Factors Considered in Material Selection Nuclear Reactors Physical

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Factors Considered in Material Selection (Nuclear Reactors) |Physical Properties - Density - Melting Point

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

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.

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) - A 501, A 508, A 533, SA 333 |Low Alloy (Bainitic) Steels (Turbine Rotors, Discs) • • • 1 Cr-1 Mo-0. 25 V 2. 25 Cr-1 Mo (Grade 22) Ni-Cr-Mo. V (A 469, Class 8) Ni-Cr-Mo. V (A 470, Class 8) Ni-Cr-Mo. V (A 471, Class 8) |Ferritic(Martensitic) Stainless Steel (turbine blades, end fittings in PHWR) - AISI 403 (S 40300) - AISI 410 (S 41000) - Sandvick Sweden HT 9 - Sandvick Sweden HT 7 - French R 8 - French EM 12 - Japanese HCM 9 M (Creep Strength, Oxidation and Corrosion Resistance)

Steels Commonly Used in Nuclear Plants • Austenitic Stainless Steels (Good Strength +Ductility +

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 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 -

Superalloys Commonly Used in Nuclear Plants |Superalloys • Inconel Alloys (Ni-Cr Series) • Inconel

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, PHWR • Nimonic PE 16 • Inconel 718 FBRs (Tie Rods, Cladding Core Cover Plate • Inconel 617 • Alloy 800 H HTGR (Heat Exchanger Tubes)

Materials Commonly Used in Nuclear Plants • Steam Generator Tubing • • • LWR

Materials Commonly Used in Nuclear Plants • Steam Generator Tubing • • • LWR PHWR HTGR PFBR FBTR : Inconel 600 SCC Resistance : Inconel 600 : Inconel 800 : Alloy 800 H - Creep Resistance : Mod. 9 Cr-1 Mo -Creep, SCC : 2 1/4 Cr-1 Mo - Low Temp. <427 o. C • Steam Condenser • Admiralty Brass - Fresh water • Aluminum - Bronze • Aluminum - Brass (SB 261) • Cupro - Nickel (SB 111, 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 E 8

ASTM Standards for Mechanical Properties Evaluation Type of Test Standard Tensile ASTM E 8 M (1994) ASTM E 21 (1992) Creep rupture and stress rupture ASTM E 139 (2000) Hardness ASTM E 10 (1984) ASTM E 18 (1984) ASTM E 92 (1984) High cycle fatigue ASTM E 466 (1999) Low cycle fatigue ASTM E 606 (1999) Impact ASTM E 23 (1999) Fracture Toughness (plane strain) ASTM E 399 (1989) Fracture toughness (JIC) ASTM E 813 (1989)

FUEL STRUCTURAL MATERIALS |Selection Criteria: - Low neutron absorption cross section - Low cost

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 316 SS (20% CW) Alloy D 9 (20% CW) (Modified 9 Cr-1 Mo) HTGR Graphite

CONTROL MATERIALS |Selection Criteria: - Neutron absorption cross section - Adequate mechanical strength -

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 B 4 C BWR (Clad in 304 SS) 80% Ag-15%In+5%Cd B 4 C PWR (Clad in CW 304 SS/Inconel 627) B 4 C LMFBR

MODERATOR MATERIALS - To slow down and moderate fast neutrons from - fission Materials

MODERATOR MATERIALS - To slow down and moderate fast neutrons from - 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 -

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 (B 4 C in Al matrix) Concrete

Major Power Reactors and their Ceramic Components Reactor Coola Fuel Type nt Primary Alternates

Major Power Reactors and their Ceramic Components Reactor Coola Fuel Type nt Primary Alternates Control Rod Primary Alterna tes BWR H 2 O UO 2 a, (UPu)O 2 a, b B 4 C, UO 2 Gd 2 O 3 PWR H 2 O UO 2 a (UPu)O 2 a, b (U-Th)O 2 a, b Al 2 O 3 -B 4 C HWR D 2 O UO 2 a (U-Pu)O 2 a B 4 C AGR CO 2 UO 2 a (U-Pu)O 2 a - HTGR He UC 2 c (Th. O 2) (U-Pu)O 2 c, (UTh. O 2)c B 4 C Gd 2 O 3 Al 2 O 3, Eu 2 O 3 GCFR He (UPu)O 2 a (U-Pu)C, a, c (U -Pu)Na, c B 4 C Eu 2 O 3 LMFBR Na (UPu)O 2 a (U-Pu)C, a, b (U -Pu)Na, (UPu)O 2 b B 4 C Eu 2 O 3 LWBR H 2 O (UTh)O 2 a - - a pellets; b sphere-pac; c coated particles UO 2 Gd 2 O 3

SCHEME OF PRESENTATION 1. Fundamental Aspects of Mechanical Testing and Various Mechanical Properties 2.

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

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