The accident at the Fukushima Daiichi Nuclear Power Plant in 2011 has been influencing design concepts for nuclear reactors with inherent safety features for decay-heat removal. By considering these issues, we obtained design parameter conditions for prismatic high-temperature gas-cooled reactors (HTGRs) with passive safety features for decay-heat removal for both underground and aboveground reactors in our previous works. To determine those conditions, we performed a parametric survey analysis using fundamental equations for residual-heat transfer mechanisms such as conduction, convection and thermal radiation. Using the obtained parametric condition, we proposed the appropriate reactor core size for a 100-MW t reactor operating at an initial core temperature of 1123 K. Subsequently, neutronic analysis was performed for the proposed reactor core using uniformly distributed fuel to validate the possibility of designing a long-lived core with dimensions decided by the conditions. The proper optimizations for the non-uniformly distributed fuel and burnable poison particles, as well as the insertion of control rods to suppress the excess reactivity and flatten the change in power peaking factor during operation, were then successfully accomplished. The calculations show that with proper optimizations, the proposed core life is about 20 years, while the maximum power-peaking factor is less than 2.0 during operation. Therefore, calculations to find the core temperature coefficient for the proposed reactor were performed; it was always negative during reactor operation, with an average of -3.92 pcm/K. This paper shows that the excess reactivity of a long-life small HTGR whose core design satisfies the conditions for passive decay-heat removal was small, and its temperature coefficient was always negative during the operation. Small excess reactivity is a significant advantage from the viewpoint of safety in the case of a reactivity accident.

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An integral kinetic model was used to perform a transient analysis for possible criticality accidents in different types of weakly coupled systems, showing the potential of this method for general coupled systems. A feature of the integral kinetic model is its ability to describe a time- and region-dependent fission density for a coupled system of any geometry. In this study, a fuel-solution-tank system and a fuel-debris system were analyzed. Obtained results were a fission density and a released energy in each region, and a coupling between regions. It was concluded from this study that the integral kinetic model can further be applied to other coupled systems and can provide important characteristics of the criticality accident of such systems.

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Given the limitations of natural uranium resources, innovative nuclear power plant concepts that increase the efficiency of nuclear fuel utilization are needed. The Pebble Bed Reactor (PBR) shows some potential to achieve high efficiency in natural uranium utilization. To simplify the PBR concept, PBR with an accumulation fuel loading scheme was introduced and the Fuel Handling System (FHS) removed. In this concept, the pebble balls are added little by little into the reactor core until the pebble balls reach the top of the reactor core, and all pebble balls are discharged from the core at the end of the operation period. A code based on the MVP/MVP-BURN method has been developed to perform an analysis of a PBR with the accumulative fuel loading scheme. The optimum fuel composition was found using the code for high burnup performance. Previous efforts provided several motivations to improve the burnup performance: First, some errors in the input code were corrected. This correction, and an overall simplification of the input code, was implemented for easier analysis of a PBR with the accumulative fuel loading scheme. Second, the optimum fuel design had been obtained in the infinite geometry. To improve the optimum fuel composition, a parametric survey was obtained by varying the amount of Heavy Metal (HM) uranium per pebble and the degree of uranium enrichment. Moreover, an entire analysis of the parametric survey was obtained in the finite geometry. The results show that improvements in the fuel composition can lead to more accurate analysis with the code.

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The wastage by Liquid Droplet Impingement (LDI) in piping system of nuclear power plant is becoming great problem in recent years because of high aging operation. In this study, by using our original fluid/material two-way coupling numerical method which considers reflection and transmission on the fluid/material interface, high-speed LDI on material surface is simulated. There are the various fluid/material factors in LDI. In this study, the influence of droplet diameter, impingement velocity, and the condition of material surface on LDI is investigated. First, in order to consider the basic condition, the LDI on dry and flat surface is analyzed. Secondly, in an actual power plant, because there may be humidity inside a steam pipe line and the material surface is wet, the LDI on wet surface is analyzed. Additionally, it is thought that the pipe inner surface is not a completely flat surface because of processing or repeating LDI. To make sure that the effect of the rough surface, the LDI on pitted surface is analyzed. From the numerical results, it was shown that the wetness of material surface has an effect on decrement and the roughness of material surface has an effect on increment of maximum equivalent stress in the material. Moreover, the evaluation of present numerical results has become in the range of the value of the existing prediction formula and experiment.

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We have been developing a new passive water-cooling system for boiling water reactors (BWRs) for application to a situation such as a long term station black out (SBO) like that which led to the Fukushima Daiichi Nuclear Power Plant accident. The new passive water-cooling system consists of a condenser and a water supply system to the reactor. Steam from the reactor pressure vessel (RPV) is condensed in the condensation tubes of the condenser, and condensate water flows out into the suppression pool in the primary containment vessel (PCV). Water temperature at the condensation tube outlet is lowered below the saturated temperature at the partial steam pressure of the maximum PCV design pressure to prevent the PCV failure. To obtain heat transfer data for the water-cooling system, we conducted heat transfer tests using the tube bundle with 4×3 tubes to investigate particular phenomena such as the inlet steam flow distribution among the tubes, dryout on the tubes under the high void fraction around them, and enhancement of boiling heat transfer on the outside of the tubes due to convection flow and steam voids. In these tests, we utilized differential pressure transducers to measure the inlet steam flow distribution at the inlet of representative tubes. The differential pressure data at the tube inlet were obtained at system pressures of 0.4 to 3.0 MPa (absolute) and averaged inlet steam velocities of 5 to 47 m/s. The steam flow velocity calculated by the test results of differential pressure between the inlet header and the tube inlet was homogeneously distributed within about 2 % under the assumed rated operation condition. A high void fraction around the upper level tubes in a real condenser was simulated by injecting steam from beneath the bundle. We confirmed no dryout occurred based on the test results of the boiling heat transfer coefficient on the tubes with the void injection that is equivalent to that of a real condenser. Furthermore, the boiling heat transfer coefficient on the tubes was increased by about 11 % for the lowest level tube and by about 6 % at other level tubes under the assumed rated operation condition.

