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A system includes a slurry depressurizing system that includes a liquid expansion system configured to continuously receive a slurry at a first pressure and continuously discharge the slurry at a second pressure. For example, the slurry depressurizing system may include an expansion turbine to expand the slurry from the first pressure to the second pressure.
Integrated small modular reactors (iSMRs) are favored because of their unique characteristics, such as enhanced safety without any large break loss of coolant accidents (LBLOCAs), low initial construction costs, and multi-use of desalination and electricity generation. Many iSMRs have been developed thus far. In Korea, an iSMR called SMART was developed after a thorough validation of the components and systems, and finally received its design certificate in 2012 for the first time in the world. To ensure the performance of its safety systems, an integral test loop, SMART-ITL, is constructed. It contains the primary system, secondary system, passive safety systems, and an auxiliary system. The passive safety systems include a passive residual heat removal system (PRHRS), passive injection system with a core makeup tank (CMT) and a safety injection tank (SIT), and an automatic depressurization system (ADS). The height is preserved while the diameter is scaled down to 1:7. A series of SBLOCA tests have been performed with 4 trains of PRHRS, 1 train of CMT and SIT, and ADS. In parallel with the experiments, simulation of a representative test has been conducted to track the underlying thermal hydraulics. SMART-ITL is an integral test loop for SMART and a series of SBLOCA tests for confirming the performance of passive safety systems have been conducted. In this paper, a system code simulation on a SBLOCA test using SMART-ITL is presented. The reactor vessel water level, break flow rate, and CMT and SIT water levels are predicted well. (authors)
A system includes a first pump having a first outlet and a first inlet, and a controller. The first pump is configured to continuously receive a flow of a slurry into the first outlet at a first pressure and to continuously discharge the flow of the slurry from the first inlet at a second pressure less than the first pressure. The controller is configured to control a first speed of the first pump against the flow of the slurry based at least in part on the first pressure, wherein the first speed of the first pump is configured to resist a backflow of the slurry from the first outlet to the first inlet.
The accidents at the Fukushima Daiichi nuclear power station in March of 2011 renewed interest globally in the nature, causes, and health and economic consequences of severe accidents involving core damage and containment impairment. As a result of these accidents, research has been conducted into the modeling and phenomenological understanding of the accidents at Fukushima and the response and effectiveness of mitigating measures such as emergency water injection, primary system depressurization and containment venting actions. An important example of this is the jointly conducted OECD-NEA organized Benchmark of the Fukushima Accidents (BSAF project) where accident reconstruction analyses using computer codes such as MELCOR and MAAP are ongoing. Even prior to the Fukushima accidents, the Sandia/NRC sponsored State of Art Reactor Consequence Analyses (SOARCA) foreshadowed the accidents in Japan where long-term and short-term station blackout accidents in boiling water reactors with Mark-1 containment were analyzed showing striking similarities to the accidents at Fukushima. These analyses also highlighted important differences in our prior understanding, in particular the importance of suppression pool thermal stratification and the extraordinary operation of the Fukushima Unit-2 reactor core isolation cooling (RCIC) system, prompting addition research into both of these areas. Additionally, the focus of the world's major computer codes on modeling the Fukushima accidents in the BSAF effort, led to several 'Crosswalk' comparisons between code predictions, highlighting subtle but important modeling differences between the codes and identifying areas where modeling improvements can be made. Collectively, these research areas are aimed at improving our understanding of severe accident progression and real-world response of safety related systems and components, and should lead to improved accident response and mitigation capabilities in the future. (authors)
This report presents an analysis of the Emergency Core Cooling System (ECCS) for a generic Boiling Water Reactor (BWR)-4 NPP. The Electric Power Research Institute (EPRI) developed Hazards and Consequences Analysis for Digital Systems (HAZCADS) process is applied to the ECCS and its subsystems to identify unsafe control actions (UCAs) which act as possible cyber events of concern. The analysis is performed for two design basis events: Small-break Loss of Coolant Accident (SLOCA) and general transients (TRANS), such as unintended reactor trip. In previous work, HAZCADS UCAs were combined with other cyber-attack analysis to develop a risk-informed approach; however, this was for a single system. This report explores advanced systems engineering modeling approaches to model the interactions between digital assets across multiple systems which may be targeted by cyber adversaries. The complex and interdependent design of digital systems has the potential to introduce emergent cyber properties that are generally not covered by hazard analyses nor formal nuclear Probabilistic Risk Assessment (PRA). The R&D and supporting analysis presented here explores approaches to predict and manage how interdependent system properties effect risk. To show the potential impact of a successful cyber-attack to formal PRA event tree probabilities, HAZCADS analysis was also used. HAZCADS was also used to model the automatic depressurization system (ADS) automatic actuation. This analysis extended to an integrated system analysis for common-cause failure (CCF). In this aspect, the HAZCADS analysis continued by analyzing plant design details for system connectivity in support of critical plant functions. A dependency matrix was developed to depict the integrated functionality of the interconnected systems. Areas of potential CCF are indicated. Future work could include adversary attack development to show how CCF could be caused, resulting in PRA events. Across the multiple systems that comprise the ECCS, the analysis shows that the change in such probabilities was very different between systems. This indicates that some systems have a larger potential risk impact from successful cyber-attack or digital failure, which indicates a need for these systems to have a higher priority for design and defensive measures. Furthermore, we were able to establish that a risk analysis using any arbitrary threat model establishes an ordering of components with regard to cyber-risk. This ordering can be used to influence the overall system design with an eye to lowering risk, or as a way to understand real-time risk to operational systems based on a current threat landscape. Expert knowledge of both the analysis process and the system being analyzed is required to perform a HAZCADS analysis. The need for a tiered risk analysis is demonstrated by the results of this report.
