Korean Standard Nuclear Power Plant Research Articles

Mixing of Hydrogen-Steam Buoyancy Jets Released into the Upper Atmosphere of the Containment Building

This study evaluates the spatial dilution of hydrogen concentration caused by steam-hydrogen buoyancy jets rising through the open top of the steam generator compartment during a loss-of-coolant accident in the OPR1000, the Korean Standard Nuclear Power Plant. The correlation of the concentration decay rate in the plume with relatively high buoyant flux was applied to estimate the hydrogen concentration in the rise distance of the buoyant jet. The MELCOR code was used to calculate the gas composition and discharge flow rate in the ruptured cold leg during the rapid cladding oxidation to determine the volume and buoyant fluxes that affect the mixing behavior. The concentration decay rate at the plume’s center decreases as the steam-hydrogen binary buoyant jet rises. Despite the assumed initial volume flux and simplified jet nozzle geometry, the decay rate correlation can assess conservatively the diluted hydrogen in a severe accident.

Loading pattern design and economic evaluation for 24-month cycle operation of OPR-1000 in Korea

Due to the tightened regulatory environment since the Fukushima accident, the capacity factor of Korean nuclear power plants has been declining since 2011. To overcome this circumstance, a shift from 18-month to 24-month cycle operation is being considered in Korea. Therefore, in this study, loading patterns(LPs) for 24-month cycle operation of the Korean standard nuclear power plant(OPR-1000) are suggested and economic evaluations are performed. A single-zone LP with 89 fresh fuels was evaluated to be optimal for 24-month operation of OPR-1000 in terms of economic gain. The 24-month operation of OPR-1000 with this LP gives a profit of 7.073 million dollars per year compared to 18-month operation.

Dynamic response of a fuel assembly for a KSNP design earthquake

Using data from the design earthquake of the Korean standard nuclear power plant, seismic analyses of a fuel assembly are conducted in this study. The modal characteristics are used to develop an input deck for the seismic analysis. With a time history analysis, the responses of the fuel assembly in the event of an earthquake are obtained. In particular, the displacement, velocity, and acceleration responses at the center location of the fuel assembly are obtained in the time domain, with these outcomes then used for a detailed structural analysis of the fuel rods in the ensuing analyses. The response spectra are also generated to determine the response characteristics in the frequency domain. The structural integrity of the fuel assembly can be ensured through this type of time history analysis considering the input excitations of various earthquakes considered in the design.

Effect of Post-Weld Heat Treatment Conditions on Mechanical Properties, Microstructures and Nonductile Fracture Behavior of SA508 Gr.1a Thick Weldments

This study analyzes the effects of post-weld heat treatment (PWHT) on the mechanical properties and microstructures of SA-508 Gr.1a welds and proposes a new PWHT exemption criterion based on nonductile fracture evaluation considering welding residual stress. The welding coupons were prepared with submerged-arc welding, gas-tungsten arc welding, and shielded-metal arc welding, using ferritic steel, SA-508 Gr.1a. The microstructure of the heat-affected zone (HAZ) was analyzed using optical microscopy, electron-back-scatter diffraction and Vickers hardness testing. The mechanical properties of the welds were evaluated by uniaxial tensile test, transverse side bend test, Charpy V-notch impact test and side bend test. Bainite and ferrite structures formed mainly in the HAZ, and the grain size became coarser with proximity to the surface and fusion line. The mechanical properties did not depend strongly on PWHT, weldment thickness or welding techniques, and they satisfied the welding procedure qualification test specified in the ASME Boiler & Pressure Vessel code. Welding residual stresses were considered in assessing structural integrity using nonductile fracture evaluation. A margin of safety against nonductile fracture with residual stress was calculated for Korean Standard Nuclear Power Plant steam-generator welds, using its design parameters and operating conditions, and this safety margin is suggested as an acceptance criterion for residual stress for exemption from PWHT.Graphic abstract

Generation of Design Time History Complying With Japanese Seismic Design Standards for Nuclear Power Plants

Seismic designs for Korean nuclear power plants (NPPs) under earthquakes’ design basis are noticed due to the recent earthquake events in Korea and Japan. Japan has developed the technologies and experiences of the NPPs through theoretical research and experimental verification with extensively accumulated measurement data. This paper describes the main features of the design-time history complying with the Japanese seismic design standard. Proper seed motions in the earthquake catalog are used to generate one set of design time histories. A magnitude and epicentral distance specify the amplitude envelope function configuring the shape of the earthquake. Cumulative velocity response spectral values of the design time histories are compared and checked to the target response spectra. Spectral accelerations of the time histories and the multiple-damping target response spectra are also checked to exceed. The generated design time histories are input to the reactor building seismic analyses with fixed-base boundary conditions to calculate the seismic responses. Another set of design time histories is generated to comply with Korean seismic design procedures for NPPs and used for seismic input motions to the same reactor containment building seismic analyses. The responses at the dome apex of the building are compared and analyzed. The generated design time histories will be also applied to subsequent seismic analyses of other Korean standard NPP structures.

