Turbulent flow and heat transfer characteristics in a square duct of 100 mm height with 45 degree square ribs of 10 mm height was analyzed numerically by using algebraic Reynolds stress model including the fixed turbulent Prandtl number and algebraic turbulent heat flux models. In this research, analytical results were compared with the experimental and predicted data reported by Bonhoff et al. [1], which were measured and analyzed turbulent flow fields at Reynolds number 5×10^4 based on bulk velocity and duct height by means of a PIV system and a Reynolds stress model. Analytical results obtained by algebraic Reynolds stress model showed a good agreement with the experimental data of streamwise and spanwise mean velocity profiles at the same Reynolds number of the experiment. As for secondary flow in the duct, analytical results showed relatively good agreement with the experimental data. Moreover, temperature distributions were obtained by using the fixed turbulent Prandtl number and algebraic turbulent heat flux models. Temperature profiles of algebraic turbulent heat flux model were not similar to velocity profiles. Considering algebraic turbulent heat flux model has potentiality to reproduce anisotropic temperature, these results suggested that correlation between momentum and heat transfer was not recognized in this turbulent field. As a result of this study, it was verified that the presented method was able to predict turbulent flow in duct with ribs through the comparison of calculated results with the experimental data.

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Creviced SSRT tests were conducted on low carbon austenitic stainless steels and their weld metals in a simulated BWR environment. When the materials were non-sensitized, the most of SCC is recognized TGSCC. From the SEM observation at high magnification indicated the presence of micro-cracks whose length is less than 5μm in all cases. TEM observation of micro-cracks shows us the Cr-rich point at the non-propagating crack tip. Examination was made of the effects of the exposure time prior to loading, applied strain, cold work, DO level, chemical compositions of base materials and volume fraction of ferrite in weld metals obtained from creviced SSRT. At less than 20% of rolling reduction, no effects of cold work could be seen, but at 30% of rolling reduction, high SCC susceptibility was apparent. A comparison of SCC susceptibility for type 304NG and 316NG indicated fewer cracks in the type 304NG. Thus possibly, Cr contents or the sub-structure may be related to SCC initiation. In weld metals, SCC occurs in the austenite phase or at the interface between austenite and ferrite. Weld metal with medium amount of ferrite has been shown to exhibit high SCC resistance since the ferrite network fuctions to be an obstacle for crack propagation.

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In the operation of the fast-breeder reactor, the sodium leak detection system plays a role of detecting a small sodium leakage at its early stage. As one of the continuous monitoring detectors, this system uses the ionization detector that monitors a small number of sodium aerosols. A conventional ionization detector (Radiative Ionization Detector, RID) has the high sensitivity for the sodium aerosols, whereas its output signal fluctuates due to several factors other than the aerosols, especially due to a change of temperature in the operation environment. To improve the temperature dependence of the output signal and thereby make full use of its high sensitivity, we developed a new ionization detector (Moving fluid Ionization Detector, MID) based on an idea of the ion transportation with the moving fluid. Experimental verification indicated that a ratio of the signal fluctuation induced by temperature to the signal level for the required sodium density of 10^<-10> g/cm^3 was -0.1 to +0.2 under the operation environment condition (0 - 50℃). Achieving this ratio of within ±1.0 (the target value) enables the accurate judgment of the sodium leakage. The MID fully satisfied this target. In conclusion, we achieved the ionization detector that makes full use of its high sensitivity and thereby contributes to the further safety enhancement of the fast-breeder reactor.

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Over several decades the nuclear energy society worldwide has developed safety assessment methodology based on probabilistic risk analysis for incorporating its benefit into design and accident prevention for nuclear reactors. Although safeguards and security communities have different histories and technical aspects compared to safety, risk assessment as a supplement to their current requirements could be developed to promote synergism between Safety, Safeguards, and Security (3S) and to install effective countermeasures in the design of complex nuclear fuel cycle facilities. Since the 3S initiative was raised by G8 countries at Hokkaido Toyako-Summit in 2008, one approach to developing synergism in a 3S By Design (3SBD) process has been the application of risk-oriented assessment methodology. In the existing regulations of safeguards and security, a risk notion has already been considered for inherent threat and hazard recognition. To integrate existing metrics into a risk-oriented approach, several mathematical methods have already been surveyed, with attention to the scarcity of intentional acts in the case of safeguards and the sparseness of actual event data. A two-dimensional probability distribution composed of measurement error and incidence probabilities has been proposed to formalize inherent difficulties in the International Atomic Energy Agency (IAEA) safeguards criteria. In particular, the incidence probability that is difficult to estimate has been explained using a Markov model and game theory. In this work, a feasibility study of 3SBD is performed for an aqueous reprocessing process, and synergetic countermeasures are presented for preliminary demonstration of 3SBD. Although differences and conflicts between individual 'S' communities exist, the integrated approach would be valuable for optimization and balance between the 3S design features as well as for effective and efficient implementation under existing regulation frameworks. In addition, advanced process monitoring is one of the synergetic methods included in this study, and simulation and modeling are used to demonstrate enhanced loss detection probability for safeguards.

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In the design of pressurized water reactor (PWR), the control rods, the burnable poison rods and the boron acid solution are used to control the core power. Among these facilities, the boron acid solution is indispensable to maintain the critical condition for normal core operation during the compensation of negative reactivity for the fuel consumption and the accumulation of the fission product. The high head safety injection of Maanshan power plant once malfunctioned and therefore the reactor coolant system obtained unnecessary boron. In this study, the validated Computation Fluid Dynamics (CFD) commercial software FLUENT is employed to investigate the boron diffusion phenomenon during the boron acid solution flow into the core, in the high head safety injection malfunction accident. The calculation scope of the model in this paper is cold leg to top of core. The diffusion behaviors within the model can be reasonably captured by the present CFD commercial code. According to the calculation result, the non-uniform boron distribution at the core support forging and the core can be found.

