The air reactor cavity cooling system (RCCS) is a safetyrelated passive decay heat removal system undergoing further development in support of the next generation of nuclear plants, specifically the very high temperature gas reactor. Recently, the University of Wisconsin (UW) designed and built a 1/4-scale air RCCS. The experimental facility at UW consists of three major components: inlet-plenum, heated riser ducts, and an outlet-plenum with exhaust ducts. Using commercial computational fluid dynamics (CFD) software, a 3-D simulation of the inlet-plenum was performed using as-built geometry and experimental data. The purpose of this study was to improve upon preliminary CFD simulations of the inlet-plenum, which initially were conducted to evaluate the fluid structure characteristics within the plenum, to assist with uncertainty evaluations of velocity transducers in the scaled experiment, and to produce CFD results for a follow-on CFD study. An informal verification and validation (V&V) methodology was followed, which included mesh refinement and iterative convergence studies. Three inlet-plenum simulations were performed as part of the mesh refinement study; the mesh sizes of the simulations ranged from 2.3 million to 9.7 million elements. Computational difficulties related to residual convergence were experienced.

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Due to its strong inherent safety feature, high temperature gas-cooled reactor (HTGR) has been seen as one of the best candidates for the next generation of nuclear plants. The safe, stable and efficient operation is necessary to the development of HTGR. Power-level control technique that strengthens the closed-loop stability and the transient responses by properly generating the insertion or withdrawal speed of control rods is beneficial to improve HTGR operational performance. In this paper, based on the physically-based control method, it is proved theoretically that proportional-differential (PD) output feedback laws can guarantee the globally asymptotic stability (GAS) for the closed-loop system if a sufficient condition is well satisfied. Numerical simulation results not only verify the theoretical result but also show the influence of PD gains to control performance. This work gives a thorough theoretic explanation to why the classical PD output feedback laws are effective for HTGR power-level regulation practically, and also shows the feasibility of applying simple PD control for load following function.

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The performance of the axial-flow helium compressor in the Power Conversion Unit of a High Temperature Gas-cooled Reactor coupled with helium turbine system is the determining factor to influence the efficiency of the generation of electricity, which indicates the high importance of the estimation of the compressor's performance. Objectively, to reduce the costs and complexities of the experimental work during the design phase and the manufacture phase of the axial-flow helium compressor, the helium should be replaced by another kind of gas which would be easy to fulfill the tasks and be acquainted with by engineers. And also in order to subjectively understand the inherent mechanism of axial-flow compressors, to present a widely applicable formula and a similitude theory revealing the relationship among the mainstream kinds of gas should be necessary. Air is chosen for the maturity of air compressors. Therefore, the studies of similitude theory between helium and air for the axial-flow compressor is significant and it should be the key to a deep understanding of flows in axial-flow compressors. At the present stage, most of the similitude schemes for axial-flow helium compressors are obtained by dimensional analysis, which are convenient but crude. With the dimensional analysis models, fundamental relationships of dimensions are required to be followed according to engineering experience. And a few studies on the similitude schemes obtained by differential equations have only simple and abridged models. In these classic differential equations models, control volume is the only study object and incidental conditions would not be considered about, of which the studies are unspecific and fuzzy. In the study, a more thoughtful model of similitude laws obtained by differential equations for axial-flow helium compressor is established, trying to display a more realistic explanation. The most important advantage of the new model is to consider the force and energy that flow takes from cascades, which is coupled with rotational speed and can be adjusted to the theoretical similar flow in the compressor. It is more logical and flexible to the physical conditions, and the comparison of the similitude laws obtained by different methods offers a better understanding of the similitude theory where the distinction of the two method is pointed out.

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The steam generator tube rupture (SGTR) accident is considered by the International Regulatory Bodies a design basis event in HLM technology because of its safety concerns in terms of impulsive dynamic load caused by the thermal interaction between the lead/lead-bismuth-water. Aim of this study is to investigate, through numerical analyses, the dynamic response induced in LIFUS5/Mod2 facility during a SGTR. In particular it is intended to furnish a post-test evaluation of the structural effects caused by the interaction between water and lead-bismuth (during a SGTR accident event) with reference to the experimental tests carried out at the ENEA Brasimone Research Centre. The interacting fluids are Lead Bismuth Eutectic (LBE) at 0.1 MPa and 400 ℃ and water at 4 MPa and 200 ℃. After a briefly description of Test A1.3 and A1.4, executed at the ENEA Brasimone Research Centre in the LIFUS 5/Mod 2 facility, the result of the numerical FEM analyses are presented and discussed. The propagation of the pressure wave was simulated by 3-dimensional Finite Element Method (FEM) models implemented in suitable FEM structural codes. The obtained results, validated by a comparison with the experimental data results, indicate that no significant dangerous effects on the LIFUS test vessel occur - as stress levels are below the allowable limit value - with the assumed geometry and in the testing conditions. The present work may also provide an useful contribution to the development of lead reactor technology by means of the test programme to be carried out in the LIFUS experimental facility (work performed in the framework of the 7^<th> FP THINS).

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The reliability of operation of a nuclear power plant substantially depends on the reliability of its individual components. Electric cables, particularly medium-voltage cables represent an important constituent of the nuclear island as well as of the balance of plant. It can't be considered as satisfactory that the redundancy of a particular subsystem is ensured. For this purpose it is important to check the cable insulation condition, which is supposed to be carried out during the overhaul activities. The lifetime of medium-voltage cables is estimated to attain somewhere around 40 years. However, due to electrical, thermal and mechanical stresses it may be significantly reduced. In order to avoid its likely failures and following outages the insulation must be tested regularly. Based on obtained testing results the condition of a cable or cable termination can be assessed and repair or replacement made, if necessary. The article presents some of most important non-destructive diagnostic measurements on cable connections recommended to be done during the overhaul of a nuclear power plant following a good practice of asset maintenance management.

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Modular high-temperature gas cooled reactor (HTGR) is known as one of the most advanced nuclear reactors because of its inherent safety and high efficiency. The power conversion system of HTGR can be steam turbine based on Rankine cycle or gas turbine based on Brayton cycle respectively. The steam turbine system is mature and the gas turbine system has high efficiency but under development. The Brayton-Rankine combined cycle is an effective way to further promote the efficiency. This paper investigated the performance of combined cycle from the viewpoint of thermodynamics. The effect of non-dimensional parameters on combined cycle's efficiency, such as temperature ratio, compression ratio, efficiency of compressor, efficiency of turbine, was analyzed. Furthermore, the optimal parameters to achieve highest efficiency was also given by this analysis under engineering constraints. The conclusions could be helpful to the design and development of combined cycle of HTGR.

