Transcript No Slide Title

Factors Considered in Material Selection
(Nuclear Reactors)
Physical Properties
- Density
- Melting Point
- Coefficient of Linear Expansion
- Thermal Conductivity
Mechanical Properties
- Yield Strength
- Tensile Strength
- Elongation at Fracture (Ductility)
- Creep Strength
- Fatigue Life
- Creep-Fatigue Interaction
- Impact Strength and Fracture Toughness
Neutronic Characteristics
- Low Neutron Capture Cross Section (Core)
- High Neutron Capture Cross Section (Control Rod)
Factors Considered in Material Selection
(Nuclear Reactors)
-Ability to Withstand Stress, Environment and
Temperature Over Life Time
-Previous Experience Under Similar
Conditions, if any
-Availability
-Affordability
-Ease of Fabrication
-Susceptibility to Chemical Attack and
Corrosion
-Guidelines for Design in Codes
-Potential for Activation Under Neutron
Bombardment
-Toxicity and Health Impact
Steels Commonly Used in
Nuclear Plants
Carbon Steels ( C: 0.10 to 0.20 %)
(Pressure Vessels of PWR, BWR, Pipings of BWR
-Primary Pressure Boundary Piping)
- A501, A508, A533, SA333
Low Alloy (Bainitic) Steels
(Turbine Rotors, Discs)
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1Cr-1Mo-0.25V
2.25Cr-1Mo (Grade 22)
Ni-Cr-MoV (A469, Class 8)
Ni-Cr-MoV (A470, Class 8)
Ni-Cr-MoV (A471, Class 8)
Ferritic(Martensitic) Stainless Steel
(turbine blades, end fittings in PHWR)
- AISI 403 (S40300)
- AISI 410 (S41000)
- Sandvick Sweden HT9
- Sandvick Sweden HT7
- French R8
- French EM12
- Japanese HCM9M
(Creep Strength, Oxidation and Corrosion
Resistance)
Steels Commonly Used in
Nuclear Plants
• Austenitic Stainless Steels
(Good Strength +Ductility + Resistance to
Corrosion at High Temperatures)
- AISI 304
- AISI 316
• AISI 304 L (Low Carbon, <0.03 %)
• AISI 316 L (Low Carbon, <0.03 %)
• AISI 304 LN (Low Carbon + Nitrogen)
• AISI 316 LN (Low Carbon + Nitrogen)
• AISI 321 (Ti – stabilised)
• AISI 347 (Nb – stabilised)
• AISI 308 (Welding electrodes)
• Primary Coolant Pipeings of BWR : 304 SS
304 SS Susceptible for IGSCC
If IGSCC is to be Avoided : 304 L, 316 L, 347
Inconel 600 can be used
Stainless Steels are extensively used in FBRs
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Superalloys Commonly Used in
Nuclear Plants
Superalloys
• Inconel Alloys
(Ni-Cr Series)
• Inconel 600
• Inconel 625
• Inconel 690
• Inconel 800
To Avoid SCC in
Steam-Water System
BWR, PWR, PHWR
• Nimonic PE16
• Inconel 718
FBRs (Tie Rods,Cladding
Core Cover Plate
• Inconel 617
• Alloy 800H
HTGR (Heat Exchanger
Tubes)
Materials Commonly Used in
Nuclear Plants
• Steam Generator Tubing
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LWR
PWR
PHWR
HTGR
PFBR
FBTR
: Inconel 600
SCC Resistance
: Inconel 600
: Inconel 800
: Alloy 800 H - Creep Resistance
: Mod. 9Cr-1Mo -Creep, SCC
: 2 1/4Cr-1Mo - Low Temp. <427 oC
• Steam Condenser
• Admiralty Brass
- Fresh water
• Aluminum - Bronze
• Aluminum - Brass (SB 261)
• Cupro - Nickel (SB111, 251)
• Titanium
• Type 304 SS
Sea Water Cooled
Condensers
(Higher Corrosion
Resistance)
Higher Life upto 40 yrs
ASTM Standards for Mechanical
Properties Evaluation
Type of Test
Standard
Tensile
ASTM E8M (1994)
ASTM E21 (1992)
Creep rupture and stress rupture
ASTM E139 (2000)
