Safety Relief Valve Research Articles

Runaway Reaction Kinetics for Emulsion Polymerization and its Consequences

AbstractFor industrial semi‐batch emulsion polymerization, it is difficult to predict the composition of the reactor contents during a runaway reaction since many recipes are involved and it actually may be a fault in the feeds to the reactor that leads to the runaway reaction, as a distributed control system cannot safely prevent wrong feeds. Therefore, in order to safeguard the reactor, very high kinetics need to be taken into account, as discussed theoretically as well as based on an example. Corresponding maximum temperature increase rates may be above 100 K · min−1. This needs to be taken into account when designing safety measures for emulsion polymerization reactions. As a result, large safety relief valves or rupture discs may need to be installed.magnified image

Mixture loss coefficient of safety valves used in nuclear plants

Theoretical and experimental investigations on the loss coefficient of gas–liquid mixture across safety relief valves have been carried out. Experiments were performed for three different types of safety valves and under different flow conditions. Using the Darcy equation and based on the presented experimental results, a new empirical correlation has been developed to calculate the loss coefficient and hence pressure loss. By consideration of flow contraction, high viscous fluids, Reynolds number and safety valve geometry, the model includes therefore the relevant primary influencing parameters. The reproductive accuracy of the proposed model and the statistical comparison, based on about 2000 measured data in the literature, demonstrated that the proposed model is the best overall agreement with the data. The standard deviation of the data is less than 27%. The model fits the data well and is sufficiently accurate for engineering purposes. The reported results of the tested safety relief valve are very important to improve the practical and safety design of the nuclear plants.

沸騰水型原子炉蒸気乾燥器の流力音響振動(流体工学,流体機械)

The boiling water reactor (BWR) in the Quad Cities Unit 2 Nuclear Power Plant experienced a dryer failure under extended power up-rate conditions. The cause of the dryer failure was considered as flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it was also desirable to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analysis to investigate the flow-induced acoustic resonance in SRV stub pipes and propagation of fluctuating pressure from SRVs to the dryer through the MSLs. We demonstrated that acoustic resonance occurred in SRV stub pipes at higher velocity flows than in the normal operation and fluctuating pressure propagated from SRVs to the dryer. The amplitude of the fluctuating pressure due to several stub pipes was much larger than that in one stub pipe because of interaction between them. The flow-induced acoustic resonance occurred for Strouhal number below 0.6 in the MSLs system of the BWR-5. Results of the tests were compared with those of acoustic analysis. The acoustic analysis could well predict the dryer loading.

Computation of nozzle flow capacities for superheated steam, subcooled water, and saturated steam/water mixtures

A new computational tool models the discharge of steam, water, and saturated steam–water mixtures from safety relief valve nozzles, employing thermodynamic-property formulas derived from the Helmholtz equation and programmed into spreadsheet-based software. The computational approach maps an isentropic process from initial conditions in a series of thermodynamic states at each of which the mass flux is computed. The user can identify the maximum mass flux from the generated mass-flux pressure graphical profile. The work describes other mass-flux predictive techniques commonly used in industry and then compares the results of the proposed method against those of the others. The data indicate a very high degree of correspondence between the proposed method and the Napier equation for saturated and superheated steam. The proposed method produces saturated-liquid results consistent with those of the other methods, particularly ASME VIII/1. For the relief of saturated liquid–vapor mixtures, the proposed method increasingly overpredicts maximum mass flux relative to the HEM-based ω-method for increasing pressures and qualities. In the low-quality range, for which there is experimental data to which to compare, the proposed method severely underpredicts measured results involving flows out of nozzles of lengths less than “relaxation” length, but for flows out of longer nozzles, the underprediction is significantly less. The proposed method's subcooled-water results compared to measured data also indicates large underpredictions for sub-relaxation lengths with improved results at generally greater degrees of subcooling at greater lengths. The proposed technique shows generally good agreement with other established predictive methods within demonstrated boundaries and has advantages in stand-alone capability and versatility.

