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| 1 | Analysis of potential water hammer at the Ignalina NPP using thermal-hydrau- lic and structural analysis codes显示文摘 | USPURAS E KALIATKA A DUNDULIS G | 2001 | Nuclear Engineering and Design2001，，203： | 1 |
| 2 | Benchmarking analysis of water hammer effects using RELAP5 code and development of RBMK -1500 reactor main circulation circuit model显示文摘 | KALIATKA A | 2007 | An- nals of Nuclear Energy2007，34，12： | 1 |
| 3 | State-of-the-art computer code RELAP5 validation with RBMK-related separate phenomena data显示文摘 | URBONAS R USPURAS E KALIATKA A | | 0，，： | 1 |
| 4 | Uncertainty and sensitivity analysis of parameters affecting water hammer pressure wave behaviour显示文摘 | KALIATKA A USPURAS E VAISNORAS M | 2006 | Kemtechnik2006，71，56： | 1 |
| 5 | Development of approach for reliability assessment of pipeline network systems显示文摘 | RIMKEVICIUS S KALIATKA A VALINCIUS M | 2012 | Applied Energy2012，94，10： | 1 |
| 6 | Analysis of potential water hammer at the Ignalina NPP using thermal-hydraulic and structural analysis codes显示文摘 | USPURAS E KALIATKA A DUNDULIS G | 2001 | Nuclear Engineering and Design2001，，203： | 1 |
| 7 | Benchmarking analysis of water hammer effects using RELAP5 code and development of RBMK-1500 reactor main circula- tion circuit model 显示文摘 | Kaliatka A Uspuras E Vaisnoras M | 2007 | Annals of Nuclear Energy2007，，34： | 1 |
| 8 | State-of-the-art computer code RELAP5 validation with REMK-related separate phenomena data显示文摘 | URBONAS R USPURAS E KALIATKA A | 2003 | Nuclear Engineering and Design2003，225，： | 1 |
| 9 | State-of-the-art computer code RELAP5 validation with REMK-related separate phenomena data显示文摘 | URBONAS R USPURAS E KALIATKA A | | 0，，： | 1 |
| 10 | Modelling of valve induced water hammer phenomena in a district heating system显示文摘 | Algirdas Kaliatka Mindaugas Vai?noras Mindaugas Valin?ius | 2014 | Computers and Fluids2014，，： | 1 |
| 11 | Analysis of the Processes in Spent Fuel Pools of Ignalina NPP in Case of Loss of Heat Removal显示文摘 | Kaliatka A Ognerubov V | 2010 | Nuclear Engineering and Desingn2010，240，： | 1 |
| 12 | Benchmarking Analysis of Water Hammer Effects Using RELAP5 Code and Development of RBMK-1500 Reactor Main Circulation Circuit Model显示文摘 | Kaliatka A Uspuras E Vaisnoras M | 2007 | Annals of Nuclear Energy2007，34，12： | 1 |
| 13 | State - of - the - art computer code RELAP5 validation with RBMK related separate phenomena data 显示文摘 | Urbonas R Uspuras E Kaliatka A | 2003 | Nuclear engineering and design2003，225，1： | 1 |
| 14 | Benchmarking analysis of water hammer effects using RELAP5 code and development of RBMK-1500 reactor main circulation circuit model显示文摘 | A. Kaliatka E. Uspuras M. Vaisnoras | 2006 | Annals of Nuclear Energy2006，，1： | 1 |
| 15 | Development of approach for reliability assessment of pipeline network systems显示文摘 | Sigitas Rimkevicius Algirdas Kaliatka Mindaugas Valincius Gintautas Dundulis Remigijus Janulionis Albertas Grybenas Inga Zutautaite | 2012 | Applied Energy2012，，： | 1 |
| 16 | Analysis of the processes in spent fuel pools of ignalina npp in case of loss of heat removal显示文摘 | KALIATKA A OGNERUBOV V VILEINI- SKIS V | 2010 | Nuclear Engineering and Design2010，240，5： | 1 |
| 17 | Integrity control of an RBMK-1500 fuel rod locally oxidized under a bounding reactivity-initiated accident显示文摘In 2007, the license for the second reactor unit of the Ignalina nuclear power plant was renewed considering the safety-related modifications introduced in this reactor.The Safety Analysis Report for this reactor unit was prepared with more strict criteria. The bounding reactivityinitiated accident(RIA) performed by the Lithuanian Energy Institute could be mentioned as an example. The performed analysis demonstrated that even when the worst initial conditions and possible uncertainties are considered,the fuel cladding remains intact. However, the analysis was performed assuming a fresh fuel assembly. In this study, an analysis of the fuel rod cladding behavior in the RBMK-1500 reactor following a bounding RIA is performed using the computational codes FEMAXI-6 and RELAP5. The analysis is extended by modeling an oxide layer(nodular corrosion) on the external surface cladding. An uncertainty and sensitivity analysis was performed using a method developed by the Society for Plant and Reactor Safety,employing the Software for Uncertainty and Sensitivity Analyses, in order to evaluate the effect of the oxide layer on the inside and outside fuel rod temperatures. The results of the thermo-mechanical analysis(stress, strain, and enthalpy) for a local oxide layer with a thickness of 70μm show that despite the exceeded limit of allowed linear power density, the fuel rod is under acceptable safety conditions. | Hamid Boucherit Algirdas Kaliatka Azzedine Lounis | 2019 | Nuclear Science and Techniques2019，30，1： | 0 |
| 18 | RELAP5—3D程序对RBMK-1500反应堆的模型验证显示文摘鉴于堆芯中子响应对RBMK.1500反应堆的重要作用,通过对伊格那林核电站(Ignalina nuclear power plant,INPP)中RBMK-1500反应堆特定瞬态的模拟,验证了RELAP5-3D程序对RBMK.1500的有效性。本文开发了一个适用于RBMK-1500的RELAP5-3D最佳估算模型,其计算结果与INPP的测量数据基本匹配。此外,基于RELAP5-3D最佳估算模型对单独主循环回路预测的热工水力参数和物理过程与RBMK-1500主回路发生的实际过程吻合良好,且计算得出的反应性和堆芯瞬态总功率与电厂的测量值也非常一致,这表明了该程序准确地模拟了堆芯中子响应过程。通过对RELAP5-3D模型的验证表明,该程序可以成功地运用于未来的RBMK-1500安全性计算。 | E. Uspuras A. Kaliatka E. Bubelis 洪钢(译) 陈军(校) | 2007 | 国外核动力2007，28，2： | 0 |
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