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The effects of time dependent failure rates caused by the aging of components are becoming increasingly important in probabilistic risk assessment (PRA) and reliability analyses of nuclear power plant systems. In the NRC Nuclear Plant Aging Research (NPAR) program, the effects of aging in nuclear systems are being evaluated through the use of time varying failure rates that are determined as a function of the age of the system. These analyses involve complex systems and include various sensitivity studies; thus, the PRAAGE88 computer code was developed to facilitate these calculations. PRAAGE88 is an IBM PC based code that computes system unavailability, component unavailability, and various importance measures for use in evaluating the effect of aging on reactor systems. This paper describes the methodology utilized in the code, its capabilities and areas of application.

Event trees are key elements in applying PRA in pre-conceptual design for safety enhancement of Oak Ridge National Laboratory's Advanced Neutron Source (ANS) research reactor which will be the world's best source of low-energy neutrons for material studies, physics research, transplutonium production and radiation effects. Event trees exhaustively explore the combinations of system failures that may follow from accident initiator through dendritic diagrams connecting initiator with plant damage state and by calculating the frequency of each accident sequence. They are the central element of PRA, but require supplementary text to relate the sequences to plant systems, explain the logic and state assumptions. These tasks are accomplished by the integration of a popular word processor and the event tree analysis code BETA. Using the worl processor, the analyst writes text that describes each sequence to a depth deemed necessary and prepares an Event Table which indicates by symbols the operability of each system in the accident sequence and the judged fuel damage state. The matrix of symbols so formed is copied to another file, edited to remove spaces, additional information is added and the file is saved in ASCII format to provide input to BETA. BETA, which may be used for further editing, displays or plots on HP laser or Epson printer, the event tree specified by the Event Table logic and computes, displays and/or prints the end or intermediate state probabilities according to damage criteria. It also provides for changing nodal probabilities within the event tree according to the preceding sequences (‘binary conditionals’). The size of event trees is essentially unlimited by its ability to branch and link with other event trees. Work sessions may be saved to floppy or hard disk for recall. This paper describes how to use BETA, and how it works.

A double guillotine break of primary piping when a reactor is at full power has long been considered a bounding accident. But just what is its probability of occurrence? This paper presents a review of attempts to assess the probability of this never occurring event from WASH-1400 through the Thomas model to efforts using probabilistic fracture mechanics as exemplified in the PRAISE code which resulted in a revision of the Code of Federal Regulations to recognize the tendency of pipe to leak-before-break. The interactive PC code PRAISDPD implements the PRAISE algorithms and was used to perform safety optimizing parametric studies of Oak Ridge National Laboratory's Advanced Neutron Source (ANS) reactor. From this, the sensitivity of the pipe break probability to the piping parameters were determined and suggestions were made to minimize the break probability. Qualitative conclusions are: 1. • Pipe lengths should be minimized. 2. • Large diameter pipe is more reliable than smaller pipe because of the greater wall thickness through which a crack must grow and the larger circumference around which a crack must grow before becoming unstable with subsequent breakage. 3. • The probability of leakage and breakage is significantly reduced by periodic non-destructive examination (NDE—specifically ultrasound testing). 4. • Aging increases the probability of leak/break.

Advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semipassive protection. Reliability assurance of the advanced reactor system is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. The authors discuss several concerns for reliability assurance of the advanced design, encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance, and test requirements. Suggested approaches are provided for addressing each of these topics. >

Advanced nuclear power reactors will use different approaches to achieving a higher level of safety than the first generation. One approach uses the technological developments in computation and electronics in the form of digital instrumentation and control (I&C) to enhance the reliability and accuracy of information for plant control, responding to the information, and controlling the plant and its systems under normal and upset environments in various states of degradation. Evaluating the reliability and safety of advanced I&C systems requires determining the reliability of the I&C used in the advanced reactors, which involves distributed processing, data pile-up, interactive systems, the man-machine interface, various forms of automatic control, and systems interactions. From these analyses will come an understanding of the potential of the new I&C, and protection from its vulnerabilities to enhance the safe operation of the new plants. Technological, safety, reliability, and regulatory issues associated with advanced I&C for the new reactors are discussed, and suggested approaches to their resolution are presented. >

