The extruded materials are routinely used in chemical, food and nuclear industry to enhance the structural integrity and reliability of such materials. The material composition, cleanliness, manufacturing route, heat treatment and many other factors influence the mechanical properties to a great extent. The uniaxial tensile test is a simple and versatile test to expose most of the mechanical properties required to model the system. In present study, the mechanical properties of an Al-Mg-Si alloy extruded cylinder have been determined with the help of uniaxial tensile test. The microstructural features were studied with the help of optical and scanning electron microscopes. Samples for tensile testing were cut and machined to final dimensions according to ASTM standard. The tests were conducted in displacement mode at different cross head speeds corresponding to strain rates ranging from 10^<-5> to 10^<-1> s^<-1>. All the tests were conducted in air atmosphere and at ambient temperature. The data thus obtained include: yield strength, ultimate tensile strength, elastic modulus, percent elongation and reduction of area. An increase in temperature with the increase of strain rate is observed in the necking section. Monotonic tests showed slight increase in yield strength upto a strain rate of 1.33 x10^<-1> s^<-1>. However, at the maximum strain rate test, this effect diminished due to high deformation rates and resulting rise in temperature. At maximum strain rate condition, an increase of 78 % in elongation and 96 % in reduction of area is measured as compared to the slowest strain rate test. Examination of the fracture surfaces of monotonic tensile specimens in SEM showed the ductile failure with dimpled fractured surfaces.

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Non-destructive method, positron annihilation lifetime spectroscopy (PALS) was applied as an evaluation tool for microstructure study of the four different Fe-Cr alloys. This paper describes ion implantation technique as a tool for creating microstructure defect in structural materials, similar to radiation damage in nuclear facilities. In present state of the research non implanted specimens were measured using PALS and TEM. The results show the dependency of positron lifetime in defects on the grain size. Measurements also confirmed presence of dislocations, typical for tempered ferritic/martensitic steels, which were the results of thermal treatment in manufacturing of the specimens.

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Every modernization measure results in failure probability improvement of a specific component, group of components or system. This failure probability improvement leads to the NPP safety improvement, by core damage frequency reduction. The calculation of this frequency by PSA requires profound knowledge of PSA theory and practice, available and installed PSA and rights to use it. This makes impossible for wide range of specialists to do quick calculation of safety change, in case of a component (group, system) reliability change. The presented software code enables core damage frequency change calculation in seconds, without using of the PSA models. Instead, the code automatically uses the PSA results, which are more available. The code can calculate not just the core damage frequency reduction, but the core damage frequency increase as well, for example in case of component (group, system) outage for planned or unplanned repair, component failure or reliability reduction, test outage etc. The code can calculate the core damage frequency changes due to common cause failure as well. There is a possibility to compare the effect of the different modernization measures (or activities within a measure), by the core damage frequency changes. The code lists the changed event trees, the changed accident scenarios and the values of their core damage frequency changes. The code calculates the core damage frequency change in the following cases: ・Failure probability change of a component (group, system); ・Seismic fragility change of a component (group, system); ・Outage of a component (group, system).

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The simultaneous occurrence of different load conditions such as gravity loads, lateral and vertical loads due to seismic event in various combinations should be considered to generate most critical design conditions. The aim of the paper is to evaluate the behaviour and the structural response in form of in-structure response spectra of the reactor building internal structures, under site specific seismic loading, and to determine whether these ones satisfy current international safety regulations. A preliminary conceptual analysis and design of a nuclear building with different foundation embedding depths for the most critical conditions were discussed in this paper with reference to solutions considered for a Near Term generation Nuclear Power Plant. The seismic input motion was considered as a free field response spectra of 0.2g PGA, while the Soil-Structure Interaction (SSI) effects were taken into account through appropriate features. To achieve the purpose of this study, the foundations can be seen as a stable base for the superstructure able to transfer safely all loads from ground to the internals. The general approach was consistent with up-to-date design conditions for evaluation and upgrade of nuclear power plant facilities. The project's major objectives may be summarized as follows: study of available data for preliminary design and as built conditions, creation of 3-D detailed finite element models, determination of dynamic characteristics, verification of capability of structure to resist relevant design load combinations, calculation of all important characteristics for specific parts of the structure, determination of possible feature and components mostly affected by the assumed seismic loads. The results of the performed analyses, the possible effects of SSI and the response of internal components (e.g. Nuclear Building, Vessel-GVs and GV-tubes system) seem to confirm the possibility to achieve an upgrading of the geometry and the performances of the proposed solutions for the considered NPP.

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The 4S (Super-Safe, Small and Simple) reactor is a sodium-cooled fast reactor aiming at an application to dispersed energy source and multi-purpose use. Its electrical output is 10MW or 50MW and the core lifetime can be varied from 10 to 30 years without refueling. An introduction of RVACS (Reactor Vessel Auxiliary Cooling System) can enhance the passive decay heat removal capability. In the present study, the RVACS performance in 4S reactor (10MWe, pool-type) was analytically evaluated under the functional loss of RVACS stack condition, which is considered as a beyond design basis event. A protected loss of heat sink accident was selected and simulated to evaluate the capability of RVACS to cool the plant under such an unusual condition. The three-dimensional thermal hydraulic analysis was conducted by PHOENICS code. Analytical results show that the functional loss of air outlet stack has more effect on RVACS performance than that of air inlet stack. The air flow rate in RVACS under the functional loss of one out of two outlet stacks decreases up to approximately 60% and the heat removal rate approximately 70%, comparing with those under the normal stack condition. However, the maximum hot plenum temperature is low enough to satisfy the safety criteria.

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In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR.

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To evaluate factors affecting irradiation assisted stress corrosion cracking (IASCC) behavior, the OECD Halden Reactor Project (HRP) is performing in-core crack growth and crack initiation studies in the Halden boiling water reactor (HBWR). In the crack growth studies, compact tension (CT) specimens are prepared from irradiated stainless steels (SSs), types 316NG, 347, 304L and 304 with neutron fluences in the range of 9.0x10^<20> - 1.2x10^<22> n/cm^2 (E > 1 MeV). The crack growth measurements are conducted in simulated boiling water reactor (BWR) and pressurized water reactor (PWR) environments. In a crack initiation study, miniature tensile specimens are machined from a type 304L SS irradiated to 8.0x10^<21> n/cm^2 (E > 1 MeV). Two specimens are installed in each test unit which has pressurized bellows for loading and a linear variable differential transformer (LVDT) for failure detection. Constant load is applied to the specimens in the 15 test units which are exposed to a simulated BWR environment. 5 signal changes indicating specimen failure have been detected since May 2002, when the test began.

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Limit load equation for a sweepolet pipe with a circumferential surface crack in weld heat-affected zone (HAZ) is proposed. Fracture tests on the scale model of sweepolet pipe were conducted under tensile loading to ensure the conservatism of proposed limit load equations. Specimens are made of SUS316 (equivalent to AISI 316) stainless steel and two specimens those have a circumferential crack with depth a/t of 0.6 and 0.8 were prepared. The maximum loads in tests are compared with the limit load calculated by the proposed equation. In the proposed model, sweepolet is regarded as a cylinder with a constant depth circumferential surface crack in which the inner radius of cylinder is equal to the minimum radius of weld line (i.e. distance from the center of branch pipe to weld line). Then the limit load equation by Kurmar et al. for a straight pipe with a circumferential crack can be applied. Proposed equation estimates limit load conservatively compared with experimental strength. For the specimen of a/t=0.8, the estimation gives over conservative result. It was considered that the over conservatism is due to the complexity of crack geometry because the crack overhanged and once closed in the initial stage of tensile loading. An alternative model is also proposed. In this model, weld line is projected on a plane perpendicular to loading axis and is assumed to be an elliptical shape. The ligament area is calculated for this elliptical weld locus and the limit load is estimated as a product of flow stress and ligament area.

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Sodium leaks and resultant fires play an important role in safe operation of a Fast Breeder Reactor. Leak Collection Tray (LCT) is a passive device which is used to collect the highly reactive liquid sodium in case of an accidental leakage. The sodium fire is contained based on the principle of oxygen starvation which effectively extinguishes the sodium fire and mitigate the consequences of it. The leak collection tray basically consists of two parts, an upper part with two sloping plates which guides the leaked sodium to a central drainpipe and a lower part which is the sodium hold up vessel. In the current work numerical simulations have been performed to analyze the flow of hot liquid sodium in the LCT. The simulation involves tracking of the interface between the hot sodium and the ambient air for which the Volume of Fluid (VOF) technique has been used. Effect of various parameters such as slope angle, drain pipe diameter and vent pipe diameter on the collection efficiency of the tray has been studied and optimized design parameters have been attained considering the constraints imposed by the combustion of liquid sodium.

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Some spent fuels generated in nuclear power plants are ongoing plan to be stored in interim storage facilities by metal casks for about 50 years in Japan, until nuclear fuel reprocessing plants can accept these spent fuels. A stainless steel and aluminum alloy containing neutron-absorbing materials is used as a basket that is a component of a partition plate for the spent fuel in the metal cask. We have developed a high-performance borated aluminum alloy. This material has not only high heat conductivity, but also high strength at high temperatures. This strengthening is caused by stable precipitation at high temperatures, created by the addition of boron as a neutron absorber, and of Mn and Cr in the matrix of alloy A6082 to forcibly create the precipitation. The yield strength of the developed alloy is 74 MPa at 523 K and heat conductivity is 185W/m・K. An accelerated test based on the Larson-Miller parameter indicates that this material retains high strength for 50 years at 523 K. The developed alloy is stronger than traditional B-Al specified ASME code case. These results indicate the developed alloy is stable to heat aging and useable for the components of metal casks.

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In the field of safety analysis of nuclear reactors horizontal stratified flows with a free and wavy surface are relevant for the injection of cold water into the hot leg. For the numerical simulation of these flows the momentum exchange term and the two-fluid Reynolds stresses of the two-fluid equations have to be closed. In this work a phase interaction model, modeling the momentum exchange, is derived which correlates the waviness of the surface flow with local turbulence quantities. The result provides a two phase layer with a thickness depending on the underlying physics. To close the phase interaction an extended k-ω model is introduced to take account of the two-fluid Reynolds stresses. The turbulence is damped at the free surface with an additional two phase turbulence damping term. The phase interaction model and the k-ω model are compared with experimental results for supercritical flow conditions gained in the WENKA test facility. The partially reversed flow conditions are compared both with the k-ω model and the k-ε model.

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Thermal fatigue of the coolant circuits of PWR plants is among the major issues for nuclear safety. The problem is very acute in mixing zones, like Tee junctions. In these zones, the conjunction of high levels of turbulence with large water temperature differences can lead to large thermal fluctuations at the wall. Studies on the subject have been already done at EDF using an unsteady conjugate heat transfer approach (Peniguel et al., 2003), (Sakiz et al., 2004). The fluid flow was solved with a Large Eddy Simulation technique in order to deal with the unsteady aspects of the problem which may cause large mechanical stresses. However at that time, due to CPU cost required by these simulations, the computational domains were limited to a small part of the T-junction, and simplified inlet conditions were used for the velocity at the inlet. Advances in computer capabilities allow now to extend the geometry upstream of the junction. The purpose of this paper is to investigate some of the effects induced by the upstream elbows. Three simulations of the former Residual Heat Remover (RHR) junction found in N4 type PWR nuclear plants, are done with the CFD tool Code_Saturne developed at EDF (conjugate heat transfer is not considered in the present study). The results of the computation including the upstream elbows are compared with the results of two simulations using simplified velocity profile and synthetic turbulent inflow conditions. In the present study, upstream elbows have little influence on the flow dynamic. As expected, the mean and RMS temperature field are twisted when elbows are taken into account, but, as it was found in previous studies (Peniguel et al., 2003), these new computations confirm that no specific frequency seems to appear downstream of the T-junction.

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Based on the power conversion system of nuclear and conventional islands of Dayabay nuclear power station, this paper models the thermal-hydraulic systems for PWR by using the best-estimate program, RELAP5. To simulate the full-scope power conversion system, not only the reactor coolant system (RCP) of nuclear island, but also the main steam system (VVP), turbine steam and drain system (GPV), bypass system (GCT), feedwater system (FW), condensate extraction system (CEX), moisture separator reheater system (GSS), turbine-driven feedwater pump (APP), low-pressure and high-pressure feedwater heater systems (ABP & AHP) of conventional island are considered and modeled. A comparison between the simulated results and the actual data of reactor under full-power demonstrates a fine match for Dayabay, and also manifests the feasibility in simulating full-scope power conversion system of PWR with RELAP5.

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We have been developing an alpha-radioactivity monitor based on ionized air transportation technology (alpha clearance monitor) for efficiently performing clearance level inspection of massive amounts of uranium-contaminated waste. This paper is one of a series related to a method for quantitatively evaluating alpha radioactivity from a measured ion current value. Using a prototype alpha clearance monitor, we measured alpha radioactivities of uranium-attached components used in back-end facilities of the nuclear fuel cycle (uranium-attached samples). We compared the measured radioactivities with reference radioactivities by assaying the samples with an inductively coupled plasma mass spectrometer and the radioactivities measured with a scintillation survey meter. The radioactivities of curved plate samples measured with the prototype monitor were highly linear with respect to the reference radioactivities (a residual standard deviation of 8%). Absolute values of the radioactivities obtained with the prototype monitor and with the survey meter were about 40% and 50% smaller than the reference radioactivities, respectively. Measurements of complex-shaped samples indicated that, to measure diverse-shaped waste, it was necessary to classify the waste by shape and determine conversion coefficients corresponding to each group in advance, experimentally and theoretically.

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The Mizunami Underground Research Laboratory (MIU) is currently being constructed. The MIU design consists of two 1,000 m-deep shafts with several research galleries. The goals of the MIU project are to establish techniques for investigation, analysis and assessment of deep geological environments, and to develop a range of engineering expertise for application in deep underground excavations in crystalline rocks such as granite. The diameter of the Main and the Ventilation Shafts are 6.5 m and 4.5 m, respectively. Horizontal tunnels to connect the shafts will be excavated at 100 m depth intervals. The Middle Stage, at about 500 m in depth, and the Main Stage, at about 1,000 m in depth, will be the main locations for scientific investigations. The Main and the Ventilation Shafts were 180 m and 191 m deep, respectively, in November 2006. During construction, water inflow into the shafts has been increasing and affecting the project progress. In order to reduce the water inflow into the shafts, pre- and post-excavation grouting has been planned. A post-excavation grouting test has been undertaken in the Ventilation Shaft and the applicability of several techniques has been evaluated. This paper describes an outline of the MIU project, its work plan and the results of the post-excavation grouting test.

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Crevice corrosion of titanium and its alloys were investigated in 10% sodium chloride at 100 ℃ simulating the environment of the overpack near the seaside. The pH and Chloride ion concentration inside the crevice were monitored by using W/WO3 and Ag/AgCl microelectrode, respectively. The pH and Cl^- concentration within the crevice were calculated from the standard potential-pH and potential-log [Cl^-] calibration curves. The effect of Mo on the crevice corrosion of titanium was mainly studied. The passivation behavior of the titanium and Ti-15%Mo alloy were also studied using electrochemical impedance studies. A marginal decrease in pH and increase in Cl^- ion concentration were observed for pure titanium at 100 ℃, where there was large increase of the crevice current. On other hand, there was no apparent change in pH and Cl^- ion activity inside the crevice for Ti-15%Mo alloy, where there was no increase of the crevice current. Based on the results, it has been documented that the Ti-15%Mo alloy was not susceptible to crevice corrosion in 10 % NaCl solutions at 100 ℃. The corrosion reaction resistance (R_t) was found to increase with addition of Mo as an alloying element and also increase with applied anodic potential. Hence, Mo is able to be an effective alloying element, which enhanced the crevice corrosion resistance of titanium.

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The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Surete Nucleaire (IRSN) and the German Gesellschaft fur Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MW PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

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This paper presents the results of numerical simulations of sub-channel flow under swirl conditions conducted with the CFD code STAR-CD. For validation purposes an experiment including detailed results for local void distribution in a heated vertical tube was chosen. The CFD results show a very good agreement for low void fractions. This work outlines possible optimization criteria for fuel assembly spacer grids. The parameters swirl, pressure loss and accumulation of bubbles on the rods were analyzed. The work is divided into three parts: single-phase calculation of the subchannel flow under swirl conditions, calculation of two-phase isothermal flow and calculations of the two-phase non-isothermal flow with heated rods. Simplified generic subchannel geometry with a prescribed rotational flow at inlet was used for the first CFD analyses. Finally, two-phase calculation for complex geometries such as real spacer grids in connected subchannels was possible. Calculations show interesting details of two-phase flow distribution in the subchannel. The results provide CHF relevant details for improved interpretation of the flow situation with respect to heat transfer. This will increase the reliability of engineering judgment significantly.

