出口集管LLOCA始发严重事故分析

采用一体化分析程序建立了包括热传输系统、慢化剂系统、端屏蔽系统、蒸汽发生器二次侧系统的重水堆核电厂的严重事故分析模型。并选取出口集管发生双端剪切断裂的大破口失水事故（LLOCA）,同时叠加低压安注失效,辅助给水强制关闭的严重事故序列进行热工水力分析。由于主热传输系统环路隔离阀的关闭,使得两个环路的热工水力响应过程不同。最终由于低压安注的失效,慢化剂系统逐渐被加热,最终导致堆芯熔化、排管容器蠕变失效。在LLOCA事故序列中叠加向排管容器中注水的缓解措施,可以终止事故进程,使堆芯保持安全、稳定的状态。     

Severe accident development modeling and evaluation for CANDU

Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents.Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 °C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.     

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.     

严重事故下蠕变诱发RCS破裂的概率评估

严重事故下一回路管道可能会发生蠕变失效,若出现蠕变诱发的蒸汽发生器传热管破裂（SGTR）,则会导致安全壳旁路失效;若出现蠕变诱发热段或波动管的失效,则产生的破口将会使一回路迅速卸压。因此,评估严重事故下蠕变诱发反应堆冷却剂系统（RCS）破裂的可能性是开展严重事故分析、特别是二级概率安全分析（PSA）的重要基础。本工作基于蠕变失效模型,考虑传热管的缺陷,建立了评价蠕变诱发RCS破裂的确定论模型。在此基础上,运用拉丁超立方体抽样方法,考虑重要参数的不确定性,开发了严重事故下蠕变诱发RCS破裂的概率评估程序。随后对典型的事故序列进行了蠕变诱发RCS破裂的概率评估。结果表明,对于高压事故序列,存在一定的蠕变诱发SGTR概率,也存在较高的蠕变诱发热段或波动管失效概率。     

An evaluation of the severe accident management strategies for CANDU-6 type plants

This paper provides an evaluation of the mitigation effects for the severe accident management strategies of the Wolsong plants which are typical CANDU-6 type reactors. The evaluation includes the effect of the following six mitigation strategies: (1) injection into the primary heat transport system (PHTS), (2) injection into the calandria vessel, (3) injection into the calandria vault, (4) reduction of the fission product release, (5) control of the reactor building condition, (6) reduction of the reactor building hydrogen. The tested scenario is a loss of coolant accident with a small out-of-core break, and the thermal hydraulic and severe accident phenomenological analyses were implemented by using the ISAAC computer program. The calculation results show that the most effective means for a primary decay heat removal is a low pressure safety injection, that for a calandria vessel integrity is an end-shield cooling injection, and that for a reactor building integrity is a pressure control via local air coolers. Besides the above, the usefulness of each safety component was evaluated in this analysis.     

基于RELAP5程序的AP1000典型事故瞬态特性研究

基于最佳估算程序RELAP5/MOD3.3,对AP1000系统进行了详细的建模分析,选取冷却剂泵卡轴事故、蒸汽发生器(SG)传热管破裂事故和直接注射管线双端断裂事故作为典型事故,获得了典型事故工况下关键参数的瞬态特性和非能动系统响应特性。结果表明:对于冷却剂泵卡轴事故,一回路最大压力为16.82 MPa,燃料包壳表面温度最大值为1 299K,满足验收准则的要求;对于SG传热管破裂事故,破损SG的水体积为231.54m3,小于AP1000蒸汽发生器255.563m3的总容积;对于直接注射管线双端断裂事故,AP1000的非能动堆芯冷却系统能对一回路进行冷却和降压,并防止堆芯裸露和燃料包壳过热。     

AP1000核电厂大量放射性释放源项分析

在AP1000核电厂的某些严重事故情景中,安全壳可能发生失效或旁通,导致大量放射性物质释放到环境中,造成严重的放射性污染。针对大量放射性释放频率贡献最大的3种释放类别（安全壳旁通、安全壳早期失效和安全壳隔离失效）,分别选取典型的严重事故序列（蒸汽发生器传热管破裂、自动卸压系统阀门误开启和压力容器破裂）,使用MAAP程序计算分析了释放到环境中的裂变产物源项。该分析结果为量化AP1000核电厂的放射性释放后果和厂外剂量分析提供了必要的输入。     

