Off-normal Conditions Research Articles

EFFECT OF SURFACE ROLLERS ON THE FORMATION OF CRESCENTIC BARS

The formation of crescentic bars is examined using a morphodynamic model based in linear stability analysis. The effect of surface rollers for off-normal wave conditions is examined. The effect of the rollers is to increase the efolding times with increasing the angle of incidence. For angles large enough the formation of crescentic bars is even inhibit. The main effect of the rollers it be through hydrodynamics. The longitudinal changes in current produced by the rollers cause the maximum of sediment concentration to be shifted towards the coast with the final effect of prevent the formation of crecscentic bars.

General and Crevice Corrosion Study of the Materials for ITER Vacuum Vessel Inwall Shield

Vacuum vessel inwall shield (IWS) (VV-IWS) which consists of block assemblies will be inserted between the inner and outer shells of the ITER vacuum vessel. The vacuum vessel will be operating at the following: 1) 100 °C and 2) 200 °C. The temperature of the vacuum vessel will be maintained by flowing water between the shells of the vessel. To achieve the temperature of the vessel, the water will be flowed at two operation phases: 1) normal operation condition (OC1), i.e., 100 °C temperature and 1.1-MPa pressure, and 2) off-normal condition, i.e., 200 °C temperature and 2.4-MPa pressure. The water flow rate is locally rather low in the VV-IWS block assemblies, and low flow rates generally promote local attacks on the stainless steel (SS). In addition, local accumulation of the aggressive compounds like chloride and other anions might be higher for ITER condition due to the different operational modes. IWS blocks are not accessible until the life of the machine after closing of the vacuum vessel. Hence, it is necessary to study the susceptibility of IWS materials to general corrosion and crevice corrosion under operations of the ITER vacuum vessel. Corrosion properties of IWS materials were studied by using the following: 1) immersion technique and 2) electrochemical polarization techniques at OC1, baking condition (OC2), and draining and drying operations of the ITER vacuum vessel. Before and after the immersion test, all the sample materials were subjected to a series of examinations, like loss/gain weight measurement, scanning electron microscopy (SEM) analysis, optical stereomicroscopy, and measurement of surface profile and hardness of materials. After the immersion test, SS 304B4 and SS 304B7 showed slight weight gains of 0.0593 and 0.0286 g, respectively, which indicate oxide layer formation on the surface of coupons. The SS 430 material showed a negligible weight loss of 0.0160 g which is equivalent to a corrosion rate of 0.00267 mm/year, indicating mild general corrosion effect. On visual observation with SEM and metallography, all materials showed pitting corrosion attack. During the electrochemical polarization test, all sample materials were subjected to a series of measurements, like open-circuit potential, cyclic polarization, pitting potential, protection potential, critical anodic current, and SEM examination. Under the OC1 condition, the ICC value (critical current density) is found lowest for SS 304B7 and highest for SS 430, and under the OC2 condition, the ICC value is found lowest for XM-19 and highest for SS 304B7 which indicates that SS 430 and SS 304B7 showed the highest general corrosion rates under the OC1 and OC2 conditions, respectively. All materials show a pitting loop in the OC2 operating condition. However, its absence in the OC1 operating condition clearly indicates the activity of chloride ion to penetrate the oxide layer on the sample surface, at higher temperature. The critical pitting temperature of all samples remains between 100 ° and 200 °C.

Evaluation of energy and particle impact on the plasma facing components in DEMO

We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacement events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Under steady-state operation heat transfer into the coolant must remain below the critical heat flux (CHF) to avoid the possible severe degradation of the coolant heat removal capability. From the plasma side it is particularly demanding to keep the bulk plasma contamination during the reactor long operational discharges below the fatal level. The possible damage of the FW materials due to the plasma sputtering erosion is estimated. The minimum thickness of the tungsten amour about 3mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions.

Autonomous monitoring of control hardware to predict off-normal conditions using NIF automatic alignment systems

The National Ignition Facility (NIF) is a high power laser system capable of supporting high-energy-density experimentation as a user facility for the next 30 years. In order to maximize the facility availability, preventive maintenance enhancements are being introduced into the system. An example of such an enhancement is a camera-based health monitoring system, integrated into the automated alignment system, which provides an opportunity to monitor trends in measurements such as average beam intensity, size of the beam, and pixel saturation. The monitoring system will generate alerts based on observed trends in measurements to allow scheduled pro-active maintenance before routine off-normal detection stops system operations requiring unscheduled intervention.

