Benchmark Tests of Gamma-Ray Production Data in JENDL-3 for Some Important Nuclides

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1 Journal of NUCLEAR SCIENCE and TECFINOLOGY, 27[9], pp. 844~852 (September 1990). TECHNICAL REPORT Benchmark Tests of Gamma-Ray Production Data in JENDL-3 for Some Important Nuclides CAI Shao-huit, Akira HASEGAWA, Tsuneo NAKAGAWA and Yasuyuki KIKUCHI Department of Physics, Tokai Research Establishment, Japan Atomic Energy Research Institute* Received February 16, 1990 Gamma-ray production data of N, Na, Al, Si, Ca, Ti, Fe, Ni and Cu in JENDL-3 have been tested by comparing the calculated g-ray spectra with the experimental ones measured at ORNL. As to g-rays arising from thermal neutron capture, structures observed in the measured spectrum could not be reproduced for most cases with the test version of JENDL-3 (JENDL-3T) whose evaluation was made mainly on the basis of theoretical calculation, while they were reproduced well with ENDF/B-IV. As to 7-rays arising from fast neutron reactions, on the other hand, JENDL-3T reproduced the measured spectra fairly well except for Ca, Fe and Ti, where JENDL-3T gave too high values in the g-ray spectra between 4 and 6 MeV. Considering the results of the benchmark tests, the g-ray spectra arising from the thermal neutron capture were reevaluated and the modified data were adopted in the final version of JENDL-3. KEYWORDS: comparative evaluations, gamma-ray production data, nitrogen, sodium, aluminium, silicon, calcium, titanium, iron, nickel, copper, benchmark tests, JENDL-3T, JENDL-3, gamma-ray spectrum, thermal neutron capture, fast neutron reactions, ORNL Tower Shielding Facility I. INTRODUCTION The compilation of the third version of Japanese Evaluated Nuclear Data Library (JENDL-3)(1)(2) was completed in October 1989, and JENDL 3 is now released to all over the world. The JENDL-3 contains more reliable neutron cross section data than JENDL-2. Furthermore for some important nuclides it provides g-ray production cross sections and angular and energy distributions of emitted photons, which are much required for shielding and g-heating calculations for both fission and fusion reactors. It is essential to verify the reliability of the contained data through various benchmark tests before release of JENDL-3. The following process was taken for the verification of the JENDL-3 data : A test version of JENDL- 3, called JENDL-3T, was provided in 1987 and has been tested from various viewpoints. If some drawbacks were pointed out from the tests, they were informed to the evaluators as feedback. The evaluators then checked their data and some modification was made if necessary, and the final version of JENDL- 3 was completed. The results of benchmark tests for the neutron cross section data were published elsewhere(2)(3). Here are reported results of the benchmark tests for the g-ray production cross sections and the g-ray spectrum data in JENDL-3T and JENDL-3. In Chap. II, experiments utilized in the present benchmark tests are briefly described. The format of g-ray data in JENDL-3 is * Tokai -mura, Ibaraki-ken A visiting scientist through t STA scientist ex - change program from Institute of Applied Phys - ics and Computational Mathematics, Beijing, The People's Republic of China. 62

2 Vol. 27, No. 9 (Sep. 1990) TECHNICAL REPORT (Cai Shao-hui et al.) 845 described in Chap. M. Methods of data processing and comparison are given in Chap. IV. In Chap. V, results and discussion are given on JENDL-3T and JENDL-3. II. EXPERIMENTS UTILIZED IN PRESENT BENCHMARK TESTS Two kinds of experiments were used in the present benchmark tests. (1) Gamma-ray spectra measured at Oak Ridge National Laboratory Tower Shieldi ng Facility (TSF) Gamma-ray spectra arising from thermal neutron capture(4) and fast neutron induced reactions(5) were measured at ORNL-TSF by using a carefully calibrated 5x5 in. Nal detector in a good geometry. The experimental results were reported in the form of energy-differential g-ray production cross section with 0.5 MeV g-ray energy bins. It is to be noted that the g-ray spectra arising the fast neutron induced reactions were measured as averaged ones over the neutron spectrum of the TSF reactor. The uncertainties were estimated to be 15% for the thermal neutron measurements and 30% for the fast neutron measurements. Table 1 lists the natural elements measured in these experiments. The oxygen data were given only for 2 discrete levels and have not been used in the present benchmark tests. Table 1 Natural elements measured at ORNL-TSF(4)(5) (2) Double-differential g-ray production cross sections measured at Oak Ridge Electron Linear Accelerator (ORELA) Gamma-ray spectra were measured(6) at a few angles at ORELA in the neutron energy range from about 1 MeV to 20 MeV. Table 2 shows the natural elements measured in the experiments. These experimental data were used supplementally in the present work. Table 2 Natural elements measured at ORELA(6) 63

