A fuelling study for CANDU reactors is conducted using natural uranium (NU) fuels doped with trace amounts of burnable neutron absorbers. The burnable absorbers of interest include gadolinium oxide (Gd_2O_3) and europium oxide (Eu_2O_3). The study incorporates fuel-lattice simulations as well as refuelling and core-following simulations to quantify the impact in the in-core behavior of the fuel. The fuel lattice simulations were conducted using the WIMS-AECL code while refuelling and core-following simulations were conducted using the Reactor Fuelling Simulation Program (RFSP). This paper presents the improvements in the safety margins gained by the use of burnable absorbers during normal operation.

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The SGTR studies correspond to the double-ended guillotine rupture of a steam generator tube, which allows unimped blowdown from both ends of the severed tube. Occurrence of the SGTR accident leads to an increase in contamination of the secondary system as well as activity release to atmosphere due to leakage of radioactive coolant from the RCS. Therefore the automatic or operator actions should be taken in time to terminate the leakage. For ACP1000, the time window of the operator action is prolonged by adding new protection signals and equipment. In present work, the maximal time interval of the operator action after the SGTR is studied with these new measures. And after the operator takes actions, in accordance with the procedure, no SG overflow takes place till the pressure of the primary side is equal to that of the secondary side.

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The IAEA fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. Therefore, in order to prove the applicability of 100 TBq Cs-137 release into environment as a safety criterion at reactor design approval stage, the limitedness of the consequences of the release to both people and the environment must be demonstrated. As it was shown by preceding study that the release generates limited health effects, this study is dedicated to demonstration of limitedness of the consequences of the release to the environment. We formed the environmental impact index based on insights from our previous studies. It is composed of decontamination index and relocation index. It is used to assess consequences to the environment under several conditions. The estimated environmental impact indices under all conditions were small enough to confirm the limitedness of the environmental impacts of the 100 TBq Cs-137 release, which ensure its applicability as a safety criterion for consequence assessment at reactor design approval stage.

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A fuelling study for CANDU reactors is conducted using natural uranium (NU) fuels doped with trace amounts of burnable neutron absorbers. The burnable absorbers of interest include gadolinium oxide (Gd_2O_3) and europium oxide (Eu_2O_3). The study incorporates fuel-lattice simulations as well as refuelling and core-following simulations to quantify the impact in the in-core behavior of the fuel. The fuel lattice simulations were conducted using the WIMS-AECL code while refuelling and core-following simulations were conducted using the Reactor Fuelling Simulation Program (RFSP). This paper presents the improvements in the safety margins gained by the use of burnable absorbers during normal operation.

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The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test.

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In sodium-cooled fast reactors (SFRs), since the coolant does not need to be pressurized, a pipe break due to the internal pressure does not occur physically. For safety margin in Japanese prototype fast breeder reactor (Monju), the guillotine pipe break accident, i.e., loss of piping integrity (LOPI) has been analyzed as an extreme assumption for beyond design basis accidents (B-DBAs) in the licensing application for the construction permit. The cooling capability of the core was re-evaluated in this paper during a large-scale, more specifically guillotine pipe break at the primary heat transport system (PHTS) in Monju, newly considering the following latest findings: a. Experimental data on sodium boiling in fuel assemblies, b. Actual PHTS pump coast-down characteristics, and c. Transient burst test data on irradiated fuel claddings. The analysis models were validated and simulations were re-performed also using the actual Monju data such as the response time to the trip signals, etc. As a result, it was clarified that the ratio of failed fuel claddings does not exceed around 3% of all of fuel assemblies, as in the past licensing analysis. The safety has been reconfirmed to be secured without significant core damage even under an extreme assumption of a double-ended guillotine pipe break at the PHTS in Monju.

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In this work, theory analysis about K-H instability of film interface in swirling flow field in cyclone separator was performed. The momentum equations and continuity equations of each phase fluid were linearized by substitution of potential function firstly. Then the dynamic boundary condition and kinematic boundary condition were obtained based on stress analysis of film interface. According to the linearized equation and boundary conditions, the dispersion relation was established. Additionally, the motion law of film was obtained based on principle of stress equilibrium. Then the criterion for interface K-H instability was obtained. A computer program was developed according to theory model. Then simulation and analysis was performed on interface K-H instability under different conditions. It was found that centrifugal force of film could constrain interface K-H instability of film but the centrifugal force of steam could cause K-H instability. Additionally, film interface tended to unstable state with the increase in film thickness. It was also found that the increase in steam velocity could constrain interface instability when rise angle of swirl-vane of separator was less than a certain value. But the increase in steam velocity could cause instability when rise angle exceeded a certain value.

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The spallation target is an important component of the Chinese Initiative Accelerator Driven Sub-critical System(CIADS). The Lead-Bismuth-Eutectic(LEB) is used as the target material for spallation reaction and coolant of the target. The heat removal from the target window is a big challenge. The thermal-hydraulics design of the LBE spallation target system were done, including the target zone and the sub-system of heat removal. Commercial CFD software FLUENT was used to simulate the flow field and temperature distribution in the target zone. Some 3-D target models with single entrance, double-entrance and four-entrance were built to analyze the flow field of the full-size target tube. An LBE target loop with outside pipe, pump and heat exchanger was designed. An electromagnetic pump was chosen to drive the LBE target loop. The best-estimated thermal-hydraulic code RELAP5 was used to simulate the loop. The results showed that the design of the LBE target system is feasible from the point of view of thermal-hydraulics.

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Wire-mesh mist eliminator is widely applied in sea water desalination apparatus because of the high economy and efficient removal of liquid droplets from vapor. Open literature on the research of demister performances with the wet working condition at atmosphere is limited. With this article, an experimental test apparatus and device were designed and established. The layered type eliminators made of stainless steel wires with the size of 0.1×0.4mm were 80mm in diameter. The study on steam vapor pressure drop of the wire mesh demister as a function of operating condition and design parameters was carried out. These factors include vapor velocity (2-10m/s), pad thickness (50-190mm), layer spacing (0.5-2mm). Experiments working medium was vapor-liquid at atmospheric pressure. The total pressure drop were found to increase with the increasing of vapor velocity, pad thickness and packing density, the specific pressure drop of each layer mesh pad declined until the vapor velocity increased and reached certain value and then increased with a further growth in vapor velocity. The optimum layer spacing of demister varies varies from 1-1.5mm under the experimental conditions.

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The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300 ℃ and total integrated dose (TID) of 1MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions.

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The Iterated Fission Probability (IFP) method, an accurate method to estimate adjoint-weighted quantities in the continuous-energy Monte Carlo criticality calculations, has been widely used for calculating kinetic parameters and nuclear data sensitivity coefficients. By using a strategy of waiting, however, this method faces the challenge of high memory usage to store the tallies of original contributions which size is proportional to the number of particle histories in each cycle. Recently, the Wielandt method, applied by Monte Carlo code McCARD to calculate kinetic parameters, estimates adjoint fluxes in a single particle history and thus can save memory usage. In this work, the Wielandt method has been applied in Rector Monte Carlo code RMC for nuclear data sensitivity analysis. The methodology and algorithm of applying Wielandt method in estimation of adjoint-based sensitivity coefficients are discussed. Verification is performed by comparing the sensitivity coefficients calculated by Wielandt method with analytical solutions, those computed by IFP method which is also implemented in RMC code for sensitivity analysis, and those from the multi-group TSUNAMI-3D module in SCALE code package.

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Reactor Boron and Water Makeup System (REA) is one of the most important auxiliary systems of PWR nuclear power plant, whose normal operating modes involve auto-filling, manual-filling, rapid dilution, slow-footed dilution and boronization. The realtime simulation on REA fluid network with high accuracy can be used to train operators. The analysis of dynamic response simulation results can provide theoretical basis for REA design and control strategy optimization. GSE Systems, Inc. is a great simulation platform, which develops and markets software-based simulation and training products. Jtopmeret is the two - phase dynamic simulation tool of GSE, which provides the graphics modeling simulation industry with high fidelity dynamic models. Moreover, models can be developed and improved manually using Compaq Visual Fortran language on the GSE platform. In this paper, we introduce the function, composition, operating conditions and control theories of REA. Furthermore, the REA model of Daya Bay nuclear power plant is built using Jtopmeret and FORTRAN language. The results indicate that the steady errors of the main parameters are less than 2% under design conditions, which verifies the availability and reliability of GSE and Jtopmeret. The dynamic response results are reasonable under various transient operating conditions.

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The behavior of droplet impacting onto dry solid surface and wetted surface has great influence on separating efficiency of the steam separator in steam and power system. Firstly, a review and comparison of published experimental results pertaining to splashing thresholds and mechanisms is summarized, mainly varying the surface condition: increasing the dimensionless surface roughness R_<nd> on dry solid surfaces or the dimensionless film H* on wetted surfaces gradually. Then impact of droplets with different velocity on wetted surfaces with three kinds of impact angles was studied experimentally using a high-speed camera at 11,000 fps. Experiments were performed at normal pressure and temperature. The impacting outcomes of deposition, crown without splashing, splashing are displayed and discussed. Finally, the thresholds between non-splashing and splashing regions are studied quantitatively.

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Kuosheng Nuclear Power Plant (NPP) is located on the northern coast of Taiwan. Its nuclear steam supply system is a type of BWR/6 designed and built by General Electric. There were three main steps in this study. First, Kuosheng NPP TRACE/SNAP model was developed in this research. The containment was also simulated in this model. In order to assess the system response of the Kuosheng NPP TRACE/SNAP model, this study used startup tests data (load rejection and a feedwater pump trip transients) to compare with the results of TRACE. Second, the transient analysis of Kuosheng NPP TRACE/SNAP model under Fukushima-like (SBO) or more severe (SBO+LOCA) conditions was performed. Third, in order to confirm the mechanical property and integrity of fuel rods, FRAPTRAN analysis was also performed in this study. In the comparison of startup tests and TRACE data, the results and sequences of TRACE were similar to startup tests data. By the above compared results, it indicated that there was a respectable accuracy in Kuosheng NPP TRACE/SNAP model. In SBO transient (no water injection case) analysis, the PCT (peak cladding temperature) was larger than 1088.7 K at 4730 sec which indicated that the zirconium-water reaction was able to generate. In SBO + LOCA transient (no water injection case), PCT reached the criteria of 1088.7 K at 1900 sec and FRAPTRAN results implied that the fuel rod burst at 2000 sec. However, if the fire water (flow rate 39 kg/sec) injected to the reactor at 800 sec in this transient, TRACE results depicted that PCT was lower than 1088.7 K and FRAPTRAN results also indicated that the integrity of fuel rod was kept. But one safety issue generated in the drywell during 1460〜2570 sec due to the drywell temperature larger than the limit (438.71 K). Finally, TRACE/SNAP analysis results were presented by the animation model of Kuosheng NPP.

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The passive containment cooling system (PCCS) is one of the important components of passive safety system, which serves the 3rd generation advanced pressurized water reactor (APWR). The containment vessel would transfer the heat released from the reactor to the ultimate heat sink through natural phenomenon. Steam condensation on the inner surface of containment is one of the most dominating mechanisms during heat removal process through PCCS. SCOPE (Steam condensation on cold plate experiment) test facility was designed to investigate the fundamentals of steam condensation with non-condensable gas on cold plate surface. Design and performance of the facility have been introduced in detail at ICONE22. Part of the preliminary results from SCOPE is described in this work. The experiment pressure range is from 0.3 to 0.5 MPa(d). The experiment results show that gas velocity plays an important role in the mass and heat transfer process and should be considered in the prediction of containment after design basic accident (DBA). More sensitive experimental work are needed for further analysis.

