RSIC DATA LIBRARY DLC-229

1. NAME AND TITLE OF DATA LIBRARY

IRDF-2002: The International Reactor Dosimetry Files.

2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS

No retrieval programs are included.

3. CONTRIBUTORS

International Atomic Energy Agency, Nuclear Data Section, Vienna, Austria, and Institute Jozef Stefan, Slovenia, through the OECD Nuclear Energy Agency Data Bank, Issy les-Moulineaux, France.

4. HISTORICAL BACKGROUND AND INFORMATION

The official website is: http://www-nds.iaea.org/irdf2002/.

IRDF-2002 is the third release in the International Reactor Dosimetry File series, following IRDF-82 and IRDF-90. This is the first time that decay parameters and abundances have been presented in IRDF. The library was created to serve as a standardized, updated and benchmarked evaluated cross section library of neutron dosimetry reactions with related uncertainty information, for use in the lifetime management assessments of nuclear power reactors and other applications. RSICC received IRDF-2002 through the NEADB, from which it is distributed with identifier: IAEA-867/04. These data were initially released by RSICC in October 2007 with package identifier D00229MNYCP00.

A library in ACE-dosimetry format for the MCNP family of codes was generated from the pointwise IRDF-2002 data at Institute Jozef Stefan in Slovenia. 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. For the ACE-format data, the NEADB identifier is NEADB ID: IAEA-867/05. These data were added to the RSICC package in July 2008. D00229MNYCP01 now contains all the data in both IRDF-2002 packages distributed by the NEADB:

NEADB ID: IAEA-867/04 -- ZZ-IRDF-2002 -- IRDF-2002 pointwise

NEADB ID: IAEA-867/05 -- ZZ-IRDF-2002-ACE -- ACE-format data

5. APPLICATION OF THE DATA

ZZ-IRDF-2002
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The International Reactor Dosimetry File (IRDF-2002) contains recommended neutron cross-section data to be used for reactor neutron dosimetry by foil activation and subsequent neutron spectrum unfolding. It also contains selected recommended values for radiation damage cross-sections and benchmark neutron spectra. Two related programs available from NEADB and RSICC are: SPECTER-ANL (PSR-263) & STAY’SL (PSR-113).

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ZZ-IRDF-2002-ACE: MCNP

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6. SOURCE AND SCOPE OF DATA

ZZ-IRDF-2002
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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-VI.
WEIGHTING SPECTRUM:

o Typical MTR spectrum used in the input of the cross-section uncertainty processing code.

o Flat weighting spectrum used in converting the pointwise cross-section data to the extended SAND-II group structure.

IRDF-2002 consists of three main data sets:

o 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), for GaAs displacement (ASTM standard E722-94).

o 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.

o Nuclear data
- decay data for all reactions of interest;
- isotopic abundances for all reactions of interest.

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ZZ-IRDF-2002-ACE

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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: -

7. DISCUSSION OF THE DATA RETRIEVAL PROGRAMS

No retrieval program is included. An html hyperlinked index is provided to view the data in ZZ-IRDF-2002. The MAKXSF program that is distributed with MCNP can be used with the ZZ-IRDF-2002-ACE data.

8. DATA FORMAT AND COMPUTER

HTML, PDF and ASCII text files; PC, UNIX Workstations. RSICC ID: D00229MNYCP01.

Card images. The expanded libraries require ~270 MB of disk space.

9. TYPICAL RUNNING TIME

Runtimes vary.

10. REFERENCES

a Included in the RSICC document file - D229.pdf:

O. Bersillon et al., “International Reactor Dosimetry File - 2002 (IRDF - 2002),”
IAEA-TECDOC-DRAFT.

b. Included in the distribution files:

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).

11. CONTENTS OF PACKAGE

The package contains the references listed above in Section 10 and the data library, which is transmitted on one CD in a GNU-compressed Linux tar file. Windows users may extract the files with WinZIP 8.0 or later.

12. DATE OF ABSTRACT

October 2007, revised July 2008.

KEYWORDS: BASED ON ENDF/B-VI; DAMAGE CROSS SECTIONS; DOSIMETRY CROSS SECTIONS; MCNP FORMAT; REACTION CROSS SECTIONS; NEUTRON CROSS SECTIONS; RADIATION ENVIRONMENT