Testing of Nuclear Data Libraries for Fission Products

Transcription

1 Testing of Nuclear Data Libraries for Fission Products A.V. Ignatyuk, S.M. Bednyakov, V.N. Koshcheev, V.N. Manokhin, G.N. Manturov, and G.Ya. Tertuchny Institute of Physics and Power Engineering, 242 Obninsk, Russia Abstract. The status of the radiative capture and neutron inelastic scattering evaluations for the most important fission products is analyzed. New BROND-3 evaluations are briefly considered. Experiments on the BFS critical assemblies, which can be used to test the available evaluations, are discussed. INTRODUCTION The importance of the fission-product neutron cross sections for the analysis of a irradiated nuclear fuel in thermal and fast reactors is well known. Development of advanced nuclear reactors with a higher burn up, but simultaneously with more severe safety criteria, requires more precise evaluations of neutron data than the available ones in data libraries nowadays. As far as the number of fission products being very large the lumped or pseudo-fission-product is usually used in most practical calculations. However, to estimate the cross sections for such a pseudo-nuclide with high accuracy the evaluated data files for more than two hundred nuclei are needed. The number of required files will be even larger if the modern MCNP or similar Monte Carlo codes are applied for the analysis of reactors or spent fuel properties. Activity of the Russian Nuclear Data Center directed towards improvement of data evaluations for fission products is briefly discussed in the present paper. NEEDS FOR DATA IMPROVEMENTS Comparison of JEF-2.2, JENDL-3.2,, and FOND-2.1 evaluations for fission products was performed by the WPEC on the basis of the fastreactor standard spectrum [1]. One-group cross sections for individual nuclides were weighted in accordance with their yields in the 239 Pu fission and the effective cross sections for the lumped fission-product were calculated. Discrepancies between different libraries for the one-group capture cross sections do not exceed 1% for most of the nuclides and the mean square deviation for the lumped fission product was about 6%. Such uncertainties of evaluated data seem to be satisfactory for the current requests on nuclear-data accuracy. During the last ten years the files for the most important fission products were improved in all national data libraries. To estimate better the present status of the fission-product data we performed a similar analysis of the ENDF/-B-VI (rel.8), JEFF-3., JENDL-3.3, and FOND-2.2 evaluations. Discrepancies for the radiative capture cross sections have been found much lower than before. In particular the mean square deviation of the capture cross sections for the pseudo-fission product was decreased from.9% to 1.6%. The higher capture cross sections belong, as a rule, to the JEFF-3. evaluations and the lower ones belong to the FOND-2.2 evaluations. The radiative capture cross section for the lumped fission nuclide is shown in Fig. 1. Discrepancies between libraries are rather small in the energy range below 1 MeV, which is most important for fast reactors. For inelastic scattering the mean square deviation of the one-group cross sections achieves now about 7.% instead of the previous 1%. The inelastic scattering cross sections for the lumped nuclide are presented in the right part of Fig. 1. Evaluation discrepancies are large enough for both the threshold energies and energies above 1 MeV. For evaluations of the (n,2n), (n,p), and (n,α) reactions the discrepancies of the one-group cross sections remain considerable. They are about 3% for the (n,2n) and 4-6% for the (n,p) and (n,α) reactions. The list of the most important fission products, estimated in accordance with a contribution of each nuclide to the lumped fission-product capture cross 14

