Mô phỏng tai nạn mất chất làm lạnh của lò phản ứng nước áp lực PWR bằng phần mềm PC TRAN

Mục lục: Chapter 1: BACKGROUND 1.1. Introduction to nuclear power 1.1.1. Nuclear power on the world . 1.1.2. Nuclear power necessity and the plan of nuclear power in Vietnam . 1.2. Principle of nuclear reactor 1.2.1. Neutron properties in nuclear reactor 1.2.2. Nuclear fission reaction 1.2.3. Radiation decay 1.2.4. Fuel material – coolant interaction 1.2.5. CoreConcrete interaction 1.3 Pressurized Water Reactor (PWR). 1.3.1. Development history 1.3.2. Operation principle 1.3.3. PWR structure Chapter 2: INTRODUCTION TO PCTRAN PWR SOFTWARE VERSION 4.0.8 2.1. PCTRAN overview 2.2. Introduction to PCTRAN PWR version 4.0.8 2.3. PCTRAN PWR interface (main micmic) 2.4. Simulation an accident 2.4.1. Malfunction setup 2.4.2. Initial conditions set up 2.4.3. Run simulation 2.5. PWR plant system control . Chapter 3: SIMULATING A LOSSOFCOOLANT ACCIDENT BY PCTRAN PWR SOFTWARE 3.1. Accident description 3.1.1. Coolant 3.1.2. LOCA 3.1.3. PWR LOCA 3.2. Set up simulation 3.3. Run simulation 3.4. Simulation analysis 3.4.1. SBLOCA properties. 3.4.2. LOCA consequences 3.4.3. Radiological consequences CONCLUSIONS AND PROPOSALS.. RFERENCES

DEPARTMENT OF NUCLEAR PHYSICS

- -

UNDERGRADUATE THESIS

STUDYING A LOSS - OF - COOLANT ACCIDENT (LOCA)

OF PRESSURIZED WATER REACTOR (PWR)

BY PCTRAN SOFTWARE

STUDENT: NGUYEN VAN THANG

SUPERVISOR: Dr VO HONG HAI REVIEWER: Dr HUYNH TRUC PHUONG

HO CHI MINH CITY – 2012

Foremost, I would like to express my sincere gratitude to my supervisor

Dr Vo Hong Hai He has been a knowledgeable, competent and helpful advisor, as well as a very kind-hearted and humble person I have greatly appreciated working and learning under his guidance

I also thank Dr Huynh Truc Phuong who is my reviewer spent a lot of time

to read and give me the honest feedbacks

Besides that, I would like to thank MSc Nguyen Quang Duy who is the predecessors and the brother He is the person who provided me the simulation software, books and the precious experience

I am grateful to Prof Chau Van Tao, the Dean of the Nuclear Physics department who facilitated my study complete on the schedule I also thank teachers

of Department of Nuclear Physics and Faculty of Physics and Engineering Physics provided me a lot of knowledge during four years at the Science University

I also thank my friends, for giving advices and help me much during the execution time of my thesis

Last but not the least; I would like to thank my parents, for given the birth

to me at the fist place and supporting me spiritually throughout my life

Ho Chi Minh City, June 17 th 2012 NGUYEN VAN THANG

CONTENTS

Page

Contents i

List of Abbreviations and Symbols iii

List of Tables iv

List of Figures v

Preface 1

Chapter 1: BACKGROUND 3

1.1 Introduction to nuclear power 3

1.1.1 Nuclear power on the world 3

1.1.2 Nuclear power necessity and the plan of nuclear power in Vietnam 4

1.2 Principle of nuclear reactor 5

1.2.1 Neutron properties in nuclear reactor 5

1.2.1.1 Neutron interaction cross section 5

1.2.1.2 Neutron flux 6

1.2.1.3 Neutron current density 7

1.2.1.4 Neutron slowing down 7

1.2.1.5 Neutron diffusion 8

1.2.2 Nuclear fission reaction 9

1.2.3 Radiation decay 11

1.2.4 Fuel material – coolant interaction 12

1.2.5 Core-Concrete interaction 14

1.3 Pressurized Water Reactor (PWR) 14

1.3.1 Development history 14

3.1.2 Operation principle 15

3.1.3 PWR structure 16

3.1.3.1 Reactor core 16

3.1.3.2 Reactor coolant 18

3.1.3.3 Moderator 18

3.1.3.4 Control rods system 18

3.1.3.5 Reactor vessel 18

3.1.3.6 Pressurizer 19

3.1.3.7 Steam generator 20

3.1.3.8 Containment system 22

Chapter 2: INTRODUCTION TO PCTRAN PWR SOFTWARE VERSION 4.0.8 23

2.1 PCTRAN overview 23

2.2 Introduction to PCTRAN PWR version 4.0.8 25

2.3 PCTRAN PWR interface (main micmic) 26

2.4 Simulation an accident 32

2.4.1 Malfunction setup 32

2.4.2 Initial conditions set up 34

2.4.3 Run simulation 34

2.5 PWR plant system control 36

Chapter 3: SIMULATING A LOSS-OF-COOLANT ACCIDENT BY PCTRAN PWR SOFTWARE 37

3.1 Accident description 37

3.1.1 Coolant 37

3.1.2 LOCA 37

3.1.3 PWR LOCA 38

3.2 Set up simulation 39

3.3 Run simulation 40

3.4 Simulation analysis 42

3.4.1 SBLOCA properties 42

3.4.2 LOCA consequences 48

3.4.3 Radiological consequences 49

3.4.3.1 Inside reactor radiation dose 49

3.4.3.2 Outside reactor radiation dose 50

CONCLUSIONS AND PROPOSALS 52

RFERENCES 54

APPENDICES 57

LIST OF ABBREVIATIONS AND SYMBOLS

PCTRAN Personal Computer Transient Analyzer

PWR Pressurized Water Reactor

BWR Boiling Water Reactor

LOCA Loss-of-Coolant Accident

ECCS Emergency Core Cooling System

HPIS High Pressure Injection System

LPIS Low Pressure Injection System

PCS Primary Coolant System

EAB Exclusion Area Boundary

LBZ Low Population Zone

IAEA International Atomic Energy Agency

INES International Nuclear and Radiological Event Scale USNRC United State Nuclear Regulatory Commission

σ Microscopic cross section (cm2 or barn)

Σ Macroscopic cross section (cm-1)

λ Mean free path of neutron (cm)

Ф Neutron flux (neutron.cm-2.s-1)

Eth Threshold energy of fission (MeV)

LIST OF TABLES

Page

Table 1.1: Threshold energy and binding energy of some fissionable nuclei 9

Table 1.2: Energy distribution of 235U fission reaction 10

Table 1.3: Components of concrete 14

Table 3.1: Comparison of LOCA frequency from studies 38

Table 3.2: Summary table of the set up accident 40

Table 3.3: Main events during the transient 42

Table 3.4: Exposure dose rate inside and out side reactor during simulation 51

LIST OF FIGURES

Figure 1.1: Nuclear electricity production from 1971 to 2009 1

Figure 1.2: Statistic of operating reactors number on the world and electricity production from 2000 to 2008 2

Figure 1.3: Schematic diagram of neutron interaction in the center of mass coordinate system 7

Figure 1.4: Schematic diagram of the fuel rod 12

Figure 1.5: Schematic diagram of pressurized water reactor system 16

Figure 1.6: Schematic diagram of fuel pellet 16

Figure 1.7: Schematic diagram of the PWR fuel rod 17

Figure 1.8: A grid of 17×17 fuel assembly 17

Figure 1.9: Schematic diagram of PWR fuel assembly 17

Figure 1.10: Schematic diagram of PWR vessel 19

Figure 1.11: Schematic diagram of presurizer operation 20

Figure 1.12: Schematic diagram of the U-Tube generator 21

Figure 1.13: Schematic diagram of containment system 22

Figure 2.1: PCTRAN interface for PWR 2 loops 26

Figure 2.2: Menu bar and the toolbar of PCTRAN 26

Figure 2.3: Status bar of PCTRAN 27

Figure 2.4: A loop of PCTRAN 27

Figure 2.5: Control operation status of reactor 28

Figure 2.6: Status of the reactor protection system (RPS) and the emergency core cooling system (ECCS) 29

Figure 2.7: Operation parameter of the core 29

Figure 2.8: Status parameter of reactor building 30

Figure 2.9: Emergency core cooling system 31

Figure 2.10: Secondary coolant system of PWR 31

Figure 2.11: Pumps and the valves operation status 32

Figure 2.12: Malfunction setup 33

Figure 2.13: Initial condition setup 33

Figure 2.14: Graphs of PCTRAN 34

Figure 2.15: Dose micmic 35

Figure 3.1: Setting up malfunction 39

Figure 3.2: Change malfunction status of the pump 39

Figure 3.3: Choose an initial condition in Initial Conditions window 40

Figure 3.4: PCTRAN display in 110s 41

Figure 3.5: Graph of pressure for SBLOCA 43

Figure 3.6: Graph of neutron flux 43

Figure 3.7: Graph of thermal power 43

Figure 3.8: Graph of the leakage coolant flow 44

Figure 3.9: Graph of temperature of fuel and cladding 44

Figure 3.10: Graph of cladding failure 45

Figure 3.11: Activity of 131I, 135Xe, 138Xe and 87Kr in coolant 45

Figure 3.12: Background of PCTRAN in 2620s 46

Figure 3.13: Dependence of temperature of fuel to HPIS activation time 47

Figure 3.14: Dependence of temperature of coolant to HPIS activation time 47

Figure 3.15: Exposure dose inside reactor 49

Figure 3.16: Dose rate EAB thyroid and whole body 50

PREFACE

2011 is the year when many countries in the world experienced the largest crisis of nuclear energy in the history That caused Fukushima Daiichi nuclear power plant disaster on March in Fukushima prefecture, Japan This accident one more time was an alarm for the safety of nuclear power plants, which are the main electric supply of many development countries After the event, the Government of several nations declared their nuclear programs to use safer energy source

In Vietnam, in recent years, the lack of electricity has been more and more serious Some of the traditional energy sources are running out Hydropower is not enough for requirement Besides that, the construction many hydropower plants always accompanied many bad effects for environment Vietnam cannot use popular infinity energy (e.g wind energy, solar energy) because of the high price Therefore, the Government of Vietnam decided to build the nuclear power plants though the society had many contrast opinions On June 17th 2010, Prime Minister makes a decision about the development orientation of the nuclear power in Vietnam henceforth to 2030 Follow that, until 2030, Vietnam will have 13 nuclear power units The important problem is the human resource Currently, the resource for nuclear power is poor, thus, the Government is planning to widen and upgrade training facilities The target is training engineers who will work for the nuclear power plants

