A sailing-type wind farm which can move freely on oceans has been proposed in Japan since 2003. In this system the wind power is turned into hydrogen using an electrolyzer and then transported to end users. Since the sailing-type wind farm is a stand-alone system and the wind is intermittent, the efficiency of hydrogen production is quite low when the electrolyzer power is below a certain value. Additionally, the electrolyzer is inevitably shutdown frequently for lack of power. The frequent electrolyzer start-up actions can also decrease the efficiency of hydrogen production and shorten the electrolyzer's lifetime. In this paper, we applied a rechargeable battery and a proper control algorithm to the system to guarantee the hydrogen production efficiency and reduce the electrolyzer's start-up times. A simulation model of the whole system was developed and wind data was used to test the validity of the method. The simulation results showed that the proposed method can effectively improve the hydrogen productivity and reduce the start-up times.

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The calculation expressions for the emissivity spectra of powdery coal ashes have been derived by fitting the measured emissivity spectra to improve the accuracy with which the radiative heat transfer rate can be estimated with a calculation tool for pulverized coal-fired boilers. The expressions are defined in three bands. First, in band 1, which corresponds to wavelengths ranging from 2 to 4 micrometers, and in band 3, which corresponds to wavelengths ranging from 8 to 16 micrometers, the expressions are formulated into the linear function of the weight ratio of five chemical compositions, which correspond to silica, alumina, calcium oxide, ferric oxide and magnesia, and the temperature of the coal ash and the wavelengths. Second, the expression in band 2, with the wavelengths ranging from 4 to 8 micrometers, is formulated into the cubic function of the wavelength. The expression predicts the emissivity spectra of powdery coal ash, whose raw coals range from 0.6 to 2.4 (-) in a fuel ratio (=mass ratio of fixed carbon to volatile matter in coal), with 0.050 (-) to 0.170 (-) predictability. The proposed expression has widespread predictability of powdery coal ashes that range from 2 to 16 micrometers in wavelength.

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Fault Tree and Event Tree analysis is almost exclusively relied upon in the assessments of seismic Core Damage Frequency (CDF). In this approach, Direct Quantification of Fault tree using Monte Carlo simulation (DQFM) method, or simply called Monte Carlo (MC) method, and Binary Decision Diagram (BDD) method were introduced as alternatives for a traditional approximation method, namely Minimal Cut Set (MCS) method. However, there is still no agreement as to which method should be used in a risk assessment of seismic CDF, especially for uncertainty analysis. The purpose of this study is to examine the efficiencies of the three methods in uncertainty analysis as well as in point estimation so that the decision of selecting a proper method can be made effectively. The results show that the most efficient method would be BDD method in terms of accuracy and computational time. However, it will be discussed that BDD method is not always applicable to PSA models while MC method is so in theory. In turn, MC method was confirmed to agree with the exact solution obtained by BDD method, but it took a large amount of time, in particular for uncertainty analysis. On the other hand, it was shown that the approximation error of MCS method may not be as bad in uncertainty analysis as it is in point estimation. Based on these results and previous works, this paper will propose a scheme to select an appropriate analytical method for a seismic PSA study. Throughout this study, SECOM2-DQFM code was expanded to be able to utilize BDD method and to conduct uncertainty analysis with both MC and BDD method.

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In this study, the core geometry of HTR-10 was simulated using four different models including: (1) model 1 - an explicit double heterogeneous geometry, (2) model 2 - a mixing of UO₂ kernel and four layers in each TRISO particle into one, (3) model 3 - a mixing of 8,335 TRISO particles and the inner graphite matrix in each fuel ball into one, and (4) model 4 - a mixing of the outer graphite shell, 8,335 TRISO particles, and the inner graphite matrix in each fuel ball into one. The associated initial core computations were performed using the MCNP version 1.51 computer code. The experimental fuel loading height of 123 cm was employed for each model. The results revealed that the multiplication factors ranged from largest to smallest with model 1, model 2, model 3, and model 4. The neutron spectrum in the fuel region of each models varied from the hardest to the softest are model 1, model 2, model 3, and model 4 while the averaged neutron spectrum in fuel ball from hardest to softest are model 4, model 3, model 2, and model 1. In addition, the CPU execution times extended from longest to shortest with model 1, model 2, model 3, and model 4.

