波动管小破口失水事故实验研究

在模块化小型反应堆非能动安全系统综合模拟实验装置上进行波动管小破口尺寸失水事故实验,研究波动管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。模块化小型反应堆发生失水事故后,压力平衡管和安注管线内流体的密度差可以驱动堆芯补水箱(CMT)内的冷流体注入反应堆压力容器,压力平衡管裸露后CMT安注流量出现波动;安注箱(ACC)的安注对事故初期的堆芯冷却效果显著;经自动卸压系统卸压后,内置换料水箱(IRWST)可以对堆芯进行持续稳定的安注和冷却。研究结果表明:波动管小破口失水事故中,非能动安注系统可以对堆芯进行有效注水,并带走堆芯衰变热量。     

核电厂堆芯冷却恶化恢复安注措施研究

本文使用RELAPSCDAPSIM3.4程序建立核电厂事故分析模型,选取了典型的中、小冷段破口事故作为分析序列,针对堆芯冷却恶化现象采取恢复安注措施进行了详细的热工水力计算。着重分析了在辅助给水有效情况下,重启安注的时间窗口、启动上充应对安注失效情况下的有效性、有无安注箱注入敏感性等。分析结果表明:当堆芯出口温度超过923K(即650℃),恢复安注建立应急堆芯冷却流量措施对于中、小破口是有效的;启动上充对较小破口效果明显;安注箱有效注入对中破口冷却恶化事故缓解有重要作用。     

主泵两相降级对大破口失水事故的影响研究

大破口失水事故过程中,主泵的工作范围覆盖了单相液、气液两相和单相气工况。在两相工况下,主泵的扬程和转矩发生降级。对于AP1000核电厂,WCOBRA/TRAC被用于大破口失水事故分析,其现有的主泵两相降级数据来源于西屋W93A主泵。为正确模拟AP1000主泵在大破口失水事故过程中的热工水力特性,需对其两相降级特性进行研究。本研究分别采用国际上广泛使用的SEMISCALE和EPRI/CE主泵的两相降级数据进行AP1000冷段双端断裂事故的计算分析,并与原有W93A的计算结果进行对比。结果表明,AP1000主泵两相降级特性对反应堆冷却剂系统压力、破口流量和安注箱流量影响不大。相比于SEMISCALE和EPRI/CE,现有的W93A的两相降级数据将导致更低的堆芯冷却流量和更高的包壳峰值温度最大值,计算结果相对偏于保守。     

非能动堆芯冷却系统LOCA下冷却能力分析

本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统（PXS）中非能动余热排出系统（PRHRS）、堆芯补水箱（CMT）、安注箱（ACC）、自动卸压系统（ADS）和安全壳内置换料水箱（IRWST）等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477 K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。     

安注箱对小型模块化压水堆LOCA的影响研究

基于小型模块化压水堆失水事故(LOCA)的现象和特征,分析燃料包壳峰值温度(PCT)出现的机理,并研究安注箱(ACC)设置对PCT和事故长期降压的影响。结果表明:在一定范围中等破口尺寸LOCA下,ACC注水可能导致堆芯更不利的后果,小型堆可合理考虑ACC的作用进行系统设计。     

压水堆核电站大破口失水事故分析

压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件,大破口失水事故是核电站运行的设计基准事故,是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算,对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值（PCT）最高,且长时间维持在较高温度,此条件下反应堆最危险。计算结果表明,事故发生后,一回路压力迅速下降,堆芯冷却剂的流动性变差,导致堆芯裸露,燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作,能保证燃料元件包壳温度不超过1204 ℃的限值。     

压水堆核电站大破口失水事故分析

压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件,大破口失水事故是核电站运行的设计基准事故,是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算,对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值(PCT)最高,且长时间维持在较高温度,此条件下反应堆最危险。计算结果表明,事故发生后,一回路压力迅速下降,堆芯冷却剂的流动性变差,导致堆芯裸露,燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作,能保证燃料元件包壳温度不超过1 204℃的限值。     

反应堆冷却剂泵水力特性对大破口失水事故的影响研究

以AP1000为研究对象,应用WCOBRA/TRAC程序对大破口失水事故进行模拟.主要分析4种不同的主泵特性曲线对系统压力、破口流量及包壳峰值温度的影响.研究结果表明,大破口失水事故下,由于主泵特性曲线的差异,导致喷放阶段及再淹没阶段的峰值包壳温度相差近150℃.通过合理优化或改进主泵特性可以为核电厂大破口失水事故带来更大的安全裕量.     

