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論文

Improvement of JASMINE code for ex-vessel molten core coolability in BWR

松本 俊慶; 川部 隆平*; 岩澤 譲; 杉山 智之; 丸山 結

Annals of Nuclear Energy, 178, p.109348_1 - 109348_13, 2022/12

シビアアクシデント時の溶融物関連事象を評価するためにFCIコードであるJASMINEの機能拡張を行った。溶融物の冷却性評価ではキャビティ床面上における粒子状・アグロメレーション・ケーキ状デブリ質量割合や最終的な幾何形状の予測が必要である。アグロメレーションモデルでは、熱を保有した粒子同士のくっつきを考慮し、組み込んだ。もう一つのモデル改良は拡がりモデルの改良である。浅水方程式を導入し、拡がり先端部のクラスト成長に基づく拡がり停止条件を組み込んだ。調整係数の最適化のためにスウェーデンKTHにおいて実施されたDEFOR-A及びPULiMS実験を参照した。JASMINEコードによる実験解析では共通のパラメータセットで良い再現性が得られ、主要な現象は適切にモデル化されたことを示した。

論文

Modelling and simulation of the source term for a sodium cooled fast reactor under hypothetical severe accident conditions; Final report of a coordinated research project

Arokiaswamy, J. A.*; Batra, C.*; Chang, J. E.*; Garcia, M.*; Herranz, L. E.*; Klimonov, I. A.*; Kriventsev, V.*; Li, S.*; Liegeard, C.*; Mahanes, J.*; et al.

IAEA-TECDOC-2006, 380 Pages, 2022/09

IAEAの共同研究プロジェクト「シビアアクシデント条件下におけるプロトタイプナトリウム冷却高速炉からの放射性物質の放出」は、シビアアクシデント条件下における、リファレンスとなるナトリウム冷却高速炉の施設内における放射性物質および燃料粒子の存在量の、異なる時間スケールにおける現実的な数値シミュレーションを目的として実施された。解析のスコープは3つに分割され、3つのワークパッケージ(WP)として定義された:(1)炉内ソースターム評価、(2)一次系/格納容器系境界のソースターム評価、および(3)格納容器内の現象分析。WP-1の参加機関の結果を比較すると、希ガスおよび放射性セシウムの放出割合、およびカバーガスへの放射性核種の放出割合はよく一致した。共通の圧力履歴を用いたWP-2の解析では、各機関の解析結果はよく一致し、いずれの機関の解析手法も同程度の精度を有することが示された。先行するWPにおける解析と分離するため、放出割合をあらかじめ設定したスタンドアロンケースがWP-3にて定義された。WP-3の全参加機関によるスタンドアロンケースの解析結果は全般的に一致した。

論文

BWR lower head penetration failure test focusing on eutectic melting

山下 拓哉; 佐藤 拓未; 間所 寛; 永江 勇二

Annals of Nuclear Energy, 173, p.109129_1 - 109129_15, 2022/08

 被引用回数:0 パーセンタイル:0.02(Nuclear Science & Technology)

Decommissioning work occasioned by the Fukushima Daiichi Nuclear Power Station (1F) accident of March 2011 is in progress. Severe accident (SA) analysis, testing, and internal investigation are being used to grasp the 1F internal state. A PWR system that refers to the TMI-2 accident is typical for SA codes and testing, on the other hand, a BWR system like 1F is uncommon, understanding the 1F internal state is challenging. The present study conducted the ELSA-1 test, a test that focused on damage from eutectic melting of the liquid metal pool and control rod drive (CRD), to elucidate the lower head (LH) failure mechanism in the 1F accident. The results demonstrated that depending on the condition of the melt pool formed in the lower plenum, a factor of LH boundary failure was due to eutectic melting. In addition, the state related to the CRD structure of 1F unit 2 were estimated.

論文

ナトリウム冷却高速炉の炉心損傷初期過程の研究(過出力時炉停止失敗事象に対するSAS4Aコードの妥当性確認)

石田 真也; 深野 義隆

日本機械学会論文集(インターネット), 88(911), p.21-00304_1 - 21-00304_11, 2022/07

炉心損傷事故(CDA)の初期の段階である起因過程の評価に係る解析コードSAS4Aに関しては、これまでにCDAの代表的な事象である流量喪失時炉停止失敗事象(ULOF)に対してPIRT手法を適用し、評価手法の信頼性向上が図られている。本研究では、PIRT手法を用いてUTOPの分析を行って物理現象を抽出するとともに、それらの物理現象にランク付けを行って8つの重要現象を抽出し、ULOFとの違いを明らかにした。さらに、抽出した重要現象に対して評価マトリクスを作成し、評価マトリクスに沿って妥当性確認を行った。評価マトリクスの作成においては、UTOPの重要現象に対してULOFの評価マトリクスで網羅されていない部分に対して妥当性確認を行った。本研究によって、SAS4Aをより広範な事故事象へ適用することが可能となり、当該コードの信頼性を大きく向上させることができた。

論文

Status of the uncertainty quantification for severe accident sequences of different NPP-designs in the frame of the H-2020 project MUSA

Brumm, S.*; Gabrielli, F.*; Sanchez-Espinoza, V.*; Groudev, P.*; Ou, P.*; Zhang, W.*; Malkhasyan, A.*; Bocanegra, R.*; Herranz, L. E.*; Berda$"i$, M.*; et al.

Proceedings of 10th European Review Meeting on Severe Accident Research (ERMSAR 2022) (Internet), 13 Pages, 2022/05

The current HORIZON-2020 project on "Management and Uncertainties of Severe Accidents (MUSA)" aims at applying Uncertainty Quantification (UQ) in the modeling of Severe Accidents (SA), particularly in predicting the radiological source term of mitigated and unmitigated accident scenarios. Within its application part, the project is devoted to the uncertainty quantification of different severe accident codes when predicting the radiological source term of selected severe accident sequences of different nuclear power plant designs, e.g. PWR, VVER, and BWR. Key steps for this investigation are, (a) the selection of severe accident sequences for each reactor design, (b) the development of a reference input model for the specific design and SA-code, (c) the selection of a list of uncertain model parameters to be investigated, (d) the choice of an UQ-tool e.g. DAKOTA, SUSA, URANIE, etc., (e) the definition of the figures of merit for the UA-analysis, (f) the performance of the simulations with the SA-codes, and, (g) the statistical evaluation of the results using the capabilities, i.e. methods and tools offered by the UQ-tools. This paper describes the project status of the UQ of different SA codes for the selected SA sequences, and the technical challenges and lessons learnt from the preparatory and exploratory investigations performed.

論文

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.

論文

Revolatilization of iodine by bubbly flow in the suppression pool during an accident

南上 光太郎; 石川 淳; 杉山 智之; Pellegrini, M.*; 岡本 孝司*

Journal of Nuclear Science and Technology, 10 Pages, 2022/04

 被引用回数:0 パーセンタイル:0.02(Nuclear Science & Technology)

To appropriately evaluate the amount of radioactive iodine released into the environment, we extended the current pool scrubbing model to consider revolatilization at bubble surfaces due to bubbly flow generated in the suppression pool, and the effect of revolatilization by bubbly flow was quantitatively evaluated using a station black out sequence in this work. Gaseous iodine species are produced in the suppression pool in an accident. They are gradually released from the pool surface, but when a large amount of gas flows from the drywell into the suppression pool, the revolatilization of gaseous iodine dissolved in the pool water is promoted by bubbly flow. The results of this study indicated that the release amount of iodine immediately after suppression chamber (S/C) vent operation increased by up to 134 times when considering the revolatilization effect associated with bubbly flow. These results were due to the increase in the gas-liquid interfacial area at bubble surfaces and the overall mass transfer coefficients under two-phase flow conditions due to bubbly flow. It was shown that caution is required for early S/C vent operation.

