The applicability of a seismic-isolation device to an Advanced Pressurized Water Reactor (APWR) plant was investigated in an attempt to rationally achieve plant safety against earthquake and promote the standardization of plant design. Consequently, it was found that the seismic-isolation device is capable of significantly mitigating the effects of seismic force on the building. As a result, in addition to a reduction in building volume, and wall thickness, the optimal design of support structures for equipment and piping can be achieved while maintaining safety against earthquakes. Finally, it has been verified that the base-isolated APWR plant will have an advantage in construction costs even after taking into account the costs of providing the seismic-isolation device.

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PWR advanced-RCC, which increase control rod worth adopting B_4C with enriched ^<10>B as RCC absorber and the B_4C/Ag-In-Cd axially hybrid design, is developed to enhance flexibility of large MOX or high enriched fuel loading capability. And the adoption of a double cladding structure result in improvement of wears resistance and rationalization of inspection. The adoption of the advanced-RCC results in the reduction of the construction costs and the maintenance costs for the significant reduction of the number of the RCCs in PWR. To verify these new designs of the advanced-RCC, some experiments of control rod flow induced vibration, RCC drop characteristics, control rod worth, the irradiation characteristics are performed. It is confirmed the advanced-RCC will be adopted in APWR in the near future.

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Based on the experienced light water reactor technology, conceptual design studies on advanced water-cooled reactors have been performed. They are named "Reduced-Moderation Water Reactor" (RMWR) with the high conversion ratio more than 1.0 and the negative void reactivity coefficients. Several concepts have been successfully established for them based on the neutronics calculations. Also the thermal hydraulic investigation has been performed for each concept, including the basic critical heat flux experiments at 15.5 MPa in the triangular tight-lattice core configuration. Through this thermal hydraulic investigation, the feasibility of each design is being confirmed.

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A new optical nuclear instrumentation system such as that in core power and temperature monitor using radiation-resistant optical fibers was demonstrated. Developed radiation-resistant optical fibers as a sensing component kept their good transmission characteristics even in a reactor irradiation. Optical fibers generated strong radio luminescence wavelength ranged 400nm to 1400nm with some of peaks and thermal radiation in infrared during irradiation in a core region of fission reactor. Intensities of radio luminescence were found to be directly proportional to the reactor power. Also, temperatures could be estimated by thermal radiation of Planck's law. The optical measuring method using optical fibers could be composed simple and convenient nuclear instrumentation system of the reactor power and temperature.

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A passive containment cooling system (PCCS) in under planning to use in a next-generation-type BWR for long-term cooling by condensing steam using horizontal heat exchangers. Heat transfer behavior in a secondary water pool is one of important phenomena governing heat removal performance of the PCCS. It was concerned that heat transfer tubes are covered by steam under high heat flux regions and the heat transfer is degraded in the region (Steam-blanket effect). This study evaluated the steam-blanket effect through analyses of thermal-hydraulic behaviors in the secondary water pool by multi-dimensional two-fluid model code ACE-3D. Before the analyses, turbulent Prandtl number was optimized using experimental data of steam jet into a water pool. It was found from the steam-blanket analyses that no any heat transfer tubes are covered only by steam and no significant heat transfer degradation is occurred.

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The Reactor Internal Pump (RIP) was adopted for the Reactor Recirculation System (RRS) of Advanced Boiling Water Reactor (ABWR) plants, and ten RIPs are located at the bottom of the reactor pressure vessel. In order to simplify the power supply system for the RIPs, a new inertia-increased RIP was developed, which allows to eliminate the Motor-Generator (M-G) sets. The rotating inertia was increased approximately 2.5 times of current RIP inertia by addition of flywheel on its main shaft. A full scale proving test of the inertia-increased RIP under actual plant operating conditions using full scale test loop was performed to evaluate vibration characteristics and coast down characteristics. From the results of this proving test, the validity of the new inertia-increased RIP and its power supply system (without M-G sets) was confirmed.

