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Myodo, Masato; Kobayashi, Tadayoshi; Tomii, Hiroyuki
JAEA-Technology 2008-001, 46 Pages, 2008/03
Dry wire-saw technology was applied for removal of penetrating pipes in annex building B of JAEA's reprocessing test facility (JRTF) and concrete blocks with penetrating pipes were successfully removed. The concrete block was cracked by a silent demolition agent and was easily separated into concrete and pipes by the secondary crushing by use of a hand-breaker. Effectiveness of the dry wire-saw technology for removal of penetrating pipes was discussed and evaluated based on various work data obtained in the present activity. The use of silent demolition agent was found to be very effective for the secondary crushing of concrete blocks and its charging condition into concrete was established so as to give effective cracks for the secondary crushing. As a result of this activity, it was confirmed that the dry wire-saw technology coupled with use of a silent demolition agent was one of hopeful candidates to remove penetrating pipes safely and effectively in reprocessing facilities.
Sekita, Kenji; Kuroha, Misao; Emori, Koichi; Kondo, Masaaki; Ouchi, Hiroshi; Shinozaki, Masayuki
JAEA-Technology 2008-002, 49 Pages, 2008/03
Graphite structures are used as one of the HTTR core internal structures. Graphite structures have high heat resistant property but its mechanical strength degrades easily by oxidization. To prevent the oxidization of graphite structures, impurity concentrations in the coolant of helium are controlled strictly. The helium sampling system is installed to measure the impurity concentrations in the helium. At gas compressor in helium sampling system, seal-oil leak at rod seal mechanism was occurred. The causes are degradation of seal material and contaminant abrasion powder of grand-packing. As these countermeasure, material of seal material was changed and contaminant was decreased. As the result long term operation is enabled. Moreover, reliable data can be obtained and efficient impurity control is enabled due to renewal of data acquisition control computer of gas chromatograph mass spectrometer and improvement of liquid nitrogen trap.
Somolova, M.*; Terada, Atsuhiko; Takegami, Hiroaki; Iwatsuki, Jin
JAEA-Technology 2008-003, 41 Pages, 2008/12
Japan Atomic Energy Agency has been conducting a conceptual design study of nuclear hydrogen demonstration plant, that is, a thermo-chemical IS process hydrogen plant coupled with the High temperature Engineering Test Reactor (HTTR-IS), which will be planed to produce a large amount of hydrogen up to 1000m/h. As part of the conceptual design work of the HTTR-IS system, preliminary analyses on small break of a hydrogen pipeline in the IS process hydrogen plant was carried out as a first step of the safety analyses. This report presents analytical results of hydrogen diffusion behaviors predicted with a CFD code, in which a diffusion model focused on the turbulent Schmidt number was incorporated. By modifying diffusion model, especially a constant accompanying the turbulent Schmidt number in the diffusion term, analytical results was made agreed well with the experimental results.
Watanabe, Masao; Ogiwara, Norio; Sawa, Soji*; Tanaka, Toshihiro*
JAEA-Technology 2008-004, 14 Pages, 2008/02
At the extraction part of J-PARC 3 GeV Rapid Cycling Synchrotron, a race-track bellows between the vacuum chamber of extraction septum magnet 1 and 2 is required, because there is no space to install a circular bellows. However, formed titanium bellows of race-track shape has never been fabricated. Smaller size race-track bellows than that of actual size were fabricated. Characteristics of them were and measured and evaluated. We performed measurements of spring force, expansion/contraction repetition life test, vacuum heating test and He-gas leak test. The characteristics of the race-track bellows are satisfied with our use conditions. It was judged from the above, we obtained prospect that using race-track bellows between the vacuum chamber of extraction septum magnet 1 and 2 is one of the solutions.
Tachibana, Mitsuo; Shiraishi, Kunio; Ishigami, Tsutomu; Tomii, Hiroyuki
JAEA-Technology 2008-005, 33 Pages, 2008/03
The PL measuring device was produced to apply to the clearance verification measurement and the radiation measurement for releasing controlled areas. The basic characteristic test and the actual test were confirmed using the PL measuring device. As a result of these tests, it was found that the evaluation value of radioactivity with the PL measuring device was accuracy equal with the existing measuring device. The PL measuring device has feature of the existing measuring device with a light weight and easy operability. The PL measuring device can correct the ray too. The PL measuring device is effective to the clearance verification measurement of concrete on buildings and the radiation measurement for releasing controlled areas.
Sone, Tomoyuki; Nonaka, Kazuharu; Sasaki, Toshiki; Yamaguchi, Hiromi
JAEA-Technology 2008-006, 23 Pages, 2008/03
Steam reforming method consists of the gasification process (GP) in which organics are vaporized and decomposed with superheated steam and the oxidation process in which vaporized organics are decomposed with heated air. Experimental study in which waste TBP/n-dodecane (WTBP) containing uranium was used was conducted with the steam reforming system (SRS) to examine the distribution ratio of uranium in the system, the feasibility of treatment of WTBP and the effect of treatment with SRS on the volume reduction of WTBP. The results of these studies are as follows: (1) Most of uranium compounds in WTBP are separated from organics in GP. (2) Only the solid waste remained in GP is the radioactive secondary waste from the treatment of WTBP with SRS. (3) The maintenance operation of the equipments installed downstream of GP become easy to perform. (4) The volume of secondary solid wastes is very small because more than 99% of the WTBP were evaporated in GP. (5) It was estimated that the effect of treatment with SRS on the volume reduction of WTBP is 30 times more than that of pyrolysis method. These results show that SRS can achieve high volume reduction of WTBP.
Sumita, Junya; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sawa, Kazuhiro
JAEA-Technology 2008-007, 23 Pages, 2008/03
In accordance with the basic policy of effectively using nuclear fuel resources, the FBR cycle, one of the most possible fuel cycle in the future, will be adapted after plu-thermal program by LWR in Japanese nuclear cycle plan. In this paper, a case study of technical investigation of HTGR fuel cycle based on HTGR fuel cycle proposed to adapt to Japanese nuclear fuel cycle plan were carried out from the viewpoint of effective utilization of uranium, fabrication technologies of MOX fuel, reprocessing technologies, amount of interim storage of HTGR fuel and graphite waste. As a result, the fuel cycle for HTGR is expected to be possible technically.
Abe, Kazuyuki; Kobayashi, Takashi*; Kajima, Hisashi*; Yoshikawa, Katsunori; Nagamine, Tsuyoshi; Nakamura, Yasuo
JAEA-Technology 2008-008, 53 Pages, 2008/03
MARICO-2 is a Testing Rig for the continuous irradiation examination of ODS ferrite steel etc.. It was necessary to re-assemble of MARICO-2 in Fuel Monitoring Facility (FMF). However, MARICO-2 is not applicable a past technology of re-assembly because it is a Rig of the total length about 11 m and its hex-tube must be welded by remote control. Then, MARICO-2 re-assembly technology development was executed, the device was designed, it produced, and the procedure of re-assembly by remote control was established.
Nomura, Yasushi*; Takahashi, Satoshi*; Okuno, Hiroshi
JAEA-Technology 2008-009, 273 Pages, 2008/03
Safety demonstration analyses were conducted under contract with the Ministry of Economic, Trade and Industry of Japan from 2001 to 2004 for the purpose of assuaging public jitters concerning the transport. The current transport routes and the past accident/incident records were surveyed, three accident scenarios, i.e., a fall from an overpass, an open fire after collision with an oil tank trailer, a fire caused by collision with 2-ton truck inside a tunnel were set up. Mechanical damages and thermal failures were analyzed using the finite element codes LS-DYNA and ABAQUS. In addition, criticality safety analyses were made using the continuous energy Monte Carlo code MVP for the transport casks damaged in reference to the previous mechanical and thermal analyses. Thus, the integrity of packaging against leakage of radioactive material was shown in the case of severe accidents anticipated to occur during transportation without any harmful effect to the public and environment.
Hayashi, Kimio; Nakagawa, Tetsuya; Onose, Shoji; Ishida, Takuya; Kodaka, Hideo; Katsuyama, Kozo; Kitajima, Toshio; Takahashi, Kozo; Tsuchiya, Kunihiko; Nakamichi, Masaru; et al.
JAEA-Technology 2008-010, 68 Pages, 2008/03
In-pile functional tests of breeding blankets for fusion reactors have been planned by Japan Atomic Energy Agency (JAEA), using a test blanket module (TBM) which will be loaded in ITER. The present report describes a conceptual investigation and a basic design of the dismantling process for irradiation capsules which were used in irradiation experiments by the Japan Materials Testing Reactor (JMTR) of JAEA. In the present design, the irradiation capsule is cut by a band saw; the released tritium is recovered safely by a purge-gas system, and is consolidated into a radioactive waste form. Furthermore, adoption of the inner-box enclosing the dismantling apparatus has brought a prospect to be able to utilize an existing hot cell (beta- cell) equipped with usual wall material permeable to tritium, without extensive refurbishing of the cell. Thus, the present study has indicated the feasibility of the present dismantling process for the irradiated JMTR capsules containing tritium.
Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; et al.
JAEA-Technology 2008-011, 46 Pages, 2008/03
It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, It is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported.
Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; et al.
JAEA-Technology 2008-012, 36 Pages, 2008/03
It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, It is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack propagation test is reported.
Ide, Hiroshi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Hanawa, Satoshi; Matsui, Yoshinori; Iwamatsu, Shigemi; Kanazawa, Yoshiharu; Miwa, Yukio; Kaji, Yoshiyuki; et al.
JAEA-Technology 2008-013, 32 Pages, 2008/03
Dissolved oxygen ions and chlorine ions concentration have been used as an evaluation index of stress corrosion cracking behavior for the light water reactor materials. In addition to these parameters, Electrochemical Corrosion Potential (ECP) was commonly used as the evaluation. Therefore, as a part of the IASCC irradiation tests, the irradiation test of the iron oxide type and the platinum type of ECP sensor were carried out under the BWR coolant condition. As a result, some measurements of ECP sensor succeed. However, it was clear that the improvement of ECP sensor is necessary. In this report, developed irradiation capsule ECP sensor is reported.