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Maanshan nuclear power plant (NPP) is the third NPP operated by Taiwan Power Company (TPC) for more than 20 years. The anticipated-transient-without-scam mitigation system and actuation circuit (AMSAC) for Maanshan NPP was implemented based on microprocessor technology, and the obsolete issue will soon become a problem. With microprocessor design, undetectable software faults and common cause failures may exist. It is worthwhile to investigate an alternative digital design approach in advance. The field programmable gate array (FPGA) technology has been used extensively for commercial applications and has also gained regulatory acceptance recently for nuclear power plant applications. The present research is to explore the feasibility and conceptual design of using triple-redundant FPGA-based AMSAC system in Maanshan NPP. The AMSAC is a diverse and backup system for reactor protection system (RPS), and prevent primary pressure exceed 3200psig if an anticipated-transient-without-scam (ATWS) event occurs. The requirements of AMSAC system are to provide turbine trip, start auxiliary feedwater, and isolate stream generators. Due to flash-based technology, superior reliability, and 128-bit write-protection-security, Actel's FPGAs are widely used in military and space applications. Actel's SmartFusion mixed signal FPGA chip is chosen to be the target FPGA platform, which integrates flash-based FPGA, 32-bit ARM Cortex-M3 MCU, programmable analog and digital peripherals, to offer customization, IP-protection, and ease-of-use development tools. SamrtFusion FPGA is ideal for true system-on-chip (SoC) solutions. Programmable analog peripherals of SamrtFusion chip include 12-bit SAR ADCs, Sigma-Delta DACs, voltage/current/temperature monitors, high-speed comparators and up to 32 analog inputs/outputs. The proposed FPGA-based board contains SRAM, UART, 10/100 Ethernet. SRAM is partitioned into two divisions, all are fully Readable/Writeable for FPGA-Fabric, and only one division is read-only for UART to transmit status. UART is connected to NPP/Simulator for real-time monitoring or portable computer for maintenance purpose. The MCU integrated with 10/100 Ethernet are reserved for connecting to NPP's plant computers for non-safety-related processes in future expansions. The proposed FPGA-based AMSAC system contents three divisions, each division includes an independent FPGA board and DC power supply. A software Verification and Validation Plan (V&VP) is developed at the beginning of the study, and will be processed during the whole development lifetime. A Maanshan NPP engineering simulator with modeling of reactor and plant systems will be developed to provides an integrated-dynamic-interactive test environment to validate the triple-redundant FPGA-based AMSAC system. The software-free FPGA-based nuclear instrumentation and control systems (I&C) can easily be used for the modernization of the TPC's nuclear power plant analog systems and thus may reduce regulatory licensing efforts and cost.

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Commercial operation of the Tsuruga nuclear power station Unit 1, owned by the Japan Atomic Power Company (JAPC), will be terminated in 2016. For safe decommissioning of the station, technologies for processing stored radioactive wastes such as spent Ion EXchange resin (IEX) and Filter Sludge (FS) have been jointly developed by JAPC and JGC Corporation. The Wet-Oxidation process (WetOx) was applied for decomposition of the spent IEX and FS, and Super Cement (SC) solidification was chosen for immobilization of the WetOx residue. Pilot scale tests for the WetOx treatment have been successfully conducted with simulated wastes. The WetOx residue is a concentrate with Suspended Solid (SS) and sodium sulfate, and was solidified by SC in lab scale cementation apparatus. The compressive strength of the solidified waste was confirmed to meet the desired level for radioactive waste disposal.

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In Japan, the underground cavern-type disposal facilities for low-level waste (LLW) with relatively high radioactivity, mainly generated from power reactor decommissioning, and for certain transuranic (TRU) waste, mainly from spent fuel reprocessing, are designed to be constructed in a cavern 50-100 m underground and to employ an engineered barrier system (EBS) made of bentonite and cement materials. To advance a disposal feasibility study, the Japanese government commissioned the Demonstration Test of Underground Cavern-Type Disposal Facilities in fiscal year (FY) 2005. Construction of a full-scale mock-up test facility in an actual subsurface environment started in FY 2007. The main test objective is to establish the construction methodology and procedures that ensure the required quality of the EBS on-site. A portion of the facility was constructed by 2010, and the test has demonstrated both the practicability of the construction and the achievement of quality standards: low permeability of less than 5×10^<-13> m/s and low-diffusion of less than 1×10^<-12> m^2/s at the completion of construction. This paper covers the test results from the construction of certain parts using bentonite and cement materials.

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In view of the practical importance of the drift-flux model for two-phase flow analyses in microgravity conditions, the distribution parameter and the drift velocity were estimated by experimentally for vertical upward bubbly flows in a 9 mm-diameter pipe in normal- and micro-gravity conditions by using a stereo imageprocessing method. The effects of gravity and liquid Reynolds number on the distribution parameter and the drift velocity were discussed in detail. It was confirmed that the drift velocity in microgravity conditions didn't become zero and the local slip effect due to the wall friction couldn't be ignored. Comparison of the previously proposed constitutive equations of the distribution parameter and the drift velocity with the experimental data showed a good agreement.

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A critical issue for building confidence in the long-term safety of geological disposal is to demonstrate the stability of the geosphere, taking into account its likely future evolution. This stability is broadly defined as the persistence of Thermal-Hydrological-Mechanical-Chemical conditions considered favourable for the long-term safety of a geological repository. This study provides the conceptualisation and preliminary results of numerical simulations of the long-term geosphere evolution, especially groundwater flow properties, in the Horonobe area, Hokkaido, northern Japan, based on data from the JAEA's underground research laboratory project. Information on natural events and processes has been integrated into a chronological conceptual model, which indicates space-time sequences of the events and processes in the Horonobe area over geological time. Spatial scale for the numerical simulations are based on the recharge and discharge area in the present and the past glacial periods, and the direction of the groundwater flow in the present of Horonobe area. Time scale for the numerical simulations over the last 1.5 million years are defined by the palaeogeography of Horonobe area, and mainly focused on the changes of the geological structure (i.e. hydrogeological structure), recharge rate, and distribution of sea and land area caused by sub-surface and earth-surface processes such as crustal movement, climatic and sea-level changes. The results of the simulation show that the range and magnitude of change in groundwater flow properties, and sensitivity of the properties to the sea-level and elevation changes due to crustal movement have close relationship with the distribution and the evolution of the hydraulic conductivities of geological formations.

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A gas entrainment rate into liquid by a vortex formed on the free surface was examined experimentally. Water flowed into a cylindrical vessel from a wall tangentially. Swirl flow was formed in the vessel, and then water left from the bottom outlet of the vessel. The flow state of the entrainment was visually observed by using a high speed video camera. The gas entrainment rate into water was measured. A stable vortex was formed in the test vessel. When the flow velocity; the velocity at the bottom outlet, was low, a single bubble was periodically torn off from the bottom tip of the vortex and the bubble-type gas entrainment was observed. As the flow rate was increased, the bottom tip of the vortex penetrated into the outlet pipe and the bubble-type gas entrainment continued. A further increase in the flow velocity resulted in the transition from the bubble-type gas entrainment to the vortex-type gas entrainment and the gas entrainment rate considerably increased with the flow velocity. After the vortex tip penetrated into the outlet pipe, the rotation speed of the vortex decayed. As a result of it, the Kelvin-Helmholtz instability wave length got long and the size of the generated bubble became large. Then, the outlet pipe was filled with the large bubbles and the flow state in the outlet pipe turned to the slug/churn flow and a large amount of gas began to be carried out form the test vessel.

