Flow blockage is difficult to analyze in reactor accidents. The multi-layer annular fuel assembly, consisted by annular-type fuel plates, has a good heat transfer performance, but multi-layer-type channels in it are of complex geometries, which are easily blocked. It is necessary to study the flow and heat transfer behavior of the assembly under flow blockage condition. Direct CFD modeling and system code simulation are two traditional ways to investigate flow blockage. In this paper, porous-jump conditions are also used to simulate the blockage. In the porous-jump treatment, the precise shape of the blockage is ignored, and the blockage is simulated by increasing the local flow resistance, just as the system code does. Besides, flow and heat redistributions are calculated in common CFD ways. Inlet flow blockage scenarios of a specified channel in the multi-layer annular flow channels are investigated. The result shows that porous-jump conditions process provides consistent mass flow rates predictions with the two traditional methods. For all the three approaches, only mass flow rates of the blocked channel decrease, indicating that flow blockage affects little on flow conditions of non-blocked channels. However, the heat transfer results of Relap5 are different from those calculated by both CFD methods. The differences result from the one-dimensional simplification used in Relap5, and show the necessity of using CFD method to investigate flow blockage accidents in multi-layer annular flow channels.

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ERVC(External Reactor Vessel Cooling) strategy is proposed to be adopted as a key issue of IVR(In-Vessel Retention) which is applied in AP1000, VVER-640 and APR1400 as an important method to avoid the consequences of severe accidents. Most researches concerned about the CHF of ERVC are based on experiments of which the scale is large and costly and only afforded by national research institutions. Recent days, some people use thermal-hydraulic cods to deal with CHF of ERVC, which cannot obtain physical mechanisms leading to the CHF. So this paper is dedicated to explore a way with CFD method to solve where CHF happens and predict the value. The framework for simulations is the commercial CFD code CFX. And Eulerian multiphase CFD models are developed to resolve wall boiling and CHF. The numerical results for CHF of ERVC are validated against existing experimental data in this paper. With the CFD simulations, the physical mechanism of CHF occurring in experiments of ERVC is to be discussed.

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The heat exchangers are frequently used as constructive elements in various plants and their dynamics is very important. Their operation is usually controlled by manipulating inlet fluid temperatures or mass flow rates. On the basis of the accepted and critically clarified assumptions, a linearized mathematical model of the cross-flow heat exchanger has been derived, taking into account the wall dynamics. The model is based on the fundamental law of energy conservation, covers all heat accumulation storages in the process, and leads to the set of partial differential equations (PDE), which solution is not possible in closed form. In order to overcome the solutions difficulties in this paper are analyzed different methods for modeling the heat exchanger: approach based on Laplace transformation, approximation of partial differential equations based on finite differences, the method of physical discretization and the transport approach. Specifying the input temperatures and output variables, under the constant initial conditions, the step transient responses have been simulated and presented in graphic form in order to compare these results for the four characteristic methods considered in this paper, and analyze its practical significance.

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This paper presents the differential discrete mathematical model of double-tube heat exchangers. Theirs operation in the face of variable loads is usually controlled by manipulating inlet fluid temperatures or mass flow rates, where the controlled variable is usually the output temperature of principal flow. The aim of this paper is to optimize the geometry of a tube with the inlet flow of principal incompressible fluid and an external cross-country flow of compressible fluid, based on performance index expressed throughout its controllability characteristics. Thus the condition number has been used to provide the necessary information on the best situation for control of the exchanger under consideration. This concept can also provide us with information about the easiest operating condition to control a particular output. A transient model of a double-tube heat exchanger is developed, where an implicit formulation is used for transient numerical solutions. The condition number performed throughout the ratio of geometric parameters of tube is optimized, subject to volume constraints, based on the optimum operation in terms of output controllability. The reported optimized aspect ratio, water mass flow rate and output controllability are studied for different external properties of the tube.

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In a nuclear power plant, one of the important issues is an evaluation of the safety of a system when an earthquake occurs. However, there is little knowledge how coolant (single and two-phase flow) response when large earthquake acceleration is added. The aim of this study is to clarify the influence of vibration on bubbly flow behavior in structures by using state-of-the-art experimental and numerical techniques. In order to investigate the influence of vibration on a bubbly flow behavior in structures, we visualized a bubbly flow in the pipeline on which a sinusoidal wave was applied. In the test section, the bubbly flow was produced by injecting nitrogen gas into the liquid flow through the horizontal circular pipe. In order to vibrate the test section, the test section was installed on an oscillating table. A high-speed video camera was fixed on the oscillating table to ignore the relative velocity between the camera and the test section. In this paper, based on observed images, bubble velocity was evaluated. The frequency and acceleration of vibration added from the oscillating table was from 1 Hz to 20 Hz and 0.4 G (=3.92 m/s^2) at each frequency, respectively. Liquid pressure was also measured at upstream and downstream of the test section. It was observed that the pressure gradient amplitude increased with the increase of the frequency of the table. Furthermore, it was confirmed that the bubble velocity amplitude also increases with the increase of the frequency of the table. It was concluded that the bubble motion was strongly affected by the pressure difference. In addition, to consider detailed effects of pressure gradient on bubble motion, numerical simulation of two-phase flow in horizontal pipe with vibration was performed by a detailed two-phase flow simulation code with an advanced interface tracking method: TPFIT.

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At the Fukushima Daiichi nuclear power plant accident, the emission of radioactive materials into the atmosphere was caused. SPEEDI was used to assess the dose emitted by this accident, and overestimation and underestimation were seen at the rained or snowed area. In this research, the diffusion prediction model of radioactive materials taking into account the wet deposition was described and the performance of this model was evaluated to improve this overestimation and underestimation. The different methods were used to accomplish this problem of SPEEDI. Atmospheric dynamic model CReSS is adopted to analyze the meteorological condition. Positions of each particle are calculated by the trajectories of many particles. Diffusion by the turbulent flow are calculated by random-walk method. Wet deposition is modeled by the change of number density. Parameters of water substance are based on the value suggested by CReSS. Analysis taking into account the wet deposition was aimed by using methods above mentioned.

