Luận án Nghiên cứu hiện tượng chuyển pha trong vùng hoạt lò phản ứng

BỘ GIÁO DỤC VÀ ĐÀO TẠO 
TRƯỜNG ĐẠI HỌC BÁCH KHOA HÀ NỘI 
HOÀNG MINH GIANG 
NGHIÊN CỨU HIỆN TƯỢNG CHUYỂN PHA TRONG VÙNG HOẠT 
 LÒ PHẢN ỨNG 
LUẬN ÁN TIẾN SĨ CƠ HỌC 
Hà Nội – 2016 
2 
LỜI CAM ĐOAN 
Văn Hiền. 
Các số liệu, những kết luận nghiên cứu được trình bày trong luận văn 
này trung thực và chưa từng được công bố dưới bất cứ hình thức nào. 
Tôi xin chịu trách nhiệm về nghiên cứu của mình. 
GV Hướng dẫn Nghiên cứu sinh 
 Nguyễn Đông 
BỘ GIÁO DỤC VÀ ĐÀO TẠO 
TRƯỜNG ĐẠI HỌC BÁCH KHOA HÀ NỘI 
HOÀNG MINH GIANG 
NGHIÊN CỨU HIỆN TƯỢNG CHUYỂN PHA TRONG VÙNG HOẠT 
 LÒ PHẢN ỨNG 
Chuyên ngành: CƠ HỌC CHẤT LỎNG 
 Mã số: 62440108 
LUẬN ÁN TIẾN SĨ CƠ HỌC 
NGƯỜI HƯỚNG DẪN KHOA HỌC: 
1. PGS.TS NGUYỄN PHÚ KHÁNH 
2. TS TRẦN CHÍ THÀNH 
Hà Nội – 2016 
3 
LỜI CAM ĐOAN 
Tôi xin cam đoan luận án là công trình nghiên cứu của bản thân tôi dưới 
sự hướng dẫn của tập thể giáo viên hướng dẫn. 
Các kết quả nêu trong luận án là trung thực, không sao chép của bất kỳ 
công trình nào và chưa từng được công bố trong bất kỳ công trình nào khác. 
Hà Nội, ngày 27 tháng 4 năm 2016 
NGHIÊN CỨU SINH 
HOÀNG MINH GIANG 
Hướng dẫn 1 
PGS. NGUYỄN PHÚ KHÁNH 
Hướng dẫn 2 
TS. TRẦN CHÍ THÀNH 
4 
 LỜI CẢM ƠN 
Trước hết, tôi xin bày tỏ lòng kính trọng và biết ơn tới: PGS Nguyễn Phú 
Khánh và TS Trần Chí Thành, những người thày đã trực tiếp hướng dẫn, giúp đỡ 
tôi trong quá trình học tập và thực hiện luận án. 
Tôi xin chân thành cảm ơn các thày cô tại Bộ môn Kỹ thuật Hàng không 
và Vũ trụ, Viện Cơ khí Động lực; cảm ơn TS Lê Văn Hồng, Viện Năng lượng 
Nguyên tử Việt Nam, chủ nhiệm đề tài độc lập cấp nhà nước (mã số 
ĐTĐL.2011-G/82) “Nghiên cứu, phân tích, đánh giá và so sánh hệ thống công 
nghệ nhà máy điện hạt nhân dùng lò VVER-1000 giữa các loại AES-91, AES-
92 và AES-2006”, các đồng nghiệp Hoàng Tân Hưng, Trung tâm An toàn hạt 
nhân, Nguyễn Hữu Tiệp, Trung tâm Năng lượng hạt nhân, Viện Khoa học và Kỹ 
thuật hạt nhân đã giúp đỡ, tạo điều kiện để tôi có thể hoàn thành luận án này. 
Tôi cũng xin trân trọng cảm ơn Ban lãnh đạo Viện Khoa học và Kỹ thuật 
hạt nhân, Viện đào tạo Sau đại học của Trường Đại học Bách Khoa Hà Nội đã 
cử tôi đi đào tạo cũng như tạo điều kiện thuận lợi trong quá trình thực hiện luận 
án. 