The use of software for simulation reactor is popular in the world Many countries use reactor simulation software for training at the universities Nowadays, there are many computer programs simulating for reactor operation (e.g CASSIM, RELAP and CATHARE) PCTRAN is also the simulation software for reactor The feature of this software is focusing to simulation of reactor accidents

In Vietnam, PCTRAN has been only used for research for the recent years PCTRAN was also the subject of a few science articles of the Vietnam Atomic Energy Institute In 2010, the master thesis “Tìm hiểu cấu trúc và mô phỏng sự cố

lò phản ứng nước áp lực hai vòng bằng phần mềm PCTRAN” of MSc Nguyen

Quang Duy was performed at Can Tho University In the thesis, he simulated two accidents of pressurized water reactor Those were “Loss of Main Feed Pumps” and

“Turbine Trip” From March 7th

, 2011 to March 9th, 2011, the International Atomic Energy Agency (IAEA) held the workshop to train for the staff of Vietnam Agency for Radiation and Nuclear Safety using PCTRAN to simulation the various accident situations

In this thesis, we also used this PCTRAN version provided license by IAEA A loss-of-coolant accident was simulated

The thesis includes three chapters:

Chapter 1: Background

This chapter provides the basic knowledge on nuclear power and nuclear reactor physics The principle operation and structure of a typical pressurized water reactor is described clearly to build the background for the simulation in chapter 3 Chapter 2: Introduction to PCTRAN PWR software version 4.0.8

This chapter introduces briefly to the PCTRAN software and guides to use PCTRAN software to simulate the various accidents

Chapter 3: Studying a loss-of-coolant accident (LOCA) by PCTRAN PWR software

In this chapter, we simulate the loss-of-coolant accident During the transient the ECCS is disabled Base on simulation results, we study the response of the reactor

to the severe accident

Chapter 1 BACKGROUND

1.1 Introduction to nuclear power

1.1.1 Nuclear power in the world [9], [10], [11], [17], [18], [25]

Nuclear power is the use of sustained nuclear fission to generate heat and electricity Nuclear power plants provide 6% of the world’s energy and 13-14% world electricity [9] In 2007, the International Atomic Energy Agency (IAEA) reported there were 439 nuclear power reactors in operation on the world, operating

at 34 countries [9] On December 2009, there were 436 nuclear power reactors operating on the world with total power is 370499 MWe There were 5 nuclear power reactors in long term shutdown and 62 nuclear power reactors under construction [10] Since the commercial nuclear power plant operated in the middle

of 1950s, 2008 was the first year that no new nuclear power plant was connected to the grid, although two were connected in 2009 [17]

Figure 1.1: Nuclear electricity production from 1971 to 2009 [25]

The accident of Three Mile Island nuclear power plant in the south of Harrisburg, Pennsylvania, United State on March 28th 1979 led to partial melting down and released a small amount of radioactive gases and radioactive Iodine into

environment [18] After that, there has been no new nuclear reactor operating in The United State The accident of Chernobyl nuclear power plant in Ukraine on April

26th, 1986 released amount of radioactive materials which are 4 times larger than their from Hiroshima atomic bomb disaster [11] The accident made the concern about safety of nuclear power plants, and delayed nuclear program in many years later Most recently, The Fukushima Daiichi nuclear power plants accident in Fukushima prefecture, Japan on March 11th, 2011 caused a nuclear crisis Many nations considered to give up nuclear power to use a safer power

Figure 1.2: Statistic of operating reactors number in the world and electricity

production from 2000 to 2008 [17]

Figure 1.2 shows the chart of nuclear power usage on the world from 2000 to

2008 The column chart is the operating reactors in the world and the line chart is the percentage of global electricity production The number of nuclear reactors has not changed more and there has been a decline of nuclear power production from 16.7% in 2000 to 13.5% in 2008 [20]

1.1.2 The necessity and the plan of nuclear power in Vietnam [6], [16]

With the development of industries, social rapidly in Vietnam, nuclear power should be considerable and necessary for the future electricity shortage, while hydro and thermal plants are not enough supply The fossil energy is also running out while nuclear resources have just mined Follow the statistic [16], total Uranium resources that were mined in 2007 were 5.5 million tons increased 17% compare

with the level of 2005 At current 2006 rates of consumption, identified resources are sufficient for about 100 years of supply Total undiscovered resources (prognosticated resources and speculative resources) in 2007 amounted to more than 10.5 million tons increasing by 485000 tons from the total reported in 2005 At the end of 2006, there were 435 commercial nuclear reactors were operating with a net generating capacity of about 370GWe, requiring about 66500 tons of Uranium

By 2030, a world nuclear capacity will grow between 509GWe and 663GWe, Uranium requirement is between 93775 and 121955 tons [16] Therefore, nuclear energy will remain the main energy of future

Vietnam plan to build the fist two nuclear power plants Ninh Thuan I and Ninh Thuan II Ninh Thuan I nuclear power plant (NPP) will be schedule to operate in

