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When total volume control is applied to toxicity in nuclear waste management it becomes a limiting factor for the permittable total operating capacity of nuclear reactors. An alternative conceptual...

To identify functions and facilities that are critical for ensuring safety in advanced reactors, for which only a limited usage history is available, it is necessary to consider the characteristics of each reactor type by utilizing more risk-related information. In this study, we compare the response capability during an accident for light water reactors and fast reactors to quantitatively evaluate the impact of the reactor type on resilience. A resilience evaluation procedure is developed for assessing the response margin and response reliability of a nuclear power plant during an accident; this procedure was then applied to a simple pressurized-water reactor and a sodium-cooled fast reactor plant model using an accident scenario involving an earthquake that generated a leakage of coolant from the primary heat-transport system. Our research indicated a relatively large difference between the response reliability with respect to time for the two reactor types.

To respond to seismic and other natural hazard events, designers of the Japan Sodium-cooled Fast Reactor (JSFR), an advanced loop-type reactor, are planning to adopt a steel-plate reinforced concre...

The authors are developing the design concept of the pool-type sodium-cooled fast reactor (SFR) that addresses Japan’s specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDG) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident. The design concepts of a reactor vessel (RV) and its internal structures have been investigated whether they could withstand severe seismic conditions in Japan and thermal loads. The design adopted enhanced RV support structure, enhanced conical-shaped core support structure, a thickened knuckle part of the RV, and a flat plenum separator with ribs. A three-dimensional steady-state thermal hydraulic analysis of the RV revealed that the temperature difference of the upper and lower surfaces of the flat plenum separator could be effectively reduced by installing layers of thermal insulation plates. The authors have also conducted a transient analysis of loss of flow type anticipated transients without scram events to evaluate the feasibility of a self-actuated shutdown system. Moreover, the configuration of the decay heat removal system has been investigated considering sufficient utilization of natural circulation capability of sodium coolant, heat removal capacity of each cooling system, conformance with design requirements, and recommendations of SDC and SDG such as diversity and redundancy of components.

Measurement of the temperature and flow rate at each fuel subassembly outlet is an effective way for a liquid metal fast breeder reactor to detect a loss of coolant accident or reactivity-i...

Development of an inspection technique in opaque liquid-metal coolant is one of the important issues to ensure the safety of the liquid-metal fast breeder reactor (LMFBR). Performance tests of an u...

The Japan Sodium-cooled Fast Reactor (JSFR) adopts a double boundary for all sodium components. In this paper, design measures are investigated against a secondary sodium fire inside the reactor building, which might be assumed as design extension conditions. Candidate sodium fire measures for the secondary sodium systems compared in terms of safety are the sodium drain, nitrogen injection, pressure release valve, catch pan, and drain system for leaked sodium. Various sodium fires in the steam generator room have been analyzed by the SPHINCS code to evaluate the performance of the candidate sodium fire measures.

In this study, we develop new indices that evaluate the response capability — response margin and response reliability — of a nuclear power plant in case of occurrence of a severe accident by considering the ability of the nuclear power plant to recover system safety functions, which may have been affected by the accident, to the minimum safety values. Further, we demonstrate a specific evaluation procedure for our proposed indices. While performing the evaluation, we consider margins with respect to a minimum safety function level and a time constraint along with their temporal changes. Additionally, for a simplified fast reactor (FR) plant model, we conduct a trial assessment of the recovery capacity of the system to ensure system safety based on the concept of new indices. Further, the applicability of the response reliability evaluation method to FR plants is discussed based on the viewpoints of reflecting the characteristics of each reactor type.

Severe external hazards on the Japan Sodium-cooled Fast Reactor (JSFR) have been analyzed and evaluated. For seismic design, safety components are confirmed to maintain their functions even against...

A phased array acoustic leak detection system for sodium heated steam generator has been proposed. The major advantage of the new system is it could provide information of acoustic source direction. An acoustic source of a sodium–water reaction is supposed to be localized while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore the new system could separate the target leak source from steam generator background noise. In the previous study, the methodology was proposed and basic performance was confirmed by numerical analysis. However, in the numerical analysis, acoustic transportation through the SG tube bundle was not modeled. In the present study, performance the proposed system has been confirmed in water experiments with mockup tube bundles.

