In relation to the design of an innovative FLexible-fuel-cycle Water Reactor (FLWR), investigation of thermal-hydraulic performance in tight-lattice rod bundles of the FLWR is being carried out at Japan Atomic Energy Agency (JAEA). The FLWR core adopts a tight triangular lattice arrangement with about 1 mm gap clearance between adjacent fuel rods. In view of importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study presents a statistical evaluation of numerical simulation results obtained by a detailed two-phase flow simulation code, TPFIT, which employs an advanced interface tracking method. In order to clarify mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is applied to simulate water-steam two-phase flow in two modeled subchannels. Attention is focused on instantaneous fluctuation characteristics of cross flow. With the calculation of correlation coefficients between differential pressure and gas/liquid mixing coefficients, time scales of cross flow are evaluated, and effects of mixing section length, flow pattern and gap spacing on correlation coefficients are investigated. Differences in mechanism between gas and liquid cross flows are pointed out.

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The flow in the vicinity of a moving contact line is characterized by a steep increase in the fluid slipping relative to the solid surface owing to fluid stress concentration at the contact line. In this study, molecular dynamics (MD) simulations are made to investigate the dependence of microscopic configuration of the fluid-fluid interface, in particular the contact angle, on the speed of contact line relative to the solid surface ΔV. While ΔV increases when the solid surface velocity relative to the bulk fluid, V, is increased from zero, ΔV starts decreasing when V exceeds a certain upper limit for the given material combinations. The contact angle dependence on ΔV (and V) was predicted from the microscopic stress balance near the contact line using the fluid distribution obtained from the continuum hydrodynamic theory. It is also shown that the stress balance prediction agrees with the MD results when appropriate slip velocities are assumed in the hydrodynamic calculation.

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Direct contact boiling heat transfer of sub-cooled water with lead-bismuth eutectic (Pb-Bi) was investigated for the evaluation of the performance of steam generation in direct contact of feed water with primary Pb-Bi coolant in upper plenum above the core in Pb-Bi-cooled direct contact boiling water fast reactor. An analytical two-fluid model was developed to estimate the heat transfer numerically. Numerical results were compared with experimental ones for verification of the model. The overall volumetric heat transfer coefficient was calculated from heat exchange rate in the chimney. It was confirmed that the calculated results agreed well with the experimental result.

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A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. This provides SI with capability to serve also as a direct-contact feed-water heater that heats up feed-water by using extracted steam from turbine. Our technology development aims to significantly simplify equipment and reduce physical quantities by applying "high-efficiency SI", which are applicable to a wide range of operation regimes beyond the performance and applicable range of existing SIs and enables unprecedented multistage and parallel operation, to the low-pressure feed-water heaters and emergency core cooling system of nuclear power plants, as well as achieve high inherent safety to prevent severe accidents by keeping the core covered with water (a severe accident-free concept). This paper describes the results of the scale model test, and the transient analysis of SI-driven passive core injection system (PCIS).

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

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Recent coupled code benchmarks identified coolant mixing in the reactor vessel as an unresolved issue in the analysis of complex plant transients with reactivity insertion. Thus, Phase 2 of the OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) was defined. The benchmark includes calculation of vessel mixing tests and main steam line break (MSLB) analysis. The reference plant is Kozloduy-6 in Bulgaria. The general objective is the assessment of system codes for VVER safety analysis and specifically for their use in the analysis of reactivity transients. A specific objective is the testing of different scale mixing models (mixing matrix, multi-1D, coarse-3D and CFD), and analysis of MSLB transients with improved vessel thermal hydraulic models. The benchmark is sponsored by CEA-France and OECD and is jointly prepared by CEA and INRNE, in collaboration with the Kozloduy NPP, IRSN and PSU. This paper summarizes CATHARE2 code assessment calculations using multi-1D vessel thermal hydraulics with cross flow. Test cases are the OECD V1000CT-1 pump start-up benchmark and the V1000CT-2 benchmarks. Emphasis is put on vessel mixing aspects. Separate effects in the lower plenum as well as component and integral system tests are considered. The comparison shows that a six-sector vessel mixing model informed by plant data or validated CFD calculations in the initial state was able to correctly reproduce the channel average temperatures at the core inlet as well as the vessel outlet temperatures. Testing at system level including code-to-experiment and CATHARE-ATHLET comparison shows that the considered CATHARE VVER-1000 system model is capable of MSLB simulation.

