SBK-L-12123, 5 License Renewal Application
| ML12178A405 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 06/19/2012 |
| From: | Walsh K NextEra Energy Seabrook |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SBK-L-12123 | |
| Download: ML12178A405 (76) | |
Text
NEXTera ENERGY June 19, 2012 SBK-L-12123 Docket No. 50-443 U.S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 Seabrook Station NextEra Energy Seabrook License Renewal Application Supplement # 25
References:
- 1. NextEra Energy Seabrook, LLC letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License," May 25, 2010. (Accession Number ML101590099)
- 2. NRC Letter, Requests For Additional Information For The Review Of The Seabrook Station, License Renewal Application-Set 17 dated May 29, 2012 (Accession Number ML12144A441)
- 3. LR-ISG-201 1-02 : Final License Renewal Interim Staff Guidance: Aging Management Program For Steam Generators (Accession Number MLI 1297A085)
- 4. NextEra Energy Seabrook, LLC letter SBK-L-11002, "Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Aging Management Programs - Set 4," January 13, 2012. (Accession Number MLI 10140809)
In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted an application for a renewed facility operating license for Seabrook Station Unit I in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.
In Reference 2, the NRC requested additional information related to the recently approved LR-ISG-201 1-01 "Aging Management of Stainless Steel Structures and Components in Treated Borated Water" and a recently installed Seal Cap Enclosures. Enclosure 1 provides NextEra's response to these RAIs.
NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874
United States Nuclear Regulatory Commission SBK-L-12123/Page 2 Reference 3, Final License Renewal Interim Staff Guidance LR-ISG-2011-02, Aging Management Program For Steam Generators was recently issued by the NRC.
contains changes related to NextEra's License Renewal Application based on the recently issued guidance.
In Reference 4, NextEra provided a response to Request for Additional Information (RAI)
B.2. 1.10-1 related to Steam Generator Tube Integrity and to RAI B.2.1.26-1 Associated with Flash Point Testing. Discussion with the NRC staff identified the need to clarify NextEra Seabrook's response to these RAIs. Enclosure 3 provides NextEra's revised response to RAIs B.2.1.10-1 and B.2.1.26-1.
In this Supplement are changes to the License Renewal Application (LRA). To facilitate understanding, the changes are explained, and where appropriate, portions of the LRA are repeated with the change highlighted by strikethroughs for deleted text and bolded italics for inserted text. In some instances the entire text of a section has been replaced or added. In these cases a note is included in the introduction indicating the replacement of the entire text of the section.
Commitment numbers 54 and 55 have been revised.
There are no other new or revised regulatory commitments contained in this letter. Enclosure 4 provides a revised LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List, updated to reflect the license renewal commitment changes made in NextEra Energy Seabrook correspondence to date.
If there are any questions or additional information is needed, please contact Mr. Richard R.
Cliche, License Renewal Project Manager, at (603) 773-7003.
If you have any questions regarding this correspondence, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.
Sincerely, NextEra Energy Seabrook, LLC.
Kevin T. Walsh Site Vice President
United States Nuclear Regulatory Commission SBK-L-12123/Page 3
Enclosures:
NextEra Reponses to NRC Requests for Additional Information dated May 29, 2012 Changes to the Seabrook Station License Renewal Application Associated with LR-ISG-2011-02, "Aging Management Program for Steam Generators Clarification to Responses to RAI B.2. 1.10-1 Associated with Steam Generator Tube-to Tubesheet Weld Inspection Plan, RAI B.2.1.26-1 Associated with Flash Point Testing, and to Steam Generator Divider Plate Inspection Plan LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List cc:
W.M. Dean, J. G. Lamb, W. J. Raymond, A.D. Cunanan, M. Wentzel, NRC Region I Administrator NRC Project Manager, Project Directorate 1-2 NRC Resident Inspector NRC Project Manager, License Renewal NRC Project Manager, License Renewal Mr. Christopher M. Pope Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399
United States Nuclear Regulatory Commission SBK-L-12123/ Page 4 NExTera 1, Kevin Walsh, Site Vice President of NextEra Energy Seabrook, LLC hereby affirm that the information and statements contained within are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.
Sworn and Subscribed Before me this
/ ý day of June, 2012 Kevin T. Walsh Site Vice President Notary Puvlic to SBK-L-12123 NextEra Reponses to NRC Requests for Additional Information dated May 29, 2012
United States Nuclear Regulatory Commission Page 1 of 48 SBK-L-12123 / Enclosure 1 RAI 3.2.1.48-1
Background:
On May 3, 2012, the staff issued License Renewal Interim Staff Guidance (LR-ISG), LR-ISG-201 1-01, "Aging Management of Stainless Steel Structures and Components in Treated Borated Water,"
(ADAMS Accession No. ML12034A047) revising the SRP-LR and GALL Report to include the following additional aging management activities:
" Add the One-Time Inspection program to verify the effectiveness of the Water Chemistry program to manage loss of material due to pitting and crevice corrosion and cracking due to stress corrosion cracking in treated borated water.
- Add reduction of heat transfer due to fouling as an aging effect for stainless steel heat exchanger tubes exposed to treated borated water, and manage this aging effect with the Water Chemistry and One-Time Inspection programs.