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Cementitious materials for the construction of the geological repository of radioactive wastes alter the pH of groundwater to the highly alkaline condition (pH&ap;13). While this alkaline groundwater dissolves silicate minerals, the soluble silicic acid polymerizes or deposits on the surface of rock with the decrease in pH by mixing with the surrounding groundwater (pH=8). Especially, the deposition of silicic acid leads a clogging effect in flow-paths, which retards the migration of radionuclide. This study estimated the clogging of silicic acid in flow-paths with one-dimensional advection-dispersion model considering the deposition rate constants evaluated in the authors' previous study. In these estimations, the initial supersaturated concentration of silicic acid and the density of deposited minerals were focused as some of the most important parameters. As a result, the aperture of flow-paths (initial width: 0.1 mm, flow-rate: 5 m/year, initial supersaturated concentration: 0.01, 0.1 and 1.0 mM) was clogged within about 200 years by the deposition of silicic acid. The period for the clogging became shorter under the conditions of higher initial supersaturated concentration and lower density of deposited minerals. In other words, the use of cementitious materials for constructing the repository might bring a retardation effect of radionuclide migration by the deposition/clogging processes of the supersaturated silicic acid.

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The influence of the Magnus and Saffman lift acting on a droplet which moves in various flow fields is studied. The steam flow is simulated by FLUENT. The Lagrangian approach is chosen to describe a single droplet moving in flows, while carrier-phase is described by Eluerian approach. All the dynamic equations are solved by Runge-Kutta method. A group of the trajectories of droplets with different diameters are performed with and without lift forces in uniform flow, simple shear flow and wave-type plate separator flow field, respectively. The result shows that the lift forces have tiny impact on the behaviors of droplets in a uniform flow field. However, in simple shear flow fields, the lift forces affect the behaviors of droplets more significantly with rising flow shear rate and the lift forces are mostly contributed by the Saffman term. In addition, the lift forces have different effects on the behaviors of droplets with different diameters in a wave-type plate separator. The droplets with larger diameters are more sensitive to the lift forces. The conclusion verifies the importance of the lift forces in the single droplet motion model,which could be used to simulate the behavior of droplets moving in a separator.

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During CFD simulations of the flows at low Mach number regime, the classical assumption which neglects the dilatable effect of gas is no longer applicable when the temperature variation or the concentration variation of the mixture's components is too large in the fluid domain. To be able to correctly predict the flow at such a regime, some authors have recourse to a Low Mach number algorithm. This algorithm is based on the well-known pressure-based algorithm or elliptic solver for incompressible flows, SIMPLE, with a modification for the treatment of the pressure which is split into two parts (the hydrodynamic pressure and the thermodynamic pressure) and a dilatable term added in the mass equation. This algorithm has been implemented in the CFD code, Code_Saturne, developed by EDF R&D, and applied for the CFD simulations of the erosion phenomena of light gas stratification by air injection. This paper is devoted to the analytical work with the Low Mach number algorithm based on the ST1 series of the SETH-2 campaign provided by the OECD project on the PANDA test facility of PSI. The first part is focused on a mesh sensitivity analysis, which is a common procedure for CFD codes validation. The second part of the paper presents a comparison between the CFD results obtained with the standard algorithms used for incompressible flows and the Low Mach number algorithm. The third part is an analysis of the CFD results obtained on the reference mesh with both different Froude numbers corresponding to the tests ST1_7 (F_r=6.04) and ST1_10 (F_r=7.95) from the ST1 series. In the last part the authors perform the knowledge of the initial light gas distribution effect on the stratification erosion and the capability of the CFD codes to predict this phenomenon with an area governed by diffusion regime (at the top of the vessel) and another one by forced convection near the injection.

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The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident (SA parameters) and are developing the Severe Accident Instrumentations and Monitoring Systems (SA-keisou) that could measure the parameters in severe accident conditions. In this study, the hydrogen sensor with solid electrolyte was developed for measurement of H_2 concentration as the Severe Accident Instrumentations and the performance tests of the developed hydrogen sensor were carried out under various temperature and pressure conditions. In the results, the basic properties of the sensor such as the electron motive force (EMF) behaviors were evaluated for the change of H_2 concentration under these conditions.

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The working area and working time are restricted for human under hazardous environment such as high radiation environment or disaster sites. For this reason, we have developed a remotely operated quadruped robot which can walk on uneven terrain such as stairs and slopes. We focused attention on using this robot for carrying various tools and materials for decommissioning work to enlarge workability in hazardous environment instead of personnel. When the arm for handling loads is mounted on the robot, the conveyable load is decreased by weight of the arm. Therefore, we realized unloading task using two of its legs as handling arms. This enables to carry the load which is equal to the maximum payload of the robot. Since the leg tip of this robot is not designed to handle objects, the lifting lug whose shape fits the leg tip was attached to the carrying tray. This unloading task was validated by simulation and experiments. Moreover, we have developed stable walking control method on unsteady or uneven terrain such as rubbles by dynamically keeping balance using posture sensors. This control method can improve robustness of walk with loads, and enhanced practicality of this robot.

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In nuclear power plant, the phenomenon of liquid film breakdown in corrugated-plate separator is common and affects its separation efficiency. The flow behavior of liquid film that fell down along the vertical plate and was driven by airflow was so complicated in the separator that the full attention was paid in this paper. The dynamics and the breakdown of liquid film were determined by such factors as gas velocity, thickness and falling velocity of liquid film and the property of fluids. Based on the conservation of energy, the criterion for predicting liquid film separation was established. Comparisons of the experimental data and the results of the theoretical model showed that the prediction of breakdown of liquid film at the corner was accurate.

