• Title/Summary/Keyword: RETRAN

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Study of the RBTRAN Code for Upper Plenum Analysis in Very Small LOCA (매우 작은 규모의 LOCA에 있어서 Upper Plenum분석을 위한 RETRAN코드의 연구)

  • Hee Cheon No
    • Nuclear Engineering and Technology
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    • v.16 no.3
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    • pp.125-130
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    • 1984

  • In the application of the RETRAN code to the analysis of very small LOCA one of main concerns is placed on use of the bubble rise model in the upper plenum, because the bubble rise model nay cause a numerical divergence problem and coefficients used to describe it are based on experimental results of large LOCA. In order to solve this problem, a method, which enables us to predict the mixture level in the upper plenum without use of the bubble rise model, was proposed. For this method the local void distribution in the core and upper plenum was derived by using a simplified slip model. It was shown that results predicted from the derived equation are in excellent agreement with experimental data. Additionally it was found that local void in the upper plenum has a uniform distribution unlike a linear distribution in large LOCA. Communication between the upper plenum and upper head was investigated. By introducing the concept of Taylor instability, it was proved that counter-current Hon between them is possible.

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APPLICATIONS OF INTEGRATED SAFETY ANALYSIS METHODOLOGY TO RELOAD SAFETY EVALUATION

  • Jang, Chan-Su;Um, Kil-Sup
    • Nuclear Engineering and Technology
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    • v.43 no.2
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    • pp.187-194
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    • 2011

  • Korea Nuclear Fuel is developing the X-GEN fuel which shows high performance and robust reliability for the worldwide supply. However, the simplified code systems such as CESEC-III which were developed in 1970s are still used in the current Non-LOCA safety analysis of OPR1000 and APR1400 plants. Therefore, it is essential to secure an advanced safety analysis methodology to make the best use of the merits of X-GEN fuel. To accomplish this purpose, the $\b{i}$ntegrated $\b{s}$afety $\b{a}$nalysis $\b{m}$ethodology (iSAM), is developed by selecting the best-estimate thermal-hydraulic code RETRAN. iSAM possesses remarkable advantages, such as generality, integrity, and designer-friendly features. That is, iSAM can be applied to both OPR1000 and APR1400 plants and uses only one computer code, RETRAN, in the whole scope of the non-LOCA safety analyses. Also the iSAM adopts the unique and automatic initialization and run tool, $\b{a}$utomatic $\b{s}$teady-$\b{s}$tate $\b{i}$nitialization and $\b{s}$afety analysis too l (ASSIST), to enable unhandy designers to use the new design code RETRAN without difficulty. In this paper, a brief overview of the iSAM is given, and the results of applying the iSAM to typical non-LOCA transients being checked during the reload design are reported. The typical non-LOCA transients selected are the single control element assembly withdrawal (SCEAW) accident, the asymmetric steam generator transients (ASGT), the locked rotor (LR) accident, and bank CEA withdrawal (BCEAW) event. Comparison to current licensing results shows a close resemblance; thus, it reveals that the iSAM can be applied to the non-LOCA safety analysis of OPR1000 and APR1400 plants.

  • https://doi.org/10.5516/NET.2011.43.2.187 인용 PDF KSCI

Development of Backup Calculation System for a Nuclear Steam Supply System Thermal-Hydraulic Model ARTS (Advanced Real-time Thermal Hydraulic Simulation) of the W/H Type NPP (W/H형 원전 시뮬레이터용 핵 증기공급 계통 열수력모델 ARTS(Advanced Real-time Thermal Hydraulic Simulation)의 보조계산체계 개발)

  • 서재승;전규동
    • Journal of Energy Engineering
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    • v.13 no.1
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    • pp.51-59
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    • 2004

  • The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited compulsational capability at that time, they usually used very simplified physical models for a real-time simulation of NSSS thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called "negative training", especially for complicated two-phase flows in the reactor coolant system. In resolve the problem, KEPRI developed a realistic NSSS T/H program ARTS which was based on the RETRAN-3D code for the improvement of the Nuclear Power Plant full-scope simulator. The ARTS (based on the RETRAN-3D code) guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there is some possibility of failing to calculate in the case of large break loss of coolant accident (LBLOCA) and low-pressure low-flow transient. In this case, the backup calculation system cover automatically the ARTS. The backup calculation system was expected to provide substantially more accurate predictions in the analysis of the system transients involving LBLOCA. The results were reasonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with FSAR and the AMSI/ANS-3.5-1998 simulator software performance criteria.

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