Compilation of the second version of Japanese Evaluated Nuclear Data Library (JENDL-2) has been completed. The aim of JENDL-2 is the applications to fast, thermal and fusion reactors, and shielding problems. The evaluated data for 181 nuclides were stored in the ENDF/B-IV format. Various benchmark tests have been performed, and applicability of JENDL-2 was confirmed for the fast and thermal reactor calculations and for the shielding problems. In this review, evaluation work for JENDL-2 is briefly described, and the results of the benchmark tests are summarized.

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Experiments were carried out on DNB (Departure from Nucleate Boiling) heat flux for both upflow and downflow in a rectangular vertical channel simulating a subchannel in the fuel element of the research reactor JRR-3, which is scheduled to be upgraded at 20 MWt with 20% low enriched uranium (LEU) fuel. The experiments were carried out for the conditions of pressure and velocity which are important for the safety design of the JRR-3. With the investigation of the data of the present experiment along with already existing data for both rectangular channels and other channels, a scheme of DNB heat flux correlations was obtained for upflow and downflow. With the all available data, the errors of the cor-relations adopted in this scheme were so evaluated as to utilize these results in the safety analysis of the JRR-3. This scheme includes a new correlation of DNB heat flux for upflow and the identification of a region of high coolant flow rate where no remarkable differences in the DNB heat flux are observed between upflow and downflow. This scheme is considered to be applicable not only to the rectangular channels of the JRR-3 but also to other channels.

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The objective of the present study is to evaluate the temperature rise due to gas release behind local blockage in LMFBR fuel subassemblies.
The experiments were conducted using six sets of electrically heated 37-, 61- and 91-pin bundles. The central or edge subchannels of each test section were blocked by a non-heat generating blockage of stainless steel plate. The effect of operational and geometrical con-ditions on the coolability of the blocked bundle were investigated from the test results. The examined factors are gas release rate, sodium flow rate, spacer type and blockage location.
The high coolant flow more than 2 m/s in the velocity at the blocked plane, forced the released gas to be accumulated within the recirculation region resulting in marked tempera-ture increase. The dependency of the temperature rise on the gas release rate was classified into two stages; (1) the temperature increases with the gas release rate, (2) after reaching a peak value, the temperature gradually decreases with the gas release rate. From these conclusions empirical correlations were derived to estimate the temperature rise under the condition of blockage with gas release. It was deduced that fission gas release in an LMFBR fuel subassembly with a local blockage has a potential to cause a limited pin-to-pin failure propagation in the recirculation region.

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An experimental study of Cylindrical Core Test Facility (CCTF) tests was performed to investigate the reflood behavior in PWR with upper plenum injection (UPI) type ECCS. The major conclusions are as follows :
(1) The ECC water injected into the upper plenum fell down to core smoothly and flowed from the core to the downcomer.
(2) This falling water flow rate was much higher than the predicted with a correlation developed on the base of the small-scale experimental results, i. e. Bankoff's CCFL cor-relation. This is considered to be because the steam flows up from the core in a limited cross section and separately the water downflow region is formed.
(3) The steam from the core was condensed by the ECC water in the upper plenum. The hydraulic behavior in the primary system was significantly different if the complete condensation was attained or not.
(4) For a UPI test with no failure low pressure coolant injection (LPCI) pump simulation, the complete condensation was attained, the water stagnated in hot legs and steam generators, and almost no fluid flowed through the primary loops. For a UPI test with single failure LPCI pump simulation, the complete condensation was not attained, and the steam with entraining the water flowed through the primary loops.

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Neutron diffraction studies have been made on a powdered NaUO₃ sample. It has been shown that an octahedron composed of six oxygen atoms around pentavalent uranium in this compound is distorted and four of six oxygen atoms are much closer to an uranium atom (U-O distance of 2.15 Å) than the other two (U-O distance of 2.24 Å). A unit of the UO₂⁺ ion does not exist in the crystal, though the pentavalent uranium has been expected to be in the form of UO₂⁺ as in the case of UO₂²⁺ of a hexavalent uranium ion. A neutron diffraction pattern at temperature of liquid helium has shown no appreciable difference from that at tem-perature of liquid nitrogen.

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The partitioning process has been developed in Japan Atomic Energy Research Institute for the removal of long-lived radionuclides from high-level waste (HLW), and was tested with actual HLW. But, the separation behavior of neptunium was not examined in the de-veloped process.
In the present study, the behavior of neptunium in the partitioning process was examined with actual HLW by means of the analysis of 237Np. At the first step of the partitioning process, 86% of neptunium was not extracted with TBP. It was suggested that most of neptunium was pentavalent in feed solution. At the following step, more than 90% of neptunium was extracted with DIDPA and was not stripped with 4 M nitric acid. It was proved in tracer experiments that most of neptunium was pentavalent before DIDPA extrac-tion and Np(V) could be extracted with DIDPA by a slow rate. Neptunium which was ex-tracted from pentavalent state was hardly stripped with nitric acid, which indicated that the extraction of Np(V) was an irreversible process. It was considered that the disproportiona-tion reaction of Np(V) was concerned in the extraction.

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A method is proposed to estimate parameters for dose calculation from inhaled activity distribution in the human body.
Inhalation of ⁶⁰Co aerosols is simulated by the metabolic model proposed in ICRP Publi-cation 30. The ranges of parameters for dose calculation are as follows : particle size 1 to 20 μm ; inhalation class Y; time after inhalation 1 to 90 d. The ratio of radionuclide retentions in the lower to upper bodies, divided by the diaphragm, is used as the index of activity distribution in the body. The results of computer simulations show that the retained activity ratio is valid to estimate intervals of possible values of particle size and time after inhalation.

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An incineration method for plutonium recovery from α contaminated organic compounds in a flow of controlled oxygen gas is stated. The species of such thermal decomposition products as hydrocarbons, free carbon, carbon monoxide and hydrogen were determined by mass spectrography. The mixture of the products which are the source of tar or soot was converted to CO₂ and H₂O in contact with copper oxide catalyst without flaming.
This incineration method is composed of two stages. The first stage is the decomposi-tion of organic compounds in the streams of gas mixtures containing oxygen in low ratios. The second stage is the incineration of the decomposition products by catalytic reaction in the streams of gas with higher oxygen ratios. Plutonium was recovered as the form of plutonium dioxide from the incineration residues of the first stage.
The behavior of oil was examined as a representative of liquid organic compounds. It was found to evaporate below ca. 500°C, but was completely incinerated by the catalytic reaction with copper oxide catalyst in the flow of gas with controlled oxygen amount and was changed to CO₂ an H₂O.

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