In order to clarify the mechanism behind the change in thermoelectric power (TEP) of Fe-Cr binary alloys and Fe-Cr-Ni ternary alloys due to Cr concentration, using these alloys with various Cr concentrations, dependence of TEP on Cr concentration was investigated. According to the Mott-Jones theory, TEP is largely related to electron density of states at the Fermi level. The electron density of states for the alloys in valence band was measured with X-ray photoelectron spectroscopy (XPS), and theoretically calculated TEP from the XPS spectrum was compared with experimentally measured TEP. Electron density of states in valence band of the both alloys changed due to Cr concentration, and the theoretically calculated TEP from XPS spectrum and the experimentally measured TEP similarly changed with Cr concentration. We think that the change in TEP with Cr concentration is due to the change in electron density of states with Cr concentration.

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A numerical analysis has been performed for three-dimensional developing turbulent flow in a rotating U-bend of strong curvature with rib-roughened walls by using an algebraic Reynolds stress model. In this calculation, an algebraic Reynolds stress model is adopted in order to predict preciously Reynolds stresses and boundary fitted-coordinate system is introduced as the method for coordinate transformation to set exact boundary conditions along complicated shape in rib-roughened walls. Calculated results of mean velocity and Reynolds stresses are compared with the experimental data in order to examine the validity of the presented numerical method and an algebraic Reynolds stress model. It has been pointed out as a characteristic features from the experimental result that positive rotation leads to a more uniform velocity distribution within bend, whereas, the high momentum fluid remains closer to the inner side with negative rotation. The present method can relatively predict such velocity profiles and reproduce the separated flow generated near the outer wall which is just located at downstream of curved duct. Besides, calculated results suggest clearly that the flow pattern of secondary flow is greatly influenced by the rotational direction and the flow pattern with negative rotation is similar to that without rotation. As for the comparison of Reynolds stresses, the present turbulent model relatively predicts the experimental data well except for the flow separated region which is located near the outlet cross section of curved duct.

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In a nuclear power plant, much knowledge on severe accidents has been acquired through PSA, and accident management (AM) guidelines are prepared by incorporating that knowledge. In PSA, it is necessary to evaluate the effectiveness of AM using the decision-making failure probability (DFP) of an emergency organization, operation failure probability of operators, success criteria of AM and reliability of AM equipment. However, to date there has been no suitable quantification method for PSA to obtain DFP.
In this study, we developed a new method for DFP quantification of an emergency organization using a cognitive analysis model, and tried to apply it to S2DC and TMLF sequence of a typical plant. As a result: (1) The methods enabled to DFP quantification appropriate to level 1.5PSA by choosing the suitable value of a basic failure probability and an error factor. (2) The DFPs of six AMs appeared to be in the range of 0.23 to 0.41 (screening method) and in the range of 0.10 to 0.19 (detailed method), and the DFP decreased about 50% as a result of sensitivity analysis of the conservative assumption. (3) The screening method was more conservative than the detailed method, and it was shown to satisfy the screening performance required by PSA.

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In order to investigate the conditions causing minute bubble emission boiling, critical heat flux experiments were conducted in the pool boiling system with different thermal capacities of heat transfer surfaces and method of heating.
Stable minute bubble emission boiling was observed for a 10 mm-thick copper cylinder heated by thermal radiation. The critical heat flux obtained was 6.0MW/mm². When the heat flux exceeded above approximately 3MW/mm², a large vapor bubble formed on the heat transfer surface, then was condensed immediately and dispersed into minute bubbles.
A 4 mm-thick silicon carbide heat transfer surface burnouted at 1.9MW/mm². When the heat transfer surface has a small thermal capacity such as the metal foil, commonly used in CHF experiments, a temporary loss of heat removal due to large bubble formation will cause a rapid temperature increase and will result in burnout.
Minute bubble emission boiling could occur stably when the time constant determined by the heating method and thermal capacity exceeds the time required for a large bubble to condense in the subcooled fluid.

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A correlation has been developed for rewetting quality in a fuel assembly under BWR (Boiling Water Reactor) transient conditions. The derivation is based on phenomenological considerations for the elementary processes in rewetting mechanism instead of empiricist approaches having been applied to previous correlations.
Two premises based on experimental observations are made: (1) The rewetting quality in steady state is in good agreement with the critical quality at the onset of boiling transition. (2) That in a transient state can be represented as the sum of a critical quality and a quality deviation corresponding to the time delay due to liquid film propagation.
As for the derivation, based on the above observations, the following hypotheses were made,
(1) Both dryout and rewetting are caused by the propagation/retrogradation of the liquid film front between two axially adjacent spacers.
(2) The maximum retrogradation length can be estimated by using mass and energy balance for liquid film.
(3) To evaluate delay in a transient, the propagation time required can basically be calculated with an empirical velociry correlation for the quench front.
The correlation has been qualified by comparisons with 30 transient post-BT tests for simulated 9×9 fuel assemblies, which leads to the applicability to 9×9 and its analogous assemblies.

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Tritium distribution in the carbon-based plasma-facing wall of JT-60U was examined by imaging plate (IP) technique and full combustion method. The highest tritium concentration was observed at the private region in the divertor. The tritium concentration of the divertor target tiles was lowest. Although significant amount of deposition were observed on the inner divertor tiles, there was no correlation between the deposition layer and the tritium retention in JT60U. The result of the triton deposition simulation using Orbit Following Monte-Carlo code was consistent with the tritium distribution obtained by the IP technique and the full combustion method. The tritium distribution can be explained by the energetic triton loss due to ripple loss and a slight modification due to high surface temperature of the divertor target tiles. According to the simulation, -30% of the tritium produced by D-D nuclear reaction in the JT-60U are lost and implanted into the first wall with high energy of up to 1 MeV.