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Thermo-hydraulic instabilities of a boiling natural circulation loop with a chimney under low and high pressure were investigated using linear stability analysis. The effect of nuclear coupling was also considered. Both in low- and high-pressure conditions, instability could occur when exit quality was relatively low. In low-pressure condition, flashing only near the exit of the chimney could induce instability, and enthalpy wave of single-phase flow was generated in the chimney. In high-pressure condition, void was generated near channel exit, and void wave propagated in the chimney. In the high pressure and high power condition, though flow could be very stable, the decay ratio of higher mode could be larger than that of lower mode. The sensitivity on decay ratio to the thermal power, inlet subcooling, void reactivity feedback coefficient and so on could be very low when there was a long chimney.

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Several plants that were the first to be constructed in Japan have been operating for more than 20 years now, and preventive maintenance is therefore a matter of great importance. This paper summarizes the status of applied laser maintenance technologies both preventive and repair. Especially for the laser peening and laser de-sensitization treatment technology, field applications were also described in detail. In future, expansion of field application area on the preventive maintenance, repair and inspection technologies will be developed.

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Hydrogen water chemistry and Noble metal chemical addition were applied to actual BWR plants to protect structural materials from IGSCC so far. The important parameter of the susceptibility of IGSCC is the corrosion potential of the materials and the recommended value to mitigate IGSCC is reported as -230mV vs. SHE. In this paper, we discussed the other method to lower the corrosion potential below -230mV vs. SHE. Some n-type semiconductors are excited by photons irradiation and generate photoelectric current. The stainless steel coated with the semiconductor, therefore, has some possibility of shifting the corrosion potential in the negative direction under the photons irradiation. From the laboratory measurements under ultraviolet rays irradiation, the corrosion potential of stainless steel coated with TiO_2 shifted in the negative direction by a few hundred mV. Evaluation of photon flux produced by Cherenkov radiation in the BWR are also carried out and the corrosion potentials of the stainless steel coated with TiO_2 are estimated.

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This paper describes a pebble-bed type nuclear reactor design of HTGR with gas-turbine plant. The main objectives of this design are to attain a thermal output of 400 MWt, whereas that of the previous design is only 300 MWt, without any changes of the core inlet/outlet temperature conditions and the material of the reactor pressure vessel. The outline of the reactor design and nuclear and thermal-hydraulic performance evaluation results are provided.

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JAERI has constructed a 30-MWt HTGR, named HTTR, of which first criticality was attained in 1998. A hydrogen production system by means of steam reforming of natural gas is planned as the first nuclear heat utilization system coupling with HTTR. A reformer is a key component to produce hydrogen by steam reforming. Heat consumption of the reformer significantly effects on the heat balance of hydrogen production system. This report describes the numerical analysis results for the reformer using mathematical model on heat and mass balance. It was investigated the effects of operating conditions, flow unbalance among catalyst tubes and choking in catalyst bed on hydrogen production rate and heat consumption of the reformer. And CO_2 reforming to reuse CO_2 as a raw material was also investigated focusing on hydrogen and CO production rates for methanol synthesis.

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Research on IS process, thermochemical cycle for hydrogen production by water-splitting process, are proceeded to utilize heat from High Temperature Gas-cooled Reactor as primary heat source. So far, the continuous hydrogen production of IS process on laboratory-scale have been successfully demonstrated through stable rate (1 liter per hour) of hydrogen production for 48 hours and stoichiometric production of hydrogen and oxygen. Further studies in three fields, the evaluation of materials in corrosion environments, the modification of HI processing scheme using membrane technologies and the art to operate closed-cycle process, are conducted. Aiming to investigate operational technique on IS process, a scaled-up glass plant is under construction; the plant has ability to produce hydrogen at a rate of 50 liters per hour and higher iodine concentration to separate solutions well.

