Verification of analytical model of MELCOR code for accident of evaporation to dryness by boiling of reprocessed high level liquid waste

Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi

JAEA-Research 2025-003, 24 Pages, 2025/06

JAEA-Research-2025-003.pdf:2.06MB

An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (RuO) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. RuO is expected to be absorbed chemically into water dissolving nitrous acid. Condensation of mixed vapor plays an important role for Ru transporting behavior in the facility building. The thermal-hydraulic behavior in the facility building is simulated with MELCOR code. The latent heat, which is a governing factor for vapor condensing behavior, has almost same value for nitric acid and water at the temperature range under 120 centigrade. Considering this thermal characteristic, it is assumed that the amount of nitric acid is substituted with mole-equivalent water in MELCOR simulation. Compensating modeling induced deviation by this assumption have been assembled with control function features of MELCOR. The comparison results have been described conducted between original simulation and modified simulation with compensating model in this report. It has been revealed that the total amount of pool water in the facility was as same as both simulations.

Microstructural observation of simulated fuel kernels for Pu-burner high temperature gas-cooled reactor in Japan

Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*

JAEA-Research 2024-012, 98 Pages, 2025/02

JAEA-Research-2024-012.pdf:32.24MB

Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.

Collection of strength characteristic data used for analysis evaluation in reactor pressure vessel and in-core structures in accident

Shimomura, Kenta; Yamashita, Takuya; Nagae, Yuji

JAEA-Data/Code 2022-012, 270 Pages, 2023/03

JAEA-Data-Code-2022-012.pdf:38.25MB

In a light water reactor, which is a commercial nuclear power plant, a severe accident such as loss of cooling function in the reactor pressure vessel (RPV) and exposure of fuel rods due to a drop in the water level in the reactor can occur when a trouble like loss of all AC power occurs. In the event of such a severe accident, the RPV may be damaged due to in-vessel conditions (temperature, molten materials, etc.) and leakage of radioactive materials from the reactor may occur. Verification and estimation of the process of RPV damage, molten fuel debris spillage and expansion, etc. during accident progression will provide important information for decommissioning work. Possible causes of RPV failure include failure due to loads and restraints applied to the RPV substructure (mechanical failure), failure due to the current eutectic state of low-melting metals and high-melting oxides with the RPV bottom members (failure due to inter-material reactions), and failure near the melting point of the structural members at the RPV bottom. Among the failure factors, mechanical failure is verified by numerical analysis (thermal hydraulics and structural analysis). When conducting such a numerical analysis, the heat transfer properties (thermal conductivity, specific heat, density) and material properties (thermal conductivity, Young's modulus, Poisson's ratio, tensile, creep) of the materials (zirconium, boron carbide, stainless steel, nickel-based alloy, low alloy steel, etc.) constituting the RPV and in-core structures to near the melting point are required to evaluate the creep failure of the RPV. In this document, we compiled data on the properties of base materials up to the melting point of each material constituting the RPV and in-core structures, based on published literature. In addition, because welds exist in the RPV and in-core structures, the data on welds are also included in this report, although they are limited.

Numerical study on effect of pressure on behavior of bubble coalescence by using CMFD simulation

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

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07

The mechanism of critical heat flux (CHF) for higher system pressure remains to be clarified, even though it is important to evaluate the CHF for the light water reactor (LWR) which is operated under the high pressure condition. In this study, the process of bubble coalescence was simulated by using a computational multi-fluid dynamics (CMFD) simulation code TPFIT under various system pressure in order to investigate the behavior of bubbles as a basic study. The growth of bubbles was simulated by blowing of vapor from a tiny orifice simulating bubble bottom. One or four orifices were located on the bottom surface in this simulation study. The numerical simulations were conducted by varying the pressure and temperature.

Progress of ITER full tungsten divertor technology qualification in Japan

Ezato, Koichiro; Suzuki, Satoshi; Seki, Yohji; Mori, Kensuke; Yokoyama, Kenji; Escourbiac, F.*; Hirai, Takeshi*; Kuznetsov, V.*

Fusion Engineering and Design, 98-99, p.1281 - 1284, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.03.009

Japan Atomic Energy Agency (JAEA) is now devoting to development of Full-W ITER divertor outer vertical target (OVT), especially, PFU that needs to withstand the repetitive heat load as high as 20MW/m. JAEA have succeeded in demonstrating that the soundness of a bonding technology is sufficient for the full-W ITER divertor. For the development of bonding technology, the load carrying capability test on the W monoblock with a leg attachment to an OVT support structure was carried out and shows that the attachment can withstand against the uniaxial load more than 20 kN which is three times higher than the IO requirement. JAEA manufactured 6 small-scale mock-ups and tested under the repetitive heat load of 10 and 20 MW/m to examine the durability of the divertor structure including W tile bonding and the cooling tube. All of the mock-ups could survived 5000 cycles at 10 MW/m and 1000 cycles 20 MW/m with no failure such as debonding of the W tile and water leak from the cooling tube. The number of cycles at 20 MW/m is three times longer than the requirement of ITER divertor.

