This paper presents a new method for calculating normalized reactor coolant flow based on the electric current, line frequency, and voltage to the reactor coolant pumps (RCPs). This method-based on the Law of Energy Conservation-uses the daily reactor coolant flow calibration to factor out core flow static error (for example, due to crud build-up in the reactor coolant flow path or steam generator tube plugging) during operation. If a low reactor coolant flow event occurs, the Reactor Coolant Flow Calculator offers important reactor protection. The Reactor Coolant Flow Calculator has the following benefits: 1. It can detect design basis events (DBEs) related to reactor coolant flow faster and more accurately than existing methods. 2. The voltage and current sensors can be located outside of the containment building (for example, at the power source) for easy access and maintenance. 3. Sensors are common commercial off-the-shelf (COTS) items, and hardware obsolescence will not be a problem.

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PHE (Process Heat Exchanger) is a key component to transfer the high temperature heat generated from a VHTR (Very High Temperature Reactor) to the chemical reaction for massive production of hydrogen. The Korea Atomic Energy Research Institute established a helium gas loop for the performance test of VHTR components and manufactured a PHE prototype in order to test in the helium gas loop. In this study, as a part of high-temperature structural integrity evaluation on the PHE prototype which is scheduled to be tested in the helium gas loop, we carried out high-temperature structural analysis modeling, thermal analysis, and a structural analysis for the PHE prototype under the fixed loop test conditions as a precedent study prior to the performance test in the helium gas loop. The results obtained in this study will be applied to the design of the performance test setup for the PHE prototype in the near future.

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MAAP5(Modular Accident Analysis Program Rev. 5.0.0), developed by Fauske & Associates, Inc.'s (FAI) based on the MAAP4 code, is a severe accident analysis code. It is a computer program capable of simulating the response and mitigation actions of light water reactor nuclear power plants (NPPs), including advanced boiling water reactor (ABWR) during severe accident. The effectiveness of the execution of emergency operation procedures (EOPs) for an advanced boiling water reactor (ABWR) during postulated accident conditions using MAAP5 code is discuss in this paper. The simulation scenarios included the loss of feedwater pumps and turbine driven Core Isolation Cooling System (RCIC), the anticipated transient without scram (ATWS), and loss of reactor pressure vessel (RPV) water level indication. Based on the comparisons of responses on different parameters for cases with and without EOP actions, we concluded the RPV emergency depressurization (ED) in the EOP could effectively mitigate the consequences of the accident. The simulation clearly reveal that the execution of ED for the condition of RPV water level unknown could reduce primary containment pressure, suppression pool temperature, and avoid large fluctuation of RPV pressure, water level, and core power. For the RPV level unknown under ATWS condition, RPV ED can give the operator the easier way to control RPV pressure, and to assure adequate core cooling.

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The purpose of this paper is the use data assimilation techniques, inspired from meteorological applications, to perform an optimal reconstruction of the neutronic field in a nuclear core. Both measurements, and information coming from a numerical model, are used in this purpose. We first study the robustness of the method when the amount of measured information varies. We then study the influence of the nature of the instruments and their spatial repartition on the efficiency of the field reconstruction. Such study allows, also enlightening the instruments providing the most information within a data assimilation procedure. The study of various network configurations of instruments in the nuclear core establishes that influence of the instruments depends both on the individual instrumentation location as well as on the chosen network.

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Several studies show that data assimilation methods, by merging information both from model and measurement, can be used to elaborate an optimal determination of neutronic activity field within a nuclear PWR reactor core. Here the problem addressed and solved is to determine an optimal repartition of the instruments within the core, to get the best possible reconstructed field using a data assimilation procedure. The position optimization is based on simulated annealing with a Metropolis-Hasting algorithm. To perform the method in the framework of data assimilation, several algebraic optimisation related to computing time of data assimilation has been developed and are presented here.

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A lifetime evaluating method of variable accelerated factor was proposed based on the characteristic of accelerated ageing test for fuse. This method, compared with the method of constant accelerated factor, was more accurate and effective to evaluate fuse's lifetime. In this paper, two methods based on the constant accelerated factor and variable accelerated factor were used to evaluate the lifetime of a type of fuse, and the results was compared to confirm the effectiveness.

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Nuclear knowledge management (KM) initiatives have been observed in many international agencies, industries and research facilities, such as the International Atomic Energy Agency (IAEA) and Canada's own Atomic Energy of Canada Limited (AECL). This paper presents a brief overview of the theory behind concept mapping, its origins and application potential in an organizational context. Successful applications of nuclear KM efforts by the IAEA, the Defence Threat Reduction Agency (DTRA), AECL and the Organization for Economic Cooperation and Development, Nuclear Energy Agency (OECD/NEA) are specifically highlighted in the report for comparison and contrast. In addition, a system based on concept mapping theory is recently designed and implemented in order to efficiently manage steam generator-related knowledge at the Canadian Nuclear Safety Commission (CNSC). This tool aims to act as a powerful information asset for storage and retrieval of explicit and implicit knowledge, and to integrate KM practices into the CNSC workplace, all while maintaining its simplicity and ease for future modifications and fine-tuning when needed.

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Thin rod bundles with tight lattice are arranged according to the equilateral triangle grid, as the proportion of fuel is large, and the power density of core is high. Based on the analysis of the performance of core, the ABV-6M reactor is taken as the example, and two objective functions, power density and flow rate of coolant are proposed for optimization calculation. Diameter and pitch of rod are optimized by using GA method respectively. The results, which are considered to be safety in security checking, show that tight lattice is effective for improving the power density and other performances of the reactor core.