論文

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

木村 郁仁*; 山村 聡太*; 藤原 広太*; 吉田 啓之; 齋藤 慎平*; 金子 暁子*; 阿部 豊*

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

 被引用回数:1 パーセンタイル:0.02(Nuclear Science & Technology)

A new three-dimensional laser-induced fluorescent (3D-LIF) technology to obtain the hydrodynamic behavior of liquid jets in a shallow pool were developed. In this technology, firstly, a refractive index matching was applied to acquire a clear cross-sectional image. Secondly, a series of cross-sectional images was obtained by using a high-speed galvanometer scanner. Finally, to evaluate the unsteady 3D interface shape of liquid jet, a method was developed to reconstruct 3D shapes from the series of cross-sectional images obtained using the 3D-LIF method. The spatial and temporal resolutions of measurement were 4.7 $$times$$ 4.7 $$times$$ 1.0 lines/mm and 25 $$mu$$s, respectively. The shape of a 3D liquid jet in a liquid pool and its impingement, spreading and atomization behavior were reconstructed using the proposed method, successfully. The behaviors of atomized particles detached from the jet were obtained by applying data processing techniques. Diameters distribution and position of atomized droplets after detachment were estimated from the results.

論文

France-Japan collaboration on thermodynamic and kinetic studies of core material mixture in severe accidents of sodium-cooled fast reactors

Quaini, A.*; Goss$'e$, S.*; Payot, F.*; Suteau, C.*; Delacroix, J.*; Saas, L.*; Gubernatis, P.*; Martin-Lopez, E.*; 山野 秀将; 高井 俊秀; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04

CEAと原子力機構は、現在の実施体制の下で新しいサブタスク、炉心混合物質における相互作用の動力学、炉心混合物質の物性、UO$$_{2}$$-Fe-B$$_{4}$$C系の高温熱力学データ、B$$_{4}$$C-SS速度論およびB$$_{4}$$C-SS共晶材料の再配置(凍結)に関する実験的研究、SIMMERコードシステムのB$$_{4}$$C/SS共晶および動力学モデル、混合物の液相化速度をモデル化するための方法論を定義した。この論文は、以前の実施協定の下での日仏協力で得られた主要な研究開発結果と、現在の協定の下での実験的および分析的ロードマップについて説明する。

論文

French-Japanese experimental collaboration on fuel-coolant interactions in sodium-cooled fast reactors

Johnson, M.*; Delacroix, J.*; Journeau, C.*; Brayer, C.*; Clavier, R.*; Montazel, A.*; Pluyette, E.*; 松場 賢一; 江村 優軌; 神山 健司

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04

シビアアクシデントに関する日仏共同実験の一環として、ナトリウム冷却高速炉の原子炉容器内下部プレナムへ溶融燃料が流出した時の燃料-冷却材相互作用について、その解明に向けた研究を実施している。MELT施設では、ナトリウム中へ流出したキログラム単位の模擬溶融炉心物質が急冷される様子をX線で可視化することができる。現在準備中のSERUA施設では、融体と冷却材の接触境界面温度が上昇した場合の沸騰熱伝達を評価するためのデータ取得を予定している。この論文では、これらの施設を活用した実験協力の現状について紹介する。

論文

Modelling and simulation of source term for sodium-cooled fast reactor under hypothetical severe accident; Primary system/containment system interface source term estimation

小野田 雄一; John Arul, A.*; Klimonov, I.*; Danting, S.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 13 Pages, 2022/04

Three Work Packages were defined in this Coordinated Research Project whose objective was to estimate fission-product-transportation behavior inside the reference pool-type sodium-cooled fast reactor. This WP, WP-2, is dedicated to estimate the primary system/containment system interface source term using improved models and tools. The mass of primary sodium instantaneously ejected via leak paths onto the top shield was evaluated as a common benchmark problem which will be the input for the subsequent WP, WP-3. The exercises were carried out for a reference pool type SFR of 1250 MWth capacity with mixed oxide fuel. The accident sequence to be considered is Unprotected Loss of Flow Accident which is assumed to result in a core damage with release of radionuclides into the primary coolant and cover gas. Four organizations, NCEPU (China), IBRAE RAN (Russian Federation), IGCAR (India) and JAEA (Japan) finally participated in this WP. Reference case calculation using a common pressure history and sensitivity study were carried out. The total amount of the ejected sodium onto the top shield for reference case was in a good agreement between the participants. The results of the sensitivity study revealed that the change of the parameters regarding uncertainty bring about the change of leaked mass in the range of several tens of %.

