Three-dimensional coupled neutronics thermal-hydraulics reactor analysis is time consuming and occupies huge memory. A one-dimensional model is preferable than the three one in nuclear system analysis, control system design and load following. In this paper, a corewide three dimensional to one dimensional equivalent method has been developed. On the basis of this method 1D axial few groups constants were obtained. The equivalent cross sections were calculated by general spatial homogenization while the transverse buckling was computed through an equivalence based on the 3D flux conservation. Three steady test cases were performed on one dimensional finite difference code ODTAC and the results were compared with TRIVAC-5. The comparison shows that the one dimensional axial power distribution computed by ODTAC correlates well with the three dimensional results calculated by TRIVAC-5. In this study, DRAGON-4 was used to generate the few-group constants of fuel assemblies and the reflector few-group parameters were calculated by WIMS-D4. These collapsed few-group constants were tabulated in a database sorted in ascending order of fuel temperature, coolant temperature and concentration of boric acid. Trilinear interpolation was adopted in cross sections feedback during the transient analysis. In this paper, G1 rod drop accident(RDA) and G1 rod ejection accident(REA) were performed on ODTAC and the computation results were consistent of the physical rules.

View full abstract

The NRC created a state-of-the-art reactor consequence analysis (SOARCA) project to develop the best estimates of the offsite consequences for potential severe reactor accidents for two pilot plants: the Peach Bottom Atomic Power Station and the Surry Power Station (USNRC, 2012). The short-term station blackout (STSBO) and long-term station blackout (LTSBO) were identified as the major groups of accident scenarios for analysis. Both types of SBOs involve a loss of all alternating current (AC) power. SOARCA-like analyses, which are limited to accident progression except offsite consequences, were conducted for an OPR-1000 PWR. This paper illustrates a preliminary assessment for the mitigative effectiveness of external cooling water injection strategies using fire trucks during a potential extended station blackout accident.

View full abstract

The Severe Accident (SA) leading to core melt had not been considered to occur in Japanese Nuclear Power Plants (NPPs) because various kinds of countermeasures were adopted for the first through third layers of Defense in Depth (DiD). The direct cause of Fukushima Daiichi accident was considered to be the loss of all AC powers caused by an unexpected huge tsunami, but the insufficient measures for the fourth and fifth layer of DiD was pointed out as a lessons learned of Fukushima Daiichi accident. From this viewpoint, foreign NPP investigations have been performed by JANSI in order to gather and analyze information how foreign NPPs are taking SA countermeasures. Before Fukushima Daiichi accident, SA countermeasures adopted in Japanese NPPs were far behind from the advanced foreign NPPs, but nowadays we think that SA countermeasures in Japanese NPPs have caught up and have reached almost the same level as advanced foreign NPPs. Moreover, JANSI is going to study additional SA countermeasures as voluntary basis for the safety improvement of Japanese NPPs based on SRS-46 and foreign NPPs investigations.

View full abstract

This study was carried out to establish the electron beam welding process for a nuclear fuel plate assembly fabrication. A preliminary investigation for plate fuel fabrication was conducted with a consideration of weld performance using AA6061-T6 aluminum alloy made by the EBW (Electron Beam Welding) process. The optimum welding parameters for the fuel plate assembly were obtained in terms of the accelerating voltage, beam current and welding time. The welds made by the optimum parameters showed slightly lower tensile strengths than those of the un-welded specimens. The integrity of the welds by the EBW process was confirmed by the results of the tensile test, an examination of the macro-cross sections and the fracture surfaces of the welded specimens.

View full abstract

Based on a 2mm vertical rectangular channel which was designed by North China Electric Power University, experiments were conducted on the experiment table to research heat transfer deterioration in natural circulation. The result is that, heat transfer deterioration can happen in natural circulation, at which heating section temperature increase rapidly and heat transfer coefficient decrease rapidly. The increase of subcooling degree and mass velocity contribute to delay heat transfer deterioration. Because of low fluid velocity in natural circulation and secondary flow influence in narrow rectangular channel, liquid film can be evaporated easily and heat transfer deterioration happens. Futuremore, liquid film can be squeezed by narrow rectangular channel, so the critical heat flux in narrow rectangular channel is lower than that in others.

View full abstract

This paper presents results related to liquid metal thermal hydraulics. The application of liquid metals in the engineering field is gaining popularity due to their beneficial thermal characteristics. The present tools for thermal hydraulic analysis usually distinguish between pipe and rod bundle flow only. These geometries are the dominant forms. Geometries deviating from the above mentioned one have been modeled by e.g. a pipe, knowing that this approach is just a coarse approximation. Nevertheless, the consideration of other channel geometries is necessary to perform comprehensive investigations of entire plants or facilities. In this study, rectangular channels will be investigated since such channels with all aspect ratios can be found in almost any thermal plant. It is therefore of crucial importance to predict the thermal hydraulic behavior correctly during any situation. A literature review is performed to derive empirical models which can be implemented into thermal hydraulic analysis tools. With the code TRACE the new models are validated by means of posttest analysis of experiments. The investigation shows that the empirical models predict accurate results. Furthermore, it is shown that geometry dependent models are necessary to perform comprehensive studies and to obtain reliable results.

View full abstract

Heat transfer fluctuation between a fluid and a solid may cause problems related to wall temperature fluctuation, such as high cycle thermal fatigue in materials. Moreover, heat transfer enhancement due to flow turbulence is similar to mass transfer enhancement, which leads to flow accelerated corrosion (FAC) downstream of an orifice in pipe flows. In order to avoid and/or predict these phenomena, quantitative information on the spatio-temporal fluctuation of the heat transfer is necessary. In the present study, a technique using high-speed infrared thermography was used to measure the spatio-temporal heat transfer to a turbulent water pipe flow around an orifice plate (bore ratio: d/D = 0.49, Re_D ≈ 12,000). The spatio-temporal distribution of the heat transfer coefficient was evaluated based on the temperature fluctuation of a heated thin-foil measured using high-speed infrared thermography (at approximately 800 Hz). As a result, it was revealed that the heat transfer downstream of the orifice fluctuated violently, and the instantaneous structure of the heat transfer was remarkably finer than the streaky structure for the fully developed pipe flow. The time-averaged value of the heat transfer had a maximum at approximately two diameters downstream of the orifice, where the rms value of the fluctuation and its characteristic frequency also became much higher than those for the fully developed pipe flow.