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Pool boiling is one of the most efficient modes of heat transfer. A passive cooling system needs to be able to remove sufficient energy following a design basis accident via steam condensation and pool boiling. This study employed a system of transient single- and twophase flow and heat transfer models to simulate the heatup process of tube bundle pool boiling in a passive cooling system. Three-dimensional simulation was performed to show the effects on temperature distribution from geometry symmetry and asymmetry in the threetube bundle system. To simulate boiling in a tube bundle water pool, the phenomena on the micro-scale have been modelled by appropriate closure relations. In this simulation, a constant heat flux is used to represent the heat release from the steam condensation inside the tubes at this stage, and an in-tube condensation model to couple with the out-tube bundle pool boiling is under development. The heat-up transient was successfully simulated with the temperature profiles of different monitor points vs. time and temperature contours of different cross sections at different time. The experimental work is undergoing, and a validation and verification of this numerical investigation will be conducted once the experimental work is completed.

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During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increasing of fuel burnup that decrease the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect fuel materials properties, power distribution and overall performance of the fuel pin. In order to investigate these important issues, we study the (U_<1-y>Pu_y)O_<2-x> fuel pellet by fully coupling the thermal transport, thermal expansion, oxygen diffusion, fission gas release and swelling and plutonium redistribution to evaluate the effects on each other with a burnup dependent thermal conductivity model, which accounts for the evolution of fuel porosity. We find the radial temperature and deformation profiles are most sensitive to O/M ratio at small values of plutonium content. For larger values of plutonium content, the effects of O/M ratio and plutonium content are equally important, which is the same results to the previous modeling work even though based on different material properties' models. Additionally, the porosity is found to be very sensitive to fuel temperature at the beginning of fuel burnup, while the fuel burnup profile dominates the degradation fuel thermal conductivity during the whole fuel irradiation.

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This paper presents the development of a light water reactor fuel performance code, which considers almost all the related physical models, including heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the equations are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. Comparisons are made for the simulation results between COMSOL and another simulation tool of BISON. The comparisons show the capability of our simulation tool to predict light water UO_2 fuel performances. In our modeling and simulation work, the performance of enhanced thermal conductivity UO_2-BeO fuel and newlyadopted corrosion resistant SiC cladding material was also studied. UO_2-BeO high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor would be improved. However, for SiC cladding, although due to its high thermal expansion, the gap closure time is delayed, irradiation induced point defects and defect-clusters in the SiC crystal will dramatically decrease SiC thermal conductivity, and cause significant increase in the fuel temperature.

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Safety injection was activated when SBLOCA happened in the steampipeline, in order to inject enough cooling water to cool and submerge the core. Temperature fluctuations in the piping system can lead to thermal fatigue. Based on the Emergency Core Cooling (ECC) of a 2nd generation plus of PWR, one CFD software was used to simulate the thermal mixing phenomenon of the sub-cooled water from safety injection nozzle and high temperature water or bubbly flow from the cold leg. For simulations of single-phase thermal mixing, the temperature profile and velocity profile were obtained which matched well with the experiment. Four different jet patterns of thermal mixing are obtained. For bubbly flow (air-water mixture), the gas volume profile, velocity profile and gas particle mean diameter were acquired, and the effect of different injection nozzle velocity and inlet gas volume fraction were also studied.

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This paper studies a new method of design verification through the VR plant, in order to perform verification & validation the design of plant conform to the requirements of accident emergency. The VR dynamic plant is established by 3D design model and digital maps that composed of GIS system and indoor maps, and driven by the analyze data of design analyzer. The VR plant could present the operation conditions and accident conditions of power plant. This paper simulates the execution of accident procedures, the development of accidents, the evacuation planning of people and so on, based on VR dynamic plant, and ensure that the plant design will not cause bad effect. Besides design verification, simulated result also can be used for optimization of the accident emergency plan, the training of accident plan and emergency accident treatment.

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In this paper, the boron transport and critical behavior of the fuel debris was studied by the three-dimensional coupled nutronic and thermal-hydraulic system dynamics code since the boron transport model is originally built in it. In this model, the effects of diffusion and boron sinking or sedimentation due to the larger specific gravity of boron are omitted, so that only the advection of boron in the liquid phase is calculated. The estimation of the boron transport conducted by the system dynamics code was compared with the results obtained from computational fluid dynamics (CFD) code, STAR-CD^[○!R] v4.20. It was found that the effect of boron sedimentation is dominant, concerning that of the diffusion, which plays only a small or negligible role. The sedimentation of boron solution, however, is not calculated in the system dynamics code. In order to consider the sedimentation in that code, the boron transport model was modified in this study. By this modification, the boron transport behavior can be evaluated by system dynamics code with a precision equivalent to that of CFD. Then, the new boron transport model was applied to the recriticality analysis of the fuel debris located on the pedestal sump, which shows that the injection of boron suppresses the criticality of the debris soon after boron reaches there. The results of this study conclude that the method of injecting boric acid into RPV has an effect on controlling the criticality of the debris.

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Nanofluids application in nuclear systems have been intensively studied for the past few years since it has been found out that this type of fluids exhibit a substantially higher critical heat flux (CHF) compared to "clean" water. Most studies done in these field usually focus on heat transfer criteria of the nanofluid itself and its industrial application including nuclear systems. In vertical upward heating flow, the turbulent heat transfer can be deteriorated if it goes from forced to mixed convection regime or also known as deteriorated turbulent heat transfer (DTHT) regime. Hence, since nanofluids application is being considered in nuclear systems, it is quite interesting to study the mixed convection phenomenon in the nanofluids flow. To this end the current paper focuses on computational fluid dynamic (CFD) analysis of mixed convection flow of the nanofluids. The nanofluids selected for the study are the ones considered for applications related to nuclear systems. Three turbulent models; namely the low-Reynolds k-ε model, the v^2-f model, and the recently developed Elliptical Blending k-ε model are first examined to find out the best model to be used in nanofluids study. The impact of particles concentration in nanofluids on the mixed convection phenomenon is also investigated. This numerical work can give insights and a better understanding of the mixed convection phenomenon in the case of nanofluids-based systems and set a basis for future experimental verification.

