[en] An assessment was previously performed to evaluate modeling capabilities and quantify preliminary biases and uncertainties associated with the modeling methods and data utilized in designing a nuclear reactor such as a beryllium-reflected, highly-enriched-uranium (HEU)-O2 fission surface power (FSP) system for space nuclear power. The conclusion of the previous study was that current capabilities could preclude the necessity of a cold critical test of the FSP; however, additional testing would reduce uncertainties in the beryllium and uranium cross-section data and the overall uncertainty in the computational models. A series of critical experiments using HEU metal were performed in the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. Of the hundreds of experiments, three were identified as fast-fission configurations reflected by beryllium metal. These experiments have been evaluated as benchmarks for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). Further evaluation of the benchmark experiments was performed using the sensitivity and uncertainty analysis capabilities of SCALE 6. The data adjustment methods of SCALE 6 have been employed in the validation of an example FSP design model to reduce the uncertainty due to the beryllium cross section data.

[en] The Standardized Computer Analysis for Licensing Evaluation (SCALE) code system developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides a 'plug-and-play' framework with nearly 80 computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE's graphical user interfaces assist with accurate system modeling and convenient access to desired results. SCALE 6.1, scheduled for release in the fall of 2010, provides improved reliability and introduces a number of enhanced features, some of which are briefly described here. SCALE 6.1 provides state-of-the-art capabilities for criticality safety, reactor physics, and radiation shielding in a robust yet user-friendly package. The new features and improved reliability of this latest release of SCALE are intended to improve safety and efficiency throughout the nuclear community.

[en] The format of the TSUNAMI-A sensitivity data file produced by SAMS for cases with deterministic transport solutions is given in Table 6.3.A.1. The occurrence of each entry in the data file is followed by an identification of the data contained on each line of the file and the FORTRAN edit descriptor denoting the format of each line. A brief description of each line is also presented. A sample of the TSUNAMI-A data file for the Flattop-25 sample problem is provided in Figure 6.3.A.1. Here, only two profiles out of the 130 computed are shown.

[en] In the criticality code validation of common systems, many paths may exist to a correct bias, bias uncertainty, and upper subcritical limit. The challenge for the criticality analyst is to select an efficient, defensible, and safe methodology to consistently obtain the correct values. One method of testing criticality code validation techniques is to use a sample system with a known bias as a test application and determine whether the methods employed can reproduce the known bias. In this paper, a low-enriched uranium (LEU) lattice critical experiment with a known bias is used as the test application, and numerous other LEU experiments are used as the benchmarks for the criticality code validation exercises using traditional and advanced parametric techniques. The parameters explored are enrichment, energy of average lethargy causing fission (EALF), and the TSUNAMI integral index ck with experiments with varying degrees of similarity. This paper is an extension of a previously published summary

[en] The SAMS module has been developed to calculate the relative change in the value of k_eff due to a change in a constituent component or cross section. The SAMS module works in conjunction with a modified version of the CSAS25 sequence of SCALE that employs an enhanced version of KENO V.a, which is capable of calculating the spherical harmonics components of the flux moments. The SAMS module performs sensitivity calculations using linear perturbation theory as implemented in the FORSS system and requires the calculation of the forward and adjoint flux moments with the enhanced version of KENO V.a. SAMS automatically selects all of the sensitivity parameters that can be calculated for each nuclide in each region of the system based on available cross-section data. Sensitivity parameters for a given nuclide may be generated for a number of parameters, including total, scatter, capture, and fission cross sections, as well as n and c. The sensitivities for any nuclide-reaction pair calculated with SAMS can be output on three bases: group-wise region dependent, energy-integrated region dependent, and energy- and region-integrated. The sensitivities generated with SAMS have been verified through comparisons with those generated with the SEN1 and SEN2 sensitivity sequences of SCALE. SAMS is capable of producing accurate sensitivities, provided the KENO V.a regions have been appropriately subdivided to allow for sufficient resolution of the flux moments throughout the problem geometry. (author)

[en] ANSI/ANS-8.1-1998;2007, Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors, and ANSI/ANS-8.24-2007, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, require validation of a computer code and the associated data through benchmark evaluations based on physical experiments. The performance of the code and data are validated by comparing the calculated and the benchmark results. A SCALE procedure has been established to generate a Verified, Archived Library of Inputs and Data (VALID). This procedure provides a framework for preparing, peer reviewing, and controlling models and data sets derived from benchmark definitions so that the models and data can be used with confidence. The procedure ensures that the models and data were correctly generated using appropriate references with documented checks and reviews. Configuration management is implemented to prevent inadvertent modification of the models and data or inclusion of models that have not been subjected to the rigorous review process. VALID entries for criticality safety are based on critical experiments documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). The findings of a criticality safety validation of SCALE 6.1 utilizing the benchmark models vetted in the VALID library at Oak Ridge National Laboratory are summarized here.

[en] Ideally, computational method validation is performed by modeling critical experiments that are very similar, neutronically, to the model used in the safety analysis. Similar, in this context, means that the neutron multiplication factors (k_eff) of the safety analysis model and critical experiment model are affected in the same way to the same degree by variations (or errors) in the same nuclear data. Where similarity is demonstrated, the computational bias calculated using the critical experiment model results is 'applicable' to the safety analysis model. Unfortunately, criticality safety analysts occasionally find that the safety analysis models include some feature or material for which adequately similar well-defined critical experiments do not exist to support validation. For example, the analyst may want to take credit for the presence of fission products in spent nuclear fuel. In such cases, analysts sometimes rely on 'expert judgment' to assign an additional administrative margin to compensate for the validation weakness or to conclude that the impact on the calculated bias and bias uncertainty is negligible. Due to advances in computer programs and the evolution of cross-section uncertainty data, analysts can use the sensitivity and uncertainty analyses tools implemented in the SCALE TSUNAMI codes to estimate the potential impact on the application-specific bias and bias uncertainty resulting from nuclides that are under-represented or not present in the critical experiments. This paper discusses the method, computer codes, and data used to estimate the potential contribution toward the computational bias of individual nuclides. The results from application of the method to fission products in a burnup credit model are presented