View full abstract

The demand for clean, non-fossil-based electricity is growing. Therefore, the world needs to develop new nuclear reactors with inherent safety and higher thermal efficiencies in order to increase electricity generation per kilogram of fuel and decrease detrimental effects on the environment. To address these issues a number of countries worldwide are developing next generation or Generation-IV nuclear-reactor concepts (six in total) and, as a result, Nuclear Power Plants (NPPs) will have significantly higher operating parameters, especially, temperatures (550-1000°C). Also, Generation-IV nuclear technology will include SuperCritical-Pressure (SCP) reactor coolants (helium and water) and/or SCP working fluids in power cycles (carbon dioxide, helium and water). Due to this reliable and accurate prediction methods for heat transfer in SC Fluids (SCFs) should be developed and verified. These methods include: 1) empirical correlations; 2) CFD software; and 3) Thermalhydraulics codes. The present paper deals with Computational Fluid Dynamics (CFD) PHOENICS software studies, which intended to predict heat transfer in SuperCritical Water (SCW) flowing upward in a 4-m bare vertical tube (D=10 mm) as an initial approach. The current study is related to lower range of mass fluxes (about 200 kg/m2s) in which heat transfer can be influenced by natural convection. In general, Heat Transfer Coefficients (HTCs) in bare tube can be considered as a conservative approach in predicting minimum possible HTCs in more complex geometries such as bundles. Compared to empirical 1-D correlations, CFD studies allow to look inside flow and to have a better picture of various phenomena related to heat transfer in SCFs. Experimental data on SCW were compared with predictions from CFD calculations in this study. The obtained results show that within some operating conditions CFD PHOENICS can predict experimental HTC values reasonably well. In other conditions, especially, within a Deteriorated-Heat-Transfer (DHT) regime more studies are required. Also, selected velocity and Turbulent Kinetic Energy (TKE) profiles across the radial and axial directions of a tube have been provided.

View full abstract

A liquid-Li free-surface stream flowing at 15 m/s under a high vacuum of 10^<-3> Pa serves as a beam target (Li target) for the planned International Fusion Materials Irradiation Facility (IFMIF). The Engineering Validation and Engineering Design Activities (EVEDA) for the IFMIF are implemented under the Broader Approach Agreement. As a major activity of the Li target facility, the EVEDA Li test loop (ELTL) was constructed by the Japan Atomic Energy Agency. The stable Li target under the IFMIF conditions (Li temperature: 250 °C, velocity: 15 m/s, vacuum pressure: 10^<-3> Pa) has been demonstrated so far by using the ELTL. This study focuses on a cavitation-like acoustic noise that was detected in a downstream conduit where the Li target flowed under vacuum conditions. This noise was investigated using acoustic-emission sensors installed at eight locations via acoustic wave guides. The sound intensity of the acoustic noise was examined against the cavitation number of the Li target. In addition, two types of frequency analysis by the fast Fourier transform (FFT) and the continuous wavelet transform (CWT) were performed to characterize the acoustic noise. The results are as follows: 1) the acoustic noise was intermittent and consisted of multiple acoustic emissions whose time-width and frequency were 0.2 ms and less than 350 kHz respectively, and 2) the development of the acoustic noise was distinguished by four stages with regards to the change of sound intensity by using the cavitation number. From these results, we concluded that acoustic noise was generated from cavitation occurring in the downstream conduit.

View full abstract

The KAERI (Korea Atomic Energy Research Institute) has been studying pyroprocess technology, and some research facilities such as the ACPF (Advanced Spent Fuel Conditioning Process Facility) and the PRIDE (PyRoprocess Integratd inactive Demonstration facility) have been developed. However, the safety regulations, standards, and associated information applicable to the development of safety documents for a pyroprocess facility dealing with spent fuels was not yet prepared. In this paper, identification of U.S. DOE nuclear safety regulations, standards, and associated information applicable to the development of safety basis documents for a nuclear fuel pyroprocess facility was summarized. Topics addressed include: facility hazard categorization, hazard and accident analysis procedures, design basis accident identification, safety system structure and component (SSC) identification, and technical safety requirement identification. This study will be used to provide insight in applying regulations, standards, and associated information used in developing the safety basis for a Korean pyroprocessing facility.

View full abstract

ABSTRACT Probabilistic Risk Assessment (PRA) for external events has been recognized as an important safety assessment method especially after the TEPCO's Fukushima Daiichi nuclear power station accident. The PRA should be performed not only for earthquake and tsunami which are especially key events in Japan, but also other external hazards (e.g. strong wind). Therefore the PRA methodology for the latter events should be developed. From this background, the present study is intended to develop the PRA methodology against the strong wind. In this study, the methodology was developed for Sodium-cooled Fast Reactors paying attention to that the ambient air is their final heat sink for removing decay heat under accident conditions. First, this study estimated hazard curves of the strong wind by using Weibull and Gumbel distributions based on weather data recorded in Japan. Second, important structures and components for decay heat removal were identified and an event tree resulting in core damage was developed in terms of wind load and missiles (i.e. steel pipes, steel beams, cars and trees) caused by strong wind. The failure probability of the reactor building is assessed by considering the possibility of damage of outer surface of the reactor building due to the wind load. The missiles that can reach those components and structures placed on high elevations were identified, and the fragility of the components and structures against the missiles was calculated as a product of two probabilities: i.e., a probability for the missiles to enter the inlet or outlet in the decay heat removal system, and a probability of failure caused by the missile impacts. Finally, conditional decay heat removal failure probabilities were quantified by introducing the fragilities into the event tree. The core damage frequency (CDF) was estimated about 4× 10^<-9>/y. A dominant sequence was led by the assumption that the operators could not extinguish tank fire caused by the missile impacts and the fire induced loss of the decay heat removal system. Through the above, this study developed the PRA methodology against the strong wind.

View full abstract

Remote handling technology has been systematically developed for nuclear fuel cycle facilities in Japan since 1970s, primarily in parallel with the development of reprocessing and HLLW (High Level Liquid Waste) vitrification process. In case of reprocessing and vitrification process to handle highly radioactive and hazardous materials, the most of components are installed in the radiation shielded hot cells and operators are not allowed to enter the work area in the cells for operation and maintenance. Therefore, a completely remote handling system is adopted for the cells to reduce radiation doses of operators and increase the availability of the facility. The hot cells are generally designed considering the scale of components (laboratory, demonstration, or full-scale), the function of the systems (chemical process, material handling, dismantling, decontamination, or chemical analysis), and the environmental conditions (radiation dose rate, airborne concentration, surface contamination, or fume/mist/dust). Throughout our domestic development work for remote handling technology, the concept of the large scale integrated cell has been adopted rather than a number of small scale separated cells, for the reasons to reduce the total installation space and the number of remote handling equipment required for the each cell as much as possible. In our domestic remote maintenance design, several new concepts have been developed, tested, and demonstrated in the Tokai Virtrification Facility (TVF) and the Rokkasho HLLW Vitrification and Storage Facility (K-facility). Layout in the hot cells, the performance of remote handling equipment, and the structure of the in-cell components are important factors for remote maintenance design. In case of TVF (hot tests started in 1995), piping and vessels are prefabricated in the rack modules and installed in two lines on both sides of the cell. These modules are designed to be remotely replaced in the whole rack. Two overhead cranes and two bilateral servo-manipulators and ITV (Industrial Television) systems for monitoring are installed for Maintenance and also Operation in the cell. These cranes and manipulators are mounted on the bridge to ensure the wide range of operations in the cell and also designed to be remotely maintained themselves by each other. In case of K-facility (active tests stated in 2007) the operating experiences at TVF were reflected to make some improvements on its remote handling system in order to ensure the availability and reduce the cost. There adopted the unilateral servo-manipulator and the auxiliary hoist with remote operation support system, the rack module design for periodically replaced components, and the direct contact maintenance for the in cell cranes and manipulators in the shielded parking space. The glass melter in the vitrification process is designed to be replaced every 5 years, so the remote replacement and dismantling technology for the spent melters have been also developed and installed in TVF and K-facility for 40 years' operation. This paper describes our development experiences on the design, construction, operation, and maintenance of the remote handling systems in nuclear fuel cycle facilities in Japan.

