For the investigation of co-current two-phase flows at atmospheric pressure and room temperature, the Horizontal Air/Water Channel (HAWAC) was built at Forschungszentrum Dresden-Rossendorf (FZD). At the channel inlet, a special device provides adjustable and well-defined inlet boundary conditions and therefore very good CFD validation possibilities. The HAWAC facility is designed for the application of optical measurement techniques, which deliver the high resolution required for CDF validation. Therefore, the 8 m long acrylic glass test-section with rectangular cross-section provides good observation possibilities. High-speed video observation was applied during slug flow. The camera images show the generation of slug flow from the inlet of the test-section. Parallel to the experiments, CFD calculations were carried out. The aim of the numerical simulations is to validate the prediction of slug flow with the existing multiphase flow models built in the commercial code ANSYS CFX. The Euler-Euler two-fluid model with the free surface option was applied to a grid of 600,000 control volumes. The turbulence was modelled separately for each phase using the k-ω based shear stress transport (SST) turbulence model. The results compare well in terms of slug formation, and breaking. The qualitative agreement between calculation and experiment is encouraging, while quantitative comparison show that further model improvement is needed.

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The TONUS code incorporates both Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) formulations, is jointly developed by the French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) for hydrogen risk safety analysis. During the cooperation between CEA and Chinese Atomic Energy Authority (CAEA) for severe accident management associated laboratory, the incorporation of a Chinese recombiner model into the TONUS code was carried out and an analysis was performed with this new recombiner model for the Chinese 300MWe Nuclear Power Plant (CNP300). The analysis of hydrogen issue for the CNP300 was performed with two recombiner models, the standard TONUS and Chinese one using several hypothetical accident scenarios. The obtained results show that the Chinese recombiner model results in a lower hydrogen concentration in CNP300 than the standard TONUS one due to its geometrical structure and high efficiency.

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This paper investigates the potential effectiveness of natural circulation-driven coolability (NCDC) as a severe accident mitigative measure. The NCDC can particularly be useful in LWR plants which employ external cavity flooding. The main idea is to provide a simple design solution that facilitates bottom feeding of coolant into the debris bed, and uses steam production in the decay-heated debris bed to drive the two-phase flow natural circulation. We use an analytical one-dimensional model to calculate characteristics of two-phase thermal-hydraulics in porous media. The model employs Lockhart-Martinelli correlations for two-phase flow friction and void fraction, and Ergun's correlation used for single-phase flow resistance. Adaptation and verification of the model are discussed in the paper. Coolability of debris beds with coolant bottom-fed is evaluated for a broad range of conditions. The analysis suggests that the dryout heat flux (DHF) in bottom-fed configurations can be increased by 80% to 160%, when compared to DHF in top-flooding beds.

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Impact loads are considered in nuclear facility design as the result of the loading effects of certain design basis accidents and design basis threats made up of natural as well as man-made hazards. Also, beyond design basis accidents and threats are considered. Typical missiles include objects caused by tornado winds, aircrafts, war or terrorist activities, dropped objects, turbine fragments and other missiles resulting from failure of rotating equipment and whipping pipes and other objects of failure of pressurized fluid systems. The aim of the work is to study the potential of tools for numerical simulations to study the local load effects of airplane missiles impacting concrete structures. Two of the leading commercial computer codes for analysis of highly dynamic events, ABAQUS/Explicit and AUTODYN, have been evaluated. Numerical simulations have been carried out for rigid as well as deformable missiles with the characteristics of airplane engines. The analysis results have been compared with test results from a test program performed in the USA at Sandia National Laboratory and in Japan at Kobori Research Complex and Central Research Institute of the Electric Power industry. Finally, numerical simulations of a large passenger airliner impacting a reactor containment has been carried out using the analysis methodology developed.

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A combined radiochemical separation method has been developed that enables the simultaneous determination of ^<89>Sr, ^<90>Sr, ^<230>Th, ^<232>Th, ^<234>U, ^<235>U, ^<238>U, ^<237>Np, ^<239,240>Pu, ^<238>Pu, ^<241>Am, ^<242>Cm, and ^<244>Cm in medium and low level radioactive wastes. The main steps of the method are sample destruction, co-precipitation on iron(II)-hydroxide and calcium-oxalate, separation by extraction chromatography using supported di-pentylpentyl phosphonate (UTEVA), supported N,N-octylphenyl-di-i-butylcarbamoylmethyl phosphine oxide (TRU) and supported bis-(t-butylcyclohexano)-crown(18,6)ether (Sr Resin), α and β source preparation. The key parameter of the method is the adjustment of the oxidation states of the actinides before adding the sample onto the UTEVA column, this can be done using KBrO_3, NaNO_2 or (NH_4)_2S_2O_8. A method for separation and determination of ^<93m>Nb and ^<94>Nb has also been developed that is based on pre-concentration of the insoluble niobium oxides by precipitation and purification of the fluoric complexes of niobium on anion exchange resin.

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Within the framework of a precautionary investigation performed in cooperation between the German Nuclear Power Plant Emsland (KKE), Forschungszentrum Juelich (FZJ) and the RWTH Aachen University deactivation effects on catalyst sheets of passive autocatalytic recombiners (PAR) were studied. The composition of the fouling and possible age-induced deactivation effects respectively were characterised and correlated with potential sources inside the reactor containment. The paper gives an overview of the project and the results achieved.