3 846 TECHNICAL REPORT (Cai Shao-hui et al.) J. Nucl. Sci. Techhol., III. SECONDARY GAMMA.RAY PRODUCTION DATA The JENDL-3 adopts the ENDF-5 format(7). According to this format, the secondary g-ray production data consist of g-ray production cross sections and angular and energy distributions of emitted photons. (1) Gamma-ray Production Cross Section, sg(en) This quantity can be specified in one of the following two ways : (a) Multiplicities or transition probability arrays The g-ray production cross section is given by I V. DATA PROCESSING In order to compare the data of JENDL-3T or JENDL-3 with the benchmark experiments made at ORNL-TSF, the following two quantities were calculated from JENDL-3T or JENDL-3. (1) Absolute g-ray spectra : (3) (4) where En is the neutron energy, sg(eg<- En) is the cross section to produce a photon of energy Eg, the subscript k designates a photon with a particular discrete energy or a photon with continuum energy, Y k(en) is the photon yield which may be given by Multiplicities or Transition Probability Arrays, and s(en) is the neutron cross section to which the multiplicities are referred. (b) sgk(en) tabulated In this case there is no need to refer to the neutron files. (2) Photon Angular Distribution, Pk(m, En) This quantity is given as (1) where p(en) is the neutron flux and MT specifies a reaction type. (2) Averaged g-production cross section : (5) where Eming corresponds to the lower limit of the detected g-energy in the experiment. In addition to the afore-mentioned quantities, the double-differential g-ray production cross sections were calculated by assuming isotropic distribution for r-ray emission : (6) In JENDL-3, the photon angular distribution is assumed to be isotropic. (3) Normalized Continuous Photon Energy Distribution, fk(eg<-en) This quantity is given as (2) which can be used to compare with the measurements at ORELA. Figure 1 describes the flow chart of the present data processing. The whole processing consists of four steps : (1) Pre-processing evaluated nuclear data file The LINEAR(8) code makes tabulated cross sections linearly interpolatable on a linearlinear scale. Pointwise cross sections are 64

4 Vol. 27, No. 9 (Sep. 1990) TECHNICAL REPORT (Cai Shao-hui et al.) 847 calculated from resolved and unresolved resonance parameters with the RECENTP(9) code and Doppler-broadened cross sections are obtained with the SIGMA10(10) code (at 300 K in the present case). Fig. 1 Flow chart of data processing (2) Generating multi-group neutron and photon cross sections The codes NJOY(11) and PROF-GROUCH(12) generate multi-group cross sections. (3) Calculating integral quantities corresponding to benchmark experiments The quantities expressed with Eqs. (4)~ (6) are calculated from the multi-group cross sections with a newly developed code GAMMA, and are stored in the ENDF format. The experimental data were also stored in the ENDF format with the GAMMA code. (4) Plotting experimental and calculating results The calculated integral quantities and the experimental results are compared by means of graphs plotted with the SPLINT(18) code. V. RESULTS AND DISCUSSION 1. Gamma-ray Spectra Arising from Thermal Neutron Capture Gamma-rays associated with thermal neutrons are caused only by the capture reaction. Its spectra show some peaks corresponding to discrete levels of the residual nucleus. It is difficult to reproduce these peaks precisely with theoretical calculations based on the statistical model. (1) JENDL-3T Most of JENDL-3T data were evaluated with the theoretical calculation. This may often result in failing to reproduce the structure observed in the measured spectra as shown in Fig. 2(a), for example, while ENDF/B-IV reproduces them fairly well. The evaluation of ENDF/B-IV may have been made by taking these measurements into consideration. The same tendency is observed for Na, Si and Ni. For Ni the data of ENDF/B-IV must have been evaluated so as to reproduce these benchmark experiments as seen in Fig. 2(b). On the other hand, the JENDL-3T data for Ca, which were evaluated also with the theoretical calculation, reproduce the structure satisfactorily as shown in Fig. 2(c). As to Fe, the JENDL-3T data were evaluated on the basis of the measurements made at Tokyo Institute of Technology(14) in the kev region. Therefore structure of g-ray spectrum is reproduced to some extent as shown in Fig. 2(d), though the difference between the thermal and kev region causes some discrepancies particularly in the g-ray energy region below 6 MeV. Both the JENDL-3T and ENDF/B-IV data 65

