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Dr. Jun Wang
Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706, USA

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0 Deep Learning
0 STEM education
0 Computational Fluid Dynamics
0 Nuclear power plant safety evaluation
0 Accident tolerant fuels/cladding/coating

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Nuclear power plant safety evaluation
Critical heat flux prediction
Severe accident analysis

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Journal article
Published: 16 June 2021 in Progress in Nuclear Energy
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The loss of coolant accident (LOCA), as one of the design basis accidents (DBAs), is a hypothetical accident that is usually considered in the design of nuclear power plant. LOCA is caused by small/large breaks in the reactor primary coolant and pressure boundary and may result in a loss of reactor coolant at a rate in excess of the reactor makeup system capability. Currently, the typical mitigation measures for LOCA are through the vessel high/low pressure coolant injection (HPCI/LPCI), core spray system, or reactor core isolation cooling (RCIC) system, to protect reactor from fuel melting and core degradation. However, is it possible to directly repair the small breaks passively? Inspired by the hemostasis and coagulation mechanism of human blood vessels, the authors are considering to develop a technology which would allow the reactor self-coagulation during the small break loss of coolant accident (SBLOCA). With the assistance of this innovative technology, the speed of losing coolant through reactor primary loop breaks could be significantly reduced or even the loss is stopped. With more and longer existing coolant in the core, the possibility of core degradation will also be significantly reduced. This innovative technology needs to be designed feasible and simple, which can be directly used into the existing nuclear power plants to improve the inherent safety of the cooling system, benefit the economy and safety of the nuclear power plant. In this paper, the authors compare the reactor primary loop of nuclear power plant with the blood circulation system of human beings from comprehensive and multi-angle perspectives, discuss the similarity, feasibility, and economy of nuclear power plant self-coagulation system based on the existing hemostasis and coagulation mechanism of human blood vessels. If possible, this technology will play a revolutionary positive role in alleviating DBAs related to SBLOCA.

ACS Style

Hui Cheng; Zijun Mai; Xiaoli Zhu; Laishun Wang; Jun Wang. An innovation idea for SBLOCA mitigation strategy: Self-coagulation system in comparison with human hemostatic mechanism. Progress in Nuclear Energy 2021, 138, 103832 .

AMA Style

Hui Cheng, Zijun Mai, Xiaoli Zhu, Laishun Wang, Jun Wang. An innovation idea for SBLOCA mitigation strategy: Self-coagulation system in comparison with human hemostatic mechanism. Progress in Nuclear Energy. 2021; 138 ():103832.

Chicago/Turabian Style

Hui Cheng; Zijun Mai; Xiaoli Zhu; Laishun Wang; Jun Wang. 2021. "An innovation idea for SBLOCA mitigation strategy: Self-coagulation system in comparison with human hemostatic mechanism." Progress in Nuclear Energy 138, no. : 103832.

Journal article
Published: 01 June 2021 in Progress in Nuclear Energy
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In recent years, micro-reactor concepts have attracted increasing attention in the nuclear industry due to the market demand for flexible, reliable, and sustainable power and heat on-site for industrial or federal installations or remote communities. To help demonstrate and validate these innovative reactor concepts, the Micro-reactor AGile Non-nuclear Experimental Test-bed (MAGNET) is being constructed at Idaho National Laboratory (INL) with an initial focus on the thermal and structural performance of heat pipe cooled micro-reactors. At this time, the preliminary design parameters of the MAGNET facility have been specified. In this work, a simulation using the System Analysis Module (SAM) code is performed for the prototypical 37-heat-pipes test article to predict its experimental facility performance and associated uncertainties. We first carry out a benchmark demonstration of our modeling method with an example provided by Argonne National Laboratory (ANL). Then, we predict the thermal performance of the MAGNET facility under steady-state operation. Moreover, several sensitivity parameters are analyzed to investigate their impact on facility thermal performance. This MAGNET experiment simulation provides valuable information for researchers to validate the facility's initial design and steady-state operation.

ACS Style

Yukun Zhou; Jun Wang; Zehua Guo; Yanan He; Yapei Zhang; Suizheng Qiu; G.H. Su; Michael L. Corradini. 3D-2D coupling multi-dimension simulation for the heat pipe micro-reactor by MOOSE&SAM. Progress in Nuclear Energy 2021, 138, 103790 .

AMA Style

Yukun Zhou, Jun Wang, Zehua Guo, Yanan He, Yapei Zhang, Suizheng Qiu, G.H. Su, Michael L. Corradini. 3D-2D coupling multi-dimension simulation for the heat pipe micro-reactor by MOOSE&SAM. Progress in Nuclear Energy. 2021; 138 ():103790.

Chicago/Turabian Style

Yukun Zhou; Jun Wang; Zehua Guo; Yanan He; Yapei Zhang; Suizheng Qiu; G.H. Su; Michael L. Corradini. 2021. "3D-2D coupling multi-dimension simulation for the heat pipe micro-reactor by MOOSE&SAM." Progress in Nuclear Energy 138, no. : 103790.

Journal article
Published: 24 May 2021 in Reliability Engineering & System Safety
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The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal operation and transients. In this study, an uncertainty analysis was employed to investigate the potential benefits of ATF Cr-coated-Zr cladding and canister for an unmitigated Short-Term Station Blackout (STSBO) sequence in a generic BWR plant using the MELCOR systems code. The MELCOR parameters that reflect the current state-of-knowledge of the relevant fuel assembly performance during core degradation were selected and characterized according to their ranges and distributions. An extensive set of simulations (240 MELCOR calculations) were performed for the Zr and Cr-coated-Zr cladding and canister materials, respectively, to determine their effect on core degradation with the associated uncertainties. The comparison between the Zr and Cr-coated-Zr calculations confirms that the use of ATF Cr-coated-Zr as cladding and canister component material in BWR might be an effective way to mitigate the accident progression and reduce the total hydrogen generation during the accident. The core degradation process was only delayed by less than a half hour, providing some additional time for compensatory actions to mitigate with the accident progression. In contrast, the effect of coated materials on total hydrogen generation was more substantial; i.e., hydrogen generation was almost reduced by half. In addition, a sensitivity analysis based on the Pearson and Spearman correlation coefficients was conducted to rank the significance of the considered parameter uncertainties. The Cr-coating failure temperature was identified as the dominant factor in the MELCOR simulations of core degradation and associated hydrogen generation. Understanding these effects will inform and guide researchers to focus on a more productive area of research and development for accident-tolerant fuel concepts and enhancement of core safety margins.

