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Post-test analyses of the CMMR-4 test

山下 拓哉; 間所 寛; 佐藤 一憲

Journal of Nuclear Engineering and Radiation Science, 8(2), p.021701_1 - 021701_13, 2022/04

Understanding the final distribution of core materials and their characteristics is important for decommissioning the Fukushima Daiichi Nuclear Power Station (1F). Such characteristics depend on the accident progression in each unit. However, boiling water reactor accident progression involves great uncertainty. This uncertainty, which was clarified by MAAP-MELCOR Crosswalk, cannot be resolved with existing knowledge and was thus addressed in this work through core material melting and relocation (CMMR) tests. For the test bundle, ZrO$$_{2}$$ pellets were installed instead of UO$$_{2}$$ pellets. A plasma heating system was used for the tests. In the CMMR-4 test, useful information was obtained on the core state just before slumping. The presence of macroscopic gas permeability of the core approaching ceramic fuel melting was confirmed, and the fuel columns remained standing, suggesting that the collapse of fuel columns, which is likely in the reactor condition, would not allow effective relocation of the hottest fuel away from the bottom of the core. This information will help us comprehend core degradation in boiling water reactors, similar to those in 1F. In addition, useful information on abrasive water suspension jet (AWSJ) cutting for debris-containing boride was obtained in the process of dismantling the test bundle. When the mixing debris that contains oxide, metal, and boride material is cut, AWSJ may be repelled by the boride in the debris, which may cut unexpected parts, thus generating a large amount of waste in cutting the boride part in the targeted debris. This information will help the decommissioning of 1F.


Effect of nitrous acid on migration behavior of gaseous ruthenium tetroxide into liquid phase

吉田 尚生; 大野 卓也; 吉田 涼一朗; 天野 祐希; 阿部 仁

JAEA-Research 2021-011, 12 Pages, 2022/01




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

岩澤 譲; 杉山 智之; 阿部 豊*

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

In severe accidents in a light water reactor, the relocated molten core (so-called corium or melt) can form a debris bed. The debris bed coolability is a critical issue for prevention and mitigation of the molten core-concrete interactions. Agglomeration has a serious impact on assessment of debris bed coolability if agglomeration forms massive debris (so-called agglomerated debris) by merging of melt particles with others when the melt particles accumulate on a floor. This paper presents the results of melt jet-breakup experiments for agglomerated debris formation using a simulant metallic melt. The experiments injected a melt jet of a low-melting point metal through a circular nozzle into a test section filled with coolant water. The particles were generated due to the melt jet-breakup accumulated on to a catcher, which is a flat plate made of stainless steel, installed in the test section. A high-speed video camera imaged particle formation and accumulation on the catcher plate. Agglomerated debris was confirmed by morphological investigation of the recovered debris. The experimental results revealed the effects of the melt jet injection conditions (melt temperature, coolant temperature, and coolant depth) on the mass fraction of agglomerated debris. On the basis of the experimental results, we proposed a simple correlation to estimate the mass fraction. The simple correlation successfully reproduced the mass fraction of agglomerated debris obtained in the DEFOR-A test [Kudinov et al., Nucl. Eng. Des., 301 (2013), 284-295]. The experimental data base presented in this paper makes further contributions to the modeling and validation of mechanistic models or simulation tools for agglomerated debris formation.


Revaporization behavior of cesium and iodine compounds from their deposits in the steam-boron atmosphere

Rizaal, M.; 三輪 周平; 鈴木 恵理子; 井元 純平; 逢坂 正彦; Gou$"e$llo, M.*

ACS Omega (Internet), 6(48), p.32695 - 32708, 2021/12

This paper presents our investigation on cesium and iodine compounds revaporization from cesium iodide (CsI) deposits on the surface of stainless steel type 304L, which were initiated by boron and/or steam flow. A dedicated basic experimental facility with a thermal gradient tube (TGT) was used for simulating the phenomena. The number of deposits, the formed chemical compounds, and elemental distribution were analyzed from samples located at temperature range 1000-400 K. In the absence of boron in the gas flow, it was found that the initial deposited CsI at 850 K could be directly re-vaporized as CsI vapor/aerosol or reacted with the carrier gas and stainless steel (Cr$$_{2}$$O$$_{2}$$ layer) to form Cs$$_{2}$$CrO$$_{4}$$ on the former deposited surface. The latter mechanism consequently gave a release of gaseous iodine that was accumulated downstream. After introducing boron to the steam flow, a severe revaporization of iodine deposit at 850 K occurred (more than 70% initial deposit). This was found as a result of the formation of two kinds of cesium borates (Cs$$_{2}$$B$$_{4}$$O$$_{7}$$$$cdot$$5H$$_{2}$$O and CsB$$_{5}$$O$$_{8}$$$$cdot$$4H$$_{2}$$O) which contributed to a large release of gaseous iodine that was capable of reaching outlet of TGT ($$<$$ 400 K). In the case of nuclear severe accident, our study have demonstrated that gaseous iodine could be expected to increase in the colder region of a reactor after late release of boron or a subsequent steam flow after refloods of the reactor, thus posing its near-term risk once leaked to the environment.