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After the development and the assessment of Three-Dimensional (3D) Neutron Kinetics (NK) - 1D Thermal-Hydraulics (TH) coupled codes analyses methods, deterministic nuclear safety technology is nowadays producing noticeable efforts for the validation of 3D NK - 3D TH coupled codes analyses methods too. Thus, the purpose of this work was to address the capability of the RELAP5-3D[○!C] 3D NK - 3D TH code to reproduce VVER1000 Nuclear Power Plant (NPP) core dynamic in simulating the mixing effects that could happen in the vessel downcomer and lower plenum during some scenarios. The work was developed in three steps. The first step dealt with the 3D TH modeling of the Kozloduy-6 VVER1000 reactor pressure vessel. Then this model was validated following a Steam Generator Isolation transient. The second step has been the development of a 3D NK nodalization for the reactor core region. Then the 3D NK model was directly coupled with the previously developed 3D TH model. The third step was the calculation of a Main Steam Line Break (MSLB) transient. The 3D NK global nuclear parameters were then compared with the 0-D results showing a good agreement; nevertheless only the 3D NK- 3D TH model allowed the calculation of each single assembly power trend for this strong NK-TH asymmetric transient.

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An experimental and numerical study was conducted on the pressure loss of flows through the bottom end piece of a nuclear fuel assembly. To determine an optimized numerical methodology using the commercial CFD code, CFX 10.0, a series of preliminary simulations of water flows through perforated plates in a square ducts were performed. A perforated plate is a predominant geometry of the bottom end piece, responsible for the majority of the flow's pressure drop. The numerical pressure loss applying an optimized mesh and the k-ε turbulence model showed good agreement when compared with a conventional methodology (Idelchik). Numerical results for the standard bottom end piece were obtained applying the previously determined mesh criteria and the k-ε turbulence model with some geometric simplifications. The agreement between the numerical simulations and experimental results can be considered satisfactory but suggests further numerical investigations with the bottom piece under real conditions of the experiment, without the geometric simplifications and with a gap between the piece and the wall of the flow channel. Additionally, other turbulence models should be appraised for this complex geometry.

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This paper describes a design process based on risk-informed probabilistic methodologies that cover a facility's life-cycle from start of conceptual design through decontamination and decommissioning. The concept uses probabilistic risk assessments to identify target reliabilities for facility systems and components. Target reliabilities are used in system and subsystem simulation analyses to determine the optimum combination of initial system and component construction reliability, maintenance frequency, and inspection frequency for both active and passive components. The target reliabilities are also used for system based code margin exchange to reduce excessive level of margins to appropriate levels resulting in a more flexible structure of codes and standards that improves facility reliability and cost. The paper includes a description of a risk informed lifecycle design process, a summary of work being done, and a discussion of work needed to implement the process.

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The necessity of multiscale numerical model is explained first to mimic the energy cascade of turbulence in the introduction of this paper. Then, the outline of a multiscale large-eddy simulation code MISTRAL is explained, including the basic equations, turbulence models, numerical formulation and numerical techniques employed. In MISTRAL, both velocity and pressure are split into the large-scale and small-scale resolvable components, and unresolvable components. Each resolvable component is approximated with large-scale and small-scale finite elements. Variational method is used not only to yield the spatially discretized equation system but also to close the turbulence terms resulted from unresolvable components. With such a formulation, multiscale turbulent viscosities can be given and the adequate magnitudes of them may be estimated by the balance of turbulent kinetic energy and model dissipation. The multiscale mass matrix lumping technique is used for an efficient time integration in MISTRAL. As an example of the application, an isothermal flow in a T-junction pipe is analyzed with MISTRAL code. For comparison, numerical results obtained by a conventional LES code are also presented. The 1st and 2nd moments of turbulent velocities as well as the power spectra are computed with these numerical results, to discuss the superiority of MISTRAL in predicting statistical quantities.

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Systems based on the Detect-and-Suppress (D&S) Long Term Stability Solution Option III have been installed in many BWR plants. Such systems issue scram signals upon detecting oscillatory neutron flux signals from Local Power Range Monitors (LPRM) grouped into several Oscillation Power Range Monitor (OPRM) cells. The licensing basis of the existing systems does not address situations leading to highly unstable conditions with multiple interacting instability modes. These conditions are anticipated with flow windows extended beyond the Maximum Extended Load Line Limit Analysis (MELLLA) domain, which are typically associated with power uprate. The Enhanced Option III (EO-III) solution presented in this paper is an evolutionary step relying on the existing methodology and hardware, and introducing measures for addressing the reduced stability associated with extended flow window conditions in general, and the higher probability of single channel hydraulic instability excitation in particular. The new elements introduced as enhancements to the existing Option III solution include the introduction of a calculated exclusion region on the power/flow map, protected by reactor scram, to preclude single channel instabilities. The calculation procedures for supporting the stability protection system are modified consistent with the introduction of the channel instability exclusion region. In particular, the limiting critical power response to power oscillations which is calculated using 3-D system codes, will always be guaranteed to be regular and bounded as a function of the oscillation magnitude when the flow in all the reactor channels oscillate coherently. In addition to the physical basis behind the improvements in the new EO-III, aspects of the engineering application and licensing will be also discussed.

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This paper deals with the control study on a flexible rotor-AMB test rig for the HTR-10GT. In order to improve the performance of system and to pass the second bending critical speed, we use 4 radial active magnetic bearings in this test rig. Through simulations and experiments, it is proved that accessorial AMBs improve the performance of system.