AP1000小破口事故下非能动氮气安注箱的瞬态特性研究

根据AP1000非能动氮气安全注入水箱的结构和工作原理建立了热工水力模型并开发了计算分析程序TACAP。利用TACAP计算得到了AP1000非能动氮气安全注入水箱在两种小破口失水事故（包括25.4 cm等效直径冷管破口和5.08 cm等效直径冷管破口）下的瞬态特性,得到了箱内水位及注入流量等关键参数的瞬态变化。计算结果表明：安注箱在小破口失水事故后能提供高效的安全注入,对一回路快速地进行冷却和降压,有效地缓解事故后果。TACAP计算结果与西屋公司NOTRUMP程序计算结果基本一致,表明了TACAP程序的适用性和正确性。     

DVI安注技术在三环路压水堆的应用研究

本文针对某三环路压水堆核电厂提出采用压力容器直接注入（DVI）技术的安注系统简化设计方案。采用RELAP5/SCDAP-SIM和CATHARE-GB程序分别分析了对安注系统功能执行要求最高的DVI管道断裂事故及反应堆冷却剂冷管段双端剪切断裂事故，研究该简化设计方案的可行性和有效性。分析结果表明，在DVI管道发生断裂的事故工况下，安注系统有足够的冗余性保证燃料温度不会明显上升。在RCS主管道发生大破口包括双端剪切断裂事故工况下，通过DVI注入可明显延长安注箱注入时间、提高安注箱水装量的有效利用率，仅利用中压安注泵和安注箱即可完成安注功能的执行，不再需要低压安注子系统。DVI技术的应用大大简化了安注系统设计，电厂的安全性和经济性得到共同提升。     

基于TRACE程序的大破口BEPU分析方法在独立审核计算中的应用

我国某大功率非能动压水堆首次采用标准设计的方法。本文基于TRACE程序,采用最佳估算及不确定性分析方法对该电厂大破口失水事故开展了独立审核计算,并探讨了该方法在核安全审评中的应用。最佳估算结果表明,该压水堆大破口失水事故下包壳峰值温度低于验收准则限值。该计算结果可作为独立审核计算的重要部分应用于核电厂安全审评中。     

事故容错燃料在大破口事故下的安全分析

事故容错燃料（ATF）系统旨在当反应堆失去冷却后,提高核燃料及包壳的安全特性,在正常工况下相比现在的UO₂-Zr系统更好。通过凭借先进材料的特性,ATF系统会明显延缓事故进程,为采取缓解措施提供更大的时间裕度。本文通过分析采用ATF的典型压水堆系统大破口事故（LBLOCA）设计基准事故以及叠加安注系统失效的极限严重事故,初步评估ATF在事故下的性能。分析结果表明,相比UO₂-Zr,ATF能降低大破口设计基准事故下的包壳峰值,延长严重事故下堆芯发生熔化的时间,具有更好的事故容错性。     

风险指引的CPR1000核电厂LBLOCA分析方法初步研究

随着核电厂安全分析方法的不断发展,结合传统确定论分析与概率风险评价（PSA）的风险指引型安全分析方法逐渐引起安审当局和核电业主的广泛关注。本文基于国际上风险指引型分析方法在其他领域的应用现状,提出了风险指引的大破口失水事故（LBLOCA）分析方法,并重新评估了CPR1000核电厂的堆芯燃料包壳峰值温度（PCT）裕量。在PSA分析中,识别并量化了LBLOCA发生后可能发生的162个事件序列,并采用确定论现实分析方法（DRM）对筛选出的18个概率较大的事件序列进行了计算分析。然后通过期望值评估法和特定序列覆盖法对LBLOCA的PCT裕量进行了评估。结果表明,本文方法下LBLOCA的PCT裕量约为36~55 ℃,相比于传统的DRM裕量提升了16~35 ℃。     

Numerical simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA.     

