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


BỘ GIÁO DỤC VÀ ĐÀO TẠO
TRƯỜNG ĐẠI HỌC BÁCH KHOA HÀ NỘI

LỜI CAM ĐOAN
Văn Hiền.

HOÀNG MINH GIANG

NGHIÊN
CỨU
HIỆNkết
TƢỢNG
CHUYỂN
PHAtrình
TRONG
VÙNG
Các số liệu,
những



LỜI CAM ĐOAN
Tôi xin cam đoan luận án là công trình nghiên cứu của bản thân tôi dƣới
sự hƣớng dẫn của tập thể giáo viên hƣớng dẫn.
Các kết quả nêu trong luận án là trung thực, không sao chép của bất kỳ
công trình nào và chƣa từng đƣợc công bố trong bất kỳ công trình nào khác.
Hà Nội, ngày 27 tháng 4 năm 2016
NGHIÊN CỨU SINH

HOÀNG MINH GIANG
Hƣớng dẫn 1

PGS. NGUYỄN PHÚ KHÁNH

Hƣớng dẫn 2

TS. TRẦN CHÍ THÀNH

3


LỜI CẢM ƠN
Trƣớc hết, tôi xin bày tỏ lòng kính trọng và biết ơn tới: PGS Nguyễn Phú
Khánh và TS Trần Chí Thành, những ngƣời thày đã trực tiếp hƣớng 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 lƣợng
Nguyên tử Việt Nam, chủ nhiệm đề tài độc lập cấp nhà nƣớc (mã số
ĐTĐL.2011-G/82) ―Nghiên cứu, phân tích, đánh giá và so sánh hệ thống công

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
6


3.3 CTF models verification and assessment with BM ENTEK tests ......................................................51


7


Abbreviations and Nomenclature
Abbreviations
VVER
VVER-1200/V491
VVER-1000/V392
VINATOM
TSO
DID

PWR
SAR
NRA
RIAs
LOFAs
LOCAs
DNB
DNBR
Castellana
EPRI
BM ENTEK

RBMK-1000

PSBT
CTF
RELAP5

Electric Power Research Institute
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
A type of Russia reactor of 1000 MWe with transliteration of Russian
characters for graphite-moderated boiling-water-cooled channel-type
reactor
OECD/NRC Benchmark based on Nuclear Power Engineering
Corporation (NUPEC, Japan) PWR sub channel and bundle tests
A version of COBRA-TF improved by Pennsylvania State University
(USA)
System code developed by Information Systems Laboratories, Inc.
Rockville, Maryland Idaho Falls, Idaho
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
Newest version of RELAP5 with coupling with COBRA-TF
Newest version of MARS with coupling with COBRA-TF
A site for nuclear power plant project in Bulgaria
A Computational Fluid Dynamics developed by Ansys
Same as Ansys CFX
A code for neutron kinetic calculation
interface tracking technique
Dimension of spatial averaging
Critical Heat Flux
Thermal hydraulics
Reynolds-averaged Navier–Stokes Simulation
8



Computational Fluid Dynamics
Deterministic Interface
Filtered Interface
Statistical Interface
Unsteady flow
Transient flow
The spatial scale with size around 1mm and less simulated with RANS
Emergency Core Cooling System
Large break for loss of coolant accident
Station black out
Steam Generator
Passive Heat Removal through Steam Generator
Secondary stage of Hydro accumulators
First stage of Hydro accumulators
Peaking temperature of cladding
Design Base Accident
Main Coolant Pipe line
Loss of offsite power
Diesel Generator
SG Active Heat Removal System
UPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark
Void fraction corresponding with critical heat flux correlation

9


Nomenclature

As
Ax

Super-heated vapor interfacial area per unit volume (m-1)
Conductor surface area in mesh cell (m2)
Mesh-cell area, X normal (m2)
Liquid specific heat, constant pressure (J/kg.K)
Vapor specific heat, constant pressure (J/kg.K)
Mixing mass flux (kg/m2.s)
Vapor saturation enthalpy (J/kg)
Sub-cooled liquid interface heat transfer coefficient (W/m2.K)
Sub-cooled vapor interface heat transfer coefficient (W/m2.K)
Super-heated liquid interface heat transfer coefficient (W/m2.K)
Super-heated vapor interface heat transfer coefficient (W/m2.K)
Chen correlation heat transfer coefficient (W/m2.K)
Liquid enthalpy (J/kg)
Liquid saturation enthalpy (J/kg)
Vapor enthalpy (J/kg)
Vapor interface heat transfer coefficient (W/m3..K)
Liquid interface heat transfer coefficient (W/m3..K)
Mass exchange due to drift model (kg/s)
Mass exchange of phase k (kg/m2.s)
Density of liquid (kg/m3)
Wall heat transfer to liquid (W)
Wall heat transfer to liquid for convection (W)
Wall heat transfer to liquid for vaporization (W)
Vapor temperature (K)
Saturated temperature (K)
Critical heat flux temperature (K)
Liquid temperature (K)
Bubble diameter (m)
Void fraction of phase k induced by sub channel i
Equilibrium quality void fraction

