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論文

Major outcomes through recent ROSA/LSTF experiments and future plans

竹田 武司; 和田 裕貴; 柴本 泰照

World Journal of Nuclear Science and Technology, 11(1), p.17 - 42, 2021/01

Many experiments have been conducted on accidents and transients of pressurized water reactor (PWR) employing the rig of safety assessment/large-scale test facility (ROSA/LSTF). Major results of the related integral effect tests with the LSTF were reviewed to experimentally identify thermal-hydraulic phenomena involved, regarding the PWR accident sequences in accordance with the new regulatory requirements for the Japanese light-water nuclear power plants. Key results of the recent integral effect tests utilizing the LSTF and future plans were presented relevant to multiple steam generator tube rupture accident with recovery operation, small-break loss-of-coolant accident (LOCA) with accident management measure on core exit temperature reliability, and small-break LOCA with thermal stratification under cold water injection from emergency core cooling system into cold legs.

論文

ROSA/LSTF test on nitrogen gas behavior during reflux condensation in PWR and RELAP5 code analyses

竹田 武司; 大津 巌

Mechanical Engineering Journal (Internet), 5(4), p.18-00077_1 - 18-00077_14, 2018/08

We conducted an experiment focusing on nitrogen gas behavior during reflux condensation in PWR with ROSA/LSTF. The primary pressure was lower than 1 MPa under constant core power of 0.7% of volumetric-scaled (1/48) PWR nominal power. Steam generator (SG) secondary-side collapsed liquid level was maintained at certain liquid level above SG U-tube height. Nitrogen gas was injected stepwise into each SG inlet plenum at certain constant amount. The primary pressure and degree of subcooling of SG U-tubes were largely dependent on amount of nitrogen gas accumulated in SG U-tubes. Nitrogen gas accumulated from outlet towards inlet of SG U-tubes. Non-uniform flow behavior was observed among SG U-tubes with nitrogen gas ingress. The RELAP5/MOD3.3 code indicated remaining problems in predictions of the primary pressure and degree of subcooling of SG U-tubes depending on number of nitrogen gas injection. We studied further non-uniform flow behavior through sensitivity analyses.

論文

ROSA/LSTF tests and posttest analyses by RELAP5 code for accident management measures during PWR station blackout transient with loss of primary coolant and gas inflow

竹田 武司; 大津 巌

Science and Technology of Nuclear Installations, 2018, p.7635878_1 - 7635878_19, 2018/00

 被引用回数:2 パーセンタイル:33.97(Nuclear Science & Technology)

Three tests were carried out with LSTF, simulating accident management (AM) measures during PWR station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total-failure of high-pressure injection system. As AM measures, steam generator (SG) depressurization was done by fully opening relief valves, and auxiliary feedwater was injected into secondary-side simultaneously. Conditions for break size and onset timing of AM measures were different. Primary pressure decreased to below 1 MPa with or without primary depressurization by fully opening pressurizer relief valve. Nonuniform flow behaviors were observed in SG U-tubes with gas ingress depending on gas accumulation rate in two tests that gas accumulated remarkably. The RELAP5/MOD3.3 code indicated remaining problems in predictions of primary pressure, SG U-tube liquid levels, and natural circulation mass flow rates after gas inflow and accumulator flow rate through analyses for two tests.

論文

RELAP5 uncertainty evaluation using ROSA/LSTF test data on PWR 17% cold leg intermediate-break LOCA with single-failure ECCS

竹田 武司; 大津 巌

Annals of Nuclear Energy, 109, p.9 - 21, 2017/11

 被引用回数:4 パーセンタイル:51.35(Nuclear Science & Technology)

An experiment was conducted for the OECD/NEA ROSA-2 Project using LSTF, which simulated a cold leg intermediate-break loss-of-coolant accident with 17% break in a PWR. Assumptions were made such as single-failure of high-pressure and low-pressure injection systems. In the LSTF test, core dryout took place because of rapid drop in the core liquid level. Liquid was accumulated in upper plenum, SG U-tube upflow-side and inlet plena because of counter-current flow limiting (CCFL). The post-test analysis by RELAP5/MOD3.3 code revealed that peak cladding temperature (PCT) was overpredicted because of underprediction of the core liquid level due to inadequate prediction of accumulator flow rate. We found the combination of multiple uncertain parameters including the Wallis CCFL correlation at the upper core plate, core decay power, and steam convective heat transfer coefficient in the core within the defined uncertain ranges largely affected the PCT.

