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Gintautas Dundulis Eugenijus Uspuras Ronald F. Kulak Algirdas Marchertas 《Nuclear Engineering and Design》2007,237(8):848-857
Stress corrosion cracks have been discovered in Group Distribution Headers (GDH) at the Ignalina and Chernobyl Nuclear Power Plants. This increases the probability that a guillotine pipe break can occur that creates a whipping pipe (GDH) with the potential to damage surrounding structures—i.e. adjacent GDH and its attached piping or adjacent reinforced concrete compartment wall. The GDH is the most important component for reactor safety in case of an accident. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the ECSS into the GDH.Presented in this paper is the transient analysis of a Group Distribution Header following a guillotine break at the blind end of the header. Using a very conservative force loading function, the transient response of a whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is obtained using finite element methodology.The results of the study, assuming that the impacted GDH does not suffer stress corrosion cracking, indicate that the structural integrity of the compartment should be maintained and failure should not propagate from GDH to GDH. 相似文献
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王智文 《上海电力学院学报》2010,(4)
针对GDH(group diffie-hellman)方案中节点可能成为系统的瓶颈以及计算复杂度、通信代价和存储复杂度远高于某些集中式方案等缺陷,提出并实现了一种基于优化GDH协商的高效安全的动态群组密钥管理方案,并对其安全性进行了证明。通过对计算量和通信量进行分析比较表明,优化GDH协商协议具有很大的优势,并且能够快速产生或更新组密钥,具有很强的实用性。 相似文献
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The integral analysis of severe accident scenario for RBMK-1500 was performed using combined approach with RELAP5, RELAP/SCDAPSIM, ASTEC and COCOSYS codes. The performed analysis covered response of the reactor core, the reactor cooling system and the confinement. There were performed several analyses: the first analysis assumed that operators take no action or their actions are not successful to provide the coolant injection to the reactor core; the other analyses were performed to investigate the accident management measures to restore the core cooling at different temperatures of the reactor core. The results of performed analyses showed that the operators have ∼5 h before the ruptures of fuel claddings occur and ∼8 h before the onset of exothermic steam-zirconium reaction. The coolant injection to the reactor core should be restored as soon as possible in order to prevent high hydrogen concentrations in the confinement and significant release of the fission products to the environment. 相似文献
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一个基于椭圆曲线的可证明安全签密方案* 总被引:1,自引:0,他引:1
签密能够在一个合理的逻辑步骤内同时完成数字签名和加密两项功能。与实现信息保密性和认证性的先签名后加密方案相比,签密具有较低的计算和通信代价。提出一个基于椭圆曲线的签密方案,能够同时完成数字签名和加密两项功能。基于可证明安全性理论,在GDH(gap Diffie-Hellman)问题难解的假设之下,该方案在随机预言机模型中被证明是安全的。该方案能够抵御自适应选择明文/密文攻击。 相似文献
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分布式协同设计系统中,信息交换的基本安全保障是信息发送者身份的真实可靠性和信息的机密性、真实性、完整性.为了在提供认证的同时保护信息发送者的匿名性,基于GDH群,提出了一个适用于分布式协同设计环境的基于身份的环签名方案,新方案采用引进两个独立PKG的方法,在一定程度上消除了现存方案中单个PKG可以随意伪造用户签名的安全隐患,实现了签名者无条件匿名性,达到了匿名认证的目的. 相似文献
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Gintautas Dundulis Renatas Karalevicius Sigitas Rimkevicius Ronald F. Kulak Algirdas H. Marchertas 《Nuclear Engineering and Design》2006,236(2):201-210
The Ignalina NPP has a pressure suppression type of confinement, which is referred to as the accident localization system (ALS). The ALS prevents the release of the radioactive material from the NPP to the environment during a loss-of-coolant accident (LOCA). Ten water pools are located in the two ALS towers (five pools in each tower), which separate the dry well from the wet well. These water pools condense the accident-generated steam and prevent high overpressures in the compartments.The steam distribution device (SDD), with the vertical vent pipes (nozzles) that are inserted under the water of the condensing pools, connects the dry well and the wet well. In case of an accident, these components must be capable of withstanding the dynamic loads generated by a LOCA for successful pressure suppression function.This paper presents the transient analysis of the SDD and their connections to the vertical steam corridors following a LOCA. A thermo-hydraulic analysis of the SDD was performed using the state-of-the-art COCOSYS code to determine pressure and temperature histories resulting from a LOCA. The finite element code NEPTUNE was used to evaluate the structural integrity of the SDD and its supporting reinforced concrete wall. Results show that, although portions of the SDD undergo plastic response and the outside surface of the vertical steam corridor reinforced concrete wall cracks, the structural integrity of the SDD and wall are maintained during a LOCA. 相似文献
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