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Sample records for httr core components

  1. Construction of the HTTR in-core components

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Jinza, K.; Miki, T.

    1996-01-01

    The reactor internals of HTTR consist of graphite and metallic core support structures and shielding blocks and are designed to support core elements and to shield neutron fluence. They also have functions to restrict by-pass flow for ensuring the core cooling performance and to maintain the temperature of metallic core support structures within their design limits. The detailed design of the HTTR core support structure was approved by the government through safety review, 1990-1991. Machining of all graphite components, which consist of about 150 large blocks, was finished in September 1994 successfully. Machining and fabricating of the metallic components were also finished in September. Prior to their installation in the reactor pressure vessel (RPV), the assembly test of actual reactor internals was performed at the works to confirm above mentioned functions. The assembly test was conducted by examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the RPV and reactor internals as well as under the core support plates with respect to structural integrity, and measuring by-pass flow rate through gaps between graphite components which may degrade core performance. The another purpose of the assembly test was to confirm the installation procedure of those components. All components were assembled at the works according to the planned procedure, and the tests were executed while assembling. As a result of the tests, measured level difference and gap width between reactor internals were negligible from core thermal and hydraulic performance point of view. Coolant flows uniformly in circumferential direction at any axial level in the RPV. By-pass flow rate was found to be suppressed sufficiently and far less than the design limit. (J.P.N.)

  2. Study on practical of eddy current testing of core and core support graphite components in HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Iyoku, Tatsuo; Ooka, Norikazu; Shindo, Yoshihisa; Kawae, Hidetoshi; Hayashi, Motomitsu; Kambe, Mamoru; Takahashi, Masaaki; Ide, Akira.

    1994-01-01

    Core and core support graphite components in the HTTR (High Temperature Engineering Test Reactor) are mainly made of nuclear-grade IG-110 and PGX graphites. Nondestructive inspection with Eddy Current Testing (ECT) is planned to be applied to these components. The method of ECT has been already established for metallic components, however, cannot be applied directly to the graphite ones, because the characteristics of graphite are quite different in micro-structure from those of metals. Therefore, ECT method and condition were studied for the application of the ECT to the graphite components. This paper describes the study on practical method and conditions of ECT for above mentioned graphite structures. (author)

  3. Acceptance test for graphite components and construction status of HTTR

    International Nuclear Information System (INIS)

    Iyoku, T.; Ishihara, M.; Maruyama, S.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs

  4. Characteristics of first loaded IG-110 graphite in HTTR core

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Sawa, Kazuhiro; Hanawa, Satoshi; Ishihara, Masahiro

    2006-10-01

    IG-110 graphite is a fine-grained isotropic and nuclear-grade graphite with excellent resistivity on both irradiation and corrosion and with high strength. The IG-110 graphite is used for the graphite components of High Temperature Engineering Test Reactor (HTTR) such as fuel and control rod guide blocks and support posts. In order to design and fabricate the graphite components in the HTTR, the Japan Atomic Energy Research Institute (the Japan Atomic Energy Agency at present) had established the graphite structural design code and design data on the basis of former research results. After the design code establishment, the IG-110 graphite components were fabricated and loaded in the HTTR core. This report summarized the characteristics of the first loaded IG-110 graphite as basic data for surveillance test, measuring material characteristics changed by neutron irradiation and oxidation. By comparing the design data, it was shown that the first loaded IG-110 graphite had excellent strength properties and enough safety margins to the stress limits in the design code. (author)

  5. PIE technology on mechanical tests for HTTR core component and structural materials developed at Research Hot Laboratory

    International Nuclear Information System (INIS)

    Kizaki, Minoru; Honda, Junichi; Usami, Kouji; Ouchi, Asao; Oeda, Etsuro; Matsumoto, Seiichiro

    2001-02-01

    The high temperature engineering test reactor (HTTR) with the target operation temperature of 950degC established the first criticality on November, 1998 based on a large amount of R and D results on fuel and materials. In such R and D works, the development of reactor materials are one of the key issues from the view point of reactor environments such as extremely high temperature, neutron irradiation and so on for the HTTR. The Research Hot Laboratory (RHL) had carried out much kind of post irradiation examinations (PIEs) on core component and pressure vessel materials for during more than a quarter century. And obtained data played an important role in development, characterization and licensing of those materials for the HTTR. This paper describes the PIE technology developed at RHL and typical results on mechanical tests such as elevated temperature tensile and creep rupture tests for Hasteloy-X, Incolloy 800H and so on, and Charpy impact, J IC fracture toughness, K Id fracture toughness and small punch tests for normalized and tempered 2 1/4Cr-1Mo steel from historical view. In addition, an electrochemical test technique established for investigating the irradiation embrittlement mechanism on 2 1/4Cr-1Mo steel is also mentioned. (author)

  6. Development of visual inspection technology for HTTR core support graphite structure

    International Nuclear Information System (INIS)

    Maruyama, So; Iyoku, Tatsuo; Inagaki, Yoshiyuki; Shiozawa, Shusaku; Masuma, Yoshitaka; Miki, Toshiya.

    1996-01-01

    The Japan Atomic Energy Research Institute is now constructing the High Temperature Engineering Test Reactor (HTTR), which employs a visual inspection of core support graphite structure, as an inservice inspection (ISI). In this inspection, TV camera will be used to investigate the alignment and integrity of the structure. Therefore, the ISI system, a combination of radiation tolerant TV camera and graphic processing system, is developed and examined its detectability and viewing angles using a simulated hot plenum of HTTR, which has artificial defects. As a result of a series of tests, it was confirmed that this system satisfied the requirements and was quite applicable for the ISI system of HTTR core support graphite structure. In addition, further improvement of the system, like a remote control procedure, will be investigated. (author)

  7. Characteristic test of initial HTTR core

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

    2004-01-01

    This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations

  8. An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

    1991-05-01

    The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

  9. Current status of HTTR project at JAERI

    International Nuclear Information System (INIS)

    Saito, Shinzo; Sudo, Yukio; Tanaka, Toshiyuki; Baba, Osamu

    1992-01-01

    The HTTR is a high temperature gas cooled test reactor with thermal output of 30 MW, outlet coolant temperatures of 850degC at rated operation and 950degC at high temperature test operation and primary coolant pressure of 4 MPa. The HTTR consists of a reactor pressure vessel with a prismatic core in it, a primary cooling loop with an intermediate helium-helium heat exchanger and a pressurized water cooler in parallel, an auxiliary cooling system, a reactor vessel cooling system and related components. The HTTR is utilized for establishing and upgrading the technology bases for advanced HTGRs including irradiation tests for fuels and materials, safety demonstration tests for HTGRs and nuclear heat application and for carrying out various kinds of innovative basic researches on high temperature technologies. Since 1969, the JAERI has carried out research and development works on block type fuel, high temperature materials, high temperature in-core instrumentations, high temperature components, reactor physics, heat transfer and fluid dynamics, plate-out of fission products etc., in order to construct the HTTR which can supply high temperature coolant of 950degC to the outside of the pressure vessel for the nuclear heat application, for the first time in the world. In November 1990, the installation permit was issued by the Government through about 20 month safety review by the Science and Technology Agency and Nuclear Safety Commission. The construction of the HTTR facility was initiated on the site in the Oarai Research Establishment, JAERI in March 1991. The excavation of ground is now finished at the HTTR site. It will take about six years for the completion of the HTTR facility and the first criticality will be attained in 1996. (author)

  10. Evaluation of local power distribution with fine-mesh core model for the HTTR

    International Nuclear Information System (INIS)

    Murata, Isao; Yamashita, Kiyonobu; Maruyama, So; Shindo, Ryuichi; Fujimoto, Nozomu; Sudo, Yukio; Nakata, Tetsuo.

    1991-01-01

    An evaluation method of the local power distribution was developed considering the radial and axial heterogeneity caused by fuel rods, BP rods and block end graphite for the High Temperature Engineering Test Reactor (HTTR) in Japan Atomic Energy Research Institute (JAERI). The evaluation method was verified through the analyses of critical assembly experiments. A good agreement was obtained between calculations and measurements and the difference was less than 3 % on the power distribution. This method was applied to the core design for the HTTR to evaluate the maximum fuel temperature. From these results, it was confirmed that this evaluation method has an enough accuracy and is able to predict the detailed power distribution of the HTTR. (author)

  11. Evaluation of aseismic integrity in HTTR core-bottom structure. Pt. 1. Aseismic test for core-bottom structure

    International Nuclear Information System (INIS)

    Iyoku, T.; Futakawa, M.; Ishihara, M.

    1994-01-01

    The aseismic tests were carried out using (1)/(5)-scale and (1)/(3)-scale models of the core-bottom structure of the HTTR to quantitatively evaluate the response of acceleration, strain, impact load etc. The following conclusions are obtained. (i) The frequency response of the keyway strain is correlative with that of the impact acceleration on the hot plenum block. (ii) It was confirmed through (1)/(5)-scale and (1)/(3)-scale model tests that the applied similarity law is valid to evaluate the seismic response characteristics of the core-bottom structure. (ii) The stress of graphite components estimated from the scale model test using S 2 -earthquake excitation was sufficiently lower than the allowable stress used as the design criterion. ((orig.))

  12. Test programmes of HTTR for future generation HTGRs

    International Nuclear Information System (INIS)

    Kunitomi, K.; Tachibana, Y.; Takeda, T.; Saikusa, A.; Shiozawa, S.

    1997-01-01

    Test programs of the High Temperature Engineering Test Reactor (HTTR) for future generation High Temperature Gas-Cooled Reactors (HTGRs) have been established considering design and development status of HTGRs in Japan and the world. Test programs are divided into six categories, thermal hydraulics, fuel, safety, high temperature components, core physics and control-instrumentations. All programs are related to the technology of future generation HTGRs and will be submitted to a new Coordinated Research Program (CRP) so that all participants from the world in test programs of the HTTR can use measured data for their future generation.HTGRs. This paper describes test programs of the HTTR for the development of future generation HTGRs after explanation of a future generation HTGR in Japan. (author)

  13. Basic data for surveillance test on core support graphite structures for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Kikuchi, Takayuki; Iyoku, Tatsuo; Fujimoto, Nozomu; Ishihara, Masahiro; Sawa, Kazuhiro

    2007-02-01

    Both of the visual inspection by a TV camera and the measurement of material properties by surveillance test on core support graphite structures are planned for the High Temperature Engineering Test Reactor (HTTR) to confirm their structural integrity and characteristics. The surveillance test is aimed to investigate the change of material properties by aging effects such as fast neutron irradiation and oxidation. The obtained data will be used not only for evaluating the structural integrity of the core support graphite structures of the HTTR but also for design of advanced Very High Temperature Reactor (VHTR) discussed at generation IV international forum. This report describes the initial material properties of surveillance specimens before installation and installed position of surveillance specimens in the HTTR. (author)

  14. Present status of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1994-01-01

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950degC at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test using HTTR. (author)

  15. Present status of High-Temperature engineering Test Reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1993-01-01

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950 deg C at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test plan using HTTR. (author)

  16. Application of new design methodologies to very high-temperature metallic components of the HTTR

    International Nuclear Information System (INIS)

    Hada, Kazuhiko; Ohkubo, Minoru; Baba, Osamu

    1991-01-01

    The high-temperature piping and helium-to-helium intermediate heat exchanger of the High-Temperature Engineering Test Reactor (HTTR) are designed to be operating at very high temperatures of about 900deg C among the class 1 components of the HTTR. At such a high temperature, mechanical strength of heat-resistant metallic materials is very low and thermal expansions of structural members are large. Therefore, innovative design methodologies are needed to reduce both mechanical and thermal loads acting on these components. To the HTTR, the design methodologies which can separate the heat-resistant function from the pressure-retaining functions and allow them to expand freely are applied to reduce pressure and thermal loads. Since these design methodologies need to verify their applicability, the Japan Atomic Energy Research Institute (JAERI) has been performing many design and research works on their verifications. The details of the design methodologies and their verifications are given in this paper. (orig.)

  17. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  18. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

  19. Development of in-service inspection system for core support graphite structures in the high temperature engineering test reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system developed from September 1996 to June 1998, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed. (author)

  20. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  1. Analysis of the European results on the HTTR's core physics benchmarks

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Ohlig, U.A.; Brockmann, H.J.; Haas, J.B.M. de; Wallerboss, E.M.

    2002-01-01

    Within the frame of the European contract HTR-N1 calculations are performed on the benchmark problems of the HTTR's start-up core physics experiments initially proposed by the IAEA in a Co-ordinated Research Programme. Three European partners, the FZJ in Germany, NRG and IRI in the Netherlands, and CEA in France, have joined this work package with the aim to validate their calculational methods. Pre-test and post-test calculational results, obtained by the partners, are compared with each other and with the experiment. Parts of the discrepancies between experiment and pre-test predictions are analysed and tackled by different treatments. In the case of the Monte Carlo code TRIPOLI4, used by CEA, the discrepancy between measurement and calculation at the first criticality is reduced to Δk/k∼0.85%, when considering the revised data of the HTTR benchmark. In the case of the diffusion codes, this discrepancy is reduced to: Δk/k∼0.8% (FZJ) and 2.7 or 1.8% (CEA). (author)

  2. Structural integrity of graphite core support structures of HTTR

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Iyoku, Tatsuo; Toyota, Junji; Sato, Sadao; Shiozawa, Shusaku

    1990-02-01

    The graphite core support structures (GCSSs) of the HTTR (High Temperature Engineering Test Reactor) are an arrangement of graphite blocks and posts that support the core and provide a lower plenum and a hot-leg path for the primary coolant. The GCSSs are designed not to be replaced by new items during plant life time (about twenty years). To maintain structural integrity of the GCSSs, conservative design has been made sufficiently on the basis of structural tests. The present study confirmed that reactor safety was still maintained even if failure and destruction of the GCSSs is supposed to occur. The GCSSs are fabricated under strict quality control and the observation and surveillance programs are planed to examine the structual integrity of the GCSSs during an operation. This paper describes the concept of design and quality control and summarizes structural tests, observation and surveillance programs. (author)

  3. Preliminary test results for post irradiation examination on the HTTR fuel

    International Nuclear Information System (INIS)

    Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

    2007-01-01

    The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

  4. Determination of hot spot factors for calculation of the maximum fuel temperatures in the core thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    Maruyama, Soh; Yamashita, Kiyonobu; Fujimoto, Nozomu; Murata, Isao; Shindo, Ryuichi; Sudo, Yukio

    1988-12-01

    The Japan Atomic Energy Research Institute (JAERI) has been designing the High Temperature Engineering Test Reactor (HTTR), which is 30 MW in thermal power, 950deg C in reactor outlet coolant temperature and 40 kg/cm 2 G in primary coolant pressure. This report summarizes the hot spot factors and their estimated values used in the evaluation of the maximum fuel temperature which is one of the major items in the core thermal and hydraulic design of the HTTR. The hot spot factors consist of systematic factors and random factors. They were identified and their values adopted in the thermal and hydraulic design were determined considering the features of the HTTR. (author)

  5. Study on system layout and component design in the HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Nishihara, Tetsuo; Shimizu, Akira [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tanihira, Masanori [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Uchida, Shoji [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2003-01-01

    The global warming becomes a significant issue in the world so that it needs to reduce the CO{sub 2} emission. It is expected that hydrogen is in place of the fossil fuels such as coal and oil, and plays the important role to resolve the global warming. There are several hydrogen making processes such as water electrolysis and steam reforming of hydrocarbon. Steam reforming of hydrocarbon is a major hydrogen making process because of economy in industry. It utilizes the fossil fuels as process heat for chemical reaction and results in a large CO{sub 2} emission. New steam reforming system without fossil fuel can contribute to resolve the global warming. High temperature gas-cooled reactor (HTGR) has a unique feature to be able to supply a hot helium gas whose temperature is approximately 950degC at the reactor outlet. This makes HTGR possible to utilize for not only power generation but also process heat utilization. JAERI constructed the high temperature engineering test reactor (HTTR) that is a sort of HTGR in Oarai establishment and starts operation. Nuclear heat utilization is one of the R and D items of the HTTR. The steam reforming system coupling to the HTTR for hydrogen production has been designed. This report represents the system layout and design specification of key components in HTTR steam reforming system. (author)

  6. Study on system layout and component design in the HTTR hydrogen production system. Contract research

    International Nuclear Information System (INIS)

    Nishihara, Tetsuo; Shimizu, Akira; Uchida, Shoji

    2003-01-01

    The global warming becomes a significant issue in the world so that it needs to reduce the CO 2 emission. It is expected that hydrogen is in place of the fossil fuels such as coal and oil, and plays the important role to resolve the global warming. There are several hydrogen making processes such as water electrolysis and steam reforming of hydrocarbon. Steam reforming of hydrocarbon is a major hydrogen making process because of economy in industry. It utilizes the fossil fuels as process heat for chemical reaction and results in a large CO 2 emission. New steam reforming system without fossil fuel can contribute to resolve the global warming. High temperature gas-cooled reactor (HTGR) has a unique feature to be able to supply a hot helium gas whose temperature is approximately 950degC at the reactor outlet. This makes HTGR possible to utilize for not only power generation but also process heat utilization. JAERI constructed the high temperature engineering test reactor (HTTR) that is a sort of HTGR in Oarai establishment and starts operation. Nuclear heat utilization is one of the R and D items of the HTTR. The steam reforming system coupling to the HTTR for hydrogen production has been designed. This report represents the system layout and design specification of key components in HTTR steam reforming system. (author)

  7. Present status and prospects of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1995-01-01

    It is essentially important in Japan, which has limited amount of natural resources, to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors. Hence, efforts are to be continuously devoted to establish and upgrade High Temperature Gas-cooled Reactor (HTGR) technologies and to make much of research resources accumulated so far. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950degC at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). The construction of the HTTR started in March 1991, with first criticality in 1998 to be followed after commissioning testing. At present the HTTR reactor building and its containment vessel have been nearly completed and its main components, such as a reactor pressure vessel, an intermediate heat exchanger, hot gas pipings and core support structures, have been manufactured at their factories and delivered to the Oarai Research Establishment of the JAERI for their installation in the middle of 1994. Fuel fabrication will be started as well. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. The IAEA Coordinated Research Programme on Design and Evaluation of Heat Utilization Systems for the HTTR, such as steam reforming of methane and thermochemical water splitting for hydrogen production, was launched successfully in January 1994. Some heat utilization system is planned to be connected to the HTTR and demonstrated at the former stage of the second core. At present, steam-reforming of methane is the first candidate. The JAERI also plans to conduct material

  8. Fuel temperature analysis method for channel-blockage accident in HTTR

    International Nuclear Information System (INIS)

    Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Hayakawa, Hitoshi

    1994-01-01

    During operation of the High Temperature Engineering Test Reactor (HTTR), coolability must be maintained without core damage under all postulated accident conditions. Channel blockage of a fuel element was selected as one of the design-basis accidents in the safety evaluation of the reactor. The maximum fuel temperature for such a scenario has been evaluated in the safety analysis and is compared to the core damage limits.For the design of the HTTR, an in-core thermal and hydraulic analysis code ppercase[flownet/trump] was developed. This code calculates fuel temperature distribution, not only for a channel blockage accident but also for transient conditions. The validation of ppercase[flownet/trump] code was made by comparison of the analytical results with the results of thermal and hydraulic tests by the Helium Engineering Demonstration Loop (HENDEL) multi-channel test rig (T 1-M ), which simulated one fuel column in the core. The analytical results agreed well with the experiments in which the HTTR operating conditions were simulated.The maximum fuel temperature during a channel blockage accident is 1653 C. Therefore, it is confirmed that the integrity of the core is maintained during a channel blockage accident. ((orig.))

  9. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  10. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

    1990-01-01

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  11. Status of HTTR and Technology Developments for Near Term Deployment of Nuclear Process Heat Applications in Japan

    International Nuclear Information System (INIS)

    Yan, Xing L.

    2017-01-01

    JAEA's High Temperature Engineering Test Reactor: Main features: Thermal power 30 MWt; Fuel SiC TRISO UO2 coated particle fuel, pin in block; Design type Prismatic core; Coolant Helium; Temperature 950 °C (Max.); Pressure 4 MPa. Proven quality of HTTR coated particle fuel - 50-day 950°C operation test in HTTR (2009): • Proved fuel production quality in commercial-scale facility - Fabricated 2 core loads of fuel for HTTR (2 t of U used); - Failure fraction: 2 orders of magnitude < licensed limits • Demonstrated fuel integrity in HTTR operation and extended burnup test: an order of magnitude < licensed limit

  12. Validation of JENDL-3.3 for the HTTR criticality

    International Nuclear Information System (INIS)

    Goto, Minoru; Nojiri, Naoki; Shimakawa, Satoshi

    2004-01-01

    Validation of JENDL-3.3 has been performed for the HTTR criticality using the MVP code with a ''lattice-cell'' of infinite models and a ''whole-core'' of finite models. It was found that the keff values calculated with JENDL-3.3 was decreased about 0.2-0.4%Δk from one with JENDL-3.2. The criticality prediction was closed to the experimental data in the critical approach situation of the HTTR. (author)

  13. Development of irradiation rig in HTTR and dosimetry method. I-I type irradiation equipment

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Taiju; Kikuchi, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi; Ogura, Kazutomo [Japan Atomic Power Co., Tokyo (Japan)

    2002-12-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium gas-cooled test reactor with a maximum power of 30 MW. The HTTR aims not only to establish and upgrade the technological basis for the HTGRs but also to perform the innovative basic research on high temperature engineering with high temperature irradiation fields. It is planned that the HTTR is used to perform various engineering tests such as the safety demonstration test, high temperature test operation and irradiation test with large irradiation fields at high temperatures. This paper describes the design of the I-I type irradiation equipment developed as the first irradiation rig for the HTTR and does the planned dosimetry method at the first irradiation test. It was developed to perform in-pile creep test on a stainless steel with large standard size specimens in the HTTR. It can give great loads on the specimens stably and can control the irradiation temperature precisely. The in-core creep properties on the specimens are measured by newly developed differential transformers and the irradiation condition in the core is monitored by thermocouples and self-powered neutron detectors (SPNDs), continuously. The irradiated neutron fluence is assessed by neutron fluence monitors of small metallic wires after the irradiation. The obtained data at the first irradiation test can strongly be contributed to upgrade the technological basis for the HTGRs, since it is the first direct measurement of the in-core irradiation environments of the HTTR. (author)

  14. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  15. R and D on thermal hydraulics of core and core-bottom structure

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Hino, Ryutaro; Kunitomi, Kazuhiko; Takase, Kazuyuki; Ioka, Ikuo; Maruyama, So

    2004-01-01

    Thermal hydraulic tests on the core and core-bottom structure of the high-temperature engineering test reactor (HTTR) were carried out with the helium engineering demonstration loop (HENDEL) under simulated reactor operating conditions. The HENDEL was composed of helium gas circulation loops, mother sections (M 1 and M 2 ) and adaptor section (A), and two test sections, i.e. the fuel stack test section (T 1 ) and in-core structure test section (T 2 ). In the T 1 test section simulating a fuel stack of the core, thermal and hydraulic performances of helium gas flowing through a fuel block were investigated for thermal design of the HTTR core. In the T 2 test section simulating the core-bottom structure, demonstration tests were performed to verify the structural integrity of graphite and metal components, seal performance against helium gas leakage among the graphite permanent blocks and thermal mixing performance of helium gas. The above test results in the T 1 and T 2 test sections were applied to the detailed design and licensing works of the HTTR and the HENDEL-loop was dismantled in 1999

  16. Cause and countermeasure for heat up of HTTR core support plate at power rise tests

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Nozomu; Takada, Eiji; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-01-01

    HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate reached higher than expected at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in the core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses. (author)

  17. Inherent safety features of the HTTR revealed in the accident condition

    International Nuclear Information System (INIS)

    Kunitomi, K.; Shinozaki, M.; Baba, O.; Saito, S.

    1992-01-01

    The High Temperature Engineering Test Reactor (HTTR) being constructed by JAERI (Japan Atomic Energy Research Institute) is a graphite-moderated and helium-cooled reactor with an outlet gas temperature of 950degC. The inherent safety characteristics in the HTTR prevent temperature increase of reactor fuels and fission product release from the reactor core in postulated accident conditions. The reactor core can be cooled by a Vessel Cooling System (VCS) indirectly, even in the case that no forced cooling is expected during the accident such as primary pipe break. The VCS consists of independent water cooling loop and cooling panel around the reactor pressure vessel. The cooling panel whose temperature of 60-90degC cools the reactor pressure vessel by radiation and removes the decay heat from the core indirectly. Furthermore, even if failure of VCS is assumed during this accident as a severe accident, the reactor core is remained safe despite the temperature increase of biological concrete shield around the reactor pressure vessel. This paper describes the inherent safety features of the HTTR specially focused on the accident condition without forced cooling. The detailed analytical results of such an accident are described together with clarifying the role of the VCS. (author)

  18. Preliminary safety analysis of the HTTR-IS nuclear hydrogen production system

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Tachibana, Yukio; Sakaba, Nariaki

    2010-06-01

    Japan Atomic Energy Agency is planning to demonstrate hydrogen production by thermochemical water-splitting IS process utilizing heat from the high-temperature gas-cooled reactor HTTR (HTTR-IS system). The previous study identified that the HTTR modification due to the coupling of hydrogen production plant requires an additional safety review since the scenario and quantitative values of the evaluation items would be altered from the original HTTR safety review. Hence, preliminary safety analyses are conducted by using the system analysis code. Calculation results showed that evaluation items such as a coolant pressure, temperatures of heat transfer tubes at the pressure boundary, etc., did not exceed allowable values. Also, the peak fuel temperature did not exceed allowable value and therefore the reactor core was not damaged and cooled sufficiently. This report compiles calculation conditions, event scenarios and the calculation results of the preliminary safety analysis. (author)

  19. Development program on HTTR heat application systems at JAERI

    International Nuclear Information System (INIS)

    Ogawa, M.; Inagaki, Y.; Nishihara, T.; Shimizu, S.; Shiozawa, S.; Miyamoto, Y.

    2000-01-01

    The High Temperature Engineering Test Reactor (HTTR), which is a Japanese High Temperature Gas-cooled Reactor (HTGR) with 30 MW thermal output at 950 deg. C of the coolant outlet temperature, was constructed at Oarai Research Establishment of Japan Atomic Energy Research Institute (JAERI). The HTTR has attained the first criticality on November 1998. In JAERI, a hydrogen production system was selected as a heat utilization system of the HTTR. The development program on the HTTR hydrogen production system consists of two parts: one is to establish technologies connecting the hydrogen production system with the HTTR, the other is to establish technologies producing hydrogen from water by using nuclear heat. Finally, hydrogen can be produced from water by using nuclear heat supplied by the HTTR. In the hydrogen production system connected to the HTTR at first, JAERI selected a steam reforming process because its technology had matured. The HTTR hydrogen production system adopting the steam reforming process is being designed to produce hydrogen of about 3800 Nm 3 /hr by using nuclear heat (10MW, 905 deg. C) supplied from the HTTR. The safety principle and criteria are also being investigated for the HTTR hydrogen production system. A facility for an out-of-pile test prior to the demonstration test with the HTTR hydrogen production system is under manufacturing to carry out tests of safety, controllability and performance. The out-of-pile test facility simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30. The tests will be started in 2001 and continued for 4 years or longer. In parallel to the tests, a hydrogen/tritium permeation test and a corrosion test of a catalyst tube of a steam reformer are being carried out to obtain data necessary for the design of the HTTR hydrogen production system. A kind of thermochemical method called IS process is under studying to produce hydrogen from water by

  20. Test plan using the HTTR for commercialization of GTHTR300C

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki; Ueta, Shohei; Aihara, Jun; Goto, Minoru; Sumita, Junya; Shibata, Taiju; Takamatsu, Kuniyoshi; Inagaki, Yoshiyuki; Kunitomi, Kazuhiko; Nishihara, Tetsuo; Hamamoto, Shimpei; Iyoku, Tatsuo

    2010-02-01

    The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA with thermal power of 30 MW and the maximum reactor outlet coolant temperature of 950degC. The HTTR achieved the first criticality in 1998, the reactor outlet coolant temperature of 950degC in 2004, and 30 days continuous operation in 2007. Since 2002, safety demonstration tests including reactivity insertion tests and coolant flow reduction tests have been conducted to show inherent safety features of the HTGRs by using the HTTR. This report describes full scope of the future feasible test plan mainly using the HTTR. The test items cover fuel performance and radionuclide transport, core physics, reactor thermal hydraulics and plant dynamics, and reactor operations, maintenance, control, etc. The test results will be utilized for realization of Japan's commercial Very High Temperature Reactor (VHTR) system, GTHTR300C. (author)

  1. Research and development of HTTR hydrogen production systems

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Ogawa, Masuro; Inagaki, Yoshiyuki; Onuki, Kaoru; Takeda, Tetsuaki; Nishihara, Tetsuo; Hayashi, Koji; Kubo, Shinji; Inaba, Yoshitomo; Ohashi, Hirofumi

    2002-01-01

    The Japan Atomic Energy Research Institute (JAERI) has constructed the High Temperature Engineering Test Reactor (HTTR) with a thermal output of 30MW and a reactor out let coolant temper at ure of 950 .deg. C. There search and development (R and D) program on nuclear production of hydrogen was started on January in 1997 as a study consigned by Ministry of Education, Culture, Sports, Science and Technology. A hydrogen production system connected to the HTTR is being designed to be able to produce hydrogen of about 4000m 3 /h by steam reforming of natural gas, using a nuclear heat of 10MW supplied by the HTTR hydrogen production system. In order to confirm controllability, safety and performance of key components in the HTTR hydrogen production system, the facility for the out-of-pile test was constructed on the scale of approximately 1/30 of the HTTR hydrogen production system. In parallel to the out-of-pile test, the following tests as essential problem, a corrosion test of a reforming tube, a permeation test of hydrogen isotopes through heat exchanger and reforming tubes, and an integrity test of a high-temperature isolation valve are carried out to obtain detailed data for safety review and development of analytical codes. Other basis studies on the hydrogen production technology of thermochemical water splitting called an iodine sulfur (IS) process, has been carried out for more effective and various uses of nuclear heat. This paper describes the present status and a future plan on the R and D of the HTTR hydrogen production systems in JAERI

  2. Study on control characteristics for HTTR hydrogen production system with mock-up test facility

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Ohashi, Hirofumi; Nishihara, Tetsuo; Sato, Hiroyuki; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji

    2005-01-01

    The Japan Atomic Energy Research Institute has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the High-Temperature Engineering Test Reactor (HTTR). Prior to the coupling of a hydrogen production plant with the HTTR, simulation tests with a mock-up test facility of the HTTR hydrogen production system (HTTR-H2) is underway. The test facility is a 1/30-scale of the HTTR-H2 and simulates key components downstream from an intermediate heat exchanger of the HTTR. The main objective of the simulation tests is the establishment and demonstration of control technology, focusing on the mitigation of a thermal disturbance to the reactor by a steam generator (SG) and on the controllability of the pressure difference between the helium and process gases at the reaction tube in a steam reformer (SR). It was confirmed that the fluctuation of the outlet helium gas temperature at the SG and the pressure difference in the SR can be controlled within the allowable range for the HTTR-H2 in the case of the system controllability test for the fluctuation of chemical reaction. In addition, a dynamic simulation code for the HTTR-H2 was verified with the obtained test data

  3. Design of high temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

    1994-09-01

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

  4. Water chemistry control in HTTR

    International Nuclear Information System (INIS)

    Sekita, Kenji; Furusawa, Takayuki; Emori, Koichi; Kuroha, Misao; Hayakawa, Masato; Ohuchi, Hiroshi; Ishii, Taro

    2008-08-01

    A carbon steel is used for the main material for the components and pipings of the pressurized water cooling system etc. that are the reactor cooling system of the HTTR. Water quality is managed by using the hydrazine in the coolant of the water cooling system to prevent corrosion of the components and deoxidize the coolant. Also, regular analysis is carried out for the confirmation of the water quality. The following results were obtained through the water quality analysis. (1) In the pressurized water cooling system, the coolant temperature rises higher due to the heat removal of the primary coolant. So, the ammonia was formed in the thermal decomposition of the hydrazine. The electric conductivity increased, while the concentration of the hydrazine decreased, there was no problem as the plan it. (2) Thermal decomposition of the hydrazine was not occurred in the auxiliary water cooling system and vessel cooling system because of the coolant temperature was low. (3) An indistinct procedure is clarified and procedure of water quality analysis was established in the HTTR. (4) It is assumed that the corrosion of the components in these water cooling system hardly occurred from measurement results of dissolved oxide and chloride ion. Thus, the water quality was managed enough. (author)

  5. Steady state HTTR model in the RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Scari, Maria E.; Reis, Patrícia A. L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A. F., E-mail: melizabethscari@yahoo.com, E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPQ (Brazil)

    2017-07-01

    The HTTR (High Temperature Engineering Test Reactor) is a high temperature gas-cooled reactor (HTGR). It is a helium-cooled and graphite-moderated with a thermal power of 30 MW, developed and operated by the Japan Atomic Energy Research Institute (JAERI). In a previous work, a thermal model for the HTTR has been developed using the RELAP5-3D code. In the present work, additional thermal analyses have been presented. Studies about the core neutronic simulation were also initiated and are described here. The code WIMSD-5B was used to generate the cross sections and the group constants necessary to the NESTLE neutron kinetic code incorporated in the RELAP5-3D. (author)

  6. Out-of-pile demonstration test of hydrogen production system coupling with HTTR

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Nishihara, Tetsuo; Takeda, Tetsuaki; Hada, Kazuhiko; Hayashi, Koji

    1999-01-01

    In Japan Atomic Energy Research Institute, a hydrogen production system is being designed to produce hydrogen by means of a steam reforming process of natural gas using nuclear heat (10 MW, 905degC) supplied by the High Temperature Engineering Test Reactor (HTTR). The safety principle and criteria are also being investigated in the HTTR hydrogen production system. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile demonstration test was planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 has a hydrogen production capacity of 110 Nm 3 /h using an electric heater as a reactor substitute. The test facility is under manufacturing aiming at completion in 2000 and followed by the test till 2004. In parallel to this, a hydrogen permeation test and a corrosion test of a catalyst tube of a steam reformer are being carried out to obtain data necessary for the design of the system. This report describes outline of the out-of-pile hydrogen production facility and demonstration test program for the HTTR hydrogen production system at present status. (author)

  7. Out-of-pile demonstration test of hydrogen production system coupling with HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Nishihara, Tetsuo; Takeda, Tetsuaki; Hada, Kazuhiko; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1999-07-01

    In Japan Atomic Energy Research Institute, a hydrogen production system is being designed to produce hydrogen by means of a steam reforming process of natural gas using nuclear heat (10 MW, 905degC) supplied by the High Temperature Engineering Test Reactor (HTTR). The safety principle and criteria are also being investigated in the HTTR hydrogen production system. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile demonstration test was planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 has a hydrogen production capacity of 110 Nm{sup 3}/h using an electric heater as a reactor substitute. The test facility is under manufacturing aiming at completion in 2000 and followed by the test till 2004. In parallel to this, a hydrogen permeation test and a corrosion test of a catalyst tube of a steam reformer are being carried out to obtain data necessary for the design of the system. This report describes outline of the out-of-pile hydrogen production facility and demonstration test program for the HTTR hydrogen production system at present status. (author)

  8. Improvement of calculation method for temperature coefficient of HTTR by neutronics calculation code based on diffusion theory. Analysis for temperature coefficient by SRAC code system

    International Nuclear Information System (INIS)

    Goto, Minoru; Takamatsu, Kuniyoshi

    2007-03-01

    The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)

  9. Control rod position and temperature coefficients in HTTR power-rise tests. Interim report

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji; Saito, Kenji; Kobayashi, Shoichi; Sawahata, Hiroaki; Kokusen, Sigeru

    2001-03-01

    Power-rise tests of the High Temperature Engineering Test Reactor (HTTR) have been carried out aiming to achieve 100% power. So far, 50% of power operation and many tests have been carried out. In the HTTR, temperature change in core is so large to achieve the outlet coolant temperature of 950degC. To improve the calculation accuracy of the HTTR reactor physics characteristics, control rod positions at criticality and temperature coefficients were measured at each step to achieve 50% power level. The calculations were carried out using Monte Carlo code and diffusion theory with temperature distributions in the core obtained by reciprocal calculation of thermo-hydraulic code and diffusion theory. Control rod positions and temperature coefficients were calculated by diffusion theory and Monte Carlo method. The test results were compared to calculation results. The control rod positions at criticality showed good agreement with calculation results by Monte Carlo method with error of 50 mm. The control position at criticality at 100% was predicted around 2900mm. Temperature coefficients showed good agreement with calculation results by diffusion theory. The improvement of calculation will be carried out comparing the measured results up to 100% power level. (author)

  10. Design of a steam reforming system to be connected to the HTTR

    International Nuclear Information System (INIS)

    Hada, K.; Nishihara, T.; Shibata, T.; Shiozawa, S.

    1996-01-01

    Top priority objective for developing the first heat utilization system to be connected to the HTTR is to demonstrate technical feasibility of a nuclear process heat utilization system for production of hydrogen for the first time in the world. Major issues to be resolved for coupling the heat utilization system to the HTTR are 1)to develop safety philosophy for reasonably and reliably ensuring safety of the nuclear reactor, 2)to develop control design concept for the total system of the nuclear reactor and heat utilization system because thermal dynamics of endothermic chemical reactor to be heated by nuclear heat is much different from the nuclear reactor, 3)to develop helium-heated components and 4)to develop enhanced hydrogen production technologies for achieving competitiveness to a fossil-fired plant. A steam reforming hydrogen production system was studied as one of the first priority candidates for an HTTR-heat utilization system due to matured technology in fossil-fired plants and since technical solutions demonstrated by the coupling of the steam reforming system to the HTTR will contribute to all other hydrogen production systems. Basic design philosophy for the HTTR-steam reforming system is that the steam reforming plant downstream of an intermediate secondary helium loop is designed at the same safety level as fossil-fired plants and therefore the secondary helium loop was selected as a safety barrier to the HTTR nuclear reactor. (J.P.N.)

  11. Design of helium-gas supplying facility of out-of-pile demonstration test for HTTR heat utilization system

    Energy Technology Data Exchange (ETDEWEB)

    Hino, Ryutaro; Fujisaki, Katsuo; Kobayashi, Toshiaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    1996-09-01

    One of the objectives of the High-Temperature Engineering Test Reactor (HTTR) is to demonstrate effectiveness of high-temperature heat utilization. Prior to connect a heat utilization system to the HTTR, a series of out-of-pile demonstration test is indispensable to improve components` performance, to demonstrate operation, control and safety technologies and to verify analysis codes for design and safety evaluation. After critical review and discussion on the out-of-pile demonstration test, a test facility have been designed. In this report, a helium-gas supplying facility simulated the HTTR system was described in detail, which supplies High-temperature helium-gas of 900degC to a steam reforming facility mocking-up the HTTR heat utilization system. Components of the Helium Engineering Demonstration Loop (HENDEL) were selected to reuse in the helium-gas supplying facility in order to decrease construction cost. Structures and specifications of new components such as a high-temperature heater and a preheater were decided after evaluation of thermal and hydraulic performance and strength. (author)

  12. Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

    2007-03-01

    The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

  13. Preliminary analyses for HTTR's start-up physics tests by Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  14. Verification of in-core thermal and hydraulic analysis code FLOWNET/TRUMP for the high temperature engineering test reactor (HTTR) at JAERI

    International Nuclear Information System (INIS)

    Maruyama, Soh; Sudo, Yukio; Saito, Shinzo; Kiso, Yoshihiro; Hayakawa, Hitoshi

    1989-01-01

    The FLOWNET/TRUMP code consists of a flow network analysis code 'FLOWNET' for calculations of coolant flow distribution and coolant temperature distribution in the core with a thermal conduction analysis code 'TRUMP' for calculation of temperature distribution in solid structures. The verification of FLOWNET/TRUMP was made by the comparison of the analytical results with the results of steady state experiments by the HENDEL multichannel test rig, T1-M, which consisted of twelve simulated fuel rods heated electrically and eleven hexagonal graphite fuel blocks. The T1-M simulated the one fuel column in the core. The analytical results agreed well with the results of the experiment in which the HTTR operating conditions were simulated. (orig.)

  15. Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1995-03-01

    In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

  16. Preliminary analyses for HTTR`s start-up physics tests by Monte Carlo code MVP

    Energy Technology Data Exchange (ETDEWEB)

    Nojiri, Naoki [Science and Technology Agency, Tokyo (Japan); Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  17. HTTR criticality calculations with SCALE6: Studies of various geometric and unit-cell options in modeling

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J. Y.; Chiang, M. H.; Sheu, R. J.; Liu, Y. W. H. [Inst. of Nuclear Engineering and Science, National Tsing Hua Univ., Hsinchu 30013, Taiwan (China)

    2012-07-01

    The fuel element of the High Temperature Engineering Test Reactor (HTTR) presents a doubly heterogeneous geometry, where tiny TRISO fuel particles dispersed in a graphite matrix form the fuel region of a cylindrical fuel rod, and a number of fuel rods together with moderator or reflector then constitute the lattice design of the core. In this study, a series of full-core HTTR criticality calculations were performed with the SCALE6 code system using various geometric and unit-cell options in order to systematically investigate their effects on neutronic analysis. Two geometric descriptions (ARRAY or HOLE) in SCALE6 can be used to construct a complicated and repeated model. The result shows that eliminating the use of HOLE in the HTTR geometric model can save the computation time by a factor of 4. Four unit-cell treatments for resonance self-shielding corrections in SCALE6 were tested to create problem-specific multigroup cross sections for the HTTR core model. Based on the same ENDF/B-VII cross-section library, their results were evaluated by comparing with continuous-energy calculations. The comparison indicates that the INFHOMMEDIUM result overestimates the system multiplication factor (k{sub eff}) by 55 mk, whereas the LATTICECELL and MULTIREGION treatments predict the k{sub eff} values with similar biases of approximately 10 mk overestimation. The DOUBLEHET result shows a more satisfactory agreement, about 4.2 mk underestimation in the k{sub eff} value. In addition, using cell-weighted cross sections instead of an explicit modeling of TRISO particles in fuel region can further reduce the computation time by a factor of 5 without sacrificing accuracy. (authors)

  18. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

    International Nuclear Information System (INIS)

    2005-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)

  19. Evaluation of heat exchange performance for the auxiliary component cooling water system cooling tower in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Kameyama, Yasuhiko; Shimizu, Atsushi; Inoi, Hiroyuki; Yamazaki, Kazunori; Shimizu, Yasunori; Aragaki, Etsushi; Ota, Yukimaru; Fujimoto, Nozomu

    2006-09-01

    The auxiliary component cooling water system (ACCWS) is one of the cooling system in High Temperature Engineering Test Reactor (HTTR). The ACCWS has main two features, many facilities cooling, and heat sink of the vessel cooling system which is one of the engineering safety features. Therefore, the ACCWS is required to satisfy the design criteria of heat removal performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the ACCWS cooling tower was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that ACCWS cooling tower has the required heat exchange performance in the design. (author)

  20. Post irradiation examinations on HTTR materials

    International Nuclear Information System (INIS)

    Sakai, Haruyuki; Ohmi, Masao; Eto, Motokuni; Watanabe, Katsutoshi

    1995-01-01

    The HTTR (High Temperature engineering Test Reactor) is being constructed at Oarai Research Establishment of the Japan Atomic Energy Research Institute. In order to develop necessary materials for the HTTR, after irradiations in the JMTR, PIEs are being carried out on these materials in the JMTRHL (JMTR Hot Laboratory). Impact test, tensile test, fatigue test, creep test, metallography and so on were performed for irradiated 2 1/4Cr 1Mo steel as the pressure vessel material and Alloy 800H as the cladding material of the control rod. A fatigue testing machine and four creep testing machines newly designed were fabricated and installed in the steel cells in order to evaluate the integrity of the HTTR materials. The development process and PIE results obtained with these machines are given in this paper

  1. Characteristics of HTTR's startup physics tests

    International Nuclear Information System (INIS)

    Nojiri, N.; Nakano, M.; Takeuchi, M.; Pohl, P.; Yamashita, K.

    1997-01-01

    The High Temperature Engineering Test Reactor (HTTR) which is under construction by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium gas-cooled reactor with an outlet temperature of 950 deg. C and a thermal output of 30MW. The first criticality is expected at the end of October 1997. The start-up physics tests (SPTs) are planned in the period from mid 1997 to the end of 1998. Characteristic items of the SPTs are: 1) Criticality approach; 2) Tests on a preliminary annual core; 3) Measurement of scram reactivity; 4) Excess reactivity test; 5) Measurements along with a 2-step-scram reactor shutdown procedure. (author)

  2. HTTR plant dynamic simulation using a hybrid computer

    International Nuclear Information System (INIS)

    Shimazaki, Junya; Suzuki, Katsuo; Nabeshima, Kunihiko; Watanabe, Koichi; Shinohara, Yoshikuni; Nakagawa, Shigeaki.

    1990-01-01

    A plant dynamic simulation of High-Temperature Engineering Test Reactor has been made using a new-type hybrid computer. This report describes a dynamic simulation model of HTTR, a hybrid simulation method for SIMSTAR and some results obtained from dynamics analysis of HTTR simulation. It concludes that the hybrid plant simulation is useful for on-line simulation on account of its capability of computation at high speed, compared with that of all digital computer simulation. With sufficient accuracy, 40 times faster computation than real time was reached only by changing an analog time scale for HTTR simulation. (author)

  3. System analysis for HTTR-GT/H2 plant. Safety analysis of HTTR for coupling helium gas turbine and H2 plant

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Yan, Xing L.; Ohashi, Hirofumi

    2017-08-01

    High Temperature Gas-cooled Reactor (HTGR) is expected to extend the use of nuclear heat to a wider spectrum of industrial applications because of the high temperature heat supply capability and inherently safe characteristics. Japan Atomic Energy Agency initiated a nuclear cogeneration demonstration project with helium gas turbine power generation and thermochemical hydrogen production utilizing the High Temperature engineering Test Reactor (HTTR), the first HTGR in Japan. This study carries out safety evaluation for the HTTR gas turbine hydrogen cogeneration test plant (HTTR-GT/H 2 plant). The evaluation was conducted for the events newly identified corresponding to the coupling of helium gas turbine and hydrogen production plant to the HTTR. The results showed that loss of load event does not have impact on temperature of fuel and reactor coolant pressure boundary. In addition, reactor coolant pressure does not exceed the evaluation criteria. Furthermore, it was shown that reactor operation can be maintained against temperature transients induced by abnormal events in hydrogen production plant. (author)

  4. Prediction of Fission Product Release during the LOFC Experiments at the HTTR

    International Nuclear Information System (INIS)

    Shi, D.; Xhonneux, A.; Verfondern, K.; Ueta, S.; Allelein, H.-J.

    2014-01-01

    Demonstration tests were conducted using the High Temperature Engineering Test Reactor (HTTR) in Oarai, Japan, to confirm the safety of HTGR technologies and assure the expected physical phenomena to occur under given conditions. As part of the OECD directed LOFC (“loss of forced cooling”) project, a series of three tests at the HTTR has been planned with tripping of all gas circulators while deactivating all reactor reactivity control to disallow reactor scram due to abnormal reduction of primary coolant flow rate. The tests fall into anticipated transient without scram (ATWS) with occurrence of reactor recriticality. They serve the important purpose to provide a valuable data base for the validation of computer models regarding neutronics, heat transfer and fluid dynamics, fuel performance and fission product transport and release behavior in HTGRs. The Source Term Analysis Code System (STACY) is a new code development at the Research Center Jülich encompassing the original verified and validated computer models for simulating fission product transport and release. For verification of the modernized and extended version, it was assured that results obtained with the original tools could be reproduced. One of the new features of STACY is its ability to also treat fuel compacts of (full) cylindrical or annular shape and a complete prismatic block reactor core, respectively, supposed sufficient input data be available. The paper will describe the new STACY tool and present the results of fission product behavior in the HTTR core under the LOFC test conditions. Calculations are based on time-dependent neutronics and fluid dynamics results obtained with the Serpent and MGT models. (author)

  5. The Assessment Of High Temperature Reactor Fuel (Characteristics Of HTTR Fuel)

    International Nuclear Information System (INIS)

    Dewita, Erlan; Tuka, Veronica; Gunandjar

    1996-01-01

    HTTR is one of the reactor type with Helium coolant and outlet coolant temperature of 950 o C. One possibility of HTTR application is the coo generation of steam in high temperature and electric power for supply energy to industry in the future. Considering to the high operating temperature of HTTR, therefore it is needed the reactor fuel which have good mechanical, chemical and physical stability to the high temperature, and stable to the influence of fission fragment and neutron during irradiation. This assessment of the HTTR fuel characteristic based on the experiment data to find information of HTTR operation feasibility. Result of the assessment indicated that fission gas release at burn-up of 3.6 % FIMA which was the same as the maximum burn up in the HTTR design was fairly lower than the maximum release estimated in the design (5 x 10 - 4), which is R/B from the fuel fabricated by the prismatic block fuel method would be low (between 10 - 9 dan 10 - 8)

  6. Preliminary design of steam reformer in out-pile demonstration test facility for HTTR heat utilization system

    Energy Technology Data Exchange (ETDEWEB)

    Haga, Katsuhiro; Hino, Ryutaro; Inagaki, Yosiyuki; Hata, Kazuhiko; Aita, Hideki; Sekita, Kenji; Nishihara, Tetsuo; Sudo, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Yamada, Seiya

    1996-11-01

    One of the key objectives of HTTR is to demonstrate effectiveness of high-temperature nuclear heat utilization system. Prior to connecting a heat utilization system to HTTR, an out-pile demonstration test is indispensable for the development of experimental apparatuses, operational control and safety technology, and verification of the analysis code of safety assessment. For the first heat utilization system of HTTR, design of the hydrogen production system by steam reforming is going on. We have proposed the out-pile demonstration test plan of the heat utilization system and conducted preliminary design of the test facility. In this report, design of the steam reformer, which is the principal component of the test facility, is described. In the course of the design, two types of reformers are considered. The one reformer contains three reactor tubes and the other contains one reactor tube to reduce the construction cost of the test facility. We have selected the steam reformer operational conditions and structural specifications by analyzing the steam reforming characteristics and component structural strength for each type of reformer. (author)

  7. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  8. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  9. Stress analysis of two-dimensional C/C composite components for HTGR's core restraint techanism

    International Nuclear Information System (INIS)

    Satoshi Hanawa; Taiju Shibata; Jyunya Sumita; Masahiro Ishihara; Tatsuo Iyoku; Kazuhiro Sawa

    2005-01-01

    Carbon fiber reinforced carbon matrix composite (C/C composite) is one of the most promising materials for HTGRs core components due to their high strength as well as high temperature resistibility. One of the most attractive applications of C/C composite is the core restraint mechanism. The core restraint mechanism is located around the reflector block and it works to tighten reactor core blocks so as to restrict un-supposition flow pass of coolant gas (bypass flow) in the core. The restriction of bypass flow reads to the high efficiency of coolant flow rate inside of the reactor core. For the future HTGRs and VHTR (Very High Temperature Reactor), it is important to develop the core restraint mechanism with C/C composite substitute for metallic materials as used for HTTR. For the application of C/C composite to core restraint mechanism, it is important to investigate the applicability of C/C composite in viewpoint of structural integrity. In the present study, supposing the application of 2D-C/C composite to core restraint mechanism, thermal stress behavior was analyzed by considering the thickness of the C/C composite and the gap between reflector block and core restraint. It was shown from the thermal stress analysis that the circumferential stress decreases with increasing the gap and that the restraint force increases with increasing the thickness. By optimizing the thickness of C/C composite and gap between reflector block and core restraint, the C/C composite is applicable to the core restraint mechanism. (authors)

  10. Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Atsushi, E-mail: shimizu.atsushi35@jaea.go.jp [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Kawamoto, Taiki [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Tochio, Daisuke [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Saito, Kenji; Sawahata, Hiroaki; Honma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Takada, Shoji [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Shinozaki, Masayuki [HTTR Operation Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393 (Japan)

    2014-05-01

    To establish the technical basis of HTGR (High Temperature Gas cooled Reactor), the long term high temperature operation using HTTR was carried out in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to establish the technical basis of operation and maintenance by demonstrating the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, it was confirmed that the temperatures and flow rate of primary and secondary coolant were well controlled within sufficiently small deviation against the disturbance by the atmospheric temperature variation in daily. Stability and reliability of the components and facility special to HTGRs was demonstrated through the long term high temperature operation by evaluating the heat transfer performance of high temperature components, the stability performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the long term high temperature operation of HTTR, the technical basis for the operation and maintenance technology of HTGRs was established.

  11. Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

    International Nuclear Information System (INIS)

    Shimizu, Atsushi; Kawamoto, Taiki; Tochio, Daisuke; Saito, Kenji; Sawahata, Hiroaki; Honma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio; Takada, Shoji; Shinozaki, Masayuki

    2014-01-01

    To establish the technical basis of HTGR (High Temperature Gas cooled Reactor), the long term high temperature operation using HTTR was carried out in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to establish the technical basis of operation and maintenance by demonstrating the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, it was confirmed that the temperatures and flow rate of primary and secondary coolant were well controlled within sufficiently small deviation against the disturbance by the atmospheric temperature variation in daily. Stability and reliability of the components and facility special to HTGRs was demonstrated through the long term high temperature operation by evaluating the heat transfer performance of high temperature components, the stability performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the long term high temperature operation of HTTR, the technical basis for the operation and maintenance technology of HTGRs was established

  12. Status of the HTTR project in Japan

    International Nuclear Information System (INIS)

    Sanokawa, K.