A COMPUTATIONAL FLUID DYNAMICS EVALUATION OF A PNEUMATIC SAFETY RELIEF VALVE

Safety relief valves are well established components preventing catastrophic failure of pressurised systems when non-normal operating conditions occur. However, it is only recently with developments in CFD techniques that the capability to predict the complex flow conditions occurring in the valves has been possible resulting in only limited studies being found in the literature. This paper presents experimental and theoretical investigations applied to a safety relief valve designed for the refrigeration industry but extended here to consider pneumatic systems since air is the compressible fluid. The discharge flow rate and valve forces are determined both theoretically and experimentally for different valve lift conditions and related to the detailed flow conditions (pressure, temperature and Mach number) in the valve predicted by CFD techniques. The CFD code FLUENT has been used with a two dimensional axisymmetric RANS approach using the k-I turbulent model to predict the highly compressible flow through the valve. The model has been validated by comparison with experimental measurements and the predicted results show good agreement, providing confidence in the use of CFD techniques for valve design and improvement.

Trend Analysis of Incidents Involving Setpoint Drift in Safety or Safety/Relief Valves at U.S. LWRs

Since the beginning of the 1980's, in the United States, there have been many licensee event reports (LERs) involving setpoint drift in safety or safety/relief valves. The United States Nuclear Regulatory Commission (NRC) has issued a lot of generic communications on this issue and the industry has made efforts to resolve the issue. However, the NRC staff recently highlighted that over 70 LERs involved instances where safety or safety/relief valves failed to meet the allowed setpoint tolerance from 2001 through August 2006. In the present study, we analyzed the U.S. experience with setpoint drift in safety/relief valves (SRVs) at BWRs, pressurizer safety valves (PSVs), and main steam safety valves (MSSVs) at PWRs by reviewing approximately 90 LERs from 2000 to 2006 and examined the trend focusing on causes and setpoint deviation ranges. This study indicates that for SRVs and MSSVs, disc-seat bonding is a dominant cause of the setpoint drifting high and has a tendency to result in a relatively large deviation of the setpoint. This means that disc-seat bonding might be a safety concern from the view point of overpressure protection. For PSVs, the deviation of setpoints is generally small, although its causes are not specified in many instances.

Best Estimated and Conservative Analyses for a Feedwater Pipe Break Accident of an Integral Type Reactor

An integral type reactor, which is an innovative design to achieve a high degree of safety, is currently being developed at the Korea Atomic Energy Research Institute. A feedwater pipe break accident is one of the most important accidents regarding the safety of an integral type reactor. A best estimated calculation, a conservative calculation, and a parameter study for a feedwater pipe break have been carried out. The sensitivity analysis in this paper performed is to establish the parameters which greatly affect the feedwater pipe break accident. A power level, an initial system pressure, a moderator reactivity coefficient and a break size are the major parameters which maximize a system pressure. The important function that must operate following a feedwater pipe break accident is an opening of the pilot operated safety relief valves, and an initiation of the passive residual heat removal system. The integral reactor safety systems function properly and thus secure the reactor to a safe condition with respect to the safety parameters.

Overpressure protection analysis of an advanced integral reactor

Overpressure protection analysis of KAERI's advanced integral reactor, which has been developed to verify the performance of the System integrated Modular Advanced ReacTor (SMART), has been performed using the Transients And Setpoint Simulation/Small and Medium Reactor (TASS/SMR) code. In the analysis, the loss of feed-water and the regulating bank withdrawal events on behalf of the decrease in the heat removal by the secondary system and the reactivity and power distribution anomalies are selected as the initiating events for the analysis because the highest peak pressures of the primary system occur during these events. Conservative assumptions and the various initial/boundary conditions have been applied to the overpressure protection analysis for the advanced integral reactor. Although the pressurization of the primary system occurs due to an unbalance between the power generation in the core and the heat removal through the steam generator, the peak pressures in the cases of using the loss of feed-water and the regulating bank withdrawal event as an initiating event are well below the acceptance criteria of 18.7 MPa, due to the reactor protection system and three pilot operated safety relief valves installed in the advanced integral reactor.