Chemical accidents consist of initiation, propagation, and termination. Eliminating the initiation of chemical accidents is the most effective step to prevent them. In spite of the best efforts, accidents do occur. However, their propagation needs to be suppressed. If that cannot be done, the accident must be terminated. To appreciate the magnitude of chemical safety, the chapter illustrates a few case studies on chemical accidents, as well as the losses that these industries incurred. The causes of accidents ranged from the lack of knowledge about explosive potential to the absence of diffuse-control system. This chapter describes the steps involved in probabilistic safety assessment of chemical processes.

Process safety management (PSM) is a systematic approach to prevent unwanted releases of hazardous materials that can affect workers and people. PSM includes process technology, operating procedures, operational and maintenance activities, nonroutine activities, emergency preparedness plans, and training. Process safety management anticipates, evaluates, and mitigates chemical releases that may result from failure of process procedures or equipment. This chapter discusses the PSM techniques and rules that can be used in nuclear and chemical reactors. The strong kinship between chemical and nuclear engineering results in the similarity of their probabilistic safety assessment techniques. The chapter describes the qualitative methods of accidental analysis, such as checklists, what-if analysis, hazards and operability (HAZOP) analysis, and failure modes and effects analysis (FMEA). The quantitative methods of accidental analysis include parts count, FMEA, reliability block diagram, fault tree, and event tree. The chapter also discusses the common causes of reactor failures and their solutions by analyzing and quantifying these causes.

The primary motivation of probabilistic safety assessment (PSA) is to assess the risk of a nuclear or chemical plant. PSA precisely provides relative risk comparisons of reactor designs and sites. The first well-known use of PSA in a regulatory process occurred in 1978, when the probabilistic analysis staff performed a study to categorize the existing technical and generic issues facing the Nuclear Regulatory Commission (NRC). The purpose of risk-based regulation is to protect people from the risk of nuclear power plant accidents. PSA makes it possible to express the risk numerically; however, the NRC regulations are proscriptive in nature as their aim is to achieve an unknown risk level. In reaching decisions, the regulator compares the credibility of the information gained from PSA with the credibility of other sources of information, including the magnitude of the estimated risk, the degree of uncertainty associated with the risk, the results of sensitivity analyses, and the net estimated effect of proposed alterations on the overall safety of the plant. The PSA insights provide the decision maker with a comprehensive and contextual definition of the problem and a better quantitative understanding of public risk with its uncertainties. The PSA objectively judges the aspects of a risk.

Probabilistic safety assessment (PSA) is an analytical method to protect public health and safety. This chapter discusses the definitions of the terms related to public safety and illustrates a few mathematical representations that help in calculation of risks, along with the shortcomings of mathematical forms. The differences between calculated risks and public perception of risks are also discussed. Furthermore, the chapter discusses the safety goals and regulatory bases required for nuclear power and chemical processing. The chapter illustrates PSA results and provides critiques of past PSAs along with the ways to improve them in the future. The chapter also gives an overview of the U.S. regulations that govern radiological and chemical materials to protect people and the process safety management (PSM) rules that provide guidelines for safe operation of chemical process facilities.

This chapter describes the way the effects of nuclear accidents are calculated as an accident progresses. The progression includes events such as radionuclide release to radionuclide migration, escape of radionuclide from retaining structures, transport of radionuclide to atmosphere, and the effects of radionuclide on public health. A reactor core's fission product inventory is the primary source of radioactivity from which people are protected by independent barriers such as fuel matrix, fuel clad, cooling water, reactor-cooling system, and emergency action. Any release of radioactive materials affecting public health requires temperature above the melting point of the materials to deform the reactor core and confining structures. The chapter describes the thermal-hydraulic calculations—before, during, and after a nuclear accident—and some other calculations done to estimate the radioactive release from containment. The transport and dilution of radioactive aerosols, vapors, and gases released into the atmosphere from a nuclear power station are functions of the atmosphere along the plume path, the topography of the region, and the characteristics of the effluents themselves. Two basic approaches are used in atmospheric diffusion modeling—the gradient-transport theory and the statistical theory.

