Removal of penetrating pipes by dry wire-saw technology

Myodo, Masato; Kobayashi, Tadayoshi; Tomii, Hiroyuki

JAEA-Technology 2008-001, 46 Pages, 2008/03

JAEA-Technology-2008-001.pdf:19.05MB

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.

Preliminary analyses on hydrogen diffusion through small break of thermo-chemical IS process hydrogen plant

Somolova, M.*; Terada, Atsuhiko; Takegami, Hiroaki; Iwatsuki, Jin

JAEA-Technology 2008-003, 41 Pages, 2008/12

JAEA-Technology-2008-003.pdf:3.77MB

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.

Production of plastic scintillation survey meter for clearance verification measurement

Tachibana, Mitsuo; Shiraishi, Kunio; Ishigami, Tsutomu; Tomii, Hiroyuki

JAEA-Technology 2008-005, 33 Pages, 2008/03

JAEA-Technology-2008-005.pdf:2.1MB

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.

An Investigation on technical feasibilities of fuel cycle for high temperature gas-cooled reactor (Case study)

Sumita, Junya; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sawa, Kazuhiro

JAEA-Technology 2008-007, 23 Pages, 2008/03

JAEA-Technology-2008-007.pdf:12.74MB

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.

Safety analysis for severe accidents anticipated during truck transport of uranium fresh fuel and raw material (Contract research)

Nomura, Yasushi*; Takahashi, Satoshi*; Okuno, Hiroshi

JAEA-Technology 2008-009, 273 Pages, 2008/03

JAEA-Technology-2008-009.pdf:8.92MB

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.

Fabrication of irradiation capsule for IASCC irradiation tests, 1; Irradiation capsule for crack growth test (Joint research)

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

JAEA-Technology-2008-011.pdf:19.39MB

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.

Fabrication of irradiation capsule for IASCC irradiation tests, 3; Irradiation capsule for ECP sensor (Joint research)

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

JAEA-Technology-2008-013.pdf:17.96MB

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.

Conceptual design of irradiation experimental device for 12 inch NTD-Si ingot in JRR-4 (Contract research)

Yagi, Masahiro; Watanabe, Masanori; Oyama, Koji; Komeda, Masao; Yamamoto, Kazuyoshi; Kashima, Yoichi

JAEA-Technology 2008-015, 91 Pages, 2008/03

JAEA-Technology-2008-015.pdf:22.92MB

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.

Confirmation tests for fabrication of low density MOX pellet for FBR

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

JAEA-Technology-2008-017.pdf:2.76MB

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.

Examination on small-sized cogeneration HTGR for developing countries

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

JAEA-Technology-2008-019.pdf:8.59MB

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.

Development of the plasma movie database system for JT-60

Sueoka, Michiharu; Kawamata, Yoichi; Kurihara, Kenichi; Seki, Akiyuki

JAEA-Technology 2008-021, 23 Pages, 2008/03

JAEA-Technology-2008-021.pdf:10.47MB

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.

Improvement of efficiency on ball-cleaning method for JRR-3 heat exchanger

Ota, Kazunori; Ikekame, Yoshinori; Owada, Minoru; Fukushima, Manabu; Oba, Toshinobu; Takeuchi, Masaki; Imahashi, Masaki; Murayama, Yoji

JAEA-Technology 2008-023, 31 Pages, 2008/03

JAEA-Technology-2008-023.pdf:3.3MB

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.

Development of emergency information clearinghouse for nuclear emergency management

Watanabe, Fumitaka; Yamamoto, Kazuya; Sajiki, Kenjiro; Yasu, Sadanori*; Igarashi, Miyuki*

JAEA-Technology 2008-025, 63 Pages, 2008/03

JAEA-Technology-2008-025.pdf:2.71MB

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.

Design of control system of neutron beam shutters for 1MW spallation neutron source at Japan Proton Accelerator Research Complex (J-PARC)

Takada, Hiroshi; Kai, Tetsuya; Oikawa, Kenichi; Sakai, Kenji

JAEA-Technology 2008-027, 34 Pages, 2008/03

JAEA-Technology-2008-027.pdf:13.17MB

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.

Development of welding techniques for assembling of IASCC Test Capsule

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

JAEA-Technology-2008-029.pdf:25.78MB

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.

Initial design study of a DC power supply system for JT-60SA

Shimada, Katsuhiro; Omori, Yoshikazu; Okano, Jun; Matsukawa, Tatsuya; Terakado, Tsunehisa; Kurihara, Kenichi

JAEA-Technology 2008-031, 38 Pages, 2008/03

JAEA-Technology-2008-031.pdf:10.84MB

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).

Thermal hydraulic design of the mercury target vessel

Haga, Katsuhiro; Wakui, Takashi; Kogawa, Hiroyuki; Harada, Masahide; Futakawa, Masatoshi; Hayashi, Kenichi*; Nakamura, Koji*

JAEA-Technology 2008-033, 40 Pages, 2008/05

JAEA-Technology-2008-033.pdf:4.9MB

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.

