Impacts and fluid-structure interaction in pressurized water reactor safety analysis

A numerical method is described for the analysis of coupled three-dimensional fluid-structure motion with impacts between structural parts at rigid or flexible supports with small clearances. The method is used for the analysis of the blowdown loadings and the response of internal structures in the vessel of a pressurized water reactor (PWR) in the hypothetical event of a sudden break of a coolant inlet pipe. The method is a generalization of the existing code FLUX which simulates the three-dimensional fluid-structure motion by means of an implicit time integration scheme. The additional supports with clearances are taken into account by applying support forces to the freely moving fluid-structure system. The forces are determined such that the kinematic constraints are enforced at each time step. Numerically, these forces are determined efficiently using a precomputed influence matrix which defines the dynamic displacement per time step at each support due to a unit force at each other support. According to the actually “active” supports the relevant influence matrix in constructed. Energy is conserved for rigid supports and for supports which are so flexible that the impact time is large in comparison to the time steps. Treatment of plastic supports is possible.An application of the new method is demonstrated by analysis of the core barrel motion in a PWR with and without impacts at the lower core barrel edge and at the upper flange. The results show the large effects of such impacts in changing the global motions. Large local impact forces and accelerations appear. The interaction with the fluid reduces these loads. By proper design of the supports the resultant stresses can be minimized. Thus the method can be used to demonstrate and enlarge nuclear reactor safety.     

Analytical investigation of a one-dimensional homogenized model for a pressurized water reactor core

A one-dimensional homogenized model for dynamic fluid-structure interaction in a pressurized water reactor core is used to study the influence of the virtual density and spacer's stiffness. The model consists of a linear system of partial differential equations for fluid velocity, rod velocity and pressure. For these equations analytical solutions are deduced for boundary conditions prescribing either periodic wall oscillations or linearly growing wall accelerations from rest. The theoretical model for the virtual density is verified by comparison to an experiment. For zero spacer stiffness, purely acoustic oscillations appear. For positive spacer stiffness, additional oscillations arise with relative rod motions. The wavelengths of the latter oscillations are small for weak spacers. Large numerical effort would be required in a more complete three-dimensional core-model to resolve such short wave lengths. In fact in a typical core the spacer's stiffness c_s is small in comparison to the fluid bulk modulus K. For it might be appropriate to neglect the influence of the spacers.     

Experimental study on the blowdown load during the steam generator feedwater line break accident in the evolutionary pressurized water reactor

Integral effect tests using the ATLAS facility were performed to obtain the thermal-hydraulic parameters such as dynamic and static pressures, local temperatures, and flow rates during a feedwater line break of a steam generator. The break of a feedwater line was simulated using a double rupture disc assembly in order to satisfy the requirements for the break opening time of around a few milliseconds. In the present study, estimated break opening time was less than 1.5 ms and broken areas were 48.1% and 93.4% of the feedwater line, respectively. The maximum dynamic pressures of about 1.57 bar were obtained inside of feedwater box that was closest to the break location of the feedwater line. After the break of the feedwater line, propagation of the pressure wave along the distance from the break location inside the steam generator was clearly and pertinently observed in all the tests. From a structural integrity point of view, however, the risk induced by this maximum dynamic load could be treated to be insignificant.     

A two-dimensional thermohydraulic calculation on the core of pressurized water reactor

Nuclear Research and Nuclear Power Institute, Bulgarian Academy of Sciences. Translated from Atomnaya Énergiya, Vol. 73, No. 5, pp. 397–400, November, 1992.     

