The heat transfer from the containment inside to the atmosphere environment via the process of the wall steam condensation with non-condensable gases, the steel wall thermal conduction and the water film evaporation outside is the advanced technique for core residual heat removal during the long term depressurized stage of a nuclear power accident. Thus, the full pressure inner steam wall condensation coupled outer evaporation experimental platform has been carried out in China to take the thermal-hydraulic studies and to validate the passive safety system used in advanced PWR. The current work mainly discussed and analyzed the important wall heat transfer phenomena which were then used for the advanced PCCS design applications and safety evaluations. Necessities of the experimental studies on these evaporation and condensation phenomena were analyzed and proposed to carry out on a experimental platform. The experimental platform requirements were analyzed and then were specified for constructions. It also gave the advantages of the coupled heat and mass transfer study platform and further study objectives. Finally, the facility construction was summarized.

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In the event of a severe accident, it is necessary to prevent the release of radioactive materials into the environment due to damage to a containment vessel. It is also important to mitigate the consequences of such a severe accident. If debris (core melt and molten structure) falls down from a reactor vessel onto a pedestal floor, concrete will be eroded by the debris as a result of thermal decomposition and fusion. In addition, concrete decomposition will cause the production of non-condensable gases such as hydrogen and carbon monoxide, resulting in pressurization of the containment vessel. Therefore, we have been developing a safety system to stabilize the core melt and avoid a molten core concrete interaction. For these reasons refractory materials with high melting points should be installed. The refractory layer can protect concrete by the system like core-catcher. In order to estimate the effect of the refractory layer, it is necessary to consider about the reaction between debris and refractory layer. In this study, we investigate a eutectic reaction using a phase diagram. To apply the phase diagram to the evaluation method of refractory layer erosion, phase diagrams of debris and candidate refractory materials, such as ZrO_2, Al_2O_3, and MgO are required. In previous works, UO_2-ZrO_2 were studied and reported by many seniors as a debris composition, but there are no data for UO_2-ZrO_2-Al_2O_3 and UO_2-ZrO_2-MgO, such containing debris and refractory material composition. However there are some methods to predict phase diagrams from a database (TDnucl) base on existing binary phase diagrams. Therefore, in this work, phase diagram analyses were carried out and experimental verification data of UO_2-ZrO_2-Al_2O_3 and UO_2-ZrO_2-MgO were collected. At the representative debris composition, quasibinary phase diagram calculations and experiments were carried out. Experiments for the phase diagram were carried out by elevated temperature test. During these test, plateaus were detected from the temperature curve. From the results, for Al_2O_3 or ZrO_2 as refractory layer, it would be possible to use TDnucl. However, for MgO as refractory layer, it would be better to consider about oxidation states of phase diagrams. Using these phase diagrams, it would be possible to include the eutectic reaction to the evaluation method and it would be possible to predict more realistic situations.

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This paper presents a possible path for developing a RANS (Reynolds Averaged Navier-Stokes simulation) based CFD (Computational Fluid Dynamics) methodology applicable to real scale 217-pin wire wrapped fuel assembly of the KAERI (Korea Atomic Energy Research Institute) PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor). In this study, a RANS (Reynolds Averaged Navier-Stokes simulation) based CFD (Computational Fluid Dynamics) analysis using the General Grid Interface (GGI) function and innovative grid generation method was implemented in the 7-pin and 37-pin wire-wrapped fuel assembly, which use the general-purpose commercial CFD code, CFX. The RANS based CFD methodology can be successfully extended to the real-scale 217-pin wire-wrapped fuel assembly of the KAERI PGSFR. Complicated and vortical flow phenomena in the wire-wrapped fuel bundles were captured by a shear stress transport (SST) turbulence model, and by a vortex structure identification technique based on the critical point theory. The CFD results show good agreement with the friction factor correlation model, which can consider the number of wire-wrapped pins in the fuel assembly. The wire spacers induce a secondary flow by up to about 16 % of the inlet mean velocity magnitude. The secondary flow in a corner and edge sub-channel is much stronger than that of an interior sub-channel. The axial velocity, which is averaged in the corner and edge sub-channels, is about 30% higher than the axial velocity averaged in the interior sub-channels. Three-dimensional multi-scale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Large-scale and small-scale vortex structures are generated in the corner and edge, and the interior sub-channel, respectively. The behavior of the large-scale vortex structures in the corner and edge sub-channel are closely related to the relative position between the hexagonal duct wall and the wire spacer. It is expected that the multi-scale vortex structures in the fuel assembly play a major role in the convective heat transfer characteristics.

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This paper deals with rewetting phenomenon and its impact on the nuclear reactor safety. The complexity of the quench front propagation along the cooled surface, especially near the geometric elements such as spacers is described in detail. This observation is based on the new experimental study with bottom flooding configuration. Initial conditions of this experimental study are follows: surface temperature of 580 ℃, coolant mass flux of 150m^<-2>.s^<-1> and coolant temperature of 50 ℃. The presented results show very high complexity of rewetting phenomenon.

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Cold trap is a device for removal of corrosion products from liquid metal (Pb, Pb-Li17). Separations of these impurities occurs at lower temperatures than the operating condition. The goal of the cold trap is to efficiently cool the eutectic and one of the possible coolants is boiling water. Steam-water mixture created above the cooling part subsequently condenses at fins of tertiary cooling circuit with cold water. The created condensate will flow back and natural circulation will be provided. The main goal of this work is study of use of water boiling model provided and implemented in ANSYS FLUENT R15 program for the cooling purposes of the cold trap secondary circuit. Results will be used in the future for validating on the actual equipment, which is currently under development.

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The modular high temperature gas-cooled reactor could realize higher power and lower capital cost by combining multiple modular reactor system with a large capacity steam turbine unit. The HTR-PM plant based on dual-module reactors has been under construction in China. Meanwhile, another multiple module reactor system is also under research, in which, six modules of reactors are combined and connected to a 600MW steam turbine to make one power unit. Due to the shared equipment or systems in the secondary loop, the coupling effects among different modules posed great challenge to the designs and operating analyses for the multi-module reactors. In the present paper, the HTR-PM and six-module reactor systems were described and compared. The different number of modules and the system compositions in the secondary loops were focused and related to the coupling effects during operations. Then, the engineering simulator for six-module reactors was introduced for the operating analyses in the advantages of real-time calculation and coupled calculation in multi-module models. Starting from the full power operation condition, a typical operation process of reducing one reactor module's power from 100% to 50% was simulated for both the HTR-PM and six-module reactor systems. The results indicated that the dynamic change of one reactor had less influence on the others in the six-module reactors than that in HTR-PM.

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Graphite cross section differs among various evaluated nuclear data libraries as a result of distinct theoretical models and experimental technologies which are utilized by different nuclear data organizations all over the world. It has been revealed that graphite cross section plays a significant role in criticality calculation of graphite-moderated reactors in which graphite is one of the most important composing components; therefore, researches on the difference of graphite cross sections and the resulting effect on core parameters are highly recommended to be carried out. In the present paper, the differences of graphite capture cross sections in ENDF/B V, ENDF/B VII and ENDF/B VII.1 libraries were investigated in a unique group structure which is employed by the VSOP code. The resulting deviation of k_<eff> has also been investigated by running exactly the same VSOP calculation routine with each considered nuclear data library for HTR-PM model. The results show that there is obvious difference among these three ENDF/B library versions. The significances of graphite capture cross section vary with energy region, which have distinguishing influence on the reactor core parameters, including k_<eff>. The calculated k_<eff> are more sensitive to graphite capture cross sections in thermal energy region than that in epithermal energy region.

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Porous media model in Fluent code, coupled with two-phase mixture flow model, resistance model of tubes and heat transfer model through tubes, is employed to develop a steam generator thermohydraulics analysis code STAF ( Steam generator Thermohydraulics Analysis code based on Fluent). In this code, the heat transfer from primary to secondary side is calculated three-dimensionally during iteration. The localized velocity, temperature, enthalpy, quality and void fraction in steam generator can be obtained by this code. STAF is validated in two ways. First, STAF is used to calculate the thermal-hydraulic parameters in steam generator of AP 1000. The calculated results are compared with designed values to prove that the coupled heat transfer calculation in STAF is accurate. Second, STAF is employed to simulate the FRIGG test to validate the localized parameter calculation performance by comparing the calculated localized void fraction with test values.

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The Japan Society of Mechanical Engineers (JSME) is developing the codes and standards to reflect the lessons learned from Fukushima event and to comply with new regulation in Japan. This paper describes the activity to develop the evaluation guidelines for containment vessel structural integrity under severe accident condition, especially for PWR, which prescribe the methodologies to evaluate pressure and temperature resistance of containment structure under severe accident condition. The guidelines are also expected to be applied in designing process of equipment inside containment vessel to enhance safety margin of containment structure under severe accident. Following the completion of the guideline for BWR steel containment, the one placed at Fukushima power station and it is the one to be evaluated at first, the guidelines for PWR pre-stressed concrete containment and steel containment have been developed. The technical basis of the PWR guideline is taken from some literature data, mainly from the international research project; conducted by Nuclear Power Engineering Corporation and Sandia National Laboratories. Especially for pre-stressed concrete containment, the experimental result by the large scale mock-up was taken into consideration as technical basis to develop the guideline. On the other hand, for PWR steel containment, since there are no experimental results available that can be used as technical basis of the guideline, JSME has made much effort to develop the guideline for PWR steel containment, in contrast with the guideline for BWR which mainly developed based on the experimental results. Especially, for methodology to evaluate ductile fracture, JSME investigated the possibility to take FEM analysis approach, analyzing the behavior of steel containment focusing on the simplicity of the PWR's containment configuration. JSME has decided to develop the methodology to evaluate ductile fracture of steel containment, by verifying it with the experimental result of pre-stressed concrete containment as much as possible. The strain concentration factor which is important in the evaluation of ductile fracture is also established based on that of pre-stressed concrete containment, with conservative assumption. The guideline may further need more improvement, but it has achieved quick response to meet the needs of industry and regulatory.

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LiH is designated as a promising moderator and shielding material because of its low density, high melting point and large fraction of H atoms. Previous work was done to produce full and partial phonon spectra of LiH in a primitive cell and then be used to generate thermal neutron scattering cross sections of H in LiH and Li in LiH using the NJOY code system. A re-evaluate work with super cells was done in this paper to modify DFT (Density Function Theory) calculations' error in First Principles. The overall agreement of the lattice constants for pure LiH with the experiments is excellent if the zero-point motion is taken into account. From the theoretical calculation for electron (n)-doped LiH, it is indicated that metallic n-doped LiH was found to be a good superconductor. The phonon spectrum was predicted to be significantly softened, with both Li and H vibrations softened with the introduction of the dopant. Analysis was done in this paper to study thermal neutron scattering cross sections of this electron (n)-doped LiH. A comparison of the generated cross sections shows that accounting for the dopant introduced in the calculations affects the cross sections mainly in some energy range, and the phonon effect mainly occurs at low temperature, which is helpful to study the neutronic properties of LiH, especially for the reactors used LiH moderator.

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In ICONE22, the authors presented the Integrated Head Assembly (IHA) design concept implemented at Callaway Nuclear Power Plant in Missouri, USA. The IHA concept is implemented to reduce the outage duration and the associated radiation exposure to the workers by reducing critical path time during Plant Refueling Outage. One of the head area components in the IHA is a steel missile shield designed to protect the Control Rod Drive Mechanism (CRDM) assembly from damaging other safety-related components in the vicinity in the Containment. Per Federally implemented General Design Criteria for commercial nuclear plants in the USA, the design of Nuclear Steam Supply System (NSSS) must provide protection from the damages caused by a postulated event of CRDM housing units and their associated parts disengaging from the reactor vessel assembly. This event is considered as a Loss of Coolant Accident (LOCA) and assumes that once the CRDM housing unit and their associated parts disengage from the reactor vessel internals assembly, they travel upward by the water jet with the following sequence of events: Per Reference 1, the drive shaft and control rod cluster are forced out of the reactor core by the differential pressure across the drive shaft with the assumption that the drive shaft and control rod cluster, latched together, are fully inserted when the accident occurs. After the travel, the rod cluster control spider will impact the lower side of the upper support plate inside the reactor vessel fracturing the flexure arms in the joint freeing the drive shaft from the control rod cluster. The control rod cluster is stopped by the upper support plate and will remain below the upper support plate during this accident. However, the drive shaft will continue to accelerate in the upward direction until it is stopped by a safety feature in the IHA. The integral missile shield as a safety feature in the IHA is designed to stop the CRDM drive shaft from moving further up in the containment and damaging other safety components in the containment that are required to be operated for safe shutdown of the reactor. The missile shield in the IHA is designed to absorb missile energy due to an impact from missiles associated with a postulated CRDM housing break. This paper provides details of the CRDM missile shield design in the IHA for Westinghouse Pressurized Water Reactors (PWR) and it can be extended to other PWRs such as VVERs.

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This paper builds simplified physical model, according to the outside structures of the passive containment cooling system, to study the accident residual heat removal by air cooling when nuclear power plant accident occurs. Numerical simulation is used to set up the air circulation outside of the steel containment and provides the heat transfer process, including air temperature, velocity, and pressure for three different sections. By analysis, it also obtains and discuss the heat flux transferred out from the steel containment and the heat flux from the baffle plate to the decline section. Finally,it obtains the main heat transfer models for the decline section, the ascent section and the dome section. Results provide supports to enhance core residual heat removal.

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For the purpose of simulating annulus of Passive Containment Cooling System (PCCS) of AP type nuclear power plant, the flow field simulated by the similar structure of test facility (sub-scaled) should be similar with the actual PCS. Compared with the actual PCS, the downward air flow path of annulus usually isn't implemented in the test facility. The flow field near the entrance of upward air flow path for actual PCS and sub-scaled test facility has been analyzed, in order to optimize the design of this structure designed in the test facility. The result shows the effect of flow deflector of upward air flow path entrance is obviously well. The flow field distribution of upward air flow path entrance is very uniform. By the simulating result, the entrance flow field of rising segment is uneven distribution, due to the flow reversal. The optimized result suggests that the deflector angle of 30° makes the entrance flow field of test and the actual PCS in good agreement, and suitable for test measurement.

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C-ring, whose sealing behavior is the key to the sealing structure, is one of the main seal-rings for Reactor Pressure Vessel. A finite element model with less simplification has been established using ANSYS APDL to simulate the sealing behavior of C-ring. The numerical results, including sealing behavior curve, specific pressure and springback value, have been compared with test results at 7.67%, 8.42% and 8.09% compression ratios respectively. It confirms the validity of this numerical simulation method. With the numerical method, the deformation and stress distribution and sealing behavior curve of C-ring can be obtained. The behavior curve can be divided into 5 sections. At 10% compression ratio, the peak contact stress is 501MPa, the contact width is 3.22mm, the total springback value is 0.460mm, the effective springback value is 0.332mm and the linear load is 788N/mm. As C-ring is composed of three parts, three models, i.e. spring-only, spring with inner lining and spring with two linings, have been simulated to study the effect of the contact states between each parts on sealing behavior.

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Liquid lithium is a promising plasma facing material in magnetic fusion devices. However, liquid lithium water interaction under the conditions of loss of coolant accidents is a principal concern to the safety of fusion devices. The prediction of explosion strength of liquid lithium water interaction is significant to the analysis and assessment of related accidents in fusion reactors. As a preliminary investigation, an experiment of lithium water interaction on the water has been conducted. The pressure and temperature in the test section were recorded during the violent interactions. The results indicate that the mass of lithium, initial lithium temperature and init ial water temperature are key factors to the explosion strength. It is a kind of complicated nonlinear relation between explosion strength and its influencing factors. Therefore, a BP neural network model for predict ing the explosion strength has been developed and the prediction results are consistent with the experimental data. The BP neural network prediction model is applicable and provides a novel method for the evaluation of liquid lithium water interaction.

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In a Nuclear Power Station, some austenitic 316H boiler platen support beams have been identified as potentially susceptible to reheat cracking, because their welds did not receive post weld heat treatment and they have been operated at sufficiently high temperatures. To ensure the integrity of these support beams, creep-fatigue crack growth assessments and creep deformation calculations for postulated extended, surface-breaking defects at the welds have been carried out using the latest R5 procedures, the plant operating temperature histories and welding residual stresses that had been calculated by finite element analysis. Limiting defect sizes have been calculated following the R6 procedure. The crack growth assessment has assumed a lifetime of 350,000 hours, and has conservatively assumed defect initiation at 510 hours in operation and an initiation crack depth of 2.3mm. Two base cases and three sensitivity studies have been investigated for all material regions including 316H parent material and Heat Affected Zone. The assessment results show that for this postulated initiation time and defect depth, integrity of the weld has been demonstrated to the lifetime of 350,000 hours.

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It is important to investigate failure mode and ultimate strength of piping components in order to evaluate seismic integrity of piping. Many failure tests of a thick wall and a high pressure piping for Light Water Reactors (LWRs) have been conducted, and the results suggest that the failure mode which should be considered in the design of a thick wall piping for LWRs under seismic loading is low cycle fatigue. On the other hands, the piping in Sodium cooled Fast Reactors (SFRs) is a thin wall configuration compared to the piping in LWRs. Failure tests of a thin wall piping is necessary because the past failure tests for the piping in LWRs are not enough to discuss failure behavior of a thin wall piping. This present work investigated the failure mode and the ultimate strength of a thin wall tees, the critical parts in seismic evaluation of the actual piping of SFRs.

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The effectiveness of in-vessel retention (IVR) by external reactor vessel cooling (ERVC) strongly depends on the critical heat flux (CHF). As long as CHF is not exceeded, the vessel lower head can be cooled sufficiently to prevent its failure. In this paper, a mathematical model of inverse heat conduction problem (IHCP) was presented to identify the critical heat flux at the outer surface of the heating block according to the measured temperature distributions of the K-type thermocouples inserted into it. An inverse numerical algorithm based on the control volume approach (CVA) was therefore proposed to solve the IHCP using temperature measurements. Unknown critical heat flux on the heated surface was estimated from knowledge of the measured temperatures. The calculated results showed that the IHCP could accurately capture the actual heat flux in comparison with the experimental data obtained from the FIRM test facility which was conducted by State Nuclear Power Technology Research and Development Center (SNPTRD).

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This research deals with the enhancement for moisture separation efficiency of moisture separator reheaters (MSRs) using wire-meshes. MSR is a facility which is used to improve the thermal efficiency for the nuclear power plant. The exhausted wet steam from the high pressure turbine exit is transported to the MSR inlet through pipes. The microdroplets in the wet steam are eliminated by the wave-shaped vanes which are installed at the MSR inlet. However, complete elimination of the moisture is difficult. Since, the small droplets are carried over (CO) through the wave-shaped vanes. The wave-shaped vanes can eliminate the droplets completely when the droplet diameter is larger than the threshold diameter, D_<th>. The CO droplets are heated through the reheaters using the extracted steam from the high-pressure turbine inlet. If the microdroplets were 100% eliminated, the phase change enthalpy is saved, of which order is Mega Watts, therefore, the thermal cyclic efficiency can be improved. We investigated installing wire-meshes at the inlet of the wave-shaped vanes to enhance the efficiency of the microdroplets capture by the enlargement of the microdroplets. In this study, we evaluated the effect of the wire-meshes upon the microdroplet diameter enlargement by experiment. The experimental parameters were the diameter of the wire and the number of the layers. The working fluids were the air and the water which simulates the wet steam and condensed water droplets in the MSR. The droplet diameters were measured by real-time image processing system which consisted of the CCD camera and single pulsed laser light source. The results showed that the microdroplet diameter increased as the number of layers increased. We evaluated the moisture defined by the mass fraction of the droplets with diameter less than D_<th>. The most effective parameters for the moisture decrease were the wire diameter 0.65 mm, the opening 1.47 mm and 6 layers in the range of present experiments. At the configuration, we conducted the visualization of the outlet flowfield using high speed camera. From the visualization results, it was observed that liquid film was formed over the wire-meshes and the surface wave of the liquid film induced the detachment of the enlarged droplets from the liquid film.

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Efficient and effective education and training of nuclear engineering students and future operators are critical for the safe operation and maintenance of nuclear power plants. Students and future operators used to receive some of the education and training at university laboratories and research reactors. With many university research reactors now shutdown, both students and future operators are deprived of this valuable training source. With an eye toward this need and to take advantage of recent developments in human machine interface technologies, we have focused on the development of 3D virtual laboratories for nuclear engineering education and training as well as to conduct virtual experiments. These virtual laboratories are expected to supplement currently available resources and education and training experiences. Resent focus is on adding interactivity and physics model to allow trainees to conduct virtual experiments. This paper reports some recent extensions to our virtual nuclear education laboratory and research reactor laboratory. These include head mounted display as well as hand tracking devices for virtual operations.

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The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology. Integrated valve (IV) is the main control unit of the whole system. Transient flow resistance of IV is the key parameter of the transient flow analysis of the CRHDS which lays the base for the vibration reduction research of the CRHDS. The structure and working principles of IV are introduced. Different flow channels are formed when the IV core piston moves to different positions during the valve shutting off or opening process. The IV differential pressures corresponding to varied inlet flow velocities were obtained through modeling and calculating different flow channels. Flow fields of the IV are discussed. The change of flow resistance coefficient of the valve opening process was found. The resistance coefficient combined with the dynamic model of valve core opening process and the water hammer equations of CRHDS result in the dynamic pressure of the IV transient flow test loop. The calculation results agree well with the experimental results and the validity of the transient resistance model is proved. Through sectional division of the flow channels, the pressure drop between different sections along with the change of the piston position was obtained and the main flow resistance part of the IV was also found. The results provide solid basis for the design and optimizing of the IV. The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology based on the hydraulic control rod driving system (HCRDS) and the commercial PWR magnetic jack (Bo, 2005). The CRHDS is composed of circulating pump, filter, integrated valve (IV) and driving mechanism in which the IV is the main control unit of the whole system. The pulse water entering into the driving mechanism controlled by the IV drives the pin claw move so as to achieve the step up, step down and scram function. The IV is a direct action solenoid valve. Because of its fast opening and closing operation, there is an obvious transient flow process in the hydraulic drive circuit. This process combined with the working process of the hydraulic cylinder will cause high pressure pulsation in the circuit. It affects the normal operation of the pump and brings pipe vibrations and noise. Therefore, it is necessary to study the transient flow of the drive system caused by the IV operation. The IV steady state flow resistance (Cai, 2008; Qin, 2010) and electromagnetic force (Liu, 2011) have been analyzed experimentally and theoretically in recent years. The IV transient flow process is the main work of this paper. Transient flow resistance of IV is the key parameter of the transient flow analysis of the CRHDS. The structure and working principles of IV are presented. The flow field of the valve core of different working positions is analyzed by FLUENT. The change of flow resistance coefficient of the valve opening process is obtained. The flow resistance coefficient as a boundary condition combined with the water hammer equations of CRHDS result in the dynamic pressure of the IV transient flow test loop. The calculation results are validated by the experimental results. On this basis, the flow resistance during IV opening process is analyzed which provides guidance for the IV design and optimization.

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The progress of the 2011 Fukushima Daiichi nuclear disaster has been calculated by severe accident analysis codes in order to understand the causes of accidents and the current status of the reactors in the Fukushima Daiichi nuclear power plants. However, effects of seawater are not considered in these calculations, although the seawater has been attempted to inject into the reactors to cool down the nuclear fuels. The seawater provides a potential to affect the heat transfer due to the changes of the physical properties of the coolant. Besides, deposition of a scale may change thermal-hydraulic performance in the reactor pressure vessels. Although it is important to clarify these effects of the seawater on cooling of a heated surface, the conventional researches which evaluated these effects of the seawater quantitatively are hardly found. In order to make clear these effects, we have conducted an experiment with an internally heated vertical annulus using manmade seawater as a working fluid to measure basic thermal-hydraulic behavior of the seawater. In the present study, we measured heat transfer coefficient, boiling behavior and flow velocity distributions in the internally heated vertical annulus. Working fluids were pure water, manmade seawater and NaCl solution. The heat transfer coefficients were estimated by measuring temperature differences between a heater rod and these working fluids. The boiling behavior was observed with backlight of LEDs. The flow velocity distributions were estimated by a PIV (Particle Image Velocimetry) with fluorescent particles. These observations were conducted with a high speed video camera. From these experiments, the heat transfer coefficients of the manmade seawater was the same as that of the pure water in the heating condition without boiling. However, in boiling conditions, the heat transfer coefficients of the manmade seawater and the NaCl solution was larger than that of the pure water. From the observations with the high speed video camera, it is considered that the differences are caused by differences of boiling behavior in each working fluid. The velocity distributions in the manmade seawater and the NaCl solution were also different from that in the pure water in the boiling conditions. Therefore, although there is no obvious difference of the thermal-hydraulic behavior in the pure water, the manmade seawater and the NaCl solution in single phase flows, the thermal hydraulic behavior of the boiling flows is different in each fluid.

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This paper focuses on a generic Pressure-Channel (PCh) SuperCritical Water-cooled Reactor (SCWR) concept, which is one of the six Generation-IV nuclear technologies, featuring two-pass (re-entrant) counter-flow fuel-channel assembly. This generic PCh SCWR contains 336 vertical pressure channels (pressure tubes) and has a thermal power of 2540 MWth. Coolant is light water, and its temperature varies within the range of 350-625°C at a pressure of about 25 MPa. The objective of this paper is to perform a onedimensional steady-state thermal-hydraulic analysis of the reentrant-channel design with different types of fuels, namely UO_2 and UN. The thermal-hydraulics analysis was conducted while the 78-element fuel-bundle design was chosen as the reference fuel bundle.