[en] The Standardized Computer Analysis for Licensing Evaluation (SCALE) code system developed at Oak Ridge National Laboratory (ORNL) includes Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI). The TSUNAMI code suite can quantify the predicted change in system responses, such as k_eff, reactivity differences, or ratios of fluxes or reaction rates, due to changes in the energy-dependent, nuclide-reaction-specific cross-section data. Where uncertainties in the neutron cross-section data are available, the sensitivity of the system to the cross-section data can be applied to propagate the uncertainties in the cross-section data to an uncertainty in the system response. Uncertainty quantification is useful for identifying potential sources of computational biases and highlighting parameters important to code validation. Traditional validation techniques often examine one or more average physical parameters to characterize a system and identify applicable benchmark experiments. However, with TSUNAMI correlation coefficients are developed by propagating the uncertainties in neutron cross-section data to uncertainties in the computed responses for experiments and safety applications through sensitivity coefficients. The bias in the experiments, as a function of their correlation coefficient with the intended application, is extrapolated to predict the bias and bias uncertainty in the application through trending analysis or generalized linear least squares techniques, often referred to as 'data adjustment.' Even with advanced tools to identify benchmark experiments, analysts occasionally find that the application models include some feature or material for which adequately similar benchmark experiments do not exist to support validation. For example, a criticality safety analyst may want to take credit for the presence of fission products in spent nuclear fuel. In such cases, analysts sometimes rely on 'expert judgment' to select an additional administrative margin to account for gap in the validation data or to conclude that the impact on the calculated bias and bias uncertainty is negligible. As a result of advances in computer programs and the evolution of cross-section covariance data, analysts can use the sensitivity and uncertainty analysis tools in the TSUNAMI codes to estimate the potential impact on the application-specific bias and bias uncertainty resulting from nuclides not represented in available benchmark experiments. This paper presents the application of methods described in a companion paper.

[en] The applicability of proposed critical experiments for the criticality code validation of a series of prototypic reactor-grade and weapons-grade mixed-oxide systems has been assessed with the TSUNAMI methodology from SCALE 5. The application systems were proposed by the Nuclear Energy Agency (NEA) Organization for Economic Cooperation and Development (OECD) Working Party on Nuclear Criticality Safety MOX Experimental Needs Working Group. Forty-eight application systems were conceived to envelope the range of conditions in processing and fabrication of reactor-grade and weapons-grade MOX fuel. The applicability of 303 existing critical benchmarks to each of the 48 applications was assessed, and validation coverage was found to be lacking for certain applications. Two series of proposed critical experiments were also considered in this analysis. The TSUNAMI analysis has revealed that both series of proposed experiments are applicable to numerous configurations of the reactor-grade and weapons-grade systems. A detailed assessment of which experiments were revealed by TSUNAMI to be most applicable to specific prototypic fuel processing systems has been performed. (author)

[en] The computational bias of criticality safety computer codes must be established through the validation of the codes to critical experiments. A large collection of suitable experiments has been vetted by the International Criticality Safety Benchmark Experiment Program (ICSBEP) and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). A total of more than 350 cases from this reference have been prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID) maintained by the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The performance of the KENO V.a and KENO-VI Monte Carlo codes within the Scale 6.1 code system with ENDF/B-VII.0 cross-section data in 238-group and continuous energy is assessed using the VALID models of benchmark experiments. The TSUNAMI tools for sensitivity and uncertainty analysis are utilized to examine some systems further in an attempt to identify potential causes of unexpected results. The critical experiments available for validation of the KENO V.a code cover eight different broad categories of systems. These systems use a range of fissile materials including a range of uranium enrichments, various plutonium isotopic vectors, and mixed uranium-plutonium oxides. The physical form of the fissile material also varies and is represented as metal, solutions, or arrays of rods or plates in a water moderator. The neutron energy spectra of the systems also vary and cover both fast and thermal spectra. Over 300 of the total cases used utilize the KENO V.a code. The critical experiments available for the validation of the KENO-VI code cover three broad categories of systems. The fissile materials in the systems vary and include high and intermediate-enrichment uranium and mixed uranium/plutonium oxides. The physical form of the fissile material is either metal or rod arrays in water. As with KENO V.a, both fast and thermal neutron energy spectra are represented in the systems considered. The results indicate generally good performance of both the KENO V.a and KENO-VI codes across the range of systems analyzed. The bias of calculated k_eff from expected values is less than 0.9% Δk in all cases. All eight categories of experiments show biases of less than 0.5% Δk in KENO V.a with the exception of intermediate enrichment metal systems using the 238-group library. The continuous energy library generally manifests lower biases than the multi-group data. The KENO-VI results show slightly larger biases, though this may primarily be the result of modeling systems with more geometric complexity, which are more difficult to describe accurately, even with a generalized geometry code like KENO-VI. Several additional conclusions can be drawn from the results of this validation effort. These conclusions include that the TSUNAMI tools can be used successfully to explain the cause of aberrant results, that some evaluations in the IHECSBE should be updated to provide more rigorous expected k_eff values and uncertainties, and that potential cross-section errors can be identified by detailed review of the results of this validation. It also appears that the overall cross-section uncertainty as quantified through the Scale covariance library is overestimated. Overall, the KENO V.a and KENO-VI codes are shown to provide consistent, low bias results for a wide range of physical systems of potential interest in criticality safety applications.