Review of nuclear data of major actinides and 56 Fe in JENDL-4.0

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1 Review of nuclear data of major actinides and 56 Fe in JENDL-4.0 Osamu Iwamoto, Nobuyuki Iwamoto Nuclear Data Center, Nuclear Science and Engineering Directorate Japan Atomic Energy Agency Ibaraki, Japan Abstract JENDL-4.0 was released in Most of the actinide data were revised from JENDL-3.3, especially at the fast energy region, using the newly developed CCONE code. Fission cross-sections of major actinides were determined using the least-squares method and the SOK code, taking available experimental data into account. In the resonance region around 1 kev for 235 U, capture cross-sections were reduced to improve integral benchmark tests. The reduction was confirmed by Jandel, et al. during a later experiment at Los Alamos. For 56 Fe, cross-sections of inelastic scattering and angular distributions of elastic scattering were revised from JENDL-3.3. Evaluation methods and comparisons with experimental data and other evaluated data will be reviewed. Introduction JENDL-4.0 was released in 2010 (Shibata, et al., 2011). The revision from JENDL-3.3 (Shibata, et al., 2002) was mainly focused on fission products and minor actinides. It is intended to improve reliability for applications such as innovative reactors, high burn-up use and MOX fuel reactors. To increase the reliability of evaluation for MA and FP, two theoretical nuclear reaction model codes, CCONE (Iwamoto, 2007)) and POD (Ichihara, et al., 2007) had been developed. All of the actinide data were reviewed and most were revised, based on available experimental data. Other significant progress of JENDL-4.0 was an increase of covariance data which were evaluated for all cross-sections, angular distributions and fission neutron spectra of all actinides. Major actinide Resonance region JENDL-4.0 adopted the resolved resonance parameters of SAMMY analyses carried out by Leal, et al. (1999), Derrien, et al. (2009) and Derrien, et al. (2007) for 235 U, 238 U and 239 Pu, respectively. The parameters of 235 U were also adopted by all of the recent nuclear data libraries of JENDL-3.3, ENDF/B-VII.1 (Chadwick, et al., 2011), JEFF-3.1 (Koning, et al., 2006), CENDL-3.1 (Ge, et al., 2011) and ROSFOND-2010 (ABBN, n.d.). However, in JENDL-4.0 the upper limit of the original resolved resonance region 2.25 kev was lowered to 500 ev and modified point-wise cross-sections were given between 500 ev to 2.25 kev to recover the worsened integral benchmark results for some uranium-fuelled fast reactors, which would be arisen by changing the resonance parameter of JENDL-3.3 from JENDL-3.2. According NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD

2 to the sensitivity analyses of the integral test, the capture cross-section would be overestimated around 1 kev and a gap of evaluated cross-section appearing at 2.25 kev in JENDL-3.3 should be decreased. The data recently measured by Jandel, et al. (2012) support JENDL-4.0 in that energy region. Evaluations of JENDL-4.0 and JENDL-3.3 are compared with experimental data in Figure 1. Figure 1: 235 U neutron capture cross-section JENDL-4.0 and JENDL-3.0 are shown by group-wise cross-section below 2.25 kev Fast energy region Fission cross-section The fission cross-sections of major actinides were evaluated by the least-squares method using measured cross-sections and ratio data simultaneously using the SOK code (Kawano, et al., 2000). The carefully checked 124 experimental data sets were included in the analysis. As well as the best-estimated cross-section, covariance data were deduced at the same time. By taking account of scatter of the experimental data, the obtained standard deviation was multiplied by a factor of 2. The 235 U fission cross-section ratios of various evaluations to JENDL-4.0 are shown in Figure 2. Above 0.1 MeV, the evaluated result of 235 U cross-sections agrees with the other evaluations. However, below 0.1 MeV, deviations from JENDL-4.0 become larger than the uncertainty. Evaluation by CCONE code With the exception of fission cross-sections, most of the data in the fast neutron region were evaluated by theoretical model calculation with the CCONE code. It had been developed for the JENDL-4.0 by integrating several nuclear reaction models such as coupled-channel optical model, distorted wave born approximation, pre-equilibrium two-component exciton model and the Hauser-Feshbach-Moldauer statistical model. The model parameters were determined to fit the experimental data, and their uncertainties were also deduced to evaluated covariance data by the least-squares method using the KALMAN code system (Kawano and Shibata, 1997) with available measured data. The covariance data, for which experimental data were scarce, were also evaluated using the parameter uncertainties. Figure 3 compares ratios of inelastic scattering cross-sections to JENDL-4.0 for 238 U and 239 Pu. For 238 U the cross-sections agree among evaluations within the uncertainty except for the lower and upper edge region. The 239 Pu cross-sections were in poor agreement especially with JEFF (=RUSFOND-2010) and CENDL-3.1 evaluations. 76 NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD 2014

