Emergency Core Cooling Research Articles

Two-Dimensional Modeling of the Direct ECC Bypass in the APR1400 Downcomer

In a nuclear reactor vessel downcomer incorporating the safety feature of direct vessel injection (DVI), the direct bypass of emergency core coolant (ECC) is activated during the reflood phase of a large-break loss-of-coolant accident due to momentum transfer between the downward liquid film and transverse gas. Direct ECC bypass is reportedly the major bypass mechanism of ECC, and various experiments have been performed to obtain detailed information about the ECC bypass in a DVI downcomer. In the present study, a model of the direct ECC bypass was developed based on two-dimensional two-fluid equations for the adiabatic two-phase flow to predict the ECC bypass flow rate. The direct ECC bypass fractions were calculated with various interfacial friction factor correlations, and the results were compared with the available experimental data. The values predicted by the current model showed reasonably good agreement with the experimental data at bypass fractions >40% when applying the interfacial friction factor model developed in a countercurrent flow condition. However, when the bypass fraction was <40%, models incorporating cocurrent annular flow provided better results than those with countercurrent flow. These results suggest that a transition occurs from a smooth film to a rough film as the gas flow rate increases, and hence, interfacial friction factor models that adequately incorporate this transition are necessary to predict the direct ECC bypass phenomenon.

Development of Technologies on Innovative-Simplified Nuclear Power Plant using High-Efficiency Steam Injectors - System Outline and Endurance Test of Low-Pressure Steam Injectors -

A Steam Injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact heater. This provides SI with capability to serve also as a direct-contact feed-water heater that heats up feed-water by using extracted steam from turbine. Our technology development aims to significantly simplify equipment and reduce physical quantities by applying "High-Efficiency SI", which are applicable to a wide range of operation regimes beyond the performance and applicable range of existing SIs and enables unprecedented multistage and parallel operation, to the low-pressure feed-water heaters and Emergency Core Cooling System of nuclear power plants, as well as achieve high inherent safety to prevent severe accidents by keeping the core covered with water (a Severe Accident-Free Concept). This paper describes the results of the endurance and performance tests of low-pressure SIs for feed-water heaters with Jet-deaerator and core injection system.

Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina Nuclear Power Plant

Stress corrosion cracks have been discovered in Group Distribution Headers (GDH) at the Ignalina and Chernobyl Nuclear Power Plants. This increases the probability that a guillotine pipe break can occur that creates a whipping pipe (GDH) with the potential to damage surrounding structures—i.e. adjacent GDH and its attached piping or adjacent reinforced concrete compartment wall. The GDH is the most important component for reactor safety in case of an accident. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the ECSS into the GDH. Presented in this paper is the transient analysis of a Group Distribution Header following a guillotine break at the blind end of the header. Using a very conservative force loading function, the transient response of a whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is obtained using finite element methodology. The results of the study, assuming that the impacted GDH does not suffer stress corrosion cracking, indicate that the structural integrity of the compartment should be maintained and failure should not propagate from GDH to GDH.

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

Three-dimensional numerical calculations have been performed for a transient moderator circulation inside the CANDU (Canada Deuterium Uranium) calandria vessel of Wolsong Units 2/3/4. The porous media approach was applied for the core region containing 380 calandria tubes. An anisotropic hydraulic resistance model for the porous media has been developed based on the empirical pressure loss correlations. The selected event was the 35% RIH (Reactor Inlet Header) break with a loss of ECC (Emergency Core Cooling) injection, which has been known to give the largest heat load to the moderator among all the DBA's (Design Basis Accidents). The calculation has been successfully done until 1,200 s after the break, when most of the considerable heat transfer procedure has been completed. During this LOCA (Loss of Coolant Accident) transient, the local subcoolings in the vicinity of any PT/CT (Pressure Tube/Calandria Tube) contact does not drop below the experimentally derived subcooling threshold of 30°C. Because the minimum subcoolings reach only a few degrees to the threshold temperature during the initial 20-40 s, future work on the CANDU moderator circulation needs to be aimed at determining whether this small subcooling margin covers the uncertainty of the moderator analysis.

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.

Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFD codes CFX-5 and Trio_U

The influence of density differences on the mixing of the primary loop inventory and the emergency core cooling (ECC) water in the downcomer of a pressurized water reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields. An experiment with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water was selected for validation of the CFD software packages CFX-5 and Trio_U. Two similar meshes with approximately 2 million control volumes were used for the calculations. The effects of turbulence on the mean flow were modeled with a Reynolds stress turbulence model in CFX-5 and a LES approach in Trio_U. CFX-5 is a commercial code package offered from ANSYS Inc. and Trio_U is a CFD tool which is developed by the CEA-Grenoble, France. The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: at higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this propagation. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. Both CFD codes were able to predict well the observed flow patterns and mixing phenomena.