The overall objective of this study is to develop a high-fidelity simulation capability for the analysis of accident progression during Station Black-Out (SBO) scenario in Boiling Water Reactor (BWR) Mark I plant. The motivation is on predicting the effectiveness of containment venting within accident mitigation strategy and how to avoid core melt by choosing a proper time to initiate primary system venting. In this work, a GOTHIC model has been developed to support characterization of reactor safety systems performance and evaluation of the venting strategy. The GOTHIC model provides a seamless coupled simulation of the reactor coolant system and the containment system based on the design of the Peach Bottom-2 plant system. Efforts are also taken on the analysis for the primary system venting strategy. The safe time intervals (so-called “safe venting window”) to initiate the primary system depressurization in order to optimize the early cooling strategy by adding fire water and avoid core melt are also studied based on the heat removal capability (HRC) of the reactor vessel coolant. This concept is instructive for the operation of the safety systems during the SBO accident mitigation.
Air-ingress following a depressurized loss-of-forced-cooling (DLOFC) event is a challenging, multiphysics safety scenario for High-Temperature Gas-Cooled Reactors (HTGRs), involving coupled gas composition transport, buoyancy-driven flow redistribution, graphite oxidation, and structural heat-up. Despite its importance — air ingress is a key scenario identified in the PIRT process for the HTGRs — existing system-level safety codes have lacked the integrated capability to simulate the complete event sequence with high confidence. This report documents the development, validation, and demonstration of three new capabilities in the SAM code to address this gap: (1) a multi-component gas mixture flow model with binary diffusion to track the helium-air composition and its effect on system density and flow; (2) a 0-D graphite oxidation model based on the Roes correlation, including oxygen consumption and exothermic heat release; and (3) an isentropic critical flow model for accurate representation of primary system depressurization through a break. These capabilities are validated against two benchmark experiments. The NSTF heavy-gas ingress experiment validates the multi-component flow model: SAM correctly reproduces the rapid buoyancydriven flow stagnation and subsequent natural circulation recovery driven by composition-dependent density changes. The NACOK graphite oxidation experiment validates the oxidation model: SAM predicts a bottom-level graphite weight loss of 25%, in close agreement with the measured 24%, and reproduces the strong axial nonuniformity and block-geometry dependence of oxidation, at a level comparable to the SPECTRA and TINTE codes. The validated capabilities are then exercised together in an integrated, reactor-scale simulation of a DLOFC air-ingress transient in a simplified HTR-PM pebble-bed reactor. In a single calculation spanning approximately 8 days, SAM reproduces the complete accident sequence: rapid depressurization, densityand diffusion-driven air ingress over ˜15 hours, onset of buoyancy-driven natural circulation, exothermic graphite oxidation with a peak fuel temperature at ˜62 hours, and eventual passive cooldown. These results demonstrate that SAM now provides the nuclear community with a preliminarily validated, modern systemlevel tool for HTGR air-ingress safety analysis, filling a recognized capability gap. Future extensions to broaden species tracking, improve oxidation chemistry, and refine the reactor model are discussed.
The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving either only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.
