36 results on '"Cecil V. Parks"'
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2. Resource Letter FuNP-1: The Future of Nuclear Power
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George E Kulynych, George F Flanagan, and Cecil V. Parks
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Physics ,Power (social and political) ,Resource (project management) ,business.industry ,Web page ,General Physics and Astronomy ,ComputerApplications_COMPUTERSINOTHERSYSTEMS ,Subject (documents) ,Nuclear power ,Telecommunications ,business - Abstract
This Resource Letter is intended to summarize the status of nuclear power in the world today, prospects of significant expansion of nuclear power over the next several decades, the planning of and forecasts for the addition of new power reactors, and issues surrounding the addition of these new reactors. Owing to the breadth of this subject, the list of references includes journal articles, web pages, and reports to guide the reader on the subject. The subject of nuclear power and its related issues are dynamic, so the most current information is likely to be found on reputable websites. more...
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- 2010
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Catalog
3. A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent-Fuel Pool Storage
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Cecil V. Parks and John C. Wagner
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Nuclear and High Energy Physics ,Nuclear fuel ,020209 energy ,Nuclear engineering ,Pressurized water reactor ,Allowance (engineering) ,02 engineering and technology ,Nuclear reactor ,Condensed Matter Physics ,Spent nuclear fuel ,law.invention ,020303 mechanical engineering & transports ,0203 mechanical engineering ,Nuclear Energy and Engineering ,Criticality ,law ,0202 electrical engineering, electronic engineering, information engineering ,Environmental science ,Spent fuel pool ,Burnup - Abstract
This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for pressurized water reactor (PWR) spent-fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub eff} estimates based on reactivity ''equivalent'' fresh fuel enrichment (REFFE) to k{sub eff} estimates using the calculated spent-fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE formore » different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity.A significant concentration ({approx}2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion ([less than or approximately equal]500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE. Consequently, the findings presented here do not represent a significant safety concern unless/until the subcritical margin associated with the soluble boron (that is not currently explicitly credited) is offset by the uncertainties associated with burnup credit and/or the expanded allowance of credit for the soluble boron.« less more...
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- 2001
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4. ORIGEN-ARP, A Fast and Easy-to-Use Source Term Generation Tool
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Cecil V. Parks, Luiz C Leal, O.W. Hermann, and Stephen M. Bowman
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Nuclear and High Energy Physics ,business.industry ,Computer science ,Pressurized water reactor ,Rapid processing ,Neutron spectra ,Spent nuclear fuel ,Term (time) ,law.invention ,Nuclear Energy and Engineering ,law ,Boiling water reactor ,Light-water reactor ,Process engineering ,business ,Nuclear chemistry ,Burnup - Abstract
ORIGEN-ARP is a new SCALE analytical sequence for spent fuel characterization and source term generation that serves as a faster alternative to the SAS2H sequence by using the Automatic Rapid Processing (ARP) methodology for generating problem-dependent ORIGEN-S cross-section libraries. ORIGEN-ARP provides an easy-to-use menu-driven input processor. This new sequence is two orders of magnitude faster than SAS2H while conserving the rigor and accuracy of the SAS2H methodology. ORIGEN-ARP has been validated against pressurized water reactor (PWR) and boiling water reactor (BWR) spent fuel chemical assay data. more...
- Published
- 2000
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5. Automatic Rapid Process for the Generation of Problem-Dependent SAS2H/ORIGEN-S Cross-Section Libraries
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Stephen M. Bowman, Luiz C Leal, O.W. Hermann, and Cecil V. Parks
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Nuclear and High Energy Physics ,Scale (ratio) ,Computer science ,020209 energy ,Process (computing) ,02 engineering and technology ,Condensed Matter Physics ,Computational science ,Nuclear physics ,Cross section (physics) ,020303 mechanical engineering & transports ,0203 mechanical engineering ,Nuclear Energy and Engineering ,Path (graph theory) ,0202 electrical engineering, electronic engineering, information engineering ,Code (cryptography) - Abstract
A methodology is described that serves as an alternative to the SAS2H path of the SCALE system to generate cross sections for point-depletion calculations with the ORIGEN-S code. Automatic Rapid Pr... more...