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.

Development of experimental metal sphere-mass map of reactor vessel scale models

Loose parts in coolant systems of a nuclear power plant (NPP) can damage structural integrity of the systems and loose part signals monitoring systems (LPMS) are installed in most of NPPs. LPMS detect impact signal of a loose part and generate alarm if amplitude of the signal is over preset level. If impact signal detected, mass estimation of the loose part must be carried out in order to see the effect of the part impact on structural integrity. A metal sphere-mass map, which shows relations between amplitude and center frequency of impact signals according to mass, is essential in mass evaluation of a loose part and the map built for US NPPs has been used so far. For the purpose of precise impact analysis suitable for Korean NPPs, this work discusses the development of a mass map of Korean Standard Nuclear Power Plant (KSNP) based on experimental data for scale models of a KSNP reactor vessel as tests for real NPP vessels under operation is actually impossible. Impact tests were carried out for scale models by using metal sphere with various mass and velocity. Amplitudes and center frequencies of impact signals were investigated based on time-frequency analysis and the mass map for each models were constructed based on the results. Blind tests were additionally implemented with metal sphere with unknown mass and velocity and verified feasibility of the proposed metal sphere mass map.

Analysis of Open-Phase Conditions in Korean Standard Nuclear Power Plants and Fault Detection Scheme

Because of partial losses of power known as “open-phase faults”, controlled reactor shutdown has occurred in nuclear power plants around world. Depending on transformer wiring and loading, such conditions can be difficult to detect, because phase voltage measurements can appear to be normal. Many of the identified events went undetected for several weeks. These findings were sufficient for staff to conclude that an open phase condition is a credible event having safety significance and must be considered in the electric power system design for nuclear power plants. This paper elaborates the processes and results of the open-phase fault studies on the standard nuclear power plants in Korea. The ETAP unbalanced load flow module is used for the system analysis. The analysis described in this paper has been performed to determine the response of unit auxiliary transformers and standby auxiliary transformers during open-phase conditions to aid in the development of system protection schemes designed to detect such conditions.

한국표준형 원전 제어봉구동장치 전원공급계통의 전동발전기 세트 안정성 개선

한국표준형 원전 제어봉구동장치 전원공급계통의 전동발전기 세트 안정성 개선

Analyses of Vertical Seismic Responses of Seismically Isolated Nuclear Power Plant Structures Supported by Lead Rubber Bearings

It is very important to assure the seismic performance of equipment as well as building structures in seismic design of nuclear power plant(NPP). Seismically isolated structures may be reviewed mainly on the horizontal seismic responses. Considering the equipment installed in the NPP, the vertical earthquake responses of the structure also should be reviewed. This study has investigated the vertical seismic demand of seismically isolated structure by lead rubber bearings(LRBs). For the numerical evaluation of seismic demand of the base isolated NPP, the Korean standard nuclear power plant (APR1400) is modeled as 4 different models, which are supported by LRBs to have 4 different horizontal target periods. Two real earthquake records and artificially generated input motions have been used as inputs for earthquake analyses. For the study, the vertical floor response spectra(FRS) were generated at the major points of the structure. As a results, the vertical seismic responses of horizontally isolated structure have largely increased due to flexibility of elastomeric isolator. The vertical stiffness of the bearings are more carefully considered in the seismic design of the base-isolated NPPs which have the various equipment inside.

한국 원자력 분야 정책추진체계 연구 - 고리원전추가건설사례로 본 위험거버넌스 구축 측면에서 -

The purpose of this study is to find ways to build a desirable risk governance to respond to these uncertainties like such as increased risk due to the rapid development of modern science and technology in S. Korea. This study is about model of risk communication with scientific technology for additional construction of nuclear power plants. This study analyzed risk communication with scientific technology through new explanatory models between approval opinions of scientific professional group and opposition opinions of the public with Kori nuclear power plant of S. Korea. And then This study investigated related system for policy promotion in the field of nuclear energy of the current S. Korea. Consequently, Governance to strengthen the negotiations on nuclear technology at the time of the interaction in risk communication have been identified and the future of policy direction of the Korean Standard Nuclear Power Plant was detected.