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Research of advanced types of burnable absorbers (BA) in nuclear fuel requires fast depletion code that would be able to calculate broad range of elements, nuclides or their combination. BAs compensate for the initial excess reactivity and consequently allow longer fuel cycles with higher fuel enrichments. Advanced types of BAs would increase fuel utilization by combining rapidly depleted BA like gadolinium with long-term effect of slowly-depleted BA like europium. State-of-art depletion codes require large amount of computational time, therefore, decision to develop fast depletion code was made. Currently developed U_WB_1 depletion code comprises of transport and burnup solver, both aiming to fast calculation of nuclear fuel depletion. In order to achieve low computational time, selected equations and data libraries describing fuel behavior was simplified, resulting in balanced accuracy and speed of the code. The paper describes validation calculations of Monte Carlo solver that is used in U_WB_1 depletion code. Reactor models with various fuel enrichment, fuel spectra and burnable absorbers were compared. Multiplication factor during fuel depletion and calculation figure of merit was compared. State-of-art MCNP6 code was used to calculate reference values.

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Long-term reactivity control consists of the compensation of fuel reactivity loss due to the fission products build up in the nuclear fuel. This reactivity loss is balanced by the reactivity gain caused by the burnable absorber (BA) depletion. Introduction of burnable absorber leads to flattening of reactivity during fuel depletion and consequently allow long fuel cycles with higher fuel enrichments. Research of fuel depletion aims at development of advanced types of BAs in nuclear fuel. The paper describes the neutronic calculation comparison of rare earth oxides that can be used as burnable absorbers. VVER nuclear fuel with 5.0 wt% enrichment was selected for the study. Content of rare earth oxides in the fuel was selected to have the same compensation of initial reactivity excess for all studied cases. Multiplication factor during depletion, pin power peaking and residual poisoning was evaluated. Comparison calculations were performed with state-of-art statistical and deterministic depletion codes SERPENT and SCALE/TRITON and with fast depletion code U_WB_1 that are currently under development. Accuracy and calculation speed of calculational tools is commented.

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The increasing mixing leads by the wake generated behind the sliding bubble can enhance heat transfer of the heated wall. Bubble sliding velocity may be affected by the varied force field caused by rolling motion. Bubble sliding experiment is carried out under rolling motion condition in order to study the influence of the periodical additional inertial force. The experimental results shows that the bubble siding velocity changes periodically with the same period of the rolling motion. A predication model is established based on the bubble forces under rolling motion. Several reasonable simplification process is adopted in the model. Bubble growth force is ignored in the model for the reason that the diameter of the bubble changes slowly and thereby produces negligible effects compared with the other forces. A good agreement between predicted and measured results is achieved.

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An effort has been made to investigate defect accumulation processes in the reactor structure steel during irradiation. Simultaneous reaction rate equations were constructed on a basis of the reaction rate theory, and solved numerically on a computer to understand such microstructural changes as the formation of voids and dislocation loops in the steel. The effect of cascade cluster formation on defect accumulation processes is especially focused. When the production of cascade clusters is considered, voids and dislocation loops created during irradiation become small in size and their number densities become large, which results in an increase in total volume swelling. The dependence of defect accumulation processes during irradiation on incident neutron energy is discussed, which may lead to an establishment of the irradiation correlation rule.

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Mass transports in a rock formation has been studied since it can work as natural barriers against the migration of radionuclides in geological disposal of high level radioactive waste. The rock formation can be categorized into crystalline rock as a fractured medium and sedimentary rock as a porous medium. However, the Wakkanai formation (siliceous mudstones) in Horonobe underground research laboratory (URL) in northern Hokkaido, Japan, has unique features having both porous and fractured medium. Therefore, matrix diffusion toward the crack-free porous media should be considered in addition to the advection-dispersion through fractures in the porous medium. In this study, in-situ dipole tracer migration tests were conducted at the G.L. -250 m gallery of the Horonobe URL. Laboratory experiments were also conducted to determine the apparent diffusivity (D_a) and sorption coefficient (K_d) of cesium and water (HTO) in the rock taken at the gallery to evaluate the performance of Wakkanai formation as a natural barrier. In the in-situ dipole tracer migration tests, a non-sorbing tracer (Uranine) and a sorbing tracer (cesium) were injected to a section in one borehole, and collected at a section in the other borehole to obtain the data on the mass transport between two parallel boreholes where a single fracture crosses. The D_a of cesium and water were determined from the laboratory tests by a non-steady, one-dimensional diffusion method. The K_d of cesium were also determined by a batch method. The breakthrough curves of non-sorbing tracer (Uranine) obtained in the in-situ dipole tracer migration tests were well described by a dual-channel model in which one-dimensional advection dispersion was taken into account. This suggests that the tracers migrate through at least two different pathways in the fracture. The breakthrough curves also indicate that the peak concentration of the sorbing tracer (cesium) was much smaller than that of the non-sorbing tracer (Uranine), suggesting that the Wakkanai Formation has a high sorptive and low diffusive properties for cesium. This specific property of the rock was confirmed at the laboratory experiments. The D_a value obtained for cesium was about 2.9×10^<-12> m^2/s, which is significantly smaller than that of water (3.4×10^<-10> m^2/s), and the K_d value of cesium was determined to be 488 ml/g. These new findings can be useful for understanding the mass transport in the fractured sedimentary rocks having unique features.

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The numerical simulation for the design of hot gas mixing structure at the reactor outlet of Pebble-bed Module High Temperature gas-cooled Reactor (HTR-PM) are carried out by using the Fluent code with the consideration of the leakage flow out of the reactor core. According to the profiles of temperature, pressure and velocity, the thermal mixing is mainly produced by the secondary flow such as the vortices which is perpendicular or parallel to the main flow direction. In addition, the pressure drop is mainly caused by the local pressure loss leading by the sudden change of flow area and direction which is also the main reason of production of the secondary flow. The numerical simulation results indicate that the design of hot gas mixing structure at HTR-PM reactor outlet can fulfill the requirement of high thermal mixing performance and low pressure drop under the rated condition by taking account into the leakage flow out of reactor.

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The condensation rate of vapor bubbles, defined by v_c=-dR/dt where R is the bubble radius (assuming that bubbles have a spherical shape) and t is time, is an important parameter to determine the heat/mass transfer in the subcooled flow boiling. Accurate measurement of v_c is required to improve the numerical simulation of the flow. In this study, using multiwave ultrasound, a novel method has been developed for the measurement of v_c. The velocity of the top and bottom interfaces of bubbles are measured by using two measurement lines. One line has an upward direction. The other has a downward direction. From the velocity difference between the two measurement lines, v_c can be calculated. By comparison with the result of the optical visualization method, the accuracy of the measured data of v_c is confirmed. Using ultrasound, measurements can be carried out for opaque fluid, non-transparent flow boundaries and extreme conditions of industrial flows (e.g. high pressure, high temperature etc.). The proposed method can be highly useful to provide validation data to improve existing correlations used in numerical simulation. Consequently, the accuracy of the numerical simulation result can be improved. That is immensely important in the thermal-hydraulic analysis in nuclear engineering.

