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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2018.3
- 수록면
- 1 - 11 (11page)
- DOI
- 10.6112/kscfe.2018.23.1.001
이용수
초록· 키워드
오류제보하기
The nodalization sensitivity analysis for the ECC(Emergency Core Cooling) bypass phenomena was performed using the SPACE(Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS(Multi-dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI(Korea Atomic Energy Research Institute) for the performance evaluation of the ECC bypass phenomenon in the DVI(Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The sensitivity study involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the MIDAS high steam flow rate tests is minimal. For computational efficiency, a 4 channels representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-predicts the bypass fraction irrespective of the nodalization scheme. The over-prediction at low steam flow may be attributed to the difficulty to accurately represent the three dimensional flow pattern in the vicinity of the broken cold leg using a one-dimensional code.
목차
1. Introduction
2. MIDAS Test
3. MIDAS Modeling for the SPACE Code
4. Results and Discussion
5. DVI Location Effect Study
6. Conclusions
References
참고문헌 (11)
참고문헌 신청
2017, KHNP Report., Topical Report on the SPACE code for Nuclear Power Plant Design, KHNP/TR-0032/2017.
2004, Lee, S.Y. and Ban, C.H., Code-Accuracy-Based Uncertainty Estimation (CABUE) Methodology for Large-Break Loss-of-Coolant Accidents, Nuclear Technology, Vol.148, Issue 3.
2010, KHNP Report., Topical Report for the LBLOCA Best-Estimate Evaluation Methodology of the APR1400 Type Nuclear Power Plant, KHNP/TR-0018/2010.
2003, Lee, S.W. and Oh, S.J., APR1400 Large Break Loss of Coolant Accident Analysis using KREM methodologies, 2003 KNS Autumn Meeting.
2007, Lee, S.W., Kim, H.G. and Oh, S.J., Assessment of APR1400 ECCS Capability against Large-Break LOCA Scenario by RELAP5/MOD3 Code, Nuclear Technology, Vol.158, Issue 3.
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UCI(KEPA) : I410-ECN-0101-2018-559-001906286