Commercial Nuclear Power Industry Research Articles

The dynamic nature of risk in DOE facilities in the surveillance and maintenance program-with observations for risk communications.

We explore three case studies of facilities at two US Department of Energy (DOE) former nuclear weapons research and production sites-the Oak Ridge National Laboratory and the Hanford site-whose risk profiles have changed during their long-term management under the DOE's surveillance and maintenance (S&M) program. These case studies provide examples of the challenges faced in communicating to external stakeholders, such as federal and state regulators, local communities surrounding the site, as well as the general public, the circumstances surrounding unexpected events or the emergence/discovery of new risk-important information at historically high-risk sites. We identify common topics of importance from these case studies and suggest a taxonomy for risk communicators to use in informing the dialogues with individuals and organizations that may not be technically oriented or fully informed on the subject matter. The taxonomy is based on technical insights from the quintessential definition of risk known as the Kaplan-Garrick "risk triplet" as well as insights from regulatory guidance documents on risk communication with external stakeholders originally developed for the commercial nuclear power industry.

CANDU FIRE PROBABILISTIC RISK ASSESSMENT (PRA) MODEL

Fire Probabilistic Risk Assessment (PRA) is being introduced to the fire protection engineering practice both locally and worldwide. The commercial nuclear power industry has also experiencing the impact of this new approach. This paper examines the work performed to assess the relative accuracy of fire models for CANDU nuclear power plant (NPP) applications. The Canadian NPP uses some portions of NUREG/CR-6850 in performing fire PRA. Canadian fire ignition frequencies have been provided by International Fire Data Exchange Project. The CANDU Fire PRA Model can quantitatively evaluate plant damage states and core damage frequencies. This model will assist fire engineers in performing CANDU Fire PRA analysis, by recognizing vulnerabilities related to fire events and will contribute to further improvement of the Canadian NPPs’ safety.

Science-based nuclear design and safety in the emerging age of data based analytics

Over the many decades of nuclear power development, pursuits of reason and observation, appearing as science- and data-based evaluations, have alternated in priority by technologists as they mutually serve to refine the understanding and utility of nuclear science. The rapid advancement of both of these methods principles for nuclear applications, among others, has been synergistic with advances in computing. Stimulated by recent data analytics achievements in the realm of computer visualization, machine learning and the internet-of-the-things, this paper reviews the historical trends of science- vs. data-based analytics with a focus on nuclear-thermal-hydraulics modeling and simulation. The benefit of this progress is then presented in terms of its value to the nuclear industry. An underlying assumption is maintained that sustainable progress relative to present-day investment in research and development is only possible where benefits for the commercial nuclear power industry can be realized.

Risk-Reduction Credit for Very Early Warning Fire Detection at Nuclear Power Plants: From FAQ to Fiction

In 2004, the U.S. Nuclear Regulatory Commission (NRC), with support from the commercial nuclear power industry, adopted the 2001 Edition of National Fire Protection Association (NFPA) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants,” as the means by which commercial nuclear power licensees could comply with Title 10 of the Code of Federal Regulations, Part 50.48(c), to replace deterministic fire protection licensing bases with ones that are risk-informed and performance-based. To facilitate licensee “transitions” to the new licensing bases via NFPA 805, a “Frequently Asked Questions” (FAQs) program, established early during the pilot-plant phase, was expanded to enable use of consensus technical “short-cuts” for fire probabilistic risk assessment (PRA) methods. These “Fire PRA FAQs” enabled licensees, with NRC approval, to bypass more traditional means of establishing acceptable PRA method enhancements on an interim basis, pending eventual confirmation by test programs and/or more detailed analyses. The NRC approved several, of which perhaps the most substantial in providing risk reduction benefits was FAQ 08-0046 on “Incipient Fire Detection Systems,” more accurately characterized as “Very Early Warning Fire Detection Systems” (VEWFDSs). Controversial from the start, the hidden story behind this FAQ’s initial adoption is relevant to examination of the NRC NUREG report that later replaced it and remains in effect today. This article examines this backstory, tracing recommendations that were proposed and bypassed, then examines alternatives to the current guidance. These alternatives, which maximize possible risk reduction credit for VEWFDSs at nuclear power plants, remain at least a factor of two less than the current peak NUREG-2180 risk-reduction factor even before the latter accounts for the possibility of fire pre-emption altogether.

Commercial Nuclear Power Industry: Assessing and Meeting the Radiation Protection Workforce Needs.