Hardness
ASTM E10 (1984)
ASTM E18 (1984)
ASTM E92 (1984)
High cycle fatigue
ASTM E466 (1999)
Low cycle fatigue
ASTM E606 (1999)
Impact
ASTM E23 (1999)
Fracture Toughness (plane strain)
ASTM E399 (1989)
Fracture toughness (JIC)
ASTM E813 (1989)
FUEL STRUCTURAL MATERIALS
Selection Criteria:
- Low neutron absorption cross section
- Low cost
- Adequate tensile strength
- Adequate creep strength
- Adequate ductility after irradiation
- Corrosion resistance
Materials:
Reactor
Cladding
BWR
Zircaloy-2 / Zircaloy-4
PWR
Stainless Steel 304
Zircaloy-4
PHWR
Zircaloy-2
Zr-2.5%Nb Alloy
LMFBR
Type 316SS (20% CW)
Alloy D9 (20% CW)
(Modified 9Cr-1Mo)
HTGR
Graphite
CONTROL MATERIALS
Selection Criteria:
- Neutron absorption cross section
- Adequate mechanical strength
- Corrosion resistance
- Chemical and dimensional stability
-
(under prevailing temperature and irradiation)
Relatively low mass to allow rapid movement
Fabricability
Availability and reasonable cost
Materials:
Boron, Cadmium, Gadolinium, Hafnium, Europium
B4C
BWR (Clad in 304 SS)
80% Ag-15%In+5%Cd
B4C
PWR (Clad in CW 304
SS/Inconel 627)
B4C
LMFBR
MODERATOR MATERIALS
- To slow down and moderate fast neutrons from
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fission
Materials with light nuclei are most effective
Materials
Moderating ratio
Light water
70
Heavy water
2100 (0.2% light water as
impurity)
12000 (100% heavy water)
Metallic Beryllium
150
Graphite
170
Beryllium oxide
180
{Moderating ratio = macroscopic scattering cross section / absorption cross section}
REFLECTOR MATERIAL
- To cut down the neutron leakage losses from core
- Desired properties same as moderators
Water
Heavy Water
Beryllium
Graphite
Thermal Reflectors
SHIELDING MATERIAL
To protect personnel and equipment from the
damaging effects of radiation
- Good moderating capability
- Reasonable absorption cross section
- Cost and space availability
- Neutron, a,b and g shielding
- Both light and heavy nuclei are preferred
WATER
PARAFFIN
POLYETHYLENE
Pb, Fe, W
Boral (B4C in Al matrix)
Concrete
Major Power Reactors and their
Ceramic Components
Reactor Coola
Fuel
Type
nt
Primary
Alternates
Control Rod
Primary
Alterna
tes
BWR
H2O
UO2a
UO2a, (UPu)O2a,b
B4C, UO2Gd2O3
PWR
H2O
UO2a
UO2a (UPu)O2a,b
(U-Th)O2a,b
Al2O3-B4C
HWR
D2O
UO2a
(U-Pu)O2a
B4C
AGR
CO2
UO2a
(U-Pu)O2a
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HTGR
He
UC2c
(ThO2)
(UO2)
(U-Pu)O2c,
(U-ThO2)c
B4C
Gd2O3Al2O3,
Eu2O3
GCFR
He
(UPu)O2a
(U-Pu)C,a,c
(U-Pu)Na,c
B4C
Eu2O3
LMFBR
Na
(UPu)O2a
(U-Pu)C,a,b
(U-Pu)Na, (UPu)O2b
B4C
Eu2O3
LWBR
H2O
(UTh)O2a
a pellets; b sphere-pac; c coated particles
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UO2Gd2O3
SCHEME OF PRESENTATION
1. Fundamental Aspects of Mechanical Testing
and Various Mechanical Properties
2. ASTM Standards for Various Mechanical
Tests
3. Factors Considered in Materials Selection
(Nuclear Reactors)
4. Types of Materials in Nuclear Reactors
5. Cladding Materials in Thermal Reactors
(Zirconium Alloys)
6. Cladding Materials in FBRs
7. Different NDT Techniques – Principles
8. Application of NDT Techniques in Nuclear
Industry
9. Different Types of Corrosion
10. Corrosion Protection Methods
11. Corrosion in Nuclear Plants