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The single channel or parallel channel density wave oscillation (DWO) type flow instability is a well-known and important issue in the operation of Boiling Water Reactors (BWR), which needs to be avoided during the reactor design. The proper simulation of the two phase flow is very important in the BWR single channel thermal-hydraulic stability analysis. The supercritical fluids are widely proposed to be used in the Gen-IV reactor designs such as Supercritical Water Cooled Reactor (SCWR) and Gas Cooled Fast Reactor (GFR). For those supercritical fluids cooled Gen-IV reactor designs, although the coolant is in single phase supercritical pressure condition during steady state operation, the two phase flow in subcritical pressure will occur during some off-normal conditions. Therefore, the appropriate two phase flow models also need to be developed for the stability analysis of the Gen-IV reactor designs. To investigate the two phase flow modeling effects on the nuclear reactor single channel thermal-hydraulic stability analysis, four two phase flow models, namely, the Homogenous-Equilibrium model (HEM), the Homogenous-Nonequilibrium model (HNEM), the Nonhomogenous-Equilibrium model (NHEM) and the Nonhomogenous-NonEquilibrium model (NHNEM), are developed and applied to a typical SCWR hot channel. The neutral stability boundaries were derived using a linear model in the frequency domain. The stability boundaries are compared and plotted in the traditional Subcooling number versus Phase change number plane. It was found that the homogenous models predict more conservative stability boundaries than the nonhomogenous models, and the difference of the stability boundaries predicted by four two phase flow models is reduced in the higher pressure conditions.

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Condition monitoring and detecting early signs of potential failure mechanisms present particular problems in vertical pumps. Most often, the majority of the pump assembly is not readily accessible for visual or audible inspection or conventional vibration monitoring techniques using accelerometers and/or proximity sensors. The root cause failure analysis of a 2-stage vertical centrifugal service-water pump at a nuclear power generating facility in the USA is presented, highlighting this long standing challenge in condition monitoring of vertical pumps. This paper will summarize the major findings of the root cause analysis (RCA), highlight the limitations of traditional monitoring techniques, and present an expanded application of motor current monitoring as a means to gain insight into the mechanical performance and condition of a pump. The "real-world" example of failure, monitoring and correlation of the monitoring technique to a detailed pump disassembly inspection is also presented. This paper will explain some of the reasons behind well known design principles requiring natural frequency separation from known forcing frequencies, as well as explore an unexpected submerged brittle fracture failure mechanism, and how such issues may be avoided.

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Once-through steam generator (OTSG) is usually used in the integrated nuclear power plants which require smaller volume and better effect of heat transfer. The double-tube OTSG component which is composed of straight tube outside and helical tube inside is presented in this paper. The primary fluid is divided into two parts, one is in the inner tube and the other is in the gap among outer tubes. The flow distribution ratio of the primary fluid obviously affects the heat transfer. Thus, the problem of optimization emerges, i.e. how to find an optimal flow distribution ratio with a maximum heat exchange. Analyzed the effects of the distribution ratio on heat transfer, the optimal distribution ratio is obtained by the constrained nonlinear optimization method. Subsequently, the optimal distribution ratio is achieved by a throttling set in the entrance of the inner tube. The result is in substantial agreement with the literature.

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Coordination control of nuclear power plant is a complex problem researchers faced. In this paper a scheme that the feed water valve opening as a power reference signal is presented. And it builds the physical models based on the lumped parameters equations, and designs the control system according to the scheme. At last, the simulation results show that this scheme is valid and the outlet pressure has less overshoot.

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The VR-1 training reactor is a pool-type light-water reactor with the low-enriched uranium and maximum thermal power of 1 kW. The reactor is mainly used for students' training. The training is aimed to areas such as the reactor physics, neutronics, dosimetry, nuclear safety and I&C systems. Since neutron flux in the VR-1 core is well measured, this work focuses on one part of the reactor - its Radial experimental Channel (RC). This paper deals with the measurement of the neutron distribution by means of gold-foil neutron-activation technique and continual measurement with 3He-filled detector. Obtained experimental results were verified with the simulation in the Monte-Carlo N-Particle Transport Code. Results and conclusions from this paper will be used for further investigation of neutrons and their spatial distribution inside the low-power training reactor. Also, the data obtained in this paper can be used as a basis for future detailed measurements of neutron flux and its distribution in other hard accessible areas inside the reactor. The paper gives a simple theoretical introduction concerning neutron measurement procedures and available techniques in this field, which is particularly important for improving training courses and a content of offered experiments in the VR-1 reactor.

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To simulate the reactor system dynamic features during density wave oscillations (DWO), both the non-linear method and the linear method can be used. Although some transient information is lost through model linearization, the high computational efficiency and relatively accurate results make the linear analysis methodology attractive, especially for prediction of the onset of instability. In the linear stability analysis, the system models are simplified through linearization of the complex non-linear differential equations, and then, the linear differential equations are generally solved in the frequency domain through Laplace transformation. In this paper, a system response matrix method was introduced by directly solving the differential equations in the time domain. By using a system response matrix method, the complicated transfer function derivation, which must be done in the frequency domain method, can be avoided. Using the response matrix method, a model was developed and applied to the single channel or parallel channel type instability analyses of the typical proposed SCWR design. The sensitivity of the decay ratio (DR) to the axial mesh size was analyzed and it was found that the DR is not sensitive to mesh size once sufficient number of axial nodes is applied. To demonstrate the effects of the inlet orificing to the stability feature for the supercritical condition, the sensitivity of the stability to inlet orifice coefficient was conducted for hot channel. It is clearly shown that a higher inlet orifice coefficient will make the system more stable. The susceptibility of stability to operating parameters such as mass flow rate, power and system pressure was also performed. And the measure to improve the SCWR stability sensitivity to operating parameters was investigated. It was found that the SCWR stability sensitivity feature can be improved by carefully managing the inlet orifices and choosing proper operating parameters.

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In a PWR plant, a steam-water two phase flow may possibly exist in the pressurizer spray pipe under a rated power operating condition since the flow rate of the spray water is not sufficient to fill the horizontal section of the pipe completely. Under such thermally stratified two phase flow conditions, the initiation of high cycle fatigue cracks is suspected to occur due to cyclic thermal stress fluctuations caused by oscillations of the water surface, which cannot be detected by the measurement of temperature on outer surface of the pipe. In order to clarify the flow and thermal conditions in the pressurizer spray pipe and assess their impact on the pipe structure, an experiment was conducted for a steam-water flow at a low flow rate using a mock-up pressurizer spray pipe. By measuring inner wall temperature fluctuations, continuous temperature fluctuations, which were around 0.2 times of the steam water temperature difference in the maximum range, were observed at the inclined section where the water surface contacted the pipe wall. Then, we investigated the causes of the fluctuations by visualization tests. As a result of the experiment, it seemed that wall temperature fluctuations were not caused by waves on the water surface, but were caused by liquid temperature fluctuations a layer below the steam-water interface. The influence of a small amount of non-condensable gas dissolved in the reactor coolant on the wall temperature fluctuations was investigated by injecting air into the experimental loop. The liquid temperature fluctuations in the layer which caused wall temperature fluctuations were attenuated after air was injected.

