RS-17-172, Commitments for Resolution of Anchor Darling Double Disc Gate Valve Part 21 Issues
| ML17354A079 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs, Dresden, Peach Bottom, Nine Mile Point, Oyster Creek, Byron, Three Mile Island, Braidwood, Limerick, Ginna, Clinton, Quad Cities, FitzPatrick, LaSalle |
| Issue date: | 12/19/2017 |
| From: | Jim Barstow Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-17-172 | |
| Download: ML17354A079 (18) | |
Text
Exelon Generation 200 Exelon Way l<ennett Square. PA 19348 www.exeloncorp.com RS-17-172 December 19, 2017 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 N RC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 N RC Docket Nos. STN 50-454 and STN 50-455 Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 N RC Docket Nos. 50-317 and 50-318 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-59 NRC Docket No. 50-333 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220 and 50-41 o Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219
U.S. Nuclear Regulatory Commission Commitments for Resolution of Anchor Darling Double Disc Gate Valve Part 21 Issues December 19, 2017 Page2 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289
Subject:
Commitments for Resolution of Anchor Darling Double Disc Gate Valve Part 21 Issues
References:
- 1) Letter from William Trafton (Exelon Generation Company, LLC), to U.S.
Nuclear Regulatory Commission, Amendment to Commitments for Resolution of Anchor Darling Double Disc Gate Valve Part 21 Issues, dated June 20, 2017
- 2) BWROG Topical Report TP16-1-112, Revision 4, Recommendations to Resolve Flowserve 10CFR Part 21 Notification Affecting Anchor Darling Double Disc Gate Valve Wedge Pin Failures, dated August 2017
- 3) Letter from Greg Krueger (NEI) to John Lubinski (U.S. Nuclear Regulatory Commission), Anchor Darling Double Disc Gate Valve Industry Resolution Plan Update (Project 689), dated August 4, 2017
- 4) Letter from James Barstow (Exelon Generation Company, LLC) to U.S.
Nuclear Regulatory Commission, Commitments for Resolution of Anchor Darling Double Disc Gate Valve Part 21 Issues, dated August 29, 2017
- 5) Letter from Joe Pollock (NEI) to Brian Holian (U.S. Nuclear Regulatory Commission), NSIAC Concurrence on Anchor Darling Double Disc Gate Valve Industry Response Actions (Project 689), dated October 26, 2017 In Reference 4, Exelon Generation Company, LLC (EGC) provided a status of the resolution of Anchor Darling Double Disk Gate Valve (ADDDGV) issues specifically related to safety-related Motor Operated Valve (MOV) applications at Oyster Creek Nuclear Generating Station and Quad Cities Nuclear Power Station, Units 1 and 2, in addition to providing commitment to valve repairs at these stations. Additionally, the Reference 4 letter committed to communicating repair schedule commitments by December 31, 2017.
U.S. Nuclear Regulatory Commission Commitments for Resolution of Anchor Darling Double Disc Gate Valve Part 21 Issues December 19, 2017 Page 3 Further review of the EGC fleet has determined that Braidwood Station, Units 1 and 2; Byron Station, Units 1 and 2; Clinton Power Station, Unit 1; and Nine Mile Point Nuclear Station, Units 1 and 2, do not use motor-operated ADDDGVs to support any active safety function and/or utilized in Generic Letter 96-05 MOV program applications.
Attached is additional information and updates regarding repair plans for the remainder of the EGC fleet. For each applicable site, Attachments 1 through 1O contain the following information for each applicable active safety-related MOV:
Plant Name, Unit, and Valve ID System Valve Functional Description Valve Size Active Safety Function (Open, Close, Both)
Are multiple design basis post-accident strokes required? (Yes/No)
Expert Panel Risk Ranking (High, Medium, Low)
Result of susceptibility evaluation (susceptible or not susceptible)
Is the susceptibility evaluation in general conformance with TP16-1-112R4 where the wedge pin applied torque must bound anticipated design basis operating torque requirements and current maximum total torque? (Yes/No)
Does the susceptibility evaluation rely on thread friction? If yes, include rotation criteria (No), (Yes, >0.10), (Yes, S0.10)
Was an initial stem-rotation check performed? If yes, include rotation criteria (No),
(Yes, S1 O deg.), (Yes, SS deg.)
Was the diagnostic test data reviewed for failure precursors described in TP16 112R4? (Yes/No)
The valve's repair status (repaired or not repaired)
EGC commits to perform all applicable activities identified in Attachment 11. EGC may modify these commitments should additional technical information or repair methods become available to justify such action. Additionally, Exelon has implemented MOV Program requirements for periodic Stem Rotation Checks for all applicable GL 96-05 motor operated ADDDGVs at the Periodic Verification Test (PVT) interval in accordance with TP16-1-112R4, Attachment 10.