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In the last two decades, CFD codes have been widely used for the design studies of NPPs. Recently high-precision simulation models have been developed to evaluate complicated phenomena, e.g. fuel melting in severe accident. The authors also are developing a high-precision CFD code with an interface tracking method to simulate the gas entrainment (GE) phenomena in sodium-cooled fast reactors (SFRs), which might be caused by a highly-intensified free surface vortex. The GE in SFRs is characterized by an elongated interfacial dent along the vortex core and the bubble pinch-off at the tip of the dent. To simulate this complicated phenomenon, the authors' simulation code has physics-basis algorithms which model accurately the interfacial dynamic behavior, the pressure jump condition at an interface and the surface tension. Several verification problems, e.g. the slotted-disk problem, have been already solved and the accuracy of each individual algorithm is confirmed. In this paper, a basic experiment of the GE is simulated to validate the developed code. In the experiment, the entrained gas flow rate is measured by image processing with a high-speed video camera. The simulation result of the entrained flow rate shows comparable value to the experimental data, that is, our simulation code is considered applicable to the evaluation of the GE in SFRs.

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Nuclear Human Resource Development Center (NuHRDeC) of JAEA has conducted nuclear human resource development for more than 50 years since its establishment in 1958. NuHRDeC conducts international nuclear human resource development, so called "Instructor Training Program (ITP)", which is a training scheme launched in 1996 in order to support Asian countries seeking peaceful use of nuclear energy. The ITP consists of 1) Instructor Training Course (ITC) in Japan, 2) Follow-up Training Course (FTC) in own countries organized by instructors trained at ITC in Japan, and 3) Nuclear Technology Seminar for bringing up nuclear trainers and leaders in Asian countries. The purpose of ITP is to develop a self-sustainable training system in Asian countries, which disseminates the knowledge and technology in their countries. After completing ITC trainings at NuHRDeC, the trainees are obliged to set up FTC in each country. They create own 1 or 2 weeks course curricula and allocate local lecturers including themselves. Two or three Japanese experts join the FTC to give technical advices and support to local lecturers. The present specialized fields of ITC are 1) Reactor engineering such as reactor physics, thermal engineering and reactor safety, 2) Environmental radioactivity monitoring, and 3) Nuclear emergency preparedness. The main feature of ITC is that the curricula places emphasis on the practical exercise with well-equipped training facilities, experimental laboratories utilizing the simulators of research reactor, and the expertise of lecturers mostly from JAEA. As of FY2014, ITC is applied to 8 countries; Indonesia, Thailand, Vietnam, Bangladesh, Kazakhstan, Malaysia, Philippines and Mongolia. The total number of participants at ITC since 1996 is approximately 300 and the participation of FTC has been increased significantly year after year with more than 3,000 in total. This result indicates that the ITP system has been effectively contributed to fostering local trainers in Asian counties. Present paper summarizes the outlines, experiences and future prospects of ITP.

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This paper presents a method of constructing nuclear power plant in-service maintenance virtual simulation scene and virtual maintenance process. Taking air baffles dismantling process of CAP1400(China Advanced Passive 1400) nuclear power plant as an instance, this paper discusses ergonomics, space analysis, time assessment based on virtual reality in the process of in-service maintenance. It demonstrates the advantage of using VR technology to design and verify in-service maintenance process of nuclear power plant compared to the conventional way.

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Probabilistic Risk Assessment (PRA) is an important tool for evaluating risk in nuclear power plants. Dynamic PRA (DPRA) is an extension of traditional PRA methods that account for dynamic and phenomenological effects associated with time-dependent complex dynamic systems. Here we investigated dynamic event trees and optimization per focus on identifying the highest probability of system failure. Using a Branch-and-Bound (B&B) algorithm that relies on and develops bounding functions to prune or delete branches that will not yield the optimal solution (i.e., clad failure). The approach also used novel LENDIT (length, energy, number, distribution, information, time) metrics and S2R2 (state, systems, resources, response) sets to support an expert-based approach that is linked to constraints per use of the B&B algorithm. Results to date indicate that this approach is effective in reducing simulation time and thus mitigating the state explosion of thermal-hydraulic states. The work demonstrated the ability to evaluate uncertainty such that a risk-informed, quantitative PIRT (phenomena identification ranking table) is generated. Quantitative PIRT (QPIRT) can be used to improve models and identify validation needs with respect to risk. Two case studies using reference PWR and BWR plant configurations under SBO were evaluated. The implementation of the B&B algorithm yielded a significant reduction (>50%) in simulation costs. QPIRT ranking for the PWR showed that lower fidelity models combined with system redundancy produces adequate results with respect to risk. For BWR SBO, the modeling uncertainty does not present a challenge with respect to risk. Recovery of the SBO requires either restoration of the AC power or activation of firewater injection combined with operator action to depressurize the system through the automatic depressurization system. The timing of both automatic and manual safety system actuation is indeed critical to the outcome reactor state. Reactor power and firewater injection capacity provided the highest degree of correlation to model success.

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Existing literature on problems of heat transfer to fluids at supercritical pressures distinguishes three heat-transfer regimes: 1) Deteriorated, 2) normal, and 3) enhanced. Conventional approach to correlate forced-convective heat transfer to supercritical fluids is applied to the normal and enhanced heat-transfer regimes. Although, there were recent attempts to correlate data for the deteriorated heat-transfer regime at mixed convection, currently there are no heat transfer correlations for DHT occurring at pure forced convection at high heat fluxes. Thus, an innovative approach was developed to correlate data without distinguishing heat transfer regimes. The approach is discussed in this paper. Using this approach, the heat transfer data to supercritical CO2 in forced convection regime, which were obtained at MR-1 Loop at Chalk River Laboratories, were correlated with an RMS of approximately 10% (corresponds to approximately 20% spread based on the 2σ-level). This result is twice less than that previously obtained at the University of Ontario Institute of Technology. The paper covers the following major topics: thermal properties of supercritical fluids, specifics of heat transfer to fluids at supercritical state, overview of existing types of correlations, conventional methodology for the development of heat transfer correlations, and innovative methodology for the development of heat transfer correlations.

