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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 37, Issue 6 - Dec 2005
Volume 37, Issue 5 - Oct 2005
Volume 37, Issue 4 - Aug 2005
Volume 37, Issue 3 - Jun 2005
Volume 37, Issue 2 - Apr 2005
Volume 37, Issue 1 - Feb 2005
Selecting the target year
REACTOR PHYSICS CHALLENGES IN GEN-IV REACTOR DESIGN
DRISCOLL MICHAEL J. ; HEJZLAR PAVEL ;
Nuclear Engineering and Technology, volume 37, issue 1, 2005, Pages 1~10
An overview of the reactor physics aspects of Generation Four(GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and ecoomics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.
STATUS AND PERSPECTIVE OF NUCLEAR DATA PRODUCTION, EVALUATION AND VALIDATION
TRKOV A. ;
Nuclear Engineering and Technology, volume 37, issue 1, 2005, Pages 11~24
A very important feature in the development of nuclear technology has been and will continue to be the flow of information from nuclear data production to the various applications fields in nuclear technology. Both, nuclear data and this communications flow are defined in this paper. Nuclear data result from specific technical activities including their production, evaluation, processing, verification, validation and applications. These activities are described, focusing on nuclear reactor calculations. Mathematical definitions of different types of nuclear data are introduced, and international forums involved in nuclear data activities are listed. Electronic links to various sources of information available on the web are specified, whenever possible.
FUNDAMENTALS AND RECENT DEVELOPMENTS OF REACTOR PHYSICS METHODS
CHO NAM ZIN ;
Nuclear Engineering and Technology, volume 37, issue 1, 2005, Pages 25~78
As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming. After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows: (1) Derivation of the multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (now almost three decades old) diffusion nodal methods. (3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues. (4) Description of the analytic function expansion nodal (AFEN) method. (5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods. (6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis. (7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.
EVOLUTION OF NUCLEAR FUEL MANAGEMENT AND REACTOR OPERATIONAL AID TOOLS
TURINSKY PAUL J. ; KELLER PAUL M. ; ABDEL-KHALIK HANY S. ;
Nuclear Engineering and Technology, volume 37, issue 1, 2005, Pages 79~90
In this paper are reviewed the current status of nuclear fuel management and reactor operational aid tools. In addition, we indicate deficiencies in current capabilities and what future research is judged warranted. For the nuclear fuel management review the focus is on light water reactors and the utilization of stochastic optimization methods applied to the lattice, fuel bundle, core loading pattern, and for BWRs the control rod pattern/core flow design decision making problems. Significant progress in addressing separately each of these design problems on a single cycle basis is noted; however, the outstanding challenge of addressing the integrated design problem over multiple cycles under conditions of uncertainty remains to be addressed. For the reactor operational aid tools review the focus is on core simulators, used to both process core instrumentation signals and as an operator aid to predict future core behaviors under various operational strategies. After briefly reviewing the current status of capabilities, a more in depth review of adaptive core simulation capabilities, where core simulator input data are adjusted within their known uncertainties to improved agreement between prediction and measurement, is presented. This is done in support of the belief that further development of adaptive core simulation capabilities is required to further significantly advance the utility of core simulators in support of reactor operational aid tools.
