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Detail analysis of fusion neutronics benchmark experiment for iron
Fusion Engineering and Design 84 (2009) 1095–1098
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Detail analysis of fusion neutronics benchmark experiment for iron Chikara Konno a,∗ , Kentaro Ochiai a , Masayuki Wada b , Kosuke Takakura a , Satoshi Sato a a b
Fusion Research and Development Directorate, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195, Japan Japan Computer System, Mito 310-0805, Japan
a r t i c l e
i n f o
Article history: Available online 14 February 2009 Keywords: Iron Benchmark FENDL-2.1 ENDF/B-VII.0 JEFF-3.1 JENDL-3.3
a b s t r a c t In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at Japan Atomic Energy Agency (JAEA)/Fusion Neutronics Source (FNS) was analyzed in detail with MCNP-4C and the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It was noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. Through the DORT calculations based on the iron data in ENDF/B-VII.0, it was found out that the first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 caused the overestimation. © 2009 Elsevier B.V. All rights reserved.
1. Introduction The nuclear data library FENDL-2.1 [1] was released for fusion reactor nuclear analyses in 2002, which was a collection of the best evaluation for fusion applications from the world-wide nuclear data libraries, ENDF/B-VI.8 [2], JEFF-3.0 [3], Brond-2.1 [4] and JENDL-3.2 [5], -3.3 [6], Fusion File [7]. It has given good results for nuclear analyses in ITER. Recently new nuclear data JEFF-3.1 (2005 May) [8] and ENDF/B-BVII.0 (2006 December) [9] were released and a new selection for the next version of FENDL [10] will be studied for more precise nuclear analyses. Iron is one of the most important elements for radiation shielding in fusion reactors and nuclear data on iron are required to be more accurate for thicker shield in future fusion reactors such as DEMO. In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron [11] at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA) is analyzed in detail in this paper. 2. Overview of fusion neutronics benchmark experiment on iron at JAEA/FNS Fig. 1 shows the experimental configuration of the fusion neutronics benchmark experiment on iron at JAEA/FNS. A large cylindrical iron assembly of 0.95 m in thickness and 1 m in diameter was placed 200 mm from the DT neutron source. Neutron spectra of almost the whole neutron energy, reaction rates for various reac-
∗ Corresponding author. Tel.: +81 29 282 6859; fax: +81 29 282 5709. E-mail address: [email protected] (C. Konno). 0920-3796/$ – see front matter © 2009 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2009.01.014
tions, gamma-heating rates and so on were measured at several points inside the experimental assembly. 3. Monte Carlo analysis The Monte Carlo code MCNP-4C [12] was used for the analysis of the experiment. Nuclear data libraries of JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0 were adopted. The ACE files of JENDL3.3, FENDL-2.1 and JEFF-3.1 were FSXLIB-J33 [13], FENDL/MC-2.1 [1], MCJEFF3.1 [14], respectively, while the ACE file of ENDF/B-VII.0 was processed with NJOY99.161 [15] by us. The experimental configuration was modeled in detail. Fig. 2 shows measured and calculated neutron spectra at the depth of 310 mm in the iron assembly. It seems that all the calculation results agree with the measured data well in general, but the calculation result with JENDL-3.3 clearly overestimates the measured neutron flux below a few keV. Ratios of calculation to experiment (C/E) for neutron flux of specified energy ranges, reaction rates and gamma heating are shown in Fig. 3 for more detailed comparison between the experimental and calculation results. The calculation result with ENDF/B-VII.0 underestimates the measured neutron flux above 10 MeV most among all the four calculation results. This is due to the angular distribution of the elastic scattering, where the forward part is underestimated in 56 Fe of ENDF/B-VII.0 as shown in Ref. [16]. The calculation results with ENDF/B-VII.0 and JEFF-3.1 agree with the measured reaction rate of the 115 In(n,n )115m In reaction, which is sensitive to neutrons above ∼0.3 MeV, very well, while the calculations with FENDL-2.1 and JENDL-3.3 tend to overestimate and underestimate the measurement with the depth, respectively. All the calculation results agree with the measured neutron flux from 0.1 to 1 MeV within 10%,
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Fig. 1. Experimental setup for iron benchmark experiments.
where the calculation result with ENDF/B-VII.0 seems to be the best. The C/Es for the neutron flux below100 keV are more scattered. As pointed out in Fig. 2, the calculation result with JENDL-3.3 overestimates the measured neutron flux below 1 keV more than those with the other libraries. The calculation results with FENDL-2.1 and
Fig. 2. Neutron spectra at depth of 310 mm inside iron slab.
