[en] The neutron total cross section and the shape of the neutron capture cross section of ²³²Th have been measured in the energy range from 0.006 eV to 18 eV using the time-of-flight method. The 100-MeV electron linear accelerator at the RPI Gaerttner Laboratory was used as the source of neutrons. The neutron total cross section was obtained by performing transmission measurements using high-purity metallic ²³²Th samples and a ⁶Li-glass neutron detector. Results above 0.1 eV are in good agreement with the recent ENDF/B-V evaluation. Below 0.1 eV the measured total cross section is significantly lower than the evaluated total cross section. The effects of Bragg scattering are seen for neutron energies less than 0.05 eV. The shape of the neutron capture cross section was obtained from 0.009 eV to 18 eV. Capture measurements were performed with a ThO₂ sample and a 1.25-m-dia. liquid scintillator detector. The shape of the measured capture cross section above 0.1 eV is in reasonable agreement with a recent measurement at Brookhaven National Laboratory. The neutron capture cross section below 0.1 eV is found to increase less rapidly with decreasing neutron energy than 1/v. A simultaneous multilevel fit to the experimental capture cross section and the experimental total cross section was attempted. It was possible to reproduce the data by using resonance parameters which are statistically improbable

[en] Sources of neutron and photon transport data are described as well as the processing of the evaluated data sets into continuous-energy and multigroup cross-section sets. The procedures for checking and validating the processed data are discussed. The question of why so many data sets are available is addressed by indicating the differences between data sets as well as their relative strengths and weaknesses. Suggestions are made to help the MCNP user in selecting appropriate cross-section sets. 31 refs

[en] Capture and transmission measurements were carried out. Eight new resonances were observed below 1.1 keV. Twenty-one new resonances were seen above 1.1 keV. Energy levels and integrated resonance yields are tabulated. The capture yield divided by sample thickness is shown; data were corrected for self-shielding and multiple scattering. ENDF data underpredict the cross section by about 15%. 1 figure, 1 table

[en] The capability to calculate effective delayed neutron fractions has now been implemented into MCNP4B and is in the testing phase. This option should prove to be most useful for multiplying systems which are not easily modeled using deterministic codes

No abstract available

[en] Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment

[en] An improved method of neutron-induced photon production has been incorporated into the Monte Carlo transport code MCNP. The new method makes use of all partial photon-production reaction data provided by ENDF/B evaluators including photon-production cross sections as well as energy and angular distributions of secondary photons. This faithful utilization of sophisticated ENDF/B evaluations allows more precise MCNP calculations for several classes of coupled neutron-photon problems

[en] A probability-table treatment recently has been incorporated into an intermediate version of the MCNP Monte Carlo code named MCNP4XS. This paper presents MCNP4XS results for a variety of uranium and plutonium criticality benchmarks, calculated with and without the probability-table treatment. It is shown that the probability-table treatment can produce small but significant reactivity changes for plutonium and ²³³U systems with intermediate spectra. More importantly, it can produce substantial reactivity increases for systems with large amounts of ²³⁸U and intermediate spectra

[en] A suite of 41 criticality benchmarks has been modeled using MCNP trademark (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes ²³³U, ²³⁵U, ²³⁸U, and ²³⁹Pu. The values of k_eff for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, ²³⁵U, ²³⁸U, and ²³⁹Pu can have a significant impact on the values of k_eff. In addition to the integral quantity k_eff, several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k_eff. Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data

[en] Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs