IIST small break LOCA experiments with passive core cooling injection

The purpose of this study is to evaluate the performance of a passive core cooling system (PCCS) with passive injection during the cold-leg small break loss-of-coolant accidents (SBLOCAs) experiments conducted at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. Four tests were performed simulating break sizes of 0.2–2% (approximately corresponding to 1.25–4″ breaks for a referenced nuclear power plant) at cold-leg for assessing the PCCS capability in accident management. The key thermal–hydraulic phenomena to core heat removal for PCCS are observed and discussed. The experimental results show that the PCCS has successfully provided a continuous removal of core heat and a long term core cooling can be reached for all cases of SBLOCA.     

Qualitative response of safety injection system for a large size break in the reactor coolant system cold leg

Safety injection system, accumulator injection system and residual heat removal system of CHASNUPP-1 were simulated using the computer code APROS. We observed the qualitative response of the simulated system during injection and re-circulation phases after LOCA. During rapid depressurization of SRC system due to leakage, these systems started coolant injection in the SRC system as per plant requirement. Different thermal-hydraulic parameters of the respective systems are presented and discussed. Results obtained are in good agreement with the reported document of the reference power plant.     

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.     

Preliminary safety analysis of ex-vessel LOCA for the European HCPB TBM system

Ex-vessel loss of coolant accident caused by a double-ended pipe break of the helium coolant system inside port cell is considered as one of the most critical accident for the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) system. The resulting rapid helium blow-down causes an immediate block of the TBM cooling, which requires a prompt plasma shutdown. Even after the plasma shutdown the temperature can increase over the design limit and the accident sequence can lead up to a break of the TBM box protection after the failure of different protection systems. Thus air ingresses in the vacuum vessel from the damaged TBM system and steam from the surrounding ITER blanket and divertor structures. The evaluation of this sequence is very important for the definition of the correct protection strategy of the system. To consider all these different events a methodology has been developed in KIT combining different codes for a complete analysis of the accident. In particular, this paper shows an application of MELCOR code to model beryllium–steam reaction in a particular accidental sequence for the long term cooling.     

Simulation of small break LOCA effects on BWR-pressure suppression systems

The GKSS-Forschungszentrum has simulated within an extensive PSS (Pressure Suppression System) program small break LOCA situations in a large scale multivent PSS test arrangement. The gained experimental information indicates that the simulated small break LOCA in a BWR-PSS which initiates steam condensation in the wetwell pool at the vent pipe outlets, gives strong cyclic pressure pulses from chugging events over a long time period.     

Investigation of BWR LOCA at ROSA-III — Effect of break configuration on system transients

The ROSA-III test facility is a volumetrically scaled ( ) BWR/6 system with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA).Six loss-of-coolant experiments with a break area of 15%, 50% or 200% at the main recirculation pump inlet line were conducted at the ROSA-III test facility with a high pressure core spray failure. A sharp-edged orifice or a long throat nozzle was used as a break plane. It was found in the experiments that the break flow differences between the orifice and the nozzle break configurations with the same flow area were observed only in the subcooled break flow region. Subcooled break flow rate through the orifice was much larger than that through the nozzle. The break configuration difference had little influence on the other system responses, especially on the peak cladding temperature. The applicability of the test results to a BWR/6 has been confirmed through analyses of the 15% break ROSA-III LOCA experiments and BWR/6 LOCAs by using RELAP4/MOD6/U4/J3 code. The experimental results of the ROSA-III LOCA experiments were calculated well by the code, and the same trends were calculated in the BWR analyses.     

非能动安注箱的设计与分析

介绍了非能动安注箱的设计与实验,并用CATHENA程序分析其特性:注入流量的峰值,高注入流量的持续时间,最低注入流量等。计算结果表明非能动安注箱设计满足主要的性能要求,CATHENA程序计算结果与实验数据基本一致,可用于概念设计与事故分析。     

Development of passive flow controlling safety injection tank for APR1400

A vortex valve, called fluidic device, is to be installed inside a Safety Injection Tank (SIT) of Advanced Power Reactor 1400 MWe (APR1400) that passively controls an Emergency Core Cooling (ECC) water discharge flow rate without any moving part or any action of the plant operator. The fluidic device was designed, and its performance was evaluated by a series of repetitive experiments using VAlve Performance Evaluation Rig (VAPER), a prototypical full-scale test facility. The passive flow controlling SIT satisfied the major performance requirements of the APR1400 plant design, in view of peak discharge flow rate, pressure loss coefficient, and duration time of the ECC water discharge.     

The effects of break location on PWR small break LOCA: Experimental study at the ROSA-IV LSTF

This paper discusses the effect of break location on the break flow rate and break flow quality transitions during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). Results from five experiments conducted at the ROSA-IV Large Scale Test Facility (LSTF) are compared for this purpose. These experiments simulated a 2-inch break at the lower plenum, upper head, pressurizer top, cold leg, and hot leg, respectively. The controlling phenomena for the break flow quality transitions in cold-leg and hot-leg break experiments are described.     