A comparison between code calculation and blowdown experiments simulating a loss-of-coolant accident in a pressurized water reactor

Within the reactor safety programme of the EURATOM Joint Research Centre at Ispra the transient heat transfer phenomena during depressurization are experimentally investigated under PWR conditions. The special closed loop DHT-1 essentially represents one subchannel and the upper and lower plenum of a pressurized water reactor. A test series simulating rupture in the hot leg of a primary cooling circuit was carried out. Pressure and test tube temperatures were measured at various rupture cross-sections. Independently from these experiments, a blowdown computer code was developed by the Groupement Atomique Alsacienne Atlantique (GAAA). The core part of this code allows calculation of the thermohydraulic history of the coolant within the core after a rupture in the primary cooling circuit. It has been checked with regard to the hypothesis and correlations applied; the experiments and calculations are compared.     

Age-related degradation of boiling water reactor vessel internals

Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection.     

Design of the advanced accumulator for the pressurized water reactor

With the increased requirement for nuclear power generation as an effective countermeasure against global warming, Mitsubishi has developed the advanced pressurized water reactor (APWR) by adopting a new component of the emergency core cooling system (ECCS), a new instrumentation and control system, and other newfound improvements. The ECCS introduces a new passive component called the advanced accumulator which integrates both functions of the conventional accumulator and the low-pressure pump without any moving parts. The advanced accumulator uses a new fluidics device that automatically regulates flow rates of injected water in case of a loss of coolant accident (LOCA). This fluidics device is referred to as a flow damper. This paper describes the design method of the flow damper and the standpipe.     

Analysis of HDR blowdown experiment for fluid-structure interactions with multiflex

The HDR experimental facility of Kahlsruhe is comprised of a full-scale pressure vessel, core barrel, and piping systems. In the blowdown experiment, V31.1, the fluid-structure interaction of the core barrel and downcomer water is significant. This experiment is analyzed in the present paper. The HDR downcomer annulus is modeled by the one-dimensional network that is equivalent to two-dimensional fluid-structure interactions. The core barrel is modeled by the projector method for combined beam and shell models. The vessel motion is taken into account by means of the relative modal analysis proposed in this paper. Computed time histories of pressure, pressure differentials, and barrel wall displacements are compared with the experimental data. Fair agreement between experiment and post-test computation is found. Effects of the vessel motion are also discussed.     

Identification of initiating events for pressurized water reactor accidents

In order to aid operators in identifying the different initiating events as defined in the Final Safety Analysis Report (FSAR), we develop a novel identification procedure. The procedure is based on the monitoring of three key system parameters in a pressurized water reactor (PWR), i.e., the pressure, the average temperature, and the temperature difference of the hot-leg and cold-leg of the reactor coolant system. By monitoring the system thermal state diagram in a pressure–temperature space, an operator can easily identify what initiating event is taking place while a static point in the diagram starts to move. The event data pool is first established by storing the transient analysis results for events of different types using the optimal estimated RELAP5 model. Since the variation ranges of system key parameters at a specific time represent the specific character for each initiating event, the identification procedure can easily determine which cases in which the event data pool can be fitted to on-line data using only variation range comparison without complex calculations. This identification method is believed to be able to help the plant operator to identify the different events and then execute the Emergency Operating Procedure more effectively.     

Dynamic analysis of a pressurized water reactor

An apparent malfunction in a pressurized water reactor system has been investigated using fluctuation analysis. Both frequency-domain and time-domain analyses have been used and the results obtained by the two methods have been compared. The recording was performed by a relatively simple, cheap system giving high recording precision and the analyses were performed on an IBM 370 digital computer. It is shown that, while considerable information can be derived from frequency-domain analyses, a misinterpretation can occur in some cases. Time-domain correlation, as normally performed, was not very informative. However, time-domain correlation on bandwidth-limited time-series proved to be very valuable and could remove the misinterpretation of the frequency-domain analyses. The bandwidth limitation was performed by digital filters.     

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment.     

