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Comparative techniques for nuclear fuel cycle waste management systems.

Description: A safety assessment approach for the evaluation of predisposal waste management systems is described and applied to selected facilities in the light water reactor (LWR) once-through fuel cycle and a potential coprocessed UO/sub 2/-PuO/sub 2/ fuel cycle. This approach includes a scoping analysis on pretreatment waste streams and a more detailed analysis on proposed waste management processes. The primary evaluation parameters used in this study include radiation exposures to the public from radionuclide releases from normal operations and potential accidents, occupational radiation exposure from normal operations, and capital and operating costs. On an overall basis, the waste management aspects of the two fuel cycles examined are quite similar. On an individual facility basis, the fuel coprocessing plant has the largest waste management impact.
Date: September 1, 1979
Creator: Pelto, P.J. & Voss, J.W.
Partner: UNT Libraries Government Documents Department

MFAULT: a computer program for analyzing fault trees. [In FORTRAN IV for CDC CYBER 74]

Description: A description and user instructions are presented for MFAULT, a FORTRAN computer program for fault tree analysis. MFAULT identifies the cut sets of a fault tree, calculates their probabilities, and screens the cut sets on the basis of specified cut-offs on probability and/or cut set length. MFAULT is based on an efficient upward-working algorithm for cut set identification. The probability calculations are based on the assumption of small probabilities and constant hazard rates (i.e., exponential failure distributions). Cut sets consisting of repairable components (basic events) only, non-repairable components only, or mixtures of both types can be evaluated. Components can be on-line or standby. Unavailability contributions from pre-existing failures, failures on demand, and testing and maintenance down-time can be handled. MFAULT can analyze fault trees with AND gates, OR gates, inhibit gates, on switches (houses) and off switches. The code is presently capable of finding up to ten event cut sets from a fault tree with up to 512 basic events and 400 gates. It is operational on the CONTROL DATA CYBER 74 computer. 11 figures.
Date: November 1, 1977
Creator: Pelto, P.J. & Purcell, W.L.
Partner: UNT Libraries Government Documents Department

Preliminary dose comparisons for the MRS Systems Study

Description: This report provides preliminary information on the radiological doses to the public and the workers for alternative system configurations proposed in the MRS Systems Study. Information published in the MRS Environmental Assessment (DOE 1986) was used as a basis for this analysis. The risk differences between alternative configurations were found to be small and should not be viewed as a major factor in selecting alternative configurations. 1 ref.
Date: April 1, 1989
Creator: Pelto, P.J. & Lavender, J.C.
Partner: UNT Libraries Government Documents Department

Heat transfer analysis of an underground storage tank containing solidified heat generating wastes

Description: Three steady-state models for heat transfer in an underground storage tank are presented. The first is a simplified 1-D model of a general symmetrical tank developed for preliminary calculations and parameteric studies. The second and third models use the finite difference computer program, HEATING4, and are based on 2- and 3-D geometries, respectively. The 2-D model describes a radially symmetrical tank and is used for relatively detailed calculations and testing the assumptions made in the 1-D model while the 3-D, 360/sup 0/ model describes the actual sludge configuration in Waste Tank 101-A. A comparison of the 1-, 2-, and 3-D models indicates that the accuracy increases as the complexity of the model increases. This increased accuracy is offset by the lost flexibility of applying the models and the increase in cost. The three models can be used together to perform a complete heat transfer analysis of an underground storage tank system. A time and cost effective study will result from applying the simpler models first and proceeding to the more complex models as the desired degree of accuracy and the use of the results require.
Date: August 1, 1976
Creator: Slate, S. C. & Pelto, P. J.
Partner: UNT Libraries Government Documents Department

