Review of ENDF/B-VI Fission-Product Cross Section

In response to concerns raised in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 93-2, the U.S. Department of Energy (DOE) developed a comprehensive program to help assure that the DOE maintain and enhance its capability to predict the criticality of systems throughout the complex. Tasks developed to implement the response to DNFSB recommendation 93-2 included Critical Experiments, Criticality Benchmarks, Training, Analytical Methods, and Nuclear Data. The Nuclear Data Task consists of a program of differential measurements at the Oak Ridge Electron Linear Accelerator (ORELA), precise fitting of the differential data with the generalized least-squares fitting code SAMMY to represent the data with resonance parameters using the Reich-Moore formalism along with covariance (uncertainty) information, and the development of complete evaluations for selected nuclides for inclusion in the Evaluated Nuclear Data File (ENDFB). The current ENDF/B library was developed for fast and thermal fission reactors and fusion reactors. Criticality safety practitioners recognize that many situations around the DOE complex are characterized by neutron spectra in the intermediate-energy region, as opposed to the high-energy region for fast reactors and fusion systems and the low-energy region for thermal reactors. Consequently, the Nuclear Data Task focuses primarily on the intermediate-energy region so that upgrades to existing evaluated data will remove deficiencies in the current ENDF/B evaluations. The ORELA allows high-resolution measurements in the intermediate-energy region and the SAMMY fitting code provides high quality resonance parameters in the resolved and unresolved energy range using the sophisticated Reich-Moore (RM) formalism for superior representation of the data in the intermediate energy region. In addition, the SAMMY fitting procedure provides covariance information for the resonance parameters that can be used in subsequent analyses to assess the uncertainty in calculated results and provide a better interpretation of criticality safety margins. Thus, the thrust of the Nuclear Data Task is to obtain high-resolution data in the intermediate energy region and provide fits to the data that utilize the modern RM formalism and covariance information for subsequent use in criticality predictability applications. As a subtask of the Nuclear Data Task, this review of the fission-product cross sections has several objectives. The first objective is a general data status review at various levels for the some 200 fission products. The second objective is a more detailed investigation of the top 20 fission products with regard to thermal- and intermediate-energy capture and scatter cross sections. The third objective is to demonstrate the revision of ENDF/B evaluations utilizing new data and evaluation techniques for 13 fission products. The fourth objective is to make recommendations for improvements, both specific and general in nature

Assessment of the available {sup 233}U cross-section evaluations in the calculation of critical benchmark experiments

In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U.S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the S{sub n} transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems

Revised evaluations of fission-product cross sections

This paper reports on revised cross-section evaluations for {sup 134}Ba, {sup 149}Sm, {sup 154}Eu, {sup 155}Eu, {sup 160}Dy, {sup 161}Dy, {sup 162}Dy, {sup 163}Dy, and {sup 164}Dy. The evaluations for {sup 134}Ba, {sup 154}Eu, and {sup 1554}Eu were previously revised for ENDF/B-VI. The other 6 evaluations, carried over from ENDF/B-V, were completed in the 1974--1980 time period. The evaluations for the dysprosium isotopes go back to ENDF/B-IV. Newer experimental data, not considered for the current ENDF/B-VI evaluations, was used in all of the revised evaluations. In the present work the primary emphasis was placed on the resolved and unresolved resonance regions, but newer measured data were also used for energies above the unresolved resonance region. Elastic, capture, and total cross sections are revised. Some important parameters from the revised evaluations are given in Table 1; corresponding values from the ENDF/B-VI evaluations are also given

Evaluation of the fission and capture cross sections of /sup 240/Pu; and /sup 241/Pu for ENDF/B-V. [10/sup -5/ eV to 20 MeV]

Since there were appreciable new data which were not available for ENDF/B-IV, new evaluations for /sup 240/Pu and /sup 241/Pu were carried out for ENDF/B-V. The evaluation of the fission and capture cross sections is reviewed and problem areas are discussed. The neutron energy range of concern was from 10/sup -5/ eV to 20 MeV. Significant changes were made over the entire neutron energy region because of the new experimental data available. The problems in the evaluations due to discrepancies in the nuclear data are emphasized, particularly the 1-eV resonance in /sup 240/Pu and the 0.3-eV resonance in /sup 241/Pu. The evaluation of the fission and capture cross sections for ENDF/B-V represents an improvement over the previous evaluation; however, there continues to be a need for accurate experimental data. 7 figures