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The JT-60SA aims to contribute and supplement ITER toward demonstration fusion reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is restricted. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor cassettes. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor cassette, which is 10 degrees wide in toroidal direction and weighs 500kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor cassette to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the cassette by a pallet installed from outside the VV.

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

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A microwave NDT method based on guided electromagnetic wave is described in order to detect longitudinal crack and evaluate its location in steel pipe. To detect longitudinal crack in pipe, suitable guided-mode should be propagated in the inspected pipe with crack. For this purpose, first and dominant circular TE₁₁-mode is sent to the system by using a mode-converter. Mode-converter is used to convert rectangular TE₁₀-mode to the circular TE₁₁-mode. A network analyzer, which is a generator and receiver of the microwaves designed to process the magnitude and phase of the reflected and transmitted waves was used to generate continues signal and to measure the amplitude of reflection coefficient. The goal of this investigation is to demonstrate the capability of this technique for detecting longitudinal cracks in piping system with high speed. More emphasis is put on the evaluation of crack location by measuring time of flight (TOF) of the reflected waves from the crack. The results for two crack locations at several plunger positions either in frequency domain or time domain for measuring TOF of the waves are presented. The experimental results of TOF are compared with calculations to show that by knowing TOF of the reflected wave, it is possible to predict crack locations. The evaluated results of TOFs are shown to agree well with the calculated ones and crack locations can be estimated with error less than 0.13%.

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Significant thermal stresses are loaded onto the structures of sodium-cooled fast reactor (SFR) due to high temperature and large temperature gradients associated with employing sodium coolant with its high thermal conductivity and low heat capacity. Therefore, it is important to monitor the temperature variation, related stress and displacement, and vibration in the cooling system piping and components in order to assure structural integrity while the reactor plant is in-service. SFR structural integrity monitoring can be enhanced by an optical fiber sensor, which is capable of continuous or dispersed distribution measurements of various properties such as radiation dose, temperature, strain, displacement and acceleration. In the experimental fast reactor Joyo, displacement and vibration measurements of the primary cooling system have been carried out using Fiber Bragg Grating (FBG) sensors to evaluate the durability and measurement accuracy of FBG sensors in a high gamma-ray environment. The data were successfully obtained with no significant signal loss up to an accumulated gamma-ray dose of approximately 4×10⁴ Gy corresponding to 120 EFPDs (effective full power days) operation. Measured displacement of the piping support was nearly equal to the calculated thermal displacement. Measured vibration power spectra of the piping support were similar to those measured with a reference acceleration sensor. The measured results indicate that the FBG sensor is suitable for monitoring the displacement and vibration aspects of fast reactor cooling system integrity in a high gamma-ray environment.

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In order to decrease the amount of aqueous liquid waste discharged from nuclear fuel reprocessing, the conversion of uranium dioxide into its nitrate using liquefied nitrogen dioxide was studied. Uranium dioxide powder was immersed in liquefied nitrogen dioxide at 313 K after a pretreatment by the oxidation of the uranium dioxide with nitrogen dioxide and air at 523 K. Seventy-nine % of the uranium dioxide, whose initial feed amount was 0.3 g, was converted into a water soluble compound. Based on an XRD analysis of the compound, uranyl nitrate trihydrate (UO₂(NO₃)₂·3H₂O) was confirmed.

