The Experts below are selected from a list of 16734 Experts worldwide ranked by ideXlab platform
Noam Lior - One of the best experts on this subject based on the ideXlab platform.
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exergy analysis of an operating boiling water Reactor nuclear power station
Energy Conversion and Management, 1995Co-Authors: William R Dunbar, Scott D Moody, Noam LiorAbstract:Abstract A Second Law analysis is performed on the LaSalle County Nuclear Station of the Commonwealth Edison Company to evaluate plant and subsystem irreversibility. The results disclose that over 80% of the exergy destroyed during plant operation is result of the highly irreversible fission and heat transport processes within the Reactor Vessel. Plant efficiency and effectiveness are found to be 34.4%, well below the 40–45% efficiencies of typical fossil-fuel-fired power generating stations. Based on these well-known numbers, and the results of the exergy analysis, one recommendation is to reevaluate the integration of fossil-fuel-fired superheat/reheat units located downstream of the Reactor Vessel. This modified plant configuration would not only improve efficiency by raising the top operating temperature, but it is also anticipated to reduce the irreversibility associated with heat transfer in the steam generators.
Jae Jun Jeong - One of the best experts on this subject based on the ideXlab platform.
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computational analysis of downcomer boiling phenomena using a component thermal hydraulic analysis code cupid
Journal of Engineering for Gas Turbines and Power-transactions of The Asme, 2011Co-Authors: Hyoung Kyu Cho, Byong-jo Yun, Ik Kyu Park, Jae Jun JeongAbstract:For the analysis of transient two-phase flows in nuclear Reactor components such as a Reactor Vessel, a steam generator, and a containment, KAERI has developed a three-dimensional thermal hydraulic code, CUPID . It adopts a three-dimensional, transient, two-phase and three-field model and includes various physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, the CUPID code and its two-phase flow models were assessed against the downcomer boiling experiment, which was performed to simulate the downcomer boiling phenomena. They may happen in the downcomer of a nuclear Reactor Vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the Reactor Vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. The computational analysis using the CUPID code showed that it can appropriately predict the multidimensional boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size model.
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computational analysis of downcomer boiling phenomena using a component thermal hydraulic analysis code cupid
18th International Conference on Nuclear Engineering: Volume 4 Parts A and B, 2010Co-Authors: Hyoung Kyu Cho, Byong-jo Yun, Ik Kyu Park, Jae Jun JeongAbstract:A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analyses of components of a nuclear Reactor, such as Reactor Vessel, steam generator, containment, etc. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. In the present paper, the two-phase models were assessed against the DOBO (DOwncomer BOiling) experiment, which was constructed to simulate the downcomer boiling phenomenon. It may happen in the downcomer of a nuclear Reactor Vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the Reactor Vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. This phenomenon has been considered as a crucial safety issue of an advanced power Reactor because it is concerned with the core cooling capability of the safety injection system. In this paper, the physical models and correlations that were incorporated into the CUPID code were introduced and the validation results against the experiment were reported. The benchmark calculation results concluded that the CUPID code can appropriately predict the boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size correlation.Copyright © 2010 by ASME
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assessment of the cobra relap5 code using the loft l2 3 large break loss of coolant experiment
Annals of Nuclear Energy, 1997Co-Authors: Jae Jun Jeong, Suk Ku Sim, C H Ban, C E ParkAbstract:Abstract The COBRA/RELAP5 code, a merged version of COBRA-TF and RELAP5/MOD3 5m5, was upgraded with RELAP5/MOD3.2. Using the “inter-process communication” technique, the realistic three-dimensional Reactor Vessel module of CIBRA-TF was linked with the RELAP5 code in the merged code. The three-dimensional capability of COBRA/RELAP5 was assessed using the LOFT L2–3 large-break loss-of-coolant accident (LOCA) experimental data. The evaluate the overall performance of the code, relatively fine three-dimensional meshes were used for the Reactor Vessel, which was simulated by CIBRA-TF, and the other systems were one-dimensionally modeled by RELAP5/MOD3.2. The simulation results showed that COBRA/RELAP5 is a promising tool for analysis of complicated, multidimensional, two-phase flow transients in pressurized water Reactors (PWRS).
