INTERNATIONAL ATOMIC ENERGY AGENCY NUCLEAR DATA SERVICES DOCUMENTATION SERIES OF THE IAEA NUCLEAR DATA SECTION IAEA-NDS-141 Rev. 2, Oct. 1993 THE INTERNATIONAL REACTOR DOSIMETRY FILE (IRDF-90 Version 2) Assembled by N.P. Kocherov, and P.K. McLaughlin Abstract: This document describes the contents of the new version of the International Reactor Dosimetry File IRDF-90 Ver. 2 which contains recommended neutron cross-section data to be used for reactor neutron dosimetry by foil activation. It also contains selected recommended values for radiation damage cross-sections and benchmark neutron spectra. This library supersedes all earlier versions of IRDF. It is available on magnetic tape or on a set of PC diskettes from the IAEA Nuclear Data Section, costfree, upon request. Nuclear Data Section e-mail: RNDS@IAEA1, BITNET International Atomic Energy Agency fax: (43-1) 234564 P.O. Box 100 cable: INATOM VIENNA A-1400 Vienna telex: 1-12645 atom a Austria telephone: (43-1) 2360-1709 The International Reactor Dosimetry File (IRDF-90) Assembled by N.P. Kocherov, and P.K. McLaughlin 1. Introduction Since the first release of the IRDF-90 v. 1 file in summer 1990 we have received many comments from its users. The main problems were identified in the covariance information (Files 33). Since then also some new evaluations appeared which were not available at the time of the release of version 1. Six new covariance files were added to the file. They were also not available before. In its present form the file contains 58 cross-sections of dosimetry reactions, all with complete covariance information. 9 new dosimetry reactions were added compared to version 1. The IRDF-90 version 2 contains 39 neutron dosimetry reaction cross-sections from the latest revisions of the ENDF/B-6 [1], 14 evaluations made by Prof. H Vonach and his co-workers at the IRK in Vienna [2] and 5 evaluations by the specialists from the Chinese Nuclear Data Center in Beijing, prepared specially for this file under contract with the IAEA [3]. The data in the original ENDF-6 format were processed to 640 group extended SANDII format in the Nuclear Data Section of the IAEA using the processing codes LINEAR, RECENT and GROUPIE by D.E. Cullen [4]. The covariance information is not processed by these codes and it is contained in IRDF-90 in the original ENDF-6 format. 2. Contents of the IRDF-90 The list of reactions and the origins of evaluations are given in Table 1. As we did not have any new sets of standard damage cross-sections or of standard and reference neutron spectra the ones from IRDF-85 were kept here with the same special notations. The damage cross-sections and neutron spectra are in the ENDF-5 format. Data Content: File 1 Cross section data in ENDF/B-VI format 25211 records for 58 reactions File 2 Damage cross sections in ENDF/B-V format 754 records for 4 materials File 3 Spectra data files in ENDF/B-V format 1598 records for 10 benchmark neutron fields In File 3 neutron spectra for the following benchmark neutron fields are given Cf-252 spontaneous fission - NBS Evaluation U-235 thermal fission - NBS evaluation U-235 thermal fission - ENDF/B-V evaluation ISNF Intermediate-energy