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A High Level Liquid Waste (HLLW) partition process, trialkyl phosphine oxides (TRPO) process has been developed by INET (Institute of Nuclear and new Energy Technology) of Tsinghua University in China to recover the Minor Actinides (MA) and Long-Lived Fission Products (LLFP) since 1980's. The feasibility of TRPO process was proved by hot tests and Pilot-plant scale experiment. HLLW liquid waste arising from spent nuclear fuel reprocessing process contains zirconium, which will deteriorate MAs and LLFPs' recovery from HLLW. Therefore, as first step 0f the TRPO process, removal of Zr is studied. In INET, studies on the adsorption of silica gel for Zr have been carried out. In this paper the progress of silica gel adsorption is reviewed and a technical process is introduced. Purchased silica gel was used to study the adsorption behavior of Zr and other nuclides in HLLW. The result of static adsorption shows that the static capacity of Zr on silica gel is about 20 mg Zr/g in 1.0〜4.0 mol/L HNO_3 solution; the adsorption distribution coefficient of Pu(IV) is about 0.7〜1.4 ml/g; U and Np are not adsorbed on silica gel; among all fission products, besides Zr only a part Mo is adsorbed on silica gel, while other fission products (Fe, Mn, Sr etc.) are not adsorbed. The result of dynamic adsorption shows that the effective capacity of silica gel column is 3.5 times of bed volume; a little Pu is adsorbed on silica gel together with Zr; the adsorbed Zr and Pu in column can be partly eluted by HNO_3; 0.2 mol/L H_2C_2O_4 can be used to elute Zr and Pu completely. In order to reduce the volume of waste silica gel, the regeneration and reuse of silica gel was studied. The 0.2 mol/L H_2C_2O_4 eluted silica gel column can be reused at least six times and the dynamic adsorption property of Zr is almost the same each time. We also studied the measurement of Pu adsorbed on silica gel and found that the radioactivity specific activity of Pu remained on the silica gel after H_2C_2O_4 elution is about 1×10^4 Bq/kg, which means the waste silica gel belongs to non-α-emitting waste. After the removal of Zr by the technical process introduced in this paper, HLLW can be treated with TRPO process. The final eluted Zr/Pu/H_2C_2O_4 mixed solution can be solidified as medium radioactivity solid waste by cement solidification method.

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This study is a contribution to the 2004 PCR-RSF program of the Centre National de la Recherche Scientifique (CNRS) devoted to research on high temperature thorium molten salt reactors. A major issue of high temperature molten salt reactors is the very large heat duty to be transferred from primary to secondary loop of the reactor with minimal thermal losses. A possible inner loop made of a series of conventional graphite filter plate exchangers, pipes and pumps was investigated. The loop was assumed to use two counter current flows of the same LiF, BeF_2, ZrF_4, UF_4 molten salt flowing through the reactor. The 3D model used the coupling of k-ε turbulent Navier-Stokes equations and thermal applications of the Heat Transfer module of COMSOL Multiphysics. For a reactor delivering 2700 MWth, the model required a set of 114 identical exchangers. Each one was optimized to limit the heat losses to 2882 W. The pipes made of a succession of graphite, ceramics, Hastelloy-N alloy and insulating Microtherm^[○!R] layers led to a thermal loss limited to 550 W per linear meter. In such conditions, the global thermal losses represent only 0.013 % of the reactor thermal power for elements covered with an insulator only 3 cm thick.

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Containment mechanical behavior characteristics are one of the important checking items during 900MW PWR containment test. This article describes these characteristics (including strain, displacement and temperature influence) systematically. Especially some intrinsic characteristics are discussed and some new viewpoints are provided. We believe that it has some reference value to other containment test.

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The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D[○!C]. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions, for 2.0% and 2.4% enriched fuel. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D[○!C] calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global reactivity versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment.

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The aim of this study is to obtain insights into the flow behavior, as well as to develop predictive capability with regards to the flow and thermal mixing, that occurs in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In this paper, numerical modeling has been used to capture qualitative phenomena observed during an experiment performed at INL, using a finite volume, thermo-fluid solver system, 'SC/Tetra' from CRADLE[○!R]. The choice of the correct turbulence model is critical to accurately predict the flow in the VHTR lower plenum. Four different turbulence models have been used in this study and the flow predictions are significantly different. A trail of marker particles and fluid temperature as a passive scalar have been used to qualitatively study the flow characteristics, specifically the turbulent mixing of water jets. The quantitative experimental data, when available, will be used to compare and improve on the available turbulence models. Preliminary numerical modeling has been carried out to address the issue of hot streaking and buoyancy effects of hot helium jets in the lower plenum.

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Stress Corrosion Cracking (SCC) test was conducted by Creviced bent beam test on type 316L and type 316 stainless Steels (SS) plate specimens machined from the longitudinal / transverse (LT), the transverse / short-transverse (TS) and the longitudinal / short-transverse (LS) planes. Each specimen was cold rolled by 10% to 30% just before machining to the final shape. After exposure to the test environment for 1000 h, each specimen was observed by a scanning electron microscope to measure SCC length on the specimen surface and SCC depth on the fracture surface. It was concluded that SCC susceptibility was higher in the LS and TS planes than the LT plane for both type 316L and type 316 SS, although the susceptibility of type 316L SS was much lower than type 316 SS. The SCC length conformed to the log-normal distribution irrespective of SS types and the cold rolling reduction. Thus a statistical analysis could be applied by assuming the distributions when a life prediction of SS components would be necessary.

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The Active controlled Magnetic Bearing (AMB) is used in the system of 10 MW High Temperature Gas Colder Reactor (HTR-10) plant with direct helium turbine cycle as its special strongpoint. Appropriate magnetic material and its working point are keys to get a well performed AMB, which is in need of parameters of magnetic material. This paper tests the dynamic properties of several soft magnetic alloys including silicon steel lamination with different thickness, permalloy and iron-cobalt alloy in different conditions. This paper studied the properties of soft magnetic alloys and chose bearing and sensor materials for AMB used in HTR-10 project, and then appointed the working points. Those data are the foundation to optimize the configuration of bearings and sensors and gain an AMB with excellent properties.