Proof of concept experiments of the multi-isotope process monitor: An online, nondestructive, near real-time monitor for spent nuclear fuel reprocessing facilities

Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the multi-isotope process (MIP) monitor, a novel approach to monitoring and safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial least squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of ±1.3M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

Assessment of coated-particle failure with the IMGA device

The coating layers of modern TRISO (tri-structural isotropic) high-temperature reactor fuel provide a barrier against the release of fission products both in normal operation and off-normal conditions. Experimental simulations of D-LOFC (depressurization and loss of forced circulation) accidents up to 1600°C have demonstrated a very low failure fraction of coated particles (CPs). In most instances, not a single particle out of the up to 15,000 CPs contained in a fuel pebble suffered mechanical failure, and all observed fission product release was due to diffusion through intact coating layers. When fuel is irradiated at temperatures well above 1000°C or when accident tests are pushed to more extreme temperatures, single CPs may fail. To selectively examine the defective particles, fuel elements must be deconsolidated and the large number of CPs contained in them analysed in an automated manner. For this purpose, an irradiated-microsphere gamma analyser (IMGA) has been developed at JRC-ITU. In this device, particles are first singularized and then assessed by gamma spectroscopy. The activity ratio of selected fission product pairs with different mobility can be used to assess whether or not a particle has suffered mechanical failure and to separate defective from intact CPs.

Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS.A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations.The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls’ temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios.Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions.To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the realizable k-ε model with two-layer all y+ wall treatment performs better than the other k-ε and k-ω turbulence models when compared to the experimental results and the Reynolds stress transport turbulence model results.A scaling analysis was developed to address the distortions introduced by the CFD model in simulating the physical phenomena inside the RCCS system with respect to the full plant configuration. The scaling analysis demonstrated that both the experimental facility and the CFD model achieve a satisfactory resemblance of the main flow characteristics inside the RCCS cavity region, and convection and radiation heat exchange phenomena are properly scaled from the actual plant.

Thermomechanical Analysis of the Revised U.S. ITER DCLL Test Blanket Module

The US Fusion Nuclear Science and Technology program selected the Dual Coolant Lead Lithium (DCLL) concept as the primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is the structural material, helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and tritium extraction.The DCLL TBM has undergone major design changes since 2005. We present here the most recent thermo-mechanical analysis of the newly revised DCLL TBM. The analysis described here is aiming to verify the thermo-mechanical response of the DCLL TBM under relevant normal operating conditions as well as during a loss of coolant accident (LOCA).A full 3-dimensional solid model of the entire DCLL TBM structure was developed, which included FW, top and bottom lids, internal supporting ribs, manifolds, plena, and flexible frame-attachment supports. A coupled thermo-mechanical analysis was performed for both normal- and off-normal operating conditions. Thermal loads included surface heat load, volumetric heating, as well as detailed position- and location dependent heat transfer along all coolant channels. Structural loads incorporated helium coolant pressure loads, self-weight, as well as the weight of the PbLi. Maximum structure temperatures of nearly 560 °C along with a maximum resultant net displacement of more than 10 mm were mapped for normal operating conditions and a number of stress concentration locations were identified. The ITER SDC-IC-1300 criteria were applied to the LOCA analysis results. It is shown that the DCLL TBM exhibits admissible behavior regarding the ITER Design Criteria and that the most recent design modifications did not compromise the structural integrity.

The use of a social network analysis technique to investigate the characteristics of crew communications in nuclear power plants—A feasibility study

Effective and reliable communications are very important in securing the safety of human-involved large process control systems because human operators have to accomplish their tasks in cooperative ways. This means that it is very important to understand the characteristics of crew communications, which can provide useful insights for preventing inappropriate communications. Unfortunately, in the nuclear industry, a systematic framework that can be used to identify the characteristics of crew communications seems to be rare. For this reason, the applicability of the social network analysis (SNA) technique to identifying the characteristics of crew communications was investigated in this study. To this end, the communication data of operating crews working in the main control room (MCR) of nuclear power plants (NPPs) were collected under two kinds of simulated off-normal conditions. Then the communication characteristics of MCR operating crews, which can be represented by the associated SNA metrics, were compared with communication characteristics that are already known from existing studies. As a result, it was found that SNA metrics could be meaningful for explaining the communication characteristics of MCR operating crews. Accordingly, it is expected that the SNA technique can be used as one of the serviceable tools to investigate the characteristics of crew communications in NPPs.

