New precision measurements of the 235U(n,γ) cross section.

The neutron capture cross section of (235)U was measured for the neutron incident energy region between 4 eV and 1 MeV at the DANCE facility at the Los Alamos Neutron Science Center with an unprecedented accuracy of 2-3% at 1 keV. The new methodology combined three independent measurements. In the main experiment, a thick actinide sample was used to determine neutron capture and neutron-induced fission rates simultaneously. In the second measurement, a fission tagging detector was used with a thin actinide sample and detailed characteristics of the prompt-fission gamma rays were obtained. In the third measurement, the neutron scattering background was characterized using a sample of (208)Pb. The relative capture cross section was obtained from the experiment with the thick (235)U sample using a ratio method after the subtraction of the fission and neutron scattering backgrounds. Our result indicates errors that are as large as 30% in the 0.5-2.5 keV region, in the current knowledge of neutron capture as embodied in major nuclear data evaluations. Future modifications of these databases using the improved precision data given herein will have significant impacts in neutronics calculations for a variety of nuclear technologies.

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.

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.

Nuclear data for radiotherapy: presentation of a new ICRU report and IAEA initiatives.

An ICRU report entitled "Nuclear Data for Neutron and Proton Radiotherapy and for Radiation Protection" is in preparation. The present paper presents an overview of this report, along with examples of some of the results obtained for evaluated nuclear cross sections and kerma coefficients. These cross sections are evaluated using a combination of measured data and the GNASH nuclear model code for elements of importance for biological, dosimetric, beam modification and shielding purposes. In the case of hydrogen both R-matrix and phase-shift scattering theories are used. In the report neutron cross sections and kerma coefficients will be presented up to 150 MeV and proton cross sections up to 250 MeV. An IAEA Consultants' Meeting was also convened to examine the "Status of Nuclear Data needed for Radiation Therapy and Existing Data Development Activities in Member States". Recommendations were made regarding future endeavours.

Neutron, proton, and photonuclear cross-sections for radiation therapy and radiation protection.

I review recent work at Los Alamos undertaken to evaluate neutron, proton, and photonuclear cross-sections up to 150 MeV (to 250 MeV for protons), based on experimental data and nuclear model calculations. These data are represented in the ENDF format and can be used in computer codes to simulate radiation transport. They permit calculations of absorbed dose in the body from therapy beams, and through use of kerma coefficients allow absorbed dose to be estimated for a given neutron energy distribution. In radiation protection, these data can be used to determine shielding requirements in accelerator environments and to calculate neutron, proton, gamma-ray, and radionuclide production. Illustrative comparisons of the evaluated cross-section and kerma coefficient data with measurements are given.