ACS Style

Zehua Guo; Ryan Dailey; Tangtao Feng; Yukun Zhou; Zhongning Sun; Michael L Corradini; Jun Wang. Uncertainty analysis of ATF Cr-coated-Zircaloy on BWR in-vessel accident progression during a station blackout. Reliability Engineering & System Safety 2021, 213, 107770 .

AMA Style

Zehua Guo, Ryan Dailey, Tangtao Feng, Yukun Zhou, Zhongning Sun, Michael L Corradini, Jun Wang. Uncertainty analysis of ATF Cr-coated-Zircaloy on BWR in-vessel accident progression during a station blackout. Reliability Engineering & System Safety. 2021; 213 ():107770.

Chicago/Turabian Style

Zehua Guo; Ryan Dailey; Tangtao Feng; Yukun Zhou; Zhongning Sun; Michael L Corradini; Jun Wang. 2021. "Uncertainty analysis of ATF Cr-coated-Zircaloy on BWR in-vessel accident progression during a station blackout." Reliability Engineering & System Safety 213, no. : 107770.

Editorial article
Published: 15 April 2021 in Frontiers in Energy Research
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Editorial on the Research Topic Safety Analysis of Nuclear Reactor Thermal-Hydraulics A large number of nuclear power plants are under construction in Asia (McDonald, 2008). Thereby, the demand for research on reactor design and thermal hydraulics is of high priority (Sovacool, 2010). Prof. Liangming Pan from Chongqing University holds the 9th China–Korea Workshop on Nuclear Reactor Thermal–Hydraulic in 2019. There are many high-quality conference papers submitted. To select and publish the top articles in the conferences, Prof. Pan leads the Research Topic “Safety Analysis of Nuclear Reactor Thermal Hydraulics” in the journal Frontiers in Energy Research. Finally, 11 articles are collected, covering the experimental and numerical research, from light water to liquid metal and supercritical coolant. There are several articles about the safety of light water reactors. Kim J. et al. contribute the “integral effect test on top-spot break scenario with 4 inches cold leg break LOCA in ATLAS Facility.” Kim H. T. et al. present their “Simulation of a Station Blackout Accident for the SMART using the CINEMA Code.” Du et al. write the “review of regime maps of steam submerged direct contact condensation.” Meanwhile, there are several pieces of research about the bubble and critical heat flux. Yu et al. work on the “experimental study of bubbly-slug flow transition criteria in a vertical circular tube by using WMS.” Liu et al. publish the “assessment of a theoretical model to predicting forced convective critical heat flux in rod bundles.” Park and Chung contribute the “simulation of critical heat flux phenomenon using a non-heating hydrogen evolving system.” There is also research on advanced reactors. Xiang et al. show their “Study on natural circulation heat transfer characteristics of different liquid metals based on factor analysis.” Lv et al. present “A Critical Flow Model For Supercritical Pressures.” Finally, this Research Topic also involves work on small modular reactors and microreactors. For example, Zhang et al. publish “Thermoelectric Conversion Performance of Combined Thermoions System for Space Nuclear Power Supply.” Yuan et al. show the work about “Numerical Simulation of Flow Boiling in Small Channel of Plate OTSG.” Wang et al. contributes to the “Rod Ejection and Drop Accident Analysis of Aqueous Homogeneous Solution Reactor.” After almost 1-year revision and update, this Research Topic finally selected those 11 articles and published them in the journal Frontiers in Energy Research. It allows showing those high-quality work to the public who do not have a chance to attend this meeting. The journal Frontiers in Energy Research will permit publishing more articles from international conferences. All the editors are open to contact for further information. LP is the leading author. JW is the contacting author. YH and K-YC are senior people. All authors contributed to the article and approved the submitted version. The authors declare that the research was conducted in the absence of any commercial or financial relationships that could be construed as a potential conflict of interest. McDonald, A. (2008). Nuclear power global status. IAEA Bull. 49:45. Available online at: https://www.iaea.org/sites/default/files/49204734548.pdf Google Scholar Sovacool, B. K. (2010). A critical evaluation of nuclear power and renewable electricity in Asia. J. Contemp. Asia 40, 369–400. doi: 10.1080/00472331003798350 CrossRef Full Text | Google Scholar Keywords: safety, nuclear reactor, thermal hydraulics, critical heat flux, small modular reactor Citation: Pan L, Wang J, Huang Y and Choi K-Y (2021) Editorial: Safety Analysis of Nuclear Reactor Thermal-Hydraulics. Front. Energy Res. 9:652233. doi: 10.3389/fenrg.2021.652233 Received: 12 January 2021; Accepted: 25 February 2021; Published: 15 April 2021. Edited and reviewed by: Wenzhong Zhou, Sun Yat-Sen University, China Copyright © 2021 Pan, Wang, Huang and Choi. This is an open-access article distributed under the terms of the Creative Commons Attribution License (CC BY). The use, distribution or reproduction in other forums is permitted, provided the original author(s) and the copyright owner(s) are credited and that the original publication in this journal is cited, in accordance with accepted academic practice. No use, distribution or reproduction is permitted which does not comply with these terms. *Correspondence: Jun Wang, [email protected]

ACS Style

Liangming Pan; Jun Wang; Yanping Huang; Ki-Yong Choi. Editorial: Safety Analysis of Nuclear Reactor Thermal-Hydraulics. Frontiers in Energy Research 2021, 9, 1 .

AMA Style

Liangming Pan, Jun Wang, Yanping Huang, Ki-Yong Choi. Editorial: Safety Analysis of Nuclear Reactor Thermal-Hydraulics. Frontiers in Energy Research. 2021; 9 ():1.