This revised guidance applies to stainless steel structures and components exposed to treated borated water environments that are not actively controlled to oxygen levels less than 5 ppb.
In the license renewal application (LRA), the applicant stated that stainless steel and steel with stainless steel cladding components exposed to treated borated water will be managed for loss of material due to pitting and crevice corrosion and cracking due to stress corrosion cracking with the Water Chemistry program for those items associated with LRA Table 3.2.1, item 3.2.1-48; Table 3.2.1, item 3.2.1-49; Table 3.3.1, item 3.3.1-90; and Table 3.3.1, item 3.3.1-91.
In its response to RAI 3.2.2.2.4.2-1-A, dated June 2,2011, the applicant stated that stainless steel heat exchanger tubes exposed to treated borated water will be managed for reduction of heat transfer with the Water Chemistry program. The associated aging management review (AMR) items added in the request for additional information (RAI) response cite generic note H.
Issue:
The LRA contains several AMR items that manage stainless steel components exposed to treated borated water for loss of material, cracking, and reduction of heat transfer with the Water Chemistry program. However, the staff noted that the associated treated borated water environments may not be controlled to less than 5 ppb dissolved oxygen, and thus, the aging effects may not be effectively managed.
Request:
Describe how the effectiveness of the Water Chemistry program will be verified for those AMR items where the Water Chemistry program is used to manage loss of material, cracking, and reduction of heat transfer for stainless steel components exposed to treated borated water with greater than 5 ppb oxygen.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 2 of 48 NextEra Energy Seabrook Response:
On May 3, 2012, the NRC issued LR-ISG-2011-01, "Aging Management of Stainless Steel Structures and Components in Treated Borated Water". This ISG provides guidance for managing the aging effects during the period of extended operation for stainless steel structures and components exposed to treated borated water within the scope of the License Renewal Rule.
In response to LR-ISG-201 1-0 1, NextEra Seabrook has made the following changes to the License Renewal Application:
- 1. The One-Time Inspection program was added to verify the effectiveness of the Water Chemistry program to manage loss of material due to pitting and crevice corrosion and cracking due to stress corrosion cracking of stainless steel components in treated borated water environment.
- 2. Reduction of heat transfer due to fouling was added as an aging effect for stainless steel heat exchanger tubes exposed to treated borated water. The Water Chemistry and One-Time Inspection programs were assigned to manage this aging effect.
The following are detailed changes to the License Renewal Application as a result of LR-ISG-2011-01, "Aging Management of Stainless Steel Structures and Components in Treated Borated Water".
- 1. In Table 3.1.2-1, on page 3.1-44, the 5 th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Flexible (Spatial)
Stainless Borated Loss of Program V.A-27 Hose Steel Water Material One-Time (EP4 1)
Pressure (Internal)
Inspection A
Boundary Program
- 2. In Table 3.1.2-1, on page 3.1-44, the 6th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Flexible (Spatial)
Stainless Borated Program V.A-28 Water Cracking (E-12) 3.2.1-48 Hose Steel
>1400 F One-Time Pressure Boundary (Internal)
Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 3 of 48
- 3.
In Table 3.1.2-1, on page 3.1-45, the 4h row is revised as follows:
Heat Water Exchanger Treated Chemistry A
Exhngr Leakage Program Components age Stainless Borated Loss of V.A-27 3.2.1-49 (RC-E-126 Boundal)
Steel Water Material Otte-Time (EP-41)
Channel (Spatial)
(Internal)
Inspection Head)
Program
- 4. In Table 3.1.2-1, on page 3.1-45, the 8th row is revised as follows:
Heat Exchanger Water Components Chemistry C
(Reactor Heat Coolant Transfer Treated Program Stainless Borated Loss of V.A-27 Pump Steel Water Material (EP-41) 3.2.1-49 Thermal Pressure Se Wter Maeil(P41 Barrier Heat Boundary (External)
One-Time Exchanger Inspection C
Cooling Program Coil)
- 5. In Table 3.1.2-1, on page 3.1-45, a new row is added after the 8th row as follows:
Heat Exchanger Water Components Chemistry A
Treated Program Coolant Transfer Stainless Borated etVIA4-4 Pump Steel Water of Heat (AP-62) 3.3.1-3 Thermal Pressure Sel Wter Transfer On-6im BrirHa Bonay(External)
One-Time Barrier Heat Bounday Inspection A
Exch anger Program Cooling Coil)
- 6. In Table 3.1.2-1, on page 3.1-48, the 6th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Loss of Program V.A-27 3.2.1-49 Fittings Steel Water Material One-Time (EP-41)
Pressure (Internal)
Inspection A
Boundary Program
United States Nuclear Regulatory Commission SBK-L-12 123 / Enclosure 1 Page 4 of 48
- 7. In Table 3.1.2-1, on page 3.1-49, the 1st row is revised as follows:
Water Leakage WA Boundary Treated Chemistry Piping and (Spatial)
Stainless Borated Program 1-4-82 Fittings Steel Water Cracking VE-1 2 Pressure