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

MELCOR 1.8.4 sensitivity analysis of the severe accident evolution during the APR 1400 LOCA

Unlike most other systems in which the emergency core cooling (ECC) water is injected into the cold-legs, the Advanced Power Reactor (APR) 1400 employs a concept of a direct vessel injection (DVI) to reduce the bypass effects of the ECC water via a break during a design basis LOCA. For this, the DVI piping is designed so that the ECC water taken from an in-containment refueling storage tank (IRWST) directly flows into the reactor pressure vessel (RPV) down-comer. The main objective of this paper is to provide the MELCOR 1.8.4 sensitivity analysis results on the evolution of the severe accidents that can be expected during the APR 1400 LOCA and the insights gained from the analysis. For this purpose, the present sensitivity analysis mainly focuses on: (1) the impact of the foregoing engineering features (i.e., DVI and IRWST) in mitigating a severe core degradation and (2) the APR 1400-specific impacts of different break locations and sizes, and an operation of the containment spray systems on the timings of the key thermal-hydraulic responses, the severe degradation of the core, and the evolution of the core materials. No significant accident management strategy that plays a great role in mitigating a further progression of severe accidents has been taken into account in the present analysis. As a result, the present analysis results can be taken as the technical basis for assessing the effectiveness of a potential severe accident management.

Experimental validation of the modified linear scaling methodology for scaling ECC bypass phenomena in DVI downcomer

In the direct vessel injection (DVI) system downcomer, the direct emergency core coolant (ECC) bypass is activated during the reflood phase of a large-break loss-of-coolant accident (LBLOCA) by the interaction between the downward-flowing liquid-film and the transverse gas flow. Direct ECC bypass is reportedly the major bypass mechanism of ECC, and various experiments have been performed to obtain detailed information about the ECC bypass in a DVI downcomer. These lead to a proposed new scaling methodology, named ‘modified linear scaling’, which is expected to preserve the phase distribution in the downcomer and the ECC bypass phenomena. In the present study, modified linear scaling was experimentally validated in air–water tests comprising Test 21-D of the upper plenum test facility (UPTF). The counterpart tests of UPTF Test 21-D were performed in 1/7.3 and 1/4.0 scale models of a UPTF downcomer, and the test results were compared with the experimental data of UPTF Test 21-D. Two important parameters of direct ECC bypass – the normalized liquid-spreading width on the downcomer wall and the direct ECC bypass fraction, which is the fraction of input water bypassed out the broken cold-leg – were considered in the validation. The comparison revealed that the scaling parameters of direct ECC bypass are well preserved in the prototype and reduced models, from which we conclude that the modified linear scaling methodology is appropriate for designing a reduced test facility and for a scaling analysis of direct ECC bypass in the reflood phase of an LBLOCA.

Experimental Study for Multidimensional ECC Behaviors in Downcomer Ann with Direct Vessel Injection Mode during the LBLOCA Reflood Phase

For the assessment of the multidimensional safety analysis codes various investigations have been performed to provide detailed information for the ECC (Emergency Core Coolant) behavior in a downcomer annulus. In the present study, the multidimensional ECC bypass phenomena in the downcomer annuli of the UPTF (Upper Plenum Test Facility) and APR1400 (Advanced Power Reactor 1400) geometries are studied to fill in the lack of knowledge about the phenomena that could occur in the DVI (Direct Vessel Injection) system downcomer. The experiments for the direct ECC bypass have been conducted in the transparent downcomer models of the UPTF and APR1400 using air and water. The flow patterns and the bypass mechanisms of the ECC bypass were identified and the characteristics of it in both downcomers were compared with each other. Based on the visual observations, the cross flow between the downward liquid film and circumferential gas was found to be the most important flow pattern of the bypass phenomena. An analysis for the flow regime was conducted and the Wallis parameters were introduced as the significant non-dimensional parameters of the multidimensional two-phase flow phenomena.

