Showing papers in "Journal of Power and Energy Systems in 2009"

Pressure Drop in Two-Phase Swirling Flow in a Steam Separator

Abstract: Pressure drop and liquid film thickness in air-water swirling flows in a one-fifth scale model of the steam separator are measured for a wide range of gas and liquid volume fluxes. Numerical simulations based on one-dimensional single-fluid and two-fluid models are also carried out to examine the feasibility of predicting the pressure drop and film thickness in swirling flows. The pressure drop in a single-phase swirling flow is about five times as large as that in a non-swirling flow due to the increase in the frictional pressure drop. The pressure gradient and liquid film thickness in a two-phase swirling annular flow at the inlet of the pick-off-ring of the separator are well evaluated by using a standard one-dimensional two-fluid model, provided that the interfacial and wall frictions in an ordinary two-phase annular flow are multiplied by appropriate constant values.

Effects of swirler shape on two-phase swirling flow in a steam separator

Abstract: Experiments on two-phase swirling flow in a separator are carried out using several swirlers having different vane angles, different hub diameters and different number of vanes to seek a way for improving steam separators of uprated boiling water reactors. Ratios of the separated liquid flow rate to the total liquid flow rate, flow patterns, liquid film thicknesses and pressure drops are measured to examine the effects of swirler shape on air-water two-phase swirling annular flows in a one-fifth scale model of the separator. As a result, the following conclusions are obtained for the tested swirlers: (1) swirler shape scarcely affects the pressure drop in the barrel of the separator, (2) decreasing the vane angle is an effective way for reducing the pressure drop in the diffuser of the separator, and (3) the film thickness at the inlet of the pick-off-ring of the separator is not sensitive to swirler shape, which explains the reason why the separator performance does not depend on swirler shape.

Numerical Analysis of Droplet Impingement on Pipe Inner Surface Using a Particle Method

Abstract: We investigated single droplet impingement to a solid wall by the MPS-AS method. The MPS-AS method was a unified particle method based on the MPS method for both compressible and incompressible flows. The speed and the profile of the shock wave from the impact point agreed well with the experimental result by Field et al. The shock reflection from the droplet surface was also captured. The same tendency as Rochester's experimental approximation was obtained with respect to the peak pressure.

Development of High Temperature Molten Salt Transport Technology for Pyrometallurgical Reprocessing

Abstract: Pyrometallurgical reprocessing technology is currently being focused in many countries for closing actinide fuel cycle because of its favorable economic potential and an intrinsic proliferation-resistant feature due to the inherent difficulty of extracting weapons-usable plutonium. The feasibility of pyrometallurgical reprocessing has been demonstrated through many laboratory scale experiments. Hence the development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for industrial realization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been very few transport studies on high-temperature fluids. In this study, a salt transport test rig was installed in an argon glove box with the aim of developing technologies for transporting molten salt at approximately 773 K. The gravitation transport of the molten salt at approximately 773 K could be well controlled at a velocity from 0.1 to 1.2 m/s by adjusting the valve. Consequently, the flow in the molten salt can be controlled from laminar flow to turbulent flow. It was demonstrated that; using a centrifugal pump, molten salt at approximately 773 K could be transported at a controlled rate from 2.5 to 8 dm3/min against a 1 m head.

Mechanism of hydrophilicity by radiation-induced surface activation

Abstract: When a metal oxide is irradiated by gamma rays, the irradiated surface becomes hydrophilic. This surface phenomenon is called as radiation-induced surface activation (RISA) hydrophilicity. In order to investigate gamma ray-induced and photoinduced hydrophilicity, the contact angles of water droplets on a titanium dioxide surface were measured in terms of irradiation intensity and time for gamma rays of cobalt-60 and for ultraviolet rays. Reciprocals of the contact angles increased in proportion to the irradiation time before the contact angles reached its super-hydrophilic state. The irradiation time dependency is equal to each other qualitatively. In addition, an effect of ambient gas was investigated. In pure argon gas, the contact angle remains the same against the irradiation time. This clearly indicates that certain humidity is required in ambient gas to take place of RISA hydrophilicity. A single crystal titanium dioxide (100) surface was analyzed by X-ray photoelectron spectrometry (XPS). After irradiation with gamma rays, a peak was found in the O1s spectrum, which indicates the adsorption of dissociative water to a surface 5-fold coordinate titanium site, and the formation of a surface hydroxyl group. We conclude that the RISA hydrophilicity is caused by chemisorption of the hydroxyl group on the surface.

Development of Probabilistic Fracture Mechanics Analysis Code for Pipes with Stress Corrosion Cracks

Abstract: Risk-Informed integrity management methodologies have been developed for Japanese nuclear power plants. One of the issues of concern is the reliability assessment of piping with flaws due to stress corrosion cracking (SCC). Therefore, the probabilistic fracture mechanics analysis code has been developed, which can perform the reliability assessment for austenitic stainless steel piping with flaws due to SCC. This paper describes technical basis of this code. This method is based on Monte-Carlo technique considering many sample cases in a piping section, where the initiation and growth of cracks are calculated and piping failures, including leaks and rapture, are evaluated. A notable feature is that multiple cracks can be treated, consequently, assessment of coalescence of cracks and intricate break evaluation of piping section have been included. Moreover, the in-service inspection (ISI) and integrity evaluation by Fitness-for-Service (FFS) code are integrated into the analysis, and the contribution to failure probability decrease can be assessed. Key parameters are determined on a probability basis with the designated probability type throughout the procedure. Size, location and time of crack initiation, coefficients of crack growth due to SCC and factors for piping failure are included in those parameters. With this method the reliability level of the piping through the operation periods can be estimated and the contribution of various parameters including ISI can be quantitatively evaluated.

Effects of Laser Peening Treatment on High Cycle Fatigue and Crack Propagation Behaviors in Austenitic Stainless Steel

Abstract: Laser peening without protective coating (LPwC) treatment is one of surface enhancement techniques using impact wave of high pressure plasma induced by laser pulse irradiation. One of the effects of the LPwC treatment is expected to reduce the tensile residual stress and to induce the compressive residual stress in the surface layer of metallic materials. As a laser has no reaction force due to irradiation and also it has easy characteristics for remote control, the LPwC treatment is practically used as a technique for preventing the stress corrosion cracking (SCC) and for improving the fatigue strength of some structural materials. In this study, high cycle fatigue tests with four-points rotating bending loading were carried out on the non-peened and the LPwC treated low-carbon type austenitic stainless steel 316L in order to investigate the effects of the LPwC treatment on the high cycle fatigue strength and the surface fatigue crack propagation behavior. Two types of specimens were prepared; one was a smooth specimen, the other was a specimen with a pre-crack by the fatigue loading from a small artificial hole. As the results of the LPwC treatment, the high compressive residual stress was induced in the surface layer on the specimens, and the region of the compressive residual stress was about 1mm depth from the surface. The fatigue strength of the LPwC treated SUS316L was remarkably improved during the whole regime of the fatigue life up to the 108 cycles compared with the non-peened materials. Through the fracture mechanics investigation of the pre-cracked materials after the LPwC treatment, it became clear that the fatigue crack propagation was restrained by the LPwC treatment on the pre-cracked region, when the stress intensity factor range ΔK on the crack tip was under the value of 7.6 MPa√m.Copyright © 2009 by ASME

Research on Flow Characteristics of Supercritical CO2 Axial Compressor Blades by CFD Analysis

Abstract: A supercritical CO2 gas turbine of 20MPa is suitable to couple with the Na-cooled fast reactor since Na - CO2 reaction is mild at the outlet temperature of 800K, the cycle thermal efficiency is relatively high and the size of CO2 gas turbine is very compact. In this gas turbine cycle, a compressor operates near the critical point. The property of CO2 and then the behavior of compressible flow near the critical point changes very sharply. So far, such a behavior is not examined sufficiently. Then, it is important to clarify compressible flow near the critical point. In this paper, an aerodynamic design of the axial supercritical CO2 compressor for this system has been carried out based on the existing aerodynamic design method of Cohen1). The cycle design point was selected to achieve the maximum cycle thermal efficiency of 43.8%. For this point, the compressor design conditions were determined. They are a mass flow rate of 2035kg/s, an inlet temperature of 308K, an inlet static pressure of 8.26MPa, an outlet static pressure of 20.6MPa and a rotational speed of 3600rpm. The mean radius was constant through axial direction. The design point was determined so as to keep the diffusion factor and blade stress within the allowable limits. Number of stages and an expected adiabatic efficiency was 14 and 87%, respectively. CFD analyses by FLUENT have been done for this compressor blade. The blade model consists of one set of a guide vane, a rotor blade and a stator blade. The analyses were conducted under the assumption both of the real gas properties and also of the modified ideal gas properties. Using the real gas properties, analysis was conducted for the 14th blade, whose condition is remote from the critical point and the possibility of divergence is very small. Then, the analyses were conducted for the blade whose conditions are nearer to the critical point. Gradually, divergence of calculation was encountered. Convergence was relatively easy for the modified ideal gas properties. Main output of calculation is a value of the mass flow rate, which was larger than the design value. However the discrepancy of mass flow rates between CFD and design reduced if the 3-dimensional effects are taken into consideration. Absolute velocity distributions, relative velocity distributions and static pressure distributions surrounding rotor blade and stator blade were obtained. The characteristics of these distributions were consistent with those of the fundamental theory and these analyses were justified.

Intergranular Corrosion for Extra High Purity Austenitic Stainless Steel in Boiling Nitric Acid with Cr(VI)

Abstract: Austenitic stainless steels suffer intergranular attack in boiling nitric acid with oxidants. The intergranular corrosion is mainly caused by the segregation of impurities at the grain. An extra high purity austenitic stainless steel (EHP alloys) was developed with conducting the new multiple refined melting technique in order to suppress the total harmful impurities less than 100ppm. The corrosion behavior of type 310 EHP alloy with respect to nitric acid solution with highly oxidizing ions (boiling 8kmol/m3 HNO3 solutions containing 1kg/m3 Cr(VI) ions) was investigated. The straining, aging and recrystallizing (SAR) treated type 310 EHP alloy showed superior corrosion resistance for intergranular attack than solution annealed (ST) type 310 EHP alloy with same impurity level. Boron segregation at the grain boundary was detected in only ST specimen using a Fission Track Etching method. It is believed that the segregated boron along the grain boundaries in type 310 EHP alloy was one of main factor of intergranular corrosion. The SAR treatment was effective to restrain the intergranular corrosion for type 310 EHP alloy with B less than 7ppm.

Development of Safety Assessment Code for Decommissioning of Nuclear Facilities (DecDose)

Abstract: A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute (now Japan Atomic Energy Agency). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning of the plant, and also evaluates the public dose at accidental situations including fire and explosion. The public dose at normal situations during decommissioning is evaluated from the amount of radionuclides discharged from the plant to the atmosphere and the ocean. The amounts of radionuclides discharged depend on which and how activated and/or contaminated components and structures are dismantled. The amount is predicted by using the radioactive inventory given by the plant. The filtration efficiency of the ventilation system and decontamination factors of the liquid waste treatment system of the plant are also considered. Both of the internal dose caused by inhalation and ingestion of agricultural crops and seafood, and the external dose by radioactive aerosols airborne and radioactive deposition at soil surfaces are calculated for all of possible pathways. Also included in the external dose are direct radiation and skyshine radiation from waste containers which are packed and temporarily stored in the in-site building. For external dose of workers, the radiation dose rate from dismantling contaminated components and structures is calculated using the dose rate library which was previously evaluated by a point kernel shielding code. In this condition, radiation sources are regarded to be consisted of two parts; one is a dismantling object of interest, and the other is the sum of surrounding objects. Difference in job type or position is taken into account; workers for cutting are situated closer to a dismantling component, other workers help them at some distance, and the supervisor watches their activities from away. For worker’s internal dose, the radionuclide concentrations in air for individual radionuclides are calculated from a dismantling condition, e.g. cutting speed, cutting length of the dismantling component and exhaust velocity. A calculation model for working time on dismantling was developed using more segmented WBS (work breakdown structure). DecDose was partially verified by comparison with measured the external dose of workers during JPDR Decommissioning Project. The DecDose is expected to contribute to utilities in formulating rational dismantling plans and to the safety authority in estimating conservativeness in safety assessment of licensing application or risk-based regulatory criteria.Copyright © 2009 by ASME