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An innovative alpha radioactivity monitor for clearance level inspection has been developed. This apparatus measures an ion current resulting from air ionization by alpha particles. Ions generated in a measurement chamber of about 1 m³ in volume are transported by airflow to a sensor and measured. This paper presents computational estimation of the ion transport efficiencies for two pipes with different lengths, the inner surfaces of which were covered with a thin layer of uranium. These ion transport efficiencies were compared with those experimentally obtained for the purpose of validating our model. Good agreement was observed between transport efficiencies from simulations and those experimentally estimated. Dependence of the transport efficiencies on the region of uranium coverage was also examined, based on which such characteristics of ion currents as anticipated errors arising from unknown contaminated positions are also discussed to clarify the effective operation conditions of this monitor.

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Calibration of recently installed proportional counter at the Hot Cells Facility of the Jozef Stefan Institute was performed. Instrument was calibrated for determination of total beta activity, Sr-90 and Pb-210. Detection efficiencies for K-40, Sr-90, Y-90, Pb-210 and Bi-210 were determined, allowing for more accurate determination of the particular nuclide as a single K-40 efficiency. In addition, self-absorption curves for different surface densities for the nuclides mentioned were derived. Two empirical equations for faster and more accurate determination of Sr-90 and Pb-210 were derived. These two equations consider differences in surface density and in-growth of Y-90 and Bi-210, respectively. The detection efficiencies obtained ranged from 10 to 52%, depending on the nuclide, surface density and chemical compositions of the salts used or precipitates obtained following radiochemical separation in the experiment. As a performance test of derived empirical equation for the determination of detection efficiency for Pb-210, specific activity of Pb-210 in IAEA 385 and IAEA 414 intercomparison materials were determined. All procedures and formulae developed include calculation of minimal detectable activities and uncertainty budgets for the determinations concerned.

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The core of Prototype Fast Breeder Reactor (PFBR) is designed to produce 1250 MWt at full power. PFBR is under construction at Kalpakkam, India. In PFBR, the core is of free standing type and one of the major safety criteria for the design of core subassemblies is that the integrity of the core subassemblies should not be impaired and they should not be lifted up from the grid plate even during seismic condition. The net downward force acting on the grid plate is less than the weight of the subassembly due to the hydraulic lifting forces acting on it. Experimental analysis has been carried out to ensure that the subassembly does not get lifted off due to vertical seismic excitation. This paper gives the details of the methodology adopted for the experimental seismic analysis carried out on a core subassembly and the upward displacement of the subassembly under the combined effect of upward fluid force and vertical seismic excitations.

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The chemical control of impurity such as hydrogen and oxygen in coolants is one of the critical issues for the development of liquid metal cooled fast reactors and self-cooled liquid breeder blankets for fusion reactors. Especially, hydrogen (isotopes) level is the key parameter for corrosion and mechanical properties of the in-reactor components. For fission reactors, the monitor of hydrogen level in the melt is important for safety operation. The control of tritium is essential for the tritium breeding performance of the fusion reactors. Therefore, on-line hydrogen sensing is a key technology for these systems. In the present study, conceptual design for the on-line hydrogen sensor to be used in liquid sodium (Na), lead (Pb), lead-bismuth (Pb-Bi), lithium (Li), lead-lithium (Pb-17Li) and molten salt LiF-BeF₂ (Flibe) was performed. The cell of hydrogen sensor is made of a solid electrolyte. The solid electrolyte proposed in this study is the CaZrO₃-based ceramics, which is well-known as proton conducting ceramics. In this concept, the cell is immersed into the melt which is containing the hydrogen at the activity of P_H1 of ambient atmosphere. Then, the cell is filled with Ar-H₂ mixture gas at regulated hydrogen activity of P_H2. The electromotive force (EMF) is obtained by the proton conduction in the electro chemical system expressed as Pt, Melt(P_H1) | Proton conductor | P_H2, Pt. The Nernst equation is used for the evaluation of the hydrogen activity from the obtained EMF. The evaluations of expected performance of the sensor in liquid Na, Pb, Pb-Bi, Pb-17Li, Li and Flibe were carried out by means of the measurement test in gas atmosphere at hydrogen activities equivalent to those for the melts in the reactor conditions. In the test, the hydrogen activity in the gas varied from 2.2x10^-14 to 1. The sensor exhibited good response, stability and reproducibility.