Loss of vacuum accident (LOVA): Comparison of computational fluid dynamics (CFD) flow velocities against experimental data for the model validation

A recognized safety issue for future fusion reactors fueled with deuterium and tritium is the generation of sizeable quantities of dust. Several mechanisms resulting from material response to plasma bombardment in normal and off-normal conditions are responsible for generating dust of micron and sub-micron length scales inside the VV (Vacuum Vessel) of experimental fusion facilities. The loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA) are types of accidents, expected in experimental fusion reactors like ITER, that may jeopardize components and plasma vessel integrity and cause dust mobilization risky for workers and public. The air velocity is the driven parameter for dust resuspension and its characterization, in the very first phase of the accidents, is critical for the dust release. To study the air velocity trend a small facility, Small Tank for Aerosol Removal and Dust (STARDUST), was set up at the University of Rome “Tor Vergata”, in collaboration with ENEA Frascati laboratories. It simulates a low pressurization rate (300 Pa/s) LOVA event in ITER due to a small air inlet from two different positions of the leak: at the equatorial port level and at the divertor port level. The velocity magnitude in STARDUST was investigated in order to map the velocity field by means of a punctual capacitive transducer placed inside STARDUST without obstacles. FLUENT was used to simulate the flow behavior for the same LOVA scenarios used during the experimental tests. The results of these simulations were compared against the experimental data for CFD code validation. For validation purposes, the CFD simulation data were extracted at the same locations as the experimental data were collected for the first four seconds, because at the beginning of the experiments the maximum velocity values (that could cause the almost complete dust mobilization) have been measured. In this paper the authors present and discuss the computer-simulation data and compare them with data collected in STARDUST.

Design finalization and start of construction of ITER vacuum vessel

The vacuum vessel (VV) design is being finalized including interface components, such as the support rails and feedthroughs of coils for mitigation of edge localized modes (ELM) and vertical stabilization (VS) of the plasma (ELM/VS coils). It was necessary to make adjustments in the locations of the blanket supports and manifolds to accommodate the design modifications in the ELM/VS coils. The lower port gussets were reinforced to keep a sufficient margin under the increased VV load conditions. The VV support design is being finalized as well, with an emphasis on structure simplification. The design of the in-wall shielding (IWS) has progressed, considering assembly and required tolerances. The layout of ferritic steel plates and borated steel plates will be optimized based on on-going toroidal field ripple analysis. The VV instrumentation was defined in detail. Strain gauges, thermocouples, displacement meters and accelerometers shall be installed to monitor the status of the VV in normal and off-normal conditions to confirm all safety functions are performed correctly. The ITER VV design was preliminarily approved, and the VV materials including 316L(N) IG were already qualified by the Agreed Notified Body (ANB) according to the procedure of Nuclear Pressure Equipment Order.

Heat transfer performance tests using a half scale horizontal storage module

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A horizontal storage module to safely store the spent nuclear fuel should be able to adequately emit the decay heat from the spent nuclear fuel. The horizontal storage module must ensure that the temperatures of the spent nuclear fuel assemblies are maintained within the allowable values for normal, off-normal, and accident conditions. Therefore, the horizontal storage module must be designed including heat removal capabilities with an appropriate reliability. However, the thermal conductivity of concrete is not good and the allowable temperature of concrete is lower than that of steel. Therefore, heat transfer performance tests were performed to evaluate the heat transfer performance in accordance with the ratio of the outlet to inlet of the air as well as the direction of the inlet and outlet of the air in the horizontal storage module. The influence that the direction of the inlet and outlet of the air reaches to the heat transfer performance was estimated to be minimal. The overall heat removal efficiencies were reduced as the mass flow rate at the outlet was reduced.

Proof of concept simulations of the Multi-Isotope Process monitor: An online, nondestructive, near-real-time safeguards monitor for nuclear fuel reprocessing facilities

The International Atomic Energy Agency will require the development of advanced technologies to effectively safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of nondestructive, near-real-time, autonomous process monitoring. This paper describes recent results from model simulations designed to test the Multi-Isotope Process (MIP) monitor, a novel addition to a safeguards system for reprocessing facilities. The MIP monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in near-real-time. Three computer models including ORIGEN-ARP, AMUSE, and SYNTH were used in series to predict spent nuclear fuel composition, estimate element partitioning during separation, and simulate spectra from product and raffinate streams using a variety of gamma detectors, respectively. Simulations were generated for fuel with various irradiation histories and under a variety of plant operating conditions. Principal component analysis was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup, and cooling time. Hierarchical cluster analysis and partial least squares (PLS) were also used in the analysis. The MIP monitor was found to be sensitive to induced variations of several operating parameters including distinguishing ±2.5% variation from normal process acid concentrations. The ability of PLS to predict burnup levels from simulated spectra was also demonstrated to be within 3.5% of measured values.

Subcooled critical heat flux: an assessment of the risk to front-end and beamline components of synchrotron light sources

X-ray absorbers in the front ends and beamlines of synchrotron light sources are exposed to very high thermal loads. Many facilities, such as the Advanced Photon Source, are investigating upgrades that will further increase the thermal load. The likelihood of exceeding the limit of subcooled critical heat flux (CHF) in these components was examined. The assessment was performed for both currently possible off-normal operational conditions, such as might occur in the event of a failure of multiple safety interlocks, and the anticipated operating conditions that may result from future upgrades. The subcooled CHF values were calculated using empirical equations frequently cited in the literature and then compared with the computed values of the heat flux at the walls of the component cooling channels in cases where the cooling wall temperature exceeded the water saturation temperature at local hydraulic conditions. Having in mind that the great majority of the available empirical correlations were developed for the conditions characteristic for the operation of heat exchangers in the nuclear power industry, the limitations of this approach are discussed and an experimental study of the subcooled CHF values in the conditions similar to those expected in the front-end and beamline components is proposed.

Topographic Control of Spinal Deformity Correction in Patients with Unilateral Shortening of the Lower Limb

Objective. To study topographical features of the trunk posture and spinal deformities in patients with unilateral shortening of the lower limb. Material and Methods. Seventy-eight patients at the age of 5 to 22 years with unilateral shortening of the lower limb were examined: Group 1 included 59 patients with anatomical shortening of the lower limb of 2.0 to 17.0 cm; Group 2 included 19 patients with functional shortening of the lower limb of 1.0 to 1.5 cm. Efficacy of the spinal correction was assessed by computer optical topography. Results. Posture disorders and spine deformities observed in lower limb length discrepancy are mainly the manifestations of the locomotor system adaptation to biomechanically off-normal conditions for function. Adaptive responses of the locomotor system such as posture disorders, postural asymmetries, and spine deformities were preserved in 39 % of cases. Depending on the side of shortening and the amount of limb length discrepancy the spinal adaptive deformities have different rates of resistance to the performed corrective manipulations: conservative therapy for deformity, surgical elimination of lower limb length discrepancy, and compensation for a short limb with supports. Conclusion. Adaptive deformities of the spine in patients with small discrepancy between the length of lower limbs turned out to be more resistant to correction, than those in patients with large anatomical shortening of the lower limb.

Thermo-physical properties of DU–10 wt.% Mo alloys

Low-enriched uranium alloyed with 10 wt.% molybdenum is under consideration by the Global Threat Reduction Initiative reactor convert program as a very high density fuel to enable the conversion of high-performance research reactors away from highly-enriched uranium fuels. As with any fuel development program, the thermo-physical properties of the fuel as a function of temperature are extremely important and must be well characterized in order to effectively model and predict fuel behavior under normal and off-normal irradiation conditions. For the alloy system under investigation, the available thermo-physical property data is relatively inconsistent and often lacks appropriate explanation. Available literature on this alloy system comes mainly from studies done during the 1960s and 1970s, and often does not include sufficient information on fabrication history or conditions to draw conclusions for the current application. The current paper has investigated specific heat capacity, coefficient of linear thermal expansion, density, and thermal diffusivity that were then used to calculate alloy thermal conductivity as a function of temperature. The data obtained from this investigation was compared to available literature on similar U–Mo alloys, and in most cases are in good agreement.

