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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Korean Nuclear Society
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Volume & Issues
Volume 45, Issue 7 - Dec 2013
Volume 45, Issue 6 - Nov 2013
Volume 45, Issue 5 - Oct 2013
Volume 45, Issue 4 - Aug 2013
Volume 45, Issue 3 - Jun 2013
Volume 45, Issue 2 - Apr 2013
Volume 45, Issue 1 - Feb 2013
Selecting the target year
OECD/NEA STUDY ON THE ECONOMICS AND MARKET OF SMALL REACTORS
Lokhov, Alexey ; Cameron, Ron ; Sozoniuk, Vladislav ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 701~706
DOI : 10.5516/NET.02.2013.517
According to the OECD/NEA estimates, nuclear power plants (NPPs), whether with a large reactor or with small modular reactors (SMRs), are competitive with many other electricity generation technologies in a significant number of cases, one of the exceptions being natural gas in the USA with the current level of prices. However, SMRs have particular features and requirements setting conditions for their deployment. This paper presents the preliminary analysis by OECD/NEA of the economics, opportunities, and market for small nuclear reactors.
DIFFUSION PIECEWISE HOMOGENIZATION VIA FLUX DISCONTINUITY RATIOS
Sanchez, Richard ; Dante, Giorgio ; Zmijarevic, Igor ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 707~720
DOI : 10.5516/NET.02.2013.518
We analyze piecewise homogenization with flux-weighted cross sections and preservation of averaged currents at the boundary of the homogenized domain. Introduction of a set of flux discontinuity ratios (FDR) that preserve reference interface currents leads to preservation of averaged region reaction rates and fluxes. We consider the class of numerical discretizations with one degree of freedom per volume and per surface and prove that when the homogenization and computing meshes are equal there is a unique solution for the FDRs which exactly preserve interface currents. For diffusion submeshing we introduce a Jacobian-Free Newton-Krylov method and for all cases considered obtain an 'exact' numerical solution (eight digits for the interface currents). The homogenization is completed by extending the familiar full assembly homogenization via flux discontinuity factors to the sides of regions laying on the boundary of the piecewise homogenized domain. Finally, for the familiar nodal discretization we numerically find that the FDRs obtained with no submesh (nearly at no cost) can be effectively used for whole-core diffusion calculations with submesh. This is not the case, however, for cell-centered finite differences.
STATUS OF THE ASTRID CORE AT THE END OF THE PRE-CONCEPTUAL DESIGN PHASE 1
Chenaud, Ms. ; Devictor, N. ; Mignot, G. ; Varaine, F. ; Venard, C. ; Martin, L. ; Phelip, M. ; Lorenzo, D. ; Serre, F. ; Bertrand, F. ; Alpy, N. ; Le Flem, M. ; Gavoille, P. ; Lavastre, R. ; Richard, P. ; Verrier, D. ; Schmitt, D. ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 721~730
DOI : 10.5516/NET.02.2013.519
Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase.
TERRAPOWER, LLC TRAVELING WAVE REACTOR DEVELOPMENT PROGRAM OVERVIEW
Hejzlar, Pavel ; Petroski, Robert ; Cheatham, Jesse ; Touran, Nick ; Cohen, Michael ; Truong, Bao ; Latta, Ryan ; Werner, Mark ; Burke, Tom ; Tandy, Jay ; Garrett, Mike ; Johnson, Brian ; Ellis, Tyler ; Mcwhirter, Jon ; Odedra, Ash ; Schweiger, Pat ; Adkisson, Doug ; Gilleland, John ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 731~744
DOI : 10.5516/NET.02.2013.520
Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.
VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE
Tak, Nam-Il ; Kim, Min-Hwan ; Lim, Hong-Sik ; Noh, Jae Man ; Drzewiecki, Timothy J. ; Seker, Volkan ; Downar, Thomas J. ; Kelly, Joseph ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 745~752
DOI : 10.5516/NET.02.2013.521
For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.
DEVELOPMENT OF CALCULATION METHOD OF SENSITIVITIES FOR LIGHT WATER REACTORS
Takeda, Toshikazu ; Foad, Basma ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 753~758
DOI : 10.5516/NET.02.2013.522
A new method of calculating sensitivity coefficients of core characteristics relative to infinite-dilution cross sections has been developed. Conventional sensitivity coefficients are evaluated for the changes of effective cross sections which are dependent on individual models of core and cell. Therefore a correction has been derived to the conventional sensitivity coefficients based on the perturbation theory. The accuracy of the present method has been verified by comparing numerical results of sensitivity coefficients with a reference Monte-Carlo method.
ASSESSMENT OF CONDENSATION HEAT TRANSFER MODEL TO EVALUATE PERFORMANCE OF THE PASSIVE AUXILIARY FEEDWATER SYSTEM
Cho, Yun-Je ; Kim, Seok ; Bae, Byoung-Uhn ; Park, Yusun ; Kang, Kyoung-Ho ; Yun, Byong-Jo ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 759~766
DOI : 10.5516/NET.02.2013.523
As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.
