International Reactor Dosimetry File 2002 - IAEA Publications

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TABLE 3.3. THERMAL NEUTRON CROSS-SECTIONS AND RESONANCE INTEGRALS FOR REACTIONS SELECTED FOR IRDF-<strong>2002</strong> (TEMPERATURE OF 300 K) Reaction Material No. MT No. Library (selected evaluation) Library cross-section (2200 m/s) σ L (m 2 ) Evaluated experimental data (2200 m/s) σ M (m 2 ) Evaluated experimental data (2200 m/s) σ H (m 2 ) Cross-section ratio Resonance integral from library data IR L (m 2 ) Evaluated resonance integral IR M (m 2 ) Evaluated resonance integral IR H (m 2 ) Resonance integral ratio σ L /σ M σ L /σ H IR L /IR M IR L /IR H 6 Li(n,t) 325 105 IRDF-90 9.42E-26 Not available 9.4(1)E-26 — 1.00 4.27E-26 Not available 4.22(4)E-26 — 1.01 10 B(n,α) 525 107 IRDF-90 3.84E-25 Not available 3.84(1)E-25 — 1.00 1.73E-25 Not available 1.73(1)E-25 — 1.00 23 Na(n,γ) a 1123 102 IRDF-90 5.29E-29 5.30(5)E-29 — 1.00 — 3.17E-29 3.11(10)E-29 3.0(6)E-29 1.02 1.06 45 Sc(n,γ) 2126 102 IRDF-90 2.73E-27 2.72(2)E-27 2.7E-27 1.00 1.01 1.20E-27 1.20(5)E-27 1.20E-27 1.00 1.00 55 Mn(n,γ) 2525 102 IRDF-90 1.34E-27 1.336(5)E-27 1.33(1)E-27 1.00 1.01 1.18E-27 1.40(3)E-27 1.40(3)E-27 0.84 0.84 58 Fe(n,γ) 2637 102 JENDL/D-99 (u) 1.30E-28 1.30(3)E-28 1.3(1)E-28 1.00 1.00 1.37E-28 1.7(1)E-28 1.3(2)E-28 0.81 1.05 59 Co(n,γ) 2725 102 IRDF-90 3.72E-27 3.718(6)E-27 3.72E-27 1.00 1.00 7.60E-27 7.59(2)E-27 7.4E-27 1.00 1.03 63 Cu(n,γ) 2925 102 IRDF-90 4.47E-28 4.52(2)E-28 4.5(2)E-28 0.99 0.99 4.96E-28 4.97(8)E-28 5.0(1)E-28 1.00 0.99 93 Nb(n,γ) a 4125 102 IRDF-90 1.16E-28 1.15(5)E-28 1.1E-28 1.01 1.05 9.91E-28 8.5(5)E-28 8.5E-28 1.17 1.17 109 Ag(n,γ) b 4731 102 IRDF-90 (u) 4.21E-28 — 4.2E-28 — 1.00 6.86E-27 — 7.0E-27 — 0.98 115 In(n,γ) a,b 4931 102 IRDF-90 1.67E-26 2.02(2)E-26 2.05E-26 — 1.04 2.59E-25 Not available 2.7E-25 — 0.96 139 La(n,γ) 5712 102 RRDF-98 (n) 9.04E-28 9.04(4)E-28 9.2(2)E-28 1.00 0.98 1.20E-27 1.21(6)E-27 1.2(1)E-27 0.99 1.00 181 Ta(n,γ) a 7328 102 JENDL/D-99 2.07E-27 2.05(5)E-27 2.01E-27 1.01 1.04 6.59E-26 6.60(23)E-26 6.504E-26 1.00 1.01 186 W(n,γ) 7452 102 RRDF-98 (n) 3.85E-27 3.85(5)E-27 3.7(2)E-27 1.00 1.04 4.80E-26 4.85(15)E-26 5.10(50)E-26 0.99 0.94 197 Au(n,γ) c 7925 102 IRDF-90 9.88E-27 9.865(90)E-27 9.87(10)E-27 1.00 1.00 1.57E-25 1.550(28)E-25 1.55(3)E-25 1.01 1.01 232 Th(n,γ) a 9040 102 IRDF-90 7.41E-28 7.35(3)E-28 7.37(4)E-28 1.01 1.00 8.56E-27 8.5(3)E-27 8.5(3)E-27 1.01 1.01 235 U(n,f) d 9228 18 IRDF-90 5.86E-26 Not available 5.86(2)E-26 — 1.01 2.72E-26 Not available 2.75(5)E-26 — 0.99 238 U(n,γ) 9237 102 IRDF-90 2.72E-28 2.680(19)E-28 2.7(1)E-28 1.01 1.00 2.77E-26 2.77(3)E-26 2.77(3)E-26 1.00 1.00 25
TABLE 3.3. THERMAL NEUTRON CROSS-SECTIONS AND RESONANCE INTEGRALS FOR REACTIONS SELECTED FOR IRDF-<strong>2002</strong> (TEMPERATURE OF 300 K) (cont.) Reaction Material No. MT No. Library (selected evaluation) Library cross-section (2200 m/s) σ L (m 2 ) Evaluated experimental data (2200 m/s) σ M (m 2 ) Evaluated experimental data (2200 m/s) σ H (m 2 ) Cross-section ratio Resonance integral from library data IR L (m 2 ) Evaluated resonance integral IR M (m 2 ) Evaluated resonance integral IR H (m 2 ) Resonance integral ratio σ L /σ M σ L /σ H IR L /IR M IR L /IR H 239 Pu(n,f) 9437 18 JENDL/D-99 7.47E-26 Not available 7.52(3)E-26 — 0.99 2.97E-26 Not available 3.0(1)E-26 — 0.99 241 Am(n,f) 9543 18 JENDL/D-99 3.03E-28 Not available 3.15(1)E-28 — 0.99 7.83E-28 Not available Not available — — b d a Diagonal matrix (for the 232 Th(n,γ) 233 Th reaction only below 15 eV). Metastable state of the product nuclide. c Uncertainty information for the 197 Au(n,γ) 198 Au reaction is not reliable; withdrawn from ENDF/B-VI (a similar old evaluation is also present in IRDF-90). Uncertainty information for the 235 U(n,f) reaction is not reliable; withdrawn from ENDF/B-VI (a similar old evaluation is also present in IRDF-90). (u): Updated data. (n): New data. Notes: Evaluated experimental cross-section data: values in brackets are ‘absolute uncertainties’ (one standard deviation). Subscript L signifies library data; subscripts M and H denote evaluated experimental data of Mughabghab [3.13] and Holden [3.14], respectively. 26
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 36 and 37: — 10 B(n,α) 7 Li and 6 Li(n,t) 4
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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