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A safety assessment needs to be conducted to analyze the damage caused by an aircraft impacting into a concrete structure at a nuclear power plant. One of the analytical methods used for this is a numerical impact simulation conducted after aircraft and reinforced concrete (RC) models are determined. We established the RC model first for this paper by conducting impact simulations of an F4 engine (GE-J79) missile crashing into three different wall thicknesses of 900, 1150, and 1600 mm. The damages to the wall in the simulations agree with the test results conducted at Sandia. We also conducted parametric impact simulations of a rigid missile crashing into a concrete wall, changing the impact speed and mass of the missile. The wall thickness required to prevent perforation in the simulations agrees with the output produced by the empirical formulae. An impact simulation of an F4 fighter crashing into a RC wall is conducted as the next step after establishing the RC model. The shape of the impact load function and the state of the F4 frames on impact are almost the same as those in the test results, which shows that the F4 model is valid. Finally, the relation between the magnitude of the impulse due to the impact and the damage to a RC wall is studied by changing the impact speed of the F4 fighter. The amounts of damage to the wall in the simulations agree with the predicted ones using an empirical formula with a reduction factor.

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The meltdown of three reactors of the TEPCO Fukushima Daiichi nuclear power station (NPS) caused by the Great East Japan Earthquake on March 11th 2011 resulted in the emission of radionuclides such as cesium-137 and strontium-90 to the environment. For example, radioactive cesium exceeding the legal discharge limit (90 Bq/L, 2×10^<-13> M) was detected in the seawater of the seawater-intake area of the NPS at the end of September 2014. Adsorbents with a high selectivity for cesium ions over other alkali metal ions such as sodium and potassium ions are required for cesium removal from seawater because sodium and potassium ions dissolve respectively at much higher concentrations of 5×10^<-1> and 1×10^<-2> M than cesium ions (2×10^<-9> M). In addition, the simple operations of the immersion in seawater and the recovery of the adsorbents from seawater are desirable at decontamination sites. We prepared a cobalt-ferrocyanide-impregnated fiber capable of specifically capturing cesium ions in seawater by radiation-induced graft polymerization and chemical modifications. First, a commercially available 6-nylon fiber was irradiated with γ-rays. Second, an epoxy-group-containing vinyl monomer, glycidyl methacrylate, was graft-polymerized onto the γ-rayirradiated nylon fiber. Third, the epoxy ring of the grafted polymer chain was reacted with triethylenediamine to obtain an anion-exchange fiber. Fourth, ferrocyanide ions, [Fe(CN)_6]^<4->, were bound to the anion-exchange group of the polymer chains. Finally, the ferrocyanide-ion-bound-fiber was placed in contact with cobalt chloride to precipitate insoluble cobalt ferrocyanide onto the polymer chains. Insoluble cobalt ferrocyanide was immobilized at the periphery of the fiber. However, the impregnation structure remains unclear. Here, we clarified the structure of insoluble cobalt ferrocyanide impregnated onto the polymer chain grafted onto the fiber to ensure the chemical and physical stability of the adsorptive fiber in various contaminated waters. The adsorption rate and capacity of the fiber for cesium ions were compared with those of a zeolite as a conventional adsorbent.

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The Great East Japan Earthquake and the tsunami that followed caused the meltdown of three reactors of the TEPCO Fukushima Daiichi nuclear power station (NPS), resulting in the emission of radionuclides such as cesium-137 and strontium-90 to the environment. Radioactive strontium was detected in seawater and groundwater at concentrations of 1.8 × 10^2 and 5.5 × 10^5 Bq/L, respectively, on October 7th 2014. Nonradioactive strontium dissolves at a concentration of 8 mg/L in seawater. No adsorbent can distinguish radioactive strontium from nonradioactive strontium; therefore, the adsorbent must collect both ions which coexist with other alkaline-earth metal ions such as magnesium and calcium ions. Inorganic compounds and chelate-forming resins are candidate adsorbents for strontium removal. However, It is difficult to use these adsorbents to process a large volume of water contaminated with radionuclides because of their granule and bead forms. We have prepared two kinds of adsorptive fiber by radiation-induced graft polymerization and subsequent chemical modifications: (1) sodium-titanate-impregnated fiber (ST fiber) and (2) iminodiacetate-group-immobilized fiber (IDA fiber). The preparation scheme of the ST fiber consisted of four steps. First, a commercially available 6-nylon fiber was irradiated with γ-rays to produce radicals. Second, sodium styrene sulfate was graft-polymerized onto the irradiated fiber. Third, a titanium species [Ti(OH)_2^<2+>]was bound to the sulfonic acid group of the grafted polymer chain. Finally, the titanium species was converted into sodium titanate with sodium hydroxide, and the resulting precipitate was impregnated onto the fiber. On the other hand, the IDA fiber was prepared as follows. An epoxy-group-containing vinyl monomer, glycidyl methacrylate, was graft-polymerized onto a previously γ-ray-irradiated 6-nylon fiber. Subsequently, the epoxy group was converted into an iminodiacetate group as a chelate-forming group by a reaction with disodium iminodiacetate. The former and latter fibers are applicable to strontium removal from seawater and groundwater, respectively.

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Electromagnetic pumps have been employed to transport the various kind of liquid metal with the high electrical conductivity including circulation of liquid sodium in a sodium fast reactor. Especially, DC electromagnetic pumps have advantage of the simple geometrical structure without any impeller and sealing part, and the continuous flow control over mechanical pumps. However, the relatively high input current is required compared with that of the induction type electromagnetic pumps. In this study, for the liquid sodium circulation in the sodium-carbon dioxide reaction experimental loop, carried out is the design analysis on the pump electromagnetic and geometrical variables including the length, area and number of coil turns considering fluid effects for the design of the rectangular and helical type pumps with the flowrate of 3 L/min.