基于PSA的压水堆LBLOCA不确定性分析

为了结合确定论与概率论分析开展更加真实的核反应堆事故工况安全分析,提出了一种结合概率安全分析（PSA）和最佳估算加不确定性（BEPU）分析的方法,并以典型三环路压水堆冷管段双端断裂大破口失水事故（LBLOCA）的极限事故为对象,首先基于PSA开展了应急堆芯冷却系统的事故失效分析,而后结合BEPU分析评估了事件树中各事故序列的包壳峰值温度（PCT）分布及条件堆芯损坏概率（CCDP）,最终确定了压水堆在该事故工况中的堆芯损坏频率（CDF）。分析结果表明,压水堆在冷管段双端断裂工况中应急堆芯冷却系统能够保证反应堆的安全,且一列低压安注系统足以排出堆芯余热及保证反应堆安全。      

Safety analysis of LBLOCA in BDBA scenarios for the Bushehr's VVER-1000 nuclear power plant

This paper presents the results of thermal-hydraulic calculations of a large break loss of coolant accident (LBLOCA) analysis for a VVER-1000/V446 unit at Bushehr nuclear power plant (BNPP). LBLOCA is analysis in two different beyond design basis accident (BDBA) scenarios using the RELAP5/MOD3.2 best estimate code. The scenarios are LBLOCA with station blackout (SBO) and LBLOCA with pump re-circulation blockage which have been evaluated in the final safety analysis report (FSAR) of BNPP. A model of VVER-1000 reactor based on Unit 1 of BNPP has been developed for the RELAP5/MOD3.2 thermal-hydraulics code consists of 4-loop primary and secondary systems with all their relevant sub-systems important to safety analysis. The analysis is performed without regard for operator's actions on accident management. The safety analysis is carried out and the results are checked against the acceptance criteria which are the possibility of using water inventory in the emergency core cooling system (ECCS) accumulators and the KWU tanks for core cooling and the available time to operators before the maximum design limit of fuel rod cladding damage is reached. These kinds of analyses are performed to provide the response of monitored plant parameters to identify symptoms available to the operators, timing of the loss of critical safety functions and timing of operator actions to avoid the loss of critical safety functions of core damage. The results of performed analyses show that the operators have 2.9 and 3.1 h for LBLOCA with SBO and LBLOCA with pump re-circulation blockage scenarios, respectively, before the fuel rod cladding rupture. The results are also compared with the BNPP FSAR data.     

Direct vessel inclined injection system for reduction of emergency core coolant direct bypass in advanced reactors

Multidimensional thermal hydraulics in the APR1400 (Advanced Power Reactor 1400 MWe) downcomer during a large-break loss-of-coolant accident (LBLOCA) plays a pivotal role in determining the capability of the safety injection system. APR1400 adopts the direct vessel injection (DVI) method for more effective core penetration of the emergency core cooling (ECC) water than the cold leg injection (CLI) method in the OPR1000 (Optimized Power Reactor 1000 MWe). The DVI method turned out to be prone to occasionally lack in efficacious delivery of ECC to the reactor core during the reflood phase of a LBLOCA, however. This study intends to demonstrate a direct vessel inclined injection (DVII) method, one of various ideas with which to maximize the ECC core penetration and to minimize the direct bypass through the break during the reflood phase of a LBLOCA. The 1/7 scaled down THETA (Transient Hydrodynamics Engineering Test Apparatus) tests show that a vertical inclined nozzle angle of the DVII system increases the downward momentum of the injected ECC water by reducing the degree of impingement on the reactor downcomer, whereby lessening the extent of the direct bypass through the break. The proposed method may be combined with other innovative measures with which to ensure an enough thermal margin in the core during the course of a LBLOCA in APR1400.     

