Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 11703

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

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

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.

Journal Articles

A Quantitative method of eutectic reaction study between boron carbide and stainless steel

Hong, Z.*; Pellegrini, M.*; Erkan, N.*; Liao, H.*; Yang, H.*; Yamano, Hidemasa; Okamoto, Koji*

Annals of Nuclear Energy, 180, p.109462_1 - 109462_9, 2023/01

A series of experiments were conducted using B$$_{4}$$C material and SUS304 tubes as a simulant of the real control rods. Reaction rate constant data in the 1450K-1500K range were obtained, and are consistent with the reference values. The reaction layer microstructure observation and the associated chemical composition analysis were also carried onto the experiment samples.

Journal Articles

Effect of nitrogen concentration on creep strength and microstructure of 9Cr-ODS ferritic/martensitic steel

Oka, Hiroshi*; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Hashimoto, Naoyuki*

Journal of Nuclear Materials, 572, p.154032_1 - 154032_8, 2022/12

9Cr oxide dispersion strengthened steels with slightly different nitrogen concentrations (0.0034 - 0.029 wt%) were prepared and their creep property at 973 K was investigated with microstructural characterization before and after the creep test. The creep strength decreased significantly as the nitrogen concentration increased. Microstructural observation revealed that, in the higher nitrogen concentration specimen, coarse Y-rich inclusions were found along the boundary between transformed ferrite region and residual ferrite region. The solubility difference of nitrogen in $$alpha$$ and $$gamma$$ phase would induce the localized increment of nitrogen concentration in the boundary region during the austenitizing process, resulting in the thermodynamic destabilization and subsequent coarsening of the dispersed oxide particles. The rows of creep voids were found near the rupture part of the crept specimen, suggesting that the coarse inclusions were the starting point of creep void formation and the subsequent premature fracture.

JAEA Reports

Assessment report on research and development activities in FY2021 Activity "Research and Development on Fast Reactor and Fuel Cycle Technologies" (Post- and pre-review peport)

Konishi, Kensuke; Kunogi, Kosuke

JAEA-Evaluation 2022-005, 106 Pages, 2022/11

Japan Atomic Energy Agency (hereafter referred to as "JAEA") consulted with the "Evaluation Committee of Research and Development Activities for Fast Reactor and Fuel Cycle" (hereinafter referred to as "Committee"), which consists of specialists in the fields of the evaluation subjects of fast reactor cycle technologies, for post-review and pre-review assessment of Research and Development (R&D) activities of fast reactor cycle in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to JAEA's request, the Committee assessed mainly the progress of the R&D project according to guidelines, which addressed the rationale behind the R&D project, the relevance of the project outcome and the efficiency of the project implementation during the period of the current and next plan. This report is issued for the purpose of actively disseminate evaluation information to the people of Japan (based on General Guideline), which lists the members of the Committee and outlines the assessment items and the review process for procedure of the assessment. The assessment reports which were issued by the Committee is attached.

Journal Articles

Evaluation of fuel reactivity worth measurement in the prototype fast reactor Monju

Ohgama, Kazuya; Takegoshi, Atsushi*; Katagiri, Hiroki; Hazama, Taira

Nuclear Technology, 208(10), p.1619 - 1633, 2022/10

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

Journal Articles

Development of safety design criteria and safety design guidelines for Generation IV sodium-cooled fast reactors

Futagami, Satoshi; Kubo, Shigenobu; Sofu, T.*; Ammirabile, L.*; Gauthe, P.*

Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs 2022 (IAEA TIC 2022) (Internet), 10 Pages, 2022/10

JAEA Reports

Fuel unloading operations -2020- in the decommissioning of the prototype fast breeder reactor "Monju"

Shiota, Yuki; Ariyoshi, Hideo; Shiohama, Yasutaka; Isobe, Yuta; Takeuchi, Ryotaro; Kudo, Junki; Hanaki, Shotaro; Hamano, Tomoharu; Takagi, Tsuyohiko