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The safety design of nuclear plants against an earthquake is an important issue in the safety of nuclear reactors. Many studies for the earthquake-resistance of the nuclear plants has been performed for a structural strength of the nuclear power plants. On the other hand, although the gas-liquid two-phase flow in the nuclear power plants has important effects on the behavior of nuclear power plants, including the power of the reactor core, there is little knowledge of the behavior of the two-phase flow under the earthquake. For example, the bubbly flow behavior under a flow rate fluctuation caused by the earthquake acceleration is unclear. It is necessary to clarify the two-phase flow behavior under the earthquake conditions. To develop the prediction technology of two-phase flow dynamics, the detailed two-phase flow simulation code with an advanced interface tracking method (TPFIT) was expanded to the two-phase flow simulation under earthquake accelerating conditions. The purpose of the study is to clarify the behavior of the gas-liquid two-phase flow under the earthquake conditions and to use the results in order to compare them with the numerical simulation results. Especially, the bubble behavior in the two-phase flow, diameter, shape, and velocity of bubbles which are affected by the oscillation of the earthquake is investigated. In the experiment, the flow was bubbly flow and/or plug flow in a horizontal pipe. The working fluids were water and nitrogen gas. The water was driven by a pump and the flow rate fluctuation was given by an oscillator attached to the main flow loop. The frequency of the flow rate fluctuation was taken between 0.5Hz and 10Hz. The behavior of the horizontal gas-liquid two-phase flow under the flow rate fluctuation was measured by image processing using a high-speed video camera and PIV at the test section. The pressure sensors were installed at the inlet of the mixer and the outlet of the test section. As a result, the bubble/gas plug behavior mechanism under the flow rate fluctuation was obtained. The deformation of the gas-liquid interface was caused by the time change of the relative velocity between the gas and liquid phases.

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The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident (SA parameters) and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. This paper discusses the GLS electro-chemical type hydrogen monitor and the multi-phase, multi-year evaluations performed for this hydrogen monitoring system. The testing was a comprehensive test program lasting almost 3 years. In the testing, the performance during SA was demonstrated. Since the conditions in SA scenarios are very severe in terms of temperature, pressure and radiation, it was necessary to ensure that the GLS hydrogen monitor utilized measurement methods and materials that could operate in these severe environments. The GLS hydrogen sensors are a chemi-resistive sensor that is intrinsically smart and tuned to hydrogen. Hydrogen measurement is rapid at typically less than one minute. The GLS hydrogen sensor was confirmed to operate at the extremely high radiation dose of 5 MGy, very high pressure of 1MPa, and extreme SA temperature of 700℃. The GLS hydrogen monitor deploys Gas Monitoring Units which measure hydrogen, temperature, pressure and oxygen present during SA conditions. The GLS hydrogen monitor has several benefits during SA, such as confirmation of hydrogen occurrence and variation of hydrogen concentration under extreme conditions. This study is a part of the results of the collaborative project by Japanese electric power companies and plant manufacturers that is carried out as the Safety Enhancement for LWRs program by Agency for Natural Resources and Energy.

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The objective of this study is to visualize the chaotic mixing patterns induced by oscillatory flow in a curved tube. The velocity field during oscillatory flow in a curved tube with inner diameter of 10 mm and tube length of 314 mm was numerically simulated using the commercial software Fluent. The working fluid was water and the flow was assumed to be 3D, laminar, unsteady, and incompressible. To clarify the effect of the curvature, the ratio of the tube radius to the curvature radius a/b was varied from 0.0125 to 0.075, and oscillatory flow frequency f was varied from 0.5 to 2.0 Hz. To visualize the fluid mixing patterns, trajectories were also calculated for the minute particles that follow convective movement. To examine the chaotic features of flow, the largest Lyapunov exponent was then evaluated. The visualized tracer patterns revealed the characteristic fluid mixing patterns with repeated stretching and folding in the radial and longitudinal cross-sections, and the length of a stretched tracer line increased with increasing a/b and with decreasing f. The average value of the largest Lyapunov exponent was positive under all flow conditions and was changed in response to the tracer line length. These results indicate that the flow mixing pattern induced by oscillatory flow in a curved tube is chaotic.

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The present work is aimed at investigating the influence of water injection into a partly-damage core as a result of hypothetical multiple failures of the safety systems during a small break (SB) LOCA in a boiling water reactor (BWR). The work is carried out in the frame of the research project WASA-BOSS (Weiterentwicklung und Anwendung von Severe Accident Codes - Bewertung und Optimierung von Storfallmassnahmen) funded by the German Federal Ministry of Economics and Technology which is devoted to the analysis of severe accident scenarios for the assessment of the safety of German nuclear power plants. The reference power plant is a generic 72-type BWR with a thermal power of 3840 MW. According to probabilistic safety analyses, a SBLOCA scenario at one of the steam lines inside the containment was defined for this work. Calculations have been carried out with the thermal-hydraulic system code ATHLET-CD up to 2 hours after reactor SCRAM, corresponding to a configuration where about half of the core material is molten in case of no water injections. The coolability of a partly-damaged core has been investigated by assuming the recovery of the safety systems operation at different core damage states. The results of the present calculations show that injection of water at different points in time has a significant impact on the produced hydrogen and on the accident progression. This study represents the starting point for further analysis, which may include reflooding at different core damage states as well as simulation of the interaction between molten material and lower plenum, and provides useful information for the improvement of the accident management measures.