View full abstract

The heat transfer efficiencies of hydrophobic micro/nano structured (HMN) surfaces that have 〜160° contact angle under atmospheric conditions were investigated experimentally. The departing diameter and the contact angle hysteresis of droplets were .measured by capturing top and tilted side views of condensation phenomena with a high speed camera and an endoscope, respectively. Condensation behaviors on the surface were observed at the micro-scale using an environmental scanning electron microscope (ESEM). Apparently-spherical droplets formed at very low heat flux q" 〜20 kW/m^2 but hemispherical droplets formed at high q" 〜440 kW/m^2. At high q", heat transfer coefficients were lower on the HMN surface than on a hydrophobic smooth (HS) surface although the HMN surface is water repellent so droplets roll off. The results of contact angle hysteresis and ESEM image revealed that the reduced heat transfer of the surface can be attributed to the large size of departing droplets caused by pinning of condensed droplets at nucleation sites. The results suggest that the effect of q" or degree of sub-cooling of a condensed wall depends on the droplet shape, which is closely related to removal rates of condensates and finally to heat transfer.

View full abstract

Based on the outside cooling of the Passive Containment Cooling System (PCCS), a simplified containment outside model was built and was then numerically solved. Air circulation driven by natural forces for a steady state along with the coupled thermal-hydraulic phenomena was then obtained and analyzed. For the PCCS, there are two encountered opposed forces. The density difference between the atmosphere and the channel is the driving force and the structure contributes to the drag forces. Thus, Variations of up-comer structure width between the steel wall and the buffer plate was carried out. The effects of its width on the residual heat removal, on the air temperature and on the air velocity during nuclear accident were obtained and discussed. The conclusion is important for PCCS flow and heat removal analysis.

View full abstract

The Advanced PWR reactor class was designed by Mitsubishi Heavy Industries (MHI) through an evolutionary approach involving steady improvements implemented through the construction of 24 nuclear power plants over a 45 year period. Conceptually based on the Advanced PWR, the EU-APWR is a reactor intended for the European market whose design was adapted to achieve compliance with the European Utility Requirements (EUR) and the Finnish regulatory guides on nuclear safety (YVL). This paper describes how the EU-APWR design evolved to be EUR compliant, as certified in 2014.

View full abstract

Many photographic studies have been conducted so far to measure the number density of active nucleation sites in subcooled flow boiling. In these studies, however, overlapping of bubbles is a serious problem in counting the number of nucleation sites accurately particularly under high void fraction condition. In the present work, we used a transparent glass heater to mitigate the influence of bubbles' overlapping in the video images and to observe bubble nucleation process in water subcooled flow boiling in detail. Using the movie data obtained in this work, important parameters in subcooled flow boiling such as the nucleation site density, bubble departure diameter, and bubble release frequency were measured. It is emphasized that the bubble departure diameter and bubble release frequency could be measured individually for all the active nucleation sites. It was found that the bubble release frequency tended to decrease with an increase in the bubble departure diameter but the bubble release frequency was exceptionally low at some nucleation sites; the number of these sites increased rapidly when the wall superheat exceeded the critical value. Based on these experimental results, active nucleation sites were divided into the two categories to investigate the appropriate functional form of correlations. The experimental data accumulated in this work were also used to evaluate the predictive performance of existing correlations.

View full abstract

Experiments were carried out to explore the mechanisms of bubble lift-off from a vertical heated surface in subcooled pool boiling. The experiments were conducted at atmospheric pressure and distilled water was used as the test fluid. A high speed camera was used to observe bubble behavior. The main experimental parameters were the static contact angle of the heated surface, the liquid subcooling, and the wall heat flux. In this study, most bubbles were lifted off the heated surface immediately after nucleation to be collapsed in subcooled liquid. It was found that the bubble lift-off velocity in the perpendicular direction to the wall is in a strong correlation with the condensation rate of a bubble after the lift-off. A preliminary correlation was developed also for the bubble liftoff diameter considering the heat transfer around the bubble.

View full abstract

Space is one of the best places for humanity to turn to keep learning and exploiting. A Nuclear Thermal Rocket (NTR) is a viable and more efficient option for human space exploration than the existing Chemical Rockets (CRs) which are highly inefficient for long-term manned missions such as to Mars and its satellites. NERVA derived NTR engines have been studied for the human missions as a mainstream in the United States of America (USA). Actually, the NERVA technology has already been developed and successfully tested since 1950s. The state-of-the-art technology is based on a Hydrogen gas (H_2) cooled high temperature reactor with solid core utilizing High-Enriched Uranium (HEU) fuel to reduce heavy metal mass and to use fast or epithermal neutron spectrums enabling simple core designs. However, even though the NTR designs utilizing HEU is the best option in terms of rocket performance, they inevitably provoke nuclear proliferation obstacles on all Research and Development (R&D) activities by civilians and non-nuclear weapon states, and its eventual commercialization. To surmount the security issue to use HEU fuel for a NTR, a concept of the innovative NTR engine, Korea Advanced NUclear Thermal Engine Rocket utilizing Low-Enriched Uranium fuel (KANUTER-LEU) is presented in this paper. The design goal of KANUTER-LEU is to make use of a LEU fuel for its compact reactor, but does not sacrifice the rocket performance relative to the traditional NTRs utilizing HEU. KANUTER-LEU mainly consists of a fission reactor utilizing H_2 propellant, a propulsion system and an optional Electricity Generation System as a bimodal engine. To implement LEU fuel for the reactor, the innovative engine adopts W-UO_2 CERMET fuel to drastically increase uranium density and thermal neutron spectrum to improve neutron economy in the core. The moderator and structural material selections also consider neutronic and thermo-physical characteristics to reduce non-fission neutron loss and reactor weight. The geometry design of fuel element and reactor focuses on protective cooling capability on its fuel and moderator, fabricability and compactness. In the preliminary design study, KANUTER-LEU shows comparable characteristics of a competent efficiency, and a compact and lightweight system despite the heavier LEU fuel utilization. The reference performance is estimated at 40.4 〜 50.4 kN thrust, 4.17 〜 5.26 thrust to weight ratio (T/W) and 904 〜 907 s specific impulse depending on the reactor powers of 200 〜 250 MW_<th>.

