A new small reactor concept named Package-Reactor has been developed through a joint research of Hitachi, Ltd., and Mitsubishi Heavy Industries. Several key design of its nuclear steam supply system have been investigated, taking into account both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) technologies. The PackageReactor is a stand-alone energy supply system, and is designed to attain high reliability, high safety, good maintainability, good operability and low construction cost. To achieve these aims, the reactor adopts natural-circulation core cooling systems. The reactor has no active devices inside its high pressure boundary. Combining a turbine electric power generation and biomass refining, which is supported by JGC Corporation or chemical heat pipe systems attains a perfect base load operation. The whole system is simple and small to be easily constructed with a very short period even at remote regions with poor infrastructures. The Package-Reactor is an innovative nuclear power plant concept to pioneer and develop new markets of the nuclear power business.

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MOX co-deposition process, which is one of the main parts of oxide electrowinning process, was studied by both sides of examination and analysis. Parameter tests using U and Pu were performed as basic examination, and fundamental data of the process such as polarization property of molten salt including U and Pu, current efficiency of electrolysis and PuO₂ concentration in recovered deposit, etc. were acquired. In addition, to analyze the process behavior, especially polarization property at electrolysis, a reaction model considering Pu⁴⁺, Pu³⁺, UO₂²⁺, UO₂⁺, PuO₂²⁺ and PuO₂⁺ in molten NaCl-2CsCl salt was developed. The validity of this model was confirmed by comparison of the experimental and analytical results. Using this model behavior of chemical species in the process was studied and control factors of MOX co-deposition process were discussed.

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Primary Water Stress Corrosion Cracking (PWSCC) which occurs on Nickel based alloy (Alloy 600) is a worldwide concern since early 1980's. Recently several significant degradations that originate from PWSCC in the reactor coolant pressure boundary (RCPB)components have been observed at U.S. PWR plants (e.g. Oconee-3, Davis Besse). The United States Nuclear Regulation Commission (NRC) has issued generic communications to address this problem and, in response to the Davis Besse event in 2002, gave the inspection order EA-03-009 for the PWR licensees to implement the inspection of the reactor vessel heads depending upon the effective degradation years. As well, in Japan, PWSCC is considered one of the safety issues, in particular, for aged nuclear power plants and actually, some plants have experienced PWSCC on RCPB components. In the present study, we analyzed the U.S. experience with Alloy 600 degradation by reviewing the licensee event reports from 1999 to 2005 and examined the trend of them mainly focusing on affected components, characteristics of cracking and inspection approaches for detecting the PWSCC. This study indicates that PWSCC is found to be occurred on the RCPB components exposed to the environment with high temperature such as the reactor vessel head, and has the tendency to happen for specific manufactures and material according to the RCPB components. As well, it is shown that for several components, the non-destructive examination is generally needed to detect and/or confirm the PWSCC after the visual inspection and different repair techniques are applied depending on the components affected.

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The Japan Atomic Energy Agency (JAEA) was entrusted "Development of Nuclear Heat Utilization Technology" by Ministry of Education, Culture, Sports, Science and Technology. In this development, the JAEA investigated the system integration technology to couple the hydrogen production system by steam reforming with the High Temperature Engineering Test Reactor (HTTR). Prior to the construction of the hydrogen production system coupling with the HTTR, a dynamic analysis code had to be developed to evaluate the system transient behaviour of the hydrogen production system because there are no examples of chemical facilities coupled with nuclear reactor in the world. This report describes the evaluation of the hydrogen production system coupling with HTTR using analysis code, N-HYPAC, which can estimate transient behaviour of the hydrogen production system by steam reforming. The results of this investigation provide that the influence of the thermal disturbance caused by the hydrogen production system on the HTTR can be estimated well.

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The probabilistic safety assessment (PSA) results can be used for decision making in various areas including risk informed regulation and safety management of nuclear plants. Since there exist uncertainties in PSA results, uncertainty evaluation is one of the most important issues in PSA. This paper describes an approach to evaluation of uncertainties in source terms. We proposed a procedure of uncertainty evaluation for source terms by the use of severe accident analysis code THALES2 and applied this procedure to the evaluation of source terms for BWR-5/Mark-II plant. Source terms analysis was performed with THALES2 for containment overpressure failure scenarios involving core damage. From this analysis, uncertainty information of source terms, such as the timing of radionuclides release to the environment and released fraction of radionuclides were estimated. As well, the dominant parameters for individual scenarios were identified: containment failure pressure, in-vessel and/or ex-vessel radionuclides release rate, deposition of aerosols in the containment and reactor building.

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The Very High Temperature Reactor (VHTR) is expected to be the best energy source for the hydrogen production. This system will handle a large amount of hydrogen which is combustible gas. It is one of the most important subjects in design with this system to assure the safety of the reactor system against the damage due to fire and the explosion accident.
This analysis provides with quantitative information about correlation between the combustible gas movement distance and various parameters which influence the gas dispersion. Based on these analytical results, we propose one of the most suitable method which is able to evaluate damage of the nuclear plant by blast overpressure. This method will be used efficiently for the safety evaluation for the future VHTR hydrogen production system.

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A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) based on the achievement of gas turbine high temperature reactor design has been carried out in Japan Atomic Energy Agency. Safety design philosophy of the GTHTR300C to keep hydrogen economy and to attract a lot of interest from non nuclear industries is discussed. The hydrogen production system which is coupled to the secondary helium loop of the intermediate heat exchanger installed upstream of the gas turbine system shall be designed as a non nuclear class system. General nuclear safety shall be ensured by the items installed in the reactor system. Functions of the secondary helium loop which are primary helium cooling, pressure control and purification of the secondary helium are required to continue normal operation. Means to maintain these functions are proposed by using equipment of the reactor system and the gas turbine system without the hydrogen production system so that the power generation can continue independently of operational state of the hydrogen production system. Means of protection against external event of flammable and/or toxic gas release are also considered.

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In the application of the probabilistic safety assessment (PSA) to a reprocessing plant, target accidents for the assessment should be identified first. The HAZOP method developed for a chemical plant was used to identify abnormal candidate events systematically and comprehensively. Among the many abnormal events thus identified, several typical accidents with high latent consequences were selected for the target accidents of PSA. Among them the PSA for complete loss of scavenging air to prevent explosion of radiolyzed hydrogen in the reservoir of concentrated PuNO₃ solution is taken up. Its occurrence frequency, uncertainty of the frequency and also relative risk-importance of each apparatus and operator's action composing its prevention were evaluated using the method originally developed for the PSA of a nuclear power plant. The frequency was calculated as 8×10^-6/yr. The results showed that the design of the scavenging system of the Rokkasho plant is adequate to prevent the accumulation of hydrogen. In addition, it was shown that even the present results with somewhat large uncertainties could be applied for practical use of risk-informed management of the reprocessing plant due to its low latent consequence compared to a nuclear power plant.

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Corrosion behavior of structural materials for thermo-chemical and electrolytic hydrogen production cycle was investigated in liquid and gaseous sulfuric acid in the temperature range of 200-500°C. The cycle is one of the hydrogen production methods using sulfuric acid and the maximum temperature through the processes is about 500°C. In this study, corrosion tests of candidate structural materials for equipment of the hydrogen production plant were performed at the conditions each equipment will be used. The concentration of sulfuric acid was 95 mass% in all experiments and maximum test duration was 500h. Only high Si cast iron had good corrosion resistance in the boiling sulfuric acid, whereas high Si cast iron and Hastelloy C276 had good corrosion resistance in the sulfurous acid gas atmosphere (vaporized sulfuric acid or mixture of sulfur dioxide and water vapor). Furthermore, post test analysis by optical microscope and SEMEDX were performed.

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Uranium cost in the annual collection of 1, 200t-U from seawater was evaluated by using the recovery system of braid type adsorbents synthesized by radiation-induced graft polymerization. The total cost was calculated by summating those in the processes of adsorbent production, uranium recovery, and elution and purification. When the adsorbent performance increased from 2g-U/kg-adsorbent (ad) to 6g-U/kg-ad, the cost of each process decreased in the same way. The increment of adsorbent durability of 6 times to 60 times reduced the process cost of adsorbent production especially. In the current state of 2g-U/kg-ad and 6 times usage of adsorbent, the uranium from seawater cost 90, 000yen/kg-U. The uranium cost becomes 25, 000yen/kg-U in the promising performance of 4g-U/kg-ad and 18 times usage of adsorbent.

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In nuclear power plant decommissioning, numerous metallic equipment, such as vessels, pipes and valves, is generated in the form of activated or contaminated waste. Such radioactive waste is expected to reduce its radioactivity via decontamination to become non-active waste.
Authors have developed a new chemical decontamination process for steel, especially carbon steel waste. The decontamination agent is formic acid, which easily decomposes into carbon dioxide and water. The optimum condition for carbon steel decontamination was proved to be 1.1mol·dm^-3 of formic acid at 80°C. In the test for a heater tube under these conditions, the activity was reduced to below 0.4Bq·g^-1 as ⁶⁰Co after decontamination.
Organic carbon in decontamination solution could be decomposed to lower than 5mg·dm^-3 (0.0004mol·dm^-3 of formic acid) with hydrogen peroxide. The iron residue could be converted to iron oxide by evaporation and drying. Subsequently, the iron oxide was solidified by cement, with the mixing of a maximum of 120kg per 200-l drum.
Consequently, the newly developed decontamination process using formic acid is concluded to be applicable for the clearance of carbon steel waste.

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Radon was discovered in the beginning of the twentieth century. Its recoil ranges and diffusion coefficients are so small in common minerals that the radon amount emanated from them has been thought to be very small. However the measured radon amount actually emanated was noticeably larger than the predicted value. Therefore, over the years, many research studies have been done to explain this mismatch. Also many measurements of radon emanation coefficients of various materials and radon concentrations in the atmosphere have been made by researchers worldwide. These activities have been done primarily because of the potential health effects brought about by radiation exposure from radon. On the other hand, the extent of radon recoil ranges in porous materials and radon diffusion lengths near solid surfaces is commensurate with various representative parameters such as pore sizes and solid surface geometry. This means that radon emanation phenomena would be applicable to study the physical properties of porous materials and solid surfaces now and in the future. This paper overviews the above mentioned fields of study.

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