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The conservative-form, pressure-based PCICE numerical method (Martineau and Berry, 2004) (Berry, 2006), recently developed for computing transient fluid flows of all speeds from very low to very high (with strong shocks), is simplified and generalized. Though the method automatically treats a continuous transition of compressibility, three distinct, limiting compressibility regimes are formally defined for purposes of discussion and comparison with traditional methods — the strictly incompressible limit, the nearly incompressible limit, and the fully compressible limit. The PCICE method's behavior is examined in each limiting regime. In the strictly incompressible limit the PCICE algorithm reduces to the traditional MAC-type method with velocity divergence driving the pressure Poisson equation. In the nearly incompressible limit the PCICE algorithm is found to reduce to a generalization of traditional incompressible methods, i.e. to one in which not only the velocity divergence effect, but also the density gradient effect is included as a driving function in the pressure Poisson equation. This nearly incompressible regime has received little attention, and it appears that in the past, strictly incompressible methods may have been conveniently applied to flows in this regime at the expense of ignoring a potentially important coupling mechanism. This could be significant in many important flows; for example, in natural convection flows resulting from high heat flux. In the fully compressible limit or regime, the algorithm is found to reduce to an expression equivalent to density-based methods for high-speed flow.

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Targeting a hydrogen production system using heat produced by a nuclear reactor at about 300°C, we are developing a dimethyl ether (DME) steam reformer and hydrogen purification systems as well as catalysts for DME reforming. The use of heat from a nuclear reactor suppresses the CO₂ concentration change in the atmosphere. In our developments, a catalyst, consisting of mixed oxides, produced hydrogen at a rate of about 1.9 Nm³/h per catalyst volume (m³) at about 300°C. Subsequently, the DME steam reformer achieved a hydrogen production rate of approximately, at least, 1.4 Nm³/h at about 300°C, by absorbing heat from the supplied steam. The aforementioned hydrogen production system via DME steam reforming is to be demonstrated using a thermal power plant. DME steam reforming by using waste heat and the utilization of the produced hydrogen within a combined cycle power plant can reduce fuel consumption, for instance, by about 17% compared to the case of direct DME combustion. The total system, with the use of DME, was compared with the methane case. If necessary, the byproduced CO₂ may be injected into coal seams, increasing CH₄ production via the substitution of CO₂ for CH₄ on coal, where CO₂ adsorption is expected to be stronger than the CH₄ adsorption.

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An in-pile creep rupture experiment was conducted in the experimental fast reactor Joyo to evaluate the creep rupture strength of oxide dispersion strengthened (ODS) ferritic steel under neutron irradiation. ODS has been developed as a most promising fuel cladding material for the next generation fast reactor because of its high temperature resistance and low swelling properties. The irradiation test device MAterial testing RIg with temperature COntrol (MARICO) was developed for the in-pile experiment in Joyo with a temperature control precision of ±4°C. Twenty four ODS specimens with no fuel were pressurized by helium gas up to 22 MPa to accelerate the creep rupture testing. The specimen temperature in the MARICO is controlled by changing the ratio of argon and helium fill gases, which changes the gas gap thermal conductivity between the double walled capsule containing the specimen. The experiment was carried out in the Joyo MK-III core from April 2006 until May 2007. Each creep rupture event was successfully detected by the temperature change at the exit of the capsule and by gamma-ray spectrometry of the reactor cover gas when the filled gas was released from the specimen. The specimen was then identified by analyzing the tag gas isotopic ratio using laser resonance ionization mass spectrometry (RIMS).

