design & fea of reactor pressure vessel
TRANSCRIPT
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Design & Finite Element Analysis of
Reactor (PWRs) Pressure Vessels
Using ASME Section III and BS 5500
Undergraduate Final Year Project
Name: Jawad Bari
Roll No: @00265031
Course Title: BEng Mechanical Engineering
Course Code: E/ME/F3
Supervisor: Dr P. Hampson
UNIVERSITY OF SALFORD
SCHOOL OF COMPUTING, SCIENCE AND ENGINEERING
DEPARTMENT OF AERONAUTICAL, MECHANICAL AND
MANUFACTURING ENGINEERING
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Declaration
I, Jawad Bari, declare that the contents of this report have been studied, evaluated and
written by myself. Any section, part or phrasing of more than twenty consecutive words
that is copied from any other work or publication has been clearly referenced at the
point of use and also fully described in the reference section of this dissertation.
I have read, understood and agree to the University Policy on the Conduct of Assessed
Work (Academic Misconduct Procedure).
Signed
Dated
.
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Abstract
A reactor pressure vessel can be designed and tested using international design codes;
these design codes provide engineers with preventative measures, which can be used in
order to avoid any catastrophic accidents. In this project, these preventative measures
are investigated and effectiveness of the elastic-plastic limit based design
methodologies of the main International Design Codes (ASME and BS) using the
ANSYS finite element program is accessed. A typical RPV of a 300MW pressurized
water reactor (PWR) was selected for the analysis. A nuclear grade steel SA-508 Gr.3
Cl.1 was used as a material of the RPV for the comparison. It has been concluded that
the application of the design by analysis allows removing the unnecessary conservatism
caused by applying the design by rule approach described in BS-5500 section 3. This
study recommends that the maximum allowable pressure of the RPV may be increased
up to 26.37 % by using design by analysis approach as described in ASME code.
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ContentsDeclaration .................................................................................................................................... 2
1. Introduction .......................................................................................................................... 8
1.1. Overview of pressure vessels ........................................................................................ 8
1.2. Pressure Vessel Codes and Standards ......................................................................... 12
1.2.1. Codes: .................................................................................................................. 12
1.2.2. Standards ............................................................................................................ 12
1.2.3. Development of pressure vessels design codes .................................................. 12
1.3. Finite Element Analysis ............................................................................................... 13
1.4. Problem statement .......................................................................................................... 14
2. Objective ............................................................................................................................. 15
2.1 Aims: .................................................................................................................................. 15
3. Literature Review ................................................................................................................ 16
3.1. Overview ..................................................................................................................... 16
3.1.1. Pressurised Water Reactor (PWR) ...................................................................... 17
3.2. Design Basis: Codes and Regulations for Reactor Pressure Vessel ............................. 18
3.2.1. Design Process..................................................................................................... 18
3.2.2. ASME Design Codes ............................................................................................. 21
3.2.3. BSI design codes .................................................................................................. 22
3.3. Materials ..................................................................................................................... 23
3.3.1. Overview.............................................................................................................. 23
3.3.2. Cladding material................................................................................................ 25
3.3.3. Properties of some noticeable reactor vessel materials ..................................... 26
3.4. Designing and Manufacturing Techniques .................................................................. 29
3.4.1. CAD Packages ...................................................................................................... 29
3.4.2. CATIA V5 .............................................................................................................. 30
3.4.3. FEA Simulation Packages ..................................................................................... 30
3.4.4. Fabrication .......................................................................................................... 32
3.5. Inspections .................................................................................................................. 35
3.5.1. Irradiation Embrittlement ................................................................................... 36
3.5.2. Corrosion ............................................................................................................. 38
3.5.3. Fracture toughness ............................................................................................. 38
3.5.4. Crack .................................................................................................................... 39
3.5.5. Creep/Stress Rupture .......................................................................................... 40
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4. Methodology ....................................................................................................................... 42
4.1. Project Plan................................................................................................................. 42
4.1.1. Semester 1........................................................................................................... 42
4.1.2. Semester 2........................................................................................................... 43
4.2. Reactor Pressure Vessel Design .................................................................................. 44
4.2.1. RPV Design Concepts........................................................................................... 44
4.2.2. Three dimensional modelling .............................................................................. 45
4.3. Material ....................................................................................................................... 47
4.4. Design Code Parameters ............................................................................................. 48
4.4.1. ASME approach ................................................................................................... 48
4.4.2. BS Approach ........................................................................................................ 50
4.5. Finite Element Method ............................................................................................... 51
4.5.1. Finite Element Modelling .................................................................................... 51
4.5.2. ANSYS 14.5 Static Structural Workbench ............................................................ 52
4.6. Results ......................................................................................................................... 54
5. Discussion ............................................................................................................................ 56
5.1. Simulation Results ....................................................................................................... 56
5.2. Design Code Comparison ............................................................................................ 58
6. Conclusion ........................................................................................................................... 61
7. Further Work ....................................................................................................................... 62
8. References ........................................................................................................................... 63
9. Appendices: ......................................................................................................................... 67
9.1. Appendix A - Introduction: .......................................................................................... 67
9.2. Appendix B - Literature Review ................................................................................... 69
9.3. Appendix CProject plan ........................................................................................... 74
9.4. Appendix DDesign & Design codes .......................................................................... 76
9.5. Appendix ESimulation ............................................................................................. 87
9.6. Appendix FResults ................................................................................................... 90
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Table of Figures
Figure 1-1 Spherical Pressure Vessels ........................................................................................... 9
Figure 1-2: Pharmaceutical Pressure Vessels ............................................................................. 10
Figure 3-1: Irradiation Embrittlement ......................................................................................... 36
Figure 3-2: Fracture Failure ......................................................................................................... 40
Figure 9-1: Schematic diagram of primary circuit of typical PWR Nuclear Power Plant ............ 67
Figure 9-2 The PWR Nuclear Power Plants schematic diagram .................................................. 67
Figure
9-3: Cutaway view of a Westinghouse Pressurized Water Reactor (PWR) ....................... 68
Figure 9-4: Typical structure of PVC committees ........................................................................ 69
Figure 9-5: Fabrication configuration of PWR beltline shells ...................................................... 72
Figure 9-6: Rolled and welded beltline shell .............................................................................. 72
