Hierarchical Bayesian modeling to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

Narukawa, Takafumi; Hamaguchi, Shusuke*; Takata, Takashi*; Udagawa, Yutaka

Nuclear Engineering and Design, 411, p.112443_1 - 112443_12, 2023/09

https://doi.org/10.1016/j.nucengdes.2023.112443

ARKADIA; For the innovation of advanced nuclear reactor design

Ohshima, Hiroyuki; Asayama, Tai; Furukawa, Tomohiro; Tanaka, Masaaki; Uchibori, Akihiro; Takata, Takashi; Seki, Akiyuki; Enuma, Yasuhiro

Journal of Nuclear Engineering and Radiation Science, 9(2), p.025001_1 - 025001_12, 2023/04

https://doi.org/10.1115/1.4054726

This paper describes the outline and development plan for ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source. ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant, including optimization of safety equipment. State-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&D are integrated with AI. In the first phase of development, ARKADIA-Design and ARKADIA-Safety will be constructed individually, with the first target of sodium-cooled reactor. In a subsequent phase, everything will be integrated into a single entity applicable not only to advanced rectors with a variety of concepts, coolants, configurations, and output levels but also to existing light-water reactors.

In-vessel source term analysis code TRACER Version 2.4.1 (User's manual)

Ono, Masahiro*; Uchibori, Akihiro; Okano, Yasushi; Takata, Takashi*

JAEA-Testing 2022-004, 193 Pages, 2023/03

JAEA-Testing-2022-004.pdf:3.31MB

A computer code TRACER (Transport phenomena of Radionuclides for Accident Consequence Evaluation of Reactor) version 2.4.1 has been developed to evaluate species and quantities of fission products (FPs) released into cover gas due to a fuel pin failure in an LMFBR. The TRACER version 2.4.1 includes the models related to NUREG-0772 and also new or modified computational program codes in order to possess a new function shown below, and partial modify of coefficient of FP transition model between coolant and cover gas. This manual includes manual conventions for TRACER Version 2.3, addition of reference such as formula, improvement of explanation of input file creation method, addition of improvement of NUREG-0772 model added to TRACER code, modification of figure of sample analysis performed in appendix. It includes modifications and additions of sample analysis.

Development of ARKADIA-Safety for severe accident evaluation of sodium-cooled fast reactors

Aoyagi, Mitsuhiro; Sonehara, Masateru; Ishida, Shinya; Uchibori, Akihiro; Kawada, Kenichi; Okano, Yasushi; Takata, Takashi

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09

Development of integrated severe accident analysis code, SPECTRA for sodium-cooled fast reactor

Uchibori, Akihiro; Sonehara, Masateru; Aoyagi, Mitsuhiro; Takata, Takashi*; Ohshima, Hiroyuki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

A new computational code, SPECTRA, has been developed for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors. The in-vessel thermal hydraulics module includes coupled analytical models for multidimensional multifluid model considering compressibility and relocation of a molten core. A lumped mass model is employed for computing behavior of ex-vessel compressible multicomponent gas including aerosols. This model is coupled with the models for ex-vessel phenomena such as sodium fire. Loss of reactor level event starting from leakage of sodium coolant was computed. Basic capability to evaluate severe accident progress was demonstrated through this analysis.

Quasi-Monte Carlo sampling method for simulation-based dynamic probabilistic risk assessment of nuclear power plants

Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Journal of Nuclear Science and Technology, 59(3), p.357 - 367, 2022/03

https://doi.org/10.1080/00223131.2021.1971119

Dynamic probabilistic risk assessment (PRA), which handles epistemic and aleatory uncertainties by coupling the thermal-hydraulics simulation and probabilistic sampling, enables a more realistic and detailed analysis than conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a station blackout sequence in a boiling water reactor and compared each method. The result indicated that quasi-Monte Carlo sampling method handles the uncertainties most effectively in the assumed scenario.