論文

Cesium chemistry in the LWR severe accident and towards the decommissioning of Fukushima Daiichi Nuclear Power Station

逢坂 正彦; Gou$"e$llo, M.*; 中島 邦久

Journal of Nuclear Science and Technology, 59(3), p.292 - 305, 2022/03

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

事故時及び長期間の2つの期間のソースタームにおけるセシウム化学について、福島第一原子力発電所(1F)事故後に行われたFP化学研究のレビューを行った。事故時についてはCsのMo, B, Siとの化学反応について、また1F固有の水相を介した長期についてはCsのコンクリートへの浸透及び燃料デブリの浸出挙動について、関連する熱力学データ整備状況とともに調べた。これらのCs化学挙動は近い将来取出し予定の燃料デブリ等1Fサンプルの分析及び評価を通して検証されるべきである。

論文

Release behaviors of elements from an Ag-In-Cd control rod alloy at temperatures up to 1673 K

永瀬 文久; 大友 隆; 上塚 寛*

Nuclear Technology, 208(3), p.484 - 493, 2022/03

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

Ag-In-Cd制御棒合金をアルゴンあるいは酸素中、1073-1673Kで60-3600s間加熱し、元素放出挙動を調べた。1123Kと1173Kの間の温度で合金の明らかな液化が起こるが、それ以下の温度では元素放出は少なかった。アルゴン中では、1173Kで3600s後に、1573Kでは60s後にほぼ全てのCdが放出されたが、AgとInの放出割合はそれぞれ3%以下及び8%以下であった。酸素中では、1573K以下でのCd放出は非常に少ないが、1673Kでは短時間に30-50%が放出された。調べた範囲では酸素中のAgとInの放出は少なかった。実験結果との比較から、従来の経験モデルはシビアアクシデント時に制御棒が破損した直後に相当する比較的低い温度範囲でCdの放出を過小評価している可能性がある。

論文

Flame structures and ignition thresholds of hydrogen jets containing sodium mist under various gas concentrations

土井 大輔; 清野 裕; 宮原 信哉*; 宇埜 正美*

Journal of Nuclear Science and Technology, 59(2), p.198 - 206, 2022/02

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

Non-premixed combustion of hydrogen jets containing sodium vapor and mist reduces threats to reactor containment integrity in sodium-cooled fast reactors (SFRs) because it gradually consumes hydrogen gas generated mainly by a reaction between sodium and concrete. Previous studies have been limited to experimentally determining ignition thresholds on the jet temperature and the sodium concentration under specific gas concentrations. In this study, ignition experiments on hydrogen jets containing sodium mist were carried out at a specific jet temperature and sodium concentration under various gas concentration conditions (1-15vol% hydrogen and 3-21vol% oxygen). As a result, a stable sodium flame was observed in the jet and then formed a lifted hydrogen flame from a fuel nozzle outlet. An attached hydrogen flame on the outlet was also formed under high hydrogen concentration conditions. These flame structures seemed to be attributed to hydrogen flame propagation, which depends on the hydrogen concentration, jet temperature, and jet velocity. Additionally, the experimental results revealed ignition thresholds on the gas concentration and indicated a flammable region where the hydrogen-sodium jet combustion was more advantageous than an explosive premixed hydrogen combustion. Our study will enable the advancement of safety assessment technology in SRFs.