View full abstract

Recently, risk management for nuclear facilities becomes more complex due to security issue addressed by IAEA. The harmonization between safety, safeguards and security is still questionable. It also challenges to nuclear engineering curriculum in the world how to appropriately lecture the new issue. This paper would like to describe how to integrate this issue in developing nuclear engineering curriculum in Indonesia. Indonesia has still no nuclear power plant, but there are 3 research reactors laid in Indonesia. As addition, there are several hospitals and industries utilizing radioisotopes in their activities. The knowledge about nuclear security of their staffs is also not enough for handling radioactive material furthermore the security officers. Universitas Gadjah Mada (UGM) is the only university in Indonesia offering nuclear engineering program, as consequently the university should actively play the role in overcoming this issue not only in Indonesia, but also in Southeast Asia. In the other hand, students has to have proper knowledge in order to compete in the global nuclear industry. After visited several universities in USA and participated in INSEN meeting, we found that most of universities in the world anticipate this issue by giving the student courses related to policy (non-technical) study based on IAEA NSS 12. In the other hand, the rest just make nuclear security as a case study on their class. Furthermore, almost all of programs are graduate level. UGM decided to enhance several present related undergraduate courses with security topics as first step to develop the awareness of student to nuclear security. The next (curriculum 2016) is to integrate security topics into the entire of curriculum including designing a nuclear security elective course for undergraduate level. The first trial has successfully improved the student knowledge and awareness on nuclear security.

View full abstract

The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor (HTGR) constructed in Japan, firstly. The operating data of the HTTR with burn-up is very important for developments of HTGRs. Many test data have been collected in the HTTR. Many tests are carried out in low power operation. On the other hand, the full power operation is not enough. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate the calculation code. Additionally, it is difficult to measure core temperature in HTTR. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle and the temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for confirm the characteristics of burn-up and validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up.

View full abstract

The European Utility Requirements organization started the compliance assessment process of the Mitsubishi Heavy Industries EU-APWR Standard Design in 2012. The EU-APWR is an Advanced PWR, 1700 MWe class, 4-loops, 14ft active core fuel length that MHI has developed for the European market. The EU-APWR is an evolution of the Advance PWR currently under the licensing process in Japan for the Tsuruga Power Station. MHI has modified the design applying improvements in safety and economy in order to be adapted for European markets and to comply with the EUR requirements. The EU-APWR Standard Design documentation has been assessed against the EUR Volume 2 - Generic Nuclear Island requirements Revision D, issued in October 2012. The assessment is divided into 20 chapters for a total of over four thousand individual requirements. Each chapter was assigned to Assessment Performers who executed the detailed analysis of the requirements. The assessment of each requirement and the Synthesis Report have been submitted to, and scrutinized by, the Coordination Group, formed by representatives of the European Utilities together with the Vendor, and reviewed by the Administration Group and by the Steering Committee. The Synthesis Reports have been collected in the Volume 3 EU-APWR Standard Design Subset and presented to the Steering Committee, which approved the final draft in October 2014. The overall results of the assessment indicated good compliance of the EU-APWR Standard Design: 77% of the requirements resulted in compliance with EUR. This percentage increases to 85% when taking into account the requirements for which the design has been evaluated in compliance with the objectives. The requirements where the design has been judged not in compliance with EUR are less than 2%. The divergences between the EU-APWR Standard Design and the EUR concern different areas like, for instance, layout, operational capability and performance, outage operations, personal protection and radiation monitoring. Some of the incongruences result from differences in approach to the design process or from differences in the rules and standards in use in Japan and in Europe. Some analyses, like the internal hazards effects, have been performed only partially because, in Japan, such analyses are considered site-specific and are carried out at the detailed design level. The analysis of the consequences of a hydrogen explosion, and the environmental qualification methodology of equipment have not been fully developed yet. While the reactor core has been designed for an operability cycle of 24 months and can be loaded with 50% MOX fuel, no other area of the plant has been designed taking into consideration MOX fuel.

View full abstract

The passive containment cooling system (PCS) is used to remove the heat out of the containment after accident in XUDAPU NPP. When PCS and containment spray system are not available after severe accident, the pressure of containment will raised to high level and the hydrogen igniters will not able to burn the hydrogen in the containment. Venting the containment through pipe link to the spent fuel storage pool is to be considered in this situation according to SAMG. The hydrogen in the containment will release to the spent fuel pool located in fuel building when venting the containment is performed, and there will be risk of hydrogen burning or explosion in the fuel building. This paper analyzed the process and phenomenon when the strategy of containment venting the containment is performed, and the hydrogen risk in the fuel building is evaluated. The results shows that there is risk of hydrogen burn in the fuel building when there is not any mitigate measure. XUDAPU NPP has added some hydrogen recombiners in the containment and the hydrogen risk has been eliminated.

View full abstract

For more than twenty years now, the European Utility Requirements (EUR) organization has been actively developing and promoting harmonized technical specifications for the new designs to be proposed by the vendors in Europe. The EUR Document consists of a comprehensive set of requirements covering the whole Nuclear Power Plant (NPP). These terms of reference can be used by the utilities (guide for design assessment, technical reference for call for bids) and by the vendors (as a technical guide). The harmonization and standardization which is sought after by the EUR aims at delivering the safest and most competitive designs based on common rules shared all over Europe. Fifteen major nuclear operators across Europe are now members of the Organization. Over the last few years, the EUR organization has been extremely active. After the publication of the Revision D of the EUR Document (October 2012), the EUR organization released in March 2013 its roadmap for the period. This paper describes the main results obtained during that period of time and the new challenges in the three following fields. First, the revision of the EUR Document in order to maintain it at a state-of-the-art level remains the highest priority for the Organization. The paper presents the technical scope of the on-going new major revision (Revision E) which is scheduled to be issued in 2016. This project will deliver significant updates of the EUR document in many fields among which: revised Safety requirements taking into account lessons learned from the Fukushima accident and consistent with the most recent international safety standards, Instrumentation & Control, Seismic Approach, Probabilistic Safety Assessments. The assessment of new designs is the second main technical activity of the EUR organization. The MHI EU-APWR design has been assessed against the revision D between 2012 and 2014 and new design assessments applications have been received by the EUR organization (namely KEPCO's EU-APR1400 and AEP's VVER TOI). The paper briefly recalls the EUR design assessment objectives and process and the progress of the different assessment projects. The third topic to be covered by this paper is the interaction between the EUR and the other stakeholders, in particular the other international organizations (ENISS, WNA/CORDEL, WENRA, IAEA, EPRI). The paper describes how the EUR organization is connected to these stakeholders and the corresponding cooperation results and future projects.

