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
VIETNAM NATIONAL UNIVERSITY – HO CHI MINH CITY UNIVERSITY OF SCIENCE FACULTY OF PHYSICS AND ENGINEERING PHYSICS 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 ACKNOWLEDGEMENTS - -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 17th 2012 NGUYEN VAN THANG [i] CONTENTS Page Contents i List of Abbreviations and Symbols iii List of Tables iv List of Figures v Preface 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 1.2.1.1 Neutron interaction cross section 1.2.1.2 Neutron flux 1.2.1.3 Neutron current density 1.2.1.4 Neutron slowing down 1.2.1.5 Neutron diffusion 1.2.2 Nuclear fission reaction 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 [ii] 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 [i] 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) [ii] LIST OF TABLES Page Table 1.1: Threshold energy and binding energy of some fissionable nuclei 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 [iii] LIST OF FIGURES Figure 1.1: Nuclear electricity production from 1971 to 2009 Figure 1.2: Statistic of operating reactors number on the world and electricity production from 2000 to 2008 Figure 1.3: Schematic diagram of neutron interaction in the center of mass coordinate system 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 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 [iv] 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 [1] 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 mơ cố lò phản ứng nước áp lực hai vòng phần mềm PCTRAN” of MSc Nguyen [2] 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 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 [61] IC 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 Acronym W3 UC MBK FRCL PSG0 LSGA LSGB SGCT LSLD IEFW ILEV ISAT ICORE ILIF KSGA IHPI IRCA IRCB ICFT ILPI ICRT X IMFW MSGA MSGB PSGA PSGB XSGA XSGB PTBVA PTBVB NSG2A NSG2B WTBA WTBB RHX WFWA WFWB WRCA WRCB WMFW0 LVCR PN2 PRB Value -5.748 0.1053 0 77.14 11.83 11.83 41350 0 0 1 0 0.09583 82000 164000 68.99 68.99 0.03239 0.03239 69 69 50 50 0 1835 7888 16520 33030 5504 3.65 44.32 1.034 Description Interconnecting flow rate (t/hr) Steam generator heat transfer coefficient at 100% power Integrated break flow in the containment (kg) Clad damage fraction (%) Steam generator pressure at zero power (Kg/cm2) A SG wide range level (M) B SG wide range level (M) Mass to level conversion constant ( = MSG0/LSG0)) Pressurizer top discharge flow (steam/liquid) EFW not initiated/initiated Pressurizer not drained/has drained RCS subcooled/saturated Reactor at power/tripped Core life beginning/middle/end of cycle Number of steam generators at the left hand A-side HPI not initiated/initiated RCP-A tripped/operating RCP-B tripped/operating CFT not