JAEA-Technology 2022-019, 95 Pages, 2022/09

JAEA-Technology-2022-019.pdf:7.59MB

In the first stage of "Monju" decommissioning project, "Fuel Unloading Operations" have been carrying out. The operations consists of two processes. The first process is "Fuel Treatment and Storage" is that the fuel assemblies unloaded from the Ex-Vessel fuel Storage Tank (EVST) are canned after sodium cleaning, and then transferred to the storage pool. The second process is "Fuel Unloading" that the fuel assemblies in the reactor core are replaced with dummy fuel assemblies and stored in the EVST. "Fuel Treatment and Storage" and "Fuel Unloading" are performed alternately until 370 fuel assemblies in the core and 160 fuel assemblies in the EVST are all transferred to the storage pool. This is a summary of "Fuel Unloading" in the third quarter of "Fuel Unloading Operation". In fiscal 2020, as "Fuel Unloading", 72 fuel assemblies and 74 blanket fuel assemblies were unloaded from the core, and stored in the EVST. From the EVST, 145 dummy fuel assemblies and 1 fixed absorber were loaded in the core instead. During these operations, a total of 36 cases alarming or equipment malfunctions classified into 4 types occurred. However, these events were estimated in advance, there were no significant events that menaces to safety of fuel assemblies and equipment. Therefore, there were no serious problem like fall of fuel assemblies and events that may affect schedule of the project like stick of gripper of ex-vessel fuel transfer machine. When equipment's work or performance fail, the operation continued with safety by elimination of causes of problem. Fuel handling system of Monju has function that is endemic to sodium cooling fast breeding reactor. Because continuous operations of fuel handling system with actual fuel assemblies start recently, we don't have as much experience as PWR and BWR. With estimation of various troubles, reduction of frequency of trouble occurrence and minimization of impacts on schedule performed.

Journal Articles

Normal spectral emissivity, specific heat capacity, and thermal conductivity of type 316 austenitic stainless steel containing up to 10 mass% B$$_{4}$$C in a liquid state

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

Journal of Nuclear Materials, 568, p.153865_1 - 153865_12, 2022/09

 Times Cited Count:0 Percentile:0.02(Materials Science, Multidisciplinary)

The normal spectral emissivity, specific heat capacity and thermal conductivity of type 316 austenitic stainless steel (SS) containing boron carbide (B$$_{4}$$C) in a liquid state were experimentally measured over the composition range of SS-$$x$$ mass% B$$_{4}$$C (up to 10%) and wide temperature ranges using an electromagnetic levitator in a static magnetic field. The normal spectral emissivity and specific heat capacity were almost constant against temperature for all SS-B$$_{4}$$C melts, and the thermal conductivities of the melts had a negligible or small positive temperature dependence. The B$$_{4}$$C-content dependence of each property at 1800 K had a different tendency across the eutectic composition (around 3 mass% B$$_{4}$$C) of the SS-B$$_{4}$$C pseudo-binary system.

Journal Articles

Thermophysical properties of molten FeO$$_{1.5}$$, (FeO$$_{1.5}$$)$$_{0.86}$$-(ZrO$$_{2}$$)$$_{0.14}$$ and (FeO$$_{1.5}$$)$$_{0.86}$$-(UO$$_{2}$$)$$_{0.14}$$

Kondo, Toshiki; Toda, Taro*; Takeuchi, Junichi*; Kargl, F.*; Kikuchi, Shin; Muta, Hiroaki*; Oishi, Yuji*

Journal of Nuclear Science and Technology, 59(9), p.1139 - 1148, 2022/09

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Preliminary analysis of core disruptive accident in sodium-cooled fast reactor using eutectic reaction model of boron-carbide control-rod material

Yamano, Hidemasa; Morita, Koji*

Nihon Kikai Gakkai 2022-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2022/09

It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B4C) as a control rod material and stainless steel (SS) during a core disruptive accident (CDA) in an advanced large-scale sodium-cooled fast reactor (SFR) designed in Japan. A physical model simulating the eutectic reaction and relocation of the eutectic melt was developed to incorporate into the fast reactor severe accident analysis code SIMMER-IV for the CDA numerical analysis of SFRs. This study applied the SIMMER-IV code with the newly developed model to the CDA analysis of the SFR. This analysis indicated that the SIMMER-IV code using the eutectic reaction model has successfully simulated the eutectic reaction and the upward motion of the eutectic melt in the molten core pool as well as the reactivity transient behavior caused by the molten core material relocation.

