Effective Aging Management of Baffle-to-Former Bolts to Assure Long-Term Reliability of Reactor Vessel Internals

Transcription

1 Effective Aging Management of Baffle-to-Former Bolts to Assure Long-Term Reliability of Reactor Vessel Internals Tiangan Lian Kyle Amberge Electric Power Research Institute Fourth Nuclear Power Plant Life Management 23 to 27 October 2017; Lyon, France

2 Aging of Reactor Pressure Vessel Internal Components Reactor pressure vessel internal components in pressurized water reactors (PWRs) can be affected by age-related degradation effects. Irradiation induced degradations such as Irradiation-assisted stress corrosion cracking (IASCC) Embrittlement Irradiation creep Void swelling Stress corrosion cracking (SCC) Fatigue, including high cycle and environmental assisted fatigues As the PWR plants age, the likelihood of these degradation mechanisms occurring in the internals and structural attachments to the vessel wall increases. The implementation of aging management plans for the reactor internals can enhance long-term safety and reliability of PWRs 2

3 Overview of Pressure Water Reactor (PWR) Internals Upper Support Plate Upper Support Column Vessel Head Hold Down Spring Control Rod Guide Tube Inlet Nozzle Outlet Nozzle Upper Core Plate Baffle Plate Lower Core Plate Lower Support Column Body Bottom-Mounted Instrumentation Column Body Core Barrel Thermal Shield Pressure Vessel Former Plate Lower Core Support Plate Highly Irradiated Region Source: Westinghouse 3

4 EPRI Generic Aging Management Methodology 1. Determine the susceptibility or non-susceptibility of PWR internals to the postulated ageing mechanisms, with consideration of material properties and operational conditions. 2. Categorize these reactor internals ranging from insignificant effects to potentially moderately significant effects to potentially significant effects. 3. Assess functionality of components and assemblies of components based on representative plant designs using irradiated and aged material properties to determine the effects of the degradation mechanisms on functionality 4. Develop aging management strategy to determine the appropriate methodologies for maintaining the long-term functions of PWR internals safely combining the results of functionality assessment with component accessibility, operating experience, existing evaluations, and prior examination results 4

5 Overview of Industry Program (US) Document descriptions Focus: Safety, Dose, Reliability Approach: Define a proactive strategy using a systematic approach aligned with the USNRC s generic license renewal elements for internals MRP-227 I&E Guideline provides what and whenis to be inspected MRP-228 is an inspection standard that describes how to inspect WCAP NP is a guideline for acceptance criteria methodology MRP-318 gives the prioritization and contingency planning options for PWR internals WCAP gives the results for the reactor internals risk ranking and response planning WCAP P Revision 1 provides Control Rod Guide Tube (CRGT) guide card inspection frequencies Note:These guidelines do not reduce, alter, or otherwise affect current American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI or plant-specific licensing inservice inspection requirements. 5

6 Operation Experience of PWR Core Internals Following the existing inspection & evaluation guidelines, many PWR owners have performed full or partial inspection of reactor internals, during their 40 th year outages Issues of greatest concern have not been seen SCC of austenitic welded components not observed o Cracking currently limited to high strength bolting (X-750) IASCC of SS welds not observed, indications limited to bolts Macroscopic effect of void swelling not observed Baffle-to-former bolts are highly susceptible to Cracking Cracking has been observed in baffle-former bolts made of both Type 347 (in US) and Type 316 (in France) stainless steels. 6

7 Baffle-Former Assembly Details Core barrel, baffle and former plates Type 304 austenitic stainless steel material Baffle-Former Bolts (BFBs) Attach the baffle plates to the former plates in the reactor lower internals assembly Type 347 solution annealed or Type 316 cold worked austenitic stainless steel material Bolt head designs and shank lengths vary from plant-to-plant, which can challenge the NDE technique Baffle Baffle Plate Edge Bolt (Baffle-to-Baffle Bolts) Core Barrel Former Baffle-Former Bolt (Long & Short) Core Barrel to Former Bolt Corner Edge Bracket Baffle to Former Bolt Source: ML15331A179 Source: ML15331A179 7

8 Operating Experience of Baffle-to-Former Bolt Operating Experience First UT baffle-former bolts (BFB) inspections in French PWR CP0 units and first cracks found First degraded baffle-former bolts found in U.S. Ginnaperforms first MRP-227 inspections DC Cook2 finds degraded bolts by visual inspection Indian Point 2, Salem 1, DCCook2 find degraded bolts (visual+ut) WCAP-13266: BFB Program for the Westinghouse Owners Group - Plant Categorization Guidance (US) NRC Information Notice on BFBs MRP publishes assessment of French BFB OE (MRP-03) MRP publishes Reactor Internals Inspection Guidelines (MRP-227) NRC reviews & approves MRP-227 NSAL-16-1 AREVA CSB Interim Guidance Westinghouse Technical Bulletin TB-12-5, related to the DC Cook OE Per EPRI Guideline (MRP-227-A), BFB UT exam is performed for WEC plants initially at 25-35EFPY and repeated every 10-years Note: UT deployed as it became available and qualified for the various sites Note: UT deployed as it becomes available and qualified for the various sites 8