g
FChen
f
Dh
Cp
Ax
As

Fo
̅
,𝑟
𝑞
𝑞
𝑞
𝑞
𝑞

̅

Evaporation rate (kg/m2.s)
Wall surface temperature (K)
Critical heat flux temperature (K)
Reynolds number
Prandtl number
Nusselt number
Wall nucleation site density (m-2)
Liquid thermal conductivity (W/m.K)
Vapor enthalpy (J/kg)
Nucleate-boiling heat transfer coefficient (W/m2.K)
Liquid enthalpy (J/kg)

Area influence factors

11


List of Tables
Table 1.1 Multiple levels of protection from DID approach (source [45]) ............................................20
Table 1.2 Content of Safety Analysis Reports (source [45]) .................................................................21
Table 1.3 Castellana 4x4 test characteristics (source [29]) ....................................................................25
Table 1.4 EPRI 5x5 characteristics for test 74 and test 75(source [29]) ................................................25
Table 1.5 Geometry and power shape for Test Assembly B5, B6, and B7 (Source [1]) .......................25
Table 2.1 Main characteristics of codes with four different scales (source [11]) ..................................36
Table 2.2 Main characteristics modeling approaches for three main types of single-phase CFD .........37
Table 3.1 Main technical characteristics of VVER-1000/V392 (source[36]) ........................................46
Table 3.2 Comparison of steady-state of VVER-1000/V392.................................................................48
Table 3.3 Boundary conditions for event number 3 (source [35]). ........................................................49
Table 3.4 Chronological sequence of Event 3 from SAR [35] and this study .......................................50
Table 3.5 Setting for base case and sensitivity cases according to test 01 and test 17 ...........................53
Table 3.6 Base case void fraction distribution calculations versus experiment for cases at 3MPa. ......54
Table 3.7 Base case void fraction distribution calculations versus experiment for cases at 7MPa. ......54
Table 3.8 Deviation of void fraction distribution calculation results versus experiment .......................55
Table 3.9 Deviation of void fraction distributions on input uncertainties ..............................................58
Table 3.10 Maximum deviation of void fraction distribution on input parameters versus base case ....58
Table 3.11 Experimental uncertainties on input parameters ..................................................................60
Table 3.12 Test Conditions for Steady-State Void Measurement of selected runs ................................61
Table 3.13 Mesh characteristics .............................................................................................................62
Table 3.14 y+ predicted by Mesh 1.........................................................................................................62
Table 3.15 y+ predicted by Mesh 2.........................................................................................................62
Table 3.16 y+ predicted by Mesh 3.........................................................................................................63
Table 3.17 Two phase flow model setting .............................................................................................63

List of Figures
Figure 1.1 Nuclear power generation by country in 2013 (source [46]) ................................................19
Figure 1.2 Multiple physical barriers in DID policy (source [45]) ........................................................22
Figure 1.3 Heal flux versus temperature difference for pool boiling heat transfers (source [31]) .........23
Figure 1.4 Types of boiling flow crisis (source [25]).............................................................................23
Figure 1.5 Critical heat flux in uniformly core (source [25]) .................................................................24
Figure 1.6 Development of VVER nuclear reactor technology chart [32] .............................................28
Figure 1.7 Multi-scale analysis of reactor thermal hydraulics (source [11]) .........................................29
Figure 1.7 (a) temperature distributions in a cylindrical fuel pin, (b) flow regime ................................31
Figure 2.1 Relations between MCNP5, system code RELAP5 and component code CTF ...................33
Figure 2.3 System code capabilities for reactor thermal hydraulics (source [7]) ...................................34
Figure 2.4 Control volume and axial flow area defined in sub channel code ........................................35
Figure 2.6 The tree of two-phase thermal hydraulic modeling approaches (source [11])......................38
Figure 2.7 Schematic of vertical flow regime map in RELAP5(source [19]) ........................................40
Figure 2.8 CTF normal-wall flow regime maps (source [38]) ...............................................................41
Figure 3.1 Side view of primary system of VVER-1000/V392 (source [36]) .......................................46
Figure 3.2 Primary system and safety system for VVER-1000/V392 (source [37])..............................47
Figure 3.3 VVER-1000/V392 nodalization schemes in this study.........................................................48
Figure 3.4 (a) Cladding temperature from calculations, (b) Cladding temperature from SAR ..............51
Figure 3.5 Test section (Heat Release Zone, φ is diameter in mm) .......................................................52
Figure 3.6 BM ENTEK modeling by CTF ............................................................................................53
Figure 3.7 Radial void distribution of the test T04 ................................................................................56
Figure 3.8 Cross mass flow due to turbulent mixing and void drift .......................................................56
Figure 3.9 Maximum and minimum voiding curves versus experiment ................................................57
Figure 3.10 Maximum and minimum voiding curves versus experiment ..............................................57
Figure 3.11 Uncertainty void fraction distributions for test T01............................................................58
Figure 3.12 Test section for central sub channel void distribution measurement (source [1]) ..............60
Figure 3.13 Cross section of three proposed meshes .............................................................................62
Figure 3.14 Base line for radial distribution investigation .....................................................................64
Figure 3.15 S14326 Radial distribution of void fraction........................................................................66

