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ENT00714B - H. Cothron Et Al., EPRI, Slides: Nrc/Industry Meeting Regarding Tube-to-Tubesheet Weld and Divider Plate Cracking Report (July 30, 2015)
	ML15222B160
Person / Time
Site:	Indian Point
Issue date:	07/30/2015
From:	; Entergy Nuclear Operations
To:	; Atomic Safety and Licensing Board Panel
	SECY RAS
References
	RAS 28139, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
	Download: ML15222B160 (30)
	v • d • e

ENT00714B Submitted: August 10, 2015 NRC Questions /Industry Responses 32 t=~~~ ~ ELECTR IC POWER

Industry's Position for Aging Management Programs for Steam Generator Channel Head Components

Summarize the industry's position/plans for aging management program for these components Discussion

SGMP has communicated the following

- Technical Report 3002002850 can be used to update Aging Management Plans

Visual inspections of present rigor are adequate to manage the Alloy 600 material in the divider plate assembly

- EPRI SGMP Chemistry Guidelines 33

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Data Removed from the Tubesheet Weld Database

Some data were removed from the database because of the tungsten weld head positioning. Why were the data removed? What is the sensitivity of the results if these data were included?

Discussion

The welds were removed from the database to avoid introducing a bias that does not represent field welds.

The tungsten positioning in the weld head was purposely offset to explore positioning effects; how~ver, with the offset applied in some welds exceed the limits where an acceptable weld could be made- welds that would have been rejected by visual inspection.

- Welding equipment is set-up on test welds to avoid these extremes E1

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SG Channel Head Inspections

What types of inspections are being referred to in Section 7.3? Are these inspections in accordance with the ASME Code or industry guidelines? Please discuss.

Discussion

Currently, utilities in the US perform visual inspection of the channel head cladding, tubesheet, and all installed plugs

- Not an ASME Code inspection

- EPRI Technical Report, 3002000473, "Steam Generator Channel Head Degradation Failure Modes and Effects Analysis"

- Westinghouse Nuclear Safety Advisory Letter 12-1 recommended visual inspections

- SGMP sent Nuclear Safety Advisory Letter to all members via email

- NRC Information Letter IN 2013-20

- Inspection recommendations are included in Revision 4 of the Integrity Assessment Guidelines draft to be published 2016 E1

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Inspections from Outside the SGs

Summarize the inspection results for the Z-weld (tubesheet to shell weld). Given the closeness of these inspections to the triple point, if something unexpected were to occur, would these inspections possibly give an opportunity to detect cracking were it to occur (e.g., detect leakage or boric acid residue during the performance of the inspection if a throughwall flaw had occurred as a result of divider plate cracking)?

Discussion

One Z-weld is required to be inspected every 10 years per ASME Section XI

- This inspection would identify leakage in the SG being examined

The area is part of routine boric acid walk downs

- These inspections would most likely identify leakage

Research conducted on full scale mockups on reactor vessel bottom mounted nozzles indicates that significant boric acid deposits can result from relatively small leaks

Research results provide a qualitative reference to support the conclusion that a leak from the channel head, even at levels below safety significance, would most likely produce boric acid residue that would be detectable through the insulation.

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Example of Boric Acid Leakage from Reactor Vessel Bottom Mounted Nozzle Full Scale Mockups 0.01gpm 37 t=~121 1 ELECTRIC POWER

NEI 03-08 Requirement for RCS Leakage Monitoring

Leakage from the steam generator channel head would most likely be detected from "online" leakage monitoring.

- The NEI 03-08 "needed" requirements detailed in PWR Owners Group letter OG-08-0400 establishes a required online leak detection limit that should provide ample margin for a plant to detect leakage from the steam generator channel head

- Utility procedures must include action levels and responses to increasing levels of RCS leakage E1

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Applicability of Conclusions Regarding Design Credit for the Divider Plate

Previously, the stresses within the tubesheet were assessed for cracks in the divider plate. For some steam generators, we believe no credit was given for the presence of the divider plate during the original design. For other steam generators (primari ly replacements), some credit may have been given for the divider plates in reducing stresses within the tubesheet. Please re-summarize the analyses performed and their results.