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Japan Nuclear Cycle Development Institute (JNC) have started the following feasibility studies (F/S) in order to develop a commercialized fast breeder reactor (FBR) cycle system satisfying the needs of the future society since JFY1999 with the electric companies. In the F/S, a number of candidate concepts will be selected from various options about the FBR system and the related fuel cycle system (reprocessing and fuel fabriction), featuring innovative technologies. The options are examined considering the attainable perspective for the followings : 1) safety, 2) economic competitiveness to future LWRs, 3) efficient utilization of uranium resources, 4) reduction of environmental burden and 5) enhancement of nuclear non-proliferation in order to determine the promising concepts making the most of the merits of the FBR cycle system and to define the necessary R&D tasks. In the end of JFY2000,we will determine the promising concepts from various options and summarize the results of the F/S (Phase I). Current status of these F/S activities was described in this paper.

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Thermal-hydraulic characteristics of the decay heat removal system (DHRS) has been evaluated by both a water test and its computational analysis. The direct reactor auxiliary cooling system (DRACS) was studied as a representative DHRS in the LMFBR design. A natural circulation flow of the DRACS is originated from the buoyancy force balanced with the pressure loss within the active core. Hence it appears that the performance of the DRACS could be fully estimated by the thermal-hydraulic study of one fuel sub-assembly under the condition of low flow rate. The inter-wrapper flow ocurring in the gap between core sub-assemblies under the natural circulation operation might have a large capacity to cool the core. An inverse flow ocurring at the outer region of the core was due to the cooling effect of the inter-wrapper flow.

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R&Ds on divertor high heat flux components are one of the key issues to realize for fusion experimental reactors such as ITER. JAERI has been developing the divertor components to meet the ITER design. The annular tube has been developed to reduce space and costs of the divertor. To optimize the annular tube design, pressure drop was measured with different end-plug structures. As the result, the lowest pressure drop of 0.1MPa at the flow velocity of 10m/s was obtained at the 7.5mm round shape end-plug section. The W/Cu hot pressing method has been also developed to improve the reliability and to reduce the cost. Thermal cycle test was performed on the W/Cu hot pressing divertor mock-up to investigate the durability of the joining interface. As the result, the surface temperature shows almost constant throughout the experiment, and no degradation of the thermal response is found through 3000 cycles.

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This paper describes a preliminary analysis of the response of an accelerator-driven system during beam trip transients. The system addressed is a dedicated transmuter consisting of a lead-bismuth cooled 800-MWt subciritical core combined with a saturated steam turbine cycle. Transients of the primary coolant temperature, the water/steam temperature, the water/steam pressure, the turbine flow rate, and the electric power output were calculated using a simple network model. The result shows that a beam trip shorter than 380s requires no active control of the system, while a prolonged beam trip demands turbine and plant trips to preventing from failure of system components. The maximum temperature swing is 185℃ in lead-bismuth at 370s after the beam trip.

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Based on the ductile fracture tests resolts of Ductile Cast Iron and Forged Steel using large-scaled flat test specimens with crack, the evaluation method of initiation of ductile fracture of a cask body with crack by J-integral were developed, and several formula was proposed. And a concrete application method of obtained formula was proposed in case of Cask drop accident. Following results were obtained. 1) By using bose Non-dimensional J-integral Φ(=J/Eεy ^2a or J/2πEεy ^2a) and Non-dimensionalizd strain (ε/εy), the differance caused by materials cames into small and J-ε overall curves become together into one curve. 2) Obtained J-integral design curves : Φ=(f^2/2)(ε/εy)^2 for ε/εy<1,Φ=f^2(ε/εy^<-0.5>) for ε/εy≧1 ε overall/εgross=8.1537(y/a)^<-0.7347> Φ : Non-dimensional J-integral (Φ=J/2πEεy ^2a), f : Shape Factor, εy : YieldStrain, ε : Strain ε overall : Overall Strain, ε gross : Gross Strain, y : GageLength×1/2,a : Equivalent Crack Length×1/2

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The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future.