Motoki, Riyozo; Aoki, Hiromichi; Uchida, Shoji; Nagaishi, Ryuji; Yamada, Reiji
JAEA-Technology 2008-014, 23 Pages, 2008/03
The study of producing hydrogen with a Sr-90/Y-90 source is planned to utilze the radioactive waste effectively. Therefore we developed two methods of caking Sr-90 and a catalyst for the production of hydrogen effectively. One is a method of caking SrTiO and TiO in a silica gel. And another is a method of caking SrSO and TiO in a silica gel. These solid matters are porous materials, which has a radiation resistant and chemical resistant. In addition, Y-90 which is a daughter nuclide of Sr-90 can be also used for hydrogen production.
Yagi, Masahiro; Watanabe, Masanori; Oyama, Koji; Komeda, Masao; Yamamoto, Kazuyoshi; Kashima, Yoichi
JAEA-Technology 2008-015, 91 Pages, 2008/03
The irradiation experimental device is designed by surveying analytically an irradiation condition to improve the neutron flux distribution of the radial direction on NTD-Si by using neutron transportation calculation Monte Carlo calculation code MCNP5 in order to develop neutron irradiation technology for the large-diameter silicon to 12 inches diameter at the maximum and an irradiation experiment will be performed in JRR-4. Thus the validity of the design technique of the irradiation device will be confirmed by this experiment. The irradiation experimental device is installed in the side of the core tank outer wall. A 12 inches silicon ingot of 60cm in height is irradiated in a reflector cover which surrounds the silicon ingot for increasing the flux. The silicon ingot is rotated during irradiation in order to require the condition of uniformly distributed thermal neutron flux over whole circumferences. A uniform irradiation condition was achieved by the pass-through core method, in which silicon ingot moved up and down with rotating. The condition was satisfied when silicon was covered with the thermal neutron filter, which is made from aluminum alloy of thickness of 2mm with natural boron element ratio of 1.5%, and was moved in the range from -42mm to 22mm for the center of the reactor core. The deviation of the Si-30 neutron absorption reaction rate was range from -3.2% to +5.3% in the intermediate volume of 40cm height and the neutron absorption reaction ratio of the circumference to the center was within 1.09 in the volume.
Hayashi, Koji; Shibata, Akira; Iwamatsu, Shigemi; Sozawa, Shizuo; Takada, Fumiki; Omi, Masao; Nakagawa, Tetsuya
JAEA-Technology 2008-016, 51 Pages, 2008/03
The irradiation capsule 74M-52J was irradiated during total 136 cycles at reactor core of JMTR and the maximum neutron dose reached on 3.910n/m at the capsule outer-tube made of a type 304 stainless steel. In order to produce mechanical test specimens from the outer tube, a punching technique was developed as a simple remote-handling method in a hot-cell. From comparison between the punching and the mechanical cutting methods, it was clarified that the punching technique was applicable to practical use. Moreover, an evaluation test of mechanical properties using specimens sampled from the 74M-52 was performed in-water high temperature condition, less than 288C. The result shows that the residual elongation is 18% at 150C and 13% at 288C. It was confirmed that the type 304 stainless steel irradiated up to such high dose shows enough ductility.
Murakami, Tatsutoshi; Suzuki, Kiichi; Hatanaka, Nobuhiro; Hanawa, Yukio; Shinozaki, Masaru; Murakami, Shinichi; Tobita, Yoshimasa; Kawasaki, Takeshi; Kobayashi, Yoshihito; Iimura, Naoto; et al.
JAEA-Technology 2008-017, 97 Pages, 2008/03
Low density MOX pellets for FBR "MONJU" have not been fabricated in Plutonium Fuel Fabricating Facility (PFPF) for these 9 years since completion of the first reload fuel for "MONJU" in 1995. In this period, about 60 % of machines in the pellet fabrication process of PFPF have been replaced with new ones, and fabrication of MOX pellets for "JOYO" has been continued using these machines. Concerning the feed MOX powders for "MONJU", the amount of decay heat has been increased with increase of accumulated Am-241 in this period. In addition, powder characteristic of recycled MOX powder which is one of feed powders, MH-MOX powder, UO powder and recycled MOX powder, was significantly changed by replacing former processing machine used for scrap recycling with improved one. Using MOX powder with increased decay heat and recycled MOX powder processed by new machine, a series of low density MOX pellet fabrication tests were conducted to confirm pellet fabrication conditions for current pellet fabrication machines from October in 2004 to August in 2006. As a conclusion, it was confirmed that low density MOX pellets could be fabricated using these feed powders and replaced machines by adjusting pellet fabrication conditions adequately. This report summarizes the results of a series of low density MOX pellet fabrication tests.
Saito, Yoshihiko; Ochs, M.*; Kunze, S.*; Kitamura, Akira; Tachi, Yukio; Yui, Mikazu
JAEA-Technology 2008-018, 116 Pages, 2008/03
In this report, the QA/classification of selected entries (1,056 Kd values) in the JNC-SDB, especially of Kd values for mudstone system to use in the Kd-derivation exercise for Horonobe rocks, was done following the approach defined in our previous report. This classification scheme made it possible to obtain quick overview of the available data, and to have suitable access to the respective Kd values.
Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; et al.
JAEA-Technology 2008-019, 57 Pages, 2008/03
The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors.
Hara, Teruo; Sakaba, Nariaki; Sato, Hiroyuki; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko
JAEA-Technology 2008-020, 28 Pages, 2008/03
Since nuclear hydrogen production system should be economically competitive compared with other hydrogen production methods, it is important to reduce the construction cost of the IS process hydrogen production plant. Japan Atomic Energy Agency (JAEA) has started the conceptual design study of components in the IS process hydrogen production system to be connected with the HTTR (HTTR-IS system). This paper describes the additional proposal for the mixer-Settler type Bunsen reactor and combined sulphuric acid decomposer of the HTTR-IS system which enhances their performance and enables reduction of fabrication cost. The enhancements of the mixing and separating performance in the Bunsen reactor using static mixers with microbubble and dams, and minimizing the size of the combined sulphuric acid decomposer by optimizing its flow path remarkably contribute to the size reduction of components.
Sueoka, Michiharu; Kawamata, Yoichi; Kurihara, Kenichi; Seki, Akiyuki
JAEA-Technology 2008-021, 23 Pages, 2008/03
A plasma movie is generally expected as one of the most efficient methods to know what plasma discharge has been conducted in the experiment. The JT-60 plasma movie is composed of video camera picture looking at a plasma, computer graphics (CG) picture, and magnetic probe signal as a sound channel. In order to use this movie efficiently, we have developed a new system having the following functions: (a) To store a plasma movie in the movie database system automatically combined with the plasma shape CG and the sound according to a discharge sequence. (b) To make a plasma movie is available (downloadable) for experiment data analyses at the Web-site. Especially, this system aimed at minimizing the development cost, and it tried to develop the real-time plasma shape visualization system (RVS) without any operating system (OS) customized for real-time use. As a result, this system succeeded in working under Windows XP. This report deals with the technical details of the plasma movie database system and the real-time plasma shape visualization system.
Shimada, Katsuhiro; Terakado, Tsunehisa; Kurihara, Kenichi
JAEA-Technology 2008-022, 24 Pages, 2008/03
In JT-60SA, the heating power with 41 MW - 100 s is assumed at the additional heating facilities including NBI and RF facilities. The present AC power system, which is composed of the motor generator (H-MG, Capacity: 400 MVA), can not supply the necessary AC power to the additional heating facilities because the necessary AC power is about 130 MW - 100 s considering the efficiency and power factor of the additional heating facilities. Therefore, new AC power system is necessary for the additional heating facilities in JT-60SA. In this report, we propose new AC power system directly powered from the commercial line for the additional heating facilities in JT-60SA. Moreover, this report describes the design study of the reactive power compensator including high harmonic current filters and shunt capacitors to be satisfied with the receiving condition for NAKA Fusion Institute using PSCAD/EMTDC simulation.
Ota, Kazunori; Ikekame, Yoshinori; Owada, Minoru; Fukushima, Manabu; Oba, Toshinobu; Takeuchi, Masaki; Imahashi, Masaki; Murayama, Yoji
JAEA-Technology 2008-023, 31 Pages, 2008/03
JRR-3 uses shell and tube heat exchangers. The secondary coolant flushes into the tubes and the primary coolant flows outside of the tubes. The heat exchangers are cleaned with the ball-cleaning method, which is a method to clean inside of the tubes by passing the sponge balls with secondary coolant. Decline in the performance of heat exchanger could rise temperature of the primary coolant and then influence the safe and stable reactor operation. The effective way of ball-cleaning for JRR-3 heat exchangers is examined based on past cleaning data. The results show the optimal ball size and the way to determine the cleaning time.
Arakawa, Kazuo; Yokota, Wataru; Agematsu, Takashi; Nakamura, Yoshiteru; Ishibori, Ikuo; Kurashima, Satoshi; Miyawaki, Nobumasa; Okumura, Susumu; Kamiya, Tomihiro; Nara, Takayuki; et al.
JAEA-Technology 2008-024, 141 Pages, 2008/03
The TIARA facilities (Takasaki Ion Accelerators for Advanced Radiation Application: TIARA) have been constructed at Takasaki Radiation Chemistry Research Establishment (Takasaki Advanced Radiation Research Institute) under 6-year program from 1988. The first phase of the construction including those of a K110 AVF cyclotron and a 3 MV tandem accelerator was complete in October 1991. Large AVF cyclotrons have been used mostly for fundamental nuclear physics and medical applications to radiation therapy and radioisotope production so far. The JAEA K110 AVF cyclotron is the first one dedicated to R&D in materials science and other irradiation purpose. This cyclotron facility consists of three external ion sources, an injection line, a K110 AVF cyclotron, eight main beam transport lines, and a control system. The first beam, 50 MeV He, was extracted from the cyclotron in March 1991. This Report describes an outline of the K110 AVF cyclotron system and results of performance tests.
Watanabe, Fumitaka; Yamamoto, Kazuya; Sajiki, Kenjiro; Yasu, Sadanori*; Igarashi, Miyuki*
JAEA-Technology 2008-025, 63 Pages, 2008/03
The larger the scale of an accident, the more closely cooperation is needed between concerned parties for appropriate and timely response, especially if they are located apart from each other. The past nuclear accidents revealed that they failed to share important information with each other and such a situation caused unnecessary confusion in public information. Based on lessons learned from accidents, JAEA developed the Emergency Information Clearinghouse (ECHO). Information is fed into a secured server from each PC. Unified information on the server can be chronologically listed through a web browser. This web-based system enables separately located parties to share proper information in a timely manner and minimize the load of managing a great deal of information. The ECHO has been in operation 24/7 on a network for nuclear emergency response and connects nationwide 22 off-site centers, and several authorities concerned. The system has been used during for training and drills.
