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