CORE DESIGN FOR HETEROGENEOUS THORIUM FUEL ASSEMBLIES FOR PWR(1)-NUCLEAR DESIGN AND FUEL CYCLE ECONOMY
BAE KANG-MOK ; KIM MYUNG-HYUN ;
Nuclear Engineering and Technology, volume 37, issue 1, 2005, Pages 91~100
Kyung-hee Thorium Fuel (KTF), a heterogeneous thorium-based seed and blanket design concept for pressurized light water reactors, is being studied as an alternative to enhance proliferation resistance and fuel cycle economics of PWRs. The proliferation resistance characteristics of the KTF assembly design were evaluated through parametric studies using neutronic performance indices such as Bare Critical Mass (BCM), Spontaneous Neutron Source rate (SNS), Thermal Generation rate (TG), and Radio-Toxicity. Also, Fissile Economic Index (FEI), a new index for gauging fuel cycle economy, was suggested and applied to optimize the KTF design. A core loaded with optimized KTF assemblies with a seed-to-blanket ratio of 1: 1 was tested at the Korea Next Generation Reactor (KNGR), ARP-1400. Core design characteristics for cycle length, power distribution, and power peaking were evaluated by HELIOS and MASTER code systems for nine reload cycles. The core calculation results show that the KTF assembly design has nearly the same neutronic performance as those of a conventional
fuel assembly. However, the power peaking factor is relatively higher than that of conventional PWRs as the maximum Fq is 2.69 at the M
equilibrium cycle while the design limit is 2.58. In order to assess the economic potential of a heterogeneous thorium fuel core, the front-end fuel cycle costs as well as the spent fuel disposal costs were compared with those of a reference PWR fueled with
. In the case of comprising back-end fuel cycle cost, the fuel cycle cost of APR-1400 with a KTF assembly is 4.99 mills/KWe-yr, which is lower than that (5.23 mills/KWe-yr) of a conventional PWR. Proliferation resistance potential, BCM, SNS, and TG of a heterogeneous thorium-fueled core are much higher than those of the
core. The once-through fuel cycle application of heterogeneous thorium fuel assemblies demonstrated good competitiveness relative to
in terms of economics.
NEUTRON CROSS SECTION DATA LIBRARY FOR PD-105, AG-109, XE-131 AND CS-133
LEE Y. D. ; CHANG J. H. ;
Nuclear Engineering and Technology, volume 37, issue 1, 2005, Pages 101~108
The neutron induced nuclear cross-section data for Pd-105, Ag-109, Xe-131, and Cs-133 were calculated and evaluated from an unresolved energy to 20 MeV. The energy dependent optical model potential parameters were extracted based on recent experimental data and applied up to 20 MeV. A spherical optical model and a statistical model for the equilibrium energy, and a multistep direct and a multistep compound model for the pre-equilibrium energy were used in the calculation. The direct capture model was recently introduced for fast neutron capture. The theoretically calculated cross-sections were compared with the experimental data and the evaluated files. The total and capture cross-sections calculated using the model were in good agreement with the reference experimental data. The evaluated cross-section results were compiled in ENDF-6 format and merged with the resonance component, already adopted in the ENDF/B-VI release 8. New data library files covering from thermal to 20 MeV were created. They are at the preliminary stage of an ENDF/B- VII release.
A STUDY ON AN ASSESSMENT METHOD FOR IMPROVING TECHNICAL SPECIFICATIONS USING SYSTEM DYNAMICS
KANG KYUNG MIN ; JAE MOOSUNG ;
Nuclear Engineering and Technology, volume 37, issue 1, 2005, Pages 109~117
Limiting conditions for operations (LCOs) are evaluated dynamically using the tool of system dynamics. The LCOs de-fine the allowed outage times (AOTs) and the actions to be taken if the repair cannot be completed within the AOT. System dynamics has been developed to analyze the dynamic reliability of a complicated system. System dynamics using Vensim software have been applied to LCOs assessment for an example system, the auxiliary feed water system of a reference nuclear power plant. Analysis results of both full power operation and shutdown operation have been compared for a measure of core damage frequency. The framework developed in this study has been shown to be very flexible in that it can be applied to assess LCOs quantitatively under any operational context of the TS in FSAR.
A VALIDATION METHOD FOR EMERGENCY OPERATING PROCEDURES OF NUCLEAR POWER PLANTS BASED ON DYNAMIC MULTI-LEVEL FLOW MODELING
QIN WEI ; SEONG POONG HYUN ;
Nuclear Engineering and Technology, volume 37, issue 1, 2005, Pages 118~126
While emergency operating procedures (EOPs) occupy an important role in the management of various abnormal situations in nuclear power plants (NPPs), current technology for the validation of EOPs still largely depends on manual review. A validation method for EOPs of NPPs is thus proposed based on dynamic multi-level flow modeling (MFM). The MFM modeling procedure and the EOP validation procedure are developed and provided in the paper. Application of the proposed method to EOPs of an actual NPP shows that the proposed method provides an efficient means of validating EOPs. It is also found that the information on state transitions in MFM models during the management of abnormal situations is also useful for further analysis on EOPs including their optimization.