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The understanding of the reflood process of a severely damaged reactor core represents a challenge in the prediction of safety margin of existing and future pressurized water reactors. After the TMI-2 accident, the understanding of coolability of severely damaged reactor core became an objective of many theoretical and experimental studies. Currently, the French Institute of Radioprotection and Nuclear Safety (IRSN) has started two experimental programs, PRELUDE and PEARL, to investigate the physical phenomena during a reflood process at high temperature and to provide relevant data in order to improve predictive models. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core. The presented model is based on the theory of heat transfer and two-phase flow in porous media and in small hydraulic diameter channels. The proposed model is implemented into the European computer code for severe accident analysis ICARE-CATHARE. The comparison of the calculations with PRELUDE experimental results is presented. Finally, the issue of transposition to the reactor scale is discussed and some answers are proposed using calculation results for a debris bed in a configuration similar to what could be expected in a severely damaged reactor core.

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The mass media has the potential to effect the utilization of nuclear power in Japan. In most cases journalists contact PR staff of the nuclear energy industry (hereinafter called "the industry") to collect information about various events of nuclear energy. The industry is always ready to distribute related information and hold a press conference timely when necessary. In terms of the organizational structure for the PR activities each electric power company organizes the PR section in-house. The PR staff provides journalists with information on a daily basis. For the purpose of grasping the mass media's awareness, the author conducted interviews with 22 journalists who had experience in reporting news on nuclear energy subjects. The result showed that the journalists recognized the necessity of nuclear energy. The interviewees suggested that a proper press launch should be needed at just the right time especially in emergency situations and a press release should be more easily understandable. This interview showed that journalists considered the media reports as reflection of citizens' opinion. Most of the journalists realize that the influence of the media coverage should not be negligible and they acknowledge commutation between the two sides is gradually improved compared to before.

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National engagement as well as bilateral or multi-national cooperation in HTGR fuel development is ongoing and is expected to further improve fuel performance and the ability to make reliable predictions. The accident condition benchmark exercise, one of the key elements within the sixth IAEAdirected Coordinated Research Project (CRP) on "Advances in HTGR Fuel Technology Development", has successfully demonstrated to be a useful basis for verification and validation in establishing the reliability of code predictions. Participants in the accident condition benchmark included France, Germany, Russia, South Africa, Korea, and the United States applying a total of eight models to all or a part of the 24 proposed benchmark cases. The benchmark consisted of three parts, a sensitivity study to examine fission product release from a fuel particle, the postcalculation of well documented irradiation and heating experiments, and finally some predictive calculations. In the sensitivity study, most codes have shown good agreement among each other. Differences can be explained by different assumptions for input data or boundary conditions. In comparison with the numerical procedure of the diffusion calculation for the kernel, the application of the analytical solution offered by the Booth model appears to be more accurate method. Time step length may also influence the calculational results. From the postcalculations of heating tests, it appears that the diffusion coefficient for cesium in silicon carbide is still varying over a broad range. In particular, strontium release data are obviously largely overpredicted and should undergo a thorough review. Silver release measurement results are often unexpected and inconsistent, and therefore extremely difficult for postcalculation. One of the most recent heating experiments, HFR-K6/3, has shown surprisingly low krypton and cesium release values, which are largely overpredicted by the model calculations. This extremely good accident condition performance of the very latest manufacture of German TRISO fuels should be the starting point of further fuel studies. Additional irradiation and postirradiation examination will be required for newly manufactured fuel, in order to improve the existing statistical data base for fuel performance analysis.

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A methodology for calculating inventories for the NBSR has been developed using the MCNPX computer code with the BURN option. A major advantage of the present methodology over the previous methodology, where MONTEBURNS and MCNP5 were used, is that more materials can be included in the model. The NBSR has 30 fuel elements each with a 17.8 cm (7 in) gap in the middle of the fuel. In the startup position, the shim control arms are partially inserted in the top half of the core. During the 38.5 day cycle, the shim arms are slowly removed to their withdrawn (horizontal) positions. This movement of shim arms causes asymmetries between the burnup of the fuel in the upper and lower halves and across the line of symmetry for the fuel loading. With the MONTEBURNS analyses there was a limitation to the number of materials that could be analyzed so 15 materials in the top half of the core and 15 materials in the bottom half of the core were used, and a half-core (east-west) symmetry was assumed. Since MCNPX allows more materials, this east-west symmetry was not necessary and the core was represented with 60 different materials. The methodology for developing the inventories is presented along with comparisons of neutronic parameters calculated with the previous and present sets of inventories.

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The sipping test devices are used to identify the defective fuel. The defective fuel can be identified by detecting the occurrence of the fission products entrained by the medium rising around the fuel rods. There are two kinds of sipping test device to inspect the tightness of the irradiated fuel assembly from nuclear power plant (NPP). The in-mast sipping taking the mast as the isolator, the air as the medium, and the Xe 133 as the indication nuclide is used for qualitative tightness test of each fuel assembly during refueling operation above the reactor. The poolside sipping taking the sipping cell as the isolator, the air and water as the medium is used for quantitative confirming the diagnosis of the in-mast sipping and identifying the tightness of the fuel at the side of the fuel storage pool after refueling. The design manufacture and calibration of three devices were successfully completed domestically step by step with serious quality assurance and quality control.

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The solution for source transfer time beyond criteria problem in LingAo project II is introduced in the paper. System is modified and criteria are updated. Analysis and tests are performed to get proper parameter. International rules are used to evaluate the new criteria. The problem is solved finally, contributing stable power supply to unit pre-operation, start-up and operation. Experience also gathered for CPR1000 new projects.