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The current methods used in the U.S. PRAs to calculate the loss of offsite power (LOOP) initiating event frequency and offsite power non-recovery probability do not adequately account for the plant-specific considerations for estimating these parameters. This is inconsistent with the ASME standard requirements for internal events PRA for risk-significant model elements. This paper presents a realistic approach to the analysis of the LOOP initiating event frequency and offsite power non-recovery probability. The U.S. industry LOOP events are collected and reviewed for applicability to the specific plant being analyzed. The initial random failures in the chain of events that eventually lead to the LOOP at other U.S. plants are combined with the switchyard design and operations of the specific plant being analyzed to determine if the same or equivalent failures would result in a LOOP event. Only those applicable events are then used for the analysis of the LOOP initiating event frequency. Similarly, the industry generic LOOP events are screened to determine if the recovery duration in the generic event is applicable to the plant being analyzed. Generic restoration duration may sometimes be adjusted for use under the emergency conditions. It is believed that the use of this plant-specific approach to refine and process the raw, LOOP events would produce much more realistic LOOP event estimates and more accurate PRA results.

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This paper introduces a method for calculating the depletion of Li/Be and inventory of tritium. A separated Be/Li2TiO3 layers blanket concept model was setup for depletion analysis. This work presents the neutronics design optimization analyses conducted as part of the development work for the modular cross cooled pebble bed (CCPB) blanket. The main objective has been to assess and optimize the tritium breeding performance to ensure tritium self-sufficiency and provide the data required for the thermal hydraulic blanket layout. After a long period of operation in fusion neutron environment, the depletion of both lithium and beryllium are of considerable effects to neutron and nuclear heat distributions. Reaction rate modification method is used to calculate lithium and beryllium depletion fractions, which can reduce the offset induced by neutron spectrum changing in the ceramic blanket. The effect of Lithium produced from beryllium is also seriously considered in the depletion calculation. Mean tritium breeding ratio (MTBR) is more suitable to represent the tritium breeding performance averaged over a period of operation. Different arrangements of ^6Li enrichment were studied in this paper. Results show that lower ^6Li enrichment of breeder in the front layers will lead to high depletion fractions, and considerable drop of MTBR can reach up to 4.93% even a packing factor of 60% is considered in breeder zones after 5 years operation. The more beryllium and the suitable higher the ^6Li enrichment of breeder placed in the front layers behind the first wall, the performance on MTBR would be better and the lower depletion fraction of lithium is achieved.

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In the present work, various national electricity generating systems associated with conventional as well as renewable energy resources are comparatively assessed in view of life-cycle multi-criteria (economic, environmental, health, and social) spaces. The essential objectives of the study are (1) to comprehensively compare the options for an electricity supply, (2) to complementarily support nuclear power's role in a national energy sector, and (3) to contribute to sustainability-oriented research and development in the energy and power sectors. Here, various national power sources including conventional as well as renewable energy systems are comparatively assessed in view of multi-criteria decision-making (MCDM) spaces. Previous MCDM approaches for energy mix policies are mostly based on risk factors or environmental factors. In the ExternE project, environmental aspects are quantified from the point of view of an externality of an energy development cycle. National energy mix policies of individual countries are still based on economic points such as power generation cost, fuel import cost, land availability, etc. In this paper a multiple aspects approach for making decisions on the selection energy generation technologies is considered. The framework of the decision making process for the energy mix alternatives in this study considered the environmental aspects, health aspects, risk aspects, social aspects, and economical aspects collectively. The AHP (analytical hierarchy process) is considered in this paper and it is demonstrated through an example work for an energy mix alternatives framework. Power source alternatives under consideration are the conventional systems such as nuclear and fossil-fuelled (coal-fired, heavy oil-fired, LNG) as well as the new and renewable energy systems (hydropower, wind power, solar photovoltaic (PV) power). These seven options are evaluated in terms of several conflicting criteria representing the generation cost, land use, global warming, etc. As a demonstration of this approach, four main criteria (Level 1) and eleven sub-criteria (Level 2) spaces are chosen after other previous work was reviewed. From an integrated point of view, overall preference of the power sources can be summarized as follows: Nuclear &sc; Wind &sc; PV &sc; Hydro &sc; LNG &sc; Oil &sc; Coal. From an integrated viewpoint of the economical, the environmentally-friendly, the socially-acceptable, and the healthy aspects, nuclear power takes first place. Renewable energy sources (i.e., PV, wind, and hydro powers) are in second place. The last one is held by the fossil-fueled power sources (i.e., LNG, heavy oil, and coal).

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For fusion material research, minichannel gas flows are designated to cool irradiated material specimens. Since the facility design requires accurate predition methods for the temperatures in the structure and the specimens, relevant experiments were conducted. This paper reports on the experimental procedures, which are specific to the small scale of the channels, and elucidates results obtained for wall friction, heat transfer and velocity profiles. Comparisons between the experimental data and engineering correlations as well as CFD calculations are presented. These comparisons reveal good accordance of the presented minichannel data with classical engineering correlations, and indicate the fitness of the v2f turbulence model for numerical flow field predictions in the scope of the considered application.

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This study investigates the transmutation properties of a critical Molten Salt Reactor (MSR) and compares them against those of three types of solid fuel reactors - Lead cooled Fast Reactor (LFR), Sodium cooled Fast Reactor (SFR) and a PWR. A consistent comparison was made of the effect of the different reactor spectra. It was found that the fast reactors spectrum gives the best transmutation performance followed by the MSR and PWR spectra. A comparison of the fractional transmutations (FT) for an infinitive recycling of actinides (Ac) in the different reactors with a 0.1% loss fraction during reprocessing shows that the MSR has the highest FT due to its high specific power, followed by the SFR and the LFR. Taking into account FT and spectra differences the MSR has preferred transmutation capability.