    1989-01-01

    In June 1987, the Japanese Atomic Energy Commission issued the revision of the Long-Term Program for Development and Utilization of Nuclear Energy, recommending that Japan should proceed with the development of more advanced new technologies for the future, parallel to the existing nuclear systems. It also emphasizes that the HTGR is one of the most promising reactors with high efficiency and inherent safety, therefore it should be explored for the broader use of nuclear energy, not only for power production. Then the early construction of a High-Temperature engineering Test Reactor (HTTR) by JAERI was proposed. Based on this program, JAERI has changed the ''VHTR program'' for ''HTTR program'' to establish the HTGR technology basis and upgrade them. The permission of the HTTR construction was made by the Science and Technology Agency (STA) in February 1989. 3 figs, 2 tabs

  13. Structural integrity assessment of intermediate heat exchanger in the HTTR. Based on results of rise-to-power test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Tachibana, Yukio; Nakagawa, Shigeaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-12-01

    A helium/helium intermediate heat exchanger (IHX) in the high temperature engineering test reactor (HTTR) is an essential component for demonstration of future nuclear process heat utilization of high temperature gas-cooled reactor (HTGR). The IHX with a heat capacity of 10 MW has 96 helically-coiled heat transfer tubes. Structural design for the IHX had been conducted through elastic-creep analysis of superalloy Hastelloy XR components such as heat transfer tubes and center pipe. In the HTTR rise-to-power test, it was clarified that temperature of the coolant in the IHX at the reactor scrams changes more rapidly than expected in the design. Effects of the IHX coolant temperature change, at anticipated reactor scram from the full power of 30 MW at high temperature test operation, on structural integrity of the heat transfer tubes and the lower reducer of the center pipe were investigated analytically based on the coolant temperature data obtained from the rise-to-power test. As results of the assessment, it was confirmed that cumulative principal creep strain, cumulative creep and fatigue damage factor of the IHX components during 10{sup 5} h of the HTTR lifetime should be below the allowable limits, which are established in the high-temperature structural design code for the HTGR Class 1 components. (author)

  14. Evaluation of parameters effect on the maximum fuel temperature in the core thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Maruyama, Soh; Sudo, Yukio; Fujii, Sadao; Niguma, Yoshinori.

    1988-10-01

    This report presents the results of quantitative evaluation on the effects of the dominant parameters on the maximum fuel temperature in the core thermal hydraulic design of the High Temperature Engineering Test Reactor(HTTR) of 30 MW in thermal power, 950 deg C in reactor outlet coolant temperature and 40 kg/cm 2 G in coolant pressure. The dominant parameters investigated are 1) Gap conductance. 2) Effect of eccertricity of fuel compacts in graphite sleeve. 3) Effect of spacer ribs on heat transfer coefficients. 4) Contact probability of fuel compact and graphite sleeve. 5) Validity of uniform radial power density in the fuel compacts. 6) Effect of impurity gas on gap conductance. 7) Effect of FP gas on gap conductance. The effects of these items on the maximum fuel temperature were quantitalively identified as hot spot factors. A probability of the appearance of the maximum fuel temperature was also evaluated in this report. (author)

  15. HTTR demonstration test plan for industrial utilization of nuclear heat

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Yan, Xing L.; Kubo, Shinji; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

    2014-09-01

    Japan Atomic Energy Agency has been conducting research and development with a central focus on the utilization of High Temperature engineering Test Reactor (HTTR), the first High Temperature Gas-cooled Reactor (HTGR) in Japan, towards the realization of industrial use of nuclear heat. Several studies have made on the integration of the HTTR with thermochemical iodine-sulfur process and steam methane reforming hydrogen production plant (H 2 plant) as well as helium gas turbine power conversion system. In addition, safety standards for coupling a H 2 plant to a nuclear facility has been investigated. Based on the past design information, the present study identified test items to be validated in the HTTR demonstration test to accomplish a formulation of safety requirement and design consideration for coupling a H 2 plant to a nuclear facility as well as confirmation of overall performance of helium gas turbine system. In addition, plant concepts for the heat utilization system to be connected with the HTTR are investigated. (author)

  16. A simplified 2D HTTR benchmark problem

    International Nuclear Information System (INIS)

    Zhang, Z.; Rahnema, F.; Pounders, J. M.; Zhang, D.; Ougouag, A.

    2009-01-01

    To access the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of relevant whole core configurations. In this paper we have created a numerical benchmark problem in 2D configuration typical of a high temperature gas cooled prismatic core. This problem was derived from the HTTR start-up experiment. For code-to-code verification, complex details of geometry and material specification of the physical experiments are not necessary. To this end, the benchmark problem presented here is derived by simplifications that remove the unnecessary details while retaining the heterogeneity and major physics properties from the neutronics viewpoint. Also included here is a six-group material (macroscopic) cross section library for the benchmark problem. This library was generated using the lattice depletion code HELIOS. Using this library, benchmark quality Monte Carlo solutions are provided for three different configurations (all-rods-in, partially-controlled and all-rods-out). The reference solutions include the core eigenvalue, block (assembly) averaged fuel pin fission density distributions, and absorption rate in absorbers (burnable poison and control rods). (authors)

  17. Investigation of water content in primary upper shield of high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Sumita, Junya; Sawa, Kazuhiro; Mogi, Haruyoshi; Itahashi, Shuuji; Kitami, Toshiyuki; Akutu, Youichi; Fuchita, Yasuhiro; Kawaguchi, Toru; Moriya, Masahiro

    1999-09-01

    A primary upper shield of the High Temperature Engineering Test Reactor (HTTR) is composed of concrete (grout) which is packed into iron frames. The main function of the primary upper shield is to attenuate neutron and gamma ray from the core, that leads to satisfy dose equivalent rate limit of operating floor and stand-pipe room. Water content in the concrete is one of the most important things because it strongly affects neutron-shielding ability. Then, we carried out out-of-pile experiments to investigate relationship between temperature and water content in the concrete. Based on the experimental results, a hydrolysis-diffusion model was developed to investigate water release behavior from the concrete. The model showed that water content used for shielding design in the primary upper shield of the HTTR will be maintained if temperature during operating life is under 110degC. (author)

  18. Production process and quality control for the HTTR fuel

    International Nuclear Information System (INIS)

    Yoshimuta, S.; Suzuki, N.; Kaneko, M.; Fukuda, K.

    1991-01-01

    Development of the production and inspection technology for High Temperature Engineering Test Reactor (HTTR) fuel has been carried out by cooperative work between Japan Atomic Energy Research Institute (JAERI) and Nuclear Fuel Industries, Ltd (NFI). The performance and the quality level of the developed fuel are well established to meet the design requirements of the HTTR. For the commercial scale production of the fuel, statistical quality control and quality assurance must be carefully considered in order to assure the safety of the HTTR. It is also important to produce the fuel under well controlled process condition. To meet these requirements in the production of the HTTR fuel, a new production process and quality control system is to be introduced in the new facilities. The main feature of the system is a computer integrated control system. Process control data at each production stage of products and semi-products are all gathered by terminal computers and processed by a host computer. The processed information is effectively used for the production, quality and accountancy control. With the aid of this system, all the products will be easily traceable from starting materials to final stages and the statistical evaluation of the quality of products becomes more reliable. (author). 8 figs

  19. Analytical method and result of radiation exposure for depressurization accident of HTTR

    International Nuclear Information System (INIS)

    Sawa, K.; Shiozawa, S.; Mikami, H.

    1990-01-01

    The Japan Atomic Energy Research Institute (JAERI) is now proceeding with the construction design of the High Temperature Engineering Test Reactor (HTTR). Since the HTTR has some characteristics different from LWRs, analytical method of radiation exposure in accidents provided for LWRs can not be applied directly. This paper describes the analytical method of radiation exposure developed by JAERI for the depressurization accident, which is the severest accident in respect to radiation exposure among the design basis accidents of the HTTR. The result is also described in this paper

  20. Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2003-03-01

    Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH{sub 4} + H{sub 2}O {yields} 3H{sub 2}O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m{sup 3}{sub N}/h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

  1. Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio

    2003-03-01

    Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH 4 + H 2 O → 3H 2 O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m 3 N /h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

  2. Radiation monitoring for the HTTR rise-to-power test (1) and (2)'

    Energy Technology Data Exchange (ETDEWEB)

    Nakazawa, Takashi; Yoshino, Toshiaki; Yasu, Katsuji; Ashikagaya, Yoshinobu; Kikuchi, Toshiki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-02-01

    The High Temperature Engineering Test Reactor (HTTR) is the first high temperature gas-cooled research reactor in Japan. This reactor is a helium-gas-cooled and graphite-moderated reactor with a thermal output of 30 MW. The rated operation temperature of the outlet coolant is 850degC. (During high temperature test operation, this reaches 950degC). The first criticality of the HTTR was attained in November 1998. The single loaded, parallel loaded operation with a thermal output of 9 MW (called the HTTR Rise-to-Power Test (1)) was completed between September 16, 1999 and July 8, 2000. The single loaded, parallel loaded continuous operation with a thermal output of 20 MW (called the HTTR Rise-to-Power Test (2)) has also been carried out, but it was shutdown at the halfway stage by a single from the reactor, when the thermal output was 16.5 MW and the reactor outlet coolant temperature was 500degC. This report describes the radiation monitoring carried out during the HTTR Rise-to-Power Tests (1) and (2)'. The data measured by the various radiation monitors is also reported. These data will be used for the estimation of radiation levels (such as the radiation dose equivalent rate, the radioactive concentration in effluents, etc.) for the next HTTR Rise-to-Power Test, and for periodic inspections. (author)

  3. Simplified two and three dimensional HTTR benchmark problems

    International Nuclear Information System (INIS)

    Zhang Zhan; Rahnema, Farzad; Zhang Dingkang; Pounders, Justin M.; Ougouag, Abderrafi M.

    2011-01-01

    To assess the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of whole core configurations. In this paper we have created two and three dimensional numerical benchmark problems typical of high temperature gas cooled prismatic cores. Additionally, a single cell and single block benchmark problems are also included. These problems were derived from the HTTR start-up experiment. Since the primary utility of the benchmark problems is in code-to-code verification, minor details regarding geometry and material specification of the original experiment have been simplified while retaining the heterogeneity and the major physics properties of the core from a neutronics viewpoint. A six-group material (macroscopic) cross section library has been generated for the benchmark problems using the lattice depletion code HELIOS. Using this library, Monte Carlo solutions are presented for three configurations (all-rods-in, partially-controlled and all-rods-out) for both the 2D and 3D problems. These solutions include the core eigenvalues, the block (assembly) averaged fission densities, local peaking factors, the absorption densities in the burnable poison and control rods, and pin fission density distribution for selected blocks. Also included are the solutions for the single cell and single block problems.

  4. Conceptual design of the HTTR-IS hydrogen production system

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Sato, Hiroyuki; Hara, Teruo; Kato, Ryoma; Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko

    2007-08-01

    Since hydrogen produced by nuclear should be economically competitive compared with other methods in a hydrogen society, it is important to build hydrogen production system to be coupled with the reactor as a conventional chemical plant. Japan Atomic Energy Agency started the safety study to establish a new safety philosophy to meet safety requirements for non-nuclear grade hydrogen production system. Also, structural concepts with integrating functions for the Bunsen reactor and sulphuric acid decomposer were proposed to reduce construction cost of the IS process hydrogen production system. In addition, HI decomposer which enables the process condition to be eased consisting of conventional materials and technologies was studied. Moreover, technical feasibility of the HTTR-IS system in which the hydrogen production rate of 1,000 Nm 3 /h by using the supplied heat of 10 MW from the intermediate heat exchanger of the HTTR was confirmed. This paper describes the conceptual design of the HTTR-IS hydrogen production system. (author)

  5. Evaluation of hydrogen production system coupling with HTTR using dynamic analysis code

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Hayashi, Koji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency (JAEA) was entrusted 'Development of Nuclear Heat Utilization Technology' by Ministry of Education, Culture, Sports, Science and Technology. In this development, the JAEA investigated the system integration technology to couple the hydrogen production system by steam reforming with the High Temperature Engineering Test Reactor (HTTR). Prior to the construction of the hydrogen production system coupling with the HTTR, a dynamic analysis code had to be developed to evaluate the system transient behaviour of the hydrogen production system because there are no examples of chemical facilities coupled with nuclear reactor in the world. This report describes the evaluation of the hydrogen production system coupling with HTTR using analysis code, N-HYPAC, which can estimate transient behaviour of the hydrogen production system by steam reforming. The results of this investigation provide that the influence of the thermal disturbance caused by the hydrogen production system on the HTTR can be estimated well. (author)

  6. HTTR demonstration program for nuclear cogeneration of hydrogen and electricity

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Sumita, Junya; Terada, Atsuhiko; Ohashi, Hirofumi; Yan, Xing L.; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

    2015-01-01

    Japan Atomic Energy Agency initiated a High Temperature Engineering Test Reactor (HTTR) demonstration program in accordance with recommendations of a task force established by Ministry of Education, Culture, Sports, Science and Technology according to the Strategic Energy Plan as of April 2014. The demonstration program is designed to complete helium gas turbine and hydrogen production system technologies aiming at commercial plant deployment in 2030s. The program begins with coupling a helium gas turbine in the secondary loop of the HTTR and expands by adding the H 2 plant to a tertiary loop to enable hydrogen cogeneration. Safety standards for coupling the helium gas turbine and H 2 plant to the nuclear reactor will be established through safety review in licensing. A system design and its control method are planned to be validated with a series of test operations using the HTTR-GT/H 2 plant. This paper explains the outline of HTTR demonstration program with a plant concept of the heat application system directed at establishing an HTGR cogeneration system with 950°C reactor outlet temperature for production of power and hydrogen as recommended by the task force. Commercial deployment strategy including a development plan for the helium gas turbine is also presented. (author)

  7. Innovative and basic researches for high temperature technologies at HTTR

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku

    1995-01-01

    The HTTR is the first HTGR which is under construction at JAERI. The objectives of the HTTR are to establish basic technologies for HTGRs, to upgrade technologies for HTGRs and to conduct innovative and basic researches for high temperature technologies. The first two are concerned with HTGR developments. The last one is not necessarily for HTGR developments, but for future innovative researches which are expected to be applied to various technologies. (author)

  8. Out-of-pile demonstration test of HTTR hydrogen production system structure and fabrication technology of steam reformer. Contract research

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Ouchi, Yoshihiro; Fujisaki, Katsuo; Kato, Michio; Uno, Hisao; Hayashi, Koji; Aita, Hideki

    1999-10-01

    A hydrogen production system by steam reforming of natural gas, chemical reaction; CH 4 +H 2 O = 3H 2 +CO, is to be the first heat utilization system of the HTTR. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test facility is presently under construction in order to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility, using an electric heater as a reactor substitute, simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 with a hydrogen production rate of 110 Nm 3 /h. A steam reformer (SR) is a key component to produce hydrogen by steam reforming of natural gas. A bayonet-type catalyst tube was applied to the SR of the out-of-pile test facility in order to enhance the heat utilization rate. Also to promote heat transfer, the thickness of the catalyst tube should be decreased to 10 mm while augmenting heat transfer by fins formed on the outer surface of the catalyst tube. Therefore, the catalyst tube was designed on the basis of pressure difference between helium and process gases instead of total pressure of them. This design method was authorized for the first time in Japan. Furthermore, a function of explosion proof was applied to the SR because it contains inflammable gas and electric heater. This report describes the structure of the SR as well as the authorization both of the design method of the catalyst tube and the explosion proof function of the SR. (author)

  9. Analytical estimation of control rod shadowing effect for excess reactivity measurement of HTTR

    International Nuclear Information System (INIS)

    Nakano, Masaaki; Fujimoto, Nozomu; Yamashita, Kiyonobu

    1999-01-01

    The fuel addition method is generally used for the excess reactivity measurement of the initial core. The control rod shadowing effect for the excess reactivity measurement has been estimated analytically for High Temperature Engineering Test Reactor (HTTR). 3-dimensional whole core analyses were carried out. The movements of control rods in measurements were simulated in the calculation. It was made clear that the value of excess reactivity strongly depend on combinations of measuring control rods and compensating control rods. The differences in excess reactivity between combinations come from the control rod shadowing effect. The shadowing effect is reduced by the use of plural number of measuring and compensating control rods to prevent deep insertion of them into the core. The measured excess reactivity in the experiments is, however, smaller than the estimated value with shadowing effect. (author)

  10. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

    International Nuclear Information System (INIS)

    2016-02-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

  11. Performance test results of mock-up test facility of HTTR hydrogen production system

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo

    2004-01-01

    For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m 3 /h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004. (author)

  12. Replaceable LMFBR core components

    International Nuclear Information System (INIS)

    Evans, E.A.; Cunningham, G.W.

    1976-01-01

    Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

  13. Development of control technology for HTTR hydrogen production system with mock-up test facility

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data

  14. Out-of-pile demonstration test of HTTR hydrogen production system structure and fabrication technology of steam reformer. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Ouchi, Yoshihiro; Fujisaki, Katsuo; Kato, Michio; Uno, Hisao; Hayashi, Koji; Aita, Hideki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1999-10-01

    A hydrogen production system by steam reforming of natural gas, chemical reaction; CH{sub 4}+H{sub 2}O = 3H{sub 2}+CO, is to be the first heat utilization system of the HTTR. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test facility is presently under construction in order to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility, using an electric heater as a reactor substitute, simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 with a hydrogen production rate of 110 Nm{sup 3}/h. A steam reformer (SR) is a key component to produce hydrogen by steam reforming of natural gas. A bayonet-type catalyst tube was applied to the SR of the out-of-pile test facility in order to enhance the heat utilization rate. Also to promote heat transfer, the thickness of the catalyst tube should be decreased to 10 mm while augmenting heat transfer by fins formed on the outer surface of the catalyst tube. Therefore, the catalyst tube was designed on the basis of pressure difference between helium and process gases instead of total pressure of them. This design method was authorized for the first time in Japan. Furthermore, a function of explosion proof was applied to the SR because it contains inflammable gas and electric heater. This report describes the structure of the SR as well as the authorization both of the design method of the catalyst tube and the explosion proof function of the SR. (author)

  15. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

    International Nuclear Information System (INIS)

    2014-12-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

  16. HTTR workshop (workshop on hydrogen production technology)

    International Nuclear Information System (INIS)

    Shiina, Yasuaki; Takizuka, Takakazu

    2004-12-01

    Various research and development efforts have been performed to solve the global energy and environmental problems caused by large consumption of fossil fuels. Research activities on advanced hydrogen production technology by the use of nuclear heat from high temperature gas cooled reactors, for example, have been flourished in universities, research institutes and companies in many countries. The Department of HTTR Project and the Department of Advanced Nuclear Heat Technology of JAERI held the HTTR Workshop (Workshop on Hydrogen Production Technology) on July 5 and 6, 2004 to grasp the present status of R and D about the technology of HTGR and the nuclear hydrogen production in the world and to discuss about necessity of the nuclear hydrogen production and technical problems for the future development of the technology. More than 110 participants attended the Workshop including foreign participants from USA, France, Korea, Germany, Canada and United Kingdom. In the Workshop, the presentations were made on such topics as R and D programs for nuclear energy and hydrogen production technologies by thermo-chemical or other processes. Also, the possibility of the nuclear hydrogen production in the future society was discussed. The workshop showed that the R and D for the hydrogen production by the thermo-chemical process has been performed in many countries. The workshop affirmed that nuclear hydrogen production could be one of the competitive supplier of hydrogen in the future. The second HTTR Workshop will be held in the autumn next year. (author)

  17. Evaluation of hot spot factors for thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    Maruyama, So; Yamashita, Kiyonobu; Fujimoto, Nozomu; Murata, Isao; Sudo, Yukio; Murakami, Tomoyuki; Fujii, Sadao.

    1993-01-01

    High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950degC in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly. (author)

  18. Assessment of damage domains of the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Flores, Alain; Izquierdo, José María; Tuček, Kamil; Gallego, Eduardo

    2014-01-01

    Highlights: • We developed an adequate model for the identification of damage domains of the HTTR. • We analysed an anticipated operational transient, using the HTTR5+/GASTEMP code. • We simulated several transients of the same sequence. • We identified the corresponding damage domains using two methods. • We calculated exceedance frequency using the two methods. - Abstract: This paper presents an assessment analysis of damage domains of the 30 MW th prototype High-Temperature Engineering Test Reactor (HTTR) operated by the Japan Atomic Energy Agency (JAEA). For this purpose, an in-house deterministic risk assessment computational tool was developed based on the Theory of Stimulated Dynamics (TSD). To illustrate the methodology and applicability of the developed modelling approach, assessment results of a control rod (CR) withdrawal accident during subcritical conditions are presented and compared with those obtained by the JAEA

  19. Nondestructive testing on graphite structures for high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Kambe, Mamoru; Tsuji, Nobumasa.

    1994-01-01

    The application of ultrasonic (for internal defects) and eddy current testing (for surface defects) were investigated on the structures of nuclear-grade IG-110 and PGX graphite for the HTTR. The equipment were developed in order to detect the specific configuration of graphite blocks and the testing conditions were defined as the practical testing methods. The established testing methods are being used for the acceptance tests of graphite structures in the HTTR. (author)

  20. Determination of the temperature coefficients and the kinetic parameters for the HTTR safety analysis

    International Nuclear Information System (INIS)

    Tokuhara, K.; Nakata, T.; Murata, I.; Yamashita, K.; Shindo, R.

    1991-01-01

    This report describes the calculational methods which were employed to determine the temperature coefficients and the kinetic parameters for the safety analysis in the HTTR (High Temperature Engineering Test Reactor). The temperature coefficients (doppler, moderator temperature) and the kinetic parameters (prompt neutron life time; l, effective delayed neutron fraction; β eff) are important for the point model core dynamic analysis and should be evaluated properly. The temperature coefficients were calculated by the whole core model. Doppler coefficient was evaluated under the conditions of all control rods withdrawn and the uniform change of fuel temperature. The minimum and the maximum value of the evaluated doppler coefficients in a burnup cycle are -4.6x10 -5 and -1.5x10 -5 ΔK/K/deg. C respectively. The moderator temperature coefficient was evaluated under the conditions of all control rods withdrawn and the uniform change of moderator temperature. The minimum and the maximum value of the evaluated moderator temperature coefficients in a burnup cycle are -17.1x10 -5 and 0.99x10 -5 ΔK/K/deg. C respectively. In spite of positive moderator temperature coefficient, the power coefficient is always negative. Therefore the HTTR possesses inherent power-suppressing feed back characteristic in all operating condition. We surveyed the effects of the Xe existence, the control rods existence, the fuel temperature and the region in which the temperature was changed on the moderator temperature coefficients. The kinetic parameters were calculated by the perturbation method with the whole core model. The minimum and the maximum value of the evaluated effective delayed neutron fraction (β eff) are 0.0047 and 0.0065 respectively. These of the evaluated prompt neutron life time (l) are 0.67 and 0.78 ms respectively. We have surveyed the effects of the fuel depletion and the core power level on these parameters, and considered these effects on the kinetic parameters. From

  1. Eddy current testing on structures of nuclear-grade IG-110 graphite for acceptance test in HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Saikusa, Akio; Iyoku, Tatsuo

    1993-09-01

    Core and core support graphite structures in the HTTR are mainly made of IG-110 graphite which is fine-grained isotropic and nuclear-grade. Nondestructive inspection with eddy current testing is planned to be applied to these graphite structures. Eddy current testing is widely applied to metallic structures and its testing method has been already established. On the other hand, the characteristics of graphite are quite different in micro-structure from these of metals. Therefore, the eddy current testing method provided for metallic structures can not be applied directly to graphite structures. Thus the eddy current testing method and condition were established for the graphite structures made of IG-110 graphite. (author)

  2. Evaluation on materials performance of Hastelloy Alloy XR for HTTR uses-5 (Creep properties of base metal and weldment in air)

    International Nuclear Information System (INIS)

    Watanabe, Katsutoshi; Nakajima, Hajime; Koikegami, Hajime; Higuchi, Makoto; Nakanishi, Tsuneo; Saitoh, Teiichiro; Takatsu, Tamao.

    1994-01-01

    Creep properties of weldment made from Hastelloy Alloy XR base metals and filler metals for the High Temperature Engineering Test Reactor (HTTR) components were examined by means of creep and creep rupture tests at 900 and 950degC in air. The results obtained are as follows: creep rupture strength was nearly equal or higher than that of Hastelloy Alloy XR master curve and was much higher than design creep rupture strength [S R ]. Furthermore, creep rupture strength and ductility of the present filler metal was in the data band in comparison with those of the previous filler metals. It is concluded from these reasons that this filler metal has fully favorable properties for HTTR uses. (author)

  3. Hydrogen production system coupled with high-temperature gas-cooled reactor (HTTR)

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku

    2003-01-01

    On the HTTR program, R and D on nuclear reactor technology and R and D on thermal application technology such as hydrogen production and so on, are advanced. When carrying out power generation and thermal application such as hydrogen production and so on, it is, at first, necessary to supply nuclear heat safely, stably and in low cost, JAERI carries out some R and Ds on nuclear reactor technology using HTTR. In parallel to this, JAERI also carries out R and D for jointing nuclear reactor system with thermal application systems because of no experience in the world on high temperature heat of about 1,000 centigrade supplied by nuclear reactor except power generation, and R and D on thermochemical decomposition method IS process for producing hydrogen from water without exhaust of carbon dioxide. Here were described summaries on R and D on nuclear reactor technology, R and D on jointing technology using HTTR hydrogen production system, R and D on IS process hydrogen production, and comparison hydrogen production with other processes. (G.K.)

  4. Design and evaluation of heat utilization systems for the HTTR through international cooperation

    Energy Technology Data Exchange (ETDEWEB)

    Lewkowicz, I. [International Atomic Energy Agency, Vienna (Austria)

    1996-07-01

    The International Atomic Energy Agency (IAEA) has the statutory function to `foster the exchange of scientific and technical information`, and `encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world`. The IAEA Co-ordinated Research Programmes (CRPs) are effective vehicles for implementing the above. The CRP on Design and Evaluation of Heat Utilization Systems for HTTR has started in September 1994 and is aimed at promoting international co-operation to identify the most promising heat utilization system(s) to be demonstrated at the HTTR, for the benefit of current operators and future designers and constructors of HTGRs. Participating Member States are collaborating by exchanging existing technical information on the technology of heat utilization systems, by developing design concepts and by performing evaluations of candidate systems for potential demonstration with the HTTR. In this report, the systems are reviewed. (J.P.N.)

  5. Design and evaluation of heat utilization systems for the HTTR through international cooperation

    International Nuclear Information System (INIS)

    Lewkowicz, I.

    1996-01-01

    The International Atomic Energy Agency (IAEA) has the statutory function to 'foster the exchange of scientific and technical information', and 'encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world'. The IAEA Co-ordinated Research Programmes (CRPs) are effective vehicles for implementing the above. The CRP on Design and Evaluation of Heat Utilization Systems for HTTR has started in September 1994 and is aimed at promoting international co-operation to identify the most promising heat utilization system(s) to be demonstrated at the HTTR, for the benefit of current operators and future designers and constructors of HTGRs. Participating Member States are collaborating by exchanging existing technical information on the technology of heat utilization systems, by developing design concepts and by performing evaluations of candidate systems for potential demonstration with the HTTR. In this report, the systems are reviewed. (J.P.N.)

  6. Investigation on structural integrity of graphite component during high temperature 950degC continuous operation of HTTR

    International Nuclear Information System (INIS)

    Sumita, Junya; Shimazaki, Yosuke; Shibata, Taiju

    2014-01-01

    Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950degC continuous operation, a high temperature continuous operation with reactor outlet temperature of 950degC for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950degC continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000degC. In addition, the programs of surveillance test and ISI using TV camera were introduced. (author)

  7. Investigation of the loss of forced cooling test by using the high temperature engineering test reactor (HTTR) (Contract research)

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Inaba, Yoshitomo; Goto, Minoru; Tochio, Daisuke

    2007-09-01

    The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur. (author)

  8. State of Art Report for the OECD-NEA Loss-of-Forced Cooling (LOFC) Test Project using HTTR Reactor

    International Nuclear Information System (INIS)

    Jun, Ji Su

    2011-05-01

    The OECD/NEA Project is planned to perform the LOFC (Loss Of Forced Cooling) test using the HTTR (High Temperature engineering Test Reactor) in Japan from 31 March 2011 to 31 March 2013 in order to obtain the data for the code validation of the VHTR safety analysis. Based on the Project Agreement Document, this report gives a description of the HTTR-LOFC test, HTTR test facility, project schedule and deliverable items as the technical state art of the project, and appends the full translation of the project agreement articles on the project management

  9. Prospects of HTGR process heat application and role of HTTR

    International Nuclear Information System (INIS)

    Shiozawa, S.; Miyamoto, Y.

    2000-01-01

    At Japan Atomic Energy Research Institute, an effort on development of process heat application with high temperature gas cooled reactor (HTGR) has been continued for providing a future clean alternative to the burning of fossil energy for the production of industrial process heat. The project is named 'HTTR Heat Utilization Project', which includes a demonstration of hydrogen production using the first Japanese HTGR of High Temperature Engineering Test Reactor (HTTR). In the meantime, some countries, such as China, Indonesia, Russia and South Africa are trying to explore the HTGR process heat application for industrial use. One of the key issues for this application is economy. It has been recognized for a long time and still now that the HTGR heat application system is not economically competitive to the current fossil ones, because of the high cost of the HTGR itself. However, the recent movement on the HTGR development, as represented by South Africa Pebble Beds Modular Reactor (SA-PBMR) Project, has revealed that the HTGRs are well economically competitive in electricity production to fossil fuel energy supply under a certain condition. This suggests that the HTGR process heat application will also possibly get economical in the near future. In the present paper, following a brief introduction describing the necessity of the HTGRs for the future process heat application, Japanese activities and prospect of the development on the process heat application with the HTGRs are described in relation with the HTTR Project. In conclusion, the process heat application system with HTGRs is thought technically and economically to be one of the most promising applications to solve the global environmental issues and energy shortage which may happen in the future. However, the commercialization for the hydrogen production system from water, which is the final goal of the HTGR process heat application, must await the technology development to be completed in 2030's at the

  10. Case study on chemical plant accidents for flow-sheet design of the HTTR-IS system

    International Nuclear Information System (INIS)

    Homma, Hiroyuki; Sato, Hiroyuki; Kasahara, Seiji; Hara, Teruo; Kato, Ryoma; Sakaba, Nariaki; Ohashi, Hirofumi

    2007-02-01

    At the present time, we are alarmed by depletion of fossil energy and adverse effect of rapid increase in fossil fuel burning on environment such as climate changes and acid rain, because our lives depend still heavily upon fossil energy. It is thus widely recognized that hydrogen is one of important future energy carriers in which it is used without emission of carbon dioxide greenhouse gas and atmospheric pollutants and that hydrogen demand will increase greatly as fuel cells are developed and applied widely in the near future. To meet massive demand of hydrogen, hydrogen production from water utilizing nuclear, especially by thermochemical water-splitting Iodine-Sulphur (IS) process utilizing heat from High-Temperature Gas-cooled Reactors (HTGRs), offers one of the most attractive zero-emission energy strategies and the only one practical on a substantial scale. However, to establish a technology based for the HTGR hydrogen production by the IS process, we should close several technology gaps through R and D with the High-Temperature Engineering Test Reactor (HTTR), which is the only Japanese HTGR built and operated at the Oarai Research and Development Centre of Japan Atomic Energy Agency (JAEA). We have launched design studies of the IS process hydrogen production system coupled with the HTTR (HTTR-IS system) to demonstrate HTGR hydrogen production. In designing the HTTR-IS system, it is necessary to consider preventive and breakdown maintenance against accidents occurred in the IS process as a chemical plant. This report describes case study on chemical plant accidents relating to the IS process plant and shows a proposal of accident protection measures based on above case study, which is necessary for flow-sheet design of the HTTR-IS system. (author)

  11. Operation, test, research and development of the high temperature engineering test reactor (HTTR). (FY2005)

    International Nuclear Information System (INIS)

    2007-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power. The full power operation of 30 MW was attained in December, 2001, and then JAERI (JAEA) received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. In fiscal 2005 year, periodical inspection and overhaul of reactivity control system were conducted, and safety demonstration tests were promoted. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2005. (author)

  12. Investigation on cause of outage of Wide Range Monitor (WRM) in High Temperature engineering Test Reactor (HTTR). Transport operation toward investigation for cause of outage

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Sawahata, Hiroaki; Kawamoto, Taiki; Saito, Kenji; Takada, Shoji; Yoshida, Naoaki; Isozaki, Ryosuke; Katsuyama, Kozo; Motegi, Toshihiro

    2012-08-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, a post irradiation examination was planned to specify the damaged part causing the event in the WRM was also planned. For the investigation, the X-ray computed tomography scanner in Fuels Monitoring Facility (FMF). This report describes the preliminary investigation on the cause of outage of the WRM. The results of study for transportation method of the irradiated WRM from HTTR to FMF is also reported with the record to complete the transport operation. (author)

  13. Improvement in oil seal performance of gas compressor in HTTR

    International Nuclear Information System (INIS)

    Oyama, Sunao; Hamamoto, Shimpei; Nemoto, Takahiro; Sekita, Kenji; Isozaki, Minoru; Emori, Kouichi; Ohta, Yukimaru; Mizushima, Toshihiko; Kaneshiro, Noriyuki; Ito, Yoshiteru; Yamamoto, Hideo

    2007-08-01

    High-Temperature engineering Test Reactor (HTTR) built by Japan Atomic Energy Agency (JAEA) has reciprocating compressor commonly used to extract and discharge helium gas into primary/secondary coolant helium loop from helium purification system. Piston rod seal of the compressor consist of several components to prevent coolant leak. However, rod seal system has weak reliability during long term operation due to repeated leakage of seal oil in operation. As a result of investigations, leakage's root is found in that seal were used in a range beyond limit sliding properties of seal material. For this reason, a lip of the seal was worn and transformed itself and was not able to sustain a seal function. Therefore, through tests using facility actual equipment for endurance of candidate materials, one seal material were chosen for long term operation. (author)

  14. Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Shindo, R.; Yamashita, K.; Murata, I.

    1991-01-01

    The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

  15. Fundamental philosophy on the safety design of the HTTR-IS hydrogen production system

    International Nuclear Information System (INIS)

    Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko

    2007-01-01

    Japan Atomic Energy Agency (JAEA) has been conducting an R and D work on the VHTR reactor system and IS hydrogen production system to realize hydrogen production using nuclear heat. As a part of this activity, JAEA is planning to connect an IS test system to the High Temperature Engineering Test Reactor (HTTR) to demonstrate its technical feasibility. This paper proposes a fundamental philosophy on the safety design of the HTTR-IS hydrogen production system including the methodology to select postulated abnormal events and its event sequences and to define safety functions of the IS system to ensure the reactor safety. Also the measure to clarify the IS system as non-reactor system is proposed. (author)

  16. Surveillance test of the JMTR core components

    International Nuclear Information System (INIS)

    Takeda, Takashi; Amezawa, Hiroo; Tobita, Kenji

    1986-02-01

    Surveillance test for the core components of Japan Materials Testing Reactor (JMTR) was started in 1966, and completed in 1985 without one capsule. Most of capsules in the program, except one beryllium specimens, were removed from the core, and carred out the post-irradiation tests at the JMTR Hot Laboratory. The data is applied to review of JMTR core components management plan. JMTR surveillance test was carried out with several kind of materials of JMTR core components, Berylium as the reflector, Hafnium as the neutron absorber of control rod, 17-4PH stainless steel as a roller spring of the control rod, and 304 stainless steel as the grid plate. Results are described in this report. (author)

  17. Test results of HTTR control system

    International Nuclear Information System (INIS)

    Motegi, Toshihiro; Iigaki, Kazuhiko; Saito, Kenji; Sawahata, Hiroaki; Hirato, Yoji; Kondo, Makoto; Shibutani, Hideki; Ogawa, Satoru; Shinozaki, Masayuki; Mizushima, Toshihiko; Kawasaki, Kozo

    2006-06-01

    The plant control performance of the IHX helium flow rate control system, the PPWC helium flow rate control system, the secondary helium flow rate control system, the inlet temperature control system, the reactor power control system and the outlet temperature control system of the HTTR are obtained through function tests and power-up tests. As the test results, the control systems show stable control response under transient condition. Both of inlet temperature control system and reactor power control system shows stable operation from 30% to 100%, respectively. This report describes the outline of control systems and test results. (author)

  18. Plant data evaluation of performance confirmation test in HTTR after Tohoku-Pacific Ocean Earthquake

    International Nuclear Information System (INIS)

    Ono, Masato; Tochio, Daisuke; Shinohara, Masanori; Shimazaki, Yosuke; Yanagi, Shunki; Iigaki, Kazuhiko

    2012-03-01

    Tohoku-Pacific Ocean Earthquake occurred on March 11th 2011 and the earthquake intensity of an upper 5 on the Japanese scale was observed in Oarai town. HTTR conducted the confirmation test on cold state in order to ensure the facilities/instruments of reactor building operate normally. In this test, the plant data in the facilities/instruments start-up phase and continue steady operation phase were measured and compared with the previous operation data, and the soundness of facilities/instruments is evaluated. As a result, in after the earthquake, the facilities/instruments operate normally and the reactor cooling function of the HTTR were ensured. (author)

  19. Plant data evaluation of performance confirmation test in HTTR after Tohoku-Pacific Ocean Earthquake

    International Nuclear Information System (INIS)

    Ono, Masato; Tochio, Daisuke; Shinohara, Masanori; Shimazaki, Yosuke; Yanagi, Shunki; Iigaki, Kazuhiko

    2012-01-01

    Tohoku-Pacific Ocean Earthquake occurred on March 11th 2011 and the earthquake intensity of an upper 5 on the Japanese scale was observed in Oarai town. HTTR conducted the confirmation test on cold state in order to ensure the facilities/instruments of reactor building operate normally. In this test, the plant data in the facilities/instruments start-up phase and continue steady operation phase were measured and compared with the previous operation data, and the soundness of facilities/instruments is evaluated. As a result, in after the earthquake, the facilities/instruments operate normally and the reactor cooling function of the HTTR were ensured. (author)

  20. Principle design and data of graphite components

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo

    2004-01-01

    The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned

  1. Evaluation of heat exchange performance for primary pressurized water cooler in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Nakagawa, Shigeaki

    2006-01-01

    In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30 MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the primary pressurized water cooler (PPWC) and the intermediate heat exchanger (IHX). The heat exchangers in the primary cooling system are required the heat exchange performance to remove reactor generated heat 30 MW under the condition of reactor coolant outlet temperature 850degC/950degC. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the PPWC in the main cooling system was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that PPWC has the required heat exchange performance in the design. (author)

  2. Evaluation of heat exchange performance for secondary pressurized water cooler in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Syuji; Saikusa, Akio; Oyama, Sunao; Nemoto, Takahiro; Hamamoto, Shinpei; Shinohara, Masanori; Isozaki, Minoru; Nakagawa, Shigeaki

    2006-02-01

    In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the intermediate heat exchanger (IHX) and the secondary pressurized water cooler (SPWC). The heat exchangers in the main cooling system are required the heat exchange performance to remove the reactor-generated-heat of 30MW under the condition of reactor coolant outlet temperature of 850degC/950degC. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance of the SPWC in the main cooling system was evaluated with the rise-to-power-up test and the in-service operation data. Moreover, evaluated value is compared with designed one, it is confirmed that the SPWC has required heat exchange performance. (author)

  3. HTTR operation monitoring with neural network in 30 days operation at 850degC

    International Nuclear Information System (INIS)

    Shimizu, Atsushi; Nabeshima, Kunihiko; Nakagawa, Shigeaki

    2009-01-01

    The High temperature engineering test reactor (HTTR) executed the rated power operation for 30days of the first time (850degC in temperature of the nuclear reactor outlet coolant) until March, 27th through April, 26th, 2007. In this operation, HTTR was observed according to the operation monitoring model with the neural network, and the detection performance of neural network was verified during slight changes of reactor state at rated power. The neural network used for the operation monitoring was an auto-associative network, where 31 input 31 outputs and the hidden layers were connected with 20 units by the hierarchy of three layer structure. Back-propagation algorithm was used for study rule. The operation monitoring model in initial study was constructed by using the power up data between 30% and rated power, which were randomly studied. The adjustment study during the operation monitoring changes the internal structure of the initial study model to follow the changes of reactor status, such as the burn-up of the nuclear fuel for the rated power operation. As a monitoring result, slight changes of reactor state by the control system operation were correctly detected, and the on-line application to an early anomaly diagnosis for HTTR facilities will be expected. (author)

  4. Verification of thermal-irradiation stress analytical code VIENUS of graphite block

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Shiozawa, Shusaku; Shirai, Hiroshi; Minato, Kazuo.

    1992-02-01

    The core graphite components of the High Temperature Engineering Test Reactor (HTTR) show both the dimensional change (irradiation shrinkage) and creep behavior due to fast neutron irradiation under the temperature and the fast neutron irradiation conditions of the HTTR. Therefore, thermal/irradiation stress analytical code, VIENUS, which treats these graphite irradiation behavior, is to be employed in order to design the core components such as fuel block etc. of the HTTR. The VIENUS is a two dimensional finite element viscoelastic stress analytical code to take account of changes in mechanical properties, thermal strain, irradiation-induced dimensional change and creep in the fast neutron irradiation environment. Verification analyses were carried out in order to prove the validity of this code based on the irradiation tests of the 8th OGL-1 fuel assembly and the fuel element of the Peach Bottom reactor. This report describes the outline of the VIENUS code and its verification analyses. (author)

  5. Demonstration of inherent safety features of HTGRs using the HTTR

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo

    2004-01-01

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) quantitatively as well as providing the core and plant transient data for validation of HTGR analysis codes for safety evaluation. The safety demonstration test are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase test simulating reactivity insertion events and coolant flow reduction events stared in FY 2002. Post-test analyses have been conducted to reproduced the test results by using the core and plant dynamics analysis code, ACCORD and Monte Carlo code, MVP. The analysis results agreed fairly well with the test results of a control rod withdrawal test simulating reactivity insertion, and gas circulators trip test simulating coolant flow reduction, at power levels of 50% and 30% of the rated power, respectively. It is shown that improvement of the ACCORD code by taking into consideration vertical and horizontal temperature distribution gives better analysis results in the control rod withdrawal test. The fist phase safety demonstration tests will continue until FY 2005, and the second phase tests are planned to be started in FY 2006. (author)

  6. PHISICS/RELAP5-3D Adaptive Time-Step Method Demonstrated for the HTTR LOFC#1 Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Baker, Robin Ivey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Balestra, Paolo [Univ. of Rome (Italy); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-05-01

    A collaborative effort between Japan Atomic Energy Agency (JAEA) and Idaho National Laboratory (INL) as part of the Civil Nuclear Energy Working Group is underway to model the high temperature engineering test reactor (HTTR) loss of forced cooling (LOFC) transient that was performed in December 2010. The coupled version of RELAP5-3D, a thermal fluids code, and PHISICS, a neutronics code, were used to model the transient. The focus of this report is to summarize the changes made to the PHISICS-RELAP5-3D code for implementing an adaptive time step methodology into the code for the first time, and to test it using the full HTTR PHISICS/RELAP5-3D model developed by JAEA and INL and the LOFC simulation. Various adaptive schemes are available based on flux or power convergence criteria that allow significantly larger time steps to be taken by the neutronics module. The report includes a description of the HTTR and the associated PHISICS/RELAP5-3D model test results as well as the University of Rome sub-contractor report documenting the adaptive time step theory and methodology implemented in PHISICS/RELAP5-3D. Two versions of the HTTR model were tested using 8 and 26 energy groups. It was found that most of the new adaptive methods lead to significant improvements in the LOFC simulation time required without significant accuracy penalties in the prediction of the fission power and the fuel temperature. In the best performing 8 group model scenarios, a LOFC simulation of 20 hours could be completed in real-time, or even less than real-time, compared with the previous version of the code that completed the same transient 3-8 times slower than real-time. A few of the user choice combinations between the methodologies available and the tolerance settings did however result in unacceptably high errors or insignificant gains in simulation time. The study is concluded with recommendations on which methods to use for this HTTR model. An important caveat is that these findings

  7. Full integrated system of real-time monitoring based on distributed architecture for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Subekti, Muhammad; Ohno, Tomio; Kudo, Kazuhiko; Takamatsu, Kuniyoshi; Nabeshima, Kunihiko

    2005-01-01

    A new monitoring system scheme based on distributed architecture for the High Temperature Engineering Test Reactor (HTTR) is proposed to assure consistency of the real-time process of expanded system. A distributed monitoring task on client PCs as an alternative architecture maximizes the throughput and capabilities of the system even if the monitoring tasks suffer a shortage of bandwidth. The prototype of the on-line monitoring system has been developed successfully and will be tested at the actual HTTR site. (author)

  8. Outline of operation and control system and analytical investigation of transient behavior of an out-of-pile hydrogen production system for HTTR heat utilization system

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Hada, Kazuhiko; Nishihara, Tetsuo; Takeda, Tetsuaki; Haga, Katsuhiro; Hino, Ryutaro.

    1997-10-01

    The hydrogen production system by steam reforming of natural gas is to be constructed to demonstrate effectiveness of high-temperature nuclear heat utilization systems with the HTTR. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test system is planned to investigate the system characteristics, to develop high-temperature components such as a reformer, a high-temperature isolation valve and so on, and to verify operation and control technologies and safety technology at accidents. This paper presents outline of operation and control systems and analytical review of transient behavior of the out-of-pile hydrogen production system. Main function of the operation and control systems is made not to give disturbance to the HTTR at transient state under start-up and stop operations. The operation modes are separated into two ones, namely normal and accident operation modes, and operation sequences are made for each operation mode. The normal operation sequence includes start-up, steady operation and stop of the out-of-pile system. The accident one deals with accident conditions at which supply of feed gas is stopped and helium gas is cooled passively by the steam generator. Transient behavior of the out-of-pile system was analyzed numerically according as the operation sequences. As the results, it was confirmed that the designed operation and control systems are adequate to the out-of-pile system. (author)

  9. Evaluation for the models of neutron diffusion theory in terms of power density distributions of the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Shimakawa, Satoshi; Nojiri, Naoki; Fujimoto, Nozomu

    2003-10-01

    In the case of evaluations for the highest temperature of the fuels in the HTTR, it is very important to expect the power density distributions accurately; therefore, it is necessary to improve the analytical model with the neutron diffusion and the burn-up theory. The power density distributions are analyzed in terms of two models, the one mixing the fuels and the burnable poisons homogeneously and the other modeling them heterogeneously. Moreover these analytical power density distributions are compared with the ones derived from the gross gamma-ray measurements and the Monte Carlo calculational code with continuous energy. As a result the homogeneous mixed model isn't enough to expect the power density distributions of the core in the axial direction; on the other hand, the heterogeneous model improves the accuracy. (author)

  10. Under sodium reliability tests on core components and in-core instrumentation

    International Nuclear Information System (INIS)

    Ruppert, E.; Stehle, H.; Vinzens, K.

    1977-01-01

    A sodium test facility for fast breeder core components (AKB), built by INTERATOM at Bensberg, has been operating since 1971 to test fuel dummies and blanket elements as well as absorber elements under simulated normal and extreme reactor conditions. Individual full-scale fuel or blanket elements and arrays of seven elements, modelling a section of the SNR-300 reactor core, have been tested under a wide range of sodium mass flow and isothermal test conditions up to 925K as well as under cyclic changed temperature transients. Besides endurance testing of the core components a special sodium and high-temperature instrumentation is provided to investigate thermohydraulic and vibrational behaviour of the test objects. During all test periods the main subassembly characteristics could be reproduced and the reliability of the instrumentation could be proven. (orig.) [de

  11. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  12. The HTTR project as the world leader of HTGR research and development

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku; Komori, Yoshihiro; Ogawa, Masuro

    2005-01-01

    As a next generation type nuclear system which will expand nuclear energy use area with high temperature nuclear heat utilization and improve economic competitiveness greatly, High Temperature Gas-cooled Reactor (HTGR) has become the R and D item of prime importance at home as well as abroad to establish hydrogen society to cope with global environmental problems. JAERI has conducted R and D on HTGR as the world leader such as to achieve a reactor outlet coolant temperature of 950 degC in the HTTR (High Temperature Engineering Test Reactor) in April 2004 as the world's first and also to succeed in continuous hydrogen production with a bench-scale apparatus of closed cycle iodine-sulfur (IS) process for six and half hours in August 2003 as the world's first. Overview and present status of HTTR program were presented in details with background and main R and D results as well as international trend of HTGR development and future program on pilot tests facilities for hydrogen production demonstration in Japan. (T. Tanaka)

  13. Construction of In-core Structure Test Section in HENDEL, (1)

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Inagaki, Yoshiyuki; Ioka, Ikuo; Kondoh, Yasuo; Nekoya, Shinichi; Miyamoto, Yoshiaki; Akisada, Toshihiro; Yamaguchi, Shigeru.

    1988-01-01

    An In-core Structure Test Section (T 2 ) in Helium Engineering Demonstration Loop (HENDEL) simulates a part of the core bottom structure with the same scale as that of a high temperature engineering test reactor (HTTR) designed in Japan Atomic Energy Research Institute. The design and construction of T 2 test section were started in March 1983, and completed in June 1986. The main objectives of the T 2 test section are to verify thermal-hydraulic performance and integrity of the core bottom structure. The report describes the general outline of T 2 test section, and experience gained from construction and preliminary test with regard to the simulated core bottom structure. (author)

  14. Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

    2003-01-01

    The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

  15. Summary of facility and operating experience on helium engineering demonstration loop (HENDEL)

    Energy Technology Data Exchange (ETDEWEB)

    Ouchi, Yoshihiro; Fujisaki, Katsuo; Kobayashi, Toshiaki; Kato, Michio; Ota, Yukimaru; Watanabe, Syuji; Kobayashi, Hideki; Mogi, Haruyoshi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1996-07-01

    The HENDEL is a test facility to perform full scale demonstration tests on the core internals and high temperature components for the High Temperature Engineering Test Reactor(HTTR). The main systems consist of Mother(M) and Adapter(A), fuel stack Test(T{sub 1}) and in-core structure Test(T{sub 2}) sections. The (M+A) section can supply high temperature helium gas to the test section. The M+A section completed in March 1982 has been operated for about 22900 hours till February 1995. The T{sub 1} and T{sub 2} sections, completed in March 1983 and June 1986, have been operated for about 19400 and 16700 hours, respectively. In this period, a large number of tests have been conducted to verify the performance and safety features of the HTTR components. The results obtained from these tests have been effectively applied to the detailed design, licensing procedures and construction of the HTTR. The operating experience of the HENDEL for more than 10 years also brought us establishment of the technique of operation of a large scale helium gas loop, handling of helium gas and maintenance of high temperature facilities. The technique will be available for the operation of the HTTR. This paper mainly describes the summary of plant facirities, operating experience and maintenance on the HENDEL. (author)

  16. Japanese HTTR program for demonstration of high temperature applications of nuclear energy

    International Nuclear Information System (INIS)

    Nishihara, T.; Hada, K.; Shiozawa, S.

    1997-01-01

    Construction works of the HTTR started in March 1991 in order to establish and upgrade the HTGR technology basis, to carry out innovative basic researches on high temperature engineering and to demonstrate high temperature heat utilization and application of nuclear heat. This report describes the demonstration program of high temperature heat utilization and application. (author). 2 refs, 4 figs, 3 tabs

  17. Modeling of fuel performance and fission product release behavior during HTTR normal operation. A comparative study of the FZJ and JAERI modeling approach

    International Nuclear Information System (INIS)

    Verfondern, Karl; Sumita, Junya; Ueta, Shohei; Sawa, Kazuhiro

    2001-03-01

    For the prediction of fuel performance and fission product release behavior in the High Temperature Engineering Test Reactor, HTTR of the Japan Atomic Energy Research Institute(JAERI), during its normal operation, calculation tools were applied as have been used at the Research Center Juelich (FZJ) in safety analyses for pebble-bed HTGR designs. Calculations were made assuming the HTTR operation with a nominal operation time of 660 efpd including a 110 efpd period with elevated fuel temperatures. Fuel performance calculations by the PANAMA code with given fuel temperature distribution in the core have shown that the additional failure level of about 5x10 -6 is expected which is about twice as much as the as-fabricated through-coatings failure level. Under the extreme safety design conditions, the predicted particle failure fraction in the core increases to about 1x10 -3 in maximum. The diffusive release of metallic fission products from the fuel primarily occurs in the core layer with the maximum fuel temperature (layer 3) whereas there is hardly any contribution from layer 1 except for the recoil fraction. Silver most easily escapes the fuel; the predicted release fractions from the fuel compacts are 10% (expected) and 50% (safety design). The figures for strontium (expected: 1.5x10 -3 ), safety design: 3.1x10 -2 ) and cesium (5.6x10 -4 , 2.9x10 -2 ) reveal as well a significant fraction to originate already from intact particles. Comparison with the calculation based on JAERI's diffusion model for cesium shows a good agreement for the release behavior from the particles. The differences in the results can be explained mainly by the different diffusion coefficients applied. The release into the coolant can not modelled because of the influence of the gap between compact and graphite sleeve lowering the release by a factor of 3 to 10. For the prediction of performance and fission product release behavior of advanced ZrC TRISO particles, more experimental work is

  18. High temperature continuous operation in the HTTR (HP-11). Summary of the test results in the high temperature operation mode

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Ueta, Shohei; Sumita, Junya; Goto, Minoru; Nakagawa, Shigeaki; Hamamoto, Shimpei; Tochio, Daisuke

    2010-11-01

    A high temperature (950 degrees C) continuous operation has been performed for 50 days on the HTTR from January to March in 2010, and the potential to supply stable heat of high temperature for hydrogen production for a long time was demonstrated for the first time in the world. JAEA has evaluated the experimental data obtained by this operation and past rated continuous one, and built the database necessary for commercial HTGRs. According to the results, the concentration of FP released from the fuels in the HTTR was a single through triple-digit lower than that in the foreign HTGRs. It became apparent that the fuels used in the HTTR are the best quality in the world. This successful operation could establish technological basis of HTGRs and show potential of nuclear energy as heat source for innovative thermo-chemical-based hydrogen production, emitting greenhouse gases on a 'low-carbon path' for the first time in the world. We have a plan to progress R and D for practical use of hydrogen production system with HTGRs in the future. (author)

  19. Safety demonstration test (SR-1/S1C-1) plan of HTTR (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Safety demonstration tests in the HTTR (High Temperature Engineering Test Reactor) will be carried out in order to verify inherent safety features of the HTGR (High Temperature Gas-cooled Reactor). The first phase of the safety demonstration tests includes the reactivity insertion test by the control rod withdrawal and the coolant flow reduction test by the circulator trip. In the second phase, accident simulation tests will be conducted. By comparison of their experimental and analytical results, the prediction capability of the safety evaluation codes such as the core and the plant dynamics codes will be improved and verified, which will contribute to establish the safety design and the safety evaluation technologies of the HTGRs. The results obtained through its safety demonstration tests will be also utilised for the establishment of the safety design guideline, the safety evaluation guideline, etc. This paper describes the test program of the overall safety demonstration tests and the test method, the test conditions and the results of the pre-test analysis of the reactivity insertion test and the partial gas circulator trip test planned in March 2003. (author)

  20. Design Basis of Core Components and their Realization in the frame of the EPR'sTM Core Component Development

    International Nuclear Information System (INIS)

    Schebitz, Florian; Mekmouche, Abdelhalim

    2008-01-01

    Rod Cluster Control Assemblies (RCCAs), Thimble Plug Assemblies (TPAs), Primary Neutron Sources (PNS) and Secondary Neutron Sources (SNS) are essential for the operation of a Nuclear Power Plant. Different functional requirements ask for different components and geometries. Therefore three different core components are used within the primary circuit: - The RCCA, which contains the absorber materials, is used to regulate and shut down the nuclear chain reaction. Under these demanding conditions different effects are determining the lifetime of the RCCA and in particular of the control rods. Several improvements like ion-nitriding of the cladding, lengthening of the bottom end plug, helium backfilling and reduction of the absorber diameter in the bottom part, which have already been introduced with the HARMONI TM RCCA, show a real improvement in terms of lifetime. - The TPAs are used at positions without RCCAs and neutron sources to limit the by-pass flow-rate in the fuel assembly guide tubes. The advanced TPA design results from a perfect combination of French and German design experience feedback. Benefits like homogenized hydraulic flow and improved manageability in terms of handling tools show the joined experience. - The neutron sources are used to enhance the flux level when the core is sub-critical so as to facilitate the core start-up control by the neutron flux detectors. Primary and secondary neutron sources are designed in a common way with reviewed and improved methodology. As there are different ways and conditions to operate core components, several designs are available. For the EPR TM , the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR TM . (authors)

  1. Radiation monitoring data on the power-up test of HTTR. Results up to 20 MW operation

    International Nuclear Information System (INIS)

    Ashikagaya, Yoshinobu; Nakazawa, Takashi; Yoshino, Toshiaki; Yasu, Katsuji

    2002-01-01

    The High Temperature Engineering Test Reactor (HTTR) have completed the Power-up test of 9 MW (the single and parallel loaded operation) in the rated operation mode. After that the Power-up test in the rated operation mode and the high-temperature test operation mode with a thermal output of 20 MW (the single and parallel loaded operation) were performed between January 16, 2001 and June 10, 2001. This report describes the radiation monitoring data carried out during the HTTR Power-up test in the rated operation mode and the high-temperature test operation mode with a thermal output of 20 MW. The followings were concluded from these radiation monitoring data. The monitoring of radioactive gaseous effluents and the radiation protection for the works will be easy to do and the exposure dose of the workers will be kept the low level. (author)

  2. Analysis of the HTTR with Monte-Carlo and diffusion theory. An IRI-ECN intercomparison

    International Nuclear Information System (INIS)

    De Haas, J.B.M.; Wallerbos, E.J.M.