2 section, includes isotope chains for ruthenium, palladium, cesium, molybdenum, neodymium, samarium, as well as some separate isotopes of 99 Tc, 13 Rh, 139 La, and some others. The improvement of the lumped fission-product cross sections can be achieved mainly by improving the evaluations for these nuclides. So the development of the BROND-3 library for fission products is concentrated on the revision of evaluations namely for the listed nuclides. Evaluations for 1 Pd and 149 Sm, shown in Fig. 2, can be considered as examples of the obtained improvements. 1.E+1. Inelastic lumped FP 1.E+ Capture lumped FP 4. Cross section, barn 1.E-1 1.E-2 FOND-2.2 ENDF/B6.7 JENDL-3.3 JEF-3.3 Cross section, barn FOND-2.2 ENDF/B6.7 JENDL-3.3 JEF E-3 1.E+3 1.E+4 1.E+ 1.E+6 1.E+7 Energy, ev. 1.E+ 1.E+6 1.E+7 1.E+8 Energy, ev FIGURE 1. Energy dependencies of the neutron radiative capture and inelastic scattering cross sections for the fission-product pseudo-nuclide 2, 2, 1, 1,,, 1 1 Pd n,n' ENDF/B-VI JENDL-3 JEFF-3. BROND (n,γ) Hockenbury 197 Macklin 1979 Cornelis 1982 n,2n,1,1 1 1 Neutron energy (MeV) n,n' Macklin 1963 Hockenbury 197 Kononov 1977 Mizumoto 1981 Macklin 1986 Wisshak Sm n,2n 1 1 ENDF/B-VI JEF-2 n,γ JENDL-3 BROND Neutron energy (MeV) FIGURE 2. Available experimental data on the neutron cross sections for 1 Pd and 149 Sm in comparison with different evaluations In comparison with ENDF/B-VI and the changes of the inelastic scattering and (n,2n) cross sections in the BROND-3 evaluation are quite essential. At the same time the improvements of the capture cross sections relate mainly to the energy region above 1 MeV, which is not very important for most practical applications. The discrepancies in the inelastic scattering and (n,2n) cross sections are mainly caused by a limited amount of experimental data for near-threshold energies and also by using models that are too simplified in previous evaluations. A revision of the available evaluations for fission products is carried out now in the frame of the WPEC Subgroup 21 collaboration [2], the result of which should be an 141

3 essential reduction of most discrepancies of evaluated data. TESTING F EVALUATIONS ON INTEGRAL EXPERIMENTS The good agreement of different evaluations does not yet guarantee data reliability, because the evaluations can be based on the same experimental data and rather similar models. For this reason it is very important to test evaluations through independent integral experiments. The best candidates for those are the critical assembly experiments. The well-known example of such experiments is the CRMF data [3], which were widely used in testing the capture cross-section evaluations for ENDF/B-V and JEF-2.. With a similar purpose in a set of experiments was carried out on the BFS critical assemblies [4], neutron spectra of which simulate the spectra of fast reactors with different nuclear fuel composition. The main characteristics of the corresponding assemblies are presented in Table 1. The reactivity coefficients in the center of the critical assembly were measured by the reactivity perturbation method for small samples of nuclides given in Table 2. The samples have a high isotopic purity and were chosen in such a way to obtain the data supplemented to the CFRMF results. Uncertainties of the measured coefficients are presented in Table 2 also, and they are lower than in the CFRMF results. TABLE 1. Experiments on BFS and CFRMF with samples of nuclides-fission products. Assembly Fuel Enrichment, % Neutron Spectrum BFS -1-1 BFS --1 BFS -2-1 BFS -4А BFS -4Б BFS BFS BFS CFRMF U PuU Pu Pu Pu U Fast reactor One zone model BN-6 model Standard fast spectrum Contribution of Neutrons with Energy E n <1 kev, % TABLE 2. Estimated uncertainties of the coefficient reactivity measurements, %. BFS Number 9 Mo 97 Mo 98 Mo 1 Mo 13 Rh 1 Pd 19 Ag 141 Pr 149 Sm 4A B Eu For comparison with calculations the reactivity coefficients, related to the boron reactivity, are used. Because the energy dependence of capture cross sections for most fission products in the considered energy region is similar to the boron one, the use of boron as a monitor enables exclusion of some systematic uncertainties connected with neutron spectrum uncertainties, heterogeneous corrections, self-shielding factors, and so on. Calculations of the neutron flux and the neutron worth for the BFS assemblies have been done with the one-dimensional CRAB code and the ABBN-93 group constants []. The corresponding reactivity coefficients and the averaged capture cross sections for the CFRMF spectra were calculated besides the ABBN-93 (ABBN-93 FOND-2.2) and also with the last versions of ENDF/-B-VI, JEF-3., JENDL-3.3, and BROND-3 libraries. The calculated neutron spectra at the center of BFS assemblies are compared in Fig. 3 with the CRMF spectrum. The analyzed assemblies can be characterized by the parameter 142

