Presentation is loading. Please wait.

Presentation is loading. Please wait.

ASTM cylindrical tension test specimen. Types of tensile fractures.

Similar presentations

Presentation on theme: "ASTM cylindrical tension test specimen. Types of tensile fractures."— Presentation transcript:

1 ASTM cylindrical tension test specimen

2 Types of tensile fractures

3 Engineering Stress-strain curve

4 Determination of Yield strength by off-set method

5 Typical stress-strain curves

6 Yield Point Behaviour in Low-Carbon Steel;

7 Typical Creep-curve

8 Andrade’s analysis of the competing processes Which determine the creep curve

9 Effect of stress on creep curves at constant temperature

10 Schematic stress-Rupture Data

11 Fatigue test curve for materials having an endurance limit

12 Methods of Plotting Fatigue data when the mean Stress is not zero

13 Alternative method of plotting the Goodman diagram

14 Response of metals to cyclic strain cycles

15 Construction of cyclic stress-strain curve

16 Parameters associated with the stress-strain hysteresis loop in LCF testing

17 Fatigue strain-life curve obtained by superposition of elastic and plastic strain equations (schematic)

18 Fatigue failure

19 Schematic representation of fatigue crack growth Behaviour in a non-aggressive environment

20 Sketch showing method of loading in Charpy and Izod impact tests

21 The method by which Izod Impact values are measured

22 Impact energy absorbed at various temperatures

23 Transition temperature curve for two steels Showing fallacy of depending on room Temperature results

24 Various criteria of transition temperature obtained from Charpy test

25 Effect of section thickness on transition temperature curves

26 PFBR heat transport flow sheet.

27 PFBR reactor assembly showing major components

28 CriterionClad TubeWrapper Tube Irradiation effects Void swelling Irradiation creep Irradiation embrittlement Void swelling Irradiation creep Irradiation embrittlement Mechanical properties Tensile strength Tensile ductility Creep strength Creep ductility Tensile strength Tensile ductility CorrosionCompatibility with sodium Compatibility with fuel Compatibility with fission products Compatibility with sodium Good workability International irradiation experience as driver or experimental fuel subassembly Availability Principal Selection Criteria for LMFBR Core Structural Materials

29 Schematic of fuel subassembly showing the cut out of fuel pins, bulging and bowing.

30 Variation with dose of the maximum diametral deformation of fuel pins

31 ReactorCountryFuel clad tube material RapsodieFrance316 SS PhenixFrance316 SS PFRU.K.M316 SS, PE 16 JOYOJapan316 SS BN-600Russia15-15Mo-Ti-Si Super Phenix-1France15-15Mo-Ti-Si FFTFU.S.A.316 SS & HT9 MONJUJapanmod 316 SS SNR-300GermanyX10 Cr Ni Mo Ti B1515 (1.4970) BN-800Russia15-15Mo-Ti-Si CRBRU.S.A.316 SS DFBRJapanAdvanced austenitic SS (PNC1520) EFREuropePE16 or 15-15- Mo-Ti-Si FBTRIndia316 SS Materials selected for cladding in major FBRs

32 General CriterionSpecific Criteria Mechanical properties Tensile Strength, Creep Low Cycle Fatigue Creep-Fatigue Interaction High cycle Fatigue DesignAvailability of Mechanical Properties Data in Codes Other important considerations Structural integrity Weldability Workability International experience Principal Selection Criteria for FBR Structural Materials

33 Comparison of creep rupture strengths of 316 and 316L(N) SS from various countries

34 General CriteriaCriteria related to use in sodium Mechanical Properties -Tensile Strength - Creep Strength -Low cycle Fatigue - High Cycle Fatigue -Creep-Fatigue Interaction -Ductility -Ageing Effects Mechanical properties in sodium Susceptibility to decarburisation Mechanical Properties Data shall be available in Pressure Vessel Codes Corrosion under normal sodium chemistry condition, fretting and wear Corrosion resistance under storage (pitting) normal and off-normal chemistry conditions Corrosion resistance in the case of sodium water reaction (Stress corrosion cracking, self enlargement of leak and impingement wastage) Other Important Considerations Workability Weldability Availability Cost Principal Selection Criteria for LMFBR Steam Generator Material

