Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Shimazaki, Yosuke; Yamazaki, Kazunori; Iigaki, Kazuhiko
Hozengaku, 18(1), p.16 - 20, 2019/04
no abstracts in English
Honda, Yuki; Inaba, Yoshitomo; Nakagawa, Shigeaki; Yamazaki, Kazunori; Kobayashi, Shoichi; Aono, Tetsuya; Shibata, Taiju; Ishitsuka, Etsuo
JAEA-Technology 2017-013, 20 Pages, 2017/06
JAEA-Technology-2017-013.pdf:2.52MB
Decay heat is one of an important factor for a safety evaluation of depressurized loss-of-forced cooling accident, a representative high consequence accident, in high temperature gas-cooled reactor (HTGR). Traditionally, a conservative decay heat curve is used for safety analysis according to the regulatory standards. On the other hand, there is growing interest in obtaining test data related to decay heat for the use of uncertainty analysis. However, such data has not been obtained for prismatic-type HTGR. Therefore, we have launched a test program to obtain the decay heat data from the HTTR. As an initial step, an applicability confirmation test of decay heat evaluation method for HTGR was conducted in February 2017 without non-nuclear heating condition. This report introduces an estimation method for the decay heat based on test data using HTTR and shows the results of validation of the reactor residual heat evaluation method which will be used to obtain the decay heat data based on test data.
Sawahata, Hiroaki; Shimazaki, Yosuke; Ishitsuka, Etsuo; Yamazaki, Kazunori; Yanagida, Yoshinori; Fujiwara, Yusuke; Takada, Shoji; Shinozaki, Masayuki; Hamamoto, Shimpei; Tochio, Daisuke
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 8 Pages, 2016/06
In the HTTR, 
Cf is loaded in the reactor core as a neutron startup source and changed at frequency. In this exchange work, there were two technical issues; slightly higher radiation exposure of workers by neutron leakage and reliability of neutron source transportation container in handling. To reduce the radiation dose by neutron leakage, detail numerical evaluations using PHITS code were carried out, the effective shielding method for fuel handling machine was proposed. Easily removable polyethylene blocks and particles were used as the neutron shielding, and installed in the cooling paths of the fuel handling machine. As a result, the collective effective dose by neutron was reduced from about 700 man-microSv to about 300 man-microSv. As to the neutron source transportation container, the handling performance was improved and the handling work was safety accomplished by downsizing.
Yamazaki, Kazunori; Kameyama, Yasuhiko; Inoi, Hiroyuki; Arakaki, Etsushi; Shinozaki, Masayuki; Ota, Yukimaru
JAEA-Testing 2008-002, 52 Pages, 2008/03
JAEA-Testing-2008-002.pdf:15.54MB
The High Temperature Engineering Test Reactor (HTTR) has the Gaseous Radwaste Treatment System (GRTS). This system appropriately collects all potentially radioactive gases discharged from the plant. After the gases are decayed with the Decay tank and decreased with the Filtering system, the gases are discharged into the atmosphere under monitoring. This system is maintained every year for keeping the performance. The maintenance is very important. Furthermore, the maintenance is profitable for designing a new High Temperature Gas cooled Reactor. This report describes the newly developed, maintenance items and improvements of the GRTS.
Tochio, Daisuke; Kameyama, Yasuhiko; Shimizu, Atsushi; Inoi, Hiroyuki; Yamazaki, Kazunori; Shimizu, Yasunori; Aragaki, Etsushi; Ota, Yukimaru; Fujimoto, Nozomu
JAEA-Technology 2006-045, 43 Pages, 2006/09
JAEA-Technology-2006-045.pdf:5.97MB
The auxiliary component cooling water system (ACCWS) is one of the cooling system in High Temperature Engineering Test Reactor (HTTR) The ACCWS has the features not only many facilities cooling but also heat sink of the vessel cooling system which is one of the engineering safety features. Therefore, the ACCWS is required to satisfy the design criteria of heat removal performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the ACCWS cooling tower was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that ACCWS cooling tower has the required heat exchange performance in the design.