Chicago/Turabian Style

Liangming Pan; Jun Wang; Yanping Huang; Ki-Yong Choi. 2021. "Editorial: Safety Analysis of Nuclear Reactor Thermal-Hydraulics." Frontiers in Energy Research 9, no. : 1.

Journal article
Published: 10 February 2021 in Journal of Fluids Engineering
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The periodic shedding of cloud cavitation in a nozzle orifice has a significant influence on the flow field and may have destructive effects. Most of the existing research on the shedding of cloud cavitation in an orifice is based on experimental visualization with focus on the 2D motion of the re-entrant jet and the shedding mechanism. However, the actual cloud cavitation shedding in an orifice is a complex 3D process. Some limited signs of three-dimensionality and asymmetry in cylindrical orifices have been detected recently, but the 3D shedding characteristics remain unclear. In this paper, the cavitation regimes and periodic shedding process in the scaled-up nozzle orifice used by Stanley experiment were simulated with Large Eddy Simulation (LES). The re-entrant jet and periodic shedding mechanism, as well as the shedding frequency, were analyzed from 2D and 3D perspectives. The main results show that the simulated cavitation regimes and the 2D periodic shedding mechanism agree fairly well with the experimental observations, but more 3D features are revealed. By analyzing the 3D shedding process and the three-dimensionality caused by the inclination of the closure line, the three-dimensional asymmetric shedding mode with phase difference p is revealed. Based upon this finding, the shedding frequency and Strouhal number are calculated. The corresponding relationships between shedding frequencies and the frequency peaks of the power spectrum density (PSD) for pressure fluctuations are also confirmed. These results extend the understanding of the unsteady cavitating flow within nozzle orifices from 2D to 3D patterns.

ACS Style

Wenjie Bai; Arris S. Tijsseling; Jun Wang; Quan Duan; Zaoxiao Zhang. Les Investigations of Periodic Cavitation Shedding with Special Emphasis On Three-Dimensional Asymmetry in a Scaled-Up Nozzle Orifice. Journal of Fluids Engineering 2021, 1 .

AMA Style

Wenjie Bai, Arris S. Tijsseling, Jun Wang, Quan Duan, Zaoxiao Zhang. Les Investigations of Periodic Cavitation Shedding with Special Emphasis On Three-Dimensional Asymmetry in a Scaled-Up Nozzle Orifice. Journal of Fluids Engineering. 2021; ():1.

Chicago/Turabian Style

Wenjie Bai; Arris S. Tijsseling; Jun Wang; Quan Duan; Zaoxiao Zhang. 2021. "Les Investigations of Periodic Cavitation Shedding with Special Emphasis On Three-Dimensional Asymmetry in a Scaled-Up Nozzle Orifice." Journal of Fluids Engineering , no. : 1.

Editorial article
Published: 26 January 2021 in Frontiers in Energy Research
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Nuclear power has been used widely around the world and is considered a highly efficient and clean energy source (Kim and Alameri, 2020). Many organizations around the globe are leading the efforts to develop new technologies to improve the economy and safety (Openshaw 2019). The research topic “Advances in Nuclear Power Engineering” aims to publish the most advanced and timely research results. The articles include research results on Gen IV reactors, small modular reactors, reactor physics, reactor thermal hydraulics, advanced safety technologies, and related topics. Variety of important areas are covered in these topics including theoretical, computational, and experimental thermal-hydraulics, multiphase heat transfer, neutronics, uncertainty analysis, aerosol transport, and scaling considerations. Zhang et al. discuss on the Gen IV sodium-cooled fast reactor heat transfer in their article “Theoretical Investigation on the Fully Developed Turbulent Heat Transfer Characteristics of Liquid Sodium.” In the article “Numerical Analysis of FLiBe Laminar Convective Heat Transfer Characteristics in Tubes Fitted with Coaxial Cross Twisted Tape Inserts,” Yang et al. present computational heat transfer results on Gen IV molten salt reactor. Zhang et al. discuss design of the small modular reactors in the article “Comparisons of Reduced Moderation Small Modular Reactors with Heavy Water Coolant.” Those works provide an insight view of the next-generation reactor design include not only high temperature Gen IV reactors but also small modular light water reactor. Song et al. share their results of computational studies in the article “GPU Based Two-Level CMFD Accelerating Two-Dimensional MOC Neutron Transport Calculation.” The neutronic work shows the multidimensional neutronic and thermal information in the reactor core. The results on thermal-hydraulics experimental work are given by Li et al. in the article “The Effect of Different Branch Angles and Different Branch Pipe Sizes on the Onset Law of Liquid Entrainment”; Gao et al. in the article “Influence of the Size and the Angle of Branches Connected to the Main Horizontal Pipe on the Onset of Gas Entrainment”; Mao et al. in the article “Natural Convection Heat Transfer of the Horizontal Rod-Bundle in a Semi-closed Rectangular Cavity”; Ren et al. in the article “Visualization Experiment of Bubble Coalescence in a Narrow Vertical Rectangular Channel”; and by Wang et al. in the article “Study on the Breakdown Mechanism of Water Film on Corrugated Plate Wall Under the Horizontal Shear of Airflow: A Short Communication.” The experimental thermal-hydraulics work aims at the improvement of the heat transfer coefficient and reactor safety. Other thermal-hydraulics related works include the work of Wang et al. in the article “Review and Prospect of the Measurement Technology of the Thickness of the Liquid Film on the Wall of the Corrugated Plate Dryer”; Saeed et al. in the article “Sensitivity Analysis of Some Key Factors on Turbulence Models for Hydrogen Diffusion Using HYDRAGON Code”; Zhou et al. in the article “Analysis of Measuring Characteristics of the Differential Pressure Water-Level Measurement System Under Depressurization Condition”; Zhou et al. in the article “Numerical Study of the Influence of Tube Arrangement on the Flow Distribution Inside the Heat Exchanger in the PCCS”; Wang et al. in the article “Numerical Study on Laminar-Turbulent Transition Flow in Rectangular Channels of a Nuclear Reactor”; Huang et al. in the article “Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET”; Sun et al. in the article “An Improved Best Estimate Plus Uncertainty Method for Small-Break Loss-of-Coolant Accident in Pressurized Water Reactor”; and Zhang et al. in the article “Uncertainty analysis on k-ε turbulence model in the prediction of subcooled boiling in vertical pipes.” The theoretical and numerical thermal-hydraulics work expands the boundary of experimental work. Reactor containment is the final safety boundary of the nuclear power plant. Articles related to reactor containment include articles by Liu et al. “Scaling Design of the Pressure Response Experimental Facility for Pressure Suppression Containment” and Tao et al., “Experimental study on natural deposition characteristics of aerosol in the containment”. Those works are important to help containment integrity during beyond design basic accident conditions. This research topic received good response from authors and successfully attracted dozens of submissions. After the peer-review and editor’s efforts, this research topic finally published nineteen articles. We want to thank all the reviewers, authors, and support from the editor office of Frontiers in Energy Research. ZM is the leading editor; JW is the contact editor; KS and SR are senior people. The authors declare that the research was conducted in the absence of any commercial or financial relationships that could be construed as a potential conflict of interest. Kim, J. H., and Alameri, S. A. (2020). Harmonizing nuclear and renewable energy: case studies. Int. J. Energy Res. 44 (10), 8053–8061. doi:10.1002/er.4987 CrossRefFull Text Google Scholar Openshaw, S. (2019). Nuclear power: siting and safety. Abingdon, United Kingdom: Routledge. CrossRefFull Text Google Scholar Keywords: gen IV reactor, small modular reactor, thermal hydraulic, neutronic, containment Citation: Meng Z, Wang J, Shi K and Revankar ST (2021) Editorial: Advances in Nuclear Power Engineering. Front. Energy Res. 8:629655. doi: 10.3389/fenrg.2020.629655 Received: 15 November 2020; Accepted: 17 December 2020;Published: 26 January 2021. Edited by: Reviewed by:...