The assembly process of probabilistic safety assessments (PSAs) brings together all assessment tasks to assume a system's risk, its significance, its sensitivity to uncertainties, its confidence limits, and the methods by which these risks can be reduced by system improvements. Not all PSAs use analysis programs such as the fault trees and the event trees; this is especially true with regard to chemical PSAs that may rely on hazards and operability (HAZOP) analysis or failure modes and effects analysis (FMEA). However, the objectives are the same, that is, accident identification, analysis, and evaluation. The validity of the risk of a nuclear reactor that is estimated by the reactor safety study (RSS) came under questions when the Three Mile Island-2 (TMI-2) accident occurred, which showed the misunderstanding of the probabilistic nature of risk. A single event neither proves nor disproves a statistical result; however, it questions the usefulness of risk as a safety measure.

This chapter gives an account of the uses of mathematics in probabilistic safety. Probabilistic safety evaluates the harmful effects of an artificial construct on human beings. This construct can be called an airplane, a plant, a facility, or a system, depending on the degree of its complexity. A logical system model consists of the important components of a system and the effects that the component failure has on the system's operability. This model treats each component as either working or not working. Hence, the state of the system may be represented by a logical equation composed of the states of its components. The chapter uses a set theory and a Venn diagram to show relationships between components and the system and to show the importance of the mincutset to calculate system probability. A system's risk is the product of its failure probability and the consequences of its failure. This chapter focuses on logic modeling and probabilities. Logic modeling uses the Boolean algebra that is based on two-state variables, whose reduction rules are explained in detail. Other concepts discussed in this chapter include definitions of probability, methods for combining probabilities, calculating failure rates from inspection and incident data by classical and Bayes statistics, treating of uncertainties as distributed variables, and calculating confidence intervals.

This chapter discusses the calculation of chemical releases and the dispersion of hazardous materials from chemical-process accident. Chemical-process materials, regardless of toxicity, are benign as long as they are confined within the structure. They may release mostly by rupturing the confinement through designed exits, such as release valves. Chemical explosions are classified as uniform or propagating explosions. An explosion in a vessel tends to be a uniform explosion, while an explosion in a long pipe is a propagating explosion. Explosions can be either deflagrating or detonating. A detonation generates greater pressures and is more destructive than a deflagration. If a vessel containing a pressurized gas or vapor ruptures, the release of the stored energy of compression produces a shock wave and flying vessel fragments. The energy that is released is in the form of (1) kinetic energy of fragments, (2) energy in shock wave, (3) heat, and 4) the energy of deformation.

Publisher Summary This chapter presents an overview of the techniques used to incorporate external events into a probabilistic safety assessment (PSA).The discussions in the chapter are primarily aimed at nuclear power plants but are equally applicable to chemical process plants. The types of external events are earthquakes, fires, and floods. The external events are the accident initiators that do not fit into the central PSA structure used for internal events. Some external events, such as fire from ignition of electrical wires or flood from a ruptured service water pipe, occur inside a plant. Other events, such as earthquakes and tornadoes, occur outside the plant. Any of the external events may cause failures in a plant like internal events. External initiators may cause multiple failures of independent equipment, thereby preventing the action of redundant protection systems. Tornadoes are like missiles that affect structures and cause failures of systems. Analysis methods are similar for all external events—probabilities of external events, probabilities of failures, effects of failures on safety systems, and estimation of the effects of failures on workers, common people, and environment.