Numerical prediction of the fission product plate-out for a VHTR application

In order to identify and quantify the key parameters affecting the predictions of fission product transport and plate-out behavior in the coolant circuits of a very high temperature reactor (VHTR) system, systematic and intensive analyses were performed based on numerical predictions as well as the existing experimental data. For the purpose, the computational module for modeling fission product transport phenomena was developed and incorporated into the system analysis code, GAMMA+ for an integrated analysis. This integration can provide more realistic boundary conditions such as velocity, temperature, etc., during off-normal conditions as well as normal operations in a given VHTR system. The developed module has been applied to predict the fission product transport and plate-out in the experimental loops of VAMPYR-I and OGL-1. A good agreement is found in general between the predictions by the present module and the measured values. The results of the sensitivity analysis have shown that the surface conditions of the coolant circuit can be major source of the uncertainty in the prediction of the plate-out activity of I-131. In the high temperature regions, in particular, it is found that the predicted plate-out activities of I-131 on the oxidized surface result in about 100 times smaller values than those on the unoxidized surface. On the other hand, Cs-137 is found to be less affected by the surface oxidation. It is also confirmed that the diffusion coefficient significantly affects the fission product distribution in the coolant and plate-out activities over the large surface area like heat exchangers.

HORIZON EXPANSION OF THERMAL-HYDRAULIC ACTIVITIES INTO HTGR SAFETY ANALYSIS INCLUDING GAS-TURBINE CYCLE AND HYDROGEN PLANT

We present three nuclear/hydrogen-related R&D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA, with which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed a tool to predict the characteristics of HTGR helium turbines based on the throughflow calculation with a Newton-Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The dynamic equations for the distillation column of HI process are described with 4 material components involved in the HI process: H2O, HI, I2, H2. For the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. The improved Neumann model predicted a total pressure with 8.6% maximum relative deviation from the data and 2.5% mean relative deviation, and liquid-liquid-separation with 9.52% maximum relative deviation from the data.

Investigation of High-Temperature Printed Circuit Heat Exchangers for Very High Temperature Reactors

Very high-temperature reactors require high-temperature (900–950°C) and high-integrity heat exchangers with high effectiveness during normal and off-normal conditions. A class of compact heat exchangers, namely, the printed circuit heat exchangers (PCHEs), made of high-temperature materials and found to have these above characteristics, are being increasingly pursued for heavy duty applications. A high-temperature helium test facility, primarily aimed at investigating the heat transfer and pressure drop characteristics of the PCHEs, was designed and is being built at Ohio State University. The test facility was designed to facilitate operation at temperatures and pressures up to 900°C and 3 MPa, respectively. Owing to the high operating conditions, a detailed investigation on various high-temperature materials was carried out to aid in the design of the test facility and the heat exchangers. The study showed that alloys 617 and 230 are the leading candidate materials for high-temperature heat exchangers. Two PCHEs, each having 10 hot plates and 10 cold plates, with 12 channels in each plate, were fabricated from alloy 617 plates and will be tested once the test facility is constructed. Simultaneously, computational fluid dynamics calculations have been performed on a simplified PCHE model, and the results for three flow rate cases of 15, 40, and 80 kg/h at a system pressure of 3 MPa are discussed. In summary, this paper focuses on the study of the high-temperature materials, the design of the helium test facility, the design and fabrication of the PCHEs, and the computational modeling of a simplified PCHE model.

On feasibility of regional frequency-based emergency control plans

Decentralized and regional load-frequency control of power systems operating in normal and near-normal conditions has been well studied; and several analysis/synthesis approaches have been developed during the last few decades. However in contingency and off-normal conditions, the existing emergency control plans, such as under-frequency load shedding, are usually applied in a centralized structure using a different analysis model. This paper discusses the feasibility of using frequency-based emergency control schemes based on tie-line measurements and local information available within a control area. The conventional load-frequency control model is generalized by considering the dynamics of emergency control/protection schemes and an analytic approach to analyze the regional frequency response under normal and emergency conditions is presented.