Thermal aging embrittlement of LWR components made of stainless cast (e.g. CF-8 and CF-8M) is a potential degradation issue, and careful attention has been paid on it. Although welds of austenitic stainless steels (SSs) have γ-δ duplex microstructure, which is similar to that of the stainless cast, examination on thermal aging characteristics of the SS welds is very limited. In order to evaluate thermal aging behavior of weld metal of austenitic stainless steel, the 316L SS weld metal has been prepared and changes in mechanical properties and in etching properties at isothermal aging at 335°C have been investigated. The hardness of the ferrite phase has increased with aging, while the hardness of austenite phase has stayed same. It has been suggested that spinodal decomposition has occurred in δ-ferrite by the 335°C aging. The etching rates of δ-ferrite at immersion test in 5wt% hydrochloric acid solution have been also investigated using an AFM technique. The etching rate of ferrite phase has decreased consistently with the increase in hardness of ferrite phase. It has been thought that this characteristic is also caused by spinodal decomposition of ferrite into chromium-rich (α') and iron-rich (α).

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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 on Nuclear Power Plant (NPP) components. The aim of this paper is to evaluate the behaviour and the structural response in form of 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 with reference to solutions considered for a Near Term generation Nuclear Power Plant. The seismic input motion was considered as a free field response spectrum with 0.2 g 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 NPP facilities. The project's major steps and 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, 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-SGs etc.) 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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A defect behavior is one of the principal concerns to be dealt with for a structural integrity since defects may lead to the failure of a SFR(Sodium-cooled Fast Reactor) structure under high temperature loading conditions. A thin-walled and large diametric piping is adopted for the IHTS (Intermediate Heat Transport System) piping system of a SFR to enhance the system's economy and to reduce the thermal stress level. The structural material for the piping is Type 316 austenitic stainless steel. In this study, a defect growth evaluation of IHTS piping was performed for a semi-elliptical surface defect subjected to a combined mechanical loading and thermal loading of a high temperature above 500°C including a cyclic loading condition corresponding to a reactor refueling cycle. The creep-fatigue defect behavior evaluations were carried out by following the French Assessment Procedure A16, the UK Assessment Procedure R5 and the JNC procedure. The results were compared with each other and their strong and weak points were discussed. Also the degree of conservatism of each procedure was reviewed.

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Since the structural integrity of thin-walled tubes in the heat exchanger is crucial from the viewpoint of safety and reliability, the integrity evaluation for flawed tubes is quite important. Accurate estimation of the failure pressure is a key element of the structural integrity assessment. With regard to the prediction of the failure pressure, most of preceding researches have been focused on the limit load approach. However, the integrity assessment scheme based on the elastic plastic fracture mechanics concept has not been settled despite of its accuracy and efficiency. In this paper, three-dimensional finite element analyses assuming elastic plastic material behavior are carried out for the thin-walled tubes with various sizes of the circumferential flaws. As for the flaw location, both the top of tube sheet and transition regions are considered. The flaw instability is evaluated by comparing the driving force with the fracture toughness of the tube material. Analysis results show that the elastic plastic fracture mechanics approach accurately predicts the failure pressures compared to the experimental data. Thus, it is thought that the elastic plastic fracture mechanics concept can be applied to the integrity assessment of the heat exchanger tubes with the circumferential through-wall flaws.

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The primary role of the grid springs in a spacer grid is to hold the fuel rods in an appropriate position using a friction force and to prevent the fuel rods from dropping during a normal reactor operation. Spring force decreases as the fuel burn-up increases since the spring stiffness is degraded by the high temperature and irradiation effects in a reactor core. So this phenomenon has to be considered when the initial spring force of a grid spring is being determined. To check whether a spring has a suitable spring force, characterization tests of a spring are conducted. In this paper, an analytical verification work, a FE model using contact definitions, is established for predicting the spring stiffness without any spring test. Finally, the test and analysis results are carefully compared to check the availability of a finite element model for investigating the spring characteristics at an in-grid boundary condition.

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Spacer grid which is one of the most important structural components in a pressurized light water reactor fuel assembly supports the fuel rods laterally and vertically. Based on design experiences and by scrutinizing the design features of advanced nuclear fuels and the international patents of spacer grids, KAERI has devised its own spacer grid shapes and acquired patents. In this study, a performance evaluation on two new spacer grid shapes devised by KAERI was carried out from mechanical/structural and thermohydraulic view points. And also a performance evaluation on two commercial spacer grid shapes was carried out for the sake of a comparison. The comparisons included the spring characteristics, fuel rod vibration characteristics, fretting wear resistance, impact strength characteristics, CHF enhancement, and pressure drop level of the spacer grid shapes. The comparison results have shown that the performances of the new spacer grid shapes are better or at least not worse than those of the commercial spacer grid shapes.