>1400 F One-Time Boundary (Internal)
Inspection A
Program I
- 8. In Table 3.1.2-1, on page 3.1-58, the 1st row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program V.A-27 Pump Casing Boundary Steel Water Material (EP-41) 3.2.1-49 (Spatial)
(Internal)
One-Time Inspection A
I Program I
I
- 9. In Table 3.1.2-1, on page 3.1-61, the 5h row is revised as follows:
Water Treated Chemistry A
Rupture Leakage Stainless Borated Loss of Program V.A-27 Disk Boundary Steel Water Material (EP-41) 3.2.1-49 (Spatial)
(Internal)
One-Time Inspection A
I Program I
I_
I
- 10. In Table 3.1.2-1, on page 3.1-61, the 8 th row is revised as follows:
Water Treated Chemistry C
Leakage Stainless Borated Loss of Program V.A-27 Tank Boundary Steel Water Material (EP-41) 3.2.1-49 (Spatial)
(SnterMal)
One-Time Inspection C
Program
- 11. In Table 3.1.2-1, on page 3.1-62, the 5 th row is revised as follows:
Water Treated Chemistry C
Thermowell Pressure Stainless Borated Loss of Program V.A-27 3.2.1-49 Boundary Steel Water Material One-Time (EP-41)
(Internal)
Inspection C
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 5 of 48
- 12. In Table 3.1.2-1, on page 3.1-64, the 1st row is revised as follows:
Leakage Water Treated Chemistry A
Valve Body (Spatial)
Borated Loss of Program V.A-27
-49 Water Material One-Time (EP-41) 3.2.1 Pressure (Internal)
Inspection A
Boundary Program
- 13. In Table 3.1.2-1, on page 3.1-64, the 2 "d row is revised as follows:
Leakage Water Leary Treated Chemistry A
o Boundary Borated Program V.A-28 Valve Body (Spatial)
CASS Water Cracking (E-12) 3.2.1-48 Pressure
>1400 F One-Time Boundary (Internal)
Inspection A
Program
- 14. In Table 3.1.2-1, on page 3.1-65, the 1st row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Valve Body (Spatial)
Stainless Borated Loss of Program V.A-27 3.2.1-49 Steel Water Material Ote-Time (EP-4 1)
Pressure (Internal)
Inspection A
Boundary Program
- 15. In Table 3.1.2-1, on page 3.1-65, the 2 nd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Valve Body (Spatial)
Stainless Water Program V.A-28 Steel Water Cracking (E-12) 3.2.1-48 Pressure
>1400 F One-Time Boundary (Internal)
Inspection A
I Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 6 of 48
- 16. In Table 3.2.1, on page 3.2-29, line item 3.2.1-48 is revised as follows (note that this table was previously revised in SBK-L-l11015, Letter dated February 3, 2011 as item 1, on page 19 of Enclosure 1):
3.2.1-48 Stainless steel or stainless-steel-clad steel piping, piping components, piping elements, and tanks (including safety injection tanks/accumulators) exposed to treated borated water
>60 0C (>140 0F)
Cracking due to stress corrosion cracking Water Chemistry No Components in the Reactor Coolant system have been aligned to this line item based on material, environment, and aging effect.
Consistent with NUREG-180 1. The Water Chemistry Program, B.2.1.2, will be used to manage cracking due to stress corrosion cracking in stainless steel piping components exposed to treated borated water
>140'F in the Reactor Coolant and Residual Heat Removal systems-,and stainless steel heat exehanger eecmpcnfents exposed to treated
ý.-
I bor-aca wa'er -i 1 I tI c
-I esl aUa Heat Removal system.
The One-Time Inspection Program, B.2.1.20, will be used to verify the effectiveness of the Water Chemistry Program, B.2.1.2, to manage cracking of stainless steel piping components exposed to treated borated water >1401F in the in the Reactor Coolant and Residual Heat Removal systems
- 17. In Table 3.2.1, on page 3.2-29, line item 3.2.1-49 is revised as follows:
3.2.1-49 Stainless steel Loss of material Water No Components in the Reactor Coolant piping, piping due to pitting Chemistry system have been aligned to this line components, piping arid crevice item based on material, environment, elements, and tanks corrosion and aging effect.
exposed to treated borated water Consistent with NUREG-1801. The Water Chemistry Program, B.2.1.2, will be used to manage loss of material due to pitting and crevice corrosion in stainless steel piping components exposed to treated borated water in the Containment Building Spray, Reactor Coolant, Residual Heat Removal, and Safety Injection systems, and stainless steel heat exchanger components exposed to treated borated water in the Containment Building Spray, Reactor Coolant, and Residual Heat Removal
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure I Page 7 of 48 systems, and stainless steel tanks exposed to treated borated water in the Containment Building Spray, Reactor Coolant, and Safety Injection systems.
The One-Time Inspection Program, B.Z.1.20, will be used to verify the effectiveness of the Water Chemistry Program, B.2.1.2, to manage loss of material of stainless steel piping components in the Containment Building Spray, Reactor Coolant, Residual Heat Removal, and Safety Injection systems, and stainless steel heat exchanger components exposed to treated borated water in the Containment Building Spray, Reactor Coolant, and Residual Heat Removal systems, and stainless steel tanks exposed to treated borated water in the Containment Building Spray, Reactor Coolant, and Safety Injection systems.