This paper will provide an overview of the process used by the commercial nuclear power industry in assessing the status of existing industry staffing and projecting future supply demand needs. The most recent Nuclear Energy Institute-developed "Pipeline Survey Results" will be reviewed with specific emphasis on the radiation protection specialty. Both radiation protection technician and health physicist specialties will be discussed. The industry-initiated Nuclear Uniform Curriculum Program will be reviewed as an example of how the industry has addressed the need for developing additional resources. Furthermore, the reality of challenges encountered in maintaining the needed number of health physicists will also be discussed.

Efficient reduction and modularization for large fault trees stored by pages

Fault Tree Analysis (FTA), an indispensable tool used in Probabilistic Risk Assessment (PRA), has been used throughout the commercial nuclear power industry for safety and reliability analyses. However, large fault tree analysis, such as those used in nuclear power plant requires significant computer resources, which makes the analysis of PRA model inefficient and time consuming. This paper describes a fault tree pre-processing method used in the reliability and probabilistic safety assessment program RiskA that is capable of generating minimal cutsets for fault trees containing more than 10,000 gates and basic events. The novel feature of this method is not only that Boolean reduction rules are used but also that a new objective of simplification is proposed. Moreover, since the method aims to find more fault tree modules by the linear-time algorithm, it can optimize fault tree modularization, which further reduces the computational time of large fault tree analysis.

How to close the US nuclear industry: Do nothing

The United States is on course to all but exit the commercial nuclear power industry even if the country awakens to the dangers of climate change and adopts measures to favor low-carbon energy sources. Nuclear power had been in economic decline for more than three decades when the Bush administration launched a program that aimed to spark a nuclear power renaissance through subsidies and a reformed reactor licensing process. But Wall Street was already leery of the historically high costs of nuclear power. An abundance of natural gas, lower energy demand induced by the 2008 recession, increased energy-efficiency measures, nuclear’s rising cost estimates, and the accident at the Fukushima Daiichi Nuclear Power Station further diminished prospects for private investment in new US nuclear plants. Without additional and significant governmental preferences for new nuclear construction, market forces will all but phase out the US nuclear fleet by midcentury.

Evolution of the Leak-Before-Break Concept in Nuclear Plants: Issues and Resolutions in the Process of Regulatory Actions

Advanced technology may be used to exclude the dynamic effects of postulated pipe ruptures from structural design consideration. However, it must first be demonstrated that the probability of pipe rupture is extremely low under conditions consistent with the design bases for the piping. Demonstration of a low probability of pipe rupture requires a deterministic fracture mechanics analysis that evaluates the stability of postulated small, through-wall flaws in piping and the ability to detect leakage through the flaws long before the flaws could grow to unstable sizes. The concept/methodology underlying such analyses is referred to as leak before break (LBB). LBB could be accepted as a technically justifiable approach for eliminating postulated double-ended primary system pipe ruptures equal to the pressurizer surge line size or larger. Large or double-ended reactor coolant system pipe ruptures equal to the pressurizer surge line size or larger need not consider the dynamic effects of pipe whipping that may result from their failure, following LBB approval of these piping systems. However, LBB may not be applied for the demonstration of adequate emergency core cooling (i.e., calculation of post-loss-of-coolant-accident peak clad temperature and cladding oxidation). Similarly, LBB may not be applied to the determination of containment building pressure and temperature responses to postulated primary and secondary system pipe ruptures or for the environmental qualification of mechanical and electrical equipment. This conclusion has resulted from extensive research and development and rigorous evaluations by the U.S. Nuclear Regulatory Commission, the German RSK, and the commercial nuclear power industry and its organizations since the early 1970s. The LBB concept can be applied to an axial flaw in a pipe, to a circumferential crack, or to when a flaw is stable under normal operating conditions and remains stable when there is a sudden dynamic event (i.e., seismic loading) as a time-dependent inertial LBB analysis. These analyses are deterministic and could be extended to probabilistic evaluations as well. This technical note describes the evolution of the LBB concept, application, issues, and resolutions raised in the process of regulatory actions globally.In this technical note, prior LBB studies in Europe and the United States, performed by various authors and organizations including the International Atomic Energy Agency, are also reviewed and presented. Also included are LBB options and licensing issues raised in the process of regulatory actions in the United States, along with the outlook and perspectives for LBB in the new generation of nuclear power plants.