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A non-linear vibration model (called VITRAN for VIbration TRansient Analysis-Nonlinear) of the dynamic response of a nuclear fuel rod and its supports has been developed and integrated to a fretting-wear analysis method to predict the performance of fuel assemblies. The approach includes the hydraulic, structural and tribological effects considered to be of sufficient importance. A general description of the software and an example of application are provided in this paper.

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Radioactive iodine is one of the most important nuclides to be prevented for release from nuclear facilities and many facilities have off-gas treatment systems to minimize the volatile nuclides dispersion to the environment. Silver impregnated inorganic adsorbents were known as inflammable and stable fixing materials for iodine and the authors started to develop 25 years ago a kind of inorganic adsorbent that has better capability compared with conventional ones. Aluminum oxide (Alumina) was selected as a carrier material and silver nitrate as an impregnated one. Pore diameters were optimized to avoid the influence of impurities such as humidity in the off-gas stream at lower temperatures. Experiments and improvements were alternately conducted for the new adsorbent. The tests were carried out in various conditions to confirm the performance of the developed adsorbent, which clarified its good ability to remove iodine. Silver nitrate impregnated alumina adsorbent (AgA) has about twice the capacity for iodine adsorption and higher iodine removal efficiency at relatively high humidity than conventional ones. The AgA chemically and stably fixes radioactive iodine and fits the storage and disposal of used adsorbent. AgA is now and will be applied to nuclear power plants, reprocessing plants, and research facilities.

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We have applied the bias factor method to a coolant void reactivity of a breeding light water reactor with use of FCA-XXII-1 experiment with introducing a concept of exponentiated experimental value into the bias factor method. We have formulated the prediction uncertainty reduction by the use of the bias factor method extended by the concept of the exponentiated experimental value. From the numerical results, it is verified that the present method can reduce the prediction uncertainty in the design calculation value while the conventional bias factor method cannot reduce it. The present method overcomes a problem caused by the conventional bias factor method in which the prediction uncertainty increases in the case that the experimental core has the opposite reactivity worth and the consequent opposite sensitivity coefficients to the real core. It is concluded that the introduction of exponentiated experimental value can effectively utilize experimental data and extend applicability of the bias factor method.

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The Very High Temperature Reactor (VHTR) system, which is one of generation IV reactors, is the high temperature gas-cooled reactor (HTGR) with capabilities of hydrogen production and high efficiency electricity generation. The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 950°C in April, 2004 during the “rise-to-power tests” confirming the reactor performance. The safety demonstration tests by using the HTTR started from 2002 and are under going to demonstrate inherent safety features of HTGRs. The experimental data obtained in these tests are inevitable to design the VHTR with high cost performance. The analytical models validated through these tests in the HTTR are applicable to precise simulation of an HTGR performance and can contribute to the research and development of the VHTR.

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The neural networks have been utilized in on-line monitoring-system of High Temperature Engineering Tested Reactor (HTTR) with thermal power of 30MW. In this system, several neural networks can independently model the plant dynamics with different architecture, input and output signals and learning algorithm. Monitoring task of each neural network is also different, respectively. Those parallel method applications require distributed architecture of computer network for performing real-time tasks. One of main task is real-time plant monitoring by Multi-Layer Perceptron (MLP) in auto-associative mode, which can model and estimate the whole plant dynamics by training normal operational data only. The basic principle of the anomaly detection is to monitor the difference between process signals measured from the actual plant and the corresponding values estimated by MLP. Other tasks are on-line reactivity prediction, reactivity and helium leak monitoring, respectively. From the on-line monitoring results at the safety demonstration tests, each neural network shows good prediction and reliable detection performances.