Two-phase countercurrent flow in a model of a pressurized water reactor hot leg

The onset of flooding or countercurrent flow limitation (CCFL) determines the maximum rate at which one phase can flow countercurrently to another phase. In the present study, the experimental data of the CCFL for gas and liquid in a horizontal pipe with a bend are investigated. The different mechanisms that lead to flooding and that are dependent on the liquid flow rate are observed. For low and intermediate liquid flow rates, the onset of flooding appears simultaneously with the slugging of unstable waves that are formed at the crest of the hydraulic jump. At low liquid flow rates, slugging appears close to the bend; at higher liquid flow rates, it appears far away from the bend, in the horizontal section. For high liquid flow rates, no hydraulic jump is observed, and flooding occurs as a result of slug formation at the end of the horizontal pipe. The effects of the inclination angle of the bends, the liquid inlet conditions and the length of the horizontal pipes are of significance for the onset of flooding. A mathematical model of Ardron and Banerjee is modified to predict the onset of flooding. Flooding curves calculated by this model are compared with present experimental data and those of other researchers. The predictions of the onset of flooding as a function of the length-to-diameter ratio are in reasonable agreement with the experimental data.     

Control rod drop hydrodynamic analysis for a pressurized water reactor

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.     

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)-ZrH_1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO₂ oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO₂ and centerline/peak for U-ZrH_1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient.In general, the thermal hydraulic performance of U-ZrH_1.6 and UO₂ fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW_th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW_th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW_th may be feasible.     

A two-dimensional model for fluid-structure interaction in curved pipes

A “channel” model was developed for the purpose of simulating the interactive fluid-structural response of curved pipes to pressure pulses. Simulation is shown to have been achieved analytically in both the axisymmetric (“breathing”) and transverse (“bending”) modes of interactive behavior.An experimental program which was aimed at the validation of the model is also described. Tests were run in both straight and curved pipe configurations. Comparisons between measurements and model calculations demonstrate the validity of the model within the range of parameters under consideration.The model was implemented into the DISCO code for nonlinear fluid-shell interaction.     

Monitoring of the VVER nuclear reactor internals

This paper presents a review of the published pressurized water reactor accidents caused by internal vibrations. These accidents are evaluated with respect to their impact on the safety and economy of nuclear power generation. Subsequently, criteria for a monitoring system which would allow the prevention of such accidents are proposed. Structures and some results obtained at NRI e in the course of such system development are presented as well. The results comprise experimental and computational analyses of the behaviour of the VVER-440/V-213 reactor internals and primary coolant.     

Identification of pressurized water reactor transient using template matching

Template matching, which is a pattern recognition method, was adopted to identify the transient in a pressurized water reactor (PWR). The transient data were generated using a plant simulation code, PCTran-PWR, and transformed into a feature vector sequence (FVS). The data set contained such FVS as the reference transients. To compare the FVS of the test transient and reference transient, a cost function was defined and dynamic programming was applied to obtain the minimum cost function value, which would indicate the degree of matching between the two transients. Considering the discrepancy between the real plant and the model to generate the transient data, the same test and reference transient may not be matched exactly. A dynamic threshold value was designed to determine if the test transient matched the reference transient of the data set. Experiments were performed and the results showed that the method was successful.     

Neutron noise monitoring of pressurized water reactor internal vibrations

Analytical and experimental results have shown that the neutron noise signals are typically the sum of a number of different noise sources. These can have significant interactions due to structural coupling and summation effects in the sensor. Analytical techniques have been developed to identify major neutron noise sources and to separate and account for some of the noise source coupling effects.

This work has demonstrated the use of various noise source models in neutron noise monitoring applications. Methods of identifying and separating the noise sources have been used to relate changes in the measured spectra to particular noise source properties. The noise source models can then be used to relate noise source properties to physical properties of the system. These techniques are used in routine surveillance applications and have provided proper evaluation of several trends and changes that have been observed in neutron noise monitoring programs.

All neutron noise measurements have shown small vibration amplitudes that are in agreement with results from preoperational measurements and analysis. Neutron noise monitoring is being continued on an optional basis in a number of plants as a means of monitoring core clamping and general long-term performance.