Thermal creep tests on Zr-2.5Nb pressure tube and related concern on the structural integrity assessment

Transcription

1 Workshop on the Prediction of Axial and Radial Creep in HWR Pressure Tubes IAEA Vienna November 2011 Thermal creep tests on Zr-2.5Nb pressure tube and related concern on the structural integrity assessment Vasile RADU Institute for Nuclear Research Pitesti, Romania

2 Outline Introduction ICN research activity ICN contribution into previous IAEA CRP s Thermal creep tests on Zr-2.5%Nb P/T Structural integrity assessment concerning on P/T flaws Possible task: thermal creep modeling at flaws Summary

3 Introduction ICN Pitesti Mission: to promote the peaceful utilization of nuclear power in according to the international agreements; to provide scientific and technologic support for the Romanian Nuclear Program; to offer scientific and technologic support for safety and economic operation of CANDU NPP Units from Cernavoda NPP; to develop technologies, methods, computer codes, experimental infrastructure, directed towards an end-product or service with applications in nuclear power plants and nuclear field.

4 Introduction Cernavoda U1&2 NPP Reactor type: CANDU 6 Electrical power: 705 (MW) Start of construction: 1980 U1-first grid connection: 1996 U2-first grid connection: 2007 Design life: 30 years Performance: Yearly more than 11 mil MWh; Capacity factor > 90%; Electricity generated 18% of the Romania overall production Unit Unit

5 Introduction CNU-Feldioara Cernavoda NPP ROMAG ICSI ICN FCN

6 ICN research activity National Programs Research, Development and Engineering Strategic Programme (managed by Romanian Authority of Nuclear Activity) National Competition (managed by Ministry of Education & Research) - Research of Excellence; - I st and II nd National Program. National Collaboration International Co-operation Units 1 & 2 CNE - Cernavoda NPP; Nuclear Fuel Plant (FCN); Heavy Water Plant (ROMAG) Center of Design for Nuclear Objectives (CITON); Other R&D Institutes Projects, Contracts IAEA Vienna ( CRP, TP); European Union Euratom FP5, FP6, FP7 COG, AECL Canada ; DoE USA; OECD Nuclear Energy Agency KAERI Rep. of Korea

7 ICN research activity Research, Development and Engineering Strategic National R&D Program Nuclear Power Programs Nuclear Safety; Fuel Channel; Nuclear Fuels; Fuel Handling; Management of Radioactive Wasted including Spent Nuclear Fuel; Environment Protection; Steam Generator; Process Systems and Equipment; Chemistry of NPP Circuits; Instrumentation and Control Analysis of NPP Operating Events, Aging, Environment Qualification and Life Extension; Advanced Nuclear Reactors and Fuel Cycles; Heavy Water and Tritium Other Programs Extension of TRIGA Reactor Performances; Irradiation Technologies and Radioisotopes; Applications of Nuclear Techniques; Informatics; Support for International Co-operation;

8 ICN research activity ICN - R&D Departments Reactor Physics and Nuclear Safety TRIGA Reactor Nuclear Materials and Corrosion Post-Irradiation Examination Radioactive Wastes Management Out-of-Pile Testing Radiation, Environmental and Civil Protection Electronics Design Services

9 ICN research activity Infrastructure Nuclear Materials & Corrosion Department

10 ICN research activity Thermo-Mechanical Testing Systems WALTER+BAI, Static & Dynamic Materials Testing System 30KN, equipped with temperature furnace ( 1000C), RT and HT extensometers, COD INSTRON Static Materials Testing System 25KN, equipped with thermal chamber ( 350C) compatible with the Walter+Bai System, RT extensometers

11 ICN research activity Creep-Testing Machine equipped with furnace (700 C) σ = MPa (standard specimen) T = 250 C -700 C; Metallographic microscope with acquisition of images and processing

12 ICN contribution into previous IAEA CRP s Romanian Participation at previous IAEA-CRP s IAEA, T12017 CRP Delayed hydride cracking (DHC) of zirconium alloy fuel cladding ( ) IAEA, T12013 CRP Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium-based Alloys ( ).

13 Thermal creep tests on Zr-2.5%Nb P/T Short overview The fuel channel of CANDU nuclear plant comprises: a Zr-2.5%Nb P/T, two end fittings (in contact with the primary coolant), a Zircaloy-2 calandria tube (in contact with the moderator) The pressure tubes are subjected to high stresses, temperatures, fast neutron fluxes which cause changes in the dimensions and material properties In addition, the long term operation induces wear with inherent flaws at the inner P/T surface To ensure the safe, reliable and economic performance of the reactor is important the changes are known and can be predicted The structural integrity analyses of CANDU pressure tubes are performed in accordance with the technical requirements stated in the Canadian standard CAN/CSA N285.8[1] [1] CAN/CSA N285.8, Technical Requirements for In-Service Evaluation of Zirconium Alloy Pressure Tubes in CANDU Reactors, 2005