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With nuclear medicine receiving greater attention due to its unique characteristics in both diagnostics and therapeutics during recent decades, finding a highly controllable fabrication method becomes more urgent. The radioisotope ¹²⁴I (T_1/2=4.18d; E_β+=2.13MeV; I_β+=25%) has gained plentiful interests in the medical usages such as functioning imaging of cell proliferation in brain tumors using [¹²⁴I]iododeoxyuridine (IUdR), imaging of immunoreactions in tumors using ¹²⁴I-labelled monoclonal antibodies, the in-vivo imaging of ¹²⁴I-labelled tyrosine derivatives, and the classical imaging of thyroid diseases with ¹²⁴I, among others. Furthermore, it is because that thermal response of target during the fabrication process may affect the production of ¹²⁴I to some extent and needs thorough investigations. Hence, the compact cyclotron located in the Institute of Nuclear Energy Research was employed in this study to generate 20MeV protons to irradiate TeO₂ solid targets in which the radioisotopes ¹²⁴I were produced through the ¹²⁵Te(p, 2n)¹²⁴I nuclear reaction. In addition, the widely-used ANSYS computer code was adopted to theoretically analyze thermal responses of TeO₂ to irradiation cases with variations in ion beam current and its thermal conductivity. The results indicate that TeO₂ temperature is strongly dependent on its thermal conductivity and ion beam current. In particular, TeO₂ surface temperature is extremely sensitive to the air-gap size between TeO₂ and target holder. Thus the target holder is suggested to be re-designed in order to prevent TeO₂ from melting and a high efficiency production of radioisotopes ¹²⁴I for nuclear medical diagnostics can be successfully achieved.

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Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MW_e) and large (1400 MW_e) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU^®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

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Aiming at proposing effective applications of seismic probabilistic safety assessment (PSA) for design and risk management of nuclear facilities, we conducted a preliminary seismic PSA study for a multi-unit site to examine core damage frequency (CDF) and core damage sequences with consideration of the effect of correlations of component failures. In addition, we also examined the effectiveness of an accident management measure, namely, cross-connections of emergency diesel generators (EDGs) between adjacent units in this study. Twin BWR-5 units of the same design were hypothesized to be located at the same site in this study and the CDF as well as the accident sequences of this two-unit site were analyzed by using SECOM2, a system reliability analysis code for seismic PSA. The results showed that the calculated CDF was dependent on the assumptions on the correlations of component failures. When the rules for assigning correlation coefficients of component responses defined in the NUREG-1150 program were adopted, the CDF of a single unit, the CDF of this two-unit site (the frequency of core damages of at least one unit at this site) and the frequency of simultaneous core damages of both units increased by factors of about 1.3, 1.2 and 2.3, respectively. In addition, it might be possible that the simultaneous core damages of both units are caused by different accident sequence pairs as well as the same sequence pairs. When cross-connections of EDGs between two units were available, the CDF of a single unit, the CDF of this two-unit site as well as the frequency of simultaneous core damages of both units decreased. In addition, the CDF of this two-unit site was smaller than the CDF of a single unit site. These results show that cross-connections of EDGs might be beneficial for a multi-unit site if the rules for assigning correlation coefficients defined in NUREG-1150 program are reasonable.

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This article discusses the specificity of RBMK (channel type, boiling water, graphite moderated) reactors and problems of Reactor Cooling System modelling employing computer codes. The article presents, how the RELAP/SCDAPSIM code, which is originally designed for modelling of accidents in vessel type reactors, is fit to simulate the phenomena in the RBMK reactor core and RCS in case of Beyond Design Basis Accidents. For this reason, use of two RELAP/SCDAPSIM models is recommended. First model with described complete geometry of RCS is recommended for analysis of initial phase of accident. The calculations results, received using this model, are used as boundary conditions in simplified model for simulation of later phases of severe accidents. The simplified model was tested comparing results of simulation performed using RELAP5 and RELAP/SCDAPSIM codes. As the typical example of BDBA, large break LOCA in reactor cooling system with failure of emergency core cooling system was analyzed. Use of developed models allows to receive behaviour of thermal-hydraulic parameters, temperatures of core components, amount of generated hydrogen due to steam-zirconium reaction. These parameters will be used as input for other special codes, designed for analysis of processes in reactor containment.

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To evaluate the irradiation behavior of the rock-like oxide fuel, irradiation experiments were carried out. Three fuels were prepared; a single phase fuel of yttria-stabilized zirconia containing UO₂ (U-YSZ) and two types of particle-dispersed fuels of U-YSZ particles in spinel or corundum matrix. U-YSZ particle sizes were about 100-200 μm. These fuels were irradiated in Japan Research Reactor No. 3 for about 280 days. The burnups were about 11% FIMA. The fission gas release rate (FGR) was determined by puncture test and gas analysis. Corundum-based fuel showed extremely high FGR (88%). The irradiation behavior of spinel-based fuels in conditions avoiding the spinel decomposition was superior to corundum-based fuels, in view of their retention of FP gases. The temperature of U-YSZ single-phase fuel pellets was highest among the fuels, because of its low thermal conductivity. Nevertheless the U-YSZ single-phase fuel showed very low FGR (5%). Microstructure analyses of irradiated fuel pellets were carried out by ceramography and electron probe micro-analysis (EPMA). The restructuring of fuel pellet was not observed in the spinel-based fuel irradiated below 1400 K. Significant appearance changes were not also observed for corundum-based fuel, in the range of about 800 K to 1900 K, where the irradiation tests were carried out.

