Modeling of the High Pressure Core Spray Systems with fuzzy cognitive maps for operational transient analysis in nuclear power reactors

In this paper the application of fuzzy cognitive maps (FCM) to model a risk scenario for Nuclear Power Plants (NPP) in a Boiling Water Reactor (BWR) is presented, specifically for failure modes and effects analysis of High Pressure Core Spray System (HPCS) during loss of reactor coolant inventory transients. A simplified model of the HPCS is analyzed with the fault tree analysis technique in order to compare this results with those obtained with the FCM and show consistency with the results, although this process is not a validation of the FCM techniques. The decision making in an NPP is a complex process, because of the numerous elements involved in its operation, and the permanent attention demanded by its maintenance. This is the first step in the development of an expert system that will help in the decision making process, through the design of the knowledge representation and the design of reasoning with FCM to automate the decision making process.     

An advanced reduced order model for BWR stability analysis

An advanced reduced order model was developed and qualified in the framework of a novel approach for nonlinear stability analysis of boiling water nuclear reactors (BWRs). This approach is called the RAM-ROM method where RAM is a synonym for system code and ROM stands for reduced order model. In the framework of the RAM-ROM method, integrated BWR (system) codes and reduced order models are used as complementary tools to examine the stability characteristics of fixed points and periodic solutions of the nonlinear differential equations describing the stability behaviour of a BWR loop. This methodology is a novel one in a specific sense: we analyse the highly nonlinear processes of BWR dynamics by applying validated system codes and by the sophisticated methods of nonlinear dynamics, e.g. bifurcation analysis. We claim and we will show that the combined application of independent methodologies to examine nonlinear stability behaviour can increase the reliability of BWR stability analysis.This work is a continuation of previous work at the Paul Scherrer Institute (PSI, Switzerland) of the second author and at the University of Illinois (USA) in this field. In the scope of a PhD work at the Technical University Dresden (Germany), the current ROM was extended to an advanced ROM by adding a recirculation loop model, a quantitative assessment of the necessity for consideration of the effect of sub-cooled boiling and a new calculation methodology for feedback reactivity. A crucial point of ROM qualification is a new calculation procedure for ROM input data based on steady-state RAM (ONA) results. The modified ROM is coupled with the BIFDD bifurcation code which performs a semi-analytical bifurcation analysis (see Appendix C). In this paper, the advanced ROM (TU Dresden ROM, TUD-ROM) is briefly described and the results of a nonlinear BWR stability analysis based on the RAM-ROM method are summarised for NPP Leibstadt, NPP Ringhals and NPP Brunsbüttel. The results show that the TUD-ROM including the new approach for ROM input data calculation is qualified for BWR stability analysis in the framework of the RAM-ROM method.     

Bagged ensemble of Fuzzy C-Means classifiers for nuclear transient identification

This paper presents an ensemble-based scheme for nuclear transient identification. The approach adopted to construct the ensemble of classifiers is bagging; the novelty consists in using supervised fuzzy C-means (FCM) classifiers as base classifiers of the ensemble. The performance of the proposed classification scheme has been verified by comparison with a single supervised, evolutionary-optimized FCM classifier with respect of the task of classifying artificial datasets. The results obtained indicate that in the cases of datasets of large or very small sizes and/or complex decision boundaries, the bagging ensembles can improve classification accuracy. Then, the approach has been applied to the identification of simulated transients in the feedwater system of a boiling water reactor (BWR).     

Emergency management for a nuclear power plant using fuzzy cognitive maps

This paper explores the application of fuzzy cognitive maps (FCM) to emergency operating procedures (EOPS), to represent the decision-making process during abnormal situations in a nuclear power plant (NPP). The decision-making process in a NPP is a complex process, due to the many elements involved in its operation, and the permanent attention demanded by its maintenance. At the present time, the decision making process in a NPP is analyzed and developed by reactor operators, based on a set of instructions as well as flow charts to mitigate the consequences of a broad range of transients, accidents and multiple equipment failures, whose main characteristic is to be linear representations of events within a scenario. One of the main objectives of this paper is to present a method based in FCM that could be applied in the development of EOPS, and show some simulations, specifically the loss of coolant accident (LOCA) scenario in a boiling water reactor (BWR) with the Mark II containment design was studied. The FCM-based method represents with high fidelity the expert reasoning (the human expert is very important) and the interpretation of the results aids instantly to the reactor operators in the surveillance of the reactor proper functionality due that they have the responsibility of the decision taking in emergency situations. The simulations results show that the FCM predict properly the phenomenon in the reactor vessel and primary containment.     

AXIAL: a system for boiling water reactor fuel assembly axial optimization using genetic algorithms

A system named AXIAL is developed based on the genetic algorithms (GA) optimization method, using the 3D steady state simulator code Core-Master-PRESTO (CM-PRESTO) to evaluate the objective function. The feasibility of this methodology is investigated for a typical boiling water reactor (BWR) fuel assembly (FA). The axial location of different fuel compositions is found in order to minimize the FA mean enrichment needed to obtain the cycle length under the safety constraints. Thermal limits are evaluated at the end of cycle using the Haling calculation; the hot excess reactivity and the shutdown margin at the beginning of cycle are also evaluated. The implemented objective function is very flexible and complete, incorporating all the thermal and reactivity limits imposed during fuel design analysis; furthermore, additional constraints can be easily introduced in order to obtain an improved solution. The results show a small improvement in the FA average enrichment obtained with the system related to the reference case that has been studied. The results show that the system converge to an optimal solution, it is observed that the mean fuel enrichment decreases while all the constraints are satisfied. A comparison was also performed using one-point and two-points crossover operator and the results of a sensitivity study for different mutation percentage are also showed.     

Hydrogen water chemistry for BWRs: a status report on the EPRI development program

Many boiling water reactors (BWRs) have experienced extensive intergranular stress corrosion cracking (IGSCC) in their austenitic stainless steel reactor coolant system piping, resulting in serious adverse impacts on plant capacity factors, O&M costs, and personnel radiation exposures. A major research program to provide remedies for BWR pipe cracking was co-funded by EPRI, GE, and the BWR Owners Group for IGSCC Research between 1979 and 1988. Results from this program show that the likelihood of IGSCC depends on reactor water chemistry (particularly on the concentrations of ionic impurities and oxidizing radiolysis products) as well as on material condition and the level of tensile stress. Tests have demonstrated that the concentration of oxidizing radiolysis products in the recirculating reactor water of a BWR can be reduced substantially by injecting hydrogen into the feedwater. Recent plant data show that the use of hydrogen injection can reduce the rate of IGSCC to insignificant levels if the concentration of ionic impurities in the reactor water is kept sufficiently low. This approach to the control of BWR pipe cracking is called hydrogen water chemistry (HWC). This paper presents a review of the results of EPRI's HWC development program from 1980 to the present. In addition, plans for additional work to investigate the feasibility of adapting HWC to protect the BWR vessel and major internal components from potential stress corrosion cracking problems are summarized.     

Comments on the application of bifurcation analysis in BWR stability analysis

Currently, BWR stability analysis is most often performed by the application of system codes which provide the time evolution of the neutron flux or thermal power at a defined operational point (OP) after imposing a system parameter perturbation. However, in general it is impossible to understand the real stability state of the BWR at a specific OP by the application of system code analysis alone. Hence, we are exploring methods developed in the nonlinear dynamics field in order to reveal the nature of the BWR stability states when power oscillations are observed. A powerful method is bifurcation analysis. In order to motivate this “nonlinear thinking” versus “linear thinking”, in this paper we will demonstrate some examples of phenomena which can only be understood in nonlinear terms by application of bifurcation theory and where linear interpretation leads to incorrect conclusions.     

Investigation of BWR LOCA at ROSA-III — Effect of break configuration on system transients

The ROSA-III test facility is a volumetrically scaled ( ) BWR/6 system with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA).Six loss-of-coolant experiments with a break area of 15%, 50% or 200% at the main recirculation pump inlet line were conducted at the ROSA-III test facility with a high pressure core spray failure. A sharp-edged orifice or a long throat nozzle was used as a break plane. It was found in the experiments that the break flow differences between the orifice and the nozzle break configurations with the same flow area were observed only in the subcooled break flow region. Subcooled break flow rate through the orifice was much larger than that through the nozzle. The break configuration difference had little influence on the other system responses, especially on the peak cladding temperature. The applicability of the test results to a BWR/6 has been confirmed through analyses of the 15% break ROSA-III LOCA experiments and BWR/6 LOCAs by using RELAP4/MOD6/U4/J3 code. The experimental results of the ROSA-III LOCA experiments were calculated well by the code, and the same trends were calculated in the BWR analyses.     

Validation of reactor noise linear stability methods by means of advanced stochastic differential equation models

The aim of this paper is to show a validation method of a stability monitor using a BWR model with multiple Wiener noise sources, of additive and multiplicative nature. This model is solved using the modern methods to integrate stochastic differential equation systems, that are based on the stochastic Îto-Taylor expansion, and developed by Kloeden and Platen (1995), Kloeden et al. (1994). The synthetic signals generated with this BWR reduced order model with multiple Wiener processes are then used to obtain what are the optimal ways of filtering the signals for the different methods to estimate the decay ration (DR) and the natural frequency (ω) of the system. Also, for each DR estimation method, we study what is the optimal combination of algorithms to obtain the order and coefficients of the AR model that yields the best prediction of the reactor stability parameters for a broad range of DR values.     

Description of void effects in a large BWR by means of a stochastic theory and simulation in a zero-power reactor

We first summarize the stochastic point model developed in previous papers to describe void effects in a large BWR and also summarize our results. The most important of these is the existence of a resonance frequency in the auto-power spectral density of the neutron noise (APSD), the position of which depends on the reactor characteristics, such as power or void coefficient.

In order to check the validity of this model, we made experiments simulating heat transfer and steam fluctuations in a BWR. A stochastic interpretation of the experiments is developed, and results are found to be similar to those obtained with the BWR model. In particular, the zero-power reactor with the simulation device exhibits a resonance frequency showing an identical behaviour to the one predicted for a BWR.

The APSD resulting from experiments in the zero-power reactor CROCUS, at a given power level, is fitted on the theoretical curves by means of the least square method, which provides the resonance frequency. The behaviour of this frequency as a function of the power level agrees fairly well with the theoretical prediction.

If we suppose that feedback mechanisms are the same in a large BWR, we can also admit that the stochastic model gives correctly the resonance frequency.     

Containment venting as a mitigation technique for BWR MARK I plant ATWS

Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it.Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure. For both cases the operators are assumed to take action to intentionally restrict injected flow such that fuel in the upper part of the core would be steam cooled. Resulting fuel temperatures are estimated with an off-line calculation and found to be acceptable.     

Design and optimization of an equilibrium reload with MOX fuel with minor actinides

In this work the design and optimization of an equilibrium core for a boiling water reactor (BWR), loaded with fuel composed of plutonium and minor actinides (Np, Am and Cm), is presented. The plutonium and minor actinides are obtained from the recycling of the spent fuel of a BWR, and are mixed with depleted uranium obtained from the enrichment tails. The design and optimization of the equilibrium reload is achieved in two steps. In the first step, the fuel assembly is adjusted and the reload pattern is designed, in order to obtain the target cycle length. In order to improve the shutdown margin, two actions were taken: to increase the boron-10 content in the control rods, and to add a burnable absorber (gadolinia) in some fuel rods. In the second step, the reload pattern, obtained in the first step, is optimized to maximize the energy, under the thermal and reactivity margins constraints; a system based on Genetic Algorithms was used in the optimization process. Results show that 5% more energy was obtained with the optimized reload.     

Application of continuous-energy Monte Carlo code as a cross-section generator of BWR core calculations

A continuous-energy Monte Carlo code is newly applied for the assembly calculations of actual BWR core analysis. Few-groups cross-sections and related constants (kinetic parameters) were generated by the continuous-energy Monte Carlo code MVP-BURN, and were tabulated for a core simulator. The commercial BWR, HAMAOKA-3 (1100MWe:BWR-5), was analyzed by a coupled neutronic-thermalhydraulic core simulator based on modified one-group diffusion theory using these assembly constants. The calculated core parameters showed good agreement with the results of the on-line core monitoring system of HAMAOKA-3. Consequently, it was confirmed that the present method is applicable to BWR core production calculations. The present method is a particularly attractive candidate for the analysis of advanced BWR fuel assemblies with exotic geometry and high Gd content, due to the features of the continuous-energy Monte Carlo code, i.e., high accuracy and generalized geometry treatment.     

Simulation of an hypothetical out-of-phase instability case in boiling water reactor by RELAP5/PARCS coupled codes

A very complex type of power instability occurring in boiling water reactor (BWR) consists of out-of-phase regional oscillations, in which normally subcritical neutronic modes are excited by thermal-hydraulic feedback mechanisms. The out-of-phase mode of oscillation is a very challenging type of instability and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, simulations of out-of-phase instabilities in a BWR obtained by assuming an hypothetical continuous control rod bank withdrawal are being presented. The RELAP5/Mod3.3 thermal-hydraulic system code coupled with the PARCS/2.4 3D neutron kinetic code has been used to simulate the instability phenomenon. Data from a real BWR nuclear power plant (NPP) have been used as reference conditions and reactor parameters. Simulated neutronic power signals from local power range monitors (LPRM) have been used to detect and study the local power oscillations. The decay ratio (DR) and the natural frequency (NF) of the power oscillations (typical parameters used to evaluate the instabilities) have been used in the analysis. The results are discussed also making use of two-dimensional plots depicting relative core power distribution during the transient, in order to clearly illustrate the out-of-phase behavior.     

Flashing-induced density wave oscillations in a natural circulation BWR—mechanism of instability and stability map

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations.     

Five percent break BWR LOCA/ECC test at ROSA-III without HPCS actuation - two-dimensional core thermal-hydraulic phenomena

The ROSA (Rig of Safety Assessment)-III facility is a volumetrically scaled (1/424) simulated boiling water nuclear reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. A recirculation pump suction line break test with a five percent break area was conducted with the assumption of high pressure core spray system (HPCS) failure. The simulated peripheral fuel rods facing the channel box wall had a tendency to be rewetted temporarily at the upper part of the core by falling water from the upper plenum before low pressure core spray system (LPCS) actuation, while the rods in the central region were not rewetted but quenched mainly from the bottom of the core after low pressure coolant injection system (LPCI) actuation. Therefore, the peak cladding temperatures of the simulated high power fuel rods were limited to lower values since they were located in the peripheral region and the temporary rewetting before LPCS actuation occurred mainly in the peripheral region. The ROSA-III five percent break test and a BWR counterpart were analyzed with the RELAP5/MOD1 (cycle 018) code. Similarity between the ROSA-III small break test and a BWR small break LOCA has been confirmed through comparison of the calculated results.     

Development of SIRIUS-N facility with simulated void-reactivity feedback to investigate regional and core-wide stability of natural circulation BWRs

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations.     

Decay ratio estimation based on time–frequency representations

A novel method based on bilinear time–frequency representations (TFRs) is proposed to determine the time evolution of the linear stability parameters of a boiling water reactor (BWR) using neutronic noise signals. TFRs allow us to track the instantaneous frequencies contained in a signal to estimate an instantaneous decay ratio (IDR) that closely follows the signal envelope changes in time, making the IDR a measure of local stability. In order to account for long term changes in BWR stability, the ACDR measure is introduced as the accumulated product of the local IDRs. As it is shown in this paper, the ACDR measure clearly reflects major long term changes in BWR stability. Last to validate our method, synthetic and real neutronic signals were used. The methodology was tested on the Laguna Verde Unit 1, two events were reported in the Forsmark stability benchmark.     

An Analytical Study of Nonlinear Reactor Dynamics Based on Bifurcation Theory with Reference to BWR Power Oscillation

A nonlinear reactor dynamics model of reduced order is derived and an analytical study on BWR power oscillation is made using this model. It provides some essential features which are not given by numerical studies, such as the explicit expressions of the linear stability condition and the weak stability condition which is related to the periodic motion. In addition, the relation between the reactivity feedback and these conditions is obtained. The application of the analytical results to the qualitative analysis of BWR dynamics is easy and quick in comparison with numerical approaches.     

Three-dimensional time-averaged void fraction distribution measurement technique for BWR thermal hydraulic conditions using an X-ray CT system

We have developed a void fraction distribution measurement technique using the three-dimensional (3D) time-averaged X-ray computed tomography (CT) system to understand two-phase flow behavior inside a fuel bundle for boiling water reactor (BWR) thermal hydraulic conditions of 7.2 MPa and 288 °C. As a first step, we measured the 3D void fraction distribution in a vertical square (5?×?5) rod array that simulated a BWR fuel bundle in the air–water test. A comparison of the volume-averaged void fractions evaluated by the developed X-ray CT system with those evaluated by a differential pressure transducer showed satisfactory agreement within a difference of 0.03. Thus, we confirmed that the developed system could be used to get 3D imaging of the vertical square rod array used in the test under the BWR operating pressure condition. In the next step, we did a verification test using the vertical pipe (11.3 mm ID) for BWR thermal hydraulic conditions. A comparison of the cross-sectional-averaged void fractions evaluated by the X-ray CT system with those evaluated by the drift-flux model showed good agreement within a difference of 0.05. We confirmed that the evaluated void fraction distribution forms in the horizontal cross section changed with the quality in response to the flow regime transition.