M. B. Chadwick

A consistent set of neutron kerma coefficients from thermal to 150 MeV for biologically important materials

Neutron cross sections for nonelastic and elastic reactions on a range of elements have been evaluated for incident energies up to 150 MeV. These cross sections agree well with experimental cross section data for charged-particle production as well as neutron and photon production. Therefore they can be used to determine kerma coefficients for calculations of energy deposition by neutrons in matter. Methods used to evaluate the neutron cross sections above 20 MeV, using nuclear model calculations and experimental data, are described. Below 20 MeV, the evaluated cross sections from the ENDF/B-VI library are adopted. Comparisons are shown between the evaluated charged-particle production cross sections and measured data. Kerma coefficients are derived from the neutron cross sections, for major isotopes of H, C, N, O, Al, Si, P, Ca, Fe, Cu, W, Pb, and for ICRU-muscle, A-150 tissue-equivalent plastic, and other compounds important for treatment planning and dosimetry. Numerous comparisons are made between our kerma coefficients and experimental kerma coefficient data, to validate our results, and agreement is found to be good. An important quantity in neutron dosimetry is the kerma coefficient ratio of ICRU-muscle to A-150 plastic. When this ratio is calculated from our kerma coefficient data, and averaged over the neutron energy spectra for higher-energy clinical therapy beams [three p (68) + Be beams, and a d (48.5) + Be beam], a value of 0.94 +/- 0.03 is obtained. Kerma ratios for water to A-150 plastic, and carbon to oxygen, are also compared with measurements where available.

Monte Carlo Simulation for Particle and γ-Ray Emissions in Statistical Hauser-Feshbach Model

Monte Carlo simulations for particle and γ-ray emissions from a compound nucleus based on the Hauser-Feshbach statistical theory with pre-equilibrium emission are performed. The simulation yields reliable nuclear-reaction-wise energy spectra, or so-called exclusive spectra, for emitted neutrons and γ-rays, which are required in particle transport calculations for nuclear applications. The Monte Carlo method is applied to neutron-induced nuclear reactions on 56Fe, and the results are compared with a traditional deterministic method. The neutron and γ-ray emission correlation is examined by gating on an 847 keV γ-ray that is produced by an inelastic scattering process. The partial γ-ray energy spectra for different γ-ray multiplicities are inferred using this Monte Carlo method. In addition, we investigate a correlation between two neutrons in the (n,2n) reaction.

Development and Validation of the 7Li(p, n) Nuclear Data Library and Its Application in Monitoring of Intermediate Energy Neutrons

Systematics have been created for neutron spectra from the 7Li(p, n) reaction at 0° in the 50-200 MeV proton energy region. The available experimental data in the continuum part of the spectra show satisfactory overall agreement with a representation based on the phase-space distribution corresponding to the three-body breakup process 7Li(p, n 3He)α, with empirical correction factors, which depend regularly on incident energy. Validation of the systematics included folding of the predicted neutron spectra with standard 238U neutron fission cross section. Modeled in this way distributions of neutron-induced fission events agree reasonably with experimental data.

Evaluation and propagation of the 239Pu fission cross-section uncertainties using a monte carlo technique

Abstract We present an approach to uncertainty quantification for nuclear applications that combines the covariance evaluation of differential cross-section data and the error propagation from matching a criticality experiment using a neutron-transport calculation. We have studied the reduction in uncertainty of 239Pu fission cross sections by using a one-dimensional neutron-transport calculation with the PARTISN code. The evaluation of 239Pu differential cross-section data is combined with a criticality measurement (Jezebel) using a Bayesian method. To quantify the uncertainty in such calculations, we generate a set of random samples of the cross sections, which represents the covariance matrix, and estimate the distribution of calculated quantities, such as criticality. We show that inclusion of the Jezebel data reduces uncertainties in estimating neutron multiplicity.

