Main Steam Isolation Valve Closure Research Articles

BWR-4 ATWS modeling with RELAP5-S3K

Anticipated Transients without Scram (ATWS) are evaluated to demonstrate the capability of the barriers to prevent fission product release. Swiss nuclear authority guidelines allow the use of best-estimate codes and methodologies for beyond design basis events analyses, such as ATWS, to simulate more realistic plant responses. Thus, realistic initial and boundary conditions may be applied so that all systems unaffected by the event are credited and no independent single failures need to be assumed.The RELAP5-S3K coupled thermal-hydraulics/neutronics code was used for KKM (Mühleberg BWR-4 NPP) ATWS analyses. The KKM RELAP5-S3K model was demonstrated to be valid for a wide range of transients including Loss of Coolant Accident (LOCA) and ATWS. Two ATWS cases, a main steam isolation valve (MSIV) closure and a loss of offsite power (LOOP), were simulated to demonstrate the KKM plant capability. The MSIV closure is a limiting ATWS event for reactor pressure vessel (RPV) overpressure and torus pressure and temperature. After closure of the MSIV, all RPV outlet flow is directed to the torus through the safety relief valves resulting in limiting torus pressures and temperatures comparable to other events. The LOOP event can be a limiting ATWS event for peak cladding temperature (PCT) if insufficient makeup water is available.The simulation results showed that the ATWS acceptance criteria for PCT, reactor peak pressure, and torus temperature and pressure are satisfied. The RELAP5-S3K results have been compared to the ATWS results from KKM’s fuel vendor, General Electric-Hitachi (GEH), using their TRACG code (TRACG being perhaps the most advanced system code available for BWRs). The plant response predicted by the RELAP5-S3K model is similar to that reported by GEH. Variations between the two are attributed to (1) code differences in modeling of the physical phenomena and (2) different levels of detail in modeling the KKM plant and core parameters, e.g. cross-sections.Considering that an independently developed model using the RELAP5-S3K code predicts plant behavior during an ATWS similar to the state-of-the-art code TRACG, it is concluded that the KKM ATWS analysis with RELAP5-S3K provides trustworthy information for comparison against safety criteria.

Analysis of the loss of heat sink transients in the secondary circuit of a VVER-1000 using RELAP5/MOD3.2

A major group of accidents related to the secondary loop of a VVER-1000 nuclear reactor are the events that lead to the reduction of heat removal from the secondary side. A reduction in the steam flow rate [due to turbine-generator (TG) load reduction] and spurious closure of a main steam isolation valve (MSIV) are two cases considered in this study.In case 1, the reduction of the TG load by 70% of nominal value is investigated. The transient condition leads to an increase in the reactor core pressure up to nearly 17.3MPa, followed by negative reactivity feedback. After about 1800s into the incident, the power level is stabilized at about 90% of its nominal value.In case 2, MSIV closure due to a control system malfunction or an operator’s error could be the causes of the initiating event. The condition leads to deterioration of heat removal from the secondary loop and the primary pressure increases up to about 16.3MPa as a consequence. The cool-down of the reactor plant (RP) is carried out through control rods and safety valves mechanisms. After 100s from the initiating event, the reactor power decreases to 70% nominal value and after about 500s into the incident complete stabilization of RP parameters is reached.Qualitative examination of the results indicates that the VVER-1000 RELAP5 model has the ability to predict the main thermo-hydraulic parameters and their behavior during the design-base accidents. Using the Fast Fourier Transform Best Estimate Method (FFTBM) analysis, the quantitative analysis of the results also demonstrates that our RELAP5 model of VVER-1000 is reliable for other transient analyzes.

Validation of the Laguna Verde Retran-03 Model for Pressurization Transients

The results obtained from the analysis of two operational transients—a load rejection and a main steam isolation valve closure, both at conditions near 100% of rated power—of the Laguna Verde nuclear power plant using RETRAN-03 (RETRAN) are presented. The methodology followed to obtain the cross section for the RETRAN-03 one-dimensional kinetics model using the fuel management system package is also discussed.