Development of General-Purpose Software to Analyze the Static Thermal Characteristic of Nuclear Power Plant

Abstract: We have developed the general-purpose software by which static thermal characteristic of the power generation system is analyzed easily. This software has the notable features as follows.-It has the new algorithm to solve non-linear simultaneous equations to analyze the static thermal characteristics such as heat and mass balance, efficiencies, etc. of various power generation systems.-It has the flexibility for setting calculation conditions.-It is able to be executed on the personal computer easily and quickly.We ensured that it is able to construct heat and mass balance diagrams of main steam system of nuclear power plant and calculate the power output and efficiencies of the system. Furthermore, we evaluated various heat recovery measures of steam generator blowdown water and found that this software could be a useful operation aid for planning effective changes in support of power stretch.

Continuous Operation Test at Engineering Scale Uranium Crystallizer

Abstract: Uranium crystallization based on solubility difference is one of the remarkable technologies which can provide simple process to separate uranium in nitric acid solution since the process is mainly controlled by temperature and concentration of solute ions. Japan Atomic Energy Agency (JAEA) and Mitsubishi Materials Corporation (MMC) are developing the crystallization process for elemental technology of FBR fuel reprocessing.[1–3] The uranium (U) crystallization process is a key technology for New Extraction System for TRU Recovery (NEXT) process that was evaluated as the most promising process for future FBR reprocessing.[4–6] We had developed an innovative crystallizer and carried out several fundamental investigations. On the basis of the results, we fabricated an engineering-scale crystallizer and have carried out continuous operation test to investigate the stability of the equipment at steady and non-steady state conditions by using depleted uranium. As for simulating typical failure events in the crystallizer, crystal accumulation and crystal blockage were occurred intentionally, and monitoring method and resume procedure were tried and selected in this work. As the test results, no significant phenomenon was observed in the steady state test. And in the non-steady state test, process fluctuation could be detected by monitoring of screw torque and liquid level in the crystallizer, and all failure events are proven to be recovered by appropriate resumed procedures.© 2009 ASME

Experimental Analyses by SIMMER-III on Molten Fuel Freezing and Boiling Pool Behavior

Abstract: This paper describes experimental analyses using the SIMMER-III computer code, which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code. Two topics of key phenomena in core disruptive accidents were presented in this paper: molten fuel freezing and dispersion; and boiling behavior of molten fuel pool. Related experimental database are reviewed to select appropriate experiments. To analyze the fuel freezing behavior, the GEYSER out-of-pile and the CABRI-EFM1 in-pile experiments were selected. The SIMMER-III calculations were in good agreement with fuel penetration lengths measured in a series of the GEYSER experiments. The fuel freezing behavior in the CABRI-EFM1 experiment was also reasonably simulated by SIMMER-III. The boiling pool consisting principally of molten fuel/steel mixtures is characterized by the heat transfer between fuel and steel. The CABRI-TPA2 experiment has suggested low transient heat flux from fuel to steel due to a steel vapor blanketing around a steel droplet. SIMMER-III well simulated the steel boiling behavior observed in the CABRI-TPA2 experiment by applying reduced heat transfer between fuel and steel.