지반-구조물 상호작용 효과를 고려한 지진격리시스템이 적용된 원전 격납건물의 지진 취약도 평가

Several researches have been studied to enhance the seismic performance of nuclear power plants (NPPs) by application of seismic isolation. If a seismic base isolation system is applied to NPPs, seismic performance of nuclear power plants should be reevaluated considering the soil-structure interaction effect. The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP structures and equipment. In this study, the seismic performance of an isolated NPP is evaluated by seismic fragility curves considering the soil-structure interaction effect. The designed seismic isolation is introduced to a containment building of Shin-Kori NPP which is KSNP (Korean Standard Nuclear Power Plant), to improve its seismic performance. The seismic analysis is performed considering the soil-structure interaction effect by using the linearized model of seismic isolation with SASSI (System for Analysis of Soil-Structure Interaction) program. Finally, the seismic fragility is evaluated based on soil-isolation-structure interaction analysis results.

Internal leakage detection for feedwater heaters in power plants using neural networks

As interest in safety and performance of power plants becomes more serious and wide-ranging, the significance of research on turbine cycles has attracted more attention. This paper particularly focuses on thermal performance analysis under the conditions of internal leakages inside closed-type feedwater heaters (FWHs) and their diagnosis to identify the locations and to quantify leak rates. Internal leakage is regarded as flow movement through the isolated path but remaining inside the system boundary of a turbine cycle. For instance, leakages through the cracked tubes, tube-sheets, or pass partition plates in a FWH are internal leakages. Internal leakages impact not only plant efficiency, but also direct costs and/or even plant safety associated with the appropriate repairs. Some types of internal leakages are usually critical to get the parts fixed and back in a timely manner. The FWHs installed in a Korean standard nuclear power plant were investigated in this study. Three technical steps have been, then, conducted: (1) the detailed modeling of FWHs covering the leakage from tubes, tube-sheets, or pass partition plates using the simulation model, (2) thermal performance analysis under various leakage conditions, and (3) the development of a diagnosis model using a feed-forward neural network, which is the correlation between thermal performance indices and leakage conditions. Since the operational characteristics of FWHs are coupled with one another and/or with other neighbor components such as turbines or condensers, recognizing internal leakages is difficult with only an analytical model and instrumentation at the inlet and outlet of tube- and shell-sides. The proposed neural network-based correlation was successfully validated for test cases.

Developing architecture for upgrading I&C systems of an operating nuclear power plant using a quality attribute-driven design method

Abstract This paper presents the architecture for upgrading the instrumentation and control (I&C) systems of a Korean standard nuclear power plant (KSNP) as an operating nuclear power plant. This paper uses the analysis results of KSNP's I&C systems performed in a previous study. This paper proposes a Preparation–Decision–Design–Assessment (PDDA) process that focuses on quality oriented development, as a cyclical process to develop the architecture. The PDDA was motivated from the practice of architecture-based development used in software engineering fields. In the preparation step of the PDDA, the architecture of digital-based I&C systems was setup for an architectural goal. Single failure criterion and determinism were setup for architectural drivers. In the decision step, defense-in-depth, diversity, redundancy, and independence were determined as architectural tactics to satisfy the single failure criterion, and sequential execution was determined as a tactic to satisfy the determinism. After determining the tactics, the primitive digital-based I&C architecture was determined. In the design step, 17 systems were selected from the KSNP's I&C systems for the upgrade and functionally grouped based on the primitive architecture. The overall architecture was developed to show the deployment of the systems. The detailed architecture of the safety systems was developed by applying a 2-out-of-3 voting logic, and the detailed architecture of the non-safety systems was developed by hot-standby redundancy. While developing the detailed architecture, three ways of signal transmission were determined with proper rationales: hardwire, datalink, and network. In the assessment step, the required network performance, considering the worst-case of data transmission was calculated: the datalink was required by 120 kbps, the safety network by 5 Mbps, and the non-safety network by 60 Mbps. The architecture covered 17 systems out of 22 KSNP's I&C systems. The architecture is implementable with the equipment developed in South Korea. The architecture can be used as a model to upgrade the existing I&C systems in a planned, large-scale, and one-shot manner. A more detailed architecture down to software level will be developed in the future.

A study on the effects of ESFAS STI modification on the unavailability of ESF actuated components

The nuclear industry has made so great efforts to attain safe and economically viable Nuclear Power Plants all over the world. Probabilistic Safety Assessment (PSA) applications are required now more than ever in order to operate Nuclear Power Plants. In this paper, the unavailability model of Engineered Safety Feature (ESF) actuated components, which is part of the Korean Standard Nuclear Power Plant (KSNPP) PSA model, is used to analyze the effects of Surveillance Test Interval (STI) extension of Engineered Safety Feature Actuation System (ESFAS) on the unavailability of the ESF actuated components. To develop a more accurate and realistic PSA model, the loop controller modules of Plant Control System (PCS) are also modeled on the KSNPP PSA model. The fault tree developed by each ESF actuated component includes the component failure, ESFAS signal failure, and loop controller module failure of PCS. When the ESFAS STI is extended from one month to three months, the unavailability of ESF actuated components partially increased by 10%. This shows the STI extension of ESFAS has minor effects on the unavailability of the ESF actuated components. The effects of ESFAS STI extension of KSNPP on the unavailability of the ESF actuated components as well as Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) is less than expected. That result indirectly indicates that the ESFAS in KSNPP is designed well enough to extend the ESFAS STI.