2021 This reactor is constructed as a pressurized water reactor following Russian technique VVER Ninh Thuan II will operate in 2020, construction follows Japanese technique Vietnam also plan for more 11 NNPs for later It expected that electric power from NNPs will reach to 15000MWe and occupy 10% total electric power [6]

1.2 Principle of nuclear reactor

A nuclear reactor may be defined is the instrument that energy is released by chain reaction between neutron and fissile nuclides Heat which is generated by fission and radiation is used to covert to electricity The system operation is researched by principle of classical gas dynamic However, the interaction among component particles is explained by the concepts of nuclear physics

1.2.1 Neutron properties in nuclear reactor

1.2.1.1 Neutron interaction cross section [3]

The interaction of neutrons in nuclear reactor is separated into 4 kinds: elastic scattering, inelastic scattering, capture radiation and fission The probability interaction of a neutron and a nucleus is measured by microscopic cross section σ If

we sign σeis the elastic scattering cross section, σiis the inelastic scattering cross

section, σcis the capture radiation cross section and σfis the fission cross section, σ may be written by formula (1.1)

σ =10barn, σ =107c barn and σ = 580f barn [3]

Σ = N σ, with N is the number of nuclei per a unit volume, is called macroscopic cross section The usual unit of Σ is cm-1

The attenuation of neutron in the distance x is presented by [3]

-Σx 0

2 - 2 0

mv 2kT

Where

3 2

mA=4π

πm

p

2kTv

m

In formula (1.8), (1.9), (1.10) and (1.11), m is the neutron mass, T is temperature in Kelvin and k is the Boltzmann constant

1.2.1.3 Neutron current density [3]

Neutron current density in a volume unit is given by equation (1.12) Equation (1.12) is the equation of Fick law

λ = 0.7 with d is the extrapolated length

1.2.1.4 Neutron slowing down [3]

In the reactor, neutrons scatter on coolant This progress make neutron lose energy

Figure 1.3: Schematic diagram of neutron interaction in the center of mass

It is assumed that E and E’ is the neutron energy before and after interaction;

c

θ is the deviation angle of neutron in the center of mass coordinate system

2 c 2

  with A is the nucleus mass

The scattering angle of neutron in laboratory system is calculated by (1.14)

c 2 c

1+Acosθcosθ =

The average energy loss in each interaction ξdepends on nucleus mass number

A For 1H, ξ = 1; 2H, ξ = 0.7261 It proves that 1H is the best slowing down material

The leak rate L per a unit volume is written

In the critical reactor, the number of neutrons per a unit volume is unchanged Thus, there is a balance among the neutron production rate S, neutron absorption rate A and neutron leak rate L

a

Equation (1.18) is the neutron diffusion equation which is the basic equation in nuclear reactor physics The solution of the equation is the neutron flux distribution

in the reactor

1.2.2 Nuclear fission reaction [1], [2], [3], [22]

Nuclear fission is a type of nuclear reactor which is main energy source of nuclear reactor A fission reaction is the division a fissile nucleus such as 235U to two or many nucleus Fist 235U absorbs a neutron and becomes an excited nucleus

2 2/3

Table 1.1: Threshold energy and binding energy of some fissionable nuclei [2]

Nucleus Threshold energy Eth

(MeV) Composite nucleus

Binding energy BE of composite nucleus

Recording to table 1.1, we can see 233U, 235U and 239Pu have BE > Eth Therefore, these nuclei get are fissionable with any kinetic energy including thermal neutron They are called fissionable nuclei 232Th and 238U are fissionable with only neutrons which have kinetic energy higher than Eth Normally, 233U, 235U and 239Pu are used as fuel in thermal reactors and 232Th and 238U are used as fuel in fast reactors 232Th, 238U and 235U exist in nature 238U occupies 0.7% natural Uranium

Reactors which are used to produce fuel like that are the breeder reactor

Each fission reaction releases energy about 200MeV The majority of fission energy concentrates at kinetic energy of fragments A little energy is from kinetic energy of prompt neutron, prompt gamma, capture gamma and decay of fission products The energy distribution of 235U fission is shown in table 1.2

Table 1.2: Energy distribution of 235U fission reaction [22]

Kinetic energy of fission products

Kinetic energy of prompt gamma

Kinetic energy of prompt neutron

Kinetic energy of capture gamma

Decay of fission products Kinetic of electrons Kinetic of positrons

1.2.3 Radiation decay [2], [5], [13]

Small part energy in reactor is energy from radiation decay Decay heat increases with operation time Most nuclides generated by fission are unstable nuclei They emit beta and gamma decay to become the stable nuclide Equations (1.25) and (1.26) are 2 decay series of 2 nuclides 140Xe and 135Te [2]

14054Xeβ14055Csβ14056Baβ14057Laβ14958Ce (stable) (1.25)

13552Teβ13553Iβ13554Xeβ13555Csβ13556Ba (stable) (1.26) Besides that, 235U and 238U also decays to 233U and 239Pu following series (1.23) and (1.24) Radiation decay progresses in a nuclear reactor are explained by basic radiation law Radiation law for the radiation series (1.27) is given by (1.28), (1.29) and (1.30) [5]

Where λ , λ and λ1 2 3are the haft life of nuclides A1, A2 and A3; R1(t), R2(t) and

R3(t) are the activity of nuclides A1, A2 and A3 at the time t sec; and N1(0) is the number of nuclei of nuclide A1 at 0 sec