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For decommissioning of Fukushima daiichi nuclear power station, reduction of the dose equivalent rates inside the reactor buildings is an important issue. Concrete core sampling from the buildings to investigate the contamination is necessary for study about effective decontamination. However, dose rate inside the reactor buildings is very high. For example, dose rate of 1^<st> floor on the Unit 1 is 1.2 - 1820 [mSv / h], the Unit2 is 2.5 - 220 [mSv / h] and Unit 3 is 2.2 - 4780 [mSv / h]. So it is difficult for workers to work long hours. Therefore, a teleoperated robot, named "MHI-MEISTeR (Mitsubishi Heavy Industries - Maintenance Equipment Integrated System of Telecontrol Robot)", has been developed to conduct operations like concrete core samples from the reactor buildings. Actually, some concrete core samples from Fukushima daiichi were taken by MHI-MEISTeR. In addition, MHI-MEISTeR is designed as a versatile robot, and so it can conduct suction / blast decontamination works as well as concrete core sampling. The above operations were performed by MHI-MEISTeR in Fukushima daiichi nuclear power station.

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For the evaluation of reactor pressure vessel (RPV) lower head rupture probably had occurred during the severe accidents in Fukushima, the thermal-hydraulics / mechanical coupling analyses are conducted in JAEA. The mechanical analyses of the RPV need to treat multi-axial deformation of the materials, and in this case the applicability of analytical model using uni-axial data for such multi-axial deformation analysis is required to be verified. In this study, internal pressure creep tests using pipe specimens made of Japanese RPV material were performed. In addition, mechanical analyses of the tests based on the finite element method (FEM) were conducted as well then the results were compared for the validation. In the experimental conditions performed in this paper, analytical results well reproduced the deformation of the specimens in internal pressure creep tests.

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When a failure occurs in a nuclear plant, a lack in the response of the controller could lead to serious consequences. The fundamental properties to be ensured by the controller is the plant stability, to be formally proved if possible, and the robustness of the control law, which means fault tolerance and parameter variation tolerance. In this paper, using a mathematical model for the primary circuit of a PWR, accurate enough to catch the nonlinear time-varying, and switching nature of the system, two controllers are designed: an inventory controller for the primary circuit and a pressurizer pressure controller. These controllers do not use direct measurements of the pressurizer pressure or temperature, but use instead pressurizer wall temperature measurements and an observer. Disturbances and parameter variations are compensated by the use of sliding mode estimators, which guarantee robustness to the control scheme. Using an event triggered control control scheme, with varying sampling, the control law has been digitalized for a possible implementation on a digital platform.

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In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1℃. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

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Within DOPAS project (Demonstration of Plugs and Seals), which involves the participation of 14 European organizations an Experimental Pressure and Sealing Plug (EPSP) is being constructed and tested by a Czech consortium made up of SURAO, CVUT and UJV Rez, a.s. at Josef underground laboratory. The aim of the experiment is to develop and demonstrate the feasibility of such plug for deep geological repository of radioactive waste in real conditions. In order to achieve this not only EPSP plug is constructed but extensive monitoring programme is carried out too. The EPSP experiment is located inside experimental gallery niche. The niche was reshaped in advance and surrounding rock has been improved by grouting. The EPSP itself consists of pressurization chamber (connected via boreholes to technology), inside plug made of low pH glass fiber shotcrete, bentonite sealing section, filter and outside concrete plug. A water, air and bentonite suspension is planned to be used as pressurization media. Possible flow through the inside plug should be stopped by bentonite sealing section. Filter at the end sealing is for monitoring of possible leakages. As part of the demonstration the bentonite emplacement technologies such as shot clay are being tested.

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iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and WENRA safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged SBO without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven ABWR design. The NSSS is exactly the same as that of the current ABWR. As for safety design it has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double FP confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a DBA, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7 day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the CV building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most PWR resulting in a smaller R/B. Each active safety division includes only one ECCS pump and one EDG. Therefore, a single failure of the EDG never causes multiple failures of ECCS pumps in a safety division. The iB1350 is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.

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Laser cutting technologies of the thick steel plates for the nuclear decommissioning were developed with a 30kW fiber laser. Plates of stainless steel and carbon steel more than 100 mm thick were successfully cut, indicating that this technology is promising for the application to the nuclear decommissioning.

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Hydrogen gas does not have any absorption lines in visible light - near infrared ray spectrum. It is hard to get the gas distribution in whole area with any contact sensors. Raman scattering shifts wavelength form the incident light due to material composition. We focused on Raman scattering light from hydrogen gas. Raman lidar is ideal to detect the leaked gas remotely and safely. In this study, Compact Raman lidar has been developed for leaked hydrogen gas detection. The minimal detection limit of hydrogen gas concentration is 1% at the observation range of 0 - 50m. When we expand our project, two directions are pointed. One is the improvement to monitor the atmosphere environment. The improved lidar added monitoring water vapor. The other is to rebuild the system to be easy-to-use and to be easy-to-install. LED based mini-lidar is developed for this purpose.

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iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. The iB1350 uses innovative passive containment cooling system (iPCCS). The iPCCS is a part of the in-containment filtered venting system (IFVS). The vent pipe is submerged in the IFVS tank in the outer well (OW) of the Mark W containment. The conventional PCCS has a suction pipe only from the dry well (DW). On the contrary, the iPCCS has two suction pipes. One is normally opened to the wet well (WW) and another normally closed to the DW. The suction pipe in the conventional design cannot be connected to the WW because the PCCS vent pipe is connected to the WW. A PCCS functions using differential pressure between two nodes to discharge noncondensable gases in a PCCS heat exchanger (Hx). A suction pipe and a vent pipe must be connected to different nodes to use differential pressure. Therefore, the conventional PCCS never can cool the S/P. Although the S/P is the in-containment heat sink, heat up of the S/P is the most unfavorable for the conventional PCCS. In order to use the PCCS the conventional design must discharge steam directly into the DW instead of the S/P. Therefore, the conventional PCCS must open depressurization valves (DPV) at a SBO if the isolation condenser (IC) fails. On the contrary, the iPCCS can cool the S/P directly using the suction pipe connected to the WW and without DPV. Instead of DPV the iB1350 has modulating valves (MV) of which discharge lines are submerged in the S/P. Even if the IC fails at a SBO, the iB1350 can cool the core using the severe accident feedwater system (SAFWS), the SRV or the MV, and the iPCCS. The SAFWS makes up the core. The decay heat is carried by steam to the S/P through the SRV or the MV. The S/P works as in-containment heat sink. Once the S/P starts boiling the iPCCS automatically initiates cooling of the steam from the S/P. In the case of a core melt accident, a certain amount of FP is released into the S/P and heats up the S/P. Once the S/P starts boiling, the noncondensable gases in the WW is purged by the steam into the DW and then into the PCCS Hx. In order to purge the stagnant gases the conventional PCCS needs an active fan in the long term. On the contrary, the iPCCS can easily purge noncondensable gases in the heat exchanger using differential pressure to the OW and need not any active fan even in the long term.

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We evaluated Cs M4,5 near edge X-ray absorption fine structure (NEXAFS) of Cs halides and clay minerals containing Cs using DFT calculation with the NEXAFS of the clay minerals measured. The evaluation of Cs halides indicated that the core-hole strength (CHS) in the NEXAFS measurement was reduced to 0.6, taking into account the strength interaction of the core-hole with the valence-band electrons. On the basis of this result, Cs M_<4,5> NEXAFS of the clay mineral was well reproduced including the dominant peak and tail structures using DFT calculation. Moreover the calculation showed that the NEXAFS clearly changed with the displacement of Si atoms to Al in the six-membered ring of clay minerals and little did with the coexist of K atoms with Cs between layers of the clay mineral. In addition, we examined the electronic state of CsCl and the clay mineral containing Cs. This examination indicated that interactions between Cs and the nearest atom were significant for the clay mineral and small for CsCl. In addition, the charge density distribution indicated that the interaction between Cs p and O p was a bonding state and that the interaction between Cs s and O s is an anti-bonding state.

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Dr. Mainte, an integrated simulator for maintenance optimization of LWRs (Light Water Reactors) is based on PFM (Probabilistic Fracture Mechanics) analyses. The concept of the simulator is to provide a decision-making system to optimize maintenance activities for typical components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability, economic rationality, environmental impact and social acceptance. For the further improvement of the safety and availability of nuclear power plants, the effect of human error and its reduction on the optimization of maintenance activities have been studied. In addition, an approach of reducing human error is proposed.

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iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation 111.7 reactor after the Fukushima Daiichi accident. The iB1350 uses the Mark W containment and the in-containment filtered venting system (IFVS). The Mark W containment is made of reinforced concrete and has double cylinder FP barriers. There are also IC/PCCS pools, a fuel pool and a dryer separator (DS) pool on the top slab of the containment. These pools work as water seal for FP leakage through the top slab. Most FP such as CsI are scrubbed in the pools. The containment head is also submerged in the reactor well pool. The pool water works as shielding for radiation from the reactor core during normal operation and water seal for FP scrubbing during a severe accident. The base mat concrete is covered with the S/P and the core catcher that is also submerged with corium flooding water during an accident. Therefore, the Mark W containment has passive double confinement barriers for FP. Moreover, the IFVS works as in-containment filtered venting system that scrubs FP from the wet well (WW) and the dry well (DW). The filtered venting tank is arranged in the outer well (OW) of the Mark W containment. Even noble gases and organic iodine going through the filtered venting system are still confined inside the containment and never released directly to the environment. The IFVS uses the innovative passive containment cooling system (iPCCS) as the pre-stage heat removal system. The iPCCS has a normal open suction line from the WW. Therefore, FP are scrubbed in the S/P at first and then vented into the filtered venting tank. After the DW suction line of the iPCCS is opened all the steam is cooled and condensed in the heat exchanger. Most FP are trapped in the condensate and returned into the WW through the condensate return line of the iPCCS. Therefore, the Mark W containment has excellent FP double confinement barriers and the in-containment filtered venting system that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. Furthermore, there is the severe accident standby gas treatment system (SASGTS). It is an active filtration system for a SA that can function during a SA. The iB1350 enables no land contamination with the SASGTS. The results of population dose evaluation using RADTRAD code show excellent performance of the FP confinement of the iB1350.

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This paper analyzes the geometrical effect of coolant channels in the tightly packed fuel rods assembly (TA) which works at operating pressures of supercritical-pressure light water cooled fast breeding reactor (Super FBR), PWR and BWR. It is well known that high breeding by light water coolant is very challenging due to its high moderating property. The TA is good for high breeding because of the small area fraction of coolant to fuel (less than 0.1). However, with the reduction of coolant area, thermal hydraulic characteristics become more limiting. By ameliorating the geometry of the coolant channel, it is expected to improve the performance on thermal hydraulics. In this study, three types of coolant channel geometry are analyzed: geometry A is circular; geometry B is triangle-like shape with rounded-corner; geometry C is also triangle-like shape but with sharper corner. A CFD code STAR-CCM+ is used to carry out the simulations for each geometry at each operating pressure. Assessments are based on the thermal hydraulic parameters such as cladding temperatures, pressure drop and CHF. It is found that geometry B is superior to others at all operating pressures because of the broad design area of power, cladding temperature and pressure drop, and is able to meet the thermal hydraulic requirements in design of high breeding reactors.

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iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. The iB1350 has the spent fuel pool in the operating dome. During normal operation the spent fuel pool is cooled by the fuel pool cooling system (FPC). If a SBO occurs, the alternate feedwater injection system (AFI) makes up the spent fuel pool. Steam from the spent fuel pool is released through the external venting system (EVS). During refueling outage, however, the PCV head and the RPV head are removed. The reactor well is flooded by refueling water and connected to the spent fuel pool. If a SBO occurs in the refueling mode the severe accident feed water system (SAFWS) makes up the refueling water with the S/P water. The overflowing water is collected into the drain pit on the operating floor. The water in the pit is returned to the S/P via the drain pipes. Decay heat of the core fuels and the spent fuels are both transferred to the S/P. In the long term the S/P starts boiling. Then the innovative passive containment cooling system (iPCCS) initiates cooling of the S/P. In this way the integrity of both the core fuels and the spent fuels is assured. If the SAFWS is also lost core melt might occur. The AFI, however, makes up the spent fuel pool and the cooling of the spent fuels is still maintained. In order to cope with this situation the operating dome is enhanced to withstand pressurization and confine FP. These enhancements are easily achieved because the operating dome is made of steel and concrete composite (SC). The enhanced containment vessel is named Mark W^+ containment. It has also a filtered dome venting system (FDVS) connected to the operating dome. If boiling of the refueling water starts steam is vented to the environment via the EVS. The oxygen in the operating dome is purged by the steam. Then if fuel damage starts the EVS is closed and the FDVS takes over. Hydrogen is purged to the environment through the FDVS. After shutdown noble gases and iodine are already well decayed and operation of the FDVS does not cause any excessive dose. After the hydrogen purging is finished the FDVS is closed. Then the connecting venting pipe is opened and the iPCCS takes over the cooling of the Mark W^+ containment. Let it go. Let it go. The core never bothers it anyway.