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China Fusion Engineering Testing Reactor (CFETR) is proposed as a good complement to ITER for demonstration of fusion energy. The main goals of CFETR are fusion power 50〜200 MW, duty cycle time 0.3〜0.5 and a tritium breeding ratio &ge; 1.2. CFETR is based on both physics and some technologies of ITER. Therefore, lower single-null (LSN) divertor (also called ITER-like divertor) is a basic choice for CFETR. The heat power from core plasma into scrapped-off layer (SOL) is estimated as 100 MW, which implies that the divertor must have enough ability to reduce the heat power via radiation before it lands on divertor targets. Therefore, the heat load onto divertor targets will not exceed the engineering limit of 10 MW/m^2. In the physical design of ITER divertor, a partly detached operating is preferred to provide enough radiation ability and to avoid undesirable influence on core plasma. It is also curial to find an appropriate operation region for CFETR divertor. In this work, to predict the operation status of CFETR divertor, the Scrapped-off Layer Plasma Simulation (SOLPS) code is used for numerical simulation. SOLPS is a coupled code constituted by multi-fluid plasma simulation code B2.5 and Monte Carlo neutral simulation code EIRENE. Based on the preliminarily designed magnetic configuration and divertor geometry of the LSN divertor for CFETR, a density scan are performed via a increase of D2 gas puffing rates in the range of 0.0 〜 5.0 ×10^<23> s^<-1> for SOLPS simulations. Along with the increase of gas puffing rate, it can be seen that a gradually change of the operation status, which is from low-recycling regime to high-recycling regime and finally to detachment. C impurity concentration is lower than 4% in the plasma core region, which shows strong impurity screening of the divertor.

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Post-installed adhesive anchor bolts are commonly used on all types projects, from standard buildings to bridges and nuclear power plants. They are exposed to varying and diverse environmental conditions which are significant successfully to make connections to concrete structures for many years. With the failure of adhesive anchors in the Boston I-90 Tunnel Project, the use of these types of anchors has been called into question in America. On the other hand, also in Japan, a section of the SASAGO tunnel's suspended concrete ceiling with adhesive anchor bolts became detached from the tunnel roof and fell onto the vehicles, 2012. In order to maintain the safety and reliability of concrete structures such as buildings or bridges on higher level over a long period time, we developed a new non-destructive inspection system of adhesive anchor bolts based on hammering tests using AE (acoustic emission) sensor. By adjusting the amount of resin of adhesive anchors, the model bolts with poor construction quality were made on concrete blocks and finite element analysis was carried out to evaluate performance of this inspection system.

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Hydrogen generated during postulated severe accidents in a sodium-cooled fast reactor may be released to the air inside a reactor containment building, resulting in the hydrogen combustion when a specified hydrogen concentration is exceeded. It is important to evaluate the pressure and temperature response of this phenomenon from the standpoint of the integrity of the reactor containment building. This paper was undertaken to seek an applicability of the hydrogen combustion models incorporated in a CONTAIN-LMR code, which is produced by applying liquid metal fast reactor-specific updates to a light water reactor version of the CONTAIN code. The paper puts its focus on the calculations of premixed combustion experiments using the CONTAIN-LMR code. The results are compared and contrasted with other stand-alone code, HECTR, which often has been used for predicting hydrogen combustion. As a result, the CONTAIN-LMR predictions mostly agree with the measured peak pressure and peak temperature and the HECTR predictions. Therefore, the results indicate that the CONTAIN-LMR code is considered to be applicable to the hydrogen combustion phenomena and available as an evaluation tool for the severe accident scenarios involving hydrogen combustion events.

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Mitigative measures against a core disruptive accident are important from the viewpoints of safety of a sodium-cooled fast reactor. In order to estimate the quench behavior of molten jets, it is important to understand how the jet breaks up in the coolant. The purpose of the present study is to clarify the jet instability and breakup behavior in immiscible liquid-liquid systems. In the liquid-gas system, jet breakup regimes can be categorized on the map described by the Ohnesorge and Reynolds numbers. Such a map for the liquid-liquid system is, however, not so much reported. In order to understand the jet instability and breakup in a liquid-liquid system, we conducted experiments using some combinations of immiscible liquids. We mapped the observed regime of jet breakup against the Ohnesorge number and the Reynolds number, and established boundaries between different jet instability regimes on the map. Finally, based on the obtained map, we discussed the jet breakup in the actual SFR condition from the hydrodynamic viewpoints.

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The leakage and combustion of liquid sodium in the sodium-cooled fast reactor could result in potential fire hazards. An experimental study of sodium fire in a columnar flow was conducted to investigate the burning characteristics by analyzing the temperature fields in the burner. Liquid sodium of 210℃ was injected into a 7.9m^3 volume stainless steel cylindrical burner via a nozzle to shape a sodium fire. The temperature data of typical positions in the burner was collected by dozens of thermocouples that were fixed in the combustion space and sodium collection plate. The sodium fire in a columnar flow was composed of the foregoing spray fire, subsequent centered columnar fire and pool fire on the collection plate. The temperature in the combustion space close to the burning sodium flow could maximally reaches up to 913.8℃. The radial temperatures apart from the sodium flow were relatively low and generally about 150℃-300℃. The maximum temperature of the burning sodium dropping on the collection plate rose in the position of Td3, to about 640.6 ℃. This study is promising to evaluate the combustion characteristics, formation process and composing forms of the sodium fire in the sodium-related facilities.

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Stress corrosion cracking (SCC) is one of the most serious problems for ensuring safety of nuclear power plants. For SCC-resistant materials, residual stress due to the welding process is an important factor of SCC. Therefore, evaluating the residual stress in welds is great important, and X-ray stress measurements are superior to other stress measurement methods because they are nondestructive and provide high-resolution results. However, it is well known that texture in welds is found in some materials such as austenitic stainless steel and Ni-based alloys, and that such texture can greatly degrade the measurement accuracy. In this study, an attempt was made to improve the accuracy of X-ray stress measurements for welds with texture. To evaluate the residual stress in welds with texture, a new method was developed for evaluating plane stress from the elastic compliance, the value fraction of the crystal, and measurements in two orthogonal directions. Moreover, in a SUS316L weld specimen, the residual stress evaluated by the proposed method agreed very well with that evaluated by the stress relief method. Thus, the proposed method has the potential to be effective for measuring residual stress in welds with texture.