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This paper presents a general approach and techniques for design and verification of Field Programmable Gates Arrays (FPGA)-based Instrumentation and Control (I&C) systems for Nuclear Power Plants (NPP). Appropriate regulatory documents used for I&C systems design, development, verification and validation (V&V) are discussed considering the latest international standards and guidelines. Typical development and V&V processes of FPGA electronic design for FPGA-based NPP I&C systems are presented. Some safety-related features of implementation process are discussed. Corresponding development artifacts, related to design and implementation activities are outlined. An approach to test-based verification of FPGA electronic design algorithms, used in FPGA-based reactor trip systems is proposed. The results of application of test-based techniques for assessment of FPGA electronic design algorithms for reactor trip system (RTS) produced by Research and Production Corporation (RPC) "Radiy" are presented. Some principles of invariant-oriented verification for FPGA-based safety-critical systems are outlined.

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In the early stages of dynamic simulations, significant effort was devoted to the development of efficient numerical methods, and Fluid Structure Interaction domain made no exception to that rule. As a matter of fact, due to limited computational capacities, great care was taken to only include significant parts of simulated structures. One consequence was that the non-structural parts - thereby including fluid domain- were systematically condensed into structurally equivalent mass, stiffness or damping. Among those methods, a series of seminal papers (Au-Yang, 1976 and Au-Yang 1981) were devoted to the coupling of coaxial shells. While the proposed method was perfectly suited to line (beam-type) models, it does not apply to the nowadays widely used surface type (plate or shell) models. Necessary developments have been made by the authors, and the results favorably benchmarked with those obtained using other rigorous but computationally intensive approaches. Therefore, the proposed method is felt to be a very valuable alternative for industrial practice. The approach does not require specific development, since it only uses generic capabilities of structural Finite Element packages. It can also be used as a reference for verification of complex, multi-physics numerical packages. Although primarily aimed at simulating the internal structure of Pressurized Water Reactor, the proposed method can be applied to any immersed fluid coupled coaxial shells.

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The dynamic aspects of loading conditions for reactor internals, piping and the other components are thought to play important roles in the initiation of failures due, for example, to stress corrosion cracking (SCC) and environmental fatigue. In the design of Light Water Reactors (LWR), however, the strain rate of components is not evaluated, because time-dependent or rate-dependent failure modes are not considered quantitatively in LWR design. The Finite Element Method (FEM) can be used, of course, to evaluate strain rate, but it is more important to know the major factors determining the strain rate magnitude. To investigate such kinds of sensitivity of transient conditions to the magnitude of strain rate, it is helpful to develop a simplified evaluation method of strain rate. For this purpose, a simple strain rate evaluation method based on Green's function was developed for a specific point of a component with a given design condition. Furthermore, a generalized simplified method was developed based on shell theory to evaluate thermal stress and strain rate.

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In this paper, it was reported that simulation design study about gas process unit, such as process gas reactor, heat exchangers and heat conditioner, for one of the heat utilization system components development of High Temperature Gas-cooled Reactor (HTGR). In simulating precisely the heat transfer control characteristics of the gas-mixing unit, a good model for turbulent mixing process of some velocity ratios is a key issue. Therefore, we have been proceeding to develop the newly CFD code. This code using algebraic Reynolds stress model has been reasonably predicted about complex turbulent flows such as flow separation in a rectangular duct with 90-degree sharp turn and strong anisotropic flow in straight duct with roughened walls. However, it has not enough been validated for gas-mixing turbulent flow. We have a lack of detailed knowledge and experimental data on the turbulent gas process in the unit. Hence, as the first step to make the air mixing process in gas unit clear, we have made an preliminary experimental study on turbulent mixing using the equipment with flow control door arranged some velocity ratios and a heating core. Flow visualization was carried out by using particle image velocimetry (PIV). As a result of this study, it was obtained representative velocity distribution data and turbulent kinetic energy distributions about separation and gas mixing section in the equipment, which is important for unit performance. Furthermore, it was found that production of turbulence in the unit were strongly formed by unsteadily long separation flow between velocity ratios and small scale vortices induced to the tip of door.

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The instability occurring in OTSG (Once-Through Steam Generator) of movable nuclear power plants is presented by the multivariable frequency domain theory. As concerning coupling interactions of OTSG tubing, it is more efficient for analyzing the instability of OTSG compared the common single variable method. A mathematical model for the system is derived from the fundamental equations by using the perturbation, Laplace-transform and the nodalization techniques. The stable boundary and parameters which influence the stability of the system are evaluated through computer simulation. Numerical examples are given in the paper and the predictions of the model agree with the experimental results.

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Parallel channel system is very typical in nuclear power plant, such as in the core and the steam generator. In this paper the author developed one model and a set of codes for the thermal hydraulic transient analysis of a core parallel channel system in QinShan II nuclear power plant. The basic model was built by Clausse and Lahey (1990) and developed by Lee and Pan (1999). In the original model the channel is divided into three parts: entrance section, heater section and riser section. The integral method is used to calculate the pressure of each section. It can reflect the whole characteristics of such system except the local parameters. And the nonuniform heating is very hard to be considered in this model. And the code based on this model is very hard to be coupled with reactor physics code in the future. Hence, the control volume concept is introduced to overcome these weaknesses. The channel can be divided into sufficient control volumes to describe the details. A development model is built based on this concept. Many classical flow and heat transfer correlations and models are included. The Gear method is chosen as the solving implement of the differential equations. Then a code is compiled. And then the transient characteristics are obtained under single and two-phase conditions. The transients of blocking, reactivity insertion and loss of flow are investigated based on imaginary data. At last a coupled method is conceived.

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In recent years, as the increasing demand of energy all over the world, and the pressure on greenhouse emissions, there's a new opportunity for the development of nuclear energy. Modular High Temperature Gas-cooled Reactor (MHTR) received recognition for its inherent safety feature and high outlet temperature. Whether the Modular High Temperature Gas-cooled Reactor would be accepted extensively, its economy is a key point. In this paper, the methods of qualitative analysis and the method of quantitative analysis, the economic models designed by Economic Modeling Working Group (EMWG) of the Generation IV International Forum (GIF), as well as the HTR-PM's main technical features, are used to analyze the economy of the MHTR. A prediction is made on the basis of summarizing High Temperature Gas-cooled Reactor module characteristics, construction cost, total capital cost, fuel cost and operation & maintenance (O&M) cost and so on. In the following part, comparative analysis is taken measures to the economy and cost ratio of different designs, to explore the impacts of modularization and standardization on the construction of multiple-module reactor nuclear power plant. Meanwhile, the analysis is also adopted in the research of key factors such as the learning effect and yield to find out their impacts on the large scale development of MHTR. Furthermore, some reference would be provided to its wide application based on these analysis.