Uncertainty Quantification of Prompt Fission Neutron Spectrum for n(0.5 MeV) + 239Pu

Abstract Uncertainties associated with the prompt fission neutron spectrum (PFNS) of n(0.5 MeV) + 239Pu evaluated for the ENDF/B-VII.0 library are estimated using known experimental information and model parameter uncertainties in the framework of the Madland-Nix model. The model parameters used for the ENDF/B-VII.0 evaluation are also used in the present work. A covariance matrix is obtained, and its eigenvalues are estimated. Sampled spectra are then used in PARTISN transport simulations to infer the impact of PFNS uncertainties on the calculation of the multiplication factor keff in the Jezebel critical assembly. The present evaluated PFNS uncertainties lead to ˜0.24% uncertainty in the Jezebel keff. Finally, multigroup covariance matrices are produced in 33- and 590-group structures.

Nuclear data for accelerator-driven systems

Abstract We review recent evaluations of neutron and proton reaction cross sections up to 150 MeV in the LA150 Library, for use in computer code simulations of accelerator-driven systems. An overview is provided of the nuclear theory together with measured cross section data used in the evaluations. The possible use of bismuth activation foils for high-energy neutron spectrometry is also discussed. We describe recent developments to the MCNPX radiation transport code, which merges MCNP and LAHET in one code and uses the LA150 evaluated data. A number of benchmark comparisons against integral experiments are described, for thick-target neutron production, neutron transmission through macroscopic slabs, and neutron kerma coefficients. The benchmarks help validate the transport code and the evaluated data for use in ADS simulations of neutron production in a spallation target (n/p), radiation shielding, heating, and damage. A brief summary is also given of future data needs for subcritical transmuters and spallation targets, in accelerator transmutation of waste technologies.

Photonuclear Physics in Radiation Transport - I: Cross Sections and Spectra

Abstract This paper describes model calculations and nuclear data evaluations of photonuclear reactions on isotopes of C, O, Al, Si, Ca, Fe, Cu, Ta, W, and Pb for incident photon energies up to 150 MeV. The calculations, using the GNASH code, include giant-dipole resonance and quasi-deuteron models for photoabsorption. The emission of secondary particles and gamma rays is computed using preequilibrium theory, together with an open-ended sequence of compound nucleus decays using the Hauser-Feshbach theory. The accuracy of the calculated and evaluated cross sections is assessed through extensive comparisons with measured cross sections, average neutron multiplicities, and energy-dependent emission spectra. The evaluated nuclear data files (ENDF) facilitate radiation transport studies of the importance of photonuclear reactions in a number of technologies including photoneutrons produced in electron/photon accelerators, shielding studies, and nondestructive detection of nuclear materials. A companion paper describes developments in the MCNP and MCNPX codes to utilize these data in transport simulations.

Photonuclear Physics in Radiation Transport - III: Actinide Cross Sections and Spectra

Abstract This paper describes model calculations and nuclear data evaluations of photonuclear reactions on actinides such as 235U, 238U, 237Np, and 239Pu for incident photon energies from the reaction threshold up to 20 MeV. The calculations are done using the GNASH code, including the giant-dipole resonance for photoabsorption. The emission of secondary particles is computed using a preequilibrium theory, together with an open-ended sequence of the compound nucleus decay using the Hauser-Feschbach theory. The accuracy of the calculated and evaluated cross sections is assessed through extensive comparison with measured cross sections. This work also summarizes evaluation methods used to create actinide photonuclear files for the forthcoming ENDF/B-VII database, which will facilitate radiation transport studies related to photonuclear reactions in a number of technologies including production of photoneutrons and photofission fragments in electron accelerators, shielding studies, and nondestructive detection of nuclear material in particular.

Predicting Total Reaction Cross Sections for Nucleon-Nucleus Scattering

Nucleon total reaction and neutron total cross sections to 300 MeV for 12C and 208Pb, and for 65 MeV scattering spanning the mass range, are predicted using coordinate space optical potentials formed by full folding of effective nucleon-nucleon interactions with realistic nuclear ground state densities. Good to excellent agreement is found with existing data.

Improved Evaluations of Neutron-Induced Reactions on Americium Isotopes

Abstract The suite of neutron-induced reactions on 241Am, 242gAm, 242mAm, and 243Am has been evaluated for incident neutron energies above the unresolved resonance region and up to 20 MeV. Modern theoretical models and computational techniques were extensively used because of the rather limited experimental information. However, when available, experimental data were used to guide and benchmark the present evaluation work. All evaluation results were formatted and distributed as ENDF data files for possible inclusion in the ENDF/B-VII library.