Figure 9-7 : Schematic diagram of laser welding ........................................................................ 73
Figure 9-8: Gantt chart for Semester 1 ....................................................................................... 74Figure 9-9: Gantt chart for Semester 2 ....................................................................................... 74
Figure 9-10: Brain Storming ........................................................................................................ 75
Figure 9-11: Westinghouse Two Loop Pressure Vessel .............................................................. 76
Figure 9-12: Design comparison of Westinghouse Reactor Pressure Vessels ............................ 77
Figure 9-13: Reactor Pressure Vessel Sketch .............................................................................. 78
Figure 9-14 : Outlet nozzle .......................................................................................................... 78
Figure 9-15: Inlet Nozzle ............................................................................................................. 79
Figure 9-16: CATIA V5 part retendering ...................................................................................... 79
Figure 9-17: Figure (a) and (b) showing the two halves of the RPV ............................................ 79
Figure 9-18: Two parts getting Aligned ....................................................................................... 80
Figure 9-19: RPV assembled ........................................................................................................ 80
Figure 9-20: Top view of the Pressure vessel .............................................................................. 81
Figure 9-21: Generating CATIA part from Assembly ................................................................... 81
Figure 9-22: 2D Detailed Drawing ............................................................................................... 82
Figure 9-23; Section A-A2, Pressure vessel 2D Drawing ............................................................. 83
Figure 9-24: Details of the part A and B, Inlet and outlet Nozzles .............................................. 84
Figure 9-25: Stress categories and limits form BS-5500 ............................................................. 85
Figure 9-26: geometries covered by the BS-5500 design by rule route ..................................... 86
Figure 9-27: Static Structure Workbench .................................................................................... 87
Figure 9-28: Engineering Data Dialogue box, Material Properties ............................................. 87
Figure 9-29: Geometry Module................................................................................................... 88
Figure 9-30: Model module ......................................................................................................... 88
Figure 9-31: solid 186, 3D higher hexahedral brick meshing ...................................................... 89
Figure 9-32: Setup, Solution and Results dialogue box ............................................................... 89
Figure 9-33: Total Deformation ................................................................................................... 90
Figure 9-34: Maximum Principal Stress ...................................................................................... 90
Figure 9-35: Middle Principal Stress ........................................................................................... 91
Figure 9-36: Peak stresses in RPV ............................................................................................... 91
Figure 9-37: Von-Mises Stress ..................................................................................................... 92Figure 9-38: Hoop Stress ............................................................................................................. 92
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Figure 9-39: Beltline Region of RPV Geometry, (b) Solid-189 3D Hexahedral Meshing ............. 93
Figure 9-40: (a) Hoop stress at 40Mpa (b) Analysis conducted on the beltline region .............. 93
List of Tables
Table 4.1: Chemical Composition of SA508 Steel ....................................................................... 47
Table 4.2: Thermal and Mechanical properties for the base material ....................................... 48
Table 9.1: Main Ferrous Materials for reactor components in Western Countries ................... 70
Table 9.2: Materials Specified For PWR Vessel Components ..................................................... 70
Table 9.3: Summary of design factors and the materials for UK codes ...................................... 71
Table 9.4: Providing the Hoop, Radial and Axial Stresses on the beltline region of RPV ............ 94
List of Graphs
Graph 9-1: Design Pressure VS Radial Displacement Graph ...................................................... 95
Graph 9-2: Design Pressure VS Radial Displacement Graph the collapse limit line.................... 96
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1. Introduction
1.1. Overview of pressure vessels
By definition, any closed container that contains fluids or gasses at a different pressure
than ambient pressure classifies as a pressure vessel. The idea of pressure vessels
emerged in 1495 by Leonardo da Vinci in Codex Madrid I (Carter, 2001). Leonardo
wrote we shall describe how air can be forced under water to lift very heavy weights,
that is, how to fill skins with air once they are secured to weights at the bottom of the
water. And there will be descriptions of how to lift weights by tying them to submerged
ships full of sand and how to remove the sand from the ships. Leonards pressurized
bags of air led the first design of pressure vessels. This idea was instrumental in the 18th
and 19th centuries for designing steam engines and boilers. Boilers, tanks and the
pipelines that carry, store, or receive fluids are also categorised as pressure vessels.
The fluid inside the vessel may undergo a change in state as is the case of steam boilers
or may combine with other reagent as in the case of chemical reactor. Pressure vessels
often have a combination with other reagents as is the case with chemical reactor.
Pressure vessels often have combinations of high pressures coupled together with high
temperatures and in some cases flammable fluid or highly reactive material. Nuclear
reactor pressure vessels are the prime example of such type. Because of such hazards it
is necessary to design them to prevent leakages. In addition vessels have to be designed
in order to cope with high operating temperatures and pressures.
Theoretically, pressure vessels can be of any shapes and sizes. The size and the
geometric form of pressure vary greatly from the large cylindrical vessels used for high
pressure gas storage to the small sized vessels commonly found in hydraulic units for
aircrafts. Practically the pressure vessels are usually found to be either mostly spherical
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Food processing
Chemical
- Pharmaceuticals
- Petrochemicals
Construction
Defence
Energy
In the energy sector, pressure vessels are usually used in nuclear power plant,
specifically in nuclear reactors to initiate and sustain the nuclear chain reaction and
more importantly to contain coolant for use in nuclear reactors. In most modern day
nuclear reactors pressure vessels hold the reactor core which endures the fuel cells of
the power plants, pressuriser and the steam generator are the integrated part of the
reactor. From the start of first designing of the Nuclear reactor there have been a
considerable reactors concepts proposed. A selected number of these have been built.
Today only three of these concepts are considered commercially viable. Two of these
concepts are based on the use of the uranium U-235 with light water employed for
cooling and neutron moderation. Of these two concepts, one is the pressurised water
reactor (PWR) developed by Westinghouse. The other is the boiling water reactor
(BWR) developed by the General Electric. The third concept is based on the use of
neutral uranium with heavy water for cooling and moderation, this plant is regulated by
the Atomic Energy of Canada (Westinghouse Electric Corporation, 1984).
The fundamental distinction between the PWR and the BWR is that in BWR the latter
the coolant moderator is allowed to boil the resulting steam passed directly to the
turbine-generating, whereas in the PWR the coolant moderator is maintained above
saturation pressure such that no significant boiling occurs in the reactor.
Figure 1-2: Pharmaceutical Pressure Vessels
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PWRs nuclear reactors have their mechanisms divided into two main circuitsprimary
and secondary circuit. Primary circuit contains the main core (fuel cell assembly) and
the primary coolant of the primary system and perhaps, classifies as one of the most
important part of PWRs nuclear plants. In particular the primary circuit consists of
pressuriser, steam generator and the core reactor vessel. In most PWR designs the
primary circuit is placed inside the pre-stressed concrete containment. As shown in the
figure 9-1 in appendix A.
The secondary circuit of PWRs include main steam system and the condensate feed
water system, turbine and the engine. The fresh water is pumped in from the feed-water
system and the passes through the steam generator, heated and converts into steam. The
steam then passes through the main steam line to the turbines where it spins the
turbines. Steam from the turbines then pass through a condenser to a condensate pumps
through low-pressure feed water heaters, then to high pressure feed-water heaters and
finally goes back to steam generator. Secondary circuit is shown in figure 9-2 in
appendix A.
The primary circuit comes in direct contact with the main core of the nuclear plant, and
is subject to high coolant water pressure of 2250 Psi (approx. 155bars) and reaches the
temperature of 600oF (315.5oC). Water heats up and circulates through the heat
exchanger by the help of the pump. The main focus of this exercise is on the primary
circuit of the PWR reactor and particularly on the core reactor vessel. The figure 9-3,
appendix A, is showing the typical PWRs type reactor vessel of a nuclear power plant.
The pressure vessel design shown in figure 9-3 is based on the Westinghouse double
loop pressure vessel. The Westinghouse reactor vessel was designed with accordance to
ASME pressure vessel design codes and standards.
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1.2. Pressure Vessel Codes and Standards
1.2.1. Codes:
The word code is defined as a group of administrative and technical rules and standards
covering any combination of materials, design, construction, installation and inspection
of equipment which is adopt into law of the legal jurisdiction. In other words, codes are
merely systematic statement of standards or regulations that are legally enforced.