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- 1999
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6. Evaluation of Shielding Analysis Methods in Spent-Fuel Cask Environments
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Jabo S. Tang, Hiroaki Taniuchi, Cecil V. Parks, B.L. Broadhead, and Robert L. Childs
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Nuclear and High Energy Physics ,Measure (data warehouse) ,Source code ,Scale (ratio) ,020209 energy ,Nuclear engineering ,media_common.quotation_subject ,02 engineering and technology ,Condensed Matter Physics ,Spent nuclear fuel ,Nuclear physics ,020303 mechanical engineering & transports ,0203 mechanical engineering ,Nuclear Energy and Engineering ,Electromagnetic shielding ,0202 electrical engineering, electronic engineering, information engineering ,Benchmark (computing) ,Environmental science ,CASK ,Analysis method ,media_common - Abstract
The three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, is applied to the analysis of a series of simple geometry benchmark experiments and prototypic spent-fuel storage cask measurements. The simple geometry experiments were performed in Japan and at the General Electric-Morris Operation facility; the cask measurements were performed at the Idaho National Engineering Laboratory. The quantification of uncertainties in a typical shielding analysis process for transport/storage casks can be accomplished by comparison of consistent trends between calculated and measured dose rate quantities in both benchmark and prototypic environments. Benchmark results typically measure the validity of cross-section data and computer code adequacy; prototypic environments, however, generally measure the overall validity of the calculational procedure. A total of five storage cask problems and two simple geometry problems were analyzed to determine the expected accur... more...
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- 1997
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7. Validation of SCALE-4 for Burnup Credit Applications
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Mark D. DeHart, Cecil V. Parks, and Stephen M. Bowman
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Nuclear and High Energy Physics ,Nuclear fuel ,020209 energy ,Nuclear engineering ,Pressurized water reactor ,02 engineering and technology ,Nuclear reactor ,Condensed Matter Physics ,Spent nuclear fuel ,law.invention ,020303 mechanical engineering & transports ,0203 mechanical engineering ,Nuclear Energy and Engineering ,Nuclear reactor core ,Criticality ,law ,0202 electrical engineering, electronic engineering, information engineering ,Environmental science ,Light-water reactor ,Burnup - Abstract
In the past, criticality analysis of pressurized water reactor (PWR) fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at Oak Ridge National Laboratory (ORNL) in support of the U.S. Department of Energy (DOE) efforts to demonstrate a validation approach of criticality safety methods to be used in burnup credit cask design. The date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The SCALE code package is a well-established code system that has been widely used in away from reactor applications. Criticality safety analyses are performed via the criticality safety analysis sequences (CSAS) and spent-fuel characterization via the shielding analysis sequence (QSAS) and spent-fuel characterization via the shielding analysis sequence (SAS2H). The SCALE 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data has been used for all calculations. The American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors of correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. Part of the work that has been performed to date to validate the SCALE-4 code system for burnup credit applications using measured critical configurations includes: 1. fresh fuel critical experiments having geometric and nuclear characteristics similar to PWR spent fuel in storage and transport configurations 2. commercial PWR hot-zero-power and hot-full-power reactor critical configurations. The ability to closely predict reactor critical conditions is important in the validation of a methodology for spent-fuel applications because input data are determined based on relatively little detail of reactor core operation. Such limited information is expected to be representative of data available when burnup credit calculations are being performed in the determination of optimum cask loadings. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications more...
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- 1995
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8. [Untitled]
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R.M. Westfall, Cecil V. Parks, R.L. Childs, and B.L. Broadhead
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Nuclear facilities ,Alarm response ,Engineering ,ALARM ,Fission products ,Criticality ,business.industry ,Nuclear engineering ,Dose rate ,business ,Calculation methods ,Simulation - Abstract
This work quantifies the expected dose rates at a series of criticality alarm locations due to several postulated criticality accidents at the Westinghouse Environmental MANAGEMENT COMPANY OF OHIO (WEMCO) Fernald site. One- and two-dimensional discrete- ordinates calculations were performed for seven different shielding configurations using leakage spectra corresponding to two specific postulated critical events. In addition, an estimate of the gaseous fission products released during the hypothetical accident was made using ORIGEN-S point-depletion code. more...
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- 2008
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9. Plutonium Production Using Natural Uranium From the Front-End of the Nuclear Fuel Cycle
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Cecil V. Parks
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Nuclear fuel cycle ,Nuclear reprocessing ,Materials science ,chemistry ,Plutonium-240 ,Nuclear engineering ,chemistry.chemical_element ,Natural uranium ,Reprocessed uranium ,MOX fuel ,Thorium fuel cycle ,Plutonium - Abstract
This report explores the potential for producing weapons-grade plutonium using conventional industrial resources, oxides of natural uranium (namely UO{sub 3}), and either heavy water or reactor-grade graphite. Using established codes and data for nuclear analysis, it is demonstrated that physics-based reactor models capable of producing kilogram quantities of weapons-grade plutonium can be readily conceived. The basic assumptions and analysis approach are discussed together with the results of the analysis. more...