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In a severe accident at a nuclear power plant, a large amount of hydrogen can be released to primary containment vessel or reactor building. Passive autocatalytic recombiner (PAR) is one of the most effective systems for hydrogen mitigation and safety accident management. Conventional PARs which have been already installed in some of nuclear power plants have considerable weight and large casing volume. In order to improve these features, the new type PAR is under developing. The catalyst of the new PAR is developed based on the monolithic substrate catalyst for automobiles, which is high quality, compact and cost competitive. In this study, the steam effect on catalyst activity was experimentally examined using gas composition analysis and X-ray Absorption Fine Structure measurement (XAFS). These results show that the steam slightly affects the reaction start up and catalyst activity.

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An iodine-sulfur based thermochemical water splitting (IS process) for hydrogen production has been developed by JAEA as application of a high-temperature gas cooled reactor. The IS process includes a severe corrosion environment which is made to boil and decompose concentrated sulfuric acid. Two kinds of brittle materials, SiC and Fe-high Si alloy, were reported as materials having enough corrosion resistance in this corrosion environment. The hybrid pipe consisting of the Fe-high Si alloy with boiling sulfuric acid-resistant and the carbon steel with the ductility was produced by centrifugal casting. An evaluation of the characteristics was carried out. The Fe-high Si alloy lining showed enough corrosion resistance in boiling concentrated sulfuric acid. Thermal cyclic tests (100-900 ℃) were executed in order to evaluate the interface between the carbon steel and the Fe-high Si alloy. There was no detachment of the interface and it was confirmed that the interface possessed the enough strength.

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Steam Injector (SI) is a passive jet pump with a converging-diverging structure. SI operates without the electrical power source by direct contact condensation of a supersonic steam flow and a subcooled water jet in a mixing nozzle of converging section. Since SI has high heat-transfer performance and discharges water at high pressure, SI is expected to apply to the safety system that is able to condense steam efficiently and inject water into the reactor when severe-accident occurs in a nuclear power plant. The objective of the present study is to clarify the operating criteria and mechanisms of generating discharge pressure. For these objects, we performed experiments. Operating in various conditions about the inlet water and steam flow rate was performed. The pressure distributions along the flow direction and the discharge pressure were measured by changing the load on the exit of test section. In addition, the discharged flow at the diffuser was observed with a high-speed camera, and an electric resistance was measured in the discharged flow in order to estimate the void fraction. The inlet conditions which SI could operate were specified experimentally. It was found that significant pressure rise position was generated in the diffuser, and this position moved toward the upstream with increasing the discharge pressure. In addition, the void fraction estimated from electric resistance in the discharged flow decreases with increasing the discharge pressure. From the above, the relation between the generation mechanism of discharge pressure and the discharged flow structure is thereby discussed.

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The pebble bed reactor (PBR), a kind of high-temperature gas-cooled reactor (HTGR), is expected to be among the next generation of nuclear reactors as it has excellent passive safety features, as well as online refueling and high thermal efficiency. Rock-like oxide (ROX) fuel has been studied at the Japan Atomic Energy Agency (JAEA) as a new once-through type fuel concept. Rock-like oxide used as fuel in a PBR can be expected to achieve high burnup and improve chemical stabilities. In the once-through fuel concept, the main challenge is to achieve as high a burnup as possible without failure of the spent fuel. The purpose of this study was to investigate the impact on burnup performance of different coated fuel particle (CFP) designs in a PBR with ROX fuel. In the study, the AGR-1 Coated Particle design and Deep-Burn Coated Particle design were used to make the burnup performance comparison. Criticality and core burnup calculations were performed by MCPBR code using the JENDL-4.0 library. Results at equilibrium showed that the two reactors utilizing AGR-1 Coated Particle and Deep-Burn Coated Particle designs could be critical with almost the same multiplication factor k_<eff>. However, the power peaking factor and maximum power per fuel ball in the AGR-1 coated particle design was lower than that of Deep-Burn coated particle design. The AGR-1 design also showed an advantage in fissions per initial fissile atoms (FIFA); the AGR-1 coated particle design produced a higher FIFA than the Deep-Burn coated particle design. These results suggest that the difference in coated particle fuel design can have an effect on the burnup performance in ROX fuel.

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In the event of a nuclear power plant accident pressure within the containment can increase. A Filtered Containment Venting System (FCVS) allows for over-pressure release through multi-scrubbers (Venturi-scrubber, bubbling-scrubber, metal fiber filter and molecular sieve) and reduces the radioactive release. However, FCVS performance changes depending on operational conditions, e.g. steam flow rate, pressure, operating time and so on. The Central Research Institute of Electric Power Industry (CRIEPI)full-height FCVS test facility is constructed to measure FCVS performance under several conditions and can evaluate the decontamination factor (DF) of three major targets (aerosol, elemental iodine (I_2) and organic iodine (CH_3I)). This project is intended to acquire a systematic database of FCVS performance and optimize the FCVS operation procedure. The CRIEPI test vessel is about 8 m high, with an internal diameter of 0.5 m. FCVS performance tests were conducted under the following conditions: maximum pressure and temperature of 0.8 MPa and 170℃, inlet gas flow of steam (〜1600 kg/h) and air (〜300 kg/h) and containing aerosol/ iodine/ organic iodine. If fission product iodine gas is released into the environment during a severe accident, it will have a major impact on public health. This paper addresses the iodine decontamination performance by the bubbling effect. Iodine is effectively soluble in an alkaline solution. Accordingly, 0.5 wt% sodium hydroxide (NaOH) or a mixture of 0.2 wt% sodium thiosulfate (Na_2S_2O_3) and 0.5 wt% NaOH is used as an iodine filter (absorber) and during the experiment, an alkaline solution with composition equivalent to the actual equipment is used. The concentration of elemental iodine is quantified with an Inductively-Coupled Plasma with Mass Spectrometry (ICP-MS), while iodine DF is defined by the concentration ratio at the inlet and outlet. Iodine DF shows low dependence on flow dynamics, but dependence on solution property. Where iodine concentration is low, DF is high (between 10^4 and 10^5) and vice versa when the iodine concentration (saturate) increases.