Conceptual study of Mo production facility in JMTR

Iimura, Koichi; Hosokawa, Jinsaku; Kanno, Masaru; Kitajima, Toshio; Nakagawa, Tetsuya; Sakamoto, Taichi; Hori, Naohiko; Kawamura, Hiroshi

JAEA-Technology 2008-035, 47 Pages, 2008/06

JAEA-Technology-2008-035.pdf:7.91MB

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.

Technology development of system chemical decontamination using iodine heptafluoride gas; Evaluation of IF treatment condition and uranium decontamination result

Ema, Akira; Yokoyama, Kaoru; Nakatsuka, Yoshiaki; Shimaike, Masamitsu; Sugitsue, Noritake

JAEA-Technology 2008-037, 50 Pages, 2008/06

JAEA-Technology-2008-037.pdf:3.15MB

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.

Preliminary study for long life as beryllium reflector, 1; Fabrication of irradiation capsule and dismounted device of capsule

Hanawa, Yoshio; Taguchi, Taketoshi; Tsuboi, Kazuaki; Saito, Takashi; Ishikawa, Kazuyoshi; Watahiki, Shunsuke; Tsuchiya, Kunihiko

JAEA-Technology 2008-039, 53 Pages, 2008/06

JAEA-Technology-2008-039.pdf:6.18MB

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.

Preliminary study for long life as beryllium reflector, 2; Development of high accuracy size measurement device

Taguchi, Taketoshi; Hanawa, Yoshio; Watahiki, Shunsuke; Tsuchiya, Kunihiko

JAEA-Technology 2008-041, 23 Pages, 2008/06

JAEA-Technology-2008-041.pdf:3.76MB

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.

Development of evaluation method on irradiation temperature for JMTR capsule; Temperature valuation of CT specimen and separated specimen holder by NISA FEM code

Tomita, Kenji; Hosokawa, Jinsaku; Matsui, Yoshinori

JAEA-Technology 2008-043, 21 Pages, 2008/07

JAEA-Technology-2008-043.pdf:9.2MB

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.

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

Nojiri, Naoki; Owada, Hiroyuki; Kato, Yasushi

JAEA-Technology 2008-045, 38 Pages, 2008/06

JAEA-Technology-2008-045.pdf:6.95MB

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.

Development of optical fiber displacement sensor for bedrocks (The Third year)

Kashiwai, Yoshio*; Daimaru, Shuji*; Sanada, Hiroyuki; Matsui, Hiroya

JAEA-Technology 2008-047, 77 Pages, 2009/07

JAEA-Technology-2008-047.pdf:31.84MB

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.

Assessment of water-cooled bleeder resistors for long pulse injection in acceleration power supply on positive-ion-based NBI System

Noto, Katsuya; Usui, Katsutomi; Kawai, Mikito; Ikeda, Yoshitaka

JAEA-Technology 2008-049, 23 Pages, 2008/06

JAEA-Technology-2008-049.pdf:8.49MB

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.

Emission characteristics of copper plasma in laser-induced breakdown spectroscopy (Contract research)

Maruyama, Yoichiro; Wakaida, Ikuo

JAEA-Technology 2008-051, 13 Pages, 2008/07

JAEA-Technology-2008-051.pdf:2.07MB

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.

Modification of power supplies in the negative-ion-based NBI system for a 100s operation in JT-60SA

Usui, Katsutomi; Noto, Katsuya; Kawai, Mikito; Oga, Tokumichi*; Ikeda, Yoshitaka

JAEA-Technology 2008-053, 35 Pages, 2008/08

JAEA-Technology-2008-053.pdf:10.56MB

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.

Improvement the transmission of ion beams and replacement of charge analysis electromagnet coil at tandem accelerator

Ishizaki, Nobuhiro; Matsuda, Makoto; Hanashima, Susumu; Nakamura, Masahiko; Kutsukake, Kenichi; Otokawa, Yoshinori; Asozu, Takuhiro

JAEA-Technology 2008-055, 24 Pages, 2008/08

JAEA-Technology-2008-055.pdf:6.28MB

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.

Water chemistry control in HTTR

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

JAEA-Technology 2008-057, 45 Pages, 2008/08

JAEA-Technology-2008-057.pdf:12.0MB

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.

Design and fabrication of an ultra-high input impedance low offset amplifier for electro-chemical measurements

Sagawa, Jun; Moriyama, Kiyofumi; Nishikizawa, Tomotoshi; Nakamura, Hideo

JAEA-Technology 2008-059, 43 Pages, 2008/09

JAEA-Technology-2008-059.pdf:8.45MB

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.

