• Title/Summary/Keyword: Pressurized-water reactor (PWR)

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Evaluation of Piping Integrity in Thinned Main Feedwater Pipes

  • Park, Young-Hwan;Kang, Suk-Chull
    • Nuclear Engineering and Technology
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    • v.32 no.1
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    • pp.67-76
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    • 2000
  • Significant wall thinning due to flow accelerated corrosion(FAC)was recently reported in main feedwater pipes in 3 Korean pressurized water reactor(PWR) plants. The main feedwater pipes in one plant were repaired using overlay weld method at the outside of pipe, while those in 2 other plants were replaced with new pipes. In this study, the effect of the wall thinning in the main feedwater pipes on piping integrity was evaluated using finite element method. Especially, the effects of both the overlay weld repair and the stress concentration in notch-type thinned area on the piping integrity were investigated. The results are as follows : (1) The piping load carrying capacity may significantly decrease due to FAC. In special, the load carrying capacity of the main feedwater pipe was reduced by about 40% during about 140 months operation in Korean PWR plants. (2) By performing overlay weld repair at the outside of pipe, the piping load carrying capacity can increase and the stress concentration level in the thinned area can be reduced.

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Welding Quality Evaluation on the LASER Welding Parts of the Spacer Grid Assembly for PWR fuel Assembly (경수로 원전연료용 지지격자의 LASER 용접품질 평가)

  • Song, Gi-Nam;Yun, Jeong-Ho;Gang, Hong-Seok;Lee, Gang-Hui;Kim, U-Gon;Kim, Su-Seong
    • Proceedings of the KWS Conference
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    • 2005.06a
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    • pp.109-111
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    • 2005
  • Nuclear fuel assemblies for pressurized water reactors(PWR) are loaded in the reactor core throughout the residence time of three to five years. A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly. The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral loads acting on the nuclear fuel assembly so as to keep the nuclear fuel assembly straight. To meet this requirement, it is necessary to weld the welding parts carefully and precisely. In this study, laser welding qualities of the spacer grid assembly welded by several welding companies, such as weld strength, weld penetration depth, and weld bead size, are examined and compared.

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Evaluation of various large-scale energy storage technologies for flexible operation of existing pressurized water reactors

  • Heo, Jin Young;Park, Jung Hwan;Chae, Yong Jae;Oh, Seung Hwan;Lee, So Young;Lee, Ju Yeon;Gnanapragasam, Nirmal;Lee, Jeong Ik
    • Nuclear Engineering and Technology
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    • v.53 no.8
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    • pp.2427-2444
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    • 2021
  • The lack of plant-side energy storage analysis to support nuclear power plants (NPP), has setup this research endeavor to understand the characteristics and role of specific storage technologies and the integration to an NPP. The paper provides a qualitative review of a wide range of configurations for integrating the energy storage system (ESS) to an operating NPP with pressurized water reactor (PWR). The role of ESS technologies most suitable for large-scale storage are evaluated, including thermal energy storage, compressed gas energy storage, and liquid air energy storage. The methods of integration to the NPP steam cycle are introduced and categorized as electrical, mechanical, and thermal, with a review on developments in the integration of ESS with an operating PWR. By adopting simplified off-design modeling for the steam turbines and heat exchangers, the results show the performance of the PWR steam cycle changes with respect to steam bypass rate for thermal and mechanical storage integration options. Analysis of the integrated system characteristics of proposed concepts for three different ESS suggests that certain storage technologies could support steady operation of an NPP. After having reviewed what have been accomplished through the years, the research team presents a list of possible future works.

PWR core calculation based on pin-cell homogenization in three-dimensional pin-by-pin geometry

  • Bin Zhang;Yunzhao Li;Hongchun Wu;Wenbo Zhao;Chao Fang;Zhaohu Gong;Qing Li;Xiaoming Chai;Junchong Yu
    • Nuclear Engineering and Technology
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    • v.56 no.6
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    • pp.1950-1958
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    • 2024
  • For the pressurized water reactor two-step calculation, the traditional assembly homogenization and two-group neutron diffusion calculation have been widely used. When it comes to the core pin-by-pin simulation, many models and techniques are different and unsettled. In this paper, the homogenization methods based on the pin discontinuity factors and super homogenization factors are used to get the pin-cell homogenized parameters. The heterogeneous leakage model is applied to modify the infinite flux spectrum of the single assembly with reflective boundary condition and to determine the diffusion coefficients for the SP3 solver which is used in the core simulation. To reduce the environment effect of the single-assembly reflective boundary condition, the online method for the SPH factors updating is applied in this paper, and the functionalization of SPH factors based on the least-squares method will be pre-made alone with the table of the group constants. The fitting function will be used to update the thermal-group SPH factors with a whole-core pin-by-pin homogeneous solution online. The three-dimensional Watts Bar Nuclear Unit 1 (WBN1) problem was utilized to test the performance of pin-by-pin calculation. And numerical results have demonstrated that PWR pin-by-pin core calculation has more accurate results compared with the traditional assembly-homogenization scheme.

