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Journal Articles

Development and validation of analysis code for spallation products behavior in LBE coolant system of ADS comparing with the distribution data in MEGAPIE spallation target

Miyahara, Shinya*; Arita, Yuji*; Nakano, Keita; Maekawa, Fujio; Sasa, Toshinobu; Obayashi, Hironari; Takei, Hayanori

Nuclear Engineering and Design, 403, p.112147_1 - 112147_17, 2023/03

It is important to evaluate the inventories and the release and transport behavior of the spallation products (SPs) in the Lead-Bismuth Eutectic (LBE) coolant system of Accelerator Driven System (ADS) for the safety studies of the radiological hazard both in the cases of normal operation and accident. University of Fukui and JAEA have been developing the computer analysis code TRAIL (Transport of RAdionuclides In Liquid metal systems) which predicts the time dependent behavior of SPs within the LBE coolant system of ADS for the wide range of operational events. The source term of both radioactive and stable SPs in the LBE coolant is given as input and the radioactive decay chain model for the radioactive SPs is implemented in the code to evaluate the effect of precursors on the SPs mobility. This paper presents the recent advancement status of the code development and the validation results comparing with the distribution data of volatile SPs in MEGAPIE spallation target.

Journal Articles

New approach for the detection of defects in the core support structure of SFRs using EMAT based on a Halbach magnet

Yamaguchi, Toshihiko; Mihalache, O.

Nuclear Engineering and Design, 401, p.112084_1 - 112084_14, 2023/01

Journal Articles

Study on evaluation method of kernel migration of TRISO fuel for High Temperature Gas-cooled Reactor

Fukaya, Yuji; Okita, Shoichiro; Sasaki, Koei; Ueta, Shohei; Goto, Minoru; Ohashi, Hirofumi; Yan, X.

Nuclear Engineering and Design, 399, p.112033_1 - 112033_9, 2022/12

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Kernel migration of TRi-structural ISOtropic (TRISO) fuel for High Temperature Gas-cooled Reactor (HTGR) has been analyzed to investigate the potential dominating effects. Kernel migration is a major fuel failure mode and dominant to determine the lifetime of the fuel for High Temperature engineering Test Reactor (HTTR). However, this study shows that the result and reliability depend on the evaluation method. The evaluation method used in this study takes into account of actual distribution of Coated Fuel Particles (CFPs) and the resulting heterogeneous fuel temperature calculation with such distribution. The result shows that the Kernel Migration Rate (KMR) is predicted to be about 10% less compared with the most conservative evaluation.

Journal Articles

Advanced thermal-hydraulic experiments and instrumentation for heavy liquid metal reactors

Pacio, J.*; Van Tichelen, K.*; Eckert, S.*; Wondrak, T.*; Di Piazza, I.*; Lorusso, P.*; Tarantino, M.*; Daubner, M.*; Litfin, K.*; Ariyoshi, Gen; et al.

Nuclear Engineering and Design, 399, p.112010_1 - 112010_15, 2022/12

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Heavy-liquid metals (HLMs), such as lead and lead-bismuth eutectic (LBE), are proposed as primary coolants in accelerator driven systems and next-generation fast reactors. In Europe, the reference systems using HLMs are MYRRHA (LBE) and ALFRED (lead). This article presents an overview of recent experiences and ongoing activities on pool-type and loop-type HLM experiments. Pool tests include the measurement of forced- and natural-circulation flow patterns in several scenarios representative of nominal and decay heat removal conditions. Loop tests are focused on the evaluation of specific components, like mockups of the fuel assembly, control rod and heat exchangers. They involve the measurement of global variables, such as flow rate and pressure difference, and local quantities like temperature, velocity and vibrations. Advanced instrumentation, capable of sustaining high temperatures and corrosion, is necessary for accurate measurements, often in compact geometries. In addition to traditional techniques, other instrumentation based on optical fibers, ultrasonic and electromagnetic methods are discussed.