standard neutron field CFRMF Coupled fast reactivity measurement facility BIG-TEN 10% enriched uranium cylindrical critical assembly (LANL) SIGMA- Coupled thermal/fast uranium and boron carbide spherical SIGMA assembly (MOL) ORR Reactor in Oak Ridge National Laboratoy YAYOI Spectrum (JAERI) Central zone flux of the NEACRP benchmark All improvements in the file became possible only through efficient cooperation between Drs. H. Nolthenius, E. Zsolnay, and E. Szondi who were testing the file [5,6] and Drs. H. Vonach, S. Tagesen and D. Hetrick who made the necessary improvements in the covariance data files. Their contribution is gratefully acknowledged. We would appreciate receiving any suggestions concerning further improvement of the quality of this file. Please send comments to: Dr. N.P. Kocherov International Atomic Energy Agency Wagramerstr. 5, P.O. Box 5 A-1400 Vienna, Austria References 1. U.S. National Nuclear Data Center, Evaluated Nuclear Data File, ENDF/B-6, BNL, Upton, N.Y. (1990) and later revisions. 2. M. Wagner, H. Vonach. A. Pavlik, B. Strohmaier, S. Tagesen, J. Martinez-Rico, "Evaluation of Cross-Sections for 14 Important Neutron Dosimetry Reactions," Physics Data, 13-5, Karlsruhe, 1990. 3. C. Dunjiu, "Evaluations of Cross-Sections for Dosimetry Reactions," Final Report on Contract 5516, INDC(CPR)-024, 1991, Vienna. 4. D.E. Cullen, "The 1992 ENDF/B Preprocessing Codes", Report IAEA-NDS-39 Rev. 7, 1992. 5. E.M. Zsolnay, H. Nolthenius, "On the Quality of the Uncertainty Information in the International Dosimetry File IRDF-90," Report ECN-1-93-019, ECN, Petten, 1993. 6. H. Nolthenius, E.M. Zsolnay, E.J. Szondi, "Testing of the IRDF-90 Cross- Section Library in Benchmark Neutron Spectra," Reactor Dosimetry ASTM 1228, Harry Farrar IV, E. Parvin Lippincott, and John G. Williams, Eds., American Society for Testing and Materials, Philadelphia, to be published in 1994. Table 1. Contents of the IRDF-90 E-6 = data taken over from ENDF/B-VI Original = data evaluated for IRDF-90 Priv. Comm. = Private Communication Nuclide IRDF Reactions and* Author & Lab ** Date Library MAT No. Uncertainties of Origin 3-Li-6 325 3 105; 33 105 G. Hale et al., LANL 1989 E-6 5-B-10 525 3 1; 3 107; G. Hale et al., LANL 1989 E-6 33 107 9-F-19 925 3 16; 33 16 M. Wagner et al., IRK 1991 Original 11-Na-23 1123 3 102; 33 102 Yu Hanrong, CNDC 1990 Priv. Comm. 12-Mg-24 1225 3 103; 33 103 M. Wagner Et al., IRK 1991 Original 13-Al-27 1325 3 103; 33 103 D. Hetrick, C.Y. Fu, ORNL 1990 Priv. Comm. 3 107; 33 107 M. Wagner et al., IRK 1991 Original 15-P-31 1525 3 103; 33 103 M. Wagner et al., IRK 1991 Original 16-S-32 1625 3 103; 33 103 D. Hetrick, C.Y. Fu, ORNL 1991 Priv. Comm. 21-Sc-45 2126 2 151; 32 151; Z. Zhao, CNDC 1991 Priv. Comm. 3 103; 33 102 22-Ti-46 2225 3 103; 33 103 D. Hetrick, C.Y. Fu, ORNL 1989 Priv. Comm. 22-Ti-47 2228 3 28; 33 28; D. Hetrick, C.Y. Fu, ORNL 1990 E-6 3 103; 33 103 22-Ti-48 2231 3 28; 33 28 C. Philis et al., ANL 1977 E-6 3 103; 33 103 D. Hetrick, C.Y. Fu, ORNL 1990 Priv. Comm. 23-V-0 2300 3 107; 33 107 A. Smith, D. Smith, ANL 1990 Priv. Comm. 24-Cr-52 2431 3 16; 33 16 M. Wagner et al., IRK 1991 Original 25-Mn-55 2525 2 151; 3 