Maruyama, Shuhei; Oki, Shigeo; Mizuno, Tomoyasu
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9300_1 - 9300_2, 2009/05
In the conceptual core and fuel design studies in Fast Reactor Cycle Technology Development Project (FaCT Project) in Japan, much interest has been taken in the fuel nuclide compositions for a transition period from light water reactor to fast breeder reactor (FBR). In this paper, the range of transuranic (TRU) nuclide composition to be provided to FBR is evaluated with extended recycling scenario calculations. The influence of TRU composition changes on FBR core characteristics are also discussed with explanations of major contributing factors.
Kato, Atsushi; Hirata, Shingo; Chikazawa, Yoshitaka; Uto, Nariaki; Obata, Hiroyuki; Kotake, Shoji
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9281_1 - 9281_6, 2009/05
One of the most important challenges to commercialize a Fast Reactor is to increase economic competitiveness. For that purpose, Japan Sodium cooled Fast Reactor (hereafter JSFR) aims to simplify the plant system and reduce the raw and processed material by adopting innovative technologies. In the JSFR design, a single rotating plug and a reactor upper inner structure (hereafter UIS) with a vertically penetrating slit are proposed, so that the fuel handling machine (hereafter FHM) can access any subassembly by horizontal movement of the FHM arm in the slit space. The feature of this FHM enables no need for the UIS removal when the rotational plug moves round above the core, which can achieve a compact reactor vessel to enhance the economic competitiveness. We fabricated the full scale FHM test equipment to perform comprehensive tests in the air for demonstrating the feasibility of the key characteristics of this FHM concept.
Uto, Nariaki; Sakai, Takaaki; Mihara, Takatsugu; Toda, Mikio*; Kotake, Shoji; Aoto, Kazumi
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9298_1 - 9298_11, 2009/05
A conceptual design for JSFR and developments of innovative technologies are implemented. A compact RV has been designed to enhance the economy. The regarding development results have been reflected to the RV design. An innovative CV design has been implemented with elemental tests to reduce the construction cost. SASS and the NC DHRS have been designed to enhance the safety, with the irradiation data acquired in Joyo and the development of a 3-dimensional thermal-hydraulic evaluation method. An approach for ISI/R has been provided to be applicable for FR characteristics, and the developmental studies on innovative inspection technologies have been progressed. Other technologies including double-walled pipes with short elbows, a pump-integrated IHX are also being developed. These results, together with a preliminary conceptual design study on a demonstrative reactor for JSFR, will be utilized as resources in 2010 to determine which innovative technologies should be adopted.
Misawa, Takeharu; Nakatsuka, Toru; Yoshida, Hiroyuki; Takase, Kazuyuki
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), P. 9487_1, 2009/05
Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9296_1 - 9296_11, 2009/05
In the present study, three-dimensional thermal-hydraulic analyses of the reactor upper plenum in JSFR were applied to evaluating the following countermeasures including the geometrical structure. The basic ideas of countermeasures are as follows. (1) In order to mitigate the thermal striping phenomenon, locations where hot sodium discharged from fuel assemblies meets with cold sodium from control rods and/or radial blanket subassemblies are kept away from un-replaceable structures above the core. (2) In order to prevent the vortex cavitations, asymmetric flow in the upper plenum due to the radial slit with UIS is mitigated by installing a cylindrical structure named as dummy plug instead of the fuel handling machine only used for refueling period. Furthermore, flow holes on perforated plates with UIS are extended to mitigate radial sodium jet from lower part of UIS. (3) In order to prevent the cover-gas entrainment, Dipped Plate (DP) is installed below the sodium free surface. Original design of DP was a double wall type with many labyrinth seals, so the manufacturing regarded as difficult. Then a single wall type DP has been newly designed and examined in this analyses.
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9451_1 - 9451_8, 2009/05
Melting temperatures of nuclear fuel are essential data to design the fuel pin. In this paper, the existing melting temperatures of the oxide fuels were reviewed, and the database of melting temperatures was made. In addition the effect of Pu content, MA content, oxygen-to-metal (O/M) ratio and burn-up on the melting temperatures were evaluated. The solidus temperatures of the oxide fuels were evaluated from the database. Average melting point of UO was estimated to be 3133 K. Melting point of PuO was obtained by evaluating the data of MOX with solid solution model and was estimated to be 2894 K. The solidus temperatures of MOX decreased with Pu content and to 3002 K at 40%Pu content.
