The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were also found to perform well for the rod bundle under BWR conditions.

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Three-dimensional calculations of ventilation air flow and thermal-hydraulic behavior in a spent fuel pit (SFP) were made using the CFD software, FLUENT6.3.26 to evaluate the heat loss and water temperature in the SFP after shutdown of its cooling systems. The air and water velocities near the water surface were evaluated from the calculated results and referred to conditions of evaporation heat transfer tests, which were carried out at Shinshu University. From the test data, a correlation for evaporation heat fluxes was introduced and incorporated into the calculation of thermal-hydraulic behavior in the SFP. Then, a three-dimensional calculation of thermal-hydraulic behavior in the SFP was done. It was confirmed that the higher the water temperature was, the larger the heat loss from water was, and that the major heat loss was the evaporation heat transfer from the water surface to ventilation air, which was about ten times larger than the heat transfer to concrete walls.

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Introducing compressive residual stress by a cavitating jet into the sub-surface of components used in nuclear power plants can mitigate stress corrosion cracking in these components. Although applying the jet is an effective method for this purpose, it should be used without causing damage to the surface from water jet droplets arising from high-pressure injection of the water jet. Thus, in introducing compressive residual stress, the injection pressure needs to be optimized. In this paper, in order to determine the optimum injection pressure, the residual stress of stainless steel treated by a jet at various injection pressures was measured using an X-ray diffraction method. The injection pressure of the jet was varied from 5 MPa to 300 MPa, and the diameter of the nozzle throat of the jet was varied from 0.35 mm to 2.0 mm. The variation of residual stress with depth was measured by alternating X-ray diffraction measurements with electropolishing. It was revealed that a cavitating jet at an injection pressure of 10 MPa with a nozzle diameter of 2.0 mm can introduce higher compressive residual stress to deeper into stainless steel compared with a jet at 300 MPa with a nozzle diameter of 0.35 mm when the downstream pressure of the nozzle was constant.

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Computational study of the sodium-water reaction at the gas (water) - liquid (sodium) interface has been carried out using the ab initio (first-principle) method. A possible reaction channel has been identified for the stepwise OH bond dissociations of a single water molecule. The energetics including the binding energy of a water molecule on the sodium surface, the activation energies of the bond cleavages, and the reaction energies, have been evaluated, and the rate constants of the first and second OH bond-breakings have been compared. It was found that the estimated rate constant of the former was much larger than the latter. The results are the basis for constructing the chemical reaction model used in a multi-dimensional sodium-water reaction code, SERAPHIM, being developed by Japan Atomic Energy Agency (JAEA) toward the safety assessment of the steam generator (SG) in a sodium-cooled fast reactor (SFR).

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Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

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Open solicitation system of research and development ideas in The Japan Atomic Power Company (JAPC) is introduced. This system is established in 1999 to solve various subjects existing in JAPC's nuclear power stations by using highly advanced technique possessed by the enterprises or the organizations limited in Fukui Prefecture, and also to contribute them to become more advanced in their skill of technology. We have improved this system to make the results more applicable. We think it is important to coexist with the local community for the company engaged in nuclear energy as to cooperate together for the improvement of the local society.

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Since the beginning of this century, the so-called 3Ss (Nuclear Safety, Nuclear Security and Safeguards) have become major regulatory areas for peaceful uses of nuclear energy. In order to rationalize the allocation of regulatory resources, interrelationship of the 3Ss should be investigated. From the viewpoint of the number of the parties concerned in regulation, nuclear security is peculiar with having “aggressors” as the third party. From the viewpoint of final goal of regulation, nuclear security in general and safeguards share the goal of preventing non-peaceful uses of nuclear energy, though the goal of anti-sabotage within nuclear security is rather similar to nuclear safety. As often recognized, safeguards are representative of various policy tools for nuclear non-proliferation. Strictly speaking, it is not safeguards as a policy tool but nuclear non-proliferation as a policy purpose that should be parallel to other policy purposes (nuclear safety and nuclear security). That suggests “SSN” which stands for Safety, Security and Non-proliferation is a better abbreviation rather than 3Ss. Safeguards as a policy tool should be enumerated along with nuclear safety regulation, nuclear security measures and trade controls on nuclear-related items. Trade controls have been playing an important role for nuclear non-proliferation. These policy tools can be called “SSST” in which Trade controls are also emphasized along with Safety regulation, Security measures and Safeguards.