JEFF-3.1 tend to underestimate the measured neutron flux below 100 keV and the measured reaction rate of the 197 Au(n,␥)198 Au reaction, which is sensitive to low energy neutrons, at the front region of the iron slab. As for the gamma-ray heating rate, the calculation with JENDL-3.3 tends to overestimate the measured data at the rear
Fig. 3. C/E for (a) neutron flux above 10 MeV, (b) neutron flux from 0.1 to 1 MeV, (c) neutron flux from 10 to 100 keV, (d) neutron flux from 0.1 to 1 keV, (e) neutron flux from 10 to 100 eV, (f) neutron flux from 1 to 10 eV, (g) reaction rate of 115 In(n,n )115m In, (h) reaction rate of 197 Au(n,␥)198 Au and (i) gamma-heating rate.
C. Konno et al. / Fusion Engineering and Design 84 (2009) 1095–1098
Fig. 4. Calculated neutron spectra at depth of 310 mm inside iron slab, where the iron isotopes in JENDL-3.3 were replaced with those in ENDF/B-VII.0 one by one.
Fig. 5. C/E for neutron flux from 0.1 to 1 keV, where the iron isotopes in JENDL-3.3 were replaced with those in ENDF/B-VII.0 one by one.
Fig. 6. Elastic cross-section data of 57 Fe in JENDL-3.3 and ENDF/B-VII.0.
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Fig. 7. The first inelastic cross-section data of 57 Fe in JENDL-3.3 and ENDF/B-VII.0.
Fig. 8. Calculated neutron spectra at depth of 310 mm inside iron slab, where the elastic or first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were replaced with those in ENDF/B-VII.0 separately.
Fig. 9. C/E for neutron flux from 0.1 to 1 keV, where the elastic or first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were replaced with those in ENDF/B-VII.0 separately.
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5. Summary In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at JAEA/FNS was analyzed in detail with the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. The iron data of ENDF/B-VII.0 will be recommended to those of the next FENDL, if the angular distribution of the elastic scattering of 56 Fe is improved. It is noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. We also investigated what of the iron data in JENDL-3.3 caused the overestimation of the measured neutrons below a few keV in the iron experiment through the DORT calculations based on the iron data in ENDF/B-VII.0. It is found out that the first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 caused the overestimation. The 57 Fe data should be revised in JENDL-4. Fig. 10. C/E of gamma-heating rate, where the elastic or first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were replaced with those in ENDF/B-VII.0 separately.
region of the iron slab, while the calculation results with the other libraries generally agree with the measured data well. It is concluded that totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It is noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. 4. Problem in iron data of JENDL-3.3 We investigated what of the iron data in JENDL-3.3 caused the overestimation of the neutron flux below a few keV based on ENDF/B-VII.0. The Sn code DORT [17] was used for this analysis because this code gave almost the same results as MCNP but with shorter computer time. Multigroup libraries of neutron 175-group structure with self-shielding correction were produced with the TRANSX code [18] from MATXS files. The MATXS files supplied from JAEA Nuclear Data Center were adopted for JENDL-3.3 [13]. Since MATXS files for ENDF/B-VII.0 were not released officially, they were produced with the NJOY99.161 code. First in order to examine which iron isotope caused the overestimation, we calculated neutron spectra of the iron experiment with DORT, where the iron isotopes in JENDL-3.3 were replaced with those in ENDF/B-VII.0 one by one. Figs. 4 and 5 show the result. These figures indicate that the 57 Fe data in JENDL-3.3 is the major source of the overestimation of the measured neutrons below a few keV. Next we compared the 57 Fe data in JENDL-3.3 with those in ENDF/B-VII.0 in every reaction. The elastic and first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were largely different from those in ENDF/B-VII.0 as shown in Figs. 6 and 7. In order to investigate which of the two reactions caused the overestimation, we calculated neutron spectra of the iron experiment with DORT, where the elastic or first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were replaced with those in ENDF/B-VII.0 separately. Figs. 8 and 9 show the result. It is concluded that the first inelastic scattering cross-section of 57 Fe in JENDL-3.3 caused the overestimation of the measured neutrons below a few keV in the iron experiment. Fig. 10 indicates that the first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 also caused the overestimation of the gamma-heating rate at the rear region of the iron slab.
