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.

On the velocity and frequency of disturbance waves in vertical annular flow with different surface tension and gas-liquid density ratio

Zhang, H.*; Umehara, Yutaro*; Yoshida, Hiroyuki; Mori, Shoji*

International Journal of Heat and Mass Transfer, 211, p.124253_1 - 124253_13, 2023/09

https://doi.org/10.1016/j.ijheatmasstransfer.2023.124253

Study on borehole sealing corresponding to hydrogeological structures by groundwater flow analysis

Sawaguchi, Takuma; Takai, Shizuka; Sasagawa, Tsuyoshi; Uchikoshi, Emiko*; Shima, Yosuke*; Takeda, Seiji

MRS Advances (Internet), 8(6), p.243 - 249, 2023/06

https://doi.org/10.1557/s43580-023-00528-y

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.

Free-surface flow simulations with floating objects using lattice Boltzmann method

Watanabe, Seiya*; Kawahara, Jun*; Aoki, Takayuki*; Sugihara, Kenta; Takase, Shinsuke*; Moriguchi, Shuji*; Hashimoto, Hirotada*

Engineering Applications of Computational Fluid Mechanics, 17(1), p.2211143_1 - 2211143_23, 2023/00

https://doi.org/10.1080/19942060.2023.2211143

In tsunami inundations or slope disasters of heavy rain, a lot of floating debris or driftwood logs are included in the flows. The damage to structures from solid body impacts is more severe than the damage from the water pressure. In order to study free-surface flows that include floating debris, developing a high-accurate simulation code of free-surface flows with high performance for large-scale computations is desired. We propose the single-phase free-surface flow model based on the cumulant lattice Boltzmann method coupled with a particle-based rigid body simulation. The discrete element method calculates the contact interaction between solids. An octree-based AMR (Adaptive Mesh Refinement) method is introduced to improve computational accuracy and time-to-solution. High-resolution grids are assigned near the free surfaces and solid boundaries. We conducted two kinds of tsunami flow experiments in the 15 and 70 m water tanks at Hachinohe Institute of Technology and Kobe University to validate the accuracy of the proposed model. The simulation results have shown good agreement with the experiments for the drifting speed, the number of trapped wood pieces, and the stacked angles.

Evolution of the reaction and alteration of granite with Ordinary Portland cement leachates; Sequential flow experiments and reactive transport modelling

Bateman, K.*; Murayama, Shota*; Hanamachi, Yuji*; Wilson, J.*; Seta, Takamasa*; Amano, Yuki; Kubota, Mitsuru*; Ouchi, Yuji*; Tachi, Yukio

Minerals (Internet), 12(7), p.883_1 - 883_20, 2022/07

https://doi.org/10.3390/min12070883

Numerical simulation of two-phase flow in fuel assemblies with a spacer grid using a mechanistically based method

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly with the spacer grid. The benchmark data of single-phase flow in the bundle with the spacers by KAERI were used to validate the simulation result by JUPITER. Moreover, the single-phase flow simulation was conducted by another simulation method, STAR-CCM+, as a supplemental analysis to consider the effect of the different simulation methods. Finally, the two-phase flow simulation for the bundle with the spacer was conducted by JUPITER. The effect of the spacer with a vane on the bubble behavior is discussed.

Long-term density-dependent groundwater flow analysis and its effect on nuclide migration for safety assessment of high-level radioactive waste disposal with consideration of interaction between fractures and matrix of rock formation in coastal crystalline groundwater systems

Park, Y.-J.*; Sawada, Atsushi; Ozutsumi, Takenori*; Tanaka, Tatsuya*; Hashimoto, Shuji*; Morita, Yutaka*

Proceedings of 3rd International Conference on Discrete Fracture Network Engineering (DFNE 2022) (Internet), 8 Pages, 2022/00

Safety analysis for underground disposal facilities for high-level radioactive waste requires thorough understanding of long-term groundwater flow and nuclide migration processes in geologic media. In the coastal subsurface systems, groundwater flow is defined by the complex interactions between freshwater of meteoric origin and denser saline water from the sea. In addition, sea levels are expected to fluctuate significantly due to a transgression and regression of the sea over the millions of years for safety analysis. This study presents long-term evolution of groundwater environment such as salinity concentration and flow velocity with focus of the interaction between fractures and matrix blocks in regional and near-field scale analysis framework for groundwater flow and nuclide migration for underground disposal facilities in hypothetical fractured crystalline coastal systems.

Flow regime and void fraction predictions in vertical rod bundle flow channels

Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 178, p.121637_1 - 121637_24, 2021/10

https://doi.org/10.1016/j.ijheatmasstransfer.2021.121637

Calculation of the amount of leaching water from concrete-pit facilities under various facility design conditions

Nagao, Rina; Namekawa, Maki*; Totsuka, Masayoshi*; Nakata, Hisakazu; Sakai, Akihiro

JAEA-Technology 2021-009, 139 Pages, 2021/06

JAEA-Technology-2021-009.pdf:13.96MB

Japan Atomic Energy Agency is the implementing body of the near surface disposal of low-level radioactive waste (LLW) generated from research facilities and other facilities. Concrete-pit disposal are considered as a method of disposing of the LLW. Since the concrete-pits are placed at deeper position than the groundwater level, we need to consider that radionuclides might migrate with the flow of groundwater. Accordingly, in order to explain the safety of the concrete-pit disposal facility, it is necessary to investigate the flow of groundwater and the volumetric flow rate of leaching water from the facility. Therefore, in this report, sensitivity analysis of the volumetric flow rate of leaching water from concrete-pit was carried out by varying the permeability of cover-soil filled with in outside of the lateral sides of the bentonite mixed soil (BMS) and the conditions of the BMS on the upper part of the concrete-pits. As a result of the analysis, when the BMS is normal condition, the volumetric flow rate of leaching water from the concrete-pits is reduced by lowering permeability of the lateral cover-soil. However, in the case of occurring the deterioration of the function of BMS on the upper part of the concrete-pit, significant reduction of the volumetric flow rate of leaching water is not seen even if the permeability of the lateral cover-soil is lowered. Therefore, taking into consideration the possibility of the deterioration of the function of BMS on the upper part of the concrete-pit, it is necessary to consider that cover-soil with low permeability is equipped on the upper part of the BMS.

Data report of ROSA/LSTF experiment SB-PV-09; 1.9% pressure vessel top small break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2021-006, 61 Pages, 2021/04

JAEA-Data-Code-2021-006.pdf:2.78MB

An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. 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 SB-PV-09.