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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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Generation IV type fast reactors (FR) are expected to be commercially deployed instead of light water reactors (LWR) from around 2050. Replacement of LWR to FR needs flexibility due to uncertain factors such as FR deployment rate which affects the FR fuel (Pu) supply amount from LWR spent fuel reprocessing and the capacity of related facilities. If the FR deployment rate is as currently planned, more Pu must be prepared by expanding LWR reprocessing. If the FR deployment rate decreases, LWR reprocessing must be reduced to avoid excess Pu. To cope with this issue we proposed the innovative system called Flexible Fuel Cycle Initiative (FFCI) that has integral reprocessing for LWR and FR spent fuels. LWR reprocessing in FFCI only carries out about 90% U recovery and residual material with Pu, U (∼5%), minor actinides (MA) and fission products (FP) goes to FR reprocessing for the planned FR deployment rate. For any decrease in the FR deployment rate temporary storage will be used. Coexistence of Pu/U with MA and FP until just before Pu/U usage in the FR provides high proliferation resistance. Preliminary evaluation revealed that FFCI can reduce the LWR reprocessing capacity and LWR spent fuel storage amount compared with current plan (reference system) if the FR deployment rate decreases. Several FR deployment scenarios and countermeasures such as FFCI were investigated.

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Most of ⁸⁵Kr in atmosphere is released from anthropogenic sources such as tests of nuclear weapons and nuclear reactors. As a chemically inert gas with a long half-life, ⁸⁵Kr will continue to accumulate in the environment. Thus, it is necessary to monitor the atmospheric ⁸⁵Kr concentrations. In this study, the authors conducted a screening test of several adsorbents to search for suitable adsorbents for adsorption of Kr at under nitrogen atmosphere, which is used for the monitoring system of ⁸⁵Kr. A screening test was carried out by focusing on adsorption characteristics at a cryogenic temperature. With regard to adsorbents selected in the screening test, more detailed adsorption characteristics were examined. As a result, it was found that carbon based adsorbents have better performance for adsorption of Kr and that the Ambersorb 572 adsorbent has the largest adsorption capacity. It was also found that the adsorption capacity of non-carbon based adsorbents is much lower than that of carbon-based adsorbents.

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Off gases produced in the reprocessing of spent nuclear fuel contain various radioactive gases and emission of these gases to the environment must be suppressed as low as possible. ¹⁴C with a long half-life, which is mainly released as the form of carbon dioxide, is one of such gaseous radioactive materials. One of the measures to capture radioactive gases from the off-gas is the utilization of adsorption technique. In this work, the adsorption behavior of carbon dioxide on various adsorbents was studied. It was found that a MS4A (Molecular Sieve 4A) adsorbent is more suitable for selective recovery of carbon dioxide. Thus, more detailed adsorption characteristics of carbon dioxide were studied for a MS4A adsorbent. Moreover, the authors investigated the influence of coexistent water vapor, which is also contained in the off-gas, on the adsorption behavior of carbon dioxide.

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Radionuclide migration experiments in quarried blocks of granite under in-situ conditions at the 240-m level in AECL's Underground Research Laboratory (URL) were performed under a five-year cooperative research program between Japan Atomic Energy Research Institute (JAERI, reorganized to Japan Atomic Energy Agency, JAEA) and Atomic Energy of Canada Ltd. (AECL). Migration experiments with Br, ³H, ⁸⁵Sr, ²³⁷Np, ²³⁸Pu, ^95mTc and synthetic colloids, and post-experimental alpha and gamma scanning of the fracture surfaces were performed using 1 m³ granite blocks, containing a single fracture, excavated from a water-bearing fracture zone. The transport of the radionuclides was affected by macroscopic mechanical dispersion, matrix diffusion and element-specific sorption on fracture surfaces. Colloid transport exhibited a complicated process that may include sedimentation and diffusion into stagnant zones.