Kobayashi, Jun; Sato, Hiroyuki; Hayashi, Kenji; Kamide, Hideki
JAEA-Technology 2008-026, 30 Pages, 2008/03
High burn-up core in a sodium cooled fast reactor may result in fuel pin deformation due to irradiation, e.g., swelling and bowing. Such deformation will influence thermal hydraulics in a core fuel subassembly, i.e., the highest temperature. Thus it is significant to estimate pin deformation effects. A water experiment was carried out to measure velocity field in a deformed pin bundle by using transparent resin and refractive index matching technique. In the experiment, the deformed pin geometry should be measured as a boundary condition. Here an optical measurement technique of pin bundle geometry was developed. An image of a horizontal cross section of the pin bundle was captured by a camera set on an inclined line to the pin bundle axis. This image was converted to a straight image to the pin bundle axis by taking accounts of refraction at the wall. The positions of deformed pins were well estimated by this method.
Takada, Hiroshi; Kai, Tetsuya; Oikawa, Kenichi; Sakai, Kenji
JAEA-Technology 2008-027, 34 Pages, 2008/03
We have designed the control systems of neutron beam shutters of the 1-MW spallation neutron source under the Japan Proton Accelerator Research Complex (J-PARC) project. The control system was designed that users were able to open and/or close shutters at experimental halles during a beam-time and facility's personnel could move shutters near the shutter drive mechanism room at the maintenance period. Moreover, it was designed to keep the primary condition that personnels were allowed to enter the instrument room when the shutter is closed. This condition was integrated to the Personnel Protection System (PPS) to ensure the safety of the personnel. Components were also designed to enhance the reliability of the control system as follows: shutter close position is detected independently by two kinds of devices of a mechanical limit switch and a rotary encorder, and the power supply for the shutter system was connected with an UPS to prevent loss of signals in case of a power failure.
Takasaki, Koji; Sagawa, Naoki; Kurosawa, Shigeyuki*; Shioya, Satoshi; Suzuki, Kazunori; Horikoshi, Yoshinori; Mizuniwa, Harumi
JAEA-Technology 2008-028, 73 Pages, 2008/03
Recently, an imaging plate (IP) was developed. Pu analysis by an IP was studied in the JAEA Cooperative Research Scheme on the Nuclear Fuel Cycle, and the availability about detection of Pu and discrimination with radon progenies was shown. In order to apply these results to the radiation control, this research aims at evaluation of radioactivity, discrimination with radon progenies, and evaluation of AMAD by an IP. In the conventional autoradiography (ARG) by a ZnS and Polaroid film, Contamination of Pu can only be grasped by viewing. Since the digital data of the radiation and location is obtained by an IP, Pu can be detected quantitatively and radioactivity can be evaluated. As a result of this research, the image of alpha emitter on an IP is the same as that of the conventional ARG, and can be applied also to the ARG of beta and emitter. In measurement of Pu, the function of Image J (image-analysis software) was very available. Pu radioactivity was evaluated by an IP.
Shibata, Akira; Kawamata, Kazuo; Taguchi, Taketoshi; Kaji, Yoshiyuki; Shimizu, Michio*; Kanazawa, Yoshiharu; Matsui, Yoshinori; Iwamatsu, Shigemi; Sozawa, Shizuo; Tayama, Yoshinobu; et al.
JAEA-Technology 2008-029, 40 Pages, 2008/03
Irradiation assisted stress corrosion cracking (IASCC) is considered to be one of the key issues from a viewpoint of the life management of core components in the aged Light Water Reactors. The in-situ crack extension examination and the in-situ constant load tensile test in the reactor are required for the study of IASCC. There are, however, some technical hurdles to be overcome for the experiments. For this in-situ IASCC test, techniques for assembling pre-irradiated specimens into an capsule in a hot cell by remote handling are necessary. In this report, I describe the establishment of those remote assembling techniques and development of new welding apparatus and the TIG upset welding for stainless tube of 3 mm in thickness. Already IASCC capsules having pre-irradiated CT specimens were remotely assembled using these techniques in the hot cell for performing crack growth tests under irradiation in JMTR. And eight in-situ IASCC capsules have been finished successfully in JMTR.
Matsuda, Norihiro; Iwamoto, Yosuke; Harada, Masahide; Teshigawara, Makoto; Meigo, Shinichiro; Maekawa, Fujio; Oguri, Tomomi*; Nakano, Hideo*; Nakao, Noriaki*; Nakane, Yoshihiro; et al.
JAEA-Technology 2008-030, 150 Pages, 2008/03
Radiation Shielding design/safety analysis required for licensing of the high intensity proton accelerator facility J-PARC (Japan Proton Accelerator Research Complex) is in progress, using various high-energy particle transport codes. Shielding benchmark problems, mainly based on experiments, on thick target neutron yield, beam dump geometry, deep penetration and radiation streaming were prepared and analyzed by the shielding design codes, in order to estimate the code's accuracy. The results of analyses show that the calculation results agree with the experiments within a factor of two on the whole.
Shimada, Katsuhiro; Omori, Yoshikazu; Okano, Jun; Matsukawa, Tatsuya; Terakado, Tsunehisa; Kurihara, Kenichi
JAEA-Technology 2008-031, 38 Pages, 2008/03
In JT-60SA, Toroidal Field coil (TF coil) and Poloidal Field coils (PF coils) are superconducting coils and a long-pulse plasma operation with more than 100s flattop is assumed. Corresponding to the superconducting TF and PF coils, new DC power supply system in JT-60SA is necessary. The DC power supply system is composed of the reused JT-60 power supply components and newly manufactured ones to reduce total cost. A quench protection circuit is newly introduced to fast discharge coil magnetic stored energy. This paper describes the initial design study of JT-60SA DC power supply system for TF and PF coils in Japan Atomic Energy Agency (JAEA).
Sone, Tomoyuki; Sasaki, Toshiki; Miyamoto, Yasuaki; Yamaguchi, Hiromi; Inoue, Haruka*; Kihara, Tsuyoshi*; Takei, Yoshihisa*; Tatekawa, Takaiki*; Fukaya, Masaaki*; Iriya, Keishiro*; et al.
JAEA-Technology 2008-032, 25 Pages, 2008/03
Reformed sulfur (RS) is superior in water interception and acid resistance compared with cement. Therefore solidified wastes with RS should have the high resistance to leaching. Unconfined compressive strength test and leaching test using solidified simulated wastes containing lead contaminated with radioactive nuclides (Lead waste) with RS and solidified simulated low level radioactive liquid waste (LLLW) with RS were conducted to examine the applicability of reformed sulfur solidification method (RSSM) as solidification technique of Lead waste and LLLW. The results of these studies show that RSSM is effective technique for stabilization of lead compared with cement solidification method because solidified lead with RS has much stronger resistance to leaching of lead than solidified lead with cement. It also show that the applicability of RSSM as solidification technique of the waste containing lead oxide and LLLW is low because the resistance to leaching of solidified lead oxide with RS and of solidified simulated LLLW with RS were equal to or lower than those of solidified products with cement respectively.
Haga, Katsuhiro; Wakui, Takashi; Kogawa, Hiroyuki; Harada, Masahide; Futakawa, Masatoshi; Hayashi, Kenichi*; Nakamura, Koji*
JAEA-Technology 2008-033, 40 Pages, 2008/05
JAEA has completed the mercury target system as the spallation neutron source of J-PARC project, which has the power of the world highest level of 1MW. The basic design of the flow channel and structure of the mercury target vessel was carried out by JAEA, and the detail design, parts fabrication and assembling has been carried out by the vendor from 2003. Taking these fabrication designs and assembling conditions into consideration, the final performance evaluation of the mercury target vessel was carried out in view of thermal hydraulics. The general thermal hydraulic analyses code, STAR-CD, was used, and the three dimensional thermal hydraulic analyses were carried out taking the nuclear heating and the heat transportation into consideration. As results, it was confirmed that the mercury target vessel fulfills the design requirements such as the fluid inlet velocity, the maximum temperature of fluid, the maximum temperature of the vessel, the pressure drop of fluid, etc.
Kikuchi, Katsumi; Akino, Noboru; Ebisawa, Noboru; Ikeda, Yoshitaka; Seki, Norikazu*; Takenouchi, Tadashi; Tanai, Yutaka
JAEA-Technology 2008-034, 25 Pages, 2008/04
The control system for auxiliary pumping facility and primary water cooling facility in JT-60 NBI was updated. To realize the cost reduction, the control system with many input and outputs of 2000 was updated by JAEA itself using commercial Programmable Logic Controllers (PLC's). JAEA also made software with 3600 ladder lines by JAEA itself based on commercial basic programs. In addition to the simple replacement of the hardware and software, the function of remote operation has been newly added. At present, the auxiliary pumping facility and the primary water cooling facility have been stably operated without troubles. The remote operation enables to collect the detailed information on the trouble more easily, resulting in a quick countermeasure for the trouble.
Iimura, Koichi; Hosokawa, Jinsaku; Kanno, Masaru; Kitajima, Toshio; Nakagawa, Tetsuya; Sakamoto, Taichi; Hori, Naohiko; Kawamura, Hiroshi
JAEA-Technology 2008-035, 47 Pages, 2008/06
At Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA) advances the plan of refurbishing Japan Materials Testing Reactor (JMTR) to start the operation in fiscal 2011. As part of effective use for JMTR, JAEA is planning to product Mo, which is a parent nuclide of Tc. Tc is most commonly used as a radiopharmaceutical in the field of nuclear medicine. Currently the supplying of Mo is only depend on imports from foreign countries, so JAEA is aiming at domestic production of a part of Mo in cooperation with the industrial circles. In this article, JAEA described the process, the choice and fabric of the irradiation facilities for Mo production, the technical study of commercializing equipment after irradiation, and the cost study for Mo production.
