At the Fukushima Daiichi nuclear power plant, zirconium in the fuel rod cladding reacted with water vapor at elevated temperatures due to a loss of cooling water, resulting in the production of a large amount of hydrogen. This hydrogen leaked from the reactor vessel and accumulated in the top of reactor building, eventually leading to an explosion. A hydrogen treatment system that re-oxidizes hydrogen to water vapor is one of the effective methods to prevent such an explosion. A prominent re-oxidation method is via a fixed bed reactor packed with metal oxide pellets. The advantages of this method are its relatively fast oxidation rate without external oxygen/air injection. In this study, experiments and complementary numerical calculations were performed on the hydrogen re-oxidation reaction by metal oxides. The oxidation of hydrogen by copper oxide is modeled by 5 interacting, elementary reactions consisting of 6 chemical species. Experiments were performed using two packed bed set-ups, with measurement of inlet/outlet gas composition and pre/post-analysis of solid composition used to determine constants of the individual reaction rates for numerical calculations. From these reaction constants, the temporal behavior of the outlet gas was predicted.

The Passive Residual Heat Removal System (PRHRS) is very important for the safety of the heating reactor after shutdown. PRHRS is a natural circulation system driven by density difference, therefore, the heat transfer performance of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) has a great impact to the heat transfer efficiency of PRHRS. However, the most research object of PRHR HX is the C-shape heat exchanger at present, which located in In-containment Refueling Water Storage Tank (IRWST). This heat exchanger is mainly used for the PRHRS of nuclear power plants. In the swimming pool-type low-temperature heating reactor (SPLTHR), the PRHR HX is placed in the reactor pool, which the pressure and temperature of the reactor pool are relatively low, and the outside heat transfer mode of tube bundle is mainly natural convection heat transfer. In this study, a miniaturized single-phase pool water cooling system was built to investigate the natural convective heat transfer coefficient of the heat exchanger under the large space and low temperature conditions. The experimental data had been compared with several correlations. The results show that the predicted value of Yang correlation is the closest to the experimental data, which the maximum deviation is about 11%.

Longevity of sensors and portable devices are severely limited by temperature, chemical instability, and electrolyte leakage issues associated with conventional electrochemical batteries. Batteries undergo self-discharge and permanent loss of capacity at high temperatures, exhibit lower voltage and capacity at low temperatures, and leak electrolyte shortening operating lives, corrosion of nearby electronics, and potential safety hazards in the form of burns and poisoning. Instabilities in lithium types often short resulting explosions, fire, and venting of hazardous gasses. Betavoltaics do not have these problems and can operate in a wide temperature range without permanent degradation and loss of capacity, will not explode, and be made safe. Betavoltaic technology is maturing with advances in radioisotope sources, semiconductor materials, and developments in low-power applications, and has been identified by Department of Defense as a disruptive technology that is needed and should be pursued. This study presents experimental and modeling research on the leading betavoltaic technology, recent developments and proposed advancements. The next generation 100 microwatt betavoltaic is introduced along with projected voltage, current, and power characteristics.

The US NRC’s post-Fukushima taskforce recommend extending the coping time up to at least 72 hours. In PWRs, the reactor coolant pump (RCP), and more particularly the shaft seal system located between the motor and the impeller, are directly impacted by these recommendations [AREVA, 2012]. Considering the importance of station black out (SBO) condition, seal performance of RCP shaft seal assembly shall be verified especially under an extended SBO condition. As a final stage of the localization program of the RCP, KAERI performed a seal LOCA test to verify seal leakage characteristics and to produce qualified experimental data for a plant safety analysis during the extended SBO condition. The seal verification test was performed using the localized seal assembly, fabricated by DOOSAN Heavy Industries & Construction Company of Korea, along with the commercial APR1400’s RCP for the Shin-Kori unit-5 nuclear plant to simulate an actual condition of the commercial nuclear power plant. Test condition was derived from a series of safety analysis for the APR1400 plant which reflected the FLEX procedure, and it consisted of three major sub-conditions such as high-pressure and high-temperature condition, saturated condition, and low-pressure and low-temperature condition. The major concerning point of the test is to measure a seal leakage flow rate during the test period. Total leakage amount and maximum leakage flow rate through the localized shaft seal assembly during the whole period of the test are in comparatively smaller ranges than the conventional input values used for a safety analysis of APR1400.