論文

Experimental study on outer surface cooling of containment vessel by using CIGMA

柴本 泰照; 石垣 将宏; 安部 諭; 与能本 泰介

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 14 Pages, 2017/09

The present paper introduces the recent outcome from the CIGMA experiments regarding containment vessel cooling, in which an external side of a vessel upper head was flooded by water. The test vessel was initially pressurized by steam and noncondensable gas (air and/or helium), and was subsequently cooled by pouring water to the outside of the vessel top. Similar experiments were performed by authors using air-steam binary system in the previous study, which showed several characteristic phenomena such as inverse temperature stratification. The experimental conditions were extended systematically in this study to investigate the effects of initial gas composition and distribution in a vessel. The measurement results indicated that natural circulation was significantly affected by distributions of each gas species. In particular, it was enhanced when the gas density became heavier after condensation on the vessel inner wall, while it was suppressed when the gas density became lighter, creating density stratification with helium-rich gas in the upper region. The results are explained by the simplified model.

論文

ROSA/LSTF test and RELAP5 analyses on PWR cold leg small-break LOCA with accident management measure and PKL counterpart test

竹田 武司; 大津 巌

Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08

 被引用回数:3 パーセンタイル:41.49(Nuclear Science & Technology)

An experiment using PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with LSTF on a cold leg small-break loss-of-coolant accident with an accident management measure in a PWR. The rate of steam generator secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

論文

ROSA/LSTF test on nitrogen gas behavior during reflux cooling in PWR and RELAP5 post-test analysis

竹田 武司; 大津 巌

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 11 Pages, 2017/07

An experiment focusing on nitrogen gas behavior during reflux cooling in a PWR was performed with the LSTF. The test conditions were made such as the constant core power of 0.7% of the volumetric-scaled PWR nominal power and the primary pressure of lower than 1 MPa. The steam generator (SG) secondary-side collapsed liquid level was maintained at a certain liquid level above the SG tube height. Nitrogen gas was injected stepwise into each SG inlet plenum at a certain constant amount. The primary pressure and the SG U-tube fluid temperatures were greatly dependent on the amount of nitrogen gas accumulated in the SG U-tubes. Non-uniform flow behavior was observed among the SG U-tubes with nitrogen gas ingress. The RELAP5/MOD3.3 code indicated remaining problems in the predictions of the primary pressure and the SG U-tube fluid temperatures after nitrogen gas inflow.

論文

Outcome of first containment cooling experiments using CIGMA

柴本 泰照; 与能本 泰介; 石垣 将宏; 安部 諭

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 10 Pages, 2016/10

The Japan Atomic Energy Agency (JAEA) initiated the ROSA-SA project in 2013 for the purpose of studying thermal hydraulics relevant to over-temperature containment damage, hydrogen risk, and fission product transport. For this purpose, the JAEA newly constructed the Containment InteGral Measurement Apparatus (CIGMA) in 2015 for the experiments addressing containment responses, separate effects, and accident managements. Recently, we successfully conducted first experiments using CIGMA to characterize the facility under typical experimental conditions. Among these experiments, the present paper focuses on the results of containment cooling tests, for which an upper part of the vessel outer surface was cooled by spray water. Several distinctive phenomena were observed in the tests, including inverse temperature stratification in the vessel due to the cooling in the upper region. The RELAP5 analysis result was also presented to roughly indicate the prediction capability of the best-estimate two-phase flow code in predicting the containment thermal hydraulics.

