• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3745-3765
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• 2022

The concrete structures related to nuclear safety are threatened by accidental impact loadings, mainly including the low-velocity drop-weight impact (e.g., spent fuel cask and assembly, etc. with the velocity less than 20 m/s) and high-speed projectile impact (e.g., steel pipe, valve, turbine bucket, etc. with the velocity higher than 20 m/s), while the existing studies are still limited in the impact resistant design of nuclear power plant (NPP), especially the primary RC slab. This paper aims to propose the numerical simulation and theoretical approaches to assist the impact-resistant design of RC slab in NPP. Firstly, the continuous surface cap (CSC) model parameters for concrete with the compressive strength of 20-70 MPa are fully calibrated and verified, and the refined numerical simulation approach is proposed. Secondly, the two-degree freedom (TDOF) model with considering the mutual effect of flexural and shear resistance of RC slab are developed. Furthermore, based on the low-velocity drop hammer tests and high-speed soft/hard projectile impact tests on RC slabs, the adopted numerical simulation and TDOF model approaches are fully validated by the flexural and punching shear damage, deflection, and impact force time-histories of RC slabs. Finally, as for the two low-velocity impact scenarios, the design procedure of RC slab based on TDOF model is validated and recommended. Meanwhile, as for the four actual high-speed impact scenarios, the impact-resistant design specification in Chinese code NB/T 20012-2019 is evaluated, the over conservation of which is found, and the proposed numerical approach is recommended. The present work could beneficially guide the impact-resistant design and safety assessment of NPPs against the accidental impact loadings.

• 한국압력기기공학회 논문집
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• 제20권1호
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• pp.32-38
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• 2024

The best estimate plus uncertainty methodologies for loss-of-coolant accident analyses make use of the best-estimate codes and relevant experimental databases. Inherently, best-estimate codes have various uncertainties in the model parameters, which can be quantified by the dedicated experimental database. Therefore, this study was devoted to establishing procedures for identifying the input parameters of predictive models and quantifying their uncertainty ranges. The rod bundle heat transfer experiments were employed as a representative reflood separate effect test, and the TRACE code was utilized as a best-estimate code. In accordance with the present procedure for uncertainty quantification, the integrated list of the influential input parameters and their uncertainty ranges was obtained through local sensitivity calculations and screening criteria. The validity of the procedure was confirmed by applying it to uncertainty analyses, which checks whether the measured data are within computed ranges of the variables of interest. The uncertainty quantification procedure proposed in this study is anticipated to provide comprehensive guidance for the conduct of uncertainty analyses.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.860-866
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• 2022

An accurate prediction of the pressure drop along the flow paths is crucial in the design of advanced passive systems cooled by heavy liquid metal coolants. To date, a generic pressure drop correlation over spacer grids by Rehme has been applied extensively, which was obtained from substantial experimental data with multiple types of components. However, a few experimental studies have reported that the correlation may give large discrepancies. To provide a more reliable correlation for ring-type spacer grids, the current numerical study aims at figuring out the most critical factor among four hypothetical parameters, namely the flow area blockage ratio, number of fuel rods, type of fluid, and thickness of the spacer grid in the flow direction. Through a set of computational fluid dynamics simulations, we observed that the flow area blockage ratio dominantly influences the pressure loss characteristics, and thus its dependence should be more emphasized, whereas the other parameters have little impact. Hence, we suggest a new correlation for the drag coefficient as C_B = C_ν,m/ε^2.7, where C_ν,m is formulated by a nonlinear fit of simulation data such that C_ν,m = -11.33 ln(0.02 ln(Re_b)).

• 한국안전학회지
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• 제37권5호
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• pp.80-88
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• 2022

The Korean nuclear industry had developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code, which adopts a two-fluid, three-field model that is comprised of gas, continuous liquid and droplet fields and has the capability to simulate three-dimensional models. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for the accident management plan of a nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification is required for the separate and integral effect experiments. Therefore, the goal of this work is to verify the calculation capability of the SPACE code for multiple failure accidents. For this purpose, an experiment was conducted to simulate a Control Element Drive Mechanism (CEDM) break with a safety injection failure using the ATLAS test facility, which is operated by Korea Atomic Energy Research Institute (KAERI). This experiment focused on the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The results of the overall system transient response using the SPACE code showed similar trends with the experimental results for parameters such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it can be concluded that the SPACE code has sufficient capability to simulate a CEDM break with a safety injection failure accident.