Dosimetry User's Perspective on Covariance Needs

The dosimetry community has a critical need for covariance matrices to characterize the uncertainty in the input nuclear data used in a range of applications. This need is growing more acute as the nuclear regulatory community increases the rigor with which dosimetry statements must be made and supported. This paper presents and prioritizes a set of needs that reflect the application of neutron dosimetry to the research reactor and commercial nuclear power industry.

A Training Performance Indicator Model For Monitoring the Effectiveness of High-Risk Training: Report of a Study in Progress

High–risk operations, including the nation's commercial nuclear power industry, are devoting extensive resources and leadership to improve human performance through training. The evaluation of the effectiveness of this training is the focus of on–going research. This report summarizes the methodology of a research study in progress at one nuclear utility and de scribes a training performance indicator model that was developed during the study. The model organizes the types of performance indicators currently thought to be worthwhile in monitoring training effectiveness. The model contains twelve categories of indicators. Each category is described and illustrated in this report.

Analysis of Material Inhomogeneity in the European Round Robin Fracture Toughness Data Set

Abstract The European fracture toughness dataset was developed by ten European laboratories in order to provide an experimental data base sufficiently large to study specimen size and temperature effects on cleavage fracture toughness in the ductile-to-brittle transition regime. The initial Master Curve analysis of this data set was presented by Wallin [Eng. Fract. Mech., Vol. 69, 2002, pp. 451–481. showing a range of T0 reference values ranging from −80°C to −110°C, with −90°C as an average and a standard deviation of 6.7°C. The initial round robin data set included all scaled C(T) specimens ranging from 1/2T to 4T in scale, but more recently additional specimens have been tested using this material including deep and shallow crack bend specimens (SE(B)) [Link, R. E., unpublished], precracked Charpy size specimens [Wallin, K., Fracture, Fatigue, and Residual stress—PVP, Vol. 393, J. Pan, Ed., ASME, New York (1999)], and even some biaxially loaded cruciform specimens [Lidbury, D. P. G., Fatigue Fract. Eng. Mater. Struct., Vol. 29, 2006, pp. 829–849]. These specimens are being analyzed to develop and test constraint analysis tools essential in the application of the Master Curve method to structural applications in the commercial nuclear power industry. The present authors have also developed fracture toughness data on this material over a range of loading rates to investigate the sensitivity of the calibrated Weibull coefficients used in the constraint correction models of Gao and Dodds [Eng. Fract. Mech., Vol., 68, 2001, pp. 263–284] to material loading rates in typical nuclear pressure vessel steels. Clear material inhomogeneity characteristics have developed in this data, and it was felt that a more complete investigation of the inhomogenity in the Euro forging data set should be conducted. Following the proposal of Wallin [Eng. Fract. Mech., Vol. 71, No. 16, 2004, pp. 2392–2346], software was written to investigate whether the subject data set could best be expressed as a bimodal distribution of two combined Master Curve distributions, and this software was then applied to the various individual data sets in the European round robin data. The output of each analysis (e.g., the 2T data set) was two Tx values and a probability value pa, where Ta and Tb are the T0 values for the two constituents and pa is the probability of Ta. Using a procedure being developed as an annex to ASTM E1921 which is based on suggestions of Wallin, standard deviations of each of these quantities are estimated and used to judge if the dataset represents inhomogeneous material. The results show that several of the data sets obtained in the original round robin test program are better described by bimodal probability distributions than by the single Weibull distribution of ASTM E1921. Both single specimen datasets, and multi-temperature analysis of the various size specimens appear to show a preference for bimodal probability distributions. Additional sets of precracked Charpy size specimens were then machined from slices of the cross section and the results of testing these specimens shows that some cross sections of the Euro forging demonstrate a clear pattern of toughness variation across the forging cross section, while other cross sections show nearly constant fracture toughness with crack position. A major conclusion of this work is that making “mechanics based” conclusions comparing constraint effects using the Euro forging material is going to be more difficult given the apparent presence of material inhomogeneity in this material as demonstrated very convincingly by the data and analysis presented in this paper.