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In this work, the flow measurement of a Two-Phase Natural Circulation Loop (2PNCL) was performed using a two-transducer pulse-echo technique. The ultrasonic system consists of a two-phase flow loop, a multiplexer, 3 panametric 10 MHz (6.4 mm) ultrasonic transducers, and a computer equipped with a Data Compuscope Digitizer, a motion controller, and Winspect Data Acquisition Software. Two transducers are attached on the two-phase flow test section separated by a distance of 3.9 cm and one transducer is mounted beneath the separation tank to measure the liquid level. The circular pipe test section is made of stainless steel with a length of 10.3 cm and an outer diameter of 13 mm with a thickness of 0.5 mm. Each transducer receives a pulse from the multiplexer and returns an echo which is received by the multiplexer. The Winspect software outputs an A-Scan which is an instantaneous display of the pulses and echoes. A C-Scan is obtained from continuous A-Scans at a scanning interval of 25 μs. Analysis of the pulse echo reflection is used to determine the water level in the separation tank. Flow measurements for two-phase flow regimes which include bubble properties, void fraction, and gas-liquid interface profile are obtained by analysis of the C-Scans. The results show flow regime transitions from bubble flow through to slug flow to an annular flow, with an increase in the inlet air flow rate.

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The phenomenon of fatigue failure due to In-line flow-induced vibration has been investigated. In the experimental program an induced-vibration experimental set-up involving water flow was developed. This apparatus was used to produce fatigue failure resulting from In-line flow-induced vibration. The fatigue test cylinders were made of medium carbon steel. A small hole was drilled onto the test cylinder surface to localize the fatigue cracking process. A strain histogram recorder was used to acquire the service strain histogram and also to detect any variation in natural frequency. The cumulative fatigue damage, D, as defined by the Modified Miner Rule, was determined by using the strain histogram of the early portion of the test record. The value of D was close to unity in the case of In-line vibration. In contrast, the value of D obtained in a previous investigation for the case of the Cross-flow vibration ranged approximately from 0.2 to 0.8.

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For the commercialization of the sodium cooled FBR system, it is desirable to develop a plant concept that has economic competitiveness compatible with the LWR system and excels in safety. Since it is difficult to perform inspection and maintenance of the core internals at FBR, such as the reactor vessel inner wall and the core support structure, measures have been taken to ensure safety and reliability by allowing a sufficiently large margin in the design reducing the burden of inspection and maintenance. To date due to problems such as the effect of radiation penetrating from the reactor core, the opacity and high chemical activity of sodium, and the fact that the inspection of reactor vessel of the FBR is performed at a high temperature, approximately 200 degree C, to avoid freezing of sodium, no method of inspecting structures inside the reactor vessel has been developed. In this study, aims to experimentally produce a real-time sensor which is used for the underwater and under-sodium test to demonstrate their applicability to the FBR system. The real-time sensor would be a piezoelectric element type sensor with a resolution of approximately 2.0mm, a resolution of sufficient quality to allow for the identification of deformation, the failure or dropping of components, and an image processing time per one image of approximately 0.5 second. Concerning the backing material, aluminum titanate (alumatite), wollastonite and hexagonal boron nitride were selected for the first screening. Based on the test result, alumatite that has an excellent damping property was finally selected for the trial production of the sensor element. We have designed and made 256 channel matrix-arrayed sensor and high speed processing system. It was shown that the real-time sensor was achieved about 2.0mm resolution and about 0.5 seconds/frame signal processing time, and succeeded taking the first under-sodium movie of the world.

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Japanese demonstration and commercialized FBR assemblies are planed to adopt the inner duct for severe accident through which melted fuel goes out of core. It is seemed that the traditional nuclear calculation scheme which generates nuclear constants with homogeneous assembly calculation cannot be adopted. From this view point, new lattice physics code for hexagonal geometry GALAXY-H has been developed based on GALAXY, which is a two dimensional transport calculation code for LWR assemblies. The methodology of flux calculation for GALAXY-H is based on the method of characteristics (MOC). For both resonance and neutron transport calculations, GALAXY can exactly treat heterogeneity of calculated configuration. GALAXY-H is a hexagonal version of GALAXY, and can exactly treat the inner duct and wrapper tube in assemblies. In this study, by using GALAXY-H, heterogeneous effect of Japanese demonstration FBR assembly on effective cross sections for inner duct is evaluated. For the first time, position dependence background cross section is calculated by MOC in assembly configuration which contains both of the inner duct and the wrapper tube. It is confirmed that the effect of this difference is not small.

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Fluid-Structure Interactions (FSIs) occurring inevitably in operating reactor component systems can cause excessive force or stress to the structures resulting in mechanical damages that may eventually threaten the structural integrity of components. In particular, thermal stratification in the pressurizer surge line has been addressed as one of the significant FSI phenomena in reactor systems. In this study, a thermal-stress simulation is performed using ANSYS FSI. The 3-dimensional transient temperature distributions in the wall of an actual PWR pressurizer surge line subjected to thermal stratification is calculated by CFD analysis either during out-surge or in-surge operation. The thermal loads from CFD analysis using ANSYS CFX are transferred to structural analysis code, ANSYS Multiphysics. From this information, thermal stresses are determined and ultimately a fatigue analysis is performed, all within the ANSYS environment. In addition, the thermal stress and fatigue analysis results obtained by applying the realistic temperature distributions from CFD analysis are compared with those by assuming the simplified temperature distributions to identify some requirements for a realistic and conservative thermal stress analysis from a safety point of view.