1.2.2. Standards
Standards are technical specification introduced to public by the corporations or general
approval of all interested affected by it on the basis of science, technology and
experience for the promotion of optimum community benefits. The standards are
usually recognised by the national and international institutes.
1.2.3. Development of pressure vessels design codes
The Boiler and Pressure vessel code establishes rules of safety design, fabrication and
inspection of boilers and pressure vessels. During 18 thand 19th centuries, steam became
the chief source of locomotive energy, which flourished in the engineering industry. By
the early 20th, numerous accidents and explosion took place which led the development
of standard design codes for the pressure vessels (Carter, 2001). American society of
Mechanical Engineers was the first one to enact the first code for the construction of
steam boiler in 1907 (Hong, 2010). In 1924, ASME and American Society for Testing
and Materials (ASTM) proposed the material specification for boiler and the pressure
vessels. In 1925, ASME published section VIII as a guidelines for unfired welded
pressure vessels in this publication the safety factor of 4 was proposed
(Chattopadhyay, 2005). Later in 1968, this section was divided into two divisionsDiv.
1 & 2. By the end of 1960s it was become necessary to issue a design code for the
nuclear pressure vessels due to the increase numbers of nuclear power plants; hence
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ASME produced the guidelines for commercial nuclear vessels under division 1 section
III. This was a design by analysis code with a theoretical safety factor of 3. It should be
noted that division 1 is specific to metallic components and division 2 provides the rules
for the construction concrete nuclear vessel (R.W.Nichols, 1987).
Similar to ASME, Great Britain has its own institute which acts as a regulatory
authority called British Standard Institute (BSI). It is the recognised body in the UK for
the preparation and issue of national standards in the fields of engineering. The main BS
code for the design of pressure vessel is BS 5500. The first edition of BS 5500 was
issued in 1976. In 1984, few distinctive features such as three categories of construction
were introduced. In 1994, BS 5500 attained the status of defacto international standards
(Morris, 2010) .
In May 2002, the first issue of European Standard EN 13445 unfired pressure vessels
was published. This standard was developed to facilitate pressure vessels subjected to
the European pressure equipment directive 97.23 commonly known as PED. Under
CEN BSI was obliged to with withdraw BS 5500 as it has the similar content, validity
and application. But later it was decided that British standard for pressure vessel should
continue to available and become a published document (PD) under the new reference
PD 5500, however in this report it is still referred as BS 5500.
1.3.
Finite Element Analysis
FEA is based on the discretization technique which engulfs the basic concept of
dividing the mathematical model into non-overlapping components of simple geometry
called finite elements. ANSYS uses a similar method. It uses a complex system to
create finite element by dividing the geometry into elements and nodes. This process is
called meshing. The mesh is programmed to takes in account the material and the
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structural properties of the model; this will determine the response of the structure to
certain loading conditions.
The rapid development in this technique has remarkably impacted the design of
pressure vessels. This technique has made the designing process extremely quick and
has led extremely difficult design getting assessed quicker and more accurate. There are
a number of FEA software packages available such as ADINA, ANSYS, SolidWorks
and PV Ellite (particular for Pressure vessel designs accordance with ASME codes). In
this project ANSYS is used for analysis of pressure vessels. The details of the software
are provided in literature and methodology sections.
1.4. Problem statement
In designing the reactor pressure vessels (RPVs) safety is the primary consideration due
to the potential effect of the radiation leakage from the core reactor. However safety
cannot be guaranteed for two reasons. Firstly, the actual form of loading during the
service can be different than what was previously anticipated during the design stages.
Secondly our knowledge is often adequate to provider answers to fracture of materials,
state of stress under certain conditions as the fundamental mechanism of failure is not
sufficiently understood (Chattopadhyay, 2005). There are number of governing bodies
around the globe such as ASME and BS, who have established preventative measures
based on semi-empirical methods. These preventative measures will be studied and
analysed in this project.
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2.Objective
The purpose of this project is to design a reactor pressure vessel RPV and perform
Finite Element Analysis in accordance to ASME and BSI standards respectively, thus
identifying the differences between ASME and BS design rules.
2.1 Aims:
The aims of this project are to:
Identify the right ASME and BS design codes for the reactor pressure vessel
Identify the materials for reactor pressure vessel (RPV)
Analyse the safety parameters for allowable working pressure using ANSYS
software package
Perform Finite Element Analysis to evaluate the differences between the ASME
and the BS codes
Study the corrosion and the radioactive embrittlement effect on the pressure
vessels
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3.Literature Review
3.1.
Overview
Nuclear power has emerged as a proven technology, since the demonstration of a
sustained reactor in 1942 and it has become the one of the fast growing method of
producing the electricity in the world. At present, there are over four hundred
operational nuclear power plants around the world, of which over 75 percent are of the
light water design with over 65 percent of the light water design furnished by
Westinghouse and its current or original licenses (Westinghouse Electric Corporation,
1984). Of the nuclear power plants in operation, the most common type is the
Pressurised Water Reactors (PWR) and the second most is the Boiling Water Reactor
(BWR). Both plant works almost on the same principle.
A Westinghouse designed RPV that is shown in figure 9.3 is fairly a typical vessel
designed used in all so called western designed RPVs. However, there are significant
differences in size, nozzle designs, penetration designs and other details among the
various suppliers. The first PWR in the WEST was the Yankee-Rowe plant in Rowe,
Massachusetts, USA. The first reactor pressure vessel, Yankee-Rowes vessel, weighed
210,000 kg and had inside diameter of 27.7 m. Depending on the design of the nuclear
steam supply system, two, three, four or six loops, the RPVs can weigh as much as
427,000 kg and have an inside diameter of 44 m.
According to (International Atomic Energy Agency , 1999) The PWR pressure vessel
is the most important pressure boundary component of the NPP because its function is
to contain the nuclear core under elevated pressures and temperature.
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(Sonaeda, 2015) emphasises on the importance of the pressure vessel stating that a
reactor vessel (RPV) is heart of the reactor core and it is also the major component that
may limit the useful life of the nuclear plant as in future due to any failure if the reactor
should replace extraordinary amount of time and money would be required. Virtually
every other component of the Nuclear power plant can be replaced cost-effectively
including steam generator except the reactor vessels.
RPV is a steel made container which comprehends nuclear fuels and bears the high-
temperature/ pressure coolant water, so the safety of this component of the nuclear
power plant should be absolutely vital. The safety goal is to avoid the catastrophic
failure of the RPVs and any consequent release of the radioactive materials to the
environment. To achieve this goal, a structural integrity evaluation based on fracture
mechanics is performed to ensure sufficient margin against failure during transients. For
the safety of the reactor vessels, reactor vessels have to pass the quality check and
follow the guidance given by the countrys design code issue authority. Most of western
NPPs reactor vessels are designed, constructed and tested by using the guidance
provided either by ASME or BS design codes.
3.1.1. Pressurised Water Reactor (PWR)
The design of reactor pressure vessel of the PWR type nuclear plant is different than the
BWR type nuclear plants. It has some 200 tube assemblies containing ceramic pellets
consisting of either enriched uranium dioxide (UO2) or a mixture of both uranium and
plutonium oxides known as MOX (mixed oxide fuel). These are encased in Zircaloy 4
cladding. Either B4C-Al2O3 pellets or borosilicate glass rods are used as burnable
poisons. Water pumped through the core at a pressure sufficient to prevent boiling. This
acts as both a coolant and a moderator to slow down the high energy neutrons. The
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water at about 600 K passes to an intermediate heat exchanger. The power is controlled
by the insertion of control rods from the top of the core and by dissolving boric acid into
the reactor water. As the reactivity of the fuel decreases, the concentration of dissolved
boron ions is reduced by passing the water through an ion-exchanger. Control rods
made of boron carbide (B4C) or an Ag-In-Cd alloy are clad in Inconel 627 or stainless
steel (304) tubes (Westinghouse Electric Corporation, 1984).
