共查询到19条相似文献,搜索用时 406 毫秒
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相比传统大型核电厂,微型反应堆各系统功能间紧密耦合且相互制约,传统的分专业解耦设计模式难以应对,需开展全范围的系统仿真。采用Modelica语言建立了气冷式微型反应堆的系统仿真模型,以未能紧急停堆的预期瞬态(ATWS)事故为例开展事故分析计算,并与专业堆芯安全分析结果对比,结果表明反应堆功率变化趋势较为一致,且ATWS事故后仅依靠堆芯温度升高引入的负反应性可实现停堆。本文研究方法为气冷式微型反应堆的全系统建模仿真打下了坚实基础,也为其他类型反应堆的系统建模仿真提供了很好的借鉴作用。 相似文献
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极限的未能紧急停堆的预期瞬态(ATWS)是核电厂二次侧热移出能力减小引起的升温瞬态。为评价AP1000核电厂在发生ATWS事故后的响应,采用LOFTRAN程序对极限的丧失主给水ATWS进行计算分析。对影响电厂系统响应的一些关键因素,如蒸汽旁排的容量、堆芯补水箱(CMT)特性和硼反应性系数、反应堆冷却剂泵(RCP)可用性、启动给水系统(STS)可用性和蒸汽发生器(SG)传热等作了一系列敏感性分析。分析结果表明:为缓解ATWS事故,应隔离蒸汽旁排,并在触发CMT的同时停运RCP。 相似文献
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压水堆核电厂高压熔堆严重事故序列分析 总被引:3,自引:3,他引:0
压水堆核电厂的高压熔堆事故覆盖了大部分的严重事故序列,且具有很大的潜在威胁。根据我国900MW压水堆核电厂的概率安全分析(PSA)结果选取了丧失厂外电、未能紧急停堆的预期瞬态、二回路管道破口、一回路小破口和蒸汽发生器传热管破裂5种典型的高压熔堆严重事故序列,并使用SCDAP/RELAP5程序对这些事故序列的进程和后果进行了计算分析。计算结果表明:5种典型高压熔堆事故序列可能导致高压熔喷和安全壳直接加热风险,可能引起安全壳早期失效,很有必要采取相应的一回路卸压措施。 相似文献
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秦山一期核电站未能紧急停堆的预期瞬变导致堆芯熔化的进程及事故缓解措施研究 总被引:1,自引:0,他引:1
选择失去主给水、失去厂外电和正常运行情况下控制棒失控提升3个典型的导致未能紧急停堆的预期瞬变(ATWS)的初因事故,采用自行研制的基于SCDAP/RELAP5/MOD3.1的核反应堆严重事故分析平台,对秦山一期核电站ATWS初因导致堆芯熔化严重事故进程进行了分析研究,对防止ATWS导致堆芯熔化进程的缓解措施的有效性进行了验证。计算分析结果表明,二回路补水和一回路卸压的事故缓解措施能有效地阻止堆芯熔化进程。 相似文献
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本文描述了在未能紧急停堆的预期瞬变(ATWS)事故工况下应急初始条件及应急行动水平在PWR核电厂和CANDU核电厂的应用,并对这两种类型核电厂在ATWS事故工况下相同应急初始条件的应急行动水平的不同进行了比较. 相似文献
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未能紧急停堆的预期瞬态(ATWS)缓解系统是保证中国先进研究堆(CARR)安全的重要系统之一。当发生预期运行瞬态,反应堆未能紧急停堆时,通过ATWS缓解系统动作实现停堆,从而保护反应堆安全。ATWS缓解系统的高可靠性是保证其完成预期功能的重要条件,因此对该系统的可靠性给予了高度重视。本文以ATWS缓解系统为研究对象,利用故障模式及影响分析和故障树等可靠性分析方法,建立相应模型,对ATWS缓解系统进行了定性和定量的分析,得到了ATWS缓解系统发生故障的概率和最小割集,找出了薄弱环节,提出了改进措施和建议,其可靠性水平已达到CARR工程的设计要求,验证了设计,为CARR其他系统分析和验证奠定了基础。 相似文献
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Performance evaluation of KAERI’s advanced integral reactor against an anticipated transient without scram has been carried out with the transients and setpoint simulation/small and medium reactor code, by considering a decrease in the heat removal by the secondary system, a loss of offsite power and an inadvertent control rod withdrawal event as an initiating event. In a decrease in the heat transfer by the secondary system and a loss of offsite power, the reactor coolant system pressures can be maintained below 110% of the design pressure during the transition period due to the effect of the large negative moderator temperature coefficient. On the other hand, in an inadvertent control rod withdrawal event, the pressure of the reactor coolant system increases up to the ASME service level C stress limit due to a high reactivity insertion into a reactor core by the adoption of a boron free core concept. Therefore, a hardware installation against an anticipated transient without scram is essential to mitigate the consequences resulting from an inadvertent control rod withdrawal event. A diverse protection system, which is an independent and diverse reactor shutdown system that is initiated by the signals of a high core power or a high pressurizer pressure, is adopted in the advanced integral reactor. According to the reassessment results by considering the diverse protection system for a reactor shutdown, the diverse protection system is helpful in mitigating the consequences of an anticipated transient without scram. 相似文献
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F. Chen Y. Dong Z. Zhang Y. Zheng L. Shi S. Hu 《Nuclear Engineering and Design》2009,239(6):1010-1018
Safety demonstration tests were conducted on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the transient data of reactor core and primary cooling system for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 3 MW power level on October 15, 2003. This paper simulates and analyzes the power transient and the thermal response of the reactor during the test by using the THERMIX code. The analytical results are compared with the test data for validation of the code.Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shut down after the stop of the helium circulator; the subsequent phenomena such as the recriticality and power oscillations are also studied. During the test a natural circulation loop of helium is established in the core and the other coolant channels and its consequent thermal response such as the temperature redistribution is investigated. In addition, temperatures of the measuring points in the reactor internals are calculated and compared with the measured values. Satisfactory agreements obtained from the comparison demonstrate the basic applicability and reasonability of the THERMIX code for simulating and analyzing the helium circulator trip ATWS test. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature during the test is always lower than 1600 °C which is the limited value for the HTGR. 相似文献
