Nuclear Power Plant Systems Research Articles

Preparation and Management of In-service Inspection Schedule Program in Nuclear Power Plants

In-service Inspection (ISI) schedule program for nuclear island equipment was of the important part of ISI program file system in nuclear power plants (NPPs), which acted as a connecting link. Based on ISI rules for the mechanical components of PWR nuclear islands (RSE-M) and the plant ISI general program, the preparation method for schedule program of CPR1000 units were presented. It focused on the analysis and discussion of the existing problems during the preparation, upgrade and management process and some targeted suggestions were put forward. It may lay a foundation for a further optimization of the ISI program file system and the smooth implementation of the ISI activities.

Seismic Response Amplification Factors of Nuclear Power Plants for Seismic Performance Evaluation of Structures and Equipment due to High-frequency Earthquakes

Analysis of the 2016 Gyeongju earthquake and the 2017 Pohang earthquake showed the characteristics of a typical high-frequency earthquake with many high-frequency components, short time strong motion duration, and large peak ground acceleration relative to the magnitude of the earthquake. Domestic nuclear power plants were designed and evaluated based on NRC's Regulatory Guide 1.60 design response spectrum, which had a great deal of energy in the low-frequency range. Therefore, nuclear power plants should carry out seismic verification and seismic performance evaluation of systems, structures, and components by reflecting the domestic characteristics of earthquakes. In this study, high-frequency amplification factors that can be used for seismic verification and seismic performance evaluation of nuclear power plant systems, structures, and equipment were analyzed. In order to analyze the high-frequency amplification factor, five sets of seismic time history were generated, which were matched with the uniform hazard response spectrum to reflect the characteristics of domestic earthquake motion. The nuclear power plant was subjected to seismic analysis for the construction of the Korean standard nuclear power plant, OPR1000, which is a reactor building, an auxiliary building assembly, a component cooling water heat exchanger building, and an essential service water building. Based on the results of the seismic analysis, a high-frequency amplification factor was derived upon the calculation of the floor response spectrum of the important locations of nuclear power plants. The high-frequency amplification factor can be effectively used for the seismic verification and seismic performance evaluation of electric equipment which are sensitive to high-frequency earthquakes.

Seismic analysis of soil-structure system of nuclear power plant on non-rock site via shaking table test

The influence analysis of the site condition is a key issue considered in the seismic design of nuclear power engineering on non-rock sites. To study the soil–structure interaction (SSI) effect of the site conditions on the structural seismic response of nuclear power engineering, we used a CAP1400 nuclear power plant as a case study subject and performed a shake table testing of a soil–structure system model. The test results indicate that there are strong nonlinear effects of non-rock site, that is, with increase of seismic input level (peak ground acceleration, PGA), the natural vibration frequency of the soil-structure system decreased significantly, and the damping ratio increased; Under all seismic input levels, the site amplification phenomena occurred in different directions, and the response accelerations of soil and structure in the high-frequency range (>3 Hz) are greatly influenced by the SSI effect; The cracks began to appear in the model soil under lower seismic input level (PGA 0.20 g) and developed, lengthen and widen under the high seismic input level (PGA≥0.30 g), and the soil cracks were almost all connected under the SME seismic input level (PGA 0.51 g), but still no micro-cracks were detected on the model structure; Under the SSE seismic input level (PGA 0.30 g) the contact surface separation appeared on the contact surfaces of the model soil and the structure foundation, and then under the ultra SME seismic input level (PGA 0.84 g) the structure foundation became unbalanced so that the soil-structure system failed. The influence of the site conditions on the seismic response of the structure was significant. The seismic response analysis and seismic design of the nuclear island structure on no-rock sites should take into account the effects of site conditions and soil-structure interaction.

Study on Cooling Water Source Seawater Filtration System of Nuclear Power Plant

Littoral nuclear power plants at home and abroad use seawater as cooling source, which is the cooling system of nuclear island and conventional island. In recent years, many marine organisms such as jellyfish, phaeocystis globosa, hyacinth melon and prawn have affected the safety of water intake of nuclear power plant, and even led to the event of unit load reduction operation, machine trip and even reactor trip. In this paper, by analysing the existing problems of cold source seawater filtration system in nuclear power plant, the plane layout of water intake project is optimized, and the sea biological interception filtration system is established. Relying on the large offshore platform, online monitoring of the incoming sea creatures and the status of interceptors can realize the mechanized automation of intercepting and cleaning, and improve the reliability and efficiency of intercepting and cleaning. Reduce manual intervention and temporary network layout, reduce the risk of marine personnel operation, and improve the ability to withstand severe weather and large-scale outbreak of marine organisms. It ensures the safe operation of nuclear power plant.

Simulation research on passive safety injection system of marine nuclear power plant based on compressed gas

In case of a large or medium LOCA accident of marine nuclear power plant, the loop pressure drops rapidly. In case of power failure of the whole ship at this time, the active safety injection system will lose its function because the safety injection pump needs AC power, and the core is at risk of melting. A passive marine safety injection system based on high-pressure compressed gas is designed for the superposition of large and medium LOCA accidents and the whole ship's power failure accidents, including the safety injection subsystem of medium and low pressure and the safety injection subsystem of recirculation. The energy required for safety injection is provided by the potential energy of compressed gas, the pressure head is provided by the technology of gas-liquid booster pump, and the collection signals and valves are controlled by the technology of intelligent fluid control unit with small power consumption. The power supply used in the system is provided by DC 24 V battery. The mathematical and physical models of accumulator, gas storage tank, gas-liquid booster pump, reactor body and its pipeline are established for the designed passive safety injection system. The operation and safety characteristics of medium pressure safety injection, low pressure safety injection and recycling safety injection under LOCA accident conditions are simulated. The safety injection time and flow rate of three safety injection pressures are simulated by medium pressure safety injection, the safety injection time and peak temperature are simulated by low pressure safety injection, and the pressure and system temperature are simulated by recirculation safety injection. The simulation results show that the marine passive safety injection system designed under LOCA accident conditions can meet the requirements of core cooling.

Improvement and validation of the heat transfer model for the containment module in the RCS-containment integral system code

Owing to improvements in containment systems of nuclear power plants, the coupling analysis of reactor system codes and containment codes is increasingly important for the thermal–hydraulic analysis of existing and future reactors. Extending reactor system code capabilities to allow them to simulate containment phenomena is of considerable current interest. In this paper, constitutive models for the containment module will be introduced in detail, with special focus on the flow regime for large spaces and the wall condensation model in the presence of noncondensable gases. The improved constitutive models have been validated by comparing simulation results with available experimental data from wall and in-pipe steam-noncondensable gas experiments and the PUMA GDLB test. The calculation results showed that the improved constitutive models for the containment module better predict the experimental phenomena with higher accuracy.

Innovative Design and Implementation of Jet Cooling of Large UPS Powered Systems in Nuclear Power Plant

In this article, it is imperative that an efficient and reliable cooling system is used for the large Uninterrupted Power Supply (UPS) panel in Nuclear Power Plant. While the UPS system can tolerate relatively high temperatures, it is advisable to keep the temperature at 25 °C or below in the room where the UPS is housed. A rule of thumb is that for every 10 °C increase in operating temperature, the expected life of the UPS components would be reduced in half. So it is absolutely necessary to maintain the temperature within the limits prescribed by the designer. The aim of this article is to devise a vastly improved cooling system by streamlining the cool air flow and targeting the hotspots of the UPS components which are particularly vulnerable to damage from high temperature. First, a computational model of the whole setup was made and then, computational fluid dynamics (fluent) technique was used to perform fluid flow and heat transfer analysis of this setup, which showed a dramatic performance increase in the heat exchange process and reduced the effective size of the UPS cabin and cost.

Study on Safety Analysis of Large Angle Rolling Test Facility for Motion Conditions

The large angle rolling test device for motion conditions is an Electromechanical-hydraulic integrated test facility, which provides a simulation environment for nuclear safety related verification test and basic technical research in laboratory. The test device is complex in composition. The method of identifying hazards from physical hazards and human behavioral hazards is used to find out the risk factors. Therefore, the test device has guiding significance for the safety and reliability design of nuclear power plant system, as well as for the personal security of operators, the configuration of test site, and the formulation of safety plans.

Island Operations of Electric Generators Connected to One Substation

<span lang="EN-GB">While increasing integration of renewable energy sources (RES), which are unregulated and difficult to predict, a large system of nuclear power plants must provide balancing peaks in the production of renewable energy. </span><span><span lang="EN">It is also important to simulate </span></span><span lang="EN-GB">the rapid changes in the power of individual </span><span lang="EN">large Nuclear Power Plant</span><span lang="EN-GB"> (NPP) units, and for these regimes to train operators of nuclear units. Therefore the paper is aimed to i<span>sland operations of more parallel electric</span></span><span><span lang="EN">synchronous generators</span></span><span lang="EN"> c<span>onnected to one</span> s<span>ubstation of a power grid.</span></span>

Orientation effect on heat transfer coefficient of a downward surface for flow boiling in a rectangular channel under low flow rate

Natural convection boiling in channel with Arc-Shaped will be encountered in the IVR-ERVC (In-Vessel Retention measure by External Reactor Vessel Cooling) system in nuclear power plant under severe accident. The flow and heat transfer characters in this situation is simulated by flow boiling of deionized water in an inclined rectangular channel under low flow rates. This paper aims to separate various parameters (such as orientation, mass flow rate and inlet quality, etc.) to investigate their individual effects on heat transfer coefficient (HTC) in a rectangular channel with cross section of 17 mm × 10 mm. By using a preheater at the inlet of the rectangular channel, the inlet quality could be controlled and the two-phase flow situation could be observed before the fluids entering into the main heater region on one side of the channel wall in downstream. Thus the characteristics of HTC on the main heater could be investigated at different flow patterns. The channel orientations vary from 15 to 90°, the mass flow rates vary from 110 to 288 kg/(m2s) and the qualities vary from 0.003 to 0.036, respectively. Experimental results show that the mass flow rate and quality effects on the HTC are very weak in this study. However, the orientation angle effect on HTC shows an transition region within 45°~60°, while it slowly changes when the orientation angle is smaller than 45° and bigger than 60°. Such tendency could be well formulated by the error function. Compared with different empirical formulas of saturated boiling HTC, it is found that the Liu & Winterton correlation can well predict the experimental HTC results in 90° orientation channel. Based on such correlation and coupled with the error function, a new model was developed by considering the orientation effect, which has an error of ±15% comparing with the experimental data.

A control-theoretic approach to detecting and distinguishing replay attacks from other anomalies in nuclear power plants

The wider use of digital systems in nuclear power plants has raised security concerns to the utilities, the regulatory entities, and the public. These digital systems provide malicious attackers with opportunities to exploit vulnerabilities in these systems to realize physical damages. Replay attacks are a particular type of attacks. They are easy to perform and when combined with other types of attacks can potentially lead to significant consequences. Therefore, timely detection of a replay attack is of vital importance for the operator to take mitigation measures. Besides, it is also important to distinguish between a replay attack and other anomalies, since they are of different nature (one intentional and the other random) and will require different responses by the operator. In this research, we propose a control-theoretic method to address this problem. The method consists of the injection of random noise (i.e. physical watermark) into the control input and two chi-squared tests. The physical watermark is used to excite the system being controlled so that the replay attack, if it exists, can be uncovered. The first chi-squared test is used to detect anomalies in the system, including replay attacks and other anomalies. If the null hypothesis in the first test is rejected, we use the second chi-squared test to determine whether the anomaly is a replay attack or any other anomaly. We demonstrate the proposed method by using a steam generator that can be found in pressurized water reactors. The results of different cases are presented and discussed.

A quantitative assessment framework for cyber-attack scenarios on nuclear power plants using relative difficulty and consequence

Digital instrumentation and control (I&C) systems have been replacing analog I&C systems in nuclear power plants (NPPs) on account of the beneficial functions digital systems provide. As a result, NPP risks have been centralized into the digital I&C systems, with the issue of cyber-attack thus emerging as one of the new threats. To achieve more effective cyber security, it is necessary to evaluate the risk of cyber-attack quantitatively. This work proposes a quantitative assessment framework to evaluate NPP risk due to cyber-attack scenarios. This framework evaluates the risk by defining the difficulty and consequence of a cyber-attack, for which the assessment methods are based on Bayesian belief network and probabilistic safety assessment methods, respectively. Application of the framework to several possible cyber-attack scenarios quantitatively assessed the difficulty and consequence of the scenarios, thereby providing risk information. It is expected that application of the proposed framework can enhance cyber security.

Orifice enabled flow stabilization of natural circulation loop at lower inclinations

Abstract Natural circulation loop (NCL) is mainly used as passive cooling system in nuclear power plants to ensure safety during pump failure and station blackout. Stable operation without flow instability is the most desirable operational mode of natural circulation loop for passive cooling. The present work deals with numerical analysis on the effect of inclination of horizontal leg of a rectangular loop working in single phase and presence of orifice on the thermohydraulic behaviour of NCL, using commercial CFD code ANSYS FLUENT 19.0. The effect of temporal variation of loop mass flow rate and temperature difference across the cooler section is compared for inclined loop with and without orifice. It is observed that the flow stability attained in the loop at inclination h = 308 can be achieved at h = 108 with orifice, with a reduction in the loop mass flow rate of 28.57%. The reduction in loop mass flow rate is accompanied by an increase in temperature difference across the cooler of 46.67%.

ОЦІНЮВАННЯ КІЛЬКОСТІ ВТОРИННИХ ДЕФЕКТІВ ПРОГРАМНИХ ЗАСОБІВ ШЛЯХОМ КОМПЛЕКСУВАННЯ МОДИФІКОВАНИХ МОДЕЛЕЙ РОСТУ НАДІЙНОСТІ ДЖЕЛІНСЬКІ-МОРАНДИ І ШИКА-ВОЛВЕРТОНА

Виконано аналіз множин моделей оцінювання надійності програмних засобів (МНПЗ) або за іноземною назвою моделей зростання надійності ПЗ (Software Reliability Growth Models – SRGM). Досліджено ймовірнісні МНПЗ з метою встановлення таких, що можуть бути використано для врахування фактору прояву вторинних дефектів. Під вторинними дефектами розуміються такі, що вносяться в ПЗ після усунення первинних, які проявляються та усуваються в процесі налагодження та тестування ПЗ. Проаналізовано припущення та аналітичні вирази моделей росту надійності програмних засобів Джелінські-Моранди, Шика-Волвертона, щодо питання їх застосування для врахування фактору внесення та прояву вторинних дефектів проектування програмних засобів. Запропоновано підхід щодо кількісної оцінки вторинних дефектів, що полягає в модифікації функцій ризику моделей внесенням до них параметра, який визначає число вторинних дефектів та комплексування модифікованих функцій ризику. Проаналізовані проблемні питання, що виникають при комплексуванні модифікованих простої експоненціальної моделі і моделі Джелінські-Моранди. Розглянуті підходи, при яких можливе комплексування модифікованих МНПЗ – узгодженість припущень, прийняття додаткових припущень, що узгоджують моделі, співставлення параметрів щодо умов здійснення аналітичних перетворень. Показані переваги комплексування модифікованих моделей росту надійності Джелінські-Моранди і Шика-Волвертона у порівнянні з комплексуванням модифікованої моделі росту надійності Джелінські-Моранди і модифікованої простої експоненціальної моделі. Проведено співставлення параметрів модифікованих моделей росту надійності Джелінські-Моранди і Шика-Волвертона, в результаті чого виявлено співпадання більшості з них. Додано припущення моделі росту надійності Шика-Волвертона про пропорційність функції ризику тривалості тестування, що не відповідає припущенню МНПЗ Джелінські-Моранди, оскільки відповідний параметр використовується в аналітичних перетвореннях при комплексуванні моделей. Обґрунтована можливість комплексування МНПЗ Джелінські-Моранди і Шика-Волвертона. Показана послідовність аналітичних перетворень об’єднаної моделі модифікованих моделей Джелінські-Моранди і Шика-Волвертона на основі яких одержана формула для оцінювання кількості вторинних дефектів програмних засобів. Одержаний вираз спрощує оцінювання кількості вторинних дефектів ПЗ, у порівнянні з їх оцінюванням на основі комплексування модифікованих моделей Джелінські-Моранди і простої експоненціальної моделі. Одержана формула у поєднанні з МНПЗ інших класифікаційних ознак дозволяє спрогнозувати значення функції ризику та далі використовувати її для комплексного оцінювання показників надійності та функціональної безпеки складних систем, у тому числі систем, які можливо віднести до класу критичних (наприклад, програмно-технічних комплексів інформаційно-керуючих систем АЕС)

Inclusion of Aircraft Crash into NPP Design Bases and Probabilistic Justification of Loads on Civil Structures and Equipment

The article is about nuclear power plant (NPP) safety analysis in case of aircraft crash. Specifically, the article considers the following problems: inclusion of aircraft crash into NPP design bases regarding calculation of frequency of an aircraft crash into NPP; aspects of justification of loads on NPP structures, systems and components (SSCs) caused by mechanical action of a primary missile – aircraft fuselage impact.
 Probabilistic characteristics of such random parameters as frequency of aircraft crash and direction of aircraft trajectory are determined by the results of analysis of world statistics of aviation accidents. Method of calculation of aircraft crash frequency on structures, buildings and NPP as a whole is presented. It takes into account options of accidental and intentional aircraft crashes and various aircraft approach scenarios.
 Procedure of probabilistic justification of loads on civil structures under aircraft impact is described. The loads are specified so as not to exceed allowable value of failure probability of NPP as a whole. Calculation of failure frequency of civil structures of existing NPP is given as an example to show analysis in case of a crash of an aircraft heavier than considered in NPP design.
 Procedure of probabilistic justification of dynamic loads on NPP equipment in case of aircraft impact is described. Method of floor response spectra (FRS) calculation with the required non-exceedance probability is given. Probabilistically justified loads in case of intentional aircraft impact (act of terrorism) are also considered. Additionally it is presented how internal forces calculated with the use of FRS with the required non-exceedance probability can be summed to provide analysis of subsystems.

Low Cycle Fatigue Behavior of Elbows with Local Wall Thinning

There have been a number of studies concerning the integrity of high-strength carbon steel pipe elbows weakened by local pipe wall thinning, the latter can be typically caused by flow accelerated erosion/corrosion. In particular, the focus of several recent studies was on low cycle fatigue behavior of damaged elbows, mainly, in relation to strength and integrity of piping systems of nuclear power plants subjected to extreme loading conditions, such as earthquake or shutdown. The current paper largely adopts the existing methodology, which was previously developed, and extends it to copper-nickel elbows, which are widely utilized in civil infrastructure in seismically active regions. FE (finite element) studies along with a full-scale testing program were conducted and the outcomes are summarized in this article. The overall conclusion is that the tested elbows with various severity of local wall thinning, which were artificially introduced at different locations, demonstrate a strong resistance against low cycle fatigue loading. In addition, elbows with wall thinning defects possess a significant safety margin against seismic loading. These research outcomes will contribute to the development of strength evaluation procedures and will help to develop more effective maintenance procedures for piping equipment utilized in civil infrastructure.

Qualification of Hardware Description Language Designs for Safety Critical Applications in Nuclear Power Plants

Field-programmable gate-array (FPGA)-based intelligent hardware modules are increasingly being used in safety systems of nuclear power plants. Qualification of these modules as per safety standards such as IEC 62566/60880 and IEEE-7.4.3.2-2010 needs considerable effort. Many of the safety standards demand high rigor in verifying that the designs of these modules meet the design intent. Use of hardware description languages such as VHDL or Verilog makes the process of code review and verification difficult due to the complex nonsequential semantics of these languages. It is now recognized that formal verification offers a complementary approach to conventional verification. Formal verification tools perform analysis of designs based on language semantics to prove/refute their functional correctness. In this article, we present the architecture of a formal verification tool for VHDL designs and our experience of using this tool on VHDL designs in nuclear applications.

The Conception of NPP Equipment Monitoring Based on Nonparametric Statistics and Trained Neural Network

The article deals with the concept of monitoring based on nonparametric statistics. It is supposed the possible application to monitor the technological systems of nuclear power plants. Modern complex plants and technogenic objects are equipped with industrial monitoring systems. The evolution of such systems in natural form is the generation of alternative concepts, constructs, and items selection for efficiency during operation

Investigation on effects of Fluid-Structure-Interaction (FSI) on the lubrication performances of water lubricated bearing in primary circuit loop system of nuclear power plant

The paper investigates the lubrication performances of water film under the consideration of different lubrication models in the primary circuit loop system of nuclear power plant. Variations between lubrication performances and rotating speeds, eccentricity ratios and lubrication models are obtained. Research results show that rotating speeds and lubrication models have significant influences on the lubrication performances. There exist optimum eccentricity ratios at which the bearing load carrying capacity and friction characteristics can strike the balance point. Research conclusions have guiding significances on the structure designs and optimization for such bearings in the primary circuit loop system.

2D PIV study of flow accelerated corrosion downstream a typical industrial gate valve

A critical degradation mechanism of piping systems in either nuclear power plant is Flow Accelerated Corrosion (FAC), which causes wall thinning of pipes and component shells from inside. The valves which are frequently used in power plant piping systems are mostly subjected to FAC. A Gate vale in the feed water piping system of a typical nuclear power plant that it’s downstream had been damaged due to the FAC was considered as a reference. In this study the effect of the mean flow velocity and the opening percentage of gate valve on the flow field downstream the valve was investigated by using 2D PIV. The locus where flow turbulences happen downstream the gate valve was determined by 2D PIV. It is found that there is a very good coincidence between the locus in which the maximum wall thinning measured by Ultrasonic Technique (UT) and that the flow turbulences happen downstream the gate valve determined by 2D PIV.It is observed that the locus of flow turbulences and intense velocity gradients is in the upper part of the pipe and over the domain of 0.2 <L/D < 2.4 along the pipe downstream the gate valves where the minimum wall thickness was measured by using UT. Flow turbulences and velocity gradients near the upper pipe wall result in higher wall shear stress and maximize the mass transfer from the pipe inner wall in turn result in wall thinning. The results of this paper can be useful for optimizing inspection plans in identifying and predicting the vulnerable FAC locations downstream the gate valves.