    2000-09-01

    In the framework of the IAEA Co-ordinated Research Program (CRP) 'Evaluation of HTGR Performance' for the start-up core physics benchmark of the High Temperature Engineering Test Reactor (HTTR) two-group cross section data for a fuel compact lattice and for a two-dimensional R-Z model have been generated for comparison purposes. For this comparison, 5.2% enriched uranium was selected. Furthermore, a simplified core configuration utilising only the selected type of fuel has been analysed with both the Monte Carlo code KENO and with the diffusion theory codes BOLD VENTURE and PANTHER. With a very detailed KENO model of this simplified core, k eff was calculated to be 1.1278±0.0005. Homogenisation of the core region was seen to increase k eff by 0.0340 which can be attributed to streaming of neutrons in the detailed model. The difference in k eff between the homogenised models of KENO and BOLD VENTURE amounts then only,Δk =0.0025. The PANTHER result for this core is k eff = 1. 1251, which is in good agreement with the KENO result. The fully loaded core configuration, with a range of enrichments, has also been analysed with both KENO and BOLD VENTURE. In this case the homogenisation was seen to increase k eff by 0.0375 (streaming effect). In BOLD VENTURE the critical state could be reached by the insertion of the control rods through adding an effective 10 B density over the insertion depth while the streaming of neutrons was accounted for by adjustment of the diffusion coefficient. The generation time and the effective fraction of delayed neutrons in the critical state have been calculated to be 1.11 ms and 0.705 %, respectively. This yields a prompt decay constant at critical of 6.9 s -1 . The analysis with PANTHER resulted in a k eff =1.1595 and a critical control rod setting of 244.5 cm compared to the detailed KENO results of: k eff = 1.1600 and 234.5 cm, again an excellent agreement. 5 refs

  3. Development of monitoring system using acoustic emission for detection of helium gas leakage for primary cooling system and flow-induced vibration for heat transfer tube of heat exchangers for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Kunitomi, Kazuhiko; Furusawa, Takayuki; Shinozaki, Masayuki; Satoh, Yoshiyuki; Yanagibashi, Minoru

    1998-10-01

    The High Temperature Engineering Test Reactor (HTTR) uses helium gas for its primary coolant, whose leakage inside reactor containment vessel is considered in design of the HTTR. It is necessary to detect leakage of helium gas at an early stage so that total amount of the leakage should be as small as possible. On the other hand, heat transfer tubes of heat exchangers of the HTTR are designed not to vibrate at normal operation, but the flow-induced vibration is to be monitored to provide against an emergency. Thus monitoring system of acoustic emission for detection of primary coolant leakage and vibration of heat transfer tubes was developed and applied to the HTTR. Before the application to the HTTR, leakage detection test was performed using 1/4 scaled model of outer tube of primary concentric hot gas duct. Result of the test covers detectable minimum leakage rate and effect of difference in gas, pressure, shape of leakage path and distance from the leaking point. Detectable minimum leakage rate was about 5 Ncc/sec. The monitoring system is promising in leakage detection, though countermeasure to noise is to be needed after the HTTR starts operating. (author)

  4. HTTR hydrogen production system. Structure and main specifications of mock-up test facility (Contract research)

    International Nuclear Information System (INIS)

    Kato, Michio; Aita, Hideki; Inagaki, Yoshiyuki; Hayashi, Koji; Ohashi, Hirofumi; Sato, Hiroyuki; Iwatsuki, Jin; Takada, Shoji; Inaba, Yoshitomo

    2007-03-01

    The mock-up test facility was fabricated to investigate performance of the steam generator for mitigation of the temperature fluctuation of helium gas and transient behavior of the hydrogen production system for HTTR and to obtain experimental data for verification of a dynamic analysis code. The test facility has an approximate hydrogen production capacity of 120Nm 3 /h and the steam reforming process of methane; CH 4 +H 2 O=3H 2 +CO, was used for hydrogen production of the test facility. An electric heater was used as a heat source instead of the reactor in order to heat helium gas up to 880degC (4MPa) at the chemical reactor inlet which is the same temperature as the HTTR hydrogen production system. Fabrication of the test facility was completed in February in 2002, and seven cycle operations were carried out from March in 2002 to December in 2004. This report describes the structure and main specifications of the test facility. (author)

  5. A preliminary study on radiation damage effect in ceramics composite materials as innovative basic research using the HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Baba, Shinichi; Aihara, Jun; Arai, T.; Hayashi, K.; Ishino, S.

    1999-01-01

    An innovative basic research concerning with the basic science and applied technology is planned using the High Temperature Engineering Test Reactor (HTTR), which provides the advantage of not only a high temperature irradiation field above 400degC but also a large irradiation space. The first irradiation experiment is to be performed in 2001. Many research themes with a wide variety of scientific and technological interests are proposed as the innovative basic research. For the purpose of demonstration of scientific feasibility and advantages in the HTTR irradiation, several research themes have been being conducted as the preliminary studies. In this paper the outline of the innovative basic research is described, and the preliminary study on the radiation damage mechanism of ceramic composite materials is presented. (author)

  6. Characteristics of DC electrical braking method of the gas circulator to limit the temperature rise at the heat transfer pipes in the HTTR

    International Nuclear Information System (INIS)

    Kawasaki, K.; Saito, K.; Iyoku, T.

    2001-01-01

    In the safety evaluation of a High Temperature Engineering Test Reactor (HTTR), it must be confirmed that the core has no chance to be damaged and the barrier against the FP release is designed properly not to be affecting the influence of radiation around the reactor site. Especially the maximum temperature of the reactor pressure boundary such as the heat transfer pipes of pressurized water cooler (PWC) must not exceed the permissible values under an anticipated accident such as pipe of rupture in PWC. A requirement for the gas circulator which circulates helium gas in the primary cooling line and the secondary cooling line, is to be braked within 10 seconds by an electrical braking method after the HTTR reactor has scrammed under the accident in PWC. The reason is that the temperature rise of the heat transfer pipe at PWC has to be suppressed when the gas circulator has stopped, the revolution of the gas circulator decreases like the free coast down so that it takes about 90 seconds to be zero and the temperature rise of the pipe in the PWC exceeds the permissible value. By braking within 10 secs., the temperature of the pipe in the PWC reaches about 368 deg. C, less than the permissible value. Using a simplified equivalent circuit of an induction motor, braking time analysis was performed with obtained electrical resistance and inductance. The obtained braking time is about 10 secs., showing close agreement with analysis values. (author)

  7. Electrosprayed core-shell polymer-lipid nanoparticles for active component delivery

    Science.gov (United States)

    Eltayeb, Megdi; Stride, Eleanor; Edirisinghe, Mohan

    2013-11-01

    A key challenge in the production of multicomponent nanoparticles for healthcare applications is obtaining reproducible monodisperse nanoparticles with the minimum number of preparation steps. This paper focus on the use of electrohydrodynamic (EHD) techniques to produce core-shell polymer-lipid structures with a narrow size distribution in a single step process. These nanoparticles are composed of a hydrophilic core for active component encapsulation and a lipid shell. It was found that core-shell nanoparticles with a tunable size range between 30 and 90 nm and a narrow size distribution could be reproducibly manufactured. The results indicate that the lipid component (stearic acid) stabilizes the nanoparticles against collapse and aggregation and improves entrapment of active components, in this case vanillin, ethylmaltol and maltol. The overall structure of the nanoparticles produced was examined by multiple methods, including transmission electron microscopy and differential scanning calorimetry, to confirm that they were of core-shell form.

  8. Study on flow-induced vibration of the fuel rod in HTTR

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    1988-03-01

    This study was performed in order to investigate flow-induced vibration characteristics of a fuel rod in HTTR (High Temperature engineering Test Reactor) from both an experiment and a numerical simulation. Two kinds of fuel rods were used in this experiment: one was a graphite rod which simulated a specification of the HTTR's fuel rod and the other was an aluminum rod whose weight was a half of the graphite one. The experiment was carried out up to Re = 31000 using air at room temperature and pressure. Air flowed downstream in an annular passage which consisted of the fuel rod and the graphite channel. Numerical simulations by fluid and frequency equations were also carried out. Numerical and experimental results were then compared. The following conclusions were drived: (1) The fuel rod amplitudes increase with the flow rate and with a decrease of the fuel rod weight. (2) The fuel rod amplitudes are obtained by δ/De = 2.22 x 10 -10 Re 1.43 , 9000 ≤ Re ≤ 31000, where δ is a vibration amplitude, De is a hydraulic diameter and Reis Reynolds number. (3) The fuel rod frequencies shift from lower natural frequency to higher as the flow rate increases. (4) The flow-induced vibration behavior of the fuel rod can simulate well by simultaneous equations which used the turbulence model for fluid and the mass model for vibration of the fuel rod. (author)

  9. Incentives and opportunities for reducing the cobalt content in reactor core components

    International Nuclear Information System (INIS)

    Ocken, H.

    1985-01-01

    Cobalt in core components contributes to radiation field buildup on out-of-core surfaces. Core components containing cobalt-base alloys and cobalt as an impurity are identified. The use of cobalt-free wear-resistant alloys and construction materials with lower impurity levels of cobalt is disused. It is argued that such measures are cost effective. Lower radiation fields and disposal costs will offset higher raw material costs. Component performance will not be affected. (author)

  10. Design Basis of Core Components and their Realization in the frame of the EPR's{sup TM} Core Component Development

    Energy Technology Data Exchange (ETDEWEB)

    Schebitz, Florian [AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Mekmouche, Abdelhalim [AREVA NP SAS, 10 rue Juliette Recamier, 69456 Lyon Cedex 06 (France)

    2008-07-01

    Rod Cluster Control Assemblies (RCCAs), Thimble Plug Assemblies (TPAs), Primary Neutron Sources (PNS) and Secondary Neutron Sources (SNS) are essential for the operation of a Nuclear Power Plant. Different functional requirements ask for different components and geometries. Therefore three different core components are used within the primary circuit: - The RCCA, which contains the absorber materials, is used to regulate and shut down the nuclear chain reaction. Under these demanding conditions different effects are determining the lifetime of the RCCA and in particular of the control rods. Several improvements like ion-nitriding of the cladding, lengthening of the bottom end plug, helium backfilling and reduction of the absorber diameter in the bottom part, which have already been introduced with the HARMONI{sup TM} RCCA, show a real improvement in terms of lifetime. - The TPAs are used at positions without RCCAs and neutron sources to limit the by-pass flow-rate in the fuel assembly guide tubes. The advanced TPA design results from a perfect combination of French and German design experience feedback. Benefits like homogenized hydraulic flow and improved manageability in terms of handling tools show the joined experience. - The neutron sources are used to enhance the flux level when the core is sub-critical so as to facilitate the core start-up control by the neutron flux detectors. Primary and secondary neutron sources are designed in a common way with reviewed and improved methodology. As there are different ways and conditions to operate core components, several designs are available. For the EPR{sup TM}, the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR{sup TM}. (authors)

  11. Turbine component casting core with high resolution region

    Science.gov (United States)

    Kamel, Ahmed; Merrill, Gary B.

    2014-08-26

    A hollow turbine engine component with complex internal features can include a first region and a second, high resolution region. The first region can be defined by a first ceramic core piece formed by any conventional process, such as by injection molding or transfer molding. The second region can be defined by a second ceramic core piece formed separately by a method effective to produce high resolution features, such as tomo lithographic molding. The first core piece and the second core piece can be joined by interlocking engagement that once subjected to an intermediate thermal heat treatment process thermally deform to form a three dimensional interlocking joint between the first and second core pieces by allowing thermal creep to irreversibly interlock the first and second core pieces together such that the joint becomes physically locked together providing joint stability through thermal processing.

  12. Data on test results of vessel cooling system of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  13. Analysis of the HTTR with Monte-Carlo and diffusion theory. An IRI-ECN intercomparison

    Energy Technology Data Exchange (ETDEWEB)

    De Haas, J.B.M. [Nuclear Research and Consultancy Group NRG, Petten (Netherlands); Wallerbos, E.J.M. [Interfaculty Reactor Institute IRI, Delft University of Technology, Delft (Netherlands)

    2000-09-01

    In the framework of the IAEA Co-ordinated Research Program (CRP) 'Evaluation of HTGR Performance' for the start-up core physics benchmark of the High Temperature Engineering Test Reactor (HTTR) two-group cross section data for a fuel compact lattice and for a two-dimensional R-Z model have been generated for comparison purposes. For this comparison, 5.2% enriched uranium was selected. Furthermore, a simplified core configuration utilising only the selected type of fuel has been analysed with both the Monte Carlo code KENO and with the diffusion theory codes BOLD VENTURE and PANTHER. With a very detailed KENO model of this simplified core, k{sub eff} was calculated to be 1.1278{+-}0.0005. Homogenisation of the core region was seen to increase k{sub eff} by 0.0340 which can be attributed to streaming of neutrons in the detailed model. The difference in k{sub eff} between the homogenised models of KENO and BOLD VENTURE amounts then only,{delta}k =0.0025. The PANTHER result for this core is k{sub eff} = 1. 1251, which is in good agreement with the KENO result. The fully loaded core configuration, with a range of enrichments, has also been analysed with both KENO and BOLD VENTURE. In this case the homogenisation was seen to increase k{sub eff} by 0.0375 (streaming effect). In BOLD VENTURE the critical state could be reached by the insertion of the control rods through adding an effective {sup 10}B density over the insertion depth while the streaming of neutrons was accounted for by adjustment of the diffusion coefficient. The generation time and the effective fraction of delayed neutrons in the critical state have been calculated to be 1.11 ms and 0.705 %, respectively. This yields a prompt decay constant at critical of 6.9 s{sup -1}. The analysis with PANTHER resulted in a k{sub eff} =1.1595 and a critical control rod setting of 244.5 cm compared to the detailed KENO results of: k{sub eff} = 1.1600 and 234.5 cm, again an excellent agreement. 5 refs.

  14. Survey report on high temperature irradiation experiment programs for new ceramic materials in the HTTR (High Temperature Engineering Test Reactor). 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-02-01

    A survey research on status of research activities on new ceramic materials in Japan was carried out under contract between Japan Atomic Energy Research Institute and Atomic Energy Society of Japan. The purpose of the survey is to provide information to prioritize prospective experiments and tests in the HTTR. The HTTR as a high temperature gas cooled reactor has a unique and superior capability to irradiate large-volumed specimen at high temperature up to approximately 800degC. The survey was focused on mainly the activities of functional ceramics and heat resisting ceramics as a kind of structural ceramics. As the result, the report recommends that the irradiation experiment of functional ceramics is feasible to date. (K. Itami)

  15. Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR. Contract research

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji; Tochio, Daisuke; Ohwada, Hiroyuki

    2005-03-01

    Safety demonstration tests using the HTTR are to be conducted from the FY2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR that is one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3, S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was verified that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram. (author)

  16. Core component vibration monitoring in BWRs using neutron noise

    International Nuclear Information System (INIS)

    Fry, D.N.; Robinson, J.C.; Kryter, R.C.; Cole, O.C.

    1975-01-01

    Neutron noise from in-core fission detectors in a BWR was investigated to determine its effectiveness as a monitor of mechanical vibrations of core components. In this study the general properties of BWR neutron noise were characterized, and a signal enhancement method was implemented to improve the measurement sensitivity. (auth)

  17. HTR core physics analysis at NRG

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

    2002-01-01

    Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

  18. Soil-structure interaction analysis of HTTR building by a simplified model

    International Nuclear Information System (INIS)

    Yagishita, F.; Suzuki, H.; Yamagishi, Y.

    1990-01-01

    For the evaluation of the design seismic forces of the embedded High-Temperature-Testing-Reactor (HTTR) structure, a sway-rocking model considering the embedment of the structure is used. As for the composition of this model; the structure is modeled into beams with lumped masses, and the soil into the horizontal side springs and the horizontal and rotational bottom springs. At the same time, the input motion to the structure which has the form of multiple excitation is calculated based on one dimensional wave propagation theory. This paper presents the concept of this modelling and evaluated results. (author). 9 refs, 11 figs

  19. Hydrogen production by high-temperature gas-cooled reactor. Conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki; Hara, Teruo; Kato, Ryoma; Kunitomi, Kazuhiko

    2008-01-01

    Nuclear hydrogen production is necessary in an anticipated hydrogen society that demands a massive quantity of hydrogen without economic disadvantage. Japan Atomic Energy Agency (JAEA) has launched the conceptual design study of a hydrogen production system with a near-term plan to connect it to Japan's first high-temperature gas-cooled reactor HTTR. The candidate hydrogen production system is based on the thermochemical water-splitting iodine sulphur (IS) process.The heat of 10 MWth at approximately 900degC, which can be provided by the secondary helium from the intermediate heat exchanger of the HTTR, is the energy input to the hydrogen production system. In this paper, we describe the recent progresses made in the conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system. A new concept of sulphuric acid decomposer is proposed. This involves the integration of three separate functions of sulphuric acid decomposer, sulphur trioxide decomposer, and process heat exchanger. A new mixer-settler type of Bunsen reactor is also designed. This integrates three separate functions of Bunsen reactor, phase separator, and pump. The new concepts are expected to result in improved economics through construction and operation cost reductions because the number of process equipment and complicated connections between the equipment has been substantially reduced. (author)

  20. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Ott, L.J.

    1997-01-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  1. Valuation of the safety concept of the combined nuclear/chemical complex for hydrogen production with HTTR

    International Nuclear Information System (INIS)

    Verfondern, K.; Nishihara, T.

    2004-06-01

    The high-temperature engineering test reactor (HTTR) in Oarai, Japan, will be worldwide the first plant to demonstrate the production of hydrogen by applying the steam reforming process and using nuclear process heat as primary energy. Particular safety aspects for such a combined nuclear/chemical complex have to be investigated to further detail. One of these special aspects is the fire and explosion hazard associated with the presence of flammable gases including a large LNG storage tank in close vicinity to the reactor building. A special focus is laid upon the conceivable development of a detonation pressure wave and its damaging effect on the reactor building. A literature study has shown that methane is a comparatively slow reacting gas and that a methane vapor cloud in the open atmosphere or partially obstructed areas is highly unlikely to result in a detonation if inadvertently released and ignited. Various theoretical assessments and experimental studies, which have been conducted in the past and which are of significance for the HTTR-steam reforming system, include the spreading and combustion behavior of cryogenic liquids and flammable gas mixtures providing the basis of a comprehensive safety analysis of the combined nuclear/chemical facility. (orig.)

  2. Understanding susceptibility of in-core components to irradiation-assisted stress corrosion cracking

    International Nuclear Information System (INIS)

    Chung, H.M.; Ruther, W.E.; Sanecki, J.E.; Kassner, T.F.

    1991-03-01

    As nuclear plants age and accumulated fluences of core structural components increase, susceptibility of the components to irradiation-assisted stress corrosion cracking (IASCC) is also expected to increase. Irradiation-induced sensitization, commonly associated with an IASCC failure, was investigated in this study to provide a better understanding of long-term structural integrity of safety-significant in-core components. Irradiation-induced sensitization of high- and commercial-purity Type 304 stainless steels irradiated in BWRs was analyzed. 7 refs., 8 figs

  3. Refractory metal component technology for in-core sensor design

    International Nuclear Information System (INIS)

    Cannon, C.P.

    1986-02-01

    Within recent years, an increasing concern over reactor safety has prompted tests that characterize reactor core environments during transient conditions. Such tests include the Loss-of-Fluid-Tests (Idaho National Engineering Lab (INEL)), Severe Fuel Damage Tests (INEL), Core Debris Rubble Tests (Sandia National Laboratories (SNL)), and similar tests performed by foreign nations. The in-core sensors for these tests require refractory metal components to be compatible with electrical insulator materials as well as materials comprising highly corrosive service mediums. This paper presents the refractory metal technology utilized to provide basic sensor designs in the above mentioned reactor tests

  4. A Delphi study to identify the core components of nurse to nurse handoff.

    Science.gov (United States)

    O'Rourke, Jennifer; Abraham, Joanna; Riesenberg, Lee Ann; Matson, Jeff; Lopez, Karen Dunn

    2018-03-08

    The aim of this study was to identify the core components of nurse-nurse handoffs. Patient handoffs involve a process of passing information, responsibility and control from one caregiver to the next during care transitions. Around the globe, ineffective handoffs have serious consequences resulting in wrong treatments, delays in diagnosis, longer stays, medication errors, patient falls and patient deaths. To date, the core components of nurse-nurse handoff have not been identified. This lack of identification is a significant gap in moving towards a standardized approach for nurse-nurse handoff. Mixed methods design using the Delphi technique. From May 2016 - October 2016, using a series of iterative steps, a panel of handoff experts gave feedback on the nurse-nurse handoff core components and the content in each component to be passed from one nurse to the next during a typical unit-based shift handoff. Consensus was defined as 80% agreement or higher. After three rounds of participant review, 17 handoff experts with backgrounds in clinical nursing practice, academia and handoff research came to consensus on the core components of handoff: patient summary, action plan and nurse-nurse synthesis. This is the first study to identify the core components of nurse-nurse handoff. Subsequent testing of the core components will involve evaluating the handoff approach in a simulated and then actual patient care environment. Our long-term goal is to improve patient safety outcomes by validating an evidence-based handoff framework and handoff curriculum for pre-licensure nursing programmes that strengthen the quality of their handoff communication as they enter clinical practice. © 2018 John Wiley & Sons Ltd.

  5. Study on in-service visual inspection using TV camera for core support graphite components in the HTTR

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Shibata, Taiju; Kikuchi, Takayuki; Mogi, Haruyoshi

    1999-01-01

    To maintain the structural integrity of graphite components during plant operation a visual inspection using a TV camera as an in-service inspection is planned in the High Temperature Engineering Test Reactor. In order to verify the in-service inspection method a preliminary analytical and experimental studies were performed. In the analytical study the harmful flaw size was determined from a viewpoint of structural integrity based on the fracture mechanics approach. Furthermore, the visible flaw size was determined by the TV camera performance test with graphite test specimens having several kinds of artificial flaws. This paper presents the analytical result on the harmful flaw size and the experimental result on the visible flaw size. From both results the applicability on the visual inspection by the TV camera as the in-service inspection is discussed in this paper. (author)

  6. Reconstitution of fuel assemblies and core components

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, Wolfgang; Langenberger, Jan [AREVA NP GmbH (Germany)

    2012-11-01

    Due to AREVA's experience and big portfolio of techniques, reconstitution of fuel assemblies and core components at light water reactors is possible within a reasonable timeframe and with interesting cost benefit. Customer feedback indicates the sustainability of such reconstitutions. As a result, a long-term maintenance of value can be assured and early waste disposal can be avoided. (orig.)

  7. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1990-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

  8. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1989-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

  9. Identification of the key parameters defining the life of graphite core components

    International Nuclear Information System (INIS)

    Mitchell, M.N.

    2005-01-01

    The Core Structures of a Pebble Bed rector core comprise graphite reflectors constructed from blocks. These blocks are subject to high flux and temperatures as well as significant gradients in flux and temperature. This loading combined with the behaviour of graphite under irradiation gives rise to complex stress states within the reflector blocks. At some point, the stress state will reach a critical level and cracks will initiate within the blocks. The point of crack initiation is a useful point to define as the end of the part's life. The life of these graphite reflector parts in a pebble bed reactor (PBR) core determines the service life of the Core Structures. The replacement of the Core Structures' components will be a costly and time consuming. It is important that the components of the Core Structures be designed for the best life possible. As part of the conceptual design of the Pebble Bed Modular Reactor (PBMR), the assessment of the life of these components was examined. To facilitate the understanding of the parameters that influence the design life of the PBMR, a study has been completed into the effect of various design parameters on the design life of a typical side reflector block. Parameters investigated include: block geometry, material property variations, and load variations. The results of this study are to be presented. (author)

  10. Ultrasonic test results for the reactor pressure vessel of the HTTR. Longitudinal welding line of bottom dome

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Ohwada, Hiroyuki; Kato, Yasushi

    2008-06-01

    This paper describes the inspection method, the measured area, etc. of the ultrasonic test of the in-service inspection (ISI) for welding lines of the reactor pressure vessel of the HTTR and the inspection results of the longitudinal welding line of the bottom dome. The pre-service inspection (PSI) results for estimation of occurrence and progression of defects to compare the ISI results is described also. (author)

  11. Effect of crosstalk on component savings in multi-core fiber networks

    DEFF Research Database (Denmark)

    Nooruzzaman, Md; Morioka, Toshio

    2017-01-01

    We estimate potential component savings in MCF-based networks by using shortest path traffic routing and compare them with the current SCF-based systems. We also investigate the potential impact of various inter-core crosstalk values on the number of needed transponders in MCF networks.......We estimate potential component savings in MCF-based networks by using shortest path traffic routing and compare them with the current SCF-based systems. We also investigate the potential impact of various inter-core crosstalk values on the number of needed transponders in MCF networks....

  12. Examination of core components removed from CANDU reactors

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Coleman, C.E.; Rodgers, D.K.; Davies, P.H.; Chow, C.K.; Griffiths, M.

    1988-11-01

    Components in the core of a nuclear reactor degrade because the environment is severe. For example, in CANDU reactors the pressure tubes must contend with the effects of hot pressurised water and damage by a flux of fast neutrons. To evaluate any deterioration of components and determine the cause of the occasional failure, we have developed a wide range of remote-handling techniques to examine radioactive materials. As well as pressure tubes, we have examined calandria tubes, garter springs, end fittings, liquid-zone control units and flux detectors. The results from these examinations have produced solutions to problems and continually provide information to help understand the processes that may limit the lifetime of a component

  13. MCNP qualification on the HTR critical configurations: HTTR, HTR10 and PROTEUS results

    Energy Technology Data Exchange (ETDEWEB)

    TRAKAS, Christos; STOVEN, Gilles [AREVA NP, Tour Areva, 92084 Paris La Defence Cedex (France)

    2008-07-01

    Recent critical experiments, including PROTEUS, HTTR and HTR-10 provide a reliable qualification base for HTR criticality predictions. The fuel tested in these experiments, be it hexagonal block or pebble type, is irradiated in a spectrum comparable to that of the HTR planned by AREVA NP. The neutron spectrum is comparable in all three cases; the mean C/M value for all critical cases is less than +350 pcm (JEF2.2), +250 pcm (JEFF3.1) and +60 pcm (ENDF BVI). The C/M obtained for the rods worth, the reaction rates and the isothermal coefficient are very satisfactory. (authors)

  14. Test program for NIS calibration to reactor thermal output in HTTR

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Shinozaki, Masayuki; Tachibana, Yukio; Kunitomi, Kazuhiko

    2000-03-01

    Rise-to-power test program for reactor thermal output measurement has been established to calibrate a neutron instrumentation system taking account of the characteristics of the High Temperature Engineering Test Reactor (HTTR). An error of reactor thermal output measurement was evaluated taking account of a configuration of instrumentation system. And the expected dispersion of measurement in the full power operation was evaluated from non-nuclear heat-up of primary coolant up to 213degC. From the evaluation, it was found that an error of reactor thermal output measurement would be less than ±2.0% at the rated power. This report presents the detailed program of rise-to-power test for reactor thermal output measurement and discusses its measurement error. (author)

  15. An approach to development of structural design criteria for highly irradiated core components

    International Nuclear Information System (INIS)

    Nelson, D.V.

    1980-01-01

    The advent of the fast breeder reactor presents novel challenges in structural design and materials engineering. For instance, the core components of these reactors experience high energy neutron irradiation at elevated temperature, which causes significant time-dependent changes in material behaviour, such as a progressive loss of ductility. New structural design criteria are needed to extend elevated temperature design-by-analysis to account for these changes. Alloys best able to cope with the demands of the core operating environment are being explored and their structural behaviour characterized. The purpose of this paper is to illustrate an approach used in the development of core component structural design criteria. To do this, several design rules, plus brief rationale, from draft RDT Standards F9-7, -8 and -9 will be presented. These recently completed standards ('Structural Design Guidelines for Breeder Reactor Core Components') were prepared for the U.S. Department of Energy and represent a consensus among most organizations participating in the U.S. breeder program. (author)

  16. Helium leak and chemical impurities control technology in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Shimizu, Atsushi; Hamamoto, Shimpei; Sakaba, Nariaki

    2014-01-01

    Japan Atomic Energy Agency (JAEA) has designed and developed high-temperature gas-cooled reactor (HTGR) hydrogen cogeneration system named gas turbine high-temperature reactor (GTHTR300C) as a commercial HTGR. Helium gas is used as the primary coolant in HTGR. Helium gas is easy to leak, and the primary helium leakage should be controlled tightly from the viewpoint of preventing the release of radioactive materials to the environment. Moreover from the viewpoint of preventing the oxidization of graphite and metallic material, the helium coolant chemistry should be controlled tightly. The primary helium leakage and the helium coolant chemistry during the operation is the major factor in the HTGR for commercialization of HTGR system. This paper shows the design concept and the obtained operational experience on the primary helium leakage control and primary helium impurity control in the high-temperature engineering test reactor (HTTR) of JAEA. Moreover, the future plan to obtain operational experience of these controls for commercialization of HTGR system is shown. (author)

  17. Temperature analysis of the control rods at the scram shutdown of the HTTR. Evaluation by using measurement data at scram test of HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Eiji; Fujimoto, Nozomu; Nakagawa, Shigeaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Matsuda, Atsuko [Toshiba Co., Tokyo (Japan)

    2003-03-01

    In the High Temperature Engineering Test Reactor (HTTR), since the primary coolant temperature become 950 degrees centigrade at the high temperature test operation, the special alloy Alloy800H is used for cladding tubes and spines of the control rods to endure the high temperature. The temperature limitation of control rod is 900 degrees centigrade according to the strength data of Alloy800H. The scram shutdown by loss of off-site electric power at the high temperature test operation was assumed as an transient of the temperature of the control rods cladding might exceed 900 degrees centigrade. In this report, the temperature of the control rods is analyzed by using the measurement data of the rise-to-power test. From the result of this analysis, it was confirmed that the control rod temperature does not exceed the limit even at the transient of the loss of off-site electric power from the high temperature test operation. (author)

  18. Electron spin resonance characterization of radical components in irradiated black pepper skin and core

    International Nuclear Information System (INIS)

    Yamaoki, Rumi; Kimura, Shojiro; Ohta, Masatoshi

    2011-01-01

    Characteristics of free radical components of irradiated black pepper fruit (skin) and the pepper seed (core) were analyzed using electron spin resonance. A weak signal near g=2.005 was observed in black pepper before irradiation. Complex spectra near g=2.005 with three lines (the skin) or seven lines (the core) were observed in irradiated black pepper (both end line width; ca. 6.8 mT). The spectral intensities decreased considerably at 30 days after irradiation, and continued to decrease steadily thereafter. The spectra simulated on the basis of the content and the stability of radical components derived from plant constituents, including fiber, starch, polyphenol, mono- and disaccharide, were in good agreement with the observed spectra. Analysis showed that the signal intensities derived from fiber in the skin for an absorbed dose were higher, and the rates of decrease were lower, than that in the core. In particular, the cellulose radical component in the skin was highly stable. - Highlights: → We identified the radical components in irradiated black pepper skin and core. → The ESR spectra near g=2.005 with 3-7 lines were emerged after irradiation. → Spectra simulated basing on the content and the stability of radical from the plant constituents. → Cellulose radical component in black pepper skin was highly stable. → Single signal near g=2.005 was the most stable in black pepper core.

  19. Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2002-07-01

    The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

  20. Performance-based improvement of the leakage rate test program for the reactor containment of HTTR. Adoption of revised test programs containing 'Type A, Type B and Type C tests'

    International Nuclear Information System (INIS)

    Kondo, Masaaki; Emori, Koichi; Sekita, Kenji; Furusawa, Takayuki; Hayakawa, Masato; Kozawa, Takayuki; Aono, Tetsuya; Kuroha, Misao; Ouchi, Hiroshi; Kimishima, Satoru

    2008-10-01

    The reactor containment of HTTR is periodically tested to confirm leak-tight integrity by conducting overall integrated leakage rate tests, so-called 'Type A tests,' in accordance with a standard testing method provided in Japan Electric Association Code (JEAC) 4203. 'Type A test' is identified as a basic one for measuring whole leakage rates for reactor containments, it takes, however, much of cost and time of preparation, implementation and restoration of itself. Therefore, in order to upgrade the maintenance technology of HTTR, the containment leakage rate test program for HTTR was revised by adopting efficient and economical alternatives including Type B and Type C tests' which intend to measure leakage rates for containment penetrations and isolation valves, respectively. In JEAC4203-2004, following requirements are specified for adopting an alternative program: upward trend of the overall integrated leakage rate due to aging affection should not be recognized; performance criterion for combined leakage rate, that is a summation of local leakage rates evaluated by Type B and Type C tests and converted to whole leakage rates, should be established; the criterion of the combined leakage rate should be satisfied as well as of the overall integrated leakage rate; correlation between the overall integrated and combined leakage rates should be recognized. Considering the historical performances, policies of conforming to the forgoing requirements and of carrying out the revised test program were developed, which were accepted by the regulatory agency. This report presents an outline of the leakage rate tests for the reactor containment of HTTR, identifies practical issues of conventional Type A tests, and describes the conforming and implementing policies mentioned above. (author)

  1. Investigation on cause of malfunction of wide range monitor (WRM) in high temperature engineering test reactor (HTTR). Sample tests and destructive tests

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Saito, Kenji; Haga, Hiroyuki; Sasaki, Shinji; Katsuyama, Kozo; Motegi, Toshihiro; Takada, Kiyoshi; Higashimura, Keisuke; Fujii, Junichi; Ukai, Takayuki; Moriguchi, Yusuke

    2012-11-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the malfunction by specifying the damaged part causing the event in the WRM. One is an experiment using a mock-up sample test which strength degradation at assembly process and heat cycle to specify the damaged part in the WRM. The other is a destructive test in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. This report summarized the results of the destructive test and the experimental investigation using the mock-up to reveal the cause of malfunction of WRM. (author)

  2. Improvement of the abnormal diagnosis technology by the development of an abnormal parts assignment system for the engineered safety features actuating system of the HTTR

    International Nuclear Information System (INIS)

    Hirato, Yoji; Kozawa, Takayuki; Saito, Kenji

    2015-01-01

    The safety protection sequence panel of HTTR is a control panel to actuate an engineering safety system for protecting the reactor core, reactor coolant pressure boundary, and containment vessel boundary at the time of an accident of the nuclear reactor facilities. The safety code stipulates that the control panel should receive safety check at a frequency of once a month during reactor operation. When abnormality has been found, it is required to eliminate its causes and restore normal operation as soon as possible. However, since this control panel is composed of a complex control circuit, the cause check during abnormality requires the confirmation by a knowledgeable person spending quite a lot of time for chart checking, which leads to a delay of restoration. To achieve a rapid restoration, the abnormal part assignment system (APAS), which can specify abnormality instantaneously even by a common operator, was developed. It has been confirmed that with this system, rapid initial response and prompt restoration can be effectively made. (A.O.)

  3. DETERMINATION OF CRYSTALLINITY INDEX OF CARBOHYDRATE COMPONENTS IN HEMP (CANNABIS SATIVA L. WOODY CORE BY MEANS OF FT-IR SPECTROSCOPY

    Directory of Open Access Journals (Sweden)

    Esat Gümüşkaya

    2005-04-01

    Full Text Available In this study; it was investigated chemical compositions of hemp woody core and changes in crystallinity index of its carbohydrate components by using FT-IR spectroscopy was investigated. It was determined that carbohyrate components ratio in hemp woody core were similar to that in hard wood, but lignin content in hemp woody core was higher than in hard wood. Crystallinity index of carbohydrate components in hemp woody core increased by removing amorphous components. It was designated that monoclinic structure in hemp woody core and its carbohydrate components was dominant, but triclinic ratio increased by treated chemical isolation of carbohydrate from hemp woody core.

  4. Refurbishment of Cirus in-core components

    International Nuclear Information System (INIS)

    Bhatnagar, A.; Sahu, A.K.; Rathore, K.K.; Subudhi, D.; Kharpate, A.V.; Tikku, A.C.

    2006-01-01

    Circus is a 40- MWt vertical tank type research reactor with natural uranium as fuel, demineralised light water as coolant, heavy water as moderator and graphite as reflector. The reactor was commissioned in the year 1960 and operated at an overage availability factor of over 70% till early nineties, when various systems, structures and components (SSCs) started showing signs of ageing. Detailed ageing studies were therefore carried out to assess the condition of various SSCs and refurbishing requirements were finalized towards extending the life of the reactor. In-core components, being non-replaceable generally, were critically examined to the extent possible. Detailed visual examination of a few reactor vessel (RV) tubes, made of aluminium, was carried out using micro video camera and in addition all the RV tubes were inspected using eddy current testing method. RV shell, also made of aluminium, was similarly visually inspected with micro video camera. To assess the effect of irradiation on the RV material, samples of similar tubes irradiated to comparable neutron fluence were tested. Towards assessment of fatigue life of RV expansion joint, made of aluminium, a finite element analysis using NISA computer code was performed. Theoretical assessment for stored Wigner energy in graphite reflector was carried out. Graphite block samples were also removed from the reactor and stored energy levels were measured to plan for any in-situ graphite annealing, if required. Visual inspection of approachable portions of steel and aluminium thermal shields was also carried out. These water-cooled thermal shields provided above and below the RV were hydro tested. The weld joint between coolant inlet pipe and top plate of upper aluminium shield showed minor leakage. A special metallic hollow plug was developed and remotely installed in the leaky pipe to isolate the leaky portion while maintaining the coolant flow in the pipe. Helium leak was found from flange joints located on

  5. Design/licensing of on-site package for core component

    International Nuclear Information System (INIS)

    Ogasawara, K.; Chohzuka, T.; Shimura, T.; Kikuchi, T.; Fujiwara, R.; Karigome, S.; Takani, M.

    1993-01-01

    For storage of used core components which are produced from reactors, Tohoku EPCO decided to construct a site bunker at Onagawa site. It was also decided to develop and fabricate one packaging to transport core components from the reactor buildings to the site bunker. The packaging will be used within the power station; therefore, it shall comply with 'The Law for the Business of Electric Power' and relevant Notification. The main requirements of the packaging are as follows: 1) The number of contents, such as channel boxes and control rods, shall be as large as possible. 2) The weight and the outer dimensions of the packaging shall be within the limitation of the reactor building and the site bunker. 3) Materials shall be selected from those which have been already applied for existing packagings and utilized without any problems. 4) It shall be considered during design of trunnions that handling equipment, such as lifting beam, can be used for not only this packaging but also for existing spent fuel packagings. The design of the packaging is completed and has been licensed. The packaging is scheduled to be utilized from November, 1993. (J.P.N.)

  6. Overview of HTGR utilization system developments at JAERI

    International Nuclear Information System (INIS)

    Miyamoto, Y.; Shiozawa, S.; Inagaki, Y.

    1997-01-01

    JAERI has been constructing a 30-MWt HTGR, named HTTR, to develop technology and to demonstrate effectiveness of high-temperature nuclear heat utilization. A hydrogen production system by natural gas steam reforming is to be the first heat utilization system of the HTTR since its technology matured in fossil-fired plant enables to couple with HTTR in the early 2000's and technical solutions demonstrated by the coupling will contribute to all other hydrogen production systems. The HTTR steam reforming system is designed to utilize the nuclear heat effectively and to achieve hydrogen productivity competitive to that of a fossil-fired plant with operability, controllability and safety acceptable enough to commercialization, and an arrangement of key components was already decided. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test is planned to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions. The out-of-pile system is an approximately 1/20-1/30 scale system of the HTTR steam reforming system and simulates key components downstream from an IHX

  7. Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

    International Nuclear Information System (INIS)

    1996-03-01

    On July 25, 1994, the US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water reactors (BWRs). This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the US and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components. This report summarizes the staff's basis for issuing GL 94-03, as well as the staff's assessment of plant-specific responses to GL 94-03. The staff is continually evaluating the licensee inspection programs and the results from examinations of BWR core shrouds and other internal components. This report is representative of submittals to and evaluations by the staff as of September 30, 1995. An update of this report will be issued at a later date

  8. Fabrication development of full-sized components for GCFR core assemblies

    International Nuclear Information System (INIS)

    Lindgren, J.R.; Flynn, P.W.; Foster, L.C.

    1980-05-01

    This paper presents the status of the development of full-sized components for gas-cooled fast reactor (GCFR) core assemblies. Methods for ribbing of the fuel rod cladding, fabrication of grid spacers of two different designs, drawing of assembly flow ducts, and fabrication of fission gas collection manifolds by several methods are discussed

  9. Evaluation of nuclear power plant component failure probability and core damage probability using simplified PSA model

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2000-01-01

    It is anticipated that the change of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components may cause the change of component failure probability and result in the change of core damage probability. It is also anticipated that the change is different depending on the initiating event frequency or the component types. This study assessed the change of core damage probability using simplified PSA model capable of calculating core damage probability in a short time period, which is developed by the US NRC to process accident sequence precursors, when various component's failure probability is changed between 0 and 1, or Japanese or American initiating event frequency data are used. As a result of the analysis, (1) It was clarified that frequency of surveillance test, preventive maintenance or parts replacement of motor driven pumps (high pressure injection pumps, residual heat removal pumps, auxiliary feedwater pumps) should be carefully changed, since the core damage probability's change is large, when the base failure probability changes toward increasing direction. (2) Core damage probability change is insensitive to surveillance test frequency change, since the core damage probability change is small, when motor operated valves and turbine driven auxiliary feed water pump failure probability changes around one figure. (3) Core damage probability change is small, when Japanese failure probability data are applied to emergency diesel generator, even if failure probability changes one figure from the base value. On the other hand, when American failure probability data is applied, core damage probability increase is large, even if failure probability changes toward increasing direction. Therefore, when Japanese failure probability data is applied, core damage probability change is insensitive to surveillance tests frequency change etc. (author)

  10. Nonlinear seismic analysis of a reactor structure with impact between core components

    International Nuclear Information System (INIS)

    Hill, R.G.

    1975-01-01

    The seismic analysis of the FFTF-PIOTA (Fast Flux Test Facility-Postirradiation Open Test Assembly), subjected to a horizontal DBE (Design Base Earthquake) is presented. The PIOTA is the first in a set of open test assemblies to be designed for the FFTF. Employing the direct method of transient analysis, the governing differential equations describing the motion of the system are set up directly and are implicitly integrated numerically in time. A simple lumped-mass beam model of the FFTF which includes small clearances between core components is used as a ''driver'' for a fine mesh model of the PIOTA. The nonlinear forces due to the impact of the core components and their effect on the PIOTA are computed. 6 references

  11. Measures ensuring safety of the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    1998-04-01

    JAERI has conducted research and development of an HTGR type reactor since 1969 under the project of the multi-purpose high-temperate gas-cooled experimental reactor, whose design was changed to the HTTR in 1985. The reactor license was granted by the Government in 1990 and the construction started next year. Various functions and performances have been tested since 1996 and the initial criticality achieved in 1998. This document consists of six chapters, describing safety matters examined in several development phases. The first chapter deals with succession of the multi-purpose experimental reactor technology and its exchange between JAERI and domestic industries. Chapter 2 reviews new technical findings after the licensing which were reflected to the current safety assessment. These technical items are given in the table form of extensive pages. Chapter 3 and 4 describe the performance tests and the criticality access, respectively. Chapter 5 and 6 deal with the detection of fuel failures and helium gas leaks, respectively. (H.Y.)

  12. Evaluation of in-core measurements by means of principal components method

    International Nuclear Information System (INIS)

    Makai, M.; Temesvari, E.

    1996-01-01

    Surveillance of a nuclear reactor core comprehends determination of assemblies' three-dimensional (3D) power distribution. Derived from other assemblies' measured values, power of non-measured assembly is calculated for every assembly with the help of principal components method (PCM) which is also presented. The measured values are interpolated for different geometrical coverings of the WWER-440 core. Different procedures have been elaborated and investigated, among them the most successful methods are discussed. Each method offers self consistent means to determine numerical errors of the interpolated values. (author). 13 refs, 7 figs, 2 tabs

  13. Analysis of irradiation-induced stresses in coating layers of coated fuel particles for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Fukuda, Kousaku; Sato, Sadao; Toyota, Junji; Shiozawa, Shusaku; Sawa, Kazuhiro; Kashimura, Satoru.

    1991-07-01

    Irradiation-induced stresses in coating layers of coated fuel particles were analyzed by the MICROS-2 code for the fuels of the High Temperature Engineering Test Reactor (HTTR) under its operating conditions. The analyses were made on the standard core fuel (A-type) and the test fuels comprising the advanced SiC-coated particle fuel (B-1 type) and the ZrC-coated particle fuel (B-2 type). For the B-1 type fuel, the stresses were relieved due to the thicker buffer and SiC layers than for the A type fuel. The slightly decreased thickness of the fourth layer for the B-1 type than for the A type fuel had no significant effect on the stresses. As for the B-2 type fuel, almost the same results as for the B-1 type were obtained under an assumption that the ZrC layer as well as the SiC layer undergoes negligible dimension change within the analysis conditions. The obtained results indicated that the B-1 and B-2 type fuels are better than the A type fuel in terms of integrity against the irradiation-induced stresses. Finally, research subjects for development of the analysis code on the fuel behavior are discussed. (author)

  14. Analysis of Hydrogen Cyanide Hyperfine Spectral Components towards Star Forming Cores

    Directory of Open Access Journals (Sweden)

    Loughnane R. M.

    2011-12-01

    Full Text Available Although hydrogen cyanide has become quite a common molecular tracing species for a variety of astrophysical sources, it, however, exhibits dramatic non-LTE behaviour in its hyperfine line structure. Individual hyperfine components can be strongly boosted or suppressed. If these so-called hyperfine line anomalies are present in the HCN rotational spectra towards low or high mass cores, this will affect the interpretation of various physical properties such as the line opacity and excitation temperature in the case of low mass objects and infall velocities in the case of their higher mass counterparts. Anomalous line ratios are present either through the relative strengths of neighboring hyperfine lines or through the varying widths of hyperfine lines belonging to a particular rotational line. This work involves the first observational investigation of these anomalies in two HCN rotational transitions, J=1→0 and J=3→2, towards both low mass starless cores and high mass protostellar objects. The degree of anomaly in these two rotational transitions is considered by computing the ratios of neighboring hyperfine lines in individual spectra. Results indicate some degree of anomaly is present in all cores considered in our survey, the most likely cause being line overlap effects among hyperfine components in higher rotational transitions.

  15. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    International Nuclear Information System (INIS)

    Guzina, Bojan; Kunerth, Dennis

    2014-01-01

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  16. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Guzina, Bojan [Univ. of Minnesota, Minneapolis, MN (United States); Kunerth, Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-30

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  17. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Bess, John D.; Fujimoto, Nozomu

    2014-01-01

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  18. Fabrication of HTTR first loading fuel

    International Nuclear Information System (INIS)

    Kato, S.; Yoshimuta, S.; Hasumi, T.; Sato, K.; Sawa, K.; Suzuki, S.; Mogi, H.; Shiozawa, S.; Tanaka, T.

    2001-01-01

    This paper summarizes the fabrication of the first loading fuel for HTTR, High Temperature engineering Test Reactor constructed by JAERI, Japan Atomic Energy Research Institute. The fuel fabrication started at the HTR fuel facility of NFI, Nuclear Fuel Industries, Ltd., June 1995. 4,770 fuel rods were fabricated through the fuel kernel, coated fuel particle and fuel compaction process, then 150 fuel elements were assembled in the reactor building December 1997. Fabrication technology for the fuel was established through a lot of R and D activities and fabrication experience of irradiation examination samples spread over about 30 years. Most of all, very high quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and the automatic compaction process. As for the inspection technology, the development of the automatic measurement equipment for coated layer thickness of a coated fuel particle and uranium content of a fuel compact contributed to the higher reliability and rationalization of the inspection process. The data processing system for the fabrication and quality control, which was originally developed by NFI, made possible not only quick feedback of statistical quality data to the fabrication processes, but also automatic document preparation, such as inspection certificates and accountability control reports. The quality of the first loading fuel fully satisfied the design specifications for the fuel. In particular, average bare uranium fraction and SiC defective fraction of fuel compacts were 2x10 -6 and 8x10 -5 , respectively. According to the preceding irradiation examinations being performed at JMTR, Japan Materials Testing Reactor of JAERI, the specimen sampled from the first loading fuel shows good irradiation performance. (author)

  19. Seismic scrammability of HTTR control rods

    International Nuclear Information System (INIS)

    Nishiguchi, I.; Iyoku, T.; Ito, N.; Watanabe, Y.; Araki, T.; Katagiri, S.

    1990-01-01

    Scrammability tests on HTTR (High-Temperature Engineering Test Reactor) control rods under seismic conditions have been carried out and seismic conditions influences on scram time as well as functional integrity were examined. A control rod drive located in a stand-pipe at the top of a reactor vessel, raises and lowers a pair of control rods by suspension cables. Each flexible control rod consists of 10 neutron absorber sections held together by a metal spine passing through the center. It falls into a hole in graphite blocks due to gravity at scram. In the tests, a full scale control rod drive and a pair of control rods were employed with a column of graphite blocks in which holes for rods were formed. Blocks misalignment and contact with the hole surface during earthquakes were considered as major causes of disturbance in scram time. Therefore, the following parameters were set up in the tests: excitation direction, combination or horizontal and vertical excitation, acceleration, frequency and block to block gaps. Main results obtained from tests are as follow. 1) Every scram time obtained under the design conditions was within 6 seconds. On the contrary, the scram times were 5.2 seconds when there were no vibration. Therefore, it was concluded that the seismic effects on scram time were not significant. 2) Scram time became longer with increase in both acceleration and horizontal excitation frequency, and control rods fell very smoothly without any jerkiness. This suggests that collision between control rods and hole surface is the main disturbing factor of falling motion. 3) Mechanical and functional integrity of control rod drive mechanism, control rods and graphite blocks was confirmed after 140 seismic scrammability tests. (author). 10 figs, 1 tab

  20. Endogenous and exogenous components in the circadian variation of core body temperature in humans

    NARCIS (Netherlands)

    Hiddinga, AE; Beersma, DGM; VandenHoofdakker, RH

    Core body temperature is predominantly modulated by endogenous and exogenous components. In the present study we tested whether these two components can be reliably assessed in a protocol which lasts for only 120 h. In this so-called forced desynchrony protocol, 12 healthy male subjects (age 23.7

  1. Development of evaluation method of fuel failure fraction during the High Temperature Engineering Test Reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Yoshimuta, Shigeharu; Tobita, Tsutomu; Sato, Masashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1997-05-01

    The High Temperature Engineering Test Reactor (HTTR) uses coated particles as fuel. During normal operation, short-lived noble gases are mainly released by diffusion from fuel particles with defects in their coating layers (i.e., failed particle). Since noble gases do not plate out on the inner surfaces of primary cooling system, their activities in primary coolant reflect fuel failure fraction in the core. An evaluation method was developed to predict failure fraction of coated fuel particles during normal operation of the HTTR. The method predicts core-average and hot plenum regionwise failure fractions based on the fractional releases, (R/B)s, of noble gases. The (R/B)s are calculated by fission gas concentration measurements in the primary cooling system of the HTTR. Recent fabrication data show that through-coatings failure fraction is extremely low. Then, fractional release from matrix contamination uranium, which is background for accurate evaluation of the fuel failure fraction, should be precisely predicted. This report describes an evaluation method of fuel failure fraction from measurements in the HTTR together with a fission gas release model from fuel compact containing failed particles and matrix contamination uranium. (author)

  2. Implication of irradiation effects on materials data for the design of near core components

    International Nuclear Information System (INIS)

    Dietz, W.; Breitling, H.