The primary pressurized water loop of a PWR carries heat from the reactor core to a
steam generator. The loop is under a working pressure of about 15 MPa which is
sufficient to allow the water in it to be heated to near 600 K without boiling. The heat is
transferred to a secondary loop generating steam at 560 K and about 7 MPa, which
generates heat that drives the turbine.
3.2. Design Basis: Codes and Regulations for Reactor Pressure
Vessel
3.2.1. Design Process
According to (Kendall, 1969) the design process involves a variety of different tasks.
He emphasises on the importance of understanding the basics of the pressure vessel
design process so that the design philosophy may be maintained throughout the life of
the plant.
According to the technical report by Thomas E. Davidson and the David Kendall
published in 1961 outline the pressure vessel operating at very high pressures as a
complex problem. As it involve many considerations including definition of the
operating and permissible stress levels, criteria of failure and material behaviour at high
pressures. For the purpose of developing the design philosophy and the relative
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operational limitations of various approaches, the elastic strength or yielding pressure of
the vessel will be used as the criterion of failure. Further in the report it is noted that the
excess pressure cab be employed to predicted the yielding of one or more components
This information then can be utilised to design, construct and inspect the pressure
vessels, and also to chart its current and future conditions. Generally, the use of vessels
beyond the yielding pressure will depend upon the amount of plastic strain permissible
and the ductility of the materials involved.
(Sonaeda, 2015) stresses on the CAD tools being used by the reactor vessel designer.
He argues that it should be recognized that the designer used the best available tools at
that time to perform the necessary analyses and calculations according to the relevant
ASME Code requirements. The key responsibilities of the designer are to:
(i) Specify an RPV material with initial properties known to have sufficient
reserve to accommodate time dependent degradation such as
embrittlement or fatigue
(ii) Make a reliable determination of the material state and quality of the
finished component
(iii)
Determine the design loading conditions
(iv) Estimate the amount of time dependent degradation that may be experienced
under service conditions for the license life of the RPV
(v)
Consider future surveillance of the RPV materials and flaw state as well as
monitoring of service conditions during plant operation
(vi) Assume worst case conditions with respect to loading, spectrum, and
material flaw state.
A reactor pressure vessel (RPV) is designed, manufactured and operated in such a
manner that it should not fail in service. To resist failure, its steel structure must remain
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ductile and not become brittle. A sudden failure of an RPV could be catastrophic even
before considering the contents of the pressure vessel, because of the sudden release of
a large amount of depressurising energy. Fracture resistance or fracture toughness is the
relevant material property in structural integrity assessment of the RPV (Ballesteros,
2014). During service, the integrity of the RPV depends on multiple factors:
The initial quality of the vessel as fabricated (for example, having a low
incidence of flaws)
Any degradation of mechanical properties with life (such as irradiation
embrittlement, thermal embrittlement, temper embrittlement, strain ageing)
The RPV's operating history, including the frequency and magnitude of the
pressure/temperature transients and the associated heat transfer and stress
distribution to which it has been exposed
Knowledge of the appropriate fatigue crack growth laws as a function of the
environment and materials structure
The frequency and effectiveness of the in-service inspections
(Harvey, 1991) states the sudden failure of pressure vessel could occur at the stress
concentration points such as the openings of pressure vessels or any area that causes the
geometric discontinuities. He defines the stress concentration effects on pressure vessels
that are somewhat like those from pricking a balloon for instance as a sharp point can
immediately rupture it. So it is not practical to design a reactor vessel without stress
concentrations as reactor vessels must have opening especially for coolant circulations.
This gives the geometric discontinuities, hence abnormal local stresses.
Reactor pressure vessels are complex geometries and essentially have openings, nozzles
and other attachments which produce geometric discontinuities. The effect of
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concentration of stresses due to geometric discontinuities is one of the basic
considerations in the design of pressure vessels. The elementary stress equations no
longer prevail in the vicinity of the geometric discontinuities. It is due to the fact that
geometric discontinuities significantly alter the stress distribution in the area. The
geometric discontinuity are called stress raisers (Hyder, 2015)
3.2.2.ASME Design Codes
Since its first issuance in 1914, ASMEs Boiler and Pressure Vessel code have
pioneered modern standards development, maintaining a commitment to enhance public
safety and technological advancement to meet the needs of a changing world. This
International Historic Mechanical Engineering Landmark now has been incorporated
into the laws of state and local jurisdictions of the United States and nine Canadian
provinces. ASME standards have been employed over 100 countries around the world,
with translations into a number of languages. The boiler and pressure vessel sections of
the BPVC have long been considered essential within such industries as electric power
generation, petrochemical, and transportation (ASME Boiler and Pressure Vessel Code,
Section III, Division , 2010)
According to (International Atomic Energy Agency , 1999) ASME has played a vital
role in supporting the nuclear industry. Since the first development of the ASME
standards and conformity assessment programs which were originally developed for
fossil fuel fired plants were later applied to nuclear power plant construction. Its widely-
adopted BPVC Section III, Rules for Construction of Nuclear Facility Components,
celebrated 50 years in 2013.
The ASME Section VIII and Section III design practices are used to limit the stresses in
the reactor pressure vessels and other components to acceptable levels. In all cases the
RPVs are designed to withstand the maximum pressure and temperature during normal
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operation or during accident conditions. Techniques such as thermal hydraulic analysis,
heat transfer analysis and stress analysis should be used to determine the pressure and
temperature as a function of a time for possible and transient events. During the
designing process the safety margins should be maintained all the time (R.W.Nichols,
1987).
Reactor pressure designers using ASME Section III also considers the stress intensity
and strain distribution in the RPV and nozzles under the assumption of a large reference
flaw (Miller, 1988). To better understand the background of RPV design, it is useful to
review the general design practices that were used by the major US vendors. For
example, a BWR designed by General Electric typically operates at 6.9 MPa and 272
C. Westinghouse, Babcock & Wilcox (B&W), and Combustion Engineering (C-E)
designed PWRs operate at a higher pressure, typically 15.5 MPa and temperatures
around 288 C. Some older Westinghouse-design PWRs have operated at
temperatures about 10 C lower for some time, and the B&W design PWRs operate
at RPV temperatures slightly higher than the Westinghouse-design PWRs.
3.2.3. BSI design codes
The British Standard Institution is the recognised body in the UK for the preparation
and issue of the national standards in the fields of engineering. According to
(R.W.Nichols, 1987) BSIs main function is to draw up voluntary standards by
agreement among all concerned, and to promote their application. This is achieved by
the through the Technical Committees made up from the members of British Standard
Institution. Societies such as Institution of Mechanical Engineers (ImechE), Royal
Aeronautical Society (RAS) etc. make up the BSI and brings up the range of national
interest.