- 18. In Section 3.2.2.2.4.2, on page 3.2-12, further evaluation discussion is revised as follows:
te. Number-3.2.1 10 is net applicable to Seabr..k Stati.n. The Engineefing Safety Features do net eontaint stainless steel heat exchanger tubes exiposed to treated water-.
Seabrook Station will implement the One-Time Inspection Program, B.2.1.20, to verify the effectiveness of the Water Chemistry Program, B.2.1.2, to manage reduction of heat transfer due to fouling in stainless steel heat exchanger tubes exposed to treated borated water in the Containment Building Spray, and, Residual Heat Removal systems.
- 19. In Table 3.2.1, on page 3.2-18, line item 3.2.1-10 is revised as follows:
3.2.1-10 Stainless steel heat exchanger tubes exposed to treated water Reduction of heat transfer due to fouling Water Chemistry and One-Time Inspection
- Yes, detection of aging effects is to be evaluated Net applicable. The Enginecring Safctly Features systems do not contain stainless steel heat exehange tubes exposed to treated water.
The One-Time Inspection Program, B.2.1.20 will be used to verify the effectiveness of the Water Chemistry Program, B.2.1.2, to manage reduction of heat transfer due to fouling in stainless steel heat exchanger tubes exposed to treated borated water in the Containment Building Spray and Residual Heat Removal systems.
See Subsection 3.2.2.2.4.2.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 8 of 48
- 20. In Table 3.2.2-2, on page 3.2-44, the 7th row is revised as follows:
Water Treated Chemistry A
Expansion Pressure Stainless Borated Loss of Program V.A-27 Joint Boundary Steel Water Material One-Time (EP-41)
(Internal)
Inspection A
Program
- 21. In Table 3.2.2-2, on page 3.2-45, the 1st row is revised as follows:
Water Treated Chemistry A
Filter Leakage Stainless Borated Loss of Program V.A-27 Housing Boundary Steel Water Material (EP-41) 3.2.1-49 (Spatial)
(Internal)
One-Time Inspection A
I Program
- 22. In Table 3.2.2-2, on page 3.2-45, the 4 th row is revised as follows:
Water Treated Chemistry A
Flexible.
Leakage Stainless Borated Loss of Program V.A-27 Hose Boundary Steel Water Material One-Time (EP-41) 3.2.149 (Spatial)
(internal)
InsecTimn InspectionA Programn
- 23. In Table 3.2.2-2, on page 3.2-46, the 2"d row is revised as follows:
Heat Water Exchanger Treated Chemistry C
Components Pressure Stainless Borated Loss of Program V.A-27 3.2.149 (l-CBS-E-16A Boundary Steel Water Material One-Time (EP-41) and 16B (Internal)
Inectioe Channel Head) nspection
________Program_________
- 24. In Table 3.2.2-2, on page 3.2-47, the 4 th row is revised as follows:
Heat Water Exchanger Heat Treated Chemistry C
Components Transfer Stainless Borated Loss of Program V.A-27 3.2.149 (1-CBS-E-16A Pressure Steel Water Material One-Time (EP-41) and 16B Boundary (Internal)
Inspection C
Tubes)
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 9 of 48
- 25. In Table 3.2.2-2, on page 3.2-47, a new row is added after the 4th row as follows:
- 26. In Table 3.2.2-2, on page 3.2-47, the 6h row is revised as follows:
Heat Water Exchanger Transfer Treated Chemistry C
Components Stainless Borated Loss of Progra V.A-27 3.2.1-49 (1-CBS-E-16A Pressure Steel Water Material One-Time (EP-41) and 16B Tube Boundary (Internal)
Inspection C
Sheet)
_Program
- 27. In Table 3.2.2-2, on page 3.2-48, the 4th row is revised as follows:
Heat Water Exchanger Heat Treated Chemistry C
Components Transfer Stainless Borated Loss of Program V.A-27 3.2.1-49 (I-CBS-P-9A Pressure Steel Water Material One-Time (EP-41) and 9B Pump Boundary (Internal)
Inspection C
Cooler Tubes)
Program C
- 28. In Table 3.2.2-2, on page 3.2-48, a new row is added after the 4th row as follows:
Heat Water Exchanger Heat Treated Chemistry A
Components Transfer Stainless Borated Reduction Program V.A-16 (1-CBS-P-9A o Heat 3.2.1-10 and 9B Pump Pressure (Internal)
Transfer One-Time Cooler Boundary Inspection A
Tubes)
Program
- 29. In Table 3.2.2-2, on page 3.2-49, the 2"d row is revised as follows:
Water Treated Chemistry A
Instrumentation Leakage Stainless Borated Loss of Program V.A-27 Element Boundary Steel Water Material (EP-41) 3.2.1-49 (Spatial)
(Internal)
One-cTime Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123.! Enclosure 1 Page 10 of 48
- 30. In Table 3.2.2-2, on page 3.2-50, the 6 th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Loss of Program V.A-27 3.2.1-49 Fittings Steel Water Material One-Time (EP-4 1)