Should you have any questions or require additional information, please contact Tom Basso, Director of Centralized Program Engineering (Thomas.Basso@Exeloncorp.com, 610-765-5910).
Respectfully, d~~
James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC
U.S. Nuclear Regulatory Commission Commitments for Resolution of Anchor Darling Double Disc Gate Valve Part 21 Issues December 19, 2017 Page 4 Attachments: 1) Calvert Cliffs Nuclear Power Plant, Units 1 and 2
- 2) Dresden Nuclear Power Station, Units 2 and 3
- 3) James A. FitzPatrick Nuclear Power Plant
- 4) R.E. Ginna Nuclear Power Plant
- 5) LaSalle County Station, Units 1 and 2
- 6) Limerick Generating Station, Units 1 and 2
- 7) Oyster Creek Nuclear Generating Station
- 8) Quad Cities Nuclear Power Station, Units 1 and 2
- 9) Peach Bottom Atomic Power Station, Units 2 and 3
- 10) Three Mile Island Nuclear Station, Unit 1
- 11) Summary of Regulatory Commitments cc: Regional Administrator - NRC Region I Regional Administrator - NRC Region Ill NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station NRC Senior Resident Inspector - Calvert Cliffs Nuclear Power Plant NRC Senior Resident Inspector - Clinton Power Station NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - James A. FitzPatrick Nuclear Power Plant NRC Senior Resident Inspector - LaSalle County Station NRC Senior Resident Inspector - Limerick Generating Station NRC Senior Resident Inspector - Nine Mile Point Nuclear Station NRC Senior Resident Inspector - Oyster Creek Nuclear Generating Station NRC Senior Resident Inspector - Peach Bottom Atomic Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station NRC Senior Resident Inspector - R.E. Ginna Nuclear Power Plant NRC Senior Resident Inspector - Three Mile Island Nuclear Station, Unit 1 S. T. Gray, State of Maryland (w/o attachments)
Illinois Emergency Management Agency - Division of Nuclear Safety -
(w/o attachments)
R. R. Janati, Bureau of Radiation Protection, Commonwealth of Pennsylvania (w/o attachments)
Manager, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection (w/o attachments)
Mayor of Lacey Township, Forked River, NJ (w/o attachments)
A. L. Peterson, NYSERDA (w/o attachments)
ATTACHMENT 1 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - Calvert Cliffs Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Plant Valve Functional susceptibility Valve repair Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Name Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
Reactor Calvert Coolant PORV Blocking Not Applicable Cliffs 1 1MOV0403 Pressurizer Valve 2.5 Both Yes Medium (Note 1) N/A N/A N/A N/A N/A Reactor Calvert Coolant PORV Blocking Not Applicable Cliffs 1 1MOV0405 Pressurizer Valve 2.5 Both Yes Medium (Note 1) N/A N/A N/A N/A N/A Instrument Air Calvert Instrument Containment Not Applicable Cliffs 1 1MOV2080 Air Isolation Valve 2.0 Both Yes Low (Note 1) N/A N/A N/A N/A N/A Reactor Calvert Coolant PORV Blocking Not Applicable Cliffs 2 2MOV0403 Pressurizer Valve 2.5 Both Yes Medium (Note 1) N/A N/A N/A N/A N/A Reactor Calvert Coolant PORV Blocking Not Applicable Cliffs 2 2MOV0405 Pressurizer Valve 2.5 Both Yes Medium (Note 1) N/A N/A N/A N/A N/A Instrument Air Calvert Instrument Containment Not Applicable Cliffs 2 2MOV2080 Air Isolation Valve 2.0 Both Yes Low (Note 1) N/A N/A N/A N/A N/A Notes (1): T-Head Connection Stem/Wedge Connection is not applicable to the Flowserve Part 21 (i.e. does not include threaded connection with wedge pin).