Hamamoto, Shimpei; Iigaki, Kazuhiko; Shimizu, Atsushi; Sawahata, Hiroaki; Kondo, Makoto; Oyama, Sunao; Kawano, Shuichi; Kobayashi, Shoichi; Kawamoto, Taiki; Suzuki, Hisashi; et al.
JAEA-Technology 2006-030, 58 Pages, 2006/03
JAEA-Technology-2006-030.pdf:10.69MB
During normal operation of High Temperature engineering Test Reactor (HTTR) in Japan Atomic Energy Agency (JAEA), the reactivity is controlled by the Control Rods (CRs) system which consists of 32 CRs (16 pairs) and 16 Control Rod Drive Mechanisms (CRDMs). The CR system is located in stand-pipes accompanied by the Reserved Shutdown System (RSS). In the unlikely event that the CRs fail to be inserted, the RSS is provided to insert B
C/C pellets into the core. The RSS shall be designed so that the reactor should be held subcriticality from any operation condition by dropping in the pellets. The RSS consists of BC/C pellets, hoppers which contain the pellets, electric plug, driving mechanisms, guide tubes and so on. In accidents when the CRs cannot be inserted, an electric plug is pulled out by a motor and the absorber pellets fall into the core by gravity. A trouble, malfunction of one RSS out of sixteen, occurred during a series of the pre-start up checks of HTTR on February 21, 2005. We investigated the cause of the RSS trouble and took countermeasures to prevent the issue. As the result of investigation, the cause of the trouble was attributed to the following reason: In the motor inside, The Oil of grease of the multiplying gear flowed down from a gap of the oil seal which has been deformed and was mixed with abrasion powder of brake disk. Therefore the adhesive mixture prevented a motor from rotating.
Isozaki, Kazunori; Saito, Takakazu; Sumino, Kozo; Yamazaki, Yuji*; Karube, Koji; Terano, Toshihiro; Sakaba, Hideo
JNC TN9410 2004-014, 172 Pages, 2004/06
JNC-TN9410-2004-014.pdf:48.0MB
This technical report describes MK-III function tests on the primary main cooling pump. MK-III function tests (SKS-1) before MK-III core configuration completed from October 17, 2001 to October 23, 2001. MK-III function tests (SKS-2) after MK-III core configuration completed from January 27, 2003 to February 13, 2003. The results of function tests were shown as follows; (1) The primary main pump was confirmed to do stable control on both CAS (cascade) mode and MAN (manual) mode in the flow control system. And also this was confirmed no-emanation trend both flow and revolution per minute against flow step response, too. (2) The main motor was shifted run-back operation in about 54 seconds after scram. Run-back operation of the main motor (A) was 167m3/h with 117 rpm. 185 m
/h with 118 rpm was the main motor (B). And they were controlled within the limit of run-back operational revolution 122 rpm 
8 rpm. The flow was confirmed to maintain in 10 percent and over of the rated flow. (3) Succeeding operation to the pony motor was confirmed to do in about 39 seconds after the primary main pump trip. The pony motor (A) operation was 180m3/h with 124rpm. 190 m/h with 123 rpm was the pony motor (B). They were enough satisfied with the forgiven revolution per minute which was 93 rpm and over. And the flow was confirmed to maintain in 10 percent and over of the rated flow. (4) Free flow coast down characteristic of the primary main pump was confirmed that time constant was 10 seconds at both the trip and run-back operation time of the primary main pump. (5) Over flow column sodium level of the main pump duty operation was NL-1,550 mm Na by column (A), NL-1,468mm Na by column (B). They were smaller than NL-1,581 mm Na by the design value. Pressure loss value of the new IHX had more conservative value than the design value. (6) The primary main pump was confirmed which the rated flow could be restored no-scram by the instantaneous power loss within 0.6 second.