Nuclear heating reactor is a new type of power plant that uses nuclear energy as heat source. Low temperature nuclear heating reactor should be the forerunner and main force for developing nuclear heating plant in China. Due to the lower water temperature required by the heating system, this dedicated, non-power generating nuclear reactor works at low temperatures and pressures with inherent safety features. The design, construction and operation of the nuclear heating reactors in various countries in the world were reviewed in this paper, and China’s new demonstration nuclear heating project and NHR-200 low-temperature heating reactor which would be used was discussed in the paper. We put forward the developing route and suggestion for the development of low-temperature heating reactor in China.

The fuel cladding is an important barrier to the release of fission products to the environment. Its integrity must be conserved during the in-reactor lifetime and during the spent fuel pool and dry cask storage. The corrosive interaction between the cladding and the water coolant in light water reactors leads to the oxidation of the zirconium-based cladding. A fraction of the hydrogen released due to those corrosive interactions or the radiolysis of the water coolant is picked-up by the fuel cladding. It diffuses inside the cladding driven by the concentration and temperature gradients. Eventually, its concentration can increase beyond a certain limit above which hydrogen precipitates as hydrides. The formation of hydrides can embrittle the cladding and leads to micro-cracks that can compromise the cladding integrity. At the spacer grids locations, the mixing vanes will create swirl flow and mixing of the coolant leading to a high temperature gradient on the fuel rod cladding. This temperature gradient is a strong driving force for hydrogen to diffuse from high to low temperature locations. Therefore, the hydrogen behavior around the spacer grids with mixing vanes is important to model. In this work, the computational fluid dynamics code START-CCM+ is used to model the effect of the mixing vanes on the temperature profile on the cladding outer surface. It ws coupled with the transport code MPACT and the fuel performance code BISON. The computational model consisted of a 5 × 5 fuel rods subassembly with a guide tube in the central location. The obtained cladding temperature profile on a fuel rod of interest was applied as a boundary condition to BISON to model the hydrogen behavior around the spacer grids in a three-dimensional manner. Three spacer grids were modeled at elevations of 217.9 cm, 270.14 cm and 322.35 cm. The hydrogen behavior at each of those locations is evaluated and compared to assess the importance order of those locations.

A depressurization accident is the one of the design-basis accidents of a Very-High-Temperature Reactor (VHTR). When a depressurization accident occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, it is important to understand the mixing processes of different kinds of gases in the stable and unstable stratified fluid layers. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Therefore, in order to predict or analyze the air ingress phenomena during a depressurization accident, it is important to develop a method for the prevention of air ingress during an accident. We carried out experiments and numerical analysis using three-dimensional (3D) CFD code to obtain the mixing process of two-component gases and the flow characteristics of localized natural convection. This study also investigated a control method for the natural circulation of air through the injection of helium gas. The numerical model consists of a storage tank and a reverse U-shaped vertical rectangular passage. They are separated by a horizontal partition plate. One sidewall of the high-temperature side vertical passage is heated and the other sidewall is cooled. The low-temperature vertical passage is cooled by ambient air. The storage tank is filled with heavy gas and the reverse U-shaped vertical passage is filled with a light gas. In the vertical passage of the high-temperature side, localized natural convection is generated by the temperature difference between the vertical walls. The results obtained from the experiments were quantitatively simulated using 3D numerical analysis. The two component gases were mixed via molecular diffusion and natural convection. After some time elapsed, natural circulation occurred through the reverse U-shaped vertical passage. These flow characteristics are the same as those of phenomena generated in the passage between a permanent reflector and a pressure vessel wall of the VHTR.

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. This can lead to the development of more compact and more efficient components, e.g. heat exchanger and compressors. A specific field of interest is a new decay heat removal system for nuclear power plants which is based on a turbine-compressor-system with supercritical CO 2 as the working fluid. In case of a station blackout this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. This scenario has already been investigated by means of the thermo-hydraulic code ATHLET, numerically demonstrating the operation of this system for more than 72 h. The practical demonstration is carried out within the Project “sCO 2 -HeRo”, funded by the European Commission, in which a small scale demonstration unit of the turbo compressor shall be installed at the PWR glass model at GfS, Essen, Germany. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger needs to be as compact and efficient as possible. Therefore, a diffusion welded plate heat exchanger (DWHE) was developed and manufactured at IKE. It has been designed with rectangular mini-channels (0.5–3 mm hydraulic diameter) to ensure high compactness and high heat transfer coefficients. Due to uncertainties the DWHE has to be tested in regard to the actual possible transferrable heat power and to the pressure loss. According to this demand a multipurpose facility has been built at IKE for various experimental investigations on supercritical CO 2 , which is in operation now. It consists of a closed loop where the CO 2 is compressed to supercritical state and delivered to the test section. The test section itself can be exchanged by other ones for various investigations. After the test section, the CO 2 pressure is reduced and the liquid is stored in storage tanks, from where it is evaporated and compressed again. The test facility is designed to carry out experimental investigations with CO 2 mass flows up to 0.111 kg/s, pressures up to 12 MPa and temperatures up to 150 °C. The first subject of interest will be the study of the thermal behavior of a DWHE using supercritical CO 2 as a working fluid close to its critical point. Experiments concerning pressure loss and heat transfer will be carried out as a start for fundamental investigation of heat transfer in mini-channels. This paper contains a detailed description of the test facility, of the first test section and first results regarding heat transfer power and pressure loss.

