Pressure Suppression Pool Research Articles

Long-term pressure and temperature analysis and suppression pool mixing of Lungmen ABWR containment

Lungmen Nuclear Power Plant in Taiwan is a twin-unit Advanced Boiling Water Reactor (ABWR) plant. The design-basis accident for the ABWR containment is a double-ended feedwater line break accident, and the long-term containment responses for this accident are calculated using GOTHIC in this study. The wetwell space is modeled in one dimension either in multi dimensions to investigate the pool mixing effect. The suppression pool is cooled by the Residual Heat Removal (RHR) system in a long-term accident. The RHR pumps have suction flow from the pool bottom. The suction flow is cooled by the heat exchangers, and then returns to the pool via the RHR return line. The long-term containment response using GOTHIC agrees with the Lungmen FSAR results. Significant stratification in the wetwell is observed using the one-dimensional model. However, the multi-dimensional wetwell models predict that the suppression pool and the wetwell airspace can be well mixed by the internal convection. The multi-dimensional results also show that using the wetwell spray to return the cooled water can reduce long-term wetwell pressure.

BWR/5 Pressure-Suppression Pool Response during an SBO

RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

Modeling pressure suppression pool hydrodynamics in the ABWR containment

This paper focuses on the assessment of pressure suppression pool hydrodynamics in the advanced boiling water reactor (ABWR) containment under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a mechanistic model for predicting various suppression pool hydrodynamics parameters. A phenomena identification and ranking table (PIRT) applicable to the ABWR containment pool hydrodynamics analysis is used as a basis for the development of the model. The highly ranked phenomena are represented by analytic equations or empirical correlations. The best estimate and several sensitivity calculations are performed for the ABWR containment using this model. Results of the sensitivity calculations are also presented that demonstrate the influence of key model parameters and assumptions on the pool hydrodynamics parameters. A comparison of model predictions to the results of the licensing analyses shows reasonable agreement. Comparison of the results of the proposed model to experimental data shows that the model predicted top vent clearance time, the pool swell height, and the bubble breakthrough elevation are within 10% of the data. The predicted pool surface velocity and the liquid slug thickness are within 30% of the measurements, which is considered adequate given the large uncertainties in the experimental measurements.

Jet-plume condensation of steam–air mixtures in subcooled water, Part 1: Experiments

A scaled-down, reduced pressure suppression pool was designed to study condensation and mixing phenomena for a LOCA (loss of coolant accident) event in a SBWR (simplified boiling water reactor) design. The scaled-down test facility represented an idealized trapezoidal cross-section, 1/10 sector of the SP (suppression pool) with scaled height ratio of 1/4.5 and volume ratio of 1/400. The facility was instrumented with thermocouples for pool temperature measurements and a high-speed camera for flow visualization. Thermal stratification data were obtained for different pool initial subcooling and steam–air mixture flow rates. A dimensionless boundary map was derived from several experimental runs of pure steam injection to determine conditions when the pool transitions from being a homogeneously mixed volume to being a thermally stratified one. Steam air mixture injection cases for single horizontal venting indicated that above a pool temperature of 40 °C with air mass fraction below 0.5% the pool can attain thermal stratification.

A three-dimensional multi-volume analysis of combustible mixture generation due to in-vessel LOCA of ITER using the gasflow code

Three-dimensional distribution calculations are performed for the ITER-FEAT vacuum vessel (VV), the connected pressure suppression pool (SP) and drain tank (DT). A first-wall coolant leak without plasma shutdown is simulated. The steam, hydrogen, and air sources for this sequence are taken from best-estimate MELCOR calculations. A new extended version of gasflow is used to model the ITER-FEAT specific phenomena in adequate detail. During the accident sequence, hydrogen initially appears only in the VV due to the steam/beryllium reaction. After opening of the valves, steam and hydrogen flow from the VV through the connecting lines to the SP and the DT. Because of the ongoing steam condensation occurring in the SP, the pressure there remains permanently at a lower level compared to the other components, resulting in a continuous flow of steam and noncondensable gases into this volume. After 10 500 s of steam flow, also air starts entering the VV, and an accumulation of N 2 and O 2 takes place in the SP cover gas. Combustible and explosive H 2–O 2–N 2 mixtures exist after 13 600 s, and at 21 000 s a stoichiometric H 2/O 2 ratio has formed, involving 13 kg of hydrogen.