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Modern Generation-III Nuclear Power Plants (NPPs) equipped with water-cooled reactors have gross thermal efficiencies of approximately 30-36%, which are significantly less than those of advanced fossil-fuel and natural-gas thermal power plants (55-62%). Therefore, global effort to progress Generation-IV reactor concepts and NPPs are required to meet the demand for clean, non-fossil-based electrical production. The main objective of this paper is to determine a pressure drop across a fuel channel of a SuperCritical Water-cooled Reactor (SCWR). A generic 1200-MWel SCWR has inlet and outlet temperatures of 350°C and 650°C, respectively, and an inlet pressure of 25 MPa. With such high operating temperatures and pressures, an SCWR NPP can achieve gross thermal efficiencies of approximately 45 - 48%, which is a substantial improvement over the currently operating Generation-III NPPs. For this purpose, a Re-Entrant-Channel design and its associated 78-element fuel bundle are selected as a basis for this analysis. An investigation of a pressure drop resulting from friction, gravity, acceleration and local losses at supercritical conditions has been performed on a basis of one-dimensional steady-state analysis. With this objective, a thermal-hydraulic code has been created with MATLAB, which calculates the pressure drop across a 5-m heated length of a vertical fuel channel. The total pressure drop across the channel was estimated to be about 60 kPa.

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Generation III and III+ reactors are currently in operation globally and Generation IV (Gen IV) reactor designs are being developed. There are six Generation IV nuclear systems under development worldwide, namely, Very High Temperature Reactor (VHTR), Molten Salt Reactor (MSR), Gas-cooled Fast Reactor (GFR), Sodium-cooled Fast Reactor (SFR), Leadcooled Fast Reactor (LFR) and SuperCritical Water-cooled Reactor (SCWR). Of these six systems, Canada has decided to pursue the SCWR as its choice for a Gen IV reactor. One major advantage of the SCWR is an increase in thermal efficiency from the 30-35% range of current nuclear power plants to approximately 40-45%. SCWRs operate well above the critical point of water at a pressure of 25 MPa and reactor outlet temperatures up to 650°C. Due to the operating conditions, current fuel-channel designs cannot be used in the SCWR and new fuel-channel design concepts are under development. One such concept is called the Re-Entrant Channel (REC). The REC is a vertical channel and consists of two tubes, the inner (flow) tube and the pressure tube. The inner tube is hollow and the fuel bundles and insulator are located in the pressure tube. An annulus is formed between the flow and pressure tubes. A perforated liner is used to protect the insulator from the fuel-string, and the pressure tube is in contact with the moderator. The coolant flows from the top of the channel to the bottom via the flow tube and then reverses its direction and flows upwards through the pressure tube. The objective of this work is to determine the heat loss from a REC for a generic 1200-MW_<el> SCWR. A onedimensional onedimensional numerical model was developed in MATLAB to calculate the temperature profiles, the heat transfer coefficients and the heat loss from the coolant to the moderator for a given set of flow, pressure and temperature boundary conditions. With the results from the numerical model, the design of the REC can be optimized to improve the efficiency.

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For the purpose of confirming failure modes and safety margin, some studies on the ultimate strength of thin-walled piping components for Sodium-cooled Fast Reactors (SFRs) under extreme loading conditions such as large earthquakes have been reported these several years. Nonlinear finite element analysis has been applied in these studies to simulate buckling and yielding with large deformation, whose accuracy is dependent on the element type, the mesh size, the elasto-plastic model and so on. It is important to check the validation of a finite element model for nonlinear analysis especially under extreme loading conditions. This paper presents static and dynamic analyses of a thin-walled elbow with large deformation under large seismic loading, and discusses the validation of the FEA models comparing with experimental results. The finite element analysis models in this study are generated by shell elements for a stainless steel pipe elbow of diameter-to-thickness ratio 59:1 similar to the main pipe of SFRs, which is used for shaking table tests. At first, a static analysis is carried out for an in-plane monotonic bending test, in order to confirm that the shell element is appropriate to the large deformation analysis and the material parameters are proper for the strain level in the experiments. And then, a dynamic in-plane bending test with the maximum acceleration of 11.7G is simulated by the nonlinear FEA with stiffness-proportional damping. The influence of mesh sizes on results is investigated, to determine proper mesh sizes and reduce the computational cost. Finally, comparing the results of the FEM analyses with those of experiments, it is concluded that the appropriately generated FEA models are effective and give accurate results for nonlinear analyses of the thin-walled elbow under large seismic loading.

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A visual experimental investigation was carried out to study the flow boiling heat transfer in a circular channel with diameter of 6.78 mm and two transparent sections (100 mm) at the two ends of the channel. A high speed camera was used to record the bubble behavior and the flow patterns at the exit of the channel with 5,000 FPS (Frame Per Second). The working fluid was deionized water, and the subcooling at the inlet was 10 K with constant system pressure of 0.101MPa. The mass flux ranged between 300 kg/m^2s and 700 kg/m^2s with different heat fluxes. It was found that the heat transfer coefficient increased with increasing vapor quality for different mass fluxes since the heat flux increased with other identical working conditions. Also, the heat transfer coefficient was inversely proportional to the liquid film thickness in annular flow. With increasing heat flux and a constant mass flux, the heat transfer coefficient of annular flow increased and fluctuated more intensely, because the liquid film thickness decreased and fluctuated intensely.

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Application of ceramics to fuel material is expected to enhance safety for light water reactors due to its chemical and mechanical stability at high temperature, compared with that for metal. Silicon carbide (SiC) ceramic is one of the candidate materials for accident tolerant fuel (ATF). The research and development projects that apply SiC/SiC composite (i.e. a SiC matrix reinforced by SiC fibers) to fuel cladding and BWR channel boxes are currently being conducted in Japan and the US. Toshiba and the partner organizations have launched the research and development efforts, which primary objective is to enhance the safety for light water reactors (LWRs) with the application of ceramics to core material. In collaboration with the partners, Toshiba has initiated a project to develop SiC composite for the application to light water reactors. The objective of the Phase I is to develop a fabrication process with the use of trial models, aiming for future production of e full-scale fuel cladding and channel box made by SiC/SiC composite. During the Phase, trial models in reduced-scale were created with SiC/SiC composite.

