Indications of Ultrasonic Testing (UT) of 4 weld joints in Primary Loop Recirculation (PLR) piping in Hamaoka unit-2 had been identified during the 20th outage (2004). Metallurgical investigation after cutting from 2 weld joins had been conducted because of its accessibility etc. The 4 weld joins had been welded at the plant construction (1978), operated and IHSI (Induction Heat Stress Improvement) had been performed during the 1st outage (1979). 25 years passed by until the metallurgical investigation. As a result of the investigation, crack growth retardation by the IHSI was confirmed since crack tip was round and plastically deformed.

The Fluoride-salt-cooled High temperature Reactors (FHRs) are an advanced concept using a novel combination of high-temperature coated-particle fuel, low-pressure fluoride-salt coolant and air-Brayton power conversion system. Prismatic fuel or pebble fuel are adopted for the conceptual core designs of FHRs like TMSR-SF, MK1 PB-FHR and SM-AHTR. The high-Prandtl-number FLiBe is mainly adopted as the primary coolant, which specifies in high melting and boiling point and high volumetric capacity. The experimental results obtained from the air, water or inert gas prove reliable for the Prandtl number vary from 0.7 to 7. Little experimental research has been conducted to prove applicability of the above results to the high-Prandtl fluid, fluoride salts in the packed pebble bed. In this paper, a pebble bed experimental facility has been designed and constructed for the FHRs to explore the thermal-hydraulic characteristics of fluoride salts in the reactor pebble bed core. Dowtherm A is adopted as a simulant fluid for the fluoride salts. The cylindrical test section is packed with steel pebbles. The electromagnetic induction heating system is used to provide internal heat source for the pebble beds. The forced flow and convective heat transfer of high-Prandtl-number fluid in the pebble bed with internal heat generation are investigated in the experiment. The fluid inlet temperature and mass flow rate are studied on the thermal-hydraulic characteristics.

Hypothetical Core Disruptive Accidents (HCDA) are dominantly concerned during safety assessment and evaluation in Sodium-cooled Fast Reactor (SFR). With molten core materials discharged into liquid sodium, positive reactivity is potentially introduced due to sodium boiling and molten core compaction, which can cause terrible recriticality. The possibility of recriticality and efficient cooling on the relocated debris bed are significantly affected by the fragmentation behavior of molten core in liquid sodium. With few available mechanism models and benchmarks, many investigations have been conducted on the fragmentation characteristics during molten fuel-coolant interaction (MFCI). In the present study, molten copper is used for molten simulant to be discharged into the liquid sodium pool through guiding tube based on a multifunctional experimental facility (COSA). The simulants are heated by electromagnetic induction system in customized ceramic crucible and the molten materials are controlled by magnetic lifting system to be drained through the guiding tube into the bottom liquid sodium pool. Temperature variation and pressure change in the liquid sodium pool are acquired against the energy release during MFCI. Furthermore, the fragments cleaned by water medium are measured and recorded for distribution and morphology analysis. Significant pressure pulses and temperature gradient almost not occur during MFCI and the molten copper is finely fragmented possibly due to hydrodynamic and thermodynamic effects. And the experimental results are helpful to confirm the prediction of fragmentation mechanism and to validate physical model, which can be applied to the development and validation of analysis code.

Passive residual heat removal system (PRHRS) is of great significance for reactor shutdown safety. The PRHRS of a small modular reactor, such as the integral pressurized water reactor (iPWR) and the modular high temperature gas-cooled reactor (MHTRG), is composed of the primary loop (PL), intermediate loop (IL) and air-cooling loop (AL). The AL is a density-difference-driven natural circulation caused by the difference of air temperature at the inlet and outlet of the air-cooling tower. Thus, it is possible to adopt the air flow in AL to generate electricity for post-shutdown reactor monitoring. In this paper, a novel residual heat electricity generation system (RHEGS), which is composed of the PRHRS and a vertical wind generator installed in the air-cooling tower, is proposed for the power supply of post-shutdown monitoring instruments. To verify the feasibility of practical implementation, the dynamical model of this newly designed RHEGS including the dynamics of PRHRS, windmill, rotor as well as doubly-fed induction generator (DFIG) are all given. Then, both steady-state and transient verification for the RHEGS of a nuclear heating reactor NHR200-II plant with a rated thermal power of 200 MW th is carried out, which shows that the output active power of RHEGS can be 20∼30kW which is about 1% the residual heat flux and can fully meet the power requirements of post-shutdown monitoring instruments.