initiated/initiated LPI not initiated/initiated LOCA flow (critical/noncritical) RCS two-phase volume quality Main feedwater status (on/tripped A SG liquid mass (kg) B SG liquid mass (kg) A SG pressure (Kg/cm2) B SG pressure (Kg/cm2) A SG quality B SG quality A SG pressure (Kg/cm2) B turbine bypass controlling pressure (Kg/cm2) A SG controlling (narrow range level (%) B SG controlling narrow range level (%) A SG turbine flow(t/hr) B SG turbine flow(t/hr) Externally inserted rod & boron reactivity ($) A SG feedwater flow (t/hr) B SG feedwater flow (t/hr) RC loop A flow (kt/hr) RC loop B flow (kt/hr) Nominal full power total feedwater flow (t/hr) Core water level from bottom of the core (M) Accumulator nitrogen pressure (Kg/cm2) Containment pressure (Kg/cm2) [62] IC 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 Acronym TRB LWRB TFSB TFPK QN SUM WSTA WSTB TPCT VH2O VN2 FWERA0 FWERB0 PERA0 PERB0 ERL0 TimeStep LVAP1 PSET1 TKLV Value 50 28 789.3 789.3 1835 3669 326.5 90 38.23 0 0 -2 158.2 3000 Description Containment temperature (°C) Containment water level (M) Submerged fuel average temperature (°C) Peak fuel temperature (°C) Normalized neutron flux) Integration term in kinetics equation A SG steam flow (t/hr) B SG steam flow (t/hr) Peak clad temperature (°C) Total CFT (accumulator) water volume Total CFT nitrogen volume (M3) Feedwater control valve A error Feedwater control valve B error SG A pressure controller error SG B pressure controller error Pressurizer level controller error Number of time steps Break elevation (dimensionless) SG header pressure set point, kg/cm2 RWST water inventory, ton [63] Appendix E: Operation characteristics of a typical PWR Characteristic Thermal power output System pressure Fuel enrichment Coolant flow Inlet temperature Outlet temperature Maximum fuel temperature Average linear heat rate Maximum linear heat rate Average heat flux Maximum heat flux Minimum Departure from nucleate boiling Active height Equivalent Active diameter Height to diameter ratio Active core volume Fuel weight Specific power Burnup Conversion ratio Number of fuel assemblies Fuel element array Assembly dimensions Number of fuel rods per assembly Total number of fuel rods Fuel element pitch Fuel element outer diameter Pitch to diameter ratio Cladding thickness Fuel pellet diameter Pellet to clad gap size Value 3,800MWth 2,250psia 1.9/2.4/2.9 1.59×108 lbs/hr 5650F 622.40F 3,4200F 3.54KW/ft 12.51KW/ft 206,000 BTU/(hr.ft2) 550,000 BTU/(hr.ft2) 1.3 150 inches 142.9 inches 1.05 1413 ft3 2,690 KW/ft3 36.9 KW/kg U 33,000 MWdays/MTU 0.5 241 16x16 8in x 8in 236 56,876 0.504 in 0.382 in 1.33 0.025 in 0.325 in 0.0035 in [64] Appendix F: Standard PWR nuclear fuel assembly (17×17) Geometry Square 17×17 matrix Fuel assembly dimension Square 214 x 214 mm