International Reactor Dosimetry File 2002 - IAEA Publications

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1. INTRODUCTION R. Paviotti-Corcuera, E.M. Zsolnay The most recently tested version of the International Reactor Dosimetry File, IRDF-90 Version 2 (IRDF-90.2), was released in 1993. Most of the evaluations used in this file were prepared in the mid-1980s, and in the meantime a large amount of new experimental data has become available, along with two new national reactor dosimetry libraries (the Russian Reactor Dosimetry File (RRDF-98) and the Japanese Evaluated Nuclear Data Library (JENDL/D-99)). The cross-sections and related uncertainties for several reactions in these libraries may be of better quality than the data in the older IRDF-90 file. These developments have resulted in different cross-section values being applied to the evaluation of experimental data, creating difficulties in comparing the results of reactor dosimetry calculations from the same types of nuclear facility. Therefore, there has been a strong demand from the reactor dosimetry community for an updated and standardized version of the IRDF. The IAEA has in the past supported similar efforts to improve the quality of data for reactor dosimetry applications; some examples are documented in Refs [1.1–1.11]. A major objective of the present data development project was to prepare and distribute a standardized, updated and tested reactor dosimetry cross-section library accompanied by uncertainty information (IRDF-2002) for use in service life assessments of nuclear power reactors. In order to achieve this objective, two technical meetings were organized. Both meetings were held at the IAEA in Vienna. The first meeting took place from 27 to 29 August 2002, the second from 1 to 3 October 2003 [1.12, 1.13]. Recommendations were made concerning the following topics and the preparation of the library: reactions to be included, requirements for new evaluations or revisions, nuclear decay data, radiation damage data, testing of the data in benchmark fields and inclusion of computer codes. The participants emphasized that good quality nuclear data for reactor dosimetry are essential to improve assessments of the service life of reactor pressure vessels. Accurate cross-section data are also essential in other neutron metrology applications such as boron neutron capture therapy, therapeutic uses of medical radioisotopes, nuclear physics measurements and reactor safety studies. The work undertaken within the project included the following tasks: (a) Detailed analyses and comparisons of the cross-section data and the related uncertainty information present in different reactor dosimetry and general purpose libraries, including IRDF-90.2, JENDL/D-99 and 1
(b) (c) (d) (e) RRDF-98, and the most recent releases of ENDF/B-VI, JEFF-3.0 and CENDL-2. Comparisons were also made of the calculated integral crosssection data with experimental reaction rates in standard neutron fields. Selection of the best quality cross-section information based on the above comparisons. Evaluation and testing of new reaction cross-sections, as requested by the reactor dosimetry community for extension of the library. Selection of evaluated and up to date nuclear decay characteristics and isotopic abundances. Testing of important dosimetry cross-sections in reference benchmark neutron fields. Although the release of IRDF-2002 and publication of the related documentation occurred after 2002, participants attending the second Technical Meeting decided to retain the title IRDF-2002, since the library has been referred to as this in the open literature. A CD-ROM containing the full contents of IRDF-2002 accompanies this report. Updated versions of this library will also be released by the IAEA on CD-ROM. 1.1. CONTENTS OF THE LIBRARY IRDF-2002 contains the best quality data for reactor dosimetry applications available at the time of preparation. These data include cross-sections and related uncertainties, nuclear decay parameters for the reaction product nuclei and abundances of the target nuclides. This is the first time that the decay parameters and abundances have been presented in the IRDF library. IRDF-2002 consists of three main data sets: (a) Multigroup data: (i) Cross-section data for 66 neutron activation (and fission) reactions, along with uncertainties in the form of covariance information. (ii) Total cross-sections of three types of cover material, boron, cadmium and gadolinium, without uncertainty information. (iii) Radiation damage cross-sections of the following elements and compounds: iron dpa cross-section (American Society for Testing and Materials (ASTM) standard E693-01); dpa cross-section for a special steel composition (Euratom); dpa cross-sections for chromium and nickel (IRDF-90), for silicon (ASTM standard E722- 94) and for GaAs displacement (ASTM standard E722-94). 2