High level waste (HLW) originating from reprocessing of spent fuel commercial power reactors contains more than 40 different elements. Vitrification into borosilicate glass at 1100 ∼1200°C is the process of choice. It is routinely used to immobilize the radioactive waste constituents in a chemically stable matrix for a final geological disposal. Melting process for commercial HLW glasses are variants of two basic designs. (1) The joule-heated ceramic-lined melter (JHCM) originally developed in the United States in 1973 and used in several nuclear sites in the world. (2) The hot-walled induction melter (HWIM), developed in France starting in 1962 and used in France and UK. These technologies, while effective, do pose limitations in waste form compositions and throughput rates. Particularly HLW originating from commercial spent fuel reprocessing usually contains noble metals elements such as ruthenium (Ru), rhodium (Rh), and palladium (Pd) which require special attention when this waste is vitrified. Recent advances to both of these baseline technologies are beginning to be used with large gains to ensure waste form flexibility, throughputs, and noble metals compatibility. The next generation JHCMs use a steeply sloped bottom and a subsidiary-heating bottom drain to allow these noble metals particles to be effectively flushed from the melter with higher waste loadings. Similar melters are being installed near Guangyuan/Sichuan province, China by German consortium team and being developed for the second K-Facility melter at Rokkasho by Japan Nuclear Fuel Limited (JNFL). As another example the advanced JHCMs will be installed in the Hanford WTP project having large glass pool surface area with rapid bubbling. Significant improvements on induction melters have also been implemented. AREVA recently installed a cold-crucible induction melter (CCIM) in combination with a rotary calciner at La Hague in France. This melter uses radio frequency induction to power the glass melt itself and water cooling of the outer surface maintains a frozen glass shell (skull) as the glass contact material. Because no permanent refractories or embedded electrodes are used, this design allows for high-temperature operation and can tolerate more corrosive melts, and uses a water-cooled, motor-driven mechanical stirrer to comply with noble metals behavior. This paper highlights some of these advances and suggests potential advantages and disadvantages of these next generation melter technologies comparing advanced JHCM with updated CCIM. In conclusion, these melters have made the technologies of choice for new HLW vitrification projects around the world.

Nuclear radiation can induce hardening and embrittlement of RPV steel. Solute defect copper rich precipitate / clusters is considered harden the base steel. Reduce or delay the irradiation hardening and embrittlement of RPV steel could be effective to ensure the safe operation of nuclear power. Furthermore, the expected target of the improved RPV steel will assist consummate nuclear power evaluation to decide whether feasible to extend the service lifetime of nuclear power plant from 40 to 60 years facing at present in China nuclear industry. In order to study the precipitate phase Cu rich clusters, but different from the traditional methods, the present work employed a new technique, electromagnetic induction heating to produce a high temperature gradient along the model rod Fe-1.2wt.%Cu alloy sample. The results exhibit that this method can speed up the research process and is convenient to study the transformation in microstructure and hard / brittle defect phase formation of copper solute rich clusters. In addition, the change of Vickers hardness along the high temperature gradient assist denote the transformation in microstructure and the Cu rich clusters in quantity roughly. And large rolling deformation can promote formation of Cu rich cluster.

Cold crucible induction melter (CCIM) technology has been widely used in melting of non-metallic refractory materials with a high melting point and a requirement of high purity. This paper discusses the application of the CCIM technology in melting of reactor corium (simulated as zirconium oxide in this experiment). In this experiment, the power supply is set at constant power control mode, with a design frequency of ∼100 kHz and a total output power of ∼ 400 kW. The cold crucible consists of copper tubes and a stainless bottom plate which has a dimension of 220 mm in diameter and 420 mm in height. The crucible is charged with zirconium oxide (ZrO 2 ) powder and surrounded by a set of water-cooling induction coils which is designed shorter than the crucible in order to reduce energy loss in the water-cooling bottom plate. The top of the crucible and the coils are at the same height. A Zirconium ring is used to initiate heating process. ∼ kg ZrO 2 powder is used for each experiment. Electrical variables of the induction melter and temperature of the melt are detected and recorded in the entire heating process. For the result, almost 3/4 of the ZrO 2 powder is melt and a thin crust (∼ 2–3 mm) is generated between the melt and the crucible. To be specific, the ZrO 2 powder above the bottom of the induction coils is melt, but others are not. In the melting process, the maximum surface temperature of the melt has reached 2230°C. Furthermore, for better understanding of the relationship between the melting process and the changing process of the electrical variables, an equivalent circuit of the heating system is established and analyzed. It can be concluded that some electrical variables such as the quality factor ( Q ) of the induction coils and the practical frequency can reflect melting status. The conclusion is confirmed by the records transmitted by sensors in the melting process. By comparison between the recorded data and the melting status, it is clear that the Q value can be accurately used in indicating the melting status due to its obvious change in the entire melting process. As the melting pool becoming larger, the Q value tends to be smaller. In contrast, the practical frequency becomes higher as the melt becoming larger. Furthermore, both of them are not influenced by the output power. In addition, the output current and the output voltage also can be used to describe the melting status. In specific, a larger melting pool corresponds to a lower output current and a higher output voltage.