Composition per assembly Total: 289 Fuel: 264 Control rod guide thimble: 24 Instrumentation thimble: Fuel material UO2 (U235,U238,Oxygen) Cladding material Zircaloy-4 (98.23 weight % zirconium with 1.45% tin, 0.21% iron, 0.1% chromium, and 0.01% hafnium) Gap filler Helium gas Fuel average density 95 – 96% Theoretical Density (TD) UO2-TD = 10.96 g/cc Moderator (coolant) light water (H2O) average density 0.7295 gr/cc H/HM ratio (hydrogen to heavy metal ratio) 1.7 – 3.4 (depends on enrichment level) Enrichment 2.5 – Wt % U235 Fuel pellet diameter 8.19 mm Pellet-clad gap 0.082 mm Clad thickness 0.572 mm Outer diameter of fuel rods 9.5 mm Pitch (center-to-center) 12.54 mm P/D 1.32 [65] Appendix G: Recoverable energy from fission of some isotopes Isotope Thermal neutron Fission neutron U-233 190.0 - U-235 192.9 - Pu-239 198.5 - Pu-241 200.3 - Th-232 - 184.2 U-234 - 188.9 U-236 - 191.4 U-238 - 193.9 Np-237 - 193.6 Pu-238 - 196.9 Pu-240 - 196.9 Pu-242 - 200.0 [66] Appendix H: Table of PWR NPPs (nuclear power plants) fission product inventory Isotope Core Inventory (Ci) Fraction in Gap Gap Inventory (Ci) I-131 8.0E7 0.10 8.0E6 I-132 1.2E8 0.10 1.2E7 I-133 1.7E8 0.10 1.7E7 I-134 1.8E8 0.10 1.8E7 I-135 1.5E8 0.10 1.5E7 Kr-83M 9.9E6 0.10 9.9E5 Kr-85M 2.2E7 0.10 2.2E6 Kr-85 5.2E5 0.30 1.6E5 Kr-87 4.1E7 0.10 4.1E6 Kr-88 5.8E7 0.10 5.8E6 Kr-89 7.2E7 0.10 7.2E6 Xe-131M 5.6E5 0.10 5.6E4 Xe-133M 2.3E7 0.10 2.3E6 Xe-133 1.6E8 0.10 1.6E7 Xe-135M 3.3E7 0.10 3.3E6 Xe-135 3.4E7 0.10 3.4E6 Xe-138 1.4E8 0.10 1.4E7 1) Based on a typical PWR plant data 2) I -127 and I-129 inventory are in Kg and have no contribution to the thyroid dose They are not included [67] Appendix K: Table of PWR NPPs design reactor coolant and SG secondary equilibrium activities Isotope RC Activity (Ci/gm) SG 2nd Activity (Ci/gm) I-131 2.3E+0 4.79E-3 I-132 2.8E+0 3.11E-3 I-133 3.7E+0 7.10E-3 I-134 5.9E-1 3.70E-4 I-135 2.1E+0 3.39E-3 Kr-83M 4.6E-1 2.01E-5 Kr-85M 2.0E+1 8.84E-5 Kr-85 7.7E+0 3.44E-4 Kr-87 1.3E+0 5.61E-5 Kr-88 3.7E+0 1.63E-4 Kr-89 1.1E-1 2.69E-6 Xe-131M 2.1E+0 9.37E-5 Xe-133M 1.7E+1 7.58E-4 Xe-133 2.6E+2 1.16E-2 Xe-135M 4.7E-1 1.80E-5 Xe-135 7.2E-1 3.20E-4 Xe-137 1.8E-1 4.78E-6 Xe-138 6.6E-1 2.49E-5 1) The RC activities are based on a typical PWR plant 2) The SG activities are based on a typical PWR plant 3) Xe-137 activity is presented in RC and SG but no value shown in core and gap inventory 4) Other isotopes, e.g Rb, Mo, Tc, Ru, Ag, Te and Tc are not included here because they not contribute to offsite dose calculation [68] Appendix L: Isotope data for dose calculation Isotope Half Life (hour) I-131 193.2s 0.381 1.49E6 I-132 2.3 2.333 1.43E4 I-133 21.0 0.608 2.69E5 I-134 0.9 2.529 3.73E4 I-135 6.7 1.635 5.60E4 Kr-83M 1.86 0.002 Kr-85M 4.48 0.159 Kr-85 93995 0.002 Kr-87 1.27 