論文

大型装置CIGMAを用いた格納容器熱水力安全研究; 重大事故の評価手法と安全対策の高度化を目指して

柴本 泰照; 与能本 泰介; 堀田 亮年*

日本原子力学会誌ATOMO$$Sigma$$, 58(9), p.553 - 557, 2016/09

日本原子力研究開発機構安全研究センターでは、シビアアクシデント対策の強化を特徴とする新しい安全規制を支援するため、2013年にROSA-SA計画を開始し、今般、本計画の中核となる大型格納容器実験装置CIGMA(Containment InteGral Measurement Apparatus)を完成させた。CIGMAは、設計温度や計測点密度において世界有数の性能を有しており、シビアアクシデント時の格納容器内の事故進展挙動や事故拡大防止に係る熱水力実験を実施することができる。本稿では、本計画と既往研究の概要を述べるとともに、CIGMA装置の特徴、及びこれまで実施した一連の実験結果を紹介する。

論文

First experiments at the CIGMA facility for investigations of LWR containment thermal hydraulics

柴本 泰照; 安部 諭; 石垣 将宏; 与能本 泰介

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 9 Pages, 2016/06

There has been an extensive reorientation of the light water reactor research in Japan since the Fukushima Dai-ichi Nuclear Power Station accident, which focuses on severe accidents and accident managements. The Japan Atomic Energy Agency (JAEA) initiated the ROSA-SA project in 2013 for the purpose of studying thermal hydraulics relevant to over-temperature containment damage, hydrogen risk, and fission product transport. For this purpose, the JAEA newly constructed the Containment InteGral Measurement Apparatus (CIGMA) in 2015 for the experiments addressing containment responses, separate effects, and accident managements. Recently, we successfully conducted first experiments using CIGMA to characterize the facility under typical experimental conditions investigating basic phenomena such as buildup of pressure by steam injection, containment cooling and depressurization by internal or external cooling, and density stratified layer mixing by impinging jet. This paper provides an overview of the research programs, the brief description of the facility specification and the outcomes obtained from the first experiments.

論文

Contributions of OECD ROSA & ROSA-2 Projects for thermal-hydraulic code validation

中村 秀夫

Proceedings of Seminar on the Transfer of Competence, Knowledge and Experience gained through CSNI Activities in the Field of Thermal-Hydraulics (THICKET 2016) (CD-ROM), 29 Pages, 2016/06

軽水炉事故時の熱水力現象の詳細解明と安全評価コードの解析性能向上を図るため、アジアでの初めてのOECD-NEA国際共同研究プロジェクトであるROSAおよびROSA-2プロジェクトが、原子力機構の主催、15ヶ国の参加で行われた。同プロジェクトでは2005年より約8年間にわたり、世界最大規模の軽水炉模擬実験装置LSTFを用いて、9課題19回の実験が行われた。一方、OECD-NEAでは、その活動によって得られた成果や経験を次代に伝えるセミナ活動THICKETを実施しており、今回その第4回において、ROSAプロジェクトで得られた成果のうち、特に安全評価コードの性能検討に焦点を当てた取り組みであるブラインド計算の成果ならびに、同時に実施されていたOECD-PKL2プロジェクトとの相互比較実験の成果など、主要な成果から得られた教訓およびLSTF実験の有効性を中心に同プロジェクトの全容を解説し、安全解析における残存課題を議論する。

論文

ROSA/LSTF experiment on a PWR station blackout transient with accident management measures and RELAP5 analyses

竹田 武司; 大津 巌

Mechanical Engineering Journal (Internet), 2(5), p.15-00132_1 - 15-00132_15, 2015/10

An experiment on a PWR station blackout transient with accident management (AM) measures was conducted using the ROSA/LSTF under an assumption of non-condensable gas inflow to primary system from accumulator tanks. The AM measures considered are SG secondary-side depressurization by fully opening safety valves in both SGs with start of core uncovery and coolant injection into the secondary-side of both SGs at low pressures. The primary pressure started to decrease when the SG primary-to-secondary heat removal resumed soon after the SG secondary-side coolant injection. The primary depressurization worsened due to the gas accumulation in SG U-tubes after accumulator completion. The RELAP5 code indicated remaining problems in the predictions of the SG U-tube collapsed liquid level and primary mass flow rate after gas ingress. The SG coolant injection flow rate was found to significantly affect the peak cladding temperature and the ACC actuation time through RELAP5 sensitivity analyses.

