Irradiated Material Welding Solutions: Potential Applications of Welding to Repair Irradiated Reactor Internals

Transcription

1 Irradiated Material Welding Solutions: Potential Applications of Welding to Repair Irradiated Reactor Internals Jon Tatman EPRI WRTC Wayne Lunceford EPRI BWRVIP Greg Frederick EPRI WRTC

2 Background - Helium-Induced Cracking Mechanism He Generation from B, Ni Reactions He Migration and Coalescence at Grain Boundaries During Welding Helium-Induced Cracking (Asano, K., JNM, 1999) Helium is generated by neutron transmutation reactions in components exposed to significant neutron fluence When welded, helium can migrate to grain boundaries and coalesce into bubbles that weaken grain boundaries Upon cooling, increased stresses along grain boundaries result in helium-induced cracking 2

3 Background - Industry State of Knowledge Observed cracking of BWR reactor internals resulted in EPRI Boiling Water Reactor Vessel and Internals Project (BWRVIP) program to develop guidance for welding on irradiated materials Initial guidance published as BWRVIP-97 in 2001 BWRVIP-97-A approved by NRC in 2011 Additional data obtained by EPRI in 2006 from studies performed in Japan and documented in BWRVIP-151, Technical Basis for Revision to BWRVIP-97 Welding Guidelines 3

4 Current Work Refinement of weldability threshold curves for 304SS and 316LSS Use of effective heat input estimation vs. theoretical heat input Review of available weldability trials to more accurately represent heat inputs Review state of knowledge for low-alloy steels and nickel-base alloys Thermal fluence calculations and He generation studies for typical B&W, Combustion Engineering and Westinghouse PWR as well as a GE designed BWR (BWR/4) 4

5 Theoretical Versus Effective Heat Input Equation Theoretical heat input equation ineffective at accurately calculating weld heat input transferred to the substrate for consumable welding processes Effective weld heat input equation has been developed to provide more accurate means to quantify heat provided to the substrate from the welding process Accounts for filler metal materials properties, arc heat used to melt filler material, and hot wire heat Equation can be tailored for all consumable welding processes Equations have been developed for GTAW and conduction-limited LBW to-date Effective Heat Input (LBW) ASME Theoretical Heat Input ( J in ) = P WP TS Q eff kj cm = μ T P WP TS τ WFS (A wire ) TS Q eff m T P WP TS Effective heat input (kj/cm) Heat transfer efficiency factor of the arc / laser beam during welding Time weighted average power input for the weld process (watts) Time weighted average weld process travel speed (cm/min) t Heat required to melt a volume of added filler metal (kj/cm 3 ) WFS Time weighted average weld process wire feed speed (cm/min) A wire Cross-sectional area of wire (cm 2 ) 5

6 Irradiated Stainless Steel Weldability - Experimental Data Irradiated weldability data originates from two Japan Nuclear Energy Safety (JNES) irradiated welding studies and one Japan Owners Group (JOG) study Gas tungsten arc welding (GTAW) and laser beam welding (LBW) processes used in JNES studies GTAW process used in JOG study Single pass and multi-pass welds performed on irradiated 304 and 316L austenitic stainless steel base materials and 308L and 316L filler materials A number of test welds in the JNES studies were performed autogenously (i.e., without filler metal addition) Termed laser surface modification (LSM) trials Trials involved performing one to three repetitive melting passes over the same area of substrate Repetitive thermal cycling of the HAZ region resulted in additional He migration and bubble growth Additional heat input from repetitive thermal cycles was accounted for by multiplying the calculated heat input by the total number of thermal cycles 6

7 Weldability of Irradiated 304 Stainless Steel *GBD Grain Boundary Deterioration *HIC Helium Induced Cracking 7

8 Weldability of Irradiated 316L Stainless Steel *GBD Grain Boundary Deterioration *HIC Helium Induced Cracking 8

9 Irradiated Stainless Steel Weldability Mechanical Testing Several mechanical test methods (i.e., side bend, tensile, micro-hardness) were performed as a part of the JNES irradiated welding studies All side bend and tensile tests were performed in compliance with Japan Industrial Standards (JIS) Z 3122 and Z 3121, respectively Comparable to ASME Section IX requirements Pre-bend metallurgical analysis of a few test specimens revealed HIC microcracking in the HAZ, however, all specimens still passed the side bend test examination In all cases, the ultimate tensile strength (UTS) value obtained from the tensile tests exceeded the minimum requirement specified per the JIS specification Micro-hardness results indicate that welding operations on irradiated austenitic material do not produce detrimental hardening to the microstructure Heat from welding annealed the irradiated base material, resulting in HAZ hardness reduction 9

10 Irradiated Ferritic Low-Alloy Steel Weldability Japanese GTAW and LBW studies 1,2 on irradiated low alloy steel did not indicate significant susceptibility to helium-induced cracking Increased porosity was noted in the LAS weld metal along the fusion boundary (weld process issue), however sound welds were achievable on irradiated LAS material No evidence of helium-induced cracking or grain boundary deterioration even at significant helium concentrations Toughness reduction is the more important concern for welding on low-alloy steels Very few locations for which welded repairs may be needed for irradiated ferritic materials Materials have performed well in service to date Absent cracking in adjacent stainless steel or nickel-base alloys, welding on irradiated lowalloy steel is unlikely Limited number of relevant locations (i.e., BWR riser brace attachment welds, PWR core support / core stop lugs) 1. K. Kazuhiko, et al., Repair Welding of Irradiated Reactor Vessel Steel by Low Heat Input GTAW and LBW, Materials Science Forum Vols (2007), pp FY2003 Safe Maintenance/Repair Welding Techniques for Nuclear Plant Irradiated Material (WIM), June 2004, Japan Nuclear Energy Safety Organization. 10

11 Conclusions Regarding Irradiated Material Weldability Weldability related to He concentration and weld heat input At He concentrations relevant to repair by welding, the dominant contributor to He generation from the 10 B + n transmutation reaction Weldability threshold plots developed for 304SS and 316LSS At He concentrations < 0.1 appm, both 304SS and 316SS are weldable without consideration of irradiation effects 304SS observed to be somewhat less susceptible to He-induced cracking than 316LSS Mechanical testing results revealed no notable detrimental effects to the weld or base metal microstructure Insufficient data to develop similar threshold plots for either nickel-base alloys or low-alloy steels Available nickel-base alloy data suggest weldability similar to austenitic stainless steels Available low-alloy steel data suggest weldability at least as good as austenitic stainless steels 11

12 BWR Helium Generation Mapping Parametric helium generation studies based on a single General Electric design BWR/4 251-inch RPV Based on extension of the BWRVIP RAMA model to include thermal neutron energies Three power cases investigated: original licensed thermal power, minor power uprate, major power uprate (EPU) Reasonable upper end estimates of initial boron content (20 wppm B for stainless steel, 50 wppm B for nickel-base alloys) 40, 60 and 80 EFPY cases Further Information: BWRVIP-97 Revision 1 12

13 BWR Helium Generation Results (80 EFPY, 50 wppm initial B) Core Shroud He Concentration Map BWR Core Shroud Helium Generation at Core Shroud Welds H1, H6a and H6b (Major Power Uprate, 80 EFPY Case) 13

14 BWR Helium Generation Study Conclusions Many locations in the beltline region remain weldable at 60 or even 80 EFPY Low fluence core shroud azimuths Most components located in the annulus region Azimuthal variation in He concentration is significant in the beltline region and must be considered Steep gradient in He generation observed below the core Core support and lower plenum structures not susceptible to He induced cracking He-Induced cracking not a concern for most all RPV locations of interest Jet Pump riser brace attachment welds located on high fluence azimuths could have He concentrations above 0.1 appm All other RPV attachment weld locations predicted to have He concentrations well below 0.1 appm He (appm) Categories for BWR RPV Attachment Welds Attachment Weld Location Low Fluence Azimuth High Fluence Azimuth Steam Dryer Lugs < 0.01 < 0.01 Core Spray Piping Brackets < 0.01 < 0.01 Jet Pump Riser Brace < 0.1 < 1.0 Shroud Support Weld (H9) < 0.01 <

15 PWR Helium Generation Mapping Parametric helium generation studies for example reactor configurations Based on OEM developed fluence models and including thermal neutron energy groups Three typical reactor configurations modeled (B&W, Combustion Engineering, and Westinghouse reactors) Multiple fuel management strategies based on assumed timing of plant transitions to low leakage or very low leakage core designs Conservative upper end estimates of initial boron content (20 to 75 wppm B) 40 and 60 EFPY Further Information: Westinghouse reactor evaluation documented in MRP-319 Rev. 1 Combustion Engineering reactor evaluation documented in MRP-346 B&W reactor evaluation documented in MRP

16 PWR Helium Generation Study Results (60 EFPY, 75 wppm initial B) B&W Reactor Westinghouse Reactor CE Reactor 16

17 PWR Helium Generation Study Conclusions Many internals locations above and below the core remain weldable regardless of accumulated EFPY or initial boron content Weldability thresholds identified above and below core relatively insensitive to accumulated EFPY Bounding initial B content assumptions can be used to define generic weldability boundaries Core guide / stop lug peak He concentration estimates range from < 0.1 appm to ~ 5 appm depending on input assumptions, azimuth, and reactor design CE Reactor, Core Centerline 17

18 Weldability Categorization (Applicable for either BWRs or PWRs) Based on observations from weldability studies and helium mapping, four weldability categories were defined: 1. [He] > 10 appm Limited weldability data exist, although it appears that 304SS may be weldable by low heat input welding processes such as LBW appm < [He] 10 appm Weldable with appropriate evaluation of helium content and application of low heat input welding processes appm < [He] 0.1 appm 0.1 appm is below the lowest He concentration found to induce helium-induced cracking by conventional welding. If the location is determined to have a He concentration 0.1 appm, welding can be performed without consideration of irradiation effects. 4. [He] 0.01 appm In regions of the reactor system where generic bounding evaluations show He concentration < 0.01 appm, welding should be permitted without the need for a detailed He generation evaluation 18

19 Key Conclusions Regarding Weldability Thresholds There are substantial data that can be applied to characterize the weldability of austenitic materials in BWRs and PWRs Austenitic materials are weldable without consideration of He-induced cracking at He concentrations less than 0.1 appm Ferritic materials appear at least as weldable as austenitic materials with regard to susceptibility to He-induced cracking Under this current work, refined weldability threshold plots were developed for 304SS and 316LSS as a function of effective heat input and He concentration Work is in progress in collaboration with ORNL to improve the fidelity of the austenitic stainless steel threshold plots and to increase the database for nickel-base alloys with the objective of creating a nickel-base alloy threshold plot Materials currently planned for irradiated weld testing are 304L, 316L, and Alloy 82 Materials are being irradiated at the ORNL High Flux Isotope Reactor (HFIR) facility to achieve helium concentrations ranging between 1 appm and 30 appm 19

20 Key Conclusions from Helium Generation Studies Initial Boron concentration and reactor operating time have relatively small influence on the weldable zones above and below the core Reactor internals components located in regions above and below the core will be weldable in many cases, even after 80 years of service Many reactor internals components can and should be generically dispositioned as not requiring consideration of helium generation or helium-induced cracking in association with welded repairs Most BWR components located in the annulus region will be weldable, even after 80 years of service By extrapolation, it can be concluded that He-induced cracking is not a plausible concern for most locations of interest with regard to RPV welded repairs including: Upper and lower head penetrations PWR inlet / outlet nozzle to safe end welds BWR Feedwater, Core Spray, and Recirculation inlet / outlet nozzle to safe end welds BWR RPV ID attachment weld locations other than jet pump riser brace attachment welds 20

21 Together Shaping the Future of Electricity 21