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The PUREX process has been progressively and continuously improved during the past three decades, and these improvements account for successful commercialization of reprocessing in a few countries. The renewed interest in nuclear energy and the international growth of nuclear electricity generation do not equate - and should not be equated - with increasing proliferation risks. Indeed, the nuclear renaissance presents a unique opportunity to enhance the culture of non-proliferation. With the recent revival of interest in nuclear technology, technical methods for prevention of nuclear proliferation are being revisited. Robust strategies to develop new advanced separation technologies are emerging worldwide for sustainability and advancement of nuclear energy with a decrease in proliferation risks. On the other hand, at this moment, advanced technologies with reduced proliferation risks are being developed. Until now proliferation resistance as it applies to reprocessing has been focused on not separating a pure stream of weapons-usable plutonium. France, as an example, has proposed a variant of the PUREX process, the COEXTM process, which does not result on a pure plutonium product stream. A further step is to implement a process based on group extraction of actinides and fission products associated with a homogeneous recycling strategy (UNEX process in the U.S., GANEX process in France). Such scheme will most likely not be deployable on an industrial scale before 2030 or so because it requires intensive R&D and robust flow-sheets. Finally, future generation recycling schemes will likely handle the used nuclear fuel in fast neutron reactors. This means that the plutonium throughput of the recycling process may increase. The need is obvious for advanced aqueous recycling technologies that have less proliferation risk than the commercial PUREX process. In this paper, we review the actual PUREX process along with the advanced recycling technologies that will enhance technical barriers that make plutonium diversion more difficult by either not isolating plutonium and/or coexistence of fission products with plutonium.

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A high-fidelity three-dimensional space time nodal method has been developed to simulate the dynamics of the reactor core for real time simulation. This three-dimensional reactor core mathematical model can be composed of six sub-models, neutron kinetics model, cay heat model, fuel conduction model, thermal hydraulics model, lower plenum model, and core flow distribution model. During simulation of each sub-model some operation data will be produced and lots of valuable, important information reflecting the reactor core operation status could be hidden in, so how to discovery these information becomes the primary mission people concern. Under this background, data mining (DM) is just created and developed to solve this problem, no matter what engineering aspects or business fields. Generally speaking, data mining is a process of finding some useful and interested information from huge data pool. Support Vector Machine (SVM) is a new technique of data mining appeared in recent years, and SVR is a transformed method of SVM which is applied in regression cases. This paper presents only two significant sub-models of three-dimensional reactor core mathematical model, the nodal space time neutron kinetics model and the thermal hydraulics model, based on which the neutron flux and enthalpy distributions of the core are obtained by solving the three-dimensional nodal space time kinetics equations and energy equations for both single and two-phase flows respectively.Moreover, it describes that the three-dimensional reactor core model can also be used to calculate and determine the reactivity effects of the moderator temperature, boron concentration, fuel temperature, coolant void, xenon worth, samarium worth, control element positions (CEAs) and core burnup status. Besides these, the main mathematic theory of SVR is introduced briefly next, on the basis of which SVR is applied to dealing with the data generated by two sample calculation, rod ejection transient and axial flux offset. Comparing with the traditional data analysis method, the main character of SVR is not enforcing a priori information or judgment before analysis begins, but obtaining the fitting result from inner relationship of data completely, which represents the main principle of data mining, letting the data speak themselves, and the result is more accurate and believable than traditional methods.

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Flow characteristics of single-phase nitrogen in the three different micro-channels of layered MEMS have been investigated numerically in view of the effect of compressibility and viscosity heating. It is shown from the computing results that the influence of viscosity heating increases with increasing Re and reducing channel size. Compressibility becomes obviously in lower Mach number with smaller channel size. In this paper, the effect of size, ratio of length to diameter, entrance effect and pressure drop on resistance characteristic are analyzed. The accuracy of calculation results has been confirmed by experimental data of Zhang[1].

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Triple-offset butterfly valve is one advanced kind of butterfly valve. It is potential in nuclear power plants because of its advantages in high temperature and high pressure occasions. There are few papers on performance of triple-offset butterfly valve. This paper is intended to predict the performance of a triple-offset butterfly valve used in a nuclear power plant using computational fluid dynamics. The flow field and aerodynamic torque on the triple-offset butterfly valve were studied at six different disk positions from 90 deg to 20 deg (where 90 deg is in the full open position). The selected six different disk positions indicated a stroke. The flow fields were predicted using the k-epsilon renormalization group theory (RNG) turbulence model. The computational results were obtained using CFX 12. The flow field is illustrated using velocity contours and disk pressure profiles, illustrating the effects of the disk position. Some results of flow field are compared to those of symmetric disk butterfly valve which have been validated by test results. Based on the flow field, valve resistance coefficient and aerodynamic torque coefficient with the disk positions are obtained, providing a better understanding of the performance of the triple-offset butterfly valve throughout a stroke.

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A new and rigorous modeling of basic transport equation and constitutive equations of turbulent transport terms of interfacial area concentration was carried out. In the basic transport equations, interfacial area concentration is transported by averaged interfacial velocity In the previous models, interfacial velocity is roughly approximated by velocity of each phase. In the present model, interfacial velocity is rigorously formulated in term of spatial correlation functions of characteristic function and velocity of each phase and their directional derivatives. In this new formulation, the averaged interfacial velocity was shown to be correlation functions of fluctuation of velocity and local instant void fraction and their derivatives which reflect the transport of interfacial area concentration due to interaction between interfacial area and turbulence of each phase. Basic conservation equations of spatial correlation functions of characteristic function and velocity of each phase were also derived based on the conservation equations momentum and its fluctuation of each phase. For practical purpose, further modeling of this turbulent transport terms of interfacial area concentration was carried out. Constitutive equations of turbulent diffusion and lateral migration of interfacial area concentration were obtained which can be applied to various flow regime of two-phase flow. Using the present model of turbulent transport of interfacial area concentration, prediction of interfacial area distributions in gas-liquid two-phase flow was carried out. Systematic prediction was carried out using turbulence model of gas-liquid tow phase flow. Distributions of void fraction, averaged velocity and turbulent velocity of liquid phase were also predicted along with radial distributions of interfacial area concentration. Experimental data of distribution of interfacial area concentration in bubbly flow were well predicted by the present model.