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Supercritical fluid DIRect Extraction (Super-DIREX) Method has been developed, and it is the technique of extracting Uranium and Plutonium elements directly from spent fuel oxides by supercritical CO_2 containing HNO_3-tributylphosphate (TBP) complex (at 313K-333K and 10-20MPa). From this study, it is concluded that the reprosessing system adopting Super-DIREX Method is more simple and unexpensive than PUREX system. The experiment was carried out by using test piece including irradiation Uranium oxide and imitative fission product (FP) elements oxide, and the results showed that Uranium could be extracted in separating from FP elements at 313K and 12MPa and that one step extraction by Super DIREX Method could get more than 100 of decommission factor (DF) of the FP elements.

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The development of the disassembly machine. which uses the laser beam. is proceed as a disassembly machine for FBR fuel reprocessing in JNC. The verification of the engineering level is proceed with the full-scale cold test device, which uses the CO_2 laser that generates maximum output 10KW manufactured in 1993. and rationalization and the upgrade examination are proceed based on the obtained findings now. In this paper we report on an outline of current research and development and results of the recent research for the upgrade.

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For the decommissioning of nuclear facilities, decontamination of an equipment and a building should be carried out to reduce the release of radioactive material and to facilitate waste management as well. Decontamination of coatings on the floor and the wall of the building and the surface of equipment is also very important issues. Laser decontamination technology has the advantage of very small secondary waste and is suitable for remote decontamination system. In the present experiment, the decontamination with pule laser and its removal ability, characteristics of decomposition and laser transmission by optical fiber have been studied, and the surface decontamination with pulse YAG laser will be available after development of high power laser in near future.

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This paper deals with decision-making methodology for decommissioning/life extension of nuclear power plant in the total plant life management (PLM). Two computer systems for economical evaluation of PLM and for planning and management of reactor decommissioning have been developed so far by the Central Research Institute of Electric Power Industry and the Japan Atomic Energy Research Institute, respectively, and the related studies on plant life extension and decommission are on going in both organizations independently. The two computer systems were then effectively combined to evaluate the PLM data elements such as power generation costs, worker dose, waste arisings for decision-making on plant life scenarios of decommission/life extension. The preliminary study on PLM indicated that the 20 years life extension would effective for lowering levelized power generation costs in whole life, and that decommissioning costs were affected by the life extension procedures if the disposal costs of replaced components were included in it. The optimum plant life scenario will be decided based on the evaluation of PLM data including other elements such as social effects. The computer systems will be customized for future study on plant life optimization.

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Among the various reprocessing wastes to be generated by future nuclear fuel cycles, the volume of hulls and end-pieces of cladded fuel structure segments must be reduced. Although are plasma melting has been considered a possible method, zircaloy, a major hull component, has an extremely high melting point, and experimental verification of are plasma melting has not been completed in Japan. In the study, we performed melting and solidification experiments in a small-sized are plasma melting furnace containing a graphite crucible, using unused zircaloy waste as simulated hull material. Its melting behavior was studied with respect to the following parameters : heating energy and heating atmosphere, and the method of introduction of the waste into the furnace. The study showed that for efficient melting, a small amount of waste at a time should be introduced into a nonwater-cooled crucible and that the waste should be briefly exposed to plasma. According to thermodynamic properties such as saturated vapor pressure of lanthanide elements and transuranic elements, surrogate elements are selected to simulate migration behavior of TRU elements during thermal plasma melting of TRU wastes. These elements are simultaneously melted by thermal plasma heating with zircaloy. The surrogate elements have a tendency not to migrate into dust.