9 Baffle-to-Former Bolt Cracking in US and France US PWRs with BFB Cracking Observed French PWRs with BFB Cracking Observed Plant Type 4-loop Westinghouse 3-loop CP0 design Core flow Down flow Up flow converted Bolt cooling hole Without bolt cooling hole Without bolt cooling hole Head-to-shank Circular Circular geometry Bolt Materials Type 347SS Type 316SS Summary The BFBs are more susceptible to degradation in older Westinghouse 4-loop reactors that have a "downflow configuration and have Type 347SS baffle former bolts without cooling holes. The BFBs are more susceptible to degradation in older 3-loop CP0 reactors that have a converted to upflow (or previously downflow) configuration and have Type 316SS baffle former bolts without cooling holes. 9

10 Potential Risk of Baffle-to-Former Bolt Degradation In the memorandum issued on October 20, 2016, US NRC concludes that: -- BFB degradation does not affect the containment function. The effect on cladding is small and limited to peripheral fuel assemblies. The RCS would remain intact except possibly for a LOCA or large seismic event, both of which are very unlikely. The BFB degradation issue does not indicate that an imminent safety concern exists from any effect on defense in depth, because it does not significantly affect the four layers of defense nor the containment fission product barrier. To date, no displacement of baffle plates has been observed, even in plates with a large majority of degraded bolts. BFB cracking has the following potential consequences: Excessive fuel grid impact loading from baffle plate detachment or deflection Baffle jetting Increase bypass flow Fatigue Loose parts Source: ML15331A179 10

11 Understand the Conditions Leading to Bolt Failure Design characteristics Power history and fluence Thermohydrolic loading Temperatures Transients Radiation induced segregation at grain boundaries Radiation induced microstructure Localized deformation under loading Irradiation creep and transmutations What is the bolt material condition? What is the correlation between material condition and failure process? What was the condition of the bolt at the onset of failure? Which mechanisms contributed to the failure of the bolts? 11

12 Investigation of Cracked Bolts What Mechanisms Contributed to the Failure of the Bolts? IASCC IASCC was seen on the fracture surface of bolts that were removed in two pieces In some cases, IASCC is only a few grains deep Transgranular Cracking TG and mixed-mode cracking is apparent on many fracture surfaces Fatigue In some regions, fatigue striations can be seen Ductile Overload The fracture of the remaining ligament shows retained ductility on most bolts consistent with tensile/pull test results 12

13 Elements of Baffle-to-Former Bolts Management Program Engineering Analysis and Evaluation Acceptable Bolting Pattern Analysis BFB degradation prediction (inspection requirement or expansion evaluation) Inspection Follow industry guideline and best practices Mitigation Upflow conversion Reduce loading and anti-clustering through bolt replacement Increase inspection frequency or scope Monitoring Operating experience and trends Inspection results of lead samples Replacements Replacement as needed based on inspection results Pre-emptive replacement and conversion to upflow configuration 13

14 Replacement Options Partial bolt replacement o Replacement of bolts with indications o Replacement to acceptable pattern o Replacement to support specific inspection interval Full bolt replacement ( 90% of the full population) Replacement staggered over multiple outages, can be applicable to full or partial replacements Replacement of additional anti-cluster bolts (1-in-3 or 1-in-5 groups) o Concentrated in high-stress and/or high-fluence regions based on analysis Replacement with alternate bolt design Full lower internals replacement 14

15 R&D Needs Improve the understanding of BFB cracking mechanisms Better understanding of IASCC mechanisms Root cause analysis on failed bolts Develop new and more reliable inspection techniques Accessibility and tooling issues Enhanced inspection techniques with better probability of detection (POD) New techniques for specific degradation modes Increase options for mitigation, repair & replacement Limited methodologies and equipment available Improve the knowledge in irradiated materials properties Availability and fidelity of irradiated materials properties important to engineering analyses and degradation prediction 15

16 Summary Baffle-to-former bolt cracking has been the main operating experience in aging of PWR reactor internals over the past two years Industry has effective program to manage baffle-former bolt aging, within the overall PWR Reactor Internal Aging Management program Baffle-former bolt degradation is not a significant safety issue for PWR plants, based on: Risk-informed evaluation The industry inspection guidance The actual baffle-former bolt degradation findings Structural analysis to date Industry and regulatory authorities continue to work together To understand baffle-former bolt degradation, and To ensure no challenges to safe operation in the longer term 16

17 Together4Shaping the Future of Electricity 17