論文

RANS analyses on erosion behavior of density stratification consisted of helium-air mixture gas by a low momentum vertical buoyant jet in the PANDA test facility, the third international benchmark exercise (IBE-3)

安部 諭; 石垣 将宏; 柴本 泰照; 与能本 泰介

Nuclear Engineering and Design, 289, p.231 - 239, 2015/08

 被引用回数:12 パーセンタイル:79.58(Nuclear Science & Technology)

Density stratification in the reactor containment vessel is an important phenomenon on an issue of hydrogen safety. The Japan Atomic Energy Agency (JAEA) has started the ROSA-SA project on containment thermal hydraulics. As a part of the activity, we participated in the third international CFD benchmark exercise (IBE-3) focused on density stratification erosion by a vertical buoyant jet in containment vessel. This paper shows our approach for the IBE-3, focusing on the turbulence transport phenomena in eroding the density stratification and introducing modified turbulence models for improvement of the CFD analyses. For this analysis, we modified the CFD code OpenFOAM by using two turbulence models; the Kato and Launder modification to estimate turbulent kinetic energy production around a stagnation point, and the Katsuki model to consider turbulence damping in density stratification. As a result, the modified code predicted well the experimental data. The importance of turbulence transport modeling is also discussed using the calculation results.

論文

Thermal hydraulic safety research at JAEA after the Fukushima Dai-ichi Nuclear Power Station accident

与能本 泰介; 柴本 泰照; 竹田 武司; 佐藤 聡; 石垣 将宏; 安部 諭; 岡垣 百合亜; 孫 昊旻; 栃尾 大輔

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5341 - 5352, 2015/08

This paper summarizes thermal-hydraulic (T/H) safety studies being conducted at JAEA based on the consideration of research issues after the Fukushima Dai-Ichi Nuclear Power Station accident. New researches have been initiated after the accident, which are related to containment thermal hydraulics and accident management (AM) measures for the prevention of core damage under severe multiple failure conditions. They are conducted in parallel with those initiated before the accident such as a research on scaling and uncertainty of the T/H phenomena which are important for the code validation. Those experimental studies are to obtain better understandings on the phenomena and establish databases for the validation of both lumped parameter (LP) and computational fluid dynamics (CFD) codes. The research project on containment thermal hydraulics is called the ROSA-SA project and investigates phenomena related to over-temperature containment damage, hydrogen risk and fission product (FP) transport. For this project, we have designed a large-scale containment vessel test facility called CIGMA (Containment InteGral Measurement Apparatus), which is characterized by the capability of conducting high-temperature experiments as well as those on hydrogen risk with CFD-grade instrumentation of high space resolution. This paper describes the plans for those researches and results obtained so far.

論文

A Study on improvement of RANS analysis for erosion of density stratified layer of multicomponent gas by buoyant jet in a containment vessel

安部 諭; 石垣 将宏; 柴本 泰照; 与能本 泰介

Journal of Energy and Power Engineering, 9(7), p.599 - 607, 2015/07

格納容器内での多成分ガスで形成される密度成層を精度よく解析することはシビアアクシデントの安全評価の上で重要である。日本原子力研究開発機構は格納容器内熱水力現象調査を目的としてROSA-SAプロジェクトを開始した。このプロジェクトの一環として、我々は浮力ジェットによる密度成層の侵食および崩壊についれ数値流体力学(CFD)解析を実行した。その解析では、既往研究でよく使われているが密度成層の侵食・崩壊を過大予測するRANS解析の改善を試みた。具体的には、低Re型k-$$varepsilon$$モデルをベースとして、ジェットの成層への貫入部分での乱流エネルギーを適切に評価、密度成層内での乱流抑制効果を再現するための改良をほどこした。RANS解析の結果は、計算コストは莫大になるものの精度が高いとされるLES解析と比較をおこなった。その結果、密度成層の侵食・崩壊について、本研究で適用した改良型のモデルは従来モデルよりもLES解析とのよく一致した。

