Improved Post-Dryout Heat Transfer Models in Developing Region and Deformed Fuel under large break loss-of-coolant accident = 대형냉각재상실사고 기간 중 발달 영역 및 변형 핵연료에서의 포스트드라이웃 열전달 모델 개선

다국어 초록 (Multilingual Abstract)

In nuclear reactor design and safety analysis, post-dryout heat transfer has been received great impetus on the thermal hydraulic concerns. The post-dryout heat transfer, which is characterized by a two-phase flow mixture of superheated vapor and dispersed droplets, exists in most of the sequence events during a hypothetical large-break loss-of-coolant accident (LB LOCA). In the present study, an improved heat transfer correlation has been developed from an extensive experimental dataset obtained from vertical tubes for the rediction of local wall temperature in the post-dryout region. The improved correlation modified the ellknown film-boiling look-up table to be applied to the developing post-dryout region. The newly-developed correlation has been validated using various postdryout datasets covering not only LB LOCA but also other pressurized conditions for different purposes. The wall temperature prediction results showed very good agreement to the experimental data except for some cases of low mass flux and heat flux inlet conditions.

At the early stage of reflood in an LB LOCA, the dispersed droplets would be mostly evaporated due to the verheating process, and the cooling water from the emergency core cooling systems has not levelled up yet. Hence, the prevailing wall-to-vapor convective heat transfer plays the most important role on the heat removal process. Under the harsh conditions in the reactor core, the fuel cladding could become ballooned due to the increase in surface temperature and the difference between the inner and outer pressures of the fuel rod. The deformation of the fuel cladding would restrict the subchannel flow area resulting in severe
flow blockage. In this work, a series of steam cooling experiments has been conducted to study the effect of flow blockage on the local wall-to-vapor convective heat transfer with consideration of fuel relocation phenomenon. The experimental results have been used to derive a new flow blockage model which is able to complement the conventional flow blockage model implemented in the COBRA-TF code. The new flow blockage model has been generalized to universally apply to different flow blockage configurations.

The upward flow inside a partially blocked core would be redistributed leading to remarkable changes in local flow pattern. The local mass flux in the intact subchannels are remarkably larger than those in the blocked subchannels due to the flow bypass effect. Additionally, the turbulence intensity and vorticity downstream of the blockage in the blocked subchannel are greatly enhanced owing to the flow separation effect. As a result, the local heat transfer in the vicinity of flow blockage may significantly altered. In order to investigate the local flow pattern in the partially blocked core, a new experimental facility has been designed by the author and constructed in Korea Atomic Energy Research Institute. The Particle Image Velocimetry measurement technique was adopted to capture the fluid motion inside a rod bundle containing partial flow blockage through a transparent test housing. The pressure drop caused by different flow blockage configurations as well as the information of local subchannel velocity have been recorded and analyzed. A new flow blockage pressure loss factor has been derived using Buckingham Pi theorem and regression technique, expressed as a function of flow blockage ratio, maximum flow blockage length, and divergence angle. The newly-developed flow blockage pressure loss factor has been used to improve the flow diversion model in the literature*.

국문 초록 (Abstract)

원자로설계 및 안전해석에서 포스트드라이아웃 열전달은 열수력 분야에서 많은 연구가 진행되었다. 포스트드라이아웃 열전달은 대형냉각재상실사고 과정 가운데 과열 증기와 분산액적의 이상유동으로 특징지을 수 있다. 본 연구에서는 포스트드라이아웃 영역에서 국소벽면온도를 예측하기 위한 수직관 실험을 수행하여 기존 연구에 확장된 실험데이터를 제공하고, 개선된 열전달상관식을 제시하였다. 개선된 상관식은 잘 알려진 막비등 룩업 테이블(Look-up table)을 수정하여 개발되었으며, 포스트드라이아웃의 발달영역에 적용 가능하다. 새로 개발된 상관식은 대형냉각재상실사고뿐 아니라 다양한 가압조건의 포스트드라이아웃 실험데이터들을 기반으로 검증되었다. 그 결과 낮은 유량 및 열속조건을 제외하고는 벽면온도를
잘 예측함을 확인하였다.

대형냉각재상실사고의 재관수 초기에 분산액적은 과열증기로 인해 대부분 기화되며, 비상노심냉각계통을 통해 제공된 냉 수의 수위는 충분히 상승하지 않는다. 따라서 이 시기에는 주로 벽면-증기 대류열전달로 인해 잔열이 제거된다. 잔열이 잘 제거되지 못하는 가혹한 조건에서는, 피복재의 표면온도 상승 및 봉내외 압력차이로 인해 피복재가 팽창할 수 있다. 피복재 변형은 부수로 유로 면적을 제한하여 유동을 부분적으로 막을 수 있다. 본 연구에서는 핵연료재배치에 따른 유동부분막힘이 국소 벽면-증기 대류열전달에 어떤 영향을 미치는지 다양한 실험을 통해 확인하였다. 실험 결과를 기반으로 새로운 모델을 개발하였으며, 이는 COBRA-TF 코드에 삽입된 기존의 유동부분막힘 모델을 보완하는데 활용될 것으로 기대된다.