Thermophysical properties of austenitic stainless steel containing boron carbide in a solid state

高井 俊秀; 古川 智弘; 山野 秀将

Mechanical Engineering Journal (Internet), 8(4), p.20-00540_1 - 20-00540_11, 2021/08



熱流動とリスク評価,1; リスク評価における熱流動解析の役割

丸山 結; 吉田 一雄

日本原子力学会誌ATOMO$$Sigma$$, 63(7), p.517 - 522, 2021/07



Evaluation of core material energy change during the in-vessel phase of Fukushima Daiichi Unit 3 based on observed pressure data utilizing GOTHIC code analysis

佐藤 一憲; 荒井 雄太*; 吉川 信治

Journal of Nuclear Science and Technology, 58(4), p.434 - 460, 2021/04

 被引用回数:2 パーセンタイル:94.32(Nuclear Science & Technology)

The vapor formation within the reactor pressure vessel (RPV) is regarded to represent heat removal from core materials to the coolant, while the hydrogen generation within the RPV is regarded to represent heat generation by metal oxidation. Based on this understanding, the history of the vapor/hydrogen generation in the in-vessel phase of Fukushima Daiichi Nuclear Power Station Unit 3 was evaluated based on the comparison of the observed pressure data and the GOTHIC code analysis results. The resultant vapor/hydrogen generation histories were then converted to heat removal by coolant and heat generation by oxidation. The effects of the decay power and the heat transfer to the structures on the core material energy were also evaluated. The core materials are suggested to be significantly cooled by water within the RPV, especially when the core materials are relocated to the lower plenum.


Phenomena identification ranking tables for accident tolerant fuel designs applicable to severe accident conditions

Khatib-Rahbar, M.*; Barrachin, M.*; Denning, R.*; Gabor, J.*; Gauntt, R.*; Herranz, L. E.*; Hobbins, R.*; Jacquemain, D.*; 丸山 結; Metcalf, J.*; et al.

NUREG/CR-7282, ERI/NRC 21-204 (Internet), 160 Pages, 2021/04

The U.S. Nuclear Regulatory Commission (NRC) is preparing to accept anticipated licensing applications for the commercial use of accident tolerant fuel (ATF) in commercial nuclear power plants in the United States. It is the objective of the NRC to evaluate the effects of ATF designs on severe accident behavior, and to determine potential changes to the NRC severe accident analysis computer codes that would simulate plant conditions using ATFs commensurate with the accuracy in accident analyses involving conventional fuels. This report documents the development of Phenomena Identification and Ranking Tables (PIRTs) for near-term ATFs under severe accident conditions in light water reactors (LWRs). The PIRTs were developed by a panel of experts for various near-term ATF design concepts (i.e., FeCrAl cladding, zirconium alloy cladding coated with chromium, and Cr$$_{2}$$O$$_{3}$$ dopants in uranium dioxide fuels) in addition to the impacts from fuel enrichment and burnup. Panel members also considered the severe accident implications of the longer-term ATF concepts. The main figures-of-merit considered in this ranking process are the amount of fission products released into the containment and the quantity of combustible gases generated during an accident. Special focus is given to whether existing severe accident codes and models would be sufficient as applied to LWRs employing these fuels, and whether additional experimental studies or model development would be warranted.


Restraint effect of coexisting nitrite ion in simulated high level liquid waste on releasing volatile ruthenium under boiling condition

吉田 涼一朗; 天野 祐希; 吉田 尚生; 阿部 仁

Journal of Nuclear Science and Technology, 58(2), p.145 - 150, 2021/02

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)



Chapter 18, Moving particle semi-implicit method

Wang, Z.; Duan, G.*; 越塚 誠一*; 山路 哲史*

Nuclear Power Plant Design and Analysis Codes, p.439 - 461, 2021/00

The Moving Particle Semi-implicit (MPS) method is one kind of particles methods which are based on Lagrangian approach. It has been developed to analyze complex thermal-hydraulic problems, including those in nuclear engineering. Since meshes are no longer used, large deformation of free surfaces or interfaces can be simulated without the problems of mesh distortion. This approach is effective in solving multiphase fluid dynamics which is subject to complex motion of free surfaces or interfaces. Since its development, MPS method has been extensively utilized for wide range of applications in nuclear engineering. In this chapter, the basic theory of the MPS method is firstly explained. Then, some examples of its application in nuclear engineering, including bubble dynamic, vapor explosion, jet breakup, multiphase flow instability, in-vessel phenomenon, molten spreading, molten core concrete interaction (MCCI) and flooding, are presented.