When a reactor is operating, the activity of fission products decreases continuously After a few days reactor operating, the energy from beta and gamma decay of fission products amount to about 7 % of total energy output reactor [13] After reactor core is shutdown by insertion of control rods, heat continues to

be generated by decay of fission products although reactor stopped to generate The heat generation of fission products after reactor shutdown is also called afterheat or decay heat If decay heat cannot be removed from the core, it would lead to fuel

damage, steam cladding interaction leads to hydrogen generation, melting even vaporization of the core

Decay heat of 235U and 238U after shutdown is calculated by formula (1.31) and (1.32) [13] Assuming that fuel is 235U, of course, it contains 238U The decay of 239U and 239Np is formed by absorbing neutrons of 238U Using the equation of radioactive decay, it is easy to calculate [13]

25 -3

25 0

- 4.91×10 t - 4.91×10 t

P = 2.28×10 CP

25 0

-7×10 (1 - e

)

1.2.4 Fuel material – coolant interaction [14], [15]

Interaction between coolant and fuel material is called the oxidation This progress takes place within the reactors at high temperature When a reactor is operating normally, the oxidation is not significantly

Nuclear fuel is covered by an alloy called cladding Between fuel and cladding

is the gap which is used to contain fission gases release from fission The cladding

is made from Zirconium Alloy because it can sustain the high temperature in reactor (over 8000C)

Zirconium alloys have high melting temperature but it is destroyed by the oxidation with coolant at temperature above 1000oC [14].

Hydrogen concentration in containment makes the pressure of the containment high and may damage to the containment, when hydrogen concentration exceed 4%, hydrogen burns with air [14]

Failure containment occurs with small probability, but very dangerous because radioactive materials will release with large activity

1.2.5 Core-Concrete interaction [14]

Core-concrete interaction only occurs in reactor in “Chinese Syndrome” Concrete which is material of the containment consists of various elements such as

Ca, Fe, Si, Al, Na, Mg, Mn, Cr, etc

Table 1.3: The components of concrete

Mass fraction of concrete Component

1.3 Pressurized Water Reactor (PWR)

1.3.1 Development history [4], [12]

Pressurized Water Reactor (PWR) is the type of reactors which are used popular in Europe and United State PWR is different to Pressurized Water Reactor which is a series of designs originally developed in Russia They are well-known with the name VVER or WWER PWR is one of three types of Light Water Reactor (LWR) The other types are Boiling Water Reactor (BWR) and Supercritical Water Reactor (SWR) PWR is the reactor which belongs to generation II reactors

The fist PWR was used as the power source for Submarine in 1954 The research and the development work were performed by Knolls Atomic Power Laboratory and Westinghouse Bettis Laboratories United State Military used

PWRs for Army Nuclear Power Program from 1954 to 1974 Three Mile Island Nuclear Generating Station initially operated two PWR plants, TMI-1 and TMI-2 The partial meltdown of TMI-2 in 1979 essentially ended the growth in new construction nuclear power plants in the United States

There is approximate 60% number of operation reactors around the world is PWR [4] Many reactors of submarines use PWR technology Some reactors in the design of PWR are [4]:

+ Westinghouse 1 loop Zorita (Spain)

+ Westinghouse 2 loops Ginna (US) and Krsko (Slovenia)

+ Westinghouse 3 loops Turkey Point (older) and North Anna (newer)

+ Westinghouse 4 loops San Onofre-1 (original and shutdown); Zion (older and shutdown); Callaway (newer)

+ Combustion Engineering 2 loops Calvert Cliffs

+ AP600, AP1000, EPR, APWR, etc

Eventually, several commercial PWR were supplied by Westinghouse, Badcock and Wilcox, and Combustion Engineering in USA; Siemens in Germany; and Framatome in France Subsequently, Mitsubishi in Japan and Agip Nucleari in Italy become PWR licensees [12]

1.3.2 Operation principle [23]

Principle of PWR operation is performed in figure 1.5 The primary coolant is pumps under high pressure to the reactor core where it is heated by energy which is generated by nuclear fission The coolant of PWR is light water (H2O) H2O is used for both coolant and neutron moderator The heated water then flows to steam generators where water transfer heat to secondary system, the secondary coolant boils and generate steam Steam flows to the turbine, spins the turbine and generates electricity

Figure 1.5: Schematic diagram of pressurized water reactor system [23] 1.3.3 PWR structure

1.3.3.1 Reactor core [7], [24], [26]

Basic units of the reactor core are the fuel pellet (or fuel block) The fuel pellet

of PWR is the cylindrical make from UO2 powder The dimension is 1 cm in diameter and 1 to 2 cm in height [7]

Figure 1.6: Schematic diagram of fuel pellet

The PWR fuel is made from Uranium dioxide (UO2) with 2.5 - 5% enriched U-235 Many fuel pellets are stuck into the metallic cylindrical which is called fuel rod or pin A pin consists of 400 fuel pellets The size is different between the various reactors For PWR, it is about 9-10 mm in diameter and 400 mm in length The fuel pellets do not occupy fully the pin The free volume (plenum) is at the top

of the pin used for containing the fission gas The pellets are kept in place by the plenum spring To contact between the plenum spring and top of the pellet an insulator pellet (Al2O3) is placed