- Published
- 2002
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10. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel
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Cecil V. Parks and I. C. Gauld
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Engineering ,Operations research ,business.industry ,Range (aeronautics) ,Experimental data ,Nuclear data ,Sensitivity (control systems) ,Decay heat ,Oak Ridge National Laboratory ,business ,Process engineering ,Spent nuclear fuel ,Burnup - Abstract
This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address more » the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental data in code validation is also discussed and demonstrated. « less more...
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- 2000
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11. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage
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Cecil V. Parks and John C. Wagner
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Criticality ,Chemistry ,Nuclear engineering ,Equating ,chemistry.chemical_element ,Reactivity (chemistry) ,Allowance (engineering) ,Boron ,Spent nuclear fuel ,Spent fuel pool ,Burnup ,Nuclear chemistry - Abstract
This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub inf} estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k{sub inf} estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration ({approximately}2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion ({le} 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE. Consequently, the findings presented here do not represent a significant safety concern unless/until the subcritical margin associated with the soluble boron (that is not currently explicitly credited) is offset by the uncertainties associated with burnup credit and/or the expanded allowance of credit for the soluble boron. more...
- Published
- 2000
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12. Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel
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Cecil V. Parks, M D DeHart, and John C. Wagner
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Background information ,Prioritization ,Engineering ,Identification (information) ,Work (electrical) ,Ranking ,Risk analysis (engineering) ,Operations research ,Process (engineering) ,business.industry ,business ,Spent nuclear fuel ,Burnup - Abstract
This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan. more...
- Published
- 2000
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13. EMSP project summary (Project ID: 60077): Development of nuclear analysis capabilities for DOE waste management activities
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Cecil V. Parks, C.M. Hopper, DeHart, B.L. Broadhead, Bradley T Rearden, and L.M. Petrie
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Engineering ,Waste management ,Fissile material ,Criticality ,business.industry ,Margin (machine learning) ,Process (engineering) ,Radioactive waste ,business ,Geometric modeling ,Reliability (statistics) ,Spent nuclear fuel - Abstract
The objective of this project is to develop and demonstrate prototypical analysis capabilities that can be used by nuclear safety analysis practitioners to: (1) demonstrate a more thorough understanding of the underlying physics phenomena that can lead to improved reliability and defensibility of safety evaluations; and (2) optimize operations related to the handling, storage, transportation, and disposal of fissile material and DOE spent fuel. To address these problems, this project has been investigating the implementation of sensitivity and uncertainty methods within existing Monte Carlo codes used for criticality safety analyses. It is also investigating the use of a new deterministic code that allows for specification of arbitrary grids to accurately model geometric details required in a criticality safety analysis. This capability can facilitate improved estimations of the required subcritical margin and potentially enable the use of a broader range of experiments in the validation process. The new arbitrary-grid radiation transport code will also enable detailed geometric modeling valuable for improved accuracy in application to a myriad of other problems related to waste characterization. Application to these problems will also be explored. more...
- Published
- 2000
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14. Assessment and recommendations for fissile-material packaging exemptions and general licenses within 10 CFR Part 71
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Cecil V. Parks, J. L. Lichtenwalter, and C.M. Hopper
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Engineering ,Actuarial science ,Fissile material ,business.industry ,Commission ,law.invention ,Consistency (negotiation) ,Consolidation (business) ,Risk analysis (engineering) ,law ,Code of Federal Regulations ,CLARITY ,Relevance (law) ,business - Abstract
This report provides a technical and regulatory assessment of the fissile material general licenses and fissile material exemptions within Title 10 of the Code of Federal Regulations Part 71. The assessment included literature studies and calculational analyses to evaluate the technical criteria; review of current industry practice and concerns; and a detailed evaluation of the regulatory text for clarity, consistency and relevance. Recommendations for potential consideration by the Nuclear Regulatory Commission staff are provided. The recommendations call for a simplification and consolidation of the general licenses and a change in the technical criteria for the first fissile material exemptions. more...
- Published
- 1998
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15. Development of nuclear analysis capabilities for DOE waste management activities. 1998 annual progress report
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Cecil V. Parks, DeHart, B.L. Broadhead, and C.M. Hopper
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Engineering ,Work (electrical) ,Criticality ,Fissile material ,business.industry ,Process (engineering) ,Systems engineering ,Radioactive waste ,business ,Geometric modeling ,Spent nuclear fuel ,Reliability (statistics) - Abstract
'The objective of this project is to develop and demonstrate prototypic analysis capabilities that can be used by the nuclear safety analysis practitioners to: (1) demonstrate a more thorough understanding of the underlying physics phenomena that can lead to improved reliability and defensibility of safety evaluations; and (2) optimize operations related to the handling, storage, transportation, and disposal of fissile material and DOE spent fuel. To address these problems, the project will investigate the implementation of sensitivity and uncertainty methods within existing Monte Carlo codes used for criticality safety analyses, as well as within a new deterministic code that allows specification of arbitrary grids to accurately model the geometry details required in a criticality safety analysis. This capability can facilitate improved estimations of the required subcritical margin and potentially enable the use of a broader range of experiments in the validation process. The new arbitrary-grid radiation transport code will also enable detailed geometric modeling valuable for improved accuracy in application to a myriad of other problems related to waste characterization. Application to these problems will also be explored. This report summarizes the progress achieved after only seven months of work on a three-year project.' more...
