Flooding in three piping systems containing multiple elbows and an orifice was experimentally investigated using a 51mm glass pipe and an air/water system. The vertical-to-horizontal geometry with multiple horizontal sections connected by 90°elbows yielded the most limiting flooding gas velocities due to the prolonged formation of the hydraulic jump caused by the added hydraulic resistance of the multiple elbows. Inclining one of the piping sections downward to 45°from the horizontal or to the vertical in the other two geometries eliminated the hydraulic-jump-induced flooding mechanism and increased the flooding gas velocities for the system at high liquid flow rates. The orifice placed in the horizontal section was seen to generally lower flooding gas velocities, with the greatest effect seen for the smallest orifice tested. A new method based on a superposition principle was also proposed for prediction of flooding in complex piping systems.

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The radial distributions of average velocity and turbulent velocity of the liquid phase are predicted based on the correlations of the bubble dispersion coefficient, turbulent diffusivity of liquid and the mixing length of two-phase flow previously derived by the authors. The radial distribution of average liquid velocity was predicted using basic equations of mass and momentum conservations for both phases. The prediction generally agrees with experimental data, which show a flattened radial distribution compared with the single-phase flow at the same liquid volume flux. The radial distribution of turbulent velocity was predicted using the basic equation of turbulent velocity for the case where turbulent velocity in two-phase flow is larger than single-phase flow turbulent velocity for the same liquid flux. The prediction successfully predicted the flattened distribution of turbulent velocity. Detailed consideration was given to the contributions of generation by velocity gradient, dissipation, diffusion and generation by a bubble to the turbulent velocity distribution, and it was shown that the turbulence generation by a bubble is predominant in the central region of a bubbly two-phase flow.

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In order to provide basic information on the multiphase flow encountered in hypothetical severe accidents of nuclear power plants, the volumetric fractions in air-water-particle three-phase bubbly flows were measured in vertical straight pipes of three different inside diameters: 20.9, 30.8 and 50.4mm. Their heights were about 10m. The mean diameters of spherical particles used were 1.15, 2.56 and 4.16mm. Their densities ranged from 2270 to 2400kg/m³. Experimental data on the volumetric fractions are offered, and effects of the volumetric flux, pipe diameter and particle diameter on the volumetric fractions are discussed. Their empirical nondimensional correlations are proposed as a universal expression in terms of dimensionless parameters related to geometrical and dynamical similarity, which are based on the multiplier method used for the estimation of frictional pressure drop. The estimation results agreed well with the experimental data.

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The concept of critical heat flux (CHF) is discussed based on the mechanism that the CHF is caused by the dryout of a liquid layer formed on a heating surface. It is suggested that a liquid macrolayer is formed due to the coalescence of bubbles for most boiling systems, and that the dryout of the macrolayer is controlled by the hydrodynamic behavior of coalesced bubbles on the macrolayer. Based on these considerations, a new CHF model is proposed for saturated pool boiling at higher pressures. The idea of this model comes from a close examination of the measured diameters of various bubbles and the photographic records obtained by Semeria (1963) for water boiling under higher pressures. In the model, a liquid macrolayer is formed due to coalescence of the secondary bubbles formed from the primary bubbles. The detachment of the tertiary bubbles formed from the secondary bubbles determines the frequency of the liquid macrolayer formation. The CHF occurs when the macrolayer is dried out before the departure of the tertiary bubbles from the heating surface. One of the formulations of the model gives the well-known Kutateladze or Zuber correlation for CHF in saturated pool boiling.

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Many heat-exchange tubes in a shell-and-tube-type heat exchanger are oscillated by means of the flow . of fluid through the heat exchanger. This flow is often cross flow and gas-liquid two-phase flow. Here, a brief review of studies on two-phase cross-flow-induced vibration is presented, and a summary of our work in this area provides a design methodology. Three types of vibration have been studied for tubes in two-phase cross-flow conditions: First, resonance of a tube due to vortex shedding is important primarily in single-phase flow, but also has been observed in homogeneous flow and even in two-phase flow. However, this vibration disappears in the slug flow or froth flow regions, which are important in numerous heat exchangers. Therefore, the vortex shedding phenomenon is not considered in this paper. Secondly, turbulent buffeting vibration is considered as the dominant phenomenon in the slug flow and froth flow regions. A method for estimating this type of random vibration is explained in this paper. Thirdly, an unstable form of vibration, so-called fluid elastic vibration, is also considered. The common method for estimating the instability boundary is similar to that used for single-phase flow; however, new insights on the method for two-phase flow are given in this paper.

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Flow fields and tube exciting forces were investigated in straight tube-bundles and in a U-bend tube-bundle for several conditions of air-water two-phase flow. The two-phase turbulent exciting forces were correlated with the local flow behavior to elucidate the excitation mechanism. In the straight tube-bundles, the excitation was found to be due to the impact force of water slugs passing around the tubes, which is proportional to the dynamic pressure of the water slug. The estimation formula thus obtained was applied to the U-bend environment and it agreed fairly well with the measured result. Thus the excitation mechanism and its relation with the local flow field has become clear.

