We describe recent nuclear model calculations and evaluations of neutron reactions on actinides in the forthcoming ENDF/B-VII US nuclear data library, with particular attention to the uranium isotopes U232-241 in the keV - 30 MeV energy range. This work makes use of extensive sets of measurements for fission, elastic, inelastic, (n, xn) and capture, as well as fission probability data. Nuclear reaction model calculations were performed for the whole suite of uranium isotopes to allow us to take advantage of the systematical properties from isotope-to-isotope, which is especially useful for nuclides where few measurements exist. In addition to improving the neutron cross sections and energy-angle distributions, new prompt fission neutron spectra and prompt/delayed neutron multiplicity evaluations are included for several isotopes. We also give some examples of integral data testing using MCNP simulations of critical assemblies, for both k-eff calculations and fission, n2n, and capture reaction rate calculations.

We describe recent nuclear model calculations and evaluations of neutron reactions on the uranium isotopes U232-241 in the keV to 30-MeV energy range. This work makes use of extensive sets of measurements for fission, elastic, inelastic, (n,xn) and capture, as well as fission probability data. The U-235(n,f) standard cross section was revised, and the fission cross sections of the uranium isotopes, as wen as Np-237 and Pu-239, were updated using the revised standard. Nuclear reaction model calculations were performed for the whole suite of uranium isotopes to allow us to take advantage of the systematical properties from isotope-to-isotope, which is especially useful for nuclides where few measurements exist. In addition to improving the neutron cross sections and energy-angle distributions, new prompt fission neutron spectra and prompt/delayed neutron multiplicity evaluations are included for several isotopes. These evaluations are among the pre-ENDF/B-VII evaluations that are currently being considered for the new ENDF file. A companion paper in this Conference by MacFarlane describes critical-assembly integral data testing results for U isotopes.

Our Nuclear Physics Group has worked over the last several years to improve the ENDF/B actinide cross sections, especially as measured against the famous series of fast-spectrum critical experiments performed at Los Alamos under such names as Godiva, Jezebel, the Flattops, and Bigten. The new evaluations include changes in the high-energy fission cross sections, nubar, elastic scattering, inelastic scattering, fission spectra, and delayed neutrons that combine to significantly improve the calculated results for the Los Alamos critical assemblies. As a happy byproduct of this work, we found that the new evaluations also removed about half of a long-standing discrepancy in calculations of thermal-reactor critical experiments using lattices of low-enriched uranium oxide rods. In the meantime, work at the Oak Ridge National Laboratory has resulted in new proposed resonance parameters for U238. When these low-energy data are combined with the new high-energy data from Los Alamos, they have the effect of removing most of the rest of the problem for thermal lattices. Although more work will be done at both Los Alamos and Oak Ridge, including making the evaluations consistent with the new standards, the results presented in this talk provide promise that ENDF/B-VII will result in good improvements for users at both high and low neutron energies.

Recent upgrades of Am-241, Am-242m, and Am-240 nuclear data in the keV-30-MeV range are described. The new evaluation takes advantage of recent measurements and advances in calculational modeling methods. The model calculations are especially important for the nuclear data of americium isotopes, because few measurements are available. The nuclear-model code GNASH is extensively used for our evaluations. The new evaluations are given for total, fission, capture, (n, 2n), and (n, 3n) reaction cross sections, and v(p) for Am-241 and Am-242m. A new evaluation for Am-240 is also given by expanding our modeling feasibility.

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.

This is the second of two companion papers. Thefirst 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 This paper describes the steps taken to process these files into transport libraries and to update the Monte Carlo N-Particle (MCNP) and M-CNPX radiation transport codes to use tabular photonuclear reaction data. The evaluated photonuclear data files are created in the standard evaluated nuclear data file (ENDF) format. These files must be processed by the NJOY data processing system into A Compact ENDF (ACE) files suitable for radiation transport calculations. MCNP and MCNPX have been modified to use these new data in a self-consistent and fully integrated manner. Verification problems were used at each step along the path to check the integrity of the methodology. The resulting methodology and tools provide a comprehensive system for using photonuclear data in radiation transport calculations. Also described are initial validation simulations used to benchmark several of the photonuclear transport tables.

The NJOY Nuclear Data Processing System is used to convert evaluated nuclear data in ENDF format into forms useful for applications. As a bridge between physics and engineering, it is best used by people with some knowledge of things like nuclear reaction theory, resonance theory, or scattering theory on one side, and some knowledge of things like particle transport codes, reactor core calculations, or radiation medicine on the other. These lectures provide an introduction you to NJOY system (using version 97.107), give a quick outline of the physics of nuclear data, and show how some of the main applications in nuclear technology link to NJOY results. This material is also available on line, including more details and exercises, at http://t2.lanl.gov/joy.

