Korea Standard Nuclear Power Plant Research Articles

Loading pattern optimization using simulated annealing and binary machine learning pre-screening

We introduce a creative approach combining machine learning with optimization techniques to enhance the optimization of the loading pattern (LP). Finding the optimal LP is a critical decision that impacts both the reload safety and the economic feasibility of the nuclear fuel cycle. While simulated annealing (SA) is a widely accepted technique to solve the LP optimization problem, it suffers from the drawback of high computational cost since LP optimization requires three-dimensional depletion calculations. In this note, we introduce a technique to tackle this issue by leveraging neural networks to filter out inappropriate patterns, thereby reducing the number of SA evaluations. We demonstrate the efficacy of our novel approach by constructing a machine learning-based optimization model for the LP data of the Korea Standard Nuclear Power Plant (OPR-1000).

Evaluation of LBB characteristics of candidate materials for main steam line piping in Korea nuclear power plants

Recently, efforts have been made to apply the leak-before-break (LBB) design concept to the main steam line (MSL) piping in secondary systems of Korea standard nuclear power plants (KSNPs). In this study, the microstructures, mechanical properties, and consequential LBB margins of five candidate materials (SA335 Gr.P11/P22, SA508 Gr.3 Cl.2, SA508 Gr.3 Cl.1, and SA508 Gr.1A steels) were evaluated, and strong/weak points of candidate materials for applying MSL piping were analyzed through the comprehensive analysis including microstructure, mechanical properties, LBB margin, usage history, economics, and so on. SA508 Gr.1A steels, which has been used extensively as a piping material in NPPs and has a low fabrication cost, had a good combination of yield strength and fracture toughness. They had LBB margin of 1.11–1.21. SA508 Gr.3 Cl.1 steel had an excellent combination of yield strength and fracture toughness with an LBB margin of 1.64. SA508 Gr.3 Cl.2 steel had the highest LBB margin (1.84) because it had the highest strength; however, it had low fracture toughness and low upper shelf energy (USE). The effect of yield strength on the LBB margin was considerably higher than the effect of fracture toughness. SA508 Gr.3 Steels have high LBB margin owing to the high yield strength, but they have never been used for piping materials in NPPs before. SA335 Gr.P11 and P22 had the LBB margin of 1.1 and 1.23, respectively. However, the elongation and J-R fracture toughness of SA335 Gr.P11 and P22 decreased significantly at 286 °C owing to dynamic strain aging (DSA).

Environmental fatigue crack growth rate of Type 347 austenitic stainless steel in simulated PWR water conditions

The pressurizer surge line in a pressurized water reactor (PWR) suffers from environmental fatigue damage, which may accelerate the fatigue crack growth. Currently, ASME code section XI provides a reference curve for the fatigue crack growth rate (FCGR) of stainless steel (SS) only under air conditions, and those curves under PWR water conditions are prescribed in ASME Code Case N-809. However, data for Type 347 stainless steel (347SS), which is used for pressurized surge lines in a Korea Standard Nuclear Power Plant (KSNPP), are rare, and most researches were focused on 304SS, 316SS, and cast austenitic stainless steel (CASS). In this study, the FCGRs of 347SS are evaluated using compact-tension (CT) specimens under simulated PWR water conditions with different levels of dissolved oxygen (DO), dissolved hydrogen (DH), and frequency (Hz). When the DO level increases and the frequency decreases, FCGR of 347SS is accelerated owing to more corrosion fatigue effects. These data are compared with the reference curves of the ASME XI, ASME Code Case N-809, and Japanese models. In this study, FCGRs of 347SS are slightly faster than the reference curves of the ASME Code Case N-809 and Japanese model under PWR water conditions.

Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4in and by only a high pressure safety injection in LOCAs of 0.8–9.4in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety assessment, but also a more realistic core damage frequency of the LOCAs using updated information. The difference between the results of the present study and the conventional knowledge can be used to modify the current emergency operator procedures and to design another nuclear power plant type.

Safety assessment of a nuclear power plant building subjected to an aircraft crash

This paper presents a safety assessment of a nuclear building subjected to an aircraft crash using numerical analysis. For impact simulation, the reinforced concrete (RC) Primary Auxiliary Building (PAB) of the Korea Standard Nuclear Power Plant (KSNP) is fully modeled and an aircraft model of a Boeing 767-400 is used. The Riera function is used to verify the analysis result of impact force–time history. The IRIS test is used to verify the structural behavior of the RC wall under impact loading. The safety assessment of the building is performed with regard to different impact scenarios. The safety of the nuclear building under aircraft crash, including (1) global structural safety, (2) local structural safety, and (3) vibration safety are evaluated and discussed. The results show that the global and local structural safety of the PAB is ensured in all impact scenarios. However, the vibration safety of the building is not ensured. In accordance, the regulatory guide of United States Nuclear Regulatory Commission (U.S. NRC), shutdown of the nuclear power plant is required.

