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.     

Response of proposed passive safety injection system for an intermediate size break LOCA on CHASNUPP-1

A passive safety injection system (PSIS) is proposed for Chashma nuclear power plant-1 (CHASNUPP-1) type nuclear power plants, for the simplification of their safety systems. This system is based upon passive components and is proposed in place of the existing safety injection system, for safety enhancement. The functionality of the proposed system is analyzed using reactor simulation. For this purpose an intermediate size break LOCA is simulated using the simulation software APROS. For this transient, different thermal-hydraulic parameters of the proposed and other safety related systems are presented and discussed. The results obtained show that the proposed system works properly by performing its role in the transient, leading to cold shutdown conditions.     

Limiting countercurrent flow phenomenon in small break LOCA transients

Countercurrent flow limit phenomena in the loops of a PWR affects small break LOCA transients significantly. This paper reviews the countercurrent flow at three different locations in the loop to identify a limiting phenomenon during a cold leg small break.It is believed that the limiting phenomenon occurs at the inclined pipe which connects the hot leg to the steam generator. Therefore this phenomenon must be simulated properly in any numerical of physical simulations of small break LOCA's. It is pointed out that data from any test loops with scaling have to be analyzed with caution. Distortions in the flow areas of the steam generator tubes and hot leg may result in nontypical transients.     

RELAP5 analyses on the influence of multi-dimensional flow in the core on core cooling during LSTF cold-leg intermediate break LOCA experiments in the OECD/NEA ROSA-2 Project

Test-2 and Test-7 of the second Organization for Economic Co-operation and Development / Nuclear Energy Agency (OECD/NEA) Rig-of-Safety Assessment (ROSA-2) Project were performed with the Large Scale Test Facility (LSTF), which simulated thermal-hydraulic responses during a pressurized water reactor cold-leg intermediate break loss-of-coolant accident (IBLOCA). Test-2 simulated 17% cold-leg break with single failure of an emergency core cooling system (ECCS). The core liquid level decreased to the bottom at loop seal clearing (LSC), causing high cladding temperature excursion. Test-7 simulated 13% cold-leg break with full injection of the ECCS. Compared to Test-2, the cladding surface temperature in Test-7 was quite low due to greater liquid level recovery after the LSC. To well understand the observed phenomena and to improve the best-estimate code predictive capability, RELAP5 post-test analyses were performed. The RELAP5 analyses employed two core models: one is a single-channel core model that simulates the whole core with one channel of a vertical stack of nine equal-height volumes, and the other is a multiple-channel core model that is composed of three channels in which adjacent vertically stacked volumes are horizontally connected with cross-flow junctions. The analyses with the multi-channel core model predicted better than with the single-channel core model for such parameters as core-collapsed liquid level and cladding surface temperature for both Test-2 and Test-7, by more realistically representing multi-dimensional flow in the core. Such a practical method for better representation of multi-dimensional flows turned out to be important to improve the IBLOCA analysis.     

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.     

BWR small break LOCA counterpart tests at ROSA-III and FIST test facilities

The Japan Atomic Energy Research Institute performed a 2.8% recirculation pump suction line break BWR LOCA test at the ROSA-III test facility. The test was a counterpart test to the 2.8% break test performed at the FIST test facility by the General Electric Company. The objective of the test was to develop a common understanding and interpretation of the controlling phenomena for a small break LOCA of a BWR. Similar phenomena were observed in the two tests in a similar time sequence and with magnitudes. These two test results and a 2.8% break reference BWR LOCA were analyzed using the THYDE-B1 computer code. It was confirmed from the analysis that the THYDE-B1 code has enough capability to analyze a BWR small break LOCA. The applicability of the tests performed at the two facilities to a BWR was also confirmed through the analyses.     

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.     

Concept of a core cooling system and experiments performed

One of the problems which must be solved in severe accidents is the melt-concrete interaction which does occur when the core debris penetrates the lower pressure vessel head and contacts the basement. To prevent these accident consequences, a core catcher concept is proposed to be integrated into a new pressurized-water reactor design. The core catcher achieves coolability by spreading and fragmentation of the ex-vessel core melt based on the process of water inlet from the bottom.In order to justify the dominant process during flooding of the melt from the bottom, prototypic experiments with thermite melts in laboratory scale have been carried out. In these experiments flooding and early coolability of the melt is demonstrated. To obtain more detailed information on the important process of water penetration into the melt, a simulant experiment has been conducted using a transparent plastic melt with the typical viscosity behaviour of an oxidic corium melt and a temperature allowing evaporation of water. In every experiment the melt is flooded, and complete freezing in the form of a porous layer occurs within a few minutes only.     