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The melting treatment is of great promise as treatment technology of volume reduction and stabilization for low-level radioactive miscellaneous solid wastes generated from nuclear facilities. Japan Atomic Energy Research Institute (JAERI) has been developing plasma melting method and carrying out melting tests of simulated miscellaneous solid wastes by this method. This paper describes heating characteristic, distribution behavior of radioactive tracer and volatilization behavior of slag component in the plasma melting. Thermal property of waste material had a great influence on heating efficiency of plasma heating. Uniformity of molten products was confirmed by radioactivity measurements. Residual fraction of Cs-137 in solidified product decreased with increasing of heating times. On the contrary, almost all of Eu-152 remained in solidified product. Volatilization of chemical components from molten slag was observed.

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Japan's first commercial nuclear power station is to be dismantled from 2008 onward. On the other hand, with the occurrence of the global environmental problem as well as the problem of natural environment conservation and wastes as a turning point, an approach is being made toward the creation of a circulation society in the future. When decommissioning the nuclear power station, therefore, ti is necessary that any plan for the disposal and recycling of dismantling wastes which are temporarily produced in large quantities be carried out under a social consensus. In Japan, there has been a lack of experience in decommissioning the nuclear power station, and in this field there are not many skilled hands. In the decommissioning plan, therefore, it is necessary to grasp the precautions which must be taken in such decommissioning and which are interwined complicatedly, and their interrelationships as well. Furthermore, a streamlined decommissioning plan must be formulated after taking int consideration the various case studies which are judged beforhand to be feasibly applicable. The objective of this research is to develop a comprehensive planning verfication tool called "Decommissioning-Recycle Simulator" in which power station locational conditions, environmental and other technical conditions, as well as their interrelationships are considered. This paper indicates the necessity of the "Deccommisioning-Recycle Simulator" and its concepts.

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To establish the technique of the recycle for concrete waste, This paper describes the recycle condition for existence of the decommissioning concrete waste in the nuclear power plant. And considered the durability of Cask Yard Concrete constructed at about twenty years ago, The authors examines the recycle system of concrete in the Power plant.

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Nuclear Power Engineering Corporation (NUPEC) has been developing decommissioning techniques, implemented under a contract with the Ministry of International Trade and Industry (MITI), to verify and improve the performance of the key decommissioning techniques. One of main themes is on concrete recycling techniques, which deals with high quality aggregate retrieval from concrete waste, high efficient usage of the by-product powder to recycling products, and effective usage of radioactive concrete to filling material for waste form. This paper describes progress and accomplishment on the concrete recycling technique development which started in 1996,

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Compressive strength and shielding performance tests of heavy weight concrete mixed with depleted Uranium (Depleted Uranium Concrete) were carried out. The depleted uranium pellets (Φ8mm, height 9.5mm) were mixed into cement paste instead of coarse aggregate. Specimens with nominal specific gravity of 3.5&acd;5.4 were manufactured. The results of the compression strength test showed that compressive strength of more than 30MPa was obtained with the specimens having the nominal specific gravity of more than 5 and it was confirmed from the shielding performance tests that Depleted Uranium Concrete has shielding corresponding to its nominal specific gravity.

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The authors have developed manufacturing method of high-volume fly ash concrete blocks named "The Super Fluidizing Method". The super fluidizing method is a completely new procedure, which hardens the mixture of cement, fly ash and a minimal amount of water by way of a series of vibrations. This method is applicable to another powdery materials. These fly ash concrete blocks (asherete blocks) are utilized s a material of the manmade sea mountain, which enable to make a fishing ground.

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Operations which operators executed with manual during the abnormal conditions were extracted as fourteen automation-systems by the event-sequences. And the automatic program which was to switch "on" or "off" a group of components, and the trimming function which was to control the flow of pumps or turbines were devised. Eight automation-systems were evaluated by computer-simulation for two abnormal events, and the time taking over from machine to operators after the automation-system trouble was confirmed to be long enough. From the viewpoint of operators' trust to automation-systems, the monitoring information was grouped by level of operational procedures. And in order to monitor the situations reliably for the cooperation between human and machine, combination of the automation-systems with the operation support system was confirmed to be necessary.