吉田 尚生; 天野 祐希; 大野 卓也; 吉田 涼一朗; 阿部 仁

JAEA-Research 2020-014, 33 Pages, 2020/12




Experimental study on transport behavior of cesium iodide in the reactor coolant system under LWR severe accident conditions

宮原 直哉; 三輪 周平; Gou$"e$llo, M.*; 井元 純平; 堀口 直樹; 佐藤 勇*; 逢坂 正彦

Journal of Nuclear Science and Technology, 57(12), p.1287 - 1296, 2020/12

 被引用回数:2 パーセンタイル:62.87(Nuclear Science & Technology)



Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF) Phase 2; Results of severe accident analyses for Unit 1

Herranz, L. E.*; Pellegrini, M.*; Lind, T.*; Sonnenkalb, M.*; Godin-Jacqmin, L.*; L$'o$pez, C.*; Dolganov, K.*; Cousin, F.*; 玉置 等史; Kim, T. W.*; et al.

Nuclear Engineering and Design, 369, p.110849_1 - 110849_7, 2020/12

 被引用回数:5 パーセンタイル:76.41(Nuclear Science & Technology)



Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2; Results of severe accident analyses for Unit 2

Sonnenkalb, M.*; Pellegrini, M.*; Herranz, L. E.*; Lind, T.*; Morreale, A. C.*; 神田 憲一*; 玉置 等史; Kim, S. I.*; Cousin, F.*; Fernandez Moguel, L.*; et al.

Nuclear Engineering and Design, 369, p.110840_1 - 110840_10, 2020/12

 被引用回数:6 パーセンタイル:84.6(Nuclear Science & Technology)



Dynamic PRA of flooding-initiated accident scenarios using THALES2-RAPID

久保 光太郎; Zheng, X.; 田中 洋一; 玉置 等史; 杉山 智之; Jang, S.*; 高田 孝*; 山口 彰*

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2279 - 2286, 2020/11

確率論的リスク評価(PRA)は巨大かつ複雑なシステムをリスクを評価する手法の1つである。従来のPRA手法を用いて外部事象のリスクを評価する場合、構造物、系統及び機器の機能喪失時刻の取扱いが困難である。この解決策として、熱水力解析と外部事象評価シミュレーションをRAPID (Risk Assessment with Plant Interactive Dynamics)コードを用いて結合した。外部事象としてPWRプラントにおけるタービン建屋内での内部溢水を選定し、溢水進展評価にはベルヌーイ則に式を用いた。また、溢水源の流量及び緩和設備の没水基準に関する不確実さを考慮した。回復操作については、運転員による溢水源の隔離とポンプによる排水を仮定とともにモデル化した。結果として、隔離操作が排水と組み合わせることによりより有効になることが示された。


Enhancement of the treatment of system interactions in a dynamic PRA tool

田中 洋一; 玉置 等史; Zheng, X.; 杉山 智之

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11

One advantage of dynamic probabilistic risk assessment (PRA) is that it can take into account the timing and ordering of event occurrences based on more explicit simulation of system dynamics. It is expected that dynamic PRA can lead us into a more realistic risk assessment, overcoming some limitations of conventional PRA. Multiple dynamic PRA tools have been developed worldwide, and applied to risk assessment of large industrial facilities such as nuclear power plants and crewed spacecrafts. Japan Atomic Energy Agency has developed the dynamic PRA tool, RAPID (Risk Assessment with Plant Interactive Dynamics), considering the interaction between accident simulation and dysfunctional models of safety-related systems. This paper introduces a recent enhancement of RAPID to treat more complicated simulation interactions from the outside of severe accident codes. It is designed to feed back and forth plant information from simulators to the accident sequence generator. It discusses how the enhancement affects the results of risk assessment, with an example analyzing thermal failure of a safety relief valve in a station blackout accident occurred at a boiling water reactor plant.