Page 1
ATTACHMENT 2 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - Dresden Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Plant Valve Functional susceptibility Valve repair Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Name Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
Reactor Head Spray Nuclear Outboard Isolation Not Applicable Dresden 2 2-0205-2-4 Boiler Valve 3.0 Close No Low (Note 1) N/A N/A N/A N/A N/A Shutdown Cooling Residual Suction Inboard Heat Containment Yes Dresden 2 2-1001-01A Removal Isolation 16 Close No Low Not Susceptible Yes No Yes, < 5 deg. Last Tested 2017 Not Repaired Shutdown Cooling Residual Suction Inboard Heat Containment Yes Dresden 2 2-1001-01B Removal Isolation 16 Close No Low Not Susceptible Yes No N/A Last Tested 2009 Not Repaired Nuclear Reactor Head Spray Not Applicable Dresden 3 3-0205-2-4 Boiler Outboard Isolation 2.5 Close No Low (Note 1) N/A N/A N/A N/A N/A Reactor RWCU Inlet Water Containment Yes Dresden 3 3-1201-1 Cleanup Isolation 8 Close No Medium Not Susceptible Yes No Yes, < 10 deg. Last Tested 2016 Not Repaired Reactor RWCU Inlet Bypass Water Containment Not Applicable Dresden 3 3-1201-1-1A Cleanup Isolation 2 Close No Medium (Note 1) N/A N/A N/A N/A N/A Reactor Water RWCU Aux Pump Yes Dresden 3 3-1201-2 Cleanup Bypass 8 Close No Medium Not Susceptible Yes No Yes, < 10 deg. Last Tested 2014 Not Repaired Notes (1): T-Head Connection Stem/Wedge Connection is not applicable to the Flowserve Part 21 (i.e. does not include threaded connection with wedge pin).
Page 2
ATTACHMENT 3 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - FitzPatrick Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
Reactor Rx Recirc Pump A Susceptible Yes Scheduled -
FitzPatrick 1 02-2MOV-53A Recirculation Discharge 28 Close No Low (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2014 Att.11, Cat. C Reactor Rx Recirc Pump B Susceptible Yes Scheduled -
FitzPatrick 1 02-2MOV-53B Recirculation Discharge 28 Close No Low (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2014 Att.11, Cat. C Residual Heat Yes FitzPatrick 1 10MOV-16A Removal RHR A Minimum Flow 4 Both Yes Low Not Susceptible Yes No No, Note 3 Last Tested: 2010 Not Repaired Residual Heat Yes FitzPatrick 1 10MOV-16B Removal RHR B Minimum Flow 4 Both Yes Low Not Susceptible Yes No No, Note 3 Last Tested: 2010 Not Repaired Residual Heat RHR Shutdown Yes FitzPatrick 1 10MOV-17 Removal Cooling OB Isolation 20 Close No Medium Not Susceptible Yes No No, Note 3 Last Tested: 2014 Not Repaired Residual Heat RHR Shutdown Yes FitzPatrick 1 10MOV-18 Removal Cooling IB Isolation 20 Close No Medium Not Susceptible Yes No No, Note 3 Last Tested: 2017 Not Repaired Residual Heat RHR Train A Susceptible Yes Scheduled -
FitzPatrick 1 10MOV-26A Removal Containment Spray 10 Both Yes High (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2017 Att. 11, Cat. A Residual Heat RHR Train B Susceptible Yes Scheduled -
FitzPatrick 1 10MOV-26B Removal Containment Spray 10 Both Yes High (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2014 Att.11, Cat. A Residual Heat RHR Train B Torus Not Susceptible Yes FitzPatrick 1 10MOV-39B Removal Cooling Isolation 16 Both Yes High (Note 2) See Note 2 Yes < 0.1 No, Note 3 Last Tested: 2013 Not Repaired Rx Water Susceptible Yes Scheduled -
FitzPatrick 1 12MOV-15 Cleanup RWCU Supply Inboard 6 Close No High (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2017 Att. 11, Cat. B Rx Water RWCU Return Susceptible Yes Scheduled -
FitzPatrick 1 12MOV-69 Cleanup Containment 4 Close No Low (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2012 Att.11, Cat. C Reactor Core RCIC Steam Supply Not Applicable FitzPatrick 1 13MOV-15 Isolation Cool Inboard Isolation 3 Close No High (Note 1) N/A N/A N/A N/A N/A High Pressure HPCI Steam Supply Not Susceptible Yes FitzPatrick 1 23MOV-15 Coolant Inject Isolation 10 Close No High (Note 2) See Note 2 Yes < 0.1 No, Note 3 Last Tested: 2014 Not Repaired Drywell Floor Drain Not Applicable FitzPatrick 1 20MOV-82 Rad Waste Sump Isolation 3 Close No Low (Note 1) N/A N/A N/A N/A N/A Drywell Equipment Not Applicable FitzPatrick 1 20MOV-94 Rad Waste Drain Sump Isolation 3 Close No Low (Note 1) N/A N/A N/A N/A N/A MS Line Drain Inboard Not Applicable FitzPatrick 1 29MOV-74 Main Steam Isolation 3 Close No Low (Note 1) N/A N/A N/A N/A N/A MS Line Drain Not Applicable FitzPatrick 1 29MOV-77 Main Steam Outboard Isolation 3 Close No Low (Note 1) N/A N/A N/A N/A N/A Notes:
(1): T-Head Connection Stem/Wedge Connection is not applicable to the Flowserve Part 21 (i.e. does not include threaded connection with wedge pin).