In response to the accident at the Fukushima Daiichi (1F) nuclear power plant, designers of the 2010 version of the Japan sodium-cooled fast reactor (JSFR) have been studying the robustness...

In this paper, the emergency power supply system for next generation sodium fast reactor (SFR) is investigated to target the latest system design with such as optimized equipment for measure of loss of heat removal system (LOHRS) events, taking into account for strengthening diversity and multiplicity. In addition to diversity and multiplicity, optimization in terms of amount of equipment is also taken into account. The emergency power supply system is consist of emergency generators for design based accident (DBA) and emergency generators to cope with the design extension condition (DEC) in order to power the additional cooling equipment of fuel handling system and reactor core cooling system as a measure for LOHRS event. Base on the load capacity and system operation, the configuration of emergency power supply system was examined aiming optimization of amount of equipment in accordance with some shared use taking into account the reliability. As a result, the emergency power supply system with 4 fixed-type gas turbine generators, 4 portable-type diesel generators and batteries was established.

A comparison of the Japan sodium-cooled fast reactor (JSFR) design with the future French sodium-cooled fast reactor (SFR) concept has been done based on the requirements of Electricite de France (...

JSFR is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a transfer pot with two fuel subassembly positions has been developed so as to shorten refueling period increasing plant availability. The pot is required to provide sufficient cooling capability in case of transportation malfunction. In this study, a three dimensional analysis model for heat transfer evaluation of the JSFR fuel transfer pot has been developed. The heat transfer models inside and outside the pot have been validated by reference experiments. Using the developed three-dimensional model, the JSFR fuel transfer pot has been analyzed. For a simpler design tool, a two dimensional analysis model has been developed. Comparison of the three and two dimensional analyses shows that two dimensional analyses could estimate pot cooling performance conservatively.

Japan sodium-cooled fast reactor is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a transfer pot with two fuel subassembly positions has been developed so as to shorten refueling period increasing plant availability. The pot is required to provide sufficient cooling capability in case of transportation malfunction. To evaluate cooling capacity of the transfer pot, a mockup pot has been fabricated and heat transfer experiments have been conducted on the mockup pot.

Japan sodium-cooled fast reactor (JSFR) is going to adopt an advanced fuel-handling system. From the viewpoint of spent fuel cleaning, a new dry-cleaning process instead of the conventional process with water rinse is under development. In this study, drain performance tests on the JSFR subassembly inner duct and dry-cleaning performance tests with a pin bundle model are summarized. Based on the experimental data of the inner duct and pin bundle model tests, residual sodium on the spent fuel subassembly after argon gas cleaning has been evaluated to be 400 g. Water alkalinity and purification performance have been evaluated and the JSFR water pool system has shown the capability to accept 400 g residual sodium on the spent fuel subassembly after argon gas cleaning.

The Commissariat a L’Energie Atomique et aux Energies Alternatives (CEA) and the Japan Atomic Energy Agency (JAEA) intend to develop prototype or demonstration sodium-cooled fa...

The joint paper presents a common view of fast reactor specific missions in the development of nuclear energy and a cross-analysis of merits and demerits of several Fast Reactors concepts studied worldwide and especially in the Generation-IV International Forum (GIF) framework. The paper provides the context for fast reactors development in the United States, France and Japan and focuses on the comparison on Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), and Lead-cooled Fast Reactor (LFR), i.e. the three fast reactor concepts that have the potential to meet the nuclear fuel cycle sustainability goals. The information provided in the article permits the reader to understand each country's objectives to see that not only the objectives searched for but also the technical orientations are converging. The authors underline that SFR technology evaluation relies significantly on the substantial base technology development programs within each country which is without comparison for the other two fast reactor technologies, e.g., SFR technology has already been developed to commercial or near commercial scale in each country whereas the performance of LFR and GFR technology is still uncertain. The main GFR merits are the potential for high temperatures and the easier possibilities for inspections and repairs. The main challenges are the fuel (fabrication, in-pile behavior), materials for high temperatures, and the implementation of mitigation means to manage severe core degradation. The main LFR merit is the lack of chemical reactivity of the lead coolant with air and water. The main challenges are the development of corrosion resistant structural and cladding materials, the implementation of mitigation means to manage severe core degradation, the density of the lead, and the comparably large core size. The selection of a reference fast reactor concept in view of possible industrialization is made on a national base, taking into account the each countries’ strategic objectives, existing technology base, the proven or expected technical performance, the R&D challenges and technical means to conduct that R&D, the possibility to share development costs and risks, etc. It is important to note that in different contexts, the U.S., French, and Japanese organizations have selected the SFR as their reference fast reactor concept.