Risk assessment method for nuclear fuel cycle operations

Description: A method is described for the identification and preliminary evaluation of potential accidents (release sequences) which could lead to the release of radioactive material from nuclear fuel cycle operations. Potential accident sequences are evaluated on the basis of risk. The basic elements of this method are presented along with its application to a conceptual high-level radioactive waste management.
Date: March 1, 1977
Creator: Pelto, P J & Winegardner, W K
Partner: UNT Libraries Government Documents Department

Survey of systems safety analysis methods and their application to nuclear waste management systems

Description: This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.
Date: November 1, 1981
Creator: Pelto, P.J.; Winegardner, W.K. & Gallucci, R.H.V.
Partner: UNT Libraries Government Documents Department

Reliability analysis of containment isolation systems

Description: This report summarizes the results of the Reliability Analysis of Containment Isolation System Project. Work was performed in five basic areas: design review, operating experience review, related research review, generic analysis and plant specific analysis. Licensee Event Reports (LERs) and Integrated Leak Rate Test (ILRT) reports provided the major sources of containment performance information used in this study. Data extracted from LERs were assembled into a computer data base. Qualitative and quantitative information developed for containment performance under normal operating conditions and design basis accidents indicate that there is room for improvement. A rough estimate of overall containment unavailability for relatively small leaks which violate plant technical specifications is 0.3. An estimate of containment unavailability due to large leakage events is in the range of 0.001 to 0.01. These estimates are dependent on several assumptions (particularly on event duration times) which are documented in the report.
Date: June 1, 1985
Creator: Pelto, P.J.; Ames, K.R. & Gallucci, R.H.
Partner: UNT Libraries Government Documents Department

Preliminary assessment of radiological doses in alternative waste management systems without an MRS facility

Description: This report presents generic analyses of radiological dose impacts of nine hypothetical changes in the operation of a waste management system without a monitored retrievable storage (MRS) facility. The waste management activities examined in this study include those for handling commercial spent fuel at nuclear power reactors and at the surface facilities of a deep geologic repository, and the transportation of spent fuel by rail and truck between the reactors and the repository. In the reference study system, the radiological doses to the public and to the occupational workers are low, about 170 person-rem/1000 metric ton of uranium (MTU) handled with 70% of the fuel transported by rail and 30% by truck. The radiological doses to the public are almost entirely from transportation, whereas the doses to the occupational workers are highest at the reactors and the repository. Operating alternatives examined included using larger transportation casks, marshaling rail cars into multicar dedicated trains, consolidating spent fuel at the reactors, and wet or dry transfer options of spent fuel from dry storage casks. The largest contribution to radiological doses per unit of spent fuel for both the public and occupational workers would result from use of truck transportation casks, which are smaller than rail casks. Thus, reducing the number of shipments by increasing cask sizes and capacities (which also would reduce the number of casks to be handled at the terminals) would reduce the radiological doses in all cases. Consolidating spent fuel at the reactors would reduce the radiological doses to the public but would increase the doses to the occupational workers at the reactors.
Date: June 1, 1986
Creator: Schneider, K.J.; Pelto, P.J.; Daling, P.M.; Lavender, J.C. & Fecht, B.A.
Partner: UNT Libraries Government Documents Department

Overview study of LNG release prevention and control systems

Description: The liquefied natural gas (LNG) industry employs a variety of release prevention and control techniques to reduce the likelihood and the consequences of accidental LNG releases. A study of the effectiveness of these release prevention and control systems is being performed. Reference descriptions for the basic types of LNG facilities were developed. Then an overview study was performed to identify areas that merit subsequent and more detailed analyses. The specific objectives were to characterize the LNG facilities of interest and their release prevention and control systems, identify possible weak links and research needs, and provide an analytical framework for subsequent detailed analyses. The LNG facilities analyzed include a reference export terminal, marine vessel, import terminal, peakshaving facility, truck tanker, and satellite facility. A reference description for these facilities, a preliminary hazards analysis (PHA), and a list of representative release scenarios are included. The reference facility descriptions outline basic process flows, plant layouts, and safety features. The PHA identifies the important release prevention operations. Representative release scenarios provide a format for discussing potential initiating events, effects of the release prevention and control systems, information needs, and potential design changes. These scenarios range from relatively frequent but low consequence releases to unlikely but large releases and are the principal basis for the next stage of analysis.
Date: March 1, 1982
Creator: Pelto, P.J.; Baker, E.G.; Holter, G.M. & Powers, T.B.
Partner: UNT Libraries Government Documents Department