論文

ROSA/LSTF experiment on accident management measures during a PWR station blackout transient with pump seal leakage and RELAP5 analyses

竹田 武司; 大津 巌

Journal of Energy and Power Sources, 2(7), p.274 - 290, 2015/07

An experiment on accident management (AM) measures during a PWR station blackout transient with leakage from primary coolant pump seals was conducted using the ROSA/LSTF under an assumption of non-condensable gas inflow to the primary system from accumulator tanks. The AM measures are steam generator (SG) secondary-side depressurization by fully opening safety valves (SVs) in both SGs and primary-side depressurization by fully opening SV in pressurizer with the start of core uncovery and coolant injection into the SG secondary-side at low pressures. The decrease was accelerated in the primary pressure when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes after accumulator completion. Remaining problems in the RELAP5 code include the predictions of pressure difference between the primary and SG secondary sides after the gas inflow.

論文

RELAP5 code study of ROSA/LSTF experiments on PWR safety system using steam generator secondary-side depressurization

竹田 武司; 大貫 晃*; 西 弘昭*

Journal of Energy and Power Engineering, 9(5), p.426 - 442, 2015/05

RELAP5 code analyses were performed on two ROSA/LSTF validation tests for PWR safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using SG (steam generator) secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place. Core collapsed liquid level recovered after the initiation of accumulator coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection system. Adjustment of break discharge coefficient for two-phase discharge flow predicted the break flow rate reasonably well. The code overpredicted the peak cladding temperature because of underprediction of the core collapsed liquid level due to inadequate prediction of the accumulator flow rate in the 8-in. break case.

論文

ROSA/LSTF experiment on AM measures during a PWR station blackout transient with pump seal leakage and RELAP5 POST-TEST analysis

竹田 武司; 大津 巌

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05

An experiment on accident management (AM) measures during a PWR station blackout transient with the TMLB' scenario and leakage from primary coolant pump seals was conducted using the ROSA/LSTF under an assumption of non-condensable gas inflow to the primary system from accumulator tanks. The AM measures are steam generator (SG) secondary-side depressurization by fully opening safety valves (SVs) in both SGs and primary-side depressurization by fully opening SV in pressurizer with the start of core uncovery and coolant injection into the SG secondary-side at low pressures. The decrease was accelerated in the primary pressure when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes. The RELAP5 code indicated remaining problems in the predictions of the primary pressure and SG U-tube collapsed liquid level.

論文

RELAP5 code study of ROSA/LSTF validation tests for PWR safety system using SG secondary-side depressurization

竹田 武司; 大貫 晃*; 西 弘昭*

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 13 Pages, 2014/12

RELAP5 code post-test analyses were performed on two ROSA/LSTF validation tests for PWR safety system that simulated cold leg small-break LOCAs using SG secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves a little after a safety injection signal. In the 8-in. break test, core uncovery and heatup took place by boil-off. Core collapsed liquid level recovered after accumulator coolant injection. In the 4-in. break test, no core uncovery and heatup happened. Adjustment of break discharge coefficient for two-phase discharge flow predicted the break flow rate reasonably well. The code overpredicted the peak cladding temperature (PCT) because of underprediction of the core collapsed liquid level in the 8-in. break case. Sensitivity analyses indicated that a time delay for SG depressurization start and break discharge coefficient for two-phase discharge flow affect the PCT significantly in the 8-in. break case.

論文

RELAP5 analyses of ROSA/LSTF experiments on AM measures during PWR vessel bottom small-break LOCAs with gas inflow

竹田 武司

International Journal of Nuclear Energy, 2014, p.803470_1 - 803470_17, 2014/00

RELAP5 code analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break LOCAs with different AM measures under an assumption of non-condensable gas inflow. Depressurization of and auxiliary feedwater (AFW) injection into both steam generators (SGs) as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI) system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow.

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