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Sustained efforts are made to improve the Nuclear Power Standard System in China for supporting current and future development of nuclear power programs. However, building a nuclear power standard system is a comprehensive project involving many subjects related to technology or administration. This paper demonstrates an approach to China nuclear power standard system. Due to the government strategy that PWR has been being the main stream of nuclear program in China, the paper focuses on the part of the standard system applicable to PWRs, and a methodology based on standard demand analysis is developed to fulfill the so called "five attributes" of system, which has been applied in the National Science and Technology Major Project (NSTMP), "Research on Developing China Advanced Nuclear Power Standard System".

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A good qualitative understanding and an accurate quantitative description of fluid flow in randomly water cooled packed bed reactors (WCPBRs) are important. In order to reflect the local phenomena, a spatially resolving three-dimensional (3D) flow simulation is needed. In this work, the packing structure in cylindrical tube was randomly packed with spherical particles and the flow through the packed bed were modeled by discrete element method (DEM) and computational fluid dynamics (CFD) respectively.For modeling an 3D representation of the geometric structure of the packing, DEM were applied to generate the packed bed with random packing of spherical particles. The obtained numerical radial porosity of the generated packed bed is compared with the empirical correlation from the literature. In this work, a method for meshing the particle-particle and particle-wall contact points in the fixed beds was presented. Compared to the shrinkage approach that commonly used in the literature, this method reduces the error of change in porosity, and achieves good quality meshing. CFD simulations were performed for water flow through randomly packed bed, in the laminar, transitional and turbulent flow regime. To validate the proposed geometry and grid, the Ergun correlation and Eisfeld and Schnitzlein correlation for predicting of pressure drop were used. The simulated averaged pressure drops were in a good agreement with the Eisfeld and Schnitzlein correlation. From the simulation, the localized and detailed flow characteristics were captured, and the stagnant regions were identified.

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We have developed Check Valve Inspection System using acoustic emission sensor and ultrasonic sensor, for the purpose of detecting degradation and detailed operating state (Full open, Tapping, Oscillation, etc.) of the swing check valve. The operating state of the swing check valve is identified as follow. First, the rough location of swing arm or disk is examined by obtaining the reflected echo from valve disk. Second, the detail motion of swing arm or disk, whether the valve disk is oscillating or is stable, is determined by analyzing the observation time of the reflected echo. Finally, it is investigated using acoustic emission sensor to see if the disk or the arm is tapping against the backstop repeatedly. Degradation in the valve is evaluated by detecting acoustic pulses collected by acoustic emission sensor. If such deterioration progresses in the driving parts, the number of acoustic pulses by collision increases. The rate of deterioration of the swing check valve is determined from the change in the number of acoustic pulses. As an example of the application of the system, the case of check valves in the seawater system line installed in a nuclear power plant were shown. The examination was carried out before and after replacing the check valve. By comparing the results of the inspection system and the results of overhaul, the detectability of the inspection system was evaluated.

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Post-installed adhesive anchor bolts are commonly used on all types projects, ranging from standard buildings to bridges and nuclear power plants. They are faster to install, generally considered to be stronger, more flexible, easier to use and more reliable in majority of applications at ambient temperatures. They are exposed to varying and diverse environmental conditions which are significant successfully to make connections to concrete structures for many years. With the failure of adhesive anchors in the Boston I-90 Tunnel Project, the use of these types of anchors has been called into question in America. On the other hand, also in Japan, a section of the Sasago tunnel's suspended concrete ceiling with adhesive anchor bolts became detached from the tunnel roof and fell onto the vehicles, 2012. In order to maintain the safety and reliability of concrete structures such as nuclear plants on higher level over a long period time, we developed a new non-destructive inspection system of adhesive anchor bolts based on hammering tests using AE (acoustic emission) sensor. By adjusting the amount of resin of adhesive anchors, the mockup anchors with poor construction quality was made on concrete blocks and mockup experiment was carried out to evaluate performance of this inspection system. As a result of the mockup experiment, peak frequency of signal obtained from AE sensor shifted to lower frequency as the amount of resin decreased. In this mockup experiment, it was confirmed that our developed inspection system based on hammering tests using AE sensor had a large potential to evaluate adhesive anchor bolts.

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A new non-destructive inspection technique with using acoustic emission (AE) sensor was developed for the purpose of overcoming problems originating in existing hammering test. The new technique enables us to obtain inspection results with no dependence on personal skills of a checker, and to judge the integrity of the bolts based on quantitative measurement results. The investigation of applying the new method to wedge anchor bolts is reported. From experimental results by using this technique, the trend of peak frequency changes obtained from concrete blocks with no crack has a clear tendency to increase with increasing tightening torque. In addition, the peak frequency obtained from the bolt installed in the cracked concrete block decreases compared with one in no-crack concrete block. These results indicate the possibility that both looseness of nut and degradations such as cracking in concrete could be detected by the new technique.

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As the important lessons learned from Fukushima-nuclear power plant accident, countermeasures and management became essential against severe accident and excessive earthquake conditions. To estimate adequate scenario of accidents, failure modes of nuclear components under severe accidents and earthquakes were investigated. Accident conditions induce some different failure modes from design conditions. Furthermore, best estimation requires clarification of their failure mechanisms. Therefore, this study focused on the loads and failure modes under beyond design conditions. To observe ultimate failure behaviors of structures, new experimental techniques were adopted with simulation materials such as lead and leadantimony alloy, which has very small yield stress. Next loading and failure modes were investigated. (1) High temperature and inner pressure loading mode Ductile fracture and local failure were investigated. At the structural discontinuities, local failure may become dominant. (2) High temperature and outer pressure loading mode Buckling and fracture were investigated. Creep bucking occurs however hardly break without structural discontinuities. (3) Excessive earthquake loading mode Ratchet deformation, collapse, fracture, and fatigue were investigated. Low-cycle fatigue is dominant for the most cases.

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Newly developed resonance interference model is implemented in the lattice physics code STREAM, and the model shows a significant improvement in computing accurate eigenvalues. Equivalence theory is widely used in production calculations to generate the effective multigroup (MG) cross-sections (XS) for commercial reactors. Although a lot of methods have been developed to enhance the accuracy in computing effective XSs, the current resonance treatment methods still do not have a clear resonance interference model. The conventional resonance interference model simply adds the absorption XSs of resonance isotopes to the background XS. However, the conventional models show non-negligible errors in computing effective XSs and eigenvalues. In this paper, a resonance interference factor (RIF) library method is proposed. This method interpolates the RIFs in a pre-generated RIF library and corrects the effective XS, rather than solving the time consuming slowing down calculation. The RIF library method is verified for homogeneous and heterogeneous problems The verification results using the proposed method show significant improvements of accuracy in treating the interference effect.