LNG annotated bibliography

Description: This document updates the bibliography published in Liquefied Gaseous Fuels Safety and Environmental Control Assessment Program: third status report (PNL-4172) and is a complete listing of literature reviewed and reported under the LNG Technical Surveillance Task. The bibliography is organized alphabetically by author.
Date: September 1, 1982
Creator: Bomelburg, H.J.; Counts, C.A.; Cowan, C.E.; Davis, W.E.; DeSteese, J.G. & Pelto, P.J.
Partner: UNT Libraries Government Documents Department

Analysis of LNG peakshaving-facility release-prevention systems

Description: The purpose of this study is to provide an analysis of release prevention systems for a reference LNG peakshaving facility. An overview assessment of the reference peakshaving facility, which preceeded this effort, identified 14 release scenarios which are typical of the potential hazards involved in the operation of LNG peakshaving facilities. These scenarios formed the basis for this more detailed study. Failure modes and effects analysis and fault tree analysis were used to estimate the expected frequency of each release scenario for the reference peakshaving facility. In addition, the effectiveness of release prevention, release detection, and release control systems were evaluated.
Date: May 1982
Creator: Pelto, P. J.; Baker, E. G.; Powers, T. B.; Schreiber, A. M.; Hobbs, J.M. & Daling, P .M.
Partner: UNT Libraries Government Documents Department

Proposed framework for the Western Area Power Administration Environmental Risk Management Program

Description: The Western Area Power Administration (Western) views environmental protection and compliance as a top priority as it manages the construction, operation, and maintenance of its vast network of transmission lines, substations, and other facilities. A recent Department of Energy audit of Western`s environmental management activities recommends that Western adopt a formal environmental risk program. To accomplish this goal, Western, in conjunction with Pacific Northwest Laboratory, is in the process of developing a centrally coordinated environmental risk program. This report presents the results of this design effort, and indicates the direction in which Western`s environmental risk program is heading. Western`s environmental risk program will consist of three main components: risk communication, risk assessment, and risk management/decision making. Risk communication is defined as an exchange of information on the potential for threats to human health, public safety, or the environment. This information exchange provides a mechanism for public involvement, and also for the participation in the risk assessment and management process by diverse groups or offices within Western. The objective of risk assessment is to evaluate and rank the relative magnitude of risks associated with specific environmental issues that are facing Western. The evaluation and ranking is based on the best available scientific information and judgment and serves as input to the risk management process. Risk management takes risk information and combines it with relevant non-risk factors (e.g., legal mandates, public opinion, costs) to generate risk management options. A risk management tool, such as decision analysis, can be used to help make risk management choices.
Date: December 1, 1994
Creator: Glantz, C. S.; DiMassa, F. V.; Pelto, P. J.; Brothers, A. J. & Roybal, A. L.
Partner: UNT Libraries Government Documents Department

Risk-based fault tree analysis method for identification, preliminary evaluation, and screening of potential accidental release sequences in nuclear fuel cycle operations