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This paper describes the critical digital review (CDR) procedure, which was developed by Institute of Nuclear Energy Research (INER), and sponsored by Taiwan Power Company (TPC). A preliminary CDR application experience which was performed by INER, is also described in this paper. Currently, CDR becomes one of the policies for digital Instrumentation and Control (I&C) system replacement in TPC. The contents of this CDR procedure include: Scope, Responsibility, Operation Procedure, Operation Flow Chart, CDR review items. The "CDR Review Items" chapter proposes optional review items, including the comparison of the design change, Software Verification and Validation (SV&V), Failure Mode and Effects Analysis (FMEA), Evaluation of Watchdog Timer, Evaluation of Electromagnetic Compatibility (EMC), Evaluation of Grounding for System/Component, Seismic Evaluation, HFE Evaluation, Witness and Inspection, Lessons Learnt from the Digital I&C Failure Events. Since CDR has become a TPC policy, Chin Shan Nuclear Power Plant (NPP) performed the CDR practice of Automatic Voltage Regulator (AVR) digital I&C replacement, even though the project had been on the half way. The major review items of this CDR were: the comparison of the design change, SV&V, FMEA, Evaluation of Watchdog Timer, Evaluation of Electromagnetic Compatibility (EMC), Evaluation of Grounding for System/ Component, Witness and Inspection, Lessons Learnt from the Digital I&C Failure Events. The experience of the CDR showed the importance of preparation of the documents by the vendor. This means the communication with the vendors for the bid preparation is crucial.

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Primary hydriding in zircaloy claddings once was one of the main causes of fuel rod failures at earlier nuclear power plants. The main sources of hydrogen leading to typical primary hydriding failure are residual moisture and hydrogen in the UO2 pellets, moisture introduced in the time of pellet loading and gas filling, and organic materials trapped inside fuel rod in the manufacturing process. During operation, the reaction between the moisture, or organic material, and the cladding and pellet results in release of gaseous hydrogen, which can be locally absorbed in the cladding and lead to fuel rod failure. Design criteria on hydrogen content in fuel rod were established to keep operation safety, such as moisture level of 2 mg of water per cubic centimeter of free volume and equivalent hydrogen content of fuel pellets before loading not more than 1.3μg/gUO2. In the course of manufacturing, equivalent hydrogen content is guaranteed by pellet drying, time control from pellet drying to loading, vacuumizing while gas filling, and so forth. In the present paper, the main sources of hydrogen in fuel rod together with their determining factors were introduced, and the corresponding design criteria were discussed. It is considered that equivalent hydrogen content per cubic centimeter of hot void in fuel rod is the most reasonable design limits for hydrogen content.

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Project Taishan is under construction in CHINA. It is a third generation nuclear power station designed with the advance EPR concept. Project Taishan's reference nuclear power station is Project FA3. For these two EPR projects, the dimension of fuel pool is the same but they are different in the storage capacity of the spent fuel required. For Project Taishan, the spent fuel pool is designed to store the spent fuels for 20 years, which means that the unloading fuel from the reactor core needs to be kept for more than 13 cycles (one cycle is 18 months). From the view of economy, more fuel assemblies are preferred to be stored in the spent fuel pool before being transported out of nuclear island. Now only 1288 effective cells on storage racks are available for fuel assemblies. 62 cells are unavailable because of the layout of equipment concerning spent fuel pool[1]. In this paper, the layout of the equipment concerning the spent fuel pool is studied. The purpose is to find out the design defect which blocks the usage of the fuel cells in spent fuel pool and to increase the available fuel cells. Several design defects are found after researching, for example, the span limit of Auxiliary Bridge in Fuel Building, the layout of the lighting under water in spent fuel pool, the specification of the penetration connection between the cooling pipes and the spent fuel pool, and etc. The analysis will be performed according to these design defects, and find out the solutions to improve the layout of equipment and increase the number of the available cells in spent fuel pool. An optimization report will be prepared to describe all improvement proposals. Increasing the containing capacity of the spent fuel is very useful for improving the economy during the power station operation stage. And the optimization proposals will be good reference for the design of the other EPR project, such as project Taishan phase II.

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The P-PIE (Probabilistic Piping Integrity Evaluation Program) code to evaluate the pipe failure probability was developed in this study under stress corrosion cracking (SCC) and fatigue condition. The validation of the P-PIE code was performed by comparison with other codes. By using P-PIE program, the pipe failure probabilities of the reactor coolant system in Korean PWR (Pressurized Water Reactor) plants were evaluated. 3 reactor types including Westinghouse 2-loop plants (WH-2), Westinghouse 3-loop plants (WH-3), and Framatome 3-loop plants (FR-3 which are operating in Korea, were considered in the evaluation. The effect of SCC-related variables such as oxygen concentration during start up and steady state operation, and operating temperature on the piping failure probabilities was investigated. The effects piping loops and reactor types on the piping failure probability were also investigated. The results show that (1) LOCA (loss of coolant accident) probability of Korean PWR plants is extremely low, (2) leak probability is sensitive to pipe thickness, operating temperature and oxygen concentration during steady state operation under SCC, (3) net section stress criterion gives more conservative results than tearing modulus criterion under fatigue, and (4) PSI and ISI can reduce pipe failure probability.