Tomita, Kenji; Tsuchiya, Kunihiko; Onuma, Yuichi; Inoue, Shuichi; Watanabe, Hiroyuki; Saito, Takashi; Kikuchi, Taiji; Hayashi, Kimio; Kitajima, Toshio
JAEA-Technology 2008-036, 61 Pages, 2008/06
The second in-situ irradiation experiment using a mock-up (ORIENT-II, JMTR capsule Number: 99M-54J) with a tritium breeder (LiTiO) pebble bed in the Japan Materials Testing Reactor (JMTR) was finished on Aug. 1, 2006. Consideration on the detaching procedure of the irradiated mock-up contaminated with tritium with pebble bed and a detaching test of this mock-up was carried out. The tritium removal properties were examined in the irradiated mock-up, the sweep gas tube, the protective tube and the junction box, for the decreasing of the tritium release to the area of detaching test. Melting/enclosed tests of sealing plug were also carried out for the prevention of tritium leakage from sweep gas lines of pebble bed. Then, the detaching test of the pebble bed was carried out. This report describes the results of tritium removal tests, examination of the detaching procedure, and the detaching test, as well as knowledge obtained from these tests and works.
Ema, Akira; Yokoyama, Kaoru; Nakatsuka, Yoshiaki; Shimaike, Masamitsu; Sugitsue, Noritake
JAEA-Technology 2008-037, 50 Pages, 2008/06
In the centrifugation method uranium enrichment plant, the UF gas was supplied for a long term. The main uranium compounds are estimated to the middle fluorides UFx(4x=5). This middle fluoride is changed to UF and IF gas again by the reaction with IF. Using this reaction, the uranium compound in the plant was decontaminated. The IF treatment tests were executed to the plant in four treatment conditions. It is necessary to clarify the relation between IF treatment condition and decontamination results for the best treatment condition setting. Therefore, the decontamination development is evaluated using the result of the weight measurement of recovery UF and IF gas, the -ray measurement, and the ICP-MS analysis at the IF treatment tests. Then the technical knowledge to clarify the decontamination characteristic features is accumulated for analyzing the treatment trend, the treatment time, decontamination level and variation of the decontamination level.
Hosokawa, Jinsaku; Kanno, Masaru; Sakamoto, Taichi
JAEA-Technology 2008-038, 24 Pages, 2008/06
At Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA) advances the plan of refurbishing Japan Materials Testing Reactor (JMTR) to start the operation in fiscal 2011. Silicon Semiconductor is used in every Country and every industrial field. Nowadays, the demand of large diameter Silicon Semiconductors are increasing. At JMTR in JAEA Oarai, the production of Silicon Semiconductors utilizing NTD (Neutron Transmutation Doping) Method is investigated. Particularly, this report describes the installation of Silicon Semiconductors producing facility on JMTR. This Report described the Conceptual Study for Silicon Semiconductor Irradiation Facility in JMTR.
Hanawa, Yoshio; Taguchi, Taketoshi; Tsuboi, Kazuaki; Saito, Takashi; Ishikawa, Kazuyoshi; Watahiki, Shunsuke; Tsuchiya, Kunihiko
JAEA-Technology 2008-039, 53 Pages, 2008/06
Beryllium has been utilized as a moderator and/or reflector in a number of material testing reactors. Beryllium frames and beryllium reflectors, which have been utilized as neutron reflector in Japan Materials Testing Reactor (JMTR) in JAEA, were fabricated beryllium metals of S-200F grade. Especially, it is necessary to exchange the beryllium frames every fixed period and there frames were exchanged six times up to the JMTR operation periods of 165 cycles. Therefore, preliminary irradiation test of beryllium metals was performed from 162 to 165 cycles of JMTR operations as a part of development on long life of beryllium reflectors. Two kinds of beryllium metals (S-200F and S-65C) were prepared in this test. This report is described the design study and fabrication of irradiation capsule of beryllium metals and dismount device of irradiation capsule.
Gorai, Shigeru; Naka, Michihiro; Izumo, Hironobu; Nagao, Yoshiharu; Kawamura, Hiroshi
JAEA-Technology 2008-040, 17 Pages, 2008/06
Various projects for improvement of JMTR are being carried out with the aim to re-start in FY2011. In the projects, it is planned to improve the JMTR core management system to execute neutronic calculations. The JMTR core management system is based on SRAC code system on mainframe, it is expected to improve the JMTR core management system by means of conversion from SRAC code system to the newest SRAC2006 code system on UNIX. Therefore neutronic calculations for JMTR using SRAC2006 code system were executed in order to estimate applicability and effectiveness of SARC2006 code system to the JMTR core management system.
Taguchi, Taketoshi; Hanawa, Yoshio; Watahiki, Shunsuke; Tsuchiya, Kunihiko
JAEA-Technology 2008-041, 23 Pages, 2008/06
Beryllium has been utilized as a reflector in a number of material testing reactors because of low parasitic capture cross section for thermal neutrons and good neutron elastic scattering characteristics. Beryllium frames and beryllium reflectors, which have been utilized as neutron reflector in Japan JMTR, were fabricated beryllium metals. Especially, it is necessary to exchange the beryllium frames every fixed period. Therefore, preliminary irradiation test of beryllium metals (S-200F and S-65C) was performed in JMTR for development on long life of beryllium reflectors. The post irradiation examinations (PIEs) were carried out for the effect on the properties of these irradiated beryllium metals. In these PIEs, size change of the irradiated beryllium was measured with the specimens for bending. This report is described development of the high accuracy measurement device.
Kobayashi, Kaoru; Hanada, Masaya; Kamada, Masaki; Akino, Noboru; Sasaki, Shunichi; Ikeda, Yoshitaka
JAEA-Technology 2008-042, 25 Pages, 2008/06
Breakdown locations of a JT-60U negative ion source were investigated to improve the voltage holding capability. The accelerator is characterized by three acceleration stages with large grids 0.45 m 1.1 m and large FRP insulators 1.8 m in inner diameter. High voltages were applied to each acceleration stage independently. Voltage holding capabilities of each stage were almost the same, 120-130 kV, which was lower than the design acceleration voltage of 167 kV. Then, in order to identify whether the breakdowns occur in the gaps between grids or on the surface of the FRP insulators, high voltages were also applied to the accelerator with the grids and their support flanges removed. The voltage holding capabilities of three FRP insulators rapidly achieved 167 kV. These results indicate that the breakdowns mainly occur in the gaps between the acceleration grids and/or their support flanges.
Tomita, Kenji; Hosokawa, Jinsaku; Matsui, Yoshinori
JAEA-Technology 2008-043, 21 Pages, 2008/07
In JMTR, highly precise-ization of irradiation temperature evaluation of material capsule is advanced towards re-operation in the 2011 fiscal year. In the conventional sub-program, rectangular form was loaded into the capsule and temperature evaluation was carried out. In this development, the CT specimen which is special form is loaded into capsule, and it enabled it to analyze a temperature distribution. Moreover, the sub-program which can also evaluate the heat stress distribution by this uneven temperature distribution was developed. And, in order to be able to perform easily extraction and re-load of irradiation specimens, the structure of separation specimen holder effective in a re-irradiation examination is adopted. However, with this structure, the inside of capsule serves as uneven temperature distribution. For this reason, the sub-program which can evaluate this temperature distribution was developed.
Watahiki, Shunsuke; Saito, Takashi; Tsuchiya, Kunihiko; Ohara, Hiroshi; Iimura, Koichi
JAEA-Technology 2008-044, 42 Pages, 2008/06
This report is described for the development of nicrosil-nisil type multi-paired thermocouple which was usable at over 1000 degrees under the neutron irradiation environment. Developed nicrosil-nisil type multi-paired thermocouple has maximum 7 hot junctions in axial direction in a sheath. Though the design, trial production and out-pile tests, its productivity and electric performances were confirmed, and the production method was established.
Nojiri, Naoki; Owada, Hiroyuki; Kato, Yasushi
JAEA-Technology 2008-045, 38 Pages, 2008/06
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.
Kashiwai, Yoshio*; Sanada, Hiroyuki; Matsui, Hiroya
JAEA-Technology 2008-046, 52 Pages, 2009/07
This work was performed by Taisei Kiso Sekkei Co., Ltd. under contract with Japan Atomic Energy Agency. The objective of this work is development of multiple stage optical fiber displacement sensors for the deformation monitoring of shafts and horizontal tunnels at the Phase 2 of Horonobe Underground Research Laboratory project. In previous year, the feasibility study was done for developing the multiple stage optical fiber displacement sensors of ten stages or more for the boreholes which diameter are 66 mm. Based on this feasibility study, two types of prototype for field measurement were developed and evaluated by laboratory tests. These prototypes were modified and field test equipments were developed which were consisted of 4 pieces of shallow part sensors and 2 pieces of deep part sensors.
Kashiwai, Yoshio*; Daimaru, Shuji*; Sanada, Hiroyuki; Matsui, Hiroya
JAEA-Technology 2008-047, 77 Pages, 2009/07
This work was performed by Taisei Kiso Sekkei Co., Ltd. under contract with Japan Atomic Energy Agency. The objective of this work is development of multiple stage optical fiber displacement sensors for the long-term deformation monitoring in rock mass of shafts and horizontal tunnels at the Phase 2 of Horonobe Underground Research project. The field test was done for installation to the borehole drilling in bedrock and continuous observation of one month. A prototype of explosion-proof fusion splice was also made for setting work in the flammable circumstance.
Shinozaki, Shinichi; Honda, Atsushi; Oshima, Katsumi; Shimizu, Tatsuo; Numazawa, Susumu*; Ikeda, Yoshitaka
JAEA-Technology 2008-048, 23 Pages, 2008/07
The modification of the JT-60U to a fully superconducting coil tokamak, JT-60SA, has been programmed as the satellite devise for the ITER and as the national centralized tokamak. The present positive-ion-based NBI system, which has employed the expensive CAMAC and has been operated for 20 years, is required to extend its pulse duration from 30 s to 100 s for JT-60SA. Recently, the frequency of troubles on the data acquisition system has increased due to its age-induced deterioration. To realize the long pulse operation and to maintain the high reliability on JT-60SA, we set to develop a new acquisition system. As a first step, we have designed and constructed a prototype acquisition system, which is combined with instruments highly available on the market, to confirm the basic performance. The result indicates that the new system allows us to construct a highly flexible and user-friendly acquisition system at low cost without highly technical software developing.