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The paper describes methodology for quantification of fuel cladding failure as a result of Loss of Coolant Accident. The methodology is based on external coupling of the RELAP5 code and TRANSURANUS code. The thermo-hydraulic response of the unit to the accident is simulated by RELAP5 code, providing initial and boundary conditions for the thermo-mechanical simulation by the TRANSURANUS code. Cladding failure criterion of the TRANSURANUS code, derived and implemented into the code in the framework of EXTRA EURATOM 5th Framework Programme is used. Cladding failure probability is evaluated by the Monte Carlo algorithm varying the outer cladding temperature. In the second part of the paper, an example of application of the methodology for typical maximum design basis accident of the VVER-440 is given, presenting every step of methodology and typical failure rate for this type of accident.

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Currently, in Pressurized Water Reactor (PWR) core seismic analysis, various testing methods are used to determine in-grid stiffness and ingrid damping. In-grid stiffness is the effective horizontal stiffness between all of the fuel rods at a grid elevation and a center point of the grid. This stiffness is determined by fuel-assembly lateralimpact tests, in air, at room temperature. In-grid stiffness and damping data are currently being derived from fuel-assembly-grid-impact data obtained with a single beam (one-beam) per assembly model. In this paper, a better correlation with the fuel-assembly lateral-impact test data is achieved by utilizing a "two-beam' fuel-assembly model. The "two-beam" model employs the rudimentary finite element technique of grouping items with common properties and boundary conditions as they would exhibit common behaviors. The first beam represents the total number of fuel rods. The second beam represents all of the guide tubes and the instrument tube. Horizontal springs representing in-grid stiffness are incorporated in the "two-beam" model between the fuel-rods beam and the guide-tubes beam at each spacer grid position. In the "one-beam" model, the fuel assembly is represented with a single vertical beam. That is, all of the fuel rods, guide tubes and instrument tube are lumped into one beam. The model extends from the top surface of the bottom nozzle to the bottom surface of the top nozzle. The upper and the lower ends of the beam are assumed fixed. In the "two-beam" model, only the guide-tube beam is assumed to be fixed at the upper and lower nozzles, while the fuel-rod beam is not fixed at the upper and lower nozzles. This set of end conditions is representative of the actual fuel assembly structure; hence, this model results in more realistic shear forces and moments on the top and bottom nozzles. With the "two-beam" linear model, it is possible to benchmark more closely to both the first natural frequency and lateral stiffness, since it represents the more realistic boundary conditions.

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Thermal fatigue may occur where the high and low temperature fluid is mixed at T-junction of branch pipe which are common piping element in nuclear power plants. In considering thermal fatigue during design phase, it is important to evaluate thermal load from design condition of flow rate, temperature difference, pipe diameter, etc. IMAT-F, an evaluation method integrating thermal hydraulic and structure analysis, was developed in this study to estimate thermal load excluding safety margins or conservative engineering judgment. In order to validate the system, numerical simulations were carried out on high-cycle thermal fatigue test SPECTRA conducted by Japan Atomic Energy Agency. It was confirmed that IMAT-F can simulate the temperature fluctuation of both fluid and structure with a fluid-structure coupled analysis and the thermal stress caused by transient temperature distribution in the pipe wall.

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When the clearance is judged by the on-site surface contamination measurement of large-scale objects by beta-ray measurement, various needs arise in such viewpoints as the background compensation for the gamma ray self-shielding or the difficulty in measurement of the narrow apertures of objects, etc. To correspond to these needs, the wavelength shifting technique combined with plastic scintillators have been applied to various types of contamination survey meters according to the waste form or the surface condition of the objects. The survey meters are found to be useful to be applied for surface contamination measurement in various situations encountered in on-site measurement

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An innovative alpha radioactivity monitor for clearance level inspection has been developed. This apparatus measures an ion current resulting from air ionization by alpha particles. Ions generated in the measurement chamber of about 1 m^3 in volume are transported by airflow to a sensor and measured. This paper presents computational estimation of ion transport efficiencies for two pipes with different lengths, the inner surfaces of which were covered with a thin layer of uranium. These ion transport efficiencies were compared with those experimentally obtained for the purpose of our model validation. Good agreement was observed between transport efficiencies from simulations and those experimentally estimated. Dependence of the transport efficiencies on the region of uranium coating was also examined, based on which anticipated errors arising from unclear positions of contamination are also discussed.

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A Steam Injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. This provides SI with capability to serve also as a direct-contact feed-water heater that heats up feed-water by using extracted steam from the turbine. We are developing technology for "Innovative Simplified Nuclear Power Plants" in order to further improve the economy and safety of nuclear power plants. Our technology development aims to significantly simplify equipment and reduce physical quantities by applying "High-Efficiency SI", which are applicable to a wide range of operation regimes beyond the performance and applicable range of existing SIs and enables unprecedented multistage and parallel operation, to the low-pressure feed-water heaters and Emergency Core Cooling Systems (ECCS) of nuclear power plants, as well as achieve high inherent safety to prevent severe accidents by keeping the core covered with water (a severe accident-free concept). The innovative-simplified nuclear power plant consists of a simplified feed-water heating system, a passive core injection system and a passive containment cooling system. This report describes the results of the scale model test, and the transient analysis of SI-driven passive core injection system. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba corporation, and seven universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.