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Structural concrete used as biological shield in nuclear power plants (NPP) become radioactive after exposure to neutron radiation. Due to radiative capture interactions, artificial radionuclides are generated to high enough concentrations causing such a concrete to become classified as low-level radioactive waste at the time of NPP's decommissioning. Disposal of this concrete adds to the expense of financing and constructing of a nuclear power plant. Three such radionuclides, namely Co-60, Eu-152, and Eu-154 are shown to account for 99% of total residual radioactivity of decommissioned concrete. The IAEA document RS-G-1.7 Application of the Concepts of Exclusion, Exemption, and Clearance, specifies clearance levels in terms of radionuclides' specific activities. The specific activity of 0.1 Bq/g for Co-60 and Eu-152, and Eu-154 defines criteria for low-activation concrete that can be recycled after decommissioning of the plant. In this paper we present the methodology and the challenges associated with the detection of trace amounts of cobalt and europium in concrete aggregates and how these detection limits affect the accuracy of application of the IAEA regulations on the clearance levels. We used the neutron activation analysis (NAA) as a method to test the detection limits on trace elements in samples of cement, coarse, and fine aggregates from the local suppliers in the State of Utah. These samples were irradiated at different reactor power levels ranging from 10kW to 90kW for time periods of 1, 3, 30, 60, and 120 minutes, with the goal to find if there is a threshold set of the NAA parameters in detecting trace amounts of these isotopes. Each of the samples is counted on a Canberra BEGe High Purity Germanium detector. Cement samples were concurrently irradiated with a NIST coal fly ash standard reference material and coarse and fine aggregates with Montana soil standard reference material in order to accurately quantify the mass concentration of the isotopes. The final results are showing that the reactor power, irradiation and detector measurement times, are heavily correlated in finding the optimum combination of these NAA parameters for detection of trace contents of cobalt and europium. This research represents the first such study in the USA. We have been able to develop the methodology with the goal to identify what sources of cement, fly ash, sand, and coarse aggregate are having cobalt and europium concentrations below the established IAEA clearance levels. Similar studies were firstly developed in Japan a few years ago, in identifying cobalt and europium to be the elements causing concrete activation at the time of power plant decommissioning.

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A three-dimensional nuclear chart is constructed with toy blocks for usage of outreach activity related on nuclear physics and atomic energy. The height of each block represents quantities like atomic mass per nucleon, the total half-life, etc. The bulk properties of the nuclei can be easily understood by using these charts. Explanations for the energy generation of nuclear fusion and fission are visually given. In addition, we newly set another chart with blocks of fission fragment mass distribution from U-235 + a thermal neutron. As an example, the origin of abundances of rather radioactive isotopes like Sr-90 and Cs-137 is explained which created in nuclear reactor and also distributed in the eastern side of Fukushima prefecture due to the accident of Fukushima-Daiichi Nuclear Power Plant. Using our charts, lectures entitled 'Alchemy of the Universe' were delivered to high schools and public places.

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In the EU and the USA, Safety Analysis Report (SAR) is made as a comprehensive document concerning plant safety, which covers the areas of design, construction, operation, maintenance as well as safety related activities (quality insurance, safety culture etc.) at the construction stage in the beginning. This document is maintained in the state of "as-is" with added modification during operation, and used as a significant tool to control the plant safety. On the other hand, Japanese plants currently have not had such a scheme to check comprehensively the safety of the plant and to maintain its report, so there are issues whether in a timely manner they have assessed the safety level and taken proper measures, and have been sufficiently accountable for the safety to public. JANSI is developing the guideline for Japanese Safety Assessment Report (JSAR) to voluntarily check the plant safety comprehensively and periodically. The development follows international standards such as R.G.1.206 (Guide for Combined license applications for nuclear power plants) of NRC as well as Japanese regulatory requirements. JSAR will have two stages. The first one is the USA type SAR, and the second one is the EU type PSR JSAR is expected to contribute the enhancement of Japanese plant safety. This development could cause the following expected results: &radic; Confirm the latest safety level and conduct effective measures in a timely manner &radic; Enhance the competency of plant personnel &radic; Sufficiently accountable to public and foreign countries &radic; Streamline the regulatory process in the future

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PRA parameter estimation has been studied in Japan since 2009. For improving component failure rate estimation, new estimation methodology was discussed and is planned to be applied for future parameter estimation. This methodology includes selection of the normal and half-Cauchy distribution as the new hyper-prior distribution. In addition, component failure rates of U.S. commercial plants are considered as prior information. Other PRA parameters to be estimated and the event database for future parameter estimation under development in Japan are discussed

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New nuclear plant technology will rely heavily, if not exclusively, on digital equipment. Obsolescence of digital I&C equipment is an inevitable part of plant technology life cycle for new and existing plants. Developing an overall strategic plan can mitigate some of the risks associated with obsolescence. Moreover, when developed as part of an overall lifecycle management plan (LCMP), a strategic obsolescence management approach can identify steps that can be taken at early stages of the technology life cycle to cope proactively with the obsolescence of equipment. Recent work within the Advanced Nuclear Technology (ANT) program at the Electric Power Research Institute (EPRI) has developed guidance and methodologies for determining when digital obsolescence is likely to occur, the extent to which it can occur, the risks and impacts due to obsolescence, and strategies that can be used to minimize its effects, all in the context of system lifecycle management planning (LCMP). Worksheets for assessing obsolescence risks and the applicability and limitations of management strategies were developed as part of this work, and can be used to create or supplement a strategic obsolescence management plan.

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To avoid failures of small bore piping connections caused by high cycle fatigue, it is important to measure the stresses around the connections. To measure such stresses, the authors have developed an easily-attachable and detachable strain measurement tool which utilizes strain gauges in combination with our patented strain gauge holder. Traditionally, strain gauges have been bonded to piping surfaces using adhesive; however, with the newly-developed measurement tool, bonding adhesive is no longer necessary. The tool can be installed quickly and easily on a piping surface and measure the strains on the piping as accurately as adhesively-bonded strain gauges. Accordingly, the new strain measurement tool significantly reduces the work time without affecting the measurement accuracy.

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The reprocessing of spent nuclear fuels at the Rokkasho reprocessing plant generates radioactive wastes in solid and liquid forms. These wastes contain a mixture of radioactive nuclides such as Transuranium (TRU) elements, fission products (FP), and corrosion products (CP). The waste processing methods prior to disposal depend upon the waste characteristics (solid, liquid, or gas), and the disposal methods (shallow land disposal at a few tens of meters below ground level, mid-depth disposal at 50 to 100 meters below ground level, or geological disposal at more than 300 meters below ground level) depend upon the type and extent of radioactivity in Japan. In the Rokkasho waste management program, radioactive waste is segregated into vitrified high-level liquid waste, iodine waste absorbed on silver, waste from cutting spent fuel assemblies, low-level liquid waste, and miscellaneous solid waste. The last one is difficult to characterize as this waste is generated from various locations in the plant. Japan Nuclear Fuel Limited (JNFL) is developing a method for determining the radioactivity of radioactive waste generated from the Rokkasho Reprocessing Plant. Miscellaneous solid waste in particular is generated in large volumes and has a wide distribution of radioactivity. Therefore, the establishment of a radioactivity determination method for this waste is an urgent issue. Important nuclides were selected from the viewpoint of waste classification for individual disposal and safety assessments. The behavior of some important nuclides at the Rokkasho Reprocessing Plant, such as C-14, Sr-90, Tc-99, I-129 and actinides, was examined. A radionuclide inventory of the spent nuclear fuel was calculated by combustion calculation code ORIGEN 2.2, and nuclide compositions at each processing line were evaluated by using the "Reprocessing Process Activity Balance Evaluation System" under development by JGC Corporation. As a result, the "Reference Nuclide Spectral Method" can be applied to miscellaneous solid waste as if the Rokkasho Reprocessing Plant is sectioned into several areas.