The comprehensive nuclear promotion plan of the Republic of Korea

Access to economic and sustainable energy is essential in every country. Rep. of Korea has insufficient supply of energy resource. Accordingly, Rep. of Korea has considered nuclear energy as the main energy source since 1970’s oil crises. In 1994, Rep. of Korea declared “Direction to Long-term Nuclear Energy Policy towards the Year 2030” and revised the Atomic Energy Act to promote the safe and peaceful use of nuclear energy.According to this Act, the government made the CNEPP (comprehensive nuclear energy promotion plan) for every 5 years since 1997. This CNEPP shall include the analysis of current status and prospect, policy targets, implementation tasks, plan to invest, and other matters for the nuclear utilization and safety. National nuclear technology has advanced a lot by 1st (1997–2001), 2nd (2002–2007), and 3rd CNEPP (2007–2011). Through these previous plans, Rep. of Korea designed its own nuclear reactor, Korea Standard Nuclear Power Plant (KSNP) and became the fifth nuclear reactor exporter after the United States, France, Russia and Canada.The Rep. of Korea has been preparing for the 4th CNEPP to advance further stage and to harmonize with the long-term policy direction. In this study, detailed process of the establishment of the 4th CNEPP was analyzed. By the analysis of the domestic and international environmental change, the achievement of the 3rd CNEPP and the major issue of the 4th CNEPP were proposed. Six promotion areas were set in the 4th CNEPP as nuclear utilization, sustainability, export/growth force, radiation, safety/public acceptance, and infra/international cooperation. It will be discussed by various social, economic, political, cultural, and technical professionals to harmonize with the national vision for the future, short- and long-term plans. By doing this, it will help to improve the nuclear safety, transparency, and effectiveness in the promotion of national nuclear technology.

한국 표준형 원자력 발전소 증기터빈 보호 및 제어를 위한 운전인자 선정과 운전반 운영

This paper contains the selection of operation parameters for protection and control of steam turbine in a Korea Standard Type Nuclear Power Plant. The safety of nuclear reactor must be ensured which generates nuclear energy and produces steam. Also, the safety of turbine, which consume the nuclear energy as a core machine, must be ensured. For the purpose of this, we describe how the operating parameters were selected, reviewed, implemented into the operator console and finally put into actual operation of the system.

Current Status of Radioactive Waste Management in Korea

Korean energy security is dependent on the peaceful and safe utilization of nuclear energy. The Korean government’s policy on nuclear energy can be summarized with two principles: promoting the peaceful uses of nuclear energy and ensuring its safe use and storage. Since April 1978 Korea has strongly relied on nuclear energy for its electricity generation. As of today, 20 nuclear power plants, 16 pressurized water reactors PWRs and 4 Canada Deuterium Uranium CANDU reactors are in operation, and 4 PWRs are under construction. The installed nuclear capacity is 17.7 GWe, representing 27% of the nation’s total installed capacity. The nuclear share in electricity remains at around 37.8%, reaching a level of 146.8 TWh. New power reactors in Korea Standard Nuclear Power Plants are fully designed by domestic technologies. A more advanced reactor named AP1400 will be commissioned in 2020. The Achilles’ heel of nuclear energy is radioactive waste. Even though proper scientific management of radioactive waste has been found and proved safe for lowand intermediate-level waste LILW disposal and an interim storage of spent nuclear fuel and even the introduction of permanent disposal of high-level radioactive waste HLW is envisaged to be feasible, the public perception on these issues is not supportive. The public’s everlasting “not in my backyard” refrain has hindered the implementation of Korea’s radioactive waste management program for the last two decades. Overcoming the safety issues involved in radioactive waste management is not a simple scientific problem. It requires attention to both state-of-the-art science and technology and the psychology of the general and local public. In fact, it is a transscientific issue, a fusion of science and socioeconomic studies. The Korean Atomic Energy Commission provided clear milestones for constructing two key facilities: the LILW repository by 2008 and the centralized storage for spent fuel by 2016. With the full support of the Korean government, institutes, and general public, Korea Hydro Nuclear Power KHNP secured a site for the LILW repository in 2005. An underground repository with the capacity for 100,000 drums is under construction, and will be in commission by 2012. Long-term safe management of spent nuclear fuel is necessary not only to store spent fuel over a long period of time but also to potentially recycle it under full observance of the nuclear nonproliferation duties. The permanent disposal of HLW is the research and development R&D topic for a future generation and involves protecting the environment and minimizing the economic burden. Since 1997 Korea Atomic Energy Research Institute KAERI has been developing a permanent disposal facility for HLW and a total system performance assessment TSPA . Its current R&D activities are focused on the preliminary conceptual design of the Korean Reference Disposal System KRS , development of the

Radiation release characterization of PWR spent fuel assemblies generated from Korean nuclear power plants

Spent nuclear fuel should be kept under safe management until it is disposed of permanently. Because of this, it is important to understand its radiation release characteristics. In this paper, the Monte Carlo method is applied to evaluate the radiation release characteristics of two types of PWR spent fuel assembly generated from the operating plants in Korea: Westinghouse and Korea Standard Nuclear Power Plant. The source terms were calculated using ORIGEN-ARP. The neutron and photon (or gamma) dose distributions along the vertical and horizontal directions of each spent fuel assembly were evaluated using MCNPX code. Compared with the two dose distributions, the photon dose was found to be about 105 times higher than the neutron dose.

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.

A Study on an Accident Diagnosis Methodology Using Influence Diagrams

In nuclear power plants, operators are allowed to follow EOPs (Emergency Operating Procedures) when reactor tripped because of accidents. But, it’s very difficult to diagnose accidents and find out appropriate procedures to mitigate current accidents in a given short time. Even if they diagnose accidents quickly, it also has possibility to misdiagnose. Methodology using Influence Diagrams has been developed and applied for representing the dependency behaviors and uncertain behaviors of complex systems. An example to diagnose the accidents such as SLOCA and SGTR with similar symptoms has been introduced. From the constructed model, operators could diagnose accidents at any states of accidents. Also, The integrated design of Thermal Hydraulics Online Monitoring Advisory System(THOMAS) is introduced in this paper. THOMAS, for improved Korea Standard Nuclear Power Plant, is an advisory monitoring system that uses the digital engineering technology such as virtual reality, and database. It is believed that the new design will improve the operability and maintainability and simplify design process of the monitoring and diagnosis system of the nuclear power plant. And this model can offer the information about accidents with given symptoms. This model might help operators to diagnose correctly and rapidly. It might be very useful to support operators for reducing human error.

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.

A model predictive controller for load-following operation of PWR reactors

The basic concept of a model predictive control method is to solve on-line, at each time step, an optimization problem for a finite future interval and to implement only the first optimal control input as the current control input. It is a suitable control strategy for time-varying systems, in particular, because the parameter estimator identifies a controller design model recursively at each time step, and also the model predictive controller recalculates an optimal control input at each time step by using newly measured signals. The proposed controller is applied to the integrated power level and axial power distribution controls for a Korea Standard Nuclear Power Plant (KSNP). The power level and the axial shape index are controlled by two kinds of the five regulating control rod banks and the two part-strength control rod banks together with the automatic adjustment of boric acid concentration. The three-dimensional reactor analysis code, Multipurpose Analyzer for Static and Transient Effects of Reactor, which models the KSNP, is interfaced to the proposed controller to verify the proposed controller for controlling the reactor power level and the axial shape index. It is known from numerical simulations that the proposed controller exhibits very fast tracking responses.

On some salient unresolved issues in severe accidents for advanced light water reactors

Some of the outstanding issues of the severe accident are evaluated from the synthesis of the results from the research programs of the simulation of naturally arrested thermal attack in the vessel (SONATA), test for real corium interaction with water (TROI), and the test for the performance of a quenching screen for a hydrogen burn. The strategy of the in vessel retention (IVR) by an external reactor vessel cooling (ERVC) is evaluated for the Korea Standard Nuclear Power Plant (KSNP) and Advanced Power Reactor-1400 (APR-1400). The favorable and adverse effects of the proposed strategy and the uncertainties of the relevant phenomena including the gap cooling and heat removal capability of the EVRC are discussed. It is shown that the thermal margin for the failure of the reactor vessel is challenged for these reactors. The observation of the energetic spontaneous steam explosions in the TROI suggests that proper measures should be taken to avoid and/or lessen the risk of an in-vessel or ex-vessel steam explosion, when we employ an ex-vessel cooling strategy and/or an in vessel injection strategy. To address the issue that multiple burns induced by the hydrogen igniter can pose a threat to the safety equipment, the use of a quenching screen is suggested. The results of the laboratory scale experiments for the performance of the quenching screen indicated the feasibility of using a quenching screen to protect the essential equipment by stopping the flame propagation.

New dynamic method to measure rod worths in zero power physics test at PWR startup

To measure and validate the worth of control (or shutdown) bank in zero power physics test at PWRs, a dynamic control rod reactivity measurement (DCRM) technique has been developed and applied to six startups of Westinghouse plants as well as Korea Standard Nuclear power Plants based on the Combustion Engineering System 80 NSSS. With this technique, just one test bank is inserted into the bottom of the core at maximum stepping rate and withdrawn immediately to the all rod-out position. Specially designed inverse point kinetics equations are used to determine the test bank worth from the measured ex-core detector signals, which are controlled by the neutron-to-response conversion factor and the dynamic-to-static conversion factor. These two parameters are predetermined by the three-dimensional neutron adjoint flux distribution for both the top and bottom ex-core detector and the three-dimensional steady and transient core power distribution for test bank movement. To eliminate the gamma-ray effect on ex-core detector signals, a simple method, using reactivity curve characteristics, was also developed. To verify the DCRM method, a total of 28 bank worths of six different PWRs was measured by the DCRM and compared with those of conventional method. Results show that the DCRM method has a similar accuracy as the conventional technique. However, with the DCRM method, it only takes ∼15 min per bank from the beginning of rod insertion to the determination of measured static worth. From its performance, one can conclude that the DCRM method is an effective replacement for the conventional rod worth measurement method.

Off-take and Slug Transition at T-junction of Vertical-up Branch in the Horizontal Pipe

The off-take and the slug transition on air-water interface are experimentally investigated at the T-junction of the horizontal pipe with a vertical upward branch to simulate the loss-of-residual-heat-removal during a mid-loop operation in the Korea standard nuclear power plant. Scaling analysis is performed to scale down the experimental facility to the reference nuclear power plant. Two different diameters of branch pipes are used to verify the scaling laws and their scale effects. Air is used as working gaseous fluid and no water flow exists. Off-take behavior on horizontal stratified and slug flows is visually observed in the horizontal pipe. The experimental data are divided into three categories; onset of liquid entrainment at T-junction, onset of slug transition in the horizontal pipe, and discharge quality in the branch pipe. It is found out that the scale effect of the branch diameter on the onset of liquid entrainment is small and the existing correlations for it are applicable. Also, the onset of slug transition shows a discrepancy with Taitel-Dukler's correlation and has a strong influence on the discharge quality. New correlations for discharge quality are developed considering the critical dependency of the onset of slugging.

An Integral Response Scaling Method for Reduced-Height Test Facilities and Its Validation Using RELAP5 for Loss of RHR During Midloop Operation

An integral response scaling method for reduced-height test facilities is described and validated for the local and the integral phenomena using the RELAP5 thermal-hydraulic code. In the present schematic scaling methodology, the scaling laws are generated from four scaling analyses: three-dimensional differential equation-based scaling analysis, one-dimensional differential equation-based scaling analysis, component-based scaling analysis, and overall integral scaling analysis. Through use of integral response function, scaling parameters related to four types of requirements are identified: time scaling, initial-condition scaling, transient scaling, and bifurcation-phenomena scaling requirements. In the present scaling method, flow velocities in the vertical channel and at break locations of the scaled-down model can be preserved setting up the scaling as τR = lR. The significance of the velocity preservation in the reduced-height facility is demonstrated by the RELAP5 code.To validate the present scaling method, similarity criteria from the integral response scaling method are applied to the scaled-down model of the Korea Standard Nuclear Power Plant simulating core boil-off process during midloop operation. The scaled-down model is basically developed with the length and area scales of 1/5 and 1/100, respectively. The integral simulation and local calculations for pot-boiling, blowdown, heat transfer in the steam generator, and off-take are conducted for a loss of RHR during the midloop operation to show the preservation of the similarity criteria using RELAP5.It turns out that the scaled-down model based on the present scaling method maintains well the similarity of the nondimensional parameters for the local phenomena. Furthermore, the integral simulation indicates that there are no noteworthy differences in the general trends between the prototype and scaled models.

A methodology for benefit assessment of using in-core neutron detector signals in core protection calculator system (CPCS) for Korea standard nuclear power plants (KSNPP)

Core Protection Calculator System (CPCS) is a digital computer based safety system generating trip signals based on the calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this work, a quantitative economic benefit assessment of using in-core neutron detector signals is carried out. In-core detector signals which directly measure the inside neutron flux of core are applied to CPCS to obtain more accurate power distribution profile, DNBR and LPD to reduce the calculation uncertainties. In order to improve axial power distribution calculation, piecewise cubic spline method is applied. Simulation is also carried out to verify its applicability to power distribution calculation in this work. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also assured that no power reduction is required while Core Operating Limit Supervisory System (COLSS) is out-of-service when the improved method is applied.