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TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Surété Nucleaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulating typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

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This study was performed at the Institute of Reactor Safety at the Forschungszentrum Karlsruhe. It is embedded in the ongoing investigations of the international code assessment and maintenance program (CAMP) for qualification and validation of system codes like TRACE⁽¹⁾ and PARCS⁽²⁾. The chosen reactor type used to validate these two codes was the Russian designed VVER-1000 because the OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 2⁽³⁾ includes detailed information of the Bulgarian nuclear power plant (NPP) Kozloduy unit 6. The post-test investigations of a coolant mixing experiment have shown that the predicted parameters (coolant temperature, pressure drop, etc.) are in good agreement with the measured data. The coolant mixing pattern, especially in the downcomer, has been also reproduced quiet well by TRACE. The coupled code system TRACE/PARCS which was applied on a postulated main steam line break (MSLB) provided good results compared to reference values and the ones of other participants of the benchmark. The results show that the developed three-dimensional nodalization of the reactor pressure vessel (RPV) is appropriate to describe the coolant mixing phenomena in the downcomer and the lower plenum of a VVER-1000 reactor. This phenomenon is a key issue for investigations of MSLB transient where the thermal hydraulics and the core neutronics are strongly linked. The simulation of the RPV and core behavior for postulated transients using the validated 3D TRACE RPV model, taking into account boundary conditions at vessel in- and outlet, indicates that the results are physically sound and in good agreement to other participant's results.

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The measurement of two-phase flow parameters for development of constitutive relationships for the HANARO/MAPLE type finned fuel using Real-Time Neutron Radiography (RTNR) is discussed in this paper. A single element finned Fuel Element Simulator (FES) was used with Freon 134a as the working fluid. To observe the effect of a spacer device on void distribution, single pin tests were performed with and without a spacer present. By analyzing the RTNR images using image processing, the effects of the spacer on the time-averaged and instantaneous void fraction distribution were studied. For the experimental results without a spacer, the time-averaged local void distribution is radially asymmetric and the degree of void fluctuation increases with a decreasing frequency along the heated channel, where the observed asymmetry may be caused by flow induced vibration. For the experimental results with a spacer, the spacer clearly limits any significant vibration and the local void distribution becomes more symmetrical. The spacer however does generate an axial void fraction maximum at the upstream of the spacer with a small depletion zone at the exit of the spacer. By analysis of the instantaneous local void distribution, void fluctuation at the heated wall due to boiling was clearly observed. Also, the agglomeration and breakup of the cold-wall void was evident. The dynamic effects of the local void transients will be discussed in detail.

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The new concept of a rotating fluidized bed in a static geometry opens perspectives for fluidized bed nuclear reactor technology and is experimentally and numerically investigated. With conventional fluidized bed technology, the maximum attainable power is rather limited and maximum at a certain fluidization gas flow rate. Using a rotating fluidized bed in a static geometry, the fluidization gas drives both the centrifugal force and the counteracting radial gas-solid drag force in a similar way. This allows operating the reactor at any chosen sufficiently high solids loading over a much wider fluidization gas flow rate range and in particular at much higher fluidization gas flow rates than with conventional fluidized bed reactor technology, offering increased flexibility with respect to cooling via the fluidization gas. Furthermore, the centrifugal force can be a multiple of earth gravity, allowing radial gas-solid slip velocities much higher than in conventional fluidized beds. The latter result in gas-solid heat transfer coefficients one or multiple orders of magnitude higher than in conventional fluidized beds. The combination of dense operation and high fluidization gas flow rates allows process intensification and a more compact reactor design.

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The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP accounting for both prompt and delayed heating components, and then adjusted using E/C for ¹⁰B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO_2-x or AmO_2-x in the (U, Pu)O_2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel.