Kugo, Teruhiko; Akie, Hiroshi; Yamaji, Akifumi; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9371_1 - 9371_8, 2009/05
Combining a nuclear reactor with thermoelectric converters is expected to be one of promising options to supply a propulsion power for deep space explorers. One of the key features of the concept is to use low enriched uranium fuels from the viewpoint of nuclear non-proliferation. Fuels of uranium oxide, nitride and metal were examined. Zirconium and yttrium hydrides, beryllium, zirconium beryllide and graphite were considered as moderators. Reflectors of beryllium, beryllium oxide, zirconium beryllide and graphite were taken into consideration. A criticality survey of the core was performed by changing the ratio of the fuel, moderator and structure, and the reflector thickness. As a result from the viewpoint of a smaller mass of reactor, it is better to use thermal spectrum cores than fast ones, and the metal hydride moderators than beryllium or graphite. For example, a combination of uranium nitride, yttrium hydride and beryllium reflector achieves a reactor mass of as low as 500kg.
Yamaji, Akifumi; Takizuka, Takakazu; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9366_1 - 9366_8, 2009/05
This study has been carried out in series with the other study, "Criticality of Low Enriched Uranium Fueled Core" to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of three different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The three systems are the solid thermal conduction system (STC), core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and cover down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept.
Yamaji, Akifumi; Suzuki, Motoe; Okubo, Tsutomu
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9092_1 - 9092_9, 2009/05
The concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been proposed and being studied at JAEA to achieve effective and flexible utilization of the uranium and plutonium resources based on the well-developed LWR technology. FLWR is a BWR type concept and it is planned to be introduced by two stages. In the first stage, the MOX fuels are irradiated in a similar condition to that of the current BWR but with a harder neutron spectrum. The core average discharge burnup is about 45 GWd/tHM. In the second stage, the neutron spectrum is further hardened to achieve a multiple recycling of plutonium with higher burnups. In order to design and evaluate the integrities of FLWR fuel rods, the uncertainties in FEMAXI-6 calculations and models for predicting LWR MOX fuel behavior need to be evaluated. As an introduction to the evaluation process, the Test-Fuel-Data-Base (TFDB) obtained from the Halden reactor experiments (IFA-514) were used for the evaluations. The maximum discharge burnup was about 40 GWd/tMOX. Based on the present investigation, the following models were found to be particularly important. Namely the FGR, pellet densification, swelling, and relocation models. These models of FEMAXI-6 have been developed and the parameters have been optimized based on the past UO irradiation test data. For predicting MOX fuel behavior, the FGR model has a relatively large uncertainty and causes a large uncertainty in the FGR calculations. On the other hand, the uncertainties in the other models are within the range expected by the property variations of typical UO fuels. Hence, the densification, swelling, and the relocation models of FEMAXI-6 can be applied to MOX fuel analyses provided the corresponding MOX property variations are taken into account in the input parameters of these models.
Horiguchi, Kenichi; Sugaya, Atsushi; Saito, Yasuo; Tanaka, Kenji; Akutsu, Shigeru; Hirata, Toshiaki
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9411_1 - 9411_9, 2009/05
The low-level radioactive Waste treatment Facility (LWTF) was constructed at the Tokai Reprocessing Plant (TRP) and cold test has been carried out since 2006. The waste which is treated in the LWTF is combustible/incombustible solid waste and liquid waste. In the LWTF, The combustible/incombustible solid waste will be incinerated. The liquid waste will be treated by the radio-nuclides removal process subsequently solidified by cement materials. This report describes the essential technologies of the LWTF and results of R&D work for the nitrate-ion decomposition technology for the liquid waste.
