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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 28, Issue 6 - Dec 1996
Volume 28, Issue 5 - Oct 1996
Volume 28, Issue 4 - Aug 1996
Volume 28, Issue 3 - Jun 1996
Volume 28, Issue 2 - Apr 1996
Volume 28, Issue 1 - Feb 1996
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A Combined Procedure of RSM and LHS for Uncertainty Analyses of CsI Release Fraction Under a Hypothetical Severe Accident Sequence of Station Blackout at Younggwang Nuclear Power Plant Using MAAP3.0B Code
Han, Seok-Jung ; Tak, Nam-Il ; Chun, Moon-Hyun ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 507~521
Quantification of uncertainties in the source term estimations by a large computer code, such as MELCOR and MAAP, is an essential process of the current Probabilistic safety assessment. The main objective of the present study is to investigate the applicability of a combined procedure of the response surface method (RSM) based on input determined from a statistical design and the Latin hypercube sampling (LHS) technique for the uncertainty analysis of CsI release fractions under a Hypothetical severe accident sequence of a station blackout at Younggwang nuclear power plant using MAAP3. OB code as a benchmark problem. On the basis of the results obtained in the present work, the RSM is recommended to be used as a principal tool for an overall uncertainty analysis in source term quantifications, while using the LHS in the calculations of standardized regression coefficients (SRC) and standardized rank regression coefficient (SRRC) to determine the subset of the most important input parameters in the final screening step and to check the cumulative distribution functions obtained by RSM. Verification of the response surface model for its sufficient accuracy is a prerequisite for the reliability of the final results that can be obtained by the combined procedure proposed in the present work.
A Method of Knowledge Base Verification for Nuclear Power Plant Expert Systems Using Extended Petri Nets
Kwon, I.W. ; Seong, P.H. ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 522~531
The adoption of expert systems mainly as operator supporting systems is becoming increasingly popular as the control algorithms of system become more and more sophisticated and complicated. The verification phase of knowledge base is an important part for developing reliable expert systems, especially in nuclear industry. Although several strategies or tools have been developed to perform potential error checking, they often neglect the reliability of verification methods. Because a Petri net provides a uniform mathematical formalization of knowledge base, it has been employed for knowledge base verification. In this work, we devise and suggest an automated tool, called COKEP(Checker Of Knowledge base using Extended Petri net), for detecting incorrectness, inconsistency, and incompletensess in a knowledge base. The scope of the verification problem is expanded to chained errors, unlike previous studies that assume error incidence to be limited to rule pairs only. In addition, we consider certainty factor in checking, because most of knowledge bases have certainty factors.
Development of Heat Transfer and Evaporation Correlations for the Turbulent Natural Convection in the Vertical Channel by Using Numerical Analysis
Kang, Han-Ok ; Lee, Un-Chul ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 532~541
Theoretical and numerical study on heat transfer and evaporation in the vertical channel has been carried out and basic correlations have been derived for the heat transfer evaluation of PCCS. Analysis program was developed with low-Reynolds-number k-
model and surface transfer rates were calculated for the turbulent natural convection in the vertical channel. In relation to dry cooling by buoyancy-driven air, first, the system parameters which govern overall heat transfer rate are determined through the adequate nondimensionalization procedure. After comparison with existing experimental data, numerical results are used to derive heat transfer correlation by sensitivity calculations. In relation to wet cooling by falling water film, numerical analysis are carried out for evaporation process with real film surface conditions and evaporation correlation is derived through analogy concept and correction factors.
Development of RETRAN-03/MOV Code for Thermal-Hydraulic Analysis of Nuclear Reactor Under Mowing Conditions
Kim, Jae-Hak ; Park, Good-Cherl ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 542~550
Nuclear ship reactors have several features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been peformed under rolling, heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removal to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions.