    1995-01-01

    For LWR's strict regulations exist for the consideration of irradiation in the design and surveillance of the reactor pressure vessel in the various codes (ASME, RCC-M, KTA) but less for near core components. For FBR's no firm rules exist either for the vessel nor the reactor internals. In this paper the German design practices for the loop type SNR-300 will be presented, and also some information from the surveillance programme of the KNK-reactor. Austenitic stainless steels have been mainly selected for the near core components. For some special applications Ni-alloys and a stabilized 2 1/4 Cr 1 Mo-alloy were specified. Considerations of the irradiation effects on material properties will be made for the various temperature and fluence levels around the core. The surveillance programmes will be described. Both, the consideration of irradiation effects in the elastic and inelastic analysis and the surveillance programmes had been a part of the licensing process for SNR-300. (author). 8 figs, 4 tabs

  3. Modeling Stress Strain Relationships and Predicting Failure Probabilities For Graphite Core Components

    Energy Technology Data Exchange (ETDEWEB)

    Duffy, Stephen [Cleveland State Univ., Cleveland, OH (United States)

    2013-09-09

    This project will implement inelastic constitutive models that will yield the requisite stress-strain information necessary for graphite component design. Accurate knowledge of stress states (both elastic and inelastic) is required to assess how close a nuclear core component is to failure. Strain states are needed to assess deformations in order to ascertain serviceability issues relating to failure, e.g., whether too much shrinkage has taken place for the core to function properly. Failure probabilities, as opposed to safety factors, are required in order to capture the bariability in failure strength in tensile regimes. The current stress state is used to predict the probability of failure. Stochastic failure models will be developed that can accommodate possible material anisotropy. This work will also model material damage (i.e., degradation of mechanical properties) due to radiation exposure. The team will design tools for components fabricated from nuclear graphite. These tools must readily interact with finite element software--in particular, COMSOL, the software algorithm currently being utilized by the Idaho National Laboratory. For the eleastic response of graphite, the team will adopt anisotropic stress-strain relationships available in COMSO. Data from the literature will be utilized to characterize the appropriate elastic material constants.

  4. Modeling Stress Strain Relationships and Predicting Failure Probabilities For Graphite Core Components

    International Nuclear Information System (INIS)

    Duffy, Stephen

    2013-01-01

    This project will implement inelastic constitutive models that will yield the requisite stress-strain information necessary for graphite component design. Accurate knowledge of stress states (both elastic and inelastic) is required to assess how close a nuclear core component is to failure. Strain states are needed to assess deformations in order to ascertain serviceability issues relating to failure, e.g., whether too much shrinkage has taken place for the core to function properly. Failure probabilities, as opposed to safety factors, are required in order to capture the bariability in failure strength in tensile regimes. The current stress state is used to predict the probability of failure. Stochastic failure models will be developed that can accommodate possible material anisotropy. This work will also model material damage (i.e., degradation of mechanical properties) due to radiation exposure. The team will design tools for components fabricated from nuclear graphite. These tools must readily interact with finite element software--in particular, COMSOL, the software algorithm currently being utilized by the Idaho National Laboratory. For the eleastic response of graphite, the team will adopt anisotropic stress-strain relationships available in COMSO. Data from the literature will be utilized to characterize the appropriate elastic material constants.

  5. Performance tests of the reactor containment structures of HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kawaji, Satoshi; Iyoku, Tatsuo

    1998-03-01

    The containment structures of the HTTR consist of the reactor containment vessel (CV), service area (SA) and emergency air purification system, which minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. The CV is designed to withstand the temperature and pressure transients and to be leak-tight within the specified leakage limit even in the case of a rupture of the primary concentric hot gas duct. The pressure of inside of the SA should be maintained slightly lower than that of atmosphere by the emergency air purification system. The radioactive materials are released from the stack to environment via the emergency air purification system under the accident condition. Then the emergency air purification system should remove airborne radio-activities and should maintain proper pressure in the SA. We established the method to measure leak rate of the CV with closed reactor coolant pressure boundary although it is normally measured under opened reactor coolant pressure boundary as employed in LWRs. The CV leak rate test was carried out by the newly developed method and the expected performance was obtained. The SA and emergency air purification system were also confirmed by the performance test. We concluded that the reactor containment structures were fabricated to minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. (author)

  6. Research on reactor physics using the Very High Temperature Reactor Critical Assembly (VHTRC)

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1988-01-01

    The High Temperature Engineering Test Reactor (HTTR), of which the research and development are advanced by Japan Atomic Energy Research Institute, is planned to apply for the permission of installation in fiscal year 1988, and to start the construction in the latter half of fisical year 1989. As the duty of reactor physics research, the accuracy of the nuclear data is to be confirmed, the validity of the nuclear design techniques is to be inspected, and the nuclear safety of the HTTR core design is to be verified. Therefore, by using the VHTRC, the experimental data of the reactor physics quantities are acquired, such as critical mass, the reactivity worth of simulated control rods and burnable poison rods, the temperature factor of reactivity, power distribution and so on, and the experiment and analysis are advanced. The cores built up in the VHTRC so far were three kinds having different lattice forms and degrees of uranium enrichment. The calculated critical mass was smaller by 1-5 % than the measured values. As to the power distribution and the reactivity worth of burnable poison rods, the prospect of satisfying the required accuracy for the design of the HTTR core was obtained. The experiment using a new core having axially different enrichment degree is planned. (K.I.)

  7. Construction of the HTTR and its testing program for advanced HTGR development

    International Nuclear Information System (INIS)

    Tanaka, T.; Baba, O.; Shiozawa, S.; Okubo, M.; Kunitomi, K.

    1996-01-01

    Concerning about global warming due to emission of greenhouse effect gas like CO 2 , it is essentially important to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors, because it can supply a large amount of energy and its plants emit only little amount of CO 2 during their lifetime. Hence, efforts are to be continuously devoted to establish and upgrade technologies of High Temperature Gas-cooled Reactor (HTGR) which can supply high-temperature heat with high thermal efficiency as well as high heat-utilizing efficiency. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950 deg. C at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). 2 refs, 2 figs, 1 tab., 2 photos

  8. Evaluation of aseismic integrity in the HTTR core-bottom structure. V. On the static and dynamic behavior of graphite HTTR key-keyway structures

    International Nuclear Information System (INIS)

    Futakawa, M.; Iyoku, T.

    1996-01-01

    For pt.IV see ibid., vol.154, p.83-95, 1995. The graphite components in high temperature gas-cooled reactors are connected to each other through a key-keyway structure that has gaps between the key and the keyway to accommodate thermal expansion. Because a dynamic load concentrates on the key-keyway structure during earthquakes, it is considered to be a crucial element for assessing the integrity of the graphite components. A combination of experiments and analyses was employed to investigate the dynamic behavior of the key-keyway structure, i.e. the equivalent stiffness associated with vibrational characteristics of the graphite components and the stress distribution under dynamic loading. The experiments were performed using a graphite scale model and a dynamic photo-elastic method. The analysis was carried out using the finite element method (FEM) code ABAQUS, taking account of the contact behavior between the key and the keyway. The following conclusions were derived. (1) The equivalent stiffness of the key-keyway structure shows nonlinearity, owing to the contact deformation. (2) The equivalent stiffness evaluated by the FEM analysis, taking account of the non-linear contact deformation, is applicable for predicting the vibrational characteristics of the key-keyway structure. (3) The stress concentration under dynamic loading is lower than or nearly equal to that under static loading. The maximum stress concentration of the seismic load can be sufficiently evaluated under static loading conditions. (orig.)

  9. The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

    International Nuclear Information System (INIS)

    Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

    2006-06-01

    Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

  10. Reciprocal and dynamic polarization of planar cell polarity core components and myosin

    Science.gov (United States)

    Newman-Smith, Erin; Kourakis, Matthew J; Reeves, Wendy; Veeman, Michael; Smith, William C

    2015-01-01

    The Ciona notochord displays planar cell polarity (PCP), with anterior localization of Prickle (Pk) and Strabismus (Stbm). We report that a myosin is polarized anteriorly in these cells and strongly colocalizes with Stbm. Disruption of the actin/myosin machinery with cytochalasin or blebbistatin disrupts polarization of Pk and Stbm, but not of myosin complexes, suggesting a PCP-independent aspect of myosin localization. Wash out of cytochalasin restored Pk polarization, but not if done in the presence of blebbistatin, suggesting an active role for myosin in core PCP protein localization. On the other hand, in the pk mutant line, aimless, myosin polarization is disrupted in approximately one third of the cells, indicating a reciprocal action of core PCP signaling on myosin localization. Our results indicate a complex relationship between the actomyosin cytoskeleton and core PCP components in which myosin is not simply a downstream target of PCP signaling, but also required for PCP protein localization. DOI: http://dx.doi.org/10.7554/eLife.05361.001 PMID:25866928

  11. Mechanical design of core components for a high performance light water reactor with a three pass core

    International Nuclear Information System (INIS)

    Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

    2007-01-01

    Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

  12. Prismatic Core Coupled Transient Benchmark

    International Nuclear Information System (INIS)

    Ortensi, J.; Pope, M.A.; Strydom, G.; Sen, R.S.; DeHart, M.D.; Gougar, H.D.; Ellis, C.; Baxter, A.; Seker, V.; Downar, T.J.; Vierow, K.; Ivanov, K.

    2011-01-01

    The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The benchmark-working group is currently seeking OECD/NEA sponsorship. This benchmark is being pursued and is heavily based on the success of the PBMR-400 exercise.

  13. Development of In-Service Inspection system for heat transfer tubes in the primary pressurized water cooler in the HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Shinozaki, Masayuki; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Wada, Shigeyuki

    1999-08-01

    The ISI (In-Service Inspection) system has been developed so as to maintain the structural integrity of heat transfer tubes in the primary pressurized water cooler in the HTTR (High Temperature Engineering Test Reactor). This system consists of eddy current probes, ultra-sonic probes, insertion and extraction units, positioning unit and so on. Verification and performance tests of the developed ISI system were carried out using mock-up heat transfer tubes in the primary pressurized water cooler. The constitution of the system, R and D results of the inspection probes, and verification and performance test results of the ISI system for heat transfer tubes are described in this paper. (author)

  14. Vectorization of nuclear codes 90-1

    International Nuclear Information System (INIS)

    Nonomiya, Iwao; Nemoto, Toshiyuki; Ishiguro, Misako; Harada, Hiroo; Hori, Takeo.

    1990-09-01

    The vectorization has been made for four codes: SONATINA-2V HTTR version, TRIDOSE, VIENUS, and SCRYU. SONATINA-2V HTTR version is a code for analyzing the dynamic behavior of fuel blocks in the vertical slice of the HTGR (High Temperature Gas-cooled Reactor) core under seismic perturbation, TRIDOSE is a code for calculating environmental tritium concentration and dose, VIENUS is a code for analyzing visco elastic stress of the fuel block of HTTR (High Temperature gas-cooled Test Reactor), and SCRYU is a thermal-hydraulics code with boundary fitted coordinate system. The total speedup ratio of the vectorized versions to the original scalar ones is 5.2 for SONATINA-2V HTTR version. 5.9 ∼ 6.9 for TRIDOSE, 6.7 for VIENUS, 7.6 for SCRYU, respectively. In this report, we describe outline of codes, techniques used for the vectorization, verification of computed results, and speedup effect on the vectorized codes. (author)

  15. Leak-tightness characteristics concerning the containment structures of the HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kondo, Masaaki; Emori, Koichi

    2004-01-01

    The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area. The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values. The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities

  16. Measuring the Core Components of Maladaptive Personality: Severity Indices of Personality Problems (SIPP-118)

    NARCIS (Netherlands)

    H. Andrea (Helene); R. Verheul (Roel); C.C. Berghout (Casper); C. Dolan (Conor); P.J.A. van der Kroft (Petra); A.W. Bateman (Anthony); P. Fonagy (Peter); J.J. van Busschbach (Jan)

    2007-01-01

    textabstractThis report describes a series of studies among 2231 subjects on the development of the Severity Indices for Personality Problems (SIPP), a self-report questionnaire measuring the core components of (mal)adaptive personality functioning. Results show that the 16 facets have good

  17. NGNP Fuel Qualification White Paper

    Energy Technology Data Exchange (ETDEWEB)

    David A. Petti

    2010-07-01

    The Japanese high temperature gas reactor program is centered on the High Temperature Engineering Test Reactor (HTTR), which has a thermal power of 30 MW and 950°C maximum coolant outlet temperature. The HTTR achieved criticality in November 1998 and has undergone a series of rise-to-power tests [Fujikawa 2004]. In December 2001, an outlet temperature of 850°C was achieved and in April 2004 a temperature of 950°C was achieved. As of July 2004, the reactor had operated for 224 effective full power days (EFPD). The planned core life cycle is 660 EFPD [Verfondern 2000]. It is planned to couple a high temperature process heat application to the HTTR through its intermediate heat exchanger in the future.

  18. Design and evaluation of heat utilization systems for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    2001-08-01

    The primary focus of this CRP was to perform detailed investigation of the high temperature industrial processes that are attainable through incorporation of an HTGR, and for their possible demonstration in the HTTR. The HTGR has the capability to achieve a core outlet temperature approaching 1,000 deg. C in a safe and effective manner. These attributes, coupled with the offer by JAERI to utilize the HTTR, resulted in the initiation of this CRP by the IAEA. High Temperature Engineering Test Reactor (HTTR) utilizes a 30 MW(th) HTGR comprised of 30 fuel columns of hexagonal pin-in-pin graphite block type fuel elements. The fuel consists of UO 2 TRISO coated particles with an enrichment of ∼ 6% wt. Relative to the demonstration of high temperature heat applications, the HTTR will be capable of producing 10 MW(th) of heat at 950 deg. C. However, the thermal power for these applications has the potential to be increased up to 30 MW(th) in the future, which may be required for demonstration of gas turbine system components. The HTTR reached initial criticality in November 1998. Initial operational plans includes a series of rise to power tests followed by tests to demonstrate the safety and operational characteristics of the HTTR. In addition to completion of the HTTR demonstration tests, it was recommended that the R and D be performed within the HTTR project. JAERI is encouraged to publicize the results of the HTTR tests and 'lessons learned' from their experiences including potential capabilities of the HTGR for heat applications. The next priority application was determined to be the generation of electricity through the use of the gas turbine. Application of the Brayton Cycle utilizing high temperature helium from a modular HTGR was chosen for development because of its projected benefits as an economic and efficient means for the production of electricity. Evaluation of the remaining high temperature heat utilization applications chosen for investigation resulted

  19. Feasibility study on silicon doping using high temperature test engineering reactor

    International Nuclear Information System (INIS)

    Seki, Masaya; Takaki, Naoyuki; Goto, Minoru; Shimakawa, Satoshi

    2011-01-01

    The feasibility study on silicon doping using the High Temperature engineering Test Reactor (HTTR) is performed by numerical simulations. The HTTR is a High Temperature Gas-cooled Reactor (HTGR) situated at JAEA Oarai research and development center. It has a 30MW thermal power and the outlet coolant temperature is 950degC. The objective of this study is to evaluate the following issues, 1. The impact of loading Si-ingots into the core on the criticality, 2. The uniformity of the neutron capture reaction rate in Si-ingots, and 3. The production rate of silicon semiconductor. In this study, six Si-ingots are loaded into the irradiation area which is located in the peripheral region of the core. They are irradiated with rotation movement around the axial direction to obtain uniform neutron capture reaction rate in the radial direction. Additionally, the neutron filter, which is made of graphite containing boron, is used to obtain uniform neutron capture reaction rate in the axial direction. The evaluations were conducted by performing the HTTR whole core calculations with the Monte Carlo code MVP-2.0. In the calculations, several tally regions were defined on the Si-ingots to investigate the uniformity of the neutron capture reaction rate. As a result, loading the Si-ingots into the core causes negative reactivity by about 0.7%dk/k. Uniform neutron capture reaction rate of Si-ingot is obtained 98% in the radial and the axial direction. In case of the target of semiconductor resistivity is set to 50 Ωcm, the required irradiation time becomes 10 hours. The HTTR is able to produce silicon semiconductor of 540kg in one-time irradiation. This study was conducted as a joint research with JAEA, Nuclear Fuel Industries, LTD, Toyota Tsusho Corporation and Tokai University. (author)

  20. Soundness confirmation through cold test of the system equipment of HTTR

    International Nuclear Information System (INIS)

    Ono, Masato; Shinohara, Masanori; Iigaki, Kazuhiko; Tochio, Daisuke; Nakagawa, Shigeaki; Shimazaki, Yosuke

    2014-01-01

    HTTR was established at the Oarai Research and Development Center of Japan Atomic Energy Agency, for the purpose of the establishment and upgrading of high-temperature gas-cooled reactor technology infrastructure. Currently, it performs a safety demonstration test in order to demonstrate the safety inherent in high-temperature gas-cooled reactor. After the Great East Japan Earthquake, it conducted confirmation test for the purpose of soundness survey of facilities and equipment, and it confirmed that the soundness of the equipment was maintained. After two years from the confirmation test, it has not been confirmed whether the function of dynamic equipment and the soundness such as the airtightness of pipes and containers are maintained after receiving the influence of damage or deterioration caused by aftershocks generated during two years or aging. To confirm the soundness of these facilities, operation under cold state was conducted, and the obtained plant data was compared with confirmation test data to evaluate the presence of abnormality. In addition, in order to confirm through cold test the damage due to aftershocks and degradation due to aging, the plant data to compare was supposed to be the confirmation test data, and the evaluation on abnormality of the plant data of machine starting time and normal operation data was performed. (A.O.)

  1. User's manual for seismic analysis code 'SONATINA-2V'

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Iyoku, Tatsuo

    2001-08-01

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

  2. User's manual for seismic analysis code 'SONATINA-2V'

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, Satoshi; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-08-01

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

  3. Development of Structural Core Components for Breeder Reactors

    International Nuclear Information System (INIS)

    Saibaba, N.

    2013-01-01

    Core structural materials: • The desire is to have only fuel in the core, structural material form 25% of the total core: – To support and to retain the fuel in position; – Provide necessary ducts to make coolant flow through & transfer/remove heat. • For 500 MWe FBR with Oxide fuel (Peak Linear Power 450 W/cm), total fuel pins required in the core are of the order 39277 pins (both inner & outer core Fuel SA); • Considering 217 pins/Fuel SA there are 181 Fuel SA wrapper tubes • These structural materials see hostile core with max temperature and neutron flux

  4. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    International Nuclear Information System (INIS)

    Griffiths, M.; Coleman, C.E.; Holt, R.A.; Sagat, S.; Urbanic, V.F.; Chow, C.K.

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of 'breakaway' growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs

  5. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, M; Coleman, C E; Holt, R A; Sagat, S; Urbanic, V F [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Chow, C K [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and {beta}-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the {alpha}-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of `breakaway` growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs.

  6. Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

    Energy Technology Data Exchange (ETDEWEB)

    Fok, Alex

    2013-10-30

    The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the model to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.

  7. Investigation on cause of outage of Wide Range Monitor (WRM) in High Temperature engineering Test Reactor (HTTR). Post Irradiation Examination (PIE) toward investigation of the cause

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Saito, Kenji; Takada, Shoji; Ishimi, Akihiro; Katsuyama, Kozo; Motegi, Toshihiro

    2012-08-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the outage by specifying the damaged part causing the event in the WRM. The one is a post irradiation examination using the X-ray computed tomography scanner in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. The other is an experiment using a mock-up simulating the WRM fabricated by the fabricator. The characteristic impedance of the damaged WRM was measured by Time Domain Reflectometry, which was compared with that of the mock-up, which could narrow down the damaged part in the WRM. This report summarized the results of the PIE and the experimental investigation using the mock-up to reveal the cause of outage of WRM. (author)

  8. Study on nuclear analysis method for high temperature gas-cooled reactor and its nuclear design (Thesis)

    International Nuclear Information System (INIS)

    Goto, Minoru

    2015-03-01

    An appropriate configuration of fuel and reactivity control equipment in a nuclear reactor core, which allows the design of the nuclear reactor core for low cost and high performance, is performed by nuclear design with high accuracy. The accuracy of nuclear design depends on a nuclear data library and a nuclear analysis method. Additionally, it is one of the most important issues for the nuclear design of a High Temperature Gas-cooled Reactor (HTGR) that an insertion depth of control rods into the reactor core should be retained shallow by reducing excess reactivity with a different method to keep fuel temperature below its limitation thorough a burn-up period. In this study, using experimental data of the High Temperature engineering Test Reactor (HTTR), which is a Japan's HTGR with 30 MW of thermal power, the following issues were investigated: applicability of nuclear data libraries to nuclear analysis for HTGRs; applicability of the improved nuclear analysis method for HTGRs; and effectiveness of a rod-type burnable poison on HTGR reactivity control. A nuclear design of a small-sized HTGR with 50 MW of thermal power (HTR50S) was performed using these results. In the nuclear design of the HTR50S, we challenged to decrease the kinds of the fuel enrichments and to increase the power density compared with the HTTR. As a result, the nuclear design was completed successfully by reducing the kinds of the fuel enrichment to only three from twelve of the HTTR and increasing the power density by 1.4 times as much as that of the HTTR. (author)

  9. Method for a reliable activation calculation of core components; Methode zur zuverlaessigen Berechnung von Aktivierungen in Kernbauteilen

    Energy Technology Data Exchange (ETDEWEB)

    Mispagel, T.; Phlippen, P.W.; Rose, J. [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany)

    2013-07-01

    During nuclear power plant operation components and materials are exposed to the neutron flux from the reactor core and radionuclides are produced. After removal of the fuel elements the radioactivity of these radionuclides in the reactor pressure vessel and the core internals provide more than 99% of the activity of the power plant. For the transport, the interim storage and the final disposal of these radioactive components the radioactive inventories have to be decoded with respect to radiation and nuclides. The declaration of the nuclide and activity inventories requires a reliable calculation of neutron induced activation of reactor components. These activation calculations describe the pile-up of nuclides due to irradiation and due to the decay of nuclides. For an optimum usage of the activity capacities of the repository Konrad it is necessary to have a qualified calculation procedure that keeps the conservatism as low as possible.

  10. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

  11. Overview of HTGR heat utilization system development at JAERI

    International Nuclear Information System (INIS)

    Miyamoto, Y.; Shiozawa, S.; Ogawa, M.; Akino, N.; Shimizu, S.; Hada, K.; Inagaki, Y.; Onuki, K.; Takeda, T.; Nishihara, T.

    1998-01-01

    The Japan Atomic Energy Research Institute (JAERI) has conducted research and development of nuclear heat utilization systems of a High Temperature Gas cooled Reactor (HTGR), which are capable to meet a large amount of energy demand without significant CO 2 emission to relax the global warming issue. The High Temperature engineering Test Reactor (HTTR) with thermal output of 30 MW and outlet coolant temperature of 950 deg C, the first HTGR in Japan, is under construction on the JAERI site, and its first criticality is scheduled for mid-1998. After the reactor performance and safety demonstration tests for several years, a hydrogen production system will be connected to the HTTR. A demonstration program on hydrogen production started in January 1997, in JAERI, as a study consigned by the Science and Technology Agency. A hydrogen production system connected to the HTTR is designed to be able to produce hydrogen by steam reforming of natural gas, using nuclear heat of 10 MW from the HTTR. The safety principle and standard are investigated for the HTTR hydrogen production system. In order to confirm safety, controllability and performance of key components in the HTTR hydrogen production system, an out-of-pile test facility on the scale of approximately 1/30 of the HTTR hydrogen production system is installed. It is equipped with an electric heater as a heat source instead of the HTTR. The out-of-pile test will be performed for four years after 2001. The HTTR hydrogen production system will be demonstratively operated after 2005 at its earliest plan. Other basic studies on the hydrogen production system using thermochemical water splitting, an iodine sulphur (IS) process, and technology of distant heat transport with microencapsulated phase change material have been carried out for more effective and various uses of nuclear heat. (author)

  12. Research and development for the high-temperature helium-leak detection system (Joint research). Part 2. Development of temperature sensors using optical fibre for the HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Sakaba, Nariaki; Nakazawa, Toshio; Kawasaki, Kozo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Urakami, Masao; Saisyu, Sadanori [Japan Atomic Power Co., Tokyo (Japan)

    2003-03-01

    In the second stage of the research and development for a high-temperature helium-leak detection system, the temperature sensor using optical fibres was studied. The sensor detects the helium leakage by the temperature increase surrounded optical fibre with or without heat insulator. Moreover, the applicability of high temperature equipments as the HTTR system was studied. With the sensor we detected 5.0-20.0 cm{sup 3}/s helium leakages within 60 minutes. Also it was possible to detect earlier when the leakage level is at 20.0 cm {sup 3}/s. (author)

  13. Reaction- and melting behaviour of LWR-core components UO2, Zircaloy and steel during the meltdown period

    International Nuclear Information System (INIS)

    Hofmann, P.

    1976-07-01

    The reaction behaviour of the UO 2 , Zircaloy-4 and austenitic steel core components was investigated as a function of temperature (till melting temperatures) under inert and oxidizing conditions. Component concentrations varied between that of Corium-A (65 wt.% UO 2 , 18% Zry, 17% steel) and that of Corium-E (35 wt.% UO 2 , 10% Zry, 55% steel). In addition, Zircaloy and stainless steel were used with different degrees of oxidation. The paper describes systematically the phases that arise during heating and melting. The integral composition of the melts and the qualitative as well as quantitative analysis of the phases present in solidified corium are given. In some cases melting points have been determined. The reaction and melting behaviour of the corium specimens strongly depends on the concentration and on the degree of oxidation of the core components. First liquid phases are formed at the Zry-steel interface at about 1,350 0 C. The maximum temperatures of about 2,500 0 C for the complete melting of the corium-specimens are well below the UO 2 melting point. Depending on the steel content and/or degree of oxidation of Zry and steel, a homogeneous metallic or oxide melt or two immiscible melts - one oxide and the other metallic - are obtained. During the melting experiments performed under inert gas conditions the chemical composition of the molten specimens generally change by evaporation losses of single elements, especially of uranium, zirconium and oxygen. The total weight losses go up to 30%; under oxidizing conditions they are substantially smaller due to the occurrence of different phases. In air or water vapor, the occurrence of the phases and the melting behaviour of the core components are strongly influenced by the oxidation rate and the oxygen supply to the surface of the melt. In the case of the hypothetical core melting accident, a heterogeneous melt (oxide and metallic) is probable after the meltdown period. (orig./RW) [de

  14. Outlines and verifications of the codes used in the safety analysis of High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Shiina, Yasuaki; Kunitomi, Kazuhiko; Maruyama, Soh; Fujita, Shigeki; Nakagawa, Shigeaki; Iyoku, Tatsuo; Shindoh, Masami; Sudo, Yukio; Hirano, Masashi.

    1990-03-01

    This paper presents brief description of the computer codes used in the safety analysis of High Temperature Engineering Test Reactor. The list of the codes is: 1. BLOOST-J2 2. THYDE-HTGR 3. TAC-NC 4. RATSAM6 5. COMPARE-MOD1 6. GRACE 7. OXIDE-3F 8. FLOWNET/TRUMP. Of described above, 1, 3, 4, 5, 6 and 7 were developed for the multi-hole type gas cooled reactor and improved for HTTR and 2 was originated by THYDE-codes which were developed to treat the transient thermo-hydraulics during LOCA of LWR. Each code adopted the models and properties which yield conservative analytical results. Adequacy of each code was verified by the comparison with the experimental results and/or the analytical results obtained from the other codes which were already proven. (author)

  15. Experience with lifetime limits for EBR-II core components

    International Nuclear Information System (INIS)

    Lambert, J.D.B.; Smith, R.N.; Golden, G.H.

    1987-01-01

    The Experimental Breeder Reactor No. 2 (EBR-II) is operated for the US Department of Energy by Argonne National Laboratory and is located on the Idaho National Engineering Laboratory where most types of American reactor were originally tested. EBR-II is a complete electricity-producing power plant now in its twenty-fourth year of successful operation. During this long history the reactor has had several concurrent missions, such as demonstration of a closed Liquid-Metal Reactor (LMR) fuel cycle (1964-69); as a steady-state irradiation facility for fuels and materials (1970 onwards); for investigating effects of operational transients on fuel elements (from 1981); for research into the inherent safety aspects of metal-fueled LMR's (from 1983); and, most recently, for demonstration of the Integral Fast Reactor (IFR) concept using U-Pu-Zr fuels. This paper describes experience gained at EBR-II in defining lifetime limits for LMR core components, particularly fuel elements

  16. EXPERIMENTAL DETERMINATION OF LONGITUDINAL COMPONENT OF MAGNETIC FLUX IN FERROMAGNETIC WIRE OF SINGLE-CORE POWER CABLE ARMOUR

    Directory of Open Access Journals (Sweden)

    I.A. Kostiukov

    2014-12-01

    Full Text Available A problem of determination of effective longitudinal magnetic permeability of single core power cable armour is defined. A technique for experimental determination of longitudinal component of magnetic flux in armour spiral ferromagnetic wire is proposed.

  17. Two-component Superfluid Hydrodynamics of Neutron Star Cores

    Energy Technology Data Exchange (ETDEWEB)

    Kobyakov, D. N. [Institute of Applied Physics of the Russian Academy of Sciences, 603950 Nizhny Novgorod (Russian Federation); Pethick, C. J., E-mail: dmitry.kobyakov@appl.sci-nnov.ru, E-mail: pethick@nbi.dk [The Niels Bohr International Academy, The Niels Bohr Institute, University of Copenhagen, Blegdamsvej 17, DK-2100 Copenhagen Ø (Denmark)

    2017-02-20

    We consider the hydrodynamics of the outer core of a neutron star under conditions when both neutrons and protons are superfluid. Starting from the equation of motion for the phases of the wave functions of the condensates of neutron pairs and proton pairs, we derive the generalization of the Euler equation for a one-component fluid. These equations are supplemented by the conditions for conservation of neutron number and proton number. Of particular interest is the effect of entrainment, the fact that the current of one nucleon species depends on the momenta per nucleon of both condensates. We find that the nonlinear terms in the Euler-like equation contain contributions that have not always been taken into account in previous applications of superfluid hydrodynamics. We apply the formalism to determine the frequency of oscillations about a state with stationary condensates and states with a spatially uniform counterflow of neutrons and protons. The velocities of the coupled sound-like modes of neutrons and protons are calculated from properties of uniform neutron star matter evaluated on the basis of chiral effective field theory. We also derive the condition for the two-stream instability to occur.

  18. The assessment of helium purification system of small power HTGR

    International Nuclear Information System (INIS)

    Siti Alimah; Sriyono

    2016-01-01

    The helium purification system (HPS) is one of safety system of High Temperature Gas-cooled Reactor. HPS removes impurities in the primary coolant, so that the impact on structure, system and component (SSC) is minimized. The two impurity types are particulates (carbon dust, fission products (Kr, Xe, Cs etc.) and the gases (O_2, N_2, H_2O, CH_4, CO, CO_2 and H_2). Every reactor has a different impurity limit during normal operation, depends on the reactor power, energy conversion system and fuel type. This paper discusses the HPS on HTR-10, HTTR and Indonesian RDE conceptual design. The purpose of this assessment is to determine the optimum HPS design as a role model for Indonesian RDE. The utilized methodology is a literature study based on the operating experiences of both HTR-10 and HTTR as well as the evaluation of RDE conceptual design. This study focuses on the impurities limit during normal operation, the main components of HPS, mass flow-rate and regeneration process. The main component that used in HPS for HTR-10, HTTR and RDE are similar i.e. filter, CuO column, water cooler, molecular sieve bed and cryogenic activated carbon bed. Refer to the HTR-10 and HTTR operational experiences, both of those reactors have a purification systems that capable to maintain the helium purity, even though the impurities limit are different. The HPS of HTTR Japan has a stricter impurities limit that N_2, CH_4, and O_2 should not be contained at all during normal operation and the pre-charcoal trap is used to adsorb the fine dust below 0.1 micron. Both of these parameters can be adopted to the RDE's HPS design to minimize the effect of impurities to SSC. (author)

  19. Artificial neural networks for dynamic monitoring of simulated-operating parameters of high temperature gas cooled engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Seker, Serhat; Tuerkcan, Erdinc; Ayaz, Emine; Barutcu, Burak

    2003-01-01

    This paper addresses to the problem of utilisation of the artificial neural networks (ANNs) for detecting anomalies as well as physical parameters of a nuclear power plant during power operation in real time. Three different types of neural network algorithms were used namely, feed-forward neural network (back-propagation, BP) and two types of recurrent neural networks (RNN). The data used in this paper were gathered from the simulation of the power operation of the Japan's High Temperature Engineering Testing Reactor (HTTR). For the wide range of power operation, 56 signals were generated by the reactor dynamic simulation code for several hours of normal power operation at different power ramps between 30 and 100% nominal power. Paper will compare the outcomes of different neural networks and presents the neural network system and the determination of physical parameters from the simulated operating data

  20. Haploinsufficiency for Core Exon Junction Complex Components Disrupts Embryonic Neurogenesis and Causes p53-Mediated Microcephaly.

    Directory of Open Access Journals (Sweden)

    Hanqian Mao

    2016-09-01

    Full Text Available The exon junction complex (EJC is an RNA binding complex comprised of the core components Magoh, Rbm8a, and Eif4a3. Human mutations in EJC components cause neurodevelopmental pathologies. Further, mice heterozygous for either Magoh or Rbm8a exhibit aberrant neurogenesis and microcephaly. Yet despite the requirement of these genes for neurodevelopment, the pathogenic mechanisms linking EJC dysfunction to microcephaly remain poorly understood. Here we employ mouse genetics, transcriptomic and proteomic analyses to demonstrate that haploinsufficiency for each of the 3 core EJC components causes microcephaly via converging regulation of p53 signaling. Using a new conditional allele, we first show that Eif4a3 haploinsufficiency phenocopies aberrant neurogenesis and microcephaly of Magoh and Rbm8a mutant mice. Transcriptomic and proteomic analyses of embryonic brains at the onset of neurogenesis identifies common pathways altered in each of the 3 EJC mutants, including ribosome, proteasome, and p53 signaling components. We further demonstrate all 3 mutants exhibit defective splicing of RNA regulatory proteins, implying an EJC dependent RNA regulatory network that fine-tunes gene expression. Finally, we show that genetic ablation of one downstream pathway, p53, significantly rescues microcephaly of all 3 EJC mutants. This implicates p53 activation as a major node of neurodevelopmental pathogenesis following EJC impairment. Altogether our study reveals new mechanisms to help explain how EJC mutations influence neurogenesis and underlie neurodevelopmental disease.

  1. Performance test results of helium gas circulator of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Akira; Kato, Michio; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2003-03-01

    Hydrogen production system by steam reforming of methane will be connected to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) against development of nuclear heat utilization system. To obtain design and safety database of the HTTR hydrogen production system, mock-up test facility with full-scale reaction was constructed in FY 2001 and hydrogen of 120m{sup 3}N{sub /}h was successfully produced in overall performance test. This report describes performance test results of a helium gas circulator in this facility. The circulator performance curves regarding to pressure-rise, input power and adiabatic thermal efficiency at standard revolution number were made based on the measured flow-rate, temperature and pressure data in overall performance test. The circulator performance prediction code was made based on these performance curves. The code can calculate revolution number, electric power and temperature-rise of the circulator using flow-rate, inlet temperature, inlet pressure and pressure-rise data. The verification of the code was carried out with the test data in FY 2002. Total pressure loss of the helium gas circulation loop was also evaluated. The circulator should be operated in conditions such as pressure from 2.7MPa to 4.0MPa and flow-rate from 250g/s to 400g/s and at maximum pressure-rise of 250 kPa in test operation. It was confirmed in above verification and evaluations that the circulator had performance to satisfy above conditions within operation limitation of the circulator such as maximum input-power of 150 kW and maximum revolution number of 12,000 rpm. (author)

  2. Work-related musculoskeletal disorders (WMDs) risk assessment at core assembly production of electronic components manufacturing company

    Science.gov (United States)

    Yahya, N. M.; Zahid, M. N. O.

    2018-03-01

    This study conducted to assess the work-related musculoskeletal disorders (WMDs) among the workers at core assembly production in an electronic components manufacturing company located in Pekan, Pahang, Malaysia. The study is to identify the WMDs risk factor and risk level. A set of questionnaires survey based on modified Nordic Musculoskeletal Disorder Questionnaires have been distributed to respective workers to acquire the WMDs risk factor identification. Then, postural analysis was conducted in order to measure the respective WMDs risk level. The analysis were based on two ergonomics assessment tools; Rapid Upper Limb Assessment (RULA) and Rapid Entire Body Assessment (REBA). The study found that 30 respondents out of 36 respondents suffered from WMDs especially at shoulder, wrists and lower back. The WMDs risk have been identified from unloading process, pressing process and winding process. In term of the WMDs risk level, REBA and RULA assessment tools have indicated high risk level to unloading and pressing process. Thus, this study had established the WMDs risk factor and risk level of core assembly production in an electronic components manufacturing company at Malaysia environment.

  3. A theoretical analysis of the response of an air-cored eddy current coil for remote oxide thickness measurements on reactor components

    International Nuclear Information System (INIS)

    Wilson, R.

    1979-10-01

    It is shown how the impedance of an air-cored eddy current coil in close proximity to an oxidised steel component may be calculated. Representative values were selected for the oxide thickness, lift off, operating frequency, conductivities and permeabilities of the oxide coating and steel base. The values of these parameters in the calculations were allowed to vary between suitable limits to quantify the effect of each one on coil impedance. The results of the calculations are used to determine the most suitable conditions for the measurement of oxide thickness on steel components using an air-cored eddy current probe. (author)

  4. Improvement works report on mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Sakaki, Akihiro; Kato, Michio; Hayashi, Koji; Fujisaki, Katsuo; Aita, Hideki; Ohashi, Hirofumi; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Maeda, Yukimasa; Sato, Hiroyuki; Hanawa, Hiromi; Yonekawa, Hideo; Inagaki, Yoshiyuki

    2005-04-01

    In order to establish the system integration technology to connect a hydrogen production system to a high temperature gas cooled reactor; the mock-up test facility with a full-scale reaction tube for the steam reforming HTTR hydrogen production system was constructed in fiscal year 2001 and its functional test operation was performed in the year. Seven experimental test operations were performed from fiscal year 2001 to 2004. On a period of each test operation, there happened some troubles. For each trouble, the cause was investigated and the countermeasures and the improvement works were performed to succeed the experiments. The tests were successfully achieved according to plan. This report describes the improvement works on the test facility performed from fiscal year 2001 to 2004. (author)

  5. Mechanical design philosophy for the graphite components of the core structure of an HTGR

    International Nuclear Information System (INIS)

    Bodmann, E.

    1987-01-01

    Parallel to the layout and design of the graphite components for THTRs and the succeeding high temperature reactor projects, the design methods for graphite components have been improved over the years. The aim of this works is to develop the design methods which take into account both the particular properties of graphite and the particular functions of the components. Because of the close relation ship between materials and design codes, this development work has progressed with the development, testing and qualification of German reactor graphite. In this paper, the experience in this field of Hochtemperatur Reaktorbau GmbH and the results of the work and approach to the design problems are reported. The example of a HTR 500 design for a 550 MWe power station is taken up, and the core structure is explained. The graphite components are divided into three classes according to the stress limits. The loading of these components is reviewed. The aim of the design is not the complete avoidance of failure, but to avoid the failure of a single component from leading to a disadvantageous consequence which is not allowable. The classification of loading events, Weibull statistics and maximum allowable stress, the formation of the permissible stress, the assessment of stress due to multiaxial loading and so on are described. (Kako, I.)

  6. Exploring cosmic origins with CORE: B-mode component separation

    Science.gov (United States)

    Remazeilles, M.; Banday, A. J.; Baccigalupi, C.; Basak, S.; Bonaldi, A.; De Zotti, G.; Delabrouille, J.; Dickinson, C.; Eriksen, H. K.; Errard, J.; Fernandez-Cobos, R.; Fuskeland, U.; Hervías-Caimapo, C.; López-Caniego, M.; Martinez-González, E.; Roman, M.; Vielva, P.; Wehus, I.; Achucarro, A.; Ade, P.; Allison, R.; Ashdown, M.; Ballardini, M.; Banerji, R.; Bartlett, J.; Bartolo, N.; Baumann, D.; Bersanelli, M.; Bonato, M.; Borrill, J.; Bouchet, F.; Boulanger, F.; Brinckmann, T.; Bucher, M.; Burigana, C.; Buzzelli, A.; Cai, Z.-Y.; Calvo, M.; Carvalho, C.-S.; Castellano, G.; Challinor, A.; Chluba, J.; Clesse, S.; Colantoni, I.; Coppolecchia, A.; Crook, M.; D'Alessandro, G.; de Bernardis, P.; de Gasperis, G.; Diego, J.-M.; Di Valentino, E.; Feeney, S.; Ferraro, S.; Finelli, F.; Forastieri, F.; Galli, S.; Genova-Santos, R.; Gerbino, M.; González-Nuevo, J.; Grandis, S.; Greenslade, J.; Hagstotz, S.; Hanany, S.; Handley, W.; Hernandez-Monteagudo, C.; Hills, M.; Hivon, E.; Kiiveri, K.; Kisner, T.; Kitching, T.; Kunz, M.; Kurki-Suonio, H.; Lamagna, L.; Lasenby, A.; Lattanzi, M.; Lesgourgues, J.; Lewis, A.; Liguori, M.; Lindholm, V.; Luzzi, G.; Maffei, B.; Martins, C. J. A. P.; Masi, S.; Matarrese, S.; McCarthy, D.; Melin, J.-B.; Melchiorri, A.; Molinari, D.; Monfardini, A.; Natoli, P.; Negrello, M.; Notari, A.; Paiella, A.; Paoletti, D.; Patanchon, G.; Piat, M.; Pisano, G.; Polastri, L.; Polenta, G.; Pollo, A.; Poulin, V.; Quartin, M.; Rubino-Martin, J.-A.; Salvati, L.; Tartari, A.; Tomasi, M.; Tramonte, D.; Trappe, N.; Trombetti, T.; Tucker, C.; Valiviita, J.; Van de Weijgaert, R.; van Tent, B.; Vennin, V.; Vittorio, N.; Young, K.; Zannoni, M.

    2018-04-01

    We demonstrate that, for the baseline design of the CORE satellite mission, the polarized foregrounds can be controlled at the level required to allow the detection of the primordial cosmic microwave background (CMB) B-mode polarization with the desired accuracy at both reionization and recombination scales, for tensor-to-scalar ratio values of rgtrsim 5× 10‑3. We consider detailed sky simulations based on state-of-the-art CMB observations that consist of CMB polarization with τ=0.055 and tensor-to-scalar values ranging from r=10‑2 to 10‑3, Galactic synchrotron, and thermal dust polarization with variable spectral indices over the sky, polarized anomalous microwave emission, polarized infrared and radio sources, and gravitational lensing effects. Using both parametric and blind approaches, we perform full component separation and likelihood analysis of the simulations, allowing us to quantify both uncertainties and biases on the reconstructed primordial B-modes. Under the assumption of perfect control of lensing effects, CORE would measure an unbiased estimate of r=(5 ± 0.4)× 10‑3 after foreground cleaning. In the presence of both gravitational lensing effects and astrophysical foregrounds, the significance of the detection is lowered, with CORE achieving a 4σ-measurement of r=5× 10‑3 after foreground cleaning and 60% delensing. For lower tensor-to-scalar ratios (r=10‑3) the overall uncertainty on r is dominated by foreground residuals, not by the 40% residual of lensing cosmic variance. Moreover, the residual contribution of unprocessed polarized point-sources can be the dominant foreground contamination to primordial B-modes at this r level, even on relatively large angular scales, l ~ 50. Finally, we report two sources of potential bias for the detection of the primordial B-modes by future CMB experiments: (i) the use of incorrect foreground models, e.g. a modelling error of Δβs = 0.02 on the synchrotron spectral indices may result in an

  7. Present status of HTGR research and development, 1995

    International Nuclear Information System (INIS)

    1996-02-01

    Based on the Long-term Program for Development and Utilization of Nuclear Energy which was revised in 1987, the Japan Atomic Energy Research Institute (JAERI) has carried out the Research and Development (R and D) on the High Temperature Gas-cooled Reactors (HTGRs) in Japan. The JAERI obtained the installation permit of the High Temperature Engineering Test Reactor (HTTR) from the Government in November 1990 and started the construction of the HTTR facility in the Oarai Research Establishment in March 1991. The HTTR is a test reactor with thermal output of 30MW and outlet coolant temperature of 850degC at the rated operation and 950degC at the high temperature test operation, using the pin-in-block type fuel, and has capability to demonstrate nuclear process heat utilization. The reactor pressure vessel and intermediate heat exchanger were installed in the reactor containment vessel in 1994, and reactor internals were also installed in the reactor pressure vessel in 1995. The first criticality will be attained in December 1997. This report describes the design outline and construction progress of the HTTR, R and D of fuel, materials and components for the HTGR and high temperature nuclear heat application, and innovative and basic researches for high temperature technologies at the HTTR. (J.P.N.)

  8. Component mode synthesis methods applied to 3D heterogeneous core calculations, using the mixed dual finite element solver MINOS

    Energy Technology Data Exchange (ETDEWEB)

    Guerin, P.; Baudron, A. M.; Lautard, J. J. [Commissariat a l' Energie Atomique, DEN/DANS/DM2S/SERMA/LENR, CEA Saclay, 91191 Gif sur Yvette (France)

    2006-07-01

    This paper describes a new technique for determining the pin power in heterogeneous core calculations. It is based on a domain decomposition with overlapping sub-domains and a component mode synthesis technique for the global flux determination. Local basis functions are used to span a discrete space that allows fundamental global mode approximation through a Galerkin technique. Two approaches are given to obtain these local basis functions: in the first one (Component Mode Synthesis method), the first few spatial eigenfunctions are computed on each sub-domain, using periodic boundary conditions. In the second one (Factorized Component Mode Synthesis method), only the fundamental mode is computed, and we use a factorization principle for the flux in order to replace the higher order Eigenmodes. These different local spatial functions are extended to the global domain by defining them as zero outside the sub-domain. These methods are well-fitted for heterogeneous core calculations because the spatial interface modes are taken into account in the domain decomposition. Although these methods could be applied to higher order angular approximations - particularly easily to a SPN approximation - the numerical results we provide are obtained using a diffusion model. We show the methods' accuracy for reactor cores loaded with UOX and MOX assemblies, for which standard reconstruction techniques are known to perform poorly. Furthermore, we show that our methods are highly and easily parallelizable. (authors)

  9. Component mode synthesis methods applied to 3D heterogeneous core calculations, using the mixed dual finite element solver MINOS

    International Nuclear Information System (INIS)

    Guerin, P.; Baudron, A. M.; Lautard, J. J.

    2006-01-01

    This paper describes a new technique for determining the pin power in heterogeneous core calculations. It is based on a domain decomposition with overlapping sub-domains and a component mode synthesis technique for the global flux determination. Local basis functions are used to span a discrete space that allows fundamental global mode approximation through a Galerkin technique. Two approaches are given to obtain these local basis functions: in the first one (Component Mode Synthesis method), the first few spatial eigenfunctions are computed on each sub-domain, using periodic boundary conditions. In the second one (Factorized Component Mode Synthesis method), only the fundamental mode is computed, and we use a factorization principle for the flux in order to replace the higher order Eigenmodes. These different local spatial functions are extended to the global domain by defining them as zero outside the sub-domain. These methods are well-fitted for heterogeneous core calculations because the spatial interface modes are taken into account in the domain decomposition. Although these methods could be applied to higher order angular approximations - particularly easily to a SPN approximation - the numerical results we provide are obtained using a diffusion model. We show the methods' accuracy for reactor cores loaded with UOX and MOX assemblies, for which standard reconstruction techniques are known to perform poorly. Furthermore, we show that our methods are highly and easily parallelizable. (authors)

  10. Core Components for a Clinically Integrated mHealth App for Asthma Symptom Monitoring.

    Science.gov (United States)

    Rudin, Robert S; Fanta, Christopher H; Predmore, Zachary; Kron, Kevin; Edelen, Maria O; Landman, Adam B; Zimlichman, Eyal; Bates, David W

    2017-10-01

    Background mHealth apps may be useful tools for supporting chronic disease management. Objective Our aim was to apply user-centered design principles to efficiently identify core components for an mHealth-based asthma symptom–monitoring intervention using patient-reported outcomes (PROs). Methods We iteratively combined principles of qualitative research, user-centered design, and “gamification” to understand patients' and providers' needs, develop and refine intervention components, develop prototypes, and create a usable mobile app to integrate with clinical workflows. We identified anticipated benefits and burdens for stakeholders. Results We conducted 19 individual design sessions with nine adult patients and seven clinicians from an academic medical center (some were included multiple times). We identified four core intervention components: (1) Invitation—patients are invited by their physicians. (2) Symptom checks—patients receive weekly five-item questionnaires via the app with 48 hours to respond. Depending on symptoms, patients may be given the option to request a call from a nurse or receive one automatically. (3) Patient review—in the app, patients can view their self-reported data graphically. (4) In-person visit—physicians have access to patient-reported symptoms in the electronic health record (EHR) where they can review them before in-person visits. As there is currently no location in the EHR where physicians would consistently notice these data, recording a recent note was the best option. Benefits to patients may include helping decide when to call their provider and facilitating shared decision making. Benefits to providers may include saving time discussing symptoms. Provider organizations may need to pay nurses extra, but those costs may be offset by reduced visits and hospitalizations. Conclusion Recent systematic reviews show inconsistent outcomes and little insight into functionalities required for mHealth asthma

  11. Casting core for a cooling arrangement for a gas turbine component

    Science.gov (United States)

    Lee, Ching-Pang; Heneveld, Benjamin E

    2015-01-20

    A ceramic casting core, including: a plurality of rows (162, 166, 168) of gaps (164), each gap (164) defining an airfoil shape; interstitial core material (172) that defines and separates adjacent gaps (164) in each row (162, 166, 168); and connecting core material (178) that connects adjacent rows (170, 174, 176) of interstitial core material (172). Ends of interstitial core material (172) in one row (170, 174, 176) align with ends of interstitial core material (172) in an adjacent row (170, 174, 176) to form a plurality of continuous and serpentine shaped structures each including interstitial core material (172) from at least two adjacent rows (170, 174, 176) and connecting core material (178).

  12. Nonspecific Organelle-Targeting Strategy with Core-Shell Nanoparticles of Varied Lipid Components/Ratios.

    Science.gov (United States)

    Zhang, Lu; Sun, Jiashu; Wang, Yilian; Wang, Jiancheng; Shi, Xinghua; Hu, Guoqing

    2016-07-19

    We report a nonspecific organelle-targeting strategy through one-step microfluidic fabrication and screening of a library of surface charge- and lipid components/ratios-varied lipid shell-polymer core nanoparticles. Different from the common strategy relying on the use of organelle-targeted moieties conjugated onto the surface of nanoparticles, here, we program the distribution of hybrid nanoparticles in lysosomes or mitochondria by tuning the lipid components/ratios in shell. Hybrid nanoparticles with 60% 1,2-dioleoyl-3-trimethylammonium-propane (DOTAP) and 20% 1,2-dioleoyl-sn-glycero-3-phosphoethanolamine (DOPE) can intracellularly target mitochondria in both in vitro and in vivo models. While replacing DOPE with the same amount of 1,2-dipalmitoyl-sn-glycero-3-phosphocholine (DPPC), the nanoparticles do not show mitochondrial targeting, indicating an incremental effect of cationic and fusogenic lipids on lysosomal escape which is further studied by molecular dynamics simulations. This work unveils the lipid-regulated subcellular distribution of hybrid nanoparticles in which target moieties and complex synthetic steps are avoided.

  13. Simulation of the thermalhydraulic behavior of a molten core within a structure, with the three dimensions three components TOLBIAC code

    Energy Technology Data Exchange (ETDEWEB)

    Spindler, B.; Moreau, G.M.; Pigny S. [Centre d`Etudes Nucleaires de Grenoble (France)

    1995-09-01

    The TOLBIAC code is devoted to the simulation of the behavior of a molten core within a structure (pressure vessel of core catcher), taking into account the relative position of the core components, the wall ablation and the crust formation. The code is briefly described: 3D model, physical properties and constitutive laws. wall ablation and crust model. Two results are presented: the simulation of the COPO experiment (natural convection with water in a 1/2 scale elliptic pressure vessel), and the simulation of the behavior of a corium in a PWR pressure vessel, with ablation and crust formation.

  14. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

    Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

  15. Cruise report on geotechnical core processing; Cruise: ATLAS-84, ISHTE Component Test, R/V Melville Sept.-Oct., 1984

    International Nuclear Information System (INIS)

    Silva, A.J.; Lipkin, J.; Brandes, H.

    1986-01-01

    The primary objectives of the geotechnical core processing program on the component test cruise were to: a) obtain additional base line physical property data of the ISHTE site sediments in MPG-I; b) compare strengths determined in the cored sediments to those obtained in situ with the In Situ Vane (ISV) system; and c) obtain samples for detailed laboratory analysis. Original plans called for processing of at least four cores obtained with the 10.2 cm APL hydrostatic corer (HLC) and at least one core obtained with the 20.3 cm WHOI corer (GC). If possible it was also planned to process one of the HLC cores at dockside in an attempt to assess the effects of ship motions on shear strength measurements. Shipboard laboratory facilities were set up for geotechnical processing with the URI sampling gear. Sandia Laboratory supplied a laboratory miniature vane device and apparatus for conducting thermal conductivity tests. Shipboard measurements included shear strength (miniature vane and Torvane) and thermal conductivity. Sampling included disturbed samples for water content, bulk density and classification tests and undisturbed samples for consolidation, permeability, strength, creep and fabric analyses. Unfortunately no GC cores were obtained. Three HLC cores, obtained on two lowerings of the large platform, were processed in considerable detail. In addition it was decided to process three of the box cores recovered by the SIO biology group. Therefore, a total of six cores were processed for geotechnical purposes. The dockside processing plan was deleted because of uncertainties caused by recovery procedures and the fact that only three HLC cores were available. Results are summarized

  16. Core components of a comprehensive quality assurance program in anatomic pathology.

    Science.gov (United States)

    Nakhleh, Raouf E

    2009-11-01

    In this article the core components of a comprehensive quality assurance and improvement plan are outlined. Quality anatomic pathology work comes with focus on accurate, timely, and complete reports. A commitment to continuous quality improvement and a systems approach with a persistent effort helps to achieve this end. Departments should have a quality assurance and improvement plan that includes a risk assessment of real and potential problems facing the laboratory. The plan should also list the individuals responsible for carrying out the program with adequate resources, a defined timetable, and annual assessment for progress and future directions. Quality assurance monitors should address regulatory requirements and be organized by laboratory division (surgical pathology, cytology, etc) as well as 5 segments (preanalytic, analytic, postanalytic phases of the test cycle, turn-around-time, and customer satisfaction). Quality assurance data can also be used to evaluate individual pathologists using multiple parameters with peer group comparison.