Alpha radioactivity measurement technology with ionized air type measurement; Applicability evaluation to verification of the clearance level

Mita, Yutaka; Matsumura, Toshihiro; Yokoyama, Kaoru; Sugitsue, Noritake

JAEA-Technology 2008-061, 35 Pages, 2008/10

JAEA-Technology-2008-061.pdf:3.87MB

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.

Leaching test of bituminized waste and waste solidified by epoxy resin

Yoshinaka, Kazuyuki; Takano, Yugo*; Kimura, Yukihiko*; Sugaya, Atsushi; Onizawa, Toshikazu

JAEA-Technology 2008-063, 135 Pages, 2008/10

JAEA-Technology-2008-063.pdf:5.88MB

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.

Study on disassembly work of the radio frequency heating system in the amplifier rooms and heating power supply building

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

JAEA-Technology-2008-065.pdf:38.83MB

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.

Removal of the liquid waste storage tank LV-2 in JRTF, 1; Preparatory works

Satomi, Shinichi; Kanayama, Fumihiko; Hagiya, Kazuaki; Myodo, Masato; Kobayashi, Tadayoshi; Tomii, Hiroyuki; Tachibana, Mitsuo

JAEA-Technology 2008-067, 53 Pages, 2008/10

JAEA-Technology-2008-067.pdf:8.66MB

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.

Modification of the drift ducts reionization loss estimation for JT-60U NBI

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

JAEA-Technology-2008-069.pdf:6.75MB

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.

Technical data on electron beam and -ray irradiation facilities; Electron accelerator and Co irradiation facility

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.

Development of production technology of ZrC-coated particle, 1 (Contract research)

Yasuda, Atsushi; Ueta, Shohei; Aihara, Jun; Takeuchi, Hitoshi; Sawa, Kazuhiro

JAEA-Technology 2008-073, 18 Pages, 2008/11

JAEA-Technology-2008-073.pdf:3.81MB

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.

Conceptual design for the next JAEA's enterprise resource planning system

Kimura, Hideo; Aoyagi, Tetsuo; Sakai, Manabu; Sato, Taiichi; Tsuji, Minoru

JAEA-Technology 2008-075, 32 Pages, 2008/11

JAEA-Technology-2008-075.pdf:8.29MB

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.

Fast Reactor Cycle Technology Development Project (FaCT Project); A Design study on an engineering-scale hot test facility (Interim report)

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

JAEA-Technology-2008-077.pdf:25.66MB

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.

Study on the method of exchanging JRR-3 cold neutron guide tubes

Hazawa, Tomoya; Tamura, Itaru; Takazawa, Hiromitsu

JAEA-Technology 2008-079, 76 Pages, 2009/01

JAEA-Technology-2008-079.pdf:6.17MB

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.

Hardness of Mod.9Cr-1Mo tube at elevated temperature

Furukawa, Tomohiro; Yoshida, Eiichi

JAEA-Technology 2008-081, 25 Pages, 2009/01

JAEA-Technology-2008-081.pdf:24.45MB

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.

Development of production technology of ZrC-coated particle, 2 (Contract research)

Yasuda, Atsushi; Ueta, Shohei; Aihara, Jun; Ishibashi, Hideharu*; Sawa, Kazuhiro

JAEA-Technology 2008-083, 11 Pages, 2009/01

JAEA-Technology-2008-083.pdf:3.14MB

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.

Evaluation of inductive heating energy of a PF Insert Coil conductor by the calorimetric method (Contract research)

Matsui, Kunihiro; Nabara, Yoshihiro; Nunoya, Yoshihiko; Koizumi, Norikiyo; Okuno, Kiyoshi

JAEA-Technology 2008-085, 7 Pages, 2009/02

JAEA-Technology-2008-085.pdf:7.18MB

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.

Development of surface temperature measurement for simulated fuel rod heater; Noise reduction temperature signal of thin-thermocouple

Sagawa, Jun; Shibamoto, Yasuteru

JAEA-Technology 2008-087, 34 Pages, 2009/03

JAEA-Technology-2008-087.pdf:4.03MB

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.

The Evaluation of the "evacuation drill with family car" conducted in the Ibaraki Prefecture Comprehensive Nuclear Disaster Prevention Exercise 2008

Yamamoto, Kazuya

JAEA-Technology 2008-089, 80 Pages, 2009/03

JAEA-Technology-2008-089.pdf:1.64MB

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.

Development of a plasma generator for a high power NBI ion source

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

JAEA-Technology-2008-091.pdf:5.82MB

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.

Basic principles on the safety evaluation of the HTGR hydrogen production system

Ohashi, Kazutaka*; Nishihara, Tetsuo; Tazawa, Yujiro; Tachibana, Yukio; Kunitomi, Kazuhiko

JAEA-Technology 2008-093, 56 Pages, 2009/03

JAEA-Technology-2008-093.pdf:2.31MB

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.