Experimental Study for Multidimensional ECC Behaviors in Downcomer Annuli with Direct Vessel Injection Mode during the LBLOCA Reflood Phase

For the assessment of the multidimensional safety analysis codes various investigations have been performed to provide detailed information for the ECC (Emergency Core Coolant) behavior in a downcomer annulus. In the present study, the multidimensional ECC bypass phenomena in the downcomer annuli of the UPTF (Upper Plenum Test Facility) and APR1400 (Advanced Power Reactor 1400) geometries are studied to fill in the lack of knowledge about the phenomena that could occur in the DVI (Direct Vessel Injection) system downcomer. The experiments for the direct ECC bypass have been conducted in the transparent downcomer models of the UPTF and APR1400 using air and water. The flow patterns and the bypass mechanisms of the ECC bypass were identified and the characteristics of it in both downcomers were compared with each other. Based on the visual observations, the cross flow between the downward liquid film and circumferential gas was found to be the most important flow pattern of the bypass phenomena. An analysis for the flow regime was conducted and the Wallis parameters were introduced as the significant non-dimensional parameters of the multidimensional two-phase flow phenomena.

RELAP5/MOD3.2 Analysis of a VVER-1000 Reactor with UO2 Fuel and Mixed-Oxide Fuel

A RELAP5/MOD3.2 model of the VVER-1000/MODEL V320 nuclear power plant was modified and a large-break loss-of-coolant accident (LBLOCA) in the cold leg was simulated. In this analysis, the core consisted of 162 UO2 assemblies and 1 mixed-oxide assembly. The results from the simulation were compared with the results from a similar study performed with the Russian computer program TECH-M. An uncertainty analysis was performed on the peak cladding temperature following a similar methodology called code scaling, applicability, and uncertainty. Monte Carlo calculations were performed using the response surface inferred from 15 runs of RELAP5 calculations. The result of this study shows that the emergency core coolant system would be sufficient to keep the cladding temperature during the LBLOCA scenario well below the required maximum limit.

Embrittlement behavior of zircaloy-4 cladding during oxidation and water quench

Simulated loss of coolant accident (LOCA) tests on Zircaloy-4 cladding were carried out to evaluate the thermal shock property during the injection of emergency core coolant. A Zircaloy-4 specimen was oxidized in a steam environment between 1000 and 1250°C followed by a flooding of the cooling water. After the test, the ductility of the thermally embrittled specimen was measured by a ring-compression test and a microstructural analysis was carried out. The results showed that the threshold equivalent cladding reacted (ECR) value to cause a failure was higher than the conventional 17% criterion calculated by the Baker–Just equation. A residual metal thickness under 0.3mm as well as a ring-compression ductility below 0.2mm for a fracture is effective to assess the thermal shock embrittlement of Zircaloy-4 in an axially unrestrained, or even in a restrained condition.

Scaling for the ECC bypass phenomena during the LBLOCA reflood phase

As one of the advanced design features of the APR1400 (Advanced Power Reactor), a direct vessel injection (DVI) system is adopted instead of the conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is underway. In this paper, a new scaling method, using the time and velocity reduced “modified linear scaling law”, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in the PWR downcomer.

An Experimental Study of Thermal-hydraulic Phenomena in the Downcomer with a Direct Vessel Injection System of APR1400 during the LBLOCA Reflood Phase

The direct vessel injection (DVI) mode is adopted as a safety injection system in the place of a conventional cold leg injection (CLI) mode in the Advanced Power Reactor 1400MW (APR1400). It is expected that “sweep-out” and “direct ECC (Emergency Core Cooling) water bypass” are two most important bypass mechanisms of ECC water injected through the DVI lines during the LBLOCA reflood phase. Using the test facility of plane-channel type scaled down to 1/7 ratio of the prototype reactor (APR1400), we carry out the following tests for the investigation of the two mechanisms: water film spreading test, sweep-out test, and direct ECC water bypass test. From the water film spreading test, it was found that the curvature effect is negligible and the present modified linear scaling law is more appropriate than the linear scaling law. In the sweep-out test, the continuous onset is used to analyze the water height in the downcomer and the amount of ECC water bypass by sweep-out is compared with the previous correlations. The direct ECC water bypass test is performed in order to understand the ECC water film behavior in the downcomer.