The variation law of motor vibration with load under saturation of magnetic circuit is one of the focus of motor vibration research. T-type equivalent circuit method is shown to be a suitable engineering analysis way. In this paper, the variation of magnetic flux density under saturation state was analyzed by using a small induction motor, and a physical model of magnetic saturation and motor vibration was established by the way of T-shaped equivalent circuit. Based on the physical model, the vibration mechanism of the motor under the saturated magnetic circuit was analyzed and the motor vibration simulation with different load conditions was carried out by the finite element method. Finally, a vibration test rig for motor under load was built and the simulation results were verified by experiments. The results show that when the motor load is less than the rated load, the amplitudes of the electromagnetic force in the middle and high frequency band change little. When the load is close to or higher than the rated value, the harmonic components of electromagnetic force in the middle and high frequency bands increase, and the amplitudes gradually increases due to the saturation of the magnetic circuit. Therefore, in the low-noise motor design, motor torque performance and vibration noise performance should be considered to optimize the saturation control of the magnetic circuit.

A simple and highly reliable motor diagnostic method was developed. The method monitors low-frequency components of leakage flux from motors. Small wear of bearing housing was detected by the method which could not be detected by the conventional vibration or electric current monitoring methods.

The volumetrically heated packed bed has been widely utilized in modern industry, however, due to the variability and randomness of packed bed channels, flow boiling heat transfer characteristics is complex. To study subcooled flow boiling heat transfer characteristics of volumetrically heated packed bed, here electromagnetic induction heating method was used to heat the oxidized carbon steel balls adopted to stack packed bed, while water was utilized as refrigerant in the experiment. The mass flux and heat flux effects on the subcooled flow boiling heat transfer were analyzed. A new correlation was developed to predict the subcooled flow boiling heat transfer coefficient in volumetrically heated packed bed.

Electromagnetic (EM) pumps, which drive liquid metal by electromagnetic force in Fast Reactors, have been developed [1] [2]. Annular Linear Induction Pump (ALIP) is one of EM pumps. There have been some reports in which developed pressure’s drop and fluctuation were described near the top of the relation curve between developed pressure and flow rate (P-Q curve). These phenomena were reported by Gailitis, Kirillov and Araseki. Araseki simulated this phenomenon by two-dimensional (2D) code and reported that it is characterized by vortexes, which were initiated by azimuthal non-uniformity of sodium flow velocity and/or magnetic field, in the sodium flow. We calculated the developed pressure of an EM pump by 2D magnetohydrodynamics (MHD) code and designed the EM pump. The code simulated the developed pressure with high accuracy in normal operations. When the flow rate was lower than the one at the top of P-Q curve, the developed pressure’s drop and fluctuation occurred. The fluctuation could disturb the stable operation of the pump. For fear of this phenomenon, the design of EM pump is sometimes too conservative. To simulate this phenomenon and examine its occurring conditions, we have developed a new three-dimensional (3D) MHD code. Clarification of these conditions for the occurrence of this phenomenon will enable to design a new structure or determine its operation conditions reasonably. This paper describes the simulation results focusing on the occurrence of developed pressure’s drop and fluctuation. We use initial conditions which have azimuthal non-uniformity of sodium velocity to simulate the initiation of vortexes in sodium. The 3D MHD code simulates the developed pressure’s drop and fluctuation by vortexes and provides more clear illustration of their occurrence.

