Suppression Pool Research Articles

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.

Mental health care and East Japan Great Earthquake

Mental health care and East Japan Great Earthquake

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 °C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 °C). Additionally, the peak drywell temperature of 155.3 °C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 °C, which is below the pool temperature used for evaluating the net positive suction head of pumps of the RHR system and the Emergency Core Cooling Systems (96.7 °C). The peak containment pressure and temperature are well below the design value (386.1 kPaG and 171.1 °C). Containment integrity of Chinshan Plant can be maintained under the SPU condition.

ICONE19-43705 SUPPRESION POOL VOID DISTRIBUTION DURING BLOWDOWN

The void fraction distribution and fluid dynamics of the Boiling Water Reactor (BWR) Suppression Pool (SP) during blowdown is studied using the SP of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. A pipe with inner diameter of 0.102 m is installed in the SP to simulate the downcomer pipe in the suppression chamber of the prototypic plant. To simulate the early blowdown stage of Loss Of Coolant Accident (LOCA), the air is injected air with a certain volumetric flow rate through the downcomer submerged in the water of SP. The void fraction distribution, bubble velocity, and bubble chord length around the downcomer exit are measured by the single-sensor and double-sensor conductivity probes mounted on the instrument cage. The sixteen tests which varied with the downcomer void condition, air volumetric flow rate and velocity ramp rate are performed and the test results are obtained. The parametric study reveals that the air volumetric flow rates have effect on the void fraction distribution and void penetration depth during the initial air injection phase and the quasi-steady phase. The initial downcomer void conditions strongly effect the void fraction distribution for the initial air injection phase while the ramp rates of air velocity have minor impact on the void fraction distribution in the both of phases.

Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

Abstract In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air–water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.

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.

Jet-plume condensation of steam–air mixtures in subcooled water. Part 2: Code model

The system-analysis thermal-hydraulics code, TRACE, was used to model jet-plume condensation of steam and steam–air mixtures in a subcooled pool. The code model was used to compare the pool temperature development in a scaled-down test facility suppression pool (SP) represented by an idealized trapezoidal cross-section and 1/10 sector of a with scaled height ratio of 1/4.5 and volume ratio of 1/400. Thermal stratification and pool mixing data obtained in the experimental test section at different pool initial subcooling and steam/air mixture flow rates were used a validation suite to benchmark against the performance of the code. Agreement of the code simulations to experiments were seen in cases of pool mixing, however, instances of pool thermal stratification and transitional regimes between stratification and mixing were not predicted correctly by the code. The experimental database described in part 1 to this paper provides the benchmark suite to be used for future developmental improvement in areas of the code which may have future application to analysis of thermal heat removal rates in large pools.

Scaling of passive condenser system for separate effect testing

The passive containment cooling system (PCCS) of the simplified boiling water reactor (SBWR) is a passive condenser system designed to remove energy from the containment for long term cooling period after a postulated reactor accident. Depending on pressure condition and noncondensable (NC) gas fraction in drywell (DW) and suppression pool (SP), three different modes are possible in the PCCS operation namely the forced flow, cyclic venting and complete condensation modes. The prototype SBWR has total of six condenser units with each unit consisting of hundreds of condenser tubes. Simulation of such prototype system is very expensive and complex. Hence a scaling analysis is used in designing an experimental model for the prototype PCCS condenser system. The motive for scaling is to achieve a homologous relationship between an experiment and the prototype which it represents. A scaling method for separate effect test facility is first presented. The design of the scaled test facility for PCCS condenser is then given. Data from the test facility are presented and scaling approach to relate the scaled test facility data to prototype is discussed.

The Impact of Severe Accident Management Guidance on the Frequency of Containment Failure Modes of a Boiling Water Reactor

Probabilistic safety assessment (PSA) employs a systematic approach to estimate the risk associated with the operation of nuclear power plants (NPPs). Severe accident management guidance (SAMG), which delineates the mitigation actions of core meltdown accidents of NPPs, is developed to support operators and staffs in the technical support centers during the emergency responses of core melt accidents. Proper execution of SAMG could lower the failure probability of containment and reduce the amount of radionuclides released to the environment during the accident. It can be expected that the implementation of SAMG will reduce the risk of NPPs. However, SAMG is not available when most of the conventional level-2 PSA analyses are performed. In the present study, the mitigation actions of SAMG are incorporated into the level-2 PSA model of the ChinShan Nuclear Power Station of the Taiwan Power Company. The NPP analyzed employs a General Electric-designed boiling water reactor-4 with Mark I containment.The effectiveness of the mitigation actions specified in SAMG to terminate the progression of the accident is verified and validated using the MAAP4 code. The containment system event trees and containment phenomenological event trees of the level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The Human Cognitive Reliability (HCR) and Technique for Human Error Rate Prediction (THERP) models are used to quantify the human error probability (HEP) of all the actions in the level-2 PSA model. The MAAP4 code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the total frequency of accident progression beyond vessel failure is reduced by 41% and the change in the probability of containment staying intact is not very significant because of the implementation of SAMG. After SAMG implementation, the frequency of containment early failure is reduced by 69.9%. The frequency of suppression pool venting is increased by 77.9%. The changes in the frequency of other containment failure modes are relatively insignificant. The most important human action is specified in Guideline RC/F of Severe Accident Guideline-1, i.e., In-Vessel Injection to Arrest Core Damage.