Analysis of natural circulation phenomena in VVER-1000

In all light water reactors (LWR), natural circulation is an important passive heat removal mechanism. In the present paper, the natural circulation phenomena are studied with reference to step-wise coolant inventory reduction and a small break loss-of-coolant-accident (SBLOCA) in the cold leg of VVER-1000. The natural circulation flow map (NCFM) approach is considered to evaluate the natural circulation performance of the VVER-1000 NPP also comparing VVER-1000 and PWR systems. Three different elevations between heat source (core) and heat sink (steam generators) zones have been considered in order to characterize the buoyancy force in a VVER-1000. The influence of power and the cold legs loop seal upon the natural circulation performance is also evaluated. In the second part, a series of SBLOCA simulations with break area ranging from 0.5 to 11.7% of the cold leg cross sectional area are performed starting with the VVER-1000 system in nominal conditions. The effect of Emergency Core Cooling System (ECCS) including passive and active parts of ECCS are evaluated. The simulations were performed by the help of the system code RELAP5. Within the framework of the qualification of the adopted computational tools, the results are compared with experimental data from Kozloduy NPP unit 6 test and PSB-VVER integral test facility available from the literature. Namely, the qualification of the adopted nodalisation in steady state conditions is achieved by using experimental data. The accuracy of selected results have been estimated in quantitative terms by applying the fast Fourier transform based method (FFTBM). Finally, the relevance and the potential for the occurrence of the reflux condensation mode, i.e., one of the Natural Circulation regimes, for cooling of reactor core in VVER-1000 are discussed.

Development, assessment and application of TRAC-BF1/v2001.2 for beyond design basis BWR LOCA transients

In preparation for the possible transition to risk based licensing, it is increasingly important to demonstrate the applicability of Best Estimate codes to more extreme conditions corresponding for example to limited equipment availability. With this idea in mind, we have reviewed the application of TRAC-BF1 to Large and Small Break LOCAs. In this context this work describes the assessment and applications of the Penn State University (PSU) version 2001.2 of TRAC-BF1 with all the PSI updates, to the analysis of hypothetical Large Break (LB) LOCAs in a BWR/4 with postulated limited ECC availability. Since in contrast to a LB-LOCA in a BWR with full ECC availability, the rod surface temperatures reach relatively high values, additional assessment of the code under such conditions is required. Hence, after analyzing bottom flooding separate-effect experiments in two different heater rod bundles and a TLTA LB-LOCA test, we shall present and discuss the results of the analysis of a LB-LOCA with highly restricted Emergency Core Cooling flow in a BWR/4. In this context, we shall also assess different descretization of some terms of the 3D momentum equations, which was found to be important in the analysis of BWR Small Break LOCAs.

Three-dimensional analysis of flow characteristics on the reactor vessel downcomer during the late reflood phase of a postulated LBLOCA

A three-dimensional CFD analysis has been performed on the flow characteristics in the reactor vessel downcomer during the late reflood phase of a postulated large-break loss-of-coolant accident (LBLOCA), in order to validate the modified linear scaling methodology that was applied in the MIDAS test facility of Korea Atomic Energy Research Institute. The vertical and circumferential velocity similarities are numerically tested for the 1/1 and 1/5 linear scale models for the APR1400 reactor vessel downcomer. The effects of scale on flow patterns, pressure and velocity distributions, and the impinging jet behavior are analyzed with the FLUENT code. In addition, a simplified half cylinder model with a single emergency core cooling (ECC) nozzle is numerically tested to investigate the scale effect on the spreading width and break-up of ECC water film. The qualitative and quantitative results indicate that the 1/5 modified linear scale model of the reactor vessel downcomer would reasonably preserve the hydrodynamic similarity with APR1400.

Effect of the yaw injection angle on the ECC bypass in comparison with the horizontal DVI

The comparison tests for the direct emergency core cooling (ECC) bypass fraction were experimentally performed with a typical direct vessel injection (DVI) nozzle and an ECC column nozzle having a yaw injection angle to the gravity axis. The ECC yaw injection nozzle is newly introduced to make an ECC water column in the downcomer region. The yaw injection angle of the ECC water relative to the gravity axis is varied from 0 to (±)90° stepped by 45°. The tests are performed in the air–water separate effect test facility (direct injection visualization and analysis (DIVA)), which is a 1/7.07 linearly scaled-down model of the APR1400 nuclear reactor. The test results show that (1) if the ECC water column is injected into the wake region which is induced by the hot leg blunt body in the downcomer annulus, the ECC bypass fraction is greatly reduced compared with the typical horizontal ECC injection which makes ECC film on the downcomer wall. At the same time, the ECC penetration toward the lower downcomer region becomes larger than those of a typical horizontal type of direct vessel injection on the downcomer wall vertically. (2) If the ECC water column is injected near the broken cold leg, the ECC water is directly bypassed. Thus, the ECC penetration fraction is greatly reduced compared with a typical film type of the horizontal ECC injection. (3) In order to minimize the ECC bypass fraction, the ECC water should be injected toward the wake region of the hot leg blunt bodies.

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold emergency core cooling (ECC) water entering the RPV through the ECC injection into the cold leg. The experimental results show an incomplete mixing with typical concentration and temperature distributions at the core inlet, which strongly depend on the boundary conditions. Computational fluid dynamics calculations were found to be in good agreement with the experiments.