Pressure suppression pool mixing in passive advanced BWR plants

In the SBWR passive boiling water reactor, the long-term post-accident containment pressure is determined by the combination of noncondensible gas pressure and steam pressure in the wetwell gas space. The suppression pool (SP) surface temperature, which determines the vapor partial pressure, is very important to overall containment performance. Therefore, the thermal stratification of the SP due to blowdown is of primary importance. This work looks at the various phases and phenomena present during the blowdown event and identifies those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests. This is important in determining which of the large body of blowdown to SP data is adequate for application to the stratification problem. The mixing by jets from the main vents is identified as the key phenomena influencing the thermal response of the suppression pool and analytical models are developed to predict the jet influence on thermal stratification. The analytical models are implemented into a system simulation code, TRACG, and used to model thermal stratification behavior in a scaled test facility. The results show good general agreement with the test data.

Containment concepts assessment for the SEAFP reactor

A simple methodology has been developed for making relative comparisons of potential containment designs for future fusion reactors. The assessment methodology requires only conceptual design information. The application of this methodology, at the early stages of a fusion reactor design, provides designers useful information regarding the suitability of various containment designs and design features. Because the radiation hazard from the operation of future fusion power reactors is expected to be low, the containment design, in addition to public safety, needs to take into account worker safety considerations, as well as factors important to the reliable and economical operation of the power plant. Several containment concepts have been assessed with a methodology that takes into account public safety, worker safety, operability and maintainability as well as cost. This paper describes this methodology and presents the results of the assessment. The paper concludes that, to obtain a containment design that is optimised with respect to safety, operational and cost factors, designers should focus on a containment that is conceptually simple – that is, one utilising a single, large containment building without relying on special features such as expansion volumes, pressure suppression pools or spray systems.

Radioactivity Confinement Studies Within the SEAL Program

In the framework of the European SEAL program, investigations have been performed with the aim of optimizing the second confinement function and plant layout with respect to normal operation as well as abnormal operation, including accident conditions. This has been done for two conceptual fusion reactor designs: one using water as the coolant and the other using helium. The starting point of these investigations was the SEAFP project design. For the water-cooled reactor design the studies were focused on design options such as pressure suppression spray system, pressure suppression pool with closed containment or with venting to gravel bed filter and stack, and separate expansion volume optionally operated with a vacuum and equipped with spray system. Similar analyses were performed for the helium-cooled reactor design. The analyses were focused on design options comprising a single, large confinement volume or a vent duct connected to an expansion volume operated at vacuum in comparison with the SEAFP Model 1. The thermal-hydraulic analyses performed with the MELCOR code provide an integrated assessment of the cooling loop and confinement system dynamics.

Containment venting as a mitigation technique for BWR MARK I plant ATWS

Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it. Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure. For both cases the operators are assumed to take action to intentionally restrict injected flow such that fuel in the upper part of the core would be steam cooled. Resulting fuel temperatures are estimated with an off-line calculation and found to be acceptable.

Analytical Simulation of Boiling Water Reactor Pressure Suppression Pool Swell

In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement.

The vent-to-vent desynchronization effects on LOCA steam condensation loads in BWR pressure suppression pool

During a loss-of-coolant accident of a boiling water reactor (BWR), direct-contact steam condensation in the pressure suppression pool will induce undesirable dynamic pressure loads onto pool boundary structures. The magnitudes of the pool boundary loads will be influenced by finite desynchronization among the condensation processes, which take place at the outlets of multiple (about one hundred) vent pipes. This paper investigates the effects of the condensation desynchronization on the pool boundary loads, with regard to test data obtained from large-scale tests which represented seven full-sized vent pipes in a BWR Mark II pressure suppression containment system. The desynchronization effects are studied by analyzing the test data in the time and frequency domains. Attempts are made to extrapolate the experimental results to the actual plant geometry. Mechanisms responsible for the desynchronization are also discussed.

The LOCA air-injection loads in BWR Mark II pressure suppression containment systems

Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.