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ATMEA, a joint-venture between AREVA and Mitsubishi Heavy Industries has developed the ATMEA1 Pressurized Water Reactor (PWR) Nuclear Island (NI), leveraging both of its shareholders' proficient technologies, innovations and experiences. The scope of the ATMEA1 PWR development covers the complete scope of engineering works necessary to develop a standard product. As a recently emergent discipline in the field of nuclear plant engineering, Human Factor Engineering (HFE) is one of the challenges which has to be integrated within new plant development process. At early design stages of ATMEA1 development, ATMEA has defined and implemented an extensive NUREG-based HFE program, encompassing HFE Preliminary Analyses, Human-Systems Interfaces (HSI) Design and Verification and Validation (V&V) activities. The HFE Preliminary Analyses are defined through Operating Experience Review (OER), Functional Requirement Analysis and Function Allocation (FRA/FA) and Task Analysis (TA). Human-System Interface (HSI) Design and related V&V activities are based on Control Center, Control Rooms and Human-Machine Interfaces (HMIs). All these steps are implemented within the ATMEA Project through a structured generic documentation basis by a HFE team composed of HF specialists from both AREVA and MHI and managed by ATMEA. ATMEA1 development project aims to develop flexible and robust design and process which can be easily adapted to be compliant to any regulations over the world. U.S. regulatory guidelines related to HFE (e.g. ref. [1] and ref. [2]) were applied as a basis for this project. This paper presents the overall ATMEA1 HFE program content and the related process used to favor its implementation within the Project; through the collaboration between MHI and AREVA under ATMEA lead. A special focus is made on the HFE team composition, roles and responsibilities, the management of interfaces with the relevant engineering disciplines and the tools used to support the HFE activities. The lessons-learnt acquired during this project, pertaining to the implementation of such an HFE program in the frame of a new product development are finally presented.

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In the Fukushima Daiichi Nuclear Power Station (NPS) accident, hydrogen was generated by oxidation reaction of the cladding and water etc,., then leaked into the NPS building, and finally led to occurrence of hydrogen explosion in the building. This resulted in serious damage to the environment. To improve the safety performance of the NPS, especially on the hydrogen safety under severe accident conditions, a simulation code system has been developed to analyze hydrogen behaviors including diffusion, combustion, explosion and structural integrity evaluation. This developing system consists of CFD and FEM tools in order to support various hydrogen user groups of students, researchers and engineers. Preliminary calculated results obtained with above mentioned tools, damage of piping induced by hydrogen combustion, agreed well with existing test data.

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Austenitic stainless steels with superior corrosion resistance such as R-SUS304ULC (Ultra Low Carbon) have been used in corrosive environments of Rokkasho Reprocessing Plant. However, austenitic stainless steels suffer an intergranular corrosion in boiling nitric acid solutions. The intergranular corrosion is caused by the segregation of impurities to the grain boundaries. R-SUS310ULC EHP^[○!R] (Extra High Purity), hereafter referred to as 310EHP, has been developed in order to suppress the intergranular corrosion by reducing contents of the harmful impurities, such as carbon, phosphorus, sulfur and boron. Slight boron in the austenitic stainless steels was regarded as the causes for the intergranular corrosion in particular. The intergranular corrosion behavior of 310EHP, which was melted and cast with a small scale melter, was examined in boiling HNO_3 solution with highly oxidizing ions to clarify the correlation between the intergranular corrosion behavior and boron content. As a result, it was confirmed that 310EHP with lower boron content, less than 50 ppm, has the superior intergranular corrosion resistance compared with that of R-SUS304ULC. And then some over 1000 kg ingots of 310EHP were melted and cast with a large scale melter, and it was confirmed that the chemical composition was controlled within target ranges and the corrosion resistance were the same as that with a small scale melter. And the mechanical properties of 310EHP were obtained and these data were added to HPIS (High Pressure Institute Standard) C108:2011 AMD1:2014 in Japan.

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The corrosion behavior of Hastelloy-N alloys in molten salt coolant containing fission product elements were investigated to determine the safety of structural materials in molten-salt reactors (MSR). Corrosion tests of Hastelloy-N in the molten fluoride salt FLiNaK in an alumina crucible and a graphite crucible under argon gas were performed at 773 to 873 K for 100 h. In the experimental condition of this study, no synergistic effect of Cs addition in molten fluoride salt were observed on the corrosion behavior of Hastelloy-N immersed in molten fluoride salt. At 773K, no effect of Te addition in molten fluoride salt could be seen. With increasing test temperature, the depth of corrosive attack increased and depletion of Ni occurred. At 873K, the Ni-Te compound layer was formed on the corroded surface of Hastelloy-N immersed in the mixture of molten fluoride salt with Te in graphite crucible. It was estimated that the formation of Ni-Te compound layer depend on electrical potential of molten fluoride salt.

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The objective of this study is to investigate the potential use of thorium-based fuels, ThO_2-UO_2 and ThO_2-PuO_2 in particular, in a small (125 MWth) fluoride-salt-cooled high-temperature reactor (FHR) with lithium-beryllium fluoride salt (Li_2Be_4F) coolant. This study investigates thorium as an alternative fuel because it offers several potential benefits including enhanced proliferation resistance, lower waste radiotoxicity, and a higher abundance in nature compared to uranium. The deterministic lattice physics code WIMS was used to find the k∞ values and reactivity (fuel Doppler and coolant temperature) coefficients and to carry out fuel depletion. The results were verified using the Monte Carlo code SERPENT. For the analysis of homogeneous ThO_2-UO_2 fuels, calculations are performed over a range of thorium volume fractions and uranium enrichments (up to 20 wt.%) that can achieve the same total burnup of 5.07 MWd/kg (with acceptable reactivity coefficients) as a 100% UO_2 4.69 wt.% enriched fuel reference case. The maximum thorium volume fraction that can be utilised to obtain same burnup is 0.668 at 20 wt.%. In ThO_2-PuO_2 analysis, reactor grade Pu with an isotopic vector taken from typical light water reactor (LWR) discharge fuel with initial 4.5 wt.% enrichment, 50 MWd/kg burnup is used. The volumetric proportions of ThO_2 and PuO_2 are varied and the impact on total burnup, plutonium incineration rates and reactivity coefficients investigated. To achieve the same total burnup, the thorium volume fraction is estimated to be 〜0.86.