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Operating experience of steam generators in nuclear power plants shows that constituent tubes are affected by diverse degradation mechanisms. The cracked tube can stay in service if sufficient structural margin is assured to preclude the risk of failure. With regard to estimation of the maximum load carrying capacity, most of preceding researches were focused on limit load concept but extensive test data would be needed to endow confidence of the relevant solution. In contrast to this, up to now, integrity assessment scheme based on the elastic-plastic fracture mechanics concept has not been settled despite of its efficiency. In this paper, elastic-plastic finite element analyses of cracked steam generator tubes are carried out. Fracture toughness of the typical tube material is also measured and the crack instability is evaluated by comparing the crack driving force with fracture toughness of the tube material. Analysis results show that the elastic-plastic fracture mechanics method predicts the load carrying capacities accurately compared to the experimental data. Thus, it is anticipated that the elastic-plastic fracture mechanics concept can be applied to integrity assessment of steam generator tubes with through-wall cracks.

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51 organizations network in SARNET (S__-evere A__-ccident R__-esearch <NET>___-work of Excellence) their capacities of research in order to resolve the most important remaining uncertainties for enhancing, in regard of Severe Accidents (SA), the safety of existing and future Nuclear Power Plants (NPPs). This project, co-funded by the European Commission (EC), has been defined in order to optimise the use of the available means and to constitute sustainable research groups in the European Union. SARNET tackles the fragmentation that exists between the different R&D national programmes, in defining common research programmes and developing common computer tools and methodologies for safety assessment. SARNET comprises most of the actors involved in SA research in Europe (plus Canada). To reach these objectives, all the organizations networked in SARNET contribute to a so-called Joint Programme of Activities (JPA), which consists in: ・Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; ・Harmonizing and re-orienting the research programmes; ・Jointly analysing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; ・Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA), which capitalizes in terms of physical models the knowledge produced within SARNET; ・Developing Scientific Databases, in which all the results of research programmes are stored in a common format (DATANET); ・Developing a common methodology for Probabilistic Safety Assessment (PSA) of NPPs; ・Developing courses and writing a text book on SA for students and researchers; ・Promoting personnel mobility between various European organizations. After the first period (2004-2008), co-funded by the EC, the network will progressively evolve toward self-sustainability. The bases for such an evolution, still under discussion, are presented in the last part of the paper.

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The hydrogen risk is one of the most important containment integrity challenge during a severe accident progression at the VVER type reactors installed at the Czech NPPs. On the basis of recent comprehensive research results the general flame acceleration (FA) and deflagration-to-detonation transition (DDT) criteria were formulated. The main objective of the new methodology developed in the NRI Rez was to prepare an analytical tool for the assessment of the hydrogen risk at the Czech NPPs in the course of a severe accident and also for the design of hydrogen removal system which should be able to prevent or at least minimize the threats of hydrogen detonation in containments. The major idea on which the development of models for the FA and DDT criteria was based is described in OECD state-of-art report (NEA, 2000). The module for the computation of FA criterion (often named σ-criterion) and DDT one (also named λ or Dorofeev criterion) was linked to the MELCOR 1.8.5 model of VVER-1000 containment. The most important feature of new model is that it makes possible to evaluate the both of the criteria for all compartments in the containment continuously during severe accident scenario progression. The containment model, which could be used for such a calculation, has to be very detailed owing to appropriate description of hydrogen distribution. New model was tested by calculation of hydrogen detonation risk in the VVER-1000 containment during severe accident scenario initiated by medium break LOCA. At present two more VVER-1000 severe accident scenarios have been analyzed. The mapping of a hydrogen detonation risk in individual compartments inside containment was performed. The results of analysis confirm that the risk of hydrogen detonation in the great majority of containment compartments during severe accidents is very high if the hydrogen removal system is not installed.

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Vibration behaviors of a U-tube bundle in air-water two-phase cross-flows are presented. A total of 39 prototypic U-bend tubes were arranged in a rotated square array with a p/d of 1.633, which is identical to U-bend tubes of the OPR1000 steam generator. The present experimental facility has 2-span U-tubes in contrast to 8-span U-tubes of OPR1000 steam generator. However, modal analysis showed that the major vibration modes and corresponding natural frequencies were almost the same as those of the full 8-span U-tubes. Twelve 3-axis accelerometers were installed inside the U-tubes to measure the vibration motions. Tube vibration responses, critical velocity for fluid-elastic instability, two-phase damping ratio, and hydrodynamic mass were obtained. Finally, the instability factor (K) of Connors' relation was evaluated based on the above measured parameters.

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Timely preventive maintenance is of great importance for safety equipment on NPP's. But preventive maintenance becomes complicated when we face systems in where these activities can be performed only during outage period and reactor shutdown must be executed in case of safety related equipment failure during operation to perform corrective actions. For preventive maintenance strategy degradation monitoring is not always sufficient and degradation trending (extrapolation) should be performed to provide predictability of the extent of degradation and timely preventive or mitigative actions. Degradation trending in proactive maintenance case plays very important role. Current paper is devoted to the analysis and application of mathematical models to the degradation assessment of systems with dependent components. The main attention in present work was paid to the mathematical description of the degradation mechanism. Developed mathematical model allows assess and perform trending of degradation processes during various time moments. Methods can be applied for assessment of various risk estimates of hazardous systems such as systems at nuclear power plants.

