For the complete nuclear fuel cycle, the development of a process for the co-conversion of Pu-U nitrate mixed solutions to mixed oxide powder has been performed along the line of non-proliferation policy of nuclear materials. A new co-conversion process using a microwave heating method has been developed and successfully demonstrated with good results using the test unit with a capacity of 2 kg MOX/d. Through the experiments and engineering test operations, several important data have been obtained concerning the feasibility of the test unit, powder characteristics and homogeneity of the product, and impurity pickups during denitration process. The results of these experimental operations show that the co-conversion process using a microwave heating method has many excellent advantages, such as good powder characteristics of the product, good homogeneity of Pu-U oxide, simplicity of the process, minimum liquid waste, no possibility of changing the Pu/U ratio and stable operability of the plant. Since August 1979, plutonium nitrate solution transported from the Tokai Reprocessing Plant has been converted to mixed oxide powder which has the Pu/U ratio=-1. The products have been processed to the ATR "FUGEN" reloading fuel. Based on the successful development of the co-conversion process, the microwave heating direct denitration facility with a 10 kg MOX/d capacity has been constructed adjacent to the reprocessing plant. This facility will come into hot operation by the fall of this year. For future development of the microwave heating method, a continuous direct denitration, a vitrification of high active liquid waste and a solidification of the plutonium-contaminated waste are investigated in Power Reactor and Nuclear Fuel Development Corp.

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The single failure tests with the ROSA-III were simulated BWR LOCA experiments by the scaled BWR test facility resulting from a 200% double-ended break at the recirculation pump suction line to evaluate the core cooling capability of a BWR ECCS under the single failure condition.
The experimental results showed that the loss of LPCS and one LPCI resulted in the highest PCT of 870 K of the single failure series tests, yet a core cooling capability by the ECCS was maintained. The REALP4/Mod 6 code was used to evaluate the predictive capability of the LOCA analysis code. The calculated results showed that the RELAP4/Mod 6 code was able to predict occurrences and sequence of major events anticipated to occur during a BWR LOCA correctly. However it was found that the code still needs to be improved in a CCFL model to better describe thermohydraulic behavior in the core.
The analyses presented in this paper are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a BWR during a LOCA.

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A computer code, which calculates the transients of heat flux from simulated nuclear fuel rods by using the transients of rod surface temperature and the heat conduction equation in the rod, was developed in order to investigate the heat transfer modes throughout the reflood phase in PWR-LOCA experiments. The code was applied to the Slab Core Reflood Tests which are part of the Large Scale Reflood Test Program at the Japan Atomic Energy Research Institute. For defining the heat transfer modes during reflood, it is important to obtain accurate heat flux from rod under a wide rod temperature change ranging higher than 1, 300 to 300 K and a rapid rod temperature change due to quench, which are principal features in heat transfer during reflood phase. Therefore, the effects of both temperature dependency on physical properties of rod and the axial heat conduction along rod on the heat flux calculation were first investigated. As the results, it was made clear that the temperature dependency on the physical properties should be taken into account and that the effect of axial heat conduction along the rod was negligible except in a very short length of rod at the quench front. The results calculated by the code for the Slab Core Tests when compared with the existing correlations could define the heat transfer modes clearly all through the reflood phase but the recommendations for further investigations were suggested.

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The present study investigates experimentally the characteristics of quench phenomena, top quench and bottom quench, in the upper part of the core during reflood phase in a PWR-LOCA. The characteristics of quench phenomena in the upper part of the core have not been predicted well even by a best estimate safety analysis code such as TRAC. The following are revealed from the experiment. (1) Quench phenomena in the upper part of the core are classified into three types ; (1) top quench of heated rods, (2) top quench of non-heated rods, and (3) bottom quench of heated rods, with respect to the relation of quench velocity vs. quench temperature. (2) Top quench of heated rods is influenced by the falling water film on the adjacent non-heated rod. (3) Top quench velocity of heated rod rises with an increase of quench temperature ; this is a tendency opposite to the prediction by the existing theory. (4) Top quench of non-heated rods has the same tendency as the existing theory.
The effects of system pressure and coolant flow rate are also investigated.

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In a series of irradiation examination on VHTR reference fuels, failure fractions of the coated particles have been measured in the acid-leaching method. All of the failure fractions of the coated particles irradiated in the form of compact were under an allowable limit designed for VHTR fuel. Comparing the failure fractions of the particles in the compact and in the loose form, the former fraction was about an order of magnitude as low as the latter, suggesting that the graphite matrix of the compact might have an effect on protection of coating layers. The metallography of the post-irradiation examination revealed irradiation-induced densification of the buffer layer, which caused debonding of this layer from inner PyC layer or occurrence of radial cracks in this layer. It was found that in the limited case the crack extended to SiC layer, then leading to total coating failure.

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Some metal iodides such as of Fe, Al, Zr and Te are known to cause stress corrosion cracking (SCC) of Zircaloy just as iodine itself does. Therefore 15 metal iodides were selected as corrodants, and SCC tests were carried out using the internal gas pressurization method.
The results showed that : (1) only those metal iodides which react thermodynamically with Zr to produce ZrI₄ cause SCC of Zircaloy-2; (2) when SCC occurs, the reaction rate between the iodide and Zr seems to be a main factor in determining the SCC susceptibility ; (3) gaseous ZrI₄ is the most corrosive agent ; and (4) some species of metal iodides, such as PbI, cause SCC of Zircaloy-2 more easily than I₂ vapor.
Scanning electron microscope (SEM) examination and electron probe microanalysis (EPMA) on the fracture surface of failed specimens revealed that ZrI, , formed as the reaction product between the metal iodides and Zr, might induce SCC of Zircaloy-2 rather than the iodides themselves.

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Mechanically blended powder of UO₂, PuO₂ and graphite and its compact were heated in a vacuum to have some informations on the formation of mixed carbide (U, Pu)C. It was experimentally confirmed that PuO₂ was reduced to sesquicarbide without forming a solid solution with UO₂ in the reduction process. Larger quantity of sesquicarbide was formed in the compacted sample than in the powder one in the intermediate products. Compacts of UO₂, PuO₂ and graphite were also heated at 1, 6654, 920 K in He, in order to investigate the effects of reduction temperature, He flow rate and blending time on the rate constant of reduction. The reaction proceeded from the surface toward the center of the compact. It was inferred that the rate-determining step was the reaction at the interfacial boundary between the reacted layer and the unreacted zone. The activation energy for the reduction was determined to be 375±20 kJ/mol. The Pu loss due to vaporization could be sufficiently suppressed, when compacted samples were reduced in He.

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