0.793 Kr-88 2.8 1.95 Xe-131M 285.6 0.02 Xe-133M 54.0 0.0416 Xe-133 127.0 0.0454 Xe-135M 0.3 0.432 Xe-135 9.15 0.247 Xe-138 0.3 1.183 Based on a typical PWR plant Average Gamma (Mev) Thyroid DCF (Rem/Ci) [69] Appendix M: The decay heat table OF 235U fission Time (s) 1.00E-01 1.50E-01 2.00E-01 3.00E-01 4.00E-01 5.00E-01 6.00E-01 8.00E-01 1.00E+00 1.50E+00 2.00E+00 3.00E+00 4.00E+00 5.00E+00 6.00E+00 8.00E+00 1.00E+01 1.50E+01 2.00E+01 3.00E+01 4.00E+01 5.00E+01 6.00E+01 8.00E+01 1.00E+02 1.50E+02 2.00E+02 3.00E+02 4.00E+02 5.00E+02 6.00E+02 8.00E+02 1.00E+03 1.50E+03 2.00E+03 3.00E+03 4.00E+03 5.00E+03 6.00E+03 8.00E+03 1.00E+04 1.50E+04 2.00E+04 3.00E+04 Beta (MeV/s/fission) 7.69E-01 7.04E-01 6.76E-01 6.51E-01 6.06E-01 5.68E-01 5.35E-01 5.06E-01 4.58E-01 4.19E-01 3.47E-01 2.97E-01 2.31E-01 1.88E-01 1.58E-01 1.35E-01 1.03E-01 8.17E-02 5.22E-02 3.73E-02 2.32E-02 1.67E-02 1.30E-02 1.06E-02 7.66E-03 5.85E-03 3.51E-03 2.43E-03 1.49E-03 1.08E-03 8.51E-04 7.06E-04 5.29E-04 4.23E-04 2.78E-04 2.01E-04 1.20E-04 8.06E-05 5.88E-05 4.57E-05 3.13E-05 2.39E-05 1.51E-05 1.10E-05 7.03E-06 Uncertainty (%) 6.9 6.4 6.1 5.9 5.5 5.3 4.8 4.6 4.3 4.1 3.9 3.9 4 4 3.9 4.1 3.6 3.7 2.3 2.1 2.2 2.4 2.5 2.8 2.9 3.1 3.1 2.8 2.4 2.3 2.2 2.3 2.4 2.5 2.5 2.4 2.3 Gamma (MeV/s/fission) 6.01E-01 5.50E-01 5.28E-01 5.08E-01 4.73E-01 4.42E-01 4.16E-01 3.93E-01 3.54E-01 3.22E-01 2.65E-01 2.25E-01 1.72E-01 1.39E-01 1.17E-01 9.99E-02 7.73E-02 6.28E-02 4.26E-02 3.23E-02 2.18E-02 1.65E-02 1.33E-02 1.12E-02 8.35E-03 6.56E-03 4.08E-03 2.86E-03 1.73E-03 1.23E-03 9.71E-04 8.08E-04 6.13E-04 4.98E-04 3.41E-04 2.57E-04 1.66E-04 1.19E-04 9.11E-05 7.25E-05 4.97E-05 3.64E-05 1.95E-05 1.22E-05 6.57E-06 Uncertainty (%) 9.9 9.4 9.1 8.8 8.5 8.4 8.3 8.2 8.1 8.1 7.9 7.8 7.3 6.8 6.3 5.8 5.2 4.9 4.4 4.1 3.5 3.1 2.8 2.7 2.5 1.8 1.9 2 1.9 1.7 1.6 1.4 1.4 1.4 1.5 1.8 2.1 2.2 2.3 2.3 2.2 1.9 1.6 Total (MeV/s/fission) 1.37E+00 1.25E+00 1.20E+00 1.16E+00 1.08E+00 1.01E+00 9.51E-01 8.99E-01 8.12E-01 7.41E-01 6.12E-01 5.22E-01 4.04E-01 3.28E-01 2.74E-01 2.35E-01 1.80E-01 1.45E-01 9.48E-02 6.96E-02 4.50E-02 3.32E-02 2.64E-02 2.18E-02 1.60E-02 1.24E-02 7.59E-03 5.29E-03 3.21E-03 2.31E-03 1.82E-03 1.51E-03 1.14E-03 9.21E-04 6.20E-04 4.58E-04 2.87E-04 2.00E-04 1.50E-04 1.18E-04 8.11E-05 6.03E-05 3.46E-05 2.32E-05 1.36E-05 Uncertainty (%) 3.7 3.2 3.1 3.1 2.9 2.9 2.9 2.8 2.8 2.8 2.8 2.7 2.6 2.5 2.5 2.4 2.3 2.3 2.2 2.1 2.1 2 1.9 1.7 1.5 1.3 1.3 1.4 1.5 1.5 1.6 1.6 1.6 1.5 1.5 1.5 1.6 1.6 1.7 1.7 1.6 1.5 1.4 [70] Time (s) 4.00E+04 5.00E+04 6.00E+04 8.00E+04 1.00E+05 1.50E+05 2.00E+05 3.00E+05 4.00E+05 5.00E+05 6.00E+05 8.00E+05 1.00E+06 1.50E+06 2.00E+06 3.00E+06 4.00E+06 5.00E+06 6.00E+06 8.00E+06 1.00E+07 1.50E+07 2.00E+07 3.00E+07 4.00E+07 5.00E+07 6.00E+07 8.00E+07 1.00E+08 1.50E+08 2.00E+08 3.00E+08 4.00E+08 5.00E+08 