View full abstract

Containment external water film flow behavior is important for traditional chemical cooling and passive containment cooling system of the third generation pressurized water reactor nuclear power plant. Fully understanding of the water-falling behavior along a flat wall with the counter-current air can provide technical supports in industrial design and nuclear safety applications. Water film thickness, transient wave features and water film average velocity with counter-current air flow were measured and obtained by using capacitance probes and two cameras for different relative humidity and temperature. The influence of the counter-current air velocity on the water film behavior was studied on a flat plate. Experimental studies also considered the test plate angle and its effects on the water film thickness. Results show that the effect of air velocity on the water film behavior with counter-current air flow is different from the steady air experiments and there is the critical velocity. When the air velocity is lower than the critical points the water film behavior is similar as the non-air velocity condition. While, for large air velocity the effects is significant. The data was well compared with the previous correlations.

View full abstract

The objective of structural analysis and seismic response analysis are well recognized and utilized as one of sophisticated analysis tools for design objects in the nuclear engineering. The way to design nuclear facilities is always in compromising with many index, such as costs, performance, robustness and so on, but the most important issues is the safety. It is true the structural analysis and seismic response analysis plays an important role to insure the safety, since it is well known that Japan is always facing to the earthquake. In this paper, a numerical analysis's controlling and managing system is implemented on a supercomputer, which controls the modelling process and data treating for structural robustness, although a numerical analysis's manager only controls a structural analysis by finite element method. The modeling process is described by the list of function ID and its procedures in a data base. The analytical modeling manager executes the process by order of the lists for simulation procedures. The manager controls the intention of an analysis by changing the analytical process one to another. Modeling process was experimentally found that may subject to the intention of designing index. The numerical experiments were carried out with static analyses and dynamic analyses. The results of its experiment introduce reasonable discussion to understand the accuracy of simulation. In the numerical experiments, K, supercomputer is utilized by using parallel computing resource with the controlling and managing system. The structural analysis and seismic response analysis are done by the FIEST, finite element analysis for the structure of an assembly, which carries out the simulation by gathering components. As components are attached to one another to build an assembly, and, therefore, the interactions between the components due to differences in material properties and their connection conditions considerably affect the behavior of an assembly. FIESTA is a structural analysis's code, which concerns interaction among components. This research and development is partially supported by HPCI strategic program of MEXT, Ministry of Education, Culture, Sports, & Science & Technology in Japan.

View full abstract

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide many advantages. However, digital control systems induce new failure modes that differ from those of analog control systems. While the cost effectiveness and flexibility of software is widely recognized, it is very difficult to achieve and prove high levels of dependability and safety assurance for the functions performed by process control software, due to the very flexibility and potential complexity of the software itself. Software safety analysis (SSA) was one way to improve the software safety by identify the system hazards caused by software failure. This paper describes the application of a software fault tree analysis (SFTA) at the software design phase. At first, we evaluate all the software modules of the reactor power regulating system in nuclear power plant and identify various hazards. The SFTA was applied to some critical modules selected from the previous step. At last, we get some new hazards that had not been identified in the prior processes of the document evaluation which were helpful for our design.

View full abstract

The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor. There are 32 control rods (16 pairs) in the HTTR. The 6 pairs of control rods are inserted into a core region and the others are inserted in a reflector region surrounding the core. The core temperature of the HTTR is too high to insert all control rods simultaneously at reactor scram near full power operation for keeping integrity of control rods metallic sleeve. Therefore, a two-step control rods insertion method for reactor scram is adopted. The reactivity inserted at the two-step control rod insertion method was measured at HTTR criticality tests. The calculated reactivity at the firststep showed larger underestimation than that of the secondstep. On the other hand, calculated results of excess reactivity at the HTTR criticality tests showed good agree with tests. It is considered that a cell model for reflector region control rod is not suitable. Therefore, this paper focuses on a new cell model for control rods in a reflector region. In a previous control rod cell model, control rod is surrounded by fuel blocks only. The surrounding condition of the new cell model corresponds to the configuration around the reflector region control rod. The calculated reactivity at the first-step using the new cell model shows better results than previous calculation. It is considered that the new cell model brings appropriate neutron flux distribution around control rods in reflector region.

View full abstract

The phenomenon and mechanism of Fuel Coolant Interaction (FCI) has been widely studied around the world in the past few decades. A series of experiments were performed and several FCI models were developed based on these experiments. However, there are still large uncertainties in the models of FCI and limitations to predict FCI process, especially the reactor scale process. A new FCI experimental facility is designed and further experiments are performed by Shanghai Jiao Tong University, China. The photography of FCI process are obtained by two high-speed cameras, recording from two different directions vertical to each other. Water level changed can also be get from images of FCI process. Pressure peak produced by intense interaction can also be recorded. To discuss the influence of different factors for FCI, numbers of variables are considered in these experiments, including jet material, melt temperature, coolant type, coolant subcooled temperature, release height, break size and interaction pool size. A series of tests in low melt temperature (400℃ to 1600℃) has been performed and a batch of results are obtained. Primary analysis is carried out based on these latest experimental results. The phenomenon of quenching and steam explosion are shown in this paper. Influence of melt temperature and material are discussed. Further and deeper study about kinds of influence factors to FCI needed to be carried out later.

View full abstract

The accident at TEPCO's Fukushima Daiichi nuclear power plant was expanded into severe accident (reactor damage and hydrogen explosion). The instrumentation systems for the severe accident of the nuclear plant are being developed1). In order to prevent the hydrogen explosion, the hydrogen concentration sensor is being developed. Although almost sensor could not be adopted by the severe accident's environment, the solid oxide electrolyte type hydrogen sensor (proton conductor) was selected for its durability to high temperature and high pressure. In the fabrication process of the sensor, the joining techniques for metal part and solid electrolyte (ceramic) is very important. After heat cycle test and pressure cycle test for joints, the test results confirmed the soundness of the ceramic parts and joints. By the hydrogen measurement in inert gas and air, it shows the theoretically predicted behavior (Nernst's Equation) in inert gas, and linear relationship between electromotive force (emf) and hydrogen concentration (logarithm) in air. It was able to confirm the feasibility of this method. Furthermore, the operating characteristics of the sensor under irradiation shows data on irradiation effect. In this paper, the performance test results of the hydrogen sensor under various conditions are presented.