3 Figure 2: Ratio of 235 U fission cross-section of various evaluations to JENDL-4.0 Evaluated uncertainty of JENDL-4.0 is shown by thick lines Figure 3: Ratio of inelastic scattering cross-section of various evaluations to JENDL-4.0 for 238 U (left) and 239 Pu (right) Evaluated uncertainty of JENDL-4.0 was shown by thick lines Prompt fission neutron spectra The prompt fission neutron spectra were calculated based on the Madland-Nix model. Below the second chance fission threshold, the method combined with multimodal random-neck rupture of the fission process was applied to JENDL-3.3 evaluation of 235 U and 239 Pu (Ohsawa, 2001). Evaluation of 238 U was performed for JENDL-3.2. JENDL-4.0 adopted those values below the neutron incident energy around 5 MeV. Above the threshold energy of the second chance fission, the fission spectra were calculated incorporating the pre-equilibrium statistical model calculations by CCONE code. The pre-scission neutron spectra and average excitation energies leading to fission were deduced from the calculations done for the cross-section evaluation. The average excitation energies were adapted to the fission spectra systematics (Iwamoto, 2008) to estimate post-scission neutron spectra. The fission spectrum for 238 U at 7 MeV is shown in Figure 4. JENDL-4.0 agrees well with the experimental data. The bump around the neutron emission energy at 0.8 MeV could be due to the pre-scission neutrons that are taken into account in JENDL-4.0. NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD

4 Figure 4: Prompt fission neutron spectrum for 238 U at neutron incident energy 7 MeV The spectra are shown by ratios to Maxwellian at T = MeV Iron-56 Total and elastic scattering cross-sections A small part of nuclear data on 56 Fe was modified from JENDL-3.3. The neutron energy range for 56 Fe data in JENDL-4.0 is from 10 5 ev to 20 MeV, while ENDF/B-VII.1 and JEFF (OECD/NEA, 2012) cover much wider energy ranges up to 150 and 200 MeV, respectively. In the resolved resonance region the Reich-Moore formula was used with resolved resonance parameters which are given up to 850 kev. The parameters were taken from Froehner s evaluation in JEF-2. The detailed measurement of total cross-section for natural Fe shows resonance-like structure above 850 kev. JENDL-3.3 (and also JENDL-4.0) followed the fine structure data, and the total cross-section on 56 Fe was generated by subtracting the contributions of the other isotopes, and thus it still has resonance features as shown in Figure 5 (left). The angular distributions are important for the criticalities of fast reactors with iron reflectors. The Legendre coefficients of angular distribution for elastically scattered neutrons were revised in the JEDNL-4.0 evaluation using fine structure data measured by Perey, et al. (1991) in the range of 40 to 850 kev and by Kinney and McConnell (1976) in the range of 850 kev to 2.5 MeV. Model calculation was done by POD in the energies above 2.5 MeV. Figure 5 (right) shows that JENDL-4.0 has stronger forward peaking than the others above 2.5 MeV. The comparisons of angular distributions in different neutron energies are made with experimental data and evaluated libraries in Figure 6. JENDL-4.0 exhibits reasonable agreement with experimental data, except for the distributions at the most backward angles in higher energies. Inelastic scattering cross-sections The cross-sections of inelastic scattering to the first excited level (846 kev) below 2.1 MeV and to the second and third excited levels (2 085 and kev, respectively) in the whole energy range were newly evaluated by POD and coupled-channel optical model calculations. The JENDL-3.3 evaluation was adopted below 2.1 MeV where the data were obtained from high resolution data of Voss, et al. (1971) by taking account of gamma-ray 78 NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD 2014

