Japanese Evaluated Nuclear Data Library(JENDL)has been developed by Japanese Nuclear Data Committee and JAERI Nuclear Data Center. The first version of JENDL, JENDL-1, which was released in 1977, has been compiled to use mainly for fast, reactor development. The second version, JENDL-2 has aimed for use in wider fields including fusion researches, shielding calculations, etc., and its compilation is in progress. As our further step we now have a plan for the compilation of JENDL-3. The outline of JENDL-3 is described in terms of its features and the user's needs as well as the history of JENDL.

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A series of simulation tests were performed with H₂, D₂ and CH₄ to study the characteristics of detritiation system composed of catalytic oxidizer packed with precious-metal-alumina catalyst, dryer with molecular sieve 5A, and so on. The amount of water vapor adsorbed on the catalyst came up to more than 10% of that on the molecular sieve 5A. Conversion ratio for D₂ as high as 99.9% were obtained in the following conditions: space velocity: 1, 800-5, 100h^-1, catalyst bed temperature: 23-100°C. The effect of the adsorbed water on the conversion ratio was not detected. The ratio for CH₄ under the constant space velocity, i.e. 2, 000h^-1, exceeded 99% at the temperature higher than 350°C. Maximum reduction ratio of the water vapor by the dryer was 99.9998%. Enclosure cleanup ratios in the high concentration range of D₂ and CH₄ were demonstrated to approach exponential dilution.
It is noticed that careful consideration on the effects of the adsorbed water should be payed for evaluating the catalyst performance, and soaking effect on the inner surface of the enclosure could not be neglected in the low concentration range of hydrogen.

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A study on post-accident core coolability of LWR is in progress based on the practical fuel failure behavior experienced in NSRR, PBF, PNS and others. The fuel failure behavior at LOCA, RIA and PCM conditions are reviewed, and seven types of fuel failure modes are extracted as the basic failure mechanism at accident conditions. Those are; (1) cladding melt or brittle failure, (2) molten UO₂ failure, (3) high temperature cladding burst, (4) low temperature cladding burst, (5) swelling of molten UO₂ and (6) cracks of embrittled cladding for irradiated fuel rods, and (7) TMI-2 core failure. The post-accident core coolability at each failure modes are discussed. The fuel failures caused actual flow blockage problems are only the types of (2), (5) and (7). A characteristic which is common among these three types is that, the fuel rods are in the conditions violated the present safety criteria for accidents, and UO₂ pellets are in melting or near melting hot conditions when the fuel rods failed.

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For the purpose of the dose evaluation of crews of a nuclear ship, the statistics of wind speed and direction relative to the ship is discussed, using wind data which are reported from ships cruising sea around Japan Island.
The analysis on the data shows that the occurrence frequency of wind speed can be fitted with the γ-distribution having parameter p around 3 and wind direction frequency can be treated as a uniform distribution.
Using these distributions and taking the ship speed u, and the long-term mean speed of natural wind u as constant parameters, frequency distribution of wind speed and direction relative to the ship was calculated and statistical quantities necessary for dose evaluation were obtained in the way similar to the procedure for reactor sites on land.
The 97% value of wind speed u₉₇, which should be used in the dose evaluation for accidental releases may give conservative doses, if it is evaluated as follows, u₉₇=0.64u_s in the cases u_s>u, and u₉₇=0.86u in the cases u_s<u including u_s=0.

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