Hà nội ngày 27/4/2016 
 Nghiên cứu sinh 
 Hoàng Minh Giang 
5 
STUDY ON PHASE CHANGE IN THE CORE OF 
NUCLEAR REACTOR 
6 
TABLE OF CONTENTS 
Abbreviations and Nomenclature ............................................................................................................... 8 
List of Tables .............................................................................................................................................. 12 
List of Figures ............................................................................................................................................. 14 
Overview .................................................................................................................................................... 17 
Chapter 1. Introduction to research work ............................................................................................... 19 
1.1 Status of nuclear power in the World and Vietnam ........................................................................... 19 
1.2 Brief overview of nuclear safety ........................................................................................................ 20 
1.3 Core thermal hydraulics safety analysis in transient condition .......................................................... 21 
1.3.1 Role of void fraction in simulation of two phase flow ................................................................ 24 
1.3.2 Experiment overview for bundle of sub channel analysis ........................................................... 25 
1.3.3 Void fraction prediction study ..................................................................................................... 26 
1.4 VVER technology understanding related to this study ...................................................................... 27 
1.5 Thesis objectives ................................................................................................................................ 29 
1.5.1 Studied object .............................................................................................................................. 30 
1.5.2 Scope of study ............................................................................................................................. 30 
1.6 Thesis outline ..................................................................................................................................... 31 
Chapter 2. Overview of phase change models in code theories with different scales ........................... 33 
2.1 Multi code and multi scales approach to PWR thermal hydraulic simulation ................................... 33 
2.1.1 Neutron codes and thermal hydraulics codes .............................................................................. 33 
2.1.2 Different scale of thermal hydraulic codes .................................................................................. 34 
2.1.3 Different thermal hydraulic modeling approaches ...................................................................... 36 
2.2 Phase change models in system code RELAP5 ................................................................................. 38 
2.3 Phase change models in sub channel code CTF ................................................................................. 40 
2.3.1 Evaporation and condensation induced by thermal phase change .............................................. 40 
2.3.2 Evaporation and condensation induced by turbulent mixing and void drift................................ 42 
2.4 Phase change models in meso scale code CFX .................................................................................. 42 
2.4.1 Evaporation at the wall ................................................................................................................ 42 
2.4.2 Condensation model in bulk of liquid ......................................................................................... 43 
2.5 Conclusions ........................................................................................................................................ 44 
Chapter 3. Phase change models verification and assessment by numerical simulation ..................... 45 
3.1 Brief information of VVER-1000/V392 ............................................................................................ 45 
3.2 Verification of RELAP5 simulation models for VVER-1000/V392 reactor with SAR ..................... 47 
3.2.1 Nodalization scheme ................................................................................................................... 48 
3.2.2 Verification of modeling through steady-state study .................................................................. 48 
3.2.3 Verification through accident case study .................................................................................... 49 
7 
3.3 CTF models verification and assessment with BM ENTEK tests ...................................................... 51 
3.3.1 ENTEK BM facility .................................................................................................................... 51 
3.3.2 Modeling by CTF ........................................................................................................................ 53 
3.3.3 Results and discussions ............................................................................................................... 53 
3.4 Verification CFX models with PSBT sub channel tests ..................................................................... 59 
3.4.1 PSBT test section for single sub channel .................................................................................... 60 
3.4.2 Mesh generation study ................................................................................................................ 61 
3.4.3 Solver convergence study ............................................................................................................ 63 
3.4.4 Mesh refinement study ................................................................................................................ 64 
3.4.5 Sensitivity study on physical models .......................................................................................... 68 