14 Thermal creep tests on Zr-2.5%Nb P/T P/T main characteristics Operating temperatures: 520 K to 540 K (inlet) and 565 to 585 K (outlet) The inlet pressure 10.5 MPa and the outlet pressure is about 9.9 MPa: axial stress 65 MPa, hoop stress varies from 130 MPa (inlet) to 122 MPa (outlet). Dimensions: 6 m long, wall thickness 4.2 mm, diameter 104 mm; Peak flux up to nm. s (E> 1MeV)

15 Thermal creep tests on Zr-2.5%Nb P/T The main factors affecting CANDU P/Ts from fuel channels are: Dimensional changes (irradiation creep and growth); Hydrogen/Deuterium ingress in P/T (corrosion); Fracture toughness reduction (embrittlement); Service induced damage (scratches from refueling); End fitting degradation.

16 Thermal creep tests on Zr-2.5%Nb P/T Short overview on the thermal creep in zirconium alloy The complex interaction between the effects of the temperature and fast neutron flux on the P/T deformation has lead to a long term steadystate deformation law which consists of the following separable, additive components: thermal creep, irradiation creep and irradiation growth [1] All three components are isotropic and contribute to length as well as diameter changes The in-reactor thermal creep component has two terms that dominate at temperature above and below 300 C [1]: dε dt k ( ) dεk k k 2 Q1 k Q3 = KC 1 1σ1+ K2C2σ2 exp K3C1 1exp dt T + σ T is strain rate in k direction (radial, transverse, axial) 1/h; σ1, σ2 effective stress in MPa for thermal creep related to radial, axial and hoop stress by means of Hill s anisotropy coefficients. [1] R. A. Holt, In-reactor deformation of cold-worked Zr-2.5%Nb pressure tubes, journal of Nuclear materials 372 (2008)

17 Thermal creep tests on Zr-2.5%Nb P/T Short overview on the thermal creep in zirconium alloy Other equation to describe thermal creep of zirconium alloys comprises two terms, first one predominant at high stress and a second one at low stress [1]: dε n Q1 Q2 = D1σ exp D2σ exp dt + RT RT The equation described by Norton law (variation of strain rate with the applied load) coupled with Arrhenius temperature dependence dε n Q = Aσ exp dt RT where n stress exponent may have values close to 1 at low stress levels, and for high levels of stress n ranges between 2 to 100. [1] F. A. Nichols, Creep of Zirconium Alloys in Nuclear Reactors, ASTM STP 815

18 Thermal creep tests on Zr-2.5%Nb P/T INR thermal creep tests objective The objective of thermal creep tests performed at INR Pitesti was to obtain the creep strain rate equation and stress exponent for transverse direction of Zr-2.5%Nb P/T in the form of Norton law dε real dt n Q = Aσrealexp RT with: A constant,q activation energy, R=8.31 J/mol. K, T in K and ε Temperature range: T from 250 C to 350 C Stress domain: real = ln 1+ ( εeng) σ from 140 to 320 MPa

19 Thermal creep tests on Zr-2.5%Nb P/T Thermal creep test Geometry and dimensions of transverse sample cut from Zr-2.5%Nb P/T Creep tensile test machine

20 Thermal creep tests on Zr-2.5%Nb P/T Results: Engineering creep strain as function of time: -transverse direction P/T, Zr-2.5%Nb, -T=300 C, -Four level of stress σ= 170 MPa, 250 MPa, 280 MPa, 300 MPa

21 Thermal creep tests on Zr-2.5%Nb P/T Results: Engineering creep strain as function of time: -transverse direction P/T, Zr-2.5%Nb, -T=350 C, -Four level of stress σ= 150 MPa, 200 MPa, 280 MPa, 320 MPa

22 Thermal creep tests on Zr-2.5%Nb P/T Results: Real creep strain as function of time: -transverse direction, Zr-2.5%Nb, -T=300 C, -Four level of stress σ= 170 MPa, 250 MPa, 280 MPa, 300 MPa

23 Thermal creep tests on Zr-2.5%Nb P/T Results: Real creep strain as function of time: -transverse direction, Zr-2.5%Nb, -T=350 C, -Four level of stress σ= 150 MPa, 200 MPa, 280 MPa, 320 MPa

24 Thermal creep tests on Zr-2.5%Nb P/T Results: Creep strain rates as function of time: -transverse direction, Zr-2.5%Nb, -T=250 C, -Four level of stress σ= 140 MPa, 200 MPa, 250 MPa, 300 MPa

25 Thermal creep tests on Zr-2.5%Nb P/T Results: Creep strain rates as function of time: -transverse direction, Zr-2.5%Nb, -T=350 C, -Four level of stress σ= 150 MPa, 200 MPa, 280 MPa, 320MPa