Anticipated Transient Without Scram Analysis of the Simplified Boiling Water Reactor Following Main Steam Isolation Valve Closure with Boron Injection

The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system.

Evaluation Tests of Event Identification Method Using Neural Network at Kashiwazaki Kariwa Nuclear Power Station Unit No.4

An event identification method using a neural network has been evaluated in an on-line environment during the plant startup test at Unit No.4 Plant in the Kashiwazaki Kariwa Nuclear Power Station. In the method, the neural network identifies the event from the change pattern of analog data, such as reactor pressure signals, and then the result is confirmed or similar events are discriminated using digital data, such as valve open signals. Before the test the neural network is trained for the events causing a reactor scram by using analysis results. For the test the method is incorporated into a prototype of the alarm handling system which is connected to the plant facilities. Five kinds of analog data are acquired and eight sampled data from each, namely a total of 40 data, are input to the neural network after normalization. The results show that the load rejection, the turbine trip and the main steam isolation valve closure events are correctly identified from 9 kinds of subject events, regardless of th...

Use of the Chinshan Plant Analyzer for Analysis of a Main-Steam-Isolation-Valve-Closure Anticipated Transient Without Scram

An anticipated transient without scram induced by main steam isolation valve closure (AMSIV) could subject a nuclear power plant to the most severe of accident conditions. The Chinshan plant analyzer contains a complete boiling water reactor system model and can be revised easily for the user`s purpose. These features make the Chinshan plant analyzer suitable for AMSIV analysis. The capability of the Chinshan plant analyzer to analyze an AMSIV is illustrated. An AMSIV is simulated, and the simulation results are similar to the results of other research. Furthermore, the AMSIV response of reducing reactor power by decreasing reactor coolant inventory is simulated, and the results of the simulation are similar to those of other research. During this transient, the reactor power is decreased. However, the margin to core uncovery is also decreased. In addition, a method of reducing the reactor power by increasing the feedwater temperature is studied. The mechanism of reducing the reactor power is associated with decreasing the inlet subcooling. Sensitivities of key parameters are also analyzed. A large negative void coefficient causes in undesirable large peak in the reactor power. A small recirculation pump moment of inertia decreases the reactor power.

Evaluation Tests of Event Identification Method Using Neural Network at Kashiwazaki Kariwa Nuclear Power Station Unit No.4.

An event identification method using a neural network has been evaluated in an on-line environment during the plant startup test at Unit No.4 Plant in the Kashiwazaki Kariwa Nuclear Power Station. In the method, the neural network identifies the event from the change pattern of analog data, such as reactor pressure signals, and then the result is confirmed or similar events are discriminated using digital data, such as valve open signals. Before the test the neural network is trained for the events causing a reactor scram by using analysis results.For the test the method is incorporated into a prototype of the alarm handling system which is connected to the plant facilities. Five kinds of analog data are acquired and eight sampled data from each, namely a total of 40 data, are input to the neural network after normalization. The results show that the load rejection, the turbine trip and the main steam isolation valve closure events are correctly identified from 9 kinds of subject events, regardless of the difference between the trained analysis results and the recognized plant data. This is owing to the data sampling and normalization methods as well as the robustness of the neural network.

Development of a Method of Selecting Important Alarms for Nuclear Power Plants

Presenting important alarms selected from a large number of activated alarms provides useful operational support under a transient status in a nuclear power plant. We have developed an alarm processing method which selects and presents important alarms depending on plant status. In this method, important alarms are selected, first, based on physical relationships between alarms and component status including alarms themselves and second, even more important alarms are selected from the previously selected alarms according to the identified initial event causing the transient. Identification of the initial event is implemented by a neural network. The identified initial event and selected important alarms are presented to show the cause and influence of the transient. A prototype based on the above alarm processing method was validated during the start-up test at Kashiwazaki Kariwa Nuclear Power Plant Unit Number 4 of Tokyo Electric Power Co. The initial events, which were load rejection, turbine trip and main steam isolation valve closure, were correctly identified and about 30% of all activated alarms were selected as important. It was verified by an operating expert that the presentation of the identified initial event and the selected important alarms were effective to understand rapidly and correctly the transient status of the plant.