Information Needs and Instrument Availability for Accident Management : Application to YGN 3&4
Kim, Jaewhan ; Park, Rae-Jun ; Suh, Kune-Yull ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 551~562
This paper introduces the five-step methodology for identifying information needs and assessing instrument availability during the course of severe accidents in nuclear power plants. The methodology is applied to the Yonggwang (YGN) 3&4 to shed light on accident management. It constructs three safety objective trees to prevent the reactor vessel failure, to prevent the containment failure, and to mitigate the fission product release from the containment. The study assesses information needs and instrument availability under severe conditions for preventing the reactor vessel failure of YGN 3&4, and recommends additional instrument that m8y prove to be of vital importance in managing the accident.
Analysis of Human Errors in Trip Cases of Korean NPPs
Lee, Jung-Woon ; Park, Geun-Ok ; Park, Jae-Chang ; Sim, Bong-Shick ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 563~575
A total of 77 cases was identified to have human errors from a total of 255 trips occurred from 1978 to 1992 in Korean NPPs. The cases were analyzed to investigate how many human errors occurred on which work conditions to find out the areas of high priority for human error reduction. For the analysis of the 77 trip cases due to human errors, classifications were made for the following four categories ; plant systems, work situation, job types, and error types. Erroneous tasks critically influencing the plant trips were carefully identified and analyzed according to the classifications. Based on the results for the individual cases, the cases were counted for the classification items in each of the four categories, then also for the group of categories to investigate the relationships among the categories in aspects of human error occurrences. As results, the plant systems, work situations, and job types, and error types that are dominant in human errors related to the trips ore identified.
Integrity of the Reactor Vessel Support System for a Postulated Reactor Vessel Closure Head Drop Event
Kim, Tae-Wan ; Lee, Ki-Young ; Lee, Dae-Hee ; Kim, Kang-Soo ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 576~582
The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor vessel closure head drop event.
A Characteristic Analysis on the Elastic Stiffness of the Tapered-width Leaf Type Holddown Spring Assembly Designed in KOFA's Design Space
Song, Kee-Nam ; Seo, Keum-Seok ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 583~593
An elastic stiffness formula of a leaf type holddown spring(HDS) assembly with a uniformly tapered width from
over the length, has been analytically derived based on Euler beam theory and Castigliano's theorem. Elastic stiffnesses of the tapered-width leaf type HDSs(TW-HDSs) designed in the same dimensional design spaces as the KOFA HDSs have been evaluated from the derived formula, in addition, a parametric study on the elastic stiffness of the TW-HDSs has been carried out. Analysis results show that, as the effects of axial and shear force on the elastic stiffness of He TW-HDSs have been 0.15~0.21% of the elastic stiffness, most of the elastic stiffness is attributed to the bending moment, and that elastic stiffnesses of the TW-HDSs have been about 32~33% higher than those of the KOFA HDSs. It is found that the number of leaves composing a HDS assembly could be lessened by one under the conditions that the TW-HDSs have been adopted in KOFA.
Verification of Safety Critical Software
Son, Ki-Chang ; Chun, Chong-Son ; Lee, Byeong-Joo ; Lee, Soon-Sung ; Lee, Byung-Chai ;
Nuclear Engineering and Technology, volume 28, issue 6, 1996, Pages 594~601
To assure quality of safety critical software, software should be developed in accordance with software development procedures and rigorous software verification and validation should be performed. Software verification is the formal act of reviewing, testing or checking, and documenting whether software components comply with the specified requirements for a particular stage of the development phase . New software verification methodology was developed and was applied to the Shutdown System No. 1 and 2(SDS1,2) for Wolsong 2, 3 and 4 nuclear power plants by Korea Atomic Energy Research Institute(KAERI) and Atomic Energy of Canada Limited(AECL) in order to satisfy new regulation requirements of Atomic Energy Control Board(AECB). Software verification methodology applied to SDS1 for Wolsong 2, 3 and 4 project will be described in this paper. Some errors were found by this methodology during the software development for SDS1 and were corrected by software designer. Output from Wolsong 2, 3 and 4 project have demonstrated that the use of this methodology results in a high quality, cost-effective product.