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The AP1000 nuclear steam supply system (NSSS) contains the reactor core and the reactor coolant system (RCS). The RCS is made up of two vertical U-tube steam generators (UTSG), four canned motor reactor coolant pumps, and one pressurizer. The energy released in the fission reaction is removed from the reactor core and transferred to the secondary side of the steam generators for power generation by RCS. In this paper, the dynamic responses and control characteristics of the AP1000 NSSS have been studied through dynamic modeling and simulation. First, a nodal core model was used to describe the core thermal power transient. A non-equilibrium two-regions-three-volumes pressurizer model and a lumped parameter dynamic model with moving boundary for the U-tube steam generator were developed based on the fundamental conversation of mass, energy and momentum. Then, the developed models, the relevant NSSS control systems, and the feedwater system of UTSG were implemented using MATLAB/SIMULINK to develop a fast simulation program of AP1000 NSSS. Based on the developed simulation program, the typical Mechanical Shim (MSHIM) load follow, 10% step load change and 5%/min ramp load change simulations were performed. The simulation results have demonstrated that the trend of transient responses agrees well with the general physical rules, and the reactor power and other key system parameters are well-controlled by the NSSS control systems. Thus, the develop simulation program can be used for dynamic simulations and control studies of AP1000 NSSS. Moreover, with the adoption of modular programming techniques, the developed simulation program facilitates easy modification, runs quickly, and can be easily applied to other large pressurizer water reactors in the futures.

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The significance and development status were introduced about the pneumatic leak test pressure tightness for irradiated fuel rods in this paper, the pneumatic leak test was conducted, and the track mediator and experiment pressure of the pneumatic leak test were defined for irradiated fuel rods. With the consideration of the factors such as sealing, fixing operation, leakage monitoring system, tracer medium and pressure of the irradiated fuel rods in hot cell, the pneumatic leak test device was designed and the gas tightness inside and outside the hot cell was verified. Through the pneumatic leak test for the artificial simulative fuel rods, the effectiveness of the pneumatic leak test device as well as the feasibility of the method were proved, the requirements of pneumatic leak test were met, the technique of pneumatic leak test for irradiated fuel rods was established, the pneumatic leak test for irradiated fuel rods under strong radioactive environment was realized, and the leakage condition and position data of irradiated fuel rods were acquired.

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The passive containment cooling system (PCCS) is one of the important passive safety systems in the advanced pressurized water reactor (APWR), which belongs to the 3rd generation nuclear reactor. In the design basis accident, the steam condensing on the inner surface of the containment shell and the cooling water evaporating from the outer surface of the containment shell. In the process, the heat is transferred from the inside of the containment to the outside. To span the expected range of conditions and provide proper model for evaluating the inner steam condensation coupled outer evaporation heat transfer process, the inner steam condensation coupled outer evaporation experimental test (ISCOE) is developed by State Nuclear Power Technology Research & Development Centre (SNPTRD). Facility components, test section structure, supplying systems and measurement technology are described in this paper, also results of some pretests is introduced to show property of the facility.

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In order to decommission nuclear reactors and to improve the safety of BWR, falling jet behavior is important. BWR lower plenum consists of with various complicated structures and the falling jet behavior is affected by these structures. Thus we are developing the numerical simulation method to estimate the molten core falling behavior. To verify the code for the case of the BWR core melt accident, it is necessary to obtain the experimental data and validate the code by comparing the numerical results with experimental results. We reconstructed the previous test section that had the downcomer to simulate ABWR. On the other hand, the present test section dose not have the downcomer to simulate BWR. We conducted visualization experiment of falling jet behavior for the case of with structures and it was clarified that jet tip behavior was not affected by the downcomer. We also visualized falling jet behavior in some arrangements of structures and it is considered that external and internal flow affected the falling jet behavior and, symmetry of structures around injected point largely affected the jet falling behavior. We investigated internal and external flow and it was clarified that external flow was faster for the case with structures.

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In the area of reprocessing and radioactive waste management, gloveboxes and cells are a kind of non-standard equipments providing an isolated room to operate radioactive material inside, while the operator outside with essential biological shield and protection. The hot cell is a typical one, which could handle high radioactive material with various operating means and tight enclosure. The dissertation is based on Vitrification Plant China, a cooperation project between China and Germany. For the sino-western difference in design philosophy, it was presented how to draft an acceptable design proposal of applicable huge hot cells by analysing the design requirements, such as radioprotection, observation, illumination, remote handling, transportation, maintenance and decontamination. The construction feasibility of hot cells was also approved. Thanks to 3D software Autodesk Inventor, digital hot cell was built to integrate all the interfaces inside, which validated the design by checking the mechanical interference.

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Hydraulic Control Rod Drive Technology(HCRDT) is a newly invented patent and Institute of Nuclear and New Energy Technology Tsinghua University owns HCRDT's independent intellectual property rights. The Integrated valve, which is made up of three direct action solenoid valves, is the key part of this technology, so the performance of the solenoid valve directly affects the function of the HCRDT. This paper describes the results of an experimental study followed by a computer simulation. First, based on the working conditions of the Control Rod Hydraulic Drive System, temperature changes resulting from increasing the current of the direct action solenoid valve were studied experimentally. Subsequently, the resulting thermal fields of the coil were analyzed by ANSYS. The results show that the temperature of the coil of the solenoid valve increases with the current increasing firstly. Second, the temperature of the inner wall of the coil is higher than that of the exterior wall. Third, the temperature of the middle of the coil is higher than that of the edge of the coil. Fourth, the coefficient of thermal conductivity of the coil is obtained. Furthermore, the design of the direct action solenoid valve can be optimized.