Noto, Katsuya; Usui, Katsutomi; Kawai, Mikito; Ikeda, Yoshitaka
JAEA-Technology 2008-049, 23 Pages, 2008/06
Water-cooled bleeder resistor, utilized in acceleration power supply on the positive-ion-based NBI system, has been designed to realize 100 s injections of intense neutral beams in JT-60 Super Advanced. The design is progressed with minimizing modification of existing electric parts. Exsiting water-cooled bleeder resistors is composed of three water vessels connected in parallel, in each of which 150 resistors of 600 are immersed and connected in series. Although the manufacturing company requires entire replacement of water-cooled bleeder, the careful assessment of thermal load allows only the replacement of the resistors inside the water vessel. The resistance of one resistors is required to be increased from 600 to 2.5 k. The total resistance of the bleeder is 140 k that includes the resistance of water. In the operation with the bleeder of 140 k, stable production of 85 keV, 55 A D beams, allowing 2 MW of the designed injection power, was confirmed without instabilization of the acceleration power supply. This modification significantly reduces the cost and the manufacturing time.
Tomita, Kenji; Inoue, Shuichi; Ishida, Takuya; Onuma, Yuichi; Tsuchiya, Kunihiko
JAEA-Technology 2008-050, 41 Pages, 2008/07
Blanket Functional Facility (BFT) for fusion blanket development was established in the Japan Materials Testing Reactor (JMTR). The irradiation tests with LiTiO pebble-bed were carried out with the BFT. The BFT was constituted a sweep gas device for tritium measurement and recover and a capsule controlled device for temperature control and neutron flux measurement of LiTiO pebble-bed. Five tritium monitors (ion chambers) for tritium measurement were established in the sweep gas device. In these tritium monitors, one tritium monitor for the measurement of tritium release property (TmIRA201) was not able to be used and it is necessary to exchange new tritium monitors. This report is described the fabrication of new tritium monitors and exchange procedure of this monitors.
Maruyama, Yoichiro; Wakaida, Ikuo
JAEA-Technology 2008-051, 13 Pages, 2008/07
Plasma emission characteristics of copper were studied by using Laser-Induced Breakdown Spectroscopy (LIBS). The intensity of plasma emission depended on the species of atmospheric gases, and the strongest plasma emission was obtained in the Ar atmosphere. And it was observed that the intensity reached its maximum at 1-2 microseconds after the ablation and decreased. The spectrum broadening due to Stark effect was observed and the spectral width varied with the observed time and the atmospheric gases, and the narrowest spectral width was obtained in He atmosphere. The plasma temperature calculated from spectral intensities reached around 10,000 K at 1-2 microseconds after the ablation and increased with increasing ablation laser energy.
Nakamura, Masahiko; Matsuda, Makoto; Nakanoya, Takamitsu; Kabumoto, Hiroshi; Kutsukake, Kenichi; Otokawa, Yoshinori; Asozu, Takuhiro; Ishii, Tetsuro; Asai, Masato; Okura, Takehisa
JAEA-Technology 2008-052, 17 Pages, 2008/07
Alpha rays (0.06 Bq/cm) were detected on the first basement passage of the Tandem Accelerator Facility. We attempted to identify the alpha-emitting nuclide with surveying this area. We found a crack of concrete at the detected spot. We inferred that radon may seep out of the crack from several phenomena observed in the survey of this area. Furthermore, we identified all nuclides emitting rays at the crack as Naturally Occurring Radioactive Materials, (NORM: K, Bi, Ac and so on) by in-situ -ray measurement. Radon is not easy to be collected and detected by its nature of noble gas. We have developed a convenient electrostatic collection method: after collection of the radon decay product (Po) using a Teflon sheet charged electrostatically, we have measured alpha rays and rays from this sheet. We concluded the leakage of radon from the soil caused a high counting-rate of alpha rays. The paper describes our survey works and identification procedures in detail.
Usui, Katsutomi; Noto, Katsuya; Kawai, Mikito; Oga, Tokumichi*; Ikeda, Yoshitaka
JAEA-Technology 2008-053, 35 Pages, 2008/08
The JT-60 negative ion-based NBI (N-NBI) system is required to extend the pulse duration from 30s to 100s in JT-60SA that is the modified JT-60U with full superconducting coils. The JT-60SA N-NBI system will have 2 ion sources, each of which will inject 5 MW at 500 keV. The present power supply system should be upgraded to operate for 100s with minimizing the modification of existing components. The protective characteristic and thermal capacities of the power supply components were assessed based on the experience of the modification for the 30s operation in 2003. The acceleration power supply is to be modified with combination of existing Gate Turnoff Thyristors (GTO) and Injection Enhanced Gate Transistors (IEGT) added newly. Five power supplies for a plasma production in the negative ion sources are to be modified by increasing the capacities of the partial resistance and cooling systems. These modifications can allow the long pulse operation of 100 for JT-60SA N-NBI system.
Kishi, Toshiaki; Motohashi, Jun; Yamamoto, Kazuyoshi; Kumada, Hiroaki; Torii, Yoshiya
JAEA-Technology 2008-054, 99 Pages, 2008/08
JRR-4 had carried out modification works for the purpose of reducing the enrichment level of fuel. About utilization facilities, followings were installed new neutron beam facility, renewal irradiation facility that was modified pneumatic irradiation facility for activation analysis of short-lived nuclides. This report describes the characteristic measurement by initial core and equilibrium core in 2001 by renewal JRR-4. Utilization facilities had been identified equal performance before modify about neutron flux and cadmium ratio on 1998 and 2001. And we have achieved less than 5% of irradiation uniformity at N-pipe. The maximum neutron flux is about 2.210ms at the New neutron beam facility and the maximum neutron flux is about 110ms at the prompt -ray analysis facility got good quality performance for medical irradiation and fundamental examination of it.
Ishizaki, Nobuhiro; Matsuda, Makoto; Hanashima, Susumu; Nakamura, Masahiko; Kutsukake, Kenichi; Otokawa, Yoshinori; Asozu, Takuhiro
JAEA-Technology 2008-055, 24 Pages, 2008/08
The JAEA-Tokai folding-type tandem accelerator has been operated for 25 years. There are many optical devices at the terminal, such as a 180-degeree bending magnet (charge-analysis electromagnet) and electrostatic quadrupole lens, used for transporting ion beams from low-energy accelerating tubes to high-energy accelerating tubes. We have replaced the aged coil inside the 180-degeree bending magnet for a maintenance period of 2007. At this time, we have carried out the project to align the whole optical devices at the terminal. Precise measurement of the positions for these devices revealed that they were misaligned by 5 mm at the maximum deviation. By aligning these devices with the standard beam axis, we have improved the transmission of ion beams and increased beam currents with stable operation.
Iyatomi, Yosuke; Ogata, Nobuhisa; Sugihara, Kozo; Seko, Noriaki; Hoshina, Hiroyuki; Okada, Kenji*; Tamada, Masao
JAEA-Technology 2008-056, 12 Pages, 2008/08
The concentrations of fluorine (7.29.5 mg/L) and boron (0.81.5 mg/L) dissolved in groundwater pumped from shafts during excavation at the Mizunami Underground Research Laboratory (MIU), Tono Geoscience Centre, are reduced to the levels below the environmental standards (fluorine: 0.8 mg/L, boron: 1 mg/L) at a water treatment facility. Coagulation treatment and ion exchange treatment are applied for fluorine and boron respectively. Consequently, we have started research on efficient groundwater treatment for fluorine and boron using radiation-induced graft polymerization adsorbent. Regarding the treatment for boron, over 95% of boron has been removed from groundwater volume of 760 times greater than the volume of the adsorbent. With respect to the fluorine removal, 95% of fluorine has been removed from groundwater volume of 320 times greater than the volume of the adsorbent. As over 90% of fluorine must be removed from the groundwater in order to meet the environmental standard, the treatment method for fluorine using radiation-induced graft polymerization adsorbent is less efficient than for boron and will need further improvement. Therefore, we are planning to perform a durability evaluation and recycling test of adsorbent using improved testing equipment for enhancing the efficiency.
Sekita, Kenji; Furusawa, Takayuki; Emori, Koichi; Ishii, Taro*; Kuroha, Misao; Hayakawa, Masato; Ouchi, Hiroshi
JAEA-Technology 2008-057, 45 Pages, 2008/08
A carbon steel used 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.
Yokokura, Kenji; Shimono, Mitsugu; Suzuki, Sadaaki; Sawahata, Masayuki; Igarashi, Koichi; Wada, Kenji; Moriyama, Shinichi
JAEA-Technology 2008-058, 103 Pages, 2008/08
This report summarized the studies on the disassemble work of radio frequency heating system in the torus hall as a preparation for the construction of JT-60SA (super advanced) which is the upgrade of the present large tokamak, JT-60U. The studies of the disassembly work were done with emphasis on the safety management because the work requires treatment of contaminated material with tritium and radiated material by neutron, and (1) object for disassemble, (2) work plan, (3) estimation of materials amount, and (4) procedure were summarized.
Sagawa, Jun; Moriyama, Kiyofumi; Nishikizawa, Tomotoshi; Nakamura, Hideo
JAEA-Technology 2008-059, 43 Pages, 2008/09
Electro-chemical electrodes including pH probes, ion-selective electrodes (ISEs) etc. generally have very high output impedances. In order to measure their outputs with generic measurement devices like data recorders, we need impedance conversion amplifiers that convert the ultra-high impedance signals of the probes into low-impedance input signals for ordinary measurement devices. Although specially designed measurement devices for the electro-chemical probes are commercially available, there are very few products that can be applied for multi-channel time series data acquisition. Thus, we designed and fabricated an ultra-high impedance low-offset amplifier fit for this purpose. The primary specification of the amplifier is, input impedance 10G, input range 1V, gain 120, response time about 1s, output range 10V, output impedance 50, and it has 5 independent channels. This report describes the originally developed design, selection of the element devices, test on the circuit characteristics, and instruction for fabrication.
Dai, H.*; Hajima, Ryoichi
JAEA-Technology 2008-060, 26 Pages, 2008/11
For future advanced energy recovery linac to generate femtosecond X-ray pulses, precise synchronization between sub-systems is highly desired. Typical synchronization methods based on direct photo detection are limited by detector nonlinearities, which lead to amplitude-to-phase conversion and introduce excess timing jitter. In this paper, we experimentally demonstrate an optical-electronic mixed phase lock loop to synchronize the RF signal and laser pulses. In this synchronism setup, a Sagnac-loop Mach-Zehnder interferometer has been used to suppress the excess noise of direct photo detection. This scheme transfers the timing information into a intensity imbalance between the two output beams of the interferometer. As experimental demonstration, the single side-band phase noise of RF signal from the VCO is locked to the mode-locked Ti:Sapphire laser in the spectrum covering the range of 10 kHz to 1 MHz. This synchronization scheme greatly reduces the phase noise and timing jitter of the RF signal.