Discussion Follows 39

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Applicability of Conclusions Regarding Design Credit for the Divider Plate

Westinghouse Model 51 was determined to be the bounding design for Westinghouse steam generators

- An investigation in 2008, EPRI Technical Report 1016552, concluded that Westinghouse-designed SGs were the only design susceptible to divider plate cracking

The divider plate function is to divide the channel head into hot and cold legs. It is not part of the primary pressure boundary

Documents reviewed with respect to impact of a degraded divider plate

- ASME Stress Reports

- LOCA and non-LOCA transients

- SG tube plugging and sleeving qualifications

- Accident analysis

- Alternate repair criteria

Conclusion:

A crack running the full length and depth of the weld is not a safety concern 40

Applicability of Conclusions Regarding Design Credit for the Divider Plate

The conclusions of the report bounds the Westinghouse steam generator fleet

AREVA and B&W have indicated that their replacement steam generators are not susceptible to divider plate cracking due to materials and design

- Report not applicable E1

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Operating Experience with Alloy 82

Please clarify the operating experience with Alloy 82 with respect to its susceptibility to cracking. The NRC staff is under the impression that welds which have exhibited cracking may have been fabricated using both Alloy 82 and Alloy 182, and hence it cannot be conclusively stated that cracks have not occurred in welds fabricated from Alloy 82.

Discussion Follows 42 t=~~~ ~ ELECTRIC POWER

Weld Design Information

Butt weld design drawings normally show "82/182" to designate the filler metal

- Permits the use of either GTAW (82) or SMAW (182) processes

- Since the welds were applied manually (more recently machine welds have been used), the weld root and hot pass normally will be applied using the GTAW process, then the process is switched to SMAW to fill the cavity.

- The design permits GTAW to complete the weld but the welding efficiency is much lower and production clearly suggests using SMAW for the fill. The Cr content of 82 is approximately 20% Cr while 182 is approximately 14o/oCr. This difference in Cr content is why 82 is more resistant to PWSCC.

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Weld Design Information

Alloy 600 steam generator channel heads use this type of butt weld for the stub runner to divider plate and the stub runner to tubesheet weld buildup.

The same SMAW procedure is used to join the divider plate to the channel head. Thus the lnconel surfaces exposed to the coolant will be 182 weld deposits and the Alloy 600 divider plate itself.

SMAW electrodes (182) are used to restore cladding around the triple point because of the need to weld over lnconel and also and to tie into the austenitic stainless steel cladding applied to the channel head surfaces

The hot and cold leg nozzles are installed in a similar progression. Small nozzles can be an exception where GTAW may be used to install the entire nozzle (e.g. drain nozzle) 44 ~~~~ ~ ELECTRIC POWER *

Operating Experience with Alloy 82

There is one report of a crack in an inlet nozzle location where a running crack initiating in Alloy 182 intersected a small portion of the Alloy 82 material

- Example of crack propagation, not initiation

Alloy 82 weld deposits have a long history of resistance to SCC initiation

- Butt welds used with lnconel 600 safe ends used Alloy 82 for the root and hot pass using GTAW (for both the Alloy 182 buttered end of the nozzle and the stainless steel pipe on the opposite end of the safe end. The butt joint was completed using Alloy 182 and the SMAW process.

- BWR environments are generally accepted to be more aggressive than PWR environments.

- IGSCC of these butt welds would begin either in the nozzle buttered region or the sensitized HAZ of the stainless steel.

- As IGSCC progressed it cracked around the Alloy 82 instead of initiating in it or propagating through it.

- It is acknowledged that laboratory studies can be set up to crack Alloy 82 deposits, but it is very difficult to initiate such cracking either in BWR or PWR environments. This is largely due to the Cr content. Theoretically, Alloy 82 will crack under the right conditions, but realistically, operating experience of many years suggests excellent resistance to PWSCC and IGSCC on these environments.

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Normal Cladding vs Structural Butter

Please clarify the difference between normal cladding and a structural butter (refer to page 2-3).

Discussion Follows

Cladding and structural butter differ in two ways

- Code qualification is different for the two applications

- The purpose of the two applications is different

Structural butter is required to be capable of transferring loads and as such the weld qualification is tested to demonstrate that capability.