Benchmark testing of {sup 233}U evaluations

In this paper we investigate the adequacy of available {sup 233}U cross-section data (ENDF/B-VI and JENDL-3) for calculation of critical experiments. An ad hoc revised {sup 233}U evaluation is also tested and appears to give results which are improved relative to those obtained with either ENDF/B-VI or JENDL-3 cross sections. Calculations of k{sub eff} were performed for ten fast benchmarks and six thermal benchmarks using the three cross-section sets. Central reaction-rate-ratio calculations were also performed

Review of ENDF/B-VI Fission-Product Cross Sections[Evaluated Nuclear Data File]

In response to concerns raised in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 93-2, the US Department of Energy (DOE) developed a comprehensive program to help assure that the DOE maintain and enhance its capability to predict the criticality of systems throughout the complex. Tasks developed to implement the response to DNFSB recommendation 93-2 included Critical Experiments, Criticality Benchmarks, Training, Analytical Methods, and Nuclear Data. The Nuclear Data Task consists of a program of differential measurements at the Oak Ridge Electron Linear Accelerator (ORELA), precise fitting of the differential data with the generalized least-squares fitting code SAMMY to represent the data with resonance parameters using the Reich-Moore formalism along with covariance (uncertainty) information, and the development of complete evaluations for selected nuclides for inclusion in the Evaluated Nuclear Data File (ENDFB)

Comparison of the ENDF/B-V and SOKRATOR evaluations of /sup 235/U, /sup 239/Pu, /sup 240/Pu and /sup 241/Pu at low neutron energies

The US and USSR's most recent evaluationsof /sup 235/U, /sup 239/Pu, /sup 240/Pu and /sup 241/Pu are compared over the thermal region and over the first few resonances. The two evaluations rest on essentially the same experimental data base and the differences reflect different approaches to the representation of the cross sections or different weightings of the experimental results. It is found that over the thermal and resolved ranges the two evaluations are very similar. Some differences in approaches are briefly discussed

Automated sensitivity analysis using the GRESS language

An automated procedure for performing large-scale sensitivity studies based on the use of computer calculus is presented. The procedure is embodied in a FORTRAN precompiler called GRESS, which automatically processes computer models and adds derivative-taking capabilities to the normal calculated results. In this report, the GRESS code is described, tested against analytic and numerical test problems, and then applied to a major geohydrological modeling problem. The SWENT nuclear waste repository modeling code is used as the basis for these studies. Results for all problems are discussed in detail. Conclusions are drawn as to the applicability of GRESS in the problems at hand and for more general large-scale modeling sensitivity studies

Isotopic dilution of {sup 233}U with depleted uranium for criticality safety in processing and disposal

The disposal of excess {sup 233}U as waste is being considered. Because {sup 233}U is a fissile material, a key requirement for processing {sup 233}U to a final waste form and disposing of it is the avoidance of nuclear criticality. For many processing and disposal options, isotopic dilution is the most feasible and preferred option to avoid nuclear criticality. Isotopic dilution is dilution of fissile {sup 233}U with nonfissile {sup 238}U. The use of isotopic dilution removes any need to control nuclear criticality in process or disposal facilities through geometry or chemical composition. Isotopic dilution allows the use of existing waste management facilities that are not designed for significant quantities of fissile materials to be used for processing and disposing of {sup 233}U. The amount of isotopic dilution required to reduce criticality concerns to reasonable levels was determined in this study to be approximately 0.53 wt % {sup 233}U. The numerical calculations used to define this limit consisted of a homogeneous system of silicon dioxide (SiO{sub 2}), water (H{sub 2}O), {sup 233}U and depleted uranium (DU) in which the ratio of each component was varied to learn the conditions of maximum nuclear reactivity. About 188 parts of DU (0.2 wt % {sup 235}U) are required to dilute 1 part of {sup 233}U to this limit in a water-moderated system with no SiO{sub 2} present. Thus for the U.S. inventory of {sup 233}U, several hundred metric tons of DU would be required for isotopic dilution

Benchmarking of EPRI-cell epithermal methods with the point-energy discrete-ordinates code (OZMA)

The purpose of the present study is to benchmark E-C resonance-shielding and cell-averaging methods against a rigorous deterministic solution on a fine-group level (approx. 30 groups between 1 eV and 5.5 keV). The benchmark code used is OZMA, which solves the space-dependent slowing-down equations using continuous-energy discrete ordinates or integral transport theory to produce fine-group cross sections. Results are given for three water-moderated lattices - a mixed oxide, a uranium method, and a tight-pitch high-conversion uranium oxide configuration. The latter two lattices were chosen because of the strong self shielding of the /sup 238/U resonances