The results of the 239PU high resolution neutron transmission measurements of Harvey et al. and of the 239Pu high resolution fission cross section measurements of Weston & Todd performed at the Oak Ridge Electron Linear Accelerator (ORELA) were analysed in the energy range from 1 keV to 2.5 keV by the Bayesian code SAMMY using the Reich-Moore approximation of the R-matrix theory. The results obtained in a previous analysis in the energy range from thermal to 1 keV were updated by taking into account the recent renormalization of the experimental fission data by Weston & Todd. The statistical properties of the parameters of the resonances identified in the energy range from thermal to 2.5 keV were examined and improved values of the average parameters were obtained. The resonance parameters are given in an ENDF-6 format file available from JAERI Nuclear Data Center and from NEA Data Bank (OECD).

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Making use of a lead neutron slowing-down spectrometer combined with an electron linear accelerator and back-to-back type double fission chambers, we measured the fission cross section of ²³⁷Np relative to that of ²³⁵U from about 1 eV to about 5 keV with energy resolution (ΔE/E) of about 40%. The experimentally obtained result has been compared with two newly evaluated data files, JENDL-3 and ENDF/B-VI, and with previously measured values by Plattard et al. and by Hoffman et al. Although the shape of the present energy dependent cross section agrees with that of ENDF/B-VI, that of JENDL-3 below 120 eV and that of Plattard et al., the absolute values of above three are from 3 to 4 times smaller than those of the present data. However the Hoffman et al.'s data are rather closer to the present data.

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The reaction cross sections of ²⁷Al(n, p)²⁷Mg, ²⁷Al(n, α)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn, ⁹⁰Zr(n, 2n)^89m+gZr and ⁹³Nb(n, 2n)^92mNb have been measured by the activation method in an energy range of 13.3 14.9 MeV using the intense D-T neutron source, FNS. Absolute flux was determined by the associated α-particle counting method incorporated with neutron spectra obtained from both a Monte Carlo calculation and a time-of-flight measurement. Corrections were extensively performed not only for the neutron flux determination, but also for the low energy neutron contribution to the reaction rates. The present data were compared with comprehensive evaluations as well as recent experimental data. The measured cross sections of ²⁷Al(n, α)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn and ⁹⁰Zr(n, 2n)^89m+gZr are generally in good agreement within experimental errors with the values in both the JENDL Dosimetry File and IRDF-90. It is also shown that there are the overestimation of the cross sections for ⁹³Nb(n, 2n)^92mNb in the JENDL Dosimetry File, and the overestimation and underestimation of the cross section for ²⁷Al(n, p)²⁷Mg in the JENDL Dosimetry File and IRDF-90, respectively.

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An experimental study is presented on the erosion process caused on the target specimens by solid particle impingement, in which precise quantifications are given as to the dependencies on collision angle and collision velocity. In order to eliminate fluid-dynamic drag force that retards the particles' motion before they impinge upon the targets, the experiments are performed in the vacuum vessel. In the process of erosion, the existence of so-called incubation period is confirmed. Measured exponents of the velocity are considerably smaller than those for the handy application and are rather close to semi-theoretical values of Bitter and Finnie, which suggests that they can be used in the numerical prediction. Regarding the dependency on collision angle, it was found clear discrepancy between the experimental results and predicted one by means of models of Bitter and Finnie as the collision angle increased. For the spherical particles, the cutting erosion actually ceases before the collision angle becomes normal.

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The Code System for Management of JPDR Decommissioning (COSMARD) was developed for use for effective planning and management of reactor decommissioning. Efforts were made to develop; (1) data-processing algorithm using a tree structure which reflects decommissioning Work Breakdown Structure (WBS), (2) a set of commands to describe WBS and specific working conditions as input data, and (3) models to evaluate various project management data using simple mathematical formulas and unit factors. In preparing these models, dismantling work activities related to the Japan Power Demonstration Reactor (JPDR) were analyzed to find out fundamental work items and make suitable formulations for evaluating the manpower needs in both manual and remote dismantling work. Manpower needs for dismantling the JPDR components were then estimated as a sample calculation. An outline of COSMARD, its methodology and calculation models are discussed.