RETRAN-3D analysis of the base case and the four extreme cases of the OECD/NRC Peach Bottom 2 Turbine Trip benchmark

This paper presents the results of RETRAN-3D calculations of the base case and the four extreme cases of phase 3 of the Peach Bottom 2 OECD/NRC Turbine Trip benchmark for coupled thermal-hydraulic and neutronic codes. The PSI-RETRAN-3D model gives good agreement with the measured data of the base case. In addition to the base case, the analysis of the extreme cases provides a further understanding of the reactor behaviour, which is the result of the dynamic coupling of the whole system, i.e., the interaction between the steam line and vessel flows, the pressure, the Doppler, void and control reactivity and power. For the extreme cases without scram the bank of safety relief valves is able to mitigate the effects of the turbine trip for short times. The 3-D nature of the core power distribution has been investigated by analysing the power density of the different thermal-hydraulic channels. In all cases prior to the reactor scram the course of the power is similar in all the channels with differences of the order of a few percent showing that, by and large, the core acts in a coherent manner. At the time of maximum power, the axial power distribution in the different channels is increased at the core centre with respect to the distribution at time zero, by an amount, which is different for the different channels.

Safety of Super LWR, (I) Safety System Design

This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis.

Fire Versus Non-Fire Contingencies: A Study of Pressure-Relief Device Sizing Risks

There are tens of thousands of industrial manufacturing facilities operating throughout the world. Each chemical plant, petroleum refinery, pharmaceutical plant and other manufacturing facility has equipment and piping systems that operate under pressure. In the event of excessive overpressure, equipment or piping failures could result in economic loss to business, environmental contamination, and health and safety risks. To reduce such risks, equipment and piping systems that operate under pressure must be protected from excessive overpressure. This is accomplished with the installation of pressure-relief devices, which must be properly sized and specified for the intended service conditions. More specifically, overpressure protection is provided by pressure-relief devices that are sized, selected, specified and installed for the postulated governing overpressure contingency. To adequately size a pressure-relief device to provide overpressure protection for equipment and piping, several relief event scenarios always should be considered. In the U.S.A., federal and state regulations require operating industrial facilities to have risk management programs in place that include the design basis for safety-relief systems installed to protect pressurized equipment from overpressure. For new installations, the pressure-relief system design philosophy should be established during the project design phase. However, for process facilities that have been in operation for many years, the original design basis and calculations for the safety-relief devices often are no longer available. For existing pressure-relieving installations, fitness-for-service assessments should include verification of the relief device size and specification, and review and substantiation of required documentation. This paper presents results from a study intended to examine which overpressure relief contingency, if any, most often governs the size of relief devices that are used to protect equipment and piping systems. The required elements of a pressure-relieving system sizing and documentation program are described. The author emphasizes seven relief contingencies to be considered when sizing pressure-relief devices. Some restrictions and limitations of the codes and standards that are applied for design guidance of pressure-relieving systems are challenged. For this study, relief device sizing data was compiled from a number of chemical and petrochemical project applications to provide a reasonable sample of contingencies that governed the sizes of existing and new safety-relief valves and rupture disks. The study results show that a significant number of pressure-relief devices presently installed in the U.S.A. likely are undersized. This further suggests that, worldwide, an alarming number of pressure-relief devices may be undersized.

Two-phase flow through pressure safety valves. Experimental investigation and model prediction

Due to the lack of experimental data regarding two-phase flow through safety relief valves, a number of different calculation methods are presently available in the literature for their sizing. All these models mainly refer to the flow through an ideal nozzle, so that, in order to match the measured values, a discharge coefficient is to be introduced in the calculation, the coefficient thus depending on the model adopted. Furthermore, most of the available data are referred to a few operating conditions. As a result, none of the available models is presently considered sufficiently accurate to be used in a wide range of operating conditions. In the present paper, new data are produced on a steam/water flashing system through a real valve with different values of the main operating parameters (vapour quality, inlet pressure, mass flow rate, and backpressure). The measurements are compared with the predictions of a commonly used homogeneous equilibrium model (HEM), the so-called ω method, and show that the model markedly underestimates the mass flow rate in the whole range of conditions investigated. This unexpectedly implies the introduction of a two-phase discharge coefficient much higher than the vapour coefficient. Finally, a new correlation for the discharge coefficient as a function of the main operating parameters is proposed.

안전방출밸브 개발과 용량인증 사례

The purpose of this study is localization of safety relief valves for Nuclear Service. The safety relief valve is the important equipment used to protect the pressure vessel, the steam generator and the other pressure facility from overpressure by discharging the operating medium when the pressure of system is reaching the design pressure of the system. We developed design technology used FEM & CFM about safety relief valve for Nuclear Service according to ASME(or KEPIC) Code and KHNP's Technical Specification. To prove validity of a design technology, actually, we manufactured and inspected and tested the sample products designed according to a developed technology. The capacity qualification test was achieved according to requirement of ASME(or KEPIC) Code by NEBI and the functional qualification test was achieved according to ASME QME-1 for operating condition in technical specification of KHNP by NLI. Therefore we have to achieve the development of safety relief valves for Nuclear Service with our own technologies.