Journal Articles

Present status of JAEA's R&D toward HTGR deployment

Shibata, Taiju; Nishihara, Tetsuo; Kubo, Shinji; Sato, Hiroyuki; Sakaba, Nariaki; Kunitomi, Kazuhiko

Nuclear Engineering and Design, 398, p.111964_1 - 111964_4, 2022/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Japan Atomic Energy Agency (JAEA) has been promoting the research and development (R&D) of High Temperature Gas-cooled Reactor (HTGR). R&D on reactor technologies is carried out by using High Temperature engineering Test Reactor (HTTR). The HTTR was resumed without significant reinforcements in 2021. On January 2022, a safety demonstration test under the OECD/NEA LOFC project was carried out. JAEA is promoting R&D on a carbon-free hydrogen production by thermochemical water splitting Iodine-Sulfur process (IS process). JAEA conducts design study for various HTGR systems toward commercialization. A new test program about demonstration of hydrogen production by the HTTR was launched. Steam methane reforming hydrogen production system was selected for the first demonstration by 2030.

Journal Articles

Friction factor and Reynolds number correlation for finned tube bundle of the air cooler of Monju reactor

Sotsu, Masutake

Nuclear Engineering and Design, 396( ), p.111893_1 - 111893_27, 2022/09

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Applicability of the analysis model of the cooling ability under a high temperature and low flowrate condition should be improved to evaluate the plant safety in case of the severe accident due to long-term station blackout. The present study reviews experiments on pressure drop behavior for complicated tube bundle geometry to the flow path and then develops a new correlation equation based on a computational fluid dynamics analysis of the Monju air cooler reflecting the actual geometry and plant data. After a basic simulation model being developed for a typical pressure drop experiment, the simulation is applied to the Monju air cooler that has a finned tube bundle. The obtained relationship between friction factors and Reynolds numbers ranging from 10 to 10,000 are fitted to a power function to derive a correlation equation of the fin tube bundle friction factor. The derived correlation equation can estimate pressure loss in the finned tube bundle more precisely than that in the Monju design. It is applicable to the future reactor design of the air cooler, especially when the cooling ability in low Reynolds number is requested.

Journal Articles

Calculation of shutdown gamma distribution in the high temperature engineering test reactor

Ho, H. Q.; Ishii, Toshiaki; Nagasumi, Satoru; Ono, Masato; Shimazaki, Yosuke; Ishitsuka, Etsuo; Goto, Minoru; Simanullang, I. L.*; Fujimoto, Nozomu*; Iigaki, Kazuhiko

Nuclear Engineering and Design, 396, p.111913_1 - 111913_9, 2022/09

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Time-resolved 3D visualization of liquid jet breakup and impingement behavior in a shallow liquid pool

Kimura, Fumihito*; Yamamura, Sota*; Fujiwara, Kota*; Yoshida, Hiroyuki; Saito, Shimpei*; Kaneko, Akiko*; Abe, Yutaka*

Nuclear Engineering and Design, 389, p.111660_1 - 111660_11, 2022/04

 Times Cited Count:2 Percentile:91.04(Nuclear Science & Technology)

Journal Articles

Improving the safety of the high temperature gas-cooled reactor "HTTR" based on Japan's new regulatory requirements

Hamamoto, Shimpei; Shimizu, Atsushi; Inoi, Hiroyuki; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Sekita, Kenji; Watanabe, Shuji; Furusawa, Takayuki; Iigaki, Kazuhiko; et al.

Nuclear Engineering and Design, 388, p.111642_1 - 111642_11, 2022/03

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Following the Fukushima Daiichi Nuclear Power Plant accident in 2011, the Japan Atomic Energy Agency adapted High-Temperature engineering Test Reactor (HTTR) to meet the new regulatory requirements that began in December 2013. The safety and seismic classifications of the existing structures, systems, and components were discussed to reflect insights regarding High Temperature Gas-cooled Reactors (HTGRs) that were acquired through various HTTR safety tests. Structures, systems, and components that are subject to protection have been defined, and countermeasures to manage internal and external hazards that affect safety functions have been strengthened. Additionally, measures are in place to control accidents that may cause large amounts of radioactive material to be released, as a beyond design based accident. The Nuclear Regulatory Commission rigorously and appropriately reviewed this approach for compliance with the new regulatory requirements. After nine amendments, the application to modify the HTTR's installation license that was submitted in November 2014 was approved in June 2020. This response shows that facilities can reasonably be designed to meet the enhanced regulatory requirements, if they reflect the characteristics of HTGRs. We believe that we have established a reference for future development of HTGR.