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This study was performed at the Institute of Reactor Safety at the Research Center Karlsruhe. It is embedded in the ongoing investigations of the international code application and maintenance program (CAMP) for qualification and validation of system codes like TRACE and PARCS. The predestinated reactor type for the validation of these two codes was the Russian designed VVER-1000 because the OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 2 includes detailed information of the Bulgarian nuclear power plant (NPP) Kozloduy unit 6. The posttest-investigations of a coolant mixing experiment have shown that the predicted parameters (coolant temperature, pressure drop, etc.) are in good agreement to the measured data. The coolant mixing pattern especially in the downcomer has been also reproduced quiet well by TRACE. The coupled code system TRACE/PARCS which was applied on a postulated main steam line break (MSLB) provides good results compared to reference values and the ones of other participants of the benchmark. It can be pointed out that the developed three-dimensional nodalisation of the reactor pressure vessel (RPV) is appropriate for the description of transients where the thermal-hydraulics and the neutronics are strongly linked.

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The ESBWR is a natural circulation Boiling Water Reactor (BWR). Natural circulation provides major simplification by removal of the recirculation pumps and associated equipment. It is also synergistic with the passive safety systems. ESBWR is designed for natural circulation with enhanced natural circulation flow rate and greatly improved stability performance relative to operating BWRs at natural circulation conditions. The performance of the ESBWR has been established using the state-of-the-art TRACG computer code, which has been validated against a wide range of test data. Conservative design criteria for stability are easily satisfied. Natural circulation in the ESBWR provides numerous benefits with large margins to thermal design limits.

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We would like to present our recently developed high-resolution gamma ray measurement system for analyzing the dry-out effect and the determination of void fraction distributions in sub-channels of electrically heated fuel element bundles at the thermal hydraulic test loop KATHY in Karlstein (AREVA NP GmbH, Germany). The instrumentation setup enables a non-invasive measurement of cross-sectional void fraction profiles through the pressure vessel for fuel element bundles under typical nuclear reaction conditions. The gamma ray tomography system consists of a Cs^<137> isotopic source with an activity of about 165 GBq and a detector arc containing 320 single elements. The source radiation is restricted to a fan beam with a tungsten collimator. The average spatial resolution of the system is 3 mm in plane and 8 mm axial. With a special gantry vertical positioning and continuous rotation of the measurement setup is realised which is necessary for a complete tomography in different planes. Typically, transversal scans require an approximate recording time of 25 minutes in which the operation conditions must be constant. Gamma ray tomography is a relative measurement method. To determine void fractions calibration measurements are recorded at zero and one hundred percent void fraction respectively. It is a challenge to develop a tomography measurement system that is non-sensitive to temperature changes, high humidity and electrical fields to scale the void measurement to the calibration data. Cross-sectional images are reconstructed by standard filtered back projection algorithms.

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Since the 1980's maintenance optimization has been centered around various formulations of Reliability Centered Maintenance (RCM). Several such optimization techniques have been implemented at the Bruce Nuclear Station. Further cost refinement of the Station preventive maintenance strategy includes evaluation of statistical optimization techniques. A review of successful pilot efforts in this direction is provided as well as initial work with graphical analysis. The present situation reguarding data sourcing, the principle impediment to use of stochastic methods in previous years, is discussed. The use of Crowe/AMSAA (Army Materials Systems Analysis Activity) plots is demonstrated from the point of view of justifying expenditures in optimization efforts.

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Development of the Erbia (Er_2O_3)-bearing super high burnup (Er-SHB) fuel for LWRs has been launched in 2005 under co-operation with Nuclear Fuel Industries, Ltd., Nuclear Engineering, Ltd., Kyoto University, Osaka University and Nagoya University. This project is financially supported by the Innovative and Viable Nuclear Energy Technology (IVNET) development framework of Ministry of Economy, Trade and Industry (METI), Japan. The Er-SHB fuel aims to extend discharge burnup by increasing ^<235>U enrichment higher than 5wt%, which is common limitation for current LWR fuels. Erbia is mixed into all UO2 powder to accommodate criticality safety issues. The present paper summarizes current development status of the Er-SHB fuel.

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The mobile-chord method is applied to the method of characteristics (MOC) to reduce spatial discretization error of ray traces. In the mobile-chord method, the offset of a ray trace in a strip depends on the azimuthal angle to cancel the spatial discretization error. Though the mobile chord method has been used in the evaluation of the collision probability, its application to MOC has not been tried yet. The mobile chord method is implemented in the AEGIS code, which is a lattice physics code based on MOC. Verification calculations are carried out in the pin-cell geometry of the C5G7 benchmark problem using UO2 and MOX fuels. Calculation results indicate that spatial discretization error of the mobile chord method is smaller than that of the equidistant ray trace method, which is commonly used in conventional MOC codes. Since the mobile chord method can be used with the cyclic ray trace method, it will be an attractive candidate for MOC codes.

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For this paper, an initial literature review was conducted to investigate the potential applications of grouting technology for geological disposal of high level radioactive waste (hereafter called geological disposal), and the potential grouting material for each application. The results show the necessity of using suspension grout, such as cement-based grout, during excavation work, especially deep underground. Next, the method to achieve highly effective seals in crystalline rock with cement grout is studied. To enhance the sealing quality, cement grout should penetrate into very fine fractures, e.g. less than 100μm aperture. In the case of suspension grout, clogging with grout at the openings of rock fractures, especially fine fractures, tends to occur, which results in poor grout penetration. A laboratory experiment was conducted to investigate the clogging phenomenon; the results suggest that high injection pressures could be effective to prevent clogging. Finally, focusing on pre-excavation grouting for horizontal tunnels in crystalline rock, the effective grout hole patterns for achieving high quality sealing was studied. A series of theoretical calculations for water inflow and cost studies were conducted. The results indicate that a dense arrangement of grout holes in a relatively narrow area around a tunnel section, as practised in the Nordic countries, is favorable in hard crystalline rock.