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Removal of Pd, Ru and RuO₂ from molten glass was studied using liquid Cu as the collecting metal. To increase the collision frequency between Cu and these platinum group metals (PGMs), copper-ruby glass containing Cu nanoparticles was used for the removal. The glass was prepared by the reduction of CuO dissolved in glass using Si as the reductant. Existence of the Cu nanoparticles was conformed by a measurement of the absorption peak of surface plasmon resonance at 595 nm. Another glass containing Cu particles of around 30 μm was prepared as a control specimen. By a heating of the two glasses with PGMs at 1473 K, separable metal buttons formed in both glasses. Metallic Pd and Ru were collected more than 90% in either metal buttons. There was no significant difference between the two glasses for the removal of metals. On the other hand, RuO₂, which was reduced to metallic Ru by Cu, was collected 83% for the ruby glass, while not more than 52% for the control glass. The use of copper-ruby glass was effective for the removal of the oxide. The results of a leaching test of the glasses indicated that dispersing smaller Cu particles in glass had an effect to decrease the leaching rate of Cu from the glass into nitric acid.

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We studied cement-like solidification process for low-level liquid waste with relatively high levels of radioactivity that contains a high concentration of sodium sulfate. For this type waste, it is important that the sulfate ion should not dissolve from the solid waste because it forms ettringite on reaction with minerals in the concrete of the planned repository, and this leads to cracking during repository storage. It is also preferable that the pH of the pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Therefore, the present solidification process has two procedures: conversion into insoluble sulfate from sodium sulfate (CIS) and formation of low pH cement-like solid (FLS). In the CIS procedure, BaSO₄ precipitation occurs with addition of Ba(OH)₂•8H₂O to the liquid waste. In the FLS procedure, silica fume and blast furnace slag are added to the liquid waste containing BaSO₄ precipitate. We show the range of appropriate Ba/SO₄ molar ratio is from 1.1 to 1.5 in the present solidification process by leaching tests for some kinds of solid waste samples. The CIS reaction yield is over 98% at a typical CIS condition, i.e. Ba/SO₄ molar ratio=1.3, reaction temperature=60°C, and time=3 hr.

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Characterization of C-14 in PWR Radioactive Wastes has been researched and formation mechanisms of C-14 have been discussed. It was found from the research results that the chemical form of C-14 existed in primary coolant was organic and was low molecule compounds which are soluble in water. On the other hand, most of C-14 components existed in condensed liquid waste and existed on solid waste were insoluble in water and chemically stable. The insoluble C-14 component was considered to be produced by activation reaction between neutron and substances with nitrogen. Those were thought to be decomposition substances escaped from high molecular organic materials, such as ion exchange resin, diaphragm seal, etc.

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In decommissioning process of nuclear facilities, large amount of radioactive isotopes are discharged as waste. Radioactive carbon isotope (¹⁴C) is one of the key nuclides to determine the upper limit of concentration in the waste disposal. In particular, ¹⁴C on the graphite reactor decommissioning should be separated from stable carbon isotopes (¹²C and ¹³C) 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 (¹²C, ¹³C and ¹⁴C) 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 ¹⁴C isotope analysis through the preliminary experimental results of ¹³C isotope analysis with a prototype system.

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R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.

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A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we describe the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with a modified TRAC-BFI code, where Japan Atomic Energy Agency (JAEA) newest critical power correlation is implemented for the BT judgement. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.

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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. However, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors 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 calculate the detailed information of the two-phase flow. We tried to verify the TPFIT code by comparing it 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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In Japan Atomic Energy Agency (JAEA), the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been developed. For thermal design of FLWR, it is necessary to develop analytical method to predict boiling transition of FLWR. Japan Atomic Energy Agency (JAEA) has been developing three-dimensional two-fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system to simulate complex shape channel flow. In this paper, as a part of development of ACE-3D to apply to rod bundle analysis, introduction of parallelization to ACE-3D and assessments of ACE-3D are shown. In analysis of large-scale domain such as a rod bundle, even two-fluid model requires large number of computational cost, which exceeds upper limit of memory amount of 1 CPU. Therefore, parallelization was introduced to ACE-3D to divide data amount for analysis of large-scale domain among large number of CPUs, and it is confirmed that analysis of large-scale domain such as a rod bundle can be performed by parallel computation with keeping parallel computation performance even using large number of CPUs. ACE-3D adopts two-phase flow models, some of which are dependent upon channel geometry. Therefore, analyses in the domains, which simulate individual subchannel and 37 rod bundle, are performed, and compared with experiments. It is confirmed that the results obtained by both analyses using ACE-3D show agreement with past experimental result qualitatively.

