IAEA0867 ZZ-IRDF.

1. NAME OR DESIGNATION OF PROGRAM:  ZZ-IRDF-82; ZZ-IRDF-90; ZZ-IRDF-2002; ZZ-IRDF-2002-ACE.

3. DESCRIPTION OF PROBLEM OR FUNCTION

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ZZ-IRDF-82
==========
FORMAT: ENDF-5 format
NUMBER OF GROUPS: 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.
ORIGIN: ENDF/B-V + Vonach + Marcinkowski + Patrick
WEIGHTING SPECTRUM:
  
ZZ-IRDF-90
==========
FORMAT: ENDF-6 format
NUMBER OF GROUPS: 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.
ORIGIN: ENDF/B-IV mainly
WEIGHTING SPECTRUM: Maxwell spectrum, 1/E spectrum and Watt fission spectrum.
    
ZZ-IRDF-2002
============
FORMAT: ENDF-6 format (pointwise cross-section data).
NUMBER OF GROUPS: SAND II 640 energy group structure (multigroupe 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.
ORIGIN:IRDF-90, RRDF-98, JENDL/D-99, JEFF 3.0, ENDF/B-IV.
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
================
FORMAT: 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.
ORIGIN:IRDF-90, RRDF-98, JENDL/D-99, JEFF 3.0, ENDF/B-IV.
WEIGHTING SPECTRUM: -
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
  
(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. Dosimetry Cross Sections
---------------------------
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. Benchmark Spectra
---------------------
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 made to IRDF-90 compared to the 1982 version of IRDF 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, in most cases accompanied with uncertainties in the form  of covariance information. Damage cross sections are given for three different materials: Fe, Cr, Ni.
- The library is written in the ENDF-6 format, which differs from the  ENDF-5 format applied in the former Reactor Dosimetry File. Therefore, difficulties may arise in processing the new data with the old computer codes interpreting the input only in the ENDF-5 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 - derived from different evaluations - is given for the following reactions NI582, CU632 and  FE DAMAGE. - No benchmark spectra are included in IRDF-90.
  
(C) ZZ-IRDF-2002
================
IRDF-2002 consists of three main data sets:
(1) Multigroup data
- cross section data for 66 neutron activation (and fission) reactions, along by uncertainties in the form of covariance information;
- total cross sections of three cover materials B, Cd and Gd, without uncertainty information;
- 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, accompanied with their uncertainty information.
(3) Nuclear data
- decay data for all reactions of interest;
- 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.

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)

ZZ-IRDF-2002-ACE:
================
MCNP distributed by the US Dept. of Energy Center:
Radiation Safety Information Computational Center
Oak Ridge National Laboratory
Post Office Box 2008
1 Bethel Valley Road
OAK RIDGE, TN 37831-6171
U. S. A.
https://rsicc.ornl.gov/Default.aspx

10. REFERENCES

- D. Garber, et al.:
  Data Formats and Procedures for the Evaluated Nuclear Data File,
  ENDF
  BNL-NCS-50496 (ENDF-102), Brookhaven (1975).
- B. Magurno:
  Private Communication, Brookhaven (1981).
- R.L. Simons, W.M. McElroy:
  Evaluated Reference Cross Section Libraries
  BNWL-1312, Richland (1970).
- S. Tagesen, H. Vonach, B. Strohmaier:
  Physics Data - Nr. 13-1 (1979) and Nr. 13-2 (1980), Vienna
- V. Pronyaev, O. Schwerer:
  Private Communcation, IAEA, Vienna (1981).
- D.E. Cullen:
  Program GROUPIE (Version 79-1): Calculation of Bondarenko self-
  shielded cross sections and multiband parameters from data in the
  ENDF/B format (UCRL-50400, Vol. 17, part D, Livermore 1980.
)
- Marcinkowski:
  Private Communication, Warsaw (1980).
- B. Patrick,
  AERE-R-8528, Harwell (1979).
- M. Mattes,
  Private Communication, Stuttgart (1981).
- W.L. Zijp,
  Private Communication, Petten (1981).
- C. Eisenhauer,
  Private Communcation, NBS, Washington (1980).
- L. Greenwood,
  Private Communcation, Argonne (1981).
- B. Goel,
  Private Communcation, Karlsruhe (1981).