(2): Initial wedge pin susceptibility evaluation relied on thread friction. Consequently, a re-evaluation performed in 2017 has determined the applicable MOVs to be susceptible and subject to repair if thread COF required is > 0.1.
(3): Until 2017, plant considered all of their MOVs as non-susceptible by wedge pin analysis; consequently no stem rotation check inspections were performed.
Page 3
ATTACHMENT 4 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - Ginna Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
RCS Pressurizer Relief Stop Not Applicable Ginna 1 515 Pressurizer Valve 3.0 Both Yes Low (Note 1) N/A N/A N/A N/A N/A RCS Pressurizer Relief Stop Not Applicable Ginna 1 516 Pressurizer Valve 3.0 Both Yes Low (Note 1) N/A N/A N/A N/A N/A Residual Heat RHR Pump 1A Cross Susceptible Yes Scheduled -
Ginna 1 704A Removal Connect 10.0 Close No Low (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2011 Att.11, Cat. C Residual Heat RHR Pump 1B Cross Susceptible Yes Scheduled -
Ginna 1 704B Removal Connect 10.0 Close No Low (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2009 Att.11, Cat. C Safety Accumulator Tank 1A Susceptible Yes Scheduled -
Ginna 1 841 Injection Shutoff 10.0 Close Yes Medium (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2014 Att. 11, Cat. C Residual Heat Yes Ginna 1 850A Removal Sump B to RHR Pumps 10.0 Both Yes High Not Susceptible Yes No Yes, 5 deg. Last Tested: 2017 Not Repaired Residual Heat Yes Ginna 1 850B Removal Sump B to RHR Pumps 10.0 Both Yes High Not Susceptible Yes No Yes, 5 deg. Last Tested: 2017 Not Repaired Residual Heat Yes Ginna 1 856 Removal RWST to RHR Pumps 10.0 Close No Medium Not Susceptible Yes No No, Note 3 Last Tested: 2011 Not Repaired Residual Heat Yes Ginna 1 857A Removal 1A RHR HX to SIP/CSP 6.0 Open No Low Not Susceptible Yes No No, Note 3 Last Tested: 2014 Not Repaired Residual Heat Yes Ginna 1 857B Removal 1B RHR HX to SIP/CSP 6.0 Open No Medium Not Susceptible Yes No Yes, 5 deg. Last Tested: 2017 Not Repaired Residual Heat Yes Ginna 1 857C Removal RHR HX to 1A Outlet 6.0 Open No Low Not Susceptible Yes No No, Note 3 Last Tested: 2015 Not Repaired Containment Yes Ginna 1 860A Spray 1A CSP Discharge 6.0 Both Yes Low Not Susceptible Yes No No, Note 3 Last Tested: 2014 Not Repaired Containment Yes Ginna 1 860B Spray 1A CSP Discharge 6.0 Both Yes Low Not Susceptible Yes No No, Note 3 Last Tested: 2015 Not Repaired Containment Yes Ginna 1 860C Spray 1B CSP Discharge 6.0 Both Yes Low Not Susceptible Yes No No, Note 3 Last Tested: 2015 Not Repaired Containment Yes Ginna 1 860D Spray 1B CSP Discharge 6.0 Both Yes Low Not Susceptible Yes No No, Note 3 Last Tested: 2015 Not Repaired Safety Accumulator Tank 1B Susceptible Yes Scheduled -
Ginna 1 865 Injection Shutoff 10.0 Close No Medium (Note 2) See Note 2 See Note 2 No, Note 3 Last Tested: 2014 Att. 11, Cat. C Notes:
(1): T-Head Connection Stem/Wedge Connection is not applicable to the Flowserve Part 21 (i.e. does not include threaded connection with wedge pin).
(2): Initial wedge pin susceptibility evaluation relied on thread friction. Consequently, a re-evaluation performed in 2017 has determined the applicable MOVs to be susceptible and subject to repair if thread COF required is > 0.1.
(3): Until 2017, Ginna was not identified as having susceptible valves due to not being included in the 2013 Flowserve Part 21; consequently no stem rotation check inspections were performed prior to 2017.