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:0.02(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

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview and progress until 2020

Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Morita, Koji*; Nakamura, Kinya*; Fukai, Hirofumi*; et al.

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09

This paper describes the project overview and progress of experimental and analytical studies conducted until 2020. Specific results in this paper are the measurement of the eutectic reaction rates and the validation of physical model describing the eutectic reaction in the analysis code through the numerical analysis of the B$$_{4}$$C-SS eutectic reaction rate experiments in which a B$$_{4}$$C pellet was placed in a SS crucible.

Journal Articles

Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions with coarse-mesh subchannel CFD model

Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09

To enhance the safety of sodium-cooled fast reactors, the natural circulation (NC) decay heat removal systems with a dipped-type direct heat exchanger (D-DHX) have been investigated. During the D-DHX operation, since the core-plenum interaction occurs, development of the reactor vessel model including the more model by using a computational fluid dynamics code (RV-CFD) is required. Previously, the CFD model based on the subchannel analysis was developed. In this study, to achieve much lower computational cost maintaining the prediction accuracy, the coarse-mesh subchannel CFD (CMSC) model has been developed and was incorporated into the core of RV-CFD. As a result of PLANDTL-1 test analysis, the RV-CFD with the CMSC model can reproduce the radial heat transfer under NC conditions.

Journal Articles

Core thermal-hydraulics analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki

Mechanical Engineering Journal (Internet), 9(4), p.21-00438_1 - 21-00438_15, 2022/08

To enhance the safety of sodium-cooled fast reactors, a dipped-type direct heat exchanger (D-DHX) has been investigated in a natural circulation decay heat removal system. During the D-DHX operation, the core-plenum interactions occurs and the thermal-hydraulics in the reactor vessel (RV) is complicated, the establishment of thermal-hydraulic analysis model in the RV for computational fluid dynamics code (RV-CFD) is required to simulate the thermal stratification in the upper plenum and thermal-hydraulics in the core. In this study, in terms of using RV-CFD for design study, the subchannel CFD model with low computational cost was adopted to the core of RV-CFD and the numerical simulation was carried out in comparison with the measured data in the sodium test facility named PLANDTL-1. As the result, the calculated sodium temperature in the core had good agreement with the experimental result and the applicability of the RV-CFD for the core-plenum interactions was confirmed.

Journal Articles

Experiment study on the effect of nozzle shape on liquid jet breakup

Sun, G.*; Zhan, Y.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Okano, Yasushi

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08

When a liquid sodium leakage accident occurs in a sodium-cooled fast reactor, the injected sodium collides with structures to produce splashing droplets, which can result in a violent combustion. According to previous studies on circular nozzles, the amount of splash is affected by the state of the jet at the moment of impact. However, the outlet shape of damaged area is hardly to be circular; and meanwhile it influences the flow pattern of jet a lot. Considering about this, in the present work, high-speed cameras were used to observe the jet discharged from oval nozzles vertically downward to investigate the falling process of the jet. The result shows that surface wave appears on the jet and within a certain range of flow velocity it can be observed obviously, meanwhile accelerate the breakup of jet.

Journal Articles

Structural analysis of a reactor vessel in a sodium-cooled fast reactor under extremely high temperature conditions

Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08

This study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR, to understand its deformation behavior under extremely high temperature conditions and to identify the areas which should be focused to mitigate impacts of failure. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum during 20 hours assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, resulting in no failure. This effect contributes to maintaining RV sodium level in a long term, thereby enhancing the RV resilience.