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To investigate the safety design criteria of heat utilization system for the HTGRs, it is necessary to evaluate the effect of fluctuation of thermal load on the reactor. The nuclear heat supply fluctuation test by non-nuclear heating was carried out by devising a new test procedure to simulate the nuclear heat supply test which is carried out in the nuclear powered operation. The test data is used to verify the numerical code to calculate the temperature of core bottom structure to carry out the safety evaluation of abnormal events in the heat utilization system. In the test, while the helium gas temperature was heated up to 120 ℃ by compression heat of gas circulators, it is necessary to impose a sufficiently high disturbance in the reactor inlet temperature. However, there was a technical restriction in the heat release from final heat sink for cooling water freeze proofing during operation in winter, which is special to the low temperature non-nuclear heating test. As the results of improvement of test procedure, a sufficiently high temperature disturbance was imposed on the reactor inlet temperature. Thus, it was found that the response of temperatures of metallic components such as side shielding blocks was faster than those of graphite blocks in the core bottom structure. Such difference of temperature transient characteristics between the metallic core support structure and the graphite blocks was emerged. It was also found that the temperature response was significantly affected by the heat capacities of components, the level of imposed disturbance and heat transfer performance.

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JAEA, JAPC and MFBR have been conducting design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. As the result of the design study and R&D activity related the innovative technologies incorporated in the design in the Fast Reactor Cycle Technology Development (FaCT) project up to 2010, basic design concept of JSFR was established and its development process to the commercialization including construction and operation of a demonstration version of JSFR was outlined. JSFR is a loop-type next generation sodium-cooled fast reactor (SFR), which is aiming at achieving development targets of Generation IV reactors concerning sustainability, safety and reliability, economics and proliferation resistance and physical protection by introducing the innovative technologies such as shortened high-chromium steel piping. The output power is assumed for the design study as 1,500MWe for the commercial version and 750MWe for the demonstration version. In FaCT phase I up to 2010, in order to evaluate feasibility to achieve the development targets, the design study has been conducted on the main components and systems. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lessons learned from the TEPCO's Fukushima Dai-ichi nuclear power plants accident, in the frame work of Generation IV International Forum (GIF), the design study is focusing on the design measures against severe external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated for the safety issues on SDC and SDG of a SFR.

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For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength of IG-430, one of important strengths for design data, is statistically evaluated, and the minimum ultimate strength for compression, Suc, one of fundamental strengths in the graphite structural design code, is also evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR. Also, the crack in IG-430 would not easy to propagate rather than IG-110. So, since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. Namely, IG-430 is applicable for VHTR graphite material from view point of compressive strength.

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Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs.

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We added models for the particle size distribution and non-local radiation heat transfer into a fuel-coolant interaction (FCI) simulation code JASMINE that was developed at Japan Atomic Energy Agency (JAEA) to extend the applicability of the code for ex-vessel melt coolability assessment. Also, modifications were made in the models related to melt particle generation and re-agglomeration of settled melt particles. The modified code was tested by simulating melt jet breakup experiments, namely selected cases of ALPHA/GPM series with alumina-zirconia mixture and steel melt by JAEA, and FARO experiments with urania-zirconia mixture by Joint Research Center (JRC) Ispra. Simulation results showed that the code reproduces the experimental results well for the cases with a deep subcooled water pool where the melt breaks up completely. On the other hand, significant underestimation of heat removal from the melt and overestimation of agglomeration of settled melt was encountered for conditions with the water pool at saturation temperature. The melt agglomeration behavior in the simulation was sensitive to model parameters on the agglomeration criterion and heat transfer depending on conditions.

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Seismic analysis of the ITER Blanket Remote Handling System (BRHS) was carried out to assess its structural integrity against a safe shutdown earthquake. Evaluating the dynamic response of the system is of essential importance because the BRHS has various configurations with the flexible articulated rail system and vehicle manipulators that move along the rail. In addition, the floor response spectrum in the vertical direction at the equatorial ports has a steep peak at 〜8 Hz due to dynamic amplification of the building therefore understanding the natural modes and frequencies is indispensable to verify the structural design. This paper reports the seismic analysis of the BRHS during blanket module handling operation. A dynamic global FE model analysis to specify the unfavorable configurations of the system via parametric study, and local FE model analyses to verify the structural strength of critical components using interface loads obtained from the global analysis were done. The seismic resistance of the BRHS was studied and the structural design confirmed.

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An experimental study was conducted to investigate the flow instability in a vertical mini-rectangular channel with distilled water as the working fluid. The rotational speed of the primary pump is gradually reduced to lower the inlet flow rate until the flow becomes unstable, while maintaining all other thermal parameters unchanged. Three types of instability, characterized by large amplitude oscillation, small amplitude oscillation and flow excursion, were identified from the experimental data. A stability map for the vertical mini-rectangular channel under forced circulation was established based on the Subcooling number and Phase Change number. The oscillation periods were correlated with the fluid transit time and the boiling delay time. A flow pattern map for vertical upward flow in a mini-rectangular channel was applied to confirm the flow patterns during the oscillation. The mechanisms of the three types of instability were obtained by considering several types of flow instabilities and comparing them with the oscillations observed in this work.

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The PEM (Prefabricated engineering barrier system module) concept is one of the promising methods of the emplacement of EBS (engineered barrier system) for high level radioactive waste in Japan. To chose this method as a most potential one, the engineering feasibility and the long-term stability must be evaluated. The key technology of assemblage and emplacement of PEM has been partly demonstrated in the laboratory. While the long-term issues such as chemical interactions between buffer materials and residual materials such as the metal shell and concrete supports by using the numerical analysis coupling the mass-transport and chemical reaction are focused in this study, because it could degrade the performance of buffer material by chemical interaction and degrade the confinement performance of overpack by passivation. In this study, The mass-transport-chemical reaction coupling analysis is performed in consideration of iron corrosion and layer formation of corrosion product on the surface of overpack/metal shell, diffusion of ferrous ions in buffer material, backfill material and concrete support, dissolution of buffer material, backfill material and concrete support, and precipitation of secondary minerals. The analytical results show that there was a little amount of the alteration of buffer material, because much of ferrous ion was precipitated as magnetite mainly on the surface of overpack/metal shell, while other ferrous ions contributes the alteration of bentonite, such as chamosite and greenalite. Because the alteration of bentonite is limited to the vicinity of the interface between the bentonite buffer and the overpack/metal shell, it was suggested that the durability of buffer material is maintained at the view point of low water permeability. Then, it was suggested that the change of EBS is not effect on confinement performance of overpack at the view point of pH.