5 Figure 5: Total cross-section in a resonance-like region (left) and first-order Legendre coefficients of elastic scattering angular distribution (right) Figure 6: Elastic scattering angular distributions angular distributions (Smith, 1976). These revisions for JENDL-4.0 were based on a shielding benchmark test. Figure 7 shows the 56 Fe(n,n1) and 56 Fe(n,n2) inelastic scattering cross-sections with uncertainties in comparison with experimental data and the other libraries. The 56 Fe(n,n1) cross-section above 7 MeV in JENDL-4.0 is somewhat small relative to experimental data. The uncertainty, however, is reasonable, compared to the measured data and even to JENDL-3.3. On the other hand, comparing with available experimental data, the 56 Fe(n,n2) cross-section in JENDL-4.0 is large below 4 MeV and small above 4 MeV, but the evaluated uncertainties still cover the experimental data. Figure 8 compares the angular distributions of inelastic scattering to the first and second excited levels with measured data and those of the evaluated libraries. The Legendre coefficients for inelastic scattering in JENDL-4.0 remain unchanged from those of JENDL-3.3. Consequently, the difference derives from each cross-section. The diversity of the measured angular distributions is relatively large, and thus we cannot judge which libraries are better to reproduce the experimental data. Much experimental effort is needed to increase the accuracy of nuclear data for the inelastic scattering angular distributions. (n,p) and (n,2n) reaction cross-sections Figure 9 shows the (n,p) and (n,2n) reaction cross-sections. The (n,p) reaction has been used as a monitor for measurements by the activation method. The cross-section in JENDL-4.0 was basically taken from JENDL-2 released in 1984 (Kikuchi, et al., 1985), but at the evaluation of JENDL-3 (Shibata, et al., 1990) it was revised by considering the data of Smith and Meadow (1975) below 7 MeV and Ikeda, et al. (1988) between 13 and 16 MeV. Covariance was based on the experimental data, and the uncertainty is 5 to 7%. The (n,2n) reaction cross-section in JENDL-4.0 was very similar to that in ENDF/B-VII.1. This is due NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD

6 Figure 7: Cross-sections of inelastic scattering to the first and second excited levels Figure 8: Angular distributions of inelastic scattering to the first and second excited levels Figure 9: (n,p) and (n,2n) reaction cross-sections to the use of the same input parameter. The cross-section above 15 MeV is somewhat uncertain due to the lack of measured data. The covariance data were evaluated by KALMAN, considering the data measured by Frehaut, et al. (1980), Corcalciuc, et al. (1978), and Wenusch and Vonach (1962). The uncertainty derived from the covariance of model parameters was 2.4 to 6% with increasing energy above 15 MeV. 80 NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD 2014

7 Neutron emission double differential cross-sections Double differential cross-sections for neutron emissions at 14.6 MeV incident energy and at 5 and 90 were compared with experimental data and the evaluated libraries in Figure 10. Since the re-evaluation for JENDL-4.0 was very limited, the difference between JENDL-3.3 and JENDL-4.0 is only seen in regions related to elastic and inelastic scattering components. The non-negligible differences among evaluated libraries especially at the most forward and backward (not shown) angles were found in the secondary neutron energy range where inelastic scattering components are important. Figure 10: Neutron emission double differential cross-sections Acknowledgements The authors are grateful to the members of Nuclear Data Center and the Japanese Nuclear Data Committee for their various contributions to the development of JENDL-4.0. References ABBN (ABBN Laboratory) (n.d.), Data Files and Substantiations, webpage on ABBN website, ABBN Laboratory, Department of Nuclear Power Plants, accessed 28 May 2011, Chadwick, M.B., et al. (2011), ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data, Nucl. Data Sheets, 112 [12], pp Corcalciuc, V., et al. (1978), A Study of the Neutron Induced Reactions for 19 F, 56 Fe and 59 Co in the Energy Interval 16 to 22 MeV, Nucl. Phys. A, 307 [3], pp Derrien, H., et al. (2007), 239 Pu Neutron Resonance Parameters Revisited and Covariance Matrix in the Neutron Energy Range from Thermal to 2.5 kev, Proc. of Int. Conference on Nuclear Data for Science and Technology, Nice, France, April 2007, article No. 176, CEA, EDP Sciences. NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD

8 Derrien, H., et al. (2009), R-Matrix Analysis of 238 U High-Resolution Neutron Transmissions and Capture Cross Sections in the Energy Range 0 to 20 kev, J. Nucl. Sci. Eng., 161 [2], pp Frehaut, J., et al. (1980), Status of (n,2n) Cross Section Measurements at Bruyères-le- Châtel, INDC(USA)-84 (1), pp Ge, Z.G., et al. (2011), The Updated Version of Chinese Evaluated Nuclear Data Library (CENDL-3.1), J. Korean Phys. Soc., 59 [2], pp Ichihara, A., et al. (2007), Program POD; A Computer Code to Calculate Cross Sections for Neutron-Induced Nuclear Reactions, JAEA-Data/Code , Japan Atomic Energy Agency. Ikeda, Y., et al. (1988), Activation Cross Section Measurements for Fusion Reactor Structural Materials at Neutron Energy from 13.3 to 15.0 MeV Using FNS Facility, JAERI-1312, Japan Atomic Energy Research Institute. Iwamoto, O. (2007), Development of a Comprehensive Code for Nuclear Data Evaluation, CCONE, and Validation Using Neutron-Induced Cross Sections for Uranium Isotopes, J. Nucl. Sci. Technol., 44 [5], pp Iwamoto, O. (2008), Systematics of Prompt Fission Neutron Spectra, J. Nucl. Sci. Technol., 45 [9], pp Jandel, M., et al. (2012), New Precision Measurements of the 235 U(n, ) Cross Section, Phys. Rev. Lett., 109, [5 pages]. Kawano, T. and K. Shibata (1997), Covariance Evaluation System, JAERI-Data/Code , Japan Atomic Energy Research Institute. Kawano, T., et al. (2000), Evaluation of Fission Cross Section and Covariances for 233 U, 235 U, 238 U, 239 Pu, 240 Pu, and 241 Pu, JAERI-Research , Japan Atomic Energy Research Institute. Kikuchi, Y., et al. (1985), Second Version of Japanese Evaluated Nuclear Data Library (JENDL-2), J. Nucl. Sci. Technol., 22 [8], pp Kinney, W.E. and J.W. McConnell (1976), High Resolution Neutron Scattering Experiments at ORELA, Proc. Int. Conf. Interactions of Neutrons with Nuclei, Lowell, p Koning, A., et al. (2006), The JEFF-3.1 Nuclear Data Library, JEFF Report 21, NEA No. 6190, OECD/NEA, Paris. Leal, L.C., et al. (1999), R-Matrix Analysis of 235 U Neutron Transmission and Cross-Section Measurements in the 0- to 2.25-keV Energy Range, J. Nucl. Sci. Eng., 131 [2], pp OECD/NEA (Organisation for Economic Co-operation and Development/Nuclear Energy Agency) (2012), JEFF Neutron Data, webpage on OECD/NEA Data Bank website, OECD/NEA, Paris, accessed 28 May 2014, Ohsawa, T. (2001), New Evaluation of Prompt Neutron Spectra of 235 U and 239 Pu for JENDL-3.3, JAERI-Conf , pp , Japan Atomic Energy Research Institute. Perey, C.M., et al. (1991), 56 Fe and 60 Ni Resonance Parameters, Conf. on Nucl. Data for Sci. and Technol., Juelich, pp Shibata, K., et al. (1990), Japanese Evaluated Nuclear Data Library, Version-3, JENDL-3, JAERI-1319, Japan Atomic Energy Research Institute. Shibata, K., et al. (2002), Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3, J. Nucl. Sci. Technol., 39 [11], pp NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD 2014

9 Shibata, K., et al. (2011), JENDL-4.0: A New Library for Nuclear Science and Engineering, J. Nucl. Sci. Technol., 48 [1], pp Smith, D.L. (1976), Fast-Neutron Gamma-Ray Production from Elemental Iron: En 2 MeV, ANL/NDM-20, Argonne National Laboratory, Argonne, IL. Smith, D.L. and J.W. Meadows (1975), Cross-Section Measurement of (n,p) Reactions for 27 Al, 46,47,48 Ti, 54,56 Fe, 58 Ni, 59 Co and 64 Zn from Near Threshold to 10 MeV, Nucl. Sci. Eng., 58, pp Voss, F., et al. (1971), Measurement of High Resolution G-ray Production Cross Sections in Inelastic Neutron Scattering on Al and Fe between 0.8 and 13 MeV, Proc. Third Conf. on Neutron Cross Sections and Technology, Knoxville, TN, pp Wenusch, R. and H. Vonach (1962), (n,2n) Cross-Section Measurements on 55 Mn, 59 Co, 52 Cr, 56 Fe and 68 Zn for 14 MeV Neutrons, Oesterr. Akad. Wiss., Math-Naturw. Kl., Anzeiger, 99, p. 1. NEMEA-7/CIELO WORKSHOP PROCEEDINGS, OECD