Safety of Super LWR, (I) Safety System Design

This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis.

A theory on the discharge coefficient for safety relief valve

Abstract Comparing the flow through an SRV to that through an orifice, a simple theory is proposed to relate the discharge coefficient in two-phase compressible flow to the value in liquid incompressible flow. This approach made use of the author’s omega method applicable for both flashing discharge and non-flashing discharge. The two-phase discharge coefficient is shown to be a smooth function of the omega parameter. For non-flashing two-phase discharge, the discharge coefficient is shown to lie between the liquid coefficient and the gas coefficient. For flashing two-phase discharge, the discharge coefficient is higher than the gas coefficient. Comparison with recently published SRV data seems to support the present recommendation.

On two-phase frozen and flashing flows in safety relief valuesRecommended calculation method and the proper use of the discharge coefficient

Abstract Many relief scenarios involve the discharge of a two-phase fluid mixture, and the proper method for sizing the relief valve for these conditions is the subject of considerable discussion. Sizing a valve is based on the flow through an isentropic nozzle, the pressure–density relation for the fluid properties, and a discharge coefficient ( K d ) to match the calculated mass flux to that measured for the flow of air or water in the actual valve. For single-phase flow, this is straightforward, since the fluid properties are simple and measured values of K d are available. For two-phase flow, the density–pressure relation is complex and no values of K d are available, so a variety of “models” have been proposed in the literature to address this problem. Since the various models produce various results, the appropriate value of K d required to match the model to the actual valve will depend on the model. This paper utilizes a simple, rigorous method for sizing relief valves for two-phase flow that utilizes the fluid properties directly and hence does not require a “model” for these properties. It is shown how this method can be applied to two-phase frozen or flashing (equilibrium or non-equilibrium) nozzle flows, and how the available values for K d for single-phase flow can be used directly with this method, depending on the critical state of flow in the nozzle, to accurately predict two-phase flow in any valve. The calculations are compared with data from the literature for frozen air/water and flashing steam/water flows in actual safety relief valves.

Updated safety relief valve

Stainless steel construction materials, both inside and out, give the new 860 Safety Relief Valve form Spence Engineering enhanced corrosion resistance across a variety of different industrial applications, including liquids handling, steam applications, and air/vacuum pumping. Visit www.worldpumps.com for the latest pump industry news

Valve meets EU safety directive

PED certification has been granted to the Figure 710 Safety Relief Valve from Spence Engineering Company Inc. The Pressure Equipment Directive (PED 92/23/EC), introduced last year by the EU, is setting new standards for product quality and safety across the entire range of pressure equipment. It is already becoming an important standard for valve manufacturers, pump makers and others involved in the industry to meet. This is a short news story only. Visit www.worldpumps.com for the latest pump industry news

Analysis and sensitivity studies of postulated SB-LOCAs in the Mühleberg (KKM) BWR/4 by TRAC-BF1

In this work, we shall report on the analysis of a number of postulated small breaks ranging from 1 to 10% of the recirculation line flow area at the Mühleberg (KKM) BWR/4 in Switzerland by using the TRAC-BF1 code. The analysis was performed by assuming both limited as well as full (nominal) emergency core cooling systems (ECCS) availability, and also limited availability of some of the safety relief valves used in the automatic depressurization system. Through these assumptions, we consider the response of the plant to multiple failures and therefore we extend our analysis beyond the normal “design basis”. Furthermore, in order to provide a measure of the “uncertainty” of the results, a sensitivity study was performed by varying some plant parameters as well as physical models in the code. To the best of our knowledge, this is the first time that such a systematic analysis of a wide spectrum of postulated small breaks in a commercial BWR has been pursued and the results of this work show that even under the most pessimistic assumptions on the availability of plant safety systems as well as of plant parameters and physical models in the code, the peak clad temperatures never exceed 1000 K.

Sizing of Safety Relief Valves for Two-Phase Flow

Sizing of Safety Relief Valves for Two-Phase Flow