References [1] D. Lopez Al-dama, A. Trokov, FENDL-2.1 update of an evaluated nuclear data library for fusion applications, IAEA report INDC (NDS)-467, 2004. [2] P.F. Rose (Ed.), ENDF-201, ENDF/B-VI Summary Documentation, 4th Edition, 1991, BNL-NCS-17541. [3] The JEFF-3.0 Nuclear Data Library, JEFF Report 19, OECD Nuclear Energy Agency, 2005. [4] A.I. Blokhin, B.I. Fursov, A.V. Ignatyuk, V.N. Koshcheev, E.V. Kulikov, B.D. Kuzminov, V.N. Manokhin, M.N. Nikolaev, Current status of Russian evaluated neutron data libraries, in: Proceedings of the International Conference on Nuclear Data for Science and Technology, vol. 2, Gatlinburg, Tennessee, USA, 9–13 May, 1994, p. p. 695. [5] T. Nakagawa, K. Shibata, S. Chiba, T. Fukahori, Y. Nakajima, Y. Kikuchi, T. Kawano, Y. Kanda, T. Ohsawa, H. Matsunobu, M. Kawai, A. Zukeran, T. Watanabe, S. Igarasi, K. Kosako, T. Asami, Japanese Evaluated Nuclear Data Library Version 3 Revision2: JENDL-3.2, J. Nucl. Sci. Technol. 32 (1995) 1259–1271. [6] K. Shibata, T. Kawano, T. Nakagawa, O. Iwamoto, J. Katakura, T. Fukahori, S. Chiba, A. Hasegawa, T. Murata, H. Matsunobu, T. Ohsawa, Y. Nakajima, T. Yoshida, A. Zukeran, M. Kawai, M. Baba, M. Ishikawa, T. Asami, T. Watanabe, Y. Watanabe, M. Igashira, N. Yamamuro, H. Kitazawa, N. Yamano, H. Takano, Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3, J. Nucl. Sci. Technol. 39 (2002) 1125–1136. [7] S. Chiba, T. Fukahori, K. Shibata, B. Yu, K. Kosako, N. Yamamuro, JENDL Fusion File 99, J. Nucl. Sci. Technol. 39 (2002) 187–194. [8] A. Koning, R. Forrest, M. Kellett, R. Mills, H. Henriksson, Y. Rugama (Eds.), The JEFF-3.1 Nuclear Data Library, JEFF report 21, OECD Nuclear Energy Agency, 2006. [9] M.B. Chadwick, P. Oblozinsky, M. Herman, N.M. Greene, R.D. McKnight, D.L. Smith, P.G. Young, R.E. MacFarlane, G.M. Hale, S.C. Frankle, A.C. Kahler, T. Kawano, R.C. Little, D.G. Madland, P. Moller, R.D. Mosteller, P.R. Page, P. Talou, H. Trellue, M.C. White, W.B. Wilson, R. Arcilla, C.L. Dunford, S.F. Mughabghab, B. Pritychenko, D. Rochman, A.A. Sonzogni, C.R. Lubitz, T.H. Trumbull, J.P. Weinman, D.A. Brown, D.E. Cullen, D.P. Heinrichs, D.P. McNabb, H. Derrien, M.E. Dunn, N.M. Larson, L.C. Leal, A.D. Carlson, R.C. Block, J.B. Briggs, E.T. Cheng, H.C. Huria, M.L. Zerkle, K.S. Kozier, A. Courcelle, V. Pronyaev, S.C. van der Marck, ENDF/BVII.0: next generation evaluated nuclear data library for nuclear science and technology, Nucl. Data Sheets 107 (2006) 2931–3059. [10] http://www-nds.iaea.org/fendl3/. [11] F. Maekawa, C. Konno, Y. Kasugai, Y. Oyama, Y. Ikeda, Data collection of fusion neutronics benchmark experiment conducted at FNS/JAEA, JAERI-Data/Code 98-021, 1998. [12] J.F. Briesmeister (Ed.), MCNP—A General Monte Carlo N-particle Transport Code, Version 4C, LA-13709-M, 2000. [13] K. Kosako, N. Yamano, T. Fukahori, K. Shibata, A. Hasegawa, The libraries FSXLIB and MATXSLIB based on JENDL-3.3, JAERI-Data/Code 2003-011, 2003. [14] OECD NEA Data Bank, Processing of the JEFF-3.1 Cross-section Library into a Continuous Energy Monte Carlo Radiation Transport and Criticality Data Library, vol. 18, NEA/NSC/DOC, 2006. [15] R.E. MacFarlane, D.W. Muir, The NJOY Nuclear Data Processing System, Version 91, LA-12740-M, 1994. [16] C. Konno, F. Maekawa, M. Wada, K. Kosako, Fusion Technol. 34 (1998) 1013– 1017. [17] DOORS3.2a: One, Two- and Three-dimensional Discrete Ordinates Neutron/Photon Transport Code System, RSICC CODE PACKAGE CCC-650, 2007. [18] R.E. MacFarlane, TRANSX 2: A Code for Interfacing MATXS Cross-section Libraries to Nuclear Transport Codes, LA-12312-MS, 1993.
