[7.7] SEKINE, T., MAEDA, S., AOYAMA, T., “Characterization of neutron field in the experimental fast reactor JOYO”, <strong>Reactor</strong> <strong>Dosimetry</strong> in the 21st Century (Proc. Int. Symp. Brussels, <strong>2002</strong>) (WAGEMANS, J., ABDERRAHIM, H.A., D’HONDT, P., DE RAEDT, C., Eds), World Scientific, Singapore (2003) 381– 388. [7.8] SHIBATA, K., “Average cross sections calculated in various neutron fields”, Summary Report of the Technical Meeting on <strong>International</strong> <strong>Reactor</strong> <strong>Dosimetry</strong> <strong>File</strong>: IRDF-<strong>2002</strong>, Rep. INDC(NDS)-435, <strong>IAEA</strong>, Vienna (<strong>2002</strong>) 49–58. [7.9] SHIMAKAWA, S., et al., “Neutron dosimetry for material irradiation tests in JMTR”, <strong>Reactor</strong> <strong>Dosimetry</strong> (Proc. Int. Symp. Prague, 1996) (ABDERRAHIM, H.A., D’HONDT, P., OSMERA, B., Eds), World Scientific, Singapore (1998) 857–864. [7.10] GRIFFIN, P.J., KELLY, J.G., VEHAR, D.W., Updated Neutron Spectrum Characterization of SNL Baseline <strong>Reactor</strong> Environments, Vol. 1: Characterization, Rep. SAND93-2554, Sandia Natl Lab. (1994). [7.11] KELLY, J.G., “Neutron spectrum adjustment with SAND-II using arbitrary trial functions”, <strong>Reactor</strong> <strong>Dosimetry</strong>: Methods, Applications, and Standardization, Rep. ASTM STP 1001, American Society for Testing and Materials, Philadelphia, PA (1989) 460–468. [7.12] STALLMAN, W., LSL-M2: A computer program for least-squares logarithmic adjustment of neutron spectra, Rep. NUREG/CR-4349, ORNL/TM-9933, Oak Ridge Natl Lab., TN (1986). [7.13] GRIFFIN, P.J., A rigorous treatment of self-shielding and covers in neutron spectra determination, IEEE Trans. Nucl. Sci. 42 (1995) 1878. [7.14] BRIESMEISTER, J., MCNP — A General Monte Carlo N-particle Transport Code, Version 4A, Rep. LA-12625-M, US 705 and US 706, Los Alamos Natl Lab., NM (1993). [7.15] McELROY, W.N., BERG, S., CROCKETT, T., HAWKINS, R., A Computerautomated Iterative Method for Neutron Flux Spectral Determination by Foil Activation, Rep. AFWL-TR-67-41, Vol. 1, Air Force Weapons Lab., Kirtland, NM (1967). 91
8. RADIATION DAMAGE FILES AND COMPUTER CODES P.J. Griffin, L.R. Greenwood Commonly used response functions can be usefully formatted so that they may be readily interfaced with neutron spectra. Therefore, the IRDF-<strong>2002</strong> library has included response functions for neutron displacement damage per atom (dpa) for iron, silicon and GaAs to support this application. The following sections detail the response functions and provide attribution for the derivation of the response. 8.1. IRON dpa (LIGHT WATER REACTOR PRESSURE VESSEL DAMAGE) The ASTM standard E693 is the source for the iron dpa response [8.1]. Iron dpa (Fig. 8.1) is used in applications supporting pressure vessel surveillance calculations, which are performed in compliance with the US Nuclear Regulatory Commission requirements. The standard incorporates the ENDF/ B-VI cross-sections in the iron dpa exposure function and recommends the use of the Norgett–Robinson–Torrens (NRT) displacement formalism. This ‘damage energy to displacement’ conversion procedure is consistent with the 10 4 10 3 Iron dpa cross-section 10 2 10 1 10 0 10 –1 10 –2 10 –10 10 –9 10 –8 10 –7 10 –6 10 –5 10 –4 10 –3 10 –2 10 –1 10 0 10 1 Neutron energy (MeV) FIG. 8.1. ENDF/B-VI based iron displacement cross-section. 92
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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
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REFERENCES TO SECTION 1 [1.1] ZIJP,
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cross-sections, and as a consequenc
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TABLE 2.1. MEASURED AND CALCULATED
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TABLE 2.2. MEASURED AND CALCULATED
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[2.24] KOBAYASHI, K., KIMURA, I., M
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Fluence rate per unit energy (m -2
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— 10 B(n,α) 7 Li and 6 Li(n,t) 4
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