The Japan Sodium-Cooled Fast Reactor (JSFR) adopts the new concept of a containment vessel called a steel-plate-reinforced concrete containment vessel (SCCV). The SCCV is considered to be effective in shortening construction periods by the elimination of rebar work at a site compared with applying a reinforced concrete CV. In addition to this advantage, the SCCV achieves high-quality building structure because steel structural parts are fabricated at a factory prior to site construction. Although the SC structure has been used for buildings at a light-water reactor (LWR), etc., the SC structure has not yet been adopted for the CV. An SFR CV has a lower pressure environment than the LWR CV, although the environmental temperature of the SFR is much higher than that of the LWR in the postulated coolant leakage accident. It is therefore important to investigate its characteristics at high temperature to adopt the SC structure to the JSFR CV because the CV keeps containment functions in accidents to be assumed in design. This paper describes the construction of the design evaluation method from design (construction period shortening) and accident management, experimental, and analytical points of view.

Key technologies for Japan Sodium-cooled Fast Reactor (JSFR) have been evaluated. The ten technologies - high-burnup core, safety enhancement, compact reactor vessel, two-loop cooling system using ...

Japan Atomic Energy Agency (JAEA) initiated “Fast Reactor Cycle Technology Development (FaCT)” project since 2005 to accomplish necessary R&Ds goals for Japan sodium-cooled fast reactor (JSFR) cycle (the loop type sodium-cooled fast reactor with MOX fuel and its related fuel cycle system). Since the concept of multi-transuranic (TRU) fuel recycling has been adopted to reduce environmental burden and to enhance proliferation resistance, decay heat and radiation of MA, etc. became major consideration in designing fresh fuel shipping cask. The use of dry cask and helium gas filling for its high thermal conductivity has been selected, since water as a coolant is not preferable in order to avoid corrosion of fuel steel by sodium–water reaction products. Hence the helium gas cask with five fuel subassemblies storage was designed. To assure the thermal integrity, the cladding temperature of the fuel assembly has to be kept below the creep temperature to prevent any creep damage before fuel loaded into the core. In this study, a whole cask analysis using three-dimensional FEM model has been carried out to confirm absence of creep damage on fuel cladding and thermal integrity complying with the IAEA safety standards. Furthermore, detailed fuel subassembly analysis using two-dimensional FEM model estimated the maximum cladding tube temperature in detail to confirm the heat transport performance. In addition, the cask filled with water is assessed by Monte Carlo method for its integrity at criticality.

The JSFR is a commercial sodium-cooled fast reactor that has been studied in the Fast Reactor Cycle Technology Development (FaCT) project since 2006. For JSFR fuel handling, various fuel-handling systems (FHSs) were investigated, and an advanced FHS with an ex-vessel storage tank (EVST) was selected. This paper summarizes the various FHS concepts and comparisons among those concepts. In the reference system, spent-fuel subassemblies are cooled in the EVST before transfer to the spent-fuel storage pool. The other FHS concepts investigated are evolutional FHSs without an EVST. The result has indicated that the construction cost of the evolutional systems does not reduce the construction cost dramatically, which is mainly due to additional safety measures that required higher decay heat handling in a gas atmosphere and separated fresh and failed fuel storage. From an economics point of view, a longer plant outage of the evolutional systems offsets its advantage of the lower construction cost. Based on the results of this comparative study, JSFR selected the FHS with an EVST.