Page 4
ATTACHMENT 5 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - LaSalle (Page 1 of 2)
Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
Main Steam MS Drain Line Not Applicable LaSalle 1 1B21-F067A Drains Isolation 1.5 Close No L (Note 1) N/A N/A N/A N/A N/A Main Steam MS Drain Line Not Applicable LaSalle 1 1B21-F067B Drains Isolation 1.5 Close No L (Note 1) N/A N/A N/A N/A N/A Main Steam MS Drain Line Not Applicable LaSalle 1 1B21-F067C Drains Isolation 1.5 Close No L (Note 1) N/A N/A N/A N/A N/A Main Steam MS Drain Line Not Applicable LaSalle 1 1B21-F067D Drains Isolation 1.5 Close No L (Note 1) N/A N/A N/A N/A N/A Main Steam MS Drain Line Not Applicable LaSalle 2 2B21-F067A Drains Isolation 1.5 Close No L (Note 1) N/A N/A N/A N/A N/A Main Steam MS Drain Line Not Applicable LaSalle 2 2B21-F067B Drains Isolation 1.5 Close No L (Note 1) N/A N/A N/A N/A N/A Main Steam MS Drain Line Not Applicable LaSalle 2 2B21-F067C Drains Isolation 1.5 Close No L (Note 1) N/A N/A N/A N/A N/A Main Steam MS Drain Line Not Applicable LaSalle 2 2B21-F067D Drains Isolation 1.5 Close No L (Note 1) N/A N/A N/A N/A N/A Notes:
(1): T-Head Connection Stem/Wedge Connection is not applicable to the Flowserve Part 21 (i.e. does not include threaded connection with wedge pin).
Page 5
ATTACHMENT 5 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - LaSalle (Page 2 of 2)
Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
High Pressure HPCS Injection Repaired June LaSalle 1 1E22-F004 Core Spray Isolation 12 Both Yes Medium Not Susceptible Note 4 N/A Note 4 N/A 2017 High Pressure Yes Scheduled -
LaSalle 1 1E22-F012 Core Spray HPCS Pump Min Flow 4 Both Yes Medium Susceptible Yes No Yes, 10 deg. Last Tested: 2016 Att.11, Cat. A High Pressure HPCS Pump Suppress Yes Scheduled -
LaSalle 1 1E22-F015 Core Spray Pool Suction 18 Close No Low Susceptible Yes No Yes, 10 deg. Last Tested: 2016 Att.11, Cat. C Reactor Core RCIC Steam Outboard Yes Scheduled -
LaSalle 1 1E51-F008 Isol. Cooling Isolation 4 Close No Medium Susceptible Yes No Yes, 10 deg. Last Tested: 2016 Att.11, Cat. B Reactor Core RCIC Injection Yes Scheduled -
Isol. Cooling LaSalle 1 1E51-F013 Outboard Isolation 6 Both No Low Susceptible Yes No Yes, 10 deg. Last Tested: 2016 Att.11, Cat. C Reactor Core RCIC Steam Supply Yes Scheduled -
LaSalle 1 1E51-F063 Isol. Cooling Inboard Isolation 10 Close No High Susceptible Yes No Yes, 10 deg. Last Tested: 2016 Att. 11, Cat. B Rx Water RWCU Inboard Yes Scheduled -
LaSalle 1 1G33-F001 Cleanup Isolation 6 Close No Medium Susceptible Yes No Yes, 10 deg. Last Tested: 2016 Att. 11, Cat. B Rx Water RWCU Outboard Yes Scheduled -
LaSalle 1 1G33-F004 Cleanup Isolation 6 Close No High Susceptible Yes No Yes, 10 deg. Last Tested: 2016 Att. 11, Cat. B High Pressure HPCS Injection Repaired LaSalle 2 2E22-F004 Core Spray Isolation 12 Both Yes Medium Not Susceptible Note 4 N/A Note 4 N/A February 2017 High Pressure Yes Scheduled -
LaSalle 2 2E22-F012 Core Spray HPCS Pump Min Flow 4 Both Yes Medium Susceptible Yes No Yes, 10 deg. Last Tested: 2015 Att. 11,Cat. A High Pressure HPCS Pump Suppress Yes Scheduled -
LaSalle 2 2E22-F015 Core Spray Pool Suction 18 Close No Low Susceptible Yes No Yes, 10 deg. Last Tested: 2015 Att. 11, Cat. C Reactor Core RCIC Steam Outboard Yes Scheduled -
LaSalle 2 2E51-F008 Isol. Cooling Isolation 4 Close No Medium Susceptible Yes No Yes, 5 deg. Last Tested: 2017 Att. 11, Cat. B Reactor Core RCIC Steam Supply Yes Scheduled -
LaSalle 2 2E51-F063 Isol. Cooling Inboard Isolation 10 Close No High Susceptible Yes No Yes, 10 deg. Last Tested: 2015 Att. 11, Cat. B Rx Water RWCU Inboard Yes Scheduled -
LaSalle 2 2G33-F001 Cleanup Isolation 6 Close No Medium Susceptible Yes No Yes, 5 deg. Last Tested: 2017 Att. 11, Cat. B Rx Water RWCU Outboard Yes Scheduled -
LaSalle 2 2G33-F004 Cleanup Isolation 6 Close No High Susceptible Yes No Yes, 5 deg. Last Tested: 2017 Att. 11, Cat. B Notes:
(4): Until repaired in 2017, these valves were identified as being susceptible. 2E22-F004 had a stem-disc separation failure in Feb 2017 during L2R16. 1E22-F004 was identified with a failed wedge pin but with the wedge/stem fully connected during June 2017 Unit 1 Maintenance Outage.
