• Nuclear Engineering and Technology
• /
• 제31권1호
• /
• pp.17-28
• /
• 1999

A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.

• Nuclear Engineering and Technology
• /
• 제56권6호
• /
• pp.2163-2173
• /
• 2024

This study introduces a transient analysis module developed for RAST-V and validates it using the Kalinin-3 benchmark problem. For the benchmark analysis, RAST-V standalone and STREAM/RAST-V calculations were performed. STREAM supplies the few-group constants and RAST-V conducts a 3D core simulation utilizing few-group cross-sectional data. To improve accuracy, the main solver was developed based on the advanced semi-analytic nodal method. To evaluate the computational capability of the transient analysis module in RAST-V, Kalinin-3 benchmark is employed. Kalinin-3 represents a coolant transient benchmark that offers experimental data during the deactivation of the Main Circulation Pumps. Consequently, the transient calculations reflected the changes in the reactor flow rate. Benchmark comprising steady-state and transient calculations. During the steady state, the STREAM/RAST-V combination demonstrated a 30 ppm root mean square difference from 0 to 128.50 EFPD. For the transient calculations, STREAM/RAST-V showed power differences within ±7 % over a range of 0-300 s. Axial offset differences were within ±3 %, and the RMS difference in radial power ranged within 2.596 % at both 0 and 300 s. Overall, this study effectively demonstrated the newly developed transient solver in RAST-V and validated it using the Kalinin-3 benchmark problem.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.72-83
• /
• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• Nuclear Engineering and Technology
• /
• 제44권7호
• /
• pp.807-816
• /
• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• 한국항공우주학회지
• /
• 제39권6호
• /
• pp.559-566
• /
• 2011

본 연구에서는 적외선 윈도우의 냉각을 위해 포펫 밸브와 방사형 오리피스로 구성된 냉각장치 유로를 설계하였다. 필요한 냉각제의 양은 운용조건에 따라 달라지므로 포펫 행정거리에 따른 유동장의 유량과 윈도우 전후단 압력 변화를 유동해석을 통해 예측하고 실험을 통해 이를 확인하였다. 설계된 포펫과 방사형 오리피스 유로는 윈도우 냉각에 필요한 유량을 공급하며 윈도우 구조 강도를 만족하도록 내부 압력을 낮추고 적외선 이미지 신호의 왜곡이 없도록 아음속으로 유지하여 요구 조건을 충족시켰다. 실험으로 측정된 유량을 이용하여 윈도우에서의 송출계수와 2차원 해석결과 사이의 보정계수를 확인하였으며 이를 냉각장치의 유량제어에 사용하였다.

• 한국소음진동공학회논문집
• /
• 제20권3호
• /
• pp.215-222
• /
• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.598-603
• /
• 1996

When locked rotor event. occurs, instantaneously affected loop and core flow were quickly reduced, which resulted in an increase in coolant temperature and system pressure. Analysis method of this event was that constant core inlet temperature and system pressure as well as change in core flow calculated from COAST code were statically used as an input variable to HERMITE code, because of no tools to simulate NSSS behavior and 1-D core neutronics transient coincidently. With employing TASS code revised with 1-D neutronics model, this event was analyzed in point of DNBR. By doing so, analysis procedure could be simplified and unreasonable conservatism might be removed in DNBR calculation by consideration of pressure increase.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2003년도 추계학술대회논문집
• /
• pp.345-349
• /
• 2003

The dynamic behavior of CANDU nuclear fuel channel was analyzed by the use of 3-dimensional finite element method, under the various fault conditions such as a fault in the end fitting support and the removal/migration of the garter spring in the fuel channel, in order to predict the dynamic behavior for a degraded symptoms of CANDU nuclear fuel channel. Moreover, the frequency response analysis for possible fault conditions was also peformed considering the effects of the pressure tube vibration and flow-induced vibration by the coolant flow. From the analysis of the frequency responses, defects in the garter spring have influenced the changes of 2nd and 3rd modes and all the important modes are varied for the failure in the journal bearing in the end fitting body.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2008년도 춘계학술대회논문집
• /
• pp.641-646
• /
• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• 한국추진공학회지
• /
• 제5권2호
• /
• pp.1-7
• /
• 2001

매우 높은 연소가스로부터 연소실을 보호하기 위하여 액체로켓에서는 재생냉각방법을 폭넓게 이용하고 있다. 재생냉각을 통한 로켓엔진의 냉각을 매우 효과적인 방법이지만, 이를 개발하기 위해서는 정확한 해석과정, 제작기술 등이 필요하다. 한다. 실제 소형 로켓엔진에 재생냉각을 이용한 엔진 냉각의 가능성을 확인하기 위하여 설계, 제작된 로켓으로 연소실험을 진행하였다. 실험에 사용한 연소실은 coolant passage 3 mm, 벽 두께 1 mm, stainless 304로 제작하였다. 최대연소압과 연소시간은 각각 400 psi와 60 sec이고, coolant 유량은 2 kg/s에서 0.12 kg/s까지 감소시키면서 실험하였다. 연소시험후 육안으로 검사한 결과 연소실에서 특별한 이상은 발견되지 않았다.