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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.

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.

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.

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

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.

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.

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.

'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.'

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 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.

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.

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.

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.

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.

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