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This paper deals with the development of a coupled PARCS/RELAP5 model of the Swedish Ringhals-3 pressurized water reactor. The standalone PARCS and RELAP5 models are first presented. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. On the thermal-hydraulic side, each of the 157 fuel assemblies is modelled. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes. Validation of the model against measured plant data at steady-state conditions is then summarized. Comparisons between calculated/measured parameters demonstrate that the model is able to correctly represent steady-state conditions of the plant. Finally, the validation of the model against measured transient plant data is described. The transient chosen for this validation task was a load rejection ("house-load") transient, which occurred on January 8, 2005.

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The void fraction of gas-liquid two-phase flow in mini channels was measured by neutron radiography. The cross sectional averaged void fraction results were compared with the drift flux models based on the flow patterns predicted by the Mishima-Ishii's flow pattern transition criteria. The drift flux model by Ishii predicted well the experimental results for relatively large diameter tubes. In small diameter tubes, the void fraction was predicted by Ishii's model except for the annular flow region with alpha < 0.8. From these results, a new flow regime transition of churn to annular flow with α = 0.8 was proposed. The void fraction in annular flow region with α <0.8 was predicted well by Ishii's drift flux model. The drift flux model by Mishima-Hibiki predicted the void fraction results for various diameter tubes except large void fraction in annular flow region.

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Human beings solve problems according to their experiences. Case-based reasoning (CBR) is a problem solving method which simulates the process of problem solving of human beings. CBR is applied various problems in engineering field. However, CBR has not been applied to the LP optimization problem so far. In this study, CBR is applied to LP optimizations and its performance is compared with Simulated Annealing (SA) and Tabu search (TS). From the results, it is demonstrated that CBR optimizes a LP faster than SA and TS. The performance of a LP that is obtained by CBR is slightly worse than that of SA and TS, but the difference of performance of LP among CBR, SA and TS is small. This indicates that CBR is a very effective method for fast and reasonable LP optimization.

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From a conflicting viewpoint, comprehensive assessment of various national power systems can be treated as a multicriteria decision-making (MCDM) problem. In reality, there are interaction phenomena among the decision elements. The main objective of this work is to propose a comprehensive framework to determinate the priority of appropriate national power sources involving various degrees of interaction among the decision elements (e.g., decision goal, decision criteria, and decision alternatives) such as inner dependence, outer dependence, and feedback effect. In the context of a generic hierarchical network (or hiernet) structure instead of one-way directional tree structure, the impact of the interaction phenomena on the grade of priority is investigated using a supermatrix technique or an analytic network process (ANP) method. Moreover, the three types of attitudes towards nuclear power system of the multiple actors are incorporated into the network structure to figure out the effect of characteristics of power systems. An illustrative example of the generic hiernet structure is demonstrated in comparison to the specific hierarchy structure without any interaction among the decision elements. The proposed framework can be applied to select the appropriate power systems, to understand the effect of its underlying decision structures, and to include risk attitudes towards a certain alternative.

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In decommissioning process of nuclear facilities, large amount of radioactive isotopes are discharged as waste. Radioactive carbon isotope (^<14>C) is one of the key nuclides to determine the upper limit of concentration in the waste disposal. In particular, ^<14>C on the graphite reactor decommissioning should be separated from stable carbon isotopes (^<12>C and ^<13>C) and monitored for the public health and safety. We propose an isotope analysis system based on cavity ring-down laser spectroscopy (CRDS) to monitor the carbon isotopes (^<12>C, ^<13>C and ^<14>C) in the isotope separation process for the graphite reactor decommissioning. This system is compact and suitable for a continuous monitoring, because the concentration of molecules including the carbon isotope is derived from its photo absorbance with ultra high sensitive laser absorption spectroscopy. Here are presented the necessary conditions of CRDS system for ^<14>C isotope analysis through the preliminary experimental results of ^<13>C isotope analysis with a prototype system.

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TiO_2 addition into boiling water reactor (BWR) primary system is being developed as a method to mitigate stress corrosion cracking (SCC) of the BWR structural materials. This technique aims for electrochemical corrosion potential (ECP) decrease of reactor materials by photo-excitation reaction under Cherenkov irradiation. Tests have been conducted in the test loop in both BWR and OECD Halden reactor to investigate the feasibility of the SCC mitigation method with TiO_2. The test results showed that the ECP of TiO_2 deposited materials was decreased to <-0.3V(vs.SHE) under both UV light irradiation in the BWR reactor water normal water chemistry (NWC) environment and in-core Cherenkov irradiation in the Halden BWR simulation loop under higher dissolved oxygen condition. This TiO_2 technique was confirmed to be feasible as a SCC mitigation method for BWR structural materials.

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SUS 304 stainless steel has been used for internal components of the light-water reactors constructed in earlier days, in which irradiation-assisted stress corrosion cracking has drawn attention. SUS 316L was regarded as less susceptible to the stress corrosion cracking and adopted in the newer reactors, although it was recently found not completely so in some Japanese reactors. Tensile residual stress introduced during manufacturing is believed to be one of the major factors in the initiation of cracking. It is, therefore, essential to assess the stress relaxation behavior under irradiation, which can be evaluated from the irradiation creep data, in conjunction with the effect of cold work. In the present study, creep experiments under 17MeV proton irradiation (2 × 10^<-7> dpa/s) at 288℃ were conducted for both SUS 316L and SUS 304 stainless steels with 5% or 25% cold work. Stress dependence of the irradiation creep was quite different between the two stainless steels, almost linear in SUS 316L while nearly quadratic in SUS 304 for both levels of cold-working. Stress relaxation under irradiation was evaluated to be different between the two stainless steels, reflecting their difference in stress dependence.