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Silicon carbide (SiC) has many exceptional properties such as thermal stability, chemical inertness, low neutron absorption cross-section and high thermal conductivity. These excellent properties make SiC competent for high temperature, high pressure and high radiation environment and suitable for direct process monitoring in nuclear industries. Due to the excellent mechanical and thermal properties, silicon carbide (SiC) has been chosen as a potential cladding material for pressurized water reactor. The presence of oxygen impurity is inevitable from the fabrication process. Oxygen in SiC may reduce the mechanical properties of SiC. However, few research works has been focused on the effect of oxygen dopant to material mechanical properties. In this paper, first principle calculations are used to study the effect of interstitial or substitutional oxygen on the mechanical property of SiC. Three kinds of situations are simulated by VASP. The strain and stress curves are calculated to evaluate those doping effects. The result shows that the oxygen atoms make SiC easier to crack, and the tensile strength of SiC is reduced.

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In the terms of nuclear power plant operational life management, current trend of components lifetime extension requires precise information of structural material degradation. Present-day conventional methods of mechanical testing are usually based on the use of large specimens and higher consumption of testing material, whose availability and volume is often limited. For determination of material properties, sampling of the necessary volume of material is in most cases connected with affecting the integrity or even destruction of the assessed component. Moreover, several components are not usually covered surveillance programs, e.g. reactor pressure vessel internals. Innovative testing methods of Small Punch Testing (SPT) and Automated Ball Indentation Test (ABIT) are based on the determination of material properties from miniaturized testing specimens and their semi-destructive approach is very promising for the possibility of present data base of irradiated materials testing results enlargement and enable the option of component in-situ testing (ABI testing). Presentation is focused on description of the process of employment of these techniques in the process of irradiated NPP materials testing and evaluation at the accredited hot cell testing laboratory of UJV Rez, Mechanical Testing Department. Comparison with testing results from conventional mechanical testing methods is also depicted.

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The CONTAIN-LMR code, which had been originally released from Sandia National Laboratories (SNL) in the U.S.A., has been developed in the Japan Atomic Energy Agency (JAEA) to evaluate ex-vessel severe accident progression in a liquid metal fast reactor (LMFR). As a part of several studies in JAEA, this paper mainly describes a state of the art of a debris-concrete interaction model. Since the debris-concrete interaction might be accompanied with concrete erosion, hydrogen generation and FP release etc., it is one of the most important phenomena in the ex-vessel accident. Stand-alone codes CORCON and VANESA, which were developed for the debris-concrete interaction calculation under a light water reactor (LWR) condition, have been incorporated into the CONTAIN-LMR code. Therefore, CORCON and VANESA have been improved to apply the LMFR condition in JAEA. Main topics of these works are followings; (1) addition of chemical reactions in the sodium pool, (2) consideration of heat conduction through the concrete, and (3) validation using SURC experimental data. As a result, improved CORCON and VANESA can represent the debris-concrete interaction behavior reasonably well.

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The Great East Japan Earthquake and subsequent tsunami on 11 March 2011 initiated accident conditions at several nuclear power stations on the northeast coast of Japan, and it developed into a very serious nuclear accident at Units 1-3 of the Fukushima Daiichi Nuclear Power Station (NPS). As a result, the IAEA set out a comprehensive program of work, called Action Plan on Nuclear Safety, in order to strengthen nuclear safety worldwide in 12 major areas. One of the action items is "Effectively utilize research and development". To assist relevant stakeholders in conducting necessary research and development in nuclear safety, technology and engineering, the IAEA has been identifying and organizing lessons learned from the accident, and has assimilated them into a comprehensive and concise set of lessons learned with emphasis on technical issues. This paper summarizes the event sequences during the Fukushima accident, taking into account recently published new information, and discusses technical lessons learned from the accident at the Fukushima Daiichi and Daini Nuclear Power Stations.

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To safely transport radioactive materials it is required that the packaging system must be able to guarantee their containment and confinement (via its integrity assurance) avoiding any additional dose exposure (respect of ALARA principle). The engineering system has thus to be designed to ensure safety features by complying/satisfy specific safety requirements, like the IAEA Safety Standards No. TS-R-1 and No. SSR-6. In this framework, the present study is intended to investigate and analyse the performance of an Italian CC-440 package, for the transport of low and intermediate level radioactive waste, subjected to the inclined free drop test. To the purpose numerical analysis simulating the drop test and performed by the FEM MSC Marc code are presented and discussed. In doing that a validation analysis of the numerical tool, based on available experimental data, has been carried out. The results obtained, in terms of stress, strains, acceleration, etc., showed that even if the packaging system components suffered local deformation, the magnitude of these effects would not determine an unacceptable loss of the safety features.

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Harmonics expansion method is employed to obtain the spatially continuous 3D on-line power distribution based on the dispersed and limited detector measurements. Power distribution is expanded by harmonics which are high-order eigenfunctions of neutron diffusion equation. The expansion coefficients are determined by using in-core detector measurements. Krylov sub-space method is employed to obtain those harmonics, while least square principle is chosen for expansion coefficients calculation. Moreover, instead of the original quarter core finite differencing method, a whole core nonlinear iteration semi-analytic nodal method is used for harmonics calculation. It has been found that the nodal harmonics calculation runs about 100 times faster than the finite differencing one without any loss in accuracy. Based on these models, an on-line monitoring system named NECP-ONION has been developed. Real detector measurements from Unit 1 reactor of DayaBay NPP, a typical PWR reactor in China, are used for code verification and validation. Numerical results show that the root-mean-square errors of assembly averaged powers are less than 1.8% for different burnup steps during the entire cycle. In addition, it has been observed that the assembly power monitoring error can still be driven down to less than 2% even if only 60% of measurements are available.