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Background of the work is the loss of coolant accident (LOCA) with release of the insulation material in a nuclear power plant (NNP). The material can be transported into the reactor containment, the building sump of the containment and into the associated systems. To secure the heat dissipation from the reactor core and the containment the cooling systems transports the water from the sump to the condensation chamber and then into the reactor pressure vessel. The function of the pumps can be affected by a high allocation of the strainers with fractionated insulation material and the heat dissipation is not guaranteed. A prediction of the routes of transport of the material is necessary to get information on the quantity of the insulation material is settled down and not transported to the strainers. To simulate the transport of the material validated CFD models have to be developed. A joint research project to investigate the problem more in detail, particularly with the aim of CFD model development, is being performed in cooperation of the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of the University in Zittau/Gorlitz with the Research Center Dresden (FZD) [1]. The main aim is the calculation of the mass of insulation material transported to the strainers. This information is the essential to affect the differential pressure on the strainers after a LOCA. The paper deals with the experimental and methodical work for the description of the transport of fragmented insulation material in a small channel. The work is divided into two steps. In the first step, the water flow was investigated on the test facility "Ring Channel" at the IPM. For this, different measurement techniques, such as ultrasonic sound for the mean velocity and an impeller sensor for the velocity in the middle of the channel (maximum velocity) as well as Particle Image Velocimetry (PIV) were used. The results were compared to theoretical calculation of the mean and maximum velocities in small channels found in the literature. Based on the geometry of the test facility, CFD-calculations of the water flow with the main focus on the turbulence model were performed. The simulation results were compared to the measured values of the experimental values. The second step was the investigation of the transport of insulation material in this channel. For this, a defined amount of material was inserted in the water flow at different water velocities. The transport of the material was captured by digital cameras followed up by the analysis using digital image processing. Algorithms for the analysis of the single particles and the particle collective behavior were developed. One of the results is a particle size distribution. Based on this the definition of the disperse phase in the CFD-Simulation is set and CFD-Simulations were performed. For the validation the volume fraction of the dispersed phase of the CFDSimulation in defined virtual planes and the particle collective behavior from the experiments were compared.

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The four basic aspects of strengths, ductility, toughness and fatigue strengths can be summarized for overall mechanical properties requirements of materials for nuclear pressure-retaining vessels in ASME-BPV code. These mechanical property indexes involve in the factors of melting, manufacture, delivery conditions, check or recheck for mechanical properties and chemical compositions, etc. and relate to degradation and damage accumulation during the use of materials. This paper specifically accounts for the basic requirements and theoretic basis of mechanical properties for nuclear pressure vessel materials in ASME-BPV code and states the internal mutual relationships among the four aspects of mechanical properties. This paper focuses on putting forward at several problems on mechanical properties of materials that shall be concerned about during design and manufacture for nuclear pressure vessels according to ASME-BPV code.

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In nuclear radiation measurement, pulse height discriminator (PHD) and single channel analyzer (SCA) are widely used to select pulses whose amplitude is within user's interesting range. PHD is derived from universal double voltage comparators by adding additional anti-coincidence circuits. Simple anti-coincidence method will generate false result, for the leading edge of upper comparator output falls behind that of lower comparator and pulse width of upper comparator output is narrower than that of lower comparator output. Anti-coincidence circuits in a conventional PHD and SCA often contain quite a few different circuits or components, such as monostable multivibrator and/or bistable flip-flop, NOR or NOT gate, etc. Such complex structure and manifold components deteriorate PHD's and SCA's response time and reliability. Great efforts have been made to simplify its structure and improve its reliability. Based on analyzing time sequence of the outputs of upper and lower voltage comparators, a novel, simple and practical structure is presented that contains only monolithic circuit (dual monostable multivibrator). Concrete circuits are fabricated and waveforms captured from digital storage oscilloscope (DSO) prove that it works properly and reliably till the input signal frequency increases up to 1MHz, regardless of sine wave, exponent falling or impulse signal. It can be utilized in the field of nuclear radiation measurement.

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We have developed a best-estimate transient analysis code TRACT, for licensing transient analyses for current BWR plants and the development of next-generation BWR system. In addition to the POST-BT correlations, capability of modeling all fuel bundles individually in a reactor core and several models required for BWR licensing transient analyses, TRACT adopts a modified one-group neutron kinetics model as well as a three-group neutron kinetics model with the advanced nodal method. For both kinetic models, an improved quasi-static method is used as it is found to be suitable for evaluation of rapidly varying phenomena in postulated transient analyses. With the modified one-group kinetic model, TRACT has been validated through a number of analyses for separate effects tests (SETs), component effects tests (CETs), integral system effects tests (IETs), and BWR plant tests. From these analyses, the results for the SPERT-III test, the Peach Bottom-2 turbine trip test, and a BWR5 load rejection test are presented. The three-group kinetic model was applied to an ABWR MSIV full closure analysis to check the operation. It is confirmed that the calculated power change during the incident was similar to the one analyzed by the modified one-group neutron kinetic model.

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The natural convection heat transfer from a horizontal cylinder placed in liquid metal pool is computationally investigated. The heat flux from surface of cylinder to the bulk liquid induces flow due to buoyancy. The effect of cylinder diameter, heat flux and bulk fluid temperature on the heat transfer coefficient is studied. The flow is computed using PHOENICS CFD software. The simulation is transient. The geometry is discretized using a 3D body fitted grid. The temperature variation of the cylinder along its periphery is computed from simulation. The validation of CFD results is performed by comparing the computed wall temperature variation with previously published experimental data. Excellent agreement of computed results with experimental data is observed for various heat fluxes. The heat transfer data is presented in the form of variation of Nusselt number with Gr・Pr^2 (1+Pr). The heat transfer coefficient is found to rise slightly with bulk temperature. However, the cylinder diameter has a more significant effect on increasing heat transfer coefficient. Simulation data is generated by varying cylinder diameter, heat flux and bulk temperatures. Based on the data, a correlation is proposed. The correlation is shown to predict better compared to other correlations which were previously proposed.