  17. Core component integration tests for the back-end software sub-system in the ATLAS data acquisition and event filter prototype -1 project

    International Nuclear Information System (INIS)

    Badescu, E.; Caprini, M.; Niculescu, M.; Radu, A.

    2000-01-01

    The ATLAS data acquisition (DAQ) and Event Filter (EF) prototype -1 project was intended to produce a prototype system for evaluating candidate technologies and architectures for the final ATLAS DAQ system on the LHC accelerator at CERN. Within the prototype project, the back-end sub-system encompasses the software for configuring, controlling and monitoring the DAQ. The back-end sub-system includes core components and detector integration components. The core components provide the basic functionality and had priority in terms of time-scale for development in order to have a baseline sub-system that can be used for integration with the data-flow sub-system and event filter. The following components are considered to be the core of the back-end sub-system: - Configuration databases, describe a large number of parameters of the DAQ system architecture, hardware and software components, running modes and status; - Message reporting system (MRS), allows all software components to report messages to other components in the distributed environment; - Information service (IS) allows the information exchange for software components; - Process manager (PMG), performs basic job control of software components (start, stop, monitoring the status); - Run control (RC), controls the data taking activities by coordinating the operations of the DAQ sub-systems, back-end software and external systems. Performance and scalability tests have been made for individual components. The back-end subsystem integration tests bring together all the core components and several trigger/DAQ/detector integration components to simulate the control and configuration of data taking sessions. For back-end integration tests a test plan was provided. The tests have been done using a shell script that goes through different phases as follows: - starting the back-end server processes to initialize communication services and PMG; - launching configuration specific processes via DAQ supervisor as

  18. An investigation of methods for neutron dose measurement in high temperature irradiation fields

    Energy Technology Data Exchange (ETDEWEB)

    Kosako, Toshisou; Sugiura, Nobuyuki [Tokyo Univ. (Japan); Kudo, Kazuhiko [Kyushu Univ., Fukuoka (Japan)] [and others

    2000-10-01

    The Japan Atomic Energy Research Institute (JAERI) has been conducting the innovative basic research on high temperature since 1994, which is a series of high temperature irradiation studies using the High Temperature Engineering Test Reactor (HTTR). 'The Task Group for Evaluation of Irradiation Dose under High Temperature Radiation' was founded in the HTTR Utilization Research Committee, which is the promoting body of the innovative basic research. The present report is a summary of investigation which has been made by the Task Group on the present status and subjects of research and development of neutron detectors in high temperature irradiation fields, in view of contributing to high temperature irradiation research using the HTTR. Detectors investigated here in the domestic survey are the following five kinds of in-core detectors: 1) small fission counter, 2) small fission chamber, 3) self-powered detector, 4) activation detector, and 5) optical fiber. In addition, the research and development status in Russia has been investigated. The present report will also be useful as nuclear instrumentation of high temperature gas-cooled reactors. (author)

  19. Core mechanics and configuration behavior of advanced LMFBR core restraint concepts

    International Nuclear Information System (INIS)

    Fox, J.N.; Wei, B.C.

    1978-02-01

    Core restraint systems in LMFBRs maintain control of core mechanics and configuration behavior. Core restraint design is complex due to the close spacing between adjacent components, flux and temperature gradients, and irradiation-induced material property effects. Since the core assemblies interact with each other and transmit loads directly to the core restraint structural members, the core assemblies themselves are an integral part of the core restraint system. This paper presents an assessment of several advanced core restraint system and core assembly concepts relative to the expected performance of currently accepted designs. A recommended order for the development of the advanced concepts is also presented

  20. Core losses of a permanent magnet synchronous motor with an amorphous stator core under inverter and sinusoidal excitations

    Science.gov (United States)

    Yao, Atsushi; Sugimoto, Takaya; Odawara, Shunya; Fujisaki, Keisuke

    2018-05-01

    We report core loss properties of permanent magnet synchronous motors (PMSM) with amorphous magnetic materials (AMM) core under inverter and sinusoidal excitations. To discuss the core loss properties of AMM core, a comparison with non-oriented (NO) core is also performed. In addition, based on both experiments and numerical simulations, we estimate higher (time and space) harmonic components of the core losses under inverter and sinusoidal excitations. The core losses of PMSM are reduced by about 59% using AMM stator core instead of NO core under sinusoidal excitation. We show that the average decrease obtained by using AMM instead of NO in the stator core is about 94% in time harmonic components.

  1. Core losses of a permanent magnet synchronous motor with an amorphous stator core under inverter and sinusoidal excitations

    Directory of Open Access Journals (Sweden)

    Atsushi Yao

    2018-05-01

    Full Text Available We report core loss properties of permanent magnet synchronous motors (PMSM with amorphous magnetic materials (AMM core under inverter and sinusoidal excitations. To discuss the core loss properties of AMM core, a comparison with non-oriented (NO core is also performed. In addition, based on both experiments and numerical simulations, we estimate higher (time and space harmonic components of the core losses under inverter and sinusoidal excitations. The core losses of PMSM are reduced by about 59% using AMM stator core instead of NO core under sinusoidal excitation. We show that the average decrease obtained by using AMM instead of NO in the stator core is about 94% in time harmonic components.

  2. Transcript specificity in yeast pre-mRNA splicing revealed by mutations in core spliceosomal components.

    Directory of Open Access Journals (Sweden)

    Jeffrey A Pleiss

    2007-04-01

    Full Text Available Appropriate expression of most eukaryotic genes requires the removal of introns from their pre-messenger RNAs (pre-mRNAs, a process catalyzed by the spliceosome. In higher eukaryotes a large family of auxiliary factors known as SR proteins can improve the splicing efficiency of transcripts containing suboptimal splice sites by interacting with distinct sequences present in those pre-mRNAs. The yeast Saccharomyces cerevisiae lacks functional equivalents of most of these factors; thus, it has been unclear whether the spliceosome could effectively distinguish among transcripts. To address this question, we have used a microarray-based approach to examine the effects of mutations in 18 highly conserved core components of the spliceosomal machinery. The kinetic profiles reveal clear differences in the splicing defects of particular pre-mRNA substrates. Most notably, the behaviors of ribosomal protein gene transcripts are generally distinct from other intron-containing transcripts in response to several spliceosomal mutations. However, dramatically different behaviors can be seen for some pairs of transcripts encoding ribosomal protein gene paralogs, suggesting that the spliceosome can readily distinguish between otherwise highly similar pre-mRNAs. The ability of the spliceosome to distinguish among its different substrates may therefore offer an important opportunity for yeast to regulate gene expression in a transcript-dependent fashion. Given the high level of conservation of core spliceosomal components across eukaryotes, we expect that these results will significantly impact our understanding of how regulated splicing is controlled in higher eukaryotes as well.

  3. Annual report on experimental operation of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2001 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Inagaki, Yoshiyuki; Kato, Michio; Fujisaki, Katsuo; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Ohashi, Hirofumi; Katanishi, Shoji; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki

    2005-06-01

    This is an annual report on the experimental operation of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2001 fiscal year. The first experimental operation was performed during two weeks from March 1, 2002 to March 13, 2002 to test on the thermal hydraulic performance of the steam reformer and also to train the operators. The thermal hydraulic performance test of the steam reformer was performed to evaluate the heat transfer characteristics between helium gas and process gas in the steam reformer. This report is summarized with an overview of the test, the results and its operation records. (author)

  4. Analysis of the methodical component of core power density field calculation error on the basis of Mochovce-1 commissioning tests

    International Nuclear Information System (INIS)

    Brik, A.

    2009-01-01

    In the first decade of June 2008, during the power commissioning of the reactor at the Mochovce NPP unit 1, the experiment with reducing the thermal power of core almost to the balance-of-plant (BOP) needs was performed. After the reactor has operated for seven hours at low power (about 200 220 MW (thermal)), its power was increased (at a rate of about 0.25% of N nom /min) to the initial level, close to 107% (1471 MW). During the experiment, core parameters, which were subsequently used for comparing the measured data with the results of experiment simulation calculations, were recorded in the reactor in-core monitoring system database. Calculated and measured levels of critical concentrations of boric acid were compared, along with power density distributions by fuel elements and assemblies obtained both by the KRUIZ in-core monitoring system and on the basis of calculations simulating reactor operation in accordance with the given core power variation schedule. The final stage consisted of assessing the methodical component of power density micro- and macro-fields calculation error in the core of Mochovce-1 reactor operating with varying load. (author)

  5. Analysis of the methodical component of core power density field calculation error on the basis of Mochovce-1 commissioning tests

    International Nuclear Information System (INIS)

    Brik, A.

    2009-01-01

    In the first decade of June 2008, during the power commissioning of the reactor at Mochovce NPP unit 1, the experiment with reducing the thermal power of core almost to the balance-of-plant needs was performed. After the reactor has operated for seven hours at low power (about 200 220 MW (thermal)), its power was increased (at a rate of about 0.25% of N nom /min) to the initial level, close to 107% (1471 MW). During the experiment, core parameters, which were subsequently used for comparing the measured data with the results of experiment simulation calculations, were recorded in the reactor in-core monitoring system's database. Calculated and measured levels of critical concentrations of boric acid were compared, along with power density distributions by fuel elements and assemblies obtained both by the KRUIZ in-core monitoring system and on the basis of calculations simulating reactor operation in accordance with the given core power variation schedule. The final stage consisted of assessing the methodical component of power density micro- and macro-fields' calculation error in the core of Mochovce-1 reactor operating with varying load. (Authors)

  6. Developing an OMERACT Core Outcome Set for Assessing Safety Components in Rheumatology Trials

    DEFF Research Database (Denmark)

    Klokker, Louise; Tugwell, Peter; Furst, Daniel E

    2016-01-01

    in such COS. The Outcome Measures in Rheumatology (OMERACT) Filter 2.0 emphasizes the importance of measuring harms. The Safety Working Group was reestablished at the OMERACT 2016 with the objective to develop a COS for assessing safety components in trials across rheumatologic conditions. METHODS: The safety......OBJECTIVE: Failure to report harmful outcomes in clinical research can introduce bias favoring a potentially harmful intervention. While core outcome sets (COS) are available for benefits in randomized controlled trials in many rheumatic conditions, less attention has been paid to safety...... that patients consider relevant so that they will be able to make informed decisions. CONCLUSION: The OMERACT Safety Working Group will advance the work previously done within OMERACT using a new patient-driven approach....

  7. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  8. Rotary core drills

    Energy Technology Data Exchange (ETDEWEB)

    1967-11-30

    The design of a rotary core drill is described. Primary consideration is given to the following component parts of the drill: the inner and outer tube, the core bit, an adapter, and the core lifter. The adapter has the form of a downward-converging sleeve and is mounted to the lower end of the inner tube. The lifter, extending from the adapter, is split along each side so that it can be held open to permit movement of a core. It is possible to grip a core by allowing the lifter to assume a closed position.

  9. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

    A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

  10. Diagnose Test-Taker's Profile in Terms of Core Profile Patterns: Principal Component (PC) vs. Profile Analysis via MDS (PAMS) Approaches.

    Science.gov (United States)

    Kim, Se-Kang; Davison, Mark L.

    A study was conducted to examine how principal components analysis (PCA) and Profile Analysis via Multidimensional Scaling (PAMS) can be used to diagnose individuals observed score profiles in terms of core profile patterns identified by each method. The standardization sample from the Wechsler Intelligence Scale for Children, Third Edition…

  11. Radio core dominance of Fermi blazars

    Science.gov (United States)

    Pei, Zhi-Yuan; Fan, Jun-Hui; Liu, Yi; Yuan, Yi-Hai; Cai, Wei; Xiao, Hu-Bing; Lin, Chao; Yang, Jiang-He

    2016-07-01

    During the first 4 years of mission, Fermi/LAT detected 1444 blazars (3FGL) (Ackermann et al. in Astrophys. J. 810:14, 2015). Fermi/LAT observations of blazars indicate that Fermi blazars are luminous and strongly variable with variability time scales, for some cases, as short as hours. Those observations suggest a strong beaming effect in Fermi/LAT blazars. In the present work, we will investigate the beaming effect in Fermi/LAT blazars using a core-dominance parameter, R = S_{core}/ S_{ext.}, where S_{core} is the core emission, while S_{ext.} is the extended emission. We compiled 1335 blazars with available core-dominance parameter, out of which 169 blazars have γ-ray emission (from 3FGL). We compared the core-dominance parameters, log R, between the 169 Fermi-detected blazars (FDBs) and the rest non-Fermi-detected blazars (non-FDBs), and we found that the averaged values are R+(2.25±0.10), suggesting that a source with larger log R has larger V.I. value. Thirdly, we compared the mean values of radio spectral index for FDBs and non-FDBs, and we obtained < α_{radio}rangle =0.06±0.35 for FDBs and < α_{radio}rangle =0.57±0.46 for non-FDBs. If γ-rays are composed of two components like radio emission (core and extended components), then we can expect a correlation between log R and the γ-ray spectral index. When we used the radio core-dominance parameter, log R, to investigate the relationship, we found that the spectral index for the core component is α_{γ}|_{core} = 1.11 (a photon spectral index of α_{γ}^{ph}|_{core} = 2.11) and that for the extended component is α_{γ}|_{ext.} = 0.70 (a photon spectral index of α_{γ}^{ph}|_{ext.} = 1.70). Some discussions are also presented.

  12. Development of carbon/carbon composite control rod for HTTR. 2. Concept, specifications and mechanical test of materials

    International Nuclear Information System (INIS)

    Eto, Motokuni; Ishiyama, Shintaro; Fukaya, Kiyoshi; Saito, Tamotsu; Ishihara, Masahiro; Hanawa, Satoshi.

    1998-01-01

    A concept and specifications of carbon/carbon composite (C/C) control rod were proposed, aiming at the application of the material to the HTTR. The outer diameter and length of the control rod were kept as the same as those of the present control rod, i.e., 113 mm and 3094 mm, respectively. According to the concept, the rod consists of ten units which are connected in series using bolts. Then, the stresses generated by dead loads in the control rod elements were estimated and compared with the design strengths which were derived from the results of measurements of tensile, compressive, bending and shear strengths of two candidate materials, AC250 (Across Co.) and CX-270 (Toyo Tanso Co.). Design strength was preliminarily determined as one-third or one-fifth of the mean strength. Ratio of the design strength to generated stress for the AC250 (2D) was : Tensile stress in the outer sleeve tube, 66, tensile and shear stresses in the M16 bolt, 8.8 and 8.5, shear stress in the plug support bolt M8, 2.43. These results are believed to indicate the mechanical integrity of the control rod structure. Data available on the candidate materials were also compiled in the Appendix. (author)

  13. Development of C/C composite for the core component of the high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H

    2005-01-15

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite.

  14. Development of C/C composite for the core component of the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H.

    2005-01-01

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite

  15. An integrated magnetics component

    DEFF Research Database (Denmark)

    2013-01-01

    The present invention relates to an integrated magnetics component comprising a magnetically permeable core comprising a base member extending in a horizontal plane and first, second, third and fourth legs protruding substantially perpendicularly from the base member. First, second, third...... and fourth output inductor windings are wound around the first, second, third and fourth legs, respectively. A first input conductor of the integrated magnetics component has a first conductor axis and extends in-between the first, second, third and fourth legs to induce a first magnetic flux through a first...... flux path of the magnetically permeable core. A second input conductor of the integrated magnetics component has a second coil axis extending substantially perpendicularly to the first conductor axis to induce a second magnetic flux through a second flux path of the magnetically permeable core...

  16. Insulating jacket for heat sensitive components

    International Nuclear Information System (INIS)

    Class, G.

    1980-01-01

    The insulating jacket for long core components of sodium-cooled reactors consists of several layers of austenitic steel, between which a woven wire mesh of the same material is fitted. It is wound in the form of a spiral bandage on the core component. (DG) [de

  17. A review of PFR core distortion experience

    International Nuclear Information System (INIS)

    Brook, A.J.

    1984-01-01

    Neutron induced voidage (NIV) swelling and irradiation creep, acting together or individually, produce deformation in core components exposed to a fast neutron flux and can lead to mechanical interaction between them. Today the nature of these processes is reasonably well understood, and reactor designers have two options in attempting to accomodate them: either by employing a flexible free standing design in which contact loadings are low but in which distortion may be high, or more commonly, by some type of restrained core in which inter-component loadings are high, but where distortion is relatively small. The aims of this paper are: a. to describe briefly the various operational limits of core and core component distortion and how they arise, for which a brief description of reactor construction is necessary; b. to outline how the problems of inter-component contact loadings are overcome for the interactive core; c. to describe some other potential problems which arise either from absolute swelling, or from differential swelling between components; of particular relevance here is the problem of contact loadings between absorber rods and their guide tubes; d. to comment on the degree of agreement with, and the feedback provided by, PIE findings; e. to show how the results of the work influence reactor operators and the reload program

  18. Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

    2012-06-01

    The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

  19. Development of the prediction technology of cable disconnection of in-core neutron detector for the future high-temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro

    2015-01-01

    Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs. (author)

  20. Selection method and device for reactor core performance calculation input indication

    International Nuclear Information System (INIS)

    Yuto, Yoshihiro.

    1994-01-01

    The position of a reactor core component on a reactor core map, which is previously designated and optionally changeable, is displayed by different colors on a CRT screen by using data of a data file incorporating results of a calculation for reactor core performance, such as incore thermal limit values. That is, an operator specifies the kind of the incore component to be sampled on a menu screen, to display the position of the incore component which satisfies a predetermined condition on the CRT screen by different colors in the form of a reactor core map. The position for the reactor core component displayed on the CRT screen by different colors is selected and designated on the screen by a touch panel, a mouse or a light pen, thereby automatically outputting detailed data of evaluation for the reactor core performance of the reactor core component at the indicated position. Retrieval of coordinates of fuel assemblies to be data sampled and input of the coordinates and demand for data sampling can be conducted at once by one menu screen. (N.H.)

  1. Multi-core fiber undersea transmission systems

    DEFF Research Database (Denmark)

    Nooruzzaman, Md; Morioka, Toshio

    2017-01-01

    Various potential architectures of branching units for multi-core fiber undersea transmission systems are presented. It is also investigated how different architectures of branching unit influence the number of fibers and those of inline components.......Various potential architectures of branching units for multi-core fiber undersea transmission systems are presented. It is also investigated how different architectures of branching unit influence the number of fibers and those of inline components....

  2. A core MRB1 complex component is indispensable for RNA editing in insect and human infective stages of Trypanosoma brucei.

    Directory of Open Access Journals (Sweden)

    Michelle L Ammerman

    Full Text Available Uridine insertion/deletion RNA editing is a unique and vital process in kinetoplastids, required for creation of translatable open reading frames in most mitochondrially-encoded RNAs. Emerging as a key player in this process is the mitochondrial RNA binding 1 (MRB1 complex. MRB1 comprises an RNA-independent core complex of at least six proteins, including the GAP1/2 guide RNA (gRNA binding proteins. The core interacts in an RNA-enhanced or -dependent manner with imprecisely defined TbRGG2 subcomplexes, Armadillo protein MRB10130, and additional factors that comprise the dynamic MRB1 complex. Towards understanding MRB1 complex function in RNA editing, we present here functional characterization of the pentein domain-containing MRB1 core protein, MRB11870. Inducible RNAi studies demonstrate that MRB11870 is essential for proliferation of both insect vector and human infective stage T. brucei. MRB11870 ablation causes a massive defect in RNA editing, affecting both pan-edited and minimally edited mRNAs, but does not substantially affect mitochondrial RNA stability or processing of precursor transcripts. The editing defect in MRB1-depleted cells occurs at the initiation stage of editing, as pre-edited mRNAs accumulate. However, the gRNAs that direct editing remain abundant in the knockdown cells. To examine the contribution of MRB11870 to MRB1 macromolecular interactions, we tagged core complexes and analyzed their composition and associated proteins in the presence and absence of MRB11870. These studies demonstrated that MRB11870 is essential for association of GAP1/2 with the core, as well as for interaction of the core with other proteins and subcomplexes. Together, these data support a model in which the MRB1 core mediates functional interaction of gRNAs with the editing machinery, having GAP1/2 as its gRNA binding constituents. MRB11870 is a critical component of the core, essential for its structure and function.

  3. HTTR(高温工学試験研究炉)の試験・運転と技術開発; 2010年度

    OpenAIRE

    高温工学試験研究炉部

    2011-01-01

    日本原子力研究開発機構大洗研究開発センターのHTTR(高温工学試験研究炉)は、熱出力30MWの黒鉛減速ヘリウムガス冷却型原子炉で、我が国初の高温ガス炉である。平成22年度は、熱出力9MWにおける安全性実証試験を行い、安全性に関するデータを取得・評価した。本報告書は、平成22年度(2010年)のHTTRの運転と保守及び各種技術開発の状況等について紹介する。

  4. Investigation of EAS cores

    Directory of Open Access Journals (Sweden)

    Shaulov S.B.

    2017-01-01

    Full Text Available The development of nuclear-electromagnetic cascade models in air in the late forties have shown informational content of the study of cores of extensive air showers (EAS. These investigations were the main goal in different experiments which were carried out over many years by a variety of methods. Outcomes of such investigations obtained in the HADRON experiment using an X-ray emulsion chamber (XREC as a core detector are considered. The Ne spectrum of EAS associated with γ-ray families, spectra of γ-rays (hadrons in EAS cores and the Ne dependence of the muon number, ⟨Nμ⟩, in EAS with γ-ray families are obtained for the first time at energies of 1015–1017 eV with this method. A number of new effects were observed, namely, an abnormal scaling violation in hadron spectra which are fundamentally different from model predictions, an excess of muon number in EAS associated with γ-ray families, and the penetrating component in EAS cores. It is supposed that the abnormal behavior of γ-ray spectra and Ne dependence of the muon number are explained by the emergence of a penetrating component in the 1st PCR spectrum ‘knee’ range. Nuclear and astrophysical explanations of the origin of the penetrating component are discussed. The necessity of considering the contribution of a single close cosmic-ray source to explain the PCR spectrum in the knee range is noted.

  5. Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

    International Nuclear Information System (INIS)

    Ortensi, J.; Pope, M.A.; Ferrer, R.M.; Cogliati, J.J.; Bess, J.D.; Ougouag, A.M.

    2010-01-01

    The Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine the INL's current prismatic reactor analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in the NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. DRAGON can generate transport solutions via Collision Probability (CP), Method of Characteristics (MOC) and Discrete Ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2-3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral

  6. Safety design

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Shiozawa, Shusaku

    2004-01-01

    JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R and D needs for establishing the safety philosophy for the future HTGRs are reported

  7. Core Stability in Chain-Component Additive Games

    NARCIS (Netherlands)

    van Velzen, S.; Hamers, H.J.M.; Solymosi, T.

    2004-01-01

    Chain-component additive games are graph-restricted superadditive games, where an exogenously given line-graph determines the cooperative possibilities of the players.These games can model various multi-agent decision situations, such as strictly hierarchical organisations or sequencing / scheduling

  8. DotU and VgrG, core components of type VI secretion systems, are essential for Francisella LVS pathogenicity.

    Directory of Open Access Journals (Sweden)

    Jeanette E Bröms

    Full Text Available The Gram-negative bacterium Francisella tularensis causes tularemia, a disease which requires bacterial escape from phagosomes of infected macrophages. Once in the cytosol, the bacterium rapidly multiplies, inhibits activation of the inflammasome and ultimately causes death of the host cell. Of importance for these processes is a 33-kb gene cluster, the Francisella pathogenicity island (FPI, which is believed to encode a type VI secretion system (T6SS. In this study, we analyzed the role of the FPI-encoded proteins VgrG and DotU, which are conserved components of type VI secretion (T6S clusters. We demonstrate that in F. tularensis LVS, VgrG was shown to form multimers, consistent with its suggested role as a trimeric membrane puncturing device in T6SSs, while the inner membrane protein DotU was shown to stabilize PdpB/IcmF, another T6SS core component. Upon infection of J774 cells, both ΔvgrG and ΔdotU mutants did not escape from phagosomes, and subsequently, did not multiply or cause cytopathogenicity. They also showed impaired activation of the inflammasome and marked attenuation in the mouse model. Moreover, all of the DotU-dependent functions investigated here required the presence of three residues that are essentially conserved among all DotU homologues. Thus, in agreement with a core function in T6S clusters, VgrG and DotU play key roles for modulation of the intracellular host response as well as for the virulence of F. tularensis.

  9. Core trénink ve florbale

    OpenAIRE

    Mašková, Alžběta

    2014-01-01

    Title: Use of core training in floorball Objectives: Present an overview of research papers regarding core training and its possible use in floorball Tasks: First aim of this thesis is to provide description and explanation of the core training by using available literature, research papers, bachelor or magister thesis and also thesis of the coaching school. Second aim is to analyze the necessary components of floorball player's performance relative to the core training. This is followed by a...

  10. Gravity Effects on the Free Vibration of a Graphite Column

    International Nuclear Information System (INIS)

    Ki, Dong-Ok; Kim, Jong-Bum; Park, Keun-Bae; Lee, Won-Jae

    2006-01-01

    The gravity effects on the free vibration of a graphite column are studied. Graphite block is a key component of a HTGR (High Temperature Gas Cooled Reactor). The major core elements, such as the fuel blocks and neutron reflector blocks, of HTTR (High Temperature Test Reactor, Japan) and GT-MHR (Gas Turbine- Modular Helium Reactor, USA) consist of stacked hexagonal graphite blocks forming a group of columns. The vibration of the columns induced by earthquakes may lead to solid impacts between graphite blocks and structural integrity problems. The study of free vibration characteristics of the graphite block column is the first step in the core internal structure dynamic analysis. Gravity force bring a negative stiffness term to the transverse vibration analysis of heavy long column structures, and results in natural frequency reductions. Generally it is not considered in the not so tall structure cases, because the gravity term makes the analysis and design complicated. Therefore it is important to check whether the gravity effect is severe or not

  11. Sprayed skin turbine component

    Science.gov (United States)

    Allen, David B

    2013-06-04

    Fabricating a turbine component (50) by casting a core structure (30), forming an array of pits (24) in an outer surface (32) of the core structure, depositing a transient liquid phase (TLP) material (40) on the outer surface of the core structure, the TLP containing a melting-point depressant, depositing a skin (42) on the outer surface of the core structure over the TLP material, and heating the assembly, thus forming both a diffusion bond and a mechanical interlock between the skin and the core structure. The heating diffuses the melting-point depressant away from the interface. Subsurface cooling channels (35) may be formed by forming grooves (34) in the outer surface of the core structure, filling the grooves with a fugitive filler (36), depositing and bonding the skin (42), then removing the fugitive material.

  12. First results from core-edge parallel composition in the FACETS project.

    Energy Technology Data Exchange (ETDEWEB)

    Cary, J. R.; Candy, J.; Cohen, R. H.; Krasheninnikov, S.; McCune, D. C.; Estep, D. J.; Larson, J.; Malony, A. D.; Pankin, A.; Worley, P. H.; Carlsson, J. A.; Hakim, A. H.; Hamill, P.; Kruger, S.; Miah, M.; Muzsala, S.; Pletzer, A.; Shasharina, S.; Wade-Stein, D.; Wang, N.; Balay, S.; McInnes, L.; Zhang, H.; Casper, T.; Diachin, L. (Mathematics and Computer Science); (Tech-X Corp.); (General Atomics); (LLNL); (Univ. of California at San Diego); (Princeton Plasma Physics Lab.); (Colorado State Univ.); (ParaTools Inc.); (Lehigh Univ.); (ORNL)

    2008-01-01

    FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

  13. First results from core-edge parallel composition in the FACETS project

    Energy Technology Data Exchange (ETDEWEB)

    Cary, J R; Carlsson, J A; Hakim, A H; Hamill, P; Kruger, S; Miah, M; Muzsala, S; Pletzer, A; Shasharina, S; Wade-Stein, D; Wang, N [Tech-X Corporation, Boulder, CO 80303 (United States); Candy, J [General Atomics, San Diego, CA 92186 (United States); Cohen, R H [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Krasheninnikov, S [University of California at San Diego, San Diego, CA 92093 (United States); McCune, D C [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Estep, D J [Colorado State University, Fort Collins, CO 80523 (United States); Larson, J [Argonne National Laboratory, Argonne, IL 60439 (United States); Malony, A D [ParaTools, Inc., Eugene, OR 97405 (United States); Pankin, A [Lehigh University, Bethlehem, PA 18015 (United States); Worley, P H [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)], E-mail: cary@txcorp.com (and others)

    2008-07-15

    FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

  14. First results from core-edge parallel composition in the FACETS project

    Energy Technology Data Exchange (ETDEWEB)

    Cary, John R. [Tech-X Corporation; Candy, Jeff [General Atomics; Cohen, Ronald H. [Lawrence Livermore National Laboratory (LLNL); Krasheninnikov, Sergei [University of California, San Diego; McCune, Douglas [Princeton Plasma Physics Laboratory (PPPL); Estep, Donald J [Colorado State University, Fort Collins; Larson, Jay [Argonne National Laboratory (ANL); Malony, Allen [University of Oregon; Pankin, A. [Lehigh University, Bethlehem, PA; Worley, Patrick H [ORNL; Carlsson, Johann [Tech-X Corporation; Hakim, A H [Tech-X Corporation; Hamill, P [Tech-X Corporation; Kruger, Scott [Tech-X Corporation; Miah, Mahmood [Tech-X Corporation; Muzsala, S [Tech-X Corporation; Pletzer, Alexander [Tech-X Corporation; Shasharina, Svetlana [Tech-X Corporation; Wade-Stein, D [Tech-X Corporation; Wang, N [Tech-X Corporation; Balay, Satish [Argonne National Laboratory (ANL); McInnes, Lois [Argonne National Laboratory (ANL); Zhang, Hong [Argonne National Laboratory (ANL); Casper, T. A. [Lawrence Livermore National Laboratory (LLNL); Diachin, Lori [Lawrence Livermore National Laboratory (LLNL); Epperly, Thomas [Lawrence Livermore National Laboratory (LLNL); Rognlien, T. D. [Lawrence Livermore National Laboratory (LLNL); Fahey, Mark R [ORNL; Cobb, John W [ORNL; Morris, A [University of Oregon; Shende, Sameer [University of Oregon; Hammett, Greg [Princeton Plasma Physics Laboratory (PPPL); Indireshkumar, K [Tech-X Corporation; Stotler, D. [Princeton Plasma Physics Laboratory (PPPL); Pigarov, A [University of California, San Diego

    2008-01-01

    FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

  15. CHEMICAL AND PHYSICAL CHARACTERIZATION OF COLLAPSING LOW-MASS PRESTELLAR DENSE CORES

    Energy Technology Data Exchange (ETDEWEB)

    Hincelin, U. [Department of Chemistry, University of Virginia, Charlottesville, VA 22904 (United States); Commerçon, B. [Ecole Normale Supérieure de Lyon, CRAL, UMR 5574 du CNRS, Université Lyon I, 46 Allée d’Italie, F-69364 Lyon cedex 07 (France); Wakelam, V.; Hersant, F.; Guilloteau, S. [Univ. Bordeaux, LAB, UMR 5804, F-33270, Floirac (France); Herbst, E., E-mail: ugo.hincelin@gmail.com [Departments of Chemistry and Astronomy, University of Virginia, Charlottesville, VA 22904 (United States)

    2016-05-01

    The first hydrostatic core, also called the first Larson core, is one of the first steps in low-mass star formation as predicted by theory. With recent and future high-performance telescopes, the details of these first phases are becoming accessible, and observations may confirm theory and even present new challenges for theoreticians. In this context, from a theoretical point of view, we study the chemical and physical evolution of the collapse of prestellar cores until the formation of the first Larson core, in order to better characterize this early phase in the star formation process. We couple a state-of-the-art hydrodynamical model with full gas-grain chemistry, using different assumptions for the magnetic field strength and orientation. We extract the different components of each collapsing core (i.e., the central core, the outflow, the disk, the pseudodisk, and the envelope) to highlight their specific physical and chemical characteristics. Each component often presents a specific physical history, as well as a specific chemical evolution. From some species, the components can clearly be differentiated. The different core models can also be chemically differentiated. Our simulation suggests that some chemical species act as tracers of the different components of a collapsing prestellar dense core, and as tracers of the magnetic field characteristics of the core. From this result, we pinpoint promising key chemical species to be observed.

  16. Control component retainer

    International Nuclear Information System (INIS)

    Walton, L.A.; King, R.A.

    1983-01-01

    An apparatus is described for retaining an undriven control component assembly disposed in a fuel assembly in a nuclear reactor of the type having a core grid plate. The first part of the mechanism involves a housing for the control component and the second part is a brace with a number of arms that reach under the grid plate. The brace and the housing are coupled together to firmly hold the control components in place even under strong flows of th coolant

  17. Evaluation of core distortion in FBR

    International Nuclear Information System (INIS)

    Ikarimoto, I.; Tanaka, M.; Okubo, Y.

    1984-01-01

    The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

  18. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  19. Modeling and analysis of the disk MHD generator component of a gas core reactor/MHD Rankine cycle space power system

    International Nuclear Information System (INIS)

    Welch, G.E.; Dugan, E.T.; Lear, W.E. Jr.; Appelbaum, J.G.

    1990-01-01

    A gas core nuclear reactor (GCR)/disk magnetohydrodynamic (MHD) generator direct closed Rankine space power system concept is described. The GCR/disk MHD generator marriage facilitates efficient high electric power density system performance at relatively high operating temperatures. The system concept promises high specific power levels, on the order of 1 kW e /kg. An overview of the disk MHD generator component magnetofluiddynamic and plasma physics theoretical modeling is provided. Results from a parametric design analysis of the disk MHD generator are presented and discussed

  20. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    Iizuka, N.

    2000-01-01

    The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

  1. AGR core safety assessment methodologies

    International Nuclear Information System (INIS)

    McLachlan, N.; Reed, J.; Metcalfe, M.P.

    1996-01-01

    To demonstrate the safety of its gas-cooled graphite-moderated AGR reactors, nuclear safety assessments of the cores are based upon a methodology which demonstrates no component failures, geometrical stability of the structure and material properties bounded by a database. All AGRs continue to meet these three criteria. However, predictions of future core behaviour indicate that the safety case methodology will eventually need to be modified to deal with new phenomena. A new approach to the safety assessment of the cores is currently under development, which can take account of these factors while at the same time providing the same level of protection for the cores. This approach will be based on the functionality of the core: unhindered movement of control rods, continued adequate cooling of the fuel and the core, continued ability to charge and discharge fuel. (author). 5 figs

  2. Nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F

    1974-07-11

    The core of the fast neutron reactor consisting, among other components, of fuel elements enriched in plutonium is divided into modules. Each module contains a bundle of four or six elongated components (fuel elements and control rods). In the arrangement with four components, one is kept rigid while the other three are elastically yielding inclined towards the center and lean against the rigid component. In the modules with six pieces, each component is elastically yielding inclined towards a central cavity. In this way, they form a circular arc. A control rod may be placed in the cavity. In order to counteract a relative lateral movement, the outer surfaces of the components which have hexagonal cross-sections have interlocking bearing cushions. The bearing cushions consist of keyway-type ribs or grooves with the wedges or ribs gripping in the grooves of the neighbouring components. In addition, the ribs have oblique entering surfaces.

  3. A Small Modular Reactor Design for Multiple Energy Applications: HTR50S

    Energy Technology Data Exchange (ETDEWEB)

    Yan, X.; Tachibana, Y.; Ohashi, H.; Sato, H.; Tazawa, Y.; Kunitomi, K. [Japan Atomic Energy Agency, Ibaraki (Japan)

    2013-06-15

    HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's 950 .deg. C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750 .deg. C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900 .deg. C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

  4. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

  5. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  6. Research program of the high temperature engineering test reactor for upgrading the HTGR technology

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

    1997-07-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

  7. An inspection standard of fuel for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao; Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

  8. Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow

    International Nuclear Information System (INIS)

    Whaley, R.L.; Sanders, J.P.

    1976-09-01

    A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection

  9. Representation of dislocation cores using Nye tensor distributions

    International Nuclear Information System (INIS)

    Hartley, Craig S.; Mishin, Y.

    2005-01-01

    This paper demonstrates how the cores of atomistically simulated dislocations in Cu and Al can be represented by a distribution of infinitesimal dislocations described by appropriate components of the Nye tensor. Components calculated from atomic positions in the dislocated crystal are displayed as contour plots on the plane normal to the dislocation line. The method provides an accurate and instructive means for characterizing dislocation core structures and calculating the total Burgers vector

  10. Torsional Oscillations of the Earths's Core

    Science.gov (United States)

    Hide, Raymond; Boggs, Dale H.; Dickey, Jean O.

    1997-01-01

    Torsional oscillations of the Earth's liquid metallic outer core are investigated by diving the core into twenty imaginary e1qui-volume annuli coaxial with the axis of ratation of the Earth and determining temproal fluctuations in the axial component of angular memonetum of each annulus under the assumption of iso-rotation on cylindrical surfaces.

  11. GCRA review and appraisal of HTGR reactor-core-design program

    International Nuclear Information System (INIS)

    1980-09-01

    The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation

  12. Modularized Functions of the Fanconi Anemia Core Complex

    Directory of Open Access Journals (Sweden)

    Yaling Huang

    2014-06-01

    Full Text Available The Fanconi anemia (FA core complex provides the essential E3 ligase function for spatially defined FANCD2 ubiquitination and FA pathway activation. Of the seven FA gene products forming the core complex, FANCL possesses a RING domain with demonstrated E3 ligase activity. The other six components do not have clearly defined roles. Through epistasis analyses, we identify three functional modules in the FA core complex: a catalytic module consisting of FANCL, FANCB, and FAAP100 is absolutely required for the E3 ligase function, and the FANCA-FANCG-FAAP20 and the FANCC-FANCE-FANCF modules provide nonredundant and ancillary functions that help the catalytic module bind chromatin or sites of DNA damage. Disruption of the catalytic module causes complete loss of the core complex function, whereas loss of any ancillary module component does not. Our work reveals the roles of several FA gene products with previously undefined functions and a modularized assembly of the FA core complex.

  13. A Study on Dielectric Properties of Cadmium Sulfide-Zinc Sulfide Core-Shell Nanocomposites for Application as Nanoelectronic Filter Component in the Microwave Domain

    Science.gov (United States)

    Devi, Jutika; Datta, Pranayee

    2018-03-01

    Complex permittivities of cadmium sulfide (CdS), zinc sulfide (ZnS), and of cadmium sulfide-zinc sulfide (CdS/ZnS) core-shell nanoparticles embedded in a polyvinyl alcohol matrix (PVA) were measured in liquid phase using a VectorNetwork Analyzer in the frequency range of 500 MHz-10 GHz. These nanocomposites are modeled as an embedded capacitor, and their electric field distribution and polarization have been studied using COMSOL Multiphysics software. By varying the thickness of the shell and the number of inclusions, the capacitance values were estimated. It was observed that CdS, ZnS and CdS/ZnS core-shell nanoparticles embedded in a polyvinyl alcohol matrix show capacitive behavior. There is a strong influence of the dielectric properties in the capacitive behavior of the embedded nanocapacitor. The capping matrix, position and filling factors of nanoinclusions all affect the capacitive behavior of the tested nanocomposites. Application of the CdS, ZnS and CdS/ZnS core-shell nanocomposite as the passive low-pass filter circuit has also been investigated. From the present study, it has been found that CdS/ZnS core-shell nanoparticles embedded in PVA matrix are potential structures for application as nanoelectronic filter components in different areas of communication.

  14. The seismic assessment of radially keyed graphite moderator cores

    International Nuclear Information System (INIS)

    Steer, A.G.; Payne, J.F.B.

    1996-01-01

    The modelling of AGR and Magnox cores has to deal with the very large number of components that make up the core, and the non-linear response due to the clearances in the keying system. This paper examines the conditions under which it is permissible to linearise the response. By comparing the results of discrete and continuum models of the core, the paper also shows that the number of components in the core is sufficiently large that the core can be approximated satisfactorily by an anisotropic solid material. The material has unusual properties, but these can be handled within the standard framework for the description of the elastic properties of an anisotropic solid. This description of the core by an equivalent solid material can readily be incorporated into finite element models of the reactor internal structure. Such models have been set up for both AGR and Magnox reactors. The models are being used to assess the seismic response of these reactors. (author). 5 refs, 6 figs

  15. Measurement of gamma ray from fuel of high temperature engineering test reactor. Method of measurement and results

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2001-02-01

    To obtain information in the HTTR core directly, gamma ray from fuel blocks was measured when fuel blocks were discharged from the core and reloaded to the core. Gamma ray was measured using GM detector, CZT semiconductor detector installed in a door valve and area monitors installed in a stand pipe compartment. The measurement was carried out for 20 fuel blocks in 4 columns considering the symmetry of uranium enrichment distribution in the core. Relative axial distribution in each column obtained by the GM detector and CZT detector agreed with calculated results. However, calculation values showed higher values than measured values in upper region of the core, lower those in lower region of the core. The axial distributions were also evaluated by the area monitors. The measured values agreed with calculated values. It became clear that it was possible to obtain the data inside the core by this method. (author)

  16. SMA millimeter observations of hot molecular cores

    Energy Technology Data Exchange (ETDEWEB)

    Hernández-Hernández, Vicente; Zapata, Luis; Kurtz, Stan [Centro de Radioastronomía y Astrofísica, Universidad Nacional Autónoma de México, Apdo. Postal 3-72 (Xangari), 58090 Morelia, Michoacán (Mexico); Garay, Guido, E-mail: v.hernandez@crya.unam.mx [Departamento de Astronomía, Universidad de Chile, Camino del Observatorio 1515, Las Condes, Santiago (Chile)

    2014-05-01

    We present Submillimeter Array observations in the 1.3 mm continuum and the CH{sub 3}CN (12 {sub K}-11 {sub K}) line of 17 hot molecular cores associated with young high-mass stars. The angular resolution of the observations ranges from 1.''0 to 4.''0. The continuum observations reveal large (>3500 AU) dusty structures with gas masses from 7 to 375 M {sub ☉}, which probably surround multiple young stars. The CH{sub 3}CN line emission is detected toward all the molecular cores at least up to the K = 6 component and is mostly associated with the emission peaks of the dusty objects. We used the multiple K-components of the CH{sub 3}CN and both the rotational diagram method and a simultaneous synthetic local thermodynamic equilibrium model with the XCLASS program to estimate the temperatures and column densities of the cores. For all sources, we obtained reasonable fits from XCLASS by using a model that combines two components: an extended and warm envelope and a compact hot core of molecular gas, suggesting internal heating by recently formed massive stars. The rotational temperatures lie in the range of 40-132 K and 122-485 K for the extended and compact components, respectively. From the continuum and CH{sub 3}CN results, we infer fractional abundances from 10{sup –9} to 10{sup –7} toward the compact inner components, which increase with the rotational temperature. Our results agree with a chemical scenario in which the CH{sub 3}CN molecule is efficiently formed in the gas phase above 100-300 K, and its abundance increases with temperature.

  17. Evaluation of thermal displacement behavior of high temperature piping system in power-up test of HTTR. No. 1 results up to 20 MW operation

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Kojima, Takao; Sumita, Junya; Tachibana, Yukio

    2002-03-01

    Temperature of the primary cooling system of the High Temperature Engineering Test Reactor, HTTR, becomes very high because the coolant temperature at the reactor outlet reaches 950degC, and 400degC at inlet of the reactor. Therefore, it is important to confirm the thermal displacement behavior of the high temperature piping system in the primary cooling system from the viewpoint of the structural integrity. Moreover, newly designed 3-dimensional floating support system is adopted to the cooling system, it is meaningful to verify the thermal displacement behavior of the piping system applied the 3-dimensional floating support system. In the power-up test (up to 20 MW operation), thermal displacement behavior of the high temperature piping system was measured. This paper describes the experimental and analytical results of thermal displacement characteristics of the high temperature piping system. The results showed that the resistance force induced from the supporting system effects to the thermal displacement behavior of cooling system, and the analytical results have a good agreement with the experimental results by optimizing the resistant force of the floating support system. Additionally, structural integrity at the 30 MW operation was confirmed by the analysis. (author)

  18. Annual report on experimental operations and maintenance of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2004 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Ohashi, Hirofumi; Morisaki, Norihiro; Kato, Michio; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Inagaki, Yoshiyuki

    2006-03-01

    This is annual report on the experimental test operations and maintenances of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2004 fiscal year. The improvement work of catalyst dust filter in combustion system was performed in May 2004, and the performance was confirmed. The sixth experimental test operation was performed from June to July 2004. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July to September 2004. The seventh experimental test operation was performed from October to December 2004 for chemical reaction shutdown test. From the results, a behavior of the helium-gas cooling system, consists of steam generator and radiator, during chemical reaction shutdown was confirmed. This report is summarized with the outline and the results of the test, maintenance works and inspections, and operation records in mentioned above. (author)

  19. Equipment for the conditioning of core components in the fuel element storage pool with particular respect to the design required by the conditions for nuclear facilities in operation and the surveillance in accordance with atomic rules and regulations

    International Nuclear Information System (INIS)

    Dumpe, J.; Schwiertz, V.; Geiser, C.; Prucker, E.

    2001-01-01

    In nuclear power plants worn out and activated parts from the reactor core (core components) which are placed into the fuel element storage pool arise on a regular basis during the technical maintenance and the review. The disposal of these core components due to radiation protection aspects is only feasible within the fuel element storage pool during the operation of the NPP using techniques of the under water conditioning. Therefore, special GNS equipment is used for the conditioning, using under water conditioning equipment, such as UWS, BZ, and ZVA, a number of lifting and auxiliary equipment for mounting and dismantling purposes and the handling of the core components and the waste casks within the fuel element storage pool. These components must meet particular safety requirements with regard to their integrity and reliability. They are designed according to the requirements on nuclear components (KTA). The manipulating equipment must be partly redundant and the protection goals for nuclear accidents must be met. The Bavarian Ministry for Development and Environment tasked the TUeV Sueddeutschland with the surveillance and control. The conditioning equipment of GNS is therefore designed in co-ordination with the examiner of the Governmental Regulating Agency, in particular respect to all safety aspects. Furthermore the examiners perform reviews of the construction and the documentation during the design and construction phase. (orig.)

  20. Mutations in SNRPB, encoding components of the core splicing machinery, cause cerebro-costo-mandibular syndrome.

    Science.gov (United States)

    Bacrot, Séverine; Doyard, Mathilde; Huber, Céline; Alibeu, Olivier; Feldhahn, Niklas; Lehalle, Daphné; Lacombe, Didier; Marlin, Sandrine; Nitschke, Patrick; Petit, Florence; Vazquez, Marie-Paule; Munnich, Arnold; Cormier-Daire, Valérie

    2015-02-01

    Cerebro-costo-mandibular syndrome (CCMS) is a developmental disorder characterized by the association of Pierre Robin sequence and posterior rib defects. Exome sequencing and Sanger sequencing in five unrelated CCMS patients revealed five heterozygous variants in the small nuclear ribonucleoprotein polypeptides B and B1 (SNRPB) gene. This gene includes three transcripts, namely transcripts 1 and 2, encoding components of the core spliceosomal machinery (SmB' and SmB) and transcript 3 undergoing nonsense-mediated mRNA decay. All variants were located in the premature termination codon (PTC)-introducing alternative exon of transcript 3. Quantitative RT-PCR analysis revealed a significant increase in transcript 3 levels in leukocytes of CCMS individuals compared to controls. We conclude that CCMS is due to heterozygous mutations in SNRPB, enhancing inclusion of a SNRPB PTC-introducing alternative exon, and show that this developmental disease is caused by defects in the splicing machinery. Our finding confirms the report of SNRPB mutations in CCMS patients by Lynch et al. (2014) and further extends the clinical and molecular observations. © 2014 WILEY PERIODICALS, INC.

  1. Thermal interaction of core melt debris with the TMI-2 baffle, core-former, and lower head structures

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Tolman, E.L.

    1987-09-01

    Recent inspection of the TMI-2 core-former baffle walls (vertical), former plates (horizontal), and lower plenum has been conducted to assess potential damage to these structures. Video observations show evidence of localized melt failure of the baffle walls, whereas fiberoptics data indicate the presence of resolidified debris on the former plates. Lower plenum inspection also confirms the presence of 20 tons or more of core debris in the lower plenum. These data indicate massive core melt relocation and the potential for melt attack on vessel structural components. This report presents analyses aimed at developing an understanding of melt relocation behavior and damage progression to TMI-2 vessel components. Thermal analysis indicates melt-through of the baffle plates, but maintenance of structural integrity of the former plates and lower head. Differences in the damage of these structures is attributed largely to differences in contact time with melt debris and pressure of water. 29 refs., 17 figs., 9 tabs

  2. Design consideration on hydrogen production demonstration plant of thermochemical IS process

    International Nuclear Information System (INIS)

    Iwatsuki, Jin; Noguchi, Hiroki; Terada, Atsuhiko; Kubo, Shinji; Sakaba, Nariaki; Onuki, Kaoru; Hino, Ryutaro

    2009-03-01

    Preliminary design study was carried out on the hydrogen production demonstration plant of thermochemical IS process. In the pilot test, hydrogen production will be examined under prototypical condition using an apparatus made of industrial materials, which is driven by the sensible heat of helium gas heated by an electric heater that simulates the High Temperature Engineering Test Reactor (HTTR). Tentative system condition was defined considering the HTTR specification and the experience on the construction and the operation of the mock-up test facility using methane reforming for hydrogen production. The process condition and the system flow diagram were discussed to meet the system condition. Based on the defined process condition, types of the main components were discussed taking the corrosion resistance of the structural materials into consideration. Applicable rules and regulations were also surveyed regarding the plant construction and operation. (author)

  3. Steady state toroidal magnetic field at earth's core-mantle boundary

    Science.gov (United States)

    Levy, Eugene H.; Pearce, Steven J.

    1991-01-01

    Measurements of the dc electrical potential near the top of earth's mantle have been extrapolated into the deep mantle in order to estimate the strength of the toroidal magnetic field component at the core-mantle interface. Recent measurements have been interpreted as indicating that at the core-mantle interface, the magnetic toroidal and poloidal field components are approximately equal in magnitude. A motivation for such measurements is to obtain an estimate of the strength of the toroidal magnetic field in the core, a quantity important to our understanding of the geomagnetic field's dynamo generation. Through the use of several simple and idealized calculation, this paper discusses the theoretical relationship between the amplitude of the toroidal magnetic field at the core-mantle boundary and the actual amplitude within the core. Even with a very low inferred value of the toroidal field amplitude at the core-mantle boundary, (a few gauss), the toroidal field amplitude within the core could be consistent with a magnetohydrodynamic dynamo dominated by nonuniform rotation and having a strong toroidal magnetic field.

  4. WARM EXTENDED DENSE GAS AT THE HEART OF A COLD COLLAPSING DENSE CORE

    International Nuclear Information System (INIS)

    Shinnaga, Hiroko; Phillips, Thomas G.; Furuya, Ray S.; Kitamura, Yoshimi

    2009-01-01

    In order to investigate when and how the birth of a protostellar core occurs, we made survey observations of four well-studied dense cores in the Taurus molecular cloud using CO transitions in submillimeter bands. We report here the detection of unexpectedly warm (∼30-70 K), extended (radius of ∼2400 AU), dense (a few times 10 5 cm -3 ) gas at the heart of one of the dense cores, L1521F (MC27), within the cold dynamically collapsing components. We argue that the detected warm, extended, dense gas may originate from shock regions caused by collisions between the dynamically collapsing components and outflowing/rotating components within the dense core. We propose a new stage of star formation, 'warm-in-cold core stage (WICCS)', i.e., the cold collapsing envelope encases the warm extended dense gas at the center due to the formation of a protostellar core. WICCS would constitute a missing link in evolution between a cold quiescent starless core and a young protostar in class 0 stage that has a large-scale bipolar outflow.

  5. Annual report on experimental operations and maintenances of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2003 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Morisaki, Norihiro; Ohashi, Hirofumi; Kato, Michio; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Inagaki, Yoshiyuki

    2006-03-01

    This is a report on the experimental operations and maintenances of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2003 fiscal year. The fourth and fifth experimental test operations were performed, from May to July and from October to December in 2003, for the following tests; (a) start-up and shutdown operation test, (b) process change test, (c) continuous hydrogen-production test and (d) chemical reaction shutdown test. From the results, a long time-range stability of the hydrogen production system was confirmed, a behavior of the helium-gas cooling system, consists of steam generator and radiator; during chemical reaction shutdown, was understood, and so on. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July 2003. This report is summarized on outlines and results of the tests, outlines and results of the periodic inspections, and operation records of the mock-up test facility. (author)

  6. The nebular spectra of the transitional Type Ia Supernovae 2007on and 2011iv: broad, multiple components indicate aspherical explosion cores

    Science.gov (United States)

    Mazzali, P. A.; Ashall, C.; Pian, E.; Stritzinger, M. D.; Gall, C.; Phillips, M. M.; Höflich, P.; Hsiao, E.

    2018-05-01

    The nebular-epoch spectrum of the rapidly declining, `transitional' Type Ia supernova (SN) 2007on showed double emission peaks, which have been interpreted as indicating that the SN was the result of the direct collision of two white dwarfs. The spectrum can be reproduced using two distinct emission components, one redshifted and one blueshifted. These components are similar in mass but have slightly different degrees of ionization. They recede from one another at a line-of-sight speed larger than the sum of the combined expansion velocities of their emitting cores, thereby acting as two independent nebulae. While this configuration appears to be consistent with the scenario of two white dwarfs colliding, it may also indicate an off-centre delayed detonation explosion of a near-Chandrasekhar-mass white dwarf. In either case, broad emission line widths and a rapidly evolving light curve can be expected for the bolometric luminosity of the SN. This is the case for both SNe 2007on and 2011iv, also a transitional SN Ia that exploded in the same elliptical galaxy, NGC 1404. Although SN 2011iv does not show double-peaked emission line profiles, the width of its emission lines is such that a two-component model yields somewhat better results than a single-component model. Most of the mass ejected is in one component, however, which suggests that SN 2011iv was the result of the off-centre ignition of a Chandrasekhar-mass white dwarf.