Page 6
ATTACHMENT 6 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - Limerick Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
Core Spray Cooling Yes Limerick 1 HV-052-1F001A Core Spray Pump A Suction 16 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2016 Not Repaired Core Spray Cooling Yes Limerick 1 HV-052-1F001B Core Spray Pump B Suction 16 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2014 Not Repaired Core Spray Cooling Yes Limerick 1 HV-052-1F001C Core Spray Pump C Suction 16 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2014 Not Repaired Core Spray Cooling Yes Limerick 1 HV-052-1F001D Core Spray Pump D Suction 16 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2015 Not Repaired Core Spray Cooling Yes Limerick 2 HV-052-2F001A Core Spray Pump A Suction 16 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2015 Not Repaired Core Spray Cooling Yes Limerick 2 HV-052-2F001B Core Spray Pump B Suction 16 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2016 Not Repaired Core Spray Cooling Yes Limerick 2 HV-052-2F001C Core Spray Pump C Suction 16 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2015 Not Repaired Core Spray Cooling Yes Limerick 2 HV-052-2F001D Core Spray Pump D Suction 16 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2016 Not Repaired Notes: None Page 7
ATTACHMENT 7 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - Oyster Creek Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
Isolation IC NEO1A INLET Yes Note 6 Oyster Creek 1 V-14-0030 Condenser ISOLATION VALVE 10 Close No Medium Susceptible Yes No Yes, 10 deg. Last Tested 2016 Att.11, Cat. B Isolation IC NEO1A INLET Yes Note 6 Oyster Creek 1 V-14-0031 Condenser ISOLATION VALVE 10 Close No Medium Susceptible Yes No Yes, 10 deg. Last Tested 2012 Att. 11, Cat. B Isolation IC NEO1B INLET Yes Note 6 Oyster Creek 1 V-14-0032 Condenser ISOLATION VALVE 10 Close No Medium Susceptible Yes No Yes, 10 deg. Last Tested 2010 Att. 11, Cat. B Isolation IC NEO1B INLET Yes Note 6 Oyster Creek 1 V-14-0033 Condenser ISOLATION VALVE 10 Close No Medium Susceptible Yes No Yes, 10 deg. Last Tested 2016 Att. 11, Cat. B Isolation IC NEO1A Outlet Yes Note 5 Oyster Creek 1 V-14-0034 Condenser ISOLATION VALVE 10 Both Yes High Susceptible Yes No Yes, 10 deg. Last Tested 2016 Att. 11, Cat. A Isolation IC NEO1B Outlet Yes Note 5 Oyster Creek 1 V-14-0035 Condenser ISOLATION VALVE 10 Both Yes High Susceptible Yes No Yes, 10 deg. Last Tested 2016 Att. 11, Cat. A Notes:
Note (5): Repair Schedule Commitment for OC1R27 in Aug 29 2017 Exelon Letter to the NRC Note (6): Test/Inspect Schedule Commitment for OC1R17 in Aug 29 2017 Exelon Letter to the NRC Page 8
ATTACHMENT 8 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - Quad Cities Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
Reactor Water RWCU Inboard Yes Note 7 Quad Cities 1 1-1201-2 Cleanup Isolation 6 Close No High Susceptible Yes No Yes, 10 deg. Last Tested 2015 Att. 11, Cat. B Reactor Water RWCU Outboard Yes Note 7 Quad Cities 1 1-1201-5 Cleanup Isolation 6 Close No Medium Susceptible Yes No Yes, 10 deg. Last Tested 2017 Att. 11, Cat. B Reactor Water RWCU Inboard Yes Note 7 Quad Cities 2 2-1201-2 Cleanup Isolation 6 Close No High Susceptible Yes No Yes, 10 deg. Last Tested 2016 Att. 11, Cat. B Reactor Water RWCU Outboard Yes Note 7 Quad Cities 2 2-1201-5 Cleanup Isolation 6 Close No Medium Susceptible Yes No Yes, 10 deg. Last Tested 2016 Att. 11, Cat. B Notes:
Note(7): Repair Schedule Commitment for Unit 1 (Q1R25 in 2019) and Unit 2 (Q2R24 in 2018) in Aug 29 2017 Exelon Letter to the NRC.