Figure 4.8 (a) Whole fuel assembly simulated as hot channel and (b) the active part ...........................95
Figure 4.9 Average void fraction calculated by RELAP5 on exit of active region in hot channel ........95
Figure 4.10 Taken twelfth of whole bundle for void fraction prediction. ..............................................96
Figure 4.11 Cross section of CTF modeling for the selected part of the whole bundle .........................97
Figure 4.12 Void fraction prediction by CTF and RELAP5 for LBLOCAs ..........................................97
Figure 4.13 Void fraction prediction by CTF and RELAP5 for SBLOCAs ..........................................98
Figure 4.14 Total vapor generation rate and vapor generation rate near wall ........................................98
Figure 4.15 Total vapor generation rate and vapor generation rate near wall ........................................99
Figure 4.16 Three meshes used to simulate geometry of single channel .............................................101
Figure 4.17 Average void fraction along channel with different meshes.............................................102
15


Figure 4.18 Axial sub channel void fraction prediction by CFX and CTF ..........................................103
Figure 4.19 (a) Overview of mesh (b) Zooming of mesh ....................................................................105
Figure 4.20 Four cases with specific timing for study by CFX...........................................................105
Figure 4.21 Improvement by CFX in left pictures and upper and lower bounds in right pictures .......107

16


Overview
Phase change in the nuclear reactor core is related to safety criteria such as Departure of
Nucleate Boiling (DNB) during normal and transient conditions. So that, a lot of computer
codes with verification and validation against experiment are used to investigation of thermal
hydraulics behavior of vertical boiling flow in core channel with system and component
scales. Until now, even many studies on boiling flow are implemented in CFD scale codes,
but their utilization to specific nuclear reactor is not yet applied. Thus, the utilization of many
codes including CFD scale (Ansys CFX) to investigate void fraction in hot channel of VVER1000/V392 reactor core is studied in this work. Due to VVER-1000/V392 nuclear reactor is a
candidate for Ninh Thuan 1 nuclear power project, so that the understanding of VVER’s

CTF and Ansys CFX void fraction curves along the channel is used as upper bound
and lower bound to predict void fraction in the core.
It is found that, in saturated boiling region, the wall boiling model built in Ansys CFX
is incorrectly partitioned heat flux to corresponding parts in convective, quenching and
evaporative. This issue causes Ansys CFX gives under prediction of void fraction in
saturated boiling region. It is proposed a calibration for bubble departure diameter and
17


maximum area fraction to improve void fraction prediction by Ansys CFX in saturated
region.

18


Chapter 1. Introduction to research work
1.1 Status of nuclear power in the World and Vietnam
Nuclear technology uses the energy released by splitting the atoms of certain elements. After
Second World War, nuclear technology turned to peaceful purposes of nuclear fission for
power generation. Today, as updated in February 2015 [46], the world produces as much
electricity from nuclear energy as it did from all sources combined in the early years of
nuclear power. Civil nuclear power now can boast over 16,000 reactor years of experience
and supplies almost 11.5% of global electricity needs. The 31 countries host over 435
commercial nuclear power reactors with a total installed capacity of over 375,000 MWe as
illustrated in Figure 1.1. About 70 further nuclear power reactors are under construction,
equivalent to 20% of existing capacity, while over 160 are firmly planned, and equivalent to
half of present capacity.