ENVIRONMENTAL HEALTH PHYSICS: 50 YEARS OF PROGRESS

Environmental health physics is an interdisciplinary field, involving study of the release, transport, and fate of radioactive material in the environment. Further, it addresses the interaction of humans with radioactive materials within the ambient (outdoor) environment and with the environments associated with modern technology and lifestyles. It also involves both naturally occurring and artificially produced radionuclides with the former generally being by far the highest source of exposure. In fact, doses from naturally occurring radionuclides are increasingly being used as a benchmark for the establishment of dose rate limits for people. Because of the pioneering work of early environmental health physicists, models exist today that can be used to assess the potential impacts of new nuclear facilities prior to their operation. In fact, these people represent the branch of the health physics profession who conducted environmental monitoring programs and performed the associated research studies that led to the identification of the principal radionuclides of interest, the major pathways and mechanisms through which they expose people, and the doses that may result from radioactive materials in the natural and technologically enhanced environments. One of their most important contributions was the identification and quantification of many of the key parameters that serve as input to such models. Monitoring of nuclear weapons development facilities used during and after World War II was the initial stimulus for the establishment of environmental health physics programs. Thereafter, these programs were expanded both nationally and globally, as a result of the atmospheric weapons testing programs of nations such as France, the People's Republic of China, the former Soviet Union, the United Kingdom, and the United States. Additional stimuli were provided by the development of the commercial nuclear power industry. Current environmental programs, particularly within the U.S., focus on decontamination and decommissioning of dormant facilities from these earlier defense and commercial programs. The range of the environmental health physics aspects of these activities is the subject of this paper. Presented at the end of the paper is a summary of some of the more important lessons that have been learned. As will be noted, this is an exciting field that will present challenges to health physicists for years to come.

Environmental health physics-50 years of progress.

Environmental health physics is an interdisciplinary field, involving study of the release, transport, and fate of radioactive material in the environment. Further, it addresses the interaction of humans with radioactive materials within the ambient (outdoor) environment and with the environments associated with modern technology and lifestyles. It also involves both naturally occurring and artificially produced radionuclides with the former generally being by far the highest source of exposure. In fact, doses from naturally occurring radionuclides are increasingly being used as a benchmark for the establishment of dose rate limits for people. Because of the pioneering work of early environmental health physicists, models exist today that can be used to assess the potential impacts of new nuclear facilities prior to their operation. In fact, these people represent the branch of the health physics profession who conducted environmental monitoring programs and performed the associated research studies that led to the identification of the principal radionuclides of interest, the major pathways and mechanisms through which they expose people, and the doses that may result from radioactive materials in the natural and technologically enhanced environments. One of their most important contributions was the identification and quantification of many of the key parameters that serve as input to such models. Monitoring of nuclear weapons development facilities used during and after World War II was the initial stimulus for the establishment of environmental health physics programs. Thereafter, these programs were expanded both nationally and globally, as a result of the atmospheric weapons testing programs of nations such as France, the People's Republic of China, the former Soviet Union, the United Kingdom, and the United States. Additional stimuli were provided by the development of the commercial nuclear power industry. Current environmental programs, particularly within the U.S., focus on decontamination and decommissioning of dormant facilities from these earlier defense and commercial programs. The range of the environmental health physics aspects of these activities is the subject of this paper. Presented at the end of the paper is a summary of some of the more important lessons that have been learned. As will be noted, this is an exciting field that will present challenges to health physicists for years to come.

Westinghouse, Master-Lee sign nuclear agreement

Westinghouse Electric Co and Master-Lee Energy Services Corp have announced a multi-year teaming agreement improving the way the two companies provide refueling services, and reactor coolant pump and motor services to the commercial nuclear power industry. Visit www.worldpumps.com for the latest pump industry news

An Evaluation of Task Network Modeling for Use in Simulating Situation Awareness

To achieve its mission of regulating the commercial nuclear power industry, the U.S. Nuclear Regulatory Commission (NRC) periodically commissions research studies to investigate potential safety issues. These include operational studies involving operator and crew performance. A main objective of these research projects is to provide information and knowledge that can be used in the development of regulation, regulatory guidance, and improved methods for collecting and analyzing personnel performance data to assist in regulatory inspections, investigations, and evaluations. The Halden Man-Machine Laboratory (HAMMLAB) in Halden, Norway recently conducted a study using professional nuclear power plant operators to investigate the effect of different alarm systems and display types on operator performance. Dependent variables in the Halden Alarm study included a mix of plant performance measures, operator task performance measures, operator cognitive processes, and subjective evaluations. Operator cognitive process measures included measures of Situation Awareness (SA) using an adaptation of the Situation Awareness Global Assessment Technique (SAGAT). Through the NRC, Micro Analysis and Design was allowed access to the work being performed at Halden in order to evaluate the use of Task Network Modeling (TNM) to simulate human behavior in a nuclear power plant. The alarm system, a turbine trip scenario and data from Halden were used as the basis for the model developed in this project. For part of this project, we also developed a model to simulate SA that was representative of how it was measured in the Halden Alarm study. This model was then integrated into our model that represented a disturbance scenario at a nuclear power plant. Data generated by the task network simulation was then compared to data collected from the nuclear power plant simulator at the HAMMLAB. Statistical tests showed no significant differences between the SA scores of actual operators and those SA scores predicted by the model. Thus, results indicate that a simple representation of SA within TNM can be used to accurately predict situation awareness given this measuring technique. Implications of this finding and recommendations for improving the model are discussed.