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Hydrogen security is an important problem of the severe accident analysis in NPP, and the security of hydrogen detonation is more important. The hydrogen risk mainly focused on the risk of flame acceleration (FA) and deflagrationto-detonation transition (DDT) processes. As the processes involved in combustion and detonation are quite complex, it is difficult to present a detailed description of the whole process. Based on the physical phenomena of flame acceleration, the present paper brief introduces two criteria for hydrogen security analysis of detonation, which are FA σ criteria and DDT λ criteria. From the present paper, effect of initial pressures and temperatures on FA and DDT has been discussed. It describes the limit condition of flame acceleration by the quench/re-ignition of flammable gas mixture theory. So as to promote the flame acceleration criteria at elevated temperatures and pressures from the former one. Last the application for hydrogen safety analysis in GASFLOW code has been recommended.

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The development of smart support system is to meet the urgent needs and strengthen requirements for the training and application for Full-Scope Severe Accident Management Guidelines (FSSAMG) among utilities, nuclear regulators and research and development (R&D) institute. One of the lessons learned after Fukushima accident is that the existing Severe Accident Management Guidelines (SAMGs) used in Nuclear Power Plants (NPPs) are not sufficient to cope with site wide catastrophic event. Therefore, the development of a full scope SAMGs that update and extend the guidelines to include comprehensive plant conditions and portable equipment is necessary. The smart support system is to alleviate the effort and challenges that encountered in FSSAMG validation, application and training. Besides, it provides intuitive insights to severe accident progression and guidelines execution, and thus to support the decision-making and mitigate the accident consequences.

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To respond the urgent needs of verification, training and drill for full scope severe accident management guidelines (FSSAMG) among nuclear regulators, utilities and research institutes, the FSSAMG verification and drill system is developed. The FSSAMG includes comprehensive scenarios under Power Condition, Shutdown Condition, Spent Fuel Pool Condition, and Refueling Conditions. This article summarized the research and development of validation and drill system for FSSAMG by using the severe accident analysis program MAAP5 (Modular Accident Analysis Program 5). Realistic accident scenarios can be verified and exercised in the developed system to support FSSAMG training, drill, examination and verification.

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When Supercritical Water Reactor (SCWR) and supercritical boiler are operating normally or under accident conditions, the fine particles produced flowing with supercritical water in the pipe, the fine particles depositing on the pipe wall will affect the heat transfer of the reactor and boiler, and it may cause a bad influence to the efficient and safe operation of the thermal power plant and SCWR. A suitable mathematical model is built to study the thermophoresis of the fine particles in SCWR and supercritical boilers. The result shows that the abrupt change of the thermo-physical properties of supercritical water in the critical temperature region will make the thermophoresis deposition efficiency of the fine particles firstly decrease and then increase with increasing temperature gradient. In the supercritical pressure region, the thermophoresis deposition of fine particles firstly increase and then decrease with the pressure of the increasing of supercritical water near the pseudo-critical temperature range, and there is a peak. For particles 0.1μm〜1μm in diameter, the decreasing diameter of particles will increase the deposition rate. However for particles larger than 1 micrometer in diameter, the decreasing of the particles size has little effect to the deposition efficiency.

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During a severe nuclear power plant (NPP) accident, large amounts of hydrogen and steam can be produced in nuclear reactor containment. As the hydrogen concentration in the containment release rates exceed 4%, hydrogen combustion may occur when the hydrogen, air and steam reach certain concentration and a flammable source is present. In the case of hydrogen combustion, there is a possibility of producing short term pressure or detonation force. Therefore, the production of these gas species could threaten containment integrity. Although after the Three Mile Island (1979) accident, study of hydrogen distribution attracted scientists attention, it was after Fukushima (2011) accident, that modeling the gas behavior became an important topic in nuclear safety analyses. In order to mitigate these accidents, we need to understand the hydrogen behavior during NPP accident. Reliable turbulence model must be used in order to have an accurate estimate of the gas concentration distribution and other physical phenomena of the gas mixture. In this study, small scale airfountain test case is selected as a benchmark. Laminar model; standard κ-ε; RNG κ-ε and realizable κ-ε turbulence models are addressed. The computations are performed with HYDRAGON code. The purpose of this study is to test the influence of turbulence models to the gas distribution, and to demonstrate the code thermal-hydraulic simulation capability during NPP accident. HYDRAGON code simulation results show that RNG and κ-ε turbulence models have achieved better agreement with the experimental data on the prediction of dimensionless density distributions. Only at the region near the air-fountain source, RNG turbulence model has better estimation values than κ-ε and realizable models. The results depict that the turbulence model choice has small influence on the result. Especially, at the region near the air-fountain source, which is the region where the turbulence model was strong during air-injection.

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Natural circulation loop (NCL) offers an efficient heat transport option for applications like heat exchanger, energy conversion and cooling systems. Due to favorable thermophysical properties and environmental conformity, supercritical CO_2 is a popular recent choice, whereas water is the most common coolant used over years. Present numerical work investigates the influence of operating pressure, sink temperature and heat flux on the steady-state behavior of a rectangular NCL using both CO_2 and water as working medium under identical operating conditions. The heat source and sink are placed at the mid position of bottom and top horizontal arms respectively. Constant temperature boundary condition is selected for the sink and constant heat flux for the source, vertical legs being perfectly adiabatic. The chosen condition is sub- to supercritical for CO_2, while water is single-phase liquid for all conditions considered. It is found that mass flow rate is directly proportional to the heating power and inversely to the coolant temperature. Performance graphs are prepared and compared to identify the favorable working fluid under the selected span of operating conditions.

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Reactor pressure vessel, usually made of steel, is the heart of a nuclear power plant. The containment is one of the significant components of the NI (Nuclear Island) in nuclear power plants. There are two mainly types of containment: prestressed concrete containment and steel containment. Concrete containment and vessel with high reinforcement ratio are introduced in this article (See Figure 1). The main contents include theoretical analysis and FEM calculation.