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The Rankine-Hugoniot relation for a detonation wave has been numerically analyzed for hydrogen-oxygen mixture to evaluate its explosion pressure taking radical products at high temperature into consideration. The present study elucidates a scheme of convergence to the Jouguet point obtained by triple iterations over temperature, specific volume, and mole fractions of product gases. The total number of product gas atoms changes monotonously in the iteration over the mole fractions under the condition of constant temperature and specific volume. The final solution is, hence, obtained by the interpolation of the last two iteration points that extend the point where the total number of gas atoms is conserved. Every term of the Rankine-Hugoniot relation also changes monotonously in the iteration over the specific volume under constant temperature condition. So the final specific volume is determined in the same way as the mole fractions. In conclusion, the Jouget point with corresponding explosion pressure obtained dose not depend on the iteration steps or the truncation conditions.

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An experimental system has been developed for aiming at supporting the simulator training of diagnosis nuclear power plant anomaly, where an expert provides the instructions to a trainee by his/her own educational and experienced point of view. This system has several characteristics as follows; (1) the expert can provide the instruction through the local area network or the Internet, (2) the training task is the detection of the primary cause of plant anomaly, which is not including the plant operation but requires only thinking, and (3) the system can automatically detects the trainee's viewpoint and think aloud protocols and provide these two types of information to the expert in real time for supporting his/her instruction. A laboratory experiment using this system was conducted, where an expert, who had really engaged in plant operation as a chief operator, was employed as the instructor for novice students. As the result of analyzing the experimental data, it was found that the two types of information, trainee's viewpoints and his verbal protocols, have some potential of effectiveness for supporting the instructor to estimate the trainees diagnostic thinking process and provide the instruction.

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The use of transport/storage cask for spent fuel storage is considered to be rational and economical. Since the storage duration may continue for 40 years or so, the function of sealing radioactive materials in the casks must be reliable for long-term. Long-term containment test of full-scale spent fuel transport/storage cask models have been in progress since 1990 in CRIEPI, Japan. It has been 11 years since it started. The results so far demonstrate and confirm very reliable containment performance of the cask lid structure with metal gaskets. Using the test data it is predicted by Larson-Miller Parameter (LMP) method that the containment system will keep its integrity at least for 40 years.

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It was made clear by critical experiments (called JUPITER) that nuclear characteristics of large-sized fast reactor cores were quite different from those of small-sized cores. For instance, radial neutron flux distributions are significantly changed by perturbations, control rod reactivity interaction effects are large, etc. These phenomena are interpreted to be commonly caused by the spatial neutronic decoupling of large core. The changeability and instability of flux distribution, which might cause a power peaking and a flux tilting, is a new problem for the development of large core. The paper shows measured results of static decoupling characteristics, interprets them physically, and considers a reactor physics system of large core. The paper also investigates a nuclear core design method, in which the decoupling characteristics are taken into account. A neutronic stability is a new requirement for the nuclear design of large core, which the design policy is to promote and secure together with the performance and the safety. The more tightly coupled the core, i.e., the smaller the degree of decoupling, the better the neutronic stability.

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A plate-fin type recuperator is needed in a high thermal efficiency power generation system in HTGR coupled with closed cycle helium gas turbine, whose pressure and temperature are 6MPa and 850°C, respectively. This paper describes the results of a series of FEM stress analyses carried out to develop a new simplified structural analysis procedure for this plate-fin type heat exchanger. The procedure consists of a combination of independent analyses under pressure load, steady thermal load and transient thermal load. Then, the stress can be evaluated by a summation of these stresses. Structural characteristics of fine plate-fin structure, in particular, two bending-stress inducing mechanisms are revealed. A proposed simplified analysis method using an equivalent stiffness is very useful to obtain the stress under the steady temperature distribution load. The stress analysis results under the transient thermal loads are also given. The combined stresses are calculated and evaluated to the primary and secondary allowable stress limits.

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Within our knowledge, there are some events which short period boiling transition (BT) increases fuel cladding temperature slightly on BWR's anticipated operational occurrences. However, there are little fuel performance data on short term BT. So, we tested to verify the fuel performance on that short period BT using Halden BWR. And also, we tested fuel cladding deformation behavior. At the HBWR tests, in which prior-β phase was observed at PIE, cladding deformation and embrittlement due to oxidation and hydride occurred. On the other hand, at the HBWR tests, in which prior-β phase was not observed at PIE, cladding deformation and recovery of irradiation hardening only occurred. From the fuel cladding deformation tests, the deformation was estimated to depend on temperature and holding time.

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For the Seminars, the authors had prepared a lecture note and a textbook for the continuous energy Monte Carlo method. In the Shielding Safety Analysis and High Energy Calculation Seminars, proposals were presented on a means of evaluating the weight window method. For all the Monte Carlo Calculation Seminars, a neutron cross section library covering 340 nuclides with 293K had been compiled from JENDL-3.2. Also for the Calculation Seminars, inputs of benchmark problems had been prepared for checking the calculation methods.

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Monte Carlo Working Group of Special Committee on Nuclear Code Evaluation at JAERI investigated present status of application of Monte Carlo calculation for large nuclear facilities in Japan. Application of Monte Carlo method has already been popular even to large facilities, but the calculated results are not compared with measured data in most of those applications. The authors believe that accuracy and precision of the method should be examined through comparisons with measured data, and that databases of measured data at experimental or commercial facilities should be developed for further comparison with the analysis results.

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