Heat transfer coefficient modeling for downward saturated boiling flows in vertical pipes

Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 249, p.127219_1 - 127219_16, 2025/10

https://doi.org/10.1016/j.ijheatmasstransfer.2025.127219

Experimental study on light gas transport during containment venting by using the large-scale test facility CIGMA

Soma, Shu; Ishigaki, Masahiro*; Shibamoto, Yasuteru

Annals of Nuclear Energy, 219, p.111455_1 - 111455_12, 2025/09

https://doi.org/10.1016/j.anucene.2025.111455

Integrated thermal power measurement in the modified STACY for the performance inspections

Araki, Shohei; Aizawa, Eiju; Murakami, Takahiko; Arakaki, Yu; Tada, Yuta; Kamikawa, Yutaka; Hasegawa, Kenta; Yoshikawa, Tomoki; Sumiya, Masato; Seki, Masakazu; et al.

Annals of Nuclear Energy, 217, p.111323_1 - 111323_8, 2025/07

https://doi.org/10.1016/j.anucene.2025.111323

JAEA has modified the STACY from a homogeneous system using solution fuel to a heterogeneous system using fuel rods in order to obtain criticality characteristics of fuel debris. The modification of the STACY was completed in December 2023. A series of performance inspections were conducted for the start of experimental operations. A new thermal power calibration is required for the performance inspections in order to operate at less than 200 W, which is the permitted thermal power. However, the thermal power measurement method and calibration data used in the former STACY is no longer available due to the modification of the modified STACY. We measured the thermal power of the STACY using the activation method that was improved to adapt to the measurement condition and calibrated the power meter system. Since the positions where activation foils could be installed were very limited, the thermal power was evaluated using numerical calculations supplemented by experimental data. Neutron flux data at the positions of the activation foil was measured by the activation method. Neutron distribution in the core was calculated by the Monte Carlo code MVP. A response function of the activation foil was calculated using the PHITS. The uncertainty of the thermal power measurement was conservatively estimated to be about 15%. Four operations were conducted for the thermal power measurement. The power meter was calibrated by using three operational data and tested with the one operational data. It was found that the indicated value of the meter adjusted by the STACY before the modification work would tend to overestimate the actual output by about 40%. In addition, the current calibration was able to calibrate the meter to within 3% accuracy.

Input data preparation for PWR large-break LOCA analysis with RELAP5/MOD3.3 code

Takeda, Takeshi

JAEA-Data/Code 2025-005, 106 Pages, 2025/06

JAEA-Data-Code-2025-005.pdf:2.93MB

JAEA has been creating input data for pressurized water reactor (PWR) analysis with RELAP5/MOD3.3 code, mainly based on design information for the four-loop PWR's Tsuruga Power Station Unit-2 as the reference reactor of the Large Scale Test Facility (LSTF). The cold leg large-break loss-of-coolant accident (LBLOCA) calculation in the flamework of the BEMUSE program is cited as a representative OECD/NEA activity related to the PWR analysis. The new regulatory requirements for PWRs in Japan include the event of loss of recirculation functions from emergency core cooling system (ECCS) in the cold leg LBLOCA. This event should be evaluated the effectiveness of measures against severe core damage. The input data for this study were made preparations to analyze the PWR LBLOCA, which is one of the design basis accidents that should be postulated in the safety design. This report describes the main features of the input data for the PWR LBLOCA analysis. The PWR model comprised a reactor vessel, pressurizer (PZR), hot legs, steam generators (SGs), SG secondary-side system, crossover legs, cold legs, and ECCS. A four-loop PWR was simulated by two loops in the LBLOCA calculation. Specifically, loop-A attached with the PZR corresponded to three loops, and loop-B mounted with the break was equal to one loop. The nodalization schemes of the PWR components were referred to those of the LSTF components. Moreover, interpretations were added to the main input data for the PWR LBLOCA analysis, and further information such as the basis for determining the input data was provided. In addition, transient analysis was performed employing the prepared input data for the loss of ECCS recirculation functions event. The present transient analysis was confirmed to be appropriate generally by comparing with the calculation in the previous study using the RELAP5/MOD3.3 code. Furthermore, sensitivity analyses were executed exploiting the RELAP5/MOD3.3 code to clarify the effects of a discharge coefficient through the break and water injection flow rate of the alternative recirculation on the fuel rod cladding surface temperature. This report explains the results of the sensitivity analyses within the defined ranges, which complement some of the content of the previous study's calculation for the loss of ECCS recirculation functions event.