Influence of Isolation Condensers on Boiling Water Reactor Performance During an Anticipated Transient Without Scram

The purpose of isolation condensers (ICs) in a boiling water reactor (BWR) is to passively control the reactor pressure by removing heat from the system during abnormal reactor operation. This type of control is expected to reduce the frequency of opening and closing of the safety relief valves (SRVs). In addition, the ICs are designed to remove the decay heat of the reactor system after reactor shutdown in case of a loss of coolant accident. Transient calculations of an anticipated transient without scram for a main steam isolation valve (MSIV) closure event were performed using the RAMONA-4B code with prescribed pressure setpoints for the SRVs and given time settings for the MSIV closure. It is demonstrated that the effect of the ICs is to reduce the rate of pressure rise and thereby decrease the cycling frequency of the SRVs. The onset of high-pressure coolant injection is delayed if an IC is functional. An analysis for natural circulation flow in a BWR indicates that the effect of an IC on the transient performance is similar to that for the forced circulation system. In this case, the MSIV closure results in a lower peak pressure because of the lower power level

Modeling Condensation Steam Quenching Effects in an MSIV Closure Transient Using RETRAN-02-MOD002 With and Without Nonequilibrium Pressurizer Model Option and TRAC-BF1

The effects of condensation steam quenching in modeling two-phase flow phenomena during a nuclear reactor transient are studied. The RETRAN-02-MOD002 code, with three field equations and a nonequilibrium pressurizer model option, and the TRAC-BF1 code, with six field equations, predicted plant response to a boiling water reactor plant test of a main steam isolation valve closure without safety relief valve opening. The basic RETRAN-02-MOD002 field equations cannot model steam quenching by condensation. However, by activatixg the nonequilibrium modeling option of the basic RETRAN-02-MOD002 code and by inputting appropriate interfacial heat transfer coefficients, steam quenching by condensation was calculated

Analysis of a Main Steam Isolation Valve Closure Anticipated Transient Without Scram in a Boiling Water Reactor

Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main stream isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4.

Retran Modeling of the Westinghouse Model D Steam Generator

The ability to model and predict the steady-state and transient thermal-hydraulic behavior of the Westinghouse model D steam generator is an important prerequisite for performing safety and licensing analyses of Commonwealth Edison’s Byron and Braidwood nuclear power plants. A RETRAN model using ten volumes to represent the primary side and ten volumes to represent the secondary side is developed. The model is validated over a range of steady-state conditions and is used to calculate the pressure and level response to a main steam isolation valve closure using operational data to drive the transient and a basis for comparison. Sensitivity studies and a sample reload licensing calculation are performed to further determine the model’s capabilities. The results of the model development show that the RETRAN model is a viable tool for analysis of the model D steam generator’s steady-state and transient behavior. Examination of the model’s behavior during rapid secondary depressurization events and confirmation of the carryover behavior is recommended.

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant. The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions. Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated. The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.

RETRAN Analysis of the San Onofre Unit 2 Turbine Trip from 100% Power

The RETRAN analysis of San Onofre Nuclear Generating Station Unit 2 (SONGS 2) turbine trip from 100% power is a continuing effort of out inhouse RETRAN model development to benchmark the calculations against the plant test data. The previous benchmarked calculations - RETRAN analyses of SONGS 2 flow measurements and low-power natural circulation test - showed very good agreement with the plant test data. The latter RETRAN model was then expanded for the current study to include steam generator secondary side, main steam line pipings, control valves, and plant control systems. The benchmarked turbine trip model has been used to analyze the station-requested plant operational transients. These are assessment of strategies for performing turbine overspeed surveillance, determining a manual position of one steam bypass valve following a reactor trip from 100% power, and plant response to one main steam isolation valve closure at 90 and 50% power levels. In addition, the RETRAN calculations employing benchmarked models can be used to verify the simulator predictions and assess the existing plant operating procedures.