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This report provides computational evaluation results demonstrating that mixed oxide waste can be safely stored in Intercell Storage Wells 2 and 3 of Building 3019 at the Oak Ridge National Laboratory. Existing, verified computational techniques are validated with applicable critical experiments and tolerance limits for safety analyses are derived. Multiplication factors for normal and credible abnormal configurations are calculated and found to be far subcritical when compared to derived safety limits.

Many avionics systems use specialized parallel architectures to speed processing and to increase system reliability. The software used therein is frequently divided into tasks and executed concurrently on multiple processors under strict real-time constraints critical to the mission's successful performance. Scheduling and planning are needed for effectively managing the computational resources on such avionics architectures. Since most real-time scheduling problems are known to be NP-hard, an approximation approach that applies heuristic methods using conventional computer algorithms has been used to solve these scheduling problems. Artificial intelligence (AI) planners have been used extensively in manufacturing scheduling and operations research. In this paper, we demonstrate the idea of using AI planners to perform scheduling through an example. We derive a solution to scheduling several image tasks on a distributed computer system, using the AI planner PRODIGY. The basic characteristics of AI planners in general and the PRODIGY solver in particular are described, the domain theory and problem specification for our problem through the PRODIGY description language PDL are presented.

A recent Oak Ridge National Laboratory report provided preliminary analyses to propose alternative design parameters for a nuclear reactor that could be fueled with natural UO{sub 3} or U{sub 3}O{sub 8} and moderated with either heavy water or reactor-grade graphite. This report provides more specific reactor design and operating parameters for a heavy water-moderated reactor only. The basic assumptions and analytical approach are discussed together with the results of the analysis.

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.

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.

The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop achievable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM), and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team`s approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some achievable scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via sorption or precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits. The analysis of SNM was restricted to {sup 235}U in the present scope of work. The outcome of the work indicates that criticality is possible given established regulatory limits on SNM disposal. However, a review based on actual disposal records of an existing site operation indicates that the potential for criticality is not a concern under current burial practices.

This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide.

The Molten Salt Reactor Experiment (MSRE) was operated from June 1965 to December 1969. The objective of the experiment was to investigate the practicality of developing a power reactor consisting of a graphite lattice with circulating molten uranium salt as fuel for application in central power stations. When the experiment was terminated in 1969, approximately 4710 kg of salt containing approximately 36.3 kg of uranium, 675 g of plutonium, and various fission products were transferred to two fuel drain tanks (FDTs). The almost 30.5 kg of Uranium 233 in the salt is the primary fissile constituent, but about 0.93 kg of Uranium 235 is also present. In April 1994, a gas sample from the MSRE off-gas system (OGS) indicated that uranium had migrated from the FDTs into the OGS. Further investigation revealed a likely accumulation of approximately 2.6 kg of uranium in the auxiliary charcoal bed (ACB), which is located in the concrete-lined charcoal bed cell (CBC) below ground level outside the MSRE building. The nuclear criticality safety (NCS) situation was further complicated by the CBC being filled with water up to the overflow pipe, which completely submerged the ACB. Thus there was not only an increased risk of criticality because of water reflection in the ACB, but also because of potential moderation in the ACB in case of water inleakage. Leakage into the ACB would result in a direct path for water between the CBC and the OGS or FDTs, thus increasing the risk of criticality in these areas. When uranium was discovered in the ACB, a number of steps, detailed in this report, were immediately taken to try to understand and ameliorate the situation. After all the actions were completed, a validation of the results obtained for the ACB was performed.

This report provides a criticality safety study of the molten salt reactor fuel currently being stored in the Fuel Drain Tank (FDT) Cell of the Molten Salt Reactor Experiment (MSRE) facility (Building 7503) located in the Melton Valley area of the Oak Ridge National Laboratory. The FDTs contain approximately 36 kg of uranium consisting primarily of {sup 233}U, some plutonium, and fission products in a solidified fluoride salt mixture. The nominal composition of the fluoride salt mixture in the FDTs is 42.16 wt % LiF, 35.79 wt % BeF{sub 2}, 21.01 wt % ZrF{sub 4}, 1.02 wt % UF{sub 4}, and 0.02 wt % PuF{sub 3}. The historic criticality safety study does not meet current standards. This work is in support of a new nuclear criticality safety analysis and approval update. Questions concerning the degree of subcriticality associated with the material in its current state and in its most reactive credible upset condition are addressed. The safety study consists of two parts. In the first part, the FDT Cell was modeled using KENO V.a and analyzed using a variety of cross-section sets. The base FDT Cell model was then modified to represent the most reactive credible upset conditions and analyzed.more » The second part consists of establishing a benchmark for the FDT Cell. Because of the lack of any other relevant benchmark experiments, the original MSRE was also modeled in KENO V.a and analyzed. The results of the reactor model were then compared with documented MSRE reactor conditions. The analysis shows that even under the most reactive credible upset conditions, the MSRE FDT Cell is significantly subcritical.« less

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.