ACS Style

Zhaoming Meng; Jun Wang; Kaiyi Shi; Shripad T. Revankar. Editorial: Advances in Nuclear Power Engineering. Frontiers in Energy Research 2021, 8, 1 .

AMA Style

Zhaoming Meng, Jun Wang, Kaiyi Shi, Shripad T. Revankar. Editorial: Advances in Nuclear Power Engineering. Frontiers in Energy Research. 2021; 8 ():1.

Chicago/Turabian Style

Zhaoming Meng; Jun Wang; Kaiyi Shi; Shripad T. Revankar. 2021. "Editorial: Advances in Nuclear Power Engineering." Frontiers in Energy Research 8, no. : 1.

Journal article
Published: 25 December 2020 in Nuclear Engineering and Design
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The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal and transient. In this study, the performance of the proposed ATF concept of Cr-coated-Zircaloy is assessed using a generic Boiling Water Reactor MELCOR plant model considering a Short-term Station Blackout (STSBO) scenario. Simulation results indicate that the use of Cr-coated-Zircaloy as cladding and canister material mitigates the core degradation process as compared to the traditional Zircaloy cladding and canister design. The onset of fuel degradation and collapse is delayed by over thirty minutes, and the extent of fuel degradation is reduced. Specifically, the gross in-vessel hydrogen generation decreased by almost a factor of three. Although the eutectic reaction between Cr-coating and Zircaloy could cause an early failure of the coating, the improvement in the delay of fuel degradation is still notable. Additionally, a thicker coating is found helpful to obtain additional coping time and to decrease hydrogen generation. In addition to the eutectic formation that may compromise Cr-coated Zr, a different failure mode is identified for the Cr-coated-Zr when compared to Zircaloy; i.e., a complete melt of base material leads to component collapse before the coating is oxidized and consumed. These findings can help the industry focus on productive areas of research and development for accident-tolerant fuel concepts and enhancement of core safety margins.

ACS Style

Zehua Guo; Ryan Dailey; Yukun Zhou; Zhongning Sun; Jun Wang; Michael L. Corradini. Effect of ATF Cr-coated-Zircaloy on BWR In-vessel Accident Progression during a Station Blackout. Nuclear Engineering and Design 2020, 372, 110979 .

AMA Style

Zehua Guo, Ryan Dailey, Yukun Zhou, Zhongning Sun, Jun Wang, Michael L. Corradini. Effect of ATF Cr-coated-Zircaloy on BWR In-vessel Accident Progression during a Station Blackout. Nuclear Engineering and Design. 2020; 372 ():110979.

Chicago/Turabian Style

Zehua Guo; Ryan Dailey; Yukun Zhou; Zhongning Sun; Jun Wang; Michael L. Corradini. 2020. "Effect of ATF Cr-coated-Zircaloy on BWR In-vessel Accident Progression during a Station Blackout." Nuclear Engineering and Design 372, no. : 110979.

Journal article
Published: 29 November 2020 in Nuclear Engineering and Design
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Since the accident at Fukushima, one major goal of reactor safety research has been the development of more accident tolerant technologies that can mitigate or delay fuel degradation during a Beyond Design Basis Accident (BDBA). One major effort has been focused on increasing the capability of the fuel to be more tolerant of damage during an accident, i.e., Accident Tolerant Fuel (ATF) materials. In this work, we present the development of a generic BWR plant model, the modification of MELCOR to model ATF materials and the use of ATF materials (specifically FeCrAl alloy) as a coating on Zircaloy cladding or as a substitute material for cladding and fuel assembly canister material and its effect on severe accident progression, specifically, a Station Blackout accident. The analysis indicates that significant fuel degradation via fuel heat-up, clad oxidation, and hydrogen generation was delayed up to an hour if FeCrAl alloy was used as a clad and canister material. And, combined with the passive safety systems (i.e., the Reactor Core Isolation Cooling system, RCIC), the extended operation of these systems delayed fuel degradation further. However, an adverse effect should be emphasized for the monolithic FeCrAl design—it generated more hydrogen than the designs based on the Zircaloy due to the high reaction rate at a high temperature of FeCrAl. The design of the FeCrAl-coated-Zircaloy avoids this defect. Therefore, it is a promising choice to combine some of the beneficial traits of both materials.