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Shih-Jen Wang Chun-Sheng Chien Sheng-Yuan Fann Show-Chyuan Chiang 《Nuclear Engineering and Design》2007,237(22):2197-2200
Boron injection initiation temperature (BIIT) provides important information for the safe shutdown of the reactor using boron injection system during anticipated transient without scram (ATWS). The purpose of this paper is to study BIIT curve of boiling water reactor owners’ group (BWROG). The unreasonable and non-conservative parts of BIIT are pointed out and suggested modifications are made. The starting reactor power of BIIT is increased in order to meet the actual application. The lower limit of suppression pool temperature of BIIT is revised for conservative operation during ATWS conditions. Analysis of the effects of maximum temperature capacity of the suppression chamber and concentration of boron in standby liquid control tank shows that BIIT is decreased by adopting a more conservative value of maximum temperature capacity of the suppression chamber. Consequently, early boron injection is anticipated. For system with automatic boron injection system, BIIT is not required. 相似文献
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Two tests initiated by unscrammed control rod withdrawal were performed on the High Temperature GasCooled Reactor-Test Module(HTR-10) in November 2003 after the reactor achieved its full power, and the test conditions represented a typical transient scenario of modular high-temperature reactors(HTRs), called pressurized loss of forced cooling, and anticipated transient without scram.Based on the test parameters, the HTR-10 thermal behaviors under the test conditions were studied with the help of the system analysis code THERMIX. The combination of the test results and the investigation results makes the HTR-10 safety potential better understood. Key phenomena, such as the helium natural circulation and the temperature redistribution in the reactor, were revealed. As the safety feature of most significance, there is a large margin between the maximum fuel temperature and its safety limit in each test. Temperatures of thermocouples in different components were calculated by THERMIX and compared with the test values. The applicability of the code was verified by good agreement obtained from the comparison. 相似文献
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Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes.Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results.Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests.The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW.The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased. 相似文献
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Recently, digital instrumentation and control systems have been increasingly installed for important safety functions in nuclear power plants such as the reactor protection system (RPS) and the actuation system of the engineered safety features. Since digital devices consist of not only electronic hardware but also software that can control microprocessors, the functions specific to digital equipment such as self-diagnostic functions have been becoming available. These functions were not realized with conventional electric components. On the other hand, it has been found that it is difficult to model the digital equipment reliability in probabilistic risk assessment (PRA) using conventional fault tree analysis technique. OECD/NEA CSNI Working Group of Risk Assessment (WGRisk) set up the task group DIGREL to develop the basis of reliability analysis method of the digital safety system and is now discussing about several issues including quantitative dynamic modeling. This paper shows that, taking account of the relationship among the RPS failures, demand after the initiating event, detection of the RPS fault by self-diagnostic or surveillance tests, repair of the RPS components and plant shutdown operation by the plant operators as a stochastic process, the anticipated transient without scram (ATWS) event can be modeled by the event logic fault tree and Markov state-transition diagrams assuming the hypothetical 1-out-of-2 digital RPS. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(3):511-517
Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large-scale liquid metal cooled fast breeder reactor (LMFBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the stability of SASS needs to be confirmed under the actual reactor-operational environment with high temperature, high neutron flux, and flowing sodium to ensure the high plant availability factor. For this purpose, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor Joyo MK-III. The rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation. 相似文献