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Intergranular oxidation (corrosion) occurred within crevice of austenitic low-carbon stainless steel (solution treated, almost no applied stress) after immersion in hightemperature water (288°C, 8.5MPa, dissolved oxygen conc. 32ppm, electrical conductivity: 1.2±0.2 μS・cm^<-1> (measured value at 25℃)) for 500h. The intergranular oxidation occurred at specific position within the crevice that is relatively distant from the crevice mouth with relatively low crevice gap. Both the grain boundary and grain matrix were oxidized. In the oxidized area, Fe and Ni were depleted and Cr was enriched compared to the matrix. Maximum penetration depth of the oxidation was approximately 50μm after 500h. In order to understand potential-pH condition within the crevice, surface oxide layer was microscopically and thermodynamically investigated. Thermodynamic properties of the surface oxides near the intergranular oxidized area indicated lowered pH of approximately 3.2 to 3.4. In-situ measurement of local solution electrical conductivity was carried out using small electrodes (dia. 800μm) imbedded into the crevice former plate. The solution pH was estimated using theoretically calculated pH vs. electrical conductivity relationship. In the area where the intergranular oxidation occurred, the solution electrical conductivity was nearly 100 times higher than that of bulk water and which indicated lowered pH of approximately 3.5. The above results suggested that, in the high temperature and relatively high purity water, acidification occurs within crevice of stainless steels and such aggressive corrosion condition result in the intergranular oxidation.

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The "Nuclear Safety Planning" has been published in Oct. 2012 in China, which stipulates the safety goals for the NPPs which will be built in the future. As for the nuclear power plants (NPPs) which will be built in China's Thirteenth Five-Year (2016-2020) and later, the high level safety goal is described as that "the possibility of the large radioactive release should be practically eliminated by design". A thorough investigation has been performed at SNERDI to explore the technical insight of this high level safety goal by using Multinational Design Evaluation Project (MDEP) hierarchical safety goal approach. The definition of large release is proposed accordingly, DID requirements and probabilistic requirements are derived from this high level safety goal.

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Hitachi-GE Nuclear Energy Ltd. (Hitachi-GE) has developed a conceptual design for the Double MS: modular simplified & medium small reactor (DMS) under the sponsorship of The Japan Atomic Power Company. The DMS is one of the small-to-medium sized reactors (SMRs) of boiling water reactor (BWR) type, which has been predicted to overcome almost all economy of scale concerns when compared to proven conventional advanced BWR (ABWR) technologies. The DMS design was dedicated to generate electric power, and in order to enhance flexibility for usage of the DMS, the University of Saskatchewan, Hitachi-GE and Hitachi Ltd. (Hitachi) have collaborated on a joint research and development (R&D) initiative to study the utilization of heat and steam from the balance of plant (BOP) for thermal utilization (TU) applications such as district heating, process heating, etc. Based on the DMS of the 300MWe class, Hitachi is developing heat balance evaluation tools for the BOP system and the intermediate heat exchanger (IHX) system which is implemented for heat transfer from the BOP to the TU application in order to detect radioactive materials from the BOP and to prevent leakage to the TU applications. In this paper, a part of the IHX system heat balance simulator for the co-generation DMS is described.

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The corrosion rate of stainless steel in boiling-nitiric-acid solutions containing oxidizing metallic ions (vanadium and ruthenium) and seawataer components were investigated by the immersion tests in order to evaluate the influence of the seawater components on the corrosion of steel in the nitric acid solution containing oxidizing metallic ions. The corrosion rates of 310Nb in the solution containing seawater components were lower than that in the solution without seawater components. These results described that contamination of seawater components in the spent fuel reprocessing fluid seems not to accelerate the corrosion of steels, which are material for the devices treating the nitric-acid reprocessing solutions.

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Nickel (Ni)-base refractory alloys such as Alloy 625 are used for structural material of reducing roasting furnace in spent nuclear fuel reprocessing plant because they have good high temperature strength and corrosion resistance in nitric acid. The operating temperature of the furnace is around 1073 K. From past studies about thermal aging effect on microstructure of Ni-base alloys, it is reported that intermetallic compounds and carbides precipitate during thermal aging. For example, δ-phase (Ni_3Nb) precipitates after 30 hr at 1073K and δ-phase transforms to γ"-phase after 100〜300hr. It is considered that precipitation and phase transformation by thermal aging cause precipitation hardening and loss of creep ductility. Therefore, it is important to evaluate thermal aging effect on creep properties of Ni-base alloys in the furnace at the operating temperature to predict its lifetime but there is few report on the effect of thermal aging for creep properties. In this study, to evaluate the effect of thermal aging on creep properties of Alloy 625, we carried out creep tests on aged and solution-treated Alloy 625 at 1073 K. According to the creep test results, time-to-rupture decreased by thermal aging when test stress was more than 100 MPa, but did not change when test stress was less than 100 MPa for any specimens. In the solution-treated alloy, creep deformation behaviors changed over 100 MPa. These results show that time-to-rupture was constant because intermetallic compounds precipitated when the test stress was less than 100 MPa in solution-treated alloy. The observed relationship between creep strain rate and test time showed that the precipitation started after 100 hr for solution treated alloys. These results suggest that intermetallic compounds precipitate immediately after furnace operation. And it is appropriate to use creep data of thermal-aged Alloy 625 for the reducing roasting furnace lifetime prediction.

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The AP1000 nuclear steam supply system (NSSS) is designed to accept full load rejection without initiating a reactor trip using the reactor power control system, pressurizer pressure and water level control systems, feedwater control system, steam dump control system and rapid power reduction (RPR) system. This paper presents the dynamic simulation and analysis of load rejection operations for the AP1000 NSSS. A real-time simulation platform is developed in Matlab/Simulink with implementation of a nodal core model, a three region nonequilibrium pressurizer model, a nonlinear steam generator model, a feedwater system model, a main steam system model, and the relevant control systems. Based on the simulation platform, two typical load rejection transients from full power (FP) are simulated, including the 50%FP load rejection without RPR and the large load rejection to 25%FP with RPR. The simulation results show that the NSSS control systems can successfully respond to load rejection transients without reactor trip or operation of the pressurizer or steam generator safety valves. Sustained stable operation can be obtained during the load rejection transients due to the proper control system responses.