報告書

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

JAEA-Research-2021-011.pdf:1.49MB

再処理施設における高レベル濃縮廃液の蒸発乾固事故について、ルテニウム(Ru)の挙動が着目されている。Ruは四酸化ルテニウム(RuO$$_{4}$$)のような揮発性の化学種を形成し、硝酸、水または窒素酸化物を含む共存ガスと共に施設外へ放出される可能性があるためである。本研究では、蒸発乾固事故に対する安全性評価に資することを目的として、事故時の蒸気凝縮を模擬した、水溶液に対する気体状RuO$$_{4}$$の液相への移行挙動を実験的に測定した。その結果、RuO$$_{4}$$のガス吸収は液相中の亜硝酸(HNO$$_{2}$$)濃度の増加により促進されたことから、化学吸収を伴う物質移動であることが示唆された。HNO$$_{2}$$を用いない対照実験では、温度が低いほど液相中のRu吸収率は大であったのに対し、HNO$$_{2}$$を用いた実験では、温度が高いほどRu吸収率が高かった。これは化学吸収に関与する化学反応が高温で活性化されたためであると考察される。

論文

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

 被引用回数:0 パーセンタイル:0.02(Nuclear Science & Technology)

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.

論文

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

 被引用回数:0 パーセンタイル:0(Chemistry, Multidisciplinary)

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.

論文

Melt impingement on a flat spreading surface under wet condition

Sahboun, N. F.; 松本 俊慶; 岩澤 譲; 杉山 智之

Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 15 Pages, 2021/10

The accident at the Fukushima Daiichi Nuclear Power Station triggered reevaluation and necessary enhancement of the accident countermeasures and safety regulations worldwide. Such actions are based on the present knowledge and evaluation techniques of the important phenomena anticipated to occur in a severe accident. The present study focused on the under-water melt spreading behavior and aimed at a formulation to predict the final geometry of the solidified melt on the floor of the containment vessel. The formulation, based on the author's previous study of the dry spreading of molten metal, considers the thermal and fluid properties of the melt, so the gap between the core and simulant materials could be filled by using adequate properties. In addition, the formulation was extended to the wet condition by considering the film boiling heat transfer at the upper side of the spreading melt. The improved formula was applied to the PULiMS experiments conducted by the Swedish Royal Institute of Technology with a simulant oxide material under wet conditions. The predicted final spreading area and thickness were in agreement with the experimental results within a twenty percent error.

論文

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

炉心損傷事故時には、制御棒材である炭化ホウ素と構造材であるステンレス鋼が共晶反応を起こし、ステンレス鋼の融点より低い温度で溶融(液化)すると考えられる。こうして生成された制御棒溶解材は流動性があるため、崩壊炉心内を広範に移行し、崩壊炉心物質に混ざり込むことで、崩壊炉心物質の反応度抑制に顕著な効果をもたらすと考えられる。しかしながら、このような制御棒溶解材の共晶溶融反応やその移行挙動については、これまでの重大事故解析では何ら考慮されていない。本研究では、シビアアクシデント解析コードの高度化に資するため、炭化ほう素溶解量の異なる制御棒溶解材について固相物性測定を実施し、温度(及び炭化ホウ素濃度)依存性を示す物性評価式として整備した結果について報告する。

論文

Numerical investigations on the coolability and the re-criticality of a debris bed with the density-stratified configuration

Li, C.; 内堀 昭寛; 高田 孝; Pellegrini, M.*; Erkan, N.*; 岡本 孝司*

第25回動力・エネルギー技術シンポジウム講演論文集(インターネット), 4 Pages, 2021/07

溶融炉心の原子炉容器内保持を達成するためには、デブリベッドの安定冷却と再臨界回避が重要である。本研究では、異なる物質から構成され、密度成層化したデブリベッドの挙動を評価するため、数値流体力学(CFD),個別要素法(DEM),モンテカルロ法を連成させた解析手法を構築した。本解析手法により、デブリベッドにおける密度成層化の挙動等を解析できることを確認した。

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