Nuclear heat supply fluctuation test by non-nuclear heating using HTTR

Takada, Shoji; Sekita, Kenji; Nemoto, Takahiro; Honda, Yuki; Tochio, Daisuke; Inaba, Yoshitomo; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

To investigate the safety design criteria of heat utilization system for the HTGRs, it is necessary to evaluate the effect of fluctuation of thermal load on the reactor. The nuclear heat supply fluctuation test by non-nuclear heating was carried out to simulate the nuclear heat supply test which is carried out in the nuclear powered operation. The test data is used to verify the numerical code to calculate the temperature of core bottom structure to carry out the safety evaluation of abnormal events in the heat utilization system. In the test, the helium gas temperature was heated up to 120C. A sufficiently high temperature disturbance was imposed on the reactor inlet temperature. It was found that the response of temperatures of metallic components such as side shielding blocks was faster than those of graphite blocks in the core bottom structure, which was significantly affected by the heat capacities of components, the level of imposed disturbance and heat transfer performance.

Experimental examination of heat removal limitation of screw cooling tube at high pressure and temperature conditions

Ezato, Koichiro; Suzuki, Satoshi; Dairaku, Masayuki; Akiba, Masato

Fusion Engineering and Design, 81(1-7), p.347 - 354, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.06.360

no abstracts in English

Critical heat flux testing on screw cooling tube made of RAFM-steel F82H for divertor application

Ezato, Koichiro; Suzuki, Satoshi; Dairaku, Masayuki; Akiba, Masato

Fusion Engineering and Design, 75-79, p.313 - 318, 2005/11

https://doi.org/10.1016/j.fusengdes.2005.06.113

no abstracts in English

Natural convection heat transfer of high temperature gas in an annulus between two vertical concentric cylinders

Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki

Heat Transfer-Asian Research, 34(5), p.293 - 308, 2005/07

https://doi.org/10.1002/htj.20070

Water cooling panels have been adopted as the vessel cooling system of the HTTR to cool the reactor core indirectly by natural convection and thermal radiation. In order to investigate the heat transfer characteristics of high temperature gas in a vertical annular space between the reactor pressure vessel and cooling panels of the HTTR, we carried out experiments and numerical analyses on natural convection heat transfer coupled with thermal radiation heat transfer in an annulus between two vertical concentric cylinders with the inner cylinder heated and the outer cylinder cooled. In the present experiments, Rayleigh number based on the height of the annulus ranged from 2.010 to 5.410 for helium gas and from 1.210 to 3.510 for nitrogen gas. The numerical results were in good agreement with the experimental ones regarding the surface temperatures of the heating and cooling walls. As a result of the experiments and the numerical analyses, the heat transfer coefficient of natural convection coupled with thermal radiation was obtained.

Study of plasma termination using high-Z noble gas puffing in the JT-60U tokamak

Bakhtiari, M.; Tamai, Hiroshi; Kawano, Yasunori; Kramer, G. J.*; Isayama, Akihiko; Nakano, Tomohide; Kamiya, Kensaku; Yoshino, Ryuji; Miura, Yukitoshi; Kusama, Yoshinori; et al.

Nuclear Fusion, 45(5), p.318 - 325, 2005/05

https://doi.org/10.1088/0029-5515/45/5/002

In the previous works we had shown that injecting a mixture of large amounts of hydrogen and small amounts of argon can terminate a tokamak discharge quickly with avoiding runaway electron generation. In this work we have done the same experiments but with different gases in addition to argon. In fact we compared the effect of the puffing of argon, krypton, and xenon gases with and without simultaneous hydrogen gas puffing on disruption mitigation. We observed that injecting all impurities in the form of an admixture in hydrogen lead to faster plasma shutdowns with less runaway electron generation. We also found that injecting krypton gas (with or without hydrogen) seems to be a good candidate for plasma shutdown purposes since it induces low heat flux to divertor plates and avoids runaway electron generation more effectively.