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An estimation of the void fraction in a tight-lattice rod bundle was needed for the R&D of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR). For this purpose, we measured the void fraction and studied the behaviors of boiling flow. The void fraction was measured by a neutron radiography, a quick-shut-valve technique, and an electro void fraction meter. The data were taken using the 7-, 14-, 19- and 37-rod bundle test sections with the rod gap of 1.0 or 1.3 mm under from atmospheric pressure to 7.2 MPa conditions. A spacer effect test was also carried out. The following estimations were conducted: (1) a similarity of the advanced analysis codes with the 3D void fraction data, (2) the comparisons of the TRAC-BF1 code and a drift-flux model with the 1D data. Followings were made clear: (a) The void fraction becomes lower at the peripheral and higher at the rod gap part of the lower core and at the center of the subchannel of the upper core, (b) the codes calculates the similar distribution to the data, and (c) the TRAC-BF1 and the drift-flux model tends to overestimate the void fraction at the lower quality region, on the other hand at the higher quality, those methods tend to same characteristics to the data. It was confirmed that several special features were existed in the tight-lattice rod bundle but the codes were applicable.

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In order to obtain the data on wall and interfacial friction forces for two-phase flows in a triangle tight lattice subchannel, adiabatic experiments were conducted for single- and two-phase flows under hydrodynamic equilibrium flow conditions. In the experiment, air was used as the test gas, while water and water with a surfactant as test liquids to know the effects of the reduced surface tension on the wall and the interfacial friction forces. The data showed that both the wall and the interfacial friction forces were higher in air-water with a surfactant system than air-water one. In the analysis, the respective data have been compared with the predicted values by existing correlations, and the existing correlations were modified to improve its prediction accuracy against the present data. The modified correlations can predict well the present data on the wall and the interfacial friction forces for both air-water and air-water with a surfactant systems.

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A three-dimensional one-way bubble tracking method is a promising numerical method for calculation of time-spatial evolution of gas-liquid interfacial configuration with use of a little computing resource. Since the method has been applied to only an adiabatic air-water bubble flow, the method is developed for the analysis of a boiling flow in this study. One-dimensional Eulerian equation of energy conservation for a continuous liquid phase and an interface heat transfer equation for dispersed bubbles are introduced. Then, radial liquid temperature distribution, wall heat transfer between a heated wall and subcooled liquid, bubble generation on a heated wall and expansion or condensation of bubbles are taken into account. The developed method is applied to the boiling flow experiment and radial void fraction distribution is compared. It is confirmed that the method can give good prediction of tendency of the void fraction distribution in the boiling flow.

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The interaction between bubble motion and its surrounding liquid motion through a collision of a pair of zigzagging bubbles in a rest water column was experimentally investigated. A pair of hypodermic needles and a bubble generator utilizing pressure oscillation were employed in order to exactly extract and highly reproduce the interaction between liquid-phase motion and bubble motion. The recursive cross-correlation PIV technique made it possible to obtain the accurate velocity field of surrounding liquid motion around a pair of bubbles. The vorticity field around a pair of bubbles was calculated from the results of velocity field. First, various kinds of interactions (e.g. bouncing and coalescence) were found out from the results of bubble motion (e.g. velocity) after a collision. We classified such interactions of a pair of bubbles using the dimensionless number l/d. Second, we investigated surrounding liquid motion about the two cases of interactions after the bouncing. One is that only horizontal velocity of bubble decreased after the collision, the other is that both horizontal and vertical velocity decreased. At the former case, the vortical region generated at the rear of a pair of bubbles. Therefore as the reason for the decrease of bubble velocity, the shedding area of hairpin-like vortex is restricted. At the latter case, bubble wake flows into the area between bubbles and vortical region doesn't generate under bubbles after the collision. Hence, bubble wake overtakes a pair of bubbles after the collision, then, bubble velocity decreased.

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The turbulent heat transfer coefficients for the flow velocities (u=4.0 to 21 m/s), the inlet liquid temperatures (T_in=296.5 to 353.4 K), the inlet pressures (P_in=810 to 1014 kPa) and the increasing heat inputs (Q₀ exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured by an experimental water loop. The Platinum test tubes of test tube inner diameters (d=3, 6 and 9 mm), heated lengths (L=32.7 to 100 mm), ratios of heated length to inner diameter (L/d=5.51 to 33.3) and wall thickness (δ=0.3, 0.4 and 0.5 mm) with surface roughness (Ra=0.40 to 0.78 μm) are used in this work. The turbulent heat transfer data for Platinum test tubes were compared with the values calculated by other workers' correlations for the turbulent heat transfer. The influence of Reynolds number (Re), Prandtl number (Pr), Dynamic viscosity (μ) and L/d on the turbulent heat transfer is investigated into details and, the widely and precisely predictable correlation of the turbulent heat transfer for heating of water in a short vertical tube is given based on the experimental data. The correlation can describe the turbulent heat transfer coefficients obtained in this work for the wide range of the temperature difference between heater inner surface temperature and average bulk liquid temperature (ΔT_L=5 to 140 K) with d=3, 6 and 9 mm, L=32.7 to 100 mm and u=4.0 to 21 m/s within ±15% difference.