4 1кэВ ε = ϕ( E) de / ϕ( E) de, which determines a contribution of neutrons with the energies below 1 kev into the assembly neutron flux. φ(e) 1.E+ 1.E-1 1.E-2 1.E-3 1.E-4 1.E+2 1.E+3 1.E+4 1.E+ 1.E+6 1.E+7 Neutron energy, ev BFS-1-1 BFS-4A BFS-49-2 CFRMF FIGURE 3. Neutron spectra of the BFS and CFRMF. Values of this parameter are given in the last column of Table 1, and they will be used as the abscess in the plots given below. The reactivity coefficients depend on the neutron absorption in a sample, which is proportional mainly to the sum of neutron capture and inelastic scattering cross sections averaged over the corresponding neutron spectrum. Contributions of the inelastic scattering, calculated with the ABBN-93 constants, are given for the studied samples in Table 3. From comparison of differences between the calculated and measured reactivity coefficients, presented in Figs. 4-6, one can conclude that the ENDF/B-VI evaluations for 9 Mo and 97 Mo, which were improved in the last revision, agree with the integral data much better than other evaluations. At the same time the and JENDL-3.3 evaluations for 1 Mo looked preferable relative to the ENDF/B-VI. For 19 Ag and 13 Eu all evaluations, besides JEFF-3., look rather similar, but the obtained fluctuating discrepancies between calculations and measurements indicate, probably, an underestimation of experimental uncertainties. If the corresponding uncertainties are increased about two times, the results of all evaluations will agree reasonably with the measured data. TABLE 3. Ratio of the inelastic scattering and the radiative capture contributions into the total reactivity, %. BFS number 9 Mo 97 Mo 98 Mo 1 Mo 13 Rh 1 Pd 19 Ag 141 Pr 149 Sm 13 Eu 4A B CONCLUSION The analysis of differences between the available evaluations of the neutron radiative capture and inelastic scattering cross sections for fission products allows clarification of nuclides whose revision could be most important for the improvement of data applied for fast-reactor calculations. New BROND-3 evaluations, considered briefly in the present paper, are concentrated mainly on such nuclides. The integral experiments performed on the BFS critical assemblies, which are supplementary to the well-known CFRMF data, provide a reasonable test of new evaluations and a selection of the best ones, which should be recommended for practical applications. REFERENCES 1. Gruppelaar H., et al. Status of pseudo-fission-product cross sections for fast reactors. Report NEA/WPEC- 17 (ECN-R-98-14), OECD, Oblozinsky P., et al. Minutes of SG21 Workshop, BNL, April 19 23, 24; 3. Hanker Y., Anderl R. Integral Cross Section Measurements on Fission Product Nuclides in Fast Neutron Fields in Proc. Meeting on Neutron Cross Sections of Fission Product Nuclei, edited by C. Coceva, G. Panini, Bologna, RTI, 1979, pp Bednyakov S.М., Manturov G.N., Sov. Atomic Energy 72(1), 9-98 (1992).. Manturov G.N., Nilolaev M.N., Tsibulya A.M., Group constant system BNAB-93, VANT, ser.: Nuclear Constants, 1996, iss. 1, pp

5 Mo-9 JENDL-3.3 = JEFF-3. = FOND JENDL-3.3 = JEFF-3. = FOND-2.2 Mo Neutron part below 1 kev, % Neutron part below 1 kev, % FIGURE 4. Difference between experimental and calculated values of the reactivity coefficients for the BFS assemblies with the samples of 9 Mo и 98 Mo = JEFF-3. MO = JEFF Rh Neutron part below 1 kev, % Neutron part below 1 kev, % FIGURE. The same as in Fig. 4 for the samples of 1 Mo и 13 Rh Eu-13 JEFF С/E - 1, % -1 C/E-1, % - = JEFF Ag Neutron part below 1 kev, % Neutron part below 1 kev, % FIGURE 6. The same as in Fig. 4 for the samples of 1 Pd и 19 Ag. 144