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BSIs involvement in pressure vessels and the boilers is through the pressure vessels
standard committee (PVE). There are six main functional committees and within these
six committees there are 14 technical committees, which may further divided into sub-
committees. The typical structure of PVC committees is shown in figure 9-4 in
appendixB.
BS 5500 design was updated in 2012 under the name of PD 5500, as BS does not use
anymore and amendments were made to the specification to those published in
September 2011. It alsoincludes details of enquiry cases and will be updated every
three years. Your BS 5500 subscription includes free updates on further amendments.
Thestart of the new three-year cycle will include updates and new enquiry cases. The
three-year subscription to BS 5500 includes free regular updates to manufacturers using
BSI design codes (BSIGroup, 2012).
In May 2002, BS 5500 was withdrawn from the list of British Standard because it was
not consistent with the European Pressure Equipment Directive (97/23/EC). The first
edition of EN 13445 is not as comprehensive as BS 5500, and due to demands from
industry it was decided that the British pressure vessel standard should continue to be
available and become a published document (PD) under the new reference PD 5500,
with equal content, validity and application to the previous BS 5500. Its principle
difference is that it does not have the status of a national standard (Nash, 2006).
3.3. Materials
3.3.1.Overview
The study of material behaviour under pressure is of interest to investigators in a wide
variety of disciplines. However, regardless of the specific area of interest, the first
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requirements of any investigator in this field are a suitable vessel to contain the required
pressure and the specific experiment, and a means of penetrating the pressure (Kendall,
1969).
The western PWR pressure vessels use different materials for the different components
(shells, nozzles, flanges, studs, etc.). Since the construction of the first nuclear plant, the
choices in the materials of construction has changed as the PWR components evolve.
The Westinghouse designers specified American Society for Testing and Materials
(ASTM) SA302 Grade B for the shell plates of earlier vessels and ASTM SA533 Grade
B Class 1 for later vessels. Other material that are common in use include American
Society of Mechanical Engineers (ASME) SA508 Class 2 plate in the USA,
22NiMoCr37 and 20MnMoNi55 in Germany, and 16MnD5 in France (Annul book of
ASTM Standards , 1989)
According to (Tenckhoff, 1992) most of the Nuclear Steam Supply System (NSSS)
vendors use forgings in the construction of the shell courses. The main ferrous materials
used for PWR vessels construction over the years and summarises their chemical
composition are listed in table 9-1, appendix B and the table 9-2 in appendix B is
showing the materials used for individual vessel components.
(Sindelar, 2000) has mentioned that mild carbon steel with specification American
Society for Testing and Materials (ASTM) A285 is a common material of construction
for vessels in the petroleum and nuclear industries. Typically this is either mild carbon
steel, various kinds of alloyed steel or pre-stressed concrete. Data providers should
choose the appropriate option from the following multiple-choice menu carbon steel,
alloyed steel, stainless steel, concrete. Extensive analyses and experimental
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investigations have demonstrated reactor vessel integrity in full consideration of
potential service induced degradation mechanisms, including stress corrosion cracking.
BSIs material properties section states that materials should be selected on the basis of
tensile properties suitable for strength, welding, corrosion, fabrication, etc. The basic
material properties utilised in the design codes are yield and the tensile strengths. When
a vessel or structure is subjected to a cyclic loading condition then material data that
will define the fatigue failure mechanism are required. For the vessels operating in the
creep range, the design strengths are based on the minimum stress to rupture in the
appropriate service life. The rupture stresses are those agreed by the International
Standards Organisation for particular steel. These properties are mentioned in BS 5500
and BS 1113 in BSI design codes (Houston, 1987). Some of these steel properties are
shown in table 9.3, appendix B.
3.3.2.Cladding material
Cladding is the outer layer of the fuel rods standing between the coolant and the nuclear
fuel as nuclear fuel cannot be allowed to make a direct contact with the coolant inside
the reactor vessel, due to the potential for radioactivity to be released into the
environment. It can also be used in the reactor pressure to prevent corrosion damage.
Cladding is made of acorrosion-resistant material with lowabsorption cross section for
thermal neutrons. Common Choices for cladding material are stainless steel in Fast
Neutron Reactors, zirconium alloy (zircoloy) in Pressurised Nuclear Reactors and
Magnox has been used in the past (University of Cambridge , 2010)
The interior surfaces of the steel vessel, closure head and flange area are typically clad
with stainless steel, usually Type 308 or 309. Cladding was used to prevent general
corrosion by borated coolant and to minimize the buildup of corrosion products in the
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reactor coolant system. The cladding was applied in one or two layers by multiple-wire,
single-wire, strip-cladding, or resistance welding processes. Some vessels have areas of
Alloy 82 or 182 weld cladding where Alloy 600 components were welded to the vessel
(International Atomic Energy Agency , 1999)
In Fukushima Daiichi nuclear plant disaster, hydrogen was built up in the reactor vessel
which caused the explosion in the nuclear as a result immense amount of the radiations
leaked. According to (John D. Stemien, 2013), hydrogen was produced as result of hot
steam coming into contact with zirconium alloy or zircoloythe material used as a fuel
rod cladding.
To prevent the disasters like Fukushima Daiichi in future new methods for fuel cladding
are being researched and develop at Massachusetts Institutes of technology. This new
method of covering active nuclear fuel pellets involves ceramic silicon carbide (SiC).
Silicon carbide as compared to zirconium alloy used in most water cooled plants,
produces up to thousand times less hydrogen when reacting with hot steam. (Chandler,
2013)
3.3.3.
Properties of some noticeable reactor vessel materials
3.3.3.1. Alloy 600
Alloy 600 (UNS designation N0660) is a nickel-chromium alloy designed for use in
applications from cryogenic to elevated temperatures in the range of 2 000 F (1 093
C). Alloy 600 is non-magnetic and readily weldable. The alloy is used in a variety of
corrosion resisting applications. The high nickel content of Alloy 600 provides a level
of resistance to reducing environments, while the chromium content of the material
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provides resistance to weaker oxidizing environments. The high nickel content of the
material provides exceptional resistance to chloride-ion stress-corrosion cracking
Alloy 600 is a nickel-based alloy which contains 72% Ni minimum, 14 - 17% Cr, and 6
-10% Fe with high general corrosion resistance that has been widely used in light water
reactor (LWR) power plants Such as PWs and BWRS. In PWR plants, alloy 600 has
been used for steam generator tubes, CRDM nozzles, pressurizer heater sleeves,
instrument nozzles and similar applications. The alloy was originally developed by the
International Nickel Corporation (INCO) and is also known as Inconel 600 which is a
trade mark now held by the Special Metals Corporation (White, 1959)
(Jeff Gorman, 2009) explains why alloy 600 was selected to use for LWRs in the 1950s
and 1960s by providing the following reasons:
a) It has good mechanical properties, similar to those of austenitic stainless steels.
b) It can be formed into tubes, pipes, bars, forgings and castings suitable for use in
power plant equipment.
c) It is weld-able to itself and can also be welded to carbon, low-alloy and austenitic
stainless steels.
d) It is a single phase alloy that does not require post weld heat treatment. Also, when
subjected to post weld heat treatments that are required for low-alloy steel parts to
which it is welded, the resulting sensitization (decreased chromium levels at grain
boundaries associated with deposition of chromium carbides at the boundaries) does
not result in the high susceptibility to chloride attack exhibited by austenitic
stainless steels that are exposed to such heat treatments.