16; K. Shibata et al., JAERI 1988 E-6 33 16; 3 102; ORNL 33 102 26-Fe-54 2625 3 103; 33 103 D. Hetrick, et al., ORNL 1989 Priv. Comm. 26-Fe-56 2631 3 103; 33 103 C. Fu et al., ORNL 1991 E-6 26-Fe-58 2637 2 151; 3 102; N. Larson et al., ORNL 1989 E-6 33 102 2 151; 3 102; A. Smith et al., ANL 1990 E-6 33 102; 3 107; 33 107 28-Ni-58 2825 3 103; 33 103 N. Larson et al., ORNL 1989 E-6 3 16; 33 16 M. Wagner et al., IRK 1990 Original 28-Ni-60 2831 3 103; 33 103 N. Larson et al., ORNL 1991 E-6 29-Cu-63 2925 3 16; 33 16 M. Wagner et al., IRK 1991 Original 2 151; 3 102; C. Fu et al., ORNL 1991 E-6 33 102; 3 107; 33 107 29-Cu-65 2931 3 16; 33 16 C. Fu et al., ORNL 1991 E-6 30-Zn-64 3025 3 103; 33 103 M. Wagner et al., IRK 1991 Original 39-Y-89 3925 3 16; 33 16 R. Howerton, A. Smith 1991 E-6 D. Smith, LLNL, ANL 40-Zr-90 4025 3 16; 33 16 M. Wagner et al., IRK 1991 Original 41-Nb-93 4125 3 16; 3 51; M. Wagner et al., IRK 1991 Original 3 102 33 16; 33 51; A. Smith et al., ANL, 1991 E-6 33 102 LLL 45-Rh-103 4525 3 51; 33 51 M. Wagner et al., IRK 1991 Original 47-Ag-109 4731 3 102; 33 102 Z. Zhao, CNDC 1990 Priv. Comm. 48-Cd-0 4800 3 1; 3 102 S. Pearlstein, BNL 1991 E-690 (translated from UK) Nuclide IRDF Reactions and* Author & Lab ** Date Library MAT No. Uncertainties of Origin 49-In-115 4931 2 151; 3 16; C. Dunjiu, CCNDC 1991 Priv. Comm. 33 16; 3, 51; 33, 51; S. Chiba et al., ANL 1990 E-6 3 102; 33 102 53-I-127 5325 3 16; 33 16 Z. Wenrong et al., CNDC 1991 Priv. Comm. 64-Gd-0 6400 3 1; 3 102 Mixed from E-6 isotopes 1990 Original by N. Kocherov, IAEA 79-Au-197 7925 2 151; 3 102 P. Young, LANL 1989 E-6 33 102 3 16; 33 16 M. Wagner et al., IRK 1991 Original 90-Th-232 9040 2 151; 3 18 M. Bhat et al., BNL, 1990 E-6 3 102; 33 18 ANL 33 102 92-U-235 9228 2 151; 3 18 L. Weston et al., ORNL, 1989 E-6 33 18 LANL 92-U-238 9237 2 151; 3 18 L. Weston et al., ORNL, 1989 E-6 33 18; 3 102 LANL 33 102 93-Np-237 9337 2 151; 3 18; F. Mann et al., HEDL, 1978 E-4 33 18 SRL 94-Pu-239 9437 2 151; 3 18 P. Young et al., LANL 1989 E-6 33 18 26-Fe-00 8000 ASTM Damage Priv. Comm. W. Zijp 1979 Priv. Comm. 26-Fe-00 8001 Eur. Damage Priv. Comm. W. Zijp 1979 Priv. Comm. Cross Sections 24-Cr-00 8002 Eur. Damage W. Zijp, Petten 1985 Priv. Comm. Cross Sections 28-Ni-00 8003 Eur. Damage W. Zijp, Petten 1985 Priv. Comm. Cross Sections Note: * The following ENDF notations for reactions are used 1-total, 16-n,2n, 18-fission, 28-n,np, parameters. 51 means total population of the 1st level from all channels (not an ENDF notation); 3 - cross-section data file; 33 - covariance data file. ** The lab codes given under "Author & Lab" are as follows: ANL - Argonne National Laboratory, Argonne Illinois BNL - Brookhaven National Laboratory, Upton, N.Y. CNDC - Chinese Nuclear Data Center IAEA - International Atomic Energy Agency, Vienna IRK - Inst. fuer Radiumforschung und Kernphysik, Vienna JAERI - Japanese Atomic Energy Research Inst., Tokai LANL - Los Alamos National Laboratory, New Mexico LLNL - Lawrence Livermore National Laborarory, California ORNL - Oak Ridge National Laboratory, Tennessee Petten - Netherland's Energy Research Foundation, Petten SRL - Savannah River Laboratory, South Carolina