- Published
- 1998
- Full Text
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16. Development of nuclear analysis capabilities for DOE waste management activities
- Author
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DeHart, L.M. Petrie, Cecil V. Parks, C.M. Hopper, and B.L. Broadhead
- Subjects
Engineering ,Fissile material ,Waste management ,Criticality ,Process (engineering) ,business.industry ,Monte Carlo method ,Radioactive waste ,Geometric modeling ,business ,Spent nuclear fuel ,Reliability (statistics) - Abstract
The objective of this project is to develop and demonstrate prototypic analysis capabilities that can be used by the nuclear safety analysis practitioners to: (1) demonstrate a more thorough understanding of the underlying physics phenomena that can lead to improved reliability and defensibility of safety evaluations; and (2) optimize operations related to the handling, storage, transportation, and disposal of fissile material and DOE spent fuel. To address these problems, the project will investigate the implementation of sensitivity and uncertainty methods within existing Monte Carlo codes used for criticality safety analyses, as well as within a new deterministic code that allows specification of arbitrary grids to accurately model the geometry details required in a criticality safety analysis. This capability can facilitate improved estimations of the required subcritical margin and potentially enable the use of a broader range of experiments in the validation process. The new arbitrary-grid radiation transport code will also enable detailed geometric modeling valuable for improved accuracy in application to a myriad of other problems related to waste characterization. Application to these problems will also be explored. more...
- Published
- 1998
- Full Text
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17. ARP: Automatic rapid processing for the generation of problem dependent SAS2H/ORIGEN-s cross section libraries
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O.W. Hermann, Cecil V. Parks, Luiz C Leal, and Stephen M. Bowman
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Cross section (physics) ,Scale (ratio) ,Computer science ,Path (graph theory) ,Rapid processing ,Code (cryptography) ,Spent nuclear fuel ,Interpolation ,Burnup ,Computational science - Abstract
In this report, a methodology is described which serves as an alternative to the SAS2H path of the SCALE system to generate cross sections for point-depletion calculations with the ORIGEN-S code. ARP, Automatic Rapid Processing, is an algorithm that allows the generation of cross-section libraries suitable to the ORIGEN-S code by interpolation over pregenerated SAS2H libraries. The interpolations are carried out on the following variables: burnup, enrichment, and water density. The adequacy of the methodology is evaluated by comparing measured and computed spent fuel isotopic compositions for PWR and BWR systems. more...
- Published
- 1998
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18. Dose consequences from a postulated criticality occurring in a low-level waste disposal facility
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B.L. Broadhead, Cecil V. Parks, R.L. Childs, and C.M. Hopper
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Radiation exposure ,Criticality ,chemistry ,Waste management ,Fissile material ,Low-level waste ,Radiochemistry ,Nuclear criticality safety ,Groundwater transport ,chemistry.chemical_element ,Radioactive waste ,Uranium ,complex mixtures - Abstract
Evaluations were done to determine conditions that could permit nuclear criticality with fissile uranium in low-level waste (LLW) facilities and to estimate potential radiation exposures to personnel if there were such an accident. Simultaneous hydrogeochemical and nuclear criticality studies were done (1) to identity realistic scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) to model groundwater transport of uranium and subsequent concentration via sorption or precipitation, (3) to evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits, and (4) to estimate potential radiation exposures to personnel resulting from criticality consequences. This paper presents the details of the radiation exposure calculations relying on the conditions as determined from the preceding studies detailed in a cited reference. more...
- Published
- 1997
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19. Recommendations for preparing the criticality safety evaluation of transportation packages
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Cecil V. Parks and H.R. Dyer
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Engineering ,Fissile material ,Land transport ,Criticality ,business.industry ,Safety standards ,business ,Civil engineering ,Construction engineering - Abstract
This report provides recommendations on preparing the criticality safety section of an application for approval of a transportation package containing fissile material. The analytical approach to the evaluation is emphasized rather than the performance standards that the package must meet. Where performance standards are addressed, this report incorporates the requirements of 10 CFR Part 71. 12 refs., 6 figs., 8 tabs. more...