This paper presents a scaling-based applicability assessment of the Integrated Systems Test (IST) facility located at the Center of Advanced Engineering Research in Forest, Virginia. The assessment draws upon both traditional and contemporary scaling methods, seeking an optimal representation of light water reactor (LWR) and small modular reactor (SMR) characteristics during design-basis events. Based on published LWR loss-of-coolant-accident phenomena identification and ranking tables compiled with a similar assessment completed by BWXT mPower, Inc., for the mPower{sup TM} SMR, the phenomena considered in the IST design include break flow, decay heat, loop thermal-hydraulics, and primary-to-secondary heat transfer. Among these, the single- and two-phase natural circulation and system depressurization, by association to loop thermal-hydraulics, were the principal challenges. (authors)
A continuous slag processing system includes a rotating parallel disc pump, coupled to a motor and a brake. The rotating parallel disc pump includes opposing discs coupled to a shaft, an outlet configured to continuously receive a fluid at a first pressure, and an inlet configured to continuously discharge the fluid at a second pressure less than the first pressure. The rotating parallel disc pump is configurable in a reverse-acting pump mode and a letdown turbine mode. The motor is configured to drive the opposing discs about the shaft and against a flow of the fluid to control a difference between the first pressure and the second pressure in the reverse-acting pump mode. The brake is configured to resist rotation of the opposing discs about the shaft to control the difference between the first pressure and the second pressure in the letdown turbine mode.
In Korea, Severe Accident Management Guidance (SAMG) for Optimized Power Reactor (OPR1000) has been developed through the research results related to the WOG SAMG, in which the operation phase shifts from the Emergency Operation Procedure (EOP) to the SAMG when the Core Exit Temperature (CET) reaches 923 K. Among 7 mitigation strategies envisioned, the third management strategy, injecting coolant into Reactor Coolant System (RCS) by depressurizing the RCS is one of the key strategies to fortify the in-vessel retention capability. As an aid to such mitigation strategy, a safety injection (SI) flow map has been developed by Lee et al. for predicting required flow rate associated with target RCS depressurization to recover in-vessel coolability. The mechanistic SI flow map used the CET increase rate and core water level decrease rate obtained from MELCOR code simulation. It should be noted that during the severe accident, the total heat generated is sum of the decay heat and the oxidation heat. Thus, it is necessary to consider additional heat sources by oxidation reaction between cladding, structure, control rod, and hot steam prior to the injection of coolant. It was believed that the SI model estimated required flow rate rather conservatively by considering the expected amount decay heat and the accumulated heat in the core based on the RCS pressure. However, it was confirmed that the SI model developed by Lee et al. could not represent the total heat removal especially since start of oxidation. Under transient conditions, the heat sources accumulated in the heat structures (fuel, cladding, supporting structures) should be embarked in the model for accurate contribution of the total heat source to the required flow rate. Therefore, the objective of this study is to improve the previous SI flow map predicting the injection flow rate associated with target RCS depressurization to satisfy the in-vessel coolability during severe accidents. Additional terms were developed to reflect accurate heat accumulating and included in the existing model. Resulting SI flow map includes the target RCS depressurization, which enables discharge of the required flow rate by utilizing the flow rate-pressure curve with operation of two high pressure safety injections (HPSIs). Also recent MELCOR simulation result of a postulated steam generator tube rupture (SGTR) accident verified the accuracy of the improved map.
Abstract Bioconversion processes offer many economic, environmental, and societal advantages for production of fuels and chemicals. Successful commercialization of any biotechnology usually requires accurate characterization of cell growth dynamics, substrate conversion and production excretion rates. Despite recent advancements in analytical equipment, obtaining accurate measurement of gas component uptake or production rates remains challenging due to their high sensitivity to system pressure or volume changes. Specifically, the consumption and production of various gases will result in changes in system pressure (for batch operations) or off-gas flow rate (for continuous operations). These changes would cause significant errors in the estimated gas component uptake and production rates if they were not accounted for. In this work, we propose two easy-to-implement experimental protocols and associated calculation procedures to obtain accurate measurements of gas component consumption and production rates; one is for batch operation and one is for continuous operation. For depressurized ( i . e ., system pressure below 1 atm) batch cultures, nitrogen (or other inert gases) is used to repressurize the system to 1 atm before taking sample; while for continuous cultures, He (or other inert gases) is used as an internal tracer to accurately measure off-gas flow rate. The effectiveness and accuracy of the two protocols and associated calculation procedures are demonstrated using several case studies with both abiotic and biotic systems.
Current oil recovery methods in hydraulically fractured shale reservoirs have a low recovery efficiency of about 10%. The objective of this work is to investigate the effectiveness of nitrogen and supercritical carbon dioxide in the huff-and-puff method for enhanced oil recovery as a means to re-energize reservoirs and improve recovery rates. Here, we conduct direct visualization experiments with a microfluidic system to reveal the mechanisms and to quantify the recovery rates of oil from fracture networks. We compared the effectiveness of water, nitrogen and supercritical carbon dioxide at reservoir conditions in a process mimicking the huff-and-puff method in both dead-end and connected fracture systems. The microfluidic chips were made of glass and placed in a confining pressure system pressurized to 10 MPa, 50 °C. The system was allowed to equilibrate, and then depressurized to simulate huff-and-puff oil recovery. Fluorescence microscope images were continuously taken to visualize and calculate residual oil saturation as a function of pressure drawdown. As the system was depressurized from 10 MPa, gas exsolution from the oil liquid phase, including bubble nucleation, growth, and coalescence, appeared to be the main driver for mobilizing oil from the fracture networks. Injection of supercritical CO 2 resulted in the highest recovery rate with an average end-point recovery of about 90% in the connected fracture network and 60% in the dead-end fracture network. N 2 has lower solubility in oil and hence showed a lower recovery rate of 40% in the connected fracture network and 25% in the dead-end fracture network. Injection of water had no effect on oil mobilization since water is insoluble, immiscible and incompressible. The main mechanism of enhanced recovery was gas exsolution from the liquid phase as pressure was decreased below the bubble point pressure. Because the gas was distributed throughout the oil phase, bubble nucleation, growth, coalescence, and elongation occurred throughout the fracture network. Expansion of the gas forced oil out of the network through piston displacement in continuous oil areas and film flow in the dispersed oil areas. Bubbles began as spheres and grew until they touched the fracture walls where they elongated along the fracture length. The bubble growth rate depended on local mass transfer from liquid to gas phase and gas volume expansion due to pressure drop. The efficiency of the huff-and-puff process is dependent on the solubility and miscibility of the injection fluid with oil. High gas solubility allows for more bubble nucleation, growth and expansion during the depressurization cycle.
In the current nuclear power plants (NPPs), most of the critical safety functions are provided by many active safety systems. However, in order to meet the futuristic goals of safety, relying on these active safety systems alone does not seem to be viable. As passive system has a strength on the mitigation of station black out accident and it also contributes to increase the diversity of mitigation way, application of passive system becomes a paramount in the nuclear industry to enhance the nuclear plant safety. Recently, many researches for developing passive systems and operation strategies have been carried out in the nuclear field. General passive systems and strategies, however, have a critical limitation, especially for passive injection systems. As these systems use gravity as their driving force, it is difficult to inject coolant in a passive way in a high-pressure condition thus depressurization has to be preferentially performed to do passive injection. The problem is that depressurization inevitably causes the loss of core inventory thus it is not preferred in order to secure plant safety. Hybrid safety injection tank (H-SIT) system was newly suggested a few years ago to passively inject the coolant into the reactor coolant system (RCS) without depressurization in a high-pressure condition. The H-SIT system is planned to be integrated into advanced power reactor plus (APR+). Generally, the H-SIT system can inject water using the pressure from nitrogen gas as a conventional safety-injection tank in low-pressure accidents, such as large and medium-break loss of coolant accidents (LOCA). Additionally, the H-SIT system can also inject water using gravitational force in high-pressure accidents. If a high pressure accident occurs, the pressure of each H-SIT is equalized with pressure of pressurizer through equalizing pipes, thus allowing the H-SITs to inject water by gravitational force. In the target plant, four H-SITs are planned to be installed. Each has a 2.5 in diameter equalizing pipe and it can be operated manually. Long-term cooling of core is an ultimate goal of all mitigation actions for plant safety and feed and bleed (F and B) operation strategy is one of the long-term cooling strategies in conventional pressurized water reactor (PWR). The important point of F and B operation is that injection for feed operation is performed by only high pressure safety injection (HPSI) pump because operation of low pressure safety injection (LPSI) pump needs the depressurization and it causes core dry out. As H-SIT is passively operated in an accident situation, it was originally designed for mitigation of station black out accident. Besides, operator can use the H-SIT system same as high pressure injection pump in any accident situations. Thus, function of high pressure injection by H-SIT can be used for feed operation when HPSI pumps fail. If it is possible, H-SIT can make up the core during depressurization then LPSI pumps can be used for F and B operation. In this study, F and B operation strategy with LPSI and HSIT is developed. This is a new concept for the long-term cooling operation. If this strategy is applied, low pressure injection pump can be successfully used for F and B operation thus operator has the additional mitigation way. As this strategy make plant safe even though HPSI and PAFS are both failed, it can effectively enhance the plant safety. For this strategy two RCGVSs and two POSRVs are needed as a depressurization system for bleed operation and only one LPSI is enough for feed operation. H-SIT operation is also needed to make up core inventory during bleed operation. For this operation, four H-SITs have to be used to make up core safely. Based on the risk analysis using PSA method, if this strategy is applied, core damage frequency is 1.868 e-6 which declined 7 percent from original model. (authors)
Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governed by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.
Axial gas communication concerns the flow along the axial axis of nuclear fuel rods during ramp and loss of coolant accident (LOCA) conditions. During power ramps, the higher linear heat generation rate may cause fuel-to-clad gap closure that may prevent transport of released fission gases to the plenum. Upon reduction in power the gas then can communicate to the plenum. This phenomenon has been experimentally observed by short power dips during ramp experiments completed at the Risø reactor. At higher burnups it is observed that the UO2 fuel and Zircaloy cladding forms a chemical bond. This bond results in complete closure of the gap. When these high burnup rods are subjected to a LOCA, the bond has implications on both the mechanical response (i.e., ballooning) of the cladding and subsequent fuel relocation and axial gas communication. In the LOCA scenario, gas communication is of interest in two different regimes: 1) pre-rupture communication from the plenum towards the lower pressure ballooning area and 2) the post-rupture depressurization of the plenum to the external system pressure. In both regimes the presence of a fuel-to-cladding bond will impact the rate of depressurization. In this work we present a fuel-to-clad bonding model that is coupled to an existing axial gas communication model framework in the BISON fuel performance code. The effect of considering the bond on fuel performance modeling predictions is presented through comparisons to existing experimental data. Experiments considered include several rods from the Halden IFA-650 test series. An evaluation on a full-length rod that explores the combined effect of plenum size and bonding status on axial gas communication behavior is also presented.
In many residential building retrofit programs, air tightening to increase energy efficiency is often constrained by safety concerns with naturally vented combustion appliances. Tighter residential buildings more readily depressurize when exhaust equipment is operated, making combustion appliances more prone to backdraft or spill combustion exhaust into the living space. Several measures, such as installation guidelines, vent sizing codes, and combustion safety diagnostics, are in place with the intent to prevent backdrafting and combustion spillage, but the diagnostics conflict and the risk mitigation objective is inconsistent. This literature review summarizes the metrics and diagnostics used to assess combustion safety, documents their technical basis, and investigates their risk mitigations. It compiles information from the following: codes for combustion appliance venting and installation; standards and guidelines for combustion safety diagnostics; research evaluating combustion safety diagnostics; research investigating wind effects on building depressurization and venting; and software for simulating vent system performance.
Here, we investigated numerical simulation strategies for a long-term test of depressurization-induced gas production from the B1 Sand of Unit B at the Hydrate-01 Stratigraphic Test Well. The main objective of this study was to estimate fluid production rates (with emphasis on water production) under a variety of conditions and production scenarios and contribute new insights to the design and management of the field test. In the first part of the study, we investigated the system response to a three-step depressurization process using two limiting sets of flow properties─the expected maximum and minimum intrinsic and effective permeabilities─for the very heterogeneous reservoir. In the second part, we investigated the effect of the production interval length and placement within the formation relative to the boundaries of the hydrate-bearing unit. The best performing well configuration was used in the third part of the study, which used the most representative subsurface flow properties to investigate the effect of the depressurization strategy on the production performance. The best overall performance (largest gas production with modest water production and a strong response at the observation wells) was obtained with a 10 m-long well situated 3 m below the top of the formation and a three-step depressurization scheme at 15-day intervals to a terminal bottomhole pressure of 2.8 MPa. The overall production performance was enhanced by a faster rate of depressurization. Estimated water production rates in all cases were limited and easily manageable. None of the tested well configurations or depressurization strategies significantly reduced water production without also severely reducing gas production. In all the investigated cases, 95% of the long-term fraction of produced water was replenished by inflows from the boundaries and could not be reduced. These substantial water inflows are an unavoidable feature of HU-B and cannot be easily mitigated by a hydraulic control.