Major findings from LBLOCA reflood tests using the ATLAS facility

A set of LBLOCA (large-break loss of coolant accident) reflood tests was performed in the first phase of the ATLAS (advanced thermal-hydraulic test loop for accident simulation) program. Their main objectives were to identify the major thermal-hydraulic characteristics during the reflood phase of a LBLOCA for APR1400 and to provide qualified data for APR1400 licensing. The ATLAS reflood test program could be divided into two phases (Phase-1 and Phase-2) according to the target period to be simulated. The Phase-1 tests were parametric effect tests for downcomer boiling in the late reflood phase of LBLOCA and the Phase-2 tests were integral effect tests for the entire reflood phase of LBLOCA. The experimental results from both Phase-1 and Phase-2 tests reproduced typical thermal-hydraulic trends expected to occur during the APR1400 LBLOCA scenario. A separate effect test was also performed under a low reflooding rate condition to provide data to validate the RELAP5 reflood models, and its experimental results showed a gradual reflooding in the core, a subsequent quenching of the core heater rods and the cooling of the reactor pressure vessel downcomer.     

Downcomer boiling phenomena during the reflood phase of a large-break LOCA for the APR1400

Downcomer boiling phenomena in a conventional pressurized water reactor has an important effect on the transient behavior of a postulated large-break LOCA (LBLOCA), because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region, a test program for a downcomer boiling (DOBO) is being progressed for the reflood phase of a postulated LBLOCA. Test facility was designed as a one side heated rectangular channel which adopts a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length reduced 47.08-fold. The test was performed by dividing it into two-phases: (I) visual observation and acquisition of the global two-phase flow parameters and (II(a)) measurement of the local bubble flow parameters on the measuring planes along five elevations. In the present paper, the test results of Phase-I and a part of Phase-II(a) were introduced.     

Integral loop test and assessment of modified RELAP5/MOD3.3 for RCS coolant inventory during a LBLOCA

This study investigates experimentally and analytically the thermal hydraulic phenomena during the transition from design basis accident (DBA) to beyond-DBA, particularly, the depletion of core coolant inventory. To investigate the overall thermal hydraulic behavior after safety injection (SI) failure during a large-break loss-of-coolant accident (LBLOCA) in a cold leg, an integral loop test was performed at the Seoul National University Integral Test Facility (SNUF), which was scaled down to 1/6.4 in length and 1/178 in area from the advanced power reactor 1400 MWe (APR1400) according to the three-level scaling method. The plant condition at 200.0 s as the base case and those at 625.0 and 1950.0 s as test cases after the initiation of LBLOCA were applied as initial conditions in the experiments, respectively. The experimental results showed that the sweepout increased the coolant flow discharged to the break depending on the steam flow rate in intact cold legs and the initial downcomer coolant level and expedited the depletion of the core coolant inventory.In the meantime, since RELAP5/MOD3.3 uses the average properties of donor volume as those of its connected junction, this scheme causes the mass and the energy flux in a junction to be calculated incorrectly if significant phase separation occurs in the donor volume such as in the downcomer during the LBLOCA. The sweepout model was developed and implemented in RELAP5/MOD3.3 to improve its calculation of coolant inventory during the LBLOCA. To assess the applicability of the modified RELAP5/MOD3.3 to the actual system, the experiments in SNUF were simulated by both the original and the modified RELAP5/MOD3.3. The original one predicted the discharge flow rate at the break larger than that of the experiment. On the other hand, the modified one calculated the discharge flow rate more similar to that of the experiment than the original one did. As a result, the modified RELAP5/MOD3.3 reduced the coolant flow discharged to the break to delay the initiation time of heater heat-up after SI failure during LBLOCA in a cold leg. This improved RELAP5/MOD3.3 will support a more realistic thermal hydraulic analysis in an integrated analysis system.