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吉本 達哉*; 石川 淳; 岡本 孝司*; 丸山 結
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 5 Pages, 2014/11
Containment Filtered Vent System (CFVS) can prevent the containment vessel failure and reduce the release amount of fission products to the environment. After the Fukushima Daiichi Nuclear Power Plant (NPP) accident, Nuclear Regulation Authority of Japan required licensees to introduce CFVS in all BWR plant in Japan to re-operate nuclear power plant. In this paper, a CFVS model was constructed and the impact of CFVS on source term during a severe accident was analyzed. A main objective in this analysis is to understand potential effect of CFVS on source terms. A series of calculations has been performed for a BWR with Mark-I containment vessel (CV) at present by considering only scrubbing efficiency in a water pool using THALES2 code. A severe accident initiated by station blackout is taken to be analyzed, which is similar with the accident at the Fukushima Daiichi NPP. Four cases; drywell CFVS case, wet well (W/W) CFVS case, W/W direct release case and CV failure case, are analyzed.
鳴戸 健一*; 西野 裕之; 栗坂 健一; 山野 秀将; 岡野 靖; 岡村 茂樹*; 衛藤 将生*
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 10 Pages, 2014/11
Seismic Probabilistic Risk Assessment (PRA) is increasingly important in assessing the safety of nuclear power plants after the TEPCO's Fukushima Daiichi Nuclear Power Station accident. In this study, a seismic PRA under a rated power operation was performed for Japan Sodium cooled Fast Reactor (JSFR) developed by Japan Atomic Energy Agency. This PRA was intended to examine seismic impacts on the JSFR by calculating the Core Damage Frequency (CDF) with the identification of all the accident sequences induced by earthquake which may have potential possibility of direct core damage. Seismic hazard data was based on assessment results for existing nuclear site locations in Japan. Seismic fragility needed to quantify the accident sequences was set based on existing assessments for similar equipment. The base-case analysis showed that the total CDF would be approximately 10
/reactor-year and JSFR is robust against the earthquake in the range of this assessment. The dominant contributor (about 80%) to the CDF is direct core damage by the sequence of simultaneous failures of reactor vessel and guard vessel. Sensitivity analysis was performed focusing on the simultaneous failures of reactor vessel and guard vessel. This result suggested that enhancement of failure probability assessment for guard vessels and/or provision of measures for maintaining coolant level following reactor vessel failure would be effective to reduce the CDF.
山野 秀将; 金子 哲也*; 相澤 康介; 田中 正暁; 江原 真司*; 橋爪 秀利*
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 10 Pages, 2014/11
本研究では、商用流体力学解析コードを用いてレイノルズ応力モデルによるU-RANSアプローチを高レイノルズ数(1.0
10
)条件でのシングル及びダブルエルボ流れに適用した。この数値シミュレーションにはJSFRのコールドレグ配管を模擬した1/7縮尺水試験を用いて、3次元的に接続されたダブルエルボ体系における流動状況を調べた。結果として、数値解析と試験結果との比較によって、U-RANSシミュレーション手法の妥当性と適用性を確認した。また、シミュレーションにより渦構造を含む非定常流動場は第1エルボと第2エルボでは非常に異なることが示された。これは1段エルボで生成された馬蹄渦が2段エルボに横向きに流入されるからである。さらに、シミュレーションによって本論文で扱ったダル部エルボ流れにおけるエルボ間距離の影響は小さいことが示された。
永武 拓; 吉田 啓之; 高瀬 和之; 倉田 正輝
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 6 Pages, 2014/11
As for improvement of Severe Accident (SA) codes, a detailed core relocation process has to be clarified because higher prediction results are obtained by numerical methods in the SA codes. Especially, it is very important to understand precisely the melting behavior of fuel elements. Then we have been developing the numerical method for analyzing fundamental melting behavior of the fuel elements based on the original POPCORN code. The POPCORN code was developed in Japan Atomic Energy Agency (JAEA) based on the Moving Particle Semi-implicit (MPS) method. In the present study, we introduced the numerical analysis methods for phase change, surface tension and multi components to the original POPCORN code. This paper shows outline of simulation methods and results of numerical simulations. From the results, melting behavior called candling can be simulated numerically.
