Ideco Ics 2.5 6 Evaluation Edition
.Part of thebook series (LNCS, volume 11098) AbstractIndustrial control systems (ICSs), and particularly supervisory control and data acquisition (SCADA) systems, are used in many critical infrastructures and are inherently insecure, making them desirable targets for attackers. ICS networks differ from typical enterprise networks in their characteristics and goals; therefore, security assessment methods that are common in enterprise networks (e.g., penetration testing) cannot be directly applied in ICSs. Thus, security experts recommend using an isolated environment that mimics the real one for assessing the security of ICSs. While the use of such environments solves the main challenge in ICS security analysis, it poses another one: the trade-off between budget and fidelity. In this paper we suggest a method for creating a digital twin that is network-specific, cost-efficient, highly reliable, and security test-oriented. The proposed method consists of two modules: a problem builder that takes facts about the system under test and converts them into a rules set that reflects the system’s topology and digital twin implementation constraints; and a solver that takes these inputs and uses 0–1 non-linear programming to find an optimal solution (i.e., a digital twin specification), which satisfies all of the constraints.
We demonstrate the application of our method on a simple use case of a simplified ICS network.
AbstractThe SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage. SCORE-EVET was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates.
No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code contains a one-dimensional steady state solution scheme to initialize the flow field, steady state and transient fuel rod conduction models, and comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions, such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage. The basic volume-averaged transient three-dimensional equations for flow in porous media are solved in their general form with constitutive relationships and boundary conditions tailored to define the porous medium as a matrix of fuel rods. By retaining generality in the form of the conservation equations, a wide range of fluid flow problem configurations, from computational regions representing a single fuel rod subchannel to multichannels, or even regions without a fuel rod, can be modeled without restrictive assumptions. The completeness of the conservation equations has allowed SCORE-EVET to be used, with modification to the constitutive relationships, to calculate three-dimensional laminar boundary layer development, flow fields in large bodies of water, and, with the addition of a turbulence model, turbulent flow in pipe expansions and tees. SCORE-EVET was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays.
The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code contains a one-dimensional steady state solution scheme to initialize the flow field, steady state and transient fuel rod conduction models, and comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions, such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage. The basic volume-averaged transient three-dimensional equations for flow in porous media are solved in their general form with constitutive relationships and boundary conditions tailored to define the porous medium as a matrix of fuel rods. By retaining generality in the form of the conservation equations, a wide range of fluid flow problem configurations, from computational regions representing a single fuel rod subchannel to multichannels, or even regions without a fuel rod, can be modeled without restrictive assumptions.
The completeness of the conservation equations has allowed SCORE-EVET to be used, with modification to the constitutive relationships, to calculate three-dimensional laminar boundary layer development, flow fields in large bodies of water, and, with the addition of a turbulence model, turbulent flow in pipe expansions and tees.CDC7600,CYBER175; FORTRAN IV; SCOPE 2.1 (CDC7600), NOS 1.3 (CDC CYBER175); 200,000 (octal) words of memory are required for execution. Results from the previously conducted Semiscale Mod-1 ECC injection test series were analyzed.
Testing in the LOFT counterpart test series was essentially completed, and the steam generator tube rupture test series was begun. Two tests in the alternate ECC injection test series were conducted which included injection of emergency core coolant into the upper plenum through use of the low pressure injection system.
The Loss-of-Fluid Test Program successfully completed nonnuclear Loss-of-Coolant Experiment L1-4. A nuclear test, GC 2-3, in the Power Burst Facility Reactor was performed to evaluate the power oscillation method of determining gap conductance and to determine the effects of initial gap size, fill gas composition, and fuel density on the thermal performance of a light water reactor fuel rod. Additional test results were obtained relative to the behavior of irradiated fuel rods during a fast power increase and during a high power film boiling transient. Fuel model development and verification activities continued for the steady state and transient Fuel Rod Analysis Program, FRAP-S and FRAP-T. A computer code known as RELAP4/MOD7 is being developed to provide best-estimate modeling for reflood during a postulated loss-of-coolant accident (LOCA). A prediction of the fourth test in the boiling water reactor (BWR) Blowdown/Emergency Core Cooling Program was completed and an uncertainty analysis was completed of experimental steady state stable film boiling data for water flowing vertically upward in round tubes.
A new multinational cooperative program to study the behavior of entrained liquid in the upper plenum and cross flow in the core during the reflood phase of a pressurized water reactor LOCA was defined. Dispersed-flow film-boiling is an important phenomenon in a postulated nuclear reactor loss-of-coolant accident (LOCA). Several experiments have been performed to investigate the location of critical heat flux (CHF) and the resulting post-CHF heat transfer. The experimental data provide necessary information to evaluate the accuracy of the predictions of a computer code. For LOCA analysis, the central issue, of course, is the prediction of fuel rod cladding temperature. The WCOBRA/TRAC code, developed from COBRA/TRAC for analysis of nuclear reactor vessel and primary loop two-phase transients, is being used as a best-estimate model for LOCA licensing calculations at Westinghouse Electric Corporation.
The code employs a two-fluid three-field (liquid, vapor, and droplet) model to represent a two-phase flow, which is necessary in simulating dispersed-flow film-boiling phenomena. Neogeo roms. Each field is treated as three-dimensional and compressible.
Ideco Ics 2.5 6 Evaluation Edition Review
The conservation equations for each field and the heat transfer from and within the solid structures in contact with the fluid are solved using a semiimplicit finite difference numerical scheme in Eulerian mesh structure. The noding scheme is very flexible and can be used to model complicated geometry. Flow blockage and spacer grid heat transfer models for rod bundle arrays have been developed for a two-phase flow situation characteristic of a PWR reflood. These models have been incorporated into COBRA-TF, which is a three-dimensional, three-field, two-fluid mechanistic two-phase flow subchannel computer code. Comparison of the predicted flow blockage heat transfer in large rod bundle arrays with test data indicates that the blockage and grid heat transfer models used with the COBRA-TF code agree quite well with the measured data.
Bias plots of the predicted and measured temperature rises from different tests indicate that, in general, the computer code calculatons tend to underpredict the heat transfer improvement observed to have been caused by grids and blockage in the experiments. The principal reason for heat transfer improvement due to blockage and grids is the breakup of the entrained liquid droplets in the superheated steam flow above the quench front. The breakup of these entrained drops results in a population of much smaller drops, which are more easily evaporated in the superheated vapor.
The enhanced heat transfer observed in and downstream of blockages and grids is also attribute to increased turbulence caused by the droplets in the steam flow. The resulting computer models and methods of modeling both grids and blockages, which are described in this report, are believed to be applicable to PWR safety analysis. Application of such models is expected to significantly reduce or eliminate the calculted peak clad temperature penalty due to flow blockage for a hypothetical PWR LOCA, using the Appendix K criteria.