6.00E+08 8.00E+08 1.00E+09 1.50E+09 2.00E+09 3.00E+09 4.00E+09 5.00E+09 6.00E+09 8.00E+09 1.00E+10 1.50E+10 2.00E+10 Beta (MeV/s/fission) 4.97E-06 3.69E-06 2.83E-06 1.81E-06 1.26E-06 6.53E-07 4.10E-07 2.22E-07 1.53E-07 1.19E-07 9.87E-08 7.42E-08 5.96E-08 4.00E-08 2.99E-08 1.93E-08 1.38E-08 1.05E-08 8.44E-09 6.03E-09 4.66E-09 2.86E-09 1.96E-09 1.16E-09 8.16E-10 6.14E-10 4.76E-10 2.99E-10 1.98E-10 9.32E-11 6.44E-11 5.09E-11 4.61E-11 4.24E-11 3.91E-11 3.35E-11 2.87E-11 1.95E-11 1.33E-11 6.21E-12 2.90E-12 1.36E-12 6.36E-13 1.41E-13 3.26E-14 1.91E-15 8.80E-16 Uncertainly (%) 2.3 2.2 2.3 2.5 2.6 2.9 3.4 3.4 3.4 3.2 2.7 2.5 2.2 2.1 1.7 1.4 0.8 0.7 0.7 0.8 0.9 1.1 1.2 1.2 1.2 1.2 1.2 1.4 1.8 1.9 1.8 1.7 1.5 1.2 1.1 1.1 1.1 1.1 1.1 1.1 1.1 1.2 1.2 1.2 1.5 Gamma (MeV/s/fission) 4.44E-06 3.34E-06 2.65E-06 1.84E-06 1.38E-06 8.05E-07 5.52E-07 3.39E-07 2.50E-07 2.02E-07 1.70E-07 1.28E-07 1.01E-07 6.39E-08 4.53E-08 2.67E-08 1.76E-08 1.26E-08 9.63E-09 6.59E-09 5.02E-09 2.83E-09 1.60E-09 5.18E-10 1.87E-10 8.61E-11 5.36E-11 3.61E-11 3.05E-11 2.49E-11 2.28E-11 2.07E-11 1.92E-11 1.78E-11 1.65E-11 1.43E-11 1.23E-11 8.57E-12 5.95E-12 2.88E-12 1.39E-12 6.71E-13 3.24E-13 7.57E-14 1.79E-14 7.42E-16 2.92E-16 Uncertainly (%) 1.5 1.4 1.3 1.3 1.3 1.4 1.6 1.8 1.7 1.7 1.5 1.2 0.8 0.7 0.8 0.9 1.1 1.3 1.7 1.9 2.1 2.2 2.1 1.9 1.6 1.4 1.2 1.1 0.8 0.8 0.8 0.8 0.9 0.9 0.8 0.8 0.8 0.8 0.8 0.8 0.8 0.8 0.8 0.8 0.9 1.9 Total (MeV/s/fission) 9.41E-06 7.02E-06 5.48E-06 3.64E-06 2.64E-06 1.46E-06 9.61E-07 5.61E-07 4.04E-07 3.21E-07 2.69E-07 2.02E-07 1.61E-07 1.04E-07 7.52E-08 4.60E-08 3.14E-08 2.31E-08 1.81E-08 1.26E-08 9.68E-09 5.68E-09 3.57E-09 1.68E-09 1.00E-09 7.00E-10 5.29E-10 3.35E-10 2.28E-10 1.18E-10 8.72E-11 7.16E-11 6.52E-11 6.02E-11 5.56E-11 4.77E-11 4.10E-11 2.81E-11 1.93E-11 9.09E-12 4.29E-12 2.03E-12 9.60E-13 2.17E-13 5.04E-14 2.65E-15 1.17E-15 Uncertainly (%) 1.4 1.3 1.3 1.4 1.4 1.5 1.6 1.7 1.7 1.6 1.5 1.3 1.1 0.9 0.9 0.8 0.8 0.8 0.9 1.1 1.1 1.1 1.1 1.1 1.1 1.2 1.3 1.4 1.3 1.2 1.1 0.9 0.8 0.8 0.8 0.8 0.8 0.8 0.8 0.8 0.8 0.9 1.2 [71] Time Beta Uncertainly Gamma Uncertainly Total Uncertainly (s) (MeV/s/fission) (%) (MeV/s/fission) (%) (MeV/s/fission) (%) 3.00E+10 4.00E+10 5.00E+10 6.00E+10 8.00E+10 1.00E+11 1.50E+11 2.00E+11 3.00E+11 4.00E+11 5.00E+11 6.00E+11 8.00E+11 1.00E+12 1.50E+12 2.00E+12 3.00E+12 4.00E+12 5.00E+12 6.00E+12 8.00E+12 1.00E+13 7.23E-16 7.10E-16 7.08E-16 7.07E-16 7.06E-16 7.04E-16 7.00E-16 6.96E-16 6.89E-16 6.81E-16 6.74E-16 6.66E-16 6.52E-16 6.38E-16 6.05E-16 5.74E-16 5.17E-16 4.68E-16 4.24E-16 3.86E-16 3.21E-16 2.70E-16 1.7 1.7 1.7 1.7 1.7 1.7 1.7 1.7 1.7 1.7 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.5 1.5 1.5 1.5 1.5 2.79E-16 2.79E-16 2.78E-16 2.78E-16 2.76E-16 2.75E-16 2.72E-16 2.69E-16 2.63E-16 2.58E-16 2.52E-16 2.46E-16 2.36E-16 