西村 正弘; 檜山 智之; 上出 英樹; 大島 宏之; 長澤 一嘉*; 今井 康友*
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11
The detailed flow velocity distribution in the edge subchannel has been obtained by PIV measurement using a wire-wrapped 3-pin bundle water model. These flow field data like flow velocity distribution and fluctuation intensity near the wrapping wire are available for code validation.
堀口 直樹; 吉田 啓之; 金川 哲也*; 金子 暁子*; 阿部 豊*
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 5 Pages, 2014/11
A Multi Venturi Scrubber System (MVSS) is one of the filtered venting devices, and used for existing reactors in Europe. One of the main components of the MVSS is a self-priming Venturi Scrubber (VS). It is considered that a dispersed or dispersed annular flow is formed in the VS by a self-priming phenomena. In the self-priming phenomena, the liquid was suctioned from a surrounding region to the inside of the VS. And a part of the radioactive materials in the gas are eliminated to the liquid through the gas-liquid interface of the dispersed or annular dispersed flow. Therefore, to consider the MVSS operation characteristics, it is important to evaluate the liquid flow rate of the self-priming of the VS including the occurrence condition of the self-priming. In this paper, to understand the self-priming phenomena of the VS for the filtered venting, theoretical preliminary analysis for evaluating liquid flow rate of the self-priming of the VS and experiment to observe self-priming phenomena and measure the liquid flow rare of the self-priming were performed. By comparing these results, we discussed about the mechanism of the self-priming phenomena. As a result, the self-priming phenomena in the VS was observed. In addition, at a higher gas mass flux, the suspension of the self-priming phenomena was confirmed both experimentally and theoretically.
小野 綾子; 田中 正暁; 小林 順; 上出 英樹
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11
The flow induced vibration (FIV) is concerned at the hot-leg piping of JSFR because of large Reynolds number in it. The pressure fluctuation in the piping causes FIV as the excitation force, which relates to velocity fluctuation. In this study, the measurement of pressure fluctuation in the elbow piping revealed that inlet velocity condition affect to the area of separation flow and the pressure fluctuation generated by the traversing of separation region. And, the measurement revealed that the pressure fluctuation with higher frequency significantly attenuated during the flow passed in the elbow.
萩原 裕之*; 江連 俊樹; 伊藤 啓; 上出 英樹
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11
A water test visualizing vortex cavitation inception have been carried out using 1/11 scale upper plenum model of Japan Sodium-cooled Fast Reactor. The influence of the splitter on the vortex cavitation has been examined. In addition to that, a measurement test for flow velocity distribution around the vortex has been performed by means of the particle imaging velocimetry (PIV). In the visualization test, the vortex cavitation inception occurs much more easily in the case without the flow splitter than that with the flow splitter. Besides, it has been observed that the influence of kinematic viscosity coefficient on the vortex cavitation inception decreases by installing flow splitter. As a result of the measurement test, the correlation has been found between the magnitude of circulation and the vortex cavitation inception.
江連 俊樹; 伊藤 啓; 亀山 祐理*; 萩原 裕之*; 上出 英樹
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11
A water experiment in the 1/22 scaled upper plenum model of JSFR was performed to clarify behavior of vortex cavitation occurrences. Velocity distributions between H/L intakes and R/V wall were obtained under various H/L inlet velocity and temperature conditions. Pressure drop at vortex center was evaluated from obtained velocity distributions based on the Burgers vortex model. Evaluated value of pressure drop in time series was also compared with temporal behavior of vortex cavitation occurrences. As the results, circulation around vortex was quantified and time-averaged normalized circulation was observed to be nearly constant independent on the variation of Reynolds number at H/L pipe intakes. The evaluated value of pressure drop showed qualitatively consistent behavior to occurrences of vortex cavitation in time series. Consequently, it is confirmed that occurrences of vortex cavitation can be predicted by means of the evaluation method based on Burgers vortex model.