새로운 유동부분막힘 열전달 모델은 다양한 부분막힘 현상에 활용될 수 있도록 일반화되었다. 부분막힘 조건에서 상향유동은 국소유동패턴이 급격하게 바뀌며 재배치된다. 건전한 부수로의 유속은 우회유동 효과로 인해 부분적으로 막힌 부수로의 유속보다 상당히 빠르다. 더욱이 유동막힘 하류에서 난류강도 및 소용돌이도(vorticity)가 유동분리효과에 의해 급격하게 증가한다. 그 결과, 유동막힘 주변의 국소열전달이 크게 바뀐다. 유동부분막힘 조건에서의 국소유동패턴을 확인하기 위해, 새로운 실험장치를 설계 및 제작하였다. 이 때 PIV (Particle Image Velocimetry) 측정기법을 이용하여 봉다발 내 유동을 투명한 시험장치를 통해 관측하였다. 또한 다양한 유동막힘 조건 및 유속 조건하에서 압력강하 값도 측정하였다. 새로운 유동막힘 압력강하인자를 Buckingham Pi 정리와 회귀 기법 (Regression Technique)을 이용하여 유도하였다. 이는 유동막힘정도, 최대 유동막힘 길이, 확산각도의 함수로 제시되었다. 새로운 유동막힘 압력강하 인자는 유동분산모델을 개선하는데 활용될 수 있다.

목차 (Table of Contents)

• 국문 초록 & Abstract...............................................................................i &iii
• List of Contents ..............................................................................................vii
• List of Tables ...................................................................................................ix
• List of Figures ..................................................................................................x
• Nomenclature ................................................................................................xiv
• Chapter 1 Introduction ................................................................................1
• 1.1 Background and motivation .................................................................1
• 1.2 Literature review..................................................................................6
• 1.3 Objectives and Research Layout...........................................................9
• Chapter 2 An improved heat transfer correlation for the developing postdryout region.................................................................................................. 11
• 2.1 Statement of the problem ................................................................... 11
• 2.2 Derivation of the improved correlation ............................................... 14
• 2.3 Wall temperature prediction results .................................................... 22
• 2.3.1 Wall temperature prediction with given CHF............................... 22
• 2.3.2 Wall temperature prediction results with unknown CHF .............. 27
• 2.3.3 Validation of the improved correlation......................................... 30
• Chapter 3 Effect of flow blockage on local heat transfer during early stage
• of reflood in LB LOCA .................................................................................. 40
• 3.1 Statement of the problem ................................................................... 40
• 3.2 ATHER facility .................................................................................. 43
• 3.2.1 Experimental setup...................................................................... 43
• 3.2.2 Experimental procedure and conditions ....................................... 46
• viii
• 3.2.3 Types of heater rod bundles......................................................... 47
• 3.3 Assessments of the COBRA-TF models............................................... 50
• 3.3.1 COBRA-TF single-phase heat transfer enhancement models... 50
• 3.3.2 COBRA-TF model assessments .................................................. 54
• 3.4 Development of new correlation......................................................... 63
• 3.5 Validation of the new correlation........................................................ 73
• Chapter 4 Flow distribution in partially blocked rod bundles.................. 81
• 4.1 Statement of the problem ................................................................... 81
• 4.2 Experimental facility.......................................................................... 86
• 4.2.1 Experimental setup...................................................................... 86
• 4.2.2 Experimental procedure and conditions ....................................... 89
• 4.2.3 Flow measurement technique ...................................................... 91
• 4.3 Experimental results........................................................................... 95
• 4.3.1 Pressure drop measurements........................................................ 95
• 4.3.2 PIV pre-tests ............................................................................... 98
• 4.3.3 Uncertainty of the PIV technique............................................... 104
• 4.4 Experimental analysis ...................................................................... 113
• 4.4.1 Local subchannel velocity analysis ............................................ 113
• 4.4.2 Pressure drop analysis ............................................................... 116
• 4.4.3 PIV pre-test analysis ................................................................. 120
• 4.4.4 Flow separation and reattachment analysis ................................ 123
• 4.4.5 Flow distribution measurement ................................................. 127
• Chapter 5 Conclusions and future works ................................................... 130
• Bibliography................................................................................................. 132