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Corrosion rate of carbon steel in neutral water containing chloride ion is mainly controlled by dissolved oxygen. Hydrogen peroxide is generated under gamma ray irradiation and it will also affect corrosion. In this study, the effect of oxygen and hydrogen peroxide on corrosion rate of carbon steel in diluted artificial seawater was investigated by electrochemical methods. Diffusion coefficient and thickness of diffusion layer for oxygen and hydrogen peroxide were measured to estimate the diffusion limiting current density. Corrosion tests of carbon steel were also conducted in diluted artificial seawater containing oxygen and/or hydrogen peroxide at 323K, and the results were compared to the estimated diffusion limiting current density. The diffusion coefficient of hydrogen peroxide was about 0.8 times lower than that of oxygen, and the thickness of diffusion layer was almost equivalent between oxygen and hydrogen peroxide. Diffusion limiting current density of hydrogen peroxide was estimated to be 0.4 times lower than that of oxygen in the same concentration at 323K. Plot of corrosion rate with the sum of concentration of oxygen and 0.4 times concentration of hydrogen peroxide showed good correlation.

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This report addresses a crack-growth driven by mixed modes and fracture failure assessment of mechanical components in nuclear power facilities. Analytic solutions to stress fields and the energy release rate of the crack are reviewed and examined for cases when the crack propagates straight-ahead (Mode I cracking) and in a kinked direction (Mixed mode cracking). It is illustrated that for cases when complex loading conditions or dissimilar materials are under consideration, the energy release rate for a straight-ahead growth can be much less than that for a kinked growth. Hence, for such cases the crack growth is inevitably driven by mixed modes and the mode mixity must be taken into account in the fracture assessment. Otherwise, the fracture failure cannot be evaluated correctly. Numerical studies are conducted and it is shown that it is not conservative to use a purely mode-I based criterion for the evaluation of the fracture failure assessment for typical problems of mixed mode driven cracking. Suggestions for accounting the effect of mixed mode cracking in connection to the application of the R6-approch for safety assessment are given.

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The CABRI E7 experiment was performed to study fuel pin failure and post-failure phenomena for an industrial pin with annular fuel pellet design and pin rupture with subsequent fuel ejection into liquid sodium. E7 used a structured Transient OverPower (TOP) starting at nominal cooling conditions leading to mechanical pin failure due to cavity pressure build-up at still constrained clad conditions. In this work we have studied the failure mechanisms and failure conditions serving as boundary conditions for the post-failure material-relocation behavior. We have also evaluated SAS-SFR code's capabilities to simulate post-failure phenomena in view of experimental results. We present comparisons of measured results with SAS-SFR results for the power operation irradiation of the fuel pin as well as for the E7 failure and post-failure behavior.

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Within the framework of the European JASMIN project, the ASTEC-Na code is being developed for safety analysis of severe accidents in SFR. In the first phase of validation of the ASTEC-Na fuel thermo-mechanical models three in-pile tests conducted in the CABRI experimental reactor have been selected to be analysed. We present here the preliminary results of the simulation of two Transient Over Power tests and one power ramp test (AGS0, LT2 and E9, respectively) where no pin failure occurred during the transient. We present the comparison of ASTEC-Na results against experimental data and other safety code results for the initial steady state conditions prior to the transient onset as well as for the fuel pin behaviour during the transients.

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This paper presents the phenomena of irradiation assisted stress corrosion cracking (IASCC) and it describes a facility dedicated to face these issues and planned to be implemented in the Jules Horowitz material testing reactor (MTR). The structural and functional integrity of the internal components of the core of light water reactor (LWR) are major concerns. Several degradations of core components have occurred in boiling water reactors (BWRs) first and then in pressurized water reactors (PWRs). IASCC was identified as the cause of these degradations. It is today an important issue for internals materials where stainless steels and nickel base alloys have shown some susceptibility. For future reactors where the flux of high energy neutrons will be stronger or where a higher burn up is aimed, IASCC would be a major concern. The first aim of the paper is to summarize the industrial issues of IASCC for PWRs and BWRs. In these LWRs, stainless steels and nickel base alloys have exhibited IASCC. Mechanisms of IASCC are then discussed together with the effects of irradiation on metallic alloys and on water chemistry (radiolysis). In the final part, a water loop dedicated to IASCC studies is proposed to be implemented in the new MTR called Jules Horowitz reactor (JHR) with the objective of improving the understanding of IASCC and the resistance of stainless steels and nickel base alloys to this phenomena. Main characteristics of JHR and of the corrosion loop, called CLOE, are then described (190 bar max, 360°C max, PWR or BWR water chemistry conditions…). In the core, the samples will be subjected to high fast neutron flux while in the beryllium reflector, the samples are exposed to high thermal neutron flux. The use of a water loop is also planned to reproduce phenomena linked to IASCC under BWR and PWR simulated conditions. The CLOE loop is aimed to perform integral experiments with applied stresses on samples under neutron flux so that IASCC studies could be performed.

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The application of the laser welding to the construction of the building frame of a nuclear plant is proposed. High dimension accuracy is required in the nuclear plant structures, but it is much difficult to satisfy the high accuracy due to welding deformation. So, it is necessary to optimize the welding process to reduce the welding deformation. Prediction of welding deformation by trial production increases the material and labor costs. So, it is necessary to predict the welding deformation and optimize the weld process to satisfy the dimension accuracy by analysis. In this study, the heat source model has been developed to predict the laser welding deformation with high accuracy. The heat source model has been developed by comparing the experimental and analysis results. In the experiment, the plate of austenitic stainless steel (Type316LN) was butt-welded using laser. The angular deformation, the transverse shrinkage, the temperature and the penetration shape of the cross section of the laser welded plate were measured. The penetration shape like a wine cup was not able to be reproduced by using only Conical heat source, but by combining the Conical and Goldak heat sources(1,2) and by optimizing the heat input efficiency. The developed analysis technique has predicted laser welding deformation with high accuracy.

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Approximately two months after the Fukushima nuclear accident, the Japan Atomic Energy Agency (JAEA) led off a series of demonstration tests to develop effective but easily applicable decontamination methods for various school facilities in Fukushima. This effort included (1) dose reduction measures in schoolyards, (2) purification of swimming pool water, and (3) removal of surface contamination from playground equipment. Through these demonstration tests, they established practical methods suitable for each situation: (1) In schoolyards, dose rates were drastically reduced by removing topsoil, which was then placed in 1-m-deep trenches at a corner of the schoolyard. (2) For the purification of pool water, the flocculation coagulation treatment was found to be effective for collecting radiocesium dissolved in the water. (3) Demonstration tests for playground equipment, such as horizontal bars and a sandbox wood frame indicated that the decontamination effectiveness considerably varied depending on the material, paint or coating condition of each equipment piece. These findings were summarized in reports, some of which were compiled in local/national guidelines or handbooks for decontaminating the living environment in Fukushima.