The Noncondensable Gas Effects on Loss-of-Coolant Accident Steam Condensation Loads in Boiling Water Reactor Pressure Suppression Pool

The noncondensable gas effects on the loss-of-coolant-accident-induced steam condensation loads in the boiling water reactor pressure suppression pool have been investigated with regard to experimental data obtained from a large-scale multivent test program. Previous studies have noted that the presence of the noncondensable gas (air), which initially fills the containment drywell space, stabilizes the direct-contact condensation in the pressure suppression pool and hampers onset of the chugging phenomenon, which induces most significant steam condensation load onto the pool boundary. This was found to be true for the tests with relatively small-break diameters, where the maximum steam mass fluxes in the vent pipe were lower than the upper threshold value for the onset of chugging. However, in the tests with the maximum vent steam mass fluxes moderately higher than the chugging upper threshold value, early depletion of the noncondensable gas tended to result in significant stabilization of steam condensation accompanied by an excursion of temperature of pool water surrounding the vent pipe outlets, which led to a delayed onset of chugging. Due to this combined influence of the noncondensable gas and nonuniform pool temperature, and due to dependence of magnitude of chugging load on the vent steam mass flux, the peak magnitude of the steam condensation load appearing in a blowdown can be very sensitive to the initial and break conditions.

Analytical method for solving fluid-structure interactions in BWR pressure suppression pool

An analytical method is developed for solving the fluid-structure interactions (FSI) associated with hydraulic transients in the BWR pressure suppression pool during a hypothetical loss-of-coolant accident. The present method assumes inviscid and infinitesimal motion of pool water for which the governing fluid-dynamic equation reduces to a linear wave equation with inhomogeneous boundary conditions of flexible pool boundaries. The present method is applied to quantitative evaluation of the FSI effect on test data obtained from a large scale pressure suppression test which was performed to investigate the pressure oscillations in the pool induced by unstationary steam condensation.

Evaluation of Dynamic Loads Induced by Chugging

A dynamic load evaluation method has been proposed for chugging phenomena which are assumed to occur and produce relatively large amplitude pressure spikes in the pressure suppression pool of a BWR containment, in case of a postulated loss of coolant accident. The proposed method is based on the analysis code developed by the authors and on theseven vent full scale tests performed at Japan Atomic Energy Research Institute (JAERI CRT), considering random nature of chugging phenomena. The dynamic loads are obtained by applying the design source functions of impulsive nature to the vent pipe exists in each BWR containment analysis model. The design source functions are defined to produce dynamic pressures which reasonably envelope the design spectrum based on JAERI CRT data in frequency domain. As an application example, the dynamic loads induced by chugging have been assessed based on the proposed method and on the reported JAERI CRT data from the view point of conservative load evaluation. The applicability of the analysis code has also been confirmed, since the simulated dynamic pressures have shown features and magnitudes similar to those observed in JAERI CRT.

Evaluation of dynamic loads induced by chugging.

A dynamic load evaluation method has been proposed for chugging phenomena which are assumed to occur and produce relatively large amplitude pressure spikes in the pressure suppression pool of a BWR containment, in case of a postulated loss of coolant accident. The proposed method is based on the analysis code developed by the authors and on theseven vent full scale tests performed at Japan Atomic Energy Research Institute (JAERI CRT), considering random nature of chugging phenomena. The dynamic loads are obtained by applying the design source functions of impulsive nature to the vent pipe exists in each BWR containment analysis model. The design source functions are defined to produce dynamic pressures which reasonably envelope the design spectrum based on JAERI CRT data in frequency domain. As an application example, the dynamic loads induced by chugging have been assessed based on the proposed method and on the reported JAERI CRT data from the view point of conservative load evaluation. The applicability of the analysis code has also been confirmed, since the simulated dynamic pressures have shown features and magnitudes similar to those observed in JAERI CRT.