5 848 TECHNICAL REPORT (Cai Shao-hui et al.) J. Nucl. Sci. Technol., (a) 27Al (b) Ni (c) Ca (d) Fe Fig. 2(a)~(f) Absolute p-ray spectra arising from thermal neutron capture for 27Al, Ni, Ca, Fe, Ti and Cu 66

6 Vol. 27, No, 9 (Sep. 1990) TECHNICAL REPORT (Cai Shao-hui et al.) 849 Fig. 2(e) Ti Fig. 2(f) Cu of Ti, which were evaluated with the theoretical calculation, fail to reproduce the measured spectrum as seen in Fig. 2(e). As to Cu, ENDF/B-IV reproduces the measured spectrum above 4.5 MeV but gives much lower values than the measured data below 4.5 MeV, while JENDL-3T gives higher values below 4.5 MeV as seen in Fig. 2(f). (2) JENDL-3 Taking account of the results of the benchmark tests for JENDL-3T, evaluators of each nuclide reexamined the thermal capture g-ray data. The data of 16 nuclides or elements were modified for JENDL-3 as results of the reexamination. The 16 nuclides or elements are 14N, 23Na, 27Al, Si, Mg, Ti, Cr, 55Mn, Fe, Ni, Cu, Zr, Cd, Eu, W and 181Ta. The data of Ca were not changed. The modification was made on the basis of available experimental data of the thermal capture g-ray. As a result of the modification the JENDL-3 data can reproduce well the structure observed in the benchmark experiments for 27Al, Fe and Ti as seen in Figs. 2(a), (d) and (e), respectively. For 14N, 23Na, Si, Ni and Cu, however, the modification was simply made so as to reproduce the structure observed in the ORNL-TSF benchmark experiments as seen in Figs. 2(b) and (f). No Cr data are available in the ORNL-TSF benchmark experiments. From the improvements in the case of 27Al, Fe and Ti, we may presume that the Cr data of JENDL-3 were also improved by considering the available experimental data. They are compared in Fig Gamma-ray Spectra Arising from Fast Neutron Reactions Gamma-rays associated with fast neutron reactions are produced by various reactions such as capture, inelastic scattering, (n, 2n), (n, a) etc. Among them, the g-rays from inelastic scattering have the highest contribution. The structures of g-ray spectra corresponding to various reactions are accumulated and the whole g-ray spectra show less structure than those arising from the thermal capture. Hence the theoretical calculation with the statistical model can reproduce the -ray spectra fairly well. g 67

7 850 TECHNICAL REPORT (Cai Shao-hui et al.) J. Nucl. Sci. Technol., Fig. 3 Absolute g-ray spectra arising from thermal neutron capture for Cr Figure 4(a) shows the g-ray spectra of Na. The JENDL-3T data agree with the measurements at ORNL-TSF better than the ENDF/ B-IV data, which are much lower. Both JENDL-3T and ENDF/B-IV data agree fairly well with the measurements for Al and Si. Figure 4(b) shows the case of Fe. Both the JENDL-3T and ENDF/B-IV data agree with each other except in the energy region between 4 and 6 MeV, where the JENDL-3T data are two times higher than the ENDF/ B-IV data. The measured data support the lower values of ENDF/B-IV. The same trend is observed for Ca and Ti. Figure 5 compares the double differential g-ray production cross sections of JENDL-3T for Fe with the measured ones(6) at ORELA for the incident neutron energy of around 7 MeV. Again the JENDL-3T data are higher than the measured data in the g-ray energy region between 4 and 6 MeV. The analogous disagreement is observed between JENDL-3T and ENDF/B-IV for Ni and Cu. In these cases, however, the higher values of JENDL-3T agree with the measured data(5) as seen in Fig. 6. It was concluded from the arguments mentioned above that the data of JENDL-3T (a) 23Na (b) Fe Fig. 4(a),(b) Absolute g-ray spectra arising from fast neutron reactions for 23No and Fe 68