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Clustering has been long known to play an important role in the structure of light nuclei. We investigate the fragmentation mechanism reflecting the cluster structure of the ^<17>C. The breakup of the neutron-rich nuclei ^<17>C into ^8Li + ^9Li has been studied using a 44MeV/nucleon ^<17>C beam inelastically excited by ^<12>Ctarget. The clustering structure of is reflected in its fragmentation as the dynamical cluster breakup into ^8Li and ^9Li , depending on the incident energy. Excited states in ^<17>C were observed at 29.3±0.1 and 34.4±0.1MeV. This result would indicate that ^<17>C have a well-established cluster structure in the ground state.

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For water-cooled ceramic breeder blankets in fusion reactors, neutron multiplier materials such as beryllium are packed into the blankets to obtain tritium breeding capability. Pipe rupture inside the blanket leading to ingress of coolant is an important event to be assessed because the reaction of water or vapor with beryllium is exothermic and produces hydrogen. In order to clarify the safety characteristics related to Ingress of Coolant Event (ICE) in the blanket, the numerical safety analyses have been performed using the modified TRAC-PF1 code based on the two-fluid model. The TRAC-PF1 was modified in the present study to treat a chemical reaction and a heat transfer models inside the packed pebble bed. A Test Blanket Module (TBM) and a cooling system developed for testing in ITER were modeled for the calculation. The coolant conditions were 15.5 MPa of pressure and about 300℃ of temperature. A scenario of the ICE and the loss of flow due to pump trip was analyzed, and the loss of flow scenario without water ingress was also analyzed for comparison. The results showed that the void fraction, the pressure and the liquid temperature in the cooling pipe increased caused by the vaporization of coolant during the ICE. It is also found that the temperature of packed beds rapidly decreased due to the heat removal by contact with the water or vapor. As a result, it is concluded that the modified TRAC-PF1 code is useful as a safety analysis for the ceramic breeder blanket.

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Experiments have been conducted using the technical scale THAI test facility to investigate the wet resuspension of aerosols from a boiling sump. One test with a water solution of potassium iodide and cesium chloride salts and two tests with different suspensions of calcium carbonate in water (primary mass-median-diameters of 0.065 μm and 0.9 μm) have been performed. Resuspension was derived by injecting steam at the bottom of the sump. The investigated superficial velocities, ranging from 0.025 m/s to 0.13 m/s, extend previous resuspension of soluble material experiments in the THAI facility to turbulent flow regime. The aerosol released from the boiling sump into the gas space has been measured independently by using an SMPS particle counter and gas scrubbers. The droplet entrainment has been quantified, based on the concentrations of salts resuspended into the gas atmospheric flow. The entrainment derived from the gas scrubber measurements is well in accordance to previous experimental findings obtained in the THAI facility and other test facilities. A significant enrichment of insoluble aerosol concentration inside the droplets has been found. The SMPS results revealed that very small aerosols are dominantly released from a boiling sump which might remain airborne for long times.

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A new systematic framework of "risk monitor system" has been recently proposed by Harbin Engineering University in order to enlarge the evaluation scope of "risk" of NPP throughout the whole process of design, operation and maintenance for normal operation of NPP and applicable for various accident situations from "prior to" to "after" core melt. The proposed risk monitor system concept is constituted by two layers systems of Plant Defense-in-Depth (DiD) risk monitor and reliability monitors: The DiD risk monitor is to predict and evaluate plausible risk state from the perspective of whole plant system, while several Reliability Monitors are to evaluate the reliability of individual subsystems to fulfill their expected functions successfully under the prescribed conditions given by the DiD risk monitor. The objective of present study is to develop the reliability monitor of AP1000 passive safety systems for dynamic reliability assessment. And how many types of failure modes and factors leading to disturbance in performance of passive safety systems should be considered in the reliability analysis. As an example, a single loop model of passive core cooling system (PXS) of AP1000 is considered for analysis. The transient behavior of PXS is summarized under the large break LOCA accident and reliability analysis is conducted by GO-FLOW. Calculated results indicated that dynamic reliability of passive safety systems can be conducted by GO-FLOW method and there are main two types of potential failure modes to be considered for reliability evaluation of passive safety systems. And AP1000 would depend on the full reliability of ADS system that it would work surely when needed.

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A theoretical model on heat transfer characteristics of axisymmetric impinging jets on flat solid surfaces is developed for stagnation region using energy integral method. The laminar boundary layer is divided into two regions of flow, Stagnation region and wall jet region. The present predictions were compared well with available experimental test data involving wide range coolant type, Reynolds number and nozzle to plate distance.

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Fukushima Daiichi nuclear power plant of Tokyo Electric Power Company is facing contaminated water issues in the aftermath of the Great East Japan Earthquake on March 11, 2011. The amount of contaminated water is continuously increasing due to groundwater leakage into the underground part of reactor and turbine buildings. Therefore, it is important to understand the groundwater flow conditions at the site and to predict the impact of countermeasures taken for isolating groundwater from the source of the contamination, i.e. the reactor buildings. Installations, such as of land-side and sea-side impermeable walls have been planned as countermeasures. In this study, groundwater flow modeling has been performed to estimate the response of groundwater flow conditions to the countermeasures. From the modeling, groundwater recharge and discharge areas, major groundwater flow direction, inflow rate into underground part of the buildings, and changes in response to implementation of the countermeasures could be reasonably estimated. The results indicate that the countermeasures will decrease the volume of inflow into the underground part of the buildings. This means that the countermeasures will be effective in reducing the discharge volume of contaminated groundwater to ocean.

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Plastic deformation of materials is localized in surface layers of the deformed item as separate spots of different intensity. Initial stages of deformation are characterized by single slip lines that transit to a dense set of parallel lines (slip packs). As deformation grows, shear bands typical of the volume microstructure are formed on the surface along with surface deformation waves. On the whole this is manifested as the material surface deformation activities. These lead to changes in the surface properties: mechanical, optical and electric. The latter can be studied by using structure-sensitive contact techniques, i.e. electric resistance measurements, thermal electromotive force and contact potential difference (Surin V., Evstjuhin N., Grisha S., 1996). The first two techniques are used to study structural failures in the volume and the contact potential difference depends primarily on conditions of the material surface. The surface deformation activity is linked with the yielding process in near-surface layers; even if the applied stress is significantly lower the yield stress. This is because process-induced stress points are on the surface (Shtremel M., 1997). Different models and assumptions are used to study into the surface deformation activity.