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e)
It has good general corrosion resistance in high temperature water environments,
resulting in low levels of corrosion products entering the coolant and resulting in
low rates of wall thinning.
f) It is highly resistant to chloride stress corrosion cracking (SCC), and has better
resistance to caustic SCC than austenitic stainless steels.
g) Its thermal expansion properties lie between those of carbon/low-alloy steels and
austenitic stainless steels, making it a good transition metal between these materials.
3.3.3.2. SA-508
There are only a few steels that have been sufficiently tested for approved use in the
construction of nuclear pressure vessels, partly because the qualification of such
materials requires an enormous amount of time consuming work. The reactor pressure
vessels (RPV) in particular have demanding requirements for tensile strength, toughness
and resistance to irradiation embrittlement over the projected service life. One of the
most popular alloys is ASME SA508. It has been in use extensively in variety of reactor
facilities, such as pressure vessels, steam generator and the pressuriser (H.
Pous_romero, 2012)
(S. Lee, 2002) states that the SA508 steel is generally given multi heat treatments
involving austenitising followed by water quenching, and tempering at temperatures as
high as 650C. The tempering treatment produces a variety of sub situationally alloyed
carbides and can relieve stresses generated by fabrication operations.
There is ongoing debate on whether to change update the SA508 grade 2 to grade 3 on
ASME as within the broad specification of SA508, there is a particular variant, Grade 3,
which exhibits better mechanical properties than earlier versions and is the material of
choice for pressure vessels in Generation III plants. There have been many studies of
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the microstructure obtained after the series of heat treatments described above, with the
conclusion that the quenching produces bainite, whereas the tempering leads to the
formation of molybdenum rich M2C precipitates in addition to residual cementite. So
the material will not lose its ductility for longer period of time (Lucas, 1999).
(Pickering, 2014) supports the idea of using the SA508 grade 3 alloy in next generation
nuclear power plants commenting that any change in the chemical composition due to
the irradiation on the inner wall of the pressure vessel influence the thermodynamics
and kinetics of phase changes and hence will affect microstructural evolution, with
corresponding changes in mechanical properties. Toughness is of particular importance
in pressure-vessel applications and the nature of carbide precipitation appears to be the
key controlling microstructural feature in ferrous steels such as SA508 Grade 3.
3.4.
Designing and Manufacturing Techniques
3.4.1. CAD Packages
With the cad software the designing professionals are offered large number of tools that
help in carrying out thorough engineering analysis of the proposed design. The tools
also help designers to consider large number of investigations. Since the cad systems
offer greater accuracy, the errors are reduced drastically in the designed product leading
to better design. Eventually, better design helps carrying out manufacturing faster and
reducing the wastages that could have occurred because of the faulty design. (Khemani,
2008)
The first the CAD graphic was developed in mid-1960s under the registered name of the
Control Data Digigraphs system. But the demand of the programme was very low only
the few copies were sold. The need of the computer-based graphics slowly started to get
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recognised by the engineers to improve the productivity. Manufacturing companies
especially those in Automotive, Defence and Aerospace started taking interest in
drafting graphics. During early years the work fell into two categories. On one hand
companies such Renault and the Ford foFcused on the mathematical definition of
complex surfaces while other Companies such as the Lockheed martin focused to
drafting packages. Eventually Renault work evolved into Dassault Systems Catia.
(Booker, 1963).
3.4.2. CATIA V5
According to Dassault Systems CATIA delivers the unique ability not only to model
any product, but to do so in the context of its real-life behaviour and they claims that
the CAD package can be used to design or draft any kind of product. It has played
major role in NASAs design of the space shuttleprogramme and has significance in
designing the fighter jets and air craft carriers. Although initially CATIA v5 was
developed for aviation industry but I has extensively being used in other engineering
industries some of these industries include; appliances, architecture, automotive,
construction, consumer goods, electronics, medical, furniture, machinery, mold and Die,
and shipbuilding because it provides advanced technologies for mechanical surfacing
&BIW. It also provides tools to complete product definition, including functional
tolerances as well askinematics definition. (Dassalt Systems, 2014)
3.4.3. FEA Simulation Packages
After Second World War when the pragmatic approach to the engineering problems
became possible to accessible by the aid of the computers, in mid 1950s, structural
engineers managed to fuse the well-established frame analysis and variational methods
into discretization method in which continuous model can be divided into elements with
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Through both significant R&D investment and key acquisitions, the richness of
technical offering has flourished throughout the years (Ansys, Inc, 2015)
There are many types engineering analysis that can be conducted on Ansys such as
structural analysis, vibrational analysis, fatigue analysis and heat transfer. In this report
the focus will be on structural analysis. Structural analysis is mainly consists of linear
and non-linear static stress analysis. Linear static stress analysis represents the most
basic type of analysis and assumes that the material remains within the elastic limit,
while the non-linear stress models consist of stressing the material past its elastic limit.
This report is only focused on linear static stress analysis.
Like many other FEA analysis software packages Ansys also provides pre-written
commercial codes. There is wide range of objective functions behind these prewritten
codes. These objective functions are usually to define; element types such as four-node
quadrilaterals and the eight-node quadrilaterals, material types such as isotropic,
orthotropic and general anisotropic, and variables within the system such as force,
displacement, pressure and heat flux.
3.4.4. Fabrication
The fabrication technics of reactor pressure vessels are constantly evolving since the
first creation of first nuclear plant. Now the new RPVs are getting fabricated by utilising
the knowledge gained from the surveillance programmes and more modern
methods such as the use of large forgings to reduce the number of welds in the
beltline (International Atomic Energy Agency , 1999)
(Ballesteros, 2014) states that before the RPV enters service it is the duty of the
manufacturer to ensure that the carefully-specified and tested, highest-quality materials
have been used and the tried-and-tested fabrication procedures have been employed. He
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also emphasises on a comprehensive quality assurance programme and extensive
ultrasonic inspection during fabrication. The importance of this last aspect is
highlighted by the economic impact and safety concerns of the hydrogen flakes
discovered in the RPVs of Doel 3 and Tihange 2.
Most of the companies such as Westinghouse, Paladin energy and NRG energy do not
construct/fabricated their own RPVs they get their RPVs fabricated from third party
manufacturers. For example in USA most RPVs in the USA were fabricated by either
combustion Engineering, Chicago Bridge and Iron, or Babcock and Wilcox. Some
vessels were fabricated in Europe by Rotterdam Dry-dock Company and by
Creusot-Loire. In some cases, vessels were constructed by more than one fabricator
because of scheduling problems in the shops.
Thick-walled cylindrical steel vessel enclosing the reactor core in a nuclear power plant
the vessel is made of special fine-grained low alloy ferritic steel, well suited for welding
and with a high toughness while showing low porosity under neutron irradiation. The
inside is lined with austenitic steel cladding to protect against corrosion. For a 1,300
MWe pressurized water reactor, the pressure vessel is usually about 12 m high, the
inner diameter is 5 m, and the wall of the cylindrical shell is about 250 mm thick. The
overall weight amounts to approx. 530 t without internals. The vessel is designed for a
pressure of 17.5 MPa (175 bar) and a temperature of 350 C
3.4.4.1. Welds
Welds are considered to be very important processes in the fabrication of the RPVs.
