RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP (original) (raw)
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Progress in Nuclear Energy, 2004
A RELAP5/MOD3.2 model of a VVER-1000 (V-320 model) nuclear power plant was updated, improved and validated against actual plant data. The data included steady-state and operational transient results from unit 6 of the Kozloduy nuclear power plant. The operational transient was initiated by a loss of flow at partial power conditions caused by the trip of a main coolant pump without reactor scram.
PSB-VVER simulation of Kozloduy NPP “loss of feed water transient”
Nuclear Engineering and Design, 2005
This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions. RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient.
International Journal on Advanced Science, Engineering and Information Technology, 2020
One of Floating Nuclear Power Plant (FNPP) designs in the world is currently being built by Russian Federation, named "Academic Lomonosov," which uses two PWR types, KLT-40S as its power unit. However, too little information regarding its detailed technical specification is available, including its thermal-hydraulics parameters. The objective of this research is to create a thermal-hydraulic model of KLT-40S reactor core use RELAP5-3D and to predict fuel and cladding temperature value at the steadystate condition, and transient condition with a variety of primary coolant mass flow rate and pressure to simulate abnormal event within the reactor. The reactor thermal-hydraulic model is created by dividing 121 coolant channels in the actual nuclear fuel assemblies into two channels: one channel to simulate coolant flow in 120 fuel assemblies with average heat generation, and the other channel to simulate coolant flow in one fuel assembly with highest heat generation in the core. The fuel structure had solid cylinder geometry and made from ceramic-metal UO2 dispersed in the inert silumin matrix. The fuel cladding is made from zirconium alloy. These fuel heat structures generate heat from fission reaction and are modelled as a heat source according to the reactor power technical data, i.e., 150 MWt. The reactor axial power distribution is approximated by cosine distribution. Operation parameter variation that represents the real reactor normal operation condition in this research is a variation that has flow loss coefficient value 8,000, radial power peaking factor 1.1, and axial power peaking factor 1.1 with axial power peaking located in the middle of the fuel rod. The fuel and cladding temperature value at the steady-state condition and several transient conditions are predicted in this research, and there is no temperature value that goes beyond the safety limit.
Progress in Nuclear Energy, 2006
As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 coolant transient benchmark phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PARCS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant, Unit 6, and the neutron kinetics models of PARCS for hexagonal geometries. Selected results of performed investigations will be presented and discussed in this paper. The overall trends of most plant parameters are in a reasonable agreement with the experimental data. Nevertheless multidimensional thermal hydraulic models are needed for a more realistic description of the coolant mixing phenomena within the reactor pressure vessel.
Scaling of the small-scale thermal-hydraulic transient to the real nuclear power plant
This paper provides a scaling methodology that was applied for scaling of the BETHSY integral test facility to the real Nuclear Power Plant (NPP). The similarity of physical phenomena between the BETHSY experimental facility and the scaled up model (representation of the real NPP) was analyzed on the Small Break Loss of Coolant Accident (SBLOCA) scenario. A comprehensive numerical analysis using the RELAP5 thermalhydraulic code was performed to evaluate the optimal scaling-up of the BETHSY facility to the real NPP. In order to investigate the phenomenological scaling-up basis, two enlarged RELAP5 input models were constructed, differing in scaling criteria for the primary cooling system: proportional volume scaling and scaling based on the Froude number. A better agreement with the physical phenomena of the SBLOCA experiment was achieved in the case of proportional volume scaling. In addition, scaling of heat structures was also analyzed. It was shown that the best predictions of the transient phenomena were obtained when the heat structures were scaled according to the tensile stress criterion. Evaluation of the prediction capability of large thermal-hydraulic codes such as RELAP5 1 and of the safety margins of light water reactors are among the objectives of some international research programs. The execution of experiments in integral test facilities 2 simulating the behaviour of NPP plays an important role in this connection both considering the system code assessment and the possibility to identify and characterize phenomena relevant during off-normal conditions. During the past years, a number of separate effects and integral test facilities have been constructed. In general, the majority of experiments related to nuclear safety have been performed in reduced scale test facilities, including full height, full pressure facilities such as BETHSY 3 or scaled based on available space and power such as PUMA Simplified Boiling Water Reactor (SBWR) experiment at Purdue University 4 . Winthin the Organisation for Economic Co-operation and Development (OECD), the International Standard Problem (ISP) program was established to increase the confidence in the validity and accuracy of computer codes used for the safety assessment of nuclear installations. In most of these ISPs, attention has been paid to the thermal-hydraulic behaviour of light water reactors during loss-of-coolant accidents and transients. The objective of this study is to evaluate the RELAP5 model of an enlarged BETHSY facility, which is scaled to dimensions of the real nuclear power plant. It may be assumed that processes involved in the transient of the experiment and of the real NPP can be satisfactory described by one-dimensional mathematical model without losing essential information 5 . Previous studies performed on this field 6,7,8 showed that the transient behaviour of the real NPP cannot be defined simply by proportional scaling of geometry of the BETHSY experimental facility to the real NPP. However, in transients where 1D physical phenomena are dominant the results of such ideal scaled-up model should be very similar to the real
Progress in Nuclear Energy, 2006
This paper provides comparisons between experimental data of ''MCP switching on when the other three MCPs are in operation'' and RELAP5 calculations with different initial levels of the reactor power 29.45% and 27.47% from the nominal. The reference power plant for this analysis is Unit 6 at the Kozloduy nuclear power plant (NPP) site. RELAP5/MOD3.2 computer code has been used to simulate the investigated transient. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. This investigation was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.