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The US-Advanced Boiling Water Reactor (ABWR), certified by the USNRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3&4) Combined License Application (COLA) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3&4 project is finishing the USNRC technical review of the COLA and the final safety evaluation report (FSER) is scheduled to be issued by the USNRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-ABWR has been further reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for USNRC Order EA-12-049 "Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE)". By virtue of the design approach, the US-ABWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-ABWR compared to most of the plants currently operating in the U.S. In addition, STP3&4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (FLEX) initiative. This paper summarizes the progress of the US-ABWR licensing and describes the technology and features of the US-ABWR design that contribute to safety post-Fukushima.

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The objective of this paper is to assess the pressure capacity of the containment in the ultimate limit state at which the structural integrity is retained. For estimating the pressure capacity of the containment structure by deterministic structural simulation with the aid of computer simulation, the static nonlinear 3D finite element analysis will be needed for predicting the global response. In the case of un-bonded tendons, the possibility of slip between tendon and tendon duct should be enabled by some suitable modelling arrangement like with the use of slot connectors modelling the contact between tendon and grout or grease in the tendon duct. The simplest method to assess the ultimate pressure capacity of the post-tensioned concrete containment vessel (PCCV) is to use analytical formulas. This approach results in pressure-displacement, which gives the mid-height radial displacement in the cylindrical part of the containment plotted against the internal pressure. This curve exhibits five significant corner points, namely: 1) pressure to overcome pre-stress, 2) concrete cracking, 3) liner yield, 4) rebar yield, 5) tendon yield (=ultimate limit state). It will be shown in this paper that the use of analytical formulas the use simple formulas gives satisfactory results for the internal pressure values at these corner point compared to nonlinear finite element simulations or to experimental results.

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The Nuclear Waste Management Organization of Japan will select the final repository site for the high level radioactive waste (HLW) and low-level radioactive waste containing long-lived radionuclide (TRU) disposal by stepwise site investigations. Borehole investigations will be conducted to characterize geological environment in future site selection campaign. Full core recovery is required in order to obtain the information of geological structure and lithology by core observation. The stable borehole wall is required to operate the in-situ tests in the borehole. Therefore, it must be important to select the optimum drilling bit and drilling mud as well as to control the bit rotation speed. The authors have conducted a demonstration test of several site investigation techniques in a coastal area in the south of the Kanto district, Japan since 2006. The geology of this area consists of Neogene mudstone, sandstone, siltstone and Quaternary sediments. Three deep boreholes have been drilled in this demonstration test. However, the collapse of borehole walls and poor core recovery often occurred in the borehole during drilling in the brittle sedimentary rocks. The results of the X-ray diffraction analysis and the cation exchange capacity measurement using rock sample indicate that the borehole troubles seem to be caused by hydration and swelling, dispersion of clay minerals. Considering such geological conditions, it is important to find an optimum drilling mud to catch the core with high core recovery rate and construct the stable borehole wall without occurred collapse of a borehole wall. Dispersion test and swelling test have been conducted in the laboratory by using the core with several kinds of mud. From a qualitative perspective, KCl-polymer mud would be suitable for drilling the brittle sedimentary rocks.

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The Fukushima-Daiichi nuclear accident has highlighted the importance of analyzing the meltdown of a Boiling Water Reactor (BWR). The core melting in a nuclear reactor is affected by different phenomena, whose the most substantial are the chemical reactions and the interactions between the core materials; these can lead to a temperature escalation and bring forward the melt progression. Moreover it has to be considered that the geometry of a BWR complicates the core modeling. The existence of channel boxes and control blades represents a significant challenge for the heat transfer calculation, in particular as regards the thermal radiation. The CORA-18 has been selected as the validation test to provide information on the damage progression of a BWR fuel element. A model has been built in SAMPSON/MCRA and a simulation of CORA-18 experiment has been performed. The zirconium oxidation assumes noteworthy importance after the temperature has reached values close to 1300 K and leads to a considerable heat release. The temperature trends appear to be similar to those in the experiment CORA-18; the hydrogen production, however, is approximately the double of the experimental data. The melting of the structures occurs at the end of the computed transient and portions of the debris are deposited on the rods under the form of crust.

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KERENA is an innovative boiling water reactor concept with passive safety systems (Generation III+) of AREVA. The reactor is an evolutionary design of operating BWRs (Generation II). In order to verify the functionality and performance of the KERENA safety concept required for the transient and accident management, the test facility "Integral Teststand Karlstein" (INKA) was built at Karlstein (Germany). It is a mock-up of the KERENA boiling water reactor containment, with integrated pressure suppression system. The complete chain of passive safety components is available. The passive components and the levels are represented in full scale. The volume scaling of the containment compartments is approximately 1:24. The reactor pressure vessel (RPV) is simulated via the steam accumulator of the Karlstein Large Valve Test Facility. This vessel provides an energy storage capacity of approximately 1/6 of the KERENA RPV and is supplied by a Benson boiler with a thermal power of 22 MW. With respect to the available power supply, the containment- and system-sizing of the facility is by far the largest one of its kind worldwide. From 2009 to 2012, several single component tests were conducted (Emergency Condenser, Containment Cooling Condenser, Core Flooding System etc.). On March 21st, 2013, the worldwide first large-scale only passively managed integral accident test of a boiling water reactor was simulated at INKA. The integral test measured the combined response of the KERENA passive safety systems to the postulated initiating event was the "Main Steam Line Break" (MSLB) inside the Containment with decay heat simulation. The results of the performed integral test (MSLB) showed that the passive safety systems alone are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them as response to an anticipated accident scenario. The test provided evidence that the INKA is worldwide the first large scale test facility to perform integral verification tests of passive safety concepts under plant-like scaling and thermodynamic conditions. Hence, the test facility also shows that it is capable to perform containment response tests for existing Generation II BWRs (with active safety systems) and advanced (passive) reactor designs besides KERENA. These test results can be used to strengthen existing containment codes with regard to heat transfer, natural circulation, gas- and temperature stratification and others.

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The adsorption velocities of molecular iodine onto chemically inactive and active aerosols estimated through the analysis of the tests in the THAI-2 project were inputted to a series of severe accident analysis for a boiling water reactor with Mark-I containment vessel by an integral severe accident analysis code, THALES2/KICHE. A severe accident initiated by long-term station blackout with drywell failure due to overpressure was selected as a reference sequence. The results of the analysis implied that the iodine source term as molecular iodine increased in cases that its adsorption onto aerosol was taken into account, which was lower than but comparable to the release fraction of iodine as cesium iodide. The fraction of the aerosol-adsorbed molecular iodine in these cases was predicted to be larger than that of the gaseous molecular iodine. The analysis also indicated that the ratio of the releases for cesium iodide, aerosol-adsorbed and gaseous molecular iodine depended on characteristics of molecular iodine such as the gas/liquid mass transfer and the adsorption onto aerosol and structure surface.