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Benchmark neutronics and safety parameter analyses have been performed for the current core configuration of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Continuous energy cross-section data from ENDF/B-VI and ENDF/B-V and S(α,β) scattering functions from the ENDF/B-IV library were used. All the neutronic parameters including effective multiplication factor, neutron flux and power distribution were evaluated and benchmarking of the reactivity experiments were performed. Individual TRIGA fuel burnup was calculated by MCNP4C-ORIGEN2.1 computer codes at different locations of the core. The MCNP predictions and the experimentally determined values are found to be in very good agreement, which indicates that the Monte Carlo model is simulating the TRIGA reactor most appropriately.

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To effectively transmute Americium (Am) in a nuclear reactor, an idea of target fuel with a high content of Am and inert matrix has been proposed by one of the authors. We evaluated the Am transmutation amount and core characteristics, such as power distribution and reactivity effect, in a fast breeder reactor (FBR) and pressurized water reactor (PWR) loaded with the present Am target fuel. The FBR loaded with Am target fuel in radial blanket region could transmute Am amount which a 1000 MWe class light water reactor generates per year. Regarding the PWR, although the same amount of Am could be transmuted, the negative reactivity effect depends on the number of target pins and their locations in a target assembly. But, since its negative reactivity effect can be reduced by optimizing the number of target pins and their locations inside the target assembly, the feasibility of the FBR and PWR loaded with the present target fuel has been proven.

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The static thermodynamic properties of the molten salt system like LiF-NaF-BeF_2 influence the design and construction of the fuel salt and coolant in the Molten Salt Reactor for the new generation. In this paper, the equation of state of the ternary system 15%LiF-58%NaF-27%BeF_2, over the temperature range of 873.15K to 1073.15K at one atmosphere pressure, is described using Peng-Robinson equation modified by us. And the density of the ternary system is evaluated by this equation directly, and compared with the experimental data. Base on the equation of state, the other static thermodynamic properties such as the enthalpy, entropy and heat capacity at constant pressure are estimated by the residual function method and the fugacity coefficient method respectively. The density calculated by Peng-Robinson equation is in highly agreement with the experimental data, and the enthalpy, entropy and heat capacity evaluated by such two different methods are consistent with each other. We could conclude that the Peng-Robinson equation modified by us could be applicable to evaluate the density of the molten salt system, and we recommend the Peng-Robinson equation to be as the fundament to estimate the enthalpy, entropy and heat capacity of the molten salt system.

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The pin-by-pin fine mesh core calculation method is considered as a candidate of next-generation core calculation method for BWR. In this study, the diffusion and the simplified P_3 (SP_3) theories are applied to the pin-by-pin core analysis of BWR. Performances of the diffusion and the SP_3 theories for cell-homogeneous pin-by-pin fine mesh BWR core analysis are evaluated through comparison with cell-heterogeneous detailed transport calculation by the method of characteristics (MOC). In this study, two-dimensional, 2x2 multi-assemblies geometry is used to compare the prediction accuracies of the diffusion and the SP_3 theories. The 2x2 multi- assemblies geometry consists of two types of 9x9 UO_2 assembly that have two different enrichment splittings. To mitigate the cell-homogenization error, the SPH method is applied for the pin-by-pin fine mesh calculation. The SPH method is a technique that reproduces a result of heterogeneous calculation by that of homogeneous calculation. The calculation results indicated that diffusion theory shows larger discrepancy than that of SP_3 theory on pin-wise fission rates. Furthermore, the accuracy of the diffusion theory would not be sufficient for the pin-by-pin fine mesh calculation. In contrast to the diffusion theory, the SP_3 theory shows much better accuracy on pin wise fission rates. Therefore, if the SP_3 theory is applied, the accuracy of the pin-by-pin fine mesh BWR core analysis will be higher and will be sufficient for production calculation.

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Liquid sodium has the excellent properties as coolant of the fast breeder reactor (FBR). On the other hand, it reacts high with water and oxygen. So an innovative technology to suppress the reactivity is desired. The purpose of this study is to control the chemical reactivity of liquid sodium by dispersing the nanometer-size metallic particles (we call them Nano-particles) into liquid sodium. We focus on the atomic interaction between Nano-particles and sodium atoms. And we try to apply it to suppress the chemical reactivity of liquid sodium. Liquid sodium dispersing Nano-particles is named "Nano-fluid". Research programs of this study are the Nano-particles production, the evaluation of reactivity suppression of liquid sodium and the feasibility study to FBR plant. In this paper, the research programs and status are described. The important factors for particle production were understood. In order to evaluate the chemical reactivity of Nano-fluid the research programs were planned. The feasibility of the application of Nano-fluid to the coolant of FBR plant was evaluated preliminarily from the viewpoint of design and operation.

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The uptake behavior of Pu (IV) has been investigated by using calcium alginate gel polymer (CaALG) and TBP microcapsules (TBP-CaALG). The characterization of CaALG and TBP-CaALG was examined by SEM and IR, and the uptake properties and distribution of Pu(IV) ions were estimated by batch method. The uptake rate of Pu(IV) on CaALG and TBP-CaALG in the presence of 5 M HNO_3 was attained within 6 and 4 h, respectively, and K_d values for CaALG and TBP-CaALG after 7 h-shaking were 50.2 and 53.2 cm^3/g, respectively. Relatively large K_d values (90.3〜425 cm^3/g) were obtained for fresh CaALG and TBP-CaALG in the presence of 0.5〜2 M HNO_3. As for CaALG contacted with 3〜7 M HNO_3, the K_d value considerably decreased with increasing HNO_3 concentration, indicating the uptake of Pu(IV) was governed by ion-exchange reaction. In the presence of 5 M HNO_3, CaALG had relatively large K_d values around 50 cm^3/g up to 7 h, while the long time contact resulted in a considerable lowering of K_d. In contrast, large K_d values above 50 cm^3/g for TBP-CaALG were kept up to 233 h. Thus CaALG and TBP-CaALG are effective for the separation of Pu(IV) in the presence of highly concentrated HNO_3.