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Although fault tree analysis has been implemented in the nuclear safety field over the past few decades, it was recently criticized for the inability to model the time-dependent behaviors. Several methods are proposed to overcome this disadvantage, and dynamic fault tree (DFT) has become one of the research highlights. By introducing additional dynamic gates, DFT is able to describe the dynamic behaviors like the replacement of spare components or the priority of failure events. Using Monte Carlo simulation (MCS) approach to solve DFT has obtained rising attention, because it can model the authentic behaviors of systems and avoid the limitations in the analytical method. In this paper, it provides an overview and MCS information for DFT analysis, including the sampling of basic events and the propagation rule for logic gates. When calculating rare-event probability, large amount of simulations in standard MCS are required. To improve the weakness, subset simulation (SS) approach is applied. Using the concept of conditional probability and Markov Chain Monte Carlo (MCMC) technique, the SS method is able to accelerate the efficiency of exploring the failure region. Two cases are tested to illustrate the performance of SS approach, and the numerical results suggest that it gives high efficiency when calculating complicated systems with small failure probabilities.

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For the design of water circulating system of a nuclear power plant (NPP), the increasing of cooling water or the decreasing of water temperature will lead to the increasing of generation capacity; while the power of water circulating pump or the scale of cooling tower will be increased, which will lead to increased investment as a result- and vice versa. This article conducts optimization and sensitivity analysis for cold end system of a nuclear power plant with sea water circulating system, using the minimum annual cost method by dynamic economical analysis. The following factors are taken into account for the optimization: the design of condenser (cooling area, the parameters of cooling tubes, etc.), the scale of cooling tower, the flow rate of circulation water and the diameter of circulating pipes, etc. In conclusion, the optimum scheme of the cold end system is obtained for the double back pressure turbine project. The variation trend of annual cost with changing circulating water amount is also concluded. Finally, this article provides the sensitivity analysis of feed-in tariff.

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China supercritical water cooled reactor CSR1000 is one of the fourth generation advanced reactors and is the first one which introduces passive safety system into the fourth generation reactor. Establish an analysis module based on Fortran language for the isolation condenser system (ICS) of CSR1000. This module mainly includes solving the flow control equation. Based on this module, sensitivity analysis for the ICS of CSR1000 is done, which mainly analysis the influence of outlet pressure and temperature, the decay power and the height difference of hot and cold end. Results show that, with the increase of the core exit temperature and the height difference between hot and cold end, the natural circulation flow rate is gradually increased, while with the increase of the outlet pressure and the decrease of heating power, the natural circulation flow rate gradually decreases. Between the core outlet pressure and temperature, the temperature has a greater influence on the natural circulation flow rate.

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Earthquakes are sources of external acceleration to components of nuclear reactors that can affect structural and material integrity. The core, that holds the fuel in which the nuclear reaction that produces heat occurs, is an intricate system that needs to be better understood in these conditions. The main objective of this work is to investigate the sloshing behavior inside a Lead-cooled Fast Nuclear Reactor (LFR) during an earthquake is conducted. The analysis of the liquid sloshing is very important in the Nuclear Power Plants (NPPs) structures to evaluate the real capacity of dynamic loads bearing and related safety levels as NPP integrity of structures, systems and components must be ensured in case of any design condition.

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The deep understanding of transient pool boiling critical heat flux (CHF) phenomena with treated surface and its correlation in water at saturated and subcooled condition are becoming increasingly important for the database of the further enhancement of the design of liquid cooling technologies in a nuclear reactor. The pool boiling CHF has been reported assuming on non-hydrodynamic instability on the effects of surface and contact angle using horizontal vertically oriented ribbons heaters in some liquids. In the present research, the steady and transient CHF have been performed due to exponentially increasing heat inputs mainly from 5 ms to 10 s on a horizontal vertically-oriented platinum ribbon in a pool of water for a range of pressure and subcooling. The three ribbon heaters with different surfaces, namely, commercial surface (CS), treated surface I (TS-I) and treated surface II (TS-II) are used. TS-I and TS-II are prepared from commercially available ribbon by finished with buff paper and emery paper, respectively. For the surface condition, surface roughness and contact were measured prior to pool boiling experiment. The non-boiling heat transfer processes are corresponding to the natural convection and the transient conduction equation, respectively. The steady-state CHFs for subcoolings and pressures, respectively, measured on TS-I are lower than corresponding values attained from commercial surface. The transient CHF, q_<cr> , on shorter period are higher than longer ones. The transient CHF, q_<cr>, of TS-II for periods in this study is higher than that of TS-I at long period. The first group of CHF on TS-II is longer than CS and TS-I, on the contrary, the second group of CHF was not observed on TS-II at atmospheric pressure under saturated condition. The increasing trend of transient CHF under saturated condition depends on pressure. The increasing trend of transient CHF under subcooling of 20 K will be become independent at pressure higher than 199 kPa. Finally, steady-state CHF were established with well-known data, respectively.