These lecture notes for criticality safety are prepared for the training of Department of Energy supervisory, project management, and administrative staff. Technical training and basic mathematics are assumed. The notes are designed for a two-day course, taught by two lecturers. Video tapes may be used at the options of the instructors. The notes provide all the materials that are necessary but outside reading will assist in the fullest understanding. The course begins with a nuclear physics overview. The reader is led from the macroscopic world into the microscopic world of atoms and the elementary particles that constitute atoms. The particles, their masses and sizes and properties associated with radioactive decay and fission are introduced along with Einstein's mass-energy equivalence. Radioactive decay, nuclear reactions, radiation penetration, shielding and health-effects are discussed to understand protection in case of a criticality accident. Fission, the fission products, particles and energy released are presented to appreciate the dangers of criticality. Nuclear cross sections are introduced to understand the effectiveness of slow neutrons to produce fission. Chain reactors are presented as an economy; effective use of the neutrons from fission leads to more fission resulting in a power reactor or a criticality excursion. The six-factor formulamore » is presented for managing the neutron budget. This leads to concepts of material and geometric buckling which are used in simple calculations to assure safety from criticality. Experimental measurements and computer code calculations of criticality are discussed. To emphasize the reality, historical criticality accidents are presented in a table with major ones discussed to provide lessons-learned. Finally, standards, NRC guides and regulations, and DOE orders relating to criticality protection are presented.« less

The lecture notes for introductory nuclear engineering are provided for Department of Energy personnel that are recent graduates, transfers from non-nuclear industries, and people with minimum engineering training. The material assumes a knowledge of algebra and elementary calculus. These notes support and supplement a three-hour lecture. The reader is led into the subject from the familiar macroscopic world to the microscopic world of atoms and the parts of atoms called elementary particles. Only a passing reference is made to the very extensive world of quarks and tansitory particles to concentrate on those associated with radioactivity and fission. The Einsteinian truth of mass-energy equivalence provides an understanding of the forces binding a nucleus with a resulting mass defect that results in fusion at one end of the mass spectrum and fission at the other. Exercises are provided in calculating the energy released in isotopic transformation, reading and understanding the chart of the nuclides. The periodic table is reviewed to appreciate that the noble elements are produced by quantum mechanical shell closings. Radioactive decay is calculated as well as nuclear penetration and shielding. The geometric attenuation of radiation is studied for personal protection; the use of shielding materials for radiation protection ismore » presented along with the buildup factor that renders the shielding less effective than might be supposed. The process of fission is presented along with the fission products and energies produced by fission. The requirements for producing a sustained chain reactor are discussed. The lecture ends with discussions of how radiation and dose is measured and how dose is converted to measures of the damage of radiation to our bodies.« less

NRC Generic Issue 43, Contamination of Instrument Air Lines'', has been unresolved since 1980. The potential seriousness of this issue was reinforced in a 1987 study by the Office for Analysis and Evaluation of Operational Data. Aging of components within compressed air systems, leading to degraded function of the system, is the subject of this study. This work was performed under the auspices of the NRC's Office of Nuclear Regulatory Research as part of the Nuclear Plant Aging Research (NPAR) Program. The objective of this study was to identify all the aging modes and their causes, which should be mitigated to achieve a reliable operation of all safety-related air equipment. Also included is an interim review of typical maintenance activities for air systems in the nuclear power industry. The Phase 2 effort of this study will make recommendations for developing an effective maintenance program industry-wide to counter the effects of aging. The analysis of operating experience data revealed that aging degradation occurs in the compressed air system, and becomes a factor as the system ages. Normal wear of the system and contamination of the air dominate the problems of system failure. Existing maintenance programs within the industry lack uniformity, andmore » quality assurance is not rigorous because the system is classified as non-safety. 16 refs., 33 figs., 11 tabs.« less

At energies of 10 Bev and higher, it is not easy to obtain physically separated beams of ? and ? mesons. We have constructed a hodoscope system which will register the identity of each of 20 particles entering a bubble chamber over a 100 microsecond time interval. Particles are localized by a scintillator hodoscope with matrix elements 1 cm square. Particle identity is determined with Cerenkov counters. The combined information is displayed on an oscilloscope and photographed at each bubble chamber expansion.