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It is necessary to automate Machine systems because they have become larger and more complicated these years. Generally speaking, humans hardly grasp the overall state in the automated systems. In fact it is reported that the accident caused by this problem occurs. To avoid such accidents, there were many studies to give human the authority of final decision making. In general it depends on circumstances whether the authority of decision making is given humans or machine systems. It is supposed therefore that humans and machine systems exchange their information each other and efficiently share their tasks. It is necessary that machine systems infer human intention in these systems. There were not enough considerations on state recognition process which is important to infer human intention. In this paper we first reconstructed human knowledge into a hierarchy and incorporated these knowledge into a Bayesian network. Next we modeled the state recognition process by using the Bayesian network.

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This paper describes the current status to develop an operator support interface system called semantic information presentation agent. The interface system uses a plant model composed of hierarchical relation among goals and functions, plant abstract structure, and component information on behavioral and operational aspects. A technique is developed to select candidates of counter action by considering operationality and rough effect estimation from the ones derived based on the functional plant model. Effect estimation is also carried out considering arrangement and rough behavior of components. A static simulation code is also developed to evaluate quantitatively the effects of a counter action. Reasonable selection and effect evaluation results are obtained by the proposed techniques in an application to an oil refinery plant.

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The authors have developed SEAMAID, which is a simulation-based evaluation and analysis support system for man-machine interface design in the domain of nuclear power plants. The SEAMAID simulates the interaction between an operator and human machine interfaces (HMI), and supports to evaluate the HMI by using the simulation results. In this paper, a case study of evaluation for conventional center control room design was conducted. The authors were confirmed that SEAMAID improves HMI design, and is a useful tool for iterative design.

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A computer-simulation-based approach has been proposed to estimate fundamental human error probability (HEP) parameters, which are required for Human Reliability Analysis (HRA) as related with Probabilistic Safety Assessment (PSA) for nuclear power plant (NPP). The target HEP parameters are the ones normally represented by time versus reliability correlation (TRC), which is called as "Time versus Cognitive Reliability" (TCR) curves. in the developed simulation system, "human model adjustment factors" were proposed to model the inherent variety and diversity in human cognitive behaviors, which are considered as the primary factors that generate different TRC. Inter-comparisons were also made between the TCR curves obtained from a laboratory experimental data and the ones obtained by the developed simulation system. It turns out that both TRCT curves agree well with each other for most cases. A discussion about the disagreements in the inter-comparison was also made to suggest the further subjects to enhance the performance of the simulation system.

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This paper describes the outline of an experiment to investigate the effect of an ecological interface for supporting situation awareness during malfunction of process. We developed an ecological interface and an conventional interface for the simulator SCARLETT of a virtual plant that have two process control modes ( automatic process control mode and manual process control mode). The purpose of this experiment is to investigate how interface and process control mode have any effects on situation awareness of human operator during malfunction of process, whether there are any interaction between interface and process control mode. We have been conducted this experiment now.

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In this paper, the contents of analysis and experiment of the contact stability of impedance control method implemented to a manipulator system are reported. To begin with, an impedance-controlled manipulator system which consists of industrial robot arm, force sensor, personal computer and impedance control software is explained. This is followed by the development of a dynamics model which analyzes the contact stability between the robot arm system and its object. Several simulation results are calculated. Furthermore, experimental results of contact tasks of the robot arm system are obtained. In conclusion, the effectiveness of simulation results of the dynamics model is verified in addition to the availability of the impedance control method implemented to the robot arm system.

