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INVESTIGATION OF A STEAM LINE BREAK ACCIDENT IN RESPECT OF THE CRITICAL SAFETY FUNCTION "SUBCRITICALITY"
Abstract
The paper presents the results of an investigation of VVER-440/V230 nuclear steam supply system to the initiating event Steam Line Break. This analysis is performed with a computer code RELAP5/Mod 3.2. Through the transient, a double-ended break in a Main Steam Header is assumed. During the Steam Line Break event at the nuclear power plant, the operator can see a rapid decrease of pressure in the faulted Steam Generator and in the Main Steam Headers. The investigated accident progression will jeopardize the reactor system due to rapidly cooling down primary coolant passing through the faulted SG and due to this an injection of positive reactivity in the reactor. As a result this could reflect in possible increasing of reactor power. The purpose of the analysis is to demonstrate the ability of the reactor system to resist to injection of positive reactivity caused by the uncontrolled cooling down of the primary coolant.
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In plant accident context the materials and operators are under complex loading conditions. As a result of the rapid and sudden change of these conditions the materials and operators suffer from shock, e.g. core damage, pressurized-thermal shock and operator stress. The paper discusses the role and function of the symptom-based emergency procedures for diagnostics, prediction and control of human and material shocks in the accident context. On the base of realistic description of the accident events in nuclear power plant, the ways to analyze for the correlation of human and material shock conditions for symptom-based emergency procedure are presented. A "Material" Context Model of the Man-Machine Interaction is proposed. INTRODUCTION A realistic description of the emergency context in which materials and operators are situated is a general precondition for phenomena study as well as for their successful agreement. This is also the ground for development of the instructions used by operators for taking control during the accident, including Symptom-based Emergency Operating Procedure (EOP). The study of accidents is most commonly dominated by determination of Critical Safety Functions (CFS) connected with core parameters (fuel temperature, clad temperature, cooldown, etc.). In recent years great attention was paid to the research of the behavior of materials in reactor pressure vessels (RPV) and pipes of the Pressurized Water Reactor (PWR) units, especially of Water Water Energy Reactor (WWER) in Eastern Europe, due to the possibility of Pressurized Thermal Shock (PTS) [IAEA, 1997 and IAEA, 1999]. The study of human behavior in the Nuclear Power Plant (NPP) emergency situation is mainly connected with the assignment of human error probability. In order to obtain this probability, the 'abstract' operator influence factors are combined. Usually the operator's performance is "fragmented" into separate actions in view of their presentation in the Event Tree (ET) and Fault Tree (FT) probability models. However, it is possible to employ a more realistic approach based on emergency context description. It is apparent that the trends in the study of any stress are connected with a thorough description of the accident context. This is also a requirement in development, verification and validation (V&V) of the NPP EOP. In our study several methods are applied: EOI (Emergency Operating Instruction) T/H (Thermal-Hydraulic) Analysis, PTS Analysis and HRA (Human Reliability Analysis), as part of PRA (Probabilistic Risk Assessment).
During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.
During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.
This paper presents the results of thermal-hydraulic calculation of “Steam Line Break” analysis at full power reactor for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of symptom based emergency operating procedures (SB EOPs) for this plant. The RELAP5/MOD 3.2 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the systems thermal-hydraulics code RELAP5/MOD 3.2 at the Institute for Nuclear Research and Nuclear Energy–Bulgarian Academy of Sciences (INRNE–BAS), Sofia.
The paper presents the results of an investigation of Steam Generator Tube Rupture (SGTR) accident. In the paper are discussed Nuclear Power Plant (NPP) behaviour and the operator actions during primary to secondary (PRISE) leakage. During the transient, it is assumed opening and stuck in open position a Steam Generator Safety Valve (SG SV) of damaged steam generator. Opening of a SG SV will be possible to happen only in case of earlier isolation of damaged SG from a Main Steam Header (MSH). In this way the investigated accident progression will jeopardize reactor system to lose all of primary coolant, including and all emergency coolant water on the NPP Unit from one side and from the other side there is a direct by passing of confinement and an earlier releases of radionuclides.
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.The assessment of the model has been performed in two stages: an initial validation against steady-state plant data and then a transient validation by comparison to operational transient data. The comparison between the plant data and the results of the calculation proved the adequacy of the model and demonstrated the capability of the code to reproduce the evolution of the main plant parameters.
The rod ejection accident (REA) and the main steam line break (MSLB) are two of the most important design basis accidents (DBA) for VVER-1000 exhibiting significant localized space-time effects. A consistent approach for assessing coupled three-dimensional (3-D) neutron kinetics/thermal-hydraulics codes for reactivity insertion accidents (RIA) is to first validate the codes using the available plant test (measured) data and after that to perform cross code comparative analysis for REA and MSLB scenarios. The coupled 3-D neutron kinetics/thermal-hydraulics benchmark presented in this paper is based on data from the Unit #6 of the Bulgarian Kozloduy Nuclear Power Plant (KNPP) and it is entitled the VVER-1000 coolant transient (V1000CT) benchmark. Two real plant transients are selected for simulation in the benchmark: main coolant pump start-up (Phase 1) and coolant mixing tests (Phase 2). In addition to these transients extreme scenarios were defined for better testing 3-D neutronics/thermal-hydraulics coupling: rod ejection simulation with control rod being ejected in the core sector cooled by the switched on MCP (Phase 1) and MSLB transient (Phase 2). The paper presents an overview of the Phase 1 (V1000CT-1) benchmark activities and describes the approach used for assessing the coupled neutron kinetics/thermal-hydraulics codes. Selected comparative analysis of currently submitted participants' results is presented with emphasis on the observed modeling issues and deviations from the measured data.
This paper presents the validation of RELAP5/MOD3.2 model of the VVER 440 for Nuclear Power Plant (NPP) in the analysis of the following transient: "Trip off one MCP". This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with measurement transient data received from Kozloduy NPP Unit #4. The baseline input deck for VVER440 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 comparison between the RELAP5 calculations and the test data indicates a good agreement. This validation was possible through the participation of leading specialists from Kozloduy NPP and with the support of the Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.
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.
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: “Control rod assembly drops to fully inserted position”.This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with experimental transient data received from Kozloduy NPP, Unit #2. The model of VVER 440 was developed at the 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.
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