Zirconium alloys remain as the main cladding materials in most water reactors. CZ zirconium alloy is recently developed for advanced PWR fuel assembly (STEP) designed by China General Nuclear Power Group (CGN). There are two kinds of zirconium alloys, designated as CZ1 and CZ2 respectively. It is well known that creep property is one of the most important characteristics to evaluate the integrated performance of new commercial alloys. In this study, we have collected the creep date of CZ alloys heat-treated at different temperatures. The result shows that the creep rates of the zirconium alloys enlarge with the creep stress in the temperature of 375 °C. The creep resistance is improved with the increasing annealing temperature. The creep resistance of CZ-RXA (high temperature) is better than CZ-SRA (low temperature). The creep rate of CZ alloys at relatively lower annealing temperature is much larger than that at higher annealing temperature. At the given stress condition, the creep behaviors of different annealed CZ alloys are found to have almost the same tendency. In addition, the creep rate of CZ1 is smaller than CZ2 at the same annealing temperature.

Thermal striping on the core instrumentation plate (CIP) at the bottom of the upper internal structure (UIS) of an advanced loop-type sodium-cooled fast reactor in Japan (Advanced-SFR) has been numerically investigated. At the top of the core below the CIP, the sodium at high temperature flows out from the fuel subassemblies (FSs) and the sodium at low temperature flows out from the primary control rod (PCR) and backup control rod (BCR) channels, and also the radial blanket fuel subassemblies (RBFSs) at the outer side of the core. In order to predict the thermal striping on the CIP caused by mixing fluids at different temperatures from the FSs, the PCR and the BCR channels, and the RBFSs, a numerical estimation method using a spatial connection methodology between the upper plenum analysis and the local area analysis for the target area has been developed. By using the connection methodology, the numerical simulation considering the influence of the transversal flow in the UIS and the external flow around the UIS in the upper plenum can be performed to improve the accuracy of the estimation results. In this paper, the outline of the spatial connection methodology including data transfer technique from the upper plenum analysis to the local area analysis was described. As a validation process, numerical simulation of the water experiment using the test apparatus named TAFUT which was 1/3-scaled 1/6 partial model of the upper plenum of the Advanced-SFR was performed to confirm applicability of the spatial connection methodology to a practical thermal striping problem. The numerical result of temperature distribution was compared with the measured result in TAFUT experiment. Additionally, mesh sensitivity of the local area analysis model to the numerical results was indicated by using a small and a large area models in order to suggest an appropriate local area analysis model.

Reactor coolant pump (RCP) is one of the most critical devices in third generation of pressurized water reactor nuclear power plant. EMD shield pump and KSB wet winding pump are two representative kinds of RCPs without complex shaft seal system. Due to cancellation of shaft seal system, the entire rotors (including the flywheel) are immersed in the coolant. The losses in RCPs take one third of the total power including rotation loss caused by rotor in the water, electromagnetic loss in the shielding sleeve,the heat transferred through high temperature coolant, and heat generated by bearing.Because of the losses listed above, bearing and winding are heated,and the losses make temperature rise. in order to ensure that the motor is working properly at low temperatures, the company EMD and KSB design the RCP internal cooling circulation which brings the heat out to ensure the normal operation of the RCPs. The RCP internal cooling circulation includes inlet flow area, auxiliary impeller, thrust bearing, the lower flywheel, motor can, upper radial bearing, upper flywheel, outlet flow area, and external heat exchanger,etc. Flow characteristics in every flow path determine the flow distribution and heat transfer, and the flow distribution determines whether the cooling performance of RCP internal cooling circulation meets the requirements. In order to control operating temperature of motor and bearing, and to optimize heat transfer, adjusting the size of flow area and changing the flow characteristics arecritical. flow field and temperature field in RCP internal cooling circulation need overall analysis. Flow distribution can be obtained theoretically through the calculation of an overall three-dimensional model.But on the one hand, the calculation time is long due to a complex three-dimensional model with a large quantity of grids, on the other hand, it is easier to casue errors in local processing and the errors are difficult to find or correct. For rapid analysis and optimization of flow and heat transfer in RCP internal cooling circulation, ensure the motor winding and bearing operate at an appropriate temperature, the local characteristics of RCP internal cooling circulation are studied, one-dimensionalanalysis method of RCP internal cooling circulation is developed. This one-dimensional analysis method can be used to predict the flow distribution of each part of RCP internal cooling circulation according to change of the channel geometry parameters, key dimensions, boundary conditions and rotor speed. The geometric parameters are optimized by analyzing the flow distribution, and the purpose of design guidance are achieved.