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Detail analysis of fusion neutronics benchmark experiment for iron Chikara Konno a,∗ , Kentaro Ochiai a , Masayuki Wada b , Kosuke Takakura a , Satoshi Sato a a b
Fusion Research and Development Directorate, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195, Japan Japan Computer System, Mito 310-0805, Japan
a r t i c l e
i n f o
Article history: Available online 14 February 2009 Keywords: Iron Benchmark FENDL-2.1 ENDF/B-VII.0 JEFF-3.1 JENDL-3.3
a b s t r a c t In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at Japan Atomic Energy Agency (JAEA)/Fusion Neutronics Source (FNS) was analyzed in detail with MCNP-4C and the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It was noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. Through the DORT calculations based on the iron data in ENDF/B-VII.0, it was found out that the first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 caused the overestimation. © 2009 Elsevier B.V. All rights reserved.
1. Introduction The nuclear data library FENDL-2.1 [1] was released for fusion reactor nuclear analyses in 2002, which was a collection of the best evaluation for fusion applications from the world-wide nuclear data libraries, ENDF/B-VI.8 [2], JEFF-3.0 [3], Brond-2.1 [4] and JENDL-3.2 [5], -3.3 [6], Fusion File [7]. It has given good results for nuclear analyses in ITER. Recently new nuclear data JEFF-3.1 (2005 May) [8] and ENDF/B-BVII.0 (2006 December) [9] were released and a new selection for the next version of FENDL [10] will be studied for more precise nuclear analyses. Iron is one of the most important elements for radiation shielding in fusion reactors and nuclear data on iron are required to be more accurate for thicker shield in future fusion reactors such as DEMO. In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron [11] at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA) is analyzed in detail in this paper. 2. Overview of fusion neutronics benchmark experiment on iron at JAEA/FNS Fig. 1 shows the experimental configuration of the fusion neutronics benchmark experiment on iron at JAEA/FNS. A large cylindrical iron assembly of 0.95 m in thickness and 1 m in diameter was placed 200 mm from the DT neutron source. Neutron spectra of almost the whole neutron energy, reaction rates for various reac-
∗ Corresponding author. Tel.: +81 29 282 6859; fax: +81 29 282 5709. E-mail address: [email protected] (C. Konno). 0920-3796/$ – see front matter © 2009 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2009.01.014
tions, gamma-heating rates and so on were measured at several points inside the experimental assembly. 3. Monte Carlo analysis The Monte Carlo code MCNP-4C [12] was used for the analysis of the experiment. Nuclear data libraries of JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0 were adopted. The ACE files of JENDL3.3, FENDL-2.1 and JEFF-3.1 were FSXLIB-J33 [13], FENDL/MC-2.1 [1], MCJEFF3.1 [14], respectively, while the ACE file of ENDF/B-VII.0 was processed with NJOY99.161 [15] by us. The experimental configuration was modeled in detail. Fig. 2 shows measured and calculated neutron spectra at the depth of 310 mm in the iron assembly. It seems that all the calculation results agree with the measured data well in general, but the calculation result with JENDL-3.3 clearly overestimates the measured neutron flux below a few keV. Ratios of calculation to experiment (C/E) for neutron flux of specified energy ranges, reaction rates and gamma heating are shown in Fig. 3 for more detailed comparison between the experimental and calculation results. The calculation result with ENDF/B-VII.0 underestimates the measured neutron flux above 10 MeV most among all the four calculation results. This is due to the angular distribution of the elastic scattering, where the forward part is underestimated in 56 Fe of ENDF/B-VII.0 as shown in Ref. [16]. The calculation results with ENDF/B-VII.0 and JEFF-3.1 agree with the measured reaction rate of the 115 In(n,n )115m In reaction, which is sensitive to neutrons above ∼0.3 MeV, very well, while the calculations with FENDL-2.1 and JENDL-3.3 tend to overestimate and underestimate the measurement with the depth, respectively. All the calculation results agree with the measured neutron flux from 0.1 to 1 MeV within 10%,
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Fig. 1. Experimental setup for iron benchmark experiments.
where the calculation result with ENDF/B-VII.0 seems to be the best. The C/Es for the neutron flux below100 keV are more scattered. As pointed out in Fig. 2, the calculation result with JENDL-3.3 overestimates the measured neutron flux below 1 keV more than those with the other libraries. The calculation results with FENDL-2.1 and
Fig. 2. Neutron spectra at depth of 310 mm inside iron slab.