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A new concept of molten salt reactor (MSR) is proposed. The new concept is a static-fuel-type MSR (S-MSR) which is composed of a cylindrical core tank filled with molten salt fuel typically such as FLiNaK added with a few % of minor actinides (MAs) and plutonium and a vessel surrounding the core tank, filled with fluoride molten salt such as FLiNaK, acting as the primary coolant. The primary coolant is cooled with heat exchangers located inside the coolant vessel. Burning / transmutation rate of TRU elements in this reactor was evaluated to show as high as over 50%, a very high value. This type MSR is extremely safe from the viewpoint of severe accident since the molten salt fuel is confined in the core tank without circulation outside of the core. So, this type MSR can be expected to be applied to drastically reduce the amount of high level radioactive waste recovered from the spent fuels of the nuclear power plants including the fuel debris from the damaged reactors of Fukushima-Daiichi Power Station.

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Co-60 is an activated metallic erosion product, which is very important for waste management and decommissioning work of pressurized water reactor (PWR) power plants. Recent measurement on the samples from the primary loop of HTR-10 indicates the existence of Co-60. In current paper, the preliminary experimental results in HTR-10 will be introduced, and the production mechanism of Co-60 in the pebble bed high temperature gas-cooled reactors will be summarized and compared with that in PWRs and Germany High Temperature Nuclear Reactor (AVR). The further experiments with decomposing the post-irradiation graphite spheres of HTR-10 are put forward, which will promote the further study to testify the production sources of Co-60 and be of great significance in the waste minimization and the decommissioning work of HTR-10.

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As in clear from Fukushima Daiichi nuclear disaster, it is expected that the removal effect of Fission Products (FP) by the scrubbing effect in the pool water in the process of BWR containment vessel. In the PWR plants, we should eliminate the steam including fission products with gas by using the pool scrubbing effect, for the case that an accident occurs in the heat-transfer-tube of steam generator. The gas in the containment vessel and primary system including the FP aerosol spread into the pool water, the aerosol in the gas is scavenged in the scrubbing water. We predict the reduction of aerosol which shifts to containment space by the aerosol removal effect of the scrubbing. Hence, the aerosol removal by the pool scrubbing is important to evaluate the source term. In terms of the scrubbing effect for describing the real phenomena, in addition to boiling under reduced pressure conditions of the pool water, in the range that non-condensable gas flowing into the pool water is suitable, the two-phase flow behavior and aerosol behavior are required to elucidate the detailed correlation. This study experimentally examines the relationship between aerosol behavior and bubbles. We obtain the basic data for evaluating quantitatively the effect of removing the nuclear product. Based on these results, we improve the model for the validity confirmation of the severe accidents predictive code such as the MELCOR under the present conditions. We thereby visualize on a rising motion of single air bubble including the aerosol visualization. In this draft paper, we conducted preliminary experiments focused on behavior of single air bubble and bubbles including aerosol. After that we confirmed problems about design of experiment and have devised experimental apparatus systems.

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The aim of this work is to use a modified version of TRACE for a preliminary study of a molten salt reactor (MSR) system behavior for the steady-state, transient and accidental conditions so as to demonstrate the suitability of the TRACE code for MSR simulation. To this end the analysis and results from the MSRE (Molten Salt Reactor Experiment) were used as a reference case. The basic approach of this work is to couple the 1D thermal-hydraulic model for 3 loops and the 0D neutronic dynamics (point-kinetic model), which was solved by the ODE-solver built by "Control Blocks" offered in TRACE. Additionally, new working fluids, namely the molten salts, were added into the source code of TRACE. Most results of the simulations show good agreement with the ORNL reports as well as the previous study in recent years and the errors were predictable and in an acceptable range. Therefore, the necessary code modification of TRACE appears to be successful and will be refined further in order to investigate new MSR designs, such as MSFR and DFR.

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For the protection of coastal Nuclear Power Plant (NPP) against the external flooding hazard, the risks caused by natural events have to be taken into account. In this article, a methodology is proposed to analyze the risk of the typical natural event in China (Typhoon). It includes the simulation of the storm surge and the strong waves due to its passage in Chinese coastal zones and the quantification of the sequential overtopping flow rate. The simulation is carried out by coupling 2 modules of the hydraulic modeling system TELEMAC-MASCARET from EDF, TELEMAC2D (Shallow water module) and TOMAWAC (spectral wave module). As an open-source modeling system, this methodology could still be enriched by other phenomena in the near future to ameliorate its performance in safety analysis of the coastal NPPs in China.

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Water-lubricated thrust bearing is one of the key parts in canned motor pump, for example, the RCP in AP1000, and it functions to balance axial loads. A calculation model which can handle all flow state lubrication performance for water-lubricated thrust bearing has been presented. The model first includes laminar and turbulent Reynolds' equation. Then to get continuous viscosity coefficients cross critical Reynolds number, a transition zone which ranges based on engineering experience is put up, through which Hermite interpolation is employed. The model is numerically solved in finite difference method with uniform grids. To accelerate the calculation process, multigrid method and line relaxation is adopted within the iteration procedure. A medium sized water-lubricated tilting pad thrust bearing is exampled to verify the calculation model. Results suggest that as rotating speed enlarges, lubrication state distribution of the thrust bearing gradually tends to turbulent lubrication from the intersection corner of pad outer diameter and pad inlet to the opposite, minimum water film thickness increases approximately linearly, maximum water film pressure has little change, meanwhile the friction power grows nearly in exponential law which could result in bad effect by yielding much more heat.

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The present study was intended to examine how the condensation heat transfer, especially the drop-wise condensation, was affected by modifying the surface nature. In the present study, condensation heat transfer experiments for steam were performed by using mirror-finished copper surface with very thin metal films by using sputtering. That is, the effects on pattern of condensation heat transfer, i.e., drop-wise condensation(DWC) or film-wise condensation(FWC), of metal-sputtered surfaces were examined experimentally and qualitatively. DWC has ten times higher heat transfer coefficient than FWC. The surface condensation with sputtered thin Pb, Cr, Ag films became DWC. On the other hand, the surface condensation with sputtered thin Ti and Cu films was FWC. In the case of DWC, it has high contact angle compared to FWC.

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