ACS Style

Jun Wang; Ryan Dailey; Zhehua Guo; Yukun Zhou; Paul Humrickhouse; Michael L. Corradini. Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC. Nuclear Engineering and Design 2020, 371, 110974 .

AMA Style

Jun Wang, Ryan Dailey, Zhehua Guo, Yukun Zhou, Paul Humrickhouse, Michael L. Corradini. Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC. Nuclear Engineering and Design. 2020; 371 ():110974.

Chicago/Turabian Style

Jun Wang; Ryan Dailey; Zhehua Guo; Yukun Zhou; Paul Humrickhouse; Michael L. Corradini. 2020. "Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC." Nuclear Engineering and Design 371, no. : 110974.

Journal article
Published: 12 September 2020 in Annals of Nuclear Energy
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The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th, 2011 put the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions and core meltdown due to station blackout. Although research efforts have been made by various parties to study the accident scenarios since the Fukushima accident, there remain unresolved issues regarding the core degradation behavior suggested by measurement data such as water level, Reactor Pressure Vessel (RPV) pressure and Primary Containment Vessel (PCV) pressure. To analyze and resolve such issues would be helpful to promote further understanding of the severe accident scenario at Fukushima units as well as the decommissioning work undergoing. The current study focuses on a detailed analysis of the RPV pressure peak event that occurred in Unit-3 at 12:00 on March 13th 2011. Sensitivity analysis cases were carried out with MELCOR 2.2 codei with sensitivity parameters that can influence the RPV pressure behavior, such as the debris quenching heat transfer coefficient, the number of opening SRVs during the RPV pressure peak event, amount of core slumping and particulate debris diameter. The cases that could reproduce the RPV pressure peak were further discussed to show likely debris bed energy history and the water mass history in the lower plenum during the RPV pressure peak event. The current study suggests that 1) Opening of SRVs equivalent to the total area of 4–6 fully-open SRVs (or equivalent leak area) could have occurred during the pressurization phase of the RPV accompanied by heavy debris quenching effect, while the opening of SRVs equivalent to a total area of at least 2 fully-open SRVs (or equivalent leak area) could have occurred during the depressurization phase of the RPV accompanied by moderate debris quenching effect. 2) The particulate debris diameter is not a very sensitive parameter when evaluating the debris quenching effect of Unit-3 in the current MELCOR modeling. 3) The current modeling suggests that around 70–110 GJ of energy can be removed by coolant during the debris quenching period with 30 tons of water reduction from the lower plenum.

ACS Style

Xin Li; Ikken Sato; Akifumi Yamaji; Mariko Regalado; Jun Wang. Sensitivity analysis of core slumping and debris quenching behavior of Fukushima Daiichi Unit-3 accident. Annals of Nuclear Energy 2020, 150, 107819 .

AMA Style

Xin Li, Ikken Sato, Akifumi Yamaji, Mariko Regalado, Jun Wang. Sensitivity analysis of core slumping and debris quenching behavior of Fukushima Daiichi Unit-3 accident. Annals of Nuclear Energy. 2020; 150 ():107819.

Chicago/Turabian Style

Xin Li; Ikken Sato; Akifumi Yamaji; Mariko Regalado; Jun Wang. 2020. "Sensitivity analysis of core slumping and debris quenching behavior of Fukushima Daiichi Unit-3 accident." Annals of Nuclear Energy 150, no. : 107819.

Journal article
Published: 07 May 2020 in Nuclear Engineering and Design
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Accident Tolerant Fuel (ATF) has the potential to improve the inherent safety of current reactors. In particular, an ATF Cr-coated Zirconium alloy (Cr-coated-Zr) cladding can be an effective short-term approach to improve the cladding oxidation resistance without changing the cladding base materials. In order to appropriately assess the effect of such an ATF concept it is important to quantify the performance of such a cladding improvement not only during normal operation but also during postulated accidents. In this work, we examine the effect of Cr-coated-Zr cladding (Cr-Zr) during a short-term station blackout (STSBO) accident as well as quantify the uncertainty in the use of ATF Cr-coated-Zr cladding. This is accomplished using a coated cladding model developed for the MELCOR severe accident systems code; i.e., MELCOR-1.8.6-UDGC model. First, we quantify the effect of the coated cladding using a set of base case simulations that show the improved oxidation resistance of the Cr-coated-Zr cladding with the UDGC model parameters. Then, we quantify the uncertainty in these simulations by a Monte-Carlo analysis that examines the effect of the ATF clad input parameters (e.g., oxidation rate and failure temperature) on key output parameters; i.e., timing of rapid hydrogen generation, clad temperatures and hot leg creep rupture. Finally, we use regression analysis methods to estimate the importance of these input parameters. These analyses show that the Cr-coated-Zr cladding can delay rapid clad oxidation and generation of hydrogen by 3600 s to 7200 s with associated delays in rapid clad temperature increases. This additional time can be of benefit for compensatory operator actions. In contrast there is little effect on the timing of hot leg creep rupture since this failure is mainly controlled by core decay heat. We also find that these results are not significantly affected by the uncertainty in input parameters, and based on the regression analysis, the high temperature Zr oxidation rate has the major influence on the selected output parameters for the Cr-coated-Zr cladding.

ACS Style

Tangtao Feng; Jun Wang; Yimin Zhou; Ping Song; Mingjun Wang; Ryan Dailey; Wenxi Tian; Michael L. Corradini. Quantification of the effect of Cr-coated-Zircaloy cladding during a short term station black out. Nuclear Engineering and Design 2020, 363, 110678 .

AMA Style

Tangtao Feng, Jun Wang, Yimin Zhou, Ping Song, Mingjun Wang, Ryan Dailey, Wenxi Tian, Michael L. Corradini. Quantification of the effect of Cr-coated-Zircaloy cladding during a short term station black out. Nuclear Engineering and Design. 2020; 363 ():110678.