  7. [caCORE: core architecture of bioinformation on cancer research in America].

    Science.gov (United States)

    Gao, Qin; Zhang, Yan-lei; Xie, Zhi-yun; Zhang, Qi-peng; Hu, Zhang-zhi

    2006-04-18

    A critical factor in the advancement of biomedical research is the ease with which data can be integrated, redistributed and analyzed both within and across domains. This paper summarizes the Biomedical Information Core Infrastructure built by National Cancer Institute Center for Bioinformatics in America (NCICB). The main product from the Core Infrastructure is caCORE--cancer Common Ontologic Reference Environment, which is the infrastructure backbone supporting data management and application development at NCICB. The paper explains the structure and function of caCORE: (1) Enterprise Vocabulary Services (EVS). They provide controlled vocabulary, dictionary and thesaurus services, and EVS produces the NCI Thesaurus and the NCI Metathesaurus; (2) The Cancer Data Standards Repository (caDSR). It provides a metadata registry for common data elements. (3) Cancer Bioinformatics Infrastructure Objects (caBIO). They provide Java, Simple Object Access Protocol and HTTP-XML application programming interfaces. The vision for caCORE is to provide a common data management framework that will support the consistency, clarity, and comparability of biomedical research data and information. In addition to providing facilities for data management and redistribution, caCORE helps solve problems of data integration. All NCICB-developed caCORE components are distributed under open-source licenses that support unrestricted usage by both non-profit and commercial entities, and caCORE has laid the foundation for a number of scientific and clinical applications. Based on it, the paper expounds caCORE-base applications simply in several NCI projects, of which one is CMAP (Cancer Molecular Analysis Project), and the other is caBIG (Cancer Biomedical Informatics Grid). In the end, the paper also gives good prospects of caCORE, and while caCORE was born out of the needs of the cancer research community, it is intended to serve as a general resource. Cancer research has historically

  8. Rules for design of nuclear graphite core components - some considerations and approaches

    International Nuclear Information System (INIS)

    Svalbonas, V.; Stilwell, T.C.; Zudans, Z.

    1978-01-01

    The use of graphite as a structural element presents unusual problems both for the designer and stress analysist. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission. (Auth.)

  9. Surface core level shifts of clean and oxygen covered Ir(111)

    Energy Technology Data Exchange (ETDEWEB)

    Bianchi, M; Cassese, D; Cavallin, A; Comin, R; Orlando, F; Postregna, L [Universita degli Studi di Trieste, Via A Valerio 2, 34127, Trieste (Italy); Golfetto, E; Baraldi, A [Dipartimento di Fisica e CENMAT, Universita degli Studi di Trieste, Via A Valerio 2, 34127, Trieste (Italy); Lizzit, S [Sincrotrone Trieste S.C.p.A., S.S. 14 Km 163.5, 34012 Trieste (Italy)], E-mail: alessandro.baraldi@elettra.trieste.it

    2009-06-15

    We present the results of high resolution core level photoelectron spectroscopy employed to investigate the electronic structure of clean and oxygen covered Ir(111) surface. Ir 4f{sub 7/2} core level spectra are shown to be very sensitive to the local atomic environment. For the clean surface we detected two distinct components shifted by 550 meV, originated by surface and bulk atoms. The larger Gaussian width of the bulk component is explained as due to experimentally unresolved subsurface components. In order to determine the relevance of the phonon contribution we examined the thermal behaviour of the core level lineshape using the Hedin-Rosengren theory. From the phonon-induced spectral broadening we found the Debye temperature of bulk and surface atoms to be 298 and 181 K, respectively, which confirms the softening of the vibrational modes at the surface. Oxygen adsorption leads to the appearance of new surface core level components at -200 meV and +230 meV, which are interpreted as due to first-layer Ir atoms differently coordinated with oxygen. The coverage dependence of these components demonstrates that the oxygen saturation corresponds to 0.38 ML, in good agreement with recent density functional theory calculations.

  10. TMI-2 core examination

    International Nuclear Information System (INIS)

    Hobbins, R.R.; MacDonald, P.E.; Owen, D.E.

    1983-01-01

    The examination of the damaged core at the Three Mile Island Unit 2 (TMI-2) reactor is structured to address the following safety issues: fission product release, transport, and deposition; core coolability; containment integrity; and recriticality during severe accidents; as well as zircaloy cladding ballooning and oxidation during so-called design basis accidents. The numbers of TMI-2 components or samples to be examined, the priority of each examination, the safety issue addressed by each examination, the principal examination techniques to be employed, and the data to be obtained and the principal uses of the data are discussed in this paper

  11. Core components for effective infection prevention and control programmes: new WHO evidence-based recommendations

    Directory of Open Access Journals (Sweden)

    Julie Storr

    2017-01-01

    Full Text Available Abstract Health care-associated infections (HAI are a major public health problem with a significant impact on morbidity, mortality and quality of life. They represent also an important economic burden to health systems worldwide. However, a large proportion of HAI are preventable through effective infection prevention and control (IPC measures. Improvements in IPC at the national and facility level are critical for the successful containment of antimicrobial resistance and the prevention of HAI, including outbreaks of highly transmissible diseases through high quality care within the context of universal health coverage. Given the limited availability of IPC evidence-based guidance and standards, the World Health Organization (WHO decided to prioritize the development of global recommendations on the core components of effective IPC programmes both at the national and acute health care facility level, based on systematic literature reviews and expert consensus. The aim of the guideline development process was to identify the evidence and evaluate its quality, consider patient values and preferences, resource implications, and the feasibility and acceptability of the recommendations. As a result, 11 recommendations and three good practice statements are presented here, including a summary of the supporting evidence, and form the substance of a new WHO IPC guideline.

  12. R and D status and requirements for PIE in the fields of the HTGR fuel and the innovative basic research on High-Temperature Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Tobita, Tsutomu; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Ishihara, Masahiro; Hayashi, Kimio; Hoshiya, Taiji; Sekino, Hajime; Ooeda, Etsurou

    1999-09-01

    The High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor (HTGR) in Japan, achieved its first criticality in November 1998 at the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI). In the field of HTGR fuel development, JAERI will proceed research and development (R and D) works by the following steps: (STEP-1) confirmation of irradiation performance of the first-loading fuel of the HTTR, (STEP-2) study on irradiation performance of high burnup SiC-coated fuel particle and (STEP-3) development of ZrC-coated fuel particle. Requirements for post-irradiation examination (PIE) are different for each R and D step. In STEP-1, firstly, hot cells will be prepared in the HTTR reactor building to handle spent fuels. In parallel, general equipments such as those for deconsolidation of fuel compacts and for handling coated fuel particles will be installed in the Hot Laboratory at Oarai. In STEP-2, precise PIE techniques, for example, Raman spectroscopy for measurement of stress on irradiated SiC layer, will be investigated. In STEP-3, new PIE techniques should be developed to investigate irradiation behavior of ZrC-coated particle. In the field of the innovative basic research on high-temperature engineering, some preliminary tests have been made on the research areas of (1) new materials development, (2) fusion technology, (3) radiation chemistry and (4) high-temperature in-core instrumentation. Requirements for PIE are under investigation, in particular in the field of the new materials development. Besides more general apparatuses including transmission electron microscopy (TEM), some special apparatuses such as an electron spin resonance (ESR) spectrometer, a specific resistance/Hall coefficient measuring system and a differential scanning calorimeter (DSC) are planned to install in the Hot Laboratory at Oarai. Acquisition of advanced knowledge on the irradiation behavior is expected in

  13. Identifying Core Mobile Learning Faculty Competencies Based Integrated Approach: A Delphi Study

    Science.gov (United States)

    Elbarbary, Rafik Said

    2015-01-01

    This study is based on the integrated approach as a concept framework to identify, categorize, and rank a key component of mobile learning core competencies for Egyptian faculty members in higher education. The field investigation framework used four rounds Delphi technique to determine the importance rate of each component of core competencies…

  14. Diffusion of cesium and iodine in compressed IG-110 graphite compacts

    International Nuclear Information System (INIS)

    Carter, L.M.; Brockman, J.D.; Robertson, J.D.; Loyalka, S.K.

    2016-01-01

    Nuclear graphite grade IG-110 is currently used in the High Temperature Engineering Test Reactor (HTTR) in Japan for certain permanent and replaceable core components, and is a material of interest in general. Therefore, transport parameters for fission products in this material are needed. Measurement of diffusion through pressed compacts of IG-110 graphite is experimentally attractive because they are easy to prepare with homogeneous distributions of fission product surrogates. In this work, we measured diffusion coefficients for Cs and I in pressed compacts made from IG-110 powder in the 1079–1290 K temperature range, and compared them to those obtained in as-received IG-110. - Highlights: • A method for analysis of fission product diffusion in graphite by ICP-MS was applied to pressed IG-110 graphite compacts containing cesium and iodine. • Diffusion coefficients for cesium and iodine were obtained. • The measurement design simulates HTGR conditions of high temperature and flowing helium.

  15. Heat transfer characteristics evaluation of heat exchangers of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki; Inagaki, Yoshiyuki; Hanawa, Hiromi; Fujisaki, Katsuo; Yonekawa, Hideo

    2005-06-01

    Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001 and hydrogen of 120 Nm 3 /h was successfully produced in overall performance test. Totally 7 times operational tests were performed from March 2002 to December 2004. A lot of operational test data on heat exchanges were obtained in these tests. In this report specifications and structures of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Heat transfer correlation equations for inside and outside tube were chosen from references. Spreadsheet programs were newly made to evaluate heat transfer characteristics from measured test data such as inlet and outlet temperature pressure and flow-rate. Overall heat-transfer coefficients obtained from the experimental data were compared and evaluated with the calculated values with heat transfer correlation equation. As a result, actual measurement values of all heat exchangers gave close agreement with the calculated values with correlation equations. Thermal efficiencies of the heat exchangers were adequate as they were well accorded with design value. (author)

  16. Synthesis and Plasmonic Understanding of Core/Satellite and Core Shell Nanostructures

    Science.gov (United States)

    Ruan, Qifeng

    Localized surface plasmon resonance, which stems from the collective oscillations of conduction-band electrons, endows Au nanocrystals with unique optical properties. Au nanocrystals possess extremely large scattering/absorption cross-sections and enhanced local electromagnetic field, both of which are synthetically tunable. Moreover, when Au nanocrystals are closely placed or hybridized with semiconductors, the coupling and interaction between the individual components bring about more fascinating phenomena and promising applications, including plasmon-enhanced spectroscopies, solar energy harvesting, and cancer therapy. The continuous development in the field of plasmonics calls for further advancements in the preparation of high-quality plasmonic nanocrystals, the facile construction of hybrid plasmonic nanostructures with desired functionalities, as well as deeper understanding and efficient utilization of the interaction between plasmonic nanocrystals and semiconductor components. In this thesis, I developed a seed-mediated growth method for producing size-controlled Au nanospheres with high monodispersity and assembled Au nanospheres of different sizes into core/satellite nanostructures for enhancing Raman signals. For investigating the interactions between Au nanocrystals and semiconductors, I first prepared (Au core) (TiO2 shell) nanostructures, and then studied their synthetically controlled plasmonic properties and light-harvesting applications. Au nanocrystals with spherical shapes are desirable in plasmon-coupled systems owing to their high geometrical symmetry, which facilitates the analysis of electrodynamic responses in a classical electromagnetic framework and the investigation of quantum tunneling and nonlocal effects. I prepared remarkably uniform Au nanospheres with diameters ranging from 20 nm to 220 nm using a simple seed-mediated growth method associated with mild oxidation. Core/satellite nanostructures were assembled out of differently sized

  17. Statistics of Shared Components in Complex Component Systems

    Science.gov (United States)

    Mazzolini, Andrea; Gherardi, Marco; Caselle, Michele; Cosentino Lagomarsino, Marco; Osella, Matteo

    2018-04-01

    Many complex systems are modular. Such systems can be represented as "component systems," i.e., sets of elementary components, such as LEGO bricks in LEGO sets. The bricks found in a LEGO set reflect a target architecture, which can be built following a set-specific list of instructions. In other component systems, instead, the underlying functional design and constraints are not obvious a priori, and their detection is often a challenge of both scientific and practical importance, requiring a clear understanding of component statistics. Importantly, some quantitative invariants appear to be common to many component systems, most notably a common broad distribution of component abundances, which often resembles the well-known Zipf's law. Such "laws" affect in a general and nontrivial way the component statistics, potentially hindering the identification of system-specific functional constraints or generative processes. Here, we specifically focus on the statistics of shared components, i.e., the distribution of the number of components shared by different system realizations, such as the common bricks found in different LEGO sets. To account for the effects of component heterogeneity, we consider a simple null model, which builds system realizations by random draws from a universe of possible components. Under general assumptions on abundance heterogeneity, we provide analytical estimates of component occurrence, which quantify exhaustively the statistics of shared components. Surprisingly, this simple null model can positively explain important features of empirical component-occurrence distributions obtained from large-scale data on bacterial genomes, LEGO sets, and book chapters. Specific architectural features and functional constraints can be detected from occurrence patterns as deviations from these null predictions, as we show for the illustrative case of the "core" genome in bacteria.

  18. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  19. Design standard issues for ITER in-vessel components

    International Nuclear Information System (INIS)

    Majumdar, S.

    1994-01-01

    Unique requirements that must be addressed by a structural design code for the ITER have been summarized. Existing codes such as ASME Section III, or the French RCC-MR were developed primarily for fission reactor out-of-core components and are not directly applicable to the ITER. They may be used either as a guide for developing a design code for the ITER or as interim standards. However, new rules will be needed for handling the irradiation-induced embrittlement problems faced by the ITER blanket components. Design standards developed in the past for the design of fission reactor core components in the United States can be used as guides in this area

  20. A Framework for Managing Core Facilities within the Research Enterprise

    OpenAIRE

    Haley, Rand

    2009-01-01

    Core facilities represent increasingly important operational and strategic components of institutions' research enterprises, especially in biomolecular science and engineering disciplines. With this realization, many research institutions are placing more attention on effectively managing core facilities within the research enterprise. A framework is presented for organizing the questions, challenges, and opportunities facing core facilities and the academic units and institutions in which th...

  1. Component mode synthesis methods for 3-D heterogeneous core calculations applied to the mixed-dual finite element solver MINOS

    International Nuclear Information System (INIS)

    Guerin, P.; Baudron, A.M.; Lautard, J.J.; Van Criekingen, S.

    2007-01-01

    This paper describes a new technique for determining the pin power in heterogeneous three-dimensional calculations. It is based on a domain decomposition with overlapping sub-domains and a component mode synthesis (CMS) technique for the global flux determination. Local basis functions are used to span a discrete space that allows fundamental global mode approximation through a Galerkin technique. Two approaches are given to obtain these local basis functions. In the first one (the CMS method), the first few spatial eigenfunctions are computed on each sub-domain, using periodic boundary conditions. In the second one (factorized CMS method), only the fundamental mode is computed, and we use a factorization principle for the flux in order to replace the higher-order Eigenmodes. These different local spatial functions are extended to the global domain by defining them as zero outside the sub-domain. These methods are well fitted for heterogeneous core calculations because the spatial interface modes are taken into account in the domain decomposition. Although these methods could be applied to higher-order angular approximations-particularly easily to an SPN approximation-the numerical results we provide are obtained using a diffusion model. We show the methods' accuracy for reactor cores loaded with uranium dioxide and mixed oxide assemblies, for which standard reconstruction techniques are known to perform poorly. Furthermore, we show that our methods are highly and easily parallelizable. (authors)

  2. Nuclear core baffling apparatus

    International Nuclear Information System (INIS)

    Cooper, F.W. Jr.; Silverblatt, B.L.; Knight, C.B.; Berringer, R.T.

    1979-01-01

    An apparatus for baffling the flow of reactor coolant fluid into and about the core of a nuclear reactor is described. The apparatus includes a plurality of longitudinally aligned baffle plates with mating surfaces that allow longitudinal growth with temperature increases while alleviating both leakage through the aligned plates and stresses on the components supporting the plates

  3. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  4. Development of standard components for remote handling

    International Nuclear Information System (INIS)

    Taguchi, Kou; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira

    1998-01-01

    The core of Fusion Experimental Reactor consists of various components such as superconducting magnets and forced-cooled in-vessel components, which are remotely maintained due to intense of gamma radiation. Mechanical connectors such as cooling pipe connections, insulation joints and electrical connectors are commonly used for maintenance of these components and have to be standardized in terms of remote handling. This paper describes these mechanical connectors developed as the standard component compatible with remote handling and tolerable for radiation. (author)

  5. Development of standard components for remote handling

    Energy Technology Data Exchange (ETDEWEB)

    Taguchi, Kou; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The core of Fusion Experimental Reactor consists of various components such as superconducting magnets and forced-cooled in-vessel components, which are remotely maintained due to intense of gamma radiation. Mechanical connectors such as cooling pipe connections, insulation joints and electrical connectors are commonly used for maintenance of these components and have to be standardized in terms of remote handling. This paper describes these mechanical connectors developed as the standard component compatible with remote handling and tolerable for radiation. (author)

  6. Stress wave nondestructive evaluation of Douglas-fir peeler cores

    Science.gov (United States)

    Robert J. Ross; John I. Zerbe; Xiping Wang; David W. Green; Roy F. Pellerin

    2005-01-01

    With the need for evaluating the utilization of veneer peeler log cores in higher value products and the increasing importance of utilizing round timbers in poles, posts, stakes, and building construction components, we conducted a cooperative project to verify the suitability of stress wave nondestructive evaluation techniques for assessing peeler cores and some...

  7. Annual report on experimental operations and maintenances of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2002 fiscal year (Contract research)

    International Nuclear Information System (INIS)

    Hayashi, Koji; Ohashi, Hirofumi; Inaba, Yoshitomo; Kato, Michio; Aita, Hideki; Morisaki, Norihiro; Takeda, Tetsuaki; Nishihara, Tetsuo; Takada, Shoji; Inagaki, Yoshiyuki

    2006-03-01

    This report describes 2002 fiscal-year experimental test operations of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system. The improvement works were performed in May 2002. The second experimental test operation was performed from June 2002 and the performances of the improved parts were confirmed. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July 2002. The third experimental test operation was performed, from October 2002, for (a) start-up and shutdown test, (b) process change test, (c) chemical reaction shutdown test and (d) characteristics test on steam reformer. It was confirmed that the changes of helium gas temperature, caused at steam reformer, could be mitigated into the target range by the steam generator. Maintenance works of high-pressure gas production facilities were also performed in February 2003. This report is summarized with the outline and the results of the test, maintenance works and inspections, and operation records in mentioned above. (author)

  8. 20180311 - High Throughput Transcriptomics: From screening to pathways (SOT 2018)

    Science.gov (United States)

    The EPA ToxCast effort has screened thousands of chemicals across hundreds of high-throughput in vitro screening assays. The project is now leveraging high-throughput transcriptomic (HTTr) technologies to substantially expand its coverage of biological pathways. The first HTTr sc...

  9. Development of carbon/carbon composite control rod for HTTR. 1. Preparation of elements and their fracture tests

    International Nuclear Information System (INIS)

    Eto, Motokuni; Ishiyama, Shintaro; Ugachi, Hirokazu

    1996-08-01

    For the High Temperature Engineering Test Reactor(HTTR) the control rod sleeve is made of Alloy 800H for which a particular process is imposed when the reactor needs to be scrammed. The less restricted operation of the reactor would be attained if there would be the control rod more resistant to high temperature and neutron irradiation. This report summarizes the results which have been obtained as of March 1996 in the course of the development of the C/C composite control rod. Materials used were pitch- or PAN-based fiber-reinforced 2-dimensional carbon composites, from which preforms of the elements of a control rod were fabricated. The preforms were carbonized at 1000degC after being impregnated with pitch. Then they were graphitized at 3000degC, followed by a purification treatment with halogen. The elements included the pellet holder, lace truck and pin. The pin was fabricated by the fiber laminating technique. A control rod is to consist of pellet holders which are connected by the lace trucks with pins. Various strength tests were carried out on these elements. An irradiation of the elements made of PAN-based material was performed in JRR-3 at 900±50degC to a neutron fluence of 1x10 25 n/m 2 (E>29fJ). As for the strength tests on the elements, there were some differences between PAN- and pitch-based composites: In general, elements made of PAN-based composite showed the more plastic behavior before they fractured, whereas those of pitch-based material behaved in the more brittle manner. Fracture tests of the irradiated elements showed that fracture load and fracture displacement enough for assuring the integrity of the control rod structure were maintained even after the irradiation. It was also found that if the applied load was parallel to the fiber felt plane both fracture load and strain increased, whereas the load increase and strain decrease were observed for the applied load against the plane. (J.P.N.)

  10. Development of carbon/carbon composite control rod for HTTR. 1. Preparation of elements and their fracture tests

    Energy Technology Data Exchange (ETDEWEB)

    Eto, Motokuni; Ishiyama, Shintaro; Ugachi, Hirokazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-08-01

    For the High Temperature Engineering Test Reactor(HTTR) the control rod sleeve is made of Alloy 800H for which a particular process is imposed when the reactor needs to be scrammed. The less restricted operation of the reactor would be attained if there would be the control rod more resistant to high temperature and neutron irradiation. This report summarizes the results which have been obtained as of March 1996 in the course of the development of the C/C composite control rod. Materials used were pitch- or PAN-based fiber-reinforced 2-dimensional carbon composites, from which preforms of the elements of a control rod were fabricated. The preforms were carbonized at 1000degC after being impregnated with pitch. Then they were graphitized at 3000degC, followed by a purification treatment with halogen. The elements included the pellet holder, lace truck and pin. The pin was fabricated by the fiber laminating technique. A control rod is to consist of pellet holders which are connected by the lace trucks with pins. Various strength tests were carried out on these elements. An irradiation of the elements made of PAN-based material was performed in JRR-3 at 900{+-}50degC to a neutron fluence of 1x10{sup 25} n/m{sup 2} (E>29fJ). As for the strength tests on the elements, there were some differences between PAN- and pitch-based composites: In general, elements made of PAN-based composite showed the more plastic behavior before they fractured, whereas those of pitch-based material behaved in the more brittle manner. Fracture tests of the irradiated elements showed that fracture load and fracture displacement enough for assuring the integrity of the control rod structure were maintained even after the irradiation. It was also found that if the applied load was parallel to the fiber felt plane both fracture load and strain increased, whereas the load increase and strain decrease were observed for the applied load against the plane. (J.P.N.)

  11. On the core-halo structure of NGC 604

    CERN Document Server

    Melnick, Yu M

    1980-01-01

    A detailed study is presented of the core-halo structure of the largest H II region in M 33, NGC 604, using newly obtained multi- aperture H/sub beta / photometry and Fabry-Perot interferometry, in conjunction with published radio continuum observations. Based on a comparison between the radio continuum and H/sub beta / luminosities of NGC 604, a dust density of rho /sub d/=6 10/sup -25/ g cm/sup -3/ is derived for the nebular core, in good agreement with published far- infrared results. By contrast, the halo of NGC 604 appears to contain virtually no dust. It is also shown that the turbulent component of the H/sub alpha /-line profile width of the halo of NGC 604 is significantly lower than that of the nebular core. This result is found to be inconsistent with models in which the highly supersonic velocities implied by the observed emission line profile widths in both nebular components are interpreted in terms of expansion motions. (14 refs).

  12. Photo-disintegration of heavy nuclei at the core of Cen A

    Energy Technology Data Exchange (ETDEWEB)

    Kundu, Esha [Tata Institute of Fundamental Research, Homi Bhabha Road, Colaba, Mumbai 400005 (India); Gupta, Nayantara, E-mail: esha.kundu@gmail.com, E-mail: nayan@rri.res.in [Raman Research Institute, Sadashiva Nagar, Bangalore 560080 (India)

    2014-04-01

    Fermi LAT has detected gamma ray emissions from the core of Cen A. More recently, a new component in the gamma ray spectrum from the core has been reported in the energy range of 4 GeV to tens of GeV. We show that the new component and the HESS detected spectrum of gamma rays from the core at higher energy have possibly a common origin in photo-disintegration of heavy nuclei. Assuming the cosmic rays are mostly Fe nuclei inside the core and their spectrum has a low energy cut-off at 52 TeV in the wind frame moving with a Doppler factor 0.25 with respect to the observer on earth, the cosmic ray luminosity required to explain the observed gamma ray flux above 1 GeV is found to be 1.5 × 10{sup 43} erg/sec.

  13. History and Systems of Psychology: A Course to Unite a Core Curriculum

    Science.gov (United States)

    Williams, Joshua L.; McCarley, Nancy; Kraft, John

    2013-01-01

    Core curricula are designed, in part, to help undergraduate students become intellectually well-rounded. To merge core curricula with the components of the scholarship of teaching and learning movement, students engaged in core curricula need capstone courses designed to aid them in retaining information over the long term and synthesizing…

  14. Numerical analysis of temperature fluctuation in core outlet region of China experimental fast reactor

    International Nuclear Information System (INIS)

    Zhu Huanjun; Xu Yijun

    2014-01-01

    The temperature fluctuation in core outlet region of China Experimental Fast Reactor (CEFR) was numerically simulated by the CFD software Star CCM+. With the core outlet temperatures, flows etc. under rated conditions given as boundary conditions, a 1/4 region model of the reactor core outlet region was established and calculated using LES method for this problem. The analysis results show that while CEFR operates under rated conditions, the temperature fluctuation in lower part of core outlet region is mainly concentrated in area over the edge components (steel components, control rod assembly), and one in upper part is remarkable in area above all the components. The largest fluctuation amplitude is 19 K and the remarkable frequency is below 5 Hz, and it belongs to typically low frequency fluctuation. The conclusion is useful for further experimental work. (authors)

  15. Structural assessment of TAPS core shroud under accident loads

    International Nuclear Information System (INIS)

    Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1996-09-01

    Over the last few years, the Core Shroud of Boiling Water Reactors (BWRs) operating in foreign countries, have developed cracks at weld locations. As a first step for assessment of structural safety of Tarapur Atomic Power Station (TAPS) core shroud, its detailed stress analysis was done for postulated accident loads. This report is concerned with structural assessment of core shroud, of BWR at TAPS, subjected to loads resulting from main steam line break (MSLB), recirculation line break (RLB) and safe shut down earthquake. The stress analysis was done for core shroud in healthy condition and without any crack since, visual examination conducted till now, do not indicate presence of any flaw. Dynamic structural analysis for MSLB and RLB events was done using dynamic load factor (DLF) method. The complete core shroud and its associated components were modelled and analysed using 3D plate/shell elements. Since, the components of core shroud are submerged in water, hence, hydrodynamic added mass was also considered for evaluation of natural frequencies. It was concluded that from structural point of view, adequate safety margin is available under all the accident loads. Nonlinear analysis was done to evaluate buckling/collapse load. The collapse/buckling load have sufficient margin against the allowable limits. The displacements are low hence, the insertion of control rod may not be affected. (author)

  16. Immobilization of Three-Mile Island core debris

    International Nuclear Information System (INIS)

    Welch, J.M.; Miller, R.L.; Flinn, J.E.

    1983-01-01

    The immobilization of Three-Mile Island core debris in iron-enriched basalt (IEB), a fused-cast nuclear waste form, was considered. The amount of zirconium clad UO 2 fuel assemblies that can be dissolved in IEB using the Zr to UO 2 ratio present in the core was bracketed between 25 and 30% at 1500 0 C. The factors controlling the rate of dissolution of fuel pellets and Inconel, a structural component of the core, were investigated. Since the UO 2 dissolved in IEB could be a valuable resource in the future, the recovery of uranium from IEB using conventional ore-dressing and leaching techniques was assessed

  17. Using sol-gel component as additive to foundry coatings to improve casting quality

    DEFF Research Database (Denmark)

    Nwaogu, Ugochukwu Chibuzoh; Poulsen, T.; Gravesen, B.

    2012-01-01

    The improvement of foundry coatings to enhance performance is important. This paper investigates the effect of using sol‐gel component as an additive to foundry coatings applied on chemically bonded sand cores. Three parameters at three levels each were investigated using Taguchi experimental...... was evaluated. The results show that the surface quality of castings obtained by adding the sol‐gel component to the coatings for cold box cores has no significant difference from castings produced with coatings without sol‐gel component. On the other hand, the addition of the sol‐gel component in coatings...

  18. Design and development of GTHTR300

    International Nuclear Information System (INIS)

    Yan, X.; Kunitomi, K.; Nakata, T.; Shiozawa, S.

    2002-01-01

    Development and recent successful operations of the HTTR in JAERI are coordinated with programs to study applied systems of the promising reactor technology. One such program being carried out from 2001 to 2008 is design and development of the Gas Turbine High Temperature Reactor, GTHTR300, with objective to allow for demonstration of the plant in 2010s in Japan. GTHTR300 features an original design of fuel cycle based on improved HTTR fuel element and of simplified plant system in pursuit of targeted economics with minimum development requirements. The fuel cycle characterizes on high burnup, low power peaking factor and extended refueling interval. The plant system applies such simplified design features as conventional steel RPV, non-intercooled cycle, horizontal single-shaft gas turbine generator and distributed modular maintenance of the overall plant. Research and development essential to validating the design includes component development and control testing. This paper describes the reference plant design and associated continuing R and D activities in JAERI. (author)

  19. Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko

    2000-03-01

    A rise-to-power test of the high temperature engineering test reactor (HTTR) started on September 28 in 1999 for establishing and upgrading the technological basis for the high temperature gas-cooled reactor (HTGR). A loss of off-site electric power test of the HTTR from the normal operation under 15 and 30 MW thermal power will be carried out in the rise-to-power test. Analytical evaluations on transient behaviors of the reactor and plant during the loss of off-site electric power were conducted. These estimations are proposed as benchmark problems for the IAEA coordinated research program on 'Evaluation of HTGR Performance'. This report describes an event scenario of transient during the loss of off-site electric power, the outline of major components and system, detailed thermal and nuclear data set for these problems and pre-estimation results of the benchmark problems by an analytical code 'ACCORD' for incore and plant dynamics of the HTGR. (author)

  20. Core design methods for advanced LMFBRs

    International Nuclear Information System (INIS)

    Chandler, J.C.; Marr, D.R.; McCurry, D.C.; Cantley, D.A.

    1977-05-01

    The multidiscipline approach to advanced LMFBR core design requires an iterative design procedure to obtain a closely-coupled design. HEDL's philosophy requires that the designs should be coupled to the extent that the design limiting fuel pin, the design limiting duct and the core reactivity lifetime should all be equal and should equal the fuel residence time. The design procedure consists of an iterative loop involving three stages of the design sequence. Stage 1 consists of general mechanical design and reactor physics scoping calculations to arrive at an initial core layout. Stage 2 consists of detailed reactor physics calculations for the core configuration arrived at in Stage 1. Based upon the detailed reactor physics results, a decision is made either to alter the design (Stage 1) or go to Stage 3. Stage 3 consists of core orificing and detailed component mechanical design calculations. At this point, an assessment is made regarding design adequacy. If the design is inadequate the entire procedure is repeated until the design is acceptable

  1. Self-Healing Many-Core Architecture: Analysis and Evaluation

    Directory of Open Access Journals (Sweden)

    Arezoo Kamran

    2016-01-01

    Full Text Available More pronounced aging effects, more frequent early-life failures, and incomplete testing and verification processes due to time-to-market pressure in new fabrication technologies impose reliability challenges on forthcoming systems. A promising solution to these reliability challenges is self-test and self-reconfiguration with no or limited external control. In this work a scalable self-test mechanism for periodic online testing of many-core processor has been proposed. This test mechanism facilitates autonomous detection and omission of faulty cores and makes graceful degradation of the many-core architecture possible. Several test components are incorporated in the many-core architecture that distribute test stimuli, suspend normal operation of individual processing cores, apply test, and detect faulty cores. Test is performed concurrently with the system normal operation without any noticeable downtime at the application level. Experimental results show that the proposed test architecture is extensively scalable in terms of hardware overhead and performance overhead that makes it applicable to many-cores with more than a thousand processing cores.

  2. Density-based and transport-based core-periphery structures in networks.

    Science.gov (United States)

    Lee, Sang Hoon; Cucuringu, Mihai; Porter, Mason A

    2014-03-01

    Networks often possess mesoscale structures, and studying them can yield insights into both structure and function. It is most common to study community structure, but numerous other types of mesoscale structures also exist. In this paper, we examine core-periphery structures based on both density and transport. In such structures, core network components are well-connected both among themselves and to peripheral components, which are not well-connected to anything. We examine core-periphery structures in a wide range of examples of transportation, social, and financial networks-including road networks in large urban areas, a rabbit warren, a dolphin social network, a European interbank network, and a migration network between counties in the United States. We illustrate that a recently developed transport-based notion of node coreness is very useful for characterizing transportation networks. We also generalize this notion to examine core versus peripheral edges, and we show that the resulting diagnostic is also useful for transportation networks. To examine the properties of transportation networks further, we develop a family of generative models of roadlike networks. We illustrate the effect of the dimensionality of the embedding space on transportation networks, and we demonstrate that the correlations between different measures of coreness can be very different for different types of networks.

  3. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  4. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  5. Properties of polyurethane foam/coconut coir fiber as a core material and as a sandwich composites component

    Science.gov (United States)

    Azmi, M. A.; Abdullah, H. Z.; Idris, M. I.

    2013-12-01

    This research focuses on the fabrication and characterization of sandwich composite panels using glass fiber composite skin and polyurethane foam reinforced coconut coir fiber core. The main objectives are to characterize the physical and mechanical properties and to elucidate the effect of coconut coir fibers in polyurethane foam cores and sandwich composite panels. Coconut coir fibers were used as reinforcement in polyurethane foams in which later were applied as the core in sandwich composites ranged from 5 wt% to 20 wt%. The physical and mechanical properties found to be significant at 5 wt% coconut coir fiber in polyurethane foam cores as well as in sandwich composites. It was found that composites properties serve better in sandwich composites construction.

  6. Properties of polyurethane foam/coconut coir fiber as a core material and as a sandwich composites component

    International Nuclear Information System (INIS)

    Azmi, M A; Abdullah, H Z; Idris, M I

    2013-01-01

    This research focuses on the fabrication and characterization of sandwich composite panels using glass fiber composite skin and polyurethane foam reinforced coconut coir fiber core. The main objectives are to characterize the physical and mechanical properties and to elucidate the effect of coconut coir fibers in polyurethane foam cores and sandwich composite panels. Coconut coir fibers were used as reinforcement in polyurethane foams in which later were applied as the core in sandwich composites ranged from 5 wt% to 20 wt%. The physical and mechanical properties found to be significant at 5 wt% coconut coir fiber in polyurethane foam cores as well as in sandwich composites. It was found that composites properties serve better in sandwich composites construction

  7. Korrelasjon mellom core styrke, core stabilitet og utholdende styrke i core

    OpenAIRE

    Berg-Olsen, Andrea Marie; Fugelsøy, Eivor; Maurstad, Ann-Louise

    2010-01-01

    Formålet med studien var å se hvilke korrelasjon det er mellom core styrke, core stabilitet og utholdende styrke i core. Testingen bestod av tre hoveddeler hvor vi testet core styrke, core stabilitet og utholdende styrke i core. Innenfor core styrke og utholdende styrke i core ble tre ulike tester utført. Ved måling av core stabilitet ble det gjennomført kun en test. I core styrke ble isometrisk abdominal fleksjon, isometrisk rygg ekstensjon og isometrisk lateral fleksjon testet. Sit-ups p...

  8. Sealing performance test for main flange of pressure vessel of T2 test section in HENDEL

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Inagaki, Yoshiyuki; Matsumoto, Kiminori; Kondou, Yasuo; Suzuki, Kunihiko; Miyamoto, Yoshiaki; Asami, Masanobu.

    1990-12-01

    A pressure vessel of T 2 test section in helium engineering demonstration loop (HENDEL) was fabricated to the same scale of the reactor pressure vessel made of 2(1/4)Cr-1Mo steel in high temperature engineering test reactor (HTTR). Also, the sealing structure of a main flange of pressure vessel in T 2 test section was composed of the double metal O-rings and Ω-seal which would be used in the sealing structure of HTTR. The sealing performance test for the main flange of the pressure vessel in T 2 test section was carried out to confirm the integrity of sealing structure of a main flange in HTTR. T 2 test section has been operated about 7700 hours in previous 18 cycles. The leakage of helium gas from inner metal O-ring was measured by the static pressurized process under the operating condition of HTTR (helium gas: 400degC, 40kg/cm 2 G, 4gk/s). The calculated leakage of helium gas was less than 9.6x10 -7 atm·cm 3 /sec. From the result, it is expected that the sealing structure of main flange in HTTR would maintain the leak tightness in the life. (author)

  9. Analysis on High Temperature Aging Property of Self-brazing Aluminum Honeycomb Core at Middle Temperature

    Directory of Open Access Journals (Sweden)

    ZHAO Huan

    2016-11-01

    Full Text Available Tension-shear test was carried out on middle temperature self-brazing aluminum honeycomb cores after high temperature aging by micro mechanical test system, and the microstructure and component of the joints were observed and analyzed using scanning electron microscopy and energy dispersive spectroscopy to study the relationship between brazing seam microstructure, component and high temperature aging properties. Results show that the tensile-shear strength of aluminum honeycomb core joints brazed by 1060 aluminum foil and aluminum composite brazing plate after high temperature aging(200℃/12h, 200℃/24h, 200℃/36h is similar to that of as-welded joints, and the weak part of the joint is the base metal which is near the brazing joint. The observation and analysis of the aluminum honeycomb core microstructure and component show that the component of Zn, Sn at brazing seam is not much affected and no compound phase formed after high temperature aging; therefore, the main reason for good high temperature aging performance of self-brazing aluminum honeycomb core is that no obvious change of brazing seam microstructure and component occurs.

  10. Core reset system design for linear induction accelerator

    International Nuclear Information System (INIS)

    Durga Praveen Kumar, D.; Mitra, S.; Sharma, Archana; Nagesh, K.V.; Chakravarthy, D.P.

    2006-01-01

    A repetitive pulsed power system based Linear Induction Accelerator (LIA-200) is being developed at BARC to get an electron beam of 200keV, 5kA, 50ns, 10-100 Hz. Amorphous core is the heart of these accelerators. It serves various functions in different subsystems viz. pulse power modulator, pulse transformer, magnetic switches and induction cavities. One of the factors that make the magnetic components compact is utilization of the total flux swing available in the core. In the present system, magnetic switches, pulse transformers, and induction cavity are designed to avail the full flux swing available in the core. For achieving this objective, flux density in the core has to be kept at the reverse saturation, before the main pulse is applied. The electrical circuit which makes it possible is called the core reset system. In this paper the details of core reset system designed for LIA-200 are described. (author)

  11. International Classification of Functioning, Disability and Health Core Sets for cerebral palsy, autism spectrum disorder, and attention-deficit-hyperactivity disorder.

    Science.gov (United States)

    Schiariti, Verónica; Mahdi, Soheil; Bölte, Sven

    2018-05-30

    Capturing functional information is crucial in childhood disability. The International Classification of Functioning, Disability and Health (ICF) Core Sets promote assessments of functional abilities and disabilities in clinical practice regarding circumscribed diagnoses. However, the specificity of ICF Core Sets for childhood-onset disabilities has been doubted. This study aimed to identify content commonalities and differences among the ICF Core Sets for cerebral palsy (CP), and the newly developed Core Sets for autism spectrum disorder (ASD) and attention-deficit-hyperactivity disorder (ADHD). The categories within each Core Set were aggregated at the ICF component and chapter levels. Content comparison was conducted using descriptive analyses. The activities and participation component of the ICF was the most covered across all Core Sets. Main differences included representation of ICF components and coverage of ICF chapters within each component. CP included all ICF components, while ADHD and ASD predominantly focused on activities and participation. Environmental factors were highly represented in the ADHD Core Sets (40.5%) compared to the ASD (28%) and CP (27%) Core Sets. International Classification of Functioning, Disability and Health Core Sets for CP, ASD, and ADHD capture both common but also unique functional information, showing the importance of creating condition-specific, ICF-based tools to build functional profiles of individuals with childhood-onset disabilities. The International Classification of Functioning, Disability and Health (ICF) Core Sets for cerebral palsy (CP), autism spectrum disorder (ASD), and attention-deficit-hyperactivity disorder (ADHD) include unique functional information. The ICF-based tools for CP, ASD, and ADHD differ in terms of representation and coverage of ICF components and ICF chapters. Representation of environmental factors uniquely influences functioning and disability across ICF Core Sets for CP, ASD and ADHD.

  12. Strength evaluation code STEP for brittle materials

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Futakawa, Masatoshi.

    1997-12-01

    In a structural design using brittle materials such as graphite and/or ceramics it is necessary to evaluate the strength of component under complex stress condition. The strength of ceramic materials is said to be influenced by the stress distribution. However, in the structural design criteria simplified stress limits had been adopted without taking account of the strength change with the stress distribution. It is, therefore, important to evaluate the strength of component on the basis of the fracture model for brittle material. Consequently, the strength evaluation program, STEP, on a brittle fracture of ceramic materials based on the competing risk theory had been developed. Two different brittle fracture modes, a surface layer fracture mode dominated by surface flaws and an internal fracture mode by internal flaws, are treated in the STEP code in order to evaluate the strength of brittle fracture. The STEP code uses stress calculation results including complex shape of structures analyzed by the generalized FEM stress analysis code, ABAQUS, so as to be possible to evaluate the strength of brittle fracture for the structures having complicate shapes. This code is, therefore, useful to evaluate the structural integrity of arbitrary shapes of components such as core graphite components in the HTTR, heat exchanger components made of ceramics materials etc. This paper describes the basic equations applying to the STEP code, code system with a combination of the STEP and the ABAQUS codes and the result of the verification analysis. (author)

  13. SBIR Final Report. Liquid Core Optical Scintillating Fibers

    International Nuclear Information System (INIS)

    Beetz, C.P.; Steinbeck, J.; Buerstler, R.

    2000-01-01

    This Phase I SBIR project focused on developing flexible scintillating liquid core optical fibers, with potential uses in high-energy calorimetry, tracking, preradiators, active targets or other fast detectors. Progress on the six tasks of the project is summarized. The technical developments involve three technology components: (1) highly flexible capillaries or tubes of relatively low n (index of refraction) to serve as cladding and liquid core containment; (2) scintillator (and clear) fluids of relatively high n to serve as a core-- these fluids must have a high light transmission and, for some applications, radiation hardness; (3) optical end plugs, plug insertion, and plug-cladding tube sealing technology to contain the core fluids in the tubes, and to transmit the light

  14. The seismic assessment of fast reactor cores in the UK

    International Nuclear Information System (INIS)

    Duthie, J.C.; Dostal, M.

    1988-01-01

    The design of the UK Commercial Demonstration Fast Reactor (CDFR) has evolved over a number of years. The design has to meet two seismic requirements: (i) the reactor must cause no hazard to the public during or after the Safe Shutdown Earthquake (SSE); (ii) there must be no sudden reduction in safety for an earthquake exceeding the SSE. The core is a complicated component in the whole reactor. It is usually represented in a very simplified manner in the seismic assessment of the whole reactor station. From this calculation, a time history or response spectrum can be generated for the diagrid, which supports the core, and for the above core structure, which supports the main absorber rods. These data may then be used to perform a detailed assessment of the reactor core. A new simplified model of the core response may then be made and used in a further calculation of the whole reactor. The calculation of the core response only, is considered in the remainder of this paper. One important feature of the fast reactor core, compared with other reactors, is that the components are relatively thin and flexible to promote neutron economy and heat transfer. A further important feature is that there are very small gaps between the wrapper tubes. This leads to very strong fluid-coupling effects. These effects are likely to be beneficial, but adequate techniques to calculate them are only just being developed. 9 refs, figs

  15. Inspection and repair of nuclear components

    International Nuclear Information System (INIS)

    Lahner, K.; Poetz, F.

    1993-01-01

    Despite careful design, manufacturing and operation, some of the important safety-relevant components show deterioration with time. Because of activation and contamination of these components, their inspection and repair has to be performed with manipulators. Some sophisticated manipulators are described, built by ABB Reaktor and used for inspection, maintenance and repair of PWR steam generators, fuel alignment pins, core baffle former bolts and reactor pressure vessel head penetrations. (Z.S.) 7 figs

  16. Identification of core components and transient interactors of the peroxisomal importomer by dual-track stable isotope labeling with amino acids in cell culture analysis.

    Science.gov (United States)

    Oeljeklaus, Silke; Reinartz, Benedikt S; Wolf, Janina; Wiese, Sebastian; Tonillo, Jason; Podwojski, Katharina; Kuhlmann, Katja; Stephan, Christian; Meyer, Helmut E; Schliebs, Wolfgang; Brocard, Cécile; Erdmann, Ralf; Warscheid, Bettina

    2012-04-06

    The importomer complex plays an essential role in the biogenesis of peroxisomes by mediating the translocation of matrix proteins across the organellar membrane. A central part of this highly dynamic import machinery is the docking complex consisting of Pex14p, Pex13p, and Pex17p that is linked to the RING finger complex (Pex2p, Pex10p, Pex12p) via Pex8p. To gain detailed knowledge on the molecular players governing peroxisomal matrix protein import and, thus, the integrity and functionality of peroxisomes, we aimed at a most comprehensive investigation of stable and transient interaction partners of Pex14p, the central component of the importomer. To this end, we performed a thorough quantitative proteomics study based on epitope tagging of Pex14p combined with dual-track stable isotope labeling with amino acids in cell culture-mass spectrometry (SILAC-MS) analysis of affinity-purified Pex14p complexes and statistics. The results led to the establishment of the so far most extensive Pex14p interactome, comprising 9 core and further 12 transient components. We confirmed virtually all known Pex14p interaction partners including the core constituents of the importomer as well as Pex5p, Pex11p, Pex15p, and Dyn2p. More importantly, we identified new transient interaction partners (Pex25p, Hrr25p, Esl2p, prohibitin) that provide a valuable resource for future investigations on the functionality, dynamics, and regulation of the peroxisomal importomer.

  17. Managing Innovation Probabilities: Core-driven vs. Bottleneck-removing Innovations

    OpenAIRE

    Tsutomu Harada

    2015-01-01

    This paper provides a simplified framework of focusing devices that generate different patterns of innovation, i.e., core-driven and bottleneck-removing innovations, and discusses the managerial implications. We show that core-driven innovation should be undertaken when technology components are independent (independent technology system), while bottleneck-removing innovation should be pursued when they are interdependent (interdependent technology system). Different types of focusing device ...

  18. Sensory nerve degeneration in a mouse model mimicking early manifestations of familial amyloid polyneuropathy due to transthyretin Ala97Ser.

    Science.gov (United States)

    Kan, H-W; Chiang, H; Lin, W-M; Yu, I-S; Lin, S-W; Hsieh, S-T

    2018-02-08

    Sensory nerve degeneration and consequent abnormal sensations are the earliest and most prevalent manifestations of familial amyloid polyneuropathy (FAP) due to amyloidogenic transthyretin (TTR). FAP is a relentlessly progressive degenerative disease of the peripheral nervous system. However, there is a lack of mouse models to replicate the early neuropathic manifestations of FAP. We established human TTR knock-in mice by replacing one allele of the mouse Ttr locus with human wild-type TTR (hTTR wt ) or human TTR with the A97S mutation (hTTR A97S ). Given the late onset of neuropathic manifestations in A97S-FAP, we investigated nerve pathology, physiology, and behavioural tests in these mice at two age points: the adult group (8 - 56 weeks) and the ageing group (> 104 weeks). In the adult group, nerve profiles, neurophysiology and behaviour were similar between hTTR wt and hTTR A97S mice. By contrast, ageing hTTR A97S mice showed small fibre neuropathy with decreased intraepidermal nerve fibre density and behavioural signs of mechanical allodynia. Furthermore, significant reductions in sural nerve myelinated nerve fibre density and sensory nerve action potential amplitudes in these mice indicated degeneration of large sensory fibres. The unaffected motor nerve physiology replicated the early symptoms of FAP patients, that is, sensory nerves were more vulnerable to mutant TTR than motor nerves. These results demonstrate that the hTTR A97S mouse model develops sensory nerve pathology and corresponding physiology mimicking A97S-FAP and provides a platform to develop new therapies for the early stage of A97S-FAP. © 2018 British Neuropathological Society.

  19. Components of WWER engineering factors for peaking factors: status and trends

    International Nuclear Information System (INIS)

    Tsyganov, S.V.

    2010-01-01

    One of the topics for discussion at special working group 'Elaboration of the methodology for calculating the core design engineering factors' is the problem of engineering factor components. The list of components corresponds to the phenomena that are taken into account with the engineering factor. It is supposed the better understanding of the influenced phenomena is important stage for developing unified methodology. This paper presents some brief overview of components of the engineering factor for VVER core peaking factors as they are in the Kurchatov Institute methodology. The evolution of some components to less conservative values is observed. Author makes some assumptions as for the further progress in components assessment. The engineering factors providing observance of design limits at normal operation, should cover, with the set probability, the uncertainty, connected with process of core design. For definition of the value of factors it is necessary to define influence of these uncertainties on the investigated parameter of the reactor. Practice consists in defining all possible sources of uncertainties, to estimate influence of each of them, and on their basis to define total influence of all uncertainties. The important stage of a technique of factor calculation is a definition of the list influencing uncertainties. It is obvious that all characteristics of VVER core are known with some uncertainty-owing to manufacturing tolerances, the measurement errors, etc. However essential influence on the parameters connected with safety, render only a part from them. At list formation those characteristics get out only, whose influence is essential to the corresponding parameter. (Author)

  20. Fast Flux Test Facility core restraint system performance

    International Nuclear Information System (INIS)

    Hecht, S.L.; Trenchard, R.G.

    1990-02-01

    Characterizing Fast Flux Test Facility (FFTF) core restraint system performance has been ongoing since the first operating cycle. Characterization consists of prerun analysis for each core load, in-reactor and postirradiation measurements of subassembly withdrawal loads and deformations, and using measurement data to fine tune predictive models. Monitoring FFTF operations and performing trend analysis has made it possible to gain insight into core restraint system performance and head off refueling difficulties while maximizing component lifetimes. Additionally, valuable information for improved designs and operating methods has been obtained. Focus is on past operating experience, emphasizing performance improvements and avoidance of potential problems. 4 refs., 12 figs., 2 tabs

  1. Lower core body temperature and greater body fat are components of a human thrifty phenotype.

    Science.gov (United States)

    Reinhardt, M; Schlögl, M; Bonfiglio, S; Votruba, S B; Krakoff, J; Thearle, M S

    2016-05-01

    In small studies, a thrifty human phenotype, defined by a greater 24-hour energy expenditure (EE) decrease with fasting, is associated with less weight loss during caloric restriction. In rodents, models of diet-induced obesity often have a phenotype including a reduced EE and decreased core body temperature. We assessed whether a thrifty human phenotype associates with differences in core body temperature or body composition. Data for this cross-sectional analysis were obtained from 77 individuals participating in one of two normal physiology studies while housed on our clinical research unit. Twenty-four-hour EE using a whole-room indirect calorimeter and 24-h core body temperature were measured during 24 h each of fasting and 200% overfeeding with a diet consisting of 50% carbohydrates, 20% protein and 30% fat. Body composition was measured by dual X-ray absorptiometry. To account for the effects of body size on EE, changes in EE were expressed as a percentage change from 24-hour EE (%EE) during energy balance. A greater %EE decrease with fasting correlated with a smaller %EE increase with overfeeding (r=0.27, P=0.02). The %EE decrease with fasting was associated with both fat mass and abdominal fat mass, even after accounting for covariates (β=-0.16 (95% CI: -0.26, -0.06) %EE per kg fat mass, P=0.003; β=-0.0004 (-0.0007, -0.00004) %EE kg(-1) abdominal fat mass, P=0.03). In men, a greater %EE decrease in response to fasting was associated with a lower 24- h core body temperature, even after adjusting for covariates (β=1.43 (0.72, 2.15) %EE per 0.1 °C, P=0.0003). Thrifty individuals, as defined by a larger EE decrease with fasting, were more likely to have greater overall and abdominal adiposity as well as lower core body temperature consistent with a more efficient metabolism.

  2. Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

    International Nuclear Information System (INIS)

    Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

    1980-01-01

    Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

  3. In-core fuel management for nuclear reactor

    International Nuclear Information System (INIS)

    Ross, M.F.; Visner, S.

    1986-01-01

    This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

  4. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector - 15271

    International Nuclear Information System (INIS)

    Saas, L.; Le Tellier, R.; Bajard, S.

    2015-01-01

    In this document, we present a simplified geometrical model (0D model) for both the in-core corium propagation transient and the characterization of the mode of corium transfer from the core to the vessel. A degraded core with a formed corium pool is used as an initial state. This initial state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate...). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4

  5. Comparison of HTGR fuel design, manufacture and quality control methods between Japan and China

    International Nuclear Information System (INIS)

    Fu Xioming; Takahashi, Masashi; Ueta, Shouhei; Sawa, Kazuhiro

    2002-05-01

    The first-loading fuel for the HTTR was started to fabricate at Nuclear Fuel Industries (NFI) in 1995 and the HTTR reached criticality in 1998. Meanwhile, 10 MW high temperature reactor (HTR-10) was constructed in Institute of Nuclear Energy Technology (INET) of Tsinghua University, and the first-loading fuel was fabricated concurrently. The HTR-10 reached criticality in December 2000. Though fuel type is different, i.e., pin-in-block type for the HTTR and pebble bed type for the HTR-10, the fabrication method of TRISO coated fuel particles is similar to each other. This report describes comparison of fuel design, fabrication process and quality inspection between them. (author)

  6. Developing an OMERACT Core Outcome Set for Assessing Safety Components in Rheumatology Trials: The OMERACT Safety Working Group.