35 Comparison of 10 5 h creep rupture strengths of several materials

36 Creep-rupture strength of eleven types of ferritic heat resistant steels

37 Materials selected in FBRs for major components ReactorCountryReactor Vessel IHXPrimary circuit piping hot leg (cold leg) # Secondary circuit piping hot leg (cold leg) RapsodieFrance316 SS 316 SS (316 SS) PhenixFrance316L SS316 SS(316 SS)321 SS (304 SS) PFRU.K.321 SS316 SS(321 SS)321 SS (321 SS) JOYOJapan304 SS 304 SS (304 SS) 2.25Cr-1Mo (2.25Cr- 1Mo) FBTRIndia316 SS 316 SS (316 SS) BN-600Russia304 SS 304 SS (304 SS) Super Phenix-1 France316L(N) SS (304L(N) SS) 316L(N) SS FFTFU.S.A.304 SS 316 SS (316 SS) 316 SS (304 SS) MONJUJapan304 SS 304 SS (304 SS) SNR-300Germany304 SS 304 SS (304 SS) BN-800Russia304 SS 304 SS (304 SS) CRBRPU.S.A.304 SS304 and 316 SS 316 SS (304 SS) 316H (304H) DFBRJapan316FR SS316 FR316FR (304 SS) 304 SS (304 SS) EFREurope316L(N) SS # for pool-type reactor, there is no hot leg piping

38 ElementASTM 304L(N) PFBR 304L(N) ASTM- 316L(N) PFBR 316L(N) RCC- MR 316L(N) RM3331 C0.030.024- 0.03 0.030.024- 0.03.03 Cr18-2018.5-2016-1817-18 Ni8-128-1010-1412-12.5 MoNS0.52-32.3-2.7 N0.1-0.160.06- 0.08 0.1-0.160.06- 0.08 Mn2.01.6- Si1. P0.0450.030.0450.030.035 S0. TiNS0.05NS0.05- NbNS0.05NS0.05- CuNS1.0NS1.0 CoNS0.25NS0.25 BNS0.002NS0.002 ElementASTM 304L(N) PFBR 304L(N) ASTM- 316L(N) PFBR 316L(N) RCC- MR 316L(N) RM3331 Comparison of PFBR specification for 304L(N) and 316L(N) SS with ASTM A240 and RCC-MR RM-3331. (single values denote maximum permissible, NS - not specified)

39 Materials Selected for Steam Generator in Fast Breeder Reactors ReactorSodium inlet (K) Steam outlet (K) Tubing material EvaporatorSuperheater Phenix 823 7852.25Cr-1Mo 2.25Cr-1Mo stabilised 321 SS PFR8137862.25Cr-1Mo stabilised Replacement unit in 2.25Cr-1Mo 316 SS Replacemen t unit in 9Cr-1Mo FBTR7837532.25Cr-1Mo stabilised BN-6007937782.25Cr-1Mo304 SS Super Phenix-1 798763Alloy 800 (once through integrated) MONJU7787602.25Cr-1Mo304 SS SNR-3007937732.25Cr-1Mo stabilised BN-8007787632.25Cr-1Mo CRBR7677552.25Cr-1Mo DFBR793768Modified 9Cr-1Mo (grade 91) (once through integrated) EFR798763Modified 9Cr-1Mo (grade 91) (once through integrated)

40 S.NoReactorMaterial 1PhenixCarbon steel (A42P2) 2Superphenix-1Carbon steel (A48P2) 3Superphenix-2Carbon steel 4PFRCarbon steel 5FFTFCarbon Steel 6CRBRLow Alloy Steel 7EFRCarbon steel (A48P2) Materials selected for Top Shield for various Fast Breeder Reactors

41 ZIRCONICUM ALLOYS : NUCLEAR APPLICATIONS Low absorption cross section for thermal neutrons Excellent corrosion resistance in water Good mechanical properties IMPORTANT PROPERTIES OF ZIRCONIUM Allotropy (  hcp  bcc ) Anisotropic mechanical and thermal properties -Unequal thermal expansions along different crystallographic directions -Strong crystallographic texture during mechanical working -high reactivity with O 2, C, N and high solubility in  -phase -Special care during melting and fabrication -Low solubility of hydrogen in  862 o C

42 DESIRABLE MECHANICAL PROPERTIES OF ZIRCONICUM ALLOYS for PRESSURE TUBES High Yield Strength-By control of Alloying Elements -Control of Texture -Proper selection of manufacturing route High Total Circumferential Elongation % - By Introducing heavy reduction in wall thickness in the last stages of pilgering High Creep Strength (out-of-pile) - By alloying with Nb Low Creep Rate during Irradiation - By Introducing Cold Work High Fracture Toughness- Control of residual Chlorine to <0.5 ppm

43 SYNERGISTIC INTERACTIONS LEADING TO DEGRADATION OF MATERIAL PROPERTIES IN ZIRCONIUM ALLOYS 1.Corrosion by Coolant Water 2.Corrosion by Fission Products 3.Hydrogen Ingress 4.Irradiation Damage 5.Dimensional Change due to Creep and Growth