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An advanced core protection system for a pressurized water reactor, Reactor Core Protection System(RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%.

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The composition of the BN film was determined using Rutherford backscattering spectrometry (RBS) and nuclear reaction analysis (NRA). RBS can provide all the constituent concentrations in the films and the absolute determination of the number of atoms/cm^2. However RBS is not suited to detection of light atoms deposited on a substrate material of higher atomic mass. On the other hand, the NRA has the advantage that it allows to measure the areal concentrations of nitrogen and boron in BN_x films on Si substrates, although calibration is required using standard specimens. These experiments were carried out on the 2 MeV Van de Graaff accelerator connected to an ultra high vacuum (UHV) chamber. For RBS measurement, a ⁴₂He+ beam at 2.0 MeV, a total scattering angle of 168° and a beam incident angle to the substrate normal or 60 deg. were used. Zr and Pt films, 1150 Å to 3300 Å in thickness, deposited on vitreous carbon plates were used as a substrate. NRA was performed using a deuteron beam of 1.7 MeV and a beam incident angle of 20 deg. A peak from ¹⁰B(d,α)⁸Be in an NRA spectrum of a standard sample appeared clearly without significant background, however a broad signal from ¹¹B(d, α)⁹Be appeared overlapping with a peak from ¹⁴N(d, α)¹²C. Therefore the ¹⁰B(d, α)⁸Be and ¹⁴N(d, α)¹²C yields were estimated, since the ratio of ¹¹B : ¹⁰B measured by RBS was 0.83 : 17, which is well consistent with the natural isotopic ratio, ¹¹B : ¹⁰B =0.802 :0.192. In the case of calculating the ¹⁴N(d, α)¹²C yields, the signal from ¹¹B(d, α)⁹Be was deconvoluted by taking into account the shape of ¹¹B(d, α)⁹Be signal. The areal ratio ¹⁴N/¹⁰B was 7.73 and the error was -3.5 to +3.2%. These values will be used for determining composition of BN films. The conversion factor allows obtaining the composition of BN thin films on Si substrate.

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A high alkali domain spreads out due to the use of cement materials for the construction of the repository of radioactive wastes. Sudden change of pH at this alkali front produces colloidal silicic acid (polymeric silicic acid) in addition to the deposition of supersaturated monomeric silicic acid onto the fracture surface of flow-pathway. The colloidal silicic acid also deposits with relatively small rate-constant in the co-presence of solid phase. Once the flow-path surface is covered with the amorphous silica, the surface seriously degrades the sorption behavior of radionuclides (RNs). Therefore, so far, the authors have examined the deposition rates of supersaturated silicic acid. This study summarized the deposition rate-constants defined by the first-order reaction equation under various conditions of co-presence of amorphous silica powder. Then, using the smallest rate-constant (1.0×10^-12 m/s in the co-presence of calcium ions of 1 mM) and a simulation code, COLFRAC-MRL, the spatial range of colloidal silicic acid deposited downstream from the alkali front was estimated. The results suggested the clogging caused by the deposition of colloidal silicic acid in flow-path. The altered spatial range in the flow-path was limited to around 30 m in fracture and to several centimeters in rock matrix.