Pressure (Internal)
Inspection A
Boundary Program
- 31. In Table 3.2.2-2, on page 3.2-52, the 3rd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Pump Casing (Spatial)
Stainless Borated Loss of Program V.A-27 Steel Water Material One-Time (EP-41)
Pressure (Internal)
Inspectioni Boundary Program
- 32. In Table 3.2.2-2, on page 3.2-53, the 4 th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Tank (Spatial)
Stainless Borated Loss of Program V.A-27 3.2.1-49 Steel Water Material One-Time (EP-4 1)
Pressure (Internal)
Inspection Boundary peon A
Program_________
th
- 33. In Table 3.2.2-2, on page 3.2-54, the 4 row is revised as follows:
Water Chemistry A
Pressure Stainless Treated Loss of Program V.A-27 Thermowell Water 3.2 1-49 Boundary Steel (Internal)
Material One-Time (EP-41)
Inspection A
Program
- 34. In Table 3.2.2-2, on page 3.2-54, the 8th row is revised as follows:
Water Treated Chemistry A
Valve Body Pressure Borated Loss of Program V.A-27 3.2.1-49 Boundary CASS Water Material One-Time (EP-4 1)
(Internal)
Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1
- 35. In Table 3.2.2-2, on page 3.2-55, the 6 th row is revised as follows:
Page I 1 of 48 Leakage Water Lekae CeityA Boundary Treated ChemistrA Valve Body (Spatial)
Stainless Borated Loss of Program V.A-27 Steel Water Material One-Time (EP-4 1)
Pressure (Internal)
Inspection A
Boundary
_Program
- 36. In Table 3.2.2-3, on page 3.2-60, the 1 st row is revised as follows:
Water Treated Chemistry A
Flexible Leakage Stainless Borated Loss of Program V.D1-30 Boundary (EP-41)49 (Spatial)
Water Material One-Time (Internal)
Inspection A
Program
- 37. In Table 3.2.2-3, on page 3.2-60, the 2nd row is revised as follows:
Water Treated Chemistry A
Flexible Leakage Stainless Borated Program V.D1-31 Hose Boundary Steel Water Cracking 3.2.1-48 (Spatial)
> 140 F One-Time (Internal)
Inspection A
Program
- 38. In Table 3.2.2-3, on page 3.2-60, the 5 th row is revised as follows:
Heat Water Exchanger Treated Chemistry C
Components Pressure Stainless Borated Loss of Program V.D1-30 (1-RH-E-9A Boundary Steel Water Material (EP-41) 3.2.1-49 and 9B One-Time Channel (Internal)
Inspection C
Head)
Program
- 39. In Table 3.2.2-3, on page 3.2-60, the 6 th row is revised as follows:
Heat Water Exchanger Treated Chemistry C
Components Pressure Stainless Borated Program V.11-31 (1-RH-E-9A Boundary Steel Water Cracking (E-12) 3.2.1-48 and 9B
>140 0 F One-Time
(
Channel (Internal)
Inspection C
Head)
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 12 of 48
- 40. In Table 3.2.2-3, on page 3.2-61, the 4 th row is revised as follows:
Heat Water Exchanger Heat Treated Chemistry C
Components Transfer Stainless Borated Program V.D1-31 (1-RH-E-9A Steel Water Cracking (E-12) 3.2.148 Pressure
> 1400 F One-Time E
Tubes)
Boundary (Internal)
Inspection Program
- 41. In Table 3.2.2-3, on page 3.2-61, the 5 t row is revised as follows:
Water Heat Heat Chmsr C
Exchanger Transfer Treated Chemistry Components Stainless Borated Loss of Program V.D1-30 3.2.1-49 (1 -RH-E-9A Pressure Steel Water Material One-Time (EP4 1) and 9B Boundary (Internal)
Inspection C
Tubes)
Program
- 42. In Table 3.2.2-3, on page 3.2-61, a new row is added after the 5th row as follows:
Heat Water Exchanger Heat Treated Chemistry A
Components Transfer Stainless Borated Reduction Program V.A-16 IoRHE9A Steel Water Of Heat (EP-34) 3.2.1-10 and 9B Pressure (Internal)
Transfer One-Time Tubes)
BIundany Inspection A
Program
- 43. In Table 3.2.2-3, on page 3.2-61, the 8th row is revised as follows:
Heat Water Exchanger Heat Treated Chemistry C
Components Stainless Borated Loss of Program V.D130 2 1-49 (1-RH-E-9A Steel Water Material Ote-Time (EP-41) and 9B Pressure (Internal)
OeTm Tube Sheet)
Boundary Inspection C
Tube Sheet)_
Program
- 44. In Table 3.2.2-3, on page 3.2-62, the 1st row is revised as follows:
Heat Water Exchanger Treated Chemistry C
Components Pressure Stainless Borated Program V.D 1-31 (1-RH-E-9A Boundary Steel Water Cracking (E-12) 3.2.1-48 and E9B b o
>1400 F One-Time Sheet)
(Internal)
Inspection C
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 13 of 48
- 45. In Table 3.2.2-3, on page 3.2-62, the 8th row is revised as follows:
Heat Water Exchanger Heat Treated Chemistry C
