5 research outputs found
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Sensitivity and uncertainty analysis for nuclear criticality safety using KENO in the SCALE code system
Sensitivity and uncertainty methods have been developed to aid in the establishment of areas of applicability and validation of computer codes and nuclear data for nuclear criticality safety studies. A key component in this work is the generation of sensitivity and uncertainty parameters for typically several hundred benchmarks experiments used in validation exercises. Previously, only one-dimensional sensitivity tools were available for this task, which necessitated the remodeling of multidimensional inputs in order for such an analysis to be performed. This paper describes the development of the SEN3 Monte Carlo based sensitivity analysis sequence for SCALE. Two options in the SEN3 package for the reconstruction of angular-dependent forward and adjoint fluxes are described and contrasted. These options are the direct calculation of flux moments versus the calculation of angular fluxes, with subsequent conversion to flux moments prior to sensitivity coefficient generation. The latter technique is found to be significantly more efficient
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Investigations and Recommendations on the Use of Existing Experiments in Criticality Safety Analysis of Nuclear Fuel Cycle Facilities for Weapons-Grade Plutonium
Sensitivity and Uncertainty (S/U) methods, recently developed at Oak Ridge National Laboratory (ORNL) have been demonstrated to determine the applicability of critical benchmark experiments to the criticality code validation of design systems. These methods, although still under development, have been recently published in several sources. Development of the techniques used in this report was conducted through joint support from the United States Department of Energy (U.S. DOE) and the Nuclear Regulatory Commission (NRC) to provide a physics-based approach for the establishment of the area of applicability of critical experiments per the requirements of ANSI/ANS-8.1. Use of these methods may allow users to interpolate and extrapolate the traditional area of applicability (AOA) of a given set of critical experiments to include new application areas that may not have been anticipated during the experiment design. The new S/U analytical tools include the SEN1 and SEN3 sensitivity analysis sequences, which will be available with the next release of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system. These analysis sequences compute the relative change in the system neutron multiplication factor, k{sub eff}, which would be observed for perturbations in the group-wise neutron cross-section data for each reaction of each nuclide in the system. The CANDE code uses sensitivity data determined separately for the design system applications and the individual experiments, along with the cross-section-covariance data, to calculate integral parameters which give a measure of the similarity between a particular design system and an experimental benchmark. A high-valued integral parameter for an experiment application pair indicates that the experiment demonstrates similar properties to the application. Thus, the experiment is applicable for the criticality code validation of the design system. A theoretical basis for the S/U techniques applied in this report is given in Sect. 2. This report pertains to two of the five AOAs identified by the licensee [Duke, Cogema, Stone and Webster (DCS)] for the validation of criticality codes in the design of the Mixed-Oxide Fuel Fabrication Facility (MFFF). The five AOAs are as follows: (1) Pu-nitrate aqueous solutions (homogeneous systems), (2) Mixed-oxide (MOX) pellets, fuel rods and fuel assemblies (heterogeneous systems), (3) PuO{sub 2} powders, (4) MOX powders, and (5) Aqueous solutions of Pu compounds (Pu-oxalate solutions). This report addresses a S/U analysis pertaining to AOA 3, PuO{sub 2} powders, and AOA 4, MOX powders. AOA 3 and AOA 4 are the subject of this report since the other AOAs (solutions and heterogeneous systems) appear to be well represented in the documented benchmark experiments used in the criticality safety community. Prior to this work, DCS used traditional criticality validation techniques to identify numerous experimental benchmarks that are applicable to AOAs 3 and 4. Traditional techniques for selection of applicable benchmark experiments essentially consist of evaluating the area of applicability for important design parameters (e.g., Pu content or average neutron energy) and ensuring experiments have similar characteristics that bound or nearly bound the range of conditions requiring design analysis. DCS provided ORNL with compositions and dimensions for critical systems used to establish preliminary mass limits for facility powder and fuel pellet handling areas corresponding to AOAs 3 and 4. ORNL has reviewed existing critical experiments to identify those, which, in addition to those provided by DCS, may be applicable to the criticality code validation for AOAs 3 and 4. A S/U analysis was then performed to calculate the integral parameters used to determine the similarity of each critical experiment to each design system provided by DCS. This report contains a review of the S/U theory, a description of the design systems, a brief description of the critical experiments evaluated for applicability, and the results of the S/U analysis determining the applicability of each experiment to each application
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Development of Nuclear Analysis Capabilities for DOE Waste Management Activities
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. RESEARCH PROGRESS AND IMPLICATION
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EMSP project summary (Project ID: 60077): Development of nuclear analysis capabilities for DOE waste management activities
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
Methods and issues for the combined use of integral experiments and covariance data Results of a NEA International Collaborative Study
International audienceThe Working Party on International Nuclear Data Evaluation Cooperation (WPEC) of the Nuclear Science Committee under the Nuclear Energy Agency (NEA/OECD) established a Subgroup (called "Subgroup 33") in 2009 on "Methods and issues for the combined use of integral experiments and covariance data." The first stage was devoted to producing the description of different adjustment methodologies and assessing their merits. A detailed document related to this first stage has been issued. Nine leading organizations (often with a long and recognized expertise in the field) have contributed ANL, CEA, INL, IPPE, JAEA, JSI, NRG, IRSN and ORNL. In the second stage a practical benchmark exercise was defined in order to test the reliability of the nuclear data adjustment methodology. A comparison of the results obtained by the participants and major lessons learned in the exercise are discussed in the present paper that summarizes individual contributions which often include several original developments not reported separately.The paper provides the analysis of the most important results of the adjustment of the main nuclear data of 11 major isotopes in a 33-group energy structure. This benchmark exercise was based on a set of 20 well defined integral parameters from 7 fast assembly experiments. The exercise showed that using a common shared set of integral experiments but different starting evaluated libraries and/or different covariance matrices, there is a good convergence of trends for adjustments. Moreover, a significant reduction of the original uncertainties is often observed. Using the a-posteriori covariance data, there is a strong reduction of the uncertainties of integral parameters for reference reactor designs, mainly due to the new correlations in the a-posteriori covariance matrix. Furthermore, criteria have been proposed and applied to verify the consistency of differential and integral data used in the adjustment. Finally, recommendations are given for an appropriate use of sensitivity analysis methods and indications for future work are provided. © 2014