8 Vol. 27, No. 9 (Sep. 1990) TECHNICAL REPORT (Cai Shao-hui et al.) 851 Circles show the experimental data(6) measured at an angle of 125- and multiplied by 4p. Fig. 5 g-ray production cross section for incident neutron energy between 6.5 and 7 MeV were satisfactory for the g-rays arising from the fast neutron reactions. Hence the data of JENDL-3T were adopted in JENDL-3 without modification. 3. Averaged Gamma-ray Production Cross Sections The averaged g-ray production cross sections expressed by Eqs. (5) are given in Table 3. The C/E values are shown in Fig. 7. As for the thermal capture g-rays, the C/E Fig. 6 Absolute r-ray spectra arising from fast neutron reactions for Ni values of JENDL-3T deviate from unity by more than 25% for 14N, Si, Ni and Cu. As a result of the modification, on the other hand, the C/E values of JENDL-3 stay near unity within 20%. The data of ENDF/B-IV are not satisfactory for Ti and Cu as predicted from Figs. 2(e) and (f). For the g-rays associated with the fast neutron reactions, the C/E values of both JENDL-3 and ENDF/B-IV agree well with each other and stay near unity, except for Na for which ENDF/B-IV gives very small Table 3 Averaged g-ray production cross section 69

9 852 TECHNICAL REPORT (Cai Shao-hui et al.) J. Nucl. Sci. Technol., evaluators of the JENDL-3 g-ray data. The first author (Cai Shao-hui) would like to express his great thanks to Science and Technology Agency of Japanese Government and Dr. N. Shikazono for giving him a chance to make the present work at Japan Atomic Energy Research Institute. REFERENCES Fig. 7 C/E values of averaged g-ray production cross section values as is predicted from Fig. 4(a). VI. CONCLUDING REMARKS The present benchmark tests have revealed the following problems of the g-ray production data in JENDL-3T. (1) The JENDL-3T could not generally reproduce the structures observed in the measured r-ray spectra arising from the thermal neutron capture. (2) Not only the spectra but also the averaged g-ray production cross sections were not satisfactory for 14N, Si, Ni and Cu. (3) The JENDL-3T reproduced well the g- ray spectra arising from the fast neutron reactions. For Ca, Fe and Ti, however, JENDL-3T gave higher values in the g-ray spectrum between 4 and 6 MeV. The present benchmark results were fed back to the evaluators and reevaluation was made for the thermal capture g-ray spectra. The modified g-ray data were much better than JENDL-3T and then were adopted in the final JENDL-3. ACKNOWLEDGMENTS The authors wish to thank Drs. S. Igarasi and K. Shibata for their helpful discussion. They appreciate valuable comments from the (1) SHIBATA, K., et al.: Japanese Evaluated Nuclear Data Library Version 3, JAERI-1319, (1990). (2) ASAMI, T., et al.: J. At. Energy Soc. Jpn., (in Japanese), 31[11], 1190 (1989). (3) TAIKANO, H., et al.: Benchmark tests of JENDL-3 for thermal and fast reactors, Proc. Int. Conf. on Physics of Reactors, April 23~ 27, 1990, Marseille, Vol. III, p. PI, 21 (1990). (4) MAERKER, R. E. : SB2. Experiment on secondary gamma-ray production cross section arising from thermal-neutron capture in each of 14 different elements plus a stainless steel, ORNL- TM-5203, (1976). (5) idem: SB3. Experiment on secondary gammaray production cross section averaged over a fast-neutron spectrum for each of the 13 different elements plus a stainless steel, ORNL- TM-5204, (1976). (6) Each element is referred with an independent document. Here we show the case of iron ; CHAPMAN, G. T., et al.: A re-measurement of the neutron-induced gamma-ray production cross sections for iron in the energy range 850 kev En<=20 MeV, ORNL/ TM-5416, (1976). <= The report numbers are given in Table 2. (7) Data formats and procedures for the Evaluated Nuclear Data File, ENDF/B-V, ENDF-102, (Edited by KINSEY, R.), (1979) ; Revised by B. A. Magurno, Nov (8) CULLEN, D. E.: Program LINEAR (Version 79-1), UCRL-50400, Vol. 17, Part A, (1979). (9) idem: Program RECENT (Version 79-1), UCRL-50400, Vol. 17, Part C (1979) ; RECENTJ is a modified version in JAERI. (10) idem: SIGMA1 (Version 79-1), UCRL-50400, Vol. 17, Part B, (1979). (11) MaCFARLANE, R. E., et al.: The NJOY nuclear data processing system, LA-9303-M, Vol. 1, Vol. 2, (1982). 2) HASEGAWA, (1 A. : Private communication, (1989). NARITA, T., et al.: SPLINT, JAERI-M (13) 5769, (in Japanese), (1974). (14) IGASHIRA, M., et al.: Measurements of kevneutron capture gamma-ray spectra of Fe and Ni, Proc. Int. Conf. Nuclear Data for Science and Technology, Mite, 1988, p. 67 (1988). 70