JEFF-3.1 tend to underestimate the measured neutron flux below 100 keV and the measured reaction rate of the 197 Au(n,␥)198 Au reaction, which is sensitive to low energy neutrons, at the front region of the iron slab. As for the gamma-ray heating rate, the calculation with JENDL-3.3 tends to overestimate the measured data at the rear
Fig. 3. C/E for (a) neutron flux above 10 MeV, (b) neutron flux from 0.1 to 1 MeV, (c) neutron flux from 10 to 100 keV, (d) neutron flux from 0.1 to 1 keV, (e) neutron flux from 10 to 100 eV, (f) neutron flux from 1 to 10 eV, (g) reaction rate of 115 In(n,n )115m In, (h) reaction rate of 197 Au(n,␥)198 Au and (i) gamma-heating rate.
C. Konno et al. / Fusion Engineering and Design 84 (2009) 1095–1098
Fig. 4. Calculated neutron spectra at depth of 310 mm inside iron slab, where the iron isotopes in JENDL-3.3 were replaced with those in ENDF/B-VII.0 one by one.
Fig. 5. C/E for neutron flux from 0.1 to 1 keV, where the iron isotopes in JENDL-3.3 were replaced with those in ENDF/B-VII.0 one by one.
Fig. 6. Elastic cross-section data of 57 Fe in JENDL-3.3 and ENDF/B-VII.0.
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Fig. 7. The first inelastic cross-section data of 57 Fe in JENDL-3.3 and ENDF/B-VII.0.
Fig. 8. Calculated neutron spectra at depth of 310 mm inside iron slab, where the elastic or first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were replaced with those in ENDF/B-VII.0 separately.
Fig. 9. C/E for neutron flux from 0.1 to 1 keV, where the elastic or first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were replaced with those in ENDF/B-VII.0 separately.
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C. Konno et al. / Fusion Engineering and Design 84 (2009) 1095–1098
5. Summary In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at JAEA/FNS was analyzed in detail with the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. The iron data of ENDF/B-VII.0 will be recommended to those of the next FENDL, if the angular distribution of the elastic scattering of 56 Fe is improved. It is noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. We also investigated what of the iron data in JENDL-3.3 caused the overestimation of the measured neutrons below a few keV in the iron experiment through the DORT calculations based on the iron data in ENDF/B-VII.0. It is found out that the first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 caused the overestimation. The 57 Fe data should be revised in JENDL-4. Fig. 10. C/E of gamma-heating rate, where the elastic or first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were replaced with those in ENDF/B-VII.0 separately.
region of the iron slab, while the calculation results with the other libraries generally agree with the measured data well. It is concluded that totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It is noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. 4. Problem in iron data of JENDL-3.3 We investigated what of the iron data in JENDL-3.3 caused the overestimation of the neutron flux below a few keV based on ENDF/B-VII.0. The Sn code DORT [17] was used for this analysis because this code gave almost the same results as MCNP but with shorter computer time. Multigroup libraries of neutron 175-group structure with self-shielding correction were produced with the TRANSX code [18] from MATXS files. The MATXS files supplied from JAEA Nuclear Data Center were adopted for JENDL-3.3 [13]. Since MATXS files for ENDF/B-VII.0 were not released officially, they were produced with the NJOY99.161 code. First in order to examine which iron isotope caused the overestimation, we calculated neutron spectra of the iron experiment with DORT, where the iron isotopes in JENDL-3.3 were replaced with those in ENDF/B-VII.0 one by one. Figs. 4 and 5 show the result. These figures indicate that the 57 Fe data in JENDL-3.3 is the major source of the overestimation of the measured neutrons below a few keV. Next we compared the 57 Fe data in JENDL-3.3 with those in ENDF/B-VII.0 in every reaction. The elastic and first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were largely different from those in ENDF/B-VII.0 as shown in Figs. 6 and 7. In order to investigate which of the two reactions caused the overestimation, we calculated neutron spectra of the iron experiment with DORT, where the elastic or first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 were replaced with those in ENDF/B-VII.0 separately. Figs. 8 and 9 show the result. It is concluded that the first inelastic scattering cross-section of 57 Fe in JENDL-3.3 caused the overestimation of the measured neutrons below a few keV in the iron experiment. Fig. 10 indicates that the first inelastic scattering cross-section data of 57 Fe in JENDL-3.3 also caused the overestimation of the gamma-heating rate at the rear region of the iron slab.
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