of the surrogate model to interpolate and extrapolate outside the training data range. The proposed method has proven capable of correcting the coarse-grid results and obtaining reasonable predictions for new cases (of different Reynolds number, different grid sizes, or larger geometries). Based on the investigated cases, we found this novel method maximizes the benefit of the available data and shows potential for a good predictive capability.
Neutron scattering study of pressure-induced antiferromagnetism in PrSb
Physical Review B, 1979
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RISMC Advanced Safety Analysis Project Plan – FY 2015 - FY 2019
In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is d... more In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advancedmore » reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (“why is this important?”) that will make the case for stakeholder’s use of the ASAP research and development (RD (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable “use case” demonstrations; (4) A proposed plan to address ASAP verification and validation (VV and (5) A proposed schedule for the multi-year ASAP.« less
Neutron-scattering study of phonon linewidths in Pd
Physical Review B, 1979
We have performed an inelastic-neutron-scattering study of the linewidths of selected longitudina... more We have performed an inelastic-neutron-scattering study of the linewidths of selected longitudinal phonons in Pd. We observe a sharp maximum in the widths of LA [zeta00] phonons for zeta~0.4. This finding is in qualitative agreement with Pinski and Butler's calculation of the contributions of electron-phonon scattering to phonon linewidths.
Safety considerations in the design of a facility for accelerator-based production of tritium
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Estimation of risk reduction from improved PORV (power operated relief values) reliability in PWRs: Final report
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The PlUS (Process Inherent Ultimate Safety) reactor is an advanced design nuclear power plant tha... more The PlUS (Process Inherent Ultimate Safety) reactor is an advanced design nuclear power plant that uses passive safety features and basic physical processes to address safety concerns. Brookhaven National Laboratory (BNL) pertbrmed a detailed study of the PIUS • design for the NRC using primarily qualitative engineering analysis techniques. Some quantitative methods were also employed. There are three key initial areas of analysis: FMECA, HAZOP, and deterministic analyses, which are described herein. Once these three analysis methods were completed, the important findings from each of the methods were assembled into the PIUS Interim Table (PIT). This table thus contains a first cut sort of the important design considerations arid features of the PIUS reactor. The table also identifies some potential initiating events and systems used for mitigating these initiators. The next stage of the analysis was the construction of event trees for each of the identified initiators. The most significant sequences were then determined qualitatively, using some quantitative input. Finally, overall insights on the PIUS design developed from the PIT and from the event tree analysis were developed and presented.
Review of the Arkansas Nuclear One Generating Station Unit No. 1. Emergency feedwater system reliability analysis
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Is a 4train support ''always'' more reliable than a 2train support
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Review of the Crystal River Nuclear Generating Station Unit No. 3, emergency feedwater system reliability analysis
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Papers by Robert Youngblood