Mita, Yutaka; Matsumura, Toshihiro; Yokoyama, Kaoru; Sugitsue, Noritake
JAEA-Technology 2008-061, 35 Pages, 2008/10
In Ningyo-toge Environmental Engineering Center. The equipments and radioactive waste which were contaminated with uranium are generated so much in future dismantling stage. In our plan, some of equipments and radioactive waste are decontamination to a clearance level, and cut down on decommission and disposal expense. This plan needs the alpha-rays measurement technology of the very low level. We think that ionized Air transfer measurement technology is promising as of clearance verification technology. The ionized Air transfer measurement technology applied to the Ionized Air Type Measurement can measure alpha radioactivity of a very low level. Moreover, as compared with a direct survey, there is the merit which can be measured in a short time. However ionized Air transfer measurement technology is new technology. Therefore, there is almost no measurement track record. Furthermore, the date about the influence of a background, a detection limit, measurement performance, and reliability is insufficient. So, this measurement test estimated applicability as clearance level verification of an Ionized Air Type Measurement.
Kondo, Masaaki; Kimishima, Satoru*; Emori, Koichi; Sekita, Kenji; Furusawa, Takayuki; Hayakawa, Masato; Kozawa, Takayuki; Aono, Tetsuya; Kuroha, Misao; Ouchi, Hiroshi
JAEA-Technology 2008-062, 46 Pages, 2008/10
The reactor containment of HTTR is tested to confirm leak-tight integrity of itself. "Type A test" has been conducted in accordance with the standard testing method in JEAC4203 since the preoperational verification of the containment was made. Type A tests are identified as basic one for measuring containment leakage rate, it costs much, however. Therefore, the test program for HTTR was revised to adopt an efficient and economical alternatives including "Type B and Type C tests". In JEAC4203-2004, following requirements are specified for adopting alternatives: upward trend of leakage rate by Type A test due to aging should not be recognized; criterion of combined leakage rate with Type B and Type C tests should be established; the criteria for Type A test and combined leakage rate test should be satisfied; correlation between the leakage rates by Type A test and combined leakage rate test should be recognized. Considering the performances of the tests, the policies of corresponding to the requirements were developed, which were accepted by the regulatory agency. This report presents an outline of the tests, identifies issues on the conventional test and summarizes the policies of corresponding to the requirements and of implementing the tests based on the revised program.
Yoshinaka, Kazuyuki; Takano, Yugo*; Kimura, Yukihiko*; Sugaya, Atsushi; Onizawa, Toshikazu
JAEA-Technology 2008-063, 135 Pages, 2008/10
This paper is reported that result of leaching tests for bituminized waste and Waste solidified by epoxy resin, done from 2003 to 2006. We've get precious knowledge and data, as follows. (1) In leaching tests for bituminized waste, it has detected iodine-129 peak, considered difficult too low energy to detect. We've get data and knowledge of iodine-129 behavior first. Leached radioactivity for 50 days calculated by peak area was equal for about 40% and 100% of including radioactivity in bituminized waste sample. And we've get data of behavior of nitric acid ion and so on, important to study for disposal, in various condition of sample shape or leaching liquid temperature. (2) In leaching test for waste solidified by epoxy resin, we've get data of behavior of TBP, radionuclides and so on.
Komeda, Masao; Yamamoto, Kazuyoshi; Yagi, Masahiro; Sagawa, Hisashi
JAEA-Technology 2008-064, 77 Pages, 2008/10
We investigated the irradiation method to irradiate 12 inch NTD silicon uniformly in JRR-3, where 6 inch NTD silicon is being irradiated at present, by using MVP of the Monte Carlo calculation code. In the case of irradiating 12 inch NTD silicon, the deviation of the doping distribution in the radial direction becomes 1.17 by the same irradiation method of 6 inch NTD silicon. Therefore the thermal neutron filter was introduced for uniform doping (the deviation is less than 1.10) in the radial direction and the effect was analyzed. As the result, it was indicated that the deviation of the doping distribution in the radial direction became less than 1.1 by using the neutron filter, which was made from aluminum alloy of 2 mm thickness including natural boron of 1%.
Yokokura, Kenji; Shimono, Mitsugu; Hasegawa, Koichi; Sawahata, Masayuki; Suzuki, Sadaaki; Terakado, Masayuki; Hiranai, Shinichi; Igarashi, Koichi; Sato, Fumiaki; Wada, Kenji; et al.
JAEA-Technology 2008-065, 98 Pages, 2008/10
Construction of the JT-60SA (super advanced) is planned as an upgrade of JT-60U as the satellite tokamak in ITER broader approach and as the national centralized tokamak facility program in Japan. The present JT-60U will be disassembled and the JT-60SA will be constructed at the same location in the JT-60 tours hall. The disassembly work will be planned in the period from 2009 to 2011. In this report, disassembly of the radio frequency heating system of JT-60U in the amplifier rooms and heating power supply building is studied on (1) object for disassembly, (2) work plan, (3) estimation of materials amount, (4) procedure.
Maeda, Shingo*; Hirano, Takahiro*; Shimada, Taro; Nakayama, Shinichi
JAEA-Technology 2008-066, 35 Pages, 2008/10
Bulk in-situ -spectroscopy is effective for a slightly and uniformly contaminated surface such as a room surrounded by concrete walls. The time-consuming scoping scanning survey for the entire surface is essential to ensure the slight and uniform contamination prior to the bulk in-situ measurement. However, the scoping scanning survey is omissible if the conservative procedure is acceptable. The count rate, cps, for the material of interest can be obtained by in-situ Ge detector will be converted to the radioactivity using conversion factor, Bq/cps, which depends on the distance from the detector to the furthest point. The radioactive concentration, Bq/g, is evaluated by dividing the radioactivity by the "measurement unit" of 100 kg. This procedure could certainly produce a conservative value. If the value obtained by this procedure is lower than the regulated clearance level, the material of interest can be cleared without the prior scoping scanning survey.
Satomi, Shinichi; Kanayama, Fumihiko; Hagiya, Kazuaki; Myodo, Masato; Kobayashi, Tadayoshi; Tomii, Hiroyuki; Tachibana, Mitsuo
JAEA-Technology 2008-067, 53 Pages, 2008/10
Dismantling activities of equipments in JAERI's Reprocessing Test Facility (JRTF) started from 1996 as a part of decommissioning of this facility. The large liquid waste storage tank LV-2 is scheduled to remove out as a whole tank without cutting in pieces from the annex building B to confirm safety and efficiency of this method from 2006. Before removal of the LV-2 tank, some preparatory works were carried out such as opening of concrete wall (LV-2 room) for the entrance of workers and materials, removal of pipes connected to the LV-2 tank, and decontamination of radioactive sludge in the LV-2 tank. Useful data were collected on manpower, radiation control and waste amount through the preparatory works, and work efficiency was analyzed by use of these data. It was compared manpower between core boring and hand-breaker crushing activities in the concrete wall opening work. It was also confirmed that local exposure of worker could be reduced in large extent by an addition of vinyl chloride cover on worker's ventilated suit.
Takai, Toshihide; Nakagiri, Toshio; Inagaki, Yoshiyuki
JAEA-Technology 2008-068, 63 Pages, 2008/10
A new experimental apparatus by the thermo-chemical and electrolytic ybrid ydrogen production in ower emperature range (HHLT) was developed and hydrogen production experiment was performed to confirm the system operability. Hydrogen production efficiency was estimated and technical problems were clarified through the experimental results. Stable operation of the SO electrolysis cell and the sulfur dioxide solution electrolysis cell were confirmed during experimental operation and any damage which would be affected solid operation was not detected under post operation inspection. To improve hydrogen production efficiency, it was found that the reduction of sulfuric acid circulation and the decrease in the cell voltage were key issues.
Kawai, Mikito; Akino, Noboru; Ikeda, Yoshitaka; Ebisawa, Noboru; Honda, Atsushi; Kazawa, Minoru; Kikuchi, Katsumi; Mogaki, Kazuhiko; Noto, Katsuya; Oshima, Katsumi; et al.
JAEA-Technology 2008-069, 32 Pages, 2008/10
The neutral beam injection system for JT-60U consists of positive-ion based type(P-NBI) and negative-ion based type(N-NBI). The reionization losses of neutral beams in the drift ducts of both P-NBI and N-NBI are estimated using the data of ambient pressure and gas flow rate into the beamlines. This system was not enough to obtain detail injection power for a long pulse operation. Modifications of the system to obtain reionization loss for a long pulse operation have been conducted. The new system has a capability to measure the pressures of drift duct during operation. The system can calculate the reionization loss automatically during the pulse from the measured pressure. More acurate injection power can be obtained by this new system.
The ad hoc Working Group of Investigating into the Crack on a Reflector Element of JRR-4
JAEA-Technology 2008-070, 121 Pages, 2008/09
A crack was ascertained on a weld area of one reflector element on December 28, 2007. The Department of Research Reactor and Tandem Accelerator set up an ad hoc working group of experts in the JAEA, and investigated cause of crack on the weld area. The following examinations were carried out; visual examination, dimensional examination, fractography examination and so on. It was concluded that the main cause of the crack is the swelling of graphite in the reflector element. The swelling must be due to neutron irradiation. We carried out a radiografical examination of the other reflector elements. As the result, we determined that many of them were not in a suitable state to be used because of swelling of graphite. The design of the new reflector elements should be carried out, based on the relation between the irradiation dose and swelling rate, which has been obtained in these investigation.
Seito, Hajime; Haruyama, Yasuyuki; Hanaya, Hiroaki; Yamagata, Ryohei; Kanazawa, Takao; Kaneko, Hirohisa; Kojima, Takuji
JAEA-Technology 2008-071, 29 Pages, 2008/11
[This article is unavailable to download the full text due to various reasons.]This report outlines useful data for users in electron beam and Co -ray irradiation facilities at JAEA, TARRI. The contents include fundamental data such as characteristics of irradiation field, mechanism of irradiator and dose measurement in irradiated materials.