Figure 1.1 Nuclear power generation by country in 2013 (source [46])


Level 2

Control of abnormal operation and
detection of failures

Level 3

Control of accidents within the design
basis
Control of severe plant conditions,
including prevention of accident
progression and mitigation of the
consequences of severe accidents
Mitigation of radiological consequences
of significant release of radioactive
materials

Level 4

Level 5

Essential Mean
Conservative design and high
quality in construction and
operation
Control, limit & protection
systems and other surveillance
feature
Engineered safety features and
accident procedures

Chapter 3 – Design of Structures, Components, Equipment, and Systems
Chapter 4 – Reactor
Chapter 5 – Reactor Coolant System and Connected Systems
Chapter 6 – Engineered Safety Features
Chapter 7 – Instrumentation and Controls
Chapter 8 – Electric Power
Chapter 9 – Auxiliary Systems
Chapter 10 – Steam and Power Conversion System
Chapter 11 – Radioactive Waste Management
Chapter 12 – Radiation Protection
Chapter 13 – Conduct of Operations
Chapter 14 – Initial Test Program and ITAAC-Design Certification
Chapter 15 – Accident Analysis
Chapter 16 – Technical Specifications
Chapter 17 – Quality Assurance
Chapter 18 – Human Factors Engineering
Chapter 19 – PSA and Severe Accidents

1.3 Core thermal hydraulics safety analysis in transient condition
It is shown that the second barrier plays a very important role due to the fact that it protects
fission product in fuel rod from release to primary system. Therefore, the peaking cladding
temperature in transient and accident conditions must be satisfied specific acceptance criteria.
The acceptance criteria introduced by IAEA for transient condition in [21] is briefly
formulated as below:
(1) The probability of a boiling crisis anywhere in the core is low. This criterion is typically
expressed by the requirement that there is a 95% probability at the 95% confidence level that
the fuel rod does not experience a departure from nucleate boiling (DNB). The DNB
correlation used in the analysis needs to be based on experimental data that are relevant to the
particular core cooling conditions and fuel design.
(2) The pressure in the reactor coolant and main steam systems is maintained below a



Figure 1.3 Heal flux versus temperature difference for pool boiling heat transfers (source [31])

The boiling crisis in coolant flow is closely related to flow pattern and void fraction. In [25], it
is mentioned about two typical phenomenon of interest in nuclear safety as presented in
Figure 1.4.

Figure 1.4 Types of boiling flow crisis (source [25])

The left picture in Figure 1.4 shows the sub cooled or low quality DNB that is caused from
detachment of bubble from boundary layer. The picture in the right of Figure 1.4 shows the
burnout at high quality region that is caused from the liquid film covering heating surface
becomes ―dry out‖. The determination of critical heat flux at which DNB occurs is very
23


complicated due to the fact that it depends on many factors such as channel sharp, surface
condition, physical properties of the coolant and flow conditions. The most popular
correlation that ensures critical heat flux not exceeded during core operation is DNB ratio. It
is defined by the minimum ratio of the critical heat flux to the heat flux achieved in the core:
(1.1)
For the PWR, the DNBR > 1.3 for insurance of DNB not occurred. Figure 1.5 shows the
DNBR along axial channel in the uniform core.

Figure 1.5 Critical heat flux in uniformly core (source [25])

It is emphasized here that thermal hydraulics safety analysis in transient condition is dealing
with finding appropriate correlation that prevent DNB occurring in flow channel of the core.
Due to DNB occurrence is a very complicated mechanism but it is strictly related to void


As mentioned in [29], the experimental Castellana allows for the evaluation of exit flow
quantities within a 4x4 test assembly. The brief information about this test is given in Table 3.
Table 1.3 Castellana 4x4 test characteristics (source [29])

Items
Rod Diameter, in. (cm)
Rod Pitch, in. (cm)
Rod to Wall spacing, in. (cm)
Ratio of heat flux
Hot rods (H)
Cold rods (C)
Heated length, in. (cm)
Assembly Dimensions in. x in., (cm x cm)


Value
0.422 (1.072)
0.555 (1.410)
0.148 (0.376)
1.0
0.86
60.0 (152.4)
2.383 x 2.383 (6.053 x 6.053)

EPRI tests

EPRI 5x5 tests allows for collection of exit flow quantities, as well as rod temperatures near
assembly exit. The brief information about this test is given in Table 4.
Table 1.4 EPRI 5x5 characteristics for test 74 and test 75(source [29])


The BM ENTEK facility [33] allows investigation void fraction along heating channel similar
to Russia RBMK-1000 bundle of fuel rods. The channel contains a 7- rod bundle made by
stainless steel with rod outer diameter of 13.5 mm, 1.25 mm wall thickness, and 7 m length.
The bundle is contained within a stainless steel pressure tube (80 mm outer diameter and 5
mm wall thickness) with inner diameter of 49 mm and 10.5 mm wall thickness.


PSBT Void Distribution Measurements

The PSBT facility [1] allows measurement of void fraction in both bundle of rods or in single
sub channel. Some brief information of PSBT rod bundle is given in Table 1.5.
Table 1.5 Geometry and power shape for Test Assembly B5, B6, and B7 (Source [1])

Item
Rod Array
Number of heater rods
Number of thimble rods
Heater rod outer diameter (mm)

Assembly B5
5x5
25
0
9.50

Assembly B6
5x5
25
0