1 cm

Figure 1.7: Schematic diagram of the PWR fuel rod [7]

The fuel rod surface is made from material which is sustainable with the high temperature of fuel The fuel is covered by is Zicaloy 4 which is the alloy of Zr and 1.45% Sn, 0.125% O, 0.21% Fe, 0.1% Cr with the melting point is 2100oK [7] PWR fuel rods are assembled in a square geometry generally ranging between (14×14) to (17×17) [7] The (17×17) fuel assembly consists of 264 fuel rods and 24 control rods

Figure 1.8: A grid of 17×17 fuel assembly [24]

Fuel assembly consists of fuel rods are bundled with grids, and the fuel assembly is equipped with top and bottom nozzles A top nozzle is designed to allow handing during loading and uploading

Figure 1.9: Schematic diagram of PWR fuel assembly [26]

A typical PWR would have 150-250 such assemblies with 80-100 tones of fuel

in all Refueling for most commercial PWRs is on an 18-24 month cycle [7]

1.3.3.2 Reactor coolant [14]

Light water (H2O) is used as primary coolant in PWR Temperature of water is different between bottom and top of reactor core The difference is about 40oC The water enters the bottom of the core at about 280oC then it flows upward the core, and it is heated to the temperature at about 320oC Although above the boiling point, the water does not boil because of high pressure of primary coolant system This pressure is approximate 155 (kg/cm2)

1.3.3.3 Moderator

PWR is a type of thermal reactor which requires the fast neutrons to be slowed down in order to interact with nuclear fuel to and sustain chain reaction In PWR, the coolant water is used for the moderator

1.3.3.4 Control rods system [15]

The thermal power of reactor is dependent on power neutron flux In order to absorb neutrons, the control rods are usually made from materials which well absorb neutron For PWR, the control rods are made from Ag-In-Cd alloy [15] The control rods insert directly into fuel bundles for the following reasons:

- To start up reactor

- To shutdown fission reactions in the reactor

- To accommodate short term transients such as changes to load the turbine

1.3.3.5 Reactor vessel [20], [27]

The reactor core is housed within the reactor vessel The reactor vessel is the cylindrical with a hemispherical bottom head and a removable hemispherical top head The top head is removable to allow for the refueling of the reactor There is an inlet nozzle (a hot leg) and an outlet nozzle (a cold leg) for each reactor coolant loop The reactor vessel is constructed of Manganese-Molybdenum-steel and inside surface is clad with stainless steel to decrease corrosion [27] Schematic of PWR vessel is shown in figure 1.10

Figure 1.10: Schematic diagram of PWR vessel [20]

1.3.3.6 Pressurizer [20]

The pressurizer is a component of the primary coolant system which is used for controlling pressure of primary coolant system The pressure is controlled by the electric heater, the pressurizer pray, power operated relief valves (PORV) and safety valves The principle schematic of the pressurizer is described in figure 1.11

If reactor coolant temperature increases, the density of coolant will decrease and the coolant will take up more space Thanks to the pressurizer pray, coolant will expand up into pressurizer The steam will be discharged by PORV

INSTRUMENTATION PORTS

THERMAL SLEEVE

HOLD DOWN SPRING

ROD TRAVEL HOUSING

LIFTING LUG

CLOSURE HEAD ASSEMBLY

CONTROL ROD GUIDE TUBE CONTROL ROD GUIDE SHAFT

INLET NOZZLE CONTROL ROD CLUSTER (WITHDRAWN

ACCESS PORT REACTOR VESSEL

LOWER CORE PLATE CORE SUPPORT

BAFFLE RADIAL

SUPPORT OUTLET NOZZLE

UPPER CORE PLATE

SUPPORT COLUMN

CORE BARREL

INTERNALS SUPPORT LEDGE

UPPER SUPPORT

PLATE CONTROL ROD DRIVE

MECHANISM

Figure 1.11: Schematic diagram of presurizer operation [20]

If reactor coolant temperature decreases, the density of coolant will increase and the coolant will take up less space The pressurizer water will decrease The electric heater boils water into steam and therefore increases pressure If pressure continues to decrease, and reaches a predetermined set point, the reactor protection system will shut down the reactor

1.3.3.7 Steam generator [20], [27]

There are two types of steam generators used for PWR designs The U-tube design is the Westinghouse and Combustion Engineering designs and the once-through design is the Badcock & Wilcox designs [27] The U-tube design is shown

in figure 1.12

Figure 1.12: Schematic diagram of the U-Tube generator [20]

The function of the steam generator is to transfer heat from primary coolant feed water at the secondary system In the steam generator, the hot water flow into many tubes which are bundle into the bundle The feed water flows outside the tube,

it absorbs heat and vapor

TUBE SHEET PRIMARY MANWAY

PRIMARY COOLANT INLET

DEMISTERS SECONDARY MOISTURE SEPARATOR

STEAM OUTLET TO TURBINE GENERATOR

TUBE LANE BLOCK

PRIMARY COOLANT OUTLET

1.3.3.8 Containment system

Figure 1.13: Schematic diagram of containment system [28]

The containment system includes the primary coolant system and the component which injects water into primary coolant system It is called emergency core cooling system (ECCS) The ECCS includes 4 main parts [27]

- Accumulators are the large vessels containing water under nitrogen water They are connected to primary coolant system by the automatic valves which open when pressure of primary system reduces below 40bar