Improvement in automated particle measurement using micromanipulation and large geometry secondary ion mass spectrometry to remove the particle mixing effect of uranium particles

Tomita, Ryohei; Tomita, Jumpei; Suzuki, Daisuke; Miyamoto, Yutaka; Yasuda, Kenichiro

Journal of Nuclear Science and Technology, 10 Pages, 2025/05

https://www.tandfonline.com/doi/full/10.1080/00223131.2025.2512812

A new automated particle measurement (APM) combined with micromanipulation using large geometry secondary ion mass spectrometry instrument was proposed and demonstrated to remove the particle mixing effect, which indicated that the aggregation of uranium particles was detected as a single uranium particle, from APM results. The results showed that the new APM method was more effective than the traditional APM method in removing the particle mixing effect from the APM results and determining the existence of minor uranium isotopes in the samples.

Heat transfer characteristics of downward saturated boiling flow in vertical round pipes

Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 239, p.126598_1 - 126598_18, 2025/04

https://doi.org/10.1016/j.ijheatmasstransfer.2024.126598

Combustion properties of glove-box panel resins under fire accidents

Tashiro, Shinsuke; Uchiyama, Gunzo; Ono, Takuya; Amano, Yuki; Yoshida, Ryoichiro; Watanabe, Koji*; Abe, Hitoshi; Yamane, Yuichi

Nuclear Technology, 211(3), p.429 - 438, 2025/03

https://doi.org/10.1080/00295450.2024.2329832

Contributing to the confinement safety evaluation of glove-box (GB) connected with high efficiency particle air (HEPA) filters for radioactive materials under fire accidents, combustion tests of a flammable polymer, Polymethyl methacrylate (PMMA), and a flame retardant polymer, Polycarbonate (PC), as typical GB panel resins have been conducted with an engineering-scale combustion apparatus. The combustion properties such as the mass loss rate (MLR) and the heat release rate (HRR) of PMMA and PC were investigated in the combustion tests. In the tests with the same shapes, it was found the followings; MLRs and HRRs of PMMA were larger than those of PC under the same supply flow rate into the combustion cell (Fv); MLRs and HRRs of PMMA and PC were constant under different Fv. Moreover, in the tests of PMMA with different cross section areas (S), MLRs and HRRs were found to be proportional to S. Using these results, the relationships of MLR and HRR to S of PMMA and PC were deduced.

Renovation and restart of STACY (Static Experiment Critical Facility)

Sono, Hiroki

Robutsuri No Kenkyu (Internet), (78), 12 Pages, 2025/03

https://rpg.jaea.go.jp/else/rpd/annual_report/pdf78/No78-03.pdf

The Static Experiment Critical Facility (STACY) was renovated from a "solution fuel reactor" to a "reactor using fuel rods and light water moderator", and restarted operation on August 2, 2024, after a hiatus of 13 years and 8 months. During that time, it took 8 years and 11 months to obtain its permission and approval, 3 years and 1 month for its construction, and 4 months for pre-operation inspections on the reactor performance. This article reports on the history of STACY from its birth to its restart of operation, as well as its future utilization.

Free outflow from the end of a horizontal circular pipe related to flow from the PWR cold leg to the downcomer

Satou, Akira; Hibiki, Takashi*; Ikeda, Ryo; Shibamoto, Yasuteru

Progress in Nuclear Energy, 180, p.105593_1 - 105593_11, 2025/02

https://doi.org/10.1016/j.pnucene.2024.105593

During a loss-of-coolant accident in a pressurized water reactor (PWR), there is a risk that pressurized thermal shock (PTS) may occur on the internal wall of the reactor pressure vessel (RPV) due to the flow of emergency core cooling (ECC) water injected into the cold leg that flows into the downcomer. PTS is caused by the rapid cooling of the downcomer wall by the ECC water and is strongly influenced by the temperature of the ECC water, the collision position and velocity of the water jet on the wall, the velocity of the liquid film on the wall, the thickness of the liquid film, and the spread of the downward flow. Therefore, the flow of ECC water discharging from the cold leg to the downcomer may strongly impact PTS events. To help understand this flow phenomenon, we reviewed studies on free outflow from a circular pipe. Experimental findings on the classification of flow conditions, transition conditions between flow conditions, end depth ratio, free surface profile of flow in the circular pipe, and shape of the nappe flowing out from the pipe have been obtained in a form that is almost consistent with each other. In contrast, when considering the flow from the cold leg to the downcomer, it is necessary to deal with the flow field in a specific situation, such as the flow into a narrow gap rather than a free space, the existence of rounded corners at the outlet of the circular pipe, and the influence of steam flow flowing from the core to the cold leg. However, few previous studies consider these factors, so we summarized them as knowledge that needs to be accumulated in the future.

Development of Cs separation methods from large amounts of soil samples to determine the Cs/Cs isotope ratio

Shimada, Asako; Tsukahara, Takehiko*; Nomura, Masao*; Takeda, Seiji

Journal of Radioanalytical and Nuclear Chemistry, 333(12), p.6297 - 6310, 2024/12

https://doi.org/10.1007/s10967-024-09749-3