Components Transfer Stainless Borated Loss of Program V.D1-30 3.2.1-49 (1-RH-E-188A Pressure Steel Water Material One-Time (EP-41) and 188B Boundary (Internal) ispection C
Tubes)
Program
- 46. In Table 3.2.2-3, on page 3.2-63, the 1st row is revised as follows:
Heat Water Exchanger Heat Treated Chemistry C
Components Transfer Stainless Borated Program V.Dr-31 (C-RH-E-188A Steel Water Cracking (E-12) 3.2.1-48 and>188 Pressure
>1400 F One-Time E
Tubes)
Boundary (Internal)
Inspection Program
- 47. In Table 3.2.2-3, on page 3.2-63, the 4 th row is revised as follows:
Water Treated Chemistry A
Instrumentation Leakage Stainless Borated Loss of Program V.D1-30 Element Boundary Steel Water Material (EP-41) 3.2.1-49 (Spatial)
(Internal)
One-Time Inspection A
Program
- 48. In Table 3.2.2-3, on page 3.2-63, the 5th row is revised as follows:
Water Treated Chemistry A
Instrumentation Leakage stainless Borated Program V.D1-31 ElementBoundary Steel Water Cracking 3.2.148 (Spatial)
>1400 F One-Time (Internal)
Inspection A
Program
- 49. In Table 3.2.2-3, on page 3.2-64, the 4h row is revised as follows:
Water Pressure Treated Chemistry A
Orifice Boundary Stainless Borated Loss of Program V.D1-30 3.2.1-49 Steel Water Material One-Time (EP-41)
Throttle (Internal)
Inspection Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 14 of 48
- 50. In Table 3.2.2-3, on page 3.2-64, the 5 th row is revised as follows:
Water Pressure Treated Chemistry A
Orifice Boundary Stainless Borated Program V.D1-31 OiieSteel Water Cracking (E-12) 3.2.1-48 Stee
>1400 F One-Time (Internal)
Inspection A
Program
- 51. In Table 3.2.2-3, on page 3.2-65, the 1st row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Loss of Program V.D I-30 3.2.1-49 Fittings Steel Water Material One-Time (EP-41)
Pressure (Internal)
Inspection A
Boundary Program
- 52. In Table 3.2.2-3, on page 3.2-65, the 2 nd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Program V.D1-31 Fittings Steel Water Cracking (E-12) 3.2.1-48 FitigsSe
>1400 F One-Time Pressure Boundary (Internal)
Inspection A
Program
- 53. In Table 3.2.2-3, on page 3.2-68, the 1 st row is revised as follows:
Water Treated Chemistry A
Pump Pressure Stainless Borated Loss of Program V.D1-30 3.2.1-49 Casing Boundary Steel Water Material One-Time (EP-41)
(Internal)
Inspection A
Program
- 54. In Table 3.2.2-3, on page 3.2-68, the 2 nd row is revised as follows:
Water Treated Chemistry A
Pressure Stainless Borated Program V.D1-31 Boundary Steel Water Cracking (E-12) 3.2.1-48
>1400 F One-Time (Internal)
Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 15 of 48
- 55. In Table 3.2.2-3, on page 3.2-68, the 5 th row is revised as follows:
Water Treated Chemistry A
Pressure Stainless Borated Loss of Program V.D1-30 Thermowell 3.2.1-49 Boundary Steel Water Material One-Time (EP-4 1)
(Internal)
Inspection A
Program
- 56. In Table 3.2.2-3, on page 3.2-68, the 6h row is revised as follows:
Water Treated Chemistry A
Pressure Stainless Borated Program V.D1-31 Thermowell Boundary Steel Water Cracking (E-12) 3.2.1-48
> 1400 F One-Time (Internal)
Inspection A
Program
- 57. In Table 3.2.2-3, on page 3.2-68, the 9th row is revised as follows:
Water Treated Chemistry A
Valve Body Pressure ASS Borated Loss of Program V.D1-30 3.2.1-49 Boundary C
Water Material One-Time (EP-41)
(Internal)
Inspection A
Program
- 58. In Table 3.2.2-3, on page 3.2-69, the 1st row is revised as follows:
Water Treated Chemistry A
Pressure Borated Valve Program V.D11-31 Valve Body Boundary CASS Water Body (E-12) 3.2.1-48
> 1400 F One-Time (Internal)
Inspection A
Program
- 59. In Table 3.2.2-3, on page 3.2-69, the 5th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Valve Body (Spatial)
Stainless Borated Loss of Program V.D1-30 3.2.1-49 Steel Water Material One-Time (EP-41)
Pressure (Internal)
Inspection A
Boundary
_Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 16 of 48
- 60. In Table 3.2.2-3, on page 3.2-69, the 6th row is revised as follows:
Water Leakage Aae Boundary Treated Chemistry A
(Spatial)
Stainless Borated Program 3
-48 Valve Body Steel Water Cracking 1
3.2.1 Steel
>1400 F One-Time Pressure Boundary (Internal)
Inspection A
Program
- 61. In Table 3.2.2-4, on page 3.2-76, the 1st row is revised as follows:
Leakage Water Boundary Chemistry A
(Spatial)
Treated Chem Orifice Stainless Borated Loss of Program V.D1-30 3.2.1-49 Pressure Steel Water Material One-Time (EP-4 1)
Boundary (Internal)
