Reactor Internals Overview

Transcription

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2 Reactor Internals Overview Mechanisms: Cracking due to Irradiation Assisted Stress Corrosion (IASCC) and Stress Corrosion (SCC) Reduction of Fracture Toughness due to Irradiation Embrittlement (IE) and Thermal Embrittlement (TE) Dimensional Changes due to Void Swelling (VS) Loss of Mechanical Closure Integrity due to Stress Relaxation (SR) Synergistic Effects of These Mechanisms Internals.ppt 2 2

3 B&W Typical Internals Layout Core Support Assembly Plenum Cover Assembly Vent Valves Core Support Shield Assembly Plenum Assembly Control Rod Guide Tube Assembly Upper Grid Assembly Thermal Shield Core Barrel Assembly Core (not in scope) Core Barrel Baffle Plates Former Plates Lower Internals Assembly Lower Grid Flow Distributor Head Incore Guide Tubes Internals.ppt 3 3

4 Typical CE Internals Layout Internals.ppt 4 4

5 Westinghouse Internals with Forged Lower Support Internals.ppt 5 5

6 Westinghouse Internals with Cast Lower Support Internals.ppt 6 6

7 Reactor Internals Lower Section Internals.ppt 7 7

8 PWR Core Internals The baffle-former assembly or shroud surrounds the core and separates the coolant down-flow from up-flow Constructed primarily from annealed 304 stainless steel Can be of either bolted or welded construction Internals.ppt 8 Baffle plates, especially at reentrant corners, will receive doses as large as 100 dpa over a 40 year lifetime

9 Baffle-former-barrel Assembly in a Typical Westinghouse PWR baffle baffle bolt core barrel Internals.ppt 9 former

10 Typical Baffle Former Assembly Internals.ppt 10 10

11 Reactor Internals Upper Section Internals.ppt 11 11

12 Component List Screening Criteria Categorization Below Screening No Credible Damage Issue Initial Screening Above Screening Probability & Consequence Analysis Moderate Resolved by Analysis Category B High Category C Functionality Analysis Analysis Aging Management Strategy Cat. A No Adverse Effects Existing Guidelines Existing Subordinate Principal New Recommendations Aging Management Program Aging Management Monitoring & Trending I&E Guidelines Internals.ppt 12 12

13 Approach for Evaluating Functionality Analysis for I&E Guidelines What? Damage mechanisms of concern? Metrics used to characterize a damage mechanism? Observable effects/consequences on functionality? Where? Location of degradation? When? Estimate the likelihood and timing of future damage? How? Inspection, monitoring or trending technique Task is to utilize representative plant results and apply to entire fleet Internals.ppt 13 13

14 The Cracking Mechanisms SCC IASCC Fatigue Produce observable cracks Most probable in regions of stress concentration Expect to manage through an integrated inspection program. Internals.ppt 14 14

15 Stress Corrosion Cracking (SCC) Austenitic stainless steel No experience with SCC in 300 series stainless steel under normal primary water conditions No model to evaluate or rank potential for SCC Large structural welds identified due to large potential residual stresses X-750 Programs for guide tube support pins in place Cast austenitic stainless steel Verify that specifications meet minimum ferrite requirements Internals.ppt 15 15

16 Irradiation-Assisted SCC (IASCC) Stainless steel bolts (316 SS) FEA intended to provide basis for ranking of time to failure Limited number of CE plants with bolted baffles Stainless steel plate (304 SS) CE shroud welds included in plate waterfall Will identify locations with IASCC susceptibility from FEA Eliminated components associated with Westinghouse lower core plate on basis of completed analysis Internals.ppt 16 16

17 Parameters Influencing IASCC Fluence IASCC in PWRs occurs above a threshold fluence of ~ 2 x n/cm 2, E > 1 MeV This fluence level is higher than in BWRs by about an order of magnitude The threshold fluence level does not correlate directly with the onset or saturation of radiation-induced materials changes such as grain boundary segregation or hardening Start of BWR IASCC Start of PWR IASCC BWR End of Life Dose *Max PWR End of Life Dose Neutron/cm 2 (E 1 MeV) dpa** Start of Start of Saturation Possible Grain Boundary Ductility of Sensitization Start of Sensitization Loss and Ductility Swelling Loss * For 32 EFPY ** Based on 15 dpa = n/cm² E 1 MeV Internals.ppt 17 17

18 Fatigue Expect that fatigue evaluation will be required to justify extended life Real vs. assumed stress history Realistic stress/strain amplitudes Potential environmental effects Two main groupings Additional evaluation required Addressed via SCC, IASCC, etc. Internals.ppt 18 18

19 The Embrittlement Mechanisms Irradiation embrittlement Thermal embrittlement Changes in material properties Strength (increase) Ductility (decrease) Toughness (decrease) Expect to manage through an industry trending program Internals.ppt 19 19

20 Irradiation Embrittlement Industry trend curves for strength and ductility are embedded in computer codes Westinghouse lower core plate Westinghouse baffle-former-barrel CE core shroud Extrapolate to remaining components based on fluence and temperature Fracture toughness estimates required for components with active cracking mechanisms Internals.ppt 20 20

21 Thermal Embrittlement Evaluate composition and temperature to determine susceptibility to thermal embrittlement Fracture toughness estimate required if there is an active cracking mechanism Internals.ppt 21 21