    Science.gov (United States)

    Klokker, Louise; Tugwell, Peter; Furst, Daniel E; Devoe, Dan; Williamson, Paula; Terwee, Caroline B; Suarez-Almazor, Maria E; Strand, Vibeke; Woodworth, Thasia; Leong, Amye L; Goel, Niti; Boers, Maarten; Brooks, Peter M; Simon, Lee S; Christensen, Robin

    2017-12-01

    Failure to report harmful outcomes in clinical research can introduce bias favoring a potentially harmful intervention. While core outcome sets (COS) are available for benefits in randomized controlled trials in many rheumatic conditions, less attention has been paid to safety in such COS. The Outcome Measures in Rheumatology (OMERACT) Filter 2.0 emphasizes the importance of measuring harms. The Safety Working Group was reestablished at the OMERACT 2016 with the objective to develop a COS for assessing safety components in trials across rheumatologic conditions. The safety issue has previously been discussed at OMERACT, but without a consistent approach to ensure harms were included in COS. Our methods include (1) identifying harmful outcomes in trials of interventions studied in patients with rheumatic diseases by a systematic literature review, (2) identifying components of safety that should be measured in such trials by use of a patient-driven approach including qualitative data collection and statistical organization of data, and (3) developing a COS through consensus processes including everyone involved. Members of OMERACT including patients, clinicians, researchers, methodologists, and industry representatives reached consensus on the need to continue the efforts on developing a COS for safety in rheumatology trials. There was a general agreement about the need to identify safety-related outcomes that are meaningful to patients, framed in terms that patients consider relevant so that they will be able to make informed decisions. The OMERACT Safety Working Group will advance the work previously done within OMERACT using a new patient-driven approach.

  7. The influence of reactor core parameters on effective breeding coefficient Keff

    Institute of Scientific and Technical Information of China (English)

    Liu Li-Po; Liu Yi-Bao; Wang Juan; Yang Bo; Zhang Tao

    2008-01-01

    The values of effective breeding coefficient Keff in a reactor core of nuclear power plant are calculated for different values of parameters (core structure, fuel assembly component) by using the Monte Carlo method. The obtained values of Keff are compared and analysed, which can provide theoretical basis for reactor design.

  8. AN OBSERVED LACK OF SUBSTRUCTURE IN STARLESS CORES

    International Nuclear Information System (INIS)

    Schnee, Scott; Johnstone, Doug; Enoch, Melissa; Culverhouse, Thomas; Leitch, Erik; Marrone, Daniel P.; Sargent, Anneila

    2010-01-01

    In this paper, we present the results of a high resolution (5'') Combined Array for Research in Millimeter-wave Astronomy and Sunyaev-Zeldovich Array survey of the 3 mm continuum emission from 11 of the brightest (at 1.1 mm) starless cores in the Perseus molecular cloud. We detect 2 of the 11 cores, both of which are composed of single structures, and the median 3σ upper limit for the non-detections is 0.2 M sun in a ∼5'' beam. These results are consistent with, and as stringent as, the low detection rate of compact 3 mm continuum emission in dense cores in Perseus reported by Olmi et al. From the non-detection of multiple components in any of the 11 cores, we conclude that starless core mass functions derived from bolometer maps at resolutions range 10''-30'' (e.g., with MAMBO, SCUBA, or Bolocam) are unlikely to be significantly biased by the blending of lower mass cores with small separations. These observations provide additional evidence that the majority of starless cores in Perseus have inner density profiles shallower than r -2 .

  9. A Testbed For Validating the LHC Controls System Core Before Deployment

    CERN Document Server

    Nguyen Xuan, J

    2011-01-01

    Since the start-up of the LHC, it is crucial to carefully test core controls components before deploying them operationally. The Testbed of the CERN accelerator controls group was developed for this purpose. It contains different hardware (PPC, i386) running various operating systems (Linux and LynxOS) and core software components running on front-ends, communication middleware and client libraries. The Testbed first executes integration tests to verify that the components delivered by individual teams interoperate, and then system tests, which verify high-level, end-user functionality. It also verifies that different versions of components are compatible, which is vital, because not all parts of the operational LHC control system can be upgraded simultaneously. In addition, the Testbed can be used for performance and stress tests. Internally, the Testbed is driven by Atlassian Bamboo, a Continuous Integration server, which builds and deploys automatically new software versions into the Test...

  10. A volatile-rich Earth's core inferred from melting temperature of core materials

    Science.gov (United States)

    Morard, G.; Andrault, D.; Antonangeli, D.; Nakajima, Y.; Auzende, A. L.; Boulard, E.; Clark, A. N.; Lord, O. T.; Cervera, S.; Siebert, J.; Garbarino, G.; Svitlyk, V.; Mezouar, M.

    2016-12-01

    Planetary cores are mainly constituted of iron and nickel, alloyed with lighter elements (Si, O, C, S or H). Understanding how these elements affect the physical and chemical properties of solid and liquid iron provides stringent constraints on the composition of the Earth's core. In particular, melting curves of iron alloys are key parameter to establish the temperature profile in the Earth's core, and to asses the potential occurrence of partial melting at the Core-Mantle Boundary. Core formation models based on metal-silicate equilibration suggest that Si and O are the major light element components1-4, while the abundance of other elements such as S, C and H is constrained by arguments based on their volatility during planetary accretion5,6. Each compositional model implies a specific thermal state for the core, due to the different effect that light elements have on the melting behaviour of Fe. We recently measured melting temperatures in Fe-C and Fe-O systems at high pressures, which complete the data sets available both for pure Fe7 and other binary alloys8. Compositional models with an O- and Si-rich outer core are suggested to be compatible with seismological constraints on density and sound velocity9. However, their crystallization temperatures of 3650-4050 K at the CMB pressure of 136 GPa are very close to, if not higher than the melting temperature of the silicate mantle and yet mantle melting above the CMB is not a ubiquitous feature. This observation requires significant amounts of volatile elements (S, C or H) in the outer core to further reduce the crystallisation temperature of the core alloy below that of the lower mantle. References 1. Wood, B. J., et al Nature 441, 825-833 (2006). 2. Siebert, J., et al Science 339, 1194-7 (2013). 3. Corgne, A., et al Earth Planet. Sc. Lett. 288, 108-114 (2009). 4. Fischer, R. a. et al. Geochim. Cosmochim. Acta 167, 177-194 (2015). 5. Dreibus, G. & Palme, H. Geochim. Cosmochim. Acta 60, 1125-1130 (1995). 6. Mc

  11. Dynamo Tests for Stratification Below the Core-Mantle Boundary

    Science.gov (United States)

    Olson, P.; Landeau, M.

    2017-12-01

    Evidence from seismology, mineral physics, and core dynamics points to a layer with an overall stable stratification in the Earth's outer core, possibly thermal in origin, extending below the core-mantle boundary (CMB) for several hundred kilometers. In contrast, energetic deep mantle convection with elevated heat flux implies locally unstable thermal stratification below the CMB in places, consistent with interpretations of non-dipole geomagnetic field behavior that favor upwelling flows below the CMB. Here, we model the structure of convection and magnetic fields in the core using numerical dynamos with laterally heterogeneous boundary heat flux in order to rationalize this conflicting evidence. Strongly heterogeneous boundary heat flux generates localized convection beneath the CMB that coexists with an overall stable stratification there. Partially stratified dynamos have distinctive time average magnetic field structures. Without stratification or with stratification confined to a thin layer, the octupole component is small and the CMB magnetic field structure includes polar intensity minima. With more extensive stratification, the octupole component is large and the magnetic field structure includes intense patches or high intensity lobes in the polar regions. Comparisons with the time-averaged geomagnetic field are generally favorable for partial stratification in a thin layer but unfavorable for stratification in a thick layer beneath the CMB.

  12. The Earth's core: its composition, formation and bearing upon the origin of the Earth

    International Nuclear Information System (INIS)

    Ringwood, A.E.

    1984-01-01

    The density of the outer core is about 3% smaller than pure iron, which implies that the core contains a substantial amount of one or more low atomic mass elements. New experimental data on the solubility of FeO in molten iron are compatible with oxygen being the other element. At atmospheric pressure FeO is extensively soluble in iron at 2500 0 C, completely miscible above 2800 0 C. Also the solubility of FeO in molten iron is considerably increased at higher pressures. The density measurements can be explained if the core contains about 35% FeO; the new data show this to be possible. A model for the formation of the core based on a high FeO content in the Bulk Earth can be explained if the Earth accreted from a mixture of two components: A, a highly reduced, metal-rich devolatilized assemblage and B, a highly oxidized, volatile-rich assemblage similar to Cl chondrites. The formation of these components in the solar nebula is discussed. The large amount of FeO now inferred to be present in the Earth was mainly produced during accretion by oxidation of metallic iron from component A by water from component B. (U.K.)

  13. Analysis of failed nuclear plant components

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1993-01-01

    Argonne National Laboratory has conducted analyses of failed components from nuclear power-generating stations since 1974. The considerations involved in working with an analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (1) intergranular stress-corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor

  14. Analysis of failed nuclear plant components

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1992-07-01

    Argonne National Laboratory has conducted analyses of failed components from nuclear power generating stations since 1974. The considerations involved in working with and analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (a) intergranular stress corrosion cracking of core spray injection piping in a boiling water reactor, (b) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressure water reactor, (c) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (d) failure of pump seal wear rings by nickel leaching in a boiling water reactor

  15. Novel type of chimera spiral waves arising from decoupling of a diffusible component

    Energy Technology Data Exchange (ETDEWEB)

    Tang, Xiaodong; Yang, Tao; Liu, Yang; Zhao, Yuemin; Gao, Qingyu, E-mail: epstein@brandeis.edu, E-mail: gaoqy@cumt.edu.cn [College of Chemical Engineering, China University of Mining and Technology, Xuzhou 221008 (China); Epstein, Irving R., E-mail: epstein@brandeis.edu, E-mail: gaoqy@cumt.edu.cn [Department of Chemistry and Volen Center for Complex Systems, MS 015, Brandeis University, Waltham, Massachusetts 02454-9110 (United States)

    2014-07-14

    Spiral waves composed of coherent traveling waves surrounding a core containing stochastically distributed stationary areas are found in numerical simulations of a three-variable reaction-diffusion system with one diffusible species. In the spiral core, diffusion of this component (w) mediates transitions between dynamic states of the subsystem formed by the other two components, whose dynamics is more rapid than that of w. Diffusive coupling between adjacent sites can be “on” or “off” depending on the subsystem state. The incoherent structures in the spiral core are produced by this decoupling of the slow diffusive component from the fast non-diffusing subsystem. The phase diagram reveals that the region of incoherent behavior in chimera spirals grows drastically, leading to modulation and breakup of the spirals, in the transition zones between 1{sup n-1} and 1{sup n} local mixed-mode oscillations.

  16. Iron snow in the Martian core?

    Science.gov (United States)

    Davies, Christopher J.; Pommier, Anne

    2018-01-01

    The decline of Mars' global magnetic field some 3.8-4.1 billion years ago is thought to reflect the demise of the dynamo that operated in its liquid core. The dynamo was probably powered by planetary cooling and so its termination is intimately tied to the thermochemical evolution and present-day physical state of the Martian core. Bottom-up growth of a solid inner core, the crystallization regime for Earth's core, has been found to produce a long-lived dynamo leading to the suggestion that the Martian core remains entirely liquid to this day. Motivated by the experimentally-determined increase in the Fe-S liquidus temperature with decreasing pressure at Martian core conditions, we investigate whether Mars' core could crystallize from the top down. We focus on the "iron snow" regime, where newly-formed solid consists of pure Fe and is therefore heavier than the liquid. We derive global energy and entropy equations that describe the long-timescale thermal and magnetic history of the core from a general theory for two-phase, two-component liquid mixtures, assuming that the snow zone is in phase equilibrium and that all solid falls out of the layer and remelts at each timestep. Formation of snow zones occurs for a wide range of interior and thermal properties and depends critically on the initial sulfur concentration, ξ0. Release of gravitational energy and latent heat during growth of the snow zone do not generate sufficient entropy to restart the dynamo unless the snow zone occupies at least 400 km of the core. Snow zones can be 1.5-2 Gyrs old, though thermal stratification of the uppermost core, not included in our model, likely delays onset. Models that match the available magnetic and geodetic constraints have ξ0 ≈ 10% and snow zones that occupy approximately the top 100 km of the present-day Martian core.

  17. Paleomagnetism of the Oman Ophiolite: New Results from Oman Drilling Project Cores

    Science.gov (United States)

    Horst, A. J.; Till, J. L.; Koornneef, L.; Usui, Y.; Kim, H.; Morris, A.

    2017-12-01

    The Oman Drilling Project drilled holes at four sites in a transect through the southern massifs of the Samail ophiolite, and recovered 1500 m of igneous and metamorphic rocks. We focus on three sites from the oceanic crustal section including lower layered gabbros (GT1A), the mid-crustal layered to foliated gabbro transition (GT2A), and the shallower transition from sheeted dikes to varitextured gabbros (GT3A). Detailed core descriptions, analyses, and paleomagnetic measurements, were made on D/V Chikyu from July to September 2017 to utilize the core laboratory facilities similar to IODP expeditions. Shipboard measurements included anisotropy of magnetic susceptibility (AMS) and alternating field and thermal demagnetization of 597 discrete samples. Sample demagnetization behavior is varied from each of the cores, with some revealing multiple components of magnetization, and others yielding nearly univectorial data. The interpretation of results from the lower crustal cores is complicated by the pervasive presence of secondary magnetite. In almost all samples, a stable component was resolved (interpreted as a characteristic remanent magnetization) after removal of a lower-coercivity or lower unblocking-temperature component. The inclinations of the stable components in the core reference frame are very consistent in Hole GT1A. However, a transition from negative to positive inclinations in GT2A suggests some structural complexity, possibly as a result of intense late faulting activity. Both abrupt and gradual transitions between multiple zones of negative and positive inclinations occur in Hole GT3A. Interpretation and direct comparison of remanence between drill sites is difficult as recovered core pieces currently remain azimuthally unoriented, and GT2A was drilled at a plunge of 60°, whereas GT1A and GT3A were both drilled vertically. Work is ongoing to use borehole imagery to reorient the core pieces and paleomagnetic data into a geographic in situ reference

  18. Neutron Fluence, Dosimetry and Damage Response Determination in In-Core/Ex-Core Components of the VENUS CEN/SCK LWR Using 3-D Monte Carlo Simulations: NEA's VENUS-3 Benchmark

    International Nuclear Information System (INIS)

    Perlado, J. Manuel; Marian, Jaime; Sanz, Jesus Garcia

    2000-01-01

    Validating state-of-the-art methods used to predict fluence exposure to reactor pressure vessels (RPVs) has become an important issue in identifying the sources of uncertainty in the estimated RPV fluence for pressurized water reactors. This is a very important aspect in evaluating irradiation damage leading to the hardening and embrittlement of such structural components. One of the major benchmark experiments carried out to test three-dimensional methodologies is the VENUS-3 Benchmark Experiment in which three-dimensional Monte Carlo and S n codes have proved more efficient than synthesis methods. At the Instituto de Fusion Nuclear (DENIM) at the Universidad Politecnica de Madrid, a detailed full three-dimensional model of the Venus Critical Facility has been developed making use of the Monte Carlo transport code MCNP4B. The problem geometry and source modeling are described, and results, including calculated versus experimental (C/E) ratios as well as additional studies, are presented. Evidence was found that the great majority of C/E values fell within the 10% tolerance and most within 5%. Tolerance limits are discussed on the basis of evaluated data library and fission spectra sensitivity, where a value ranging between 10 to 15% should be accepted. Also, a calculation of the atomic displacement rate has been carried out in various locations throughout the reactor, finding that values of 0.0001 displacements per atom in external components such as the core barrel are representative of this type of reactor during a 30-yr time span

  19. Predictable Particle Engineering: Programming the Energy Level, Carrier Generation, and Conductivity of Core-Shell Particles.

    Science.gov (United States)

    Yuan, Conghui; Wu, Tong; Mao, Jie; Chen, Ting; Li, Yuntong; Li, Min; Xu, Yiting; Zeng, Birong; Luo, Weiang; Yu, Lingke; Zheng, Gaofeng; Dai, Lizong

    2018-06-20

    Core-shell structures are of particular interest in the development of advanced composite materials as they can efficiently bring different components together at nanoscale. The advantage of this structure greatly relies on the crucial design of both core and shell, thus achieving an intercomponent synergistic effect. In this report, we show that decorating semiconductor nanocrystals with a boronate polymer shell can easily achieve programmable core-shell interactions. Taking ZnO and anatase TiO 2 nanocrystals as inner core examples, the effective core-shell interactions can narrow the band gap of semiconductor nanocrystals, change the HOMO and LUMO levels of boronate polymer shell, and significantly improve the carrier density of core-shell particles. The hole mobility of core-shell particles can be improved by almost 9 orders of magnitude in comparison with net boronate polymer, while the conductivity of core-shell particles is at most 30-fold of nanocrystals. The particle engineering strategy is based on two driving forces: catechol-surface binding and B-N dative bonding and having a high ability to control and predict the shell thickness. Also, this approach is applicable to various inorganic nanoparticles with different components, sizes, and shapes.

  20. Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

    CERN Document Server

    2002-01-01

    This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

  1. French R&D on Materials for the Core Components of SFRs

    International Nuclear Information System (INIS)

    Le Flem, M.; Séran, J.L.; Blat-Yrieix, M.; Garat, V.

    2013-01-01

    ASTRID demonstrator 480-700°C, 110 dpa. • Use of reference materials benefiniting from a large feed-back from the previous French SFRs (Rapsodie, Phénix, SuperPhénix) • Austenitic steels (cladding), Martensitic steels (wrapper tube), B4C (absorbers). • Improving the description of their behavior (swelling, high temperature) • Qualifying the materials regarding the specificities of ASTRID core. Future SFRs 530-750, 180 dpa. • Use of advanced materials with improved properties • ODS ferritic/martensitic steels (cladding), Other metallic solutions as V alloys (cladding), SiC/SiC composites (wrapper tube), Innovative absorbers and reflectors. • R&D to develop/fabricate suitable grades • Qualifying these materials in ASTRID

  2. Pattern recognition on X-ray fluorescence records from Copenhagen lake sediments using principal component analysis

    DEFF Research Database (Denmark)

    Schreiber, Norman; Garcia, Emanuel; Kroon, Aart

    2014-01-01

    Principle Component Analysis (PCA) was performed on chemical data of two sediment cores from an urban fresh-water lake in Copenhagen, Denmark. X-ray fluorescence (XRF) core scanning provided the underlying datasets on 13 variables (Si, K, Ca, Ti, Cr, Mn, Fe, Ni, Cu, Zn, Rb, Cd, Pb). Principle......, Fe, Rb) and characterized the content of minerogenic material in the sediment. In case of both cores, PC2 was a good descriptor emphasized as the contamination component. It showed strong linkages with heavy metals (Cu, Zn, Pb), disclosing changing heavy-metal contamination trends across different...

  3. Multiple age components in individual molybdenite grains

    Science.gov (United States)

    Aleinikoff, John N.; Creaser, Robert A.; Lowers, Heather; Magee, Charles W.; Grauch, Richard I.

    2012-01-01

    Re–Os geochronology of fractions composed of unsized, coarse, and fine molybdenite from a pod of unusual monazite–xenotime gneiss within a granulite facies paragneiss, Hudson Highlands, NY, yielded dates of 950.5 ± 2.5, 953.8 ± 2.6, and 941.2 ± 2.6 Ma, respectively. These dates are not recorded by co-existing zircon, monazite, or xenotime. SEM–BSE imagery of thin sections and separated grains reveals that most molybdenite grains are composed of core and rim plates that are approximately perpendicular. Rim material invaded cores, forming irregular contacts, probably reflecting dissolution/reprecipitation. EPMA and LA-ICP-MS analyses show that cores and rims have different trace element concentrations (for example, cores are relatively enriched in W). On the basis of inclusions of zircon with metamorphic overgrowths, we conclude that molybdenite cores and rims formed after high-grade regional metamorphism. The discovery of cores and rims in individual molybdenite grains is analogous to multi-component U-Pb geochronometers such as zircon, monazite, and titanite; thus, molybdenite should be carefully examined before dating to ensure that the requirement of age homogeneity is fulfilled.

  4. Theoretical and experimental methods to determine the properties of molten core components and reaction products. Pt. 2

    International Nuclear Information System (INIS)

    Nazare, S.; Ondracek, G.; Schulz, B.

    1975-10-01

    In the course of a loss of coolant accident, a sequence of events would be initiated that ultimately could lead to core melting. The course of these events and the consequences of core meltdown would in part be determined by the properties of the core materials and the products of their interaction. On the basis of available theoretical and experimental results, the report attempts an estimation of properties such as: 1) work of adhesion between UO 2 - and (U,Zr) liquid phase, 2) heat of fusion of some melts, 3) heat capacity of liquid reaction products, 4) viscosity of liquid reaction products, 5) thermal conductivity of liquid reaction products. Experimental work is suggested for those cases, where the estimates need to be improved or verified. (orig.) [de

  5. Phase Equilibrium Experiments on Potential Lunar Core Compositions: Extension of Current Knowledge to Multi-Component (Fe-Ni-Si-S-C) Systems

    Science.gov (United States)

    Righter, K.; Pando, K.; Danielson, L.

    2014-01-01

    Numerous geophysical and geochemical studies have suggested the existence of a small metallic lunar core, but the composition of that core is not known. Knowledge of the composition can have a large impact on the thermal evolution of the core, its possible early dynamo creation, and its overall size and fraction of solid and liquid. Thermal models predict that the current temperature at the core-mantle boundary of the Moon is near 1650 K. Re-evaluation of Apollo seismic data has highlighted the need for new data in a broader range of bulk core compositions in the PT range of the lunar core. Geochemical measurements have suggested a more volatile-rich Moon than previously thought. And GRAIL mission data may allow much better constraints on the physical nature of the lunar core. All of these factors have led us to determine new phase equilibria experimental studies in the Fe-Ni-S-C-Si system in the relevant PT range of the lunar core that will help constrain the composition of Moon's core.

  6. Nonlinear Dynamic Model of PMBLDC Motor Considering Core Losses

    DEFF Research Database (Denmark)

    Fasil, Muhammed; Mijatovic, Nenad; Jensen, Bogi Bech

    2017-01-01

    The phase variable model is used commonly when simulating a motor drive system with a three-phase permanent magnet brushless DC (PMBLDC) motor. The phase variable model neglects core losses and this affects its accuracy when modelling fractional-slot machines. The inaccuracy of phase variable mod...... on the detailed analysis of the flux path and the variation of flux in different components of the machine. A prototype of fractional slot axial flux PMBLDC in-wheel motor is used to assess the proposed nonlinear dynamic model....... of fractional-slot machines can be attributed to considerable armature flux harmonics, which causes an increased core loss. This study proposes a nonlinear phase variable model of PMBLDC motor that considers the core losses induced in the stator and the rotor. The core loss model is developed based...

  7. Higher-Order Components for Grid Programming

    CERN Document Server

    Dünnweber, Jan

    2009-01-01

    Higher-Order Components were developed within the CoreGRID European Network of Excellence and have become an optional extension of the popular Globus middleware. This book provides the reader with hands-on experience, describing a collection of example applications from various fields of science and engineering, including biology and physics.

  8. Ranking related entities: components and analyses

    NARCIS (Netherlands)

    Bron, M.; Balog, K.; de Rijke, M.

    2010-01-01

    Related entity finding is the task of returning a ranked list of homepages of relevant entities of a specified type that need to engage in a given relationship with a given source entity. We propose a framework for addressing this task and perform a detailed analysis of four core components;

  9. Heysham II/Torness power stations: Seismic qualification of core structures and boilers

    International Nuclear Information System (INIS)

    Shepherd, D.J.

    1990-01-01

    For the advanced gas cooled reactors at Heysham II and Torness the seismic qualification of the core and support structures and boilers posed special problems. In each case the response was highly non-linear due to impacting. Within the core itself there are many thousands of degrees of freedom each dominated by impacting during the seismic event and these impact forces are transmitted to the support structure. The boilers, although supported and located in the design case by linear systems, have their motion during the seismic event controlled by seismic restraints and other components which introduce substantial impacting during seismic excitation. For both these important components a substantial programme of testing was carried out to validate an analysis approach. This testing and correlation with analysis is described in detail for both components. In the case of the core the qualification was based upon a non-linear code AGRCORE which was specifically developed to handle the large number of impact degrees of freedom for this component. The implementation of this code is also described together with a brief summary of results. The boiler analysis was ultimately carried out using conventional finite difference codes and the implementation of these together with a summary of results is also presented. (author). 13 figs, 1 tab

  10. Storage, handling and movement of fuel and related components at nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    The report describes in general terms the various operations involved in the handling of fresh fuel, irradiated fuel, and core components such as control rods, neutron sources, burnable poisons and removable instruments. It outlines the principal safety problems in these operations and provides the broad safety criteria which must be observed in the design, operation and maintenance of equipment and facilities for handling, transferring, and storing nuclear fuel and core components at nuclear power reactor sites

  11. Analysis of failed nuclear plant components

    Science.gov (United States)

    Diercks, D. R.

    1993-12-01

    Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

  12. Final report on cost estimate of forward superconducting air core toroid

    International Nuclear Information System (INIS)

    Fields, T.

    1992-12-01

    An independent cost-estimate for key components of the forward superconducting air core toroid (ACT) was obtained in May 1992 from an experienced manufacturer of large cryogenic vessels. This new cost estimate is summarized in this report. It implies that a suitably designed ACT may have a cost which is approximately equal to that of the presently designed SDC forward iron core toroid

  13. Equation of State of Fe3C and Implications for the Carbon Content of Earth's Core

    Science.gov (United States)

    Davis, A.; Brauser, N.; Thompson, E. C.; Chidester, B.; Greenberg, E.; Prakapenka, V. B.; Campbell, A.

    2017-12-01

    Carbon is a common component in protoplanetary cores, as represented by iron meteorites. Therefore, along with silicon, oxygen, and other light elements, it is likely to be an alloying component with iron in Earth's core. Previous studies of the densities of iron carbides have not reached the combined pressure and temperature conditions relevant to Earth's core. To better understand the geophysical implications of carbon addition to Earth's core, we report P-V-T measurements of Fe3C to pressures and temperatures exceeding 110 GPa and 2500 K, using synchrotron X-ray diffraction in a laser heated diamond anvil cell. Fitting these measurements to an equation of state and assuming 1.5% density change upon melting and a 4000 K core-mantle boundary temperature, we report a value of 6 wt% carbon necessary to match the PREM density in the outer core. This value should be considered an upper bound due to the likely presence of other light elements.

  14. Thermal-hydraulic characteristics of double flat core HCLWR

    International Nuclear Information System (INIS)

    Sugimoto, Jun; Iwamura, Takamichi; Okubo, Tsutomu; Murao, Yoshio

    1989-02-01

    A thermal-hydraulic characteristics of double flat core high conversion light water reactor (HCLWR) is described. The concept of flat core proposed by Ishiguro et al. is to achieve negative void reactivity coefficient in tight lattice core, and at the same time, high conversion ratio and high burnup can be obtainable. The proposed double flat core HCLWR, based on these physical advantages and the consideration of safety assurance, aims at efficient use of the pressure vessel space to produce comparable thermal output as current 3-loop PWRs. The present work revealed the following items concerning the thermalhydraulic feasibility of the double flat core HCLWR: (1) Main thermal-hydraulic parameters of the plant can be almost the same as current PWRs, showing the use of PWR standard components without major modifications except in core region. (2) Heat removal from the fuel rod in a steady operational condition has enough margin to the critical heat flux (CHF) limit, which is evaluated with the existing CHF correlations. (3) The calculation by REFLA code shows that the maximum cladding temperature in LOCA-reflood is estimated to be far lower than the licensing criteria. It is therefore considered that the proposed double flat core HCLWR is feasible from the point of thermal-hydraulics. Since the available data base has certain applicational limit to the very short core as the present double flat core HCLWR, further detailed assessment is required. (author)

  15. Fire resistance of extruded hollow-core slabs

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl; Sørensen, Lars Schiøtt; Giuliani, Luisa

    2017-01-01

    to the structural codes with data derived from a standard fire test and from a thorough examination of the comprehensive test documentation available on fire exposed hollow-core slabs. Findings – Mechanisms for loss of load-bearing capacity are clarified, and evidence of the fire resistance is found. Originality......Purpose – Prefabricated extruded hollow-core slabs are preferred building components for floor structures in several countries. It is therefore important to be able to document the fire resistance of these slabs proving fulfilment of standard fire resistance requirements of 60 and 120 min found...... in most national building regulations. The paper aims to present a detailed analysis of the mechanisms responsible for the loss of loadbearing capacity of hollow-core slabs when exposed to fire. Design/methodology/approach – Furthermore, it compares theoretica calculation and assessment according...

  16. Structural models of the different trimers present in the core of phycobilisomes from Gracilaria chilensis based on crystal structures and sequences.

    Directory of Open Access Journals (Sweden)

    Jorge Dagnino-Leone

    Full Text Available Phycobilisomes (PBS are accessory light harvesting protein complexes that directionally transfer energy towards photosystems. Phycobilisomes are organized in a central core and rods radiating from it. Components of phycobilisomes in Gracilaria chilensis (Gch are Phycobiliproteins (PBPs, Phycoerythrin (PE, and Phycocyanin (PC in the rods, while Allophycocyanin (APC is found in the core, and linker proteins (L. The function of such complexes depends on the structure of each component and their interaction. The core of PBS from cyanobacteria is mainly composed by cylinders of trimers of α and β subunits forming heterodimers of Allophycocyanin, and other components of the core including subunits αII and β18. As for the linkers, Linker core (LC and Linker core membrane (LCM are essential for the final emission towards photoreaction centers. Since we have previously focused our studies on the rods of the PBS, in the present article we investigated the components of the core in the phycobilisome from the eukaryotic algae, Gracilaria chilensis and their organization into trimers. Transmission electron microscopy provided the information for a three cylinders core, while the three dimensional structure of Allophycocyanin purified from Gch was determined by X-ray diffraction method and the biological unit was determined as a trimer by size exclusion chromatography. The protein sequences of all the components of the core were obtained by sequencing the corresponding genes and their expression confirmed by transcriptomic analysis. These subunits have seldom been reported in red algae, but not in Gracilaria chilensis. The subunits not present in the crystallographic structure were modeled to build the different composition of trimers. This article proposes structural models for the different types of trimers present in the core of phycobilisomes of Gch as a first step towards the final model for energy transfer in this system.

  17. Reactivity control system of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio

    2004-01-01

    The reactivity control system of the high temperature engineering test reactor (HTTR) consists of a control rod system and a reserve shutdown system. During normal operation, reactivity is controlled by the control rod system, which consists of 32 control rods (16 pairs) and 16 control rod drive mechanisms except for the case when the center control rods are removed to perform an irradiation test. In an unlikely event that the control rods fail to be inserted, reserve shutdown system is provided to insert pellets of neutron-absorbing material into the core. Alloy 800H is chosen for the metallic parts of the control rods. Because the maximum temperature of the control rods reaches about 900 deg. C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Observing the guideline, temperature and stress analysis were conducted; it can be confirmed that the target life of the control rods of 5 years can be achieved. Various tests conducted for the control rod system and the reserve shutdown system are also described

  18. A Student Selected Component (or Special Study Module) in Forensic and Legal Medicine: Design, delivery, assessment and evaluation of an optional module as an addition to the medical undergraduate core curriculum.

    Science.gov (United States)

    Kennedy, Kieran M; Wilkinson, Andrew

    2018-01-01

    The General Medical Council (United Kingdom) advocates development of non-core curriculum Student Selected Components and their inclusion in all undergraduate medical school curricula. This article describes a rationale for the design, delivery, assessment and evaluation of Student Selected Components in Forensic and Legal Medicine. Reference is made to the available evidence based literature pertinent to the delivery of undergraduate medical education in the subject area. A Student Selected Component represents an opportunity to highlight the importance of the legal aspects of medical practice, to raise the profile of the discipline of Forensic and Legal Medicine amongst undergraduate medical students and to introduce students to the possibility of a future career in the area. The authors refer to their experiences of design, delivery, assessment and evaluation of Student Selected Components in Forensic and Legal Medicine at their respective Universities in the Republic of Ireland (Galway) and in the United Kingdom (Oxford). Copyright © 2017. Published by Elsevier Ltd.

  19. Materials for 300 to 5000C magnetic components

    International Nuclear Information System (INIS)

    Weichold, M.H.; Pandey, R.K.; Palmer, D.W.

    1980-01-01

    Core materials and winding wire for audio and rf transformers have been investigated to 500 0 C. Audio cores of 2 V Permendur had parameter stability from 25 to 500 0 C and during aging at 450 0 C. High frequency ferrite material, Mix 63, displayed usefulness up to 300 0 C. Both anodized aluminum and ceramic-coated copper wire function to 500 0 C in low voltage or large gauge applications. Components based on these materials operate reliably to 500 0 C

  20. Methodological aspects of core meltdown accidents frequency estimates

    International Nuclear Information System (INIS)

    Matthis, P.

    1984-01-01

    A survey is given of the work of the ecological institute relating to models and methods used in the German Risk Study for the assessment of core meltdown accident frequency. A statistical model used by the ecological institute for the estimation of the outage behaviour of components is taken as a comparison, which leads to the conclusion that no appropriate methods for the assessment of component reliability are available to date. Furthermore, there are no secured methods for error propagation computation. The lower limits for the ranges of reliability of components are calculated by approximation. As a result of imperfect modelling and of a number of methodical inaccuracies and neglects, the German Risk Study underestimates the ranges of component reliability by a factor of 3 to 70 (depending on the type of component). (RF) [de

  1. Multi-core System Architecture for Safety-critical Control Applications

    DEFF Research Database (Denmark)

    Li, Gang

    and size, and high power consumption. Increasing the frequency of a processor is becoming painful now due to the explosive power consumption. Furthermore, components integrated into a single-core processor have to be certified to the highest SIL, due to that no isolation is provided in a traditional single...... certification cost. Meanwhile, hardware platforms with improved processing power are required to execute the applications of larger size. To tackle the two issues mentioned above, the state of the art approaches are using more Electronic Control Units (ECU) in a federated architecture or increasing......-core processor. A promising alternative to improve processing power and provide isolation is to adopt a multi-core architecture with on-chip isolation. In general, a specific multi-core architecture can facilitate the development and certification of safety-related systems, due to its physical isolation between...

  2. Optical properties of core-shell and multi-shell nanorods

    Science.gov (United States)

    Mokkath, Junais Habeeb; Shehata, Nader

    2018-05-01

    We report a first-principles time dependent density functional theory study of the optical response modulations in bimetallic core-shell (Na@Al and Al@Na) and multi-shell (Al@Na@Al@Na and Na@Al@Na@Al: concentric shells of Al and Na alternate) nanorods. All of the core-shell and multi-shell configurations display highly enhanced absorption intensity with respect to the pure Al and Na nanorods, showing sensitivity to both composition and chemical ordering. Remarkably large spectral intensity enhancements were found in a couple of core-shell configurations, indicative that optical response averaging based on the individual components can not be considered as true as always in the case of bimetallic core-shell nanorods. We believe that our theoretical results would be useful in promising applications depending on Aluminum-based plasmonic materials such as solar cells and sensors.

  3. Data Acquisition Backbone Core DABC

    International Nuclear Information System (INIS)

    Adamczewski, J; Essel, H G; Kurz, N; Linev, S

    2008-01-01

    For the new experiments at FAIR new concepts of data acquisition systems have to be developed like the distribution of self-triggered, time stamped data streams over high performance networks for event building. The Data Acquisition Backbone Core (DABC) is a software package currently under development for FAIR detector tests, readout components test, and data flow investigations. All kinds of data channels (front-end systems) are connected by program plug-ins into functional components of DABC like data input, combiner, scheduler, event builder, analysis and storage components. After detailed simulations real tests of event building over a switched network (InfiniBand clusters with up to 110 nodes) have been performed. With the DABC software more than 900 MByte/s input and output per node can be achieved meeting the most demanding requirements. The software is ready for the implementation of various test beds needed for the final design of data acquisition systems at FAIR. The development of key components is supported by the FutureDAQ project of the European Union (FP6 I3HP JRA1)

  4. Suggested Methods for Preventing Core Saturation Instability in HVDC Transmission Systems

    Energy Technology Data Exchange (ETDEWEB)

    Norheim, Ian

    2002-07-01

    In this thesis a study of the HVDC related phenomenon core saturation instability and methods to prevent this phenomenon is performed. It is reason to believe that this phenomenon caused disconnection of the Skagerrak HVDC link 10 August 1993. Internationally, core saturation instability has been reported at several HVDC schemes and thorough complex studies of the phenomenon has been performed. This thesis gives a detailed description of the phenomenon and suggest some interesting methods to prevent the development of it. Core saturation instability and its consequences can be described in a simplified way as follows: It is now assumed that a fundamental harmonic component is present in the DC side current. Due to the coupling between the AC side and the DC side of the HVDC converter, a subsequent second harmonic positive-sequence current and DC currents will be generated on the AC side. The DC currents will cause saturation in the converter transformers. This will cause the magnetizing current to also have a second harmonic positive-sequence component. If a high second harmonic impedance is seen from the commutation bus, a high positive-sequence second harmonic component will be present in the commutation voltages. This will result in a relatively high fundamental frequency component in the DC side voltage. If the fundamental frequency impedance at the DC side is relatively low the fundamental component in the DC side current may become larger than it originally was. In addition the HVDC control system may contribute to the fundamental frequency component in the DC side voltage, and in this way cause a system even more sensitive to core saturation instability. The large magnetizing currents that eventually will flow on the AC side cause large zero-sequence currents in the neutral conductors of the AC transmission lines connected to the HVDC link. This may result in disconnection of the lines. Alternatively, the harmonics in the large magnetizing currents may cause

  5. Enhanced Core Noise Modeling for Turbofan Engines

    Science.gov (United States)

    Stone, James R.; Krejsa, Eugene A.; Clark, Bruce J.

    2011-01-01

    This report describes work performed by MTC Technologies (MTCT) for NASA Glenn Research Center (GRC) under Contract NAS3-00178, Task Order No. 15. MTCT previously developed a first-generation empirical model that correlates the core/combustion noise of four GE engines, the CF6, CF34, CFM56, and GE90 for General Electric (GE) under Contract No. 200-1X-14W53048, in support of GRC Contract NAS3-01135. MTCT has demonstrated in earlier noise modeling efforts that the improvement of predictive modeling is greatly enhanced by an iterative approach, so in support of NASA's Quiet Aircraft Technology Project, GRC sponsored this effort to improve the model. Since the noise data available for correlation are total engine noise spectra, it is total engine noise that must be predicted. Since the scope of this effort was not sufficient to explore fan and turbine noise, the most meaningful comparisons must be restricted to frequencies below the blade passage frequency. Below the blade passage frequency and at relatively high power settings jet noise is expected to be the dominant source, and comparisons are shown that demonstrate the accuracy of the jet noise model recently developed by MTCT for NASA under Contract NAS3-00178, Task Order No. 10. At lower power settings the core noise became most apparent, and these data corrected for the contribution of jet noise were then used to establish the characteristics of core noise. There is clearly more than one spectral range where core noise is evident, so the spectral approach developed by von Glahn and Krejsa in 1982 wherein four spectral regions overlap, was used in the GE effort. Further analysis indicates that the two higher frequency components, which are often somewhat masked by turbomachinery noise, can be treated as one component, and it is on that basis that the current model is formulated. The frequency scaling relationships are improved and are now based on combustor and core nozzle geometries. In conjunction with the Task

  6. High-temperature, high-pressure bonding of nested tubular metallic components

    International Nuclear Information System (INIS)

    Quinby, T.C.

    1980-01-01

    This invention is a tool for effecting high-temperature, high compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators, the target assembly comprising a uranium foil and an aluminum-alloy substrate. The tool preferably is composed throughout of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus with the member. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend respectively into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hotpress evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity

  7. High-temperature, high-pressure bonding of nested tubular metallic components

    Science.gov (United States)

    Quinby, T.C.

    A tool is described for effecting high-temperature, high-compression bonding between the confronting faces of nested, tubular, metallic components. In a typical application, the tool is used to produce tubular target assemblies for irradiation in nuclear reactors or particle accelerators. The target assembly comprising a uranum foil and an aluninum-alloy substrate. The tool is composed of graphite. It comprises a tubular restraining member in which a mechanically expandable tubular core is mounted to form an annulus. The components to be bonded are mounted in nested relation in the annulus. The expandable core is formed of individually movable, axially elongated segments whose outer faces cooperatively define a cylindrical pressing surface and whose inner faces cooperatively define two opposed, inwardly tapered, axial bores. Tapered rams extend into the bores. The loaded tool is mounted in a conventional hot-press provided with evacuation means, heaters for maintaining its interior at bonding temperature, and hydraulic cylinders for maintaining a selected inwardly directed pressure on the tapered rams. With the hot-press evacuated and the loaded tool at the desired temperature, the cylinders are actuated to apply the selected pressure to the rams. The rams in turn expand the segmented core to maintain the nested components in compression against the restraining member. These conditions are maintained until the confronting faces of the nested components are joined in a continuous, uniform bond characterized by high thermal conductivity.

  8. Introducing FACETS, the Framework Application for Core-Edge Transport Simulations

    International Nuclear Information System (INIS)

    Cary, John R.; Candy, Jeff; Cohen, Ronald H.; Krasheninnikov, Sergei I.; McCune, Douglas C.; Estep, Donald J.; Larson, Jay W.; Malony, Allen; Worley, Patrick H.; Carlsson, Johann Anders; Hakim, A.H.; Hamill, P.; Kruger, Scott E.; Muzsala, S.; Pletzer, Alexander; Shasharina, Svetlana; Wade-Stein, D.; Wang, N.; McInnes, Lois C.; Wildey, T.; Casper, T.A.; Diachin, Lori A.; Epperly, Thomas; Rognlien, T.D.; Fahey, Mark R.; Kuehn, Jeffery A.; Morris, A.; Shende, Sameer; Feibush, E.; Hammett, Gregory W.; Indireshkumar, K.; Ludescher, C.; Randerson, L.; Stotler, D.; Pigarov, A.; Bonoli, P.; Chang, C.S.; D'Ippolito, D.A.; Colella, Philip; Keyes, David E.; Bramley, R.

    2007-01-01

    The FACETS (Framework Application for Core-Edge Transport Simulations) project began in January 2007 with the goal of providing core to wall transport modeling of a tokamak fusion reactor. This involves coupling previously separate computations for the core, edge, and wall regions. Such a coupling is primarily through connection regions of lower dimensionality. The project has started developing a component-based coupling framework to bring together models for each of these regions. In the first year, the core model will be a 1 dimensional model (1D transport across flux surfaces coupled to a 2D equilibrium) with fixed equilibrium. The initial edge model will be the fluid model, UEDGE, but inclusion of kinetic models is planned for the out years. The project also has an embedded Scientific Application Partnership that is examining embedding a full-scale turbulence model for obtaining the crosssurface fluxes into a core transport code.

  9. Preliminary Uncertainty Analysis for SMART Digital Core Protection and Monitoring System

    International Nuclear Information System (INIS)

    Koo, Bon Seung; In, Wang Kee; Hwang, Dae Hyun

    2012-01-01

    The Korea Atomic Energy Research Institute (KAERI) developed on-line digital core protection and monitoring systems, called SCOPS and SCOMS as a part of SMART plant protection and monitoring system. SCOPS simplified the protection system by directly connecting the four RSPT signals to each core protection channel and eliminated the control element assembly calculator (CEAC) hardware. SCOMS adopted DPCM3D method in synthesizing core power distribution instead of Fourier expansion method being used in conventional PWRs. The DPCM3D method produces a synthetic 3-D power distribution by coupling a neutronics code and measured in-core detector signals. The overall uncertainty analysis methodology which is used statistically combining uncertainty components of SMART core protection and monitoring system was developed. In this paper, preliminary overall uncertainty factors for SCOPS/SCOMS of SMART initial core were evaluated by applying newly developed uncertainty analysis method

  10. CFD to modeling molten core behavior simultaneously with chemical phenomena

    International Nuclear Information System (INIS)

    Vladimir V Chudanov; Anna E Aksenova; Valerii A Pervichko

    2005-01-01

    Full text of publication follows: This paper deals with the basic features of a computing procedure, which can be used for modeling of destruction and melting of a core with subsequent corium retaining into the reactor vessel. The destruction and melting of core mean the account of the following phenomena: a melting, draining (moving of the melt through a porous layer of core debris), freezing with release of an energy, change of geometry, formation of the molten pool, whose convective intermixing and distribution influence on a mechanism of borders destruction. It is necessary to take into account that during of heating molten pool and development in it of convective fluxes a stratification of a multi-component melt on two layers of metal light and of oxide heavy components is observed. These layers are in interaction, they can exchange by the separate components as result of diffusion or oxidizing reactions. It can have an effect considerably on compositions, on a specific weight, and on properties of molten interacting phases, and on a structure of the molten stratified pool. In turn, the retaining of the formed molten masses in reactor vessel requires the solution of a matched heat exchange problem, namely, of a natural convection in a heat generating fluid in partially or completely molten corium and of heat exchange problem with taking into account of a melting of the reactor vessel. In addition, it is necessary to take into account phase segregation, caused by influence of local and of global natural convective flows and thermal lag of heated up boundaries. The mathematical model for simulation of the specified phenomena is based on the Navier-Stokes equations with variable properties together with the heat transfer equation. For modeling of a corium moving through a porous layer of core debris, the special computing algorithm to take into account density jump on interface between a melt and a porous layer of core debris is designed. The model was

  11. The caCORE Software Development Kit: Streamlining construction of interoperable biomedical information services

    Directory of Open Access Journals (Sweden)

    Warzel Denise

    2006-01-01

    Full Text Available Abstract Background Robust, programmatically accessible biomedical information services that syntactically and semantically interoperate with other resources are challenging to construct. Such systems require the adoption of common information models, data representations and terminology standards as well as documented application programming interfaces (APIs. The National Cancer Institute (NCI developed the cancer common ontologic representation environment (caCORE to provide the infrastructure necessary to achieve interoperability across the systems it develops or sponsors. The caCORE Software Development Kit (SDK was designed to provide developers both within and outside the NCI with the tools needed to construct such interoperable software systems. Results The caCORE SDK requires a Unified Modeling Language (UML tool to begin the development workflow with the construction of a domain information model in the form of a UML Class Diagram. Models are annotated with concepts and definitions from a description logic terminology source using the Semantic Connector component. The annotated model is registered in the Cancer Data Standards Repository (caDSR using the UML Loader component. System software is automatically generated using the Codegen component, which produces middleware that runs on an application server. The caCORE SDK was initially tested and validated using a seven-class UML model, and has been used to generate the caCORE production system, which includes models with dozens of classes. The deployed system supports access through object-oriented APIs with consistent syntax for retrieval of any type of data object across all classes in the original UML model. The caCORE SDK is currently being used by several development teams, including by participants in the cancer biomedical informatics grid (caBIG program, to create compatible data services. caBIG compatibility standards are based upon caCORE resources, and thus the caCORE SDK has

  12. A simple model for induction core voltage distributions

    International Nuclear Information System (INIS)

    Briggs, Richard J.; Fawley, William M.

    2004-01-01

    In fall 2003 T. Hughes of MRC used a full EM simulation code (LSP) to show that the electric field stress distribution near the outer radius of the longitudinal gaps between the four Metglas induction cores is very nonuniform in the original design of the DARHT-2 accelerator cells. In this note we derive a simple model of the electric field distribution in the induction core region to provide physical insights into this result. The starting point in formulating our model is to recognize that the electromagnetic fields in the induction core region of the DARHT-2 accelerator cells should be accurately represented within a quasi-static approximation because the timescale for the fields to change is much longer than the EM wave propagation time. The difficulty one faces is the fact that the electric field is a mixture of both a ''quasi-magnetostatic field'' (having a nonzero curl, with Bdot the source) and a ''quasi-electrostatic field'' (the source being electric charges on the various metal surfaces). We first discuss the EM field structure on the ''micro-scale'' of individual tape windings in Section 2. The insights from that discussion are then used to formulate a ''macroscopic'' description of the fields inside an ''equivalent homogeneous tape wound core region'' in Section 3. This formulation explicitly separates the nonlinear core magnetics from the quasi-electrostatic components of the electric field. In Section 4 a physical interpretation of the radial dependence of the electrostatic component of the electric field derived from this model is presented in terms of distributed capacitances, and the voltage distribution from gap to gap is related to various ''equivalent'' lumped capacitances. Analytic solutions of several simple multi-core cases are presented in Sections 5 and 6 to help provide physical insight into the effect of various proposed changes in the geometrical parameters of the DARHT-2 accelerator cell. Our results show that over most of the gap

  13. Simulation Based Optimization of Complex Monolithic Composite Structures Using Cellular Core Technology

    Science.gov (United States)

    Hickmott, Curtis W.

    Cellular core tooling is a new technology which has the capability to manufacture complex integrated monolithic composite structures. This novel tooling method utilizes thermoplastic cellular cores as inner tooling. The semi-rigid nature of the cellular cores makes them convenient for lay-up, and under autoclave temperature and pressure they soften and expand providing uniform compaction on all surfaces including internal features such as ribs and spar tubes. This process has the capability of developing fully optimized aerospace structures by reducing or eliminating assembly using fasteners or bonded joints. The technology is studied in the context of evaluating its capabilities, advantages, and limitations in developing high quality structures. The complex nature of these parts has led to development of a model using the Finite Element Analysis (FEA) software Abaqus and the plug-in COMPRO Common Component Architecture (CCA) provided by Convergent Manufacturing Technologies. This model utilizes a "virtual autoclave" technique to simulate temperature profiles, resin flow paths, and ultimately deformation from residual stress. A model has been developed simulating the temperature profile during curing of composite parts made with the cellular core technology. While modeling of composites has been performed in the past, this project will look to take this existing knowledge and apply it to this new manufacturing method capable of building more complex parts and develop a model designed specifically for building large, complex components with a high degree of accuracy. The model development has been carried out in conjunction with experimental validation. A double box beam structure was chosen for analysis to determine the effects of the technology on internal ribs and joints. Double box beams were manufactured and sectioned into T-joints for characterization. Mechanical behavior of T-joints was performed using the T-joint pull-off test and compared to traditional

  14. Contribution of CRUST2.0 components to the tri-axiality of the Earth and equatorial flattening of the core

    Directory of Open Access Journals (Sweden)

    Sun Rong

    2013-08-01

    Full Text Available Equatorial flattening of the core were previously estimated to be 5 × 10−4 by using seismically derived density anomaly, and 1.7748280 × 10−5 by assuming that the ratio of polar flattening to equatorial flattening of the core is the same as that of the whole Earth. In this study, we attempted to explain the difference by applying a density-contrast stripping process to the crust in the second method. We use the CRUST2.0 model to estimate the inertia-moment contribution resulted from the density-contrast structure in the crust to a tri-axial Earth. The contribution of the density contrast in the crust was removed layer by layer. The layers include topography, bathymetry, ice, soft sediment, hard sediment, upper crust, middle crust, lower crust and the reference crust. For the boundaries of the topography and bathymetry layers, we used ETOPO5 values with a resolution of 5'. For boundaries of other layers, we used values from the CRUST2.0 model with a resolution of 2°. After the contribution of density contrast is stripped, the equatorial flattening of the core was found to be 6.544 × 10−5, which is still one order of magnitude smaller than the result given by the first method. This suggests that at least one of the methods is not correct. The influence of the uncertainty in the equatorial flattening of the core on the Free Core Nutation frequency is small, but its effect on the gravitational torque acting on the tri-axial inner core cannot be ignored. So an accurate determination of the equatorial flattening of the core is still necessary.

  15. Spring unit especially intended for a nuclear reactor core

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, Wilhelm.

    1977-01-01

    This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

  16. HTGR nuclear heat source component design and experience

    International Nuclear Information System (INIS)

    Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

    1982-05-01

    The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

  17. Core-shell polymer nanorods by a two-step template wetting process

    International Nuclear Information System (INIS)

    Dougherty, S; Liang, J

    2009-01-01

    One-dimensional core-shell polymer nanowires offer many advantages and great potential for many different applications. In this paper we introduce a highly versatile two-step template wetting process to fabricate two-component core-shell polymer nanowires with controllable shell thickness. PLLA and PMMA were chosen as model polymers to demonstrate the feasibility of this process. Solution wetting with different concentrations of polymer solutions was used to fabricate the shell layer and melt wetting was used to fill the shell with the core polymer. The shell thickness was analyzed as a function of the polymer solution concentration and viscosity, and the core-shell morphology was observed with TEM. This paper demonstrates the feasibility of fabricating polymer core-shell nanostructures using our two-step template wetting process and opens the arena for optimization and future experiments with polymers that are desirable for specific applications.

  18. Component studies of psychological treatments of adult depression: A systematic review and meta-analysis.

    NARCIS (Netherlands)

    Cuijpers, P.; Cristea, I.A.; Karyotaki, E.; Reijnders, M.; Hollon, S.D.

    2017-01-01

    Objectives: A recent report from the US Institute of Medicine indicated that identifying core elements of psychosocial interventions is a key step in successfully bringing evidence-based psychosocial interventions into clinical practice. Component studies have the best design to examine these core

  19. Final Report for the "Fusion Application for Core-Edge Transport Simulations (FACETS)"

    Energy Technology Data Exchange (ETDEWEB)

    Cary, John R; Kruger, Scott

    2014-10-02

    The FACETS project over its lifetime developed the first self-consistent core-edge coupled capabilities, a new transport solver for modeling core transport in tokamak cores, developed a new code for modeling wall physics over long time scales, and significantly improved the capabilities and performance of legacy components, UEDGE, NUBEAM, GLF23, GYRO, and BOUT++. These improved capabilities leveraged the team’s expertise in applied mathematics (solvers and algorithms) and computer science (performance improvements and language interoperability). The project pioneered new methods for tackling the complexity of simulating the concomitant complexity of tokamak experiments.