Large vessels are fabricated by two methods. In the first method, rolled and welded
plates are used to form separate steel courses. Such a vessel has both longitudinal and
circumferential weld seams, shown in figure 9.6 in appendix B. According to
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(International Atomic Energy Agency , 1999) report in some older vessels before 1972,
the longitudinal welds are of particular concern with regard to vessel integrity because
they contain high levels of copper and phosphorous.
In the second method, large ring forgings are used as shown in figure 9.7. This method
improves component reliability because of the lack of longitudinal welds. Weld
seams are located to avoid intersection with nozzle penetration weldments. Weldments
within the beltline region were minimized once research showed that weld metal could
be more sensitive to neutron radiation than base material. In general, parts of the
longitudinal shell course welds are within the beltline region when the RPV is
fabricated using plate material. At least one circumferential weld is near or marginally
within the beltline region when the RPVs are fabricated from either plates or ring
forgings. Recently, NSSS vendors are designing the RPV such that the beltline region
does not contain any weldments. This is accomplished by utilizing very large ring
forgings to fabricate the shell course.
There is an extensive research on going for the new efficient and the most economical
ways of welding the plates or the rings of the RPV. SA508 steels are typically used in
civil nuclear reactors for critical components such as the reactor pressure vessel.
Nuclear components are commonly joined using arc welding processes, but with design
lives for prospective new build projects exceeding 60 years, new welding technologies
are being sought. One of these new technologies introduced and implanted by the
institute of china (Wei Guo, 2015) is Autogenous laser welding. the schematic of the
laser welding configuration is shown in figure 9.8 in appendix B.
3.4.4.2. Surveillance programmes
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According to (Sonaeda, 2015) the reactor pressure vessels are fabricated in accordance
with strict quality assurance (QA) programmes. Information about how to produce a
RPV is well documented. All phases are covered, beginning with the technical
requirements and ending with the monitoring of all work performance activities. During
fabrication activities, the RPV undergoes non-destructive examinations (NDE) and
concludes fabrication with a shop hydrostatic test at some given value above operating
limits. Further, once a NPP is in operation, the RPV is subjected to comprehensive
periodic in service inspection, including material radiation damage assessment via the
surveillance programme.
(Ericksonkirk, 2007) states that there are number surveillance programme running for a
number of older water moderated, water cooled energy reactor (WWER) power plants
and some western pressurise water reactor which can be used in future in design stage
of reactor vessels. This information would also be helpful to predict the radiation
damage on the pressure vessel.
3.5. Inspections
The most important task of every utility operating a nuclear power plant is the
continuously keeping of the desired safety and reliability level. This is achieved by the
performance of numerous inspections of the components, equipment and system of the
nuclear power plant in operation and in particular during the scheduled maintenance
periods at re-fuelling time. Periodic non-destructive in-service inspections provide most
relevant criteria of the integrity of primary circuit pressure components. The task is to
reliably detect defects and realistically size and characterize them. One of most
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important and the most extensive examination is a reactor pressure vessel in-service
inspection (Picek, 2014).
Inspection demand high standards of technology and quality and continual innovation in
the field of non-destructive testing (NDT) advanced technology. The Inspection in
service nuclear reactors inspects and takes into account the following characteristics of
the nuclear plant.
3.5.1. Irradiation Embrittlement
Reactor pressure vessels which contain
the main core of the nuclear reactor are
made of thick steel plates that are welded
together. Neutrons from the fuel in the
reactor irradiate the vessel as the reactor is
operated. This can embrittle the steel,
make it less tough and less capable of
withstanding flaws which may be present.
Embrittlement usually occurs at a vessels
beltline as this section is closest to the
reactor fuel. Small fractures only nano-
meter in sizes produced due to the irradiation effect cause the hardening of the beltline
of the pressure vessels. The key embrittlement processes is illustrated on the figure 3-1.
On the figure 3-1; (a) shows the generation of lattice defects in displacement cascades
by high-energy recoil atoms from neutrons scattering which cause the primary radiation
damage. The primary defects are in the form of single and small clusters of vacancies
and self-interstitials: (b) the small cluster leads to enhanced solute diffusion and
Figure 3-1: Irradiation Embrittlement
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formation of nanoscale defect-solute cluster complexes (iron atoms not shown); (c)
these complexes of nanoscale cause the dislocation pinning and hardening of metal.
According to (Gunter, 1996) the reactor vessel is by far the single largest safety-related
component in the reactor. The pressure vessel is the principle boundary for the reactor
core cooling capability and radiation containment system. A NRC report emphasizes the
importance of this component stating integrity of the reactor pressure vessel is
essential in ensuring the reactor safety. He argues that if the reactor pressure vessel
fails there is no back up available to cool down the main reactor core, thus the effects of
the radiation embrittlement should be emphasized during the design process.
(G.E.Lucas, 2001) states that the neutron irradiation embrittlement could limit the
service life of some of the reactor-pressure vessels in existing commercial nuclear
power plants. Improved understanding of the underlying causes of embrittlement has
provided regulators and power plant operator better estimates of vessel-operating
margins. He further argues that emphasizes on the status of mechanistic understanding
of models, and their role in increasing the reliability of vessel-integrity assessments can
improves the life expectancy of reactor pressure vessels. Vessel assessment that
includes a new fracture toughness master curve method can improve the material
selection phase for the pressure vessels.
(Fahim Hashim, 2006) states that the degradation of reactor pressure vessel steels due to
neutron irradiation embrittlement is directly related to safety and life of the nuclear
power plant. In order to ensure structural integrity and safe operation of nuclear power
plants surveillance programs should be conducted more often to monitor and predict the
changes in reactor pressure vessel material.
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3.5.2. Corrosion
(Harvey, 1991) says that corrosion can reduce the fatigue life through the surface
damage effect of roughening and also by pitting which reduces the cross-sectional area,
thereby increasing the magnitude of the applied stress. He also argues that a more
serious type of damage can occurs if corrosion and radiation embrittlement act
simultaneously.
(R.W.Nichols, 1987) Describe the use of the ASME and British standard codes to
estimate the growth of cracks driven mainly by thermal shocks. Reapeated application
of the thermal shocks may lead to crack ignition and crack growth. The ability to use
current codes and standards to describe this type of crack is desirable. If the standards of
the design codes are fully applied during the designing processes thermal shock can be
avoided.
3.5.3. Fracture toughness
Fracture toughness is an indication of the amount of stress required to propagate a pre-
existing flaw. It is a very important material property since the occurrence of flaws is
not completely avoidable in the processing, fabrication or service of a
material/component. Flaws may appear as cracks, voids, metallurgical inclusions, weld
defects, design discontinuities, or some combination thereof. Since engineers can never
be totally sure that a material is flaw free, it is common practice to assume that a flaw of
some chosen size will be present in some number of components and use the linear
elastic fracture mechanics (LEFM) approach to design critical components. This
approach uses the flaw size and features, component geometry, loading conditions and
the material property called fracture toughness to evaluate the ability of a component
containing a flaw to resist fracture.