22 Dimensional Stability Mechanisms Void swelling Irradiation induced stress relaxation/creep Component distortion Modify stress/strain distribution Affects SCC, IASCC and fatigue Expect to manage through industry trending and inspecting for distortion Internals.ppt 22 22

23 Void Swelling FEA analysis provide ranking based on swelling model in computer codes Westinghouse baffle-former-barrel Westinghouse lower core plate CE core shroud Components not included in FEA that can be easily compared to analyzed components Westinghouse lower core support structure CE baffle bolts Internals.ppt 23 23

24 Irradiation Induced Stress Relaxation/ Creep FEA model incorporates stress relaxation and creep effects (can rank effect) Stress relaxation may have significant impact on other stress related mechanisms (e.g., IASCC) Loss of bolt preload must be considered as contributing to wear and fatigue waterfalls Internals.ppt 24 24

25 Wear Mechanism Difficult to compare or rank wear potential in identified components Match inspection/trending monitoring program to component requirements Internals.ppt 25 25

26 Wear Existing wear management programs Westinghouse flux thimbles and tubes CE thermal shield positioning pins CE In-core Instrumentation thimble tubes Monitored through control rod drop times Inspect & monitor neutron noise Inspection requirements combined with integrated crack monitoring programs Internals.ppt 26 26

27 What is a Reactor Internals Aging Management Program (AMP)? A document (procedure, instruction, specification) that describes a plant s program to ensure the long-term integrity and safe operation of PWR internal components Why is it required? Previously required only for plants applying for license renewal With publication of MRP-227A, now required for all plants (Mandatory requirement under NEI 03-08) Internals.ppt 27 27

28 Contents of an AMP What are the required contents of an AMP? MRP-227A, Appendix A defines the 10 elements which constitute an acceptable AMP These elements are from NUREG-1801 (Generic Aging Lessons Learned [GALL] Report) Internals.ppt 28 28

29 Requirements for Reactor Internals AMPs GALL report NUREG-1801, PWR Vessel Internals identified 10 Attributes/Elements necessary for the Evaluation and Technical Basis 1. Scope of Program 2. Preventive Actions 3. Parameters Monitored/Inspected 4. Detection of Aging Effects 5. Monitoring and Trending 6. Acceptance Criteria 7. Corrective Actions 8. Confirmation Process 9. Administrative Controls 10. Operating Experience Internals.ppt 29 29

30 First MRP-227 Inspections Started in 2011 First plant was Ginna in May 2011 Baffle-former bolts Baffle-former assembly Baffle former edge bolts Lower guide tube flange weld Upper core barrel flange to shell weld Thermal shield flexures Control rod guide cards Type X-750 split pins replaced using coldworked 316 stainless steel Internals.ppt 30 30

31 Ginna Inspection Internals.ppt 31 Alignment and interfacing for internals hold-down spring; at Ginna spring material is 410 stainless steel, therefore, only EVT-1 scan on the external surface was needed In addition to MRP-227 requirements, ASME Code Section XI ISI examinations require visual exam of lower core plate and fuel pins Only problem area was baffle-former bolts 31

32 Ginna Baffle-Former Bolt Inspections Proactive inspection in 1999 found 59 bolts with indications (347 SS) 56 replaced (316 SS) and 5 broke when extracted Original plan for 2011 was to replace 126 old bolts Big problem bolts could not easily be removed and replaced (25 of 28 successful) Shifted to UT inspection approach of 98 bolts Internals.ppt 32 32

33 Year License Renewal Reactor Internals MRP 227-A Inspections Internals.ppt 33

34 Reactor Vessel Internals Aging Managing Program Pure Water Stress Corrosion Cracking PWSCC Irradiated Assisted Stress Corrosion Cracking IASCC Stress Corrosion Cracking SCC Fatigue Irradiation Enhanced Stress Relaxation SR CASS TE/IE CASS Irradiation Embrittlement IE Irradiated Induced Void Swelling VS Wear Use of Material Data, Fluence and Generic Analyses to Screen and Select Most-Affected Internals Components Baffle-Bolts Welds Thermal Shield & Core Barrel Bolts Bolted Joints Baffle-Former Plates and Bolts CASS Components (Synergistic Effect) Internals Components (Increased Strength & Loss of Ductility) Guide Cards Guide Lugs Use Plant-Specific Analyses to Determine Critical Locations, Critical Crack Sizes and Flaw Tolerance for Actual Stresses, Fluences and Temperatures UT Inspections MRP 227-A Inspection Guidelines Enhance VT-1 Inspections Embrittlement Analysis with Stress and Deflection Due to LOCA and SSE Dimensional Changes and Functionality Improved Inspection Techniques Proposed After Functionality Assessment. 1. Scope 6. Acceptance Criteria 2. Preventive Actions 7. Corrective Action 3. Parameters Monitored 8. Confirmation Process 4. Detection of Aging Effects 9. Administrative Controls 5. Monitoring and Trending 10. Operation Experience 1. Scope 6. Acceptance Criteria 2. Preventive Actions 7. Corrective Action 3. Parameters Monitored 8. Confirmation Process 4. Detection of Aging Effects 9. Administrative Controls 5. Monitoring and Trending 10. Operation Experience Integrated Baffle Bolt Microstructural Evaluation Possible Lead Irradiation of SS at Bounding Temp? Meet Current License Basis Meet Current License Basis Meet Current License Basis Internals.ppt 34