The Analysis for Ex-Vessel debris coolability of BWR

松本 俊慶; 岩澤 譲; 安島 航平*; 杉山 智之

Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 10 Pages, 2020/11

本研究では、事前注水した格納容器内デブリの冷却確率を評価した。まず、落下溶融物条件を求めるため、シビアアクシデント解析コードMELCORによる不確かさ解析を行った。この解析では炉心の溶融・移行過程に関連する5つの不確かさパラメータを選択し、仮定された確率分布を用いて、ラテン超方格法(LHS)により入力パラメータセットを生成した。これを用いたMELCORによる多ケース解析の結果から落下溶融物条件を抽出した。次に、MELCOR解析結果をもとに、パラメータの確率分布を決定し、LHSにより生成した59個のパラメータセットを用いてJASMINEコードによる水中の溶融物挙動の解析を行った。水位の条件は0.5m, 1.0m及び2.0mとした。広がり半径とデブリ質量の解析結果からデブリの堆積高さを求め、判定基準と比較することで冷却の成否判定を行った。以上の一連の解析の結果、デブリ冷却の成功確率を求めた。また、MELCOR及びJASMINEを組み合わせた冷却性解析の課題について論じた。


Main findings, remaining uncertainties and lessons learned from the OECD/NEA BSAF Project

Pellegrini, M.*; Herranz, L.*; Sonnenkalb, M.*; Lind, T.*; 丸山 結; Gauntt, R.*; Bixler, N.*; Morreale, A.*; Dolganov, K.*; Sevon, T.*; et al.

Nuclear Technology, 206(9), p.1449 - 1463, 2020/09

 被引用回数:11 パーセンタイル:98.02(Nuclear Science & Technology)

The OECD/NEA Benchmark Study at the Accident of Fukushima Daiichi Nuclear Power Station (BSAF) project, which started in 2012 and continued until 2018, was one of the earliest responses to the accident at Fukushima Daiichi. The project, divided into two phases addressed the investigation of the accident at Unit 1, 2 and 3 by Severe Accident (SA) codes until 500 h focusing on thermal-hydraulics, core relocation, Molten Corium Concrete Interaction (MCCI) and fission products release and transport. The objectives of BSAF were to make up plausible scenarios based primarily on SA forensic analysis, support the decommissioning and inform SA codes modeling. The analysis and comparison among the institutes have brought up vital insights regarding the accident progression identifying periods of core meltdown and relocation, Reactor Pressure Vessel (RPV) and Primary Containment Vessel (PCV) leakage/failure through the comparison of pressure, water level and CAMS signatures. The combination of code results and inspections (muon radiography, PCV inspection) has provided a picture of the current status of the debris distribution and plant status. All units present a large relocation of core materials and all of them present ex-vessel debris with Unit 1 and Unit 3 showing evidences of undergoing MCCI. Uncertainties have been identified in particular on the time and magnitude of events such as corium relocation in RPV and into cavity floor, RPV and PCV rupture events. Main uncertainties resulting from the project are the large and continuous MCCI progression predicted by basically all the SA codes and the leak pathways from RPV to PCV and PCV to reactor building and environment. The BSAF project represents a pioneering exercise which has set the basis and provided lessons learned not only for code improvement but also for the development of new related projects to investigate in detail further aspects of the Fukushima Daiichi accident.


最先端の研究開発,日本原子力研究開発機構,4; 今こそ、高速炉の話; 持続性あるエネルギー供給へ

根岸 仁; 上出 英樹; 前田 誠一郎; 中村 博文; 安部 智之

日本原子力学会誌ATOMO$$Sigma$$, 62(8), p.438 - 441, 2020/08



Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 2; Thermophysical properties of eutectic mixture containing of high concentration boron in a solid state

高井 俊秀; 古川 智弘; 山野 秀将

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

Eutectic melting behavior between boron carbide (B$$_{4}$$C) as control rod material and stainless steel (SS) as structural material and subsequent relocation behavior plays an important role to achieve an in-vessel retention concept which ensures long-term coolability of degraded core under core disruptive accident, because these behaviors are expected to reduce the neutronic reactivity significantly. However, these behaviors have never been simulated in severe accident computer codes before. Since 2016, JAEA has been conducting a research project to develop physical models that describe these behaviors. For the physical models' development, it is necessary to obtain thermophysical properties of SS-B$$_{4}$$C eutectic mixture with various B$$_{4}$$C concentration and maintain them as a database. In this work, the density and specific heat of SS-17 mass%B$$_{4}$$C in a solid state are obtained and compared with these of stainless steel containing 0 and 5 mass%B$$_{4}$$C.

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