Immortal experimental loop at JAEA; Post-process and validation

Watanabe, Nao; Obayashi, Hironari; Sugawara, Takanori; Sasa, Toshinobu; Nishihara, Kenji; Castelliti, D.*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.248 - 261, 2019/08

For the R&D for lead-bismuth eutectic alloy (LBE) cooled Accelerator Driven System (ADS), installation of experimental facility using LBE spallation target dedicated for ADS materials irradiation under flowing high temperature LBE environment is planned within the J-PARC project. JAEA has recently finalized the construction of "IMMORTAL", a demonstrative test loop representing a 1:1 model of the above LBE spallation target system. Such facility pursues several different objectives such as studies on the thermal-hydraulic behavior of the target, and validation of the LBE physical properties, the pressure drops and the heat transfer correlations. These results will be useful for design of the facility and LBE target/cooled ADS. In the frame of a bi-lateral collaboration between JAEA and SCK-CEN, a benchmark exercise on the experimental results from IMMORTAL has been carried out. The calculations have been performed with RELAP5-3D System Thermal-Hydraulic code. To assess and predict the thermal-hydraulic behavior of its primary loop, RELAP5-3D calculation models have been configured. The calculated results from these models represented that a valid contribution towards the validation of the LBE properties and empirical correlations present in RELAP5-3D code.

CFD analysis of the CIGMA experiments on the heated JET injection into containment vessel with external surface cooling

Hamdani, A.; Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.5463 - 5479, 2019/08

Analysis for the accident at Unit 3 of the Fukushima Daiichi NPS with THALES2/KICHE Code in BSAF2 project

Ishikawa, Jun; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.536 - 547, 2019/08

Steady-state and transient experiments in mock-up of J-PARC LBE spallation target system using mock-up loop "IMMORTAL"

Obayashi, Hironari; Yamaki, Kenichi; Kita, Satoshi; Yoshimoto, Hidemitsu; Wan, T.; Saito, Shigeru; Sasa, Toshinobu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.262 - 274, 2019/08

Construction of an experimental facility using LBE spallation target is under planning in the framework of J-PARC project to understand the compatibility of candidate ADS structural materials under flowing high temperature LBE environment with proton / neutron irradiation. We constructed a mock-up test loop, IMMORTAL to verify the feasibility of a primary cooling system in the target system. Including the layout considering remote maintainability, the same design as the actual JAEA's LBE target is reflected in most devices of IMMORTAL such as an electromagnetic pump (EMP), a heat exchanger and instrumentations. During the integrated tests, the individually developed devices of LBE loop system worked without significant malfunction. Several experimental data to verify the safety analysis model for LBE loop system using RELAP5-3D was acquired by simulation experiment of transient events such as beam over power and decrease of forced cooling.

Analysis for the accident at unit 2 of the Fukushima Daiichi NPS with THALES2/KICHE code in BSAF2 project

Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.100 - 111, 2019/08

JAEA participated in the OECD/NEA BSAF2 project with our integrated severe accident analysis code, THALES2/KICHE, in order to analyze and discuss the accident progression and source term of the accident at the Fukushima Daiichi NPS. One of important characteristics of THALES2/KICHE code is that it has the capability of predicting iodine chemistry based on reaction kinetics in the aqueous phase. JAEA performed the three week analysis for the accident at unit 2 on the basis of the boundary conditions and assumptions proposed by the BSAF2 project and our own assumptions. One of focusing points in the BSAF2 project was the trend of measured data of reactor vessel from 20:00 March 14 to 02:00 March 15. An assumption was made that the lower part of the suppression chamber failed to form a water leakage path. The released iodine and cesium within three weeks after the earthquake were predicted to be approximately 3% and 0.1% of the initial inventory, respectively.

Application of unstructured mesh-based sodium-water reaction analysis code SERAPHIM

Uchibori, Akihiro; Shiina, Yoshimi*; Watanabe, Akira*; Takata, Takashi

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.5315 - 5324, 2019/08

A computational fluid dynamics code SERAPHIM for a sodium-water reaction phenomenon has been developed. An unstructured mesh-based analysis method has been integrated into this code for evaluation in a complex-shaped domain including multiple heat transfer tubes in our recent studies. In this study, a numerical analysis of an experiment on sodium-water reaction in a tube bundle domain was performed to investigate applicability of the unstructured mesh-based SERAPHIM code. The formulated high-temperature region and its temperature level agreed with an experimental result.

Establishment of guideline for credibility assessment of nuclear simulations in the Atomic Energy Society of Japan

Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08

Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.

Ultrasound measurement of upward liquid film flow in vertical pipe

Wada, Yuki; Satou, Akira; Shibamoto, Yasuteru; Yonomoto, Taisuke; Sagawa, Jun*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.4518 - 4531, 2019/08

Liquid film detection under boiling transition (BT) condition is one of the important issues to develop models on dry out and rewet including physical characteristics of liquid film behavior. Although a heater surface temperature has been often used in previous studies to detect the position of liquid film front, it is difficult to accurately identify the position from the temperature measurement. Therefore, we are developing a nonintrusive measurement technique for detecting thin liquid film thickness under BT and rewet condition using ultrasound. In this study, we focus on high accuracy measurement for liquid film thinner than 0.1 mm by using high frequency ultrasound of 15 MHz and developing a signal processing method. Liquid film measurement results were found to agree with liquid film thickness correlations. Based on a comparison with constant current method, it is concluded that the present technique gives more reasonable liquid film thickness than constant current method.

Two-phase flow structure in a particle bed packed in a confined channel

Ito, Daisuke*; Kurisaki, Tatsuya*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.6430 - 6439, 2019/08

In core disruptive accident of sodium-cooled fast reactor, cooling of residual fuel debris formed in the reactor core is one of important factors to achieve in-vessel retention of the fuel. To clarify the feasibility of the cooling which is called "in-place cooling", characteristics of gas-liquid two-phase flow in the debris bed must be well understood. Since the debris bed can be formed in a confined flow channel in the core, effect of the channel wall cannot be neglected. Thus, this study aims to clarify the effect of the wall on two-phase flow characteristics in the debris bed, which was simulated as a particle bed packed in a pipe. The pressure drop was measured and compared with results by previous models, and porosity and void fraction distributions were measured by X-ray radiography. Then, the pressure drop evaluation model was modified considering the wall effect, and the applicability of the models was discussed.