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Suppression of gas entrainment (GE) phenomena caused by free surface vortices are very important to establish an economically superior design of the sodium-cooled fast reactor in Japan (JSFR). However, due to the non-linearity and/or locality of the GE phenomena, it is not easy to evaluate the occurrences of the GE phenomena accurately. In other words, the onset condition of the GE phenomena in the JSFR is not predicted easily based on scaled-model and/or partial-model experiments. Therefore, the authors are developing a CFD-based evaluation method in which the non-linearity and locality of the GE phenomena can be considered. In the evaluation method, macroscopic vortex parameters, e.g. circulation, are determined by three-dimensional CFD and then, GE-related parameters, e.g. gas core (GC) length, are calculated by using the Burgers vortex model. This procedure is efficient to evaluate the GE phenomena in the JSFR. However, it is well known that the Burgers vortex model tends to overestimate the GC length due to the lack of considerations on some physical mechanisms. Therefore, in this study, the authors develop a turbulent vortex model to evaluate the GE phenomena more accurately. Then, the improved GE evaluation method with the turbulent viscosity model is validated by analyzing the GC lengths observed in a simple experiment. The evaluation results show that the GC lengths analyzed by the improved method are shorter in comparison to the original method, and give better agreement with the experimental data.

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The paper considers a Rod Ejection Accident (REA) which has been calculated by the coupled-code system ATHLET-QUABOX/CUBBOX. For the present study, a MOX/UOX mixed core loading was developed on the basis of a generic PWR. The results are particularly focused on the fuel enthalpy rise which is the main safety criterion for such transient. A parametric REA study has been performed, showing the influence of some important thermal-hydraulic and neutron-physical parameters. Simulations have been performed using realistic or artificially decreased delayed neutron fractions for two different core states (HZP and 30% of the nominal power). Effective fuel rod temperature influence (i.e. Doppler coefficient) has been studied by using different correlations (0.5/0.5 weighting factors or the typical T_Doppler = 0.7 T_Surface + 0.3 T_Center) or by changing the fuel gap conductance. It is shown that the maximum enthalpy (and enthalpy increase) does not always appear in the affected fuel assembly but can also appear in the neighboring ones. This result is a direct consequence of the burn up dependence of the enthalpy. The paper also considers the case of local delayed neutron parameters and briefly describes the future REA studies foreseen at GRS such as an investigation of quantitative uncertainty propagation from the nuclear data to the transient behavior.

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In order to meet fast development of nuclear power, China should establish its own nuclear power standards system. This paper gives some opinions on establishment of Chinese nuclear island systems and components design and construction standards. It is suggested to draft “Chinese Nuclear Power Utility Requirements Document”

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In the experimental fast reactor Joyo, loading of irradiation test rigs causes a decrease in excess reactivity because the rigs contain less fissile materials than the driver fuel. In order to carry out duty operation cycles using as many irradiation rigs as possible, it is necessary to upgrade the core performance to increase its excess reactivity and irradiation capacity. Core modification plans have been considered, such as the installation of advanced radial reflectors and reduction of the number of control rods. To implement such core modifications, it is first necessary to improve the prediction accuracy in core design and to optimize safety margins. In the present study, verification of the JUPITER fast reactor standard analysis method was conducted through a comparison between the calculated and the measured Joyo MK-III core characteristics, and it was concluded that the accuracy for a small sodium-cooled fast reactor with a hard neutron spectrum was within 5 % of unity. It was shown that, the performance of the irradiation bed core could be upgraded by the improvement of the prediction accuracy of the core characteristics and optimization of safety margins.

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In a PWR plant, a steam-water two-phase flow may possibly exist in the pressurizer spray pipe under a normal operating condition since the flow rate of the spray water is not sufficient to fill the horizontal section of the pipe completely. Initiation of high cycle fatigue cracks is suspected to occur under such thermally stratified two phase flow conditions due to cyclic thermal stress fluctuations caused by oscillations of the water surface. Such oscillations cannot be detected by the measurement of temperature on outer surface of the pipe. In order to clarify the flow and thermal conditions in the pressurizer spray pipe and assess their impact on the pipe structure, an experiment was conducted for a steam-water flow at a low flow rate using a mock-up pressurizer spray pipe. The maximum temperature fluctuation of about 0.2 times of the steam-water temperature difference was observed at the inner wall around water surface in the test section. Visualization tests were conducted to investigate the temperature fluctuation phenomena. It was shown that the fluid temperature fluctuations were not caused by the waves on the water surface, but were caused by liquid temperature fluctuations in water layer below the interface. The influence of small amount of non-condensable gas dissolved in the reactor coolant on the liquid temperature fluctuation phenomena was investigated by injecting air into the experimental loop. The air injection attenuated the liquid temperature fluctuations in the water layer since the condensation was suppressed by the non- condensable gas. It is not expected that wall temperature fluctuation in the actual PWR plant may exceed the temperature equivalent to the fatigue limit stress amplitude when it is assumed to be proportional to the steam-water temperature difference.