Iris small break loca phenomena identification and ranking table (PIRT)

The international reactor innovative and secure (IRIS) is a modular pressurized water reactor with an integral configuration (all primary system components – reactor core, internals, pumps, steam generators, pressurizer, and control rod drive mechanisms – are inside the reactor vessel). The IRIS plant conceptual design was completed in 2001 and the preliminary design is currently underway. The pre-application licensing process with the United States Nuclear Regulatory Commission (USNRC) started in October 2002.The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. If it is not possible to eliminate certain accidents altogether, then the design inherently reduces their consequences and/or decreases their probability of occurring. One of the most obvious advantages of the IRIS Safety-by-Design™ approach is the elimination of large break loss-of-coolant accidents (LBLOCAs), since no large primary penetrations of the reactor vessel or large loop piping exist.While the IRIS Safety-by-Design™ approach is a logical step in the effort to produce advanced reactors, the desired advances in safety must still be demonstrated in the licensing arena. With the elimination of LBLOCA, an important next consideration is to show the IRIS design fulfills the promise of increased safety also for small break LOCAs (SBLOCAs). Accordingly, the SBLOCA phenomena identification and ranking table (PIRT) project was established. The primary objective of the IRIS SBLOCA PIRT project was to identify the relative importance of phenomena in the IRIS response to SBLOCAs. This relative importance, coupled with the current relative state of knowledge for the phenomena, provides a framework for the planning of the continued experimental and analytical efforts.To satisfy the SBLOCA PIRT project objectives, Westinghouse organized an expert panel whose members were carefully selected to insure that the PIRT results reflect internationally recognized experience in reactor safety analysis, and were not biased by program preconceptions internal to the IRIS program.The SBLOCA PIRT Panel concluded that continued experimental data and analytical tool development in the following areas, in decreasing level of significance, are perceived as important with respect to satisfying the safety analysis and licensing objectives of the IRIS program: (1) steam generator; (2) pressure suppression system, containment dry well and their interactions; (3) emergency heat removal system; (4) core, long-term gravity makeup system, automatic depressurization system, and pressurizer; (5) direct vessel injection system and reactor vessel cavity.     

IRIS反应堆严重事故下堆内自然循环及下封头失效分析

使用REALP5/SCDAP分析了IRIS堆汽轮机停机和部分失流事故导致的严重事故进程及缓解措施。分析结果表明IRIS堆内水装量大,使得堆芯较长时间处于淹没状态,事故发生后近7个小时堆芯开始裸露,10小时后堆芯开始损坏。对于不卸压不安注的情况,压力容器会完全干涸,堆芯和蒸汽发生器之间形成蒸汽自然循环流动,堆芯温度缓慢升高,低熔点的控制棒金属首先熔化落入下腔室并加热下封头,使得下封头底部区域发生蠕变断裂失效。在不卸压的情况下一个上充泵的安注流量就能够缓解事故。     

Steam generator tube rupture (SGTR) scenarios

The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000–2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER).     

Experimental Study on Heat Removal during Cold Leg Small Break LOCAs in PWRs

Experiments which simulated small break loss-of-coolant accidents (SBLOCAs) resulting from 2.1–0.13% break in the cold leg of a PWR were conducted with an apparatus of 1/270 scale in volume. In the large break size case, the decay heat was mainly removed by the break flow and in the case of a small break, the steam generator played an important role. In this case, thermal hydraulic behaviors such as natural circulation and reflux condensation cooling were important during the transient. Depressurization in the secondary system due to bleeding steam from the steam generator by an operator action was so effective to make the accident to come to an end. The operation to depressurize the secondary system was also efficient to rewet the core which had been uncovered due to a loop seal formation in a cross-over leg.

No effects of initial 200 ppm dissolved gas in the coolant were observed on the cooling performance of the steam generator. It was considered that it was because the gas which came from the coolant into the steam during the depressurization transient did not remain in the tubes of the steam generator.     

