Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.
A unified cross-section set, ADJ2000, which combined wide range of integral experimental information with differential nuclear data JENDL-3.2, has been developed by using the cross-section adjustment technique based on the Bayesian parameter-estimation theory. The ADJ2000 unified cross-section set has obtained more reliability and generality to predict the various core characteristics of fast reactors accurately, since it is a compilation of the latest sensitivity and adjustment methodology, the newest nuclear data library and the consistent covariance file in Japan, and the wide set of integral experimental database from the world. The ADJ2000 set is open to the public, and is expected to be utilized in the study and design work of future fast reactors.
Covariances of neutron nuclear data have been estimated for 16 nuclides contained in JENDL-3.2. The physical quantities for which covariances are deduced are cross sections, resolved and unresolved resonance parameters, the first order Legendre-polynomial coefficients for the angular distribution of elastically scattered neutrons, and fission neutron spectra. As for 233,235, 238 U and 239, 241 pU, covariances were obtained also for the average number of neutrons emitted in fission. Least-squares methods were applied to obtain the covariances of those cross sections which were based on experimental data. A simultaneous evaluation method yielded uncertainties in the fission cross sections of 235U, 238U, 239pU, 240pU and 241pU. Covariances ofnuc\ear model calculations were determined from uncertainties in model parameters. The covariance file thus obtained is processed by a system which has been developed, and used for the adjustment of group cross sections.
A cell calculation code SLAROM-UF has been developed for fast reactor analyses to produce effective cross sections with high accuracy in practical computing time, taking full advantage of fine and ultra-fine group calculation schemes.The fine group calculation covers the whole energy range in a maximum of 900-group structure. The structure is finer above 52.5 keV with a minimum lethargy width of 0.008. The ultra-fine group calculation solves the slowing down equation below 52.5 keV to treat resonance structures directly and precisely including resonance interference effects. Effective cross sections obtained in the two calculations are combined to produce effective cross sections over the entire energy range.Calculation accuracy and improvements from conventional 70-group cell calculation results were investigated through comparisons with reference values obtained with continuous energy Monte Carlo calculations. It was confirmed that SLAROM-UF reduces the difference in k-infinity from 0.15 to 0.01% for a JOYO MK-I fuel subassembly lattice cell calculation, and from À0:21% to less than a statistical uncertainty of the reference calculation of 0.03% for a ZPPR-10A core criticality calculation.
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