44 Important steps in fabrication flow sheets of Zirconium components for PHWR and BWR

45 Long term, in reactor, oxidation and hydrogen Pick-up behaviour of zircaloy-2 and Zr-2.5Nb pressure tubes,

46 (a)Stress reorientation of circumferential zirconium hydride platelets(left hand side) at 250 MPa stress level in the direction shown (b) A hydride blister in the zirconium alloy pressure tube section

47 Irradiation creep rate in zircaloy-2 under biaxial loading (150 MPa and 300 o C) and a schematic diagram to show the growth rate of cold-worked and recrystallization (RX) zircaloy 2

48 Change in room temperature tensile properties of mild steel produced by neutron irradiation

49 Stress-strain curves for polycrystalline copper tested at 20 o C after irradiation to the does indicated

50 Accelerated in-reactor creep in zircaloy-2

51 Impact energy vs. temperature curves for ASTM 203 grade D steel A.Unirradiated B.Irradiated to a fluence of 3.5 x 10 19 -2 C.Irradiated to a fluence of 5 x 10 18 -2 D.Annealed at 300 o C for 15 days after irradiation to a fluence of 3.5 x 10 19 -2

52 Schematic illustration of the Ludwig-Davidenkov Criterion for NDTT and its shift with irradiation

53 Element Incre- ases NDTT Redu- ces Ductile Shelf Forms Precip- itates Reduc- es surface energy Increa- ses flow stress Restri- cts cross slip P  (S) -  Cu  (S) --   S- -- V  (M)  Al  (S) Increases  (S)  Si  (M)  (S)  Effects of residual elements on sensitivity to irradiation embrittlement of steel S – Strong Effect; M – Mild Effect

54 Extra Slides Follow

55 Effects of fast reactor irradiation on the tensile properties of solution annealed 316 stainless steel

56 Irradiation creep results from pressurized tube of 20% cold worked 316 stainless steel

57 Linear stress dependence of irradiation Creep in 316 stainless steel at 520 o C and a fluence of 3 x 10 22 -2

58 Temperatur e T/T m DefectSize 0 0.1 0.3 0.5 Point defects Vacancies and interstitials One atomic diameter Multiple point defects Cluster of point defects Complexes of vacancies and interstitials with solutes A few atomic diameter Vacancies clusters and loops Diameter < 7 nm Interstitial loopsDiameter > 7 nm Rafts (agglomerates of clusters and small loops) 6-10 nm thick, 100-200 nm in length and width Voids10-60 nm Helium bubbles3-30 nm Transmutation atoms (produced at all temperatures but agglomerates at T/T m > 0.5 Defects Produced by Irradiation

59 Summary of results of dislocation dynamics In irradiated materials Lattice typeRate-controlling obstacle Un-irradiatedIrradiated BCCP-N Barrier Interstitial Solutes P-N Barrier Solutes Solute-defect complexes Clusters or loops Divacancies FCC and HCP, c/a >ideal (basal slip) Intersection of forest dislocations Depleted zones Faulted loops HCP c/a < ideal (prism slip) Interstitial solutes P-N Barrier Interstitial solutes Irradiation induced defects

60 Crack-deformation modes

61 Relation between fracture toughness and allowable stress and crack size

62 Effect of specimen thickness on stress and mode of fracture

63 Common specimens for K Ic testing

64 Load displacement curves (slope Op s is exaggerated fir clarity)

65 (a)J vs.  a curve for establishing J ic (b)Sketch of a specimen fracture surface showing how  a is determined

66 K Q = Fracture toughness P Q = Maximum recorded load B = Specimen thickness W = Specimen Width a = Crack length

67 Drop-weight test (DWT)

68 Element316L(N) SS (EFR) 316FR (DFBR) 316L(N) SS (Superphenix) C0.030.020.03 Cr17-1816-1817-18 Ni12-12.510-1411.5-12.5 Mo2.3-2.72-32.3-2.7 N0.06-0.080.06-0.120.06-0.08 Mn1.6- Si0.51.00.5 P0.0250.015-0.040.035 S0.005-.010.030.025 TiNS 0.05 NbNS 0.05 Cu.3NS1.0 Co.250.25 B.0020.0010.0015-0.0035 Nb+Ta+Ti0.15 Chemical composition specified for 316L(N), 316FR and 316LN used/proposed in EFR, DFBR and Superphenix, respectively.

69 Texture developed due to pilgering, sheet rolling and wire drawing (cold working) operations

70 Fracture appearance vs. temperature for explosion crack starter test

Download ppt "ASTM cylindrical tension test specimen. Types of tensile fractures."

Similar presentations

Ads by Google