- The high pressure injection system (HPIS) injects water when pressure is about 100bar but low rate

- The low pressure injection system (LPIS) injects water when pressure is below 30bar

- The containment prays use to quench any steam released in the accidents

Micro-The first version of PCTRAN was introduced in 1985 by Dr Li-Chi Cliff Po

He is the founder of PCTRAN and the director of Micro-Simulation Technology Since that, PCTRAN has been selected by IAEA as a training platform for its annual Advanced Reactor Simulation workshop In this program, PC TRAN license

is provided to individuals or organizations having requirement for studying, researching or operating IAEA also holds training courses at nuclear power plants and institutions using available version of PCTRAN This course took place in Vietnam on March, 2011 Specific PCTRAN models have been installed at nuclear power plants and institutions all over the world for practical application in training, analysis, probabilistic safety assessment, and emergency exercises

Micro-Simulation Technology (MST) has just released two versions BWR4 and BWR5 simulating all four units of Fukushima electric station Every year, MST releases many PCTRAN versions and receives many orders MST confirms their products are the most widely-used educational simulators (and perhaps the only one) in the world Several universities, colleges, institute and nuclear power plants use PCTRAN

PCTRAN licensees in 2011:

- The Center of Nuclear Technology Education Consortium (NTEC) in United Kingdom at University of Manchester has acquired our AP1000 and EPR modules

- Nanking University of Aeronautics and Astronautics (NUAA) and Northeast DianLi (Electricity) University have acquired AP1000 module

- Fukui University of Technology in Japan has got the latest versions of PCTRAN

of PWR and BWR It expands the capability to cold startup in normal operation and Fukushima-type severe accidents

- Tsinghua University at Beijing, East China Institute of Technology, Shanghai Jiaotong and Xian Jiaotong have obtained AP-1000 for their education and research

- Tsinghua University in Taiwan has obtained the high-temperature gas reactor

model for its joint project between Taiwan and United State on Generation reactor development

- Stevens Institute of Technology in New Jersey got both PWR and BWR for its new nuclear engineering program

- IAEA Award of a VVER-1000 Simulator for Vietnam

The United Nations agency has awarded a contract to MST to prepare a PC-based simulator of the Russian-designed NPP for Vietnam Agency for Radiation and Nuclear Safety (VARANS) It was delivered and a training class was conducted in Hanoi in March of 2011

Recent PCTRAN licensees 2012:

- University of Massachusett at Lowell has acquired AP1000

- Southern Polytechnic State University at Marrietta, Georgia has acquired AP1000 and RadPuff

- Southeast Community College at Licoln, Nebreaska has acquired Cooper BWR1 Mark I) model

(GE-2.2 Introduction to PCTRAN PWR version 4.0.8 [13]

PCTRAN PWR version 4.0.8 simulates the pressurized water reactor 2 loops with the U-bend steam generators and dry containment system It is type of reactor which is used popularly in Europe It could be a Westinghouse, Framatome or KWU design with thermal output is 1800MWt (600MWe) A single loop with the pressurizer is modeled separately from the other loop There are a number of PWR plants in the world belong to this category: Point Beach, Kewaunee, Prairie Island and Ginna in the US, Mihama 1 in Japan, Krsko in Slovenia and ChinShan 2 in China

The Windows version is a state of the art product taking full advantage of bit PC technology The source code is completely re-written using Microsoft Visual Basic 6.0 Operation follows the Microsoft Windows 95, 98, 2000 or NT operating system’s environment strictly Data input and output are in Microsoft Office 97’s Access database files Since 2003, International Atomic Energy Agency (IAEA) and International Centre for Theoretical Physics (ICTP) upgraded PCTRAN system

32-In addition, based on workshop experience, there are strong demand from the participants to incorporate real-life plant events and severe accident features in the program The following features are thus included:

- Severe accident features such as hydrogen burns at high concentration in the tainment

con Radiological release source term generation for offsite dose projection

- A complete revision of the User’s Manual using Word 2000 with color graphics This manual presents the typical PWR Final Safety Analysis Report (FSAR)

- An independent spent fuel pool simulator

2.3 PCTRAN PWR interface (main micmic)

Figure 2.1 is the PCTRAN interface recorded when the reactor operates with 100% full power at 0 sec, no malfunction setup The structure of the realistic reactor

is very complex Therefore, just main components are arranged scientifically on the computer desktop The details of components will in turn be described

Figure 2.1: PCTRAN interface for PWR 2 loops

Similar with programs which run on Window environment, all control functions are on the menu bar and the tool bar, and the operation status is on the status bar

Figure 2.2: The menu bar and the tool bar of PCTRAN

(9) View PT plot (10) View transient report (11) PCTRAN options (12) Help contents

Figure 2.3: The status bar of PCTRAN

(1) Click to change run time

(2) Transient time

(3) Transient and real time

(4) Initial condition number (5) Malfunction active (6) Status of plottingThe center of the main micmic is presented in figure 2.4 The reactor vessel is

a cylindrical vessel with a hemispherical bottom head and a removable hemispherical top head The top head is removable to allow the refueling of the reactor There are two inlets (or cold legs) nozzle and two outlets (or hot legs) nozzle The temperature of coolant is 320.50C within hot legs and 281.50C within cold legs The fuel assembly is arranged typically for PWR types The fuel rods are

at the lower position of vessel and the control rods are at the higher position of the vessel When the reactor shut down, fuel rods will be inserted among fuel rods