Inspection A
Throttle Program
- 62. In Table 3.2.2-4, on page 3.2-77, the 3rd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Loss of Program V.D1-30 3.2.1-49 Fittings Steel Water Material One-Time (EP-4 1)
Pressure (Internal)
Inspection A
Boundary Program
- 63. In Table 3.2.2-4, on page 3.2-80, the 4 th row is revised as follows:
Water Treated Chemistry A
Pump Casing Pressure Stainless Borated Loss of Program V.D1-30 3.2.1-49 Boundary Steel Water Material One-Time (EP-41)
(Internal)
Inspection A
Program
- 64. In Table 3.2.2-4, on page 3.2-81, the 2 nd row is revised as follows:
Water Steel Treated Chemistry A
Pressure with Borated Loss of Program V.D1-30 Tank Stainless Water Material 3.2.1-49 Boundary Sel Wtr Mtra On-me (EP-41)
Steel (Internal)
OeTm Cladding Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 17 of 48
- 65. In Table 3.2.2-4, on page 3.2-8 1, the 6th row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program V.D1-30 Thermowell Boundary (EP41) 3.2.1-49 Throel (Spatial)
Steel Water Material One-Time (pta)(Internal)
One-Time Inspection A
Programn
- 66. In Table 3.2.2-4, on page 3.2-82, the 2nd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Valve Body (Spatial)
CASS Borated Loss of Program V.D1-30 3.2.149 Water Material One-Tine (EP-41)
Pressure (Internal)
Inspection A
Boundary Program
- 67. In Table 3.2.2-4, on page 3.2-82, the 6 th row is revised as follows:
- 68. In Section 3.3.2.2.2, on page 3.3-67, the 2nd paragraph of the Reduction of Heat Transfer Due to Fouling discussion is revised as follows:
ID 1 1 A
-lg
~tpn N,..n,~p..
4.1 I
.1
.,,~t
~
t,.r fln...I.n.-.,
'.2.,ntp,-.,.r,
~
~
Jt.flhti¶Chflhl tJ*.~i
.J.3.
1
.3 149 fll9.Jt IAFFIItaIJI9..'
19.31 1 LLLflIIILLI J LSJ JL*.'11149
.d9.JIIIy9.J11%flILO at 1J1/44flJIXJLfl
~
I-
'TTT"T1 J1N 1 fl l
I*--
TTTY~
I'
- 1 buttl0ft. tRiS ittle assfeetat 7
T r..
i l-i Vm ii.r
,3.
is
.AAA!
PA"W
+A-+---I----------~~-
-AA-----F ------------'4-Y---------------
Seabrook Station will implement the One-Time Inspection Program, B.2.1.20, to verify the effectiveness of the Water Chemistry Program, B.2.1.2, to manage reduction of heat transfer due to fouling in stainless steel heat exchanger tubes exposed to treated borated water in the Chemical and Volume Control, Spent and Fuel Pool Cooling systems and Reactor Coolant systems. The One-Time Inspection and Water Chemistry programs are described in Appendix B.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 18 of 48
- 69. In Table 3.3.1, on page 3.3-86, line item 3.3.1-3 is revised as follows:
3.3.1-3 Stainless steel heat exchanger tubes exposed to treated water Reduction of heat transfer due to fouling Water Chemistry and One-Time Inspection Yes, detection of aging effects is to be evaluated Not applicable for Auxiliary Systems eomponents at Seabrol Station. This line item is asso-iated w'ith NUREG 1801 line item X'H.E3 6, whieh is applicable to BWR Reactor-Water-Cleanup System heat exchanger-s.
The One-Time Inspection Program, B.2.1.20 will be used to verify the effectiveness of the Water Chemistry Program, B.2.1.2, to manage reduction of heat transfer due to fouling in stainless steel heat exchanger tubes exposed to treated borated water in the Containment Building Spray, Chemical and Volume Control, Residual Heat Removal, Spent Fuel Pool Cooling, and Reactor Coolant systems.
See subsection 3.3.2.2.2.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 19 of 48
- 70. In Table 3.3.1, on page 3.3-121, line item 3.3.1-90 is revised as follows:
3.3.1-90 Stainless steel and steel with stainless steel cladding piping, piping components, piping elements, tanks, and fuel storage racks exposed to treated borated water >60'C
(>140 0F)
Cracking due to stress corrosion cracking Water Chemistry No Consistent with NUREG-1801. The Water Chemistry Program, B.2.1.2, will be used to manage cracking due to stress corrosion cracking of the stainless steel piping components exposed to treated borated water >60'C
(> 140'F) in the Chemical and Volume Control, Sample, and Valve Stem Leak-Off systems and stainless steel tanks in the Chemical and Volume Control system.
In addition The Water Chemistry Program, B.2.1.2, will be used to manage cracking due to stress corrosion cracking of the stainless steel fuel storage rack support exposed to treated borated water >60'C (> 140'F) in Section 3.5, table 3.5.2.6, Supports.
