ZZ BUGLE-96, Multigroup Coupled Neutron Gamma Cross-Section for LWR Shielding Calculation

A - Description of program or function: Format: ANISN; Number of groups: 47 Neutron, 20 Gamma-Ray; Nuclides: Ag, Al, Am, Au, B, Ba, Be, Bi, C, Ca, Cd, Cl, Cm, Co, Cr, Cu, Eu, F, Fe, Ga, H, He, Hf, In, K, Li, Mg, Mn, Mo, N, Na, Nb, Ni, Np, O, P, Pa, Pb, Pu, Re, S, Si, Sn, Ta, Th, Ti, U, V, W, Y, Zr. Origin: ENDF/B-IV data; Weighting spectrum: The concrete-spectrum-weighted cross sections have been shown to be generally applicable to a wide range of shielding problems. Flux spectra from five specific locations were used, corresponding to: 1) off-center in a BWR core region, 2) off-center in a PWR core region, 3) the downcomer region in a PWR model, 4) within the pressure vessel at a depth of one-fourth the total thickness, and 5) within the concrete shield surrounding a PWR reactor vessel. 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, designated DLC-185/BUGLE-96, is intended to replace the DLC-75/BUGLE-80 and DLC-76/SAILOR libraries, which are both based on ENDF/B-IV data. It also replaces the DLC-175/BUGLE-93 library by correcting some deficiencies and adding several additional data sets. The processing methodology for BUGLE-96 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. An extensive integral data testing effort was performed to qualify the data and to assess its impact on LWR shielding applications. In general, results using the new data show significant improvements relative to earlier ENDF data. The BUGLE-96 cross sections are intended for use in LWR shielding and pressure vessel dosimetry applications. The multigroup data have been collapsed, and in some cases self-shielded, using flux spectra typical of PWR and BWR reactor models. Flux spectra from five specific locations within these models were used, corresponding to: 1) off-center in a BWR core region, 2) off-center in a PWR core region, 3) the downcomer region in a PWR model, 4) within the pressure vessel at a depth of one-fourth the total thickness, and 5) within the concrete shield surrounding a PWR reactor vessel. The concrete-spectrum-weighted cross sections have been shown to be generally applicable to a wide range of shielding problems. B - Method of solution: BUGLE-96 contains 120 nuclides which have been processed as infinitely dilute and collapsed using an LWR concrete shield spectrum. Additionally, it contains 105 nuclides which have been energy self-shielded and collapsed using LWR-specific material compositions and flux spectra. A separate set of data are included which combines the isotopic BUGLE-96 nuclides into natural elements and provides them with material identifiers which are identical to the original DLC-76/SAILOR library. In addition to the BUGLE-96 and SAILOR-96 data sets, which have been processed without up-scatter in the thermal groups, two new data sets are provided which retain the up-scatter reactions for groups below 5 eV. These data sets are designated as BULGE-96T and SAILOR-96T. Nuclides with Z < 30 (hydrogen through copper) are given in a P7 Legendre expansion while P5 expansion is available for all other nuclides. Several dosimetry and standard response functions are included with the library along with kerma factors for all nuclides. The library was collapsed from the VITAMIN-B6 fine-group library using the AMPX-77 processing code system. VITAMIN-B6 is derived from ENDF/B-VI nuclear data, except for two nuclides (Sn obtained from LENDL and Zirc2 obtained from ENDF/V-IV). The responses and kerma factors were also derived primarily from ENDF/B-VI