Detection and Mode Identification of Axial Cracks in the Steam Generator Tube of the Nuclear Power Plant Using Ultrasonic Guided Wave

For those people who are involved in NDE, there is a growing concern regarding the significant traveling distance of a guided wave in a structure, which ensures the inspection of a large area of the structure from a single location. A significant number of studies on the guided wave have therefore been made to apply the foregoing to a nondestructive evaluation in many different industries and resulted in an increase in the efficiency of practical guided wave inspection. Unlike the previous studies based mainly on the detection of circumferential flaws, this study is focused on the axial flaw detection in the steam generator tubes of Korean standard nuclear power plants by generating the guided wave by changing frequency and selecting the applicable mode from the dispersion curve for the steam generator tube calculated in this study, where the dispersion-based short-time Fourier transform (D-STFT) algorithm is used to enhance mode identification. In conclusion, the L (0,1) mode at 2.25MHz is found to be most sensitive in detecting axial flaws in a steam generator tube.

A steam generator model identification and robust H ∞ controller design with ν-gap metric for a feedwater control system

A robust H ∞ controller for the feedwater system of KSNP (Korean Standard Nuclear Power Plant) vertical U-tube type steam generators (UTSG) is presented herein. As the first step of the controller development, a precise thermal–hydraulic model for the steam generator is built. A series of model experiments are performed using the developed thermal–hydraulic model in order to acquire the input–output data sets which represent steam generator characteristics. These data sets are utilized to build simplified steam generator models for control through a system identification algorithm, Simple Process Models. Among the developed steam generator models, the representative models for the designated power ranges are selected by a criterion of ν-gap metric. The representative robust controllers for the selected models are designed utilizing the loop-shaping H ∞ design technique. Finally, the robustness and performance of the proposed controllers are validated and compared against those of PI (proportional–integral) controller. The validation results demonstrated that the proposed H ∞ robust controller has a superior robustness and an enhanced control performance.

Detection and Mode Identification of Axial Cracks in the Steam Generator Tube of the Nuclear Power Plant Using Ultrasonic Guided Wave

For those people who are involved in NDE, there is a growing concern regarding the significant traveling distance of a guided wave in a structure, which ensures the inspection of a large area of the structure from a single location. A significant number of studies on the guided wave have therefore been made to apply the foregoing to a nondestructive evaluation in many different industries and resulted in an increase in the efficiency of practical guided wave inspection. Unlike the previous studies based mainly on the detection of circumferential flaws, this study is focused on the axial flaw detection in the steam generator tubes of Korean standard nuclear power plants by generating the guided wave by changing frequency and selecting the applicable mode from the dispersion curve for the steam generator tube calculated in this study, where the dispersion-based short-time Fourier transform (D-STFT) algorithm is used to enhance mode identification. In conclusion, the L (0,1) mode at 2.25MHz is found to be most sensitive in detecting axial flaws in a steam generator tube.

내충격성을 고려한 사용후연료 수송용기 내부구조물의 설계 연구

A simple preliminary analysis is often useful to check a validity of design alternatives before the detailed analysis phase in the viewpoint of efficiency. This paper describes a preliminary analysis procedure for the selection among basket design candidates for the spent fuel shipping cask of Korean standard nuclear power plant. As the cask should maintain the structural integrity in hypothetical accident condition, the case of 9 m drop is significantly considered as the worst scenario among the accident conditions in structural design viewpoint in this paper. As basket design options, totally four different types are considered and analyzed in the point of structural integrity at drop impact and weldability for fabrication. As a result, an insertion round plate type with densely spaced supports turns out to be the best in both of the viewpoints, though the weld plate type shows a bit more design margin.

Head Loss Coefficient Evaluation Based on CFD Analysis for PWR Downcomer and Lower Plenum

Nuclear vendors and utilities perform numerous simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes, most of which were developed based on one-dimensional lumped parameter models. During the past decade, however, computers, parallel computation methods, and three-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. The use of advanced commercial CFD codes is considered beneficial in the safety analysis and design of NPPs. The present work analyzes the flow distribution in the downcomer and lower plenum of Korean standard nuclear power plants using STAR-CD. The lower plenum geometry of a pressurized water reactor (PWR) is very complicated, as there are many reactor internals, which hinders CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of a PWR is used. The results give a clear figure about flow fields in the downcomer and lower plenum of a PWR, which is one of major safety concerns. The calculated pressure drop across the downcomer and lower plenum appears to be in good agreement with the data in engineering calculation. An algorithm that can evaluate the head loss coefficient, which is necessary for thermal-hydraulic system code running, was suggested based on these CFD analysis results.