F B Cheung - One of the best experts on this subject based on the ideXlab platform.
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development of a downward facing nucleate boiling correlation for thermal hydraulics analysis
Experimental Thermal and Fluid Science, 2021Co-Authors: F B Cheung, Faruk A Sohag, Michael RileyAbstract:Abstract The concept of in-Vessel retention (IVR) by passive external Reactor Vessel cooling (ERVC) in a flooded cavity during a severe accident has now been recognized as a viable approach to retain the radioactive core melt (i.e., corium) within the Reactor Vessel. During the process of long-term in-Vessel cooling of core melt, the heat flux must remain below the critical heat flux (CHF) level to maintain a regime of nucleate boiling such that the integrity of the Reactor pressure Vessel (RPV) is not compromised. In this study, steady-state data for downward facing boiling (DFB) obtained under simulated IVR-ERVC conditions in the Subscale Boundary Layer Boiling (SBLB) test facility is critically analyzed based upon which a new DFB correlation is derived mathematically from theoretical considerations with the coefficients resulting from the scaling analysis being determined from the DFB data. The new correlation which adequately predicts the local variation of the nucleate boiling heat flux along the outer surface of the RPV under IVR-ERVC conditions can be used to describe the long-term cooling behavior of the corium within the RPV of advanced nuclear power plants.
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enhancement of downward facing saturated boiling heat transfer by the cold spray technique
Nuclear Engineering and Technology, 2017Co-Authors: Faruk A Sohag, F B Cheung, Faith R Beck, Lokanath Mohanta, A E Segall, Timothy J Eden, John K PotterAbstract:Abstract In-Vessel retention by passive external Reactor Vessel cooling under severe accident conditions is a viable approach for retention of radioactive core melt within the Reactor Vessel. In this study, a new and versatile coating technique known as “cold spray” that can readily be applied to operating and advanced Reactors was developed to form a microporous coating on the outer surface of a simulated Reactor lower head. Quenching experiments were performed under simulated in-Vessel retention by passive external Reactor Vessel cooling conditions using test Vessels with and without cold spray coatings. Quantitative measurements show that for all angular locations on the Vessel outer surface, the local critical heat flux (CHF) values for the coated Vessel were consistently higher than the corresponding CHF values for the bare Vessel. However, it was also observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit vary appreciably along the outer surface of the test Vessel. Nonetheless, results of this intriguing study clearly show that the use of cold spray coatings could enhance the local CHF limit for downward-facing boiling by > 88%.
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chf enhancement by Vessel coating for external Reactor Vessel cooling
Nuclear Engineering and Design, 2006Co-Authors: J Yang, J L Rempe, F B Cheung, M B Dizon, Kune Y Suh, Sang B KimAbstract:Abstract In-Vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and proposed for some advanced light water Reactors (ALWRs). One viable means for IVR is the method of external Reactor Vessel cooling (ERVC) by flooding the Reactor cavity during a severe accident. As part of a joint Korean–United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate Vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the subscale boundary layer boiling (SBLB) facility at the Pennsylvania State University using test Vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady-state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady-state boiling, the Vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the Vessel outer surface but with very little effect of aging.
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corium retention for high power Reactors by an in Vessel core catcher in combination with external Reactor Vessel cooling
Nuclear Engineering and Design, 2004Co-Authors: J L Rempe, D L Knudson, Keith G Condie, F B CheungAbstract:Abstract If there were inadequate cooling during a Reactor accident, a significant amount of core material could become molten and relocate to the lower head of the Reactor Vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the Vessel lower head remains intact so that relocated core materials are retained within the Vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-Vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, IVR of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water Reactors. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power Reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for Reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.
Doyoung Ko - One of the best experts on this subject based on the ideXlab platform.