直流蒸汽发生器传热管破裂事故分析

通过对直流蒸汽发生器传热管破裂（SGTR）事故的分析,可看出RELAP5瞬态分析程序能较好地模拟一体化反应堆在SGTR事故后的事件响应序列及主要热工水力现象,例如环路的不对称效应、主回路的自然循环等。一体化反应堆在发生SGTR事故后,可通过一系列安全与保护系统的动作得到有效缓解,并最终能应用非能动余热排出系统（PRHRS）的自然循环导出堆芯余热,使反应堆处于安全状态。同时,受事故影响蒸汽发生器压力在PRHRS投入运行后会快速升高,最终与一回路压力相平衡,此后,破口处的泄漏也会终止。此外,本文还研究了破口处临界流量及其积分流量结果不确定性的影响因素,其中主要考虑了采用不同的临界流模型和破口建模方式等两个方面。     

Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

Accident sequences which lead to severe core damage and to possible radioactive fission products into the environment have a very low probability. However, the interest in this area increased significantly due to the occurrence of the small break loss-of-coolant accident at TM1–2 which led to partial core damage, and of the Chernobyl accident in the former USSR which led to extensive core disassembly and significant release of fission products over several countries. In particular, the latter accident raised the international concern over the potential consequences of severe accidents in nuclear reactor systems. One of the significant shortcomings in the analyses of severe accidents is the lack of well-established and reliable scaling criteria for various multiphase flow phenomena. However, the scaling criteria are essential to the severe accident, because the full scale tests are basically impossible to perform. They are required for (1) designing scaled down or simulation experiments, (2) evaluating data and extrapolating the data to prototypic conditions, and (3) developing correctly scaled physical models and correlations. In view of this, a new scaling method is developed for the analysis of severe accidents. Its approach is quite different from the conventional methods. In order to demonstrate its applicability, this new stepwise integral scaling method has been applied to the analysis of the corium dispersion problem in the direct containment heating.     

Core coolability of an ATR by heavy water moderator in situations beyond design basis accidents

In the case of a loss-of-coolant accident (LOCA) with coincident loss of emergency coolant injection (LOECI), core cooling is generally very severe. However, as the ATR plant has heavy water at about 60°C in the core, decay heat can be removed by the heavy water cooling system. Separate-effects tests relating to heavy water cooling were conducted with each setup. The important thermal hydraulics was radiation heat transfer, ballooning of a pressure tube, contact conductance between the pressure tube and a calandria tube and critical heat flux of the calandria tube. Constants and correlations obtained by the tests were incorporated into several codes to assess the core cooling. Long term core cooling capability with the heavy water cooling system was assessed. The core was cooled without melting under the postulated events due to inherent characteristics of the ATR.     

SMART behavior under over-pressurizing accident conditions

SMART (System-integrated Modular Advanced ReacTor) is an integral reactor of 330 MW capacity with passive safety features under development in Korea. The design is developed by combining the firmly-established commercial reactor technologies with new and advanced technologies such as industry proven KOFA (Korea Optimized Fuel Assembly) based nuclear fuels, self-pressurizing pressurizer, helically coiled once-through steam generators, and new control concepts. The design of SMART focuses on enhancing the safety and reliability of the reactor by employing inherent safety features such as low core power density, elimination of large break loss of coolant accident, etc. In addition, in order to prevent the progression of emergency situations into accidents, the SMART is provided with a number of engineered safety features such as Passive Residual Heat Removal System, Passive Emergency Core Cooling System, Safeguard Vessel, and Passive Containment Over-Pressure Protection System. This paper presents an overview of the SMART design, characteristics of it’s safety systems, and results of over-pressure accident analyses. The results of the accident analyses show that the SMART provides the inherent over-pressure protection capability for design basis accidents without actuation of any protection devices such as safety valves, rupture disks, etc.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.     

Detailed Evaluation of RCS Boundary Rupture during High-Pressure Severe Accident Sequences

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture.     

Cernavoda CANDU severe accident evaluation

The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment.     

MAAP5 simulation of the PWR severe accident induced by pressurizer safety valve stuck-open accident

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines.     

Safety consideration and economic advantage of a new underground nuclear power plant design

A conceptual design of an underground nuclear power plant is proposed to make undergrounding of nuclear reactors not only environmentally desirable but also economically feasible. Expedient to the underground environment, this design capitalizes on the pressure-containing and radiation filtering characteristics of the new underground boundary conditions. Design emphasis is on the containment of a catastrophic accident — that of a reactor vessel rupture caused by external means. The igh apacity apid nergy issipation nderground ontainment (HiC—REDUCE) system which efficiently contains loss-of-coolant accidents (LOCAs) and small break conditions is described. The end product is a radiation-release-proof plant which, in effect, divorces the safety of the public from the safety of the reactor.