A Probabilistic Risk Assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan, but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is intended to develop PRA methodology for a combination of low temperature and snow for a Sodium-cooled Fast Reactor (SFR) that uses the ambient air as its ultimate heat sink for decay heat removal under accident conditions. Annual excess probabilities of low temperature and of snow are statistically estimated based on the meteorological records of low temperature, snow depth and daily snowfall depth. To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., plugged intake and/or outtake for the DHRS and for the emergency diesel generator (EDG), unopenable door on the access routes due to accumulated snow, failure of the intake filters due to accumulated snow, possibility of freezing of the water in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around the intake and outtake induced by loss of the access routes, this study has investigated effects of electric heaters installed around the intake and outtake as an additional countermeasure. By using the annual excess probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence is failure of the electric heaters and loss of the access routes for snow removal against the combination hazard at daily snowfall depth of 2 m/day, duration time (snow and low temperature) of 1 day.

Although no one would like to see, a severe nuclear reactor accident may result in reactor core melting, the fuel melt dropping into water in the reactor vessel, and then interacting with coolant into steam explosion. Steam explosion is a result of very rapid and intense heat transfer and violent interaction between the high temperature melt and low temperature coolant. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. Steam explosion may endanger the reactor vessel and surrounding structures. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the containment integrity. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, fist, small to intermediate scale experiments related to premixing, triggering and propagation stages are reviewed and summarized in tables. Then the intermediate to large scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities’ work is also discussed. The studies on steam explosion are vital for reactor severe accident management, and will lead to improved reactor safety.

In order to gain high strength, fine grain size, stronger anti-corrosion property, and especially low permeability, the material 022Cr19Ni10N was chosen to manufacture the Rod Travel Housing Forging (RTHF) for Control Rod Drive Mechanism (CRDM). But, cracks were found in some forgings failing to meet the requirements of ultrasonic testing (UT). The causes of the forging cracks of this austenitic stainless steel forging were investigated by means of metallography, scanning electron microscopy (SEM) and other experimental methodology. The results indicated that the second δ-ferrite phase leads to the forging cracks between γ-δ interface during the low temperature forging process, and finally leads to the forging failure. It’s found that the cracks are distributing along the stripe δ-ferrite, and almost distributing in the same area as the large size δ-ferrite by metallography & SEM microstructure observation. The δ-ferrite is firstly found in the electroslag ingot, and in which, the distribution and size is different from the case to the core. The largest size δ-ferrite is around the core area, and this characteristic passes on to the final forging microstructure, although the shape, quantity & distribution of the δ-ferrite changed during the manufacturing process. Most forging cracks were found around the core area of the forging by UT examination. In the final forging process, when the forging temperature drops to 750∼850°C, the δ-ferrite have been forged to stripe shape and hundreds-micron size while the plasticity of the austenite reduce. What’s more, there are large hot plasticity differences between the δ-ferrite and the austenite, so the forging cracks initiate between γ-δ interface and extend to the area around to be a long crack in the low temperature forging process. In order to avoid the forging cracks in the Rod Travel Housing Forging, it’s necessary to reduce the content of δ-ferrite or improve the final forging temperature. Improving the final forging temperature, to guarantee the plasticity of the δ-ferrite and austenite, is another process to reduce the cracks. But while the temperature improves, the grain size grows rapidly, and may form mixed structure. So the most effective mean to reduce the content of δ-ferrite is to redesign the chemical components, mainly by increasing the nitrogen content from 0.06 (wt, %) to 0.12(wt, %), which makes the low temperature forging process for fine grain size possible. In the high-nitrogen-content forging, the δ-ferrite distributed sporadically and no δ-ferrite strip is found. By increasing the austenite forming elements (especially nitrogen), the cracks during low temperature forging process are avoided. What’s more, owning to the optimization of chemical compositions and manufacturing processes, the Rod Travel Housing Forging got fine grain size, low relative permeability, and good comprehensive mechanical properties with the ultimate tensile strength up to 570MPa.