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Air ingress is considered as one of the severe hypothetical accidents for high temperature gas-cooled reactors, because it possibly results in graphite oxidation reaction of reflectors and fuel elements so as to weaken the structural strength, impact the fission-production-retention capacity of coated particle and produce flammable gas mixtures. The research of air ingress is significant for the study of the reactor accidental characteristics and the improvement of the safety design. Based on the preliminary design of the 200 MWe High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM), the air ingress into the core, due to the chimney effect caused by the simultaneous rupture of both upper and lower fuel discharging pipes connected to the primary loop, was analyzed in detail. The mechanism and the characteristics of the air ingress are firstly introduced in this paper. Then the key parameters, such as natural convection flow rate, core graphite corrosion rate, oxygen consumption quantity and fuel element temperatures, are studied by the help of TINTE code. The calculation results indicate that, during this kind of air ingress accident, the graphite corrosion in the reactor core and reflector is very slow, which will not lead to the deterioration of heat transfer and the excess over the design limitation of the maximum fuel temperature. Therefore, the integrity of the fuel particles and the ability of retaining fission product will be kept well. Besides, the very low natural convection flow rate will not damage the structure strength in a considerably long term, so that there is enough time to adopt appropriate measures to cut off the air source to impede the continuous graphite corrosion and the reactor safety will not be endangered any more.

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According to the relationship between the reactor power, the core power distributions and the measured signals of power range channels of the ex-core detectors (RPN), the mathematical formula of calibration for parameters KU, KL and α can be deduced. On the Unit 3 of LingAo Phase 2, during the plant's first Startup, some tests were carried out at various power levels of the reactor. At each power level, full or partial flux map measurements were done by the in-core neutron measurement system (RIC), during the measurements of flux map, several heat balance measurements were done on the special equipment named KME system, and at the same time the signals of the RPN power range channels were recorded. From these tests, exact reactor powers, the core power distributions and the signals of the RPN detectors can be known. And then, the parameters KU, KL and α calibration results can be calculated by putting the Data into the formula deduced by us. According to the result of the calibration, the biggest error of the power is -0.59%FP and the biggest error of axial power deviation is -0.59%FP which was very good to the criterion.

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In order to reduce the crack growth rate on the welding of penetration pipe, Pressurized Water Reactor (PWR) of Maanshan nuclear power plant (NPP) uses vessel head injection to cool vessel lid and control rod driving components. The injection flow from the cold leg is drained by the pressure difference between cold leg and upper internal components. In this study, 10 million meshes model with 4 sub-models have been developed to simulate the thermal-hydraulic behavior by commercial CFD program FLUENT. The results indicate that the injection nozzles can provide good cooling ability to reduce the maximum temperature for lid on the vessel head. The maximum temperature of vessel lid is about 293.81℃. Based on the simulated temperature, ASME CODE N-729-1 was further used to recount the effective degradation years (EDY) and reinspection years (RIY) factors. It demonstrates that the EDY and RIY factors are still less than 1.0. Therefore, the re-inspection period for Maanshan PWR would not be significantly affected by the miner temperature difference.

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Pool boiling heat transfer experiments were preformed for the diameter D of 1.0 mm 〜 20.0 mm by using ethanol at 0.1 MPa. The trend of the nucleate boiling was well expressed with the Rohsenow correlation for D = 20, 3 and 2 mm. In the case of D = 1.0 mm, the increasing rate of the wall super-heat with an increase in the heat flux was larger than that of the Rohsenow correlation. The heat transfer surface was covered with a large bubble as the heat flux was increased. The heat flux at which the large bubble began to cover the heat transfer surface started to decrease as the diameter of the heat transfer surface became smaller than 2 mm. The CHF of D = 20 mm agreed well with the value predicted with the Kutateladze correlation. The CHF began to decrease as the diameter of the heat transfer surface became smaller than 2 mm.

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The Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) is proceeding as one of the ITER Broader Approach (ITER-BA). A Li circulation loop for testing hydraulic stability of the Li target (high speed free-surface flow of liquid Li as a beam target) and Li purification traps are under construction in the Japan Atomic Energy Agency as a major Japanese activities in the EVEDA. This paper presents specification of an electro-magnetic pump (EMP) for the EVEDA Li Test Loop (ELTL) and evaluation of the pressure drop in the main loop of the ELTL. The EMP circulates the liquid Li at a large flow rate up to 0.05 m^3/s (3000 l/min) under a vacuum cover gas (Ar) pressure of 10^<-3> Pa, thus the evaluation of cavitation generation is a crucial issue. The EMP used in the ELTL consists of two EMPs aligned in series through a U-tube whose size of one EMP is 0.8 m square and 2.6 m in length. The calculation of the pressure drop in the main Li loop to the EMP is approx. 25 kPa at the design maximum flow rate of 0.05 m^3/s. On the other hand the height from the EMP to a Li tank to supply Li to the EMP is designed to be 9.72 m, and secures a static pressure and the cavitation number of 18 kPa and 3.4 respectively at the maximum flow rate in a vacuum condition. As a result, it is confirmed to prevent cavitation at the inlet of the EMP in this design.

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An interface capturing scheme, implemented within the framework of the conservative phase-field method, is applied to simulate the incompressible two-phase flows with large density ratio and surface tension. For accurately predicting the solution, the finite difference dispersionrelation-preserving advection schemes are utilized to accommodate the true dispersion relation. The solutions computed from the resulting dispersion-relation-preserving advection schemes can minimize the dispersion error. The surface tension force is calculated by the continuum surface tension force (CSF) formulation of Brackbill et al.. (Brackbill, J. U., et al.., 1992) For the sake of programming simplicity, the proposed incompressible two-phase flow solver will be discretized on the semi-staggered grids without incurring velocity-pressure checkerboard oscillations. To verify the proposed method, four benchmark problems, including Rayleigh-Taylor instability, bubble rising and coaxial/oblique bubble merging problems, are numerically investigated. All predicted results have been shown to compare fairly well with the experimental and/or other numerical results. It also shows that the satisfactions of mass conservations are guaranteed for the proposed two-phase flow solver. Finally, the two-phase flow solver is applied to numerically predict the Kevin-Helmholtz instability in a tilted horizontal channel. The predicted results show the applicability to the stratified two-phase flow analysis in the nuclear engineering field, like the slug flow, droplet impacting or counter-current flow problems

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An advanced core protection system for a pressurized water reactor (PWR), RCOPS, was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS used a safety class programmable logic controller POSAFE-Q PLC that was developed as a controller for the nuclear reactor safety system. It consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis (FMEA) demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for departure from nucleate boiling ratio (DNBR) by approximately 2%.