3.4.6 Assessment of CFX and CTF modeling results in comparison with PSBT single channel ........ 79 
3.4.7 Discussion on CTF and CFX void fraction predictions .............................................................. 82 
3.4.8 Improvement of CFX void fraction prediction in saturated region ............................................. 84 
3.5 Conclusions ........................................................................................................................................ 86 
Chapter 4. Void fraction prediction in hot channel of VVER-1000/V392 ............................................ 88 
4.1 Calculation Diagram .......................................................................................................................... 88 
4.2 Power distribution calculation by MCNP5 code ................................................................................ 90 
4.3 LOCAs simulation by RELAP5 code ................................................................................................ 93 
4.4 Void fraction prediction in hot channel during transient by CTF code .............................................. 96 
4.4.1 VVER-1000/V392 void fraction prediction by CTF ................................................................... 96 
4.4.2 Discussion on RELAP5 and CTF void fraction predictions ....................................................... 98 
4.5 Void fraction prediction in single channel by CFX code ................................................................. 100 
4.5.1 Mesh refinement study .............................................................................................................. 101 
4.5.2 Void fraction prediction calculated by CFX along sub channel ................................................ 102 
4.6 Void fraction prediction in bundle of channel calculated by CFX code .......................................... 104 
4.7 Conclusions ...................................................................................................................................... 107 
Conclusions and proposals ...................................................................................................................... 108 
Achievements and new findings given by the thesis .............................................................................. 108 
Proposal of future work .......................................................................................................................... 110 
References ................................................................................................................................................. 112 
List of Author’ papers and report .......................................................................................................... 116 
8 
Abbreviations and Nomenclature 
Abbreviations 
VVER A Type of Pressurized Water Reactor developed by Russia 
VVER-1200/V491 A type of Russia reactor with capability of 1200 MWe 
VVER-1000/V392 A type of Russia reactor with capability of 1000 MWe 
VINATOM Vietnam Atomic Energy Institute 
TSO Technical Support Organization 
DID Defend in depth policy in nuclear power plant design 
PWR Pressurized Water Reactor 
SAR Safety Analysis Report of nuclear power plant 
NRA Nuclear Regulatory Authority 
RIAs Reactivity insertion accident 
LOFAs Loss of coolant flow 
LOCAs Loss of coolant accident 
DNB Departure of nucleate boiling 
DNBR Departure of nucleate boiling ratio 
Castellana The 4 x 4 square rod bundle test for fuel rod in Columbia University 
(USA) 
EPRI Electric Power Research Institute 
BM ENTEK The BM Facility at the Research and Development Institute of Power 
Engineering (RDIPE; a.k.a., ENTEK and NIKIET) models the forced 
circulation circuit of RBMK type reactors 
RBMK-1000 A type of Russia reactor of 1000 MWe with transliteration of Russian 
characters for graphite-moderated boiling-water-cooled channel-type 
reactor 
PSBT OECD/NRC Benchmark based on Nuclear Power Engineering 
Corporation (NUPEC, Japan) PWR sub channel and bundle tests 
CTF A version of COBRA-TF improved by Pennsylvania State University 
(USA) 
RELAP5 System code developed by Information Systems Laboratories, Inc. 
Rockville, Maryland Idaho Falls, Idaho 
COBRA-TF Coolant-Boiling in Rod Arrays—Two Fluids (COBRA-TF) is a Thermal 
Hydraulic (T/H) simulation code designed for Light Water Reactor (LWR) 
vessel analysis developed by Pacific Northwest Laboratory 
RELAP-3D Newest version of RELAP5 with coupling with COBRA-TF 
MARS-3D Newest version of MARS with coupling with COBRA-TF 
Belene A site for nuclear power plant project in Bulgaria 
Ansys CFX A Computational Fluid Dynamics developed by Ansys 
CFX Same as Ansys CFX 
PARCS A code for neutron kinetic calculation 
ITT interface tracking technique 
0D, 1D, 2D Dimension of spatial averaging 
CHF Critical Heat Flux 
TH Thermal hydraulics 
RANS Reynolds-averaged Navier–Stokes Simulation 
9 
LES Large Eddy Simulation 
MSLB Main steam line break 
PTS Pressurize Thermal shock 
CFD Computational Fluid Dynamics 
DI Deterministic Interface 
FI Filtered Interface 
SI Statistical Interface 
U-RANS Unsteady flow 
T-RANS Transient flow 
meso scale The spatial scale with size around 1mm and less simulated with RANS 
ECCS system Emergency Core Cooling System 
LBLOCAs Large break for loss of coolant accident 
SBO Station black out 
SG Steam Generator 
SG PHRS Passive Heat Removal through Steam Generator 
HA-2 Secondary stage of Hydro accumulators 
HA-1 First stage of Hydro accumulators 
PCT Peaking temperature of cladding 
DBA Design Base Accident 
MCPL Main Coolant Pipe line 
LOOP Loss of offsite power 
DG Diesel Generator 
SAR SG SG Active Heat Removal System 
OECD/NRC BFBT UPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark 
αcrit Void fraction corresponding with critical heat flux correlation 
10 
Nomenclature 
 Sub-cooled vapor interfacial area per unit volume (m
-1
) 
 Super-heated liquid interfacial area per unit volume (m
-1
) 
 Super-heated vapor interfacial area per unit volume (m
-1
) 
As Conductor surface area in mesh cell (m
2
) 
Ax Mesh-cell area, X normal (m
2
) 
Cpl Liquid specific heat, constant pressure (J/kg.K) 
Cpv Vapor ... lizing multi code and multi scale including MCNP5, RELAP5, CTF and CFX for 
analysis of void fraction behavior in the core during transient. 