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It is a time-consuming and error-prone task to prepare neutronics model for the discrete ordinates transport codes (S_N codes) in manual way. A more efficient solution is presented in this paper, which shift geometric modeling to computer aided design (CAD) system, and to use an interface program for S_N codes to convert the CAD model to neutronics model, and then generate the input file of S_N code automatically. The detailed conversion method is described and some kernel algorithms are implemented in SNAM, an interface program between CAD system and S_N codes. The method has been used to convert the ITER benchmark model to the input file of S_N code successfully. It is shown that the conversion method is a correct, efficient and potential solution for S_N code modelling.

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The digital safety instrumentation and control (I&C) system has been developed and approved in Japan. Based on this proven technology, the digital platform will be also applied to the I&C system in US-APWR. The fully computerized system provides more benefit such as reduction of operator's load, potential for human error and maintenance workload, and improvement of reliability and availability. In addition, the I&C system in US-APWR has some specific features to meet U.S. requirement. For example 4 train systems is completely applied for safety and safety related systems, and the other remarkable feature is that non-safety consoles are used for control of safety systems and components, which is generally called multi-channel operation. Thus this system is slightly improved and modified from the previous digital system architecture, and the approved digital platform is the same in Japan. In this work, we provide the system description and design basis for US-APWR and some discussion for the specific design features.

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Non-destructive testing (NDT) techniques for reactor internals and recirculation piping are essential to keeping integrity of boiling water reactors (BWRs). Inspections of shroud and piping of BWRs have been applied and detection and sizing of stress corrosion cracks (SCCs) have been done in field applications. Recently an inspection system for the shroud support welded to a reactor pressure vessel (RPV) has been developed. Inspection of shroud supports is very important and challenging because the weld joints are pressure boundaries consisting of Ni-based alloys through which it is difficult to propagate ultrasonic beams because of the large anisotropic grains. In this paper, a new inspection system is introduced using point-focusing phased array ultrasonic testing (UT) sensors which provide access from the outside surface of the RPV. The UT detectability and sizing accuracy of the system are confirmed by experiments using mock-up test pieces of shroud supports.

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Two methods of applying Fracture Research Institute Theoretical Stress Corrosion Cracking model (FRI model), the theoretical SCC model developed by Shoji and Suzuki, to SCC growth analysis in a cylindrical component with stress distribution are proposed. FRI model has rarely been adapted to estimate SCC growth behavior in a real structural component in which the changing rate of stress intensity factor with time (dK/dt) naturally changes as the crack grows. One of the reasons might be the difficulty of solving the equation consisting of many parameters including dK/dt. In this paper, the authors propose two methods to conduct SCC growth analysis using FRI model. In Method 1 (Once-through method), the authors successfully applied FRI model to SCC growth analysis of 360 degree circumferential crack in a cylinder under the stress distribution by replacing dK/dt with the changing rate of stress intensity factor with crack length (dK/da). Method 1 is so easy that it can be achieved only using Worksheet functions of EXCEL. Method 2 (Iteration method) is relatively difficult solving the complex equation at each crack growth step. The authors performed Method 2 using originally developed codes with VBA, Visual Basic for Applications, for EXCEL. Method 2 is applicable to two-dimensional crack (semi-elliptical crack) contrary to Method 1 that is only for a one-dimensional crack. The details of Method 1 and Method 2, an example of application of FRI model to SCC growth analysis and the parameter study for FRI model are shown in this paper.

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Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test that simulates its design is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear rectors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method: TPFIT is developed to get the detailed information of the two-phase flow. We tried to verify the TPFIT code comparing with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.

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Noise analysis is applied to the plant data of the commercial BWRs in Japan to evaluate the validity of noise analysis for plant diagnosis. The plant diagnosis has successfully detected and distinguished two cases of anomalies; partial blockage of a sensing line for main steam flow and breakage of a sensing line for jet-pump flow.

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The preliminary conceptual design study of prismatic-type Very High Temperature Reactor (VHTR) has been performed with 950℃ outlet coolant temperature for higher efficient hydrogen and electricity production. First, the core internals that enable higher outlet temperature are considered in the viewpoint of reduction of core bypass flow. Three-dimensional thermal and hydraulic analyses are carried out and show that the 950℃ outlet temperature requires approximately 90% fuel flow fraction and it can be achieved with the installation of the seals in bottom blocks, the coolant tubes in the permanent side reflectors and the core restraint devices. Next, the core and fission product release analyses are performed. The analysis methods that have been developed for the pin-in-block fuel, one type of prismatic VHTR cores, can be applied to multi-hole fuel, another type of the cores, with some adjustments of the analytical models.

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Design studies of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) are being carried out at the Japan Atomic Energy Agency (JAEA) as one candidate for the future reactors. In actual core design, it is precondition to prevent fuel rods contact due to fuel rod bowing. However, the FLWR cores have nonconventional characteristics such as a hexagonal tight lattice arrangement and a high enrichment fuel loading. Therefore, as conservative evaluation, it is important to investigate influence of fuel rod bowing upon the boiling transition In the JAEA, a 37-rod bundle experiments (base case test section (1.3mm gap width), gap width effect test section (1.0mm gap width), and rod bowing test section) were performed in order to investigate the thermal hydraulic characteristics in the tight lattice bundle. In this paper, the rod bowing effect test is paid attention. It is suspected that the actual fuel rod positions in the rod bowing test section may be different from the design-based positions. Even a slight displacement from the design-based position of fuel rod may occur variation of flow area, and give influence upon the thermal hydraulic characteristics in the rod bundle. Therefore, if the critical power in the rod bundle is evaluated by an analytical approach, the analysis based on more correct input can be performed by using actual fuel rod position data. In this study, the rod positions in the rod bowing test section were measured using the high energy X-ray computer tomography (Xray-CT). Based on the measured rod positions data, the subchannel analysis by the NASCA code was performed, in order to investigate applicability of the NASCA code to BT estimation of the rod bowing test section, and influence of displacement from design-based rod position upon BT estimation by the NASCA code. . The predicted critical powers are agreement with those obtained by the experiment. The analysis based on the design-based rod positions is also performed, and the result is compared with the analysis based on the rod positions measured by the Xray-CT. As the result, the calculated result based on the design-based rod positions gives close agreement with the result based on the rod positions measured by the Xray-CT. This indicates that, in the rod [figure] [figure] Fig. 1 The rod bowing effect test section [figure] Fig. 2 Imaged position by the Xray-CT [figure] Fig. 3 CT picture (Z=880mm) bundle with rod bowing, the effect of rod bowing gives larger influence upon boiling transition than the effect of displacement of the measured rod position, From above, it is conclude that NASCA code can be applicable to thermal hydraulic analysis in the rod bundle with rod bowing, and that the displacement from the design-based rod position is hardly influenced upon the BT estimation by the NASCA code.