Prediction of Heat Transfer Rates to Spray Water Droplets in a High Pressure Mixture Composed of Saturated Steam and Noncondensable Hydrogen Gas (고압의 포화수증기-비응축성 수소기체 혼합기 속에서 분무수적으로의 열전달을 예측)

  • Lee, S.K.;Jo, J.C.;Cho, J.H.
    • Korean Journal of Air-Conditioning and Refrigeration Engineering
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    • v.3 no.5
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    • pp.337-349
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    • 1991
  • Heat and mass transfer rates to spray water droplets for spray transients in a high pressure vessel have been predicted by two different droplet models: the complete mixing model and the non-mixing model. In this process, the ambient fluid surrounding the droplets is a real-gas mixture composed of saturated steam and noncondensable hydrogen gas at high pressure. The physical properties of the mixture are estimated by applying the concept of compressibility factor and using appropriate correlations. A computer program, DROPHMT, to calculate the heat and mass transfer rates for two different droplet models has been developed. As an illustrative application of the computer program to engineering practices, heat and mass transfer rates to spray water droplets for spray transients in a Pressurized Water Reactor (PWR) pressurizer have been calculated, and the typical results have been provided.

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PWSCC and System Engineering Development of Internal Inspection and Maintenance Methodology for RCS

  • Abdallah, Khaled Atya Ahmed;Mesquita, Patricia Alves Franca de;Yusoff, Norashila;Nam, GungIhn;Jung, JaeCheon;Lee, YoungKwan
    • Journal of the Korean Society of Systems Engineering
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    • v.12 no.1
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    • pp.89-103
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    • 2016
  • Due to safety of the plant, it became very clear the importance of study occurrence reactor coolant system (RCS) issues specially the primary water stress corrosion cracking (PWSCC). The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. Robotic devices have been used for internal inspection, maintenance and performing remote welding and inspection in high-radiation areas. In this paper, PWSCC overview and inlay and over lay welding methodology introduced, concept of robotic device that can be inserted into the piping via Steam Generator (SG) main way to access to primary piping of pressurized water reactor (PWR) is developed based on SE methodology. A 3D model of the inspection system was developed along with the APR1400 (Advanced Power Reactor)reactor coolant systems (RCS) and internals with virtual 3D simulation of the operation for visualization to prove the validity of the concept.

Validation of UNIST Monte Carlo code MCS using VERA progression problems

  • Nguyen, Tung Dong Cao;Lee, Hyunsuk;Choi, Sooyoung;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • v.52 no.5
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    • pp.878-888
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    • 2020
  • This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

On the Reconstruction of Pointwise Power Distributions in a Fuel Assembly From Coarse-Mesh Nodal Calculations (노달계산결과로부터 핵연료 집합체내의 출력분포를 재생하는 방법에 관하여)

  • Jeong, Hun-Young;Cho, Nam-Zin
    • Nuclear Engineering and Technology
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    • v.20 no.3
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    • pp.145-154
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    • 1988
  • This paper is a study on an accurate and computationally efficient method for reconstructing pointwise power distributions from coarse-mesh nodal calculations. The modern nodal codes can calculate global reactor power shapes and criticality very efficiently and accurately. But inherent in the nodal procedures, there is inevitable loss of information on local heterogeneous quantities. In this study, an improved form function method which reflects the exponential transition of the thermal flux near the assembly surface is developed for the reconstruction of the heterogeneous fluxes. Use of the new form function method in several pressurized water reactor (PWR) benchmark problems reduces the maximum errors in the reconstructed thermal flux to those in the reconstructed fast flux. Even for assemblies adjacent to the steel baffle in realistic PWR cores, use of this method also results in improved pointwise power reconstruction.

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Crush Strength Analysis of a Spacer Grid for PWR Nuclear Fuel Considering Mechanical Properties in Weld Zone (용접부 기계적 물성치를 고려한 경수로 핵연료 지지격자의 충격해석)

  • Song, Kee-Nam;Lee, Sang-Hoon
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.8 no.2
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    • pp.7-13
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    • 2012
  • A spacer grid which is one of the most important structural components in a pressurized water reactor fuel is an interconnected array of slotted grid straps, welded at the intersections to form an egg-crate structure. The spacer grid is required to not only protect fuel rods stably but also have sufficient lateral crush strength for the sake of enabling shut-down of the nuclear reactor during abnormal operating environments. Then, the lateral crush strength of the spacer grid is closely related with welding quality of the spacer grid. Previous research on the crush strength analysis of the spacer grid had been performed using only parent material properties. In this study, to investigate the effect on the crush strength of the spacer grid when used mechanical properties in weld zone instead of parent material properties, crush strength analysis considering mechanical properties in weld zone obtained from the instrumented indentation technique was performed and compared the results with the previous research.

Assessment of thermal fatigue induced by dryout front oscillation in printed circuit steam generator

  • Kwon, Jin Su;Kim, Doh Hyeon;Shin, Sung Gil;Lee, Jeong Ik;Kim, Sang Ji
    • Nuclear Engineering and Technology
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    • v.54 no.3
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    • pp.1085-1097
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    • 2022
  • A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.