• Title/Summary/Keyword: MARS code

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Application of a combined safety approach for the evaluation of safety margin during a Loss of Condenser Vacuum event

  • Shin, Dong-Hun;Jeong, Hae-Yong;Park, Moon-Ghu;Sohn, Jung-Uk
    • Nuclear Engineering and Technology
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    • v.54 no.5
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    • pp.1698-1711
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    • 2022

  • A combined safety approach, which uses a best-estimate computer code and adopts conservative assumptions for safety systems availability, is developed and applied to the safety margin evaluation for the Loss of Condenser Vacuum (LOCV) of the 1000 MWe Korean Nuclear Power Plant. The Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code is selected as a best-estimate code and the PAPIRUS program is used to obtain different initial operational conditions through random sampling of control variables. During an LOCV event, fuel integrity is not threatened by the increase in Departure from Nuclear Boiling Ratio (DNBR). However, the high pressure in the primary coolant system and the secondary system might affect the system integrity. Thus, the peak pressure becomes a major safety concern. Transient analyses are performed for 124 cases of different initial conditions and the most conservative case, which results in the highest system pressure is selected. It is found the suggested methodology gives similar peak pressures when compared to those predicted from existing methodologies. The proposed approach is expected to minimize the time and efforts required to identify the conservative plant conditions in the existing conservative safety methodologies.

  • https://doi.org/10.1016/j.net.2021.11.006 인용 PDF KSCI

THE CUPID CODE DEVELOPMENT AND ASSESSMENT STRATEGY

  • Jeong, J.J.;Yoon, H.Y.;Park, I.K.;Cho, H.K.
    • Nuclear Engineering and Technology
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    • v.42 no.6
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    • pp.636-655
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    • 2010

  • A thermal-hydraulic code, named CUPID, has been being developed for the realistic analysis of transient two-phase flows in nuclear reactor components. The CUPID code development was motivated from very practical needs, including the analyses of a downcomer boiling, a two-phase flow mixing in a pool, and a two-phase flow in a direct vessel injection system. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations are solved over unstructured grids with a semi-implicit two-step method. This paper presents an overview of the CUPID code development and assessment strategy. It also presents the code couplings with a system code, MARS, and, a three-dimensional reactor kinetics code, MASTER.

  • https://doi.org/10.5516/NET.2010.42.6.636 인용 PDF KSCI

THE DEVELOPMENT AND ASSESSMENT STRATEGY OF A THERMAL HYDRAULICS COMPONENT ANALYSIS CODE (열수력 기기해석용 CUPID 코드 개발 및 평가 전략)

  • Park, I.K.;Cho, H.K.;Lee, J.R.;Kim, J.;Yoon, H.Y.;Lee, H.D.;Jeong, J.J.
    • Journal of computational fluids engineering
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    • v.16 no.2
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    • pp.30-48
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    • 2011

  • A three-dimensional thermal-hydraulic code, CUPID, has been developed for the analysis of transient two-phase flows at component scale. The CUPID code adopts a two-fluid three-field model for two-phase flows. A semi-implicit two-step numerical method was developed to obtain numerical solutions on unstructured grids. This paper presents an overview of the CUPID code development and assessment strategy. The governing equations, physical models, numerical methods and their improvements, and the systematic verification and validation processes are discussed. The code couplings with a system code, MARS, and, a three-dimensional reactor kinetics code, MASTER, are also presented.

  • https://doi.org/10.6112/kscfe.2011.16.2.030 인용 PDF KSCI

Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor

  • Zhao, Pengcheng;Liu, Zijing;Yu, Tao;Xie, Jinsen;Chen, Zhenping;Shen, Chong
    • Nuclear Engineering and Technology
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    • v.52 no.12
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    • pp.2789-2802
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    • 2020

  • Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

  • https://doi.org/10.1016/j.net.2020.05.023 인용 PDF KSCI

MULTI-SCALE THERMAL-HYDRAULIC ANALYSIS OF PWRS USING THE CUPID CODE

  • Yoon, Han Young;Cho, Hyoung Kyu;Lee, Jae Ryong;Park, Ik Kyu;Jeong, Jae Jun
    • Nuclear Engineering and Technology
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    • v.44 no.8
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    • pp.831-846
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    • 2012

  • KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

  • https://doi.org/10.5516/NET.02.2012.716 인용 PDF KSCI