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The flow around a circular cylinder is a traditional problem of fluid dynamics, knowledge of which is essential for basic understanding as well as for technical applications, such as large buildings, bridges, standpipes, heat exchanger tubes, rods, transport pipelines, poles and cables, all of which attracted widespread attention. A circular cylinder usually experiences boundary layer separation. In certain Reynolds number range, a periodic flow motion develops in the wake as a result of boundary layer vortices being shed alternatively from either side of the cylinder leading to unwanted structural vibrations. In order to calculate the cylinder response to the flow, a computational method to solve the flow around the body and its resultant vibration using Fluent^® is previously developed and validated. The present study details the extension and verification using the same method by incorporating flow turbulence through its modeling. The incoming free stream flow is uniform with Reynolds number based on diameter of 3.8 and 12.7mm. Results for the unsteady shedding flow behind a circular cylinder and its vibration are presented with experimental comparisons, along with a comparison of two-dimensional laminar as well as turbulent models of the flow for fully coupled interaction. The Strouhal number and structural displacements are in good comparison with the experimental data of [2] showing the capability of FSI method using Fluent^® to tackle laminar as well as turbulent flows.

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In boiling water reactors (BWRs), traversing incore probes (TIPs) are used for the continuous measurement of the axial distribution of neutron flux in a reactor and the calibration of local power range monitors (LPRMs). One of candidates for a substitute for TIP is gamma thermometers (GTs). Some researches have been conducted so far for the verification of applying GT to BWR plants. In the research initiated in 1996, the applicability of hardware of GT was confirmed. In the next research initiated in 2000, the applicability of GT to Core Monitoring System (CMS) in BWR plants was confirmed. Considering the results of these two researches, we manufactured GT assembly with improved structures and installed 2 of them in Kashiwazaki-Kariwa 6 (ABWR) to conduct a verification test and acquired GT signal data for 1 cycle of the plant operation. We compared the reactor power distribution calculated from TIP data with that calculated from the acquired GT signal data. The average Root Mean Square (RMS) value between these two reactor power distributions was 3.5%, and it is well consistent compared with the result of the past two researches. So we had verified the applicability of the improved GT assembly from the result of the in-plant test.

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

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Methods to systematically analyse existing nuclear power plants (NPP) regarding the adequacy of their existing protection equipment against external hazards, e.g. flooding, can be of deterministic as well as probabilistic nature. In the past the adequacy of the protection measures has been assessed only on a deterministic basis. The German regulatory body has issued probabilistic safety assessment (PSA) guidelines, which had been elaborated for a comprehensive integrated safety review of all NPP in operation. Amongst others the guidelines imply, that probabilistic considerations regarding external flooding are required. This paper presents a newly developed graded approach for the probabilistic assessment of external flooding. Main aspects are explained such as the underlying probabilistic considerations and the mathematical procedures for the calculation of exceedance frequencies, which have recently been developed and issued as part of the German Nuclear Safety Standard. Exemplarily it has been investigated if extreme events such as tsunami waves could be a hazard for NPP at coastal sites in Germany. Here it could be shown that due to limited source mechanisms and the specific morphological conditions in the North Sea no dedicated measures for protection against tsunamis in the German Bight are necessary.

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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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Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved “steam” parameters (outlet temperatures up to 625°C and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the “nuclear” heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

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The preliminary conceptual design study of prismatic-type Very High Temperature Reactor (VHTR) has been performed with 950°C 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°C 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 (FP) 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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A power uprate of Forsmark's Unit 3 from 109 % to 125 % will be implemented during the 2010 refuelling outage. This implies an increase in gamma heating of the core shroud which could lead to temperatures higher than the design thickness-mean temperature, 300°C, according to ASME regulations. The CFX 5 model to estimate the core shroud temperature distribution consists of the core bypass, from the lower core support plate up to the core grid, and the upper plenum, limited from above by the core shroud cover with steam separator inlets. The conjugate heat transfer at the core shroud inner wall comprises gamma heating from the core, considered as volume distributed heat sources, and subcooled boiling of the bypass flow. The effect of subcooled boiling has been included by using the model by Kurul and Podowski. Using a conservative gamma heat source distribution, leads a maximum thickness-mean temperature that exceeds the temperature limit by approximately 4°C. If a higher temperature limit is accepted, the ASME regulations are not fulfilled, but the consequence is a minor change in the design stress intensity value, S_m. Using a somewhat realistic gamma heat source distribution, the maximum thickness-mean temperature is well below the design temperature. Decreasing the core inlet temperature from 275.4°C to 272.85°C, i.e. the core inlet temperature used today, leads to no subcooled boiling at the core shroud inner wall and to a different location of the maximum temperature.

