Actual Nuclear Power Plant Research Articles

Probability of restoring power to the transmission power system and the time to restore power

Probabilistic safety assessment requires inputs related to the probability of restoring certain equipment versus the available time interval available. The data about the history of events include descriptions of events and the time for restoring the failed equipment. The objective is to determine the approximation function, which can be understood also as a cumulative distribution function for the probability of restoring equipment and to apply it considering the database of previous events. The method includes selection of the function and its fit in order to find the coefficients that match the function to the data. The root mean square method is used for getting the best fit regarding the parameters of the function. The resulted function gives the probability of restoring power before the determined time, e.g. the coping time. The resulted probability can be used in the event trees within the probabilistic safety assessment models. The application of the method was placed to three regions and the resulted functions for probability of restoring power before the determined time were compared to the probabilistic safety assessment of actual nuclear power plant. The results show the differences in core damage frequency and thus in nuclear safety for selected three regions.

Nuclear reactor vessel water level prediction during severe accidents using deep neural networks

Abstract Acquiring instrumentation signals generated from nuclear power plants (NPPs) is essential to maintain nuclear reactor integrity or to mitigate an abnormal state under normal operating conditions or severe accident circumstances. However, various safety-critical instrumentation signals from NPPs cannot be accurately measured on account of instrument degradation or failure under severe accident circumstances. Reactor vessel (RV) water level, which is an accident monitoring variable directly related to reactor cooling and prevention of core exposure, was predicted by applying a few signals to deep neural networks (DNNs) during severe accidents in NPPs. Signal data were obtained by simulating the postulated loss-of-coolant accidents at hot- and cold-legs, and steam generator tube rupture using modular accident analysis program code as actual NPP accidents rarely happen. To optimize the DNN model for RV water level prediction, a genetic algorithm was used to select the numbers of hidden layers and nodes. The proposed DNN model had a small root mean square error for RV water level prediction, and performed better than the cascaded fuzzy neural network model of the previous study. Consequently, the DNN model is considered to perform well enough to provide supporting information on the RV water level to operators.

Numerical investigation on the thermal stratification in a pressurizer surge line

Thermal stratification is a common phenomenon in the surge line of Pressurized Water Reactors (PWRs). A study using Computational Fluid Dynamics (CFD) has been performed to analyze the thermal stratification in the pressurized surge line with different surge line and main pipe velocity conditions. This study is typical of considering both the flow subjected to the buoyancy and the conjugate heat transfer between the fluid and the pipe wall. What makes the research complex is exactly the impact of buoyancy on thermal stratification. To validate the accuracy of CFD calculations, necessary comparison is performed with experiment results. Models with different computational conditions resulting from various similarity criterion parameters are used. More meaningful information than that from the experimental results can be obtained. At the same time, many actual operating conditions which are difficult for experiment can be also accomplished by applying CFD mode. Temperature distributions inside and on the outer surface of the surge line are provided by the CFD mode with different turbulent models. Compared to the standard k-ɛ turbulence model and Reynolds stress model (RSM), the SST k-ω turbulent model is more adapted for this study. Based on the conclusion, the numerical results with the SST k-ω turbulence model for the prototype on the actual nuclear power plant unit were mainly described.

Fault-tree-based instantaneous risk computing core in nuclear power plant risk monitor

Risk monitor is an application of Probabilistic Safety Assessment (PSA) methodology, which has been widely used to determine the instantaneous risk based on the actual configuration of the nuclear power plant (NPP). The instantaneous risk model which has been converted from the baseline risk model of PSA is the fundament of instantaneous risk analysis in a risk monitor. However, the instantaneous risk model is usually very huge and complex, which is hard to be calculated and analyzed by hand. A fault-tree-based instantaneous risk computing core was developed by FDS Team. This code adopts the fast fault tree analysis method based on modified Zero-suppressed Binary Decision Diagram (ZBDD), which includes a set of algorithms such as fault tree importing, fault tree reduction, ZBDD variable ordering, ZBDD parallel manipulation and cut-sets-based ZBDD reconstruction, etc. Benchmarks of actual NPP instantaneous risk models compared with similar codes showed that this code was correct and efficient. This computing core was applied in the Third Qinshan NPP risk monitor TQRM, and some cases based on the actual NPP instantaneous risk models are given.

Interaction and Coordination among Nuclear Power Plants, Power Grids and Their Protection Systems

Nuclear power plants (NPPs) have recently undergone rapid development in China. To improve the performance of both NPPs and grids during adverse conditions, a precise understanding of the coordination between NPPs and grids is required. Therefore, a new mathematical model with reasonable accuracy and reduced computational complexity is developed. This model is applicable to the short, mid, and long-term dynamic simulation of large-scale power systems. The effectiveness of the model is verified by using an actual NPP full-scope simulator as a reference. Based on this model, the interaction and coordination between NPPs and grids under the conditions of over-frequency, under-frequency and under-voltage are analyzed, with special stress applied to the effect of protection systems on the safe operation of both NPPs and power grids. Finally, the coordinated control principles and schemes, together with the recommended protection system values, are proposed for both NPPs and grids. These results show that coordination between the protection systems of NPPs and power networks is a crucial factor in ensuring the safe and stable operation of both NPPs and grids. The results can be used as a reference for coordination between NPPs and grids, as well as for parameter optimization of grid-related generator protection of NPPs.

Prediction of hydrogen concentration in nuclear power plant containment under severe accidents using cascaded fuzzy neural networks

Recently, severe accidents in nuclear power plants (NPPs) have attracted worldwide interest since the Fukushima accident. If the hydrogen concentration in an NPP containment is increased above 4% in atmospheric pressure, hydrogen combustion will likely occur. Therefore, the hydrogen concentration must be kept below 4%. This study presents the prediction of hydrogen concentration using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model was developed using data on severe accidents in NPPs. The data were obtained by numerically simulating the accident scenarios using the MAAP4 code for optimized power reactor 1000 (OPR1000) because real severe accident data cannot be obtained from actual NPP accidents. The root-mean-square error level predicted by the CFNN model is below approximately 5%. It was confirmed that the CFNN model could accurately predict the hydrogen concentration in the containment. If NPP operators can predict the hydrogen concentration in the containment using the CFNN model, this prediction can assist them in preventing a hydrogen explosion.

Assessment of thermal fatigue damage caused by local fluid temperature fluctuation (part I: characteristics of constraint and stress caused by thermal striation and stratification)

This study was aimed at identifying the constraint conditions under local temperature fluctuation by thermal striping at a mixing tee and by thermal stratification at an elbow pipe branched from the main pipe. Numerical and analytical approaches were made to derive the thermal stress and its fluctuation. It was shown that an inhomogeneous temperature distribution in a straight pipe caused thermal stress due to a membrane constraint even if an external membrane constraint did not act on the pipe. Although the membrane constraint increased the mean stress at the mixing tee, it did not contribute to fluctuation of the thermal stress. On the other hand, the membrane constraint played an important role in the fatigue damage accumulation near the stratification layer of the branched elbow. Based on the constraint and stress conditions analyzed, the characteristics of the cracking observed in actual nuclear power plants were reasonably explained. Namely, at the mixing tee, where thermal crazing has been found, the lack of contribution of the membrane constraint to stress fluctuation caused a stress gradient in the thickness direction and arrested crack growth. On the other hand, at the branched elbow, where axial through-wall cracks have been found, the relatively large hoop stress fluctuation was brought about by movement of the stratified layer together with the membrane constraint even under a relatively low frequency of stress fluctuation.

福島原子力発電所の東日本大震災被害からみた1980年の国土庁によるエコロジカル・プランニング調査の現代的意義

One of advantage in ecological planning method of Ian L. McHarg is synthesis of respective optimal landuse (non optimal landuse). The earthquake and tsunami damage in March 11, 2011 recalled us the importance of the comprehensive optimal landuse evaluation. Especially, damages of the Fukushima nuclear plants were catastrophic. This study focused on the criteria of land evaluation paper data in report of "Land Use Aptitude Evaluation Technique Investigation based on Ecological Planning" by National Land Agency Japan in 1980. We traced 5 paper maps (geological map, landform classification map, slope distribution map, vegetation map, and soil cartography map) for digital map, and inclination in coastal areas in where the two Fukushima nuclear plants exist. In addition, we made risk hazard maps (landslide, earthquake resistant, and flood, tsunami damage estimation) on each disaster with above 5 divisions maps and the table risk scores in the Land Agency Report 1980. Finally, we developed a comprehensive hazard map with overlaid a previous evaluation result and evaluation of optimum area of nuclear plant. An actual nuclear power plant site was selected to the coast for the reasons of barren ground bought easily. However, our results indicate that the nuclear power plant site was relative dangerous location in the sight of overall disasters vulnerability by the overlay analysis in 1980 data.

Positron annihilation in neutron irradiated iron-based materials

The hardening and embrittlement of reactor pressure vessel steels is of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, like vacancies, interstitials, solutes and their clusters. But the reason for the embrittlement of the material is not yet totally known. The real nature of the irradiation damage should thus be examined as well as its evolution in time. Positron annihilation spectroscopy has been shown to be a powerful method for analyzing some of these defects. In fact, both vacancy type clusters and precipitates can be visualized by positrons. Recently, at SCK·CEN, a new setup has been constructed, calibrated and optimized to measure the coincidence Doppler broadening and lifetime of neutron irradiated materials. To be able to compare the results obtained by the positron studies, with those of other techniques (such as transmission electron microscopy, atom probe tomography and small angle neutron scattering), quantitative estimations of the size and density of the annihilation sites are needed. Using the approach proposed by Vehanen et al., an attempt is made to calculate the needed quantities in Fe and Fe-Cu binary alloys that were neutron irradiated to different doses. The results obtained are discussed highlighting the difficulties in defining the annihilation centres even in these simple model alloys, in spite of using both lifetime and Doppler broadening measurements in the same samples.

Proposal and Field Practice of a “Hiyarihatto” Activity Method for Promotion of Statements of Participants for Nuclear Power Plant Organization

In a “hiyarihatto” activity, workers report and discuss incident cases related to their work. Such an activity is particularly effective for cultivating participants' attitudes about safety. Nevertheless, a conventional face-to-face hiyarihatto activity includes features that are inappropriate for conduct in a nuclear power plant organization. For example, workers at nuclear power plants are geographically distributed and busy. Therefore, they have great difficulty in participating in a face-to-face hiyarihatto activity. Furthermore, workers' hesitation in discussing problems inhibits the continuation of their active participation. This study is conducted to propose a hiyarihatto activity with an asynchronous and distributed computer-mediated communication (CMC) for a nuclear power plant organization, with the demonstration of its effectiveness through field practice. The proposed method also involves the introduction of special participants who follow action guidelines for the promotion of the continuation of the activity. The method was used in an actual nuclear power plant organization. Results showed that the method is effective under some conditions, such as during periods without facility inspection. Special participants promoted the activity in some cases. Moreover, other factors affecting the activity and some improvements were identified.

OPTIMIZATION OF THE PARAMETERS OF FEEDWATER CONTROL SYSTEM FOR OPR1000 NUCLEAR POWER PLANTS

In this study, the parameters of the feedwater control system (FWCS) of the OPR1000 type nuclear power plant (NPP) are optimized by response surface methodology (RSM) in order to acquire better level control performance from the FWCS. The objective of the optimization is to minimize the steam generator (SG) water level deviation from the reference level during transients. The objective functions for this optimization are relationships between the SG level deviation and the parameters of the FWCS. However, in this case of FWCS parameter optimization, the objective functions are not available in the form of analytic equations and the responses (the SG level at plant transients) to inputs (FWCS parameters) can be evaluated by computer simulations only. Classical optimization methods cannot be used because the objective function value cannot be calculated directely. Therefore, the simulation optimization methodology is used and the RSM is adopted as the simulation optimization algorithm. Objective functions are evaluated with several typical transients in NPPs using a system simulation computer code that has been utilized for the system performance analysis of actual NPPs. The results show that the optimized parameters have better SG level control performance. The degree of the SG level deviation from the reference level during transients is minimized and consequently the control performance of the FWCS is remarkably improved.

Experimental Investigation into the Effects of Decision Support Systems on Operator Performance

As human error is recognized as one of the main causes of nuclear power plant (NPP) accidents, there have been efforts to reduce and prevent human error by developing various decision support systems. Before adapting these support systems to actual NPPs, it is necessary to validate their reliability and to evaluate their effect on operator performance. Particularly for safety-critical systems such as NPPs, the validation and evaluation of decision support systems is as important as the design of good systems. Such evaluations may be carried out through a theoretical modeling or experimentation. The objective of this study is to experimentally investigate the effects of decision support systems on operator performance. Several trends were observed in the results of the experiments in this study; the support systems reduced both human error and workload, and the impact of the systems was most readily observed during complex and highly workloaded situations in which an operator must handle a high volume of information.

Actual flow calibration of a feedwater flowmeter using a high Reynolds number facility at NMIJ

The results of calibration tests of the feedwater flowrate of ultrasonic flowmeters used in a nuclear power plant for variety of upstream conditions obtained using the new high Reynolds number calibration facility at NMIJ are described. In this examination, the measurements are performed for five pattern pipe layouts with one or two elbows. The flow conditioners installed upstream of the flowmeter are the tube bundle type and the Mitsubishi, which are normally used in nuclear power plants. The calibration result for each flowmeter are largely different for each flow conditioner and each upstream pipe layout, except in some special cases. Moreover, the trend of the correction factor with Reynolds number is not uniform for each case. Furthermore, some differences were observed for individual flowmeters. It is recommended that the feedwater flowmeter, especially when used to perform measurement uncertainty recapture, is calibrated based on the actual pipe layout and the Reynolds number corresponding to the actual nuclear power plant conditions.

An Experimental Study on the Cooling Characteristics of Passive Containment Cooling Systems

The condensation heat transfer coefficients with noncondensable gases are measured to obtain the condensation heat transfer correlation to be applied to the design of passive containment cooling systems for light water nuclear power plants. The derived condensation heat transfer correlation is applicable in a wider range of subcooling than existing correlations. The dependence of the derived correlation on the length of condenser tubes is weak. The derived correlation based on experimental heat transfer results can be applied to the design of the passive containment cooling systems at actual nuclear power plants.

Experimental Investigation into the Effects of Decision Support Systems on Operator Performance

As human error is recognized as one of the main causes of nuclear power plant (NPP) accidents, there have been efforts to reduce and prevent human error by developing various decision support systems. Before adapting these support systems to actual NPPs, it is necessary to validate their reliability and to evaluate their effect on operator performance. Particularly for safety-critical systems such as NPPs, the validation and evaluation of decision support systems is as important as the design of good systems. Such evaluations may be carried out through a theoretical modeling or experimentation. The objective of this study is to experimentally investigate the effects of decision support systems on operator performance. Several trends were observed in the results of the experiments in this study; the support systems reduced both human error and workload, and the impact of the systems was most readily observed during complex and highly workloaded situations in which an operator must handle a high volume of i...

革新的簡素化原子力発電プラント用蒸気インジェクタ駆動静的炉心注水システムの実験と解析評価(転水炉・次世代炉,<特集>第13回動力・エネルギー技術シンポジウム)

A Steam Injector (SI) is a simple, compact and passive pump. We are developing this innovative concept by applying the SI system for core injection system in emergency core cooling systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent severe accidents by keeping the core covered with water (Severe accident-free concept). The passive core injection system (PCIS) driven by high-efficiency SI is a system that, in an accident such as a LOCA, attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI. This report describes the experiments and the analytical simulation on a SI-driven PCIS for innovative-simplified nuclear power plant. In addition, we conducted the analytical simulations of SI, which grew in size for the actual nuclear power plant.

Positron annihilation study of neutron irradiated model alloys and of a reactor pressure vessel steel

The hardening and embrittlement of reactor pressure vessel steels are of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, and positron annihilation spectroscopy has been shown to be a suitable method for analysing most of these defects. In this paper, this technique (both positron annihilation lifetime spectroscopy and coincidence Doppler broadening) has been used to investigate neutron irradiated model alloys, with increasing chemical complexity and a reactor pressure vessel steel. It is found that the clustering of copper takes place at the very early stages of irradiation using coincidence Doppler broadening, when this element is present in the alloy. On the other hand, considerations based on positron annihilation spectroscopy analyses suggest that the main objects causing hardening are most probably self-interstitial clusters decorated with manganese in Cu-free alloys. In low-Cu reactor pressure vessel steels and in (Fe, Mn, Ni, Cu) alloys, the main effect is still due to Cu-rich precipitates at low doses, but the role of manganese-related features becomes pre-dominant at high doses.

B112 革新的簡素化原子力発電プラント用蒸気インジェクタ駆動静的炉心注水システムの実験と解析評価(軽水炉,OS-8 軽水炉・新型炉・核燃料サイクル(1),一般講演,地球温暖化防止と動力エネルギー技術)

A Steam Injector (SI) is a simple, compact and passive pump. We are developing this innovative concept by applying the SI system for core injection system in emergency core cooling systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent severe accidents by keeping the core covered with water (Severe accident-free concept). The passive core injection system (PCIS) driven by high-efficiency SI is a system that, in an accident such as a LOCA (loss of coolant accident), attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI using water stored in the pool as the water supply source and steam contained in the reactor as the source of pressurization energy. This report describes the experiments and the analytical simulation on a SI-driven PCIS for innovative-simplified nuclear power plant. In addition, we conducted the analytical simulations of SI, which grew in size for the actual nuclear power plant.

Positron annihilation study of neutron irradiated pure Fe and Fe‐Cu binary alloys

AbstractThe hardening and embrittlement of Reactor Pressure Vessel (RPV) steels is of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation‐induced damage, like vacancies, interstitials, solutes and their clusters. Fe‐Cu binary alloys are often used to mimic the behaviour of such steels. Their study allows identifying some of the defects responsible of the hardening, especially when compared to pure iron or C‐micro‐alloyed iron. Owing to their self‐seeking and selective trapping, positrons are used to determine the nature of these defects. Recently, at SCK·CEN, a new Positron Annihilation Spectroscopy (PAS) setup has been built, calibrated and optimized to measure the Coincidence Doppler Broadening (CDB) and Lifetime (LT) of neutron irradiated materials. This set‐up has been used to measure the CDB and LT of n‐irradiated pure Fe and Fe–Cu alloys. It is found that the clustering of Cu take place at the very early stages of irradiation using the CDB while LT measurement are showing much more vacancy clustering for low Cu alloys than in the higher ones. Increasing the neutron dose up to 1.3 × 1020 n/m2, allows the follow up of the kinetic of Cu and V‐clustering especially in Fe‐Cu alloys. It is found that both copper and carbon decrease the size of vacancy‐cluster, when added to iron. (© 2007 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

An analytical approach to quantitative effect estimation of operation advisory system based on human cognitive process using the Bayesian belief network

The design of instrumentation and control (I&C) systems for nuclear power plants (NPPs) is rapidly moving towards fully digital I&C systems and is trending towards the introduction of modern computer techniques into the design of advanced main control rooms (MCRs) of NPPs. In the design of advanced MCRs, human–machine interfaces have improved and various types of decision support systems have been developed. It is important to design highly reliable decision support systems in order to adapt them in actual NPPs. In addition, to evaluate decision support systems in order to validate their efficiency is as important as to design highly reliable decision support systems. In this paper, an operation advisory system based on the human cognitive process is evaluated in order to estimate its effect. The Bayesian belief network model is used in the evaluation of the target system, and a model is constructed based on human reliability analysis event trees. In the evaluation results, a target system based on the operator's cognitive process showed better performance compared to independent decision support systems.