International Reactor Dosimetry File 2002 - IAEA Publications

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— 10 B(n,α) 7 Li and 6 Li(n,t) 4 He: The uncertainty of the library crosssections in the intermediate neutron energy region is too small (not realistic) compared with the corresponding C/E values (or the library data deviate significantly from the experimental values). — 23 Na(n,γ) 24 Na: The uncertainty information contains only a diagonal matrix — a new evaluation is required. — 55 Mn(n,γ) 56 Mn: The C/E value for the resonance integral deviates by 16% from unity — excessive when compared with the related uncertainty values; a new cross-section evaluation is needed in the intermediate neutron energy region. — 58 Fe(n,γ) 59 Fe: The C/E value for the resonance integral with the Mughabghab data deviates by 19% from unity. A large difference is also found between the recommended experimental data for the sources considered. Clarification of this discrepancy is necessary because this reaction is one of the most frequently used detectors in reactor dosimetry. A new cross-section evaluation in the intermediate neutron energy region should also be considered. — 93 Nb(n,γ) 94 Nb: The C/E value for the resonance integral deviates by 17% from unity; furthermore, the uncertainty information contains only a diagonal matrix — a new evaluation is required. — 109 Ag(n,γ) 110 Ag m : Mughabghab data represent the sum of the reaction cross-sections leading to 110 Ag m+g , while the evaluated data libraries contain only cross-section data for the 109 Ag(n,γ) 110 Ag m reaction. Therefore, no comparison with the Mughabghab data was possible. — 115 In(n,γ) 116 In m : In the present library the cross-section leading to the metastable states of the product nucleus 116 In is given; uncertainty information contains only a diagonal matrix — a new evaluation is required. — 181 Ta(n,γ) 182 Ta: The uncertainty information contains only a diagonal matrix — a new evaluation is required. — 197 Au(n,γ) 198 Au: The available uncertainty information for this reaction is not reliable; similar data have been withdrawn from ENDF/B-VI. Uncertainty data in IRDF-90 were derived from the same source — a new evaluation is required. — 232 Th(n,γ) 233 Th: The uncertainty information below 15 eV is defined only in terms of a diagonal matrix — a new evaluation is required. — 235 U(n,f): The uncertainty information has been declared to be unreliable, and has been withdrawn from ENDF/B-VI. Data in IRDF-90 have the same origin — a new evaluation is required. 23
— 241 Am(n,f): No up to date experimental data are available for this reaction, therefore the corresponding C/E values could not be derived. 3.3. CONCLUSIONS Based on the results of the cross-section selection procedure outlined above, the following principal conclusions can be drawn related to the data in Tables 3.3 and 3.4: (a) Very few new cross-section evaluations accompanied by complete uncertainty information have been undertaken in the low neutron energy region over the previous decade, except the 139 La(n,γ) 140 La and 186 W(n,γ) 187 W reactions evaluated for the RRDF [3.4]. (b) Integral values of the selected cross-sections in the thermal neutron region exhibit very good agreement in most cases with the corresponding recommended experimental values. (c) Resonance integrals of the 55 Mn(n,γ) 56 Mn, 58 Fe(n,γ) 59 Fe and 93 Nb(n,γ) 94 Nb reactions deviate significantly (>10%) from the corresponding experimental data. Further investigation (e.g. testing the data in benchmark neutron fields) and new cross-section evaluations are required. (d) For the 23 Na(n,γ) 24 Na, 93 Nb(n,γ) 94 Nb, 115 In(n,γ) 116 In m and 181 Ta(n,γ) 182 Ta reactions and the 232 Th(n,γ) 233 Th reaction below 15 eV, the uncertainty information is quantified in terms of diagonal covariance matrices only. New evaluations with complete covariance information are required. (e) Unreliable uncertainty data are present in all the investigated crosssection libraries for the 197 Au(n,γ) 198 Au and 235 U(n,f) reactions (withdrawn from ENDF/B-VI). New cross-section evaluations with complete covariance information are required. (f) Selected cross-sections in Table 3.3 should be subjected to a consistency test by comparing the relevant integral data with the corresponding experimental values in benchmark neutron fields. 24
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 14 and 15: 1. INTRODUCTION R. Paviotti-Corcuer
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 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
Page 68 and 69: — 238 U(n,γ) 239 U(β - ) 239 Np
Page 70 and 71: (i) (ii) (iii) (iv) (v) Only diagon
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
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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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International Reactor Dosimetry File 2002 - IAEA Publications

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