A medium fast time-of-flight front end for use with an on-line computer is described. Principal features of this unit are: (1) 31.25 nsec minimum channel width. (2) 1 ?sec minimum deadtime. (3) Controllable deadtime (for synchronization purposes). (4) 3 stage output derandomization buffer. (5) Simple low cost construction almost entirely of commercially available computer logic. (6) Crystal clock for stability. (7) Output pulse for clock synchronized accelerator triggering. (8) 220 channels available (1,048,576). (9) 3 tag bits for data identification. (10) Deadtime, clock livetime and lost data outputs.

The R.P.I. Linac, one of the most powerful electron accelerators in the world, has been in operation for several months. It has been operated at 130 megawatts peak power and at average beam power of 50 kilowatts for sustained runs. Evidence of pulse shortening has been observed only at currents of 0.8 ampere for pulse widths of 4.6 microseconds. A detailed discussion of performance is presented as well as current improvements.

A method of calibration of the Moxon-Rae type gamma-ray detectors used in neutron capture crosssection measurements with underground nuclear explosions is described. The method consists of measuring the activation and detector signal in individual resonances in samples of 197Au and 238U exposed to a neutron beam from an underground nuclear explosion. The resonances are resolved by time of flight. Since both the activation and the corresponding detector signal are known, the efficiency can be determined without measuring the incident neutron flux or sample capture cross section.

Operating experience in U. S. light water reactors (LWRs) has shown that the impact of refueling outages on plant unavailability is much higher than has been previously anticipated. The purpose of this summary is to identify the principal causes of the extensions of refueling outages, the effect of these outages on plant productivity, and an alternate refueling cycle to reduce their impact. In order to provide a perspective on the effects of refueling outages, major outages greater than 100 hours in duration are also included in this assessment. Both the refueling outages and other major outages are displayed as a function of plant age; this method allows identification of trends in these outages as a plant matures. The result of the evaluation indicates that utilities can improve plant availability by up to six percent per year by increasing the time between refuelings (i.e., from an annual to an eighteen-month refueling cycle).

A study of the aging effects on Residual Heat Removal (RHR) systems in Boiling Water Reactors (BWRs) was performed as part of the Nuclear Plant Aging Research (NPAR) program. The objectives of the NPAR program are to provide a technical basis for the identification and evaluation of degradation caused by age in nuclear power plant applications. The information from this and other NPAR studies will be used to assess the impact of aging on plant safety and to develop effective mitigation actions. The effects of aging in the RHR system were characterized using the Aging and Life Extension Assessment Program (ALEAP) Systems Level Plan developed by Brookhaven National Laboratory. Failure data from various national data bases were reviewed and analyzed to identify predominant failure modes, causes and mechanisms. Time-dependent failure frequencies for major components were calculated to identify aging trends. Plant specific information was also reviewed to supplement data base results. A computer program (PRAAGE-1988) was developed and implemented to model a typical RHR design and perform time-dependent Probabilistic Risk Assessment (PRA) calculations. Time-dependent failure probabilities were input to the PRAAGE program to evaluate the effects of aging on component importance and system unavailability. 21 refs., 63 figs., 26 tabs.

Publisher Summary This chapter discusses the linac facility at Rensselaer Polytechnic Institute, which has been established to carry on applied and fundamental research related to the nuclear field; it is a part of the Nuclear Center, administered through the Department of Nuclear Engineering and Science. The high power linear accelerator, which is the core of the facility, was procured under an Atomic Energy Commission contract, which also provides substantial equipment and operating support. To make use of the accelerator, it is necessary to direct the electrons from the output of the accelerator to experimental stations. High energies are somewhat incidental to high intensity, and to this end, a new type or another generation of accelerators has come about. The electron linear accelerator was chosen over other machines because it satisfies the requirements and incorporates a degree of versatility not obtained in other accelerator.