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It is required for nuclear power plant construction to reduce construction cost and shorten construction period. An early and accurate construction planning including schedule coordination among the companies has recently become more important and it is possible to obtain necessary information for construction planning in early stage. In this situation, we have been developing a visualization system for construction engineering for nuclear power plants. This system has an interface with the existing Plant Layout 3D-CAD system and consists of three sub systems : (1) Scheduling and simulation system, (2) Yard planning system and (3) Scaffolding planning system. This paper describes overview of this system. This visualization system is very helpful for construction engineers to easily understand situation and environment around installation area, to easily plan a work sequence and confirm the planned schedule, and it is also effective for customers and workers to understand the planning. As a result, this visualization system enables safety and high quality construction.

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This paper proposes a TBNM (Team Behavior Network Model) that can simulate and analyze behaviors of an operator team in a dynamic and context-sensitive scenario. Case of FAB (Feed and Bleed) in PWR (Pressurized water reactors) under emergency situation has been studied. The result was compared with the experimental result and then applied to analyze the real event in nuclear power plant.

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We have developed HSI evaluation method based on a model of human cognitive processes in order to provide the viewpoint of the evaluation on the operability of interface. The model describes systematically the human error categories of GEMS by Reason. Based on the model together with reference to the other published information such as NUREG-0700Rev.1,the evaluation items for HSI have embodied in the electronic handbook HIBISCUS. The applicability of HIBISCUS have been assessed by evaluating experimental results using a simulator equipped with control panel using CRTs and touch operations. From the results, the usefulness of the handbook has been confirmed.

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This paper describes a query system to extract the structure of the conception of plant safety functions as a mental model of plant safety operation. The plant safety functions defined empirically as means to maintain plant safety. They are effective in a symptom-based operation in an emergency plant situation because the plant safety is maintained if the do not lose. The query system uses a cause-means analysis to derive plant safety functions where the relations between cause and means are analyzed from a top undesirable state of plant. The system performs the cause-means analysis by making questions corresponding to the answers of users about plant safety. The query system is composed of databases of sentences and vocabularies, query control subsystem and display subsystem of cause-means structure.

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The passive autocatalytic recombiner for flammable gas control in the light water reactor containment during postulated accident, has the unique feature which does not require any external driving force. However, the system performance has close interaction with thermal-hydraulic condition inside containment, and empirical correlations have been applied to date. Authors have developed mechanistic model which is applicable to extended atmospheric condition such as inerted containment. The new model has been incorporated into 3-dimensional CFD code, and verified against existing performance test data ranges from small to large scale. The analytical result shows good agreement with test data, and supports the model capability of recombiner performance prediction.

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JASMINE-pre is a thermohydraulics simulation code specifically developed for the prediction of pre-mixing process in FCIs. It consists of a newly developed melt model and a two-phase flow model ACE3D. The melt model has three components, jet, pool and particles. The jet and pool are modeled one-dimensionally in the vertical and radial direction, respectively, while the particles are modeled with Lagrangian grouped-particle concept. Molten jet breakup in non-boiling and boiling conditions were simulated to test the basic capability of JSAMINE-pre code. The calculation results showed that the code can simulate the molten jet breakup and multiphase mixing with and without steam generation.

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The Four years of the IMPACT, Integrated Modular Plant Analysis and Computing Technology'project Phase 1 have been completed. At the end of Phase 1,the Basic Single-, Two-, Multi-Phase Flow Analysis Modules of various coordinates have been parallelized. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The work is sponsored by the Ministry of International Trade and Industry, Japan.

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Some spent fuels from light-water reactors have been reprocessed in the U.K and France, and some of the high-level radioactive wastes generated by such reprocessing have bee returned to Japan. In order to ensure the safety sea transport of vitrified high-level radioactive wastes, thermal analyses of the packages were conducted under sea surface fire accidents. According to thermal analyses results of an exclusive ship using the thermal characteristic test results for materials which compose hatch cover members in a cargo hold, the thermal integrity of packages containing vitrified high-level radioactive wastes under sea surface fire accidents is consequently maintained both in the cases that the emergency water flooding system operates and does not operate.

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For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel is transported safety on the sea. To gain the public acceptance for the transport, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unknown reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1mSv/year).

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