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Flow behavior in a sloping bottom cavity is observed to study the effect of cavity shape on flow behavior for Joule-heating flow. In the former study, a simple cubic cavity is applied to study the chaotic flow behavior of Joule-heating convection due to simplification as the real melter case is complicated. In this study, a sloping bottom cavity of the dimension one-fifth of the actual melter is applied to study the detail flow behavior. Carbon electrodes and top cooling surface are placed to make Joule-heating and the chaotic flow behavior. The working fluid is 80%wt Glycerol-water solution with LiCl as electrolyte. To observe the chaotic flow behavior spatio-temporally, Ultrasonic Velocity Profiler (UVP) is applied in this experiment to obtain the one-dimensional continuous velocity profiles in the center line of cavity. Particle Image Velocity (PIV) method is also applied to observe the two-dimensional flow behavior and to examine the cross-check between UVP and PIV for the chaotic flow behavior with temperature distribution. The flow profiles of the former cubic cavity and the sloping bottom cavity are compared changing voltage magnitude and cooling temperature of the electrodes side to analyze the effect of cavity shape under Joule-heating condition. The flow behavior in the upper part of the sloping bottom cavity is similar to that in the cubic cavity in the experiment in whole cavity, the range down-flow achieved is larger than the cubic cavity.

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The inert matrix plutonium and uranium fuels, Pu- and U-Rock like Oxide (Pu-ROX and U-ROX) fuels partial loading PWR core calculations were carried out, and the power distribution, fuel temperature reactivity coefficient (FTC) and spent fuel composition were estimated. In comparison with the MOX-UO_2 partial loading core, the Pu transmutation is higher in Pu-ROX fuel than in MOX fuel, and the Pu production in U-ROX is less than UO_2, because the ROX fuel matrix does not contain fertile U-238. As for minor actinides (MAs), theMA production in Pu-ROX is more than in MOX, because the higher order plutonium isotopes are produced more in Pu-ROX than in MOX due to lack of U-238 conversion to Pu-239. In U-ROX, MA production is nearly as much as that in UO_2, because the dominant MA isotope in these fuels is Np-237 transmuted from U-235. There are also found Pu and MA compositions differences. These data will be used in the analysis of nuclear energy phase-out scenario in Japan. Without containing U-238 in fuel matrix, both the Pu-ROX and U-ROX fuels tend to have larger power peakings than MOX-UO_2 core, due to larger burnup reactivity changes of the ROX fuels. The power distribution flattening is the important issue of the Pu- and U-ROX partial loading core.

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Human Factors Engineering (HFE), like other engineering disciplines involved in plant design, cannot be considered retroactively. The engineering principles and methods derived from deep knowledge of the cognitive and perceptual capabilities and limitations of the plant's "human element" are applied instead throughout plant design. Focusing HFE efforts on the plant's I&C, the plant's HMI is designed to ensure effective and error-free performance of the monitoring, control, and administrative tasks allocated to the control room crew. Generally speaking, a project's HFE program prescribes three main steps: (i) the analyses of plant monitoring and control functions in order to identify those to be performed manually (all others are performed automatically while still manually monitored) and determine in turn the HMI inventory of information displays, controls, alarms, and operating procedures required to support their performance, (ii) the guided design of the plant's HMI, ensuring its compliance with HFE principles and the completeness and correctness of the task support it provides, and (iii) the subsequent evaluation of operators performance, trained to follow the operating procedures and use the HMI referred to. The I&C systems designed to monitor and control the plant processes and implement, among other functions, the plant's HMI, are likely validated, governed by I&C norms and the project's V&V guidelines. Past experience shows that the three following obligatory steps pose challenges to project execution: (i) the acquisition and analysis of the multidisciplinary functional requirements (related to plant monitoring and control); (ii) the likely interdisciplinary analysis whether and how fulfillment of these requirements shall be allocated to I&C automation systems or operators (or both), and (iii) the HFE-guided HMI design and validation. A timely and cost-effective application of HFE to I&C engineering can be achieved by adequate planning and project management. This paper aims to summarize some of our experiences

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Helium circulator is one of the key equipment of High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). In order to simulate most normal and accident operating conditions of helium circulator in HTR-PM, a full scale, rated flow rate and power, engineering test loop, which was called Engineering Test Facility - Helium Circulator (ETF-HC), was designed and established. Two prototypes of helium circulator, which was supported by Active Magnetic Bearing (AMB) or sealed by dry gas seals, would be tested on ETF-HC. Therefore, special interchangeable design was under consideration. ETF-HC was constructed compactly, which consisted of eleven sub-systems. In order to reduce the flow resistance of the circuit, special ducts, elbows, valves and flowmeters were selected. Two stages of heat exchange loops were designed and a helium - high pressure pure water heat exchanger was applied to ensure water wouldn't be vaporized while simulating accident conditions. Commissioning tests were carried out and operation results showed that ETF-HC meets the requirement of helium circulator operation. On this test facility, different kinds of experiments were supposed to be held, including mechanical and aerodynamic performance tests, durability tests and so on. These tests would provide the features and performance of helium circulator and verify its feasibility, availability and reliability.

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For improvement of Severe Accident (SA) codes, a detailed core relocation process has to be clarified because the numerical methods in SA codes for analyzing the core relocation process is simplified. Especially, it is very important to understand precisely the melting behavior of fuel elements including fuel rods, fuel channel boxes and control rods. Then we have been developing the numerical method for analyzing fundamental melting behavior of the fuel elements based on the original POPCORN code. The POPCORN code was developed in Japan Atomic Energy Agency (JAEA) based on the Moving Particle Semi-implicit (MPS) method. In the present study, we introduced the numerical analysis models for phase change, surface tension and multi components to the original POPCORN code. This paper shows an outline of the simulation methods and results of numerical simulations, it was confirmed from the present results that melting behavior called candling can be simulated numerically.

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In this paper the lumped parameter containment code system COCOSYS has been applied to evaluate the performance of Passive Containment Cooling System (PCCS) during accidents. The original physical model describing water film behaviors in the code has been modified by taking into consideration the film breakup and subsequent phenomena as well as the effect of film interfacial shear stress created by countercurrent air flow. Moreover, the phenomenon of rivulet hysteresis and the effect of wave on the surface have also been taken into account. The results of simulation, which is conducted based on the geometry of Generic Containment, indicate that without the modification the original film model may overestimate performance of PCCS and therefore bring about inaccuracies. Sensitivity analyses regarding contact angle have been carried out and discussed as well.

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In an attempt to investigate Cs or Cs-Te corrosion dependence on chromium or carbon content in Fe-Cr steel, cesium and Cs-Te corrosion test were performed to three specimens, Fe-9Cr-0C, Fe-9Cr-0.14C and Fe-13Cr-0.14C, for 100 hours at 973K in simulated high burn-up fuel pin environment. Cesium corrosion depth has no dependence on chromium or carbon content in Fe-Cr steel. Cs-Te corrosion was appeared in only Fe-13Cr-0.14C which has chromium carbides ranged along grain boundary. Appearance of the Cs-Te corrosion was determined by distribution or arrangement of chromium carbides which depends on chromium and carbon content

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A depressurization accident is one of the design-basis accidents of the VHTR. When the pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. In order to predict or analyze the process of air ingress during the depressurization accident of the VHTR, it is very important to develop computer programs and to validate them by experiments. This study is to investigate the effect of one-dimensional natural circulation on the mixing process of two component gases by evaluating the onset time of natural circulation through the apparatus under the stable density stratified fluid layer. The experimental apparatus consists of a reverse U-shaped vertical slot and a storage tank. The left side vertical slot consists of the heated wall and the cooled wall. The right side vertical slot consists of the two cooled walls. These experimental results show that generation time of natural circulation was affected by molecular diffusion and localized natural convection. When the two components of gases have large density ratios and large Gr numbers, the mixing process of two components of gases was affected by more intensively molecular diffusion than localized natural convection when temperature difference was 50K. The mixing process of two component gas was affected by more intensively localized natural convection than molecular diffusion when temperature difference was 70 to 100K. However, two component gases were affected by more intensively molecular diffusion than localized natural convection at small density ratios and small Gr numbers.

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This paper presents a formulation of a nonlinear model predictive controller for steering of a nuclear reaction in a pressurized water reactor via control rod displacement. The predictive model, based on point kinetic equations, includes an averaged delayed neutron group, thermal dependence on reactivity, heat exchange between temperature and moderator and Xenon poisoning. From the model, a finite horizon optimal control problem is cast in the form of a control Hamiltonian minimization. A basic gradient descent scheme is derived to solve the optimal control problem. The predictive model and gradient descent are used to form the basis from which the model predictive controller is formulated.

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This work experimentally determined the viscosity and thermal conductivity of In-Bi-Sn eutectic alloy (Field's metal), which possesses a melting temperature of 333 K. The Field's metal was fabricated at the University of Ontario Institute of Technology (UOIT) in Oshawa, Canada. The work will include a discussion of the non-Newtonian behaviour of Field's metal and the thermal dependence of both viscosity and thermal conductivity. The work used a rotational viscometer to measure the viscosity of the metal in a liquid state up to 363 K. The work used an axial type experimental configuration to measure the thermal conductivity up to 400 K. The results will compare the viscosity and thermal conductivity to the traditional liquid metal coolants of sodium and lead-bismuth eutectic (LBE). Finally a prediction of the expected Prandtl number in low velocity natural circulation will be determined based on the geometry of the liquid metal natural circulation experimental loop at UOIT.

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In the case of a severe reactor accident, knowledge of the coolant level and the state of the core, including the progress of a possible core melt, would be of crucial interest. In such a scenario, the in-core instrumentation will most likely not be available. In this work, we explore the possibility of using the ex-core neutron detectors to gain information about the state of the reactor pressure vessel (RPV) inventory for a light water reactor. These detectors, which are typically implemented as ionization chambers, are located inside the biological shield and might still be operational during a severe accident. Stationary Monte Carlo calculations using the radiation transport code MCNP were performed to simulate the transport of neutrons outside the RPV and the reactions of the ionization chambers. These detector signals are computed for different model reactor states which might occur during a severe accident with core meltdown. The reactor model is based on data from a typical German Pressurized Water Reactor. The results indicate that a change in coolant level should be detectable. Due to the core's neutron self-shielding, deformations in the inner core region, such as the formation of a cavity, do not yield different signal rates in the ex-core neutron chambers. Changes not confined to the centre of the core, such as the relocation of corium into the lower head, are detectable by their change in the ionization chambers' reaction rates.

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During the 70' and 80' of the last century 12 units of Russian VVER-440 type of reactors were started to build on the territory of Czech and Slovak Republic. Nowadays 2 units were already shutdown, 8 are operates and 2 are still under construction. During theirs operation period wide range of modifications and upgrades were performed to strengthen the nuclear safety and reactors operability. In scope of strengthening the reactor's core monitoring and surveillance the original Russian VK3 system was completely replaced with advanced computer based core monitoring and surveillance system SCORPIO-VVER developed by the local organizations with international support. Since 1999 up to today the SCORPIO-VVER system were installed and still operating on four units of Dukovany NPP (Czech Republic), on two units of Bohunice NPP (Slovak Republic) and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). With the first installation in 1998 at Dukovany NPP the SCORPIO-VVER system has proved to be a valuable tool for the reactor operators and reactor physicists and was licensed by both Czech and Slovak Nuclear Regulatory Bodies as a Plant Technical Specification Surveillance Tool. The development of SCORPIO-VVER core monitoring system continues along with the changes in VVER reactors operation. Within the planned upgrades the system is being adapted according the utility needs. Between the most significant upgrades and changes belongs the modifications in connection with implementation of a new digital I&C system, adaptation of the system to up-rated unit conditions, loading of the optimized Gadolinium bearing Gd2M+ fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), in design and methodology of the limit and technical specifications checking (on-line shutdown margin calculation) and improvements in the predictive part of the system (Strategy Generator). The latest upgrade of the SCORPIO-VVER system at Dukovany NNP is under progress with the planned end in December 2015. The system is being completely renewed, re-hosted to the new hardware, being implemented the latest advanced neutron-physical and thermo-hydraulical codes with enhanced accuracy in accordance with local, national and international standards and requirements.

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To understand the mechanism of maximum energy released in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs), we investigated behavior of whole core molten fuel pool using a fast reactor safety analysis code. Several high reactivity insertions and power peaks due to super prompt criticality appears by radial pool sloshing motion. We noticed that the pool motion suppressed at the super prompt criticality in series of analysis. We called this suppression phenomenon as "mitigation effect". It is caused by fuel vapor pressure due to a rise in fuel temperature. By this effect, the reactivity ramp rate and inserted reactivity are restrained. It is expected that the existence of the mitigation effect becomes a clue in quantifying an upper limitation of the released energy.

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In a core disruptive accident of a sodium-cooled fast reactor, the considerable amount of fuel in the core region may melt, and be discharged into the coolant like a jet. Hence, it is necessary to understand the jet breakup behavior from the viewpoint of the post-accident heat removal. In order to clarify the mechanism of jet breakup caused by the hydrodynamic interaction, we simulated the injection of a jet into a coolant using the lattice Boltzmann method for two-phase fluid which can represent the spontaneous interfacial behavior. In the simulation, we changed the Reynolds number of the jet and the viscosity ratio between the jet and the coolant, and observed their influence to the interfacial behavior. By observing the interfacial behavior, we found that the wave formation and fragmentation at the side of the jet were suppressed when the Reynolds number becomes small, and cognized the importance of considering the influence of the viscosity of a coolant to the jet behavior.

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Oxygen control in lead-bismuth eutectic (LBE) is one of the key issues for the development of lead-cooled fast breeder reactor, particularly suppression of lead oxidation and dissolution of metal elements in structural material such as Fe and Ni into LBE. A potentiometric oxygen sensor has been developed to measure the oxygen concentration in LBE. However, the detail fabrication and condition of reference electrode and the effect of impurity to electromotive force (EMF) output signal of oxygen sensor were not clear and need to be investigated. Two types of reference electrode of the sensor were tested: the mixture of Bi granule and Bi_2O_3 powder; and the mixture of Bi powder and Bi2O3 powder. The sensors were tested in two different conditions of LBE: one with the high purity of LBE and the other one with impurity inside the LBE. The characteristics of the sensor output signal, electromotive force (EMF), was investigated under Ar+3%H_2 gas injection at 550℃ and the theoretical expected value of EMF after 150 hours should be near 500 mV or oxygen concentration equals to 10^<-9> wt%. The result of sensor with mixture of Bi granule-Bi_2O_3 powder did not agree with the theoretical expected value. The EMF value of sensor with mixture of Bi powder-Bi_2O_3 powder agreed with the theoretical expected value when it was tested in high purity of LBE. However, the EMF values were lower than expected value when it was tested in LBE with impurity. The influence of Fe powder as an impurity in LBE has effect on EMF of oxygen sensor under gas mixture injection into LBE.

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Small and major accidents and near misses are still occurring in nuclear power plants (NPPs). Risk level has increased with the degradation of NPP equipment and instrumentations. In order to achieve NPP safety, it is important to continuously evaluate risk for all potential hazard and fault propagation scenarios and map protection layers to fault/failure/hazard propagation scenarios to be able to evaluate and verify safety level during NPP operation. There are major limitations in current real time safety verification tools, as it is mainly offline and with no integration to NPP simulation tools. The main goal of this research is to develop real time safety verification with co-simulation tool to be integrated with plant operation support systems. This includes the development of static and dynamic fault semantic network (FSN) to model all possible fault propagation scenarios and the interrelationships among associated process variables. Safety and protection layers along with their reliability are mapped to FSN so that safety levels can be verified during plant operation. Errors between multiphysics models and real time data are modeled to accurately and dynamically tune FSN for each fault propagation scenario. The detailed methodology will show how to integrate process models, construction of static FSN with fault propagation scenarios, and evaluation and tuning of dynamic FSN with probabilistic and process variable interaction values. Principle Component Analysis method is used reduce dimensionality and reduce process variables associated with each fault scenario. Then map independent protection layers (IPL) to FSN with estimated reliability measures of each protection layer to accurately verify safety for different operational scenarios. Intelligent algorithms is used with multivariate techniques to accurate define the interrelation among process variables, in terms of signal strength and time delay, using Genetic Programming (GP), which will provide basis for fault detection and tuning of FSN, as well as fault diagnosis to understand the closest state of fault scenario. Intelligent algorithm for Bayesian Believe Networks (BBN) is developed to estimate probabilities associated with dynamic FSN with priori and posteriori probabilities. This will dynamically tune FSN with probabilities and real time and simulation data. Probabilistic risk are estimated for each propagation scenario along with the reliabilities of associated IPLs. This will accurately verify safety for all propagation scenarios during plant operation and maintenance. And in order to fine tune propagation scenarios within FSN, rules are synthesized using fuzzy logic using real time and simulation data.

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On March 11, 2011, a severe nuclear accident occurred at Tokyo Electric Power Company (TEPCO)'s Fukushima Daiichi Nuclear Power Plant (hereinafter called as F1). After the accident, the Council for the Decommissioning was established, mainly by the government and TEPCO, and a road map for the F1 decommissioning was drawn up. In the road map, the fuel debris removal from the reactors is scheduled to launch around 2020. In this study, the characteristics and technological issues of each potential treatment scenario were extracted, and the scenarios were prioritized in advance of formal evaluations in the future. The preliminary evaluation results show that long term storage and direct disposal have more positive aspects in terms of economic efficiency and radioactive waste generation. On the other hand, stabilizing processing, aqueous processing, and pyrochemical processing have been estimated to have more disadvantages in such aspects.

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At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units-bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500 ℃. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500℃. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant.

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In the Russian nuclear power industry, developments of reactor plants running on lead and lead-bismuth coolants are currently under way. Such projects include reactor plans SVBR-100, BREST-OD-300 and BREST-1200.

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In the conceptual modular high temperature gas-cooled reactor (MHTGR), the reactor cavity cooling system (RCCS) is designed for decay heat removal. Following the occurrence of a design basis accident such as a loss of forced circulation, the RCCS operation is designed to transition from the active to passive convective heat transfer modes. The RCCS is a cross-cutting concept for any reactor type and in practical design can be air or water-cooled. Here the water-cooled RCCS design consists of multiple standpipes surrounding the RPV and as such provides natural circulation decay heat removal to the ultimate heat sink, the ambient environment. The decay heat load on the full-scale RCCS, of a reference MHTGR, during a transient event varies over time with an estimated peak power of 1.5 MW. This work experimentally investigates the two-phase performance using a 1/4-scaled, three channel natural circulation loop at different radiative-convective thermal loads representing decay heat, and the corresponding 'linear to oscillatory' thermalhydraulics as anticipated under single to two-phase flashing flow.

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