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- 1997
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20. Statistical analysis of radiochemical measurements of TRU radionuclides in REDC waste
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J. Beauchamp, F. Schultz, D. Downing, L. Yong, Cecil V. Parks, J. Chapman, L. Nguyen, and V. Fedorov
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Radionuclide ,Chemistry ,Radiochemistry ,Radioactive waste ,Statistical analysis ,Radiochemical analysis ,Transuranium element - Abstract
This report summarizes results of the study on the isotopic ratios of transuranium elements in waste from the Radiochemical Engineering Development Center actinide-processing streams. The knowledge of the isotopic ratios when combined with results of nondestructive assays, in particular with results of Active-Passive Neutron Examination Assay and Gamma Active Segmented Passive Assay, may lead to significant increase in precision of the determination of TRU elements contained in ORNL generated waste streams. more...
- Published
- 1996
- Full Text
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21. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results
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M.C. Brady, Cecil V. Parks, and DeHart
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Set (abstract data type) ,Credit analysis ,Fission products ,Engineering ,Failure mode, effects, and criticality analysis ,Criticality ,business.industry ,Nuclear engineering ,Benchmark (computing) ,Operations management ,business ,Spent nuclear fuel ,Burnup - Abstract
In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155. more...
- Published
- 1996
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22. Shielding and criticality analyses of phase I reference truck and rail cask designs for spent nuclear fuel
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B.L. Broadhead, Cecil V. Parks, and R.L. Childs
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Truck ,Engineering ,Criticality ,business.industry ,Nuclear engineering ,Electromagnetic shielding ,Depleted uranium ,Shields ,CASK ,business ,Spent nuclear fuel ,Burnup - Abstract
Results are presented herein to determine the adequacy with respect to shielding regulations of reference designs for a truck cask containing 2 PWR or 5 BWR assemblies of standard burnup (45 GWd/MTU for PWR, 40 GWd/MTU for BWR) and 1 PWR assembly with extended burnup (55 GWd/MTU). The study also includes reference and modified rail cask designs with projected payloads of 8, 10, or 12 PWR assemblies. The burnup/age trends are analyzed in one dimension for both Pb and depleted uranium (DU) gamma-ray shields. The results of the two-dimensional shielding analysis uphold the one-dimensional results as being an appropriate means of studying the burnup/age trends for the truck cask. These results show that the reference design for the Pb-shield truck cask is inadequate for all cases considered, while the DU-shield truck cask is capable of carrying the desired payloads. The one-dimensional shielding analysis results for the reference Pb and DU rail casks indicate substantial margins exist in the side doses for reasonable burnup/age combinations. For a Pb-cask configuration, margins exist primarily for long-cooled (15 years) fuel. For the modified Pb and DU rail casks, the 2-m dose rates offer substantial margins below the regulatory limits for all burnup values considered provided the spent fuel has cooled for {>=}10 years. The modified Pb and DU casks yield essentially identical results and, hence, could be considered equivalent from a shielding perspective. The criticality analyses that were performed indicate that a truck basket can be designed to provide an adequate subcritical margin for 2 PWR assemblies enriched to 5 wt%. While the 10- and 12- assembly rail cask designs are very close to the regulatory limit of 0.95 for k{sub eff}, after accounting for a 0.01 {Delta}k bias and 2 standard deviations, the limit is exceeded by about 3%. It is believed that a combination of decreased enrichments and/or increased water gaps should allow these baskets to be acceptable. more...
- Published
- 1996
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23. Adequacy of the 123-group cross-section library for criticality analyses of water-moderated uranium systems
- Author
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W. C. Jordan, R. Q. Wright, and Cecil V. Parks
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Nuclear physics ,Physics ,Cross section (physics) ,Criticality ,chemistry ,Group (periodic table) ,Analytical chemistry ,chemistry.chemical_element ,Scale (descriptive set theory) ,Uranium ,Enriched uranium ,Resonance (particle physics) ,Original data - Abstract
In a recent criticality analysis for an array of water-moderated packages containing highly enriched uranium, the 123-group cross-section library in the SCALE system was observed to have a nonconservative discrepancy of approximately 3 to 3.5% when compared with more recently developed libraries. A simple representative system of UO{sub 2}F{sub 2}-H{sub 2}O was used to identify that the problem results from a lack of resonance data for {sup 235}U. Only a single set of self-shielded cross sections, most likely corresponding to a water-moderated infinite dilute system, was provided with the original data. The UO{sub 2}F{sub 2}-H{sub 2}O study indicates that this limitation may cause nonconservative discrepancies as high as 5.5% for some water-moderated, highly enriched uranium systems. Characteristics of the systems where the discrepancy is evident are identified and discussed. more...