2.26E-16 2.02E-16 1.81E-16 1.45E-16 1.17E-16 9.37E-17 7.52E-17 4.85E-17 3.12E-17 2 2 2 2 2 2 2 2 2 2 2 1.00E-15 9.89E-16 9.86E-16 9.85E-16 9.82E-16 9.79E-16 9.72E-16 9.65E-16 9.52E-16 9.39E-16 9.26E-16 9.13E-16 8.88E-16 8.64E-16 8.07E-16 7.55E-16 6.63E-16 5.85E-16 5.18E-16 4.61E-16 3.70E-16 3.01E-16 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.3 1.2 1.2 1.2 1.2 1.1 1.1 1.1 2.4 0 44.3 13 18.1 92.4 51.8 38.3 24.5 22 428.7 158.7 12.5 15.6 0 291.5 Armenia Bangladesh Belarus Belgium Brazil Bulgaria Canada China China: Taiwan Czech Republic Egypt Finland France Germany Hungary India Indonesia Iran Israel Japan Kazakhstan Argentine NATION 30 0 2.6 38 32 78 28 31 20 1.9 16 44 3.3 54 0 42 NUCLEAR ELECTRICITY GENERATION 2006 Billion %e KWh 7.2 6.9 55 0 17 17 59 6 11 18 2 0 No 47577 0 3779 1826 20339 63473 2696 3472 4884 8587 12652 1906 1901 5728 0 376 935 MWe REACTORS OPERABLE Jan-08 0 0 1500 4540 2600 0 1600 1630 0 2976 915 2285 0 0 0 1 0 0 692 MWe 0 No REACTORS under CONSTRUCTION Jan-08 11 2 10 0 0 0 30 2 0 No 14945 1900 2000 8560 0 0 0 32000 4000 1900 1245 2000 0 740 MWe REACTOR PLANNED Jan-08 Appendix N: World nuclear power reactors 2006-08 and uranium requirements 2008 1 1 1 86 0 1 No 300 1100 1200 300 4800 2000 1600 1000 1000 1900 68000 2200 4000 0 2000 1000 740 MWe REACTOR PROPOSED Jan-08 7569 143 978 271 3332 10527 1051 619 832 1396 1665 261 303 1011 0 51 123 Tones U URANIUM REQUIRED In 2008 [72] 3.5 2.7 16 57 40 4.4 20 48 37 0 48 18 19 16 10.4 3.3 2.6 5.2 144.3 16.6 5.3 10.1 57.4 65.1 26.4 0 84.8 69.2 787.2 2658 Mexico Netherlands Pakistan Romania Russia Slovakia Slovenia South Africa Spain Sweden Switzerland Thailand Turkey Ukraine United Kingdom USA Vietnam WORLD 4.9 39 69 141.2 Korea RO (South) Lithuania 0 Korea DBA (North) NATION NUCLEAR ELECTRICITY GENERATION 2006 Billion %e KWh 439 104 19 15 0 10 31 2 20 No 372059 99049 11035 13168 0 3220 9086 7442 1842 696 2064 21743 1310 400 485 1310 17533 1185 MWe REACTORS OPERABLE Jan-08 34 0 0 0 0 0 0 0 No 27789 0 0 0 0 0 840 4920 300 0 3000 0 MWe REACTORS under CONSTRUCTION Jan-08 93 0 0 0 2 0 No 100595 10180 1900 0 0 165 0 9600 1310 600 0 6600 950 MWe REACTOR PLANNED Jan-08 222 25 20 0 24 20 2 No 193095 2000 32000 27000 4500 4000 1000 0 4000 1000 18200 655 2000 2000 3200 MWe REACTOR PROPOSED Jan-08 64615 18918 2199 1974 0 537 1418 1398 303 141 313 3365 174 65 98 246 3109 225 Tones U URANIUM REQUIRED In 2008 [73] [74] Appendix O: The International Nuclear and Radiological Event Scale (INES) GENERAL DESCRIPTION OF INES LEVELS INES Level Major Accident Level Serious Accident Level People and Environment Radiological Barriers and Control Major release of radio active material with widespread health and environmental effects requiring implementation of planned and extended countermeasures