Journal Articles

Experiments of melt jet-breakup for agglomerated debris formation using a metallic melt

Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Abe, Yutaka*

Nuclear Engineering and Design, 386, p.111575_1 - 111575_17, 2022/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Design of a portable backup shutdown system for the high temperature gas cooled reactor

Hamamoto, Shimpei; Ho, H. Q.; Iigaki, Kazuhiko; Goto, Minoru; Shimazaki, Yosuke; Sawahata, Hiroaki; Ishitsuka, Etsuo

Nuclear Engineering and Design, 386, p.111564_1 - 111564_8, 2022/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The experience of Fukushima Daiichi Nuclear Power Plant accident caused by the great earthquake that occurred in eastern Japan in 2011 showed the importance of preparing for the loss of function of the engineered safety features. Increasing the strength of equipment to prevent loss of function in an accident is effective, but the possibility of loss of function remains. Therefore, it is important to have an alternative to lost functions in order to put the accident under control early. Thus, this study designed an alternative shutdown system, namely a portable backup shutdown system (PBSS), to make countermeasures in the event of a loss of shutdown function more robust without impairing economic efficiency of the High Temperature Gas-cooled Reactor (HTGR). The PBSS is portable and capable of being installed manually so that it can operate in a total loss of off-site electricity. Various neutron absorber materials for the PBSS were also considered from the viewpoints of technical and cost-effective properties. As results of optimization, the boron nitride (BN) was selected as it shows a good neutronic property as well as a reasonable cost in comparison with other materials.

Journal Articles

Seismic classification of high temperature engineering test reactor

Ono, Masato; Shimizu, Atsushi; Ohashi, Hirofumi; Hamamoto, Shimpei; Inoi, Hiroyuki; Tokuhara, Kazumi*; Nomoto, Yasunobu*; Shimazaki, Yosuke; Iigaki, Kazuhiko; Shinozaki, Masayuki

Nuclear Engineering and Design, 386, p.111585_1 - 111585_9, 2022/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In the late 1980s during the design stage, the seismic classification of the high temperature engineering test reactor (HTTR) was formulated. Owing to the lack of operation experiences of the HTTR to sufficiently understand the safety characteristics of high temperature gas cooled reactors (HTGR) at that time, the seismic classification of commercial light water reactors (LWR) was applied to HTTR. However, the subsequent operation experiences and test results using HTTR made it clear that the seismic classification of commercial LWR was somewhat too conservative for the HTGR. As a result, Class S facilities were downgraded compared to the commercial LWR. Moreover, the validity of the new seismic classification is confirmed. In June 2020, the Nuclear Regulatory Authority approved that the result of the seismic classification conformed to the standard rules of the reactor installation change.

Journal Articles

Comparison of sodium fast reactor core assembly seismic evaluation using the Japanese and French simulation tools

Yamamoto, Tomohiko; Matsubara, Shinichiro*; Harada, Hidenori*; Saunier, P.*; Martin, L.*; Gentet, D.*; Dirat, J.-F.*; Collignon, C.*

Nuclear Engineering and Design, 383, p.111406_1 - 111406_14, 2021/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Japan-France collaboration on ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is launched in 2014. In this project, Japan-France evaluates core assemblies with interferences on seismic event. The object of this study is to verify the seismic evaluation method on core assemblies between Japan and France by comparing the results. The analysis of this benchmark calculation shows a satisfactory agreement between the Japanese and French tools and the figures show a good behavior of the core in horizontal direction under French seismic condition.