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Flow induced vibration in primary cooling system of the Japan Sodium cooled Fast Reactor (JSFR) has been investigated. The primary cooling system consists of a large diameter pipe and a pipe elbow with short curvature radius corresponding to its diameter (short-elbow). Flow-induced vibration by flow through the short-elbow is an important issue in design study of the JSFR, because it may affect to structural integrity of the piping. In this paper, numerical simulations for several pipe elbows with different pipe diameters and curvature radii in literature were conducted at Reynolds number conditions from Re=500 to 1.47x10⁷ to investigate unsteady flow behavior through the short-elbow, including validation study of an in-house LES code (MUGTHES). Numerical results in each condition were compared with the experimental results in literature. Unsteady flow characteristics and pressure fluctuation generation mechanism in the short-elbow were clarified in relation to the large-scale eddy motion.

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The effect of turbulence on the dynamics of three-dimensional dam break flow is numerically investigated on the basis of the incompressible Reynolds-averaged Navier-Stokes (RANS) equations with the volume of fluid (VOF) function. It is found that the tip velocity over the ground and the impact pressure on the vertical wall in the Launder-Gibson (LG) model are in good agreement with experimental results. % The dynamics of the dam break flow is subject to the viscous dissipation during the collapse of the flow, which is underestimated in the laminar model and overestimated in the realizable k-ε (RKE) model. The turbulent viscous dissipation near the free surface is comparable to that in the water in the LG model.

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Thermodynamic consequence in liquid sodium leak and fire accident is one of the important issues to be evaluated when considering the safety aspect of fast reactor plant building. The authors are therefore initiating systematic verification and validation (V&V) activity to assure and demonstrate the reliability of numerical simulation tool for sodium fire analysis. The V&V activity is in progress with the main focuses on already developed sodium fire analysis codes SPHINCS and AQUA-SF. The events to be evaluated are hypothetical sodium spray, pool, or combined fire accidents followed by thermodynamic behaviors postulated in a plant building. The present paper describes that the ‘Phenomena Identification and Ranking Table (PIRT)’ is developed at first for clarifying the important validation points in the sodium fire analysis codes, and that an ‘Assessment Matrix’ is proposed which summarizes both separate effect tests and integral effect tests for validating the computational models or whole code for important phenomena. Furthermore, the paper shows a practical validation with a separate effect test in which the spray droplet combustion model of SPHINCS and AQUA-SF predicts the burned amount of a falling sodium droplet with the error mostly less than 30%.

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This paper describes the achievements of a program in which technology education is provided to cultivate practical core engineers for low-level radiation. It was made possible by means of (1) an introductory education program starting at an early age and a continuous agenda throughout college days and (2) regional collaboration. First, with regard to the early-age introductory education program and the continuous education agenda, the subjects of study related to atomic energy or nuclear engineering were reorganized as “Subjects related to Atomic Power Education” for all grades in all departments. These subjects were included in the syllabus and the student guide book, emphasizing a continuous and consistent policy throughout seven-year college study, including the five-year system and additional two-year advanced course. Second, to promote practical education, the contents of lectures, experiments, and internships were enriched and realigned in collaboration with the Japan Atomic Energy Agency, Okayama University and The Cyugoku Electric Power Co., Inc. In addition to the expansion and rearrangement of atomic power education, research on atomic power conducted for graduation thesis projects were undertaken to enhance the educational and research activities. In consequence, it has been estimated that there is now a total of fourteen subject areas in atomic energy technology, more than eight-hundred registered students in the department, and thirteen members of the teaching staff related to atomic energy technology. Furthermore, the “Tsuyama model” is still being developed. This program was funded by the Ministry of Education, Culture, Sports, Science and Technology.