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With the development of scaling analysis approach, the reduced height test facility, such as APEX and PUMA in US, and ATLAS in Korea, have been successfully used for the NPP safety test. In this paper, the distortion originated from the reduced height effect is analyzed for the long term cooling phase based on the Natural Circulation. First, based on the NC scaling performed at a loop level and a component level, it shows the pressure drop through a specific component is scaled down, which could change the fluid property especially during the long term cooling phase for its low atmospheric pressure. Then, the scaling analysis of long term cooling loop is performed by cutting the loop into the three control volume sections, and it shows the reduced height will change the pressure at the core, which causes the fluid property slightly bias from the similitude condition and is conservative for the test result.

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In this paper, Neutron Activation Technique (NAT) is used for neutron flux mapping in homogenous boron concentrated paraffin wax. For this purpose, activator detectors were prepared by irradiating process in an Americium-Beryllium (^<241>Am-^9Be) neutron field for the required time period to induce approximate saturation activity. After this, the induced activity was measured by time coincidence method using a Sodium iodide (thallium) NaI (T1) crystal detector and organic scintillation "Anthracene" detector for gamma ray and beta particle detection respectively. The activator detectors were activated at different radii from the irradiator neutron source in boron mixed paraffin wax. The flux profile was mapped by measuring induced activity. Errors in measurements and calculations were estimated by Poisson distribution.

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When a crack is detected in a stainless steel piping system of a nuclear power plant during in-service inspections, failure bending moment can be predicted by limit load criterion in accordance with Appendix E-8 in the JSME S NA-1-2008 or Appendix C in the ASME Code Section XI. However, in these current codes, the limit load criterion is only provided for the case of a pipe containing a single crack with uniform depth, although independent multiple cracks, such as stress corrosion cracks, have actually been detected in the same cross section of stainless steel piping systems and their actual geometrical shapes are generally complex. In this paper, a method is proposed to predict the failure bending moment for a pipe containing multiple circumferential cracks with complex shapes and several numerical examples are given to show the effectiveness of this method.

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A key problem in the application of a supercritical carbon dioxide (CO_2) turbine cycle to a fast breeder reactor is the corrosion of structural materials brought about by supercritical CO_2 at high temperatures. In this study, high-temperature oxidation tests on the structural materials were performed in carbon dioxide pressurized at 0.2 and 1 MPa, and in air, and the oxidation behavior were compared. Results of investigating the effect of CO_2 pressure including the previous reports tested at 10MPa and at 20MPa, the effect was hardly observed for all steels. In air environment, weight gain caused by high temperature oxidation was much lower than that in CO_2.

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The present paper describes the CFD analysis on the primary-side of an intermediate heat exchanger (IHX) which has the similar configurations as the IHX for the fast breeder reactor "Monju". The IHX is precisely modeled based on the discussion about meshing system. The present model is used for the heat transfer analysis under low-flowrate and natural circulation conditions. The IHX is a shell-and-tube type and counter-flow heat exchanger which has more than 3000 heat transfer tubes on the secondary side. Therefore, the flow pattern on the primary side gets complex. Measurements of flow pattern and temperature distribution on the primary-side of the real IHX are almost impossible. Since the heat transfer tubes of approximately 5 m in length are fixed at 7 plates with many flow holes and placed on the 23 circles with an appropriate lattice pitch, the number of meshes becomes enormous size. In order to overcome these problems, a separate model is discussed. In the present study, two models are discussed. The first one is a precise full-sector model with one flow entrance, 6 windows on the primary-side. The flow distributions are calculated changing inlet flow rate from 100% to 0.1% which is equivalent to 106 to 103 in the Reynolds numbers. The other model is a sector model with 8 chamber separated by 7 flow-rectifying plats. Pressure losses at each plate and chamber are calculated using this model. As a result of the analysis, since there is only a small flow deviation between the flow from the 6 windows under turbulent flow and laminar flow conditions, the sector model with one window is possible model in the calculation. The small radial velocity gradient is calculated from 23rd layer (outer heat transfer tube) to 10th layer. The distribution is not dependent on the flow rate. Axial flow distributions through the rectifying plates are unified from the entrance to the down-stream. The sector model is applicable to calculate the primary-side flow distributions through the above mentioned discussion.

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In the framework of the investigation on the thermal-fluid dynamics phenomena in helical pipes of the innovative nuclear power reactor steam generator, at the Department of Nuclear Engineering at the University of Palermo, various research activities were performed relating to validations works of the models implemented in RELAP5/MOD3.2b thermal-hydraulic advanced code in order to simulate two phase flow phenomena taking place in these systems, though the one-dimensional nature of it. In this paper it is shown that the results obtained by the analyses of the experiments performed in different international laboratories and related to various helical pipes geometries, allows to prove the good performance of the so modified RELAP5 code.

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The present study relating to corrosion in piping under the high-temperature and high-pressure conditions has been conducted for turbine and feedwater systems of the advanced thermal reactor (ATR) "Fugen", according to the request of the Japan Atomic Energy Agency (JAEA). The influence of the upstream bends on the flow field is important in the corrosion evaluation. Therefore, the 3D analysis has been carried out using the computational fluid dynamics (CFD) product code. The scope of the CFD modeling is limited in single-phase flow piping. The specification of the piping is approximately 14 m in length, and approximately 0.2 m in inner diameter. The pressure of the fluid is 9.5 MPa, and temperature approximately 115 ℃. The Reynolds number (Re) is approximately 6.4×10^6. The number of meshes of calculation scheme is about 2.7 million with hexahedral elements. Important parameters in terms of the corrosion evaluation are distributions of velocity and turbulent kinetic energy. Therefore, those values around the targeted piping are visualized and evaluated. Moreover, the thinning speed can be calculated by using those analytical results based on velocity magnitude. In the present study, an evaluation method using the turbulent energy is not practiced, because this method has not been established yet. The patterns of global flow and flow field such as velocity magnitude and turbulent kinetic energy at several regions near the wall are clearly understood by the present CFD analysis relating the flow passing the multiple bends.

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