Chicago/Turabian Style

Tangtao Feng; Jun Wang; Yimin Zhou; Ping Song; Mingjun Wang; Ryan Dailey; Wenxi Tian; Michael L. Corradini. 2020. "Quantification of the effect of Cr-coated-Zircaloy cladding during a short term station black out." Nuclear Engineering and Design 363, no. : 110678.

Editorial article
Published: 10 March 2020 in Frontiers in Energy Research
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Editorial: Nuclear Safety Design and Innovation

ACS Style

Jun Wang; Kaiyi Shi; Zhaoming Meng; Shripad T. Revankar. Editorial: Nuclear Safety Design and Innovation. Frontiers in Energy Research 2020, 8, 1 .

AMA Style

Jun Wang, Kaiyi Shi, Zhaoming Meng, Shripad T. Revankar. Editorial: Nuclear Safety Design and Innovation. Frontiers in Energy Research. 2020; 8 ():1.

Chicago/Turabian Style

Jun Wang; Kaiyi Shi; Zhaoming Meng; Shripad T. Revankar. 2020. "Editorial: Nuclear Safety Design and Innovation." Frontiers in Energy Research 8, no. : 1.

Technical papers
Published: 22 August 2019 in Nuclear Technology
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Since the accident at Fukushima, one major goal of reactor safety research has been the development of accident tolerant technologies that can mitigate or delay fuel degradation during a beyond-design-basis accident. One major effort has been focused on the development of coatings for light water reactor fuel cladding. Chromium-coated zirconium-alloy clad is one of the leading options. In this work, the MELCOR systems code (version 1.8.6 User-Defined Generalized Coating) is used to evaluate the performance of Cr-coated Zr-alloy clad as compared to Zr-alloy clad and APMT FeCrAl-coated Zr-alloy clad for a pressurized water reactor (i.e., Surry) for a station blackout (SBO) accident scenario. Our focus is primarily on the accident progression behavior depending on oxidation kinetics and the assumed failure criterion for the coated cladding material. Our simulation and comparison indicate that the presence of the coating material can significantly reduce the initial rate of hydrogen generation and delay the time when hydrogen generation becomes significant. This decrease in the rate of oxidation and delay in timing can provide additional coping time for compensatory operator actions. We also note that the effect of extended auxiliary feedwater system operation (long-term SBO) can increase this additional coping time in combination with Cr-coated Zr-alloy, but it is limited by other primary system failures (e.g., hot-leg creep rupture) that will occur driven by core decay heat and independent of coated cladding effects. Finally, we observe that while the initial suppression of hydrogen generation for Cr-coated Zr-alloy clad compared to Zr-alloy is notable, the overall amount of hydrogen produced is similar since hydrogen can also be produced through competing oxidation of stainless steel components during the accident progression. Our future work is focused on the uncertainty analysis of the oxidation rate data, coating failure criteria, and severe accident modeling limitations in order to better quantify accident tolerant fuel clad benefits.

ACS Style

J. Wang; H. Yeom; P. Humrickhouse; K. Sridharan; M. Corradini. Effectiveness of Cr-Coated Zr-Alloy Clad in Delaying Fuel Degradation for a PWR During a Station Blackout Event. Nuclear Technology 2019, 206, 467 -477.

AMA Style

J. Wang, H. Yeom, P. Humrickhouse, K. Sridharan, M. Corradini. Effectiveness of Cr-Coated Zr-Alloy Clad in Delaying Fuel Degradation for a PWR During a Station Blackout Event. Nuclear Technology. 2019; 206 (3):467-477.

Chicago/Turabian Style

J. Wang; H. Yeom; P. Humrickhouse; K. Sridharan; M. Corradini. 2019. "Effectiveness of Cr-Coated Zr-Alloy Clad in Delaying Fuel Degradation for a PWR During a Station Blackout Event." Nuclear Technology 206, no. 3: 467-477.

Journal article
Published: 30 May 2019 in Nuclear Engineering and Design
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Five published Liquid Sublayer Dryout (LSD) models are assessed using the Groeneveld Look-up table 2006 (LUT-2006) and critical heat flux (CHF) experimental data in circular tube under both subcooled boiling conditions and saturated boiling conditions. The flow regime transition from inverted annular flow to dispersed flow at post-CHF is postulated as the upper limit of the LSD model application range in saturated boiling region. Lee and Mudawar’s LSD model modified by W.X. Liu shows good agreement when void fractions is lower than 0.7. The modified L&M’s LSD model is coupled into subchannel code, COBRA-TF, to predict DNB type CHF in rod bundles. In present study, the gap clearance or thermal equivalent diameter of rod bundles is adopted as characteristic length in the modified L&M’s LSD model. For rod bundles with simplified grid spacers, Karman velocity distribution equation is utilized to calculate the velocity distribution normal to pin wall. The rod bundle CHF predictions by COBRA-TF coupled with LUT-2006, Bowring’s CHF correlation and the modified L&M’s LSD model, are compared with the CHF experimental data in 2 × 2 rod bundles with non-uniform power profile. Based on reasonable predictability on both the input power and axial position at CHF, the deviations of CHF predictions by the modified L&M’s LSD model, which adopts gap clearance instead of thermal equivalent diameter as characteristic length, and Bowring’s CHF correlation are within ±20% as void fraction approaching to 0.7.

ACS Style

Dawei Zhao; Juliana P. Duarte; Wenxing Liu; Michael L. Corradini; Jun Wang; Jingliang Bi. DNB type critical heat flux prediction in rod bundles with simplified grid spacer based on Liquid Sublayer Dryout model. Nuclear Engineering and Design 2019, 351, 94 -105.

AMA Style

Dawei Zhao, Juliana P. Duarte, Wenxing Liu, Michael L. Corradini, Jun Wang, Jingliang Bi. DNB type critical heat flux prediction in rod bundles with simplified grid spacer based on Liquid Sublayer Dryout model. Nuclear Engineering and Design. 2019; 351 ():94-105.