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Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

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Critical heat fluxes (CHFs) in a pool of Fluorinert FC-72 were measured for period from transient heat inputs up to steady-state ones. Measurements were made using 1.0 mm diameter of platinum and gold horizontal cylinders in wide ranges of liquid subcoolings and pressures. The steady-state CHFs for saturated condition almost agree with the hydrodynamic instability (HI) model. However, in higher subcoolings, the increasing rate of steady-state CHFs is lower than HI model and it was suggested due to the heterogeneous spontaneous nucleation (HSN). The CHFs for period are clearly categorized into the first, second, and third groups for long, short, and intermediate periods, respectively. Transient CHFs on semi-direct and direct transitions from non-boiling heat conduction to film boiling, exist predominantly during short periods. Those transition processes were assumed due to the explosive-like HSN in originally flooded cavities on cylinder surface. Each of the steady-state and transient CHFs that were obtained from both heaters in various liquid subcoolings and pressures, generally well agree each other.

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For a joule-heated glass melter, we have developed a numerical simulation scheme, which couples electric field, thermal fluid dynamics and platinum group metal particles behavior. Glass properties, especially the viscosity and electrical conductivity, widely change due to the platinum group metals included in high-level liquid waste. It is necessary to estimate the distribution of the platinum group metals in the glass, because operational conditions of the melter are strongly affected by them. In this study, the platinum group metals behavior was treated as the transport of the particle concentration by the Eulerian method in order to reduce the computational loading. Databases of glass properties with the platinum group metals were obtained by measurements of glass samples. Numerical simulation of 18 batch cycles showed that the platinum group metals not only were transported by the glass flow but also settled down and deposited on the bottom walls of the melter. Additionally, the electric current intensively converged along the bottom walls due to the increase of the electrical conductivity in the deposited layer of the platinum group metals. Numerical results agreed reasonably well with experimental data. It has been clarified that this numerical method is useful for the design and operation.

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The static thermophysical properties of the molten salt system like LiF-NaF-BeF₂ influence the design and construction of the fuel salt and coolant in the Molten Salt Reactor for the new generation. In this paper, the equation of state of the ternary system 0.15LiF-0.58NaF-0.27BeF₂, over the temperature range from 873.15K to 1073.15K at one atmosphere pressure, is described by using modified Peng-Robinson equation. The density of the ternary system is evaluated by this equation directly, and compared with the experimental data. Base on the equation of state, the other static thermophysical properties such as the enthalpy, entropy and heat capacity at constant pressure are evaluated by the fugacity coefficient and residual function methods respectively. The density calculated by Peng-Robinson equation is in highly agreement with the experimental data, and the enthalpy, entropy and heat capacity evaluated by such two different methods are consistent with each other. It could be concluded that the modified Peng-Robinson equation could be applicable to estimate the density of the molten salt system, and the Peng-Robinson equation is recommended to be as the fundamental to evaluate the enthalpy, entropy and heat capacity of the molten salt system.

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In case of welding for pressure retaining parts on nuclear components, the verifications of heat affected zone (HAZ) impact properties are required according to application codes such as ASME Sec. III, RCC-M, KEPIC (Korea Electric Power Industry Code) MN, and JEA (Japan Electric Association) Code. Especially in case of Charpy V-notch tests of HAZ, the requirements of notch location and specimen direction have greatly impact on the reliability and consistency of the test results. For the establishment of newly adequate impact test requirements, the requirements about the HAZ impact tests of ASME Section III, RCC-M, KEPIC MN and JEA code were researched in this study. And also the HAZ impact test requirements about surveillance tests in nuclear reactor vessels were compared and investigated. For the effects of the notch location and specimen direction on the impact properties, SA-516 Gr.70 materials were investigated. The specimens were fabricated with using shielded metal-arc welding, and maximum heat inputs were controlled within the range of 16∼27 kJ/cm. Especially, this research showed the lateral expansion values and absorbed energies were not compatible and the impact test results were varied depending on notch location and specimen direction. Based on this study, newly adequate impact test requirements of HAZ were proposed.

