Pressure Suppression Containment Research Articles

Scaling analysis of the pressure suppression containment test facility for the small pressurized water reactor

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

Scaling Design of the Pressure Response Experimental Facility for Pressure Suppression Containment

Pressure suppression containment has been adopted extensively in the BWR reactor design due to the remarkable pressure suppression capacity. In order to study the pressure response characteristics of the containment in the initial stage of the loss of coolant accident, it is necessary to build the scaling facility to simulate the prototype. The analysis of the blow down source term and the steam condensation occurred in the suppression pool are conducted respectively for the prototype PWR. The suppression pipe system resistance term is also scaled with a certain ratio. The length and the diameter of the mian pipe and the branch pipe as well as the orifice diameter are calculated. At last, the prototype and the experiment facility are modeled by MELCOR to scale the source term. The results show that the scaling method in this paper is effective to predict the pressure evolution in the prototype containment during the loss of coolant accident.

Experimental study on the condensation of sonic steam in the underwater environment

Abstract Steam jet condensation is of great importance to pressure suppression containment and automatic depressurization system in nuclear power plant. In this paper, the condensation processes of sonic steam jet in a quiescent subcooled pool are recorded and analyzed, more precise understanding are got in direct contact condensation. Experiments are conducted at atmospheric pressure, and the steam is injected into the subcooled water pool through a vertical nozzle with the inner diameter of 10 mm, water temperature in the range of 25–60 °C and mass velocity in the range of 320–1080 kg/m2s. Richardson number is calculated based on the conservation of momentum for single water jet and its values are in the range of 0.16–2.67. There is no thermal stratification observed in the water pool. Four condensation regimes are observed, including condensation oscillation, contraction, expansion-contraction and double expansion-contraction shapes. A condensation regime map is present based on steam mass velocity and water temperature. The dimensionless steam plume length increase with the increase of steam mass velocity and water temperature, and its values are in the range of 1.4–9.0. Condensation heat transfer coefficient decreases with the increase of steam mass velocity and water temperature, and its values are in the range of 1.44–3.65 MW/m2°C. New more accurate semi-empirical correlations for prediction of the dimensionless steam plume length and condensation heat transfer coefficient are proposed respectively. The discrepancy of predicted plume length is within ± 10% for present experimental results and ± 25% for previous researchers. The discrepancy of predicted condensation heat transfer coefficient is with ± 12%.

The Integral Test Facility Karlstein

The Integral Test Facility Karlstein (INKA) test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1 : 24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein—a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event—for example pipe rupture.

Design of the Safe Integral Reactor

The Safe Integral Reactor (SIR TM) is an innovative pressurized water reactor (PWR) design with enhanced safety which has been developed jointly by ABB Combustion Engineering, Inc., Rolls-Royce and Associates Ltd., Stone & Webster Engineering corporation and AEA Technology. The concept is a unitized primary coolant system in which the reactor core, the pressurizer, the reactor coolant pumps, and the steam generators are contained in a single reactor pressure vessel. This simplified reactor coolant system is coupled with a reliable, easily operated and maintained balance of plant utilizing a pressure suppression containment. The safety systems are primarily passive, relying on natural circulation and a large heat capacity rather than active, AC-powered equipment. The design is intended to meet the requirements for passive light water reactors being developed in the United States and jointly sponsored by the Electric Power Research Institute and U.S. Department of Energy, and requirements for small, safer light water reactors (LWRs) in the United Kingdom.

Future directions in boiling water reactor design

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proved features from BWR designs in Europe, Japan, and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 in. last stage buckets; and advanced radwaste technology. The ABWR is being developed as the next generation Japan standard BWR under the guidance and leadership of the Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. During 1987, the Tokyo Electric Power Company, Inc. announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa Nuclear Power Station, with commercial operation of the first unit in 1996 and the second unit in 1998. The units will be supplied by a joint venture of General Electric, Hitachi and Toshiba, with General Electric selected to supply the nuclear steam supply systems, fuel and turbine/generators. In the United States it is being adapted to the needs of U.S. utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the U.S. Nuclear Regulatory Commission for certification as a preapproved U.S. Standard BWR under the U.S. Department of Energy's ALWR Design Verification Program. These cooperative Japanese and U.S. Programs are expected to establish the ABWR as a world class BWR for the 1990s. International cooperative efforts are also underway aimed at development of a simplified BWR employing natural circulation and passive safety systems. This BWR concept, while only in the conceptual design stage, shows significant technical and economic promise.

LOCA steam condensation loads in BWR Mark II pressure suppression containment system

Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates ( ∼ 30 kg/m 2·s ) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.

Boundaries between regimes of pressure oscillation induced by steam condensation in pressure suppression containment

Boundaries between various regimes of pressure oscillation, occurring when steam is made to condense in pool water through vent tubes, were analyzed. “Bubbling”, characterized by low-frequency oscillation only, occurs between temperature boundaries of low- and high-frequency components which are obtained by the linear stability analysis for movement of the steam-water interface. The boundary between small chugging (consisting of low and high frequencies) and condensation oscillation (high frequency only) correlates well with the Hodgson Number to control pulsating gas flow. The threshold of internal chugging, characterized by periodic back flow of pool water into vent tubes, appears when the eq. term relating to the sucking up of pool water due to steam condensation balances the term relating to the pushing down of the interface due to supplied steam.

Fluid and pressure oscillations occuring at direct contact condensation of steam flow with cold water

Pressure and fluid oscillations at pressure suppression containment and cold leg flow oscillations at Emergency Core Cooling Water Injection in water-cooled nuclear power reactors are studied concerning flow oscillations occuring at direct contact condensation of steam flow with cold water. Classifications and mechanisms of these oscillation patterns are presented. Linear frequency analyses are developed in explaining the oscillation frequency and oscillation threshold of both condensation oscillation and plug oscillation. Oscillation frequencies at chugging and at ON-OFF oscillation are examined with analytical models developed for such large amplitude oscillations. Analytical results with these models are compared with the data by simulation experiments. Discussions are made focusing on the applicability of the analyses.

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.

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.

Vent Clearing during a Simulated Loss-of-Coolant Accident in a Mark I Boiling Water Reactor Pressure-Suppression System

The response of the pressure-suppression containment system of Mark I boiling water reactors to a large break, design basis loss-of-coolant accident was examined on a 1/5-scale experimental facility. A logical interrelationship between measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) is established. Peak downward forces on the wetwell occur at about the time of vent clearing. Peak upward forces occur shortly before bubble breakthrough. Quantitative values of forces and pressures from our 1/5-scale experiment, at times up to and including the peak download, can be applied to full-scale plants using established scaling laws. Only qualitative relationships to full-scale plants are determined at later times, because substantial quantities of steam (not simulated in our 1/5-scale experiment) would have entered the wetwell.

The swedish underground containment studies: State of art

Since 1973 studies of underground siting for nuclear power plants have been going on in Sweden. War protection, being the primary aim in accordance with the instructions, the first containment study has lead to siting in rock or in a pit. Rock siting gives better war protection than pit siting and also has less effect on the landscape, the cost being about equal. The second study was aimed at surveying the advantages and disadvantages of a rock sited 1000 MW BWA nuclear power plant from a reactor safety standpoint, compared to a plant above ground. Based on the instructions and considerations within the study group, the following criteria for the plant design have been established. (1) The plant should be designed to give protection against external acts of war with conventional weapons. (2) The plant should have a safety level equal to that of an above ground plant. It should fulfil the demands set by the authorities for above ground plants with respect to normal operation and accidents. No accidents that can be dealt with above ground may be permitted to result in more serious consequences, nor may they have a higher probability in a plant sited in rock. (3) The design of the plant should moreover utilize the possibilities of improving the safety afforded by rock siting. The criterion about war protection leads to siting in rock or pit, as shown in a previous CDL study. The study group has concentrated its work on rock siting. To clarify the two other criteria, the study group has outlined four alternative designs of a rock sited plant: (1) Reactor with complete pressure suppression (PS) containment placed together with central auxiliary equipment in a closed cavern with 48 m span that is placed about 50 m under the rock surface. (2) Reactor with complete PS containment placed together with auxiliary equipment in similar cavern as alternative (1) but open to the atmosphere. (3) Reactor placed ment. (4) Reactor placed in a containment directly surrounded by the rock. Auxiliary equipment placed in separate caverns. Standardization and quality improvement today are preferred to choosing new systems and more advanced technical solutions. Also, considering the desire from a safety standpoint, a rock sited plant should as much as possible exploit established technology. A consequence of the desire to use established technology is that the reactor cavern should be open to the atmosphere. If the cavern is closed, certain pipe rupture accidents may give overpressures that are difficult to master without large design alterations. The criterion about utilizing the possibilities for increasing the safety leads the interest to extreme improbable accidents, where certain advantages seem to be attainable with rock siting. A tight, strong cavern around the reactor would thereby be an ideal solution. This design appears, however, difficult to combine with the criterion about equal safety level as above ground. This criterion that controls the safety in the circumstances normally considered for nuclear power plants must be satisfied primarily, since extreme accidents have such very low probability. The tight cavern has therefore had to stand back for the open. The study shows, however, that an open reactor cavern can also be designed to significantly increase the protection of the surroundings in extreme improbable accidents compared to above ground plants. The chosen technical design of the reactor plant demands a cavern with a 45–50 m span. Caverns without strengthening efforts with such spans are used in mines, but have not previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant. The third study was delivered to the government in 1977. One part in this study is going deeper in certain questions (safety, operation, maintenance, sabotage, war protection, cost and decommissioning). Another part aims to a broader view of risks and consequences in peace and war and also advantages and disadvantages of nuclear power plant for district heating.

Pressure Suppression Containment Design—Current State of the Art

A summary of the thermodynamic design limits for pressure suppression containments for nuclear reactors is presented. Those parameters which must adhere to tested values are tabulated and discussed. An analytical model is described and is shown to accurately predict the existing test data. A graphical technique for predicting the transient peak containment pressure, based on the model, is presented for use in containment design.

Foundation caisson provides underground containment for nuclear power plant

A reinforced concrete caisson provides an economical underground structure for housing a portion of the Humboldt Bay Nuclear Plant. In addition to housing the nuclear steam supply system, the caisson also serves as part of the pressure suppression containment. The unique differences between this caisson and the conventional marine caissons are pointed out. The caisson was sunk immediately adjacent to an existing steam power plant. Special sinking techniques were used to reduce the surface friction, soil subsidence, and sinking weight. Upon completion, the underground structure served as a foundation for the plant refueling building and 250 ft high ventilation stack. The use of a caisson structure proved to be a sound solution to the varied requirements of plant layout, containment, construction, adjacent foundations, and budget.