Unusual low-temperature ductility increase mediated by dislocations alone

Naeem, M.*; Ma, Y.*; Tian, J.*; Kong, H.*; Romero-Resendiz, L.*; Fan, Z.*; Jiang, F.*; Gong, W.; Harjo, S.; Wu, Z.*; et al.

Materials Science & Engineering A, 924, p.147819_1 - 147819_10, 2025/02

https://doi.org/10.1016/j.msea.2025.147819

Data report of ROSA/LSTF experiment SB-PV-03; 0.2% pressure vessel bottom break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2024-014, 76 Pages, 2024/12

JAEA-Data-Code-2024-014.pdf:4.0MB

An experiment denoted as SB-PV-03 was conducted on November 19, 2002 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-03 simulated a 0.2% pressure vessel bottom small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system of emergency core cooling system (ECCS) and noncondensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 55 K/h in the primary system was initiated 10 min after the generation of a safety injection signal, and continued afterwards. Auxiliary feedwater injection into the secondary-side of both SGs was started for 30 min with some delay after the onset of the AM action. The AM action was effective on the primary depressurization until the ACC tanks began to discharge nitrogen gas into the primary system. The core liquid level recovered in oscillative manner because of intermittent coolant injection from the ACC system into both cold legs. Therefore, the core liquid level remained at a small drop. The pressure difference between the primary and SG secondary sides became larger after nitrogen gas ingress. Core uncovery occurred by core boil-off during reflux condensation in the SG U-tubes under nitrogen gas influx. When the maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 908 K, the core power was automatically reduced to protect the LSTF core. After the automatic core power reduction, coolant injection from low pressure injection (LPI) system of ECCS into both cold legs led to the whole core quench. After the continuous core cooling was confirmed through the actuation of the LPI system, the experiment was terminated.

neutron diffraction study to elucidate hydrogen effect on the deformation mechanism in Type 310S austenitic stainless steel

Ito, Tatsuya; Ogawa, Yuhei*; Gong, W.; Mao, W.*; Kawasaki, Takuro; Okada, Kazuho*; Shibata, Akinobu*; Harjo, S.

Proceedings of the 7th International Symposium on Steel Science (ISSS 2024), p.237 - 240, 2024/11

https://doi.org/10.2355/isijisss.2024.0_237

France-Japan collaboration on severe accident studies in sodium-cooled fast reactors, 2; Methodologies and calculations of severe accident phases

Sogabe, Joji; Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Kamiyama, Kenji; Onoda, Yuichi; Matsuba, Kenichi; Yamano, Hidemasa; Kubo, Shigenobu; Kubota, Ryuzaburo*; et al.

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

In the frame of France-Japan collaboration, the calculational methodologies were defined and assessed, and the phenomenology and the severe accident consequences were investigated in a pool-type sodium-cooled fast reactor.

Interaction of solute manganese and nickel atoms with dislocation loops in iron-based alloys irradiated with 2.8 MeV Fe ions at 400 C

Nguyen, B. V. C.*; Murakami, Kenta*; Chena, L.*; Phongsakorn, P. T.*; Chen, X.*; Hashimoto, Takashi; Hwang, T.*; Furusawa, Akinori; Suzuki, Tatsuya*

Nuclear Materials and Energy (Internet), 39, p.101639_1 - 101639_9, 2024/06

https://doi.org/10.1016/j.nme.2024.101639

Comparison on safety features among HTGR's Reactor Cavity Cooling Systems (RCCSs)

Takamatsu, Kuniyoshi; Funatani, Shumpei*

Nuclear Engineering and Technology, 56(3), p.832 - 845, 2024/03

https://doi.org/10.1016/j.net.2023.09.036

The objectives of this study are as follows: to understand the characteristics, degree of passive safety features for heat removal were compared for RCCSs based on atmospheric radiation and based on atmospheric natural circulation under the same conditions. Therefore, the authors concluded that the proposed RCCS based on atmospheric radiation has the advantage that the temperature of the RPV can be stably maintained against disturbances in the outside air (ambient air). Moreover, methodology to utilize all the heat emitted from the RPV surface for increasing the degree of waste-heat utilization was discussed.

Hierarchical deformation heterogeneity during Lders band propagation in an Fe-5Mn-0.1C medium Mn steel clarified through scanning electron microscopy

Koyama, Motomichi*; Yamashita, Takayuki*; Morooka, Satoshi; Yang, Z.*; Varanasi, R. S.*; Hojo, Tomohiko*; Kawasaki, Takuro; Harjo, S.

Tetsu To Hagane, 110(3), p.205 - 216, 2024/02

https://doi.org/10.2355/tetsutohagane.TETSU-2023-052

Impact of interatomic structural characteristics of aluminosilicate hydrate on the mechanical properties of metakaolin-based geopolymer

Kim, G.*; Cho, S.-M.*; Im, S.*; Suh, H.*; Morooka, Satoshi; Shobu, Takahisa; Kanematsu, Manabu*; Machida, Akihiko*; Bae, S.*

Construction and Building Materials, 411, p.134529_1 - 134529_18, 2024/01

https://doi.org/10.1016/j.conbuildmat.2023.134529

Data report of ROSA/LSTF experiment TR-LF-15; Accident management actions during station blackout transient with pump seal LOCA

Takeda, Takeshi

JAEA-Data/Code 2023-012, 75 Pages, 2023/10

JAEA-Data-Code-2023-012.pdf:4.45MB

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.

Investigations and consideration on conditions of contamination and measures of decontamination for motor vehicles at a nuclear emergency

Togawa, Orihiko; Hokama, Tomonori; Hiraoka, Hirokazu

JAEA-Review 2023-013, 48 Pages, 2023/08

JAEA-Review-2023-013.pdf:2.11MB

When radionuclides are released into the atmospheric environment at a nuclear emergency, protective measures such as evacuation and temporal relocation are carried out using motor vehicles such as private cars and buses to reduce radiation exposure to residents. To confirm conditions of contamination for the evacuated or relocated residents, contamination inspection is conducted, in which it is important not to spoil its rapidity. In the present inspection, wipers and tires are designated to first measuring parts, and they are basically inspected by persons using GM survey meters. Utilization of portable radiation portal monitors is also being considered for rapid and efficient inspection of motor vehicles. In order to contribute to rapid and efficient operation of contamination inspection, this report investigated conditions of contamination and measures of decontaminations for motor vehicles at a nuclear emergency. Although available documents and information were quite few, results of the investigations described in the related documents were extracted and rearranged according to the objectives of this report. Furthermore, these results were considered from a viewpoint of rapid and efficient operation of contamination inspection.