An electromagnetic pump (EMP) has superior potential to improve the economic performance and ease of maintenance of sodium-cooled fast reactors. This study investigates the adequateness of a modular-type EMP system for large-sized (1,500MWe class) sodium-cooled fast reactors. A flow rate of over 500 m3/min is required for the main circulating pump of such reactors. There is concern that such a large EMP will cause flow instability. A modular-type EMP system can solve this issue since smaller EMPs are arranged in parallel and the flow rate of each EMP is reduced. Parallel-module EMP systems have been investigated as the primary and secondary circulating pumps. The results of the design study and electromagnetic analysis of the primary main pump confirmed that flow instability does not occur under all operational conditions. From a safety viewpoint, a reliable flow-coast-down system has been proposed, comprising an electric supply system with a permanent magnet synchronous motor and a reliable circuit break...

Comparative evaluation of fast reactor pool/loop configurations in the JAEA feasibility study (FS) from 1996 to 2006 has been re-evaluated from the new point of view. In the FS, both pool and loop configurations have been investigated. The FS loop concept (FS-loop) is an advanced two-loop design which is being developed in the Fast Reactor Cycle Technology Development (FaCT) project as Japan Sodium-cooled Fast Reactor (JSFR). In this study, a brief description of the FS pool concept (FSpool) has been provided. The FS-pool pursues a compact reactor vessel structure using a flask shape intermediate heat exchanger. The original FS pool/loop comparison in 2000 concluded that material amounts of those systems were about similar between pool and loop configurations. However, the present comparison based on the reactor vessel structure and primary cooling system pointed out relatively good economic potential of the FS-loop concept, because differences in the secondary cooling systems of the pool/loop configurations are not essential from the viewpoint of pool/loop comparison. In addition, a rough estimation method of a reactor vessel diameter against power has been proposed and the result has been compared with RV diameters of various pool/loop concepts. The discussion on the reactor vessel diameter shows that the FS-pool reactor vessel is the most compact of the recent commercial reactor designs such as SPX-2, SNR-2, EFR, CDFR and BN-1600. This study points out better economic potential of FS-loop (present JSFR) as a result of competition with the most compact pool concept.

The Japan Sodium-cooled Fast Reactor (JSFR) has adopted an in-vessel fuel handling system that consists of a single rotating plug, an upper inner structure (UIS) with a vertically penetrating slit, and a fuel handling machine (FHM) with a pantograph arm enhancing a compact reactor vessel design. Since the reactor vessel design depends on the in-vessel fuel handling system, the feasibility of the JSFR compact reactor vessel design is directly related to the feasibility of the new FHM. In this study, we have fabricated a full-scale mock-up of the JSFR FHM and performed tests in air. From the tests, the FHM mock-up shows sufficient performance in terms of positioning accuracy, motion speed, and stiffness to ensure durability for practical use in commercial plants. Structural analyses have been conducted to validate and improve the seismic analysis model and the positioning control of the FHM. The numerical results are in good agreement with the vibration and positioning tests, showing that there is a suffici...

A new acoustic leak detection system for sodium-cooled reactor steam generators using a delay-andsum beamformer is proposed. The major advantage of the delay-and-sum beamformer is that it could provide information on the acoustic source direction. An acoustic source of a sodium-water reaction is supposed to be localized, while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore, the delay-and-sum beamformer could distinguish the acoustic source of the sodium-water reaction from the steam generator background noise. In this paper, results of numerical analyses are provided to show the fundamental feasibility of the new method.

In remote areas, a small power source with a capacity less than 50MW electricity without refueling is attractive since fuel transfer cost is high. In a previous 50MW sodium-cooled reactor study, a concept with a long-life core and a simple plant system without refueling was proposed. The 50MW plant decay heat removal system adopts direct reactor auxiliary cooling systems (DRACSs). To enhance passive safety features, the 50MW plant DRACSs adopt flow diodes instead of valves which need active signals to become activated. In this study, two full-scale flow diode models, Types A and B, were manufactured and water tests were conducted. The Type A flow diode is a conventional vortex flow diode and Type B is a modified vortex flow diode which provides easy maintenance. Under the test condition, the former showed good performance while the latter lacked sufficient performance. Then, flow diodes for the 50MW plant DRACSs were designed according to the experimental results. An optimized Type A flow diode and a modi...