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In the course of a severe reactor accident, molten core will drop into water as a coherent jet and the jet will be broken into droplets. Both the melt jet and droplets are surrounded with a vapor film because of high temperature of the melt. The existence of this vapor film could affect both the jet breakup behavior and movement characteristics of a melt droplet. The work of this paper is to perform numerical simulation of interfacial phenomena to investigate the micro-physics of these multiphase interactions. A computer code using Volume-of-Fluid (VOF) method to track interface is established for this purpose. Some special strategy is used to deal with the rapid vaporization on the liquid-vapor interface. Simulation of film boiling around a hot static sphere in a flowing field using body fitted coordinate is performed in order to give some insight into the premixing phase of a steam explosion. The drag force exerted on the sphere with vapor film in the whole course is analyzed. The code is extended to treat with three fluid representing melt, vapor and volatile liquid to simulate high temperature jet entering to volatile liquid. The leading edge behavior due to boundary layer stripping is described by the code.

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The inter-face tracking method and the particle interaction method can simulate detailed two phase flow in the rod bundle without any physical model based on experiment. However, evaluation of liquid films and small scale bubbles requires high resolution. Therefore, it is difficult to apply such methods to two phase flow analysis in the rod bundle, because of huge computational cost. [figure] [figure] Fig.1 Core design of the FLWR On the other hand, the two fluid model can simulate two phase flow less computational cost than the inter-face tracking method and the particle interaction method. Therefore, the two fluid model is useful for thermal hydraulic analysis in the rod bundle Japan Atomic Energy Agency (JAEA) develops three dimensional two fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system in order to simulate complex shape channel flow, and which can divide the subchannel in the rod bundle into more than one block, and calculate with interaction between blocks. In order to simulate the two phase flow in the large scale domain such as the rod bundle, large number of meshes are necessary. Therefore, parallelization based on Message Passing Interface (MPI) is introduced to ACE-3D for the purpose of calculation time reduction. In order to investigate the parallelization efficiency, the computation with about 9 million meshes has been performed. It is confirmed that parallelization efficiency over 80% can be obtained with 252 CPU. In order to confirm that ACE-3D can be applied to the thermal hydraulic analysis in the rod bundle, the two phase flow analysis in the channel which simulates the rod bundles is performed. The obtained results are agreement with experimental result qualitatively. Therefore, it is concluded that ACE-3D can be applied to the thermal hydraulic analysis in the rod bundle.

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In order to estimate inert gas behavior in a circulating system of a sodium-cooled fast reactor (SFR), it is necessary to develop a computational code for a dynamics of the gas in the primary system. In the present study, multi-dimensional analysis in the upper plenum of the SFR is performed using a numerical method for gas bubble transportation developed based on one-way-coupling method. As a result of the analyses, non-dimensional correlation model for gas behavior has been derived. The model is employed in the VIBUL code that is a plant dynamics code based on one point approximation. In order to investigate the usefulness of the present model in the plant dynamics, analyses in the rated condition are carried out. We discuss the difference of the two models in predicting the bubble number density in the upper plenum.

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In the pressurized-water reactor (PWR) and the boiling water reactor (BWR) power plant, The anticipated transients without scram (ATWS) accident does possibly happen, so the containment (pressure-resistant airtight space) is used as an important safety barrier. At the same time, for assuring the structure integrity of containment in a severe accident, a filtered vented containment system (FVCS) is usually mounted. But in the high temperature gas-cooled reactor (HTR) power plant, the functions of spherical fuel elements have been equivalent to two barriers, the first is the coated particles, the second is the graphite shell. The pressure vessel (pressure-resistant airtight space) is the third safety barrier. In addition, the ATWS accident will not occur in HTR, so the vented confinement is adopted instead of the containment. For the heating, ventilating and air conditioning (HVAC) systems in a reactor, the containment of Qinshan Phase I nuclear power plant includes five sub-systems and the configuration is complex, but the confinement of HTR-10 built in Tsinghua University has only the safety class negative pressure exhaust system rather than the supply air system. By comparison, the vented confinement design brings to a great advance and simple in designing and building HTR.

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With nearly 30 year's development and improvement, spent fuel dry storage (SFDS) has become a mature technology and the quantities being placed into dry storage are increasing significantly. MACSTOR module is one of the latest SFDS design developed by AECL, and it is selected to be applied for Qinshan Phase III CANDU-6 PHWR NPP (TQNPC) which is the first time to apply interim spent fuel dry storage technology in the commercial NPP in China. Natural cooling performance has been analyzed with FLUENT 3D computational fluid dynamics (CFD) analysis code for MACSTOR SFDS. Sensitivity study have also been performed to determine the effect of internal geometric parameters on the maximum temperatures of internal wall and outlet air, including the air outlet size and the distance from the cylinder to either the wall or to the adjacent cylinder, etc. The analysis results are provided and it further indicates that MACSTOR-400 design could be applied to TQNPC.