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Hot-pressed boron carbide (B_4C) pellet will be used as shielding material in China Experimental Fast Reactor (CEFR) which is the first fast reactor in China. In this paper, two types of B_4C sample provided by two different units in China were investigated on out-reactor properties, and one type of sample was studied on irradiation performance. Finally, the approval of domestic B_4C pellet as shielding material in CEFR was given.

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The present paper documents the CFD code validation activity carried out at the University of Pisa. In particular, the ANSYS CFX-10.0 code was used to simulate one of the experiments conducted at the ROCOM mixing test facility (FZD, Germany), that reproduced the injection of a de-borated slug in one cold leg of a pressurized water reactor (simulated by a salt tracer) with all circulation pumps at steady-state operation. The calculations were run on several grids obtained through different meshing strategies and having different sizes. The numerical results, in terms of normalized concentration of the transported passive scalar in the downcomer and at the core inlet, were compared against corresponding values obtained through experimental measurements of electrical conductivity in the ROCOM facility. Such comparison resulted in a general good qualitative agreement between simulations and experiments, while some discrepancies were evidenced from a quantitative point of view.

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The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since SIMMER-III is expected to become a standard tool for fast reactor safety analysis with likely application to licensing calculations, the code must be demonstrated to be sufficiently robust and reliable. For this purpose, a systematic assessment program of the code has been conducted in cooperation with European partners. The philosophy behind the code development was to generate a versatile and flexible tool, thus SIMMER-III has been applied to safety analyses of various reactor types with different neutron spectra and coolants. Since the three-dimensional representation of the core enables realistic distribution of the materials constituting the core including control rods, the development of SIMMER-IV has been performed to be a direct extension of SIMMER-III to three dimensions with retaining exactly the same physical models as SIMMER-III. Recently, the parallelization of SIMMER-IV has been made to allow applications to reactor calculation within available computational resource. A three-dimensional calculation with SIMMER-IV is presented to indicate more realistic accident scenario.

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The present paper investigates the interaction between a bubbly flow and an obstacle (sphere or cylinder). The goal is to calculate the force exerted by the bubbly flow on a transverse obstacle. To do this, the bubbly fluid is described with the help of the Kogarko, Iordanski and Van Wijngaarden model. The flow is assumed to be potential. The resolution of the motion equations leads to a "wave equation" satisfied by the velocity potential. Coupled with the boundary conditions, we solve this equation in two situations. First, in permanent regime, we prove that the force exerted on the obstacle is equal to zero, generalizing the d'Alembert paradox. Second, we consider the harmonic vibrations of the obstacle. Then, we derive an analytic expression of the force which is the sum of two terms, the added mass and a friction force due to the compressibility of the gas bubbles. More, we prove that this force increases with the void fraction.

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Steam separators consist of a part of reactor core coolant recirculation system in Boiling Water Reactors (BWRs), and have the function to separate water and steam from two-phase mixture out of core. Reduction of the pressure loss of steam separators is advantageous in that it makes the required recirculation pump power lower and hence increases net electrical power output. The development program of a steam separator with more than 20 percent lower pressure loss than that of the current steam separator while keeping steam separation performance has been done. The low pressure loss steam separator was developed by means of the three-dimensional two-phase flow simulations verified by 1/2.22 scale-model experiment. The performance tests for the full-scale steam separator were conducted under actual reactor operating conditions, 7.17MPa and 287 degree Celsius. The pressure loss of the steam separator, carryover water in separated steam and carryunder steam in separated water from steam separator were measured under rated flow conditions. The results demonstrate that pressure loss decreases about 25 percent and almost the same separation performance is achieved compared to the current steam separator at the reactor rated conditions. The reduction of the recirculation pump power as an advantageous effect is appraised in case of applying to the advanced BWR (ABWR).

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The present paper is related to the design of the principal control rod system for the reference large (2400 MW_<th>) Generation IV Gas-cooled Fast Reactor (GFR), which makes use of CERCER plate-type fuel-assemblies with carbide fuel contained within a SiC inert matrix. For the calculations, the deterministic code system ERANOS-2.0 has been used in association with a RZ core model, with particular attention given to the heat generation within the control rods. Based upon the results obtained, a heterogeneous control rod pattern has been developed taking into account both neutronic and thermal-hydraulic constraints. Thus, the rod temperature was computed using the thermal-hydraulic code COPERNIC, for different helium flow rates and heat transfer correlations. It is found that it is necessary to dedicate a significant coolant mass flow rate (〜2.5% of the total for the core) to maintain acceptable cladding temperatures. Then, the control rod worths were computed by applying a methodology based on reactivity equivalence, in order to account fully for the heterogeneity of the control assembly design. Furthermore, the heterogeneity effects between the absorber pins leads to some rather large rod worth reductions in large sodium cooled fast reactors (〜20% in the case of Superphenix, for example). Therefore the design was set up in a manner to minimize this shadowing effect and as a consequence, the explicit consideration of the heterogeneity of the new design leads to a significantly lower rod worth reduction of 〜13%, as compared to the result obtained with the homogeneous model.

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The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment "11% upper plenum break", conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

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