- N.P. Kocherov and H.K. Vonach:
  International Reactor Dosimetry File (IRDF-90) Status and Testing
  INDC/P(90)-22.
- N.P. Kocherov and P.K. McLaughlin:
  The International Reactor Dosimetry File (IRDF-90)
  IAEA-NDS-141, Rev. 2 (Oct, 1993).
- E.M. Zsolnay and H.J. Nolthenius:
  Test of the IRDF-90.V.1 Reactor Dosimetry Library
  ECN-I--91-004 (January 1991).
- H.D. Lemmel:
  Index to BROND-2, CENDL-2, ENDF/B-6, JEF-2, JENDL-3, IRDF
  IAEA-NDS-107, Rev. 8 (November 1993).
- N.P. Kocherov, P.K. McLaughlin ;
"The International Reactor Dosimetry File (IRDF-90)
IAEA-NDS-141, Rev.0 (August 1990)
- D.E. Cullen, P.K. Mc Laughlin:
"The International Reactor Dosimetry File (IRDF-85)
IAEA-NDS-11, Rev.1 (April 1985)
- O. Bersillon et al.:
International Reactor Dosimetry File - 2002 (IRDF - 2002)
IAEA-TECDOC-DRAFT
- L. R. Greenwood and R. Paviotti-Corcuera:
Summary Report of the Technical Meeting on 'International Reactor Dosimetry
File: IRDF-2002' IAEA Headquarters, Vienna, Austria 27-29 August 2002
INDC(NDS)-435 (September 2002)
- Patrick J. Griffin and R. Paviotti-Corcuera:
Summary Report of the Final Technical Meeting on 'International Reactor
Dosimetry File: IRDF-2002' IAEA Headquarters Vienna, Austria 1-3 October 2003
INDC(NDS)-448 (October 2003)
- H. K. Vonach, A. Trkov, V. Pronyaev, M. Herman, R. Paviotti-Corcuera:
Minutes of an Informal Consultants' Meeting IAEA-NDS, Vienna, 12 July 2001
- E. Zsolnay, E. Szondi, A. Trkov, R. Paviotti-Corcuera
Minutes of an Informal Consultants' Meeting IAEA-NDS, Vienna, 7 August 2001
- O. Gritzay, M. Vlasov, L. Chervonna, V. Zerkin, N. Klimova, G. Kolota:
Neutron Excitation Function Guide For Reactor Dosimetry
INDC(UKR)-005 (January 2002)
- O. Bersillon et al.:
International Reactor Dosimetry File - 2002 (IRDF - 2002)
IAEA-TECDOC-DRAFT

15. NAME AND ESTABLISHMENT OF AUTHOR

ZZ-IRDF-82
==========
Assembled by: D.E. Cullen, N. Kocherov, P.M. McLaughlin
              International Atomic Energy Agency
              Nuclear Data Section
              P.O. Box 100
              Vienna, Austria
and           H.K. Vonach
              Institut fuer Radiumforschung und Kernphysik
              P.O. Box 100
              A-1400 Vienna, Austria
  
ZZ-IRDF-90
==========
Assembled by: N.P. Kocherov, P.K. McLaughlin
              International Atomic Energy Agency
              Nuclear Data Section
              P.O. Box 100
              Vienna, Austria
  
ZZ-IRDF-2002
============
              Nuclear Data Section
              Division of Physical and Chemical Sciences
              International Atomic Energy Agency
              Wagramer Strasse 5
              P.O. Box 100
              A-1400 Vienna, Austria
  
ZZ-IRDF-2002-ACE
================
              Dr. Andrej TRKOV
              Institute Jozef Stefan
              Jamova 39
              1000 Ljubljana
              Slovenia
  
              for the Nuclear Data Section
              Division of Physical and Chemical Sciences
              International Atomic Energy Agency
              Wagramer Strasse 5
              P.O. Box 100
              A-1400 Vienna, Austria

Damage cross sections FILE
Decay data
Standard spectra
Dosimetry cross sections in pointwise ENDF-6 format (compressed)
Dosimetry cross sections in groupwise ENDF-6 format
Dosimetry cross sections in metrology format
SPECTER: a code for radiation damage parameter calculations
STAYSL: a code for least squares neutron spectral adjustment
Documentation in electronic format

IRDF2002   the IRDF-2002 library in ACE-dosimetry format
IRDF2002.xsd   the 'xsdir' section that has to be pasted into the active
'xsdir' file of MCNP to make the dosimetry reaction cross sections accessible
IRDF2002.lst listing of all materials and reactions in the dosimetry library
Electronic documentation

File name	File description	Records
IAEA0867_03.001	Information file	32
IAEA0867_03.002	IRDF-90/G Version-2 ENDF/B-IV x-sec, part 1	12682
IAEA0867_03.003	IRDF-90/G Version-2 ENDF/B-IV x-sec, part 2	12531
IAEA0867_03.004	Damage x-sec in ENDF/B-V format	754
IAEA0867_03.005	Spectra data files in ENDF/B-V format	1598