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The general understanding of transient critical heat fluxes in a pool of water at various pressures due to exponentially increasing heat inputs is necessary as the database for the correct prediction of the severe nuclear reactor accident due to a power burst. The transitions from non-boiling regime to film boiling occur instantaneously at the critical heat fluxes and then the film boiling leads to the actual burnout of the fuel rod. The purpose of the present work is to investigate the effect of surface conditions of platinum ribbon heaters with different surface conditions such as commercial and rough surfaces finished by Emery paper on the steady and transient critical heat flux (CHF) in saturated condition under atmospheric pressure due to the exponentially increasing heat inputs with the wide range of exponential periods. Namely the heat inputs correspond to those with the increasing rates from quasi-steadily to rapid ones. The dynamic heat transfer processes including boiling incipience and transition to film boiling due to exponential heat input were measured for a horizontal vertically oriented ribbon with 4 mm in width in a pool of water for the pressures from 101.3 kPa to 984 kPa, the periods from 10 ms to 20 s and for the subcoolings from 0 to 50 K. The steady-state critical heat fluxes for the subcoolings ranging from 0 to 50 K at pressures were measured with different surface conditions. The data obtained for lower and higher subcoolings were compared with ones given by Sakurai and Fukuda. The typical trend of critical heat fluxes for the exponential periods is that the critical heat flux gradually increases from the steady-state value, then decreases and again increases with the decrease in period from the longest one tested here, though the trend was not observed completely for the short period range at high pressures. The critical heat fluxes for both cases which increase with the decrease in period from the minimum critical heat flux become almost in agreement with each other for short periods. The mechanism of the transition at the critical heat flux from non-boiling regime to film boiling regime was assumed to be a consequence of the heterogeneous spontaneous nucleation in originally flooded cavities on the cylinder surface. The effects of subcooling and pressures on the transition to film boiling due to the heat inputs with a wide range of increasing rate were investigated for the both surfaces.

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The process of bubbles sliding along a wall under the effect of buoyancy and gravitation in vertical rectangular channels is studied using the VOF model in FLUENT along with the three-dimensional numerical simulation method. The impact of bubble size, flow field and other factors on the bubble behavior was studied considering the effects of surface tension, fluid viscosity and contact angle. The numerical simulation was verified using empirical data. The empirical data were collected using high-speed photography technique and post-processing software to capture the bubble slip process. The numerical simulation proved to be very close to the results obtained from the experiment. The simulation results showed that the bubble is half ellipsoid while sliding along the wall and its size fluctuates in all directions. The slip velocity becomes smaller with the increasing size of bubble diameter, which shows that the increasing rate of resistance is faster than that of flotation in the vertical direction. The system pressure has little impact on slip velocity and the shape of the bubble. Fluid viscosity mainly affects the bubble shape; the higher the fluid viscosity, the more stable the bubble. Surface tension mainly affects the slip velocity of the bubble and higher surface tension causes the bubble to become spherical. The larger the surface tension, the larger bubble slip velocity.

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A new risk monitor system is under development which can be applied not only to prevent severe accident in daily operation but also to serve as to mitigate the radiological hazard just after severe accident happens and long term management of post-severe accident consequences. The fundamental method for the new risk monitor system is first given on how to configure the Plant Defense in-Depth (DiD) Risk Monitor by object-oriented software system based on functional modeling approach. In this paper, software system for the plant DiD risk monitor is newly developed by object oriented method utilizing Unified Modeling Language (UML).Usage of the developed DiD risk monitor is also introduced by showing examples for LOCA case of AP1000.

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In order to investigate the process of nuclear power plant accidents, an accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 1 is analyzed using data available to the public. A phase equilibrium process model and adiabatic expansion model were introduced. Original data reported in the first stage of the accident were examined to clarify the behavior of the isolation condensers (ICs), which are generally believed to have become nonfunctional after the tsunami and station blackout. The original data and observation reports verified that the so called "fail safe" system to close the valves in the IC did not work properly owing to the shutdown of AC power. The reports also showed evidence that the operators injected water into the IC reservoir tank. In the author's previous report, the initial leak from the reactor pressure vessel (RPV) was assumed just after the earthquake. This report assumes that the leakage of the RPV occurred at 3/11 20:26 owing to small dry out of nuclear fuel clusters. In the present accident scenario, the leakage of the primary containment vessel (PCV) occurred at 3/12 3:30. A large break in the RPV occurred at 6:20 and again at 16:00. It is estimated that most of the fuel remains in the RPV, whereas, Tokyo Electric Power Company, Inc. (TEPCO) estimates that most of the fuel melted out through the RPV. The present analysis model and the accident scenario describe the data measured at the accident, many evidences and witnesses reported at the early stage of the accident.

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The direct contact condensation (DCC) is a significant phenomenon in nuclear reactor and its balance facilities, together with other chemical engineering systems. DCC occurs when the vapor is injected into the subcooled water, contacts and condenses on the interface directly. The DCC phenomenon accompanied with the heat transfer and mass transfer will result in the temperature and pressure fluctuations in the tank. This paper discusses the DCC in the subcooled water tank from the view of computation fluid dynamics. The numerical simulation was performed with the CFD software, Fluent, in which the direct contact condensation phenomenon was modeled with the Euler-Euler framework for two phases flow, and the evaporation & condensation model was adopted to simulate the mass transfer process. During the simulation process, the flow field and temperature field are derived. In addition, the shape and size of plume jet are also investigated.

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The existing fleet of U.S. nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper demonstrates how Idaho National Laboratory (INL) researchers use the RISMC Toolkit to investigate complex nuclear plant phenomena using RAVEN and RELAP-7. The analysis focused on a highly relevant topic currently facing some nuclear power plants - specifically flooding issues. This research and development looked at challenges to a hypothetical pressurized water reactor, including: (1) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (2) earthquake induced station-blackout, and (3) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) - these are currently under development at INL. Using RAVEN, we were able to perform multiple RELAP-7 simulation runs by changing specific parts of the model in order to reflect specific aspects of different scenarios, including both the failure and recovery of critical components. The simulation employed traditional statistical tools (such as Monte-Carlo sampling) and more advanced machine-learning based algorithms to perform uncertainty quantification in order to understand changes in system performance and limitations as a consequence of power uprate. Qualitative and quantitative results obtained gave a detailed picture of the issues associated with potential accident scenarios. These types of insights can provide useful material for decision makers to perform risk-informed margins management.

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