View full abstract

The 1D and 3D coupled simulation technical method is put forward in this paper to overcome the analysis difficulties of the flow field in the steam manifold system. The method is verified via coupled analysis for the steam manifold system of a PWR Nuclear Power Plant. According to the simulation results of steam manifold system with different structuring forms, the flow and pressure distribution properties were compared and analyzed, which guided the optimization research of the steam manifold system.

View full abstract

Taking the CSR1000 reactor as the research object, the SCAC-CSR1000 transient safety analysis code was made. The safety analysis of CSR1000 was done. The results showed that there is a flow reversal phenomenon in the reactor coolant pump seizure accident. When the inlet flow decreased gradually, the flow distribution of upper plenum and the 1st coolant channel will reduce, that of the lower dome and both the 1st and the 2nd moderator channel will increase gradually. The 1st coolant channel will have the flow reversal phenomenon most likely. The 1st and the 2nd moderator channel require a large orifice mode, the lower dome and the 1st coolant channel require a smaller orifice mode.

View full abstract

Large eddy simulation(LES) of the swirling flow in the rod bundle subchannels with spacer grids are presented. According to the rod bundle flow benchmark experiment, the 5×5 rod bundle with a split-type spacer grid is simulated by this transient simulation with a fine hybrid mesh. Mean velocity components, RMS (root mean square) velocity components and turbulent kinetic energy profiles at different locations downstream the mixing vanes are compared between the experiment and simulation results. LES results can reproduce the mixing turbulent flow velocity field well. We also have studied the influence of the mixing vanes' bending degree on the subchannel flow structures by using LES to simulate the 3×3 rod bundle with the typical PWR spacer grid. The mixing vanes' bending degree is modified to 35°and 40°based on the original spacer grid, whose mixing vanes' bending degree is 30°. The larger bending degree produces a stronger mean coolant mixing effect, while larger pressure loss and fluctuation come together with this effect. The increased pressure loss would generate additional flow resistance, and larger fluctuation of coolant flow would make the mixing effect unstable, as well as would cause flow-induced vibration and grid-to-rod fretting phenomena. The LES method is a viable tool to optimize the spacer grid design and provide transient flow field data for mechanical analysis.

View full abstract

Changes in the average brightness, non-uniformity and color channel intensity (include R, G and B) of dark output images and natural light image is analyzed. These parameters was captured by the running Complementary Metal Oxide Semiconductor (CMOS) image sensors which exposed in low dose rate environment at 1000Gy/h, 5Gy/h, 15Gy/h and 30Gy/h. Average brightness and non-uniformity increase sharply with an increasing dose and reaches to an equilibrium. The steady region reaches faster with the dose rate increase. Hot pixels appears due to excess radiation generated electrons move into the potential well of the photodiode and become signal current. Color cast is also induced by γ-ray irradiation due to the surface recombination velocity change and the color centers introduced in pixel material. CIS (CMOS Image Sensors) samples destroyed when expose to high dose rate at 1000Gy/h immediately, but still function when exposed to low dose rate of γ-ray irradiation at 5Gy/h, 15Gy/h and 30Gy/h. A possible experiment is presented.

View full abstract

Austenitic steel is a candidate material for Supercritical Water-Cooled Reactor. This study is to investigate stress corrosion cracking behavior of HR3C under effect of supercritical water chemistry. A transition phenomenon of water parameters was monitored during a pseudo critical region by water quality experiments at 650 °C and 30 MPa.The stress-strain curves and fracture time of HR3C were obtained by slow strain rate tensile tests in supercritical water at 620 °C and 25 MPa. The concentration of dissolved oxygen was 200-1000 μg/kg and the strain rate was 7.5×10^<-7>/s. Recent results showed the failure mode was dominated by intergranular brittle fracture. The relations of oxygen concentration and fracture time were nonlinear. 200-500 μg/kg of oxygen accelerated the cracking but a longer fracture time was measured when oxygen concentration was increased to 1000 μg/kg. Chromium depletion occurred in the oxide layer at the tip of cracks. Grain size increased and chain precipitated phases was observed in the fractured specimens. These characteristics were considered to be contributive to the intergranular stress corrosion cracking.

View full abstract

If a Loss-of-Coolant-Accident (LOCA) inside the reactor building were to occur, it could generate debris. Debris that could accumulate on the sump screens would increase head loss across the resulting debris bed and sump screen. Furthermore, materials within containment could dissolve or corrode, and result in precipitation or corrosion products which could add to the debris load and further increase head loss across the debris bed and sump screen. This head loss might be sufficiently large such that it might exceed the net positive suction head (NPSH) margin of the RIS and EAS pumps, which would challenge the safety function of RIS and EAS system.This paper describes the impact of chemical effect on sump strainer head loss in a power plant in China. The test fluid used representative plant coolant chemistry, and the test fluid temperature was controlled to simulate the post-LOCA plant coolant temperature profile. Water samples were taken periodically during the test and then analyzed for elemental concentrations of the Al, Si, Na, Ca. Following the tests, we can conclude that no visible precipitation was noted during the test. The aluminium coupon was corroded, but did not form precipitation in the solutions during the test. The available Ca ions in TSP environment were potentially related to the Ca_3(PO_4)_2 precipitation. Overall, Ca ICP levels were reducing over time which may indicate that the Ca was used to form precipitate. Although the chemical reaction occurred and precipitation formed during the test, it didn`t have remarkable impact on the sump strainer head loss.

View full abstract

The U.S. Nuclear Regulatory Commission (NRC) requires that each application for a standard design certification be referred to the Advisory Committee on Reactor Safeguards (ACRS) for a review and report on those portions of the application which concern safety. This paper begins with perspectives on the role of the ACRS in the design certification review process. It then summarizes the ACRS observations and recommendations made in the Committee's reports during the General Electric Nuclear Energy (GENE) U. S. Advanced Boiling-Water Reactor (ABWR) design certification reviews.

View full abstract

A customized Computational Fluid Dynamics (CFD) model of supercritical water (SCW) heat transfer in a vertical tube upward flow is developed and partially validated using experimental data obtained under the operating conditions typical for SCW cooled reactors (SCWRs): at a pressure of 24 MPa, an inner tube diameter of 10 mm, an inlet temperature of 320 or 350 °C and a heated tube length of 4 m. The three values of mass flux (500, 1000 and 1500 kg/m^2s) and various values of wall heat flux (from 141 to 729 kW/m^2) are considered. Physical properties of SCW are calculated by using the REFPROP software from National Institute of Standards and Technology (NIST). The model has been incorporated into the commercial general-purpose CFD software, PHOENICS. Various turbulence models and numerical grid settings are tested. The study has demonstrated a good agreement between the CFD predictions and the experimental data on the inside tube wall temperature and heat transfer coefficient with use of a two-layer low-Reynolds-number k-ε turbulence model. However, a further model development is required under the conditions of significant effects of buoyancy force on heat transfer characteristics (the conditions of low values of mass flux and high values of wall heat flux). Practical recommendations are made regarding potential model applications in 3D analyses of SCWRs.