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The transient turbulent heat transfer coefficients in a short vertical Platinum test tube were systematically measured for the flow velocities (u=4.0 to 13.6 m/s), the inlet liquid temperatures (T_<in>=296.93 to 304.81 K), the inlet pressures (Pin=794.39 to 858.27 kPa) and the increasing heat inputs (Q_0 exp(t/τ), exponential periods, τ, of 18.6 ms to 25.7 s) by an experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The Platinum test tubes of test tube inner diameters (d=3 and 6 mm), heated lengths (L=66.5 and 69.6 mm), effective lengths (L_<eff>=56.7 and 59.2 mm), ratios of heated length to inner diameter (L/d=22.16 and 11.6), ratios of effective length to inner diameter (L_<eff>/d=18.9 and 9.87) and wall thickness (δ=0.5 and 0.4 mm) with average surface roughness (Ra=0.40 and 0.45 μm) were used in this work. The surface heat fluxes between the two potential taps were given the difference between the heat generation rate per unit surface area and the rate of change of energy storage in the test tube obtained from the faired average temperature versus time curve. The heater inner surface temperature between the two potential taps was also obtained by solving the unsteady heat conduction equation in the test tube under the conditions of measured average temperature and heat generation rate per unit surface area of the test tube. The transient turbulent heat transfer data for Platinum test tubes were compared with the values calculated by authors' correlation for the steady state turbulent heat transfer. The influence of inner diameter (d), ratio of effective length to inner diameter (L_<eff>/d), flow velocity (u) and exponential period (τ) on the transient turbulent heat transfer is investigated into details and the widely and precisely predictable correlation of the transient turbulent heat transfer for heating of water in a short vertical tube is given based on the experimental data and authors' studies for the transient critical heat fluxes (CHFs) of subcooled water flow boiling in a short vertical tube caused by exponentially, ramp-wise and stepwise increasing heat inputs. The correlation can describe the transient turbulent heat transfer coefficients obtained in this work for the wide range of the temperature difference between heater inner surface temperature and average bulk liquid temperature (ΔT_L=10 to 160 K) with the inner diameters (d=3 to 6 mm), L_<eff>/d (=9.87 to 18.9), the flow velocities (u=4.0 to 13.6 m/s) and the exponential periods (τ=18.6 ms to 25.7 s) within ±15 % difference.

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Design the nuclear reactor power control system in this paper to cater to a nonlinear nuclear reactor. First, calculate linear power models at five power levels of the reactor as five local models and design controllers of the local models as local controllers. Every local controller consists of an optimal controller contrived by the toolbox of Optimal Controller Designer (OCD) and a proportion-integration-differentiation (PID) controller devised via Genetic Algorithm (GA) to set parameters of the PID controller. According to the local models and controllers, apply the principle of flexibility model developed in the paper to obtain the flexibility model and the flexibility controller at every power level. Second, the flexibility model and the flexibility controller at a level structure the power control system of this level. The set of the whole power control systems corresponding to global power levels is to approximately carry out the power control of the reactor. Finally, the nuclear reactor power control system is simulated. The simulation result shows that the idea of flexibility model is feasible and the nuclear reactor power control system is effective.

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The feasibility study of a liquid lithium type proton beam target was performed for the neutron source of the boron neutron capture therapy (BNCT). As the candidates of the liquid lithium target, a thin sheet jet and a thin film flow on a concave wall were chosen, and a lithium flow experiment was conducted to investigate the hydrodynamic stability of the targets. The surfaces of the jets and film flows with a thickness of 0.5mm and a width of 50 mm were observed by means of photography. It has been found that a stable sheet jet and a stable film flow on a concave wall can be formed up to certain velocities by using a straight nozzle and a curved nozzle with the concave wall, respectively.

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The transient critical heat flux experiments with forced sinusoidal inlet flow oscillation have been conducted in a vertical tube. The experimental range was as follows: the system pressure P=0.5〜3.0MPa, the averaged inlet mass flux G_<av>=90〜200kg/m^2s, the inlet subcooling T_<sub>=55℃, normalized amplitude of flow oscillation ΔG_<ma>x/G_<av>=0〜3.0, and oscillation period =1.04, 2.1, 5.2sec. The analysis on parametric trends of normalized oscillation amplitude, oscillation period, pressure and averaged inlet mass flux on normalized periodic dryout critical heat flux, F_p show that: the F_p declines remarkably and tends to a reduction limit, F_<p,lim>, with an increase of ΔG_<max>/G_<av>; the F_<p,lim> decrease with an increase of oscillation period τ and averaged inlet mass flux, G_<av>, respectively; the pressure has no obviously effects on the decrease of F_p in 0.5〜3.0MPa. A normalized oscillation period, τ/t_<tr>, was introduced to built a correlation of the reduction limit of periodic dryout, F_<lim>. Based on the F_<lim> correlation and the analysis of parametric trend, a semi-empirical correlation was proposed to predict the F_p under flow oscillation condition in vertical tubes. The comparisons between the prediction of the new correlation and experimental data show a reasonable agreement.

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Tokai-1 (GCR, Gas Cooled Reactor) nuclear power plant of JAPC (the Japan Atomic Power Company) started commercial operation in 1966 as the first commercial nuclear power plant in Japan The unit had helped introduction and establishment of the construction and operation technologies regarding nuclear power plant at early stage in Japan by its construction and operating experiences. However, The Japan Atomic Power Company (JAPC), the operator and owner of Tokai-1, decided to cease its operation permanently because of a fulfillment of its mission and economical reason. The unit was finally shut down in March 1998 after about 32 year operation. It took about three years for removal of all spent fuels from the site, and then decommissioning started in 2001. JAPC, always on the forefront of the nation's nuclear power generation, is now grappling Japan's first decommissioning of a commercial nuclear power plant, striving to establish effective, advanced decommissioning. The decommissioning for Tokai-1 was scheduled as 20 years project. At the beginning, the reactor was started to be in a static condition ("safe storage period"). While the reactor had been safely stored, the phased decommissioning works started from non-radioactive or low radioactive equipment toward high radioactive equipment. First five years of the project, JAPC concentrated to drain and clean spent fuel cartridge cooling pond and to remove conventional equipments such as turbine, feed water pump and fuel charge machine as planed and budgeted. From 2006, the project came into a new phase. JAPC has been trying to remove four Steam Raising Units (SRUs). The SRUs are huge component (750ton, φ6.3m, H24.7m) of the Gas Cooling Reactor (GCR) and inside of the SRUs are radioactively contaminated. Major concerns are workers safety and minimizing contamination areas during SRU removal. Therefore, JAPC is developing and introducing Jack-down method and remote control multi-functional dismantling system. This method is to cut and remove the SRUs in turn from the bottom to top remotely while lifting the SRU by a large jack system. The system enables cutting and holding not only the SRU body but also internals. This technology and experiences would be useful for the reactor removal in the near future. Skirt part of No.2 SRU cutting work was done carefully by well trained JAPC staff from August 2010 to December, 2010,

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Cracks were found at shroud supports of Tokai Daini Power Station (Tokai-2) in the 24^<th> refueling outage. Flaw evaluation of these cracks was performed applying Codes for Nuclear Power Generation Facilities - Rules on Fitness-for-Service - (FFS Code) of Japan Society of Mechanical Engineers (JSME) and Guideline for Inspection and Evaluation of Reactor Internals of Japan Nuclear Technology Institute (JANTI). Evaluation methods and results are outlined in this report. Issues to be studied for better improvement of JSME FFS Code and JANTI Guideline will be discussed.