Pressure distribution due to a steam bubble collapse in a BWR pressure suppression pool

At the end of a steam blowdown into a pressure suppression pool severe steam bubble collapses occur. Because the condenser pipes are arranged in a regular pattern. There are two limit loading conditions: the simultaneous collapse with the same intensity at each condenser pipe which is mostly discussed and a strong collapse only at one pipe or a small group of pipes. Since the influence of the wall elasticity on the pressure input due to the steam bubble collapse is unessentially small, the complete bubble-fluid-structure problem may be solved in two steps: First the bubble-fluid interaction in the rigid containment and later on the fluid-structure interaction in the elastic containment due to the pressure input by the bubble-fluid system. The bubble-fluid problem is also separated in the analysis for the pressure time history and that for the pressure distribution. If compressibility effects are neglected, the steam bubble surface is to idealize as a sphere during the whole collapse time for analyzing the pressure time history. The pressure distribution is fixed then by the potential due to a stationary source in the bubble center. The structure system is solved as usual in structural dynamics by the eigenmode response analysis in which the fluid-structure system is substituted by some uncoupled mass spring oscillators. The aim of this paper is to clarify the behaviour of the bubble-fluid-structure system by simple models and by separating the problem into the important global behaviour and some additional localized effects. This seems to be necessary since the pool geometry especially for steel containments and also the analysis for the structural system, which consists of different shells is too complicated. The influence of different idealisations on the pressure distribution in the rigid containment and on the eigenmodes of the fluid-structure system is presented. Formulas are derived for estimating the eigenfrequencies, the eigenmodes and the pressure distribution as well as the pressure reduction by the wall elasticity. Furthermore the similarity rules for transfering the results from small experimental facilities to the complex ring containment are explained. Analytical and experimental results are compared and how the maximum pressure distribution in steel and concrete containments differ from each other. Finally the following engineering concept for a stepwise analysis is confirmed: The resulting wall loads due to the fluid-pressure and the inertia wall forces may be determined by a simple fluid-structure model and then the stresses due to these loads by a refined shell analysis.

Numerical Simulation of Boiling Water Reactor Vent-Clearing Hydrodynamics

Pressure suppression pools used in nuclear reactors are subject to hydrodynamic processes involving complicated free surface configurations. A new numerical method, SOLA-VOF, developed to handle such problems is described and evaluated through comparisons with laboratory test data. Results from numerous computations provide a detailed understanding of the hydrodynamic phenomena associated with boiling water reactor vent-clearing processes. In addition, calculations show the sensitivity of the results to variations in the water vapor content, vent submergence depth, vent orifice size, and to the influence of fluid-structure interactions.

Vent clearing analysis of a Mark III pressure suppression containment

An analysis of the vent clearing transient in a Mark III pressure suppression containment after a hypothetical LOCA is carried out. A two-dimensional numerical model solving the transient fluid dynamic equations is used. The geometry of the pressure suppression pool is represented and the pressure and velocity fields in the pool are obtained from the moment the LOCA occurs until the first vent in the drywell wall clears. The results are compared to those obtained with the one-dimensional model used for containment design, with special interest on two-dimensional effects. Some conclusions concerning the effect of the water discharged into the suppression pool through the vents on submerged structures are obtained. Future improvements to the model are suggested.

Auslegungskriterien und Konstruktive Gestaltung der Sicherheitsumschliessung bei Siedewasserreaktoren

A survey on design, essential criteria and function of the containment system for AEG—Telefunken's present boiling water reactor product line is given. Safety enclosures of other BWR's are in most aspects similar with respect to the basic design principles, so this review might serve as a guide. The system presented consists of a steel primary containment with pressure suppression pool protected by reinforced concrete, of a leak-tight shell around the primary containment with a reventing system, of a reinforced concrete reactor building as secondary containment and of auxiliary systems for air conditioning and filtering and for decay heat removal. A study reveals that the smallest primary containment is the most economic, because primary containment size governs reactor building dimensions. Since the same fuel elements are being used for all plants — the power being proportional to their number —, the hight of the reactor pressure vessel is almost independent of thermal power. There seems to be a potential to size pressure suppression containments nearly independently of thermal power. Therefore the limits of compact construction are shown. Design criteria presented for the overall containment system consider access to components for operation, inspection and recurring tests in order to maintain high plant availability. They take into account system lay-out under all possible loads and operating conditions as well as quality assurance. They consider independence, separation and diversification of systems as well as redundancy. The criteria are aimed at system and — as a consequence — environment protection under accident and, if necessary, external hazard conditions. A flash-light is thrown on some new test results and on conservatively predicted performance of the containment system under reactor accident conditions, showing margins to keep the system safely under control even under highly degraded conditions.