Page 9
ATTACHMENT 9 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - Peach Bottom Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
Peach Main MSL Drain Inboard Yes Bottom 2 MO-2-01A-074 Steam Isolation to Condenser 3 Close No Low Not Susceptible Yes No Yes, 10 deg. Last Tested 2014 Not Repaired Peach Main MSL Drain Outboard Yes Bottom 2 MO-2-01A-077 Steam Isolation to Condenser 3 Close No Low Not Susceptible Yes No Yes, 10 deg. Last Tested 2014 Not Repaired Peach Main MSL Drain Inboard Yes Bottom 3 MO-3-01A-074 Steam Isolation to Condenser 3 Close No Low Not Susceptible Yes No Yes, 10 deg. Last Tested 2013 Not Repaired Peach Main MSL Drain Outboard Yes Bottom 3 MO-3-01A-077 Steam Isolation to Condenser 3 Close No Low Not Susceptible Yes No Yes, 5 deg. Last Tested 2017 Not Repaired Peach Reactor Reactor Recirc Pump A Yes Scheduled Bottom 2 MO-2-02-53A Recirc Discharge 24 Close No Low Susceptible Yes No Yes, 10 deg. Last Tested 2016 Att. 11, Cat. C Peach Reactor Reactor Recirc Pump B Yes Scheduled Bottom 2 MO-2-02-53B Recirc Discharge 24 Close No Low Susceptible Yes No Yes, 10 deg. Last Tested 2016 Att. 11, Cat. C Peach Reactor Reactor Recirc Pump A Yes Scheduled Bottom 3 MO-3-02-53A Recirc Discharge 24 Close No Low Susceptible Yes No Yes, 5 deg. Last Tested 2017 Att. 11, Cat. C Peach Reactor Reactor Recirc Pump B Yes Scheduled Bottom 3 MO-3-02-53B Recirc Discharge 24 Close No Low Susceptible Yes No Yes, 5 deg. Last Tested 2017 Att. 11, Cat. C Notes: None Page 10
ATTACHMENT 10 Exelon Generation Plant ADDDGV Listing with Active Safety Related Applications - Three Mile Island Is the susceptibility Does the susceptibility Was an initial stem-Result of Valve Active Are multiple design Expert evaluation in evaluation rely on thread rotation check Was the diagnostic test data Valve Functional susceptibility Valve repair Plant Name Unit Valve ID System Size Safety basis post-accident Panel Risk general friction? performed? reviewed for failure precursors Description evaluation status (inches) Function strokes required? Ranking conformance with If yes, was the COF greater If yes, include rotation described in TP16-1-112R4?
TP16-1-112R4? than 0.10? criteria (Open, (High, (No), (No),
(susceptible or (repaired or Close, Medium, (Yes/No) (Yes, >0.10), (Yes, 10 deg.), (Yes/ No)
(Yes/No) not susceptible) not repaired)
Both) Low) (Yes, 0.10) (Yes, 5 deg.)
DECAY HEAT DROP Three Mile Decay LINE CONTAINMENT Yes, Island 1 DH-V-3 Heat ISOL 12 Open No High Not Susceptible Yes No Yes, 5 deg. Last Tested 2017 Not Repaired Notes: None Page 11
ATTACHMENT 11 -
SUMMARY
OF REGULATORY COMMITMENTS (Page 1 of 3)
The following table identifies the no later than completion date/outage commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRCs information and are not regulatory commitments.)
COMMITMENT TYPE COMMITTED COMMITMENT DATE OR ONE-TIME Programmatic "OUTAGE" ACTION (Yes/No)
(Yes/No)
Category A MOVs Category A MOVs to be repaired at the Oyster Creek Nuclear Generating:
MOV Number Outage(Year)
V-14-0034 OC1R27 (2018) Yes No V-14-0035 OC1R27 (2018)
Category A MOVs to be repaired at the James A.
