ZZ IRDF-2002, 640-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-6 Format. ZZ IRDF-2002-ACE, Cross-Section Library and Spectra for Dosimetry Calculation in ACE Format for Monte Carlo methods. ZZ IRDF-90, 640-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-6 Format. ZZ IRDF-82, 620-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-5 Format

1 - Description of problem or function: - ZZ-IRDF-82: ENDF-5 Format; 620 group (SAND II) Dosimetry Library. Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, Mn, Fe, Co, Ni, Cu, Zn, Zr, Nb, Rh, In, I, Au, Th, U, Np, Pu, Am. - ZZ-IRDF-90: ENDF-6 Format; 640 groups extended SAND II structure. Nuclides: Li, B, F, Mg, Al, P, S, Sc, Ti, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Nb, Rh, Cd, Ir, Gd, Au, Th, U, Np, Pu, V. Damage cross section for Fe, Cr, Ni. Weighting spectrum: Maxwell spectrum, 1/E spectrum and Watt fission spectrum. - ZZ-IRDF-2002: ENDF-6 Format (pointwise cross-section data). SAND II 640 energy group structure (multigroup data). Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Y, Zr, Nb, Rh, Ag, In, I, La, Pr, Tm, Ta, W, Au, Hg, Pb, Th, U, Np, Pu, Am, Cd, Gd. Damage cross section for Fe, Cr, Ni, Si, GaAs displacement. Weighting spectrum: - Typical MTR spectrum used in the input of the cross-section uncertainty processing code. - Flat weighting spectrum used in converting the pointwise cross-section data to the extended SAND-II group structure. - ZZ-IRDF-2002-ACE: ACE Format (continuous energy cross-section data for Monte Carlo). Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Y, Zr, Nb, Rh, Ag, In, I, La, Pr, Tm, Ta, W, Au, Hg, Pb, Th, U, Np, Pu, Am, Cd, Gd. Damage cross section for Fe, Cr, Ni, Si, GaAs displacement. - (A) ZZ-IRDF-82: The 1982 version of the International Reactor Dosimetry File is composed of two different parts. The first part is made up of a collection of dosimetry Cross sections and the second part contains a collection of benchmark spectra. For ease of use in dosimetry applications both Cross sections and spectra are distributed in multigroup form. Each of these two parts is in the ENDF/B-V Format as a separate computer file. I) The dosimetry cross section library contains the following data: (1) The entire ENDF/B-V Dosimetry Library (Mod. 1) in the form of 620 group averaged Cross sections, using the SAND-II group structure. (2) The reactions 19F(n,2n), 24Mg(n,p), 31P(n,p), 29Cu(n,2n), 64Zn(n,p), 90Zr(n,2n), 93Nb(n,n') and 103(Rh(n,n'), supplied by Vonach. This data was converted to the ENDF/B-V Format, which in turn was converted to 620 group form. (3) The reaction 23Na(n,2n) provided by Marcinkowski. This data was converted to the ENDF/B-V Format (5) and then converted to 620 group Format. (4) The reaction 241Am(n,f) as supplied by Patrick. This data was converted to the ENDF/B-V Format at Stuttgart and then converted to 620 group form. (5) ASTM and EUR standards damage Cross sections for iron as provided by Zijp (10) in the form of 620 group Cross sections. This data was converted to the ENDF/B-V Format. II) The Benchmark Spectra library contains ten benchmark spectra, including: (1) The NBS 252Cf spontaneous fission; the NBS 235U and ENDF/B-V 235U thermal fission, the Intermediate-Energy Standard Neutron Field (ISNF), the Coupled Fast Reactivity Measurement Facility (CFRMF), the 10 % Enriched Uranium Cylindrical Critical Assembly (BIG-TEN) and the Coupled Thermal/Fast Uranium and Boron Carbide Spherical Assembly (SIGMA-SIGMA) spectra, all of which were provided by Eisenhauer in 620 group form. (2) The ORR and YAYOI Spectra, which were provided by Greenwood in 100 group form. (3) The Central Zone Flux of the NEACRP Benchmark Spectra provided by Goel in 208 group form. - (B) ZZ-IRDF-90: The following changes compared to the 1982 version should be noted: - IRDF-90 is mainly based on the ENDF/B-IV data and its present version contains cross section values for 51 different dosimetry reactions. Damage Cross sections are given for three different materials: Fe, Cr, Ni. - The library is written in the ENDF-6 Format. - Gas production reactions are completely missing from IRDF-90.V.1 and other important dosimetry reactions are also not yet present. At the same time, double cross section information is given for the following reactions Ni582, Cu632 and Fe damage. - No benchmark spectra are included in IRDF-90. - (C) ZZ-IRDF-2002: consists of three main data sets: (1) Multigroup data: - cross section data for 66 neutron activation (and fission) reactions; - total Cross sections of three cover materials B, Cd and Gd; - radiation damage Cross sections of the following elements and compounds: Fe dpa cross section (ASTM standard E693-01); dpa cross section for a special steel composition (EURATOM); dpa Cross sections for Cr and Ni IRDF-90), for Si (ASTM standard E722-94), and for GaAs displacement (ASTM standard E722-94). (2) Pointwise data: - all cross sections listed above, except radiation damage cross sections; - total Cross sections of all neutron capture and fission reactions in the library. (3) Nuclear data: - decay data and isotopic abundances for all reactions of interest. - (D) ZZ-IRDF-2002-ACE: A library in ACE-dosimetry Format for the MCNP family of codes derived from from the pointwise IRDF-2002 data. The ACE reaction MT* numbers are related to the ENDF MT numbers as MT* = MT +1000*(10+LFS) where LFS is the metastable state designator of the reaction product. 2 - Related or auxiliary programs: IAEA0867/04: - ZZ-IRDF-2002: Two programs are related to this library. Both programs are distributed by the NEA Computer Program Service. - SPECTER-ANL, Neutron Damage for Material Irradiation (PSR-0263); - STAY-SL, Dosimetry Unfolding with Activation, Dosimetry, Flux Error Calculation (PSR-0113). IAEA0867/05: - ZZ-IRDF-2002-ACE: MCNP. 3 - Status: IAEA0867/01: 18-FEB-1983 Tested at NEADB; IAEA0867/02: 19-Jan-1994 Obsolete; IAEA0867/03: 01-FEB-1994 Screened; IAEA0867/04: 10-Oct-2005 Screened; IAEA0867/05: 25-FEB-2008 Screened. 3 - Programming language(S) used: IAEA0867/01: FORTRAN-IV