Results of evaluation of neutron cross sections are presented for 27 fission product nuclides selected as being most important for fast reactor calculation. The cross sections considered are total, elastic scattering, inelastic scattering and capture cross sections in the energy range from thermal to 15 MeV. Thermal and resonance cross sections were calculated from resonance parameters. The calculated thermal capture cross section was adjusted by the measured value by adding a background cross section of 1/υ form. A modified multi-level Breit-Wigner formula was developed to avoid the well-known occur-rence of negative values in elastic scattering cross section. Smooth cross sections above resonance region were calculated with the spherical optical model and the statistical theory, taking account of neutron width fluctuation and level interference. The calcula-tion was adjusted by capture data when available. The joining between the resonance and smooth cross sections was performed with the aid of statistical examination using Monte Carlo method. Present results are discussed in comparison with other evaluated data sets. Numerical results are stored on magnetic tape in the ENDF/B format.

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A computer system for performing quick survey and optimization has been developed to aid in-core fuel management of sodium cooled fast reactors. The method utilizes the conversational mode of computer operation, to perform on-line computation, display of results and acceptance of commands, thus permitting rapid exchange of information be-tween the machine and the person in charge. To improve the overall efficiency of the system as problem-solver, an algorithm is introduced to provide for automated shuffling to determine the fuel loading patterns of successive cycles. This algorithm serves to model the refueling and shuffling arrangement after a prescribed optimized standard core pattern according to the extent of burnup of each fuel sub-assembly.
The proposed system is applied to problems of : (1) straightforward simulation of fuel performance in given in-core fuel management programs, covering core and radial blanket refueling and control rod positioning, (2) optimization, and (3) modification of in-core fuel management programs.
Several numerical examples are treated, to confirm the applicability of this conver-sational-mode problem-solving system to in-core fuel management.

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The time-dependent behavior of monochromatic neutrons from a pulsed cold moderator system provides important information for application of the system as a pulsed cold neutron source. The time distribution of neutron pulses from 20°K methane has been studied using time-of-flight and crystal monochromator techniques. The first time-moments of the neutron pulses and the thermalization time constant have been obtained.

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Small leak sodium-water reaction tests were conducted to develop various kinds of leak detectors for the sodium-heated steam generator in FBR. The super-heated steam was injected into sodium in a reaction vessel having a sodium free surface, simulating the steam generator. The level gauge in the reaction vessel generated the most reliable signal among detectors, as long as the leak rates were relatively high. The level gauge signal was estimated to be the sodium surface oscillation caused by hydrogen bubbles produced in sodium-water reaction.
Experimental correlation was derived, predicting the amplitude as a function of leak rate, hydrogen dissolution ratio, bubble rise velocity and other parameters concerned, assuming that the surface oscillation is in proportion to the gas hold-up. The noise amplitude under normal operation without water leak was increased with sodium flow rate and found to be well correlated with Froud number. These two correlations predict that a water leak in a "MONJU" class (300 MWe) steam generator could possibly be detected by level gauges at a leak rate above 2 g/sec.

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The absolute fission rates of ²³⁵U, ²⁸⁷Np, ²⁵⁸U and ²³²Th were measured in four types of spherical blanket assemblies containing lithium and/or natural uranium and/or gra-phite. The results of measurement are compared with those of one-dimensional transport calculations employing 100-group neutron cross-sections obtained from the ENDF/B-IV data file. It is shown that the ratios between calculated and experimental values of ²³²Th, ²³⁸U and ²³⁷Np fission rates decrease with distance from the assembly center, where D-T neutrons are generated. An overestimation of about 50% is observed in the calcu-lated ²³⁵U fission rate for the graphite reflector region.
One of the main sources of the disagreement proves to lie in the inability of the codes adopted for generating the multi-group cross-section to take account of the angular distributions of the secondary neutrons emanating from nonelastic reactions. The results of the analysis indicate that the method of calculation currently employed in fusion reactor neutronics overestimates the reflection of neutrons and underestimates the pene-tration of fast neutrons when a graphite reflector is used.

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The NSRR programme is in progress in JAERI using a pulsed reactor to investigate fuel behaviors under the reactivity-initiated accident conditions. Pulsing characteristics and experimental capability, especially heat deposition in test fuel rods given by a single pulse are key parameters to this purpose.
In pulsing performance tests, it has been ascertained that the maximum pulsing with 4.67 $ (=3.41%Δk) brings peak reactor power of 21, 100 MW and core energy release of 117 MW -sec. The calculated time responses of reactor power, fuel temperature and cladding surface temperature as well as these maximum values at various pulse sizes agreed well with measured data. In addition, it has been also ascertained by measure-ment as well as analysis that there are no essential differences in pulsing characteristics between the pulsing from critical and that from subcritical.
The heat deposition in a test fuel rod given by a single pulse is much enough as predicted, and a 2.6% enriched BWR type fuel rod gains about 230 cal/g•UO₂ in the maximum pulsing. In case of irradiation of clustered five test fuel rods by a single pulse, heat deposition reduces by about 20% for a surrounding rod and about 40% for a center rod in comparison with that in a single rod irradiation.

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