Cladding is used as a corrosion barrier and the requirement is to be corrosion resistant and to be bonded to the surface.

Thus the testing is for deposit chemistry and to demonstrate bond strength so it will not delaminate 46

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Utility Mockup

On page 2-19, there is reference to a mockup from another utility. Please discuss whether inclusion of this data would affect the conclusions of this report.

Discussion

Several tube-to-tube sheet welds were removed from a vendor mockup that was used for NDE testing.

The removed section was quartered then polished and etched to document that the test welds performed by EPRI produced representative weld profiles as obtained as produced by a vendor.

The vendor mockup welds were not included in the database because of a lack of specific records for the materials used.

This avoided the inclusion of data that might bias the dilution distributions.

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EdF Experience

lntergranular stress corrosion cracking was reported on both the hot- and cold-leg sides of the channel head at Dam pierre. Please clarify where this cracking occurred on the cold-leg (in the tubes, in the divider plate, etc.). Refer to page 3-22.

Discussion

Both hot and cold legs of the Dam pierre original steam generator #27 retired from loop 3 were found to display intergranular SCC. A large section was removed of the cracked locations and studied extensively in the laboratory.

Shallow surface intergranular cracking was observed along the top and bottom surfaces of the stub plate adjacent to the base metal on both sides of the divider plate. Some interdendritic crack paths were observed along the welds joining them to the divider plate and the tube sheet.

- Cracking was shallow in all cases and appeared to be restricted to a thin surface layer believed to be associated with cold working (grinding) during manufacturing.

It was noted that cracking was intermittent but extended over a significant length.

- The studies assumed that crack initiation in welds might be linked to some manufacturing aspects (grinding or arc striking ... ).

Tubes were not affected

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EdF Inspections Since 2002

EdF committed to inspect and has not challenged the commitment

- They do not view this as a safety concern

- Performed inspections since 2002 on divider plate for 900MW plants -

Alloy 600 51 Bs thin divider plate (1 .3")

Identified SCC indications

- Performed inspections since 2002 on divider plate for 1300MW plants

- Alloy 600 thicker divider plate

1 steam generator with small defects associated with markings welded on the stub

- Not related to 51 B experience or any of the scenarios of the final report

- Performed inspections since 2002 on divider plate for 1400MW plants

- Alloy 690 material

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Model 51 Bounding Steam Generator

This report potentially applies to all steam generators. On page 4-3, it was indicated that the model 51 steam generator was considered the most limiting steam generator. Given the current fleet of steam generators in operation in the U.S.,

is this still the case?

Discussion

The conclusions of the report bounds the susceptible Westinghouse steam generator fleet

AREVA and B&W have indicated that their replacement steam generators are not susceptible to divider plate cracking due to materials and design

- Report not applicable E1

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Applicability to a Once Through Steam Generator Design

Discuss the applicability of the information in this report to once-through steam generators (recognizing there is no divider plate in once through steam generators)?

Discussion

The information related to the divider plates would not be applicable to OTSGs since they don't have divider plates.

The potential for cracking in the cladding or tube to tubesheet welds would be similar for OTSGs or recirculating SGs. All AREVA B&W ROTSGs have 52/152 tubesheet cladding and therefore they are not susceptible to this issue.

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Loose Parts Impacting Cladding

Page 5-17 (and page 5-1) implies that possible loose parts may result in steam generator channel head degradation (refer to Figure 5-26). Where was this degradation located (channel head cladding, divider plate, etc.). Please discuss the applicability of this report to plants in which loose parts have impacted the tubesheet cladding or divider plate complex.

Discussion

The loose part impact at the Chi non 84 steam generator shown in Figure 5-26 was described as PWSCC in the hot water box but did not identify specific location. It did indicate the impact was from a loose part of several kilograms. Reference 5-2 was cited in the report.

The main focus of the cited study was on the stub plate and its welds, and on the low yield strength mismatch between the stub plate and the divider plate material causing strain in the stub plate.