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In order to evaluate the effect of chemical forms of Cs and I on source terms and solution pH in a severe accident of a BWR, parametric analysis was performed with an integral severe accident code, THALES2, developed by JAEA. In the present analysis, THALES2 code was modified to take into account CsOH, Cs_2MoO_4 and CsBO_2 as Cs chemical forms, and CsI and HI as I chemical forms. The severe accident sequence examined was similar to that occurred at unit 3 of the Fukushima Dai-ichi NPP. Due to the effective scrubbing, approximately 90% of the initial core inventory of Cs and I was predicted to be retained in the water pool of the suppression chamber, resulting in limited influence of Cs chemical form on Cs source term. On the other hand, the present analysis indicated that solution pH of the water pool was strongly affected by chemical forms of Cs and I. This outcome implies that chemical forms of Cs and I influence I source term since the formation of volatile I species such as I_2 and organic iodine in the water pool depends strongly on solution pH.

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A Prompt Gamma Neutron Activation Analysis (PGNAA) system with a LaBr_3 detector was built at the #1 radial beam port of Xi'an Pulsed Reactor. This system was designed to test the feasibility of the application of LaBr_3 detector in nuclide identification using PGNAA method. The prompt gamma spectrum of a sample irradiated by neutron beams was measured. The results were analyzed by a self-developed program, and the nuclides (^<35>Cl, ^<159>La, ^1H, ^<207>Pb, ^<63>Cu and ^<56>Fe) in the sample were identified. It indicates that the LaBr_3 detector can be expected to do material inspection using PGNAA method in further application.

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PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace a conventional active auxiliary feedwater system. PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the PCCT (Passive Condensation Cooling Tank). PAFS-CIV-01 test was performed for validating cooling rate of the reactor according to the CIV opening stroke at the FLB accident, which was analyzed as the most severe case in the APR+ SSAR (Standard Safety Analysis Report). With an aim of simulating a FLB+CIV accident of the APR+ as realistically as possible, the three-level scaling methodology was taken into account to determine the test conditions of the steadystate and the transient. From the experiment, major thermalhydraulic phenomena such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate in PAFS were investigated and discussed. With decreasing the flow area of the CIV, the gradient of system pressure and temperature were reduced due to the decrease of the heat removal performance. As the CIV was re-opened, the two-phase natural convection flow in the loop of the PAFS was recovered. From the present experimental result, it could be concluded that the cooling rate of the core was controlled by the adjustment of the CIV opening stroke when the APR+ PAFS was operating.

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In response to the Fukushima accident, the European Commission ordered to perform stress tests to all European Nuclear Power Plants (NPPs). Due to shortage of time a number of conclusions in national stress tests reports were based on engineering judgment only. In the Czech Republic, as a follow up, a consortium of Research Organizations and Universities has decided to simulate selected stress tests scenarios, in particular Station Black-Out (SBO) and Loss of Ultimate Sink (LoUS), with the aim to verify conclusions made in the national stress report and to analyse time response of respective source term releases. These activities are carried out in the frame of the project "Prevention, preparedness and mitigation of consequences of Severe Accident (SA) at Czech NPPs in relation to lessons learned from stress tests after Fukushima" financed by the Ministry of Interior. The Research Centre Rez has been working on the preparation of a MELCOR model for VVER1000 NPP starting with a plant systems nodalization. The basic idea of this paper is to benchmark the MELCOR model with the validated TRACE model, first comparing the steady state and continuing in a long term SBO plus another event until the beginning of the severe accident. The presented work focuses mainly on the preliminary comparison of the thermo-hydraulics of the two models created in MELCOR and TRACE codes. After that, preliminary general results of the SA progression showing the hydrogen production and the relocation phenomena will be shortly discussed. This scenario is considered closed after some seconds to the break of the lower head.

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In response to the Fukushima accident the Research Centre Rez plans to simulate accident scenarios of NPP Temelin and NPP Dukovany analyzed in the stress tests ordered by the European Commission to all European NPPs. While in a number of safety analysis of various accident scenarios it is sufficient to use one point reactor kinetics there are selected types of accidents in which it is useful to model the space (3D) behaviour of neutron kinetics, in particular control rod ejections, boron dilution scenarios, including transitions from design basis to beyond design basis accidents. For example in order to analyse control rod ejections, boron dilution scenarios and local changes of moderator temperature one has to see the local changes (3D) of reactivity and power. In this paper the current model of VVER1000/V320 reactor created in PARCS 3.2 is benchmarked with MCNP6 code and some reference data from NPP Temelin. The geometry of fuel in the core used for the analyses comes from fuel assemblies of NPP Temelin (hexagonal shape). This benchmark may serve as one of validation tests of the PARCS code for VVER1000/V320.

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The main aim of this paper is to simulate a safety analysis related to the SB LOCA (Small Break Loss of Coolant Accident) using a CFD (Computational Fluid Dynamics) code, where the investigated part is a section of VVER 440 reactor. SB LOCA can be regarded as one of the accidents disrupting the normal operation and its investigation is essential for the safety evaluation of the entire plant. Solved scenario assumes a small break on the hot leg where the coolant leaks. Lost coolant is compensated by safety systems injecting water from the emergency tanks with significantly lower temperature in comparison with the coolant. A major part of this paper is dedicated to the modelling and to the mesh generation. This part deals with the modelling of the involved components, which was done using Design Modeler. Complex components were adjusted, simplified or omitted in order to generate a suitable mesh. This process was done using ANSYS-Meshing by the application of the appropriate algorithms. Attention was also paid to the number of cells and their quality as the crucial parameter affecting the accuracy and the length of calculation. ANSYS-Fluent r14 was chosen as the CFD solver that allows 3D solution of a transient scenario. Input parameters representing boundary conditions were obtained from the simulation tool RELAP where the investigated analysis was performed for the entire primary circuit. The result of this analysis principally provides of the determination of the maximal and minimal temperature in the regions, that are prone to the brittle fracture, namely welds at the inlet nozzles and the weld nearby the core.

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