Development of reacting jet evaluation model based on engineering approaches with particle method for improvement of LEAP-III code

Kosaka, Wataru; Uchibori, Akihiro; Takata, Takashi; Yanagisawa, Hideki*; Watanabe, Akira*; Jang, S.*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 11 Pages, 2022/03

For the safety assessment of a steam generator (SG) in a sodium-cooled fast reactor, the analysis code LEAP-III can evaluate the water leak rate during the long-term event progress including the tube failure propagation triggered by an occurrence of a small water leak in a failed heat transfer tube in SG. The LEAP-III has the advantage in completing the calculation with low computational cost since it consists of semi-empirical formulae and one-dimensional equations of conservation. However, an evaluation model of temperature distribution by the reacting jet provides wider high temperature region than the experimental data. As a result, LEAP-III shows excessive conservativeness in some case. A Lagrangian particle method code based on engineering approaches has been developed in order to improve this model to get more realistic temperature distribution. In this method, the jet behavior and chemical reaction are simulated using Newton's equation of motion with several engineering approximations instead of solving multi-dimension multiphase thermal hydraulic equations with sodium-water reaction. In this study, interparticle interaction force model was added, and also the chemical reaction and gas-liquid heat transfer evaluation models were improved. We conducted a test analysis, and compared the results by this particle method with the ones by SERAPHIM, that is a mechanistic analysis code for multi-dimensional multiphase flow considering compressibility and sodium-water reaction. Through this test analysis, it confirmed that this particle method has the basic capability to get a realistic temperature distribution with low computational cost, and also to predict tube failure occurrence by coupled with LEAP-III.

Study on sodium-water reaction jet evaluation model based on engineering approaches with particle method

Kosaka, Wataru; Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Jang, S.*

Nihon Kikai Gakkai Rombunshu (Internet), 88(905), p.21-00310_1 - 21-00310_9, 2022/01

https://doi.org/10.1299/transjsme.21-00310

If a pressurized water/water-vapor leaks from a heat transfer tube in a steam generator (SG) in a sodium-cooled fast reactor (SFR), sodium-water reaction forms high-velocity, high-temperature, and corrosive jet. It would damage the other tubes and might propagate the tube failure in the SG. Thus, it is important to evaluate the effect of the tube failure propagation for safety assessment of SFR. The computational code LEAP-III can evaluate water leak rate during the tube failure propagation with short calculation time, since it consists of empirical formulae and one-dimensional equations of conservation. One of the empirical models, temperature distribution evaluation model, evaluates the temperature distribution in SG as circular arc isolines determined by experiments and preliminary analyses instead of complicated real distribution. In order to improve this model to get more realistic temperature distribution, we have developed the Lagrangian particle method based on engineering approaches. In this study, we have focused on evaluating gas flow in a tube bundle system, and constructed new models for the gas-particles behavior around a tube to evaluate void fraction distribution near the tube. Through the test analysis simulating one target tube system, we confirmed the capability of the models and next topic to improve the models.

Numerical assesment of sodium fire incident

Takata, Takashi; Aoyagi, Mitsuhiro; Sonehara, Masateru

IAEA-TECDOC-1972, p.224 - 234, 2021/08

Sodium fire is one of the key issues for plant safety of sodium-cooled fast reactor (SFR) regardless of its size. In general, a concrete structure, which includes free and bonging water inside, is used in a reactor building. Accordingly, water vapor will release from the concrete during sodium fire incident due to temperature increase resulting in a hydrogengeneration even in a dry air condition. The probability of hydrogen generation will increase in accordance with a decrease of a dimension of compartment that corresponds to a small and medium sized or modular reactor (SMR). A numerical investigation of a small leakage sodium pool fire has been carried out by changing a dimension of compartment. Furthermore, numerical challenges to enhance a prediction accuracy of hydrogen generation during sodium fire has also been discussed in the paper.

Droplet entrainment by high-speed gas jet into a liquid pool

Sugimoto, Taro*; Kaneko, Akiko*; Abe, Yutaka*; Uchibori, Akihiro; Kurihara, Akikazu; Takata, Takashi; Ohshima, Hiroyuki

Nuclear Engineering and Design, 380, p.111306_1 - 111306_11, 2021/08

https://doi.org/10.1016/j.nucengdes.2021.111306

Liquid droplet entrainment by a high-speed gas jet is a key phenomenon for evaluation of sodium-water reaction. In this study, a visualization experiment for liquid droplet entrainment by an air jet in a water pool by using frame-straddling method was carried for development of an entrainment model in a sodium-water reaction analysis code. This experiment successfully provided clear images that captured generation and movement of droplets. Droplet diameter and moving speed were obtained at different locations and gas jet velocities from image processing. The measured data contributes phenomena elucidation and model development.