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Thermal recovery of radiation defects and microstructural change in UO₂ fuels irradiated under LWR conditions (burnup : 25 and 44 GWd/t) have been studied after annealing at temperature range of 4501, 800°C by X-ray diffractometry and transmission electron microscopy (TEM). The lattice parameter of as-irradiated fuels increase with higher burnup, which was mainly due to the accumulation of fission induced point defects. The lattice parameter for both fuels began to recover around 450650°C with one stage and was almost completely recovered by annealing at 850°C for 5 h. Based on the recovery of broadening of X-ray reflections and TEM observations, defect clusters of dislocations and small intragranular bubbles began to recover around 1, 1501, 450°C. Complete recovery of the defect clusters, however, was not found even after annealing at 1, 800°C for 5 h. The effect of irradiation temperature on microstructural change of sub-grain structure in high burnup fuels was assessed from the experimental results.

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Two transmutation methods of ¹³⁷Cs using a proton accelerator were evaluated in terms of the effective half life and the transmutation energy. One was the proton method which mainly used high energy proton spallation reaction for transmutation, and the other was the spallation neutron method which mainly used thermal neutron capture reaction. The transmutation energies and the effective half lives for the two transmutation methods were calculated by Monte Carlo simulation codes for particle transport, the NMTC/JAERI code and the MCNP code. The calculated transmutation energies were 510 MeV and 570 MeV for the spallation neutron method and the proton method, respectively, for an effective half life of 2 yr for ¹³⁷Cs.

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Hydrogen and oxygen gases radiolytically generated in a nitrate system by α-radiation were investigated. The following points were elucidated :
(1) Plutonium (IV) is involved in the G-value for hydrogen generation. The difference in G-values in ²⁴⁴Cm-HNO₃ solution and Pu(IV)-HNO₃ solution can be explained by the presence of Pu.
(2) No significant effect of the difference of oxidation states, Pu(IV)/Pu(VI), resulting from heating solution on the hydrogen generation was observed.
(3) Radiolytic evolution of hydrogen in plutonium nitrate solutions depends upon a height of the solution in a vessel, although this dependence is apparently less than that observed under γ-radiation.
(4) The effect of the solution height would be explained by a reaction between OH and H₂ because the effect can be approximately described by concentrations of undissociated HNO₃ and Pu(IV), which are thought to be primary scavengers of OH radicals.
(5) The G-value for oxygen generation in plutonium nitrate solutions was roughly in accord with that observed in the ²⁴⁴Cm/²¹⁰Po-HNO₃ system.

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The power-up test of the Nuclear Ship Mutsu was restarted in March 1990 and completed successfully in February 1991. The experimental voyages were carried out for about one year and all experiments were completed in February 1992. A comparison between the measured and design dose rate equivalents on board is described with showing a modified method in the shielding design. The measured values were obtained extensively in the cavity between the primary and secondary shields, in the double bottom, outside the secondary shield, and on the surface of the main coolant loop. The shielding design calculations were made with the most conservative geometries and material compositions within the allowed tolerance. In addition, a conservative model was adopted in case of performing the approximation due to the geometrical restriction of calculation code. The computational accuracies were evaluated based on various experimental analyses. The evaluated value was used as the design value. The shield structures were determined with a judgement that the real value does not exceed the design value. The adequacy of the judgement was confirmed by measurements on board. The measured dose rate equivalents in all positions on board satisfied the design criteria.

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A measuring method of core flow rate in the Advanced Boiling Water Reactor (ABWR), which equips Reactor Internal Pumps (RIPs) for reactor coolant recirculation, were evaluated by 1/5 scaled model tests and turbulent analyses. The conclusions obtained are as follows :
(1) An interrelationship among RIPs was so weak that ABWR core flow was derived accurately as the summation of the individual RIP flow rate, which was estimated from each pump O-H characteristics (Q: pump flow rate, H: differential pressure head along the RIP's upstream and downstream). Measurement accuracy of core flow was confirmed to meet the requirements for expected ABWR operating conditions. Furthermore, it appeared likely that the requirements should be satisfied even in hypothetical operating states, namely operation with one pump out of service.
(2) It was also confirmed that the turbulent flow analysis was useful to estimate the flow pattern near the RIPS.

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