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The behavior of the graphite dust is important to the safety of High Temperature Gas-cooled Reactors. The present study focuses on the forces which make the graphite dust attach or detach from the surface in HTGR. The effect of graphite dust size, the fluid velocity and the surface energy between the particles and the substrate were investigated. The result showed that van der Waals adhesion force is the main factors affecting the dust attach on the surface, the gravity force and the electrostatic force were much smaller than it. For small particles, both the aerodynamic lift and drag are smaller than van der Waals adhesion force. While for the large particles, the coupled effects of aerodynamic lift and drag can make the dust detach from the substrate easier. Both the aerodynamic lift and drag forces will increase quickly as the fluid velocity increases. The surface energy is an important parameter for van der Waals adhesion force, which will decrease as the surface energy decreases.

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Two-phase flow parameters were obtained for air-water flow in a rectangular channel which is 15 cm in hydraulic diameter. The test section includes horizontal, inclined and vertical channel sections in a loop test facility. A double conductivity probe method was employed to measure the flow parameters. Air was injected from the top side of the channel which induced the natural circulation of water inside the test section. The natural circulation flow rate of liquid was measured by a paddle type flow meter installed at the downcomer side in which only a single phase liquid flow existed. The probes mounted on the top of the test section were set at various locations along the channels. From the raw probe data, the experimental results are presented on the statistical averaged data of local void fraction, gas velocity and bubble chord length distribution. In addition, the turbulence intensity was deduced from fluctuation of gas velocity. The flow regime observed through high speed camera indicated the existence of elongated bubble flow in the inclined section and bubbly or transition from bubbly to churn turbulent flow in the vertical section. The data on flow parameters associated with its behavior in the channels facilitates the understanding of twophase flow in large diameter pipes which has an extensive application in nuclear engineering.

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Subcooled heating/boiling process under pool condition is widely seen in present engineering applications such as nuclear spent fuel pool and reactor core under maintenances and/or accident scenarios. In order to understand the heat transfer characteristics and possible enhancement mechanism of using ultrasonic vibration to improve the heat removal performance and safety operating ranges, detail investigations are required. In this study, experimental tests are carried out to investigate the heat transfer performance under pool natural convection and subcooled boiling conditions, and ultrasonic vibration is introduced to evaluate the possible enhancement performance. The commercial stainless-steel circular-rod heaters and ultrasonic vibrators are utilized to generate the heat flux of 4 × 10^3 × 2 × 10^5 W/m^2 and ultrasonic vibration waves with a frequency of 40 kHz and a power of 150W. Experimental tests are divided into two groups: (1) pool natural convection and subcooled boiling tests without vibration (stationary), and (2) pool natural convection and subcooled boiling tests with ultrasonic vibration. These tests are carried out under atmosphere pressure (1atm) with three subcooled pool temperatures. Detailed experimental data such as surface temperature, fluid temperature and heater power are recorded during the tests, and the heat transfer coefficients are analyzed for different operating conditions. Through this study, an experimental database of pool-subcooled heat transfer from single phase natural convection to pool boiling is established, and the heat transfer enhancement performance is identified and determined under with and without ultrasonic vibration conditions. The results shows that the heat transfer enhancement is higher in single phase than in subcooled boiling. The heat transfer augmentation is better in high heat flux condition. This paper can provide clear experimental evidences and better understanding about the influences of ultrasonic waves on heat transfer performance in pool conditions, and the experimental database can be useful for future safety design of reactor core and fuel storage.

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Loss-of-Coolant-Accident (LOCA) is a classical design basis accident which would be considered in LWR safety analyses. The development of LOCA simulation technique can benefit from thorough understanding of local thermal-mechanical states of the axial load on the cladding. This paper describes some analysis results of the updated RANNS code developed by Japan Atomic Energy Agency (JAEA) as well as LOCA quench experiments. This update includes equations which are newly derived in a study based on published data regarding the thermal diffusivity and the thermal expansion. In addition, the measured axial load gain of a single-sided oxidized cladding was compared with the simulation results using the updated RANNS simulation, and they agreed quite well. Finally, the effect of the axial temperature profile of the cladding on the axial load gains is discussed and it is found that this effect is small.

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This paper describes the ongoing activities for the development of structural codes in Japan Society of Mechanical Engineers (JSME) for fast reactors including the Japanese prototype reactor Monju, and the demonstration reactor JSFR (Japan Sodium-cooled Fast Reactor), which is under conceptual design study. The existing design and construction code and welding code will be upgraded to support the design and construction of JSFR which involves various new features, among which are 60-year design and the application of new materials. Moreover, a fitness-for-service code, leak-beforebreak evaluation code and guidelines for structural reliability evaluation of passive components will be newly developed to be applied to the operation and maintenance of Monju and JSFR. Major revisions to incorporate the development are planned 2016 and 2020. Also discussed is an international perspective of codes and standards development and where the JSME Fast Reactor Codes stand in it.

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In the present study, the steady state performance of a natural circulation loop with inclined heat source is investigated. The rectangular single phase natural circulation loop consists of a vertical cooling pipe, an inclined heating pipe and two horizontal pipes made of steel, connected by means of four 90° bends. The loop has an imposed heat flux in the heating section. The parameters investigated were: power transferred to the fluid, mass flow rate and inclinations of heating pipe. The generalized dimensionless groups are established which are useful in comparing the performances of different loops and to extend data from small scale loops to the prototype. The capability of the computer code ATHLET, developed by GRS, for predicting the stable behavior of single phase natural circulation loops is tested by simulating rectangular NCLs facilities. Then, the dimensionless groups of the present apparatus are developed based on Vijayan's model taking into account of the heating pipe inclinations. The steady-state analysis results show a good agreement with the correlation developed by this study, which suggests that simulation of the steady state flow in single phase NCLs can be well achieved by the non-dimensional parameter Re, Gr_m and N_G.

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One limitation today in simulating horizontal segregated flows is that there is no treatment of droplet formation mechanisms at wavy surfaces. For self-generating waves and slugs, the interfacial momentum exchange and the turbulence parameters have to be modelled correctly. Furthermore, understanding the mechanism of droplet entrainment for heat and mass transfer processes is of great importance in the chemical and nuclear industry. A further step of improvement of modelling interfaces is the consideration of droplet entrainment mechanisms. The proposed entrainment model assumes that due to liquid turbulence the interface gets rough and wavy leading to the formation of droplets. The new approach is validated against existing horizontal two-phase flow data from the HAWAC channel.