Page 2 and 3: INTERNATIONAL REACTOR DOSIMETRY FIL
Page 4 and 5: TECHNICAL REPORTS SERIES No. 452 IN
Page 6 and 7: FOREWORD An accurate and complete k
Page 9 and 10: Contributing authors O. Bersillon C
Page 11 and 12: 5.1. Plots of experimental data and
Page 16 and 17: (b) (c) Pointwise data: (i) All dos
Page 18 and 19: REFERENCES TO SECTION 1 [1.1] ZIJP,
Page 20 and 21: cross-sections, and as a consequenc
Page 22 and 23: TABLE 2.1. MEASURED AND CALCULATED
Page 24 and 25: TABLE 2.2. MEASURED AND CALCULATED
Page 26 and 27: [2.24] KOBAYASHI, K., KIMURA, I., M
Page 28 and 29: from ENDF/B-VI Release 8, JEFF-3.0
Page 30 and 31: TABLE 3.1. REACTIONS FROM THE VARIO
Page 32 and 33: Fluence rate per unit energy (m -2
Page 34 and 35: TABLE 3.2. COMPARISON OF THE INTEGR
Page 36 and 37: — 10 B(n,α) 7 Li and 6 Li(n,t) 4
Page 38 and 39: TABLE 3.3. THERMAL NEUTRON CROSS-SE
Page 40 and 41: TABLE 3.4. RELATIVE STANDARD DEVIAT
Page 42 and 43: [3.14] HOLDEN, N.E., “Neutron sca
Page 44 and 45: The rigorous inclusion of all uncer
Page 46 and 47: TABLE 4.1. C/E VALUES IN THE CALIFO
Page 48 and 49: TABLE 4.3. IRDF-90.2 ( DATA IN A CA
Page 50 and 51: TABLE 4.3. IRDF-90.2 ( DATA IN A CA
Page 52 and 53: TABLE 4.4. JENDL/D-99 ( DATA IN A C
Page 54 and 55: TABLE 4.4. JENDL/D-99 ( DATA IN A C
Page 56 and 57: TABLE 4.5. UPDATED RRDF-98 ( DATA I
Page 58 and 59: REFERENCES TO SECTION 4 [4.1] KOCHE
Page 60 and 61: neutron yield substantially. Furthe
Page 62 and 63: TABLE 5.1. CROSS-SECTION EVALUATION
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TABLE 5.1. CROSS-SECTION EVALUATION
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— 58 Ni(n,2n) 57 Ni: The IRDF, JE
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— 238 U(n,γ) 239 U(β - ) 239 Np
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(i) (ii) (iii) (iv) (v) Only diagon
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TABLE 6.1. INTEGRAL CHARACTERISTICS
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TABLE 6.1. INTEGRAL CHARACTERISTICS
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TABLE 6.2. CROSS-SECTIONS IN IRDF-2
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7. CONSISTENCY TEST OF THE CROSS-SE
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TABLE 7.1. REFERENCE NEUTRON FIELDS
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Fluence, (E) (MeV -1 ) 10 7 10 6 10
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TABLE 7.2. SPECTRUM AVERAGED DOSIME
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testing of electronic components an
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Table 7.3 details the 34 dosimetry
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7.2. RESULTS OF CONSISTENCY TESTING
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TABLE 7.4. RATIO OF SPECTRUM AVERAG
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[7.7] SEKINE, T., MAEDA, S., AOYAMA
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ecommendations found in ASTM practi
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10 2 Damage (MeV· mb) 10 1 10 0 10
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10 2 Damage (MeV· mb) 10 1 10 0 10
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9. DECAY DATA AND ISOTOPIC ABUNDANC
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9.1.4. Data processing The required
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Protactinium-231 is stated to have
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TABLE I.1. METROLOGY REACTIONS (ACT
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For the convenience of the metrolog
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Such approximate calculations are n
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of each neutron energy group. These
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activation rate for the product iso
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III.1. THERMAL CROSS-SECTIONS Gener
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TABLE III.1. COMPARISON OF THERMAL
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TABLE III.2. COMPARISON OF CAPTURE
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TABLE III.3. COMPARISON OF THE Q 0
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— 50 Cr: Compared with the Mughab
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REFERENCES TO APPENDIX III [III.1]
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CONTRIBUTORS TO DRAFTING AND REVIEW
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CONTRIBUTORS TO DRAFTING AND REVIEW
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