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9078_1 - 9078_6, 2009/05
A new acoustic steam generator leak detection system using delay-and-sum beamformer is proposed. The major advantage of the delay-and-sum beamformer 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 delay-and-sum beamformer could distinguish the acoustic source of the sodium-water reaction from steam generator background noise. In this paper, results from numerical analyses are provided to show fundamental feasibility of the new method. The analysis shows that the new array system can detect a leak sound even if the background noise is as strong as the leak sound.
Furukawa, Tomohiro; Otsuka, Satoshi; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*; Kimura, Akihiko*
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9221_1 - 9221_7, 2009/05
As fuel cladding material for lead bismuth-cooled fast reactors and supercritical pressurized water-cooled fast reactors, our research group has been developing highly corrosion-resistant oxide dispersion strengthened ferritic steels with superior high-temperature strength. In this study, the mechanical properties of super ODS steel candidates at elevated temperature have been evaluated. Tensile tests, creep tests and low cycle fatigue tests were carried out for a total of 21 types of super ODS steel candidates which have a basic chemical composition of Fe-16Cr-4Al-0.1Ti-0.35YO, with small variations. The testing temperatures were 700C (for tensile, creep and low cycle fatigue tests) and 450C (for tensile test). The major alloying parameters of the candidate materials were the compositions of Cr, Al, W and the minor elements such as Hf, Zr and Ce etc. The addition of the minor elements is considered effective in the control of the formation of the YAl complex oxides, which improves high-temperature strength. The addition of Al was very effective for the improvement of corrosion resistance. However, the addition also caused a reduction in high-temperature tensile strength. Among the efforts aimed at increasing high-temperature strength, such as the low-temperature hot-extrusion process, solution strengthening by W and the addition of minor elements, a remarkable improvement of strength was observed in ODS steel with a basic chemical composition of 2W-0.6Hf steel (SOC-14) or 2W-0.6Zr steel (SOC-16). The same behavior was also observed in creep tests, and the creep rupture times of SOC-14 and SOC-16 at 700C - 100MPa were greater than 10,000 h. The strength was similar to that of no-Al ODS steels. No detrimental effect by the additional elements on low-cycle fatigue strength was observed in this study. These results showed that the addition of Hf/Zr to ODS-Al steels was effective in improving high-temperature strength.
Ishikawa, Nobuyuki; Okubo, Tsutomu
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9071_1 - 9071_8, 2009/05
In this study, the potential of mixed oxide fueled light water reactors (MOX-LWRs), especially focusing on the high conversion type LWRs (HC-LWRs) such as FLWR are evaluated in terms of both economic aspect and effective use of plutonium. For economics consideration, relative economics positions of MOX-LWRs are clarified comparing the cost of electricity for uranium fueled LWRs (U-LWRs), MOX-LWRs and fast breeder reactors (FBRs) assuming future natural uranium price raise and variation of parameters such as construction cost and capacity factor. Also the economic superiority of MOX utilization against the uranium use is mentioned from the view point of plutonium credit concerning to the front-end fuel cycle cost. In terms of effective use of plutonium, comparative evaluations on plutonium mass balance in the cases of HC-LWR and high moderation type LWRs (HM-LWRs) taking into account plutonium quality (ratio of fissile to total plutonium) constraint in multiple recycling are performed as representative MOX utilization cases. Through this evaluation, the advantageous features of plutonium multiple recycling by HC-LWR are clarified.
Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9022_1 - 9022_9, 2009/05
An advanced LWR concept, FLWR, is a BWR type reactor, in which the moderation of neutron in the core is reduced by use of tight-lattice fuel rod configuration. It aims at realizing effective and flexible utilization of uranium and plutonium resources by two stages, corresponding to the advancement of the fuel cycle technologies and related infrastructure. The present paper has proposed advanced fuel and core designs for realization of breeder-type operation in the first stage of FLWR. To achieve a high fissile-plutonium conversion ratio over 1.0, a new design concept of the MOX fuel assembly has been developed, in which MOX rods are arranged in the central region of the assembly, while enriched UO rods in the peripheral region of the assembly. Performance evaluation shows that the proposed concept is feasible and promising under the framework of the UO and MOX fuel technologies and related infrastructures, which have been established for the current LWR-MOX utilization.