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The passive safety PWR, such as AP1000 and CAP1400, becomes a mainstream model to be built around the world. The largest part of coolant inventory of passive core cooling system (PXS) is stored by In-containment refueling water storage tank (IRWST). And the stability of its draining to the core is very important for the core cooling under accident condition. And the relationship between flow rate and pressure drop plays an important role on the two-phase flow oscillation. This study will investigate the relationship and its impact on the gravity draining stability. An analytic model is built for the purpose of analysis based on a quasi-steady state, in which the flow path is divided into two sides which are IRWST injection single-phase side and ADS4 riser two-phase side. Through the analysis, the pressure and the flow rate is substantially a concave curve, and this possesses the system a potential for oscillation instability. The stability is also determined by the system boundary conditions, which includes the core decay power, injection and venting line resistances, coolant subcooled temperature and other parameters. The analysis result shows a good agreement with the ACME test result.

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It is one of the efficient approach to reduce the memory consumption in Monte Carlo based reactor physical simulations by using the On-the-fly Doppler broadening for temperature dependent nuclear cross sections. RXSP is a nuclear cross sections processing code being developed by REAL team in Department of Engineering Physics in Tsinghua University, which has an excellent performance in Doppler broadening the temperature dependent continuous energy neutron cross sections. To meet the dual requirements of both accuracy and efficiency during the Monte Carlo simulations with many materials and many temperatures in it, this work enables the capability of on-the-fly pre-Doppler broadening cross sections during the neutron transport by coupling the Fast Doppler Broaden module in RXSP code embedded in the RMC code also being developed by REAL team in Tsinghua University. Additionally, the original OpenMP-based parallelism has been successfully converted into the MPI-based framework, being fully compatible with neutron transport in RMC code, which has achieved a vast parallel efficiency improvement. This work also provides a flexible approach to solve Monte Carlo based full core depletion calculation with many temperatures feedback in many isotopes.

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The KA facility in Rokkasho Reprocessing Plant started the active tests to solidify HAW into the glass in 2007 which was the examination of the final stage before the operation, but the active test had to be discontinued due to the trouble of glass melter operation with down of pouring by deposit of noble metals on the melter bottom. After the K-melter equipment and operating conditions were improved in response to the result of the mock-up tests, a series of active tests were restarted including pre-confirmation tests in May, 2012 by operating each one of two K-melters for the purpose of confirming the stable condition before having the pre-use inspection. One year later, these tests were finished with enough confirmation of stability in the state such as glass temperature and controlling the noble metals, and also confirmation of the processing capacity. This paper gives the successful situation in latest active tests of the KA facility by processing the actual waste including the implementation contents in the full scale mock-up tests using KMOC-melter that examined to identify the cause and measure for the troubles in K-melter operation.

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For the scenario of boil-off in the spent fuel pool (SFP) initiated with loss of pool cooling, when the pool water level drops below the top of the spent fuel (SF), the steam is heat up by the upper uncovered part of the SF and becomes super-heated. The cladding temperature mainly depends on the steam temperature before it rises up to the oxidation heating dominate bound. For an ideal adiabatic condition without radial heat transport, the steam temperature only depends on the decay heat proportion between the evaporation section (water covered part) and supper heating section (uncovered part), not on total decay heat or decay time. It can be easily deduced that for one SFP stored with different decay time SFs, if the axial decay heat profiles are same, the SFs' outlet steam temperatures are identical. Thus, four kinds of axial decay heat profiles with peak heat at upper, lower, middle parts of and uniform along the SFs are applied to investigate the effect on the steam and cladding temperatures in adiabatic condition with RELAP5/MOD3.3 code. The calculation results are showed and discussed. Obviously, SFs with upper peak decay heat suffered the worst deficient cooling and could burn at a much higher water level.

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This paper describes recent activity on development of the Micro-Physics Nuclear Reactor Simulator^<TM> and its application to introductory educations of nuclear engineering at high schools and university. The simulator has been continuously improved with active feedbacks from existing and potential users through its applications to exercises in classes/seminars. A newly developed reactor core transient analysis code, RAMBO-T has been adopted in the simulator along with SIMULATE-3K by Studsvik Scandpower Inc. (Borkowski, 1994) The internal data structure has been revised so that any combinations of the target reactor type, the core transient analysis code and the display language can be established. A new graphical user interface was implemented to realize the intuitive and easy-to-understand operations by novice users. The improved version of the Micro-Physics Nuclear Reactor Simulator has been practically used at educational institutions. In order to contribute to the activities on human resource development in the field of nuclear engineering, it is planned to donate the Micro-Physics Simulator^<TM> Lite, a variation of the simulator that supports the only transient core analysis with RAMBO-T, to IAEA, the International Atomic Energy Agency. It will be included into the "NPP Simulators suite for Education" where complimentary copies are distributed to the member states countries.

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Drift phenomena are caused by gas accumulating in pressure transmitters in nuclear power plants. We found factors contributing to drift were radiolysis of silicone oil as well as hydrogen permeation, as a result of analyzing drift factors. We propose a method of suppressing drift utilizing hydrogen absorptive material. The effects of suppressing drift were evaluated through radiation decomposition experiments. Our proposed pressure transmitter outperformed a conventional pressure transmitter without hydrogen absorption material in the experiments.

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In NPP, the digital control system which integrated software and hardware are increasingly used to improve dependability and introduce new functionality. Traditional safety analysis can get a good result when handling accidents caused by component failures, but software does not fail in this way. STPA is a new hazard analysis technique based on systems theory rather than reliability theory. It considers the system as a whole (include the hardware and software) to analyze failure and causality of systems and treats safety as a control problem rather than a failure problem. Being a safety-critical system, RPS in NPP needs to be considered carefully in system safety. So, we adopt this new approach to analyze the design process. From the analysis results, we found that causal factors leading to safety accidents identified by STPA included all the hazards identified by the fault tree analysis. Furthermore, there are some causal factors that were identified by STPA only. We utilize these results of the analysis on causation factor to refine the safety requirements and reduce the occurrences of the hazardous scenarios.

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According to lessons learned from the Fukushima Accident, it is crucial in nuclear emergency response to organize emergency staff effectively. Furthermore, every response organization is supposed to know its responsibility exactly and keep reliable communication links to cooperate properly. However, traditional decision-making systems focus more on protection actions rather than organization management. This article introduces a new decision support system based on emergency plans, which can digitalize the nuclear emergency plans from different emergency organizations as a "rule base" in an expert system, get real-time information from the accident as an "environment variables base", and integrate the traditional consequence assesment at the same time, in order to give reasonable suggestions to the emergency staff. We have already developed a system based on above ideas, which has been used in some off-site emergency centers in China.

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