Journal Articles

Event tree analysis for material relocation on core catcher in a sodium-cooled fast reactor

Yamano, Hidemasa; Kubo, Shigenobu; Kan, Taro*; Shibata, Akihiro*; Hourcade, E.*; Dirat, J. F.*

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08

In this paper, the approach to event tree development and the scope of the event tree analysis were described with key points on core catcher loading. For the analytical conditions, two core catcher loading conditions were given as bounding and conservative cases. For important heading of the event tree, key important phenomena were included: strong back design, fuel-coolant interaction and quench in the sodium plenum design, jet attack, criticality and coolability on the core catcher. In this paper, preliminary trial quantification was attempted using a probability ranking table which is based on engineering judgement. This event tree analysis has identified the dominant sequence, and clarified the effect of the core catcher loading and effectiveness of design measures. This study suggests that the criticality measure is very important for the core catcher study.

Journal Articles

Coolability evaluation of the debris bed on core catcher in a sodium-cooled fast reactor with a whole vessel model

Yamano, Hidemasa; Kubo, Shigenobu; Sasa, Kyohei*; Shibata, Akihiro*; Hourcade, E.*; Dirat, J. F.*

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08

This paper describes coolability evaluations of a debris bed with a variety of decay heat removal system (DHRS) operating conditions with a whole vessel model assuming fuel accumulation on the core catcher in a short term. The evaluation tool is a one-dimensional plant dynamics code, Super-COPD, with a debris bed module. The coolability evaluations have indicated that the current core catcher design secures sufficient natural circulation flows around the core catcher to ensure the debris bed cooling when at least one circuit of DHRS was activated. Sensitivity analyses under a pessimistic condition have shown that the debris bed is coolable with at least one circuit of improved DHRS even if most of fuel accumulates on the core catcher in a short term.

Journal Articles

Proposal of detailed procedures of determining rational in-service inspection requirements based on system based code concept

Yada, Hiroki; Takaya, Shigeru; Morohoshi, Kyoichi*; Yokoi, Shinobu*

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08

In order to develop rationalized maintenance plans of nuclear power plants, it is necessary to consider characteristics of each plant. For sodium-cooled fast reactor (SFR) plants, there are constraints on inspections due to the specialty that sodium equipment needs to be handled, which is one of the important points when considering rationalization of maintenance. Therefore, we previously proposed a maintenance optimization scheme based on the System Based Code (SBC) concept. One of proposed scheme goals is to develop detailed procedures of preparing a rationalized maintenance plan. In this study, the procedures to determine inspections for potential degradation and additional inspections in terms of defense-in-depth have been further clarified. Furthermore, the modified maintenance optimization scheme is also illustrated by a quantitative trial evaluation of the core support structure of the next SFR under development in Japan.

Journal Articles

Development of evaluation method of gas entrainment on the free surface in the reactor vessel in pool-type sodium-cooled fast reactors; Gas entrainment judgment based on three-dimensional evaluation of vortex center line and distribution of pressure decrease

Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/08

Development of evaluation method for cover gas entrainment (GE) by vortices generated at free surface in upper plenum of sodium-cooled fast reactor (SFR) is required. GE evaluation tool, named StreamViewer, based on method using numerical results of three-dimensional computational fluid dynamics analysis for loop-type SFRs has been developed. In this study, modification of evaluation method of StreamViewer to rationalize conservativeness in evaluation results was examined by identifying vortex center lines and calculating three-dimensional distribution of pressure decrease along vortex center lines. The applicability of modified method was checked using water experimental result in rectangular open channel where unsteady vortices are generated. As the result, it was indicated that evaluation results on gas core depth which were excessive in current method were improved in modified method, and it is confirmed that modified method may discriminate onset of GE with appropriate criteria.

11703 (Records 1-20 displayed on this page)