Improvement of Basic Fluid Dynamics Models for the COMPASS Code

Abstract: The COMPASS code is a new next generation safety analysis code to provide local information for various key phenomena in core disruptive accidents of sodium-cooled fast reactors, which is based on the moving particle semi-implicit (MPS) method. In this study, improvement of basic fluid dynamics models for the COMPASS code was carried out and verified with fundamental verification calculations. A fully implicit pressure solution algorithm was introduced to improve the numerical stability of MPS simulations. With a newly developed free surface model, numerical difficulty caused by poor pressure solutions is overcome by involving free surface particles in the pressure Poisson equation. In addition, applicability of the MPS method to interactions between fluid and multi-solid bodies was investigated in comparison with dam-break experiments with solid balls. It was found that the PISO algorithm and free surface model makes simulation with the passively moving solid model stable numerically. The characteristic behavior of solid balls was successfully reproduced by the present numerical simulations.

Development of Observation Techniques in Reactor Vessel of Experimental Fast Reactor Joyo

Abstract: In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. And several IVO equipments for an SFR are developed. However, in order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident that occurred with Joyo, IVO using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8mm in thickness could be discriminated. The performance of IVO equipments under the actual reactor environment was successfully confirmed. And the results provided useful information on incident investigations. In addition, fundamental findings and the experience gained during this study, which included the design of equipment, operating procedures, resolution, lighting adjustments, photograph composition and the durability of the RRF under radiation exposure, provided valuable insights into further improvements and verifications for IVO techniques to secure the safety and integrity of the future SFRs.

Effect of Heater Configurations on Transient Heat Transfer for Various Gases Flowing Over a Twisted Heater

Abstract: Forced convection transient heat transfer coefficients were measured for helium gas and carbon dioxide gas flowing over a twisted heater due to exponentially increasing heat input (Q0 exp(t/τ)). The twisted platinum plate with a thickness of 0.1 mm was used as test heater and heated by electric current. The heat generation rate was exponentially increased with a function of Q0 exp(t/τ). The gas flow velocities ranged from 1 to 10 m/s, the gas temperatures ranged from 313 to 353 K, and the periods of heat generation rate ranged from 46 ms to 17 s. The surface temperature difference and heat flux increase exponentially as the heat generation rate increases with the exponential function. Transient heat transfer coefficients increase with increasing gas flow velocity. The geometric effect of twisted heater in this study shows an enhancement on the heat transfer coefficient. Empirical correlation for quasi-steady-state heat transfer was obtained based on the experimental data. The data for heat transfer coefficient were compared with those reported in authors’ previous paper.Copyright © 2009 by ASME

Development and Verification of Unstructured Adaptive Mesh Technique With Edge Compatibility

Abstract: In a design study of the large-sized sodium-cooled fast reactors in Japan (JSFR), one key issue to establish an economically superior design is suppression of a gas entrainment (GE) phenomenon at a free surface in the reactor vessel. However, the GE phenomenon is highly non-linear and too difficult to be evaluated theoretically. Therefore, we are developing a high-precision CFD method to evaluate the GE phenomenon accurately. The CFD method is formulated on an unstructured mesh to establish an accurate modeling for a complicated shape of the JSFR system. As a two-phase flow simulation method, a high-precision volume-of-fluid algorithm is employed in the CFD method. In addition, physically appropriate formulations at gas-liquid interfaces are introduced into the CFD method. The developed CFD method is already applied to the simulation of a GE phenomenon in a basic GE experiment and the simulation results show good agreement with experimental results. Therefore, it is confirmed that the proposed CFD method can reproduce a GE phenomenon. However, for the simulation of the GE phenomenon in the JSFR, we still have one problem on a mesh subdivision. Though a fine mesh subdivision has to be applied to the regions where the GE occurs, it is difficult to preliminarily know the regions because the GE occurrence is strongly affected by a local instant flow pattern, i.e. a vortex generation. Therefore, an adaptive mesh technique is necessary to apply a fine mesh subdivision automatically to only the local GE occurrence regions in the large-sized JSFR. In this study, as one part of an adaptive mesh development, a two-dimensional unstructured adaptive mesh technique is developed and verified. In the proposed two-dimensional adaptive mesh technique, each cell is isotropically subdivided to reduce distortions of the mesh. In addition, a connection cell is formed to eliminate the edge incompatibility between a refined and a non-refined cells. A connection cell has several subdivision patterns and one of them is selected to be compatible with adjacent cells on every cell edge. Finally, the present unstructured adaptive mesh technique is verified by solving well-known driven cavity problem. As the result, the present unstructured adaptive mesh technique succeeds in providing a high-precision solution, although we employ a poor-quality distorted mesh at the initial state. In addition, the simulation error on the unstructured adaptive mesh at the steady state is much less than the error on the structured mesh consisting of a larger number of cells.Copyright © 2009 by ASME