View full abstract

The Japan Atomic Energy Agency (JAEA) has been developing the centrifugal contactor for spent fuel reprocessing. In this study, we investigated the sludge behavior in centrifugal contactors at three different scales. The operational conditions (the flow rate and rotor speed) were varied. Most insoluble particles such as sludge remained in the rotor via centrifugal force. The capture ratio of sludge in the contactor was measured as a function of particle size at various flow rates, rotor speeds, and contactor scales. The sludge adhered and accumulated inside the rotor as the operational time increased, and the operational conditions influenced the capture ratio of the sludge; a lower flow rate and higher rotor speed increased the capture ratio. The results confirmed that Stokes' law can be applied to estimate the experimental result on the behavior of the capture ratio for centrifugal contactors with different scales.

View full abstract

Temper bead welding is applied as one of repair methods for reactor vessels made of low alloy steel. It is desirable that hardness of the weld heat affected zone in the temper bead weld is less than 350HV to prevent the cold cracking. As diffusible hydrogen in a factor of cold cracking hardly diffuses in austenitic weld metal cold, cracking is available to be prevented. Cold cracking and the mechanical properties of the weld by gas tungsten arc welding using welding material of Ni-based alloy was evaluated. Even if maximum hardness in the weld heat affected zone is more than 400HV, cold cracking don't occur, and also the mechanical properties are good.

View full abstract

The flow instability has been extensively investigated in the past as it can lead to a premature critical heat flux (CHF) at the heat flux level much lower than that for stable conditions, but the relationship between the flow instability and CHF is still unclear. In view of this, an analysis model of CHF under flow instability in forced circulation has been proposed in this paper, which adopts the classical instability theory analysis of the pressure difference resulting from a small disturbance perpendicular to the liquid-vapor interface, and the flow instability characterized by periodic mass flow fluctuation is hypothesized to exert upon the small disturbance, which results in the change of interfacial curvature. When the heat flux closes to the CHF, the intense vapor production will lift the liquid-vapor interface away from the heater surface, preventing the supply of liquid from contacting the heater surface. Soon after, the CHF occurs. The results of the model indicate that the amplitude and period of the flow instability significantly influence the CHF, which is in the form of q_<CHF>〜q_0[1-0.43(ΔG)/(G_0)^<0.427>(τ/T)^<0.263>], i.e. the CHF under flow oscillation decreases with increasing amplitude and period of flow instability, and the model also can satisfactorily predict the experimental data available in literature.

View full abstract

In this paper, evaluation method is studied for condition that the large damping forces of viscous damper act on structure locally. The response spectrum analysis method by complex modal analysis is proposed as a simplified method to evaluate seismic responses of structures with viscous dampers. A steam generator is selected as an example of structure. Seismic response accelerations by proposed method are evaluated. The evaluation results are in good agreement with those of direct integration method in the practical range of damping coefficient of viscous dampers 1〜100MNs/m. From this fact, it is confirmed that the proposed method is effective to evaluate the seismic responses of structures with large capacity viscous dampers.

View full abstract

It is important for utilities to achieve a stable operation in nuclear power plants. To achieve it, plant anomalies that affect a stable operation must be found out and eliminated. Therefore, the advanced condition monitoring program was developed. In this program, a sophisticated heat balance model based on the actual plant data is adopted to identify plant anomalies at an incipient stage and the symptoms of plant anomalies are found by heat balance changes from the model calculation. The model calculation results have shown precise prediction for actual plant parameters. Moreover, this program has the diagnostic engine that helps operators derive the cause of plant anomalies. By using this monitoring program, the component reliability in the turbine system can be periodically monitored and assessed, and as a result the stable operation of nuclear power plants can be achieved.

View full abstract

Field Programmable Gate Arrays (FPGAs) have recently garnered significant interest for certain applications within the nuclear field. Some applications of these devices include Instrumentation and Control (I&C) systems, pulse measurement systems, particle detectors and health physics purposes. In CANada Deuterium Uranium (CANDU) nuclear power plants, the use of heavy water (D2O) as the moderator leads to the increased production of Tritium, which poses a health risk and must be monitored by Tritium-In-Air Monitors (TAMs). Traditional TAMs are mostly designed using microprocessors. More recent studies show that FPGAs could be a potential alternative to implement the electronic logic used in radiation detectors, such as the TAM, more effectively. In this paper, an FPGA-based TAM is designed and constructed in a laboratory setting using an FPGA-based cRIO system. New functionalities, such as the detection of Carbon-14 and the addition of noble gas compensation are incorporated into a new FPGA-based TAM. Additionally, all of the standard functions included in the original microprocessor-based TAM, such as tritium detection, gamma compensation, pump and air flow control, and background and thermal drift corrections were also implemented. The effectiveness of the new design is demonstrated through simulations as well as laboratory testing on the prototype system.

View full abstract

In this paper, we study the technique for monitoring the incore neutron flux in an accelerator driven sub-critical (ADS) system, where a heavy metal spallation target located vertically at the centre of a sub-critical core is bombarded vertically by the high-energy protons from an accelerator. In an ADS system, the incore neutron flux is affected dramatically by the neutrons from the spallation target. We propose a multi-point measurement technique as follows. During the operation of the reactor, the incore neutron flux should be measured at multiple vertical locations. The detectors for the measurements of incore neutron flux may either be left in a fixed location or provided with a motorized drive to allow move vertically within the reactor core. To evaluate the proposed technique, we have studied the neutron production from spallation targets with the Geant4-based Monte Carlo simulations. In the simulations, two targets, lead and bismuth, have a cylindrical shape with the radius of 10 cm and various lengths. The proton beam with the energy of 250 MeV vertically impinges on the top of the cylindrical targets. The neutron detector with a length of 10 cm moves vertically from the top to the bottom within the reactor core to measure the incore neutrons at seven locations. The simulation results have indicated that the neutron flux at the central location is more than two orders of magnitude higher than the flux at the lower locations. Therefore, during the reactor startup, as the neutrons from the spallation target dominate, the incore neutron detectors should be put at the central location which is close to the target for the commissioning measurements of the proton accelerator.

View full abstract

Spider, the key component of the AFA3G cluster control assemblies (RCCA), is brazed with body, vanes and fingers. Vacuum brazing is crucial in the spider process and it is directly relevant to the final product quality. This paper analyze the deformation of the AFA3G spider in vacuum brazing procedure based on a large amount of data. The results indicate that the parallelism of the finger is most affected by the brazing and its deformation has obvious regularity. Deformation is mainly caused by the different contraction directions of components along with the interactions among them during cooling process. An optimized design of the brazing fixture based on the regularity and the value of the deformation greatly improves the parallelism of the fingers. Besides, the vacuum brazing procedure also affects the hole diameter of the finger, however, we could reduce the deformation by using columnar pin on the brazing fixture.