Journal Articles

Analysis of Fukushima-Daiichi Nuclear Power Plant Unit 3 pressure data and obtained insights on accident progression behavior

Sato, Ikken

Nuclear Engineering and Design, 383, p.111426_1 - 111426_19, 2021/11

 Times Cited Count:1 Percentile:33.13(Nuclear Science & Technology)

Journal Articles

An Approach toward evaluation of long-term fission product distributions in the Fukushima Daiichi Nuclear Power Plant after the severe accident

Uchida, Shunsuke; Karasawa, Hidetoshi; Kino, Chiaki*; Pellegrini, M.*; Naito, Masanori*; Osaka, Masahiko

Nuclear Engineering and Design, 380, p.111256_1 - 111256_19, 2021/08

 Times Cited Count:2 Percentile:56.5(Nuclear Science & Technology)

It is essential to grasp the long-term distributions of FP as well as fuel debris all over the Fukushima Daiichi Nuclear Power Plant (1F) for safe completion of its decommissioning projects. The fuel debris is going to be removed from the plant under the severe conditions of FP being scattered during major decommissioning work, and then, the decommissioning projects are going to be terminated by storing safely the removed debris as recovered fertile materials or as materials for final radioactive disposal. In order to determine the FP distribution in the plant for the long period from the accident occurrence to the termination of the plant decommissioning, procedures for analyzing multi-term FP behaviors were proposed. The proposed procedures should be improved by applying the FP data measured in the plant and validated based on the feedback data. Then, the accuracy-improved procedures should be applied to estimate FP distribution during each period of the decommissioning projects.

Journal Articles

Conversion factors bridging radioactive fission product distributions in the primary containment vessel of Fukushima Daiichi NPP and dose rates measured by the containment atmosphere monitoring system

Uchida, Shunsuke; Pellegrini, M.*; Naito, Masanori*

Nuclear Engineering and Design, 380, p.111303_1 - 111303_11, 2021/08

 Times Cited Count:1 Percentile:33.13(Nuclear Science & Technology)

Multi-term FP analysis procedures were developed to determine FP distribution all over F1 not only for analyzing accident propagation but also for planning its decommissioning projects. They should be validated based on the measured FP data. One of the useful tools for their validation was application of the dose rate data monitored by the containment atmosphere monitoring system (CAMS). However, in order to compare the data with different characteristics and dimensional units, e.g., FP distribution (kg, Bq) and dose rate (Sv/h), application of the conversion factors bridging them would be effective and useful. In order to prepare speedy, easy-to-handle and tractable procedures to calculate radiation dose rates at the CAMS detector locations, dose rate conversion factors were determined for major source locations and major radionuclides. The dose rates could be easily calculated by multiplying FP amounts obtained with the multiterm FP analysis procedures by the conversion factors.

Journal Articles

Numerical modeling of radiation heat transfer from combusting droplets for a sodium fire analysis

Aoyagi, Mitsuhiro; Takata, Takashi; Uno, Masayoshi*

Nuclear Engineering and Design, 380, p.111258_1 - 111258_11, 2021/08

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

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

 Times Cited Count:3 Percentile:71.19(Nuclear Science & Technology)

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.

Journal Articles

Preparation for restarting the high temperature engineering test reactor; Development of utility tool for auto seeking critical control rod position

Ho, H. Q.; Fujimoto, Nozomu*; Hamamoto, Shimpei; Nagasumi, Satoru; Goto, Minoru; Ishitsuka, Etsuo

Nuclear Engineering and Design, 377, p.111161_1 - 111161_9, 2021/06

 Times Cited Count:1 Percentile:33.13(Nuclear Science & Technology)

Journal Articles

Estimation of the core degradation and relocation at the Fukushima Daiichi Nuclear Power Station Unit 2 based on RELAP/SCDAPSIM analysis

Madokoro, Hiroshi; Sato, Ikken

Nuclear Engineering and Design, 376, p.111123_1 - 111123_15, 2021/05

 Times Cited Count:3 Percentile:71.19(Nuclear Science & Technology)

259 (Records 1-20 displayed on this page)