APPLICATION OF UNCERTAINTY ANALYSIS TO MAAP4 ANALYSES FOR LEVEL 2 PRA PARAMETER IMPORTANCE DETERMINATION
Roberts, Kevin ; Sanders, Robert ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 767~790
DOI : 10.5516/NET.02.2013.524
MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.
ESTABLISHMENT OF A MAINTENANCE PROGRAM TO PREVENT LOSS OF OFFSITE POWER IN NUCLEAR POWER PLANTS
Lee, Eun-Chan ; Na, Jang-Hwan ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 791~794
DOI : 10.5516/NET.02.2013.525
Since the Fukushima accident in 2011, the importance of the electrical systems in nuclear power plants (NPPs) has been emphasized. The result has been that NPP regulators are enhancing their monitoring of loss of offsite power (LOOP) events. Korea Hydro & Nuclear Power Co. (KHNP) is reviewing the status and issues related to LOOPs, and is attempting to establish specific countermeasures to prevent LOOPs, because they can have severe consequences in the complicated maintenance schedule during an outage. A starting point for preventing LOOPs is the control of the loss of voltage (LOV)-initiating components. In order to reflect this in the risk assessment program, an LOV monitor is being developed for use during plant outages.
FAULT-TOLERANT DESIGN FOR ADVANCED DIVERSE PROTECTION SYSTEM
Oh, Yang Gyun ; Jeong, Kin Kwon ; Lee, Chang Jae ; Lee, Yoon Hee ; Baek, Seung Min ; Lee, Sang Jeong ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 795~802
DOI : 10.5516/NET.02.2013.526
For the improvement of APR1400 Diverse Protection System (DPS) design, the Advanced DPS (ADPS) has recently been developed to enhance the fault tolerance capability of the system. Major fault masking features of the ADPS compared with the APR1400 DPS are the changes to the channel configuration and reactor trip actuation equipment. To minimize the fault occurrences within the ADPS, and to mitigate the consequences of common-cause failures (CCF) within the safety I&C systems, several fault avoidance design features have been applied in the ADPS. The fault avoidance design features include the changes to the system software classification, communication methods, equipment platform, MMI equipment, etc. In addition, the fault detection, location, containment, and recovery processes have been incorporated in the ADPS design. Therefore, it is expected that the ADPS can provide an enhanced fault tolerance capability against the possible faults within the system and its input/output equipment, and the CCF of safety systems.
U.S. FUEL CYCLE TECHNOLOGIES R&D PROGRAM FOR NEXT GENERATION NUCLEAR MATERIALS MANAGEMENT
Miller, M.C. ; Vega, D.A. ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 803~810
DOI : 10.5516/NET.02.2013.527
The U.S. Department of Energy's Fuel Cycle Technologies R&D program under the Office of Nuclear Energy is working to advance technologies to enhance both the existing and future fuel cycles. One thrust area is in developing enabling technologies for next generation nuclear materials management under the Materials Protection, Accounting and Control Technologies (MPACT) Campaign where advanced instrumentation, analysis and assessment methods, and security approaches are being developed under a framework of Safeguards and Security by Design. An overview of the MPACT campaign's activities and recent accomplishments is presented along with future plans.
THERMAL SHOCK FRACTURE OF SILICON CARBIDE AND ITS APPLICATION TO LWR FUEL CLADDING PERFORMANCE DURING REFLOOD
Lee, Youho ; Mckrell, Thomas J. ; Kazimi, Mujid S. ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 811~820
DOI : 10.5516/NET.02.2013.528
SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to
) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.
EFFECTS OF HEAT TREATMENTS ON MICROSTRUCTURES AND MECHANICAL PROPERTIES OF DUAL PHASE ODS STEELS FOR HIGH TEMPERATURE STRENGTH
Noh, Sanghoon ; Choi, Byoung-Kwon ; Han, Chang-Hee ; Kang, Suk Hoon ; Jang, Jinsung ; Jeong, Yong-Hwan ; Kim, Tae Kyu ;
Nuclear Engineering and Technology, volume 45, issue 6, 2013, Pages 821~826
DOI : 10.5516/NET.02.2013.529
In the present study, the effects of various heat treatments on the microstructure and mechanical properties of dual phase ODS steels were investigated to enhance the high strength at elevated temperature. Dual phase ODS steels have been designed by the control of ferrite and austenite formers, i.e., Cr, W and Ni, C in Fe-based alloys. The ODS steels were fabricated by mechanical alloying and a hot isostatic pressing process. Heat treatments, including hot rolling-tempering and normalizing-tempering with air- and furnace-cooling, were carefully carried out. It was revealed that the grain size and oxide distributions of the ODS steels can be changed by heat treatment, which significantly affected the strengths at elevated temperature. Therefore, the high temperature strength of dual phase ODS steel can be enhanced by a proper heat treatment process with a good combination of ferrite grains, nano-oxide particles, and grain boundary sliding.