Figure 2.4: A loop of PWR

Pressurizer Relief Tank

Fuel Rods

The U-bend steam generator is the feature of Westinghouse PWRs The pressure

of steam generators water is 55 kg/cm2, and water occupies 50 % steam generators The primary coolant flows within reactor vessel and U-tubes by circulation pumps The coolant flow is large to ensure that all heat is removed from the core On figure 2.4, the coolant flow reaches 15265 t/hr Although the temperature of coolant exceeds the water boiling point, water remains fluid The pressure is kept constantly

by a pressurizer which is connected to the cold leg When the reactor operates normally, the pressure of primary coolant is kept at 155 kg/cm2

The operation status of the reactor core, steam generators and pressurizer are presented in figure 2.5 (a) and figure 2.5 (b) The status could be changed on the main micmic by clicking on the button “A” or “M” and is set up as figure 2.5 (c) The power demand is set default 100% and can be changed 0-120% The power demand is set default 10% and can be changed 0-20%/min The steam generator level (SG LvL Stpt) is set default 50% and can be changed 0-100% The steam generator pressure (SG Press Stpt) is set default 55 kg/cm2 and can be changed 0-

100 kg/cm2 The pressurizer level (Level Stpt) is set default 55.6% and can be changed 0-100% The pressurizer pressure (Press Stpt) is set default 155 kg/cm2 and can be changed 20-160%

Figure 2.5: Control operation status of reactor

(a) Control the core parameters (b) Control pressurizer parameters (c) Set value window

A miniature control room records operation parameters during the transient in order to operators can observe easier The table in the figure 2.6 shows the status of the Reactor Protection System (RPS) and the Emergency Core Cooling System (ECCS) The reactor will be scrammed automatically upon conditions exceeding any of the RPS set points and the corresponding symbol will turn into red For example, if the reactor pressure is below the setpoint for low-pressure trip, 127 bar, the symbol “Lo RxP” and the button “reactor” will turn into red

The table 1 consists of low set points (Lo) and high set points (Hi) of RPS The reactor will be shutdown if it exceeds one of set points of reactor coolant pressure (RxP), coolant flow (Flw), steam generator level (SGL), neutron flux (Fx), Over-temperature delta-T (OTDT), Over-power delta-T (OPDT) and pressurizer level (PzL)

Figure 2.6: The status of the Reactor Protection System (RPS) and the Emergency

Core Cooling System (ECCS) The table 2 consists of low-low set points (LL), low set points (Lo) and high set points The ECCS will be activated automatically if operation status parameters exceed one of set points of RxP, PzL, steam generator pressure (SGP), and reactor building pressure (RBP)

Figure 2.7: The operation parameter of the core

Figure 2.7 shows the table of the core operation parameters when the reactor is operating with 100% power (1800 MWth) The average temperature of coolant

Table 1

Table 2

(Tavg) is 3010C; average temperature of fuel (Avg Fuel Temp) is 788.90C and the maximum temperature of cladding (Max Clad Temp) is 320.50C The void is the volume percentage of vapor in the primary coolant system When the PWR operates normally, the void is 0%

Figure 2.8: The status parameter of reactor building

Reactor building or containment contains all of primary systems, reactor protection systems and emergency core cooling systems The PCTRAN emergency core cooling system (ECCS) shown in figure 2.9 includes the high pressure injection system (HPI/CVC), the accumulator, the low pressure injection system (LPI/RHR) and the reactor building pray (RB pray) ECCS activation is upon to the pressure of primary coolant system Pumps of ECCS operate by electric source of the plant and diesel powers in the emergency situations Coolant of ECCS is gotten from Refueling Water Storage Tank (RWST) The volume of RWST (RWST LvL)

is 2600 m3

Figure 2.9: Emergency core cooling system

Figure 2.10: Secondary coolant system of PCTRAN

Turbine

Condenser Main Feed Pumps

Low Pressure Injection

Accumulator

High Pressure Injection

Electric generation system or secondary coolant system is the component that heat is converted to electricity The electric generation system of PCTRAN is shown in figure 2.10 including 2 turbines The turbine which is operating is pink and not operating is blue The condensers are connected to the turbines, and they are the place that steam condenses before return the steam generators In order to water flows within the secondary coolant system, the main feed water pumps (MFWPs) with high capacity are used The feed flow is set default 1772.2 t/hr

Figure 2.11: The pumps and the valves operation status

On the main micmic, the pumps and the valves status is separated by the color

as figure 2.11 (a) and 2.11 (b) The red are the operating instruments and the white

is the non-operating instruments The status can be changed by right clicking on the buttons and setting 0-100% as figure 2.11 (c)

2.4 Simulation an accident

2.4.1 Malfunction setup

Accidents of a reactor are separated into two types: design basic accidents and severe accidents A design basic accident is a postulated accident that a nuclear facility must be designed and built to withstand without loss to the systems, structures, and components necessary to ensure public health and safety [31] A severe accident is a type of accident that may challenge safety systems at a level much higher than expected [32] PCTRAN simulates twenty malfunctions which is design basic accidents of PWR The list of malfunctions is listed in Appendix A