Chicago/Turabian Style

Dawei Zhao; Juliana P. Duarte; Wenxing Liu; Michael L. Corradini; Jun Wang; Jingliang Bi. 2019. "DNB type critical heat flux prediction in rod bundles with simplified grid spacer based on Liquid Sublayer Dryout model." Nuclear Engineering and Design 351, no. : 94-105.

Journal article
Published: 20 November 2018 in Progress in Nuclear Energy
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A new simple core degradation experiment will be conducted to investigate the distribution of mass and energy during a core degradation process, and the International Standard Problem (ISP) No.31 is chosen as a pre-simulation for the new experiment. The pre-simulation has been conducted using the widely accepted severe accident analysis software MELCOR. Firstly, the numerical analysis model and oxidation model are described, and all the input parameters are in accord with experiment conditions. Then, numerical results are validated by experimental measurements and SCDAP/RELAP5 results. Simulations results agree well with measured data. It is indicated that MELCOR has the capability of predicting the behaviors of fuel elements in reflood correctly. Finally, the visually spatial temperature distribution is obtained by TECPLOT, and the behaviors of molten fuel elements are directly reflected. The evolution of peak temperature in fuel rods during the experiment period can be visible. The peak temperature firstly appeared in outer heated rods of the fourth ring and later showed in a heated fuel rod of the second ring. The behaviors of molten fuel elements are visible in the figures of spatial temperature distribution, and it does help researchers to understand the migration behaviors of molten material accompanied by effective mitigation measures for a severe accident.

ACS Style

Tangtao Feng; Wenxi Tian; Ping Song; Jun Wang; Mingjun Wang; Longze Li; G.H. Su; Suizheng Qiu. Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment. Progress in Nuclear Energy 2018, 111, 174 -182.

AMA Style

Tangtao Feng, Wenxi Tian, Ping Song, Jun Wang, Mingjun Wang, Longze Li, G.H. Su, Suizheng Qiu. Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment. Progress in Nuclear Energy. 2018; 111 ():174-182.

Chicago/Turabian Style

Tangtao Feng; Wenxi Tian; Ping Song; Jun Wang; Mingjun Wang; Longze Li; G.H. Su; Suizheng Qiu. 2018. "Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment." Progress in Nuclear Energy 111, no. : 174-182.

Journal article
Published: 10 September 2018 in Journal of Nuclear Engineering and Radiation Science
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Accident tolerant fuels (ATF) and steam generator (SG) auxiliary feedwater (AFW) extended operation are two important methods to increase the coping time for nuclear power plant safety response. In light of recent efforts to investigate such methods, we investigate both FeCrAl cladding oxidation kinetics and SG AFW sensitivity analyses, for the Surry nuclear power plant Short-Term Station Blackout simulation using the MELCOR YR 1.8.6 systems code. The first part describes the effects of FeCrAl cladding oxidation kinetics. Zircaloy cladding and two different oxidation models of FeCrAl cladding are compared. The initial hydrogen generation time (>0.5 kg) is used as the evaluation criterion for fuel degradation in a severe accident. Results showed that the more recent oxidation correlation by ORNL predicts much less hydrogen generation than Zircaloy cladding. The second part investigates the effects of three different methods of AFW injection into the SG secondary side. We considered three different methods of water injection; i.e., constant water injection into the secondary side (case 1); water injection based on secondary side water level in boiler region (case 2); water injection based on secondary side water level in the downcomer region (case 3). The case of constant water injection is the most straightforward, but it would have the tendency to overfill the SG with excess water. Water injection with downcomer level control is more reasonable but requires DC power to monitor level and to control AFW injection rate.

ACS Style

Jun Wang; Mckinleigh McCabe; Troy Christopher Haskin; Yingwei Wu; Guanghui Su; Michael L. Corradini. Iron–Chromium–Aluminum (FeCrAl) Cladding Oxidation Kinetics and Auxiliary Feedwater Sensitivity Analysis—Short-Term Station Blackout Simulation of Surry Nuclear Power Plant. Journal of Nuclear Engineering and Radiation Science 2018, 4, 041002 .

AMA Style

Jun Wang, Mckinleigh McCabe, Troy Christopher Haskin, Yingwei Wu, Guanghui Su, Michael L. Corradini. Iron–Chromium–Aluminum (FeCrAl) Cladding Oxidation Kinetics and Auxiliary Feedwater Sensitivity Analysis—Short-Term Station Blackout Simulation of Surry Nuclear Power Plant. Journal of Nuclear Engineering and Radiation Science. 2018; 4 (4):041002.

Chicago/Turabian Style

Jun Wang; Mckinleigh McCabe; Troy Christopher Haskin; Yingwei Wu; Guanghui Su; Michael L. Corradini. 2018. "Iron–Chromium–Aluminum (FeCrAl) Cladding Oxidation Kinetics and Auxiliary Feedwater Sensitivity Analysis—Short-Term Station Blackout Simulation of Surry Nuclear Power Plant." Journal of Nuclear Engineering and Radiation Science 4, no. 4: 041002.

Articles
Published: 05 September 2018 in Energy Sources, Part A: Recovery, Utilization, and Environmental Effects
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In order to improve the flotation performance of the coking coal particles, the flotation tests of the coking coal particles were conducted in the direct flotation, reserve flotation, and reverse-direct flotation processes. It was found that the concentrate ash content of coking coal particles was higher than 20%, which cannot be effectively reduced using the direct and reserve flotation processes. However, the flotation concentrate with the ash content of 12.53% can be obtained from the reverse-direct flotation process. In the reverse-direct flotation process, the surface hydrophobicity was reduced with the dextrin and 1-dodecylamine (DDA) addition at the reverse flotation process stage. For the addition of diesel collectors at the direct flotation process stage, the surface hydrophobicity of the coking coal samples was improved.

ACS Style

Kaiyi Shi; Jun Wang; Shiwei Wang; Zhongming Yu; Peng Chen; Shuai Li. Improving the flotation performance of coking coal using the reverse-direct flotation process. Energy Sources, Part A: Recovery, Utilization, and Environmental Effects 2018, 40, 2886 -2894.