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HTGR as a high temperature (950oC) nuclear heat source can efficiently generate variable energy products including electricity, hydrogen, and process heat. This paper describes the plant design of a modular (600MWt) HTGR system cogenerating electricity and process heat and discusses its operational feasibility of electric load follow using a new control scheme. The system generates electricity by direct cycle gas turbine and process heat through a topping high temperature intermediate heat exchanger. The process heat is transported via a secondary heat transport loop to such an industrial plant as hydrogen or steam production plant. The electricity generated supplies external grid output while meeting in-house power consumption in the reactor and the production plant. The load follow operations are performed by controlling the reactor coolant inventory while keeping the primary system thermal conditions including reactor power, reactor temperature, and turbine temperatures unchanged. This control strategy is designed to achieve high thermal efficiency in a wide range of part electric load and, by minimizing thermal transient of the reactor, to enable response to rapid load follow demand. The newly proposed control strategy is evaluated by simulation of a bounding load-follow event. The results of the simulation are discussed.

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The HDR specimen was a large scale representation of a typical pressurized water reactor. The specimen had a core support barrel with a mass ring attached on its bottom to simulate the core weight. The core barrel and surrounding subcooled water were instrumented to measure displacement, strain, pressure difference, absolute pressure and temperature at numerous locations during pipe break test. In this proceeding a procedure to calculate the responses of the core barrel with acoustic and (structural) shell and solid finite elements is presented. The analysis was carried out using acoustic-structural coupling feature available in Abaqus/Standard software. The natural boundary condition needed for the pipe break was derived from the mass flow results of the system analysis code APROS. The theoretical basis of the boundary condition derivation and its use in acoustic analysis is explained. The displacement and strain results obtained for the core barrel showed good agreement with the measurements up to 0.15 seconds. It is concluded that the presented procedure yields same order of accuracy as the FSI analysis with computational dynamics codes presented in the literature.

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This proceeding presents the finite element (FEM) modeling of a soft missile impact benchmark case called Meppen II/4. The benchmark was launched by OECD/NEA IAGE in 2010. The target in the test case II/4 was 6.5 m x 6.0 m x 0.7 m reinforced concrete slab. The missile was ca. 6 m long steel pipe with the mass of 1016 kg. The impact speed was 248 m/s. The FEM analysis was carried out using Abaqus/Explicit-6.10 software. A so called microplane material model, to model concrete, was adopted from the literature and it was implemented as a user subroutine to the Abaqus/Explicit. The missile load was modeled in this study by using well-known Riera method. The concrete slab reaction force and displacement results achieved in this study were satisfactory. The FEM model predicted cone formation and the radial cracking of the concrete slab.

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The reaction-time response of a large commercial aircraft is defined. The aircraft is examined as a thin-walled tubular missile. The impact is assumed soft, and the target's effect on the reaction-time response is neglected. The reaction-time response is defined assuming a normal impact on a rigid wall. The reaction-time response is defined with the analytical Riera method and with the numerical explicit finite element method. The Riera force history is solved with the finite difference method. For the finite element method, two codes are used: Abaqus/Explicit and LS-DYNA. Focus is on the sensitivity study of the used methods. The outer shell of the aircraft is modeled, and an approximation for the mass-distribution is made. Sensitivity to modeling assumptions is studied in order to get information on the adequacy of modeling. The results indicate relatively small sensitivity to modeling assumptions. The wings should be modeled more accurately in order to obtain the dominant frequency response in global structural analysis.

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To investigate the transient characteristic of the passive emergency feedwater system (PEFS), a small test facility is designed to do scaling simulation and support the code development. Scaling analysis was performed to find the criteria for the simulation study and the test facility design. The results show, natural circulation, single and two-phase heat transfer, and single and two-phase flow have important effect on the characteristic of the PEFS. Based on the scaling analysis, a small scale test facility was designed to carry out the static and transient investigation.

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A new resonance calculation method, SDGM (Spatially Dependent Gray-Resonance-Shielding Method), is developed for calculating spatially dependent effective cross sections within fuel pellet. The present method has been implemented to GALAXY, which is the MHI's new lattice physics code. SDGM accurately generates burnup dependent radial power profiles within pellet. The formulation of SDGM is carried out by extending the resonance self-shielding method of GALAXY with the combination of Stoker-Weiss method and SDDM. If pellet is not sub-divided, SDGM is completely consistent with the method used in the basic design calculations of GALAXY. For verifications and validations of SDGM, Monte Carlo benchmark and post irradiation examination (PIE) analysis are carried out. In Monte Carlo benchmark, reaction rate distributions within pellet are compared between GALAXY with SDGM and continuous energy Monte Carlo code MVP. On the other hand, for the PIE analysis, burnup and nuclide composition distributions within irradiated UO2 pellet are compared between GALAXY with SDGM and measurements. From these calculations, the applicability of SDGM is totally confirmed. SDGM is so efficient and practical method for generating power profiles within pellet (used in fuel integrity evaluation), accurate neutronics analyses for fuel with spatially high gradient of neutron flux and nuclide compositions, etc.

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Models of a liquid-metal target of an acceleratorcontrolled system have been experimentally studied at the Nizhny Novgorod State Technical University to develop an optimal design of the flow part of the target. The main explored variants of liquid-metal targets are: - Design with a diaphragm (firm-and-impervious plug) mounted on the pipe tap of particle transport from the accelerator cavity to the working cavity of the liquid-metal target. - Design without a diaphragm on the pipe tab of particle transport from the accelerator. The study was carried out in a high-temperature liquidmetal test bench under the conditions close to full-scale ones: the temperature of the eutectic lead-bismuth alloy was 260℃- 400℃, the coolant mass flow was 5-80 t/h, and the rarefaction in the gas cavity was 10^5 Pa, the coefficient of geometric similarity equal to 1. The experimental studies of hydrodynamic characteristics of flow parts in the designs of targets under full-scale conditions indicated high efficiency of a target in triggering, operating, and deactivating modes. Research and technology instructions for designs of the flow part of the liquid-metal target, the target design as a whole, and the target circuit of accelerator-controlled systems were formulated as a result of the studies.