FitzPatrick Nuclear Power Plant:
MOV Number Outage(Year) 10MOV-26A FPR23 (2018) Yes No 10MOV-26B (Note A) FPR23 (2018)
Category A MOVs to be repaired at the LaSalle County Station, Units 1 and 2:
MOV Number Outage(Year) 1E22-F012 L1R17 (2018) Yes No 2E22-F012 L2R17 (2019)
Notes:
Note A: James A. FitzPatrick Nuclear Power Plant will repair 10MOV-26B either during the FPR23 outage or in a 2018 RHR Work window.
Page 1
ATTACHMENT 11 -
SUMMARY
OF REGULATORY COMMITMENTS (Page 2 of 3)
COMMITMENT TYPE COMMITTED DATE ONE-TIME COMMITMENT Programmatic OR "OUTAGE" ACTION (Yes/No)
(Yes/No)
Category B MOVs Category B MOVs to be repaired at the James A. FitzPatrick Nuclear Power Plant:
MOV Number Outage(Year) 12MOV-15 FPR23 (2018) Yes No Category B MOVs to be repaired at the LaSalle Generating Station, Units 1 and 2:
MOV Number Outage(Year) 1E51-F008 L1R17 (2018) Yes No 1E51-F063 L1R17 (2018) 1G33-F001 L1R17 (2018) 1G33-F004 L1R17 (2018) 2E51-F008 L2R17 (2019) 2E51-F063 L2R17 (2019) 2G33-F001 L2R17 (2019) 2G33-F004 L2R17 (2019)
Perform diagnostic testing and stem rotation checks with contingent repairs on the Group B MOVs at the Oyster Creek Nuclear Station:
MOV Number Outage(Year)
V-14-30 OC1R27 (2018) Yes No V-14-31 OC1R27 (2018)
V-14-32 OC1R27 (2018)
V-13-33 OC1R27 (2018)
Category B MOVs to be repaired at the Quad Cities Nuclear Power Station, Units 1 and 2:
MOV Number Outage(Year) 1-1201-2 Q1R25 (2019) Yes No 1-1201-5 Q1R25 (2019) 2-1201-2 Q2R24 (2018) 2-1201-5 Q2R24 (2018)
Page 2
ATTACHMENT 11 -
SUMMARY
OF REGULATORY COMMITMENTS (Page 3 of 3)
COMMITMENT TYPE COMMITTED DATE ONE-TIME COMMITMENT Programmatic OR "OUTAGE" ACTION (Yes/No)
(Yes/No)
Category C MOVs Category C MOVs to be repaired at the James A.
FitzPatrick Nuclear Power Plant:
MOV Number Outage(Year) 02-2MOV-53A (Note C) FPR25 (2022) Yes No 02-2MOV-53B (Note C) FPR25 (2022) 12MOV-69 (Note C) FPR25 (2022)
Category C MOVs to be repaired at the R. E. Ginna Nuclear Power Plant (Note B):
MOV Number Outage(Year) 704A G1R42 (2020) Yes No 704B G1R42 (2020) 841 G1R42 (2020) 865 G1R42 (2020)
Category C MOVs to be repaired at the LaSalle County Generating Station, Units 1 and 2:
MOV Number Outage(Year) 1E22-F015 L1R17 (2018) Yes No 1E51-F013 L1R17 (2018) 2E22-F015 L2R17 (2019)
Category C MOVs to be repaired at the Peach Bottom Atomic Power Station, Units 2 and 3:
MOV Number Outage(Year)
MO-2-02-053A (Note D) P2R24 (2022) Yes No MO-2-02-053B (Note D) P2R24 (2022)
MO-3-02-053A (Note D) P3R23 (2021)
MO-3-02-053B (Note D) P3R23 (2021)
Notes:
Note B: Assuming no degradation is found, R.E. Ginna Nuclear Power Plant is planning a high strength wedge pin replacement. If stem/wedge degradation is found, a full repair in accordance with BWROG TP16 112 Att. 6 will be performed.
Note C: These James A. FitzPatrick Nuclear Power Plant Group C MOVs (if not repaired) will also undergo Stem Rotation Checks and Diagnostic Testing during FPR23 (2018) and FPR24 (2020).
Note D: These Peach Bottom Atomic Power Station, Units 2 and 3 Group C MOVs (if not repaired) will also undergo Stem Rotation Checks and Diagnostic Testing during P2R22 (2018) and P2R23 (2020) for MO-2-02-053A & 53B; P3R22 (2019) for MO-3-02-053A & 53B.
Page 3