Kawashima, Katsuyuki; Maruyama, Shuhei; Oki, Shigeo; Mizuno, Tomoyasu
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9288_1 - 9288_7, 2009/05
750 MWe MOX fuel fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters.
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9040_1 - 9040_8, 2009/05
Pulse irradiation tests; simulation of reactivity initiated accident (RIA) were conducted on un-irradiated silicide mini-plate fuels. A failure threshold and its mechanism as well as a water channel closure (plate bowing) were studied as a function of deposited energy and peak cladding surface temperature. For the latter, not only single plate configuration but also triplet plate configuration with a restricted cross flow was conducted. Obtained data is revealed to be useful for licensing of research reactor fuels.
Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9304_1 - 9304_9, 2009/05
Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a low moderation water reactor which can realize Pu breeding and multiple recycling. And for the introduction stage, a high conversion (HC) type FLWR is also proposed to keep technical continuity from current LWRs. When the HC type core is shifted to the breeder (BR) core, there exist both types of fuel assemblies in the same core configuration. The power distribution in the HC + BR assemblies mixed core configuration is studied, because there might appear a power peaking in the adjacent region between HC and BR assemblies due to the difference in neutron spectrum. As a result, though a power peaking can be very large in the adjacent regions between the assemblies, the power distribution can be effectively flattened by considering a rod-wise fuel enrichment distribution and by optimizing the fuel assembly loading pattern. It is expected that FLWR can be shifted from HC type to BR type without major neutronic difficulties.
Takada, Shoji; Funatake, Yoshio; Inagaki, Yoshiyuki
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9225_1 - 9225_5, 2009/05
A design method of a multi-input multi-output controller, which links magnetic forces of multiple magnetic bearings by feedback of multiple measurement values of vibration of a rotor, was proposed for the radial magnetic bearings for the generator rotor of helium gas turbine with a power output of 300 MWe. The generator rotor is a flexible rotor, which passes over the forth critical speed. A controller transfer function was derived to reduce a modeling error at the forth critical speed, in which the bending vibration mode is similar to the one which is excited by unbalance mass. A 1404th order un-symmetric coefficient matrix of equation of the rotating rotor affected by Jayro effect was reduced to 24th by modal decomposition using Schur decomposition to reduce a reduction error. The numerical results showed that unbalance response of rotor was 53 and 80E-6 m (peak to peak), respectively, well below the allowable limits both at the rated and critical speeds.
Koshizuka, Seiichi*; Liu, J.*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; et al.
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), 1 Pages, 2009/05
A computer code, named COMPASS, is developed for multi-physics analysis of core disruptive accidents of sodium-cooled fast reactors (SFRs). A meshless method, called MPS method, is employed since complex thermal-hydraulics and structural problems with various phase change processes have to be analyzed. Verification for separeted basic processes and validation for practical phenomena are carried out. COMPASS is also expected to investigate molten fuel discharge to avoid re-criticality in large size SFR cores. Both MOX and metal fuels are considered. Eutectic reactions between the metal fuel and the cladding material are investigated by phase diagram calculation, classical and first-principles molecular dynamics. Basic studies relevant to the numerical methods support the code development of COMPASS. Parallel processing is implemented by OpenMP to treat large-scale problems. A visualization tool is also prepared by using AVS.
Futagami, Satoshi; Hayafune, Hiroki; Fujimura, Ken; Sato, Mitsuru*
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9169_1 - 9169_8, 2009/05
The major target performance of the SG for commercialized FBR is not only economic performance but also property protection performance. The straight double wall tube SG is one of the SG candidates for commercialized FBR, and other SG concepts were studied in this paper. In proposing an alternative SG, alternative technological measures with a double wall tube were investigated and included reinforcing the tube against wastage and quick detection of initial tube leaks. Alternative SG concept candidates for preventing tube failure propagation and mitigation of water leak accidents were proposed through a combination of technological measures. At the end of JFY 2010, the straight double wall tube SG will be decided upon as the result of R&D activities, and alternative SGs evaluated in feasibility studies. A plan for studying feasibility with the technological issues of the alternative SG was proposed.