26 Thermal creep tests on Zr-2.5%Nb P/T Results: Creep strain rates as function of applied load: -transverse direction, Zr-2.5%Nb, -T=250, 300, 350 C,

27 Thermal creep tests on Zr-2.5%Nb P/T Results: dε real dt n = Aσrealexp Q RT The Norton law thermal creep rate is obtained with the following parameters: A= ^(-6) 1/s; n=1.80 Q=68170 J/mol Comparison between experimental and prediction

28 Structural integrity assessment of P/T flaws Thermal Creep at flaws The principle that creep can relax stresses at flaws has been clearly demonstrated experimentally. The stress concentrations at flaw tips in pressure tubes can initiate DHC under certain conditions of stress, temperature history and hydrogen concentration. The flaws are usually present in P/T for many operating hours before the conditions exist which could initiate cracking. In the vicinity of a notch the stress distribution and its evolution with time depend strongly on creep response of the material which is highly anisotropic in Zr-2.5%Nb [1]. The creep rate is fastest at flaw tip and some authors found that steady state creep rate over temperature range 373 K-596 K and stresses of MPa was proportional to the 4 th power of stress.[2] [1] A.R. Causey, Anisotropy of irradiation creep of Zr-2.5 wt% Nband Zircaloy-2 Alloys, JNM, vol. 98, issues 3, June 1981, pages [2] N. Christodoulou, et al., Analysis of Steady-State Thermal Creep of Zr-2.5 NbPressure Tube Material, Metallurgical and Material Transactions A, Volume 33A, April,

29 Structural integrity assessment of P/T flaws Flaws in Pressure Tubes The flaws found during ISI of Zr-2.5%Nb CANDU P/Ts include: The fuel bundle bearing pad fretting flaws (BPF); The debris fretting flaws (DFF).

30 Structural integrity assessment of P/T flaws FEA of BPFF for stress analyses at flaw root Von Mises stress field by FEA at BPFF, h=0.4mm, p=12 MPa Hoop stress by FEA at BPFF, h=0.4mm, p=12 MPa

31 Structural integrity assessment of P/T flaws On-going research at INR: crack initiation from DHC and fatigue Radial-transversal section in P/T Zr-2.5%Nb, BPFF model, h=0.4 mm, hydrogen concentration about 100 ppm

32 Structural integrity assessment of P/T flaws DHC crack initiation Unirradiated: Failed Pre-irradiated: Failed FFSG Acceptance Criteria Unirradiated: Run-outs Pre-irradiated: Run-outs DHC crack initiation at a blunt flaw (DFF or BPFF) may be assessed in terms of a threshold peak flawtip stress σ th and the following inequality must be satisfied [1] Peak Stress (M Pa) σp< σth σ p with peak principal flaw-tip stress due to applied load [1] Number of Thermal Cycles Peak stress criterion for DHC initiation in P/T as function of thermal cycles [1] CAN/CSA N285.8, Technical Requirements for In-Service Evaluation of Zirconium Alloy Pressure Tubes in CANDU Reactors, 2005

33 Structural integrity assessment of P/T flaws Fatigue crack initiation Crack initiation by fatigue at blunt flaws in pressure tubes may result if the maximum alternating von Mises peak stress at the flaw tip exceeds the allowable alternating peak stress or if the number of load cycles exceeds the allowable cycles [1]. Case by case, also a cumulative usage factor should be evaluated in accordance with [1] and must be less than 1.0. Fatigue crack initiation evaluation curve for blunt flaws in irradiated cold worked Zr-2.5%Nb P/T material [1] [1] CAN/CSA N285.8, Technical Requirements for In-Service Evaluation of Zirconium Alloy Pressure Tubes in CANDU Reactors, 2005

34 Possible task: thermal creep modeling at flaws The principle that creep can relax stresses at flaws has been clearly demonstrated experimentally. The susceptibility of notched specimens of pressure tube material is significantly less if the notched specimens is held at temperature, under load, to allow creep to relax the notch-tip stress prior for testing DHC susceptibility. Some models for localized creep at the high stress regions near flaws are based on the limited data set and this certainly needs to be augmented. The model of creep at flaw-tip could improve the structural integrity assessments of flaws found by in-service inspections, in order to evaluate the crack initiation and growth due to DHC or fatigue phenomena.

35 Summary Few results of experimental works performed at INR Pitesti, concerning on thermal creep of Zr-2.5%Nb pressure tube has been outlined. The obtained results are in line with those from open literature and will be used in the application on the structural integrity of CANDU pressure tube. Concerning on the crack initiation and growth at the blunt flaws, the modeling of the thermal creep in high stresses spots, could play an important role in the structural integrity assessment of CANDU P/Ts by considering the stress relaxation process. By using of an appropriate stress exponent in the creep rate equation valid for high stress ranges, the assessments could give more reliable results with effect on plant operation and safety.

36 Thank you for your attention!