Advanced Recycling Core Accommodating Oxide Fuel and Metal Fuel for Closed Fuel Cycle

Abstract: This report presents a unique TRU burning core capable of accommodating oxide fuel and metal fuel and easy to change oxide core to metal core conforming to the design requirements. For the homogeneous oxide fueled core containing transuranics (TRU) fuel with 12% of the moderator pins, the results of calculation show the TRU conversion ratio (ratio of loss of TRU to loss of heavy metal) of 0.33 and the TRU burning capability (ratio of loss of TRU per electric generation) of 67 kg/TWeh. On the other hand, the calculations replacing from oxide fuel assemblies to metal fuel assemblies have indicated the TRU transmutation capability of 69 kg/TWeh with the TRU conversion ratio of 0.30. As the result of simulation calculations, three ordinary fuel exchanges transform the oxide equilibrium core to the full metal core by way of transitional cores, where the maximum linear heat rates are still equal to the metal equilibrium core or less. With this, the presented core concept is concluded that a full oxide core, a full metal core, mixed fueled cores can be materialized in the presented first unit of Advanced Recycling Reactor (ARR1).

Moisture Separator Reheater for NPP Turbines

Abstract: This paper introduces the development of the current model Moisture Separator Reheater (MSR) for nuclear power plant (NPP) turbines, commercially placed in service in the period 1984-1997, focusing on the mist separation performance of the MSR along with drainage from heat exchanger tubes. A method of predicting the mist separation performance was devised first based on the observation of mist separation behaviors under an air-water test. Then the method was developed for the application to predict under the steam conditions, followed by the verification in comparison with the actual results of a steam condition test. The instability of tube drainage associated with both sub-cooling and temperature oscillation might adversely affect the seal welding of tubes to tube sheet due to thermal fatigue. The instability was measured on an existing unit to clarify behaviors and the development of a method to suppress them. Both methods were applied to newly constructed units and the effectiveness of the methods was demonstrated.

Current Status and Performance of the J-PARC Accelerators

Abstract: J-PARC (Japan Proton Accelerator Research Complex) is a multi-purpose research facility for materials and life sciences, nuclear and particle physics, and nuclear engineering with extremely high power proton beams of 1 MW. The accelerator complex consists of a 400-MeV linac, a 3-GeV Rapid Cycling Synchrotron (RCS), and a 50-GeV Main Ring synchrotron (MR). Its goals are to provide MW-class beams at 3 GeV and at several 10 GeV, while it is a challenge to localize and suppress beam loss to the level to allow hands-on maintenance of accelerator components. The RCS scheme is adopted to realize them, which is advantageous over conventional Accumulation Ring (AR) regarding less beam loss problems due to lower beam current and easier construction and operation of a linac. RCS, however, required various challenging technologies such as ceramic ducts to reduce eddy current effects, high field Radio Frequency (RF) system, and paint injection technique (an injection scheme to reduce phase space density of the beam) to reduce space charge effects. The linac has also unique technologies to minimize beam loss, such as compact electromagnet Drift Tube Quadrupoles (DTQ’s) to control beam envelopes precisely, and a fast beam suspending system in Machine Protection System (MPS) with Radio Frequency Quadrupole linac (RFQ). The beam commissioning of the linac started in Nov. 2006, and its design energy of 181 MeV in the first construction phase was achieved in Jan. 2007. RCS beam commissioning started in Sep. 2007 and the beam was accelerated to the designed energy of 3 GeV in Oct. 2007. MR beam commissioning started in May 2008, and the beam acceleration to 30 GeV was established in Dec. 2008. The first neutron and muon beams were produced in May and Sep. 2008, respectively, at Materials and Life science experimental Facility (MLF). The linac commissioning has resulted in very stable beam with short down time. RCS commissioning quickly achieved beam acceleration and extraction, and paint injections are being studied intensively. RCS recorded the highest beam power of 0.21 MW in Sep. 2008 with beam loss well localized at the collimators. The linac beam energy will be upgraded to 400 MeV with Annular Coupled Structure linac (ACS) in order to increase the beam power to 1 MW. In the second construction phase, upgrade of the linac with 600-MeV Super-Conducting Linac (SCL) for Accelerator-Driven nuclear waste transmutation System (ADS) and upgrade of MR energy from 30 to 50 GeV are planned.Copyright © 2009 by ASME