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed more » so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling. « less

The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues ofhandling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonablysized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled- reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the sodium coolant. The cladding temperature requirement is maintained below the creep temperature limit to avoid any damage before core installation. The thermal analysis shows that a helium gas-filled cask can accommodate ABR-1000 fresh minor actinide-bearing fuel with 700-W decay heat. The above analysis results revealed the overall requirement for minor actinide-bearing metal fuel handling. The information is thought to be helpful in the design of the ABR-1000 and future sodium-cooled-reactor fuel-handling system.

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclea...

A small modular fast reactor is thought to be one of the solutions to meet future energy security with low research and development (R&D) risk. In the present study, a new small reactor concept for a modular power source is proposed. A minimum configuration with a compact reactor vessel, one-loop main cooling system, and simple fuel-handling system is adopted, enhancing cost reduction. In the present one-loop main cooling system, there are double electromagnetic pumps in series considering pump failure. To show the reliability of the one-loop main cooling system, pipe-break transient analyses have been carried out. In addition, the construction cost of a set of a first-of-a-kind reactor and small fuel cycle plant is evaluated to show the economical potential at the demonstration stage. A major advantage of the present concept is that the demonstration reactor and fuel cycle plant can be directly appropriated for first commercial modules and the power plant can easily increase its capacity adding reactor and electrorefiner modules. Commercialization of the nuclear fuel cycle fusing the present modular concept is thought to reduce R&D risk since the total budget for demonstration is small and the facilities for demonstration are directly appropriated to commercial use.

A small reactor has the potential to be utilized as a power source to meet diverse social needs and reduce capital risks. In remote areas, populations tend to be small, and an economic power grid may not be available. In such situations, a small power source with a capacity of less than 50 MW(electric) without refueling is attractive since the costs for fuel transfer to such a site are expensive. In the present study, a metal fuel core with a lifetime of 30 yr and a simple reactor plant design has been proposed. The local burnup reactivity change in every core region is minimized by adjusting the zirconium content and the smear density of the three-core region to achieve a 550°C core outlet temperature. At the end of the cycle, the burnup reactivity is evaluated to be 1.1% of (dk/kk'), achieving a 30-yr core life. The reactor vessel is dramatically simplified by eliminating a fuel-handling system. The number of main cooling loops is reduced to one by installing dual electromagnetic pumps in the primary sodium circuit. The nuclear steam supply system mass, at 309 tonnes, shows that the present loop-type concept can more dramatically reduce material mass than that of the previous pool-type concept of 484 tonnes. The rough estimation of the electricity cost shows that this concept will be competitive for remote sites. Transient analyses show that a self-actuated shutdown system enhances the passive safety features, thus ensuring reactor integrity in anticipated transient without scram events.

Hydrogen production with a fast breeder reactor (FBR) may be attractive as a long-term energy source with nuclear fuel breeding. The thermochemical and electrolytic hybrid process is one of the hyd...

A conceptual design study of a small sized sodium cooled reactor with 165 MWe output with a metallic fuel, which aimed at the application for the diversified power supply has been carried out. A metal fuel core has been developed with 550°C core outlet temperature and 20 years core life time by utilizing the three zone core having different Zr contents in U-Pu-Zr of metal fuel. Major components in the nuclear steam supply system has been design and safety analyses has been performed to evaluate economical and safety potential of the concept. ATWS analyses show the passive safety feature of this concept adopting control rod driver-line (CRD) expansion reactivity and prolongation of electromagnetic pump (EMP) coastdown. Enhancing passive safety features, a improved upper inner structure enhancing CRD expansion and a reliable power source for EMP has been proposed. Though construction cost in first of a kind (FOAK) does not satisfy the economical goal, the present concept has potential for achieving the econ...

A small fast reactor has the potential to be utilized as a power source applicable to diversified social needs and to reduce capital risks. At remote sites where the population is small and plants ...