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A pipe-wall thinning phenomenon behind an orifice in a circular pipe is studied by measuring the mass transfer using the naphthalene sublimation method. An attention is placed on an asymmetric increase of local mass transfer behind the orifice, which is combined with the influence of swirling flow and the orifice bias. The present results indicate that the mass transfer is generally increased with an increase in swirl intensity. However, it becomes asymmetric by the combined effect of the orifice bias for the bias larger than 0.4% of the pipe diameter and the swirl intensity larger than approximately 0.2. The enhanced mass transfer is found on the shorter orifice side behind a biased orifice, and the maximum mass transfer is increased 40% in comparison with the case without swirl and orifice bias.

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In the joint research and development of treatment systems of cooling water for cutting asphalt pavement surface with our authors' group, the liquid-solid separation technology by flocculating and settling technology, and the flocculants for the use of systems were developed. In this paper, the developed flocculating and settling technology and the flocculants are discussed first. Next, the demonstration tests of decontamination technology on the contaminated water in swimming pools in an elementary school located at Motomiya City, Fukushima Prefecture had been conducted by use of the stationary purification system of contaminated water and the flocculants compounding with or without iron ferrocianide developed by the preliminary test. It was clarified from the results that ionized cesium (Cs) rarely exists in the stagnant water in pools, ponds, lakes and so on at the time when nine months have passed since Fukushima Dai-ichi nuclear power plant accidents. Further, it is necessary to use the flocculants compounding iron ferrocianide in the case where ionized Cs exists in water. From the above-mentioned results, the following problems were pointed out: One problem was cyanide dissolution in the purified water and the other one was the dissolution from the dehydration sludge. Finally, the high-performance mobile purification units of contaminated water which is capable for carrying with trucks have been developed, and the demonstration test was performed in Minami-soma City, Fukushima Prefecture to purify the contaminated water in a pond and generated by the high-pressure water washing in a Public Hall. From the test results, it was made clear that the dehydration sludge separated by liquid-solid settling of the contaminated water of around 1,000Bq/l became a high radiation dose of about 185,000Bq/l.

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In order to evaluate the heat loss and water temperature in a spent fuel pit (SFP) after the shutdown of its cooling systems, three dimensional thermal hydraulic behavior was calculated using the CFD software, FLUENT 6.3.26. In the calculations, the decay heat calculated by using the burn-up calculation software, ORIGEN 2.2, and the previously proposed correlation for evaporation heat fluxes from the water surface to air flow were used. Water temperatures were almost uniform in the SFP, and this indicated that one-regional calculation model might be able to obtain an average water temperature accurately. Comparisons of calculated results with measured temperatures showed that ORIGEN 2.2 gave good prediction of the decay heat for fuel assemblies with large decay heat which had been taken relatively recently from the shutdown reactor core. But ORIGEN 2.2 overestimated for longtime cooled fuel assemblies with small decay heat. The previously proposed evaporation heat flux correlation overestimated the temperature increasing rate of the SFP water.

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This paper describes simple flow rate monitoring using the Ultrasonic Velocity Profile method (UVP), for disturbed flow just after double elbows and large pipes with high Reynolds numbers. This method utilizes the linearity between the flow rate and the velocity at the pipe center under the following flow conditions. The Reynolds number is from 4.1×10⁶ to 5.1×10⁶, the dihedral angle of double elbows is 30° and the distance between double elbows is six diameters. The measuring position is set at two diameters distance downstream of the second elbow. The Computational Fluid Dynamics (CFD) simulations and experimental results indicate that it can monitor flow rate changes with an accuracy of less than 1% under these flow conditions.