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In the present study, liquid-phase turbulence modulation induced by a bubble swarm ascending in arbitrary turbulence was experimentally investigated. Liquid-phase homogeneous isotropic turbulence was formed using an oscillating grid in a cylindrical acrylic vessel of 149 mm in inner diameter. A bubble swarm consisting of 19 bubbles of 2.8 mm in equivalent diameter was examined; the bubble size and generating time were completely controlled using a bubble generating device through audio speakers. This bubble generating device was able to repeatedly control the bubble swarm arbitrarily and precisely. The bubble swarm was generated at a frequency of 4 Hz. The liquid phase motion was measured via two LDA (Laser Doppler Anemometer) probes. The turbulence intensity, spatial correlation and integral scale were calculated from LDA data obtained by the two spatially-separate-point measurement. When the bubble swarm was added, the turbulence intensity dramatically changed. The original isotropic turbulence was modulated to the anisotropic turbulence by the mutual interaction between the bubble swarm and the ambient isotropic turbulence. The integral scales were calculated from the spatial correlation function. The increase in turbulence intensity and the decrease in integral scale were observed by injecting the bubble swarm in oscillating-grid turbulence.

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Nanofluids, colloidal dispersions of nanoparticles in a base fluid such as water, can afford very significant Critical Heat Flux (CHF) enhancement. Such engineered fluids potentially could be employed in reactors as advanced coolants in safety systems with significant safety and economic advantages. However, a satisfactory explanation of the CHF enhancement mechanism in nanofluids is lacking. To close this gap, we have identified the important boiling parameters to be measured. These are the properties (e.g., density, viscosity, thermal conductivity, specific heat, vaporization enthalpy, surface tension), hydrodynamic parameters (i.e., bubble size, bubble velocity, departure frequency, hot/dry spot dynamics) and surface conditions (i.e., contact angle, nucleation site density). We have also deployed a pool boiling facility in which many such parameters can be measured. The facility is equipped with a thin indium-tin-oxide heater deposited over a sapphire substrate. An infra-red high-speed camera and an optical probe are used to measure the temperature distribution on the heater and the hydrodynamics above the heater, respectively. The first data generated with this facility already provide some clue on the CHF enhancement mechanism in nanofluids. Specifically, the progression to burnout in a pure fluid (ethanol in this case) is characterized by a smoothly-shaped and steadily-expanding hot spot. By contrast, in the ethanol-based nanofluid the hot spot pulsates and the progression to burnout lasts longer, although the nanofluid CHF is higher than the pure fluid CHF. The presence of a nanoparticle deposition layer on the heater surface seems to enhance wettability and aid hot spot dissipation, thus delaying burnout.

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The Supercritical Water-Cooled Reactor (SWCR) is one of the most promising concepts for Generation IV candidate systems [Kataoka et al., 2002; USDOE, 2002; Buongiorno, 2004]. The SCWR has several advantages compared to the existing light water reactor (LWR) systems, including the use of direct cycle combined with single-phase working fluid, high thermal efficiency, and the existing experience with the proven technology used in fossil power plants. A common feature of most Supercritical Water-Cooled Reactor (SWCR) designs that have been proposed to date is a highly nonuniform temperature distribution inside the reactor core. This is mainly due to the combined effects of core peaking factors and limits imposed on coolant flow rate. Furthermore, statistical uncertainties in the evaluation of hot spot factors normally contribute to an increase in the range of temperature distribution that must be considered in reactor design. The purpose of this paper is to present the results of analysis on the SCWR in-core temperature distribution, aimed at identifying possible methods of reducing the maximum coolant temperature and improving the thermal-hydraulic characteristics of the proposed reactor system.

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Both in-phase (core wide mode) instability and out-of-phase (regional mode) instability are of great concerns in BWR stability issues. Normally, decay ratios for regional mode oscillations are much less than those under core wide conditions. However, under certain observation mode, the regional mode instability has the phenomenon of power increasing in one half of the core and at the same time, it decrease in the other half, so it looks like that the average power remains essentially constant. This research presents a study of fractional change of decay ratio to evaluate parametric effects of regional mode instability on reload core design power/flow stability boundary for the Chinshan Nuclear Power Plant Unit 2 Cycle 21 (BWR4). Making use of LAPUR5.2 and SIMULATE-3 codes, we have established a methodology to conduct such out-of-phase stability analysis. Many important parameters, such as system pressure, core flow rates, moderator void fraction, fuel physical and geometrical properties, have strong influences on regional mode stability. Current investigations have shown that at some operation points along the stability boundary, certain parameters present more sensitive characteristics.