Yagi, Masahiro; Horiguchi, Hironori; Yokoo, Kenji; Oyama, Koji; Kusunoki, Tsuyoshi
JAEA-Technology 2008-072, 79 Pages, 2008/09
A crack had been found on the weld of one reflector element in JRR-4. A survey revealed that the cause for the crack was the expansion of graphite reflector in the reflector element. It appeared that the expansion of graphite reflector was caused by fast neutron irradiation at low temperature. The survey confirmed radiographically that graphite reflectors in the other reflector elements without the crack expanded similarly by the irradiation growth. Irradiated graphite reflectors were carefully observed and were precisely measured the three dimensions after dismantling the irradiated reflector elements in order to understand quantitatively the irradiation growth behavior of IG-110 graphite under the JRR-4 operation condition. As the results, it was confirmed that growth of graphite reflectors increased with increasing of fast neutron fluence. The maximum irradiation growth per fast neutron fluence was 7.1310%m/n, the minimum was 4.2110%m/n, the average was 5.7110%m/n in the range of fast neutron fluence below 2.510n/m.
Yasuda, Atsushi; Ueta, Shohei; Aihara, Jun; Takeuchi, Hitoshi; Sawa, Kazuhiro
JAEA-Technology 2008-073, 18 Pages, 2008/11
As the conventional SiC-coated fuel particle, the ZrC-coated particle is proposed as a nuclear fuel for the Very High Temperature Reactor (VHTR) which is one of Generation IV nuclear reactors. Therefore it is examined by ZrC-coating equipment to get a ZrC-coating condition of C/Zr ratio 1.0, e.g., Zr and C atomic ratio equal to 1:1. Raw materials as surrogated kernel are Stabilized Zirconium Oxide (SZR) particle and PyC-coated SZR particle. For getting the basic production technology for mass production, the ZrC-coating parametric exanimation (coating gas flow rate, coating temperature and so on) is done up to 100 g as the equipment inventory. As the result of parameter examination, finally it could make the ZrC-coated particle with a thickness of ZrC layer of 0.030 mm and high quality in quantity of the particle inventory 100 g.
Tanaka, Nobuyuki; Suwa, Hirokazu*; Furukawa, Tomohiro; Inagaki, Yoshiyuki
JAEA-Technology 2008-074, 20 Pages, 2008/12
The thermochemical hydrogen production IS process utilizes corrosive chemicals such as sulfuric acid and hydriodic acid. Corrosion tests in IS process environments have been carried out to get the corrosion data of materials. In the corrosion test in sulfuric acid at 400C, the leak of sulfuric acid was observed in a pipe connected with a reflux condenser. The cause of the leakage is an important significant knowledge for the operation of the test apparatus. Therefore the cause was investigated. A through hole was detected in the pipe around the welding bead. By visual observation after cutting the pipe, the wall thickness of the pipe became thin at the inside welding bead around the through hole. In addition, EPMA showed that the inhomogeneous distribution of the constituent elements of the pipe was observed around the through hole. For these reasons, it is estimated that the lowering of the corrosion resistance by the sensitization at the welding caused the leakage.
Kimura, Hideo; Aoyagi, Tetsuo; Sakai, Manabu; Sato, Taiichi; Tsuji, Minoru
JAEA-Technology 2008-075, 32 Pages, 2008/11
JAEA developed the ERP (Enterprise Resource Planning) system at the establishment in 2005, aiming to support and enhance its business-critical task such as financial accounting and contract management. We considered the conceptual design of the next ERP system, and we implemented the prototype system to validate its effectiveness. Moreover, we implemented the simple add-on tool for rapid and easy development. At the result, we gauged the future prospects that the XML-centric system which we designed will offer high modularity, flexibility, connectivity between other systems, independence among subsystems. The simple add-on tool also demonstrated its effectiveness.
Tanzawa, Sadamitsu; Hiroki, Seiji; Abe, Tetsuya; Shimizu, Katsusuke*; Inoue, Masahiko*; Watanabe, Mitsunori*; Iguchi, Masashi*; Sugimoto, Tomoko*; Inohara, Takashi*; Nakamura, Junichi*
JAEA-Technology 2008-076, 99 Pages, 2008/12
The primary vacuum pumping system of the International Thermonuclear Experimental Reactor (ITER) exhausts a helium (He) ash resulting from the DT-burn with excess DT fueling gas, as well as performing a variety of functions such as pump-down, leak testing and wall conditioning. A mechanical based vacuum pumping system has some merits of a continuous pumping, a much lower tritium inventory, a lower operational cost and easy maintenance, comparing with a cryopump system, although demerits of an indispensable magnetic shield and insufficient performance for hydrogen (H) pumping are well recognized. To overcome the demerits, we newly fabricated and tested a helical grooved pump (HGP) unit suitable for H pumping at the ITER divertor pressure of 0.1-10 Pa. Through this R&D, we successfully established many design and manufacturing databases of large HGP units for the lightweight gas pumping. Based on the databases, we conceptually designed the ITER vacuum pumping system mainly comprising the HGP with an optimal pump unit layout and a magnetic shield. We also designed conceptually the reduced cost (RC)-ITER pumping system, where a compound molecular pump combining turbine bladed rotors and helical grooved ones was mainly used. The ITER mechanical based primary pumping system proposed has eventually been a back-up solution, whereas a cryopump based one was formally selected to the ITER for construction.
Nakamura, Hirofumi; Nagai, Toshihisa; Suto, Toshiyuki; Kosaka, Ichiro; Nakazaki, Katsutoshi; Suto, Shinya; Nakamura, Tomotaka; Nakabayashi, Hiroki; Hayashi, Naoto; Sumida, Daisaku
JAEA-Technology 2008-077, 276 Pages, 2008/12
Japan Atomic Energy Agency (JAEA) has been conducting "Fast Reactor Cycle Technology Development Project (FaCT Project)" for the purposes of researching and developing the technologies for the fast breeder reactor cycle commercialization since Japanese fiscal year (JFY) 2007. Based on the above R&D program for reprocessing system, the engineering-scale hot test would provide demonstration data on the specification, operation and maintenance of the adapted innovative technologies, system and plant. And more, these results would be fed to the design of the demonstration facility planning on the FaCT project road map. This report is the interim report of design studies about the engineering-scale hot test facility and includes not only design of the equipment and facility, but also consideration for design principle, requirements and facility basic planning.
Onuma, Yuichi; Ishida, Takuya; Sakata, Ikuma*; Kodaira, Akira*; Sakai, Jun*; Oba, Seiichiro*; Kanno, Masaru; Saito, Takashi; Kinase, Muneyuki*; Ishitsuka, Etsuo
JAEA-Technology 2008-078, 39 Pages, 2008/12
Decommissioning of the water loop irradiation facility polluted by fission products and cruds was studied, and the reasonable waste classification occurring by the decommissioning was also studied. A out-pile equipment of the irradiation loop facility installed in JMTR is considered as a decommissioning object. Measurement of ambient dose rate in the out-pile facility and evaluation of the deposited radionuclide concentration in the cooling pipe were carried. In result, it was clear that the significant radionuclide is Co, and that occurred waste can classify as the shallow-ground trench disposal level, clearance level, non-radioactive waste. Furthermore, through the investigation of the cutting method for minimizing secondary waste generation, plumbing cutting machine with preventing scattering function was developed by trial manufactured cutter that surrounds the cutting pipe by box.
Hazawa, Tomoya; Tamura, Itaru; Takazawa, Hiromitsu
JAEA-Technology 2008-079, 76 Pages, 2009/01
The Department of Research Reactor and Tandem Accelerator has planned the project of increasing cold neutron intensities by ten times by upgrading the cold neutron guide tubes with super-mirrors also and the moderator cell of cold neutron source. An exchange of the guide tubes requires taking the beam shutter of about 6 ton out of the narrow guide tunnel. The cold neutron guide tubes in the front end are 3 times longer then thermal ones and set very close to the reactor core. Therefore, the work has to be done under a high radiation dose as well as paying close attention to setting the tubes precisely. This report describes the result of study on replacing the Ni-guide tubes with super-mirror guide tubes as the first step of the 10-times upgrading project for cold neutrons.
Tochio, Daisuke; Shinohara, Masanori; Fujimoto, Nozomu
JAEA-Technology 2008-080, 56 Pages, 2009/01
The HTTR is block-type high-temperature gas-cooled reactor composed graphite-block piled-up structure. In the viewpoint of core structure, it is very difficult to measure fuel temperature of the HTTR directory. Therefore, power distribution is calculated by nuclear characteristics estimation code, and the fuel temperature is estimated by fuel temperature estimation code with obtained power distribution data in the HTTR. This report describes the estimation results of power distribution and fuel temperature for 850 C operation through a burn-up period.
Furukawa, Tomohiro; Yoshida, Eiichi
JAEA-Technology 2008-081, 25 Pages, 2009/01
High temperature hardness measurement for the steam generator tube made of Mod.9Cr-1Mo has been performed in the temperature range from R.T. to 1200C in inert gas and in vacuum, as one of the basic experiments for the establishment of the mechanism analysis of sodium - water reaction. The major results are as follows. (1) The hardness of the material was decreased with the increasing of the testing temperature. In the temperature range of 800 to 900C, stagnation of the value caused by Alfa - phase transformation was observed, and then the value was decreased with the increasing of the temperature again. The effect of examination atmosphere was not observed under this research. (2) Good correlation between hardness and tensile strength was observed in the temperature range from R.T. to 1200C. (3) It is considered that the knowledge about the strength of the steam generator tube in high temperature can be reflected as base data to the solution of a wastage mechanism.
Shimizu, Atsushi; Nabeshima, Kunihiko; Nakagawa, Shigeaki
JAEA-Technology 2008-082, 44 Pages, 2009/01
The High temperature engineering test reactor (HTTR) executed the rated power driving for 30 days of the first time (850C in temperature of the nuclear reactor exit 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 is 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 are 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 combustion of the nuclear fuel for the rated power driving. 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.