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selection criteria of measurement locations for advanced power Reactor 1400 Reactor Vessel internals comprehensive vibration assessment program
Transactions of The Korean Society for Noise and Vibration Engineering, 2011Co-Authors: Doyoung KoAbstract:ABSTRACT U.S. nuclear regulatory commission(NRC) regulatory guide(RG) 1.20 requires a comprehensive vi-bration assessment program(CVAP) for use in verifying the structural integrity of Reactor Vessel in-ternals(RVI) for flow-induced vibrations prior to commercial operation. The CVAP program consist of vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. One of the main purposes of the analysis program is to select measure-ment locations, however measurement locations can not be determined by only analysis results, there-fore we developed selection criteria of measurement locations for advanced power Reactor 1400(APR1400) RVI CVAP, It will be used to select measurement locations and instrument types for APR1400 RVI CVAP. * 1. 서 론 원자로내부구조물(RVI : Reactor Vessel internals)은 원자로(RV : Reactor Vessel)의 정상(steady) 및 과도(transient) 운전조건하에서 원자로 냉각재 유동에 의해 진동을 겪게 되므로 원전의 전 수명기간 동안 그 건전성이 유지됨과 안전여유도가 확보되고 있음을 입증해야 하며, 이를 위해서 원자로내부구조물에 대한 종합진동평가계획(CVAP : comprehensive †교신저자; 정회원, 한국수력원자력(주) 중앙연구원E-mail : kodoyoung@khnp.co.krTel : (042)870-5774, Fax : (042)870-5779* 한국수력원자력(주) 중앙연구원# 이 논문의 일부는 2011년 춘계 소음진동 학술대회에서 발표되었음.vibration assessment program)은 U.S. NRC(nuclear regulatory commission) R.G.(regulatory guide) 1.20
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a review of measuring sensors for Reactor Vessel internals comprehensive vibration assessment program in advanced power Reactor 1400
Transactions of The Korean Society for Noise and Vibration Engineering, 2011Co-Authors: Doyoung KoAbstract:Reactor Vessel internals comprehensive vibration assessment program(RVI CVAP) is one of the necessary tests to ensure the safety of nuclear power plants. RVI CVAP of U.S. nuclear regulatory commission regulatory guide 1.20(U.S. NRC R.G. 1.20) consists of the analysis, measurement and inspection. One of the core technologies of the measurement program for RVI CVAP is to select suitable sensors because the measurement is conducted during the critical path of the construction period of nuclear power plants. Therefore, we analyzed RVI thermal-hydraulic and structure design data of Palo Verde nuclear power plant(U.S.), Yonggwang nuclear power plant(Korea) and APR1400 and researched measuring sensors used in them; moreover, we investigated sensors used for measurement of RVI CVAP for the last 20 years throughout the world. Based on these results, we selected suitable measuring sensors for RVI CVAP in advanced power Reactor 1400(APR1400).
Michael Riley - One of the best experts on this subject based on the ideXlab platform.
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development of a downward facing nucleate boiling correlation for thermal hydraulics analysis
Experimental Thermal and Fluid Science, 2021Co-Authors: F B Cheung, Faruk A Sohag, Michael RileyAbstract:Abstract The concept of in-Vessel retention (IVR) by passive external Reactor Vessel cooling (ERVC) in a flooded cavity during a severe accident has now been recognized as a viable approach to retain the radioactive core melt (i.e., corium) within the Reactor Vessel. During the process of long-term in-Vessel cooling of core melt, the heat flux must remain below the critical heat flux (CHF) level to maintain a regime of nucleate boiling such that the integrity of the Reactor pressure Vessel (RPV) is not compromised. In this study, steady-state data for downward facing boiling (DFB) obtained under simulated IVR-ERVC conditions in the Subscale Boundary Layer Boiling (SBLB) test facility is critically analyzed based upon which a new DFB correlation is derived mathematically from theoretical considerations with the coefficients resulting from the scaling analysis being determined from the DFB data. The new correlation which adequately predicts the local variation of the nucleate boiling heat flux along the outer surface of the RPV under IVR-ERVC conditions can be used to describe the long-term cooling behavior of the corium within the RPV of advanced nuclear power plants.