The Good Lessons of Bad Experience

New scientific and political realities make it necessary for the United States to continue research and development (R & D) in commercial nuclear power. However, this approach must be based upon a new institutional management structure that recognizes commercial nuclear power as a “Faustian technology,” the management of which requires new forms of institutional controls. This new institutional design should involve a government-private consortium that is aimed at correcting past mistakes made by the institutional management of the commercial nuclear power industry from 1950 to 1975. The new institutional design must (a) involve a multiple partnership between government, industry, environmental interests, and other significant stakeholders; (b) begin with small-scale start-up technologies; (c) include an open public involvement process throughout its duration; (d) aspire to a new standard of institutional constancy; and (e) explicitly socialize management to the political aspects of risk communication as a basic technical preparation. The institutional constancy standard is particularly significant because it involves provisions for managing the intergenerational equities associated with commercial nuclear technology and an explicit recognition of problems associated with the social and temporal distribution of risk.

Comprehensive Guidance for the Evaluation of Human-Systems Interfaces in Complex Systems

Advanced human-system interface (HSI) technologies are being developed in the commercial nuclear power industry. These HSIs may have significant implications for plant safety in that they will affect the ways in which the operator interacts with and supervises an increasingly complex system. The U.S. Nuclear Regulatory Commission (NRC) reviews the HSI aspects of nuclear plants to ensure that operator performance and reliability are supported. The NRC is developing guidance to support its review of these advanced designs. The guidance consists of an evaluation methodology and an extensive set of human factors guidelines which are used in one aspect of the evaluation. The paper describes the guidance development of the evaluation methodology and the guidelines. While originally developed for nuclear plant evaluation, the methodology is applicable to other types of complex human-machine systems as well.

Work Practices, Fatigue, and Nuclear Power Plant Safety Performance

This paper focuses on work practices that may contribute to fatigue-induced performance decrements in the commercial nuclear power industry. Specifically, the amount of overtime worked by operations, technical, and maintenance personnel and the 12-h operator shift schedule are studied. Although overtime for all three job categories was fairly high at a number of plants, the analyses detected a clear statistical relationship only between operations overtime and plant safety performance. The results for the 12-h operator shift schedule were ambiguous. Although the 12-h operator shift schedule was correlated with operator error, it was not significantly related to the other five safety indicators. This research suggests that at least one of the existing work practices--the amount of operator overtime worked at some plants--represents a safety concern in this industry; however, further research is required before any definitive conclusions can be drawn.

The State of Practice of Computerized Operating Procedures in the Commercial Nuclear Power Industry

The State of Practice of Computerized Operating Procedures in the Commercial Nuclear Power Industry

Site-specific Parameter Values For the Nuclear Regulatory Commissionʼs Food Pathway Dose Model

Routine operations at the Savannah River Site (SRS) in Western South Carolina result in radionuclide releases to the atmosphere and to the Savannah River. The resulting radiation doses to the off-site maximum individual and the off-site population within 80 km of the SRS are estimated on a yearly basis. These estimates are currently generated using dose models prescribed for the commercial nuclear power industry by the Nuclear Regulatory Commission (NRC). The NRC provides default values for dose-model parameters for facilities without resources to develop site-specific values. A survey of land- and water-use characteristics for the Savannah River area has been conducted to determine site-specific values for water recreation, consumption, and agricultural parameters used in the NRC Regulatory Guide 1.109 (1977) dosimetric models. These site parameters include local characteristics of meat, milk, and vegetable production; recreational and commercial activities on the Savannah River; and meat, milk, vegetable, and seafood consumption rates. This paper describes how parameter data were obtained at the Savannah River Site and the impacts of such data on off-site dose. Dose estimates using site-specific parameter values are compared to estimates using the NRC default values.