Yasuda, Atsushi; Ueta, Shohei; Aihara, Jun; Ishibashi, Hideharu*; Sawa, Kazuhiro
JAEA-Technology 2008-083, 11 Pages, 2009/01
The Very-High-Temperature Reactor (VHTR) is one of the candidates for the Generation IV nuclear energy system. ZrC coated fuel particles are expected to make the performance of the VHTR higher. Therefore, we are investigating the ZrC-coating process. From April 2007 to March 2008, ZrC-outer pyrolytic carbon (OPyC) layer continuous coating tests were carried out with the nonnuclear particles and we succeeded to coat continuously the ZrC layer and the OPyC layer with the thicknesses up to about 27 and about 48 m, respectively, in the batch scale of 100 g.
Takahashi, Kuniaki; Meguro, Yoshihiro; Kawato, Yoshimi; Kuroda, Kazuhiko*; Ogawa, Naoki*
JAEA-Technology 2008-084, 12 Pages, 2009/02
Low level liquid waste discharged from a Reprocessing Facility includes sodium nitrate. In the case that it is directly solidified with cement and so on and then the solidified waste are disposed under the ground, sodium nitrate soaks into the environment through underground water layer. We planned to apply the biological treatment system that many ordinary industrial plants are running in the field of waste water treatment to reduce nitrate. We carried out degradation experiments of nitrate for 4wt% sodium nitrate solution by biological method. To solve the assignments that biological treatment technology has, we tested and obtained the results as shown below; (1) The amount of sludge ash could be cut down a tenth as much as usual. The disposal cost reduction of secondary waste is just in sight. (2) Treatment performance could be improved up to 7 kg-N/m/d from 4 kg-N/m/d. It could be expected the more compact system by improvement of the membrane set into the biological treatment tanks.
Matsui, Kunihiro; Nabara, Yoshihiro; Nunoya, Yoshihiko; Koizumi, Norikiyo; Okuno, Kiyoshi
JAEA-Technology 2008-085, 7 Pages, 2009/02
PF Insert Coil is a single layer solenoid coil using the superconducting conductor for ITER Poloidal field coil. The stability test of the conductor will be performed under the magnetic field. In this test, the inductive heating by the inductive heater attached on the conductor will be applied to originate initial normal zone in the conductor. Since the conductor for the PF Insert Coil is the cable-in-conduit conductor, it is quite difficult to estimate the inductive heating energy theoretically. Thus, the inductive heating energy is experimentally evaluated by the calorimetric method. The proportional constants of the inductive heating generated in the conductor, cable and conduit are evaluated at 0.129 [J/As], 0.019 [J/As] and 0.109 [J/As], respectively.
Mozumi, Yasuhiro; Ueta, Shohei; Aihara, Jun; Sawa, Kazuhiro
JAEA-Technology 2008-086, 16 Pages, 2009/02
Fuel for the Very High Temperature Reactor (VHTR) is required to be used under severer irradiation conditions and higher operational reactor temperatures than those of present high temperature gas cooled reactors. Japan Atomic Energy Agency has developed the advanced silicon carbide (SiC) -coated fuel particles having thicker layer thicknesses, and zirconium carbide (ZrC)-coated particles that are expected to preserve their integrity at higher temperatures and burnup conditions than current conventional coated fuel particles. These particles have been fabricated successfully in order to perform irradiation tests at experimental reactors. This paper is summarized fabrication data of irradiation samples.
Sagawa, Jun; Shibamoto, Yasuteru
JAEA-Technology 2008-087, 34 Pages, 2009/03
In experimental facilities to investigate a system integral-response and/or to verify fuel rod integrity of nuclear reactors, the electrical heater specially manufactured to simulate the real nuclear fuel rod with the same scale have been used in the core of the experimental facility. This type of the electrical heater, so-called "simulated fuel rod", is a kind of a sheath heater which involves Nichrome coiled wire as a heat generation element in the metal cladding tube. An alternating current power is supplied for heat generation source in this heater and thin thermocouples were embedded on the cladding surface to measure the fuel surface temperature. It means that a switching regulator by silicon-controlled rectifier is used to control the AC electrical power and undesirable electrical noises are superimposed on the thermocouples' signals by the time variation of the heater current. Although a low-pass-filter with a low cut-off frequency was commonly applied to remove the noises in the previous steady-state experiment, the problem have arisen that a fast temperature transient could not be followed due to a time-delay accompanied by the filter in the transient experiment.
Kondo, Keitaro; Ochiai, Kentaro; Kutsukake, Chuzo; Konno, Chikara
JAEA-Technology 2008-088, 90 Pages, 2009/03
In order to utilize the d-D neutron source with a titanium deuteride target of the FNS facility for fusion neutronics researchs, the specification of the d-D neutron source was investigated. The characteristic of neutrons produced by the d-D reaction was described based on the reaction kinematics and the target assembly of the accelarator was detailedly modeled for the MCNP calculation. In order to validate the MCNP model, the angular distribution of the neutron strength was measured with the foil activation method. The measured reaction rates were well predicted by the MCNP calculation and the validity of the present model was confirmed. Based on the model, a MCNP source term was prepared for usual analysis calculations of experiments with DD neutrons.
Yamamoto, Kazuya
JAEA-Technology 2008-089, 80 Pages, 2009/03
The first "evacuation drill with family car" in Japan had been conducted by the Ibaraki Prefecture in the Ibaraki Prefecture Comprehensive Nuclear Disaster Prevention Exercise on September 30, 2008. In this work, dynamic traffic flow of vehicles of evacuees was analyzed using three kinds of data, that is, questionnaires for the participants of the evacuation drill, GPS tracking data of vehicles of evacuees, and an aerial video record by the Police of Ibaraki Prefecture. An opinion poll was also conducted to the participants of the evacuation drill to gain insight into how the residents respond in evacuation with their own cars in a nuclear disaster.
Yamada, Keisuke; Okoshi, Kiyonori; Saito, Yuichi; Orimo, Takao*; Omae, Akiomi*; Yamada, Naoto*; Mizuhashi, Kiyoshi
JAEA-Technology 2008-090, 95 Pages, 2009/03
The 400 kV ion implanter at the Takasaki Ion Accelerators for Advanced Radiation Application facility (TIARA) provides ion beams for various experiments of research and development (R&D's) mainly on materials science and biotechnology using Freeman ion sources. Two methods of ionization are generally used to a Freeman ion source. In one method, a sample gas is directly fed to the plasma chamber, and in the other, vapor of a solid material having high vapor pressure is vaporized by an oven and fed. Those methods, however, can supply limited number of ion species of those required from various R&D's. We have developed new methods available to materials which are difficult to vaporize by an oven: the disc method for high melting point materials, the SF plasma method for high melting point and low vapor pressure materials, the filament method for metals with melting point higher than 2400 C, and one of them is chosen according to the nature of a material. Forty four ion species from hydrogen to bismuth are generated by using the Freeman ion sources to date. Furthermore, a small ECR (electron cyclotron resonance) ion source is also developed to generate multiply charged ions for acceleration to higher energies.
Dairaku, Masayuki; Watanabe, Kazuhiro; Tobari, Hiroyuki; Kashiwagi, Mieko; Inoue, Takashi; Sakamoto, Keishi; Hanada, Masaya; Akino, Noboru; Ikeda, Yoshitaka; Yamamoto, Takumi*
JAEA-Technology 2008-091, 23 Pages, 2009/03
A plasma generator whose inner dimensions are 25 cm in width, 59 cm in length, and 31 cm in depth for a high power and long pulse ion source in neutral beam injector has been designed and fabricated. The plasma generator has a beam extraction area of 12 cm in width and 46 cm in length. A target of the output beam using the plasma generator is to produce deuterium positive ion beams up to 120 keV, 65 A for longer than 200 s pulses. Arrangement of the permanent magnets and filaments has been designed by using an electron trajectory simulation code to produce uniform and high density plasma with high proton yield. Cooling channels have been also designed to operate the long pulse plasma generation with a 100 kW arc discharge power.
Kitagishi, Shigeru; Inaba, Yoshitomo; Tsuchiya, Kunihiko; Ishitsuka, Etsuo
JAEA-Technology 2008-092, 37 Pages, 2009/03
In the internal environment of nuclear reactors, it is considered that the stress corrosion cracking (SCC) of a structural material is caused with the dissolved oxygen and hydrogen peroxide produced by radiation degradation of water. Therefore, it is necessary to measure the concentrations of these corrosive chemical species in the cooling water for assessing health and lifetime extension of the structural material. The in-situ water analyzer has been developed for the concentration measurement of corrosive chemical species. This device is composed of the spectrometer and chemical sensor, and the concentrations are evaluated by the intensity of the absorption and luminescence of the chemical species in the reactor water. In this study, the preliminary test of the in-situ water analyzer was carried out under the normal temperature and pressure, and optical transmission through the optical fiber was measured with the guide tube.
Ohashi, Kazutaka*; Nishihara, Tetsuo; Tazawa, Yujiro; Tachibana, Yukio; Kunitomi, Kazuhiko
JAEA-Technology 2008-093, 56 Pages, 2009/03
As HTGR hydrogen production systems, such as HTTR-IS system or GTHTR300C currently being developed by Japan Atomic Energy Agency, consists of nuclear reactor and chemical plant, which are without a precedent in the world, safety design philosophy and regulatory framework should be newly developed. In this report, phenomena to be considered and events to be postulated in the safety evaluation of the HTGR hydrogen production systems were investigated and basic principles to establish acceptance criteria for the explosion and toxic gas release accidents were provided. Especially for the explosion accident, quantitative criteria to the reactor building are proposed with relating sample calculation results. It is necessary to treat abnormal events occurred in the hydrogen production system as external events to the nuclear plant in order to classify the hydrogen production system as no-nuclear facility and basic policy to meet such requirement was also provided.
Takagi, Takashi*; Makita, Taiyo*; Kunimoto, Eiji; Shibata, Taiju; Sawa, Kazuhiro
JAEA-Technology 2008-094, 22 Pages, 2009/03
To advance the performance and safety of High Temperature Gas-cooled Reactors (HTGRs), it is expected to use heat-resistant materials substitute for metallic materials in the core internal structural components of HTGRs at higher temperature. Carbon fiber reinforced carbon-carbon composite (C/C composite) is one of candidates as heat-resistant materials, and investigations are under going to apply the core internal structural components such as the control rod for the Very High Temperature Reactor (VHTR). This report describes the experimental results of irradiation effects on dimensional change, coefficient of thermal expansion and elastic modulus obtained by PIE (Post-Irradiation Examinations) for C/C composite irradiated in 03M-47AS capsule at the Japan Materials Testing Reactor (JMTR).