• 대한기계학회논문집
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• 제19권6호
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• pp.1451-1459
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• 1995

The objective of this paper is to evaluate the material properties of SA312 TP316 and SA312 TP304 stainless steels and their associated welds manufactured for safety injection system of Yonggwang 3,4 nuclear generating stations. A total of 62 tensile tests and 46 fracture toughness tests were conducted and the effects of various parameters such as pipe size, crack plane orientation, tests were conducted and the effects of various parameters such as pipe size, crack plane orientation, test temperature, welding on material properties were discussed. Test results show that the effect of test temperature on fracture toughness was significant while the effects of pipe size and crack plane orientation on fracture toughness were negligible. Fracture toughness of the weld metal was in general higher than that of the base metal.

• 비파괴검사학회지
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• 제28권5호
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• pp.427-435
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• 2008

Inspection of dissimilar metal welds using phased array ultrasound is not easy at all, because crystalline structure of dissimilar metal welds cause deviation and splitting of the ultrasonic beams. Thus, in order to have focusing and/or steering phased array beams in dissimilar metal welds, proper time delays should be determined by ray tracing. In this paper, we proposed an effective approach to solve this difficult problem. Specifically, we modify the Oglivy's model parameters to describe the crystalline structure of real dissimilar metal welds in a fabricated specimen. And then, we calculate the proper time delay and incident angle of linear phased array transducer in the anisotropic and inhomogeneous material for focusing and/or steering phased array ultrasonic beams on the desired position.

• Structural Engineering and Mechanics
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• 제76권3호
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• pp.367-378
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• 2020

This paper presents a new design concept for ocean nuclear power plants (ONPPs) using a tension leg platform (TLP). The system-integrated modular advanced reactor, which is one of the successful small modular reactors, is mounted for demonstration. The authors define the design requirements and parameters, modularize and rearrange the nuclear and other facilities, and propose a new total general arrangement. The most fundamental level of design results for the platform and tendon system are provided, and the construction procedure and safety features are discussed. The integrated passive safety system developed for the gravity based structure-type ONPP is also available in the TLP-type ONPP with minor modifications. The safety system fully utilizes the benefits of the ocean environment, and enhances the safety features of the proposed concept. For the verification of the design concept, hydrodynamic analyses are performed using the commercial software ANSYS AQWA with the Pierson-Moskowitz and JONSWAP wave spectra that represent various ocean environments and the results are discussed.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.37-44
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• 2023

Using the Kalman filtering technique, we propose a novel method for estimating the missing video frames to monitor the activities inside the control room of a nuclear power plant (NPP). The purpose of this study is to reinforce the existing security and safety procedures in the control room of an NPP. The NPP control room serves as the nervous system of the plant, with instrumentation and control systems used to monitor and control critical plant parameters. Because the safety and security of the NPP control room are critical, it must be monitored closely by security cameras in order to assess and reduce the onset of any incidents and accidents that could adversely impact the safety of the NPP. However, for a variety of technical and administrative reasons, continuous monitoring may be interrupted. Because of the interruption, one or more frames of the video may be distorted or missing, making it difficult to identify the activity during this time period. This could endanger overall safety. The demonstrated Kalman filter model estimates the value of the missing frame pixel-by-pixel using information from the frame that occurred in the video sequence before it and the frame that will occur in the video sequence after it. The results of the experiment provide evidence of the effectiveness of the algorithm.

• Journal of Radiation Protection and Research
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• 제40권3호
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• pp.181-186
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• 2015

방사성 제논 탐지는 공기 중 $^{131m}Xe$, $^{133}Xe$, $^{133m}Xe$ 및 $^{135}Xe$를 저준위 백그라운드 계측 시스템으로 검출하여 지하 핵실험 여부를 규명하는 핵심기술 중 하나이다. 방사성 제논 감시는 공기 포집, 제논 추출, 측정 및 분석을 통해 수행되며, $^{135}Xe$의 최소검출가능농도는 비교적 짧은 반감기로 인해 포집, 추출 및 측정시간에 따라 큰 차이를 보이게 된다. 본 연구에서는 방사성 제논 계측 시스템의 정해진 시료 포집 및 전처리 조건에서 최적의 방사성 제논 측정시간을 도출하기 위해 이론적 접근 및 SAUNA 시스템을 이용한 실험을 통해 최소의 MDC를 보이는 측정시간을 결정하고 이론적 계산과 실험결과에 대하여 비교 평가하였다.