New accelerator-driven technologies that utilize spallation neutrons, such as the production of tritium and the transmutation of radioactive waste, require accurate nuclear data to model the performance of the target/blanket assembly and To predict neutron production, activation, heating, shielding requirements, and material damage. To meet these needs, nuclear-data evaluations and libraries up to 150 MeV have been developed for use in transport calculations to guide engineering design. By using advanced nuclear models that account for details of nuclear structure and the quantum,nature of the nuclear scattering, significant gains in accuracy can be achieved below 150 MeV where intranuclear cascade calculations become less accurate. Evaluations are in ENDF-6 format for important target/blanket and shielding materials (isotopes of H, C, N, O, Al, Si, P, Ca, Ci; Fe, Ni, Cu, Nb, W: Hg, and Pb) for both incident neutrons and incident protons. The evaluations are based on measured da:ta as well as predictions from the GNASH nuclear model code, which calculates cross sections using Hauser-Feshbach, exciton, and Feshbach-Kerman-Koonin preequilibrium models. Elastic scattering distributions and direct reactions are calculated from the optical model. All evaluations specify production cross sections and energy-angle correlated spectra of secondary light particles as well as production cross sections and energy distributions of heavy recoils and gamma rays. A formalism developed to calculate recoil energy distributions is presented. The use of these nuclear data in the MCNPX radiation transport code is also briefly described. This code merges essential elements of the LAHET and MCNP codes and uses these new data below 150 MeV and intranuclear cascade collision physics at higher energies. Extensive comparisons are shown between the evaluated results and experimental cross-section data to benchmark and validate the evaluated library. In addition, integral benchmarks of calculated and measured kerma coefficients for neutron energy deposition and neutron transmission through an iron slab compared with MCNPX calculations are provided. These evaluations have been accepted into the ENDF/B-VI library as Release 6.

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.

The CINDER'90 code and library continue to be developed for the description of nuclide inventories produced in a wide range of radiation environments. Although generally applicable, CINDER'90 is typically applied with ancillary codes and the LAHET Code System to describe a variety of accelerator transmutation problems. The first release version of the CINDER'90 code and library LibC90a permit the calculation of inventories of 3400 nuclides of Z less than or equal to 103, using evaluated neutron reaction data below 20 MeV. Data extension efforts are now progressing to extend reaction data to 150 MeV for neutrons and protons, limited to targets of Z less than or equal to 84, in LibC90b. Public release of the code and libraries is now planned for the Fall of 1997.

A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library is designated BUGLE-93. The processing methodology is consistent with ANSI/ANS 6.1.2, since the ENDF data were first processed into a fine-group, “pseudo problem-independent” format and then collapsed into the final broad-group format. The fine-group library is designated VITAMIN-B6. An extensive integral data testing effort was also performed. In general, results using the new data show significant improvements relative to earlier ENDF data.

This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem.

"Response to “Limitations on the Use of the THOR Critical Assembly for Validation of n + 239Pu Cross Sections”." Nuclear Science and Engineering, 92(4), p. 610

Atomic displacement, gas production, transmutation, and nuclear heating can all be calculated with the NJOY nuclear data processing system using evaluated data in ENDF/B format. Using NJOY helps assure consistency between damage cross sections and those used for transport, and NJOY provides convenient interface formats for linking data to application codes. Unique features of the damage calculation include a simple momentum balance treatment for radiative capture and a new model for (n,particle) reactions based on statistical model calculations. Sample results for iron and nickel are given and compared with the results of other methods.

Accurate calculations of atomic displacement damage in materials exposed to neutrons require detailed spectra for primary recoil nuclei. Such data are not available from direct experimental measurements. Moreover, they cannot always be computed accurately starting from evaluated nuclear data libraries such as ENDF/B-V that were developed primarily for neutron transport applications, because these libraries lack detailed energy-and-angle distributions for outgoing charged particles. Fortunately, a new generation of nuclear model codes is now available that can be used to fill in the missing spectra. One example is the preequilibrium statistical-model code GNASH. For heating and damage applications, a supplementary code called RECOIL has been developed. RECOIL uses detailed reaction data from GNASH, together with angular distributions based on Kalbach-Mann systematics to compute the energy and angle distributions of recoil nuclei. The energy-angle distributions for recoil nuclei and outgoing particles are written out in the new ENDF/B “File 6” format. The result is a complete set of nuclear data that can be used to calculate displacement-energy production, heat production, gas production, transmutation and activation. Sample results for iron are given and compared to the results of conventional damage models such as those used in NJOY.

A major revision of the ENDF/B-V evaluation of neutron-induced nuclear data for 239Pu has been completed for neutron energies between 8 keV and 20 MeV. The most important changes to the evaluation include incorporation of a comprehensive new theoretical analysis based on recent experimental data to replace part of the total cross-section file and all of the elastic and inelastic cross sections and secondary distributions, reevaluation of the prompt and total average neutron multiplicities from fission for incident energies between 0.4 and 11.5 MeV to correct discrepancies of almost 3% with new experimental data, and the replacement of all secondary neutron energy spectra from fission with improved shapes based on approximations to a new theoretical method. The results have been validated by calculating measured quantities for five fast critical assemblies. The evaluation is being distributed as Revision 2 of ENDF/B-V by the National Nuclear Data Center at Brookhaven National Laboratory.