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The flexural properties and fracture toughness of several SiC nanowire/fibers reinforced SiC matrix composites were investigated. The volume fractions of the SiC nanowires in the composites are 0, 1.6, 5.7 and 6.1%, respectively. The results show that SiC nanowires can be very effective reinforcement materials in SiC/SiC composites. The reinforcement efficiency depends on the carbon coating thickness on the nanowires.

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The geological environment has spatially heterogeneous characteristics with varied host rock types, fractures and so on. In this case the generic disposal tunnel layout, which has been designed by JNC, is not the most suitable for HLW disposal in Japan. The existence of spatially heterogeneous characteristics means that in the repository region there exist sub-regions that are more favorable from the perspective of long-term safety and ones that are less favorable. In order that the spatially heterogeneous environment itself may be utilized most effectively as an NBS, an alternative design of disposal tunnel layout is required. Focusing on the geological environment with spatially heterogeneous characteristics, the authors have developed an alternative design of disposal tunnel layout. The alternative design adopts an optimization approach using a "variable disposal tunnel layout". The optimization approach minimizes the number of locations where major water conducting fractures are intersected, and maximizes the number of emplacement locations for waste packages. This paper will outline the variable disposal tunnel layout and its applicability.

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A thermal-hydraulic design of the high-conversion (HC) type core of the innovative water reactor for flexible fuel cycle (FLWR) was constructed. HC-FLWR is required to proceed to the breeder type of FLWR with no change of any reactor systems. Although tightness of the fuel pin arrangement is significantly different between the two types of cores, the natural circulation cooling is adopted in both cores. TRAC analyses were performed under the condition that chimney length for natural circulation and the setting of the inlet orifice were common to the both types of cores. Form loss coefficients of lower tie-plate were differently set to control the natural circulation flow rate and the feed water temperature were adjusted to realize preferable value of average void fraction of HC-FLWR core. The analyses showed that both types of the FLWR could be cooled by the same reactor system.

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In order to ensure the sustainable energy supply in the future based on the matured light water reactor (LWR) technologies, a concept of innovative water reactor for flexible fuel cycle (FLWR) has been investigated in JAEA. The concept utilizes the tight-lattice core loaded with the plutonium mixed oxide (MOX) fuel, and consists of two steps in the chronological sequence. The first step realizes a high conversion type one (HC-FLWR), which is basically intended to keep the smooth technical continuity from the current LWR / MOX-LWR technologies. The second represents the reduced-moderation water reactor (RMWR) concept, which realizes a high conversion ratio over 1.0 and is preferable for the long-term sustainable energy supply through plutonium multiple recycling. In the present paper, investigation results on the FLWR conceptual design are presented mainly from the neutronics point of view. The design of the HC-FLWR core has been recently improved, and detailed core properties have been evaluated by neutronics and thermal-hydraulics coupled calculations. The core can achieve the average burn-up around 55GWd/t as well as the negative void reactivity coefficients. The size of its fuel assembly is the same as in RMWR, and hence, it can be replaced with the highly tight one for RMWR.

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The study of the effects behind the degradation of components and materials is becoming increasingly important for the safe operation of aged plants especially when it comes to life-extension. Since the Russian nuclear community began to examine life extension issues nearly fifteen years ago, there is much to learn from these Russian pioneering studies, a portion of which were performed under the TACIS (Technical Assistance for Commonwealth of Independent States) international collaboration program with EU countries. At the Ninth International Conference, recent data were introduced regarding the ageing effects of mechanical properties of various kinds of steel and the welding joints of Russian NPP components. The full title of the conference was Material Issues in Design, Manufacturing and Operation of Nuclear Power Plants Equipment. The meeting was organized by the Central Research Institute of Structural Materials (CRISM) "Prometey" in cooperation with the IAEA and JRC-EU. In reviewing the recent data presented at the Ninth Conference, the authors think that the paradigms of structural integrity issues in aged plants are now reasonably well established in (1) fracture mechanics and irradiation hardening of reactor vessels and core internals and (2) thermal ageing and annealing effects. Yet even when considering these well established paradigms, the current direction of study is not adequately addressing the effects of a corrosive environment. The first author believes that the current approach of low cycle fatigue is far from able to prevent and predict environmentally assisted cracks. This fundamental flaw stems from design codes, which do not incorporate the basic knowledge of corrosion mechanisms. Our focus in researching aged plants should be re-directed toward environmentally assisted cracking, typically the film rupture-slip dissolution mechanism of crack propagation under the effect of long cell action on local cells, as discussed by the first author in the other article which will also be presented at this conference (ICONE15-10559). The local cell action corrosion hypothesis is a practical application of the more fundamental theory of corrosion mechanism (called cathodic de-polarization theory) developed by a Russian academician Akimov as early as in 1945. Without finding a way to mitigate this corrosion process, the first author believes that the most current mechanical approaches towards the assurance of structural integrity are basically flawed as they do not take into consideration the electrochemical processes of corrosion.

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Vibration-based packing of sphere-pac fuel is a key technology in a nuclear fuel manufacturing. In the production process of sphere-pac fuel, a Mixed Oxide (MOX) fuel is formed to spherical form and is packed in a cladding tube by adding a vibration force. In the present study, we have developed a numerical simulation method to investigate the behavior of the particles in a vibrated tube using the Distinct Element Method (DEM). In general, the DEM requires a significant computational cost. Therefore we propose a new approach in which a small particle can move through the space between three larger particles even in the two-dimensional simulation. We take into account an equivalent three-dimensional effect in the equations of motion. Thus it is named pseudo three-dimensional modeling.

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