 For system analysis by RELAP5 code for VVER-1000/V392, it is found that 
temperature near heated wall is not defined with enough accuracy due to large 
equivalent diameter if simulation a whole of fuel assembly, so the phase change 
models in RELAP5 do not give appropriate value of void fraction. 
 From verification and validation of CTF results with ENTEK BM experiment, it is 
observed that CTF tends to give under prediction of void in the region of sub cooled 
boiling and flow regime in small bubble (αg < 0.2) and CTF tends to give over 
prediction of void in nucleate boiling region, corresponding with small-to-large bubble 
in flow regime. 
 From verification with PSBT single sub channel experiment, CFX with model setup 
proposed in this thesis is converged with RMS of 1e-6 and stabilized in term of 
average void fraction prediction with physical sensitivity study. For the sub cooled 
boiling region corresponding with small bubble of flow regime (αg < 0.2), CFX gives 
the appropriate void fraction prediction with accuracy around ±0.03 of void. 
 In saturated boiling region, the wall boiling model built in CFX is incorrectly 
partitioned heat flux to corresponding parts in convective, quenching and evaporative. 
This issue causes CFX gives under prediction of void fraction in saturated boiling 
region. 
 It is proposed a calibration for bubble departure diameter and maximum area fraction 
to improve void fraction prediction by CFX in saturated region. 
 It is established a procedure of utilizing CTF and CFX codes for void fraction 
prediction as following: (a) at sub cooled region, corresponding with small bubble 
flow regime, CFX results is used; (b) in saturated boiling region, CTF and CFX void 
fraction curves along the channel is used as upper and lower bound to predict void 
fraction in the core. 
Proposal of future work 
Utilization of CFD codes for investigation of void fraction in the core is still a challenge. This 
comes from complexity of boiling phenomena and the lack of experiment with similar PWR 
condition to verification and validation CFD models. Based on study in the thesis, several 
following issues are proposed to study. 
 Study on modification of RPI wall boiling model built in CFX (and FLUENT) in 
saturated boiling region. Due to the fact that, in saturated boiling model, liquid 
111 
temperature is the same saturated one everywhere, even near wall, so that only 
evaporation and quenching phenomena can occur. 
 Implement more experiment in similar PWR conditions which provides with void 
fraction distribution that can be used to validate evaporation and condensation 
models in CFX 
 Study on more accuracy of void fraction prediction of CFX based on focusing on 
condensation such as the correlation of Nusselt number in different boiling 
conditions. 
112 
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116 
List of Author’ papers and report 
[1] Hoang Minh Giang, Nguyen Phu Khanh, Le Van Hong., (2014) “Capability analysis of passive 
systems in typical design extension conditions for nuclear reactor VVER-1000/V392”, Journal of 
Science and Technology 52 (2C) (2014) pp. 81-92. 
[2] Hoang Minh Giang, Nguyen Phu Khanh, Le Van Hong, Le Dai Dien.,(2014) “Study on 
improvement of convergence for PWR sub channel void distribution benchmark”, Journal of Science 
and Technology 52 (2C) (2014) pp. 184-197. 
[3] Hoang Minh Giang, Nguyen Phu Khanh., (2014) “Numerical investigation of departure bubble 
diameter for wall boiling model in PWR sub channel”, proceeding in AUN/SEED-Net Regional 
Conference on Mechanical and Manufacturing Engineering, 9-10 October, 2014 (RCMME-2014). 
[4] Hoang Minh Giang, Hoang Tan Hung, Nguyen Phu Khanh., (2015)“Investigation of CTF void 
fraction prediction by ENTEK BM experiment data”, Nuclear Science and Technology (ISSN 1810-
5408),Vol5, No1, 2015 pp.8 -17 
[5] Hoang Minh Giang, Hoang Tan Hung, Nguyen Huu Tiep., (2015) “Multi codes and multi-scale 
analysis for void fraction prediction in hot channel for VVER-1000/V392”, Nuclear Science and 
Technology (ISSN 1810-5408), Vol5, No3, 2015 pp.19-31. 
[6] Hoang Minh Giang, Hoang Tan Hung, Nguyen Phu Khanh., (2015) “Numerical study on CTF 
code to predict void fraction in PWR sub channel conditions”, Nuclear Science and Technology (ISSN 
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