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The thermal fragmentation process of melt droplets in an energetic fuel coolant interaction is investigated. Boling effect is considered to be the triggering event, during the transition of heat transfer mode from film boiling to nucleate boiling and coolant periodically contacting with the droplet surface. The vapor bubble around droplet becomes unstable and collapses toward the surface of the droplet which induces the fragmentation of melt droplets. The vapor bubble collapse is modeled by writing a momentum equation for vapor bubble dynamics, an energy equation for each region of the droplet, coolant vapor and liquid and linking each region by the appropriate boundary conditions. And then a thermal fragmentation model triggered by boiling effect is developed and verified. By using the developed model, the fragmented mass of the droplet triggered by boiling effect is calculated. The result shows that the fragmentation rate is larger than that given by hydrodynamic model.

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Several options have been identified that could further enhance the lifetime and reliability of thermocouples developed by the Idaho National Laboratory (INL) for in-pile testing, allowing their use in higher temperatures applications (up to at least 1700 °C). A joint project between the INL and the University of Idaho (UI) is underway to investigate these options and, ultimately, provide recommendations for an enhanced thermocouple design. This paper presents preliminary results from this UI/INL effort. Tests show that unalloyed, but doped, molybdenum wires (ODS-Momolybdenum doped with lanthanum oxide and KW-Mo-molybdenum doped with silicon, tungsten and potassium) better retain ductility at higher temperatures than evaluated candidate undoped developmental alloys (Mo-1.6%Nb and Mo-3%Nb). Thermocouples that contain unalloyed molybdenum were also observed to have better high temperature resolution. Candidate niobium alloys (Nb-1%Zr, Nb-4%Mo, Nb-6%Mo, and Nb-8%Mo) became brittle at lower heating temperatures and shorter durations than any of the wires primarily containing molybdenum. Hence, results indicate that a combination of either ODS-Mo or KW-Mo with Nb-1%Zr appear to be the most favorable configuration. Initial results also show that thermocouple stability can also be improved by using larger diameter wires.

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Hot-pressed boron carbide (B₄C) pellet will be used as shielding material in China Experimental Fast Reactor (CEFR) which is the first fast reactor in China. In this paper, two types of B₄C sample provided by two different units in China were investigated on out-reactor properties, and one type of sample was studied on irradiation performance. Finally, the evaluation of domestic B₄C pellet as shielding material in CEFR was given: the B₄C pellets produced in China could satisfy the requirements of CEFR.

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Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

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Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model does not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions.

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Due to the large plutonium (Pu) throughput and high burn-up fuel in an advanced reprocessing facility, we are faced with the inevitable increasing burden of nuclear material accountancy (NMA) to meet the International Atomic Energy Agency (IAEA) safeguards criteria. A large volume of sampling analysis and inspectors' activities result in a great cost for facility operation. Therefore, it is increasingly important to evaluate a cost-effective performance for the safeguards system. In order to design an advanced safeguards system, we have initiated the development of a safeguards system simulator. The simulator is composed of several interrelated cores and a separate core is planned to develop. The NMA core is a near-real-time accounting (NRTA) code that had been originally developed more than ten years ago and has been improved on an objective-driven pre- and post-processor. A multivariate and multi-scale core based on a principle component analysis with a wavelet technique has been developed to provide an algorithm of process monitoring. The time and frequency decomposition was verified to be an effective technique to detect an abnormal event. In addition, a multiple optimization core has been developed with a fuzzy-linear-programming technique to investigate the cost-effective performance of the conceptual safeguards system. It is shown that a combination of flow-meter and non-destructive assay can be applied to the system in a cost-effective manner. In the future, a virtual design core will be developed to support a walk-through and three dimensional visible plant model.

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