Significant release of radioactive material likely to require implementation of planned countermeasures - Limited release of radioactive material Accident with likely to require Wider implementation of some Consequences planned countermeasures Level - Several deaths from radiation - Minor release of radioactive material Accident with unlikely to result in Local implementation of Consequences planned countermeasures other Level than local food controls - At least one death from radiation - Severe damage to reactor core - Release of large quantities of radioactive material within an installation with a high probability of significant public exposure This could arise from a major criticality accident or fire - Fuel melt or damage to fuel resulting in more than 0.1% release of core inventory - Release of significant quantities of radioactive material within an installation with a high probability of significant public exposure Defence-in-Depth [75] INES Level Serious Incident Level Incident Level People and Environment - Exposure in excess of ten times the statutory annual limit for workers - Non-lethal deterministic health effect (e.g., burns) from radiation - Exposure of a member of the public in excess of 10 mSv - Exposure of a worker in excess of the statutory annual limits Radiological Barriers and Control - Exposure rates of more than Sv/h in an operating area - Severe contamination in an area not expected by design, with a low probability of significant public exposure - Radiation levels in an operating area of more than 50 mSv/h - Significant contamination within the facility into an area not expected by design Defence-in-Depth - Near accident at a nuclear power plant with no safety provisions remaining - Lost or stolen highly radioactive sealed source - Misdelivered highly radioactive sealed source without adequate procedures in place to handle it - Significant failures in safety provisions but with no actual consequences - Found highly radioactive sealed orphan source, device or transport package with safety provisions intact - Inadequate packaging of a highly radioactive sealed source - Overexposure of a member of the public in excess of statutory annual limits Anomaly Level - Minor problems with safety components with significant defence-indepth remaining - Low activity lost or stolen radioactive source, device or transport package NO SAFETY SIGNIFICANCE (BELOW SCALE/LEVEL 0) ... 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 mơ cố lò phản ứng nước áp lực hai vòng phần mềm PCTRAN”... 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... The containment prays use to quench any steam released in the accidents [23] Chapter INTRODUCTION TO PCTRAN PWR SOFTWARE VERSION 4.0.8 2.1 PCTRAN overview [28] PCTRAN (Personal Computer Transient