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In this paper, fatigue verification of Class 1 nuclear power piping according to ASME Boiler and Pressure Vessel Code, Section III, NB-3600, and relevant issues that are often discussed in connection to the power uprate of several Swedish BWR reactors in recent years, are dealt with. Key parameters involved in the fatigue verification, i.e. the alternating stress intensity S_<alt>, the penalty factor K_e and the cumulative damage factor U, and relevant computational procedures applicable for the assessment of low-cycle fatigue failure using straincontrolled data, are particularly addressed. A so-calle simplified elastic-plastic discontinuity analysis for alternative verification when basic fatigue requirements found unsatisfactory, and the procedures provided in NB-3600 for evaluating the alternating stress intensity S_<alt>, are reviewed in detail. Our emphasis is placed on other procedures alternative to the simplified elastic-plastic discontinuity analysis. A more in-depth discussion is given to an alternative suggested earlier by the authors using nonlinear finite element analyses. This paper is a continuation of our work presented in ICONE16/17/18, which attempted to categorize design rules in the code into linear design rules and non-linear design rules and to clarify corresponding design requirements and finite element analyses, in particular, those non-linear ones.

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On the basis of successful validation of the MELCOR code on several experiments from the international OECD/NEA programme THAI, a numerical parametric study has been conducted at NRI Rez evaluating the influence of PAR's vertical position within a vessel on its performance. Simulations were carried out for an Areva PAR unit which was considered to be placed at 5 different elevations in the THAI test vessel, the sixth simulation comprised a model of blower, simulating thus forced convection. The initial conditions were those of the THAI HR-12 test, which was characterized by steam-saturated atmosphere at elevated pressure and temperature. The results show that the overall hydrogen mass recombined monotonically decreases with PAR elevation. This behavior is due to hot, light and hydrogen-lean plume, coming out from the PAR outlet, which, due to buoyancy forces, eventually fills the upper part of the vessel and prevents thus the PAR unit from efficient operation (if the latter is placed near the top). It was also demonstrated that the effect of forced convection is favorable since it breaks the gas stratification and increases thus hydrogen concentration at the PAR inlet.

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The water cooled pebble bed reactor with the pebble diameter from 2-10 mm has both the characteristics of high temperature gas-cooled pebble bed reactor and the traditional light water one. In the conceptual design of high temperature gas-cooled reactor and supercritical water-cooled reactor, pebble bed reactor has attracted much interest because of its outstanding advantages in the safety and fuel cycle. As far as china is concerned, it is worth researching and designing the water cooled pebble bed reactor, and the formers' researches were focused on the water cooled reactor. So it is great important to investigate the thermal-hydraulic characteristics of coolant in a pebble bed reactor core. The 3-D physical and geometrical models of the water cooled pebble bed reactor were built using the commercial software Pro/E and CFX, in which the thermal-hydraulic characteristics of the reactor core were studied. The velocity and temperature fields and pressure distribution of the coolant as well as the distribution of internal temperature of the fuel are obtained and analyzed in the case of diameter of 3mm and diameter of 6mm, and its effects on the core safety are also analyzed. The results indicate that the tendency of the same parameter values under different diameters are nearly identical, and it is larger with the diameter of 6mm, contrary, the flow disturbance with the diameter of 6mm is weaker than that of 3mm. The two maximum temperature of the pebble were found at the gap area. The maximum velocities of the first layer pebbles were found at the gap area, thereafter the maximum velocity was found between the gap areas. The pressure gradually reduces from bottom to top and the minimum values of pressure were generated at the gap area.

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The challenges related to problem of assessment of actual diversity level and evaluation of diversity-oriented NPP I&C systems safety are analyzed. There are risks of inaccurate assessment and problems of insufficient decreasing probability of CCFs. CCF probability of safetycritical systems may be essentially decreased due to application of several different types of diversity (multidiversity). Different diversity types of FPGA-based NPP I&C systems, general approach and stages of diversity and safety assessment as a whole are described. Objectives of the report are: (a) analysis of the challenges caused by use of diversity approach in NPP I&C systems in context of FPGA and other modern technologies application; (b) development of multi-version NPP I&C systems assessment technique and tool based on check-list and metric-oriented approach; (c) case-study of the technique: assessment of multi-version FPGA-based NPP I&C developed by use of Radiy^<TM> Platform.

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The Japan Atomic Energy Agency has been conducting study on thermo-chemical Iodine-Sulfur(IS) process for water splitting hydrogen production. Based on the test results and know-how obtained through the bench-scale test, a hydrogen test equipment made of industrial materials is being designed conceptually as the next step of the IS process development. In design of the IS test equipment, it is important to design ceramics chemical reactors with high performance from the viewpoint of thermal hydraulics and structural integrity. A new simple hydraulic analytical code has been developed for considering sulfuric acid solution as a mixture of two components especially in the H_2SO_4 decomposer. Complex flow with liquid-vaporized interaction involving chemical reaction will be characterized in the H_2SO_4 decomposer. Based on the preliminary analytical results obtained with above mentioned code, conceptual investigation of structural integrity of ceramics decomposer was carried out by a general commercial FEM code.

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Noble Metal Chemical Addition (NMCA) [1-2] is planned to be applied in Tokai Daini Power Station (Tokai-2, BWR, 1,100MWe, commenced commercial operation on November in 1973) after the 25th outage for environmental mitigation for Stress Corrosion Cracking (SCC) of BWR core internals, PLR pipes and so on. In addition, it is also planned to measure Electrochemical Corrosion Potential (ECP) at some locations such as core internals in order to check the mitigation effect of NMCA. This paper describes the outline of this proactive project, as well as explaining challenges of Japanese Code on Fitnessfor- Service (FFS Code) [3] to reflect ECP evaluation results on SCC assessment. Orientation for solutions of these challenges of FFS Code is also discussed.

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In the reflux condensation, steam generated in the reactor core and water condensed in the pressurizer form a countercurrent flow in a surge line. The flow is highly complicated because the surge line consists of a vertical pipe, a vertical elbow and an inclined pipe including plural elbows. Detailed measurements of countercurrent flow limitation (CCFL), however, have not been reported. Therefore, in this study, CCFL characteristics in a scale-down model of the pressurizer surge line were measured using air and water as working fluids. As a result, the following conclusions were obtained. (1) Supplied water was limited at either the upper junction, the inclined pipe or the lower junction in the surge line, depending on the inclination angle and air flow rates. (2) The smaller the inclination angle of the surge line is, the stronger the limitation of falling water flow rates in the inclined pipe is.

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In the phase II of HTR-10, it will focus on experimental studies of helium turbine cycle. Secondary helium pipe with temperature 750℃ and internal pressure 1.2MPa should be designed in the second design. This paper will introduce the design of high-temperature secondary helium pipe. First, the design principle for this kind pipe is presented. The performance of resistance to elevated temperature and pressure are taken into consideration separately. Then, the key parts of the high-temperature pipe, which includes pipe, bellows compensator, penetration assembly, are introduced in detail. After heat transfer and structural analysis, it is shown that the high temperature pipe could satisfy the requirements of the phase II of HTR-10.