An Fe-Cr-W-V-Ti-N steel named SCRAM (super-clean reduced-activation martensitic) steel was designed for the first wall and blanket structure of fusion power plants. Compared with the Fe-Cr-W-V-Ta steel, TiN can precipitate first at 1650°C rather than TaC by the Thermal-cal Calculation. And we take vacuum induction melting (VIM) and electro-slag re-melting (ESR) together to manufacture the SCRAM steel, which can make the TiN fine and the steel pure. Mechanical properties and microstructures of SCRAM steels irradiated with single-beam (Fe) and sequential-beam (Fe plus He and Fe plus H) at 300°C were studied. The results show that, SCRAM steel can have better mechanical properties before and after irradiation while Ti was doped into the SCRAM steel. It has been reported that the precipitation formed in the steel has effect on the mechanical properties, irradiation properties and the ductile brittle transition temperature (DBTT). The effect of intermediate heat treatment on precipitation behavior and mechanical properties of SCRAM steel was investigated in order to obtain dispersed fine M23C6 carbides. The results indicated that MX carbonitrides precipitated first in the steel with intermediate heat treatment at 870°C rather than M23C6, which led to a decrease of carbon concentration in the supersaturated martensitic matrix and correspondingly a reduced volume fraction and mean size of M23C6. The intermediate heat treatment was beneficial to the mechanical properties, and proposed for reduction on the ductile brittle transition temperature (DBTT).

The article presents the results of the experimental study carried on molten corium and its crystallization products resulted from contact with sacrificial concretes. Experiments were carried out at COMETA installation using induction melting in cold crucible technology in air. The results of the physical-chemical analysis of the examined composite system U 4+ -U 6+ -Zr 4+ -Fe 3+ -Fe 2+ -Si 4+ -Al 3+ -Ca 2+ -O2 − presented in this paper are based on computational diagram of fusibility of quasi ternary oxide systems. This work was carried out within the framework program KONTAKT-II project, LH12224 “Modeling and investigation of hypothetical non-standard situations at nuclear reactors of new type”, experimental work has been partly supported by the SUSEN Project CZ.1.05/2.1.00/03.0108 (European Regional Development Fund (ERDF)). Recommendations for further investigation are presented in this paper.

A fixed bed reactor that operates in gas-liquid co-current down flow is called Trickle Bed Reactor (TBR). It is widely used in chemical engineering. And, recently used in purification of radioactive contamination from contaminated water generated in the Fukushima Daiichi nuclear power plant. There are several flow conditions that occur in the TBR due to gas and liquid flow rate. Since mass and heat transfer rate and particles wetting depend on flow condition, it is necessary to establish the visualization techniques to understand flow condition, transition boundary and properties of gas liquid flow in TBR. In this study, authors employed the lab-scale TBR, made of 100mm inner diameter acrylic column, packed with particles of two sizes (3, 5 mm) that are used in the actual reactor. Water and air were injected from the top of the column and cross-sectional liquid distribution was captured at the bottom of the column by electrical resistance tomography (ERT). ERT is a tomographic technique that provides the cross-sectional conductivity distribution at the rate of about 50 frames per second by injecting current and measuring voltages between the 16 electrodes that are attached around the column. By analyzing the spatial and temporal characteristics of the liquid distribution obtained by ERT, it was found that particle size has only little impact on induction of pulsating flow and larger particle causes distinct pulses. Smaller particle causes blurred tiny pulses due to higher flow resistance. Larger particle (5 mm) is advantageous for pulsating flow.

The linear annular induction electromagnetic (EM) pump was designed for the purpose of transporting the electrically conducting liquid sodium that is used as a coolant in the Sodium cooled Fast Reactor (SFR) in the high temperature of 550 °C. The EM pump was designed by using an equivalent circuit method which is commonly employed to the design of the induction machines. The mathematical equations on the developing pressure and efficiency was found out by using Laithewaite’s standard design formula. The main geometrical parameters for the design were a pump core length, a diameter of the pump and an annular gap size. The dominant electromagnetic variables were the input frequency, magnetic pole pitch and number, and coil turns. Especially, it was shown that the developing pressure and efficiency were maximized at the range of the low frequency less than 60 Hz of the commercial frequency. Also, the size of an annular gap was suitably selected considering the hydraulic frictional loss at the narrow annular channel. The P-Q characteristic was theoretically predicted according to the change of the input current, voltage and power. The functional and structural components of the pump consisted of the material compatible with the high temperature and chemical reactivity of the liquid sodium. The designed pump was predicted to have the nominal flowrate of 10 L/min and the developing pressure of 4 bar.

