The thermal-hydraulic characteristics of the transition core was analyzed and the related thermal-hydraulic safety was investigated by evaluating DNBR at design overpower transient.
In the course of this work, the establishment of the design DNBR limit using Improved Thermal Design Procedure(ITDP) was inevitable, so the DNBR sensitivity analysis was performed. The DNBR sensitivity factors were achieved with respect to core inlet mass flux, core inlet flow temperature, system pressure, core power level and nuclear enthalpy rise hot channel factor. Besides, flow redistribution effects on DNBR in the transition core were considered quantitatively by introducing the transition core DNBR penalty.
COBRA-IV-I code was used for these calculations. In addition, R correlation was applied for the DNB calculation.
From these calculations, it is found that the MDNBR values of the transition core through Standard Thermal Design Procedure(STDP) violate the safety limit and the possibility of DNB occurrence is higher in OFA than in STD fuel assembly. Exit void fraction of OFA is smaller than that of STD fuel assembly, implying less voiding in OFA, as expected.
Meanwhile, the design DNBR limits of thimble cell and typical cell are established for OFA and STD fuel assembly, respectively. And the transition core DNBR penalty to compensate the fact that OFA lose flow in the transition core is achieved.
Finally, the fact that the MDNBR values of the transition core through ITDP satisfy the safety analysis DNBR limits is found.
From above results, it can be concluded that firstly, the thermal-hydraulic characteristics of OFA slightly worse than that of STD fuel assembly, secondly, STDP has too excess conservatism in the DNB calculation compared to ITDP and finally, the transition core is confirmed to be safe with respect to the problem of the DNB.
최적 및 표준 핵연료 집합체의 혼합노심에 대한 열·수력 특성을 분석하였다. 아울러 설계 과출력 상태에서의 DNBR을 계산함으로써 열·수력 안전성을 평가하였다.
열설계 절차로는 ITDP를 사용하였으며 설계 DNBR한계치를 설정하기 위하여 DNBR민감도 분석을 수행하였다. DNBR 민감도 분석에서, 노심입구 유속, 노심 입구 냉각재 온도, 원자로 압력, 노심 출력 정도, 핵적 엔탈피 상승 고온수로 계수 등의 DNBR민감도 계수를 결정하였다. 이밖에 혼합노심에서의 유량 재분배 현상을 정량적으로 평가하기 위하여 혼합노심 DNBR페날티를 도입하였다. 사용된 코드는 COBRA-IV-I이며 DNB 계산을 위하여 R상관식을 선택하였다.
계산결과, 혼합노심을 STDP 로 분석한 경우의 MDNBR값들은 안전한계치를 만족하지 못한다는 사실과 핵비등이탈 가능성은 최적 핵연료 집합체가 표준 핵연료 집합체에 비하여 크다는 사실을 확인하였다.
혼합노심을 ITDP로 분석하기 위하여 먼저 설계 DNBR한계치와 안전성 분석 DNBR한계치를 결정하였다. 최적 핵연료 집합체에 대한 혼합노심 DNBR 페날티를 계산한 후, 혼합노심의 DNBR 을 평가한 결과, DNBR 한계치를 만족함을 확인하였다.