AMA Style

Kaiyi Shi, Jun Wang, Shiwei Wang, Zhongming Yu, Peng Chen, Shuai Li. Improving the flotation performance of coking coal using the reverse-direct flotation process. Energy Sources, Part A: Recovery, Utilization, and Environmental Effects. 2018; 40 (23):2886-2894.

Chicago/Turabian Style

Kaiyi Shi; Jun Wang; Shiwei Wang; Zhongming Yu; Peng Chen; Shuai Li. 2018. "Improving the flotation performance of coking coal using the reverse-direct flotation process." Energy Sources, Part A: Recovery, Utilization, and Environmental Effects 40, no. 23: 2886-2894.

Journal article
Published: 01 September 2018 in Experimental Thermal and Fluid Science
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ACS Style

Zhiyuan Lu; Zhaoming Meng; Ningxin Gu; Jinpu Wang; Feifei Nian; Jun Wang; Ming Ding. Development of correlations for liquid entrainment through a large-scale inclined branch pipe connected to the main horizontal pipe. Experimental Thermal and Fluid Science 2018, 96, 128 -136.

AMA Style

Zhiyuan Lu, Zhaoming Meng, Ningxin Gu, Jinpu Wang, Feifei Nian, Jun Wang, Ming Ding. Development of correlations for liquid entrainment through a large-scale inclined branch pipe connected to the main horizontal pipe. Experimental Thermal and Fluid Science. 2018; 96 ():128-136.

Chicago/Turabian Style

Zhiyuan Lu; Zhaoming Meng; Ningxin Gu; Jinpu Wang; Feifei Nian; Jun Wang; Ming Ding. 2018. "Development of correlations for liquid entrainment through a large-scale inclined branch pipe connected to the main horizontal pipe." Experimental Thermal and Fluid Science 96, no. : 128-136.

Editorial article
Published: 10 August 2018 in Frontiers in Energy Research
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Editorial: Nuclear Thermal Hydraulic and Two-Phase Flow

ACS Style

Jun Wang; Kaiyi Shi; Zhaoming Meng; Shripad T. Revankar. Editorial: Nuclear Thermal Hydraulic and Two-Phase Flow. Frontiers in Energy Research 2018, 6, 1 .

AMA Style

Jun Wang, Kaiyi Shi, Zhaoming Meng, Shripad T. Revankar. Editorial: Nuclear Thermal Hydraulic and Two-Phase Flow. Frontiers in Energy Research. 2018; 6 ():1.

Chicago/Turabian Style

Jun Wang; Kaiyi Shi; Zhaoming Meng; Shripad T. Revankar. 2018. "Editorial: Nuclear Thermal Hydraulic and Two-Phase Flow." Frontiers in Energy Research 6, no. : 1.

Short communication
Published: 23 July 2018 in Annals of Nuclear Energy
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ACS Style

Xiaomeng Dong; Juliana P. Duarte; Dong Liu; Jun Wang; Zhijian Zhang; Zhaofei Tian. Numerical investigation of azimuthal heat conduction effects on CHF phenomenon in rod bundle channel. Annals of Nuclear Energy 2018, 121, 203 -209.

AMA Style

Xiaomeng Dong, Juliana P. Duarte, Dong Liu, Jun Wang, Zhijian Zhang, Zhaofei Tian. Numerical investigation of azimuthal heat conduction effects on CHF phenomenon in rod bundle channel. Annals of Nuclear Energy. 2018; 121 ():203-209.

Chicago/Turabian Style

Xiaomeng Dong; Juliana P. Duarte; Dong Liu; Jun Wang; Zhijian Zhang; Zhaofei Tian. 2018. "Numerical investigation of azimuthal heat conduction effects on CHF phenomenon in rod bundle channel." Annals of Nuclear Energy 121, no. : 203-209.

Technical papers
Published: 11 June 2018 in Nuclear Technology
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Accident-tolerant fuel (ATF) cladding materials have been a focus of recent work to provide a greater resistance to fuel degradation, oxidation, and melting in light water reactors for beyond-design accident scenarios such as a station blackout (SBO). In a previous study, researchers at The University of Wisconsin–Madison used the Surry Nuclear Plant as the pilot plant to examine the effect of ATF substitute clad materials with the short-term SBO as the postulated accident, examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. In this work, we examine the effect of recovery actions for an SBO in Surry as a follow-on topic. Specifically, we selected two kinds of core cladding materials (Zircaloy and FeCrAl), and then conducted comparative analysis of the effect of water injection; first with a delay in water injection start times into the reactor pressure vessel (RPV) and then with steam generator (SG) steam-side AFW end times. We find that alternative cladding materials (FeCrAl) can effectively delay fuel degradation and system failures for both water injection strategies. One finds that RPV water injection can prevent such severe accident effects if restored in a few hours into the SBO. Conversely, SG steam-side AFW flow with alternative cladding materials (FeCrAl) can delay the fuel degradation and system failure processes by hours. We mainly focus on analyzing the severe accident progression by different quantitative signals, such as the onset of rapid hydrogen production, hot-leg creep rupture failure, and core slump. Analyses are now underway to consider the effects of proposed coating materials on Zircaloy cladding and if such coatings can afford similar benefits.

ACS Style

J. Wang; H. J. Jo; M. L. Corradini. Potential Recovery Actions from a Severe Accident in a PWR: MELCOR Analysis of a Station Blackout Scenario. Nuclear Technology 2018, 204, 1 -14.

AMA Style

J. Wang, H. J. Jo, M. L. Corradini. Potential Recovery Actions from a Severe Accident in a PWR: MELCOR Analysis of a Station Blackout Scenario. Nuclear Technology. 2018; 204 (1):1-14.

Chicago/Turabian Style

J. Wang; H. J. Jo; M. L. Corradini. 2018. "Potential Recovery Actions from a Severe Accident in a PWR: MELCOR Analysis of a Station Blackout Scenario." Nuclear Technology 204, no. 1: 1-14.