For nuclear power stations, the main function of Essential Service Water System (ESWS) is to discharge the waste heat from reactor core and spent fuel pool to the environment controllably, which is directly related to the safety and economy of nuclear power stations. Usually ESWS use open water from sea, rivers, lakes, reservoirs, as heat transfer medium. Extremely harsh environmental conditions may disable system functions and even lead to ESWS failure, directly reduce the safety and economy of nuclear power stations, and cause serious nuclear accidents. Failure of ESWS is one of the main reasons that lead to the Fukushima nuclear accident because of the loss of electricity after the earthquake and tsunami. Based on the typical ESWS configuration and conditions of serving nuclear power stations in China, the influence of environmental conditions on the function of water system is studied, and the corresponding measures are analyzed. These conditions can be divided into three categories: temperatures, water levels, and physical and chemical characteristics. Temperatures affect cooling characteristic of ESWS mainly. Nuclear power stations in tropical areas need to focus on cooling capacity might be reduced by high temperature. Those in cold region need attention to excessive cooling and freezing problems caused by low temperature. The influence of water levels is mainly fluid transport capacity and selection of equipment to ESWS. When the range of natural water level is too wide, designers shall consider measures to narrow it, such as the construction of highly reliable reservoir. Inland nuclear power stations shall try to ensure the reliability of ESWS; prevent water level changes beyond the scope of design caused by drought and flood disasters. The effects of physical and chemical properties are derived from the open water characteristics, including high salinity, high chloride ion concentration, carrying solid particles, suspended solids, and aquatics, and so on. These characteristics will cause the equipment and pipeline eroded or even damaged, aqueducts of intake and output jammed, heat exchangers of the final heat sink weakened and other negative effects, resulting in ESWS performance decline. Some of these factors are the characteristics of station site natural environment, some others are changes caused by human activities. Some factors are sustained, long-term; some others may be sudden, temporary. Influence on these factors need to be taken measures from many aspects, including structure, biological disinfection, special materials and equipment, environmental protection measures around the nuclear power station, and so on. On the whole, the environmental factors that affect ESWS in the nuclear power stations are wide, and the influence mechanism is more complex. These factors ultimately act on ESWS, but most of them cannot be banished inside of ESWS or the final heat sink system. Against the negative effects from environmental conditions, it has to be considered from all steps in the engineering of nuclear power stations, including design, construction and operation. All the measures shall be suitable to local conditions, in order to ensure the safety and economy of nuclear power stations.

The passive containment cooling system (PCCS) of the 3rd generation APWR utilizes natural phenomena to transfer the heat released from the reactor to the environment during postulated designed basic accidents. Steam condensation on the inner surface of the containment shell is one of the most dominate mechanism to keep the ambient conditions within the design limits. Extensive experiment and theoretical research shows condensation is a complex process, gas pressure, film temperature and velocity of the gas have impact on the heat transfer coefficient. To span the expected range of conditions and provide proper model for evaluating the condensation heat transfer process, SCOPE test facility was designed by State Nuclear Power Technology Research & Development Centre (SNPTRD) in various conditions anticipated the operating range of CAP1400 in accident conditions. Pressurized test section with a rectangular flowing channel was used, with one of the walls cooled to maintain low temperature for condensing, supplying systems was designed for different pressures, gas temperatures, velocities and coolant water temperatures. Facility components, test section structure, supplying systems and measurement technology were described in this paper, also results of some pre-tests was introduce to show property of the facility.