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Two-phase critical flow in the nozzle tube is analyzed numerically by the best estimate code TRACE and the CFD code FLUENT, and the performance of the mass flow rate estimation by the numerical codes is discussed. For the best estimate analysis by the TRACE code, the critical flow option is turned on. The mixture model is used with the cavitation model and the evaporation-condensation model for the numerical simulation by the FLUENT code. Two test cases of the two-phase critical flow are analyzed. One case is the critical flashing flow in a convergent-divergent nozzle (Super Moby Dick experiment), and the other case is the break nozzle flow for a steam generator tube rupture experiment of pressurized water reactors at Large Scale Test Facility of Japan Atomic Energy Agency. The calculation results of the mass flow rates by the numerical simulations show good agreements with the experimental results.

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Several studies show that data assimilation methods, by merging information both from model and measurement can be used to elaborate an optimal determination of neutronic activity field within a nuclear PWR reactor core. Here the problem addressed and solved is to determine an optimal repartition of the instruments, used to make measurements within the core, to get the best possible reconstructed field using a data assimilation procedure. The position optimization is based on simulated annealing, with a Metropolis-Hasting algorithm. To perform the method in the framework of data assimilation, algebraic optimisation related to computing time of data assimilation has been developed and is used here.

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Inspired from meteorological applications, data assimilation techniques can be used to perform an optimal reconstruction of the neutronic field in a nuclear reactor core. Measurements and simulation information, coming from a numerical model, are merged together to build a better estimation of the whole field. In this paper, we first study the robustness of the method when the amount of measured information varies. We then study the influence of the nature of the instruments and their spatial repartition on the efficiency of the field reconstruction. This study also highlights the instruments providing most information within a data assimilation procedure. The study of various network configurations of instruments in the nuclear core establishes that the influence of each instrument depends both on the individual instrumentation location as well as on the chosen network.

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In this work Density Wave Oscillations (DWO) and Ledinegg instabilities are analyzed for boiling and condensing systems in a single tube. The analysis is based on a numerical model solved with a least squares spectral element method which is characterized by negligible numerical diffusion and high accuracy. Stability limits are constructed and compared with available correlations. The analysis is extended to sub-cooled, saturated and over-heated inlet conditions. Finally a discussion regarding the occurrence of these phenomena in condensing systems is presented.

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As a fundamental study for the direct contact heat exchange which was employed for in-vessel heat exchange in the Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR) and for the steam generator tube rupture (SGTR) accident in lead alloy-cooled fast reactor (LFR), ethanol jet was injected into high temperature fluorinert (FC-3283) as a simulation experiment in order to investigate the jet boiling phenomena just after volatile water contacting with the high temperature continuous lead alloy liquid. Two series of tests (no-boiling and boiling) were initiated to evaluate the ethanol vapor volume which generated around the ethanol jet. From synchronized temperature measurement around ethanol jet, the overview of the boiling behavior showed that jet boiling occurred at bottom part of jet first and developed to the upper part within very narrow area around jet.

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The feasibility study of a liquid lithium type proton beam target was performed for the neutron source of the boron neutron capture therapy (BNCT). As the candidates of the liquid lithium target, a thin sheet jet and a thin film flow on a concave wall were chosen, and a lithium flow experiment was conducted to investigate the hydrodynamic stability of the targets. The surfaces of the jets and film flows with a thickness of 0.5 mm and a width of 50 mm were observed by means of photography. It has been found that a stable sheet jet and a stable film flow on a concave wall can be formed up to certain velocities by using a straight nozzle and a curved nozzle with the concave wall, respectively.

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The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal- hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method.

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