  20. Present Status of HTGR Utilization System Development in Japan

    International Nuclear Information System (INIS)

    Miyamoto, Yoshiaki

    2000-01-01

    Efforts are to be continuously devoted to establish and upgrade HTGR technology in the world. Japan Atomic Energy Research Institute (JAERI) has conducted the R and D of HTGRs since the 1960's in Japan, focusing on mainly the construction of High Temperature engineering Test Reactor (HTTR) which is an HTGR with a maximum helium gas temperature of 950 o C at the reactor outlet and HTGR utilization systems. The HTTR achieved first criticality on November 10, 1998 and will restart from January in 2001. In the R and D program of HTGR utilization systems, JAERI has conducted hydrogen production systems with HTGR to demonstrate the applicability of nuclear heat for extensive energy demands besides the electric power generation. JAERI has developed a hydrogen production system by steam reforming process of natural gas using nuclear heat supplied from the HTTR. Prior to the demonstration test of HTTR hydrogen production system, a 1/30-scale out-of-pile test facility is under construction for safety review and detailed design of the system. The out-of-pile test facility will be started in 2001 and will be continued about 4 years. The hydrogen permeation and corrosion tests have been carried out since 1997. Check and review for the demonstration program in the HTTR hydrogen production system will be made in 2001. Then the HTTR hydrogen production system is scheduled to be constructed from 2003 and demonstratively operated from around 2006. In parallel with the R and D of the HTTR hydrogen production system, hydrogen production method by thermochemical water splitting, so-called IS process, has been studied in JAERI. The IS process is placed as one of future candidates of the heat utilization systems of the HTTR following the steam reforming system. Continuous and stoichiometric production of hydrogen and oxygen for 48 hours was successfully achieved with a laboratory-scale apparatus mainly made of glass. Following this achievement, the study has been continued with a larger

  1. Core Hunter 3: flexible core subset selection.

    Science.gov (United States)

    De Beukelaer, Herman; Davenport, Guy F; Fack, Veerle

    2018-05-31

    Core collections provide genebank curators and plant breeders a way to reduce size of their collections and populations, while minimizing impact on genetic diversity and allele frequency. Many methods have been proposed to generate core collections, often using distance metrics to quantify the similarity of two accessions, based on genetic marker data or phenotypic traits. Core Hunter is a multi-purpose core subset selection tool that uses local search algorithms to generate subsets relying on one or more metrics, including several distance metrics and allelic richness. In version 3 of Core Hunter (CH3) we have incorporated two new, improved methods for summarizing distances to quantify diversity or representativeness of the core collection. A comparison of CH3 and Core Hunter 2 (CH2) showed that these new metrics can be effectively optimized with less complex algorithms, as compared to those used in CH2. CH3 is more effective at maximizing the improved diversity metric than CH2, still ensures a high average and minimum distance, and is faster for large datasets. Using CH3, a simple stochastic hill-climber is able to find highly diverse core collections, and the more advanced parallel tempering algorithm further increases the quality of the core and further reduces variability across independent samples. We also evaluate the ability of CH3 to simultaneously maximize diversity, and either representativeness or allelic richness, and compare the results with those of the GDOpt and SimEli methods. CH3 can sample equally representative cores as GDOpt, which was specifically designed for this purpose, and is able to construct cores that are simultaneously more diverse, and either are more representative or have higher allelic richness, than those obtained by SimEli. In version 3, Core Hunter has been updated to include two new core subset selection metrics that construct cores for representativeness or diversity, with improved performance. It combines and outperforms the

  2. Gravitational torque on the inner core and decadal polar motion

    Science.gov (United States)

    Dumberry, Mathieu

    2008-03-01

    A decadal polar motion with an amplitude of approximately 25 milliarcsecs (mas) is observed over the last century, a motion known as the Markowitz wobble. The origin of this motion remains unknown. In this paper, we investigate the possibility that a time-dependent axial misalignment between the density structures of the inner core and mantle can explain this signal. The longitudinal displacement of the inner core density structure leads to a change in the global moment of inertia of the Earth. In addition, as a result of the density misalignment, a gravitational equatorial torque leads to a tilt of the oblate geometric figure of the inner core, causing a further change in the global moment of inertia. To conserve angular momentum, an adjustment of the rotation vector must occur, leading to a polar motion. We develop theoretical expressions for the change in the moment of inertia and the gravitational torque in terms of the angle of longitudinal misalignment and the density structure of the mantle. A model to compute the polar motion in response to time-dependent axial inner core rotations is also presented. We show that the polar motion produced by this mechanism can be polarized about a longitudinal axis and is expected to have decadal periodicities, two general characteristics of the Markowitz wobble. The amplitude of the polar motion depends primarily on the Y12 spherical harmonic component of mantle density, on the longitudinal misalignment between the inner core and mantle, and on the bulk viscosity of the inner core. We establish constraints on the first two of these quantities from considerations of the axial component of this gravitational torque and from observed changes in length of day. These constraints suggest that the maximum polar motion from this mechanism is smaller than 1 mas, and too small to explain the Markowitz wobble.

  3. Historical Variations in Inner Core Rotation and Polar Motion at Decade Timescales

    Science.gov (United States)

    Dumberry, M.

    2005-12-01

    Exchanges of angular momentum between the mantle, the fluid core and the solid inner core result in changes in the Earth's rotation. Torques in the axial direction produce changes in amplitude, or changes in length of day, while torques in the equatorial direction lead to changes in orientation of the rotation vector with respect to the mantle, or polar motion. In this work, we explore the possibility that a combination of electromagnetic and gravitational torques on the inner core can reproduce the observed decadal variations in polar motion known as the Markowitz wobble. Torsional oscillations, which involve azimuthal motions in the fluid core with typical periods of decades, entrain the inner core by electromagnetic traction. When the inner core is axially rotated, its surfaces of constant density are no longer aligned with the gravitational potential from mantle density heterogeneities, and this results in a gravitational torque between the two. The axial component of this torque has been previously described and is believed to be partly responsible for decadal changes in length of day. In this work, we show that it has also an equatorial component, which produces a tilt of the inner core and results in polar motion. The polar motion produced by this mechanism depends on the density structure in the mantle, the rheology of the inner core, and the time-history of the angle of axial misalignment between the inner core and the mantle. We reconstruct the latter using a model of torsional oscillations derived from geomagnetic secular variation. From this time-history, and by using published models of mantle density structure, we show that we can reproduce the salient characteristics of the Markowitz wobble: an eccentric decadal polar motion of 30-50 milliarcsecs oriented along a specific longitude. We discuss the implications of this result, noting that a match in both amplitude and phase of the observed Markowitz wobble allows the recovery of the historical

  4. Manufacture of a four-sheet complex component from different titanium alloys by superplastic forming

    Science.gov (United States)

    Allazadeh, M. R.; Zuelli, N.

    2017-10-01

    A superplastic forming (SPF) technology process was deployed to form a complex component with eight-pocket from a four-sheet sandwich panel sheetstock. Six sheetstock packs were composed of two core sheets made of Ti-6Al-4V or Ti-5Al-4Cr-4Mo-2Sn-2Zr titanium alloy and two skin sheets made of Ti-6Al-4V or Ti-6Al-2Sn-4Zr-2Mo titanium alloy in three different combinations. The sheets were welded with two subsequent welding patterns over the core and skin sheets to meet the required component's details. The applied welding methods were intermittent and continuous resistance seam welding for bonding the core sheets to each other and the skin sheets over the core panel, respectively. The final component configuration was predicted based on the die drawings and finite element method (FEM) simulations for the sandwich panels. An SPF system set-up with two inlet gas pipe feeding facilitated the trials to deliver two pressure-time load cycles acting simultaneously which were extracted from FEM analysis for specific forming temperature and strain rate. The SPF pressure-time cycles were optimized via GOM scanning and visually inspecting some sections of the packs in order to assess the levels of core panel formation during the inflation process of the sheetstock. Two sets of GOM scan results were compared via GOM software to inspect the surface and internal features of the inflated multisheet packs. The results highlighted the capability of the tested SPF process to form complex components from a flat multisheet pack made of different titanium alloys.

  5. Deterministic Modeling of the High Temperature Test Reactor

    International Nuclear Information System (INIS)

    Ortensi, J.; Cogliati, J.J.; Pope, M.A.; Ferrer, R.M.; Ougouag, A.M.

    2010-01-01

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL's current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green's Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2-3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control

  6. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

    Directory of Open Access Journals (Sweden)

    Saas Laurent

    2017-01-01

    Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…. During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.

  7. IEA-R1 reactor core simulation with RELAP5 code

    International Nuclear Information System (INIS)

    Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

    2005-01-01

    This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

  8. Assessing Fidelity of Core Components in a Mindfulness and Yoga Intervention for Urban Youth: Applying the CORE Process

    Science.gov (United States)

    Gould, Laura Feagans; Mendelson, Tamar; Dariotis, Jacinda K.; Ancona, Matthew; Smith, Ali S. R.; Gonzalez, Andres A.; Smith, Atman A.; Greenberg, Mark T.

    2014-01-01

    In the past years, the number of mindfulness-based intervention and prevention programs has increased steadily. In order to achieve the intended program outcomes, program implementers need to understand the essential and indispensable components that define a program's success. This chapter describes the complex process of identifying the core…

  9. Noise variation by compressive stress on the model core of power transformers

    Energy Technology Data Exchange (ETDEWEB)

    Mizokami, Masato, E-mail: mizokami.g76.masato@jp.nssmc.com; Kurosaki, Yousuke

    2015-05-01

    The reduction of audible noise generated by cores for power transformers has been required due to environmental concern. It is known that compressive stress in the rolling direction of electrical steel affects magnetostriction and it can result in an increase in noise level. In this research, the effect of compressive stress to noise was investigated on a 3-phase 3-limb model core. Compressive stress was applied in the rolling direction of the limbs from the outside of the core. It increased the sound pressure levels and the slope of the rise was about 2 dBA/MPa. Magnetostriction on single sheet samples was also measured under compressive stress and the harmonic components of the magnetostriction were compared with those of noise. It revealed that the variation in magnetostriction with compressive stress did not entirely correspond to that in noise. In one of the experiments, localized bending happened on one limb during compressing the core. While deformation of the core had not been intended, the noise was measured. The deformation increased the noise by more than 10 dBA and it occurred on most of the harmonic components. - Highlights: • Audible noise was measured on a model core to which compressive stress was applied. • The stress in the rolling direction of the steel causes a rise in noise level. • The slope of the rise in sound pressure level up to 2.5 MPa is about 2 dBA/MPa. • Variation in magnetostriction by stress does not entirely agree with that in noise. • Bend arisen in the core causes an extreme increase in noise.

  10. Noise variation by compressive stress on the model core of power transformers

    International Nuclear Information System (INIS)

    Mizokami, Masato; Kurosaki, Yousuke

    2015-01-01

    The reduction of audible noise generated by cores for power transformers has been required due to environmental concern. It is known that compressive stress in the rolling direction of electrical steel affects magnetostriction and it can result in an increase in noise level. In this research, the effect of compressive stress to noise was investigated on a 3-phase 3-limb model core. Compressive stress was applied in the rolling direction of the limbs from the outside of the core. It increased the sound pressure levels and the slope of the rise was about 2 dBA/MPa. Magnetostriction on single sheet samples was also measured under compressive stress and the harmonic components of the magnetostriction were compared with those of noise. It revealed that the variation in magnetostriction with compressive stress did not entirely correspond to that in noise. In one of the experiments, localized bending happened on one limb during compressing the core. While deformation of the core had not been intended, the noise was measured. The deformation increased the noise by more than 10 dBA and it occurred on most of the harmonic components. - Highlights: • Audible noise was measured on a model core to which compressive stress was applied. • The stress in the rolling direction of the steel causes a rise in noise level. • The slope of the rise in sound pressure level up to 2.5 MPa is about 2 dBA/MPa. • Variation in magnetostriction by stress does not entirely agree with that in noise. • Bend arisen in the core causes an extreme increase in noise

  11. Meterwavelength Single-pulse Polarimetric Emission Survey. IV. The Period Dependence of Component Widths of Pulsars

    Science.gov (United States)

    Skrzypczak, Anna; Basu, Rahul; Mitra, Dipanjan; Melikidze, George I.; Maciesiak, Krzysztof; Koralewska, Olga; Filothodoros, Alexandros

    2018-02-01

    The core component width in normal pulsars, with periods (P) > 0.1 s, measured at the half-power point at 1 GHz, has a lower boundary line (LBL) that closely follows the P ‑0.5 scaling relation. This result is of fundamental importance for understanding the emission process and requires extended studies over a wider frequency range. In this paper we have carried out a detailed study of the profile component widths of 123 normal pulsars observed in the Meterwavelength Single-pulse Polarimetric Emission Survey at 333 and 618 MHz. The components in the pulse profile were separated into core and conal classes. We found that at both frequencies, the core, as well as the conal component widths versus period, had a LBL that followed the P ‑0.5 relation with a similar lower boundary. The radio emission in normal pulsars has been observationally shown to arise from a narrow range of heights around a few hundred kilometers above the stellar surface. In the past the P ‑0.5 relation has been considered as evidence for emission arising from last open dipolar magnetic field lines. We show that the P ‑0.5 dependence only holds if the trailing and leading half-power points of the component are associated with the last open field line. In such a scenario we do not find any physical motivation that can explain the P ‑0.5 dependence for both core and conal components as evidence for dipolar geometry in normal pulsars. We believe the period dependence is a result of a currently unexplained physical phenomenon.

  12. Magnetohydrodynamic Convection in the Outer Core and its Geodynamic Consequences

    Science.gov (United States)

    Kuang, Weijia; Chao, Benjamin F.; Fang, Ming

    2004-01-01

    The Earth's fluid outer core is in vigorous convection through much of the Earth's history. In addition to generating and maintaining Earth s time-varying magnetic field (geodynamo), the core convection also generates mass redistribution in the core and a dynamical pressure field on the core-mantle boundary (CMB). All these shall result in various core-mantle interactions, and contribute to surface geodynamic observables. For example, electromagnetic core-mantle coupling arises from finite electrically conducting lower mantle; gravitational interaction occurs between the cores and the heterogeneous mantle; mechanical coupling may also occur when the CMB topography is aspherical. Besides changing the mantle rotation via the coupling torques, the mass-redistribution in the core shall produce a spatial-temporal gravity anomaly. Numerical modeling of the core dynamical processes contributes in several geophysical disciplines. It helps explain the physical causes of surface geodynamic observables via space geodetic techniques and other means, e.g. Earth's rotation variation on decadal time scales, and secular time-variable gravity. Conversely, identification of the sources of the observables can provide additional insights on the dynamics of the fluid core, leading to better constraints on the physics in the numerical modeling. In the past few years, our core dynamics modeling efforts, with respect to our MoSST model, have made significant progress in understanding individual geophysical consequences. However, integrated studies are desirable, not only because of more mature numerical core dynamics models, but also because of inter-correlation among the geophysical phenomena, e.g. mass redistribution in the outer core produces not only time-variable gravity, but also gravitational core-mantle coupling and thus the Earth's rotation variation. They are expected to further facilitate multidisciplinary studies of core dynamics and interactions of the core with other

  13. Development of Special Tools for the Straightness Measurement of JRTR Core Inner Shell

    International Nuclear Information System (INIS)

    Sinjlawi, Abdullah; Cho, Yeong-Garp; Chung, Jong-Ha

    2014-01-01

    Jordan Research and Training Reactor (JRTR) is an open pool type nuclear research reactor, 5 MW power, JRTR core made from Zircaloy. The JRTR will be used for nuclear applications such as isotopes production, nuclear researches, neutron transmutation doping (NTD), and training. JRTR core structures will be exposed to a large amount of neutron irradiation during the life time of the reactor. The core inner shell also will be exposed to a pressure that comes from heavy water system. JRTR core inner shell will deform due to the neutron irradiation and the mechanical stress. Therefore, the dimensional change of the core inner shell should be periodically (every 10 years) measured as an in-service inspection to confirm the structural integrity. As a result of neutron irradiation, pressure difference of the heavy water vessel, and the mechanical stress, the reactor core will deform as shown in figure 2 to figure 4. The maximum deformation to the normal direction of inner shell wall is 0.75 mm as shown in figure 3. This study discusses development of special tools that will be used for pre-service and in-service inspection of JRTR inner shell. The performance and procedure for the measurements tools will be verified using by the real inner shell of the heavy water vessel at factory before shipping to Jordan.. There will be very delicate working procedure for the measurement in the limited space in JRTR core. Therefore, we will develop the detail procedures to cover the removal of the core components, installation of the measurement tools, measurement, and re-installation of the core components. The measurement of the inner shell at JAEC site during commissioning stage will be the first remote measurement at the same conditions of pool water and heavy water system

  14. Components of Effective Cognitive-Behavioral Therapy for Pediatric Headache: A Mixed Methods Approach.

    Science.gov (United States)

    Law, Emily F; Beals-Erickson, Sarah E; Fisher, Emma; Lang, Emily A; Palermo, Tonya M

    2017-01-01

    Internet-delivered treatment has the potential to expand access to evidence-based cognitive-behavioral therapy (CBT) for pediatric headache, and has demonstrated efficacy in small trials for some youth with headache. We used a mixed methods approach to identify effective components of CBT for this population. In Study 1, component profile analysis identified common interventions delivered in published RCTs of effective CBT protocols for pediatric headache delivered face-to-face or via the Internet. We identified a core set of three treatment components that were common across face-to-face and Internet protocols: 1) headache education, 2) relaxation training, and 3) cognitive interventions. Biofeedback was identified as an additional core treatment component delivered in face-to-face protocols only. In Study 2, we conducted qualitative interviews to describe the perspectives of youth with headache and their parents on successful components of an Internet CBT intervention. Eleven themes emerged from the qualitative data analysis, which broadly focused on patient experiences using the treatment components and suggestions for new treatment components. In the Discussion, these mixed methods findings are integrated to inform the adaptation of an Internet CBT protocol for youth with headache.

  15. Training in Vocational Assessment: Preparing Rehabilitation Counselors and Meeting the Requirements of the CORE Standards

    Science.gov (United States)

    Tansey, Timothy N.

    2008-01-01

    Assessment represents a foundational component of rehabilitation counseling services. The revised Council on Rehabilitation Education (CORE) standards implemented in 2004 resulted in the redesign of the knowledge and outcomes under the Assessment standard. The author reviews the current CORE standard for training in assessment within the context…

  16. Impact of the dynamical core on the direct simulation of tropical cyclones in a high-resolution global model

    International Nuclear Information System (INIS)

    Reed, K. A.

    2015-01-01

    Our paper examines the impact of the dynamical core on the simulation of tropical cyclone (TC) frequency, distribution, and intensity. The dynamical core, the central fluid flow component of any general circulation model (GCM), is often overlooked in the analysis of a model's ability to simulate TCs compared to the impact of more commonly documented components (e.g., physical parameterizations). The Community Atmosphere Model version 5 is configured with multiple dynamics packages. This analysis demonstrates that the dynamical core has a significant impact on storm intensity and frequency, even in the presence of similar large-scale environments. In particular, the spectral element core produces stronger TCs and more hurricanes than the finite-volume core using very similar parameterization packages despite the latter having a slightly more favorable TC environment. Furthermore, these results suggest that more detailed investigations into the impact of the GCM dynamical core on TC climatology are needed to fully understand these uncertainties. Key Points The impact of the GCM dynamical core is often overlooked in TC assessments The CAM5 dynamical core has a significant impact on TC frequency and intensity A larger effort is needed to better understand this uncertainty

  17. Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

    International Nuclear Information System (INIS)

    Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

    2001-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

  18. Core-to-core uniformity improvement in multi-core fiber Bragg gratings

    Science.gov (United States)

    Lindley, Emma; Min, Seong-Sik; Leon-Saval, Sergio; Cvetojevic, Nick; Jovanovic, Nemanja; Bland-Hawthorn, Joss; Lawrence, Jon; Gris-Sanchez, Itandehui; Birks, Tim; Haynes, Roger; Haynes, Dionne

    2014-07-01

    Multi-core fiber Bragg gratings (MCFBGs) will be a valuable tool not only in communications but also various astronomical, sensing and industry applications. In this paper we address some of the technical challenges of fabricating effective multi-core gratings by simulating improvements to the writing method. These methods allow a system designed for inscribing single-core fibers to cope with MCFBG fabrication with only minor, passive changes to the writing process. Using a capillary tube that was polished on one side, the field entering the fiber was flattened which improved the coverage and uniformity of all cores.

  19. The Recent Performance of the Traditional Measure of Core Inflation in G7 Countries

    OpenAIRE

    Luciana Lo Bue; Antonio Ribba

    2009-01-01

    In this paper we undertake an empirical investigation concerning the performance of the traditional measure of core inflation in recent years. We consider the group of G7 countries and explore both the high-frequency and the low-frequency relations between overall inflation and core inflation. We find that the traditional core measure, obtained by subtracting from the overall index those components which exhibit high volatility and which are responsible for the short-run variability of inflat...

  20. Assessment of core structural materials and surveillance programme of research reactors. Report of the consultants meeting. Working material

    International Nuclear Information System (INIS)

    2009-01-01

    A series of presentations on the assessment of core structural components and materials at their facilities were given by the experts. The different issues related to degradation mechanisms were discussed. The outputs include a more thorough understanding of the specific challenges related to Research Reactors (RRs) as well as proposals for activities which could assist RR organizations in their efforts to address the issues involved. The experts recommend that research reactor operators consider implementation of surveillance programs for materials of core structural components, as part of ageing management program (TECDOC-792 and DS-412). It is recognised by experts that adequate archived structural material data is not available for many RRs. Access to this data and extension of existing material databases could help many operating organisations extend the operation of their RRs. The experts agreed that an IAEA Technical Meeting (TM) on Assessment of Core Structural Materials should be organised in December 2009 (IAEA HQ Vienna). The proposed objectives of the TM are: (i) exchange of detailed technical information on the assessment and ageing management of core structural materials, (ii) identification of materials of interest for further investigation, (iii) proposal for a new IAEA CRP on Assessment of Core Structural Materials, and (iv) identification of RRs prepared to participate in proposed CRP. Based on the response to a questionnaire prepared for the 2008 meeting of the Technical Working Group for Research Reactors, the number of engineering capital projects related to core structural components is proportionally lower than those related to,for example, I and C or electrical power systems. This implies that many operating research reactors will be operating longer using their original core structural components and justifies the assessment and evaluation programmes and activities proposed in this report. (author)

  1. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  2. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

  3. Reduction of outgas from the components of the J-PARC kicker magnet

    International Nuclear Information System (INIS)

    Kamiya, Junichiro; Ogiwara, Norio; Kinsho, Michikazu; Takayanagi, Tomohiro

    2005-01-01

    The extraction kickers in Rapid Cycling Synchrotron (RCS) of Japan Proton Accelerator Research Complex (J-PARC) are installed in the vacuum chamber against electrical discharge. Therefore, outgas from the components have large effects on the vacuum system of the accelerator. We have succeeded in reducing the outgas from the components, which are made of ferrite core and aluminum alloy, by baking them before construction of the magnet. The ferrite cores were baked at 200degC in the vacuum about 300 hours, while the components made of aluminum alloy at 150degC about 70 hours. Main outgas from both materials was known to be water by mass spectroscopy, and the pressure after baking has been decreased by two or three order of magnitude. We also report the reduction method for outgas while the magnet is stored in. (author)

  4. Challenges in design of zirconium alloy reactor components

    International Nuclear Information System (INIS)

    Kakodkar, Anil; Sinha, R.K.

    1992-01-01

    Zirconium alloy components used in core-internal assemblies of heavy water reactors have to be designed under constraints imposed by need to have minimum mass, limitations of fabrication, welding and joining techniques with this material, and unique mechanisms for degradation of the operating performance of these components. These constraints manifest as challenges for design and development when the size, shape and dimensions of the components and assemblies are unconventional or untried, or when one is aiming for maximization of service life of these components under severe operating conditions. A number of such challenges were successfully met during the development of core-internal components and assemblies of Dhruva reactor. Some of the then untried ideas which were developed and successfully implemented include use of electron beam welding, cold forming of hemispherical ends of reentrant cans, and a large variety of rolled joints of innovative designs. This experience provided the foundation for taking up and successfully completing several tasks relating to coolant channels, liquid poison channels and sparger channels for PHWRs and test sections for the in-pile loops of Dhruva reactor. For life prediction and safety assessment of coolant channels of PHWRs some analytical tools, notably, a computer code for prediction of creep limited life of coolant channels has been developed. Some of the future challenges include the development of easily replaceable coolant channels and also large diameter coolant channels for Advanced Heavy Water Reactor, and development of solutions to overcome deterioration of service life of coolant channels due to hydriding. (author). 5 refs., 13 figs., 1 tab

  5. Various reactivity effects value for assuring fast reactor core inherent safety

    International Nuclear Information System (INIS)

    Belov, S.B.; Vasilyev, B.A.

    1991-01-01

    The paper presents the results of temperature and power reactivity feedback components calculations for fast reactors with different core volume when using oxide, carbide, nitride and metal fuel. Reactor parameters change in loss of flow without scram and transient over power without scram accidents was evaluated. The importance of various reactivity feedback components in restricting the consequences of these accidents has been analyzed. (author)

  6. Integration of Magnetic Components in a Step-Up Converter for Fuel Cell

    DEFF Research Database (Denmark)

    Klimczak, Pawel; Munk-Nielsen, Stig

    2009-01-01

    converter is a critical part. The input voltage of the converter decreases while the output power increases. It creates challenges in design of the converter's magnetic components. Scope of this paper is integration of the dc inductor and the transformer on a single core. Such integration improve...... utilization of the core and windings. It leads to size reduction of the converter....

  7. Toxicological Assessment of a Lignin Core Nanoparticle Doped with Silver as an Alternative to Conventional Silver Core Nanoparticles

    Directory of Open Access Journals (Sweden)

    Cassandra E. Nix

    2018-05-01

    Full Text Available Elevated levels of silver in the environment are anticipated with an increase in silver nanoparticle (AgNP production and use in consumer products. To potentially reduce the burden of silver ion release from conventional solid core AgNPs, a lignin-core particle doped with silver ions and surface-stabilized with a polycationic electrolyte layer was engineered. Our objective was to determine whether any of the formulation components elicit toxicological responses using embryonic zebrafish. Ionic silver and free surface stabilizer were the most toxic constituents, although when associated separately or together with the lignin core particles, the toxicity of the formulations decreased significantly. The overall toxicity of lignin formulations containing silver was similar to other studies on a silver mass basis, and led to a significantly higher prevalence of uninflated swim bladder and yolk sac edema. Comparative analysis of dialyzed samples which had leached their loosely bound Ag+, showed a significant increase in mortality immediately after dialysis, in addition to eliciting significant increases in types of sublethal responses relative to the freshly prepared non-dialyzed samples. ICP-OES/MS analysis indicated that silver ion release from the particle into solution was continuous, and the rate of release differed when the surface stabilizer was not present. Overall, our study indicates that the lignin core is an effective alternative to conventional solid core AgNPs for potentially reducing the burden of silver released into the environment from a variety of consumer products.

  8. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    1977-09-01

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

  9. OS and Runtime Support for Efficiently Managing Cores in Parallel Applications

    OpenAIRE

    Klues, Kevin Alan

    2015-01-01

    Parallel applications can benefit from the ability to explicitly control their thread scheduling policies in user-space. However, modern operating systems lack the interfaces necessary to make this type of “user-level” scheduling efficient. The key component missing is the ability for applications to gain direct access to cores and keep control of those cores even when making I/O operations that traditionally block in the kernel. A number of former systems provided limited support for these c...

  10. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  11. Sulfur in Earth's Mantle and Its Behavior During Core Formation

    Science.gov (United States)

    Chabot, Nancy L.; Righter,Kevin

    2006-01-01

    The density of Earth's outer core requires that about 5-10% of the outer core be composed of elements lighter than Fe-Ni; proposed choices for the "light element" component of Earth's core include H, C, O, Si, S, and combinations of these elements [e.g. 1]. Though samples of Earth's core are not available, mantle samples contain elemental signatures left behind from the formation of Earth's core. The abundances of siderophile (metal-loving) elements in Earth's mantle have been used to gain insight into the early accretion and differentiation history of Earth, the process by which the core and mantle formed, and the composition of the core [e.g. 2-4]. Similarly, the abundance of potential light elements in Earth's mantle could also provide constraints on Earth's evolution and core composition. The S abundance in Earth's mantle is 250 ( 50) ppm [5]. It has been suggested that 250 ppm S is too high to be due to equilibrium core formation in a high pressure, high temperature magma ocean on early Earth and that the addition of S to the mantle from the subsequent accretion of a late veneer is consequently required [6]. However, this earlier work of Li and Agee [6] did not parameterize the metalsilicate partitioning behavior of S as a function of thermodynamic variables, limiting the different pressure and temperature conditions during core formation that could be explored. Here, the question of explaining the mantle abundance of S is revisited, through parameterizing existing metal-silicate partitioning data for S and applying the parameterization to core formation in Earth.

  12. Results and future plans for the innovative basic research on high temperature engineering

    International Nuclear Information System (INIS)

    2001-05-01

    The High Temperature Engineering Test Reactor (HTTR) is under the rise-to-power stage at the Oarai Research Establishment of Japan Atomic Energy Research Institute (JAERI). This reactor is aimed not only at establishment of the infrastructural technology on high temperature gas-cooled reactor and its upgrading, but also at promotion of the innovative basic research on high temperature engineering. The research is a series of innovative high-temperature irradiation studies, making the best use of the characteristic of the HTTR that it provides a very wide irradiation space at high temperatures. The JAERI has been conducting preliminary tests of the innovative research since 1994, in collaboration with universities and other research institutes, in the fields of 1) new materials development, 2) high temperature radiation chemistry and fusion-related research, and 3) high temperature irradiation techniques and other nuclear research. The HTTR Utilization Research Committee has been examining the results and methodology of the preliminary tests and the future plans, as well as examining the preparatory arrangements of facilities for the HTTR irradiation and post-irradiation examinations. This report presents a summary of results of the preliminary tests and preparatory arrangements for about seven years, together with an outline of the future plans. (author)

  13. Coupling between core and cladding modes in a helical core fiber with large core offset

    International Nuclear Information System (INIS)

    Napiorkowski, Maciej; Urbanczyk, Waclaw

    2016-01-01

    We analyzed the effect of resonant coupling between core and cladding modes in a helical core fiber with large core offset using the fully vectorial method based on the transformation optics formalism. Our study revealed that the resonant couplings to lower order cladding modes predicted by perturbative methods and observed experimentally in fibers with small core offsets are in fact prohibited for larger core offsets. This effect is related to the lack of phase matching caused by elongation of the optical path of the fundamental modes in the helical core. Moreover, strong couplings to the cladding modes of the azimuthal modal number much higher than predicted by perturbative methods may be observed for large core offsets, as the core offset introduces higher order angular harmonics in the field distribution of the fundamental modes. Finally, in contrast to previous studies, we demonstrate the existence of spectrally broad polarization sensitive couplings to the cladding modes suggesting that helical core fibers with large core offsets may be used as broadband circular polarizers. (paper)

  14. The Promises and Challenges of Teaching from an Intersectional Perspective: Core Components and Applied Strategies

    Science.gov (United States)

    Jones, Susan R.; Wijeyesinghe, Charmaine L.

    2011-01-01

    This chapter explores how the framework of intersectionality can be used by faculty in course development and classroom teaching. An overview of intersectionality, highlighting core assumptions and tenets of the framework, is presented first. These assumptions and tenets are then applied to classroom dynamics and the practice of teaching in…

  15. Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

    International Nuclear Information System (INIS)

    S, Tukiran; MS, Tagor; P, Surian

    2003-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

  16. Fire resistance of extruded hollow-core slabs

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl; Giuliani, Luisa; Sørensen, Lars Schiøtt

    2016-01-01

    Prefabricated extruded hollow-core slabs are preferred building components for floor structures in several countries. It is therefore important to be able to document the fire resistance of these slabs proving fulfilment of standard fire resistance requirements of 60- and 120 minutes found in most...... a standard fire test and from a thorough examination of the comprehensive test documentation available on fire exposed hollow-core slabs. Mechanisms for loss of load-bearing capacity are clarified, and evidence of the fire resistance is found. For the first time the mechanisms responsible for loss of load......-bearing capacity are identified and test results and calculation approach are for the first time Applied in accordance with each other for assessment of fire resistance of the structure....

  17. Applications of simulation experiments in LMFBR core materials technology

    International Nuclear Information System (INIS)

    Appleby, W.K.

    1976-01-01

    The development of charged particle bombardment experiments to simulate neutron irradiation induced swelling in austenitic alloys is briefly described. The applications of these techniques in LMFBR core materials technology are discussed. It is shown that use of the techniques to study the behavior of cold-worked Type-316 was instrumental in demonstrating at an early date the need for advanced materials. The simulation techniques then were used to identify alloying elements which can markedly decrease swelling and thus a focused reactor irradiation program is now in place to allow the future use of a lower swelling alloy for LMFBR core components

  18. Do protostellar fountains shape the regional core mass function?

    International Nuclear Information System (INIS)

    Li Jin-Zeng; Huang Ya-Fang; Carlos Mallamaci Claudio; César Podestà Ricardo; Actis Vicente Eloy; Maria Pacheco Ana

    2013-01-01

    The emerging massive binary system associated with AFGL 961 signifies the latest generation of massive star and cluster formation in the Rosette Molecular Complex. We present the detection of a compact cluster of dusty cores toward the AFGL 961 region based on continuum imaging at 1.3 mm by the Submillimeter Array. The binary components of AFGL 961 are associated with the most intensive millimeter emission cores or envelopes, confirming that they are indeed in an early stage of evolution. The other massive cores, however, are found to congregate in the close vicinity of the central high-mass protostellar binary. They have no apparent infrared counterparts and are, in particular, well aligned transverse to the bipolar molecular outflows originating from AFGL 961. This provides evidence for a likely triggered origin of the massive cores. All 40 individual cores with masses ranging between 0.6 and 15 Msun were detected above a 3 σ level of 3.6 mJy beam −1 (or 0.4 Msun), based on which we derive a total core mass of 107 Msun in the AFGL 961 region. As compared to the stellar initial mass function, a shallow slope of 1.8 is, however, derived from the best fit to the mass spectrum of the millimeter cores with a prestellar and/or protostellar origin. The flatter core mass distribution in the AFGL 961 region is attributed here to dynamic perturbations from the massive molecular outflows that originated from the massive protostellar binary, which may have altered the otherwise more quiescent conditions of core or star formation, enhanced the formation of more massive cores and, as a result, influenced the core mass distribution in its close vicinity.

  19. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Borio Di Tigliole, A.; Schaaf, Van Der; Barnea, Y.; Bradley, E.; Morris, C.; Rao, D. V. H. [Research Reactor Section, Vianna (Australia); Shokr, A. [Research Reactor Safety Section, Vienna (Australia); Zeman, A. [International Atomic Energy Agency, Vienna (Australia)

    2013-07-01

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from

  20. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    International Nuclear Information System (INIS)

    Borio Di Tigliole, A.; Schaaf, Van Der; Barnea, Y.; Bradley, E.; Morris, C.; Rao, D. V. H.; Shokr, A.; Zeman, A.

    2013-01-01

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from the MS

  1. Constraints on The Coupled Thermal Evolution of the Earth's Core and Mantle, The Age of The Inner Core, And The Origin of the 186Os/188Os Core(?) Signal in Plume-Derived Lavas

    Science.gov (United States)

    Lassiter, J. C.

    2005-12-01

    Thermal and chemical interaction between the core and mantle has played a critical role in the thermal and chemical evolution of the Earth's interior. Outer core convection is driven by core cooling and inner core crystallization. Core/mantle heat transfer also buffers mantle potential temperature, resulting in slower rates of mantle cooling (~50-100 K/Ga) than would be predicted from the discrepancy between current rates of surface heat loss (~44 TW) and internal radioactive heat production (~20 TW). Core/mantle heat transfer may also generate thermal mantle plumes responsible for ocean island volcanic chains such as the Hawaiian Islands. Several studies suggest that mantle plumes, in addition to transporting heat from the core/mantle boundary, also carry a chemical signature of core/mantle interaction. Elevated 186Os/188Os ratios in lavas from Hawaii, Gorgona, and in the 2.8 Ga Kostomuksha komatiites have been interpreted as reflecting incorporation of an outer core component with high time-integrated Pt/Os and Re/Os ( Brandon et al., 1999, 2003; Puchtel et al., 2005). Preferential partitioning of Os relative to Re and Pt into the inner core during inner core growth may generate elevated Re/Os and Pt/Os ratios in the residual outer core. Because of the long half-life of 190Pt (the parent of 186Os, t1/2 = 489 Ga), an elevated 186Os/188Os outer core signature in plume lavas requires that inner core crystallization began early in Earth history, most likely prior to 3.5 Ga. This in turn requires low time-averaged core/mantle heat flow (<~2.5 TW) or large quantities of heat-producing elements in the core. Core/mantle heat flow may be estimated using boundary-layer theory, by measuring the heat transported in mantle plumes, by estimating the heat transported along the outer core adiabat, or by comparing the rates of heat production, surface heat loss, and secular cooling of the mantle. All of these independent methods suggest time-averaged core/mantle heat flow of ~5

  2. STRIP1, a core component of STRIPAK complexes, is essential for normal mesoderm migration in the mouse embryo.

    Science.gov (United States)

    Bazzi, Hisham; Soroka, Ekaterina; Alcorn, Heather L; Anderson, Kathryn V

    2017-12-19

    Regulated mesoderm migration is necessary for the proper morphogenesis and organ formation during embryonic development. Cell migration and its dependence on the cytoskeleton and signaling machines have been studied extensively in cultured cells; in contrast, remarkably little is known about the mechanisms that regulate mesoderm cell migration in vivo. Here, we report the identification and characterization of a mouse mutation in striatin-interacting protein 1 ( Strip1 ) that disrupts migration of the mesoderm after the gastrulation epithelial-to-mesenchymal transition (EMT). STRIP1 is a core component of the biochemically defined mammalian striatin-interacting phosphatases and kinase (STRIPAK) complexes that appear to act through regulation of protein phosphatase 2A (PP2A), but their functions in mammals in vivo have not been examined. Strip1 -null mutants arrest development at midgestation with profound disruptions in the organization of the mesoderm and its derivatives, including a complete failure of the anterior extension of axial mesoderm. Analysis of cultured mesoderm explants and mouse embryonic fibroblasts from null mutants shows that the mesoderm migration defect is correlated with decreased cell spreading, abnormal focal adhesions, changes in the organization of the actin cytoskeleton, and decreased velocity of cell migration. The results show that STRIPAK complexes are essential for cell migration and tissue morphogenesis in vivo. Copyright © 2017 the Author(s). Published by PNAS.

  3. Ultrasonic imaging of projected components of PFBR

    Energy Technology Data Exchange (ETDEWEB)

    Sylvia, J.I., E-mail: sylvia@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Jeyan, M.R.; Anbucheliyan, M.; Asokane, C.; Babu, V. Rajan; Babu, B.; Rajan, K.K.; Velusamy, K.; Jayakumar, T. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2013-05-15

    Highlights: ► Under sodium ultrasonic scanner in PFBR is for detecting protruding objects. ► Feasibility study for detecting Absorber rods and its drive mechanisms. ► Developed in-house PC based ultrasonic imaging system. ► Different case studies were carried out on simulated ARDM's. ► Implemented the experimental results to PFBR application. -- Abstract: The 500 MWe, sodium cooled, Prototype Fast Breeder Reactor (PFBR) is under advanced stage of construction at Kalpakkam in India. Opacity of sodium restricts visual inspection of components immersed in sodium by optical means. Ultrasonic wave passes through sodium hence ultrasonic techniques using under sodium ultrasonic scanners are developed to obtain under sodium images. The main objective of such an Under Sodium Ultrasonic Scanner (USUSS) for Prototype Fast Breeder Reactor (PFBR) is to detect and ensure that no core Sub Assembly (SA) or Absorber Rod or its Drive Mechanism is protruded in the above core plenum before starting the fuel handling operation. Hence, it is necessary to detect and locate the object, if it is protruding the above core plenum. To study the feasibility of detecting the absorber rods and their drive mechanisms using direct ultrasonic imaging technique, experiments were carried out for different orientations and profiles of the projected components in a 5 m diameter water tank. The in-house developed PC based ultrasonic scanning system is used for acquisition and analysis of data. The pseudo three dimensional color images obtained are discussed and the results are applicable for PFBR. This paper gives the details of the features of the absorber rods and their drive mechanisms, their orientation in the reactor core, experimental setup, PC based ultrasonic scanning system, ultrasonic images and the discussion on the results.

  4. Core-6 fucose and the oligomerization of the 1918 pandemic influenza viral neuraminidase

    International Nuclear Information System (INIS)

    Wu, Zhengliang L.; Zhou, Hui; Ethen, Cheryl M.; Reinhold, Vernon N.

    2016-01-01

    The 1918 H1N1 influenza virus was responsible for one of the most deadly pandemics in human history. Yet to date, the structure component responsible for its virulence is still a mystery. In order to search for such a component, the neuraminidase (NA) antigen of the virus was expressed, which led to the discovery of an active form (tetramer) and an inactive form (dimer and monomer) of the protein due to different glycosylation. In this report, the N-glycans from both forms were released and characterized by mass spectrometry. It was found that the glycans from the active form had 26% core-6 fucosylated, while the glycans from the inactive form had 82% core-6 fucosylated. Even more surprisingly, the stalk region of the active form was almost completely devoid of core-6-linked fucose. These findings were further supported by the results obtained from in vitro incorporation of azido fucose and "3H-labeled fucose using core-6 fucosyltransferase, FUT8. In addition, the incorporation of fucose did not change the enzymatic activity of the active form, implying that core-6 fucose is not directly involved in the enzymatic activity. It is postulated that core-6 fucose prohibits the oligomerization and subsequent activation of the enzyme. - Graphical abstract: Proposed mechanism for how core-fucose prohibits the tetramerization of the 1918 pandemic viral neuraminidase. Only the cross section of the stalk region with two N-linked glycans are depicted for clarity. (A) Carbohydrate–carbohydrate interaction on non-fucosylated monomer allows tetramerization. (B) Core-fucosylation disrupts the interaction and prevents the tetramerization. - Highlights: • Expressed 1918 pandemic influenza viral neuraminidase has inactive and active forms. • The inactive form contains high level of core-6 fucose, while the active form lacks such modification. • Core fucose could interfere the oligomerization of the neuraminidase and thus its activation. • This discovery may explain why

  5. Core-6 fucose and the oligomerization of the 1918 pandemic influenza viral neuraminidase

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Zhengliang L., E-mail: Leon.wu@bio-techne.com [Bio-Techne Inc., 614 McKinley Place NE, Minneapolis, MN 55413 (United States); Zhou, Hui [Gregg Hall, UNH Glycomics Center, University of New Hampshire (United States); Ethen, Cheryl M. [Bio-Techne Inc., 614 McKinley Place NE, Minneapolis, MN 55413 (United States); Reinhold, Vernon N., E-mail: Vernon.Reinhold@unh.edu [Gregg Hall, UNH Glycomics Center, University of New Hampshire (United States)

    2016-04-29

    The 1918 H1N1 influenza virus was responsible for one of the most deadly pandemics in human history. Yet to date, the structure component responsible for its virulence is still a mystery. In order to search for such a component, the neuraminidase (NA) antigen of the virus was expressed, which led to the discovery of an active form (tetramer) and an inactive form (dimer and monomer) of the protein due to different glycosylation. In this report, the N-glycans from both forms were released and characterized by mass spectrometry. It was found that the glycans from the active form had 26% core-6 fucosylated, while the glycans from the inactive form had 82% core-6 fucosylated. Even more surprisingly, the stalk region of the active form was almost completely devoid of core-6-linked fucose. These findings were further supported by the results obtained from in vitro incorporation of azido fucose and {sup 3}H-labeled fucose using core-6 fucosyltransferase, FUT8. In addition, the incorporation of fucose did not change the enzymatic activity of the active form, implying that core-6 fucose is not directly involved in the enzymatic activity. It is postulated that core-6 fucose prohibits the oligomerization and subsequent activation of the enzyme. - Graphical abstract: Proposed mechanism for how core-fucose prohibits the tetramerization of the 1918 pandemic viral neuraminidase. Only the cross section of the stalk region with two N-linked glycans are depicted for clarity. (A) Carbohydrate–carbohydrate interaction on non-fucosylated monomer allows tetramerization. (B) Core-fucosylation disrupts the interaction and prevents the tetramerization. - Highlights: • Expressed 1918 pandemic influenza viral neuraminidase has inactive and active forms. • The inactive form contains high level of core-6 fucose, while the active form lacks such modification. • Core fucose could interfere the oligomerization of the neuraminidase and thus its activation. • This discovery may explain

  6. Cores, Joins and the Fano-Flow Conjectures

    Directory of Open Access Journals (Sweden)

    Jin Ligang

    2018-02-01

    Full Text Available The Fan-Raspaud Conjecture states that every bridgeless cubic graph has three 1-factors with empty intersection. A weaker one than this conjecture is that every bridgeless cubic graph has two 1-factors and one join with empty intersection. Both of these two conjectures can be related to conjectures on Fano-flows. In this paper, we show that these two conjectures are equivalent to some statements on cores and weak cores of a bridgeless cubic graph. In particular, we prove that the Fan-Raspaud Conjecture is equivalent to a conjecture proposed in [E. Steffen, 1-factor and cycle covers of cubic graphs, J. Graph Theory 78 (2015 195–206]. Furthermore, we disprove a conjecture proposed in [G. Mazzuoccolo, New conjectures on perfect matchings in cubic graphs, Electron. Notes Discrete Math. 40 (2013 235–238] and we propose a new version of it under a stronger connectivity assumption. The weak oddness of a cubic graph G is the minimum number of odd components (i.e., with an odd number of vertices in the complement of a join of G. We obtain an upper bound of weak oddness in terms of weak cores, and thus an upper bound of oddness in terms of cores as a by-product.

  7. Teaching a Geographical Component in World History Curriculum

    Science.gov (United States)

    Kachina, Olga A.

    2011-01-01

    This article is devoted to the topic of teaching a geographical component in World History curriculum in American public high schools. Despite the fact that the federal legislation entitled "No Child Left Behind" (2001) declared geography as a "core" academic subject, geography was the only subject dropped from federal funding.…

  8. Earth's inner core: Innermost inner core or hemispherical variations?

    NARCIS (Netherlands)

    Lythgoe, K. H.; Deuss, A.|info:eu-repo/dai/nl/412396610; Rudge, J. F.; Neufeld, J. A.

    2014-01-01

    The structure of Earth's deep inner core has important implications for core evolution, since it is thought to be related to the early stages of core formation. Previous studies have suggested that there exists an innermost inner core with distinct anisotropy relative to the rest of the inner core.

  9. Dependence of Core and Extended Flux on Core Dominance ...

    Indian Academy of Sciences (India)

    Abstract. Based on two extragalactic radio source samples, the core dominance parameter is calculated, and the correlations between the core/extended flux density and core dominance parameter are investi- gated. When the core dominance parameter is lower than unity, it is linearly correlated with the core flux density, ...

  10. Triple-Component Drug-Loaded Nanocomposites Prepared Using a Modified Coaxial Electrospinning

    Directory of Open Access Journals (Sweden)

    Wei Qian

    2013-01-01

    Full Text Available Triple-component nanocomposites for improved sustained drug release profiles are successfully fabricated through a modified coaxial electrospinning process, in which only organic solvent N,N-dimethylacetamide was used as sheath fluid. Using polyacrylonitrile (PAN as filament-forming matrix, ibuprofen (IBU as functional ingredient, and polyvinylpyrrolidone (PVP as the additional component, the IBU/PVP/PAN triple-component nanocomposites had uniform structure and an average diameter of 620±120 nm and 650±120 nm when the contents of PVP in the nanofibers were 10.5% and 22.7%, respectively. The optimal sheath-to-core flow rate ratio was 0.11 under a total sheath and core fluid flow rate of 1.0 mL/h. Compared with the IBU/PAN composite nanofibers, the triple-component composites could release 92.1% and 97.8% of the contained IBU, significantly larger than a value of 73.4% from the former. The inclusion of PVP in the IBU/PAN could effectively avoid the entrapment of IBU in the insoluble PAN molecules, facilitating the in vitro release of IBU. The modified coaxial process and the resulted multiple component nanocomposites would provide new way for developing novel drug sustained materials and transdermal drug delivery systems.

  11. Design Procedure of Graphite Components by ASME HTR Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ji-Ho; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet.

  12. Design Procedure of Graphite Components by ASME HTR Codes

    International Nuclear Information System (INIS)

    Kang, Ji-Ho; Jo, Chang Keun

    2016-01-01

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet

  13. On-line core monitoring with CORE MASTER / PRESTO

    International Nuclear Information System (INIS)

    Lindahl, S.O.; Borresen, S.; Ovrum, S.

    1986-01-01

    Advanced calculational tools are instrumental in improving reactor plant capacity factors and fuel utilization. The computer code package CORE MASTER is an integrated system designed to achieve this objective. The system covers all main activities in the area of in-core fuel management for boiling water reactors; design, operation support, and on-line core monitoring. CORE MASTER operates on a common data base, which defines the reactor and documents the operating history of the core and of all fuel bundles ever used

  14. Component Repair Times Obtained from MSPI Data

    Energy Technology Data Exchange (ETDEWEB)

    Eide, Steven A. [Curtiss-Wright/Scietech, Ketchum, ID (United States); Cadwallader, Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    Information concerning times to repair or restore equipment to service given a failure is valuable to probabilistic risk assessments (PRAs). Examples of such uses in modern PRAs include estimation of the probability of failing to restore a failed component within a specified time period (typically tied to recovering a mitigating system before core damage occurs at nuclear power plants) and the determination of mission times for support system initiating event (SSIE) fault tree models. Information on equipment repair or restoration times applicable to PRA modeling is limited and dated for U.S. commercial nuclear power plants. However, the Mitigating Systems Performance Index (MSPI) program covering all U.S. commercial nuclear power plants provides up-to-date information on restoration times for a limited set of component types. This paper describes the MSPI program data available and analyzes the data to obtain median and mean component restoration times as well as non-restoration cumulative probability curves. The MSPI program provides guidance for monitoring both planned and unplanned outages of trains of selected mitigating systems deemed important to safety. For systems included within the MSPI program, plants monitor both train UA and component unreliability (UR) against baseline values. If the combined system UA and UR increases sufficiently above established baseline results (converted to an estimated change in core damage frequency or CDF), a “white” (or worse) indicator is generated for that system. That in turn results in increased oversight by the US Nuclear Regulatory Commission (NRC) and can impact a plant’s insurance rating. Therefore, there is pressure to return MSPI program components to service as soon as possible after a failure occurs. Three sets of unplanned outages might be used to determine the component repair durations desired in this article: all unplanned outages for the train type that includes the component of interest, only

  15. Polymer Magnetic Composite Core Based Microcoils and Microtransformers for Very High Frequency Power Applications

    Directory of Open Access Journals (Sweden)

    Saravana Guru Mariappan

    2016-04-01

    Full Text Available We present a rapid prototyping and a cost effective fabrication process on batch fabricated wafer-level micro inductive components with polymer magnetic composite (PMC cores. The new PMC cores provide a possibility to bridge the gap between the non-magnetic and magnetic core inductive devices in terms of both the operating frequency and electrical performance. An optimized fabrication process of molding, casting, and demolding which uses teflon for the molding tool is presented. High permeability NiFeZn powder was mixed with Araldite epoxy to form high resistive PMC cores. Cylindrical PMC cores having a footprint of 0.79 mm 2 were fabricated with varying percentage of the magnetic powder on FR4 substrates. The core influence on the electrical performance of the inductive elements is discussed. Inductor chips having a solenoidal coil as well as transformer chips with primary and secondary coils wound around each other have been fabricated and evaluated. A core with 65% powder equipped with a solenoid made out of 25 µm thick insulated Au wire having 30 turns, yielded a constant inductance value of 2 µH up to the frequency of 50 MHz and a peak quality factor of 13. A 1:1 transformer with similar PMC core and solenoidal coils having 10 turns yielded a maximum efficiency of 84% and a coupling factor of 96%. In order to protect the solenoids and to increase the mechanical robustness and handling of the chips, a novel process was developed to encapsulate the components with an epoxy based magnetic composite. The effect on the electrical performance through the magnetic composite encapsulation is reported as well.

  16. Further HTGR core support structure reliability studies. Interim report No. 1

    International Nuclear Information System (INIS)

    Platus, D.L.

    1976-01-01

    Results of a continuing effort to investigate high temperature gas cooled reactor (HTGR) core support structure reliability are described. Graphite material and core support structure component physical, mechanical and strength properties required for the reliability analysis are identified. Also described are experimental and associated analytical techniques for determining the required properties, a procedure for determining number of tests required, properties that might be monitored by special surveillance of the core support structure to improve reliability predictions, and recommendations for further studies. Emphasis in the study is directed towards developing a basic understanding of graphite failure and strength degradation mechanisms; and validating analytical methods for predicting strength and strength degradation from basic material properties

  17. Induced supersolidity in a mixture of normal and hard-core bosons

    International Nuclear Information System (INIS)

    Mishra, Tapan; Das, B. P.; Pai, Ramesh V.

    2010-01-01

    We present a scenario where a supersolid is induced in one of the components of a mixture of two species bosonic atoms where there are no long-range interactions. We study a system of normal and hard-core boson mixture with only the former possessing long-range interactions. We consider three cases: the first where the total density is commensurate and the other two where it is incommensurate to the lattice. By suitable choices of the densities of normal and hard-core bosons and the interaction strengths between them, we predict that the charge density wave and the supersolid orders can be induced in the hard-core species as a result of the competing interatomic interactions.

  18. Performance of core modifications to reduce the reactor pressure vessel fluence

    International Nuclear Information System (INIS)

    Kiehlmann, H.D.; Lisdat, R.; Sommer, D.

    1997-01-01

    It's often discussed that nuclear power plants (NPP) are designed for an operation of 40 years equivalent to 32 full power years (FPY) assuming a load factor of 0.8. Such fixed plant life times are subjects of US operating licenses but not, as in most other countries, in the Federal Republic of Germany. Here the operating licenses are issued for an indefinite period. However, the German utilities are continuously upgrading their plants to attain a safety level that meets all current requirements. These upgrading measures also include the replacement of bigger components like e.g. the steam generator. The reactor pressure vessel (RPV), however, has a special status. Unlike most other components of a NPP which most likely will be exchanged during its service life a replacement or annealing treatment of the RPV certainly require more efforts to be economically justified. Thus the embrittlement of the RPV has an essential impact on the life time of a NPP. The end-of-life (EOL) RPV material toughness in essential depends on the steel quality and the accumulated neutron fluence. For a given NPP the reduction of the neutron flux at the inner surface of the RPV is the only way to limit its embrittlement. The resulting modifications for the core loadings in combination with the insertion of additional core components like steel elements are described and the impact on core performance and RPV fluence considered. (UK)

  19. Restraint system for core elements of a reactor core

    International Nuclear Information System (INIS)

    Class, G.

    1975-01-01

    In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

  20. Gravity driven emergency core cooling experiments with the PACTEL facility

    International Nuclear Information System (INIS)

    Munther, R.; Kalli, H.; Kouhia, J.

    1996-01-01

    PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs