The design and construction codes for nuclear power plants worldwide have been established based on ASME Sec. III. Since the requirements of fatigue evaluation were first introduced in the ASME codes in 1963, the fatigue evaluation methods have been used with few changes. In 2006, the world's first consensus codes for the environmental fatigue evaluation method have been established in JSME. The codes incorporate the results of international research projects mainly led by Japan on the environmental fatigue, that is, the reduced fatigue life of components in a high-temperature reactor coolant environment. This is a good example of international collaboration involving code engineers with expertise in the field of codes and standards. In these projects, the Japanese team contributed to the establishment of the world-leading codes for environmental fatigue evaluation. This paper describes the activities leading to the establishment of the codes for environmental fatigue evaluation, development of the evaluation method, roles of code engineers and remaining technical issues to be addressed.

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  An advanced microchannel heat exchanger (MCHE) with S-shaped fins can realize high heat transfer performance and reduced pressure drop performance. Simulation calculations by the three-dimensional computational fluid dynamics (3D-CFD) code were examined for a MCHE with S-shaped fins using carbon dioxide (CO₂) as the working fluid in the wide range of inlet pressure and temperature conditions, which are from subcritical conditions to supercritical conditions. From the numerical results, a Nusselt number correlation was obtained using multivariable analysis. The standard deviation between the correlation and numerical results was 6.1%. A MCHE test piece with the same flow channel configuration as that for numerical simulation was manufactured, and experimental verification of the thermal-hydraulic performance of the test piece was examined. The heat transfer performance of the test piece was calculated using the Nusselt number correlation and was compared with experimental results. Numerical heat transfer performance showed good coincidence with a deviation of 2.5%. The conditions used in numerical simulation include the conditions of a recuperator and an intermediate heat exchanger of CO₂ gas turbine cycle. Using the obtained Nusselt number correlation, designing these heat exchangers becomes easier.

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  It is important to understand the behavior of fission products (FPs) for the evaluation of fuel performance. For example, in high-burnup oxide fuels, some FPs dissolve in the fuel matrix and others form oxide or metallic inclusions, which would affect the physical and chemical properties of the fuels. Here, we investigated the thermal conductivity (λ) of oxide inclusions; in particular, we focused on Cs-Mo-O and (Sr or Ba)-Mo-O ternary systems. The λ value of Cs₂MoO₄ is quite low (around 0.6 Wm⁻¹ K⁻¹ at 300 K) compared with that of UO₂ (around 8.5 Wm⁻¹ K⁻¹ at 300 K). In addition, we found that the λ value of (Sr or Ba)MoO₃ is approximately 10 times higher than that of (Sr or Ba)MoO₄. This high λ value of (Sr or Ba)MoO₃ is due to not only a high electronic contribution but also an intrinsically high lattice thermal conductivity (λ_lat). This high λ_lat could be explained using the general lattice thermal conductivity theory; that is, a strong interatomic bonding within a simple crystal structure is realized in (Sr or Ba)MoO₃, leading to an exceptionally high λ_lat compared with that of (Sr or Ba)MoO₄.

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  On the basis of the background of the global warming and energy resource problems, the programs for utilization of nuclear power are now being revised upward around the world. This situation leads to expansion in utilization of uranium-plutonium mixed oxide (MOX) fuels in existing thermal reactors. The MOX fuels include larger amounts of heavy actinide nuclides as the initial components than conventional uranium fuels. This means that the decay heat of spent fuels becomes much larger than that of uranium fuels. In the present study, prediction accuracy is evaluated for actinide decay heat in one of the typical cases of spent MOX fuels in light water reactors on the basis of covariance data in JENDL-3.3. There are few studies on burnup sensitivity analyses focused on the actinide decay heat of spent fuels; the present study provides a good starting point for future works. The present results show that the errors of actinide decay heat are 35.1±0.9 kW/t (2.7%), 3.27±0.12 kW/t (3.7%), and 2.37±0.05 kW/t (1.9%) at a 68% confidence level in the cases of MOX fuels of 0.1, 10, and 50 year cooling after reactor shutdown, respectively. In order to improve the prediction accuracy of the MOX decay heat, it is important to reduce the uncertainty of the cross section of actinide nuclides including the capture reaction in Pu-238, -239, -242, Am-241, -243, and the fission in Pu-241 followed by capture in Cm-244.

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  A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility.

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  A special committee on “Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)” was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for probabilistic safety assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the useful information related to establishing quantitative performance objectives and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the analysis method of consequences for postulated accidents with potentially large consequences in NFFs, e.g. events of criticality, leakage of molten glass, hydrogen explosion, boiling of radioactive solution and fire (including rapid decomposition of TBP complexes), resulting in release of radioactive materials to the environment. The results of the research were summarized in a series of six reports. This report aims to provide common backgrounds of the events studied in order to promote the understanding of the other five technical reports and shows overviews of abnormal events postulated in a reprocessing plant and their features.

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  A special committee on “Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)” was organized by the Atomic Energy Society of Japan under the entrustment of Japan Atomic Energy Agency for research on the state-of-the-art consequence analysis method for Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the basic useful information related to the establishment of the quantitative performance requirement and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on accidents with more severe consequences than design basis, such as events of criticality, explosion, fire, and boiling of radioactive solution postulated in NFFs resulting in the release of radioactive materials into the environment. The research results are summarized in this technical report about basic experimental data related to key physical and chemical phenomena postulated in a boiling event of a radioactive solution storage tank caused by the loss of the cooling function.

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  A special committee on “Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)” was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the useful information related to the establishment of quantitative performance objectives and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the analysis method of consequences for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution, and fire (including rapid decomposition of TBP complexes), resulting in the release of radioactive materials into the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this technical report, the research results about basic experimental data related to the consequence of the radiolytically generated hydrogen gas explosion postulated in the radioactive solution reserve tank caused by the loss of dilution air supply were summarized.

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  A special committee on “Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)” was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the useful information related to the establishment of quantitative performance objectives and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the analysis method of consequences for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution, and fire (including rapid decomposition of TBP complexes), resulting in the release of radio active materials into the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this technical report, the research results about basic experimental data and the method for safety evaluation of fire and explosion incidents were summarized.

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  A special committee of “Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)” was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for the Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objectives of this research are to obtain information useful for establishing quantitative performance objectives and to demonstrate risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the consequence analysis method for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution and fire (including the rapid decomposition of TBP complexes), resulting in the release of radioactive materials to the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this report, the evaluation methods of criticality accident, such as simplified methods, one-point reactor kinetics codes and quasi-static method, were investigated and their features were summarized to provide information useful for the safety evaluation of NFFs. In addition, several trial evaluations were performed for a hypothetical scenario of criticality accident using the investigated methods, and their results were compared. The release fraction of volatile fission products in a criticality accident was also investigated.

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