JAEA-Data/Code 2023-012, 75 Pages, 2023/10
An experiment denoted as TR-LF-15 was conducted on June 11, 2014 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment TR-LF-15 simulated accident management (AM) actions during a station blackout transient with TMLB' scenario with pump seal loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). This scenario is featured by loss of auxiliary feedwater functions. The pump seal LOCA was simulated by a 0.1% cold leg break. The test assumptions included total failure of both high pressure injection system and low pressure injection system of emergency core cooling system (ECCS). Also, it was presumed non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. When steam generator (SG) secondary-side collapsed liquid level dropped to a certain low liquid level, the primary pressure turned to rise. After the SG secondary-side became voided, the safety valve of a pressurizer cyclically opened which led to loss of primary coolant. Core uncovery thus took place owing to core boil-off at high pressure. When an increase of 10 K was confirmed in cladding surface temperature of simulated fuel rods, SG secondary-side depressurization was started as the first AM action. At that time, the safety valves in both SGs were fully opened. Primary depressurization was initiated by completely opening the pressurizer safety valve as the second AM action with some delay after the first AM action onset. When the SG secondary-side pressure lowered to 1.0 MPa following the first AM action, water was injected into the secondary-side of both SGs via feedwater lines with low-head pumps as the third AM action. A reduction in the primary pressure was accelerated because the heat removal from the SG secondary-side system resumed shortly after the third AM action initiation.
JAEA-Data/Code 2023-007, 72 Pages, 2023/07
An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment IB-HL-01.
Sawaguchi, Takuma; Takai, Shizuka; Sasagawa, Tsuyoshi; Uchikoshi, Emiko*; Shima, Yosuke*; Takeda, Seiji
MRS Advances (Internet), 8(6), p.243 - 249, 2023/06
In the intermediate depth disposal of relatively high-level radioactive waste, a method to confirm whether the borehole for monitoring is properly sealed should be developed in advance. In this study, groundwater flow analyses were performed for the hydrogeological structures with backfilled boreholes, assuming sedimentary rock area, to understand what backfill design conditions could prevent significant water pathways in the borehole, and to identify the confirmation points of borehole sealing. The results indicated the conditions to prevent water pathways in the borehole and BDZ (Borehole Disturbed Zone), such as designing the permeability of bentonite material less than or equal to that of the host rock, and grouting BDZ.
Nemoto, Takahiro; Arakawa, Ryoki; Kawakami, Satoru; Nagasumi, Satoru; Yokoyama, Keisuke; Watanabe, Masashi; Onishi, Takashi; Kawamoto, Taiki; Furusawa, Takayuki; Inoi, Hiroyuki; et al.
JAEA-Technology 2023-005, 33 Pages, 2023/05
During shut down of the HTTR (High Temperature engineering Test Reactor) RS-14 cycle, an increasing trend of filter differential pressure for the helium gas circulator was observed. In order to investigate this phenomenon, the blower of the primary helium purification system was disassembled and inspected. As a result, it is clear that the silicon oil mist entered into the primary coolant due to the deterioration of the charcoal filter performance. The replacement and further investigation of the filter are planning to prevent the reoccurrence of the same phenomenon in the future.
Ogata, Sho*; Yasuhara, Hideaki*; Aoyagi, Kazuhei; Kishida, Kiyoshi*
Proceedings of 53rd US Rock Mechanics/Geomechanics Symposium (USB Flash Drive), 6 Pages, 2019/06
Noguchi, Hiroki; Takegami, Hiroaki; Kamiji, Yu; Tanaka, Nobuyuki; Iwatsuki, Jin; Kasahara, Seiji; Kubo, Shinji
International Journal of Hydrogen Energy, 44(25), p.12583 - 12592, 2019/05
JAEA has been conducting R&D on thermochemical water-splitting hydrogen production IS process to develop one of heat applications of high-temperature gas-cooled reactor. A test facility was constructed using corrosion-resistant industrial materials to verify integrity of the IS process components and to demonstrate continuous and stable hydrogen production. The performance of components installed in each section was confirmed. Subsequently, a trial operation of integration of the processing sections was successfully carried out for 8 hours with hydrogen production rate of approximately 10 NL/h. After that, hydrogen production operation was extended to 31 hours (approximately hydrogen production rate of 20 NL/h) by introducing a corrosion-resistance pump system with a developed shaft seal technology.
Jo, Mayumi*; Ono, Makoto*; Nakayama, Masashi; Asano, Hidekazu*; Ishii, Tomoko*
Geological Society Special Publications, 482, 16 Pages, 2018/09
Tokuyasu, Kayoko; Furuta, Sadaaki*; Kokubu, Yoko; Umeda, Koji
Nihon Hoshasen Anzen Kanri Gakkai-Shi, 15(1), p.80 - 87, 2016/07
An optically stimulated luminescence reader (Riso TL/OSL DA-20) was installed in Toki Research Institute of Isotope Geology and Geochronology (Toki-shi, Gifu Prefecture), Japan Atomic Energy Agency (JAEA) for dating the geological sample. An accumulated dose of the sample is obtained using the reader. Sealed beta source of strontium-90 is required to be mounted on the reader because repeated artificial irradiation is necessary for the accumulated dose estimation. However, there are not many introduction examples for the reader domestically, and the information as to radiation control of the reader is limited. We therefore report here the process of source loading on the reader and radiation control associated with the use of the source.
JAEA-Data/Code 2015-022, 58 Pages, 2016/01
The SB-HL-12 test simulated PWR 1% hot leg SBLOCA under assumptions of total failure of HPI system and non-condensable gas (nitrogen gas) inflow. SG depressurization by fully opening relief valves in both SGs as AM action was initiated immediately after maximum fuel rod surface temperature reached 600 K. After AM action due to first core uncovery by core boil-off, the primary pressure decreased, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before LSC induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after nitrogen gas inflow. Third core uncovery by core boil-off occurred during reflux condensation. The maximum fuel rod surface temperature exceeded 908 K.