- Published
- 1995
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24. Investigation of nuclide importance to functional requirements related to transport and long-term storage of LWR spent fuel
- Author
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DeHart, J.S. Tang, B.L. Broadhead, Cecil V. Parks, and J.C. Ryman
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Radionuclide ,Materials science ,Waste management ,Criticality ,Isotope ,Radioactive waste ,Nuclide ,Decay heat ,Spent nuclear fuel ,Burnup - Abstract
The radionuclide characteristics of light-water-reactor (LWR) spent fuel play key roles in the design and licensing activities for radioactive waste transportation systems, interim storage facilities, and the final repository site. Several areas of analysis require detailed information concerning the time-dependent behavior of radioactive nuclides including (1) neutron/gamma-ray sources for shielding studies, (2) fissile/absorber concentrations for criticality safety determinations, (3) residual decay heat predictions for thermal considerations, and (4) curie and/or radiological toxicity levels for materials assumed to be released into the ground/environment after long periods of time. The crucial nature of the radionuclide predictions over both short and long periods of time has resulted in an increased emphasis on thorough validation for radionuclide generation/depletion codes. Current radionuclide generation/depletion codes have the capability to follow the evolution of some 1600 isotopes during both irradiation and decay time periods. Of these, typically only 10 to 20 nuclides dominate contributions to each analysis area. Thus a quantitative ranking of nuclides over various time periods is desired for each of the analysis areas of shielding, criticality, heat transfer, and environmental dose (radiological toxicity). These rankings should allow for validation and data improvement efforts to be focused only on the most important nuclides. This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to 100,000 years. Ranking plots for each of these analysis areas are given in an Appendix for completeness, as well as summary tables in the main body of the report. Summary rankings are presented in terms of high (greater than 10% contribution to the total), medium (between 1% and 10% contribution), and low (less than 1% contribution) for both short- and long-term cooling. When compared with the expected measurement accuracies, these rankings show that most of the important isotopes can be characterized sufficiently for the purpose of radionuclide generation/depletion code validation in each of the analysis areas. Because the main focus of this work is on the relative importances of isotopes associated with LWR spent fuel, some conclusions may not be applicable to similar areas such as high-level waste (HLW) and nonfuel-bearing components (NFBC). more...
- Published
- 1995
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25. Validation of the scale system for PWR spent fuel isotopic composition analyses
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Stephen M. Bowman, Cecil V. Parks, M.C. Brady, and O.W. Hermann
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Fission products ,Neutron transport ,Design data ,Scale (ratio) ,Chemistry ,Fission ,Nuclear engineering ,Actinide ,Isotopic composition ,Spent nuclear fuel - Abstract
The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant. more...
- Published
- 1995
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26. Technical support for a proposed decay heat guide using SAS2H/ORIGEN-S data
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Cecil V. Parks, O. W. Hermann, and J. P. Renier
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Technical support ,Engineering ,Operations research ,business.industry ,Nuclear engineering ,Decay heat ,business - Published
- 1994
- Full Text
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27. Application of Differential Sensitivity Theory to a Neutronic/Thermal-Hydraulic Reactor Safety Code
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P.J. Maudlin and Cecil V. Parks
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Nuclear and High Energy Physics ,Computer science ,020209 energy ,Nuclear engineering ,02 engineering and technology ,Condensed Matter Physics ,Thermal hydraulics ,020303 mechanical engineering & transports ,0203 mechanical engineering ,Nuclear Energy and Engineering ,0202 electrical engineering, electronic engineering, information engineering ,Code (cryptography) ,Sensitivity (control systems) ,Reactor safety ,Differential (mathematics) - Abstract
A recently proposed sensitivity technique called differential sensitivity theory is applied to the neu-tronic/thermal-hydraulic fast reactor safety code MELT-IIIB. This application centers on the d... more...
- Published
- 1981
- Full Text
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28. Criticality Analysis Support for the Three Mile Island Unit 2 Fuel Removal Operations
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Cecil V. Parks, B. L. Broadhead, and Robert M. Westfall
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Nuclear and High Energy Physics ,020209 energy ,02 engineering and technology ,Oak Ridge National Laboratory ,Condensed Matter Physics ,Civil engineering ,Design team ,Unit (housing) ,020303 mechanical engineering & transports ,Failure mode, effects, and criticality analysis ,0203 mechanical engineering ,Nuclear Energy and Engineering ,Criticality ,0202 electrical engineering, electronic engineering, information engineering ,Environmental science ,Mile - Abstract
Beginning in 1984, the Three Mile Island Unit 2 Defueling Design Team requested Oak Ridge National Laboratory to supply criticality safety analyses in support of the licensing activities for all fu... more...
- Published
- 1989
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29. Authors
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Rayford L. Patterson, Michael L. Estabrook, D. C. Wilson, Gregory L. Calhoun, Lawrence H. Porter, William E. Austin, Paul M. Shearer, Sander Levin, Robert J. Wolfgang, R. H. Fillnow, P. R. Bengel, David L. Giefer, Leona E. Champeny, William L. Whittaker, Andre B. Jeffries, Michael S. McGough, George J. Knetl, Cecil V. Parks, Robert M. Westfall, B.L. Broadhead, Gerald L. Palau, James J. Byrne, Robert E. Rogan, Richard D. Schauss, David K. Cowser, and Michael J. Kelley more...
- Subjects
Nuclear and High Energy Physics ,Nuclear Energy and Engineering ,Condensed Matter Physics - Published
- 1989
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30. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications. [Radiation dose rates from shielded spent fuels and high-level radioactive waste]
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O.W. Hermann, R.W. Roussin, J.S. Tang, B.L. Broadhead, B.L. Kirk, Cecil V. Parks, S.N. Cramer, and J.C. Gauthey
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Engineering ,Waste management ,business.industry ,Radiation dose ,Radioactive waste ,Spent nuclear fuel ,law.invention ,High-level waste ,law ,Electromagnetic shielding ,Shielded cable ,Radiation monitoring ,business ,Energy source - Published
- 1988
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31. Thermal-hydraulic differential sensitivity theory. [LMFBR]
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Cecil V. Parks, C.F. Weber, and P.J. Maudlin
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Thermal hydraulics ,Nonlinear system ,Engineering ,Mathematical model ,business.industry ,Nuclear engineering ,Heat transfer ,Applied mathematics ,Fluid mechanics ,Sensitivity (control systems) ,Space (mathematics) ,business ,Differential (mathematics) - Abstract
Differential sensitivity theory is applied to a nonlinear, three-dimensional space- and time-dependent description of the thermal-hydraulic conservation equations. The resulting sensitivity equations, which are derived using adjoint functions, can be readily utilized for input parameter sensitivity analysis of large or long-running thermal-hydraulic computer codes without any prior engineering judgement. The procedure for applying this sensitivity theory is illustrated using several classical analytical problems. more...
- Published
- 1981
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32. Parametric study of radiation dose rates from rail and truck spent fuel transport casks
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Cecil V. Parks, J.R. Knight, and O.W. Hermann
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Truck ,Engineering ,Land transport ,Waste management ,business.industry ,Nuclear engineering ,Dosimetry ,Neutron ,CASK ,Neutron radiation ,business ,Energy source ,Spent nuclear fuel - Abstract
Neutron and gamma dose rates from typical rail and truck spent fuel transport casks are reported for a variety of spent PWR fuel sources and cask conditions. The IF 300 rail cask and NLI 1/2 truck cask were selected for use as approprite cask models. All calculations (cross section preparation, generation of spent fuel source terms, radiation transport calculations, and dose evaluation) were performed using various modules of the SCALE computational system. Conditions or parameters for which there were variations between cases include detector distance from cask, spent fuel cooling time, the setting of fuel or neutron shielding cavities to either wet or dry, the cobalt content of assembly materals, normal fuel assemblies and consolidated cannisters, the geometry mesh interval size, and the order of the angular quadrature set. more...
- Published
- 1985
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33. Multicode comparison of selected source-term computer codes
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J. P. Renier, J.W. Roddy, Cecil V. Parks, R.J. LaBauve, W.B. Wilson, R.C. Ashline, and O.W. Hermann
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Fission products ,Radionuclide ,Engineering ,business.industry ,Nuclear engineering ,Radioactive waste ,Decay heat ,Energy source ,business ,Radioactive decay ,Burnup ,Waste disposal - Abstract
This report summarizes the results of a study to assess the predictive capabilities of three radionuclide inventory/depletion computer codes, ORIGEN2, ORIGEN-S, and CINDER-2. The task was accomplished through a series of comparisons of their output for several light-water reactor (LWR) models (i.e., verification). Of the five cases chosen, two modeled typical boiling-water reactors (BWR) at burnups of 27.5 and 40 GWd/MTU and two represented typical pressurized-water reactors (PWR) at burnups of 33 and 50 GWd/MTU. In the fifth case, identical input data were used for each of the codes to examine the results of decay only and to show differences in nuclear decay constants and decay heat rates. Comparisons were made for several different characteristics (mass, radioactivity, and decay heat rate) for 52 radionuclides and for nine decay periods ranging from 30 d to 10,000 years. Only fission products and actinides were considered. The results are presented in comparative-ratio tables for each of the characteristics, decay periods, and cases. A brief summary description of each of the codes has been included. Of the more than 21,000 individual comparisons made for the three codes (taken two at a time), nearly half (45%) agreed to within 1%, and an additional 17% fellmore » within the range of 1 to 5%. Approximately 8% of the comparison results disagreed by more than 30%. However, relatively good agreement was obtained for most of the radionuclides that are expected to contribute the greatest impact to waste disposal. Even though some defects have been noted, each of the codes in the comparison appears to produce respectable results. 12 figs., 12 tabs.« less more...