Validation of RELAP5/MOD3.2 model of VVER440 on reactor scram transient
Progress in Nuclear Energy, 2006
This paper describes validation of a computer model that has been developed for VVER 440 Nuclear Power Plant (NPP) for use with RELAP5/MOD 3.2 computer code in the analysis of the following transient: 'Reactor SCRAM'. This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with the experimental transient data received from Kozloduy NPP, Unit #2. The model of VVER 440 was developed at the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events, and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP. The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The comparisons between the RELAP5 results and the test data indicate good agreement.
Transient Thermal Hydraulic Studies of Turbine Trip For VVER-1000 Reactor
7th International and 45th National Conference on Fluid Mechanics and Fluid Power (FMFP), 2018
Turbine Trip, which is one of the Anticipated Operational Occurring event, has been analysed using Thermal Hydraulic computer code RELAP-5/MOD 3.2 for Kudankulam Nuclear Power Plant (KKNPP) having two operating VVER-1000 reactors. VVER-1000 belongs to series of Pressurized Water Reactor. RELAP uses a one-dimensional, non-equilibrium mass, energy momentum equation for liquid and vapour phase. Turbine Trip results in the decrease in heat removal through the secondary circuit. Decrease in heat removal from secondary side can lead to increase in primary coolant parameters (pressure, temperature, etc.) at full power operation. This analysis follows realistic approach and aims to find the most probable thermal hydraulic response of the NPP to this event, considering functionality of all systems. The steady state conditions for this transient are assumed based on realistic plant condition. It is found that the thermal hydraulic parameters are within acceptance limits.
A RELAP5 model for the thermal-hydraulic analysis of a typical pressurized water reactor
Thermal Science, 2010
This study de scribes a RELAP5 com puter code for ther mal-hy drau lic anal y sis of a typ i cal pres sur ized wa ter re ac tor. RELAP5 is used to cal cu late the ther mal hy draulic char ac ter is tics of the re ac tor core and the pri mary loop un der steady-state and hy po thet i cal ac ci dents con di tions. New de signs of nu clear power plants are di rected to in crease safety by many methods like re duc ing the de pend ence on ac tive parts (such as safety pumps, fans, and die sel gen er a tors) and re plac ing them with pas sive fea tures (such as grav ity draining of cool ing wa ter from tanks, and nat u ral cir cu la tion of wa ter and air). In this work, high and me dium pres sure in jec tion pumps are re placed by pas sive in jec tion com po nents. Dif fer ent break sizes in cold leg pipe are sim u lated to an a lyze to what de gree the plant is safe (with out any op er a tor ac tion) by us ing only these pas sive com po nents. Also sta tion black out ac ci dent is sim u lated and the time re sponse of op er a tor ac tion has been dis cussed.
Annals of Nuclear Energy, 2004
This paper summarizes RELAP5-3D code validation activities carried out at the Lithuanian Energy Institute, which was performed through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data, as well as with the calculation results obtained using the Russian STEPAN/KOBRA code. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters, as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors. Future activities are discussed.