One of the most crucial issues in development of the lead-bismuth cooled fast reactors (LFR), and the accelerator-driven system (ADS) is the compatibility of structural steels with liquid lead-bismuth eutectic (LBE) at high temperature range from 500°C to 650°C. In the primary circuit of LBE, metallic impurities such as Fe, Cr and Ni dissolve from structural steels into LBE at high temperature region, are transported to low temperature region, and precipitate as solid metals on channel wall. In this circulation, the data bases of diffusion coefficients of the metallic impurities are required for the estimate of mass transfer rates. In order to investigate the characteristics of diffusion of metallic impurities, the measurement of diffusion coefficients of Ni and Fe in liquid LBE was carried out using a capillary method. The concentrations of the diffusing species in LBE were determined by means of the inductively coupled plasma mass spectrometry (ICP-MS). From the measured axial distribution of Ni and Fe concentration in the LBE, the diffusion coefficient of Ni and Fe were obtained. The results are expressed by D N i = 1.7 × 10 -3 exp ⁡ -3.63 × 10 -4 / R T 500 ° C ≤ T ≤ 650 ° C cm 2 / s D Fe = 3.5 × 10 -3 exp ⁡ -4.15 × 10 -4 / R T 550 ° C ≤ T ≤ 650 ° C cm 2 / s It was found that the value of diffusion coefficient of Ni is similar to the diffusion coefficient of Fe in liquid LBE.

PRHR system for low temperature and low pressure pool-type LWR, AHR400 is designed by two-phase closed thermosyphon and experimental validation is conducted. AHR400 is dedicated only to heat generation used in seawater desalination and operation temperature and pressure. LBLOCA is not considered as DBA due to no pipeline in primary system. There are LOHS and SBO for DBAs and PRHR system reduces damage during DBAs. Design of the PRHR system follows Direct Reactor Auxiliary Cooling System (DRACS) type. Two-phase closed thermosyphon, which uses phase change of working fluid, is applied to the PRHR system and the heat transfer in thermosyphon are analyzed by thermal resistance calculation model. Experimental thermosyphon that has similar thermal condition with the thermosyphon in designed PRHR system was explored for validation. The results show that the evaporation model overestimates heat transfer rate on the evaporator region.

In this paper, the low temperature sintering technology to UO 2 pellets has been introduced, and we have studied the high temperature creep properties of the pellets which manufactured by low temperature sintering. The sintering temperatures are 1073K, 1273K, 1473K and 1673K, sintering time are 1 hour, 2 hours and 3 hours respectively. We obtained the highest sintering density of pellets at 1673K with 3 hours, which is 10.41g/cm 3 (94.98% theoretical density). The grain size of pellets which prepared by low temperature sintering technology and traditional technology are 9.0μm and 23.8μm respectively. So the high temperature creep properties of the two kinds of pellet must be studied. They were performed at 20–50 MPa, 1673K and 1773K respectively, under a nitrogen atmosphere to shorten the experimental time. According to the results, the creep rates of sintered UO 2 with the grain sizes of 9.0μm and 23.8μm under the load of 10MPa are almost the same. The creep process is controlled by both Nabarro-Herring creep and Hamper-Dorn creep for uranium dioxide with grain size of 9.0μm; while Hamper-Dorn creep is the dominant mechanism for uranium dioxide with grain size of 23.8μm.

Three dimensional numerical simulation is conducted on the CANDU Moderator Test Facility (MTF). Heat generation inside the tank is modeled through surface heating. Transient variations of the temperature and velocity distributions inside the tank are determined. The results show that the flow and temperature distributions inside the moderator tank are three dimensional and no symmetry plane can be identified. A high temperature zone, located on the top left corner of the tank, is identified. The inlet jets are found to flow along the walls of the tank and impinge on each other at the top of the tank. This impingement point is located more towards the right side of the tank. The impingement of these two flow result in a secondary downward moving jet, which penetrates into the tube bundle. This secondary jet divides the tank into two sides. One side contains high temperature liquid and the other side contains low temperature liquid. The temperature contours along the length of the tank have a saddle shape, with high temperatures towards the edges of the saddle. This is due to strong wall jet flows in the middle planes pushing the hotter fluid towards the end walls of the tank. Competition between the upward moving buoyancy driven flows and the downward moving momentum driven flows, results in the formation of circulation zones inside the tank. The numerical results for MTF indicate that the moderator tank operates in the buoyancy driven mode. Any small disturbances in the flow or temperature can make the system unstable and asymmetric. Once the system comes out of symmetry, it cannot go back to symmetry. This results in circulating buoyancy driven flow at one side and a momentum driven flow at the other side of tank. Different types temperature fluctuations are noted inside the tank: (i) large amplitude temperature fluctuations are mainly at the boundaries between the hot and cold; (ii) low amplitude temperature fluctuations are mainly in the core of the tank with more uniform temperature distributions; (iii) high frequency fluctuations are in the regions with high velocities; and (iv) low frequency fluctuations are in the regions with lower fluid velocities.