International Reactor Dosimetry File 2002 - IAEA Publications

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TABLE 3.1. REACTIONS FROM THE VARIOUS LIBRARIES WITH CROSS-SECTIONS SUITABLE FOR IRDF-<strong>2002</strong> (cont.) Reactions from RRDF-98 F192 TI462 TI46P TI47NP TI48NP TI48P TI49NP V51A FE542 FE54A FE56P CO59A CU63A AS752 NB932 a NB93N a LA139G PR1412 W186G PB204N a AL27P d NI58P d RH103N a,d IN115N a,d NP237F 25 crosssections Reactions from ENDF/B-VI Release 8 CR522 NI58P NI60P CU632 CU63G CU652 NA23G b NB93G b IN115G a,b TH232G b 10 crosssections Reactions from JEFF-3.0 FE56P NI582 NI58P NI60P 4 crosssections Σ = 117 cross-sections Problematic reactions NA23G b TI0XSC47 c CR50G MN552 FE57NP NB93G b IN115G a,b EU151G TA181G b TH232G b 10 crosssections a b c d Metastable state of the reaction product nuclide. Diagonal covariance matrix. TI0XSC-46, -47 and -48 indicate the reactions on a natural titanium target leading to the products 46 Ti, 47 Ti and 48 Ti, respectively. New evaluations or updates, 2003. Notes: (a) SAND type short reaction notation: chemical symbol and mass number of the target nucleus are followed by the name of the reaction product; A, G, F, 2, N, P, NP and T represent (n,α), (n,γ), (n,f), (n,2n), (n,n′), (n,p), (n,np) and (n,t) reactions, respectively. (b) No suitable cross-section data have been found in the libraries investigated for reactions nat Ti(n,x) 47 Sc, 50 Cr(n,γ) 51 Cr, 55 Mn(n,2n) 54 Mn, 57 Fe(n,np) 56 Mn and 151 Eu(n,γ) 152 Eu. (c) Only diagonal covariance matrices were found for the following reactions: 23 Na(n,γ) 24 Na, 93 Nb(n,γ) 94 Nb, 115 In(n,γ) 116 In m , 181 Ta(n,γ) 182 Ta and 232 Th(n,γ) 233 Th (below 15 eV). 17
eaction by the Japanese evaluators [3.12] so as to substitute for earlier erroneous data in JENDL/D-99. 3.2.1.1. Selection procedure The procedure for selecting the cross-sections for IRDF-<strong>2002</strong> was based on comparisons with each other of the integral values of the cross-sections and related uncertainty information from the libraries of interest, and with experimental data obtained in standard neutron fields. This implies that the experimental data in a Maxwellian thermal neutron spectrum and in a 1/E neutron field had to be considered [3.2]. A review of the literature identified two sources that were adequate for the purpose: Mughabghab [3.13] and Holden [3.14]. The thermal neutron cross-sections in both experimental evaluations refer to a neutron energy of 0.0253 eV (v 0 = 2200 m/s), while the resonance integrals were calculated by Mughabghab with a lower energy limit of 0.5 eV and an upper energy limit corresponding to the upper resonance with known scattering width [3.15]. Holden calculated the resonance integrals from 0.5 eV to 0.1 MeV. The recommended cross-section values refer to room temperature in both cases (293.43 K and 300 K for Holden and Mughabghab, respectively). Thermal neutron cross-sections in this work refer to 0.0253 eV neutron energy (v 0 = 2200 m/s), while the resonance integrals were calculated from 0.5 eV to 1.05 MeV (preliminary analyses are found in Refs [3.7, 3.8]). In the comparison of the corresponding cross-section data, 0 K was used, while in characterizing the selected data, 300 K neutron temperature was used. The difference between the corresponding data at the two neutron temperatures was less than 1%, including the uncertainty deriving from the data processing (see below). A similar comparison of the thermal neutron cross-sections and resonance integrals with the corresponding Mughabghab data [3.13] was made for a series of capture cross-sections from different dosimetry libraries by Trkov (Appendix III); these results agree with those presented in this report. The uncertainty information for the cross-sections of interest is represented by their relative standard deviation values (calculated in a three energy group structure as described in Section 3.1), weighted with a typical MTR spectrum (Fig. 3.1). 3.2.1.2. Results The results of the cross-section comparison are listed in Table 3.2. As shown for some of these reactions, the same cross-section data are given in both 18
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 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
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 78 and 79: TABLE 6.2. CROSS-SECTIONS IN IRDF-2
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TABLE 6.2. CROSS-SECTIONS IN IRDF-2
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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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