Shen, X.*; 日引 俊*; 孫 昊旻; 中村 秀夫
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 10 Pages, 2014/11
新型沸騰水型原子炉ESBWRの炉心上部のチムニーなど大口径の垂直矩形流路では、気液2相流の液相速度勾配や気泡分布は大口径円管と異なる可能性がある。一方、大口径の矩形流路内については、ボイド率や液相速度、気泡径などの分布に関する報告はあるが、3次元気泡速度や界面積濃度等の流れ構造を詳細に示すパラメータについては報告がなされていない。本研究では、安全解析に用いる流動モデルの信頼性向上等に資するため、それらの詳細計測を行った。実験では、1辺が100mmの正方形断面テスト部を用いて鉛直上昇流のボイド率、界面積濃度、気泡径や3次元気泡速度など主要パラメータの断面内分布を4センサ光プローブを用いて計測し、液流量の増加に伴って、ボイド率と界面積濃度の分布が壁面ピークから管中心ピークに変化することや、主流方向の気泡速度が管中心ピーク分布を持つ結果を得た。更に、断面内の気泡速度分布の結果から、断面内循環流が対称8分の1三角形領域に存在し、液流量に伴って速度が増加することを明らかにした。
内堀 昭寛; 大島 宏之
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 6 Pages, 2014/11
Na冷却高速炉の蒸気発生器において伝熱管破損時に形成される隣接伝熱管周りのウェステージ環境を評価するため、Na側で生じる圧縮性多成分多相流及びNa-水化学反応を対象とした機構論的数値解析コードSERAPHIMを開発している。本研究では、ウェステージ環境評価モデルの一つとして液滴エントレインメント・輸送モデルを開発し、その基本検証のため基礎的な液滴エントレインメント実験の解析を実施した。その結果、本解析モデルが液滴エントレインメント終了までの時間や液滴発生時の圧力変動等の実験結果を概ね良好に再現し、モデルの基本的な妥当性を確認した。
伊藤 啓; 江連 俊樹; 大島 宏之; 河村 拓己*; 中峯 由彰*
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 6 Pages, 2014/11
ナトリウム冷却高速炉における液中渦キャビテーション研究の一環として、半径方向圧力分布と蒸気圧、気液界面における圧力ジャンプ条件に基づいてキャビティ半径を計算する表面張力モデルの開発を行い、液中渦キャビテーション評価手法を改良した。基本検証として基礎実験を対象とした評価を行い、流速の増加によって渦中心圧力が低下してキャビティ半径が大きくなるという現象が定性的に再現できることを確認した。また、液中渦キャビテーション挙動に大きな影響を与える粘性について評価を行った。
高田 孝*; 小中 祐至*; 山口 彰*; 伊藤 啓; 大野 修司; 大島 宏之
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11
本研究では、高速炉炉心のエントランスノズル部における不活性ガス気泡挙動を理論および数値解析によって評価する。まず、FLUENTコードを用いて3次元流動場の計算を行い、低濃度気泡流に対して適用できるOne-way気泡追跡法による3次元気泡追跡解析を行った。その結果に基づいて、ナトリウム冷却高速炉の高圧プレナム部における気泡蓄積挙動を定量的に評価できるモデルを開発し、フローネットワークコードSYRENAによって評価を行った。さらに、気泡除去(ガス抜き)装置について検討を行うため、様々な対策構造の効果について定量的な評価を行った。
Jang, S.*; 高田 孝; 山口 彰*; 内堀 昭寛; 栗原 成計; 大島 宏之
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 8 Pages, 2014/11
ナトリウム冷却高速炉の蒸気発生器において伝熱管の微小亀裂が徐々に拡大するセルフウェステージ現象について、多次元解析コードSERAPHIMを用いた数値解析(定量化)を実施し、拡大後の亀裂サイズを調査した。まず、ナトリウム-水反応によって生じるセルフウェステージ現象を再現するための複数ステップの数値解析手法を考案した。これに基づき、亀裂近傍の化学反応を伴う熱流動特性を得るための2次元解析を実施した。伝熱管壁のウェステージ量は温度と水酸化ナトリウム濃度をパラメータとしたアレニウス型評価式から求め、ウェステージ後の構造部形状に整合するように解析メッシュを再構築した。以上の手順を繰り返すことで評価した亀裂拡大後サイズ及び形状が実験結果と一致することを確認した。