Description: A method is described for identification, preliminary evaluation, and screening of potential accident sequences leading to uncontrolled release of radioactive materials. Included is a procedure for estimating the risk sum of all identified sequences. In addition, portions of the procedures have been developed for detailed analysis of the dominant (highest risk) sequences so screened. This method was developed for the ERDA-sponsored risk analysis of systems for managing high-level waste, part of the Waste Fixation Program (WFP). The method begins with certain preliminary analyses. The facility and operation are described and analysis bounds are established. A type of fault tree construction, the ''to/through'' approach, was chosen for the WFP waste management system. The to/through fault tree approach offers advantages over others in several respects. The analysis is considered more complete because the system is treated as a whole. The screening process was successfully demonstrated on a conceptual waste management system for the Waste Fixation Program. Fault trees were constructed and evaluated for processing, handling, transporting, and storing high-level waste. Trees of up to 14,000,000 release sequences (BICS-Boolean-indicated cut sets) were screened and the top few hundred or thousand sequences preliminarily ranked. An estimate of the total risk represented in the fault tree was also obtained. (auth)
Date: January 1, 1976
Creator: Smith, T. H.; Pelto, P. J.; Stevens, D. L.; Seybold, G. D.; Purcell, W. L. & Kimmel, L. V.
Partner: UNT Libraries Government Documents Department

PNL Technical Review of Pressurized Thermal Shock Issues Supplement 1: Technical Critique of the NRC Near-Term Screening Criteria

Description: Pacific Northwest Laboratory (PNL) provided a technical critique of the draft report, NRC Staff Evaluation of Pressurized Thermal Shock, dated September 13, 1982. This report provided the basis for the NRC near-term regulatory position on pressurized thermal shock {PTS) and recommended a generic screening criteria for welds in the vessel beltline region. The PNL staff concluded that the screening criteria were adequate to meet the intent of the NRC safety goal and to retain past predictions of vessel reliability. The conclusion was based on selecting the plant-specific nilductility transition reference temperature (RT{sub NDT}) in the conservative manner described within the staff report. Conservative and unconservative factors were mentioned throughout the NRC staff report. The PNL staff has listed these factors together with unknown (may be either conservative or unconservative) factors and estimated, where possible, the range in °F RT{sub NDT}. The unknown factors were so widespread that the PNL staff recommended that specific conservatisms not be reduced until the unknowns are further resolved.
Date: May 1, 1983
Creator: Pederson, L. T.; Apley, W. J.; Bian, S. H.; Pelto, P. J.; Simonen, E. P.; Simonen, F. A. et al.
Partner: UNT Libraries Government Documents Department

Waste Isolation Safety Assessment Program scenario analysis methods for use in assessing the safety of the geologic isolation of nuclear waste.

Description: The relative utility of the various safety analysis methods to scenario analysis for a repository system was evaluated by judging the degree to which certain criteria are satisfied by use of the method. Six safety analysis methods were reviewed in this report for possible use in scenario analysis of nuclear waste repositories: expert opinion, perspectives analysis, fault trees/event trees, Monte Carlo simulation, Markov chains, and classical systems analysis. Four criteria have been selected. The criteria suggest that the methods: (1) be quantitative and scientifically based; (2) model the potential disruptive events and processes, (3) model the system before and after failure (sufficiently detailed to provide for subsequent consequence analysis); and (4) be compatible with the level of available system knowledge and data. Expert opinion, fault trees/event trees, Monte Carlo simulation and classical systems analysis were judged to have the greatest potential appliation to the problem of scenario analysis. The methods were found to be constrained by limited data and by knowledge of the processes governing the system. It was determined that no single method is clearly superior to others when measured against all the criteria. Therefore, to get the best understanding of system behavior, a combination of the methods is recommended. Monte Carlo simulation was judged to be the most suitable matrix in which to incorporate a combination of methods.
Date: November 1, 1978
Creator: Greenborg, J.; Winegardner, W.K.; Pelto, P.J.; Voss, J.W.; Stottlemyre, J.A.; Forbes, I.A. et al.
Partner: UNT Libraries Government Documents Department

PNL technical review of pressurized thermal-shock issues. [PWR]

Description: Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.
Date: July 1, 1982
Creator: Pedersen, L.T.; Apley, W.J.; Bian, S.H.; Defferding, L.J.; Morgenstern, M.H.; Pelto, P.J. et al.
Partner: UNT Libraries Government Documents Department