International Reactor Dosimetry File 2002 - IAEA Publications

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— 238 U(n,γ) 239 U(β – ) 239 Np: The IRDF provides the only evaluated data file, and fits the data reasonably well, apart from one high data point near 14.5 MeV. — 238 U(n,f): JENDL and the IRDF are nearly identical, and both sets of data are slightly lower than the available experimental data. — 239 Pu(n,f): JENDL gives a better fit to the data than the IRDF. — 241 Am(n,f): JENDL is the only available file, and gives a good fit to the average of the available experimental data. ACKNOWLEDGEMENT K.I. Zolotarev of the Institute of Physics and Power Engineering, Obninsk, Russian Federation, kindly made available his evaluations and renormalizations of the experimental data for many of the reactions, as noted in the text. Such evaluations provide the best basis for assessment of the various cross-section files, and were usually adopted and taken into account during the review process. REFERENCES TO SECTION 5 [5.1] KOCHEROV, N.P., McLAUGHLIN, P.K., The International Reactor Dosimetry File (IRDF-90), Rep. IAEA-NDS-141, Rev. 2, IAEA, Vienna (1993). [5.2] KOBAYASHI, K., et al., JENDL Dosimetry File (JENDL/D-99), Rep. JAERI 1344, Japan Atomic Energy Research Institute, Tokai-mura (2002). [5.3] ZOLOTAREV, K.I., IGNATYUK, A.V., MAHOKHIN, V.N., PASHCHENKO, A.B., RRDF-98, Russian Reactor Dosimetry File, Rep. IAEA-NDS-193, IAEA, Vienna (1999). [5.4] Joint European Fusion File 3.0 (JEFF-3.0); JEFF-3 is available on CD-ROM on request from the IAEA, Nuclear Data Section, http://www.nea.fr/html/dbdata/projects/nds_jef.htm; JACQMIN, R., et al., Status of the JEFF project, J. Nucl. Sci. Technol, Suppl. 2 (2002) 54–55. [5.5] ZERKIN, V.V., et al., “EXFOR-CINDA-ENDF: Migration of databases to give higher-quality nuclear data services”, Nuclear Data for Science and Technology (Proc. Int. Conf. Santa Fé, NM, 2004), American Institute of Physics, NY (2005) 586–589. 55
6. FINAL SELECTION OF CROSS-SECTIONS FOR IRDF-2002, AND CHARACTERIZATION OF THE SELECTED DATA E.M. Zsolnay, H.J. Nolthenius The final selection of cross-section data for IRDF-2002 was made during a technical meeting held at the IAEA in Vienna from 1 to 3 October 2003 [6.1]. This selection procedure was based on the following considerations: (a) (b) (c) Comparison of the integral values of the candidate cross-sections with the corresponding experimental results in the four standard neutron fields (thermal Maxwellian, 1/E slowing down, 252 Cf fission and 14 MeV neutron field) recommended for the purpose of cross-section selection [6.2]. Quality of the uncertainty information. Consistency of the data (C/E values compared with the corresponding uncertainty information). The cross-section and uncertainty data described in Sections 3, 4 and 5 were used. However, in addition, spectrum averaged cross-sections were calculated for the theoretical function of the Watt fission spectrum [6.3–6.6]. These data are presented in Table 6.1 for the candidate cross-sections of the fast neutron threshold reactions [6.6]. Such integral cross-section data are also published in standard neutron cross-section tables. The uncertainty information for the cross-sections is represented by the corresponding standard deviations above 1.05 MeV, weighted with a typical MTR spectrum [6.7]. The results in Table 6.1 show good agreement with those of Mannhart that were obtained in the 252 Cf fission spectrum (Section 4). The observed variations may be attributed to the differences between the two spectrum functions. Based on the results outlined in Sections 3, 4 and 5, and following the recommendations of Ref. [6.6], the final selection procedure resulted in the data files being chosen as summarized in Table 6.2. This table lists the crosssections included in IRDF-2002, together with their integral characteristics and the ratios of the corresponding calculated and experimental cross-section data (C/E). Uncertainties of the C/E values involve the standard deviations of both the calculated and experimental cross-sections. The following shortcomings occur in the chosen cross-section data: 56
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INTERNATIONAL REACTOR DOSIMETRY FIL
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TECHNICAL REPORTS SERIES No. 452 IN
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FOREWORD An accurate and complete k
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Contributing authors O. Bersillon C
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5.1. Plots of experimental data and
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1. INTRODUCTION R. Paviotti-Corcuer
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(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
Page 64 and 65: TABLE 5.1. CROSS-SECTION EVALUATION
Page 66 and 67: — 58 Ni(n,2n) 57 Ni: The IRDF, JE
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Page 72 and 73: TABLE 6.1. INTEGRAL CHARACTERISTICS
Page 74 and 75: TABLE 6.1. INTEGRAL CHARACTERISTICS
Page 76 and 77: TABLE 6.2. CROSS-SECTIONS IN IRDF-2
Page 84 and 85: 7. CONSISTENCY TEST OF THE CROSS-SE
Page 86 and 87: TABLE 7.1. REFERENCE NEUTRON FIELDS
Page 88 and 89: Fluence, (E) (MeV -1 ) 10 7 10 6 10
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Page 94 and 95: Table 7.3 details the 34 dosimetry
Page 96 and 97: 7.2. RESULTS OF CONSISTENCY TESTING
Page 98 and 99: TABLE 7.4. RATIO OF SPECTRUM AVERAG
Page 104 and 105: [7.7] SEKINE, T., MAEDA, S., AOYAMA
Page 106 and 107: ecommendations found in ASTM practi
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Page 110 and 111: 10 2 Damage (MeV· mb) 10 1 10 0 10
Page 112 and 113: 9. DECAY DATA AND ISOTOPIC ABUNDANC
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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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