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The steady-state turbulent heat transfer coefficients in a vertical circular Platinum (Pt) test tube for the flow velocities (u=4.22 to 21.45 m/s), the inlet liquid temperatures (T_in=308.20 to 311.77 K), the inlet pressures (P_in=834.04 to 910.23 kPa) and the increasing heat inputs (Q₀exp(t/τ), exponential periods, τ, of 6.04 to 32.13 s) were systematically measured by an experimental water loop comprised of a multistage canned-type circulation pump with high pump head. Measurements were made on Pt test tubes of 3, 6 and 9 mm inner diameters, 32.7, 69.6 and 49.6 mm heated lengths and 0.5, 0.4 and 0.3 mm thicknesses, respectively. Theoretical equations for turbulent heat transfer in circular tubes of 3, 6 and 9 mm in diameter and 492, 636 and 616 mm long were numerically solved for heating of water with heated sections of 3, 6 and 9 mm in diameter and 33, 70 and 50 mm long by using PHOENICS code under the same condition as the experimental one considering the temperature dependence of thermo-physical properties concerned. The surface heat flux, q, and the average surface temperature, T_s,av, on the circular tubes solved theoretically under the flow velocities, u, of 4.22 to 21.45 m/s were compared with the corresponding experimental values on heat flux, q, versus the temperature difference between average inner surface temperature and liquid bulk mean temperature, ΔT_L [=T_s,av-T_L, T_L=(T_in+T_out)/2], graph. The numerical solutions of q and ΔT_L are almost in good agreement with the corresponding experimental values of q and ΔT_L with the deviations less than ±10 % for the range of ΔT_L tested here. The numerical solutions of local surface temperature, (T_s)_z, and local average liquid temperature, (T_f,av)_z, are within ±10 % of the corresponding experimental data on (T_s)_z and (T_f,av)_z. The thickness of the conductive sub-layer, δ_CSL [=(Δr)_out/2], and the non-dimensional thickness of conductive sub-layer, y⁺_CSL [=(f_F/2)^0.5ρ_luδ_CSL/µ_l], for the turbulent heat transfer in various vertical tubes are clarified based on the numerical solutions. It was confirmed in this study that authors’ steady-state turbulent heat transfer correlation based on the experimental data [Hata and Noda, 2008] can not only describe the experimental data of steady-state turbulent heat transfer but also the numerical solutions within ±10 % difference for the wide ranges of tube inner diameters (3, 6 and 9 mm), temperature differences between average inner surface temperature and liquid bulk mean temperature (ΔT_L=8 to 145 K) and flow velocities (u=4.22 to 21.45 m/s).

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MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops that can supply 280 MW of electricity. The limiting conditions of operation defined in the safety regulations for MONJU given the allowed outage time were evaluated by a probabilistic safety assessment technique in our previous study. If a function failure is found in a control rod, certain measures are required by the safety regulations. In this case, if it is confirmed within 24 h that no other control rods are stuck, reactor operation is allowed to continue. To assess the validity of the 24 h allowable time in view of core damage risk, it is necessary to analyze the conditions to be changed when a stuck rod is discovered. Furthermore, to develop a method for evaluating the probability of a control rod insertion failure, it is necessary to re-estimate the frequency of core damage under control rod insertion failure conditions. This paper describes a method for this re-estimation. The probability of an insertion failure of one control rod has been calculated by applying Jeffrey's noninformative prior distribution by considering insertion times based on the results of a mock-up test. The necessary rod insertion numbers for the main reactor shutdown and backup shutdown systems were considered. The results showed that a completion time of 24 h gives a safety margin comparable to that of the Incremental Conditioned Core Damage Probability, that is, an acceptable risk threshold represented by the U.S. Nuclear Regulatory Commission RG 1.177. Thus, the timeframe defined in the present safety regulations was concluded to be appropriate.

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