A thermal source for hydrogen production is an attractive utilization of nuclear energy. Hydrogen production from natural gas is a promising method in an early stage of hydrogen society, though hydrogen production with water splitting without carbon dioxide emission is the final goal. Steam methane reforming is a well-known method for producing hydrogen from natural gas. A hydrogen separation membrane makes the reforming temperature much lower than that of the equilibrium condition, and a sodium-cooled fast reactor, which supplies heat at {approx}500 deg. C, can be used as a heat source for hydrogen production.In this study, a hydrogen production plant with the membrane reforming method using a sodium-cooled reactor as a thermal source has been designed, and its economic potential is roughly evaluated. The hydrogen production cost is estimated to be about $1.67/kg, achieving the economic target of $1.7/kg. The construction cost is largely shared by the reformers' cost, and it can be decreased using a more efficient hydrogen separation membrane. This shows that steam methane reforming hydrogen production with a sodium-cooled reactor has high economical potential.

Aiming at enabling maintenance and repair of almost all components in JSFR demonstration reactor to a level equivalent to that attained for the light water reactors, we identified a number of parts which have difficulty in maintenance and repair in main components of the reactor structure and the primary/secondary main coolant system. And we defined the criteria for design improvement and then provided candidates of improvement measures for the identified parts. Furthermore, we made a modification of the plant design in a consistent manner integrating the improvements investigated for each major component. A series of evaluations were conducted to check the feasibility as a power plant. As the result, we found that the concept could be adopted not only to the demonstration reactor (750MWe) but to the commercial one (1500MWe).

Moving particle semi-implicit (MPS) method was developed for analyzing incomplessible flow with particle models. In the present study, a numerical model for thin walls is developed and the MPS method is extended to analyze fluid-structure interactions with large deformation of free surfaces and walls. First, sloshing in a rigid rectangular tank is calculated. A finite amplitude wave occurs in a shallow tank, while a standing wave appears in a deep tank. Results are in good agreement with the experiments by Hayama et al., though the calculated sloshing amplitude is smaller than the experimental one in the deep tank. This is attributed to unknown dissipation mechnisms in the present method. Next, sloshing in a tank with elastic walls is calculated. The calculated resonance peak is shifted to a lower frequency due to large deformation of the walls.

A seismic analysis has been performed showing that the seismic interaction between the UIS and the FHM can be avoided eliminating the internal clearance of bearing at the FHM arm joint. An angular contact ball bearing is suitable for the new FHM since it can eliminate gaps by preload pressure. A major problem of the FHM bearings is lubrication since the contact pressure between the steel rings and ball of the ball bearing is larger than that of the roller bearing used in the previous design. Additionally, FHM operating temperature is about 200 deg-C and normal grease is not applicable under argon gas with sodium vapor to prevent contamination of grease in the primary sodium coolant. An endurance test with 1/10 scale bearings in the air has been performed to show applicability of angular contact ball bearings to the FHM arm joint. The results with 20,000 cycle showed that bearings with combination of MoS{sub 2} coating steel rings and ceramics balls can be tolerable as the FHM operating condition. A full-scale bearing test in argon gas with sodium vapor has also been demonstrated to reveal bearing size and sodium vapor effects. (authors)

In phase II of a feasibility study on commercialized fast reactor cycle system of Japan Nuclear Cycle Development Institute, we are finding a concept of a multi-purpose small sized fast reactor with various requirements, such as, safety and improved economical competitiveness. In this study, a membrane reforming hydrogen production plant using a small sized sodium cooled reactor was designed as one of promising concepts. In the membrane reformer, methane and steam are reformed into carbon dioxide and hydrogen with sodium heat at a temperature 500deg-C. In the equilibrium condition, steam reforming proceeds with catalyst at a temperature more than 800deg-C. Using membrane reformers, the steam reforming temperature can be decreased from 800 to 500deg-C because the hydrogen separation membrane remove hydrogen selectively from catalyst area and the partial pressure of hydrogen is kept much lower than equilibrium condition. In this study, a hydrogen and electric co-production plant has been designed. The reactor thermal output is 375MW and 25% of the thermal output is used for hydrogen production (70000Nm3 /h). The hydrogen production cost is estimated but it is still higher than the economical goal. The major reason of the high cost comes from the large size of hydrogen separation reformers because of the limit of hydrogen separation efficiency of palladium membrane. A new highly efficient hydrogen separation membrane is needed to reduce the cost of hydrogen production using membrane reformers. There is possibility of multi-tube failure in the membrane reformers. In future study, a design of measures against tube failure and elemental experiments of reaction between sodium and reforming gas will be needed.Copyright © 2004 by JSME