Numerical Simulation of Bubble Motion about a Grid Spacer in a Rod Bundle

Abstract: Numerical simulations based on a three-dimensional two-way bubble tracking method are carried out to predict bubble motions in a square duct with an obstacle and in a two-by-three rod bundle with a grid spacer. Comparisons between measured and predicted bubble motions demonstrate that the two-way bubble tracking method gives good predictions for trajectories of small bubbles in the upstream side of the grid spacer in the rod bundle geometry. The predicted bubble trajectories clearly show that bubbles are apt to migrate toward the rod surface in the vicinity of the bottom of the grid spacer. Analysis of forces acting on the bubbles confirms that pressure gradient force induced by the presence of the spacer is the main cause of the bubble lateral migration toward the rod surface. Motions of steam bubbles at a nominal operating condition of a pressurized water reactor (PWR) are also predicted by using the bubble tracking method, which indicates that steam bubbles also migrate toward the rod surface at the upstream side of the spacer due to the spacer-induced pressure gradient force.

Experimental Research on the Effect of Axial Power Distribution on Critical Power

Abstract: This paper concerns experimental research to ascertain the effect of axial power distribution on critical power in the positive quality region. Experiments took place at atmospheric pressure in a circular tube. Axial uniform heating and two other axial non - uniform heating cases were selected for detailed evaluation. The effects of relative power ratio on critical power, critical quality and critical boiling length were ascertained in detailed evaluations. Using the experimental data, we evaluated existing correlating concepts with critical power. Result showed a combination of the overall power concept (χBT - LB) and the local conditions concept (χBT - qBT) appearing to be promising in correlating present critical power data in axial non - uniform heating conditions.

A Design Study of High Breeding Ratio Sodium Cooled Metal Fuel Core without Blanket Fuels

Abstract: The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650 oC, and the maximum fuel pin bundle pressure drop of 0.4MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40.

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan

Abstract: According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs) During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core The increased burnup reactivity may reduce the cycle length of an FBR We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs The reference specifications were selected as follows Output of 1500MWe and average discharge fuel burnup of about 150GWd/t Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5% The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5% This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core

Development of Short Stroke Shearing Technology for FBR Fuel Pins

Abstract: The short stroke shearing tests with simulated fuel pin bundle were carried out in engineering scale. The shearing device was designed to handle the simulated Monju (FBR prototype reactor) type fuel pin bundle. Monju type and Commercial reactor type simulated fuel pins were used for the test. The length of sheared pin and the opening ratio of sheared section were measured under several shearing settings such as the pressure to hold pin bundle, the shearing speed and the filling-ratio of pins in the pin magazine. Both types of fuel pin were able to be sheared accurately at the length of about 10mm, and the opening ratio of sheared section was not significantly reduced. As the results, fundamental data of the short stroke shearing characteristics were obtained and that shearing method was confirmed to be promising with the reliable shearing device.Copyright © 2009 by ASME

Evaluation of MONJU Core Damage Risk with Change of AOT Using Probabilistic Method

Abstract: MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops which can supply 280 MW of electricity. Limiting conditions of operation (LCO) defined in the safety regulations in MONJU given the allowed outage time (AOT) are evaluated using a PSA technique. The result indicates the possibility of limit extension and some prospects that we should examine.Copyright © 2009 by ASME