The One-Time Inspection Program, B.2.1.20, will be used to verify the effectiveness of the Water Chemistry Program, B.2.1.2, to manage cracking of stainless steel piping and heat exchanger components exposed to treated borated water >601C (>140 °F) in the Chemical and Volume Control, Sample, and Valve Stem Leak-Off systems and stainless steel tanks in the Chemical and Volume Control system
- 71. In Table 3.3.1, on page 3.3-121, line item 3.3.1-91 is revised as follows (note that this line item was previously revised in SBK-L-1 1069 dated April 22, 2011, item b of Enclosure 2, on page 11):
3.3.1-91 Stainless steel and steel with stainless steel cladding piping, piping components, and piping elements exposed to treated borated water Loss of material due to pitting and crevice corrosion Water Chemistry No Components in the Auxiliary Steam, Chemical and Volume Control System, Sample, Spent Fuel Pool Cooling, and Waste Processing Liquid Drains systems have been aligned to this line item based on material, environment, and aging effect.
Consistent with NUREG-1801. The Water Chemistry Program, B.2.1.2, will be used to manage loss of material due to pitting and crevice corrosion of the following stainless steel components exposed to treated borated water:
a) Stainless steel piping components exposed to treated borated water in the
United States Nuclear Regulatory Commission Page 20 of 48 SBK-L-12123 / Enclosure 1 Auxiliary Steam, Boron Recovery, Chemical and Volume Control, Nitrogen Gas, Reactor Make-Up Water, Release Recovery, Resin Sluicing, Sample, Spent Fuel Pool Cooling, Valve Stem Leak-Off, Vent Gas, Waste Gas, and Waste Processing Liquid Drains systems, b) Stainless steel heat exchanger components exposed to treated borated water in the Chemical and Volume Control, Spent Fuel Pool Cooling, and Waste Processing Liquid Drains
- system, c) Stainless steel tanks exposed to treated borated water in the Chemical and Volume Control, Sample, Spent Fuel Pool Cooling, and Waste Processing Liquid Drains.
The One-Tine Inspection Program, B.2.1.20, will be used to verify the effectiveness of the Water Chemistry Program, B.2.1.2, to manage loss of material due to pitting and crevice corrosion of thefollowing stainless steel components exposed to treated borated water:
a) Stainless steel piping components exposed to treated borated water in the Auxiliary Steam, Boron Recovery, Chemical and Volume Control, Nitrogen Gas, Reactor Make-Up Water, Release Recovery, Resin Sluicing, Sample, Spent Fuel Pool Cooling, Valve Stem Leak-Off, Vent Gas, Waste Gas, and Waste Processing Liquid Drains systems, b) Stainless steel heat exchanger components exposed to treated borated water in the Chemical and Volume Control, Spent Fuel Pool Cooling, and Waste Processing Liquid Drains
- system, c) Stainless steel tanks exposed to treated borated water in the Chemical and Volume Control, Sample, Spent Fuel Pool Cooling, and Waste Processing Liquid Drains systems.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1
- 72. In Table 3.3.2-2, on page 3.3-136, the 4 th row is revised as follows:
Page 21 of 48 Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.El-17 Flexible Hose Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
Otte-Time Inspection A
Program
- 73. In Table 3.3.2-2, on page 3.3-137, the 2nd row is revised as follows:
- 74. In Table 3.3.2-2, on page 3.3-138, the 4 th row is revised as follows:
Water Treated Chemistry A
Leakage Borated Loss of Program VII.E1-17 Valve Body Boundary CASS Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
Oae-Time Inspection A
Program
- 75. In Table 3.3.2-2, on page 3.3-138, the 8th row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.E1-17 Valve Body Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program
- 76. In Table 3.3.2-3, on page 3.3-141, the 1st row is revised as follows:
Water Treated Chemistry A
Pressure Stainless Borated Program VII.El-20 Filter Housing Boundary Steel Water Cracking (AP-82) 3.3.1-90
> 140OF One-Time (Internal)
Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 22 of 48
- 77. In Table 3.3.2-3, on page 3.3-141, the 3rd row is revised as follows:
- 78. In Table 3.3.2-3, on page 3.3-143, the 4th row is revised as follows:
- 79. In Table 3.3.2-3, on page 3.3-146, the 6th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VILE 1-17 Components (AP-79) 3.3.1-91 (CS-E-3 Boundary Steel Water Material One-Time Channel Head)
(Internal)
Inspection C
Program
- 80. In Table 3.3.2-3, on page 3.3-147, the 6 th row is revised as follows:
Water Heat Heat Treated Chemistry C
Exchanger Transfer Stainless Borated Loss of Program Vis.EI-I7 Components Steel Water Material (AP-79) 3.3.1-91 (CS-E-3 Pressure Se Wter One-Time Tubes)
Boundary (Internal)
Inspection C
Program
- 81. In Table 3.3.2-3, on page 3.3-147, a new row is added after the 6th row as follows:
Water Heat Heat Treated Chemistr.
A Exchanger Transfer Stainless Borated Reduction Program VII.A4-4 Sanes Brtd of Heat3..-
Components Steel Water (AP-62)
(CS-E-3 Pressure SteelrWate Transfer One-Time Tubes)
Boundary (Iternal Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 23 of 48
- 82. In Table 