Performance evaluation of a new signal processing system design to improve CANDU SDS1 trip response during large break LOCA events

Performance of a recently developed signal processing system for CANDU (Canada Deuterium Uraniu) reactor shutdown system 1 (SDS1) is evaluated in this paper. The evaluation is carried out in MATLAB/Simulink software environment as well as with an existing power measurement and signal processing system. The new signal processing algorithm is obtained based on the synthesis of several first order low pass filters with different delayed time constants. Throughout this paper, a special attention has been paid to compare the new signal processing system with the existing one. The dynamic behavior of the new signal processing system in the practical large loss of coolant accidents (LLOCA) events has also been examined. Simulation results show that during the LLOCA event, the reactor trip time, as well as the peak power, is decreased remarkably. Through the simulation studies, it has convincingly demonstrated that the new signal processing system has significant advantages over the existing system in terms of the improved trip response and accommodation of the spurious trip immunity. This advantage will significantly enhance the safety margin, or will bring economical benefits to nuclear power plants.     

SG LOCA摇晃动力响应数值分析

经过合理的简化与等效处理，建立了国内某3代核电站的蒸汽发生器（SG）非线性有限元模型，将其与反应堆冷却剂环路（RCL）串联，开展了SG失水事故（LOCA）摇晃动力响应数值分析，得到了作用在SG传热管上的应力极值及其随管径的变化规律，并获得了作用在上部支承上的载荷。将本文方法与传统解耦法进行对比，结果表明:SG的解耦对摇晃动力响应有较大影响，应采用与RCL耦联的计算方式。      

Assessment of scaling principles for the simulation of small break LOCA experiments in PWRs

The choice of the scaling laws to be applied for the simulation of nuclear reactor behaviour and, more particularly, the extrapolation of data measured in experimental facilities to real plants, remains an important unresolved issue in nuclear safety.After the analysis of scaling principles adopted in the design of four PWR simulators, the above problem is dealt with in this paper.The definition of a counterpart test and a code analysis, comparing LOFT measured data with calculated trends in the PWR-PUN plant and in LOBI/MODI, LOBI/MOD2 and SEMISCALE facilities, make it possible to check the validity of the criteria utilized in the design of the experimental loops and to reduce uncertainty margins in predicting PWR behaviour.     

Gravity-driven injection experiments and direct-contact condensation regime map for passive high-pressure injection system

A gravity-driven injection experiment of a passive high-pressure injection system with a pressurizer pressure balance line (PRZ PBL) is conducted by using a small-scale test facility to identify the parameters affecting the gravity-driven injection and the major condensation regimes. It turns out that the larger the water subcooling is, the more the injection initiation is delayed. A sparger and natural circulation of the hot water from the steam generator accelerate the gravity-driven injection. The condensation regimes identified through the experiments are divided into three distinct ones: sonic jet, subsonic jet, and steam cavity. The steam cavity regime is a unique regime of downward injection with the present geometry not previously observed in other experiments. The condensation regime map is constructed using Froude number and Jacob number. It turns out that the buoyancy force has a larger influence on the regime map transition because the regime map using the Froude number better fits data with different geometries than other dimensionless parameters. Simple correlations for the regime boundaries are proposed using the Froude number and the Jacob number.     

Analyses of PACTEL passive safety injection experiments GDE-21 through GDE-25

In advanced light water reactors (ALWR), gravity-driven passive safety injection systems (PSIS) replace pump-driven emergency core cooling systems. PSISs often rely on small density differences and driving forces for natural circulation. In a typical loss-of-coolant accident (LOCA), interactions between different parts of the emergency core cooling system also take place. VTT Energy in Finland, in co-operation with the Lappeenranta University of Technology (LUT), performed five experiments in the PACTEL loop to study PSIS performance during SBLOCAs. The purpose of the PSIS, a passive core make-up tank (CMT), was to provide high-pressure safety injection water to the primary circuit. The purpose of these experiments was to produce data to validate the current thermal-hydraulic safety codes, and to study the effects of break size on the PSIS behaviour. In all experiments the CMT ran as planned. No problems with rapid condensation in the CMT, as seen in earlier passive safety injection experiments in PACTEL. The main reason was the new CMT arrangement, with a flow distributor (sparger) installed. The analyses of the test data supported the use of McAdams correlation for calculating the heat transfer from the hot liquid layer to the CMT wall. The use of Nusselt film condensation correlation for condensation at the CMT walls seems correct. The APROS code simulated successfully the overall primary system behaviour in the GDE-24 experiment, such as timing of the core heat-up at the end of the experiment. The code had some problems, in the simulation of thermal stratification in the CMT.     

反应堆堆内构件在地震加失水事故下的结构反应分析研究

对反应堆堆内构件进行地震加失水事故作用下详细的动力分析与评定,是核电厂设计规范和安全审查的要求.压水反应堆堆内构件具有较强的非线性,它的非线性体现在部件间存在的间隙和部件的变刚度结构.本文应用ANSYS程序中的时程分析法和模态叠加分析法解决带有间隙的梁之间的碰撞,以模型试验结果考核计算方法的可靠性.在此基础上建立堆内构件的有限元非线性计算模型,应用ANSYS程序进行求解,得出一些有重要意义的结论.     

核电厂主设备在地震加失水事故下的结构反应分析研究

核电厂主设备是核电厂的关键设备.对反应堆堆内构件、控制棒驱动系统、燃料组件和蒸汽发生器传热管等设备进行地震加失水事故联合作用下详细的动力分析与评定,是核电厂设计规范和安全审查的要求.上海核工程研究设计院在主设备的地震加失水事故下反应分析和试验研究的基础上,将主设备作为一个总体进行分析,从而形成一个完整的分析和评定系统.该研究成果已应用于秦山、PC两座核电厂的设计分析和安全评审中,对我国自主开展百万级先进压水堆核电厂主设备在地震加失水事故下的设计和安全分析具有良好的推广和应用前景.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones.In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes).The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.     

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.