View full abstract

Field programmable gate arrays (FPGAs) are gaining increased attention worldwide for application in the instrumentation and control systems for nuclear power plants. In this paper, we study potential application of FPGAs in the radiation environment of an accelerator driven sub-critical (ADS) system, where a spallation target located at the centre of a sub-critical core is bombarded by the high-energy protons from an accelerator. In comparison with a commercial reactor used in nuclear industry, more control electronics are required in the reactor building of an ADS system to exactly couple the high-energy beam from the accelerator to the spallation target in the reactor core. By estimating the neutron flux in the reactor building of China initiative ADS system which should be able to demonstrate the ADS concept at 10 MW power level, we observe that the neutron-induced displacement damage and single event effect are serious. Thus, we suggest to shield these semiconductor devices with shielding materials, such as polyethylene or concrete. Neutron shielding simulations by using GEANT4 with G4 neutron data library (G4NDL) have indicated that the polyethylene material with a thickness of 30 cm can reduce the neutron dose up to three order of magnitudes and thus reduce dramatically the neutron-induced radiation effects. By assessing the displacement damages and soft errors caused by the neutrons with various energies, we observe that the most effective neutrons in creating radiation damages are those fast neutrons with the energy of more than 0.1 MeV. Therefore, we suggest to add heavy metals in the shielding material to further shield these fast neutrons.

View full abstract

The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident (SA parameters) and are developing the Severe Accident Instrumentations and Monitoring Systems (SA-keisou) that could measure the parameters in severe accident conditions. For an example, reactor water level was selected as one of the SA parameters. The reactor water level is considered to be one of the most important parameters for reactor operators to understand the status of the reactor-core fuel. A new reactor water-level instrument was developed that complements the existing differential pressure method. The instrument is essentially a simple gamma thermometer which consists of several differential thermocouples and a heater. The gamma thermometer had been developed to monitor the in-core gamma-ray distribution, can be installed in the in-core housing, and measures the reactor water level by using the difference of the thermal conductivity of water and the air. This water-level instrument can measure the water-level condition of cooling reactor-core fuel. The instrument was manufactured, was verified at a testing facility, and the feasibility was confirmed. Another selected SA parameter was the hydrogen concentration in the primary containment vessel (PCV). An instrument that measures electric resistance of palladium was developed to monitor hydrogen concentration. In principle, the resistance of palladium increases with hydrogen absorption. The resistance of palladium is almost independent of vapor, whereas the temperature shift of resistance requires to be corrected. Platinum resistance was used for the correction: the hydrogen sensor consisted of palladium and platinum wires and is assembled in double spirals. This instrument was tested in our facility where its specification was confirmed. Also, the PCV water-level instrument and the Containment Atmospheric Monitoring System (ion chamber) were developed for the severe accident.

View full abstract

The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident (SA parameters) and are developing the Severe Accident Instrumentations and Monitoring Systems (SA-keisou) that could measure the parameters in severe accident conditions. Hitachi-GE has also been developing some instrumentation systems for use under severe environmental conditions during a severe accident including a fiber optic radiation monitor and a fiber optic water level monitoring system. The fiber optic radiation monitor uses a neodymium doped radiation luminescence element which emits infrared wavelength light when irradiated by gamma-rays. We confirmed that the monitor has a capability for measuring dose rates from 10mGy/h to 100kGy/h. The fiber optic water level monitoring system consists of a pressure sensor which is set on the tip of the optical fiber and a fiber optic temperature sensor. We are presently confirming that it can measure the water level properly under severe accident conditions.

View full abstract

Under postulated accident conditions in a pressurized water reactor (PWR), steam and condensate water form countercurrent gas-liquid flows in a hot leg and a pressurizer surge line, so that countercurrent flow limitation (CCFL) may occur. There are many studies for CCFL in hot leg models, but there are only a few studies for CCFL in a pressurizer surge line (consisting of a vertical pipe, a vertical elbow, and a slightly inclined pipe with elbows). In our previous studies, we measured CCFL characteristics in a 1/10-scale model of a pressurizer surge line using air and water, developed a one-dimensional (1D) computation model, and also did three-dimensional (3D) simulations for the inclination angle of 0.6 deg (slope of 1/100) to validate simulation capability. 1D computations and 3D simulations gave good agreement with the 1/10-scale air-water data for the inclination angle of 0.6 deg. In the present study, we did 1D computations and 3D simulations for air-water countercurrent flows in the 1/10-scale model of the pressurizer surge line to validate them for effects of inclination angles on CCFL. Although 1D computations and 3D simulations gave good agreement with measured data for the inclination angle of 0.6 deg, they slightly underestimated effects of inclination angles on CCFL for the inclination angles of 0 deg and 1.0 deg.

View full abstract

The prevention of excessive deformation by thermal ratcheting is important in the design of high-temperature components of fast breeder reactors (FBR). This includes evaluation methods for a new type of thermal ratcheting caused by an axial traveling of temperature distribution, which corresponds to moving-up of liquid sodium surface in startup phase. Long range traveling of the axial temperature distribution brings flat plastic deformation in wide range. Therefore, at the center of this range, residual stress that brings shakedown behavior does not accumulate. As a result, repeating of this temperature traveling brings continuous accumulation of the plastic strain, even if there is no primary stress. In contrast, in the case with short range traveling, residual stress is caused by constraint against elastic part, and finally it results in shakedown. Because of this mechanism, limit for the shakedown behavior depends on distance from the elastic part (i.e. half length of region with plastic deformation). Igari et al. proposed a mechanism-based evaluation method that focuses the traveling range of the temperature distribution. In this method, temperature difference was assumed to constant in the traveling phase, in other words, the temperature distribution moves subsequently to temperature rise. According to this assumption, the traveling range is equal to the plastic deformation range. However, in the actual design of the fast reactor vessel nearby liquid sodium surface, the temperature distribution moves up synchronizing with the temperature rise, without any intentional control. Because the moving up of the liquid sodium surface results from the heat expansion of the liquid sodium the assumption to isolate the temperature increase rise and traveling may be too conservative. In the actual design, the plastic deformation range becomes smaller than the traveling range of the coolant level. In this paper, we examined characteristics of the accumulation of the plastic strain caused by realistic heat transients, namely, traveling of temperature distribution synchronizing with temperature rise. This examination was based on finite element analyses using elastic-perfectly plastic material. As a result, we confirmed that the shakedown limit depends on not the traveling range of the temperature distribution but the plastic deformation range, which was predicted by the elastic analysis. We can control the plastic deformation range by changing rate of the moving-up of liquid sodium surface.

View full abstract