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The integrity of the reactor pressure vessel (RPV) is essential for the continued
operation of a nuclear power plant (NPP). Most studies related to long term operation,
beyond typical design life, have identified the RPV as the most critical component of
the NPP. Essentially all commercial light water reactors use ferritic low alloy steels for
the construction of the RPV, so structural integrity relies upon accurate knowledge of
the change in fracture toughness of the RPV materials over the time of operation
(IAEA, 2009)
ASME has recommended the guide lines for the fracture toughness of the RPVs. The
new criteria are introduced in section III of nuclear codes. It states that the Nuclear
Power Plant Components, to provide assurance against brittle failure. The criteria
required the component materials to satisfy certain fracture toughness requirements
(NB-2330 of the Code). The criteria also introduced non-mandatory Appendix G,
"Protection against Non-Ductile Failure", into the ASME Code. Appendix G of Section
III presents a procedure for obtaining the allowable loading for ferrous pressure-
retaining materials in Class 1 components. The procedure is based on the principles
of linear elastic fracture mechanics (LEFM) (ASME Boiler and Pressure Vessel
Code, Section III, Division , 2010)
3.5.4. Crack
During the fabrication of some RPVs it was discovered that small cracks were present
in the base metal beneath the cladding of the steel. The first incident of underclad
cracking was discovered in the early 1970s in Europe and later in the USA. This
cracking was defined as "reheat cracking" because the cracks appeared after the final
stress relief heat treatment of the RPVs. Reheat cracking was limited to RPVs fabricated
from A508 Class 2 forging steel or the equivalent European grades. Reheat cracking
only occurred when the cladding was applied utilizing a high heat input welding
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procedure. During the cladding process, grain coarsing occurred due to the high heat
input of the welding procedure, thus weakening the underclad grain boundaries. Then
the subsequent post-weld stress relief heat treatment at elevated temperature
resulted in decohension of the grain boundaries, e.g., small cracking occurred.
Underclad reheat cracks are approximately 2 to 3 mm in depth and can be detected
during the preserve NDE by using straight beam transducers. However, it is virtually
impossible to size these cracks with NDT. Reheat cracking is, for the most part,
confined to the cylindrical portion of the RPV. The beltline region can contain many
millions micro cracks (International Atomic Energy Agency , 1999).
3.5.5. Creep/Stress Rupture
Virtually all PWRs are experiencing accelerating deterioration of steam generator tubes
because of the susceptibility of the metal used in this component and other safety-
related parts. The Nuclear Regulatory
Commission (NRC) reports that cracking
of steam generator tubes is surging in US
reactors. Cracking has been identified by
Nuclear Reaction Commission (NRC) as
having serious safety implications
because the thousands steam generator
tubes constitute a major reactor coolant
pressure boundary. A multiple tube rupture in this system could result in a rapid loss of
coolant accident in the reactor beyond the ability of the Emergency Core Cooling
System to control. This could result in a meltdown of the reactor core. Because the
steam generators are equipped with relief valves, a rupture of the primary coolant loop
Figure 3-2: Fracture Failure
http://www.nationalboard.org/index.aspx?pageID=164&ID=187
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results in a radioactivity release that bypasses the containment structure with significant
risk of a catastrophic accident.
(David N, 1991) states that Indeed, theASME Boiler and Pressure Vessel
Code recognizes creep and creep deformation as high-temperature design limitations
and provides allowable stresses for all alloys used in the creep range. One of the criteria
used in the determination of these allowable stresses is 1% creep expansion, or
deformation, in 100,000 hours of service. Thus, the code recognizes that over the
operating life, some creep deformation is likely. And creep failures do display some
deformation or tube swelling in the immediate region of the rupture.
(R.W.Nichols, 1987) Describe the use of the ASME and British standard codes to
estimate the growth of cracks driven mainly by thermal shocks. Reapeated application
of the thermal shocks may lead to crack ignition and crack growth. The ability to use
current codes and standards to describe this type of crack is desirable. If the standards of
the design codes are fully applied during the designing processes thermal shock can be
avoided.
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4.MethodologyThis section provides the details on how the project is conducted. Project plan of
semester one and two is provided. The engineering drawing of pressure vessels are
produced and the details are provided. Results from simulation packages are
graphically and numerically given. The comparison is made between ASME andBSI codes. For the design codes only the relevant information is provided and
where it thought to be necessary the information is referred to relevant code
tables.
4.1. Project Plan
The project is conduct on the basis of design by analysis rule and design by rule as per
design codes determined. These two rules are elaborated further in this section. Before
starting the project detailed project plan was made. The work load was divided between
two semesters. The details of these two semesters are provided below.
4.1.1. Semester 1
The Gantt chart for semester one is provided in the appendix C, figure 9-9. In the first
semester most of the tasks that were conducted were research based. Research is
conducted on the history of the pressure vessels. Then on the types of pressure vessels,
reactor pressure vessel was selected after rigorous brain storming for the project see
appendix C. figure 9-10. All kind of pressure vessels are regulated by design codes and
standards. These codes were then researched in details. Some of the details of the design
code is presented in the literature review. After conducting the research on the design
codes the name of the project was planned. The most important task of semester one
was the literature review. Different methods of designing, materials, FEA packages,
manufacturing and inspecting the Reactor pressure vessels were researched, however
the literature review took longer than it was planned and then the researched was carried
on in the second semester. The parameter of the reactor pressure vessels were
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determined by the help of design codes. The software packages for the project were
decided.
4.1.2.
Semester 2The Gantt chart for the semester 2 is presented in the appendix C, figure 9-8. In
semester two some of research in the literature was conducted again and the relevant
research was added in the report. The core of the report is based on the design of the
pressure vessel as this is what supposed to get tested in the simulation package for
Finite Element Analysis. The case study was done on the RPV designs and the design
was finalised. The 3D design was then modelled on the CAD software and the detailed
engineering drawing sets were prepared, the details are provided further in this section.
At this time some more relevant research material was added in the literature review.
The modelled was then tested in the simulation software and results were then
compared using different design codes. The discussion and reporting compiling was
finished within the three weeks.
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4.2. Reactor Pressure Vessel Design
4.2.1. RPV Design Concepts
The Design of the RPV in this project is inspired by the two loop Westinghouses
reactor pressure vessel design shown in the appendix D, figure 9-11. The vessel
shown in the figure 9-11 is fairly typical of the reactor vessels used in almost all
western designed RPV with small number of modifications. However there are
significant differences in the size, nozzle designs and penetration designs among the
various RPV suppliers.
The Westinghouse pressure vessels were researched in detail and the various types of
Westinghouse RPVs then evaluated to deduce the right dimensions of the pressure
vessels. Westinghouse has produced various designs and models over the course of
several decades. Comparison between two models is shown in the figure 9-12, appendix
D. Based on these models the hand sketch was drawn to get the better picture of
reactor pressure vessel and some of dimensions were approximated. The hand sketch is
shown in appendix, figure 9-13. The nozzles are shown in the figures 9-14 and the 9-15.
Since there are no detailed dimensions available of reactor pressure vessels most of the
dimensions were derived using reverse engineering method.
Since the design is based on the Westinghouse RPV the same design parameter has
been chosen. The PWR pressure vessel design pressure is 17.24 MPa (2500 psi) and the
operating pressure is 15.51 MPa (2250 psi). The usual vessel pre-service hydrostatic
pressure is 21.55 MPa (1.25 x design pressure). The PWR pressure vessel design
temperature is 350C, while the operating temperature is typically 280 to 350C.
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4.2.2. Three dimensional modelling
To produce the detailed drawing of the reactor pressure vessel the model was 3D model
was prepared using the CAD software CATIA V5, the description of the software