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Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes⁽¹⁾, which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility ⁽²⁾, operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed ⁽³⁾. The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

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The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für 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 MWe 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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Within the nuclear reactor safety analysis, one of the events that could potentially lead to a re-criticality accident in case of a Small Break Loss of Coolant Accident (SBLOCA) in a Pressurized Water Reactor (PWR) is a boron dilution scenario followed by a coolant mixing transient. Some UPTF experiments can be interpreted as generic boron dilution experiments. In fact, the UPTF experiments were originally designed to conduct separate effects studies focused on multi-dimensional thermal hydraulic phenomena. However, in the case of experimental program TRAM, some studies are realized on the boron mixing: tests C3. Some of these tests have been used for the validation and assessment of the 3D module of CATHARE code. Results are very satisfying; CATHARE 3D code is able to reproduce correctly the main features of the UPTF TRAM C3 tests, the temperature mixing in the cold leg, the formation of a strong stratification in the upper downcomer, the perfect mixing temperature in the lower downcomer and the strong stratification in the lower plenum. These results are also compared with the CFX5 and TRIO-U codes results on these tests.

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Since the revision of the 10 CFR50.46 in 1988, the best-estimate codes may be used in safety demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters comes principally from the lack of knowledge of the input parameters (initial and boundary conditions of the calculated transient, empirical models of the code, etc.). The application of the best estimate plus uncertainty analysis can be made in three steps in a statistical evaluation:
1. The determination of the input parameters (IP) statistical characteristics.
2. The modelling and understanding of the output parameters.
3. The evaluation of the 95^th percentile with a high degree of confidence.
The first step is generally done manually according to expert judgments and comparing experimental data. The user effect is also very important in the determination of the statistical characteristics: range of variation and probability law. To reduce this user effect and to help the experts in their evaluation, IRSN has been developing a fully automated methodology and approach. This methodology is called DIPE: Determination of Input Parameters uncertaintiEs. This paper presents the advantages and limits of the application of DIPE for the physical models input parameters of CATHARE V2.5_1, a thermal-hydraulics code used in safety demonstration. DIPE will be applied to separate effect tests data used in the code qualification such as CANON and MARVIKEN.

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In order to investigate the stability of a nuclear reactor core with an oxide mixture of uranium and plutonium (MOX) fuel installed, channel stability and regional stability tests were conducted with the SIRIUS-F facility. The SIRIUS-F facility was designed and constructed to provide a highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermalhydraulics-neutronics instabilities of the Advanced Boiling Water Reactors (ABWRs). A real-time simulation was performed by modal point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility.
A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying auto regressive (AR) methods to the time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among the five common AR methods are within 0.03 and 0.01 Hz, respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of the Yule-Walker method with the model order of 30.

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This work describes the application of noise analysis techniques to study the unsteady flow patterns in the core of nuclear reactor plants. The study is focused on the determination of the position of the flow perturbations and their causes. The analysis was performed using sampled data collected from a four loop plant instrumentation that included the loop flow rates, the ex-core neutron detectors and the core exit thermocouples. As part of the analysis, the thermocouple signals were cross-correlated to the flow data to determine approximate flow perturbation mappings in the core. The results of the analysis illustrate the value of noise analysis as a tool for the characterization of the flow conditions in the core.

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A computational fluid dynamics model on a joule-heated glass melter, which is used for the vitrification process of high-level liquid waste, has been developed to simulate the operation of actual melters. Electric field simulation is coupled with the thermal and flow field calculation to investigate the effects of the joule-heating on the buoyancy-driven flow of the molten glass. The melter is operated in a cyclic sequence including a bottom-heating phase, a drain phase and a bottom-cooling phase. In order to estimate the transient behavior of actual melters, unsteady simulations for several tens hours are necessary. It is found that there is non-uniformity of the heating at the bottom of the melter during bottom-heating phases. It is generated by the asymmetrical electric current distribution due to the arrangement of the electrodes. The temperature distribution at the bottom, however, is kept almost symmetrically due to the mixing effect by natural convection of the glass. Computational results are compared with experimental data obtained by actual operations of a prototype melter. The temperature distribution and electric resistances between electrodes predicted by the simulations indicate that the characteristics of the simulated melter show good agreements with the actual melter.

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