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The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA, with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950℃ (Saito, 1994). Test researches are being conducted using the HTTR to improve HTGR technologies and to collaborate with domestic industries to contribute to foreign projects for acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are being developed using data obtained with the HTTR, which include reactor kinetics, thermal-hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). A three-gas-circulators trip test and a vessel-cooling-system stop test were planned as a loss-of-forced-cooling test and demonstrate the inherent safety features of HTGR. The vessel-cooling-system stop test consists of stopping the vessel-cooling-system located outside the reactor pressure vessel (RPV), to remove the residual heat of the reactor core as soon as the three-gas-circulators are tripped. All three-gas-circulators is tripped at 9 MW. The primary coolant flow rate is reduced from the rated 45 t/h to 0 t/h. The control rods are not inserted into the core and the reactor power control system does not operated. A core dynamics analysis of the loss-of-forced-cooling test of the HTTR is performed. Analytical results for the reactor transient during the test are presented in this report. It is determined that the reactor power immediately decreases to the decay heat level due to the negative reactivity feedback effect of the core, even though the reactor shutdown system is not operational, and that the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Furthermore, the relation between the reactivities (namely, the Doppler, moderator temperature, and xenon reactivities) affecting re-critical time and reactor peak power level and total reactivity is addressed. The analytical results will be utilized for the design and construction of the Kazakhstan High Temperature Reactor (KHTR) and the realization of commercial Very High Temperature Reactor (VHTR) systems.

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In this paper the thermal-hydraulic characteristics of the primary loop of China Experimental Fast Reactor (CEFR) are calculated and analyzed. A one-dimension, single-phase flow model is used to establish the system control equations. The single channel model is adopted in the reactor core, and a dynamic model of intermediate heat exchanger is built. At the same time, the property of sodium and flow and heat transfer correlations or models of sodium are collected and compiled. The discussion of the sensitivity of different flow and heat transfer correlations is given. The validation of the code developed in this paper shows that the code can be adopted to do some typical transient and accident analysis. The model and code presented in this paper can be used not only in the safety analysis of pool-type sodium cooled fast reactor, but also in the development of CEFR simulation platform.

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In the project of the Engineering Simulation System (ESS) of Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the model of steam generator (SG) was designed as a module matrix composed of several general modules, which could simulate heat exchange and flow characteristics under different conditions and in different sections of steam generator. Heat balance equations of the primary helium side and the secondary steam side were established separately. Heat transfer coefficients of both single-phase flow and two-phase flow were calculated using empirical correlations derived from experimental data, and pressure and flow rate were calculated by a flow network simulation solver. It concluded that the mathematic model presented in this paper could correctly describe the dynamic characteristic of steam generator in normal reactor power operation range.

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For supercritical pressure natural circulation loops, explicit correlation for steady state flow are not available. While using the subcritical natural circulation flow correlation for supercritical pressure data, it has been observed that subcritical flow correlation is not able to predict the steady state flow accurately near supercritical pressure condition. A generalized correlation has been proposed to estimate the steady state flow in supercritical pressure natural circulation loop based on a relationship between dimensionless density and dimensionless enthalpy reported in literature. This generalized correlation has been tested with the steady state supercritical pressure CO_2 data and found to be in good agreement. Subsequently supercritical pressure data for different working fluids reported in literature has also been compared with the proposed correlation. It is observed that the same generalized correlation is applicable for other fluids also. The present paper deals with the details of the test facility, the derivation of the generalized correlation and comparison with experimental data.

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Plant-specific analyses of the typical pressurized water reactor in Korea are performed to assure the structural integrity of the reactor pressure vessel during transient which is expected to initiate pressurized thermal shock event. The deterministic analysis is performed to determine the critical time interval in the transient during which mitigating action can be effective. Also, the failure probability is obtained by performing probabilistic fracture mechanics analysis. The probabilistic reactor integrity evaluation code, named R-PIE code, is developed in this study. In addition, several sensitivity analyses are performed for warm pre-stressing, plastic zone correction and simulation option to calculate fracture toughness etc. to assess its effect on the failure probability. The critical crack depth and vessel failure probabilities from the deterministic and probabilistic fracture mechanics analyses are obtained, generating the following conclusions. (1) By including WPS effect, critical crack depth is not changed but critical time interval increased about 100 seconds, giving more time for the operator take some mitigating actions. (2) The warm prestressing has a significant effect on the failure probability for SBLOCA by lowering it by more than 50%. (3) Consideration of internal pressure and plastic zone corrections increases about 10% of stress intensity factor and generates more than 200% of increase in the failure probability depending on the fluence level. (4) The simulation option to calculate RTNDT is a very important factor to get the failure probability with the difference of more than 3 orders depending on the simulation option.

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In order to plan a decommissioning, amount investigation of waste materials and residual radioactivity inventory evaluation must be performed at the first stage of preparatory tasks. These tasks are called radiological characterization. Reliable information from radiological characterization is crucial for specification of decommissioning plan. With the information, we can perform radiological safety analysis and optimize decommissioning scenario. Japan Atomic Power Company (JAPC) has already started preparatory tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1) that is the first commercial Small Type Light Water Reactor in Japan. To obtain reliable information about residual radioactivity inventory, we improved radioactivity inventory evaluation procedure. The procedure consists of neutron flux distribution calculation and radioactivity distribution calculation. We need a better understanding about characteristics of neutron transport phenomena in order to obtain reliable neutron flux distribution. Neutron flux was measured in Primary Containment Vessel (PCV) at 30 locations using activation foils. We chose locations where characteristic phenomena can be observed. Three dimensional (3D) neutron flux calculation was also performed to simulate continuous changes of neutron flux distribution. By assessing both the measured values and 3D calculation results, we could perform the calculation that simulates the phenomena well. We got knowledge about how to perform an appropriate neutron flux distribution calculation and also became able to calculate a reliable neutron flux distribution [2][3]. Using the neutron flux distribution, we can estimate a reliable radioactivity distribution. We applied network-parallel-computing method to the estimation [4][5]. And further we developed "flux level approximation method" which use linear or parabola fitting method to estimation. Using these new methods, radioactivity by neutron irradiation, which is radioisotope formation, was calculated at all space mesh points determined for neutron flux distribution calculation. Continuous changes of radioisotope concentration, that is radioisotope concentration distribution, became able to be observed. By using this distribution, a quantity of radioactive waste can be estimated more accurately and dismantling designs can be optimized.

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