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An experimental study on boric acid penetration into the crevices between tube and tube-support-plate, and intergranular-attack (hereinafter called "IGA") cracks in crevices has been performed to obtain the optimum boron soaking procedure in operating steam generators with IGA. IGA is a corrosion mechanism involving the dissolution of the metal grain boundary. It consists of several cracks which propagate along metal grain boundaries from the outer surface to the inside of the tube. The term "IGA crack" refers to these cracks. Two experimental crevice models were set up. One was of the packed-crevice type where the crevice gap is completely packed by sludge, and the other was of the open-crevice type where the crevice gap is not packed, but reduced by sludge. The boron in IGA cracks was investigated using ion microanalysis in order to confirm the existence of anticorrosive film in IGA propagation. The optimum reactor power for effective boric acid penetration into the tube surface and into the IGA cracks within the crevice was found to be a power level of about 5% and 30%, which are applicable to both the packed-and open-crevice types.

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Understanding of the thermal-hydraulic behavior in the secondary side of the PWR (Pressurized Water Reactor) steam generator was required to improve the reliability of steam generator. Therefore, the thermal-hydraulic computer code FIT-III has been developed to predict the thermal-hydraulic behavior in the secondary side of steam generator. The FIT-III has been verified by using the two-dimensional and three-dimensional thermal-hydraulic test results. The verified FIT-III is used for the modification and design of the steam generator.

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Heat-treat (HT) Zr-2.5 wt%Nb pressure tubes which have high-temperature strength are used for the Advanced Thermal Reactor Fugen (boiling-light-water-cooled heavy-water-moderated pressure-tube-type reactor) in Japan. In general, zirconium alloys are considered to be degraded in the fracture toughness by picking up hydrogens from reactor coolant during reactor operation. It is reported in CANDU (Canada Deuterium Uranium) reactors that the deuterium pickup rate of cold-worked (CW) Zircaloy-2 pressure tubes is very high but that the deuterium pickup rate of CW Zr-2.5 wt%Nb pressure tubes is very low. Taking these results and those of Fugen pressure tube surveillance specimens (present work) into account, the hydrogen pickup for HT Zr-2.5 wt%Nb pressure tubes was estimated as about 44 ppm/30 years, which is very small and lower than that of the hydrogen solubility limit at 285°C, to yield almost no degradation of the pressure tubes under the operational temperature. Furthermore, a qualitative equation was proposed for the hydrogen pickup of the pressure tube.

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Analytical and experimental studies were conducted on the fatigue strength reduction factor (FSRF) of the weldments for breeder reactor structural design, focusing on the relationship between the strain constraint effect and the ratio of the welded metal breadth to the welded joint wall thickness. As the result, it was found that the FSRF for a 304SS-308 hot TIG welded joint is expected to be about 1. 1 for the wall thickness of the Japanese demonstration LMFBR vessel.

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Pseudodynamic tests of cylindrical shells under high temperature were performed in order to study elasto-plastic shear-bending buckling and the nonlinear response of fast breeder reactor (FBR) main vessels under earthquake loading. The test results showed a response reduction effect due to pre-buckling plasticity and a large seismic margin, which is the difference in seismic excitation amplitudes between the earthquake design conditions and the ultimate state, and is attributed to the energy absorption of the cylinders before and after buckling. A simple expression of the response reduction effect was proposed, as a contribution to the safe and effective seismic design of FBRs. A seismic margin evaluation method was also proposed, and it was shown that appropriate seismic margins can be ensured when the response reduction effect is incorporated into seismic design.

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Although plastic shear-bending buckling of the cylindrical part of Fast Breeder Reactor (FBR) main vessels under horizontal earthquake loading is one of the most critical problems in structural design, the evaluation method of buckling strength is not specified in related standards in Japan. Central Research Institute of Electric Power Industry (CRIEPI), commissioned by the Ministry of International Trade and Industry of the Japanese government, is carrying out verification tests of fast breeder technologies (from FY 1987 through 1993). The Demonstration Test and Research Program of Buckling of FBR is part of the commissioned research program, and the first half was accomplished after establishing a seismic buckling design guideline (a tentative draft) in FY 1989. The purpose of this paper is to describe the results of static buckling tests and numerical analyses, and to present formulae for determining the buckling strength of FBR main vessels. Shear buckling and bending buckling are evaluated under the assumption that both buckling modes slightly interact with each other. Regarding shear buckling, shape imperfection within the plate thickness is considered in the formulae originally, and a correction factor for larger imperfections is provided

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The Japan Power Demonstration Reactor (JPDR) decommissioning program is in progress : new technology for reactor decommissioning is being developed and a database is being established. Various data pertaining to dismantling of the JPDR have been accumulated in the database. The data collected in the program are being used for (1) managing ongoing dismantling activities, (2) verifying the code system for management of reactor decommissioning (COSMARD) developed for the forecasting of management information, and (3) planning future decommissioning of commercial nuclear power reactors. COSMARD and the database are expected to be particularly useful in the last of these three.

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In a reactor dismantling technology development project being conducted by the Japan Atomic Energy Research Institute under contract from the Science and Technology Agency, Kajima Corporation has been performing research and development on an abrasive water jet cutting system. This system is to be applied to the cutting of biological shields, which cannot be performed by humans due to the high level of radioactivity of the shields. A mock-up test of dismantling the biological shield of the Japan Power Demonstration Reactor (JPDR) has been performed to demonstrate that the abrasive water jet cutting system is fast, safe, and reliable under actual working conditions. The mock-up test provided valuable experience and various reference data useful for the dismantlement of the JPDR, which started in 1991. The abrasive water jet cutting system is expected to prove very useful in the dismantlement of atomic power plants.

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