- Published
- 1989
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34. ADJOINT SENSITIVITY ANALYSIS OF EXTREMUM-TYPE RESPONSES IN REACTOR SAFETY
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P.J. Maudlin, Cecil V. Parks, and D. G. Cacuci
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State variable ,010308 nuclear & particles physics ,0211 other engineering and technologies ,Fast Flux Test Facility ,02 engineering and technology ,Scram ,01 natural sciences ,Maxima and minima ,Nonlinear system ,Nuclear Energy and Engineering ,Control theory ,Saddle point ,0103 physical sciences ,021108 energy ,Sensitivity (control systems) ,Transient (oscillation) ,Mathematics - Abstract
A recently developed sensitivity theory for nonlinear systems with responses defined at critical points, e.g., maxima, minima, or saddle points, of a function of the system's state variables and parameters is applied to a protected transient with scram on high-power level in the Fast Flux Test Facility. The single-phase segment of the fast reactor safety code MELT-IIIB is used to model this transient. Two responses of practical importance, namely, the maximum fuel temperature in the hot channel and the maximum normalized reactor power level, are considered. For the purposes of sensitivity analysis, a complete characterization of such responses requires consideration of both the numerical value of the response at the maximum, and the location in phase space where the maximum occurs. more...
35. Eigenvalue sensitivity theory for resonance-shielded cross sections
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M. L. Williams, Cecil V. Parks, and B.L. Broadhead
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010308 nuclear & particles physics ,Mathematical analysis ,0211 other engineering and technologies ,02 engineering and technology ,01 natural sciences ,Self shielding ,law.invention ,Nuclear physics ,Nuclear Energy and Engineering ,law ,0103 physical sciences ,Shielded cable ,021108 energy ,Eigenvalues and eigenvectors ,Mathematics - Abstract
A method is presented to compute sensitivity coefficients for the eigenvalue of a critical assembly, including implicit effects associated with changes in resonance-shielded multigroup cross sectio... more...
36. Sensitivity- and Uncertainty-Based Criticality Safety Validation Techniques
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Cecil V. Parks, J. J. Wagschal, B.L. Broadhead, Bradley T Rearden, and C.M. Hopper
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010308 nuclear & particles physics ,0211 other engineering and technologies ,Data validation ,02 engineering and technology ,Function (mathematics) ,01 natural sciences ,Nuclear Energy and Engineering ,Criticality ,Similarity (network science) ,0103 physical sciences ,Statistics ,Benchmark (computing) ,Applied mathematics ,Cutoff ,021108 energy ,Limit (mathematics) ,Sensitivity (control systems) ,Mathematics - Abstract
The theoretical basis for the application of sensitivity and uncertainty (S/U) analysis methods to the validation of benchmark data sets for use in criticality safety applications is developed. Sensitivity analyses produce energy-dependent sensitivity coefficients that give the relative change in the system multiplication factor keff value as a function of relative changes in the cross-section data by isotope, reaction, and energy. Integral indices are then developed that utilize the sensitivity information to quantify similarities between pairs of systems, typically a benchmark experiment and design system. Uncertainty analyses provide an estimate of the uncertainties in the calculated values of the system k eff due to cross- section uncertainties, as well as correlation in the k eff uncertainties between systems. These uncertainty correlations provide an additional measure of system similarity. The use of the similarity measures from both S/U analyses in the formal determination of areas of applicability for benchmark experiments is developed. Furthermore, the use of these similarity measures as a trending parameter for the estimation of the computational bias and uncertainty is explored. The S/U analysis results, along with the calculated and measured keff values and estimates of uncertainties in the measurements, were used in this work to demonstrate application of the generalized linear-least-squares methodology (GLLSM) to data validation for criticality safety studies. An illustrative example is used to demonstrate the application of these S/U analysis procedures to actual criticality safety problems. Computational biases, uncertainties, and the upper subcritical limit for the example applications are determined with the new methods and compared to those obtained through traditional criticality safety analysis validation techniques. The GLLSM procedure is also applied to determine cutoff values for the similarity indices such that applicability of a benchmark experiment to a criticality safety design system can be assured. Additionally, the GLLSM procedure is used to determine how many applicable benchmark experiments exceeding a certain degree of similarity are necessary for an accurate assessment of the computational bias. more...
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