Analysis of large break LOCA experiment at ROSA-III

Analysis of the ROSA-III test RUN 704 was performed by using the computer codes RELAP4J, RELAP4/MOD6 and RELAP5/MOD0 to verify the predictive capability of the codes for a BWR LOCA. The ROSA-III facility is a volumetrically scaled (1/424) BWR system with an electrically heated core, designed for in tegral LOCA/ECCS tests. The RUN 704 experiment at the ROSA-III test facility simulated a 200% double-ended offset shear break on the inlet side of the pump in the recirculation loop. From present analyses, key parameters which are important to predict major behavior during a BWR large break LOCA have been clarified and the promising predictive capability of the advanced code RELAP5 has been verified.     

Analysis of the OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

OECD/NEA ROSA Project experiment with the large scale test facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR LOCA analysis using JAEA-developed coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1 with detailed core model. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode due to high vapor velocity. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment. The code, however, overpredicted the break flow rate especially during two-phase flow discharge period probably because of the failure in the correct simulation of the cold leg liquid level due to late decrease in the primary loop flow rate.     

Five percent break BWR LOCA/ECC test at ROSA-III without HPCS actuation - two-dimensional core thermal-hydraulic phenomena

The ROSA (Rig of Safety Assessment)-III facility is a volumetrically scaled (1/424) simulated boiling water nuclear reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. A recirculation pump suction line break test with a five percent break area was conducted with the assumption of high pressure core spray system (HPCS) failure. The simulated peripheral fuel rods facing the channel box wall had a tendency to be rewetted temporarily at the upper part of the core by falling water from the upper plenum before low pressure core spray system (LPCS) actuation, while the rods in the central region were not rewetted but quenched mainly from the bottom of the core after low pressure coolant injection system (LPCI) actuation. Therefore, the peak cladding temperatures of the simulated high power fuel rods were limited to lower values since they were located in the peripheral region and the temporary rewetting before LPCS actuation occurred mainly in the peripheral region. The ROSA-III five percent break test and a BWR counterpart were analyzed with the RELAP5/MOD1 (cycle 018) code. Similarity between the ROSA-III small break test and a BWR small break LOCA has been confirmed through comparison of the calculated results.     

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.     

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.     

Boiloff pattern transition during core uncovery in a small break loca

Through the core uncovery experiments, it has been known that the low power and high power core boiloff patterns are observed in the high pressure core uncovery following a small-break loss-of-coolant accident. A criterion for the prediction of annular to intermittent flow transition in vertical flow is developed and applied to the classification of low power boiloff and high power boiloff patterns. The instability of the interface wave on the liquid film is considered as the real physical mechanism for the flow pattern transition and the instability is explained by the concept of the hyperbolicity breaking in the characteristic equation. The applicability of the developed criterion to the rod-bundle geometry is demonstrated using the flow pattern transition data taken by Bergles et al. and Venkateswararao. Finally, it is shown that the developed criterion well predicts the boundary between low power boiloff and high power boiloff through the comparisons of the predicted annular to intermittent flow transition conditions with the boiloff experiment data taken by Anklam and Osakabe.     

AP1000非能动堆芯冷却系统及系统设计瞬态研究

AP1000在标准设计中革新性重大改进之一就是采用了独特的非能动堆芯冷却系统(PXS)。目前世界上在役核电厂和在建核电工程中,AP1000非能动堆芯冷却系统是第一个完全采用非能动手段来达到堆芯冷却、冷却剂补充以及限制放射性释放等安全功能的安全相关系统。文章结合AP1000非能动堆芯冷却系统设计与运行,应用包络方法对一些重要的设计瞬态进行研究分析,从而得出系统设计的合理性和系统功能实现的可行性,为自主研发ACP100、ACP600、ACP1000等第三代核电技术提供借鉴和参考。     

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.     

Analyses of PACTEL passive safety injection experiments with APROS, CATHARE and RELAP5 codes

The European Commission fourth framework programme project ‘Assessment of passive safety injection systems of advanced light water reactors’ involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of passive safety injection systems (PSIS) of advanced light water reactors (ALWRs) in small break loss-of-coolant accident (SBLOCA) conditions. The PSIS consisted of a core make-up tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the phenomena in the PSIS.     

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.