Takeda, Takeshi; Otsu, Iwao
Journal of Energy and Power Sources, 2(7), p.274 - 290, 2015/07
JAEA-Data/Code 2014-021, 59 Pages, 2014/11
Experiment SB-CL-32 was conducted on May 28, 1996 using the LSTF. The experiment SB-CL-32 simulated 1% cold leg small-break LOCA in PWR under assumptions of total failure of HPI system and no inflow of non-condensable gas from ACC tanks. Secondary-side depressurization of both SGs as AM action to achieve the depressurization rate of 200 K/h in the primary system was initiated 10 min after break. Core uncovery started with liquid level drop in crossover leg downflow-side. The core liquid level recovered rapidly after first LSC. The surface temperature of simulated fuel rod then increased up to 669 K. Core uncovery took place before second LSC induced by steam condensation on ACC coolant. The core liquid level recovered rapidly after second LSC. The maximum fuel rod surface temperature was 772 K. The continuous core cooling was confirmed because of coolant injection by LPI system. This report summarizes the test procedures, conditions and major observation.
Hiratsuka, Hajime; Ichige, Hisashi; Kizu, Kaname; Honda, Masao; Miya, Naoyuki
JAERI-Tech 2002-076, 37 Pages, 2002/10
no abstracts in English
Isozaki, Futoshi*; Kikuchi, Taiji; Ioka, Ikuo; Ishikawa, Kazuyoshi; Hirata, Yuji*
JAERI-Tech 2002-074, 22 Pages, 2002/09
The pressurized tube specimens which enclosed high pressure inert gas were produced for the irradiation creep test. The pressurized tube specimen with 7mm outer diameter and 0.5mm wall thickness must be sealed by the welding, after the helium gas was impressed in the inside of tube. In this process, there was a technical problem of welding under high pressure, and it is difficult to seal the pressurized tube specimen in the present facility of our group. The production process was examined by taking shortening in production period and reduction in the cost into consideration. The sealing technology to enclose the helium gas up to 5.5MPa was established by new technique using the present facility and the mock-up test. And, it is necessary to measure the outer diameter of the pressurized tube specimen with high accuracy in order to predict irradiation creep deformation arising from neutron radiation and internal pressure. Therefore, the method for measuring at the 0.01mm measurement accuracy was established, which combined laser measuring instrument with the lathe.
Matsui, Yoshinori; Kikuchi, Taiji; Kakuta, Tsunemi
JAERI-Tech 2002-060, 23 Pages, 2002/08
Recently, the number of in-situ tests using optical fibers as signal transmission lines is increasing in the JMTR, because of their noise-tolerant capability and recent development of irradiation proof type fibers. The convensional seal methods, however, have the problem of trust when they applied to the penetration of the irradiation capsule using optical fibers. As the solution of the problem, we selected soldering seal method that was already applied to the deep-sea optical fiber cables and, as its modification, the silver alloy brazing method. From the result of the out-of-pile test, it was confirmed that both soldering seal and silver alloy brazing seal would be applicable to the plug seal of irradiation capsules. Then the soldering seal was applied to a capsule with optical fibers and was irradiated for 5 operation cycles of the JMTR (about 3000 hours). No deterioration was found at the seal after irradiation. This suggested that in-core use of optical fiber was thought to be promising.
Fujimoto, Nozomu; Takada, Eiji*; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatsuo
JAERI-Tech 2001-090, 69 Pages, 2002/01
HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate showed higher results than expected value at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in a core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses.
Matsui, Yoshinori; Ide, Hiroshi; Itabashi, Yukio; Kikuchi, Taiji; Ishikawa, Kazuyoshi; Abe, Shinichi; Inoue, Shuichi; Shimizu, Michio; Iwamatsu, Shigemi; Watanabe, Naoki*; et al.
KAERI/GP-195/2002, p.33 - 40, 2002/00
Studies on the irradiation damage of the material of the RPV are inevitable for the LWR. Recently, the researches of annealing effect on the irradiation damage of RPV material were extensively carried out using specimens irradiated in the JMTR of the JAERI. As the next step, an annealing test of irradiated specimens and re-irradiation of annealed specimens were planned. The aim of the test is to evaluate the effect of annealing by comparing the damage of irradiated specimen, its recovery by annealing and the damage after re-irradiation. For the re-irradiation test of this study, JAERI developed a new capsule in which the specimens can be exchanged before and after annealing, and, re-irradiated afterward. The development of the capsule consisted of the design and fabrication of airtight connector for thermocouples and mechanical seal device which was fit to remote handling. Remote operation procedures for handling the radioactive capsule and for exchanging specimens were carefully performed. The results of the re-irradiation proved that the development was technically successful.
Gulden, W.*; Cook, I.*; Marbach, G.*; Raeder, J.*; Petti, D.*; Seki, Yasushi
Fusion Engineering and Design, 51-52(Part.B), p.419 - 427, 2000/11
no abstracts in English
Suzuki, Mitsuhiro; Anoda, Yoshinari
JAERI-Tech 2000-016, p.173 - 0, 2000/03
no abstracts in English
Nakahira, Masataka; Kakudate, Satoshi; Oka, Kiyoshi; Takeda, Nobukazu; *; *; *; Tada, Eisuke; Shibanuma, Kiyoshi; T.Burgess*; et al.
Fusion Technology, 34(3), p.1160 - 1164, 1998/11
no abstracts in English