Korea Electric Power Industry Code (KEPIC) has been developed on the basis of referring to the prevailing U.S. codes and standards which had been applied to the electric power facilities in Korea. KEPIC has been applied to the construction of Ulchin Nuclear Units 5&6 since 1997 as per the endorsement of Ministry of Science and Technology, and is being completely applied to the construction of Shin-Kori and Shin-Wolsung Nuclear Power Plants (NPPs) in Korea and of Barakah NPPs in UAE. Although the technical requirements of KEPIC related to design and inspections for pressure vessels have been developed on the basis of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Korea Electric Association (KEA) that is KEPIC maintenance organization is trying to find and improve unreasonable requirements based on the acquired experiences for the continuous construction and operation of NPPs. In the administrative requirements, KEPIC, unlike ASME, adopted not only mechanical field but also electrical field related to NPPs. Furthermore, the qualifying system for services such as Nondestructive Examination, Heat Treatment, Design and Equipment Qualification Organization was adopted to improve safety and reliability for NPPs. In the technical and administrative requirements, KEA has been performing betterment studies such as ‘integration of welding coordinator in welding quality system of KEPIC’, ‘induction of Korean national standards (KS) materials equivalent to ASME/ASTM materials into KEPIC-MH codes’. The result of study for development of Steel-plate Concrete (SC) structure and modularization technique for NPPs has been adopted to 2010 edition of KEPIC-SNG.

SNPTRD was planning to employ cold crucible induction heating technique to investigate the corium pool heat transfer, using the UO 2 and ZrO 2 mixture as the simulated corium pool. Compared with the actual situation, the primary problem of the cold crucible induction heating was the additional introduction of electromagnetic field. To investigate the cold crucible applicability in simulating the corium pool heat transfer, the distribution of joule heat inside cold crucible and the magnitude of electromagnetic force were carried out. To be more suitable for heat transfer test research, the shape of cold crucible was changed to hemisphere, and the cooling water channels were set horizontally. The simulation results indicated that joule heat distribution would be more uniform with lower frequency power but considering power efficiency, there was only one best frequency. The magnitude of electromagnetic force can be ignored compared to the gravity on the natural convection research.

One of the main research topics of the Chair for Reactor Safety and Technology at RWTH Aachen University and the Institute for Nuclear Waste Management and Reactor Safety (IEK-6) deals with accident scenarios of gas-cooled High-Temperature Reactors, especially the air ingress scenario. Two experimental facilities have just started operation providing experimental data for the validation and improvement of fluid mechanics codes being developed and applied at IEK-6. The INDEX (INDuction EXperiment) facility is able to heat up single spheres inductively up to 1200°C while exposed to defined gas atmospheres and gas flow conditions. This experimental setup is well suited to study pebble / gas flow interactions as well as graphite corrosion phenomena in detail. The NACOK II (NAturzug im COre mit Korrosion) facility is an integral experiment for fluid mechanics and graphite corrosion processes under natural convection effects. It will examine spherical fuel element samples as well as prismatic blocks. In addition the instrumentation is suitable for CFD validation calculations, for example because PIV (particle image velocity) is applied. The data obtained are used to validate and improve computational fluid dynamics (CFD) models for pebble bed reactors or reactor dynamics code like MGT-3D, which are able to simulate air ingress scenarios. The CFD model shall be able to simulate the fluid mechanics as well as the corrosion processes of after a total pressure release. In this paper we report on the status of the experimental facilities as well as on advances in modelling the fluid mechanics of HTR pebble cores.

The 1.2MV, 70ns FWHM induction cell is developed for a 3MV Induction Voltage Adder (IVA) accelerator with three series connected cavities through a high voltage, vacuum insulating transmission line (VITL) driving rod-pinch diode (RPD) for radiography. The experimentally measured maximum relative permeability of IVA used amorphous material under pulse excitation is consistent with saturation wave model with flux density changing rate dB/dt greater than 10T/μs up to 32 T/μs. The remanence ratio B r /B s is experimentally measured for IVA pre-annealed amorphous cores. Upon experimental results, cores in each cell are designed and determined according to current transfer efficiency and volt-second integral necessary for each induction cavity, and peak field preventing VITL bore negative surface from electron emission. The field analysis is carried out for designing VITL vacuum stack and oil cavity that contains cores and azimuthal transmission line with operation field stress being about 50% critical breakdown field. The prototype cell tests validate electric field safety and magnetic core performance. With magnetization inductance and resistance of eddy current calculated for inductive cell, the IVA accelerator circuit model is set up, and simulation predictions approximately according with experimental results are presented.

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.