Validation of RCS Bleed-and-Feed Strategy for the SAMG of the Kuosheng Nuclear Power Plant

The severe accident management guideline (SAMG) was developed and implemented at the Kuosheng nuclear power plant (NPP) in Taiwan at the end of 2003. The Kuosheng NPP is a Mark-III boiling water reactor (BWR)-6 located in the north of Taiwan. The SAMG of the Kuosheng NPP is developed based on the BWR Owners Group Emergency Procedure and Severe Accident Guidelines. In this paper, MELCOR 1.8.5 is used to investigate the effectiveness of the Kuosheng SAMG and to analyze the reactor coolant system (RCS) bleed-and-feed strategy in a postulated station blackout (SBO) accident since this accident is a dominant sequence that can induce core damage. These simulation results can be applied for the typical BWR-6 NPPs. Based on the simulation results, the major events of an SBO accident without any operator actions, including core uncovery, cladding oxidation, hydrogen generation, fuel relocation, vessel failure, and containment failure, are well presented. The RCS bleed-and-feed strategy can cool down the hot core and bring the vessel into the stable condition in the SBO accident with high-pressure core spray injection according to the Kuosheng SAMG. The reactor pressure vessel depressurization before the reactor liquid level reaches one-fourth of the core liquid level can prevent damage to the core fuel. However, the water temperature of the suppression pool will reach saturation temperature, and the containment pressure may challenge the containment integrity. These phenomena were not considered while developing the SAMG.

A study on the boron injection initiation temperature curve of BWR

Boron injection initiation temperature (BIIT) provides important information for the safe shutdown of the reactor using boron injection system during anticipated transient without scram (ATWS). The purpose of this paper is to study BIIT curve of boiling water reactor owners’ group (BWROG). The unreasonable and non-conservative parts of BIIT are pointed out and suggested modifications are made. The starting reactor power of BIIT is increased in order to meet the actual application. The lower limit of suppression pool temperature of BIIT is revised for conservative operation during ATWS conditions. Analysis of the effects of maximum temperature capacity of the suppression chamber and concentration of boron in standby liquid control tank shows that BIIT is decreased by adopting a more conservative value of maximum temperature capacity of the suppression chamber. Consequently, early boron injection is anticipated. For system with automatic boron injection system, BIIT is not required.

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10−2 (mean) and 3.9x 10−2 (median) for the BWR suppression pool case, 2.2x10−3 (mean) and 2.8x10−10 (median) for the BWR pedestal case, and 6.8X10−2 (mean) and 1.4x10−2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.

Enhancing the Utility of ProstaScint SPECT Scans for Patient Management

This project investigated reducing the artifact content of In-ill ProstaScint SPECT scans for use in treatment planning and management. Forty-one patients who had undergone CT or MRI scans and simultaneous Tc-99m RBC/In-111 ProstaScint SPECT scans were included. SPECT volume sets, reconstructed using Ordered Set-Expectation Maximum (OS-EM) were compared against those reconstructed with standard Filtered Back projection (FBP). Bladder activity in Tc-99m scans was suppressed within an ellipsoidal volume. Tc-99m voxel values were subtracted from the corresponding In-111 after scaling based on peak activity within the descending aorta. The SPECT volume data sets were merged with the CT or MRI scans before and after processing. Volume merging, based both on visual assessment and statistical evaluation, was not affected. Thus iterative reconstruction together with bladder suppression and blood pool subtraction may improve the interpretation and utility of ProstaScint SPECT scans for patient management.

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200. The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations. Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.

Safety System Design Concept and Performance Evaluation for a Long Operating Cycle Simplified Boiling Water Reactor

The long operating cycle simplified boiling water reactor is a reactor concept that pursues both safety and the economy by employing a natural circulation reactor core without a refueling, a passive decay heat removal, and an integrated building for the reactor and turbine. Throughout the entire spectrum of the design basis accident, the reactor core is kept covered by the passive emergency core cooling system. The decay heat is removed by the conventional active low-pressure residual heat removal system. As for a postulated severe accident, the suppression pool water floods the lower part of the reactor pressure vessel (RPV) in the case when core damage occurs, and the in-vessel retention that keeps the melt inside the RPV is achieved by supplying the coolant. The containment adopts a parallel–double-steel-plate structure similar to a hull structure, which contains coolant between the inner and outer walls to absorb the heat transferred from the inside of the containment. Consequently, the containment structure functions as a passive containment cooling system (PCCS) to remove the decay heat in case of an accident. This paper describes the PCCS performance evaluation by using TRAC code to show one of the characteristic plant features. The core damage frequency for internal events was also evaluated to examine the safety level of the plant and to show the adequacy of the safety system design.

Dynamics of contained liquids: A personal odyssey

Dynamics of contained liquids: A personal odyssey

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.

10.2172/1079302

This is the fourth in a series of reports on the response of a BWR/5 boiling water reactor to anticipated transients without reactor scram (ATWS) when operating in the expanded operating domain MELLLA+. In this report ATWS events initiated by closure of main steam isolation valves are analyzed at beginning-of-cycle, and end-of-full-power-life, conditions. The objective is to understand the sensitivity of ATWS-ED events to the intial operating core flow and to the spectrally corrected moderator density history (void history). Different water level control strategies are considered. The ATWS events are simulated for a sufficiently long time (2500 s) to understand the response of key components and the potential for fuel damage or damage to the wetwell (suppression pool). These events lead to the automatic trip of recirculation pumps; and operator actions to activate the emergency depressurization system when the wetwell has reached the heat capacity temperature limit, and to control power through water level control and the injection of soluble boron. The simulations were carried out using the TRACE/PARCS code system and models developed for a previous study with all relevant BWR/5 systems.