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Loose Parts Impacting Cladding

When plants identify loose parts in the channel head of the steam generator, they evaluate the damage for continued operation

The cracking in Chi non 84 was local to the impact area from the loose part and never reported to have propagated to the triple point

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Cold Working

The first and second paragraph on page 5-24 are not clear on what is necessary for the stress corrosion cracking to occur. For example, can cracking occur in a divider plate with no cold-working? Is cold-working considered a "fabrication defect?"

Discussion

The stress corrosion cracking requires three concurrent conditions-some threshold tensile stress state, susceptible material, and supporting en vi ron ment.

The stress can be cumulative from many sources including those applied by cold working.

Heavy cold working (such as severe surface grinding) also influences the susceptibility of the material itself.

Surface grinding also smears and tears the surface of stainless steel thus introducing sites more vulnerable to crack initiation.

Finally, cold-working is not considered a fabrication defect because many of the metal working processes are necessary for fabrication.

Surface cold working can be mitigated if the design so designates.

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Design Guidance for New Steam Generators

As a result of this effort, does the industry plan to make any recommendations regarding material or design considerations for the design of new (including replacement) steam generators?

Discussion

SGMP currently has guidance for manufacturing steam generator tubing

No current plans for developing guidance for designing steam generators

New designs incorporate advanced materials and improved manufacturing processes 55 ~~121 1 ELECTR IC POWER

Fatigue Growth Analysis Based on 40 Years

Some of the analyses performed in support of your assessments for divider plate cracking appear to assume 40 years of steam generator operation. Some units may operate their steam generators for a longer period of time, for example up to a maximum of 80 years of plant operation for subsequent license renewal. Please discuss which analysis were based on 40 years of operation and whether the conclusions would still be valid for steam generator operation for a longer period of time (e.g., up to 80 years or the maximum steam generator operating term for a plant license period of 80 years).

Discussion Follows 56 t=~121

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Fatigue Growth Analysis Based on 40 Years

40-year is the typical growth period in an ASME fatigue assessment

This was assumed adequate in that the cracking identified at EdF plants after more than 20 years of operation did not extend to the triple point

Based on operating experience, it is not likely that any operating plant has divider plate cracks that extend the full length of the divider plate and have propagated into the low alloy steel

- This was the beginning condition of the fatigue growth analysis

- To require 60-year assessment, the assumption has to be made that PWSCC grew into the low alloy steel year one of SG life 57

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ELECTR IC POWER RESEARCH INSTITUTE

Loads Evaluated

Because each plant has a unique set of design/licensing bases for their steam generators, did the analyses in this report address the limiting load levels from the entire range of design/licensing basis loadings for the population of plants that this report covers? For any design/licensing basis conditions that may have loads greater than those addressed in this report, please discuss how these greater loading levels would change the conclusions in this report, both for a fully degraded divider plate (through-wall flaw) and for one with less serious degradation.

Discussion Follows 58

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Loads Evaluated

Peak pressures and temperatures were assumed to exist for the duration of the transients; therefore evaluation of dynamic loads were not required.

- Time varying or transient temperatures and pressures would reduce the pressure and thermal loads.

A fully degraded divider plate will not result in an increased pressure differential across the tubesheet in any postulated SG operating condition. *Mixing of fluid temperatures through the crack opening area due to bypass flow would result in a less critical result for the applied stress on the tubesheet junctions because the hot leg temperature would decrease.

- This would bound crack scenarios that would result in less communication between hot and cold leg E1

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Conclusions

SGMP has communicated the following to utility members

- EPRI Technical Report 3002002850 can be used to update Aging Management Plans

Visual inspections of present rigor are adequate to manage the Alloy 600 material in the divider plate assembly

- EPRI Technical Report, 3002000473, "Steam Generator Channel Head Degradation Failure Modes and Effects Analysis" recommends inspecting channel head each time primary side inspections are performed

EPRI SGMP Chemistry Guidelines

Defense in Depth

- Boric acid walkdowns would most likely identify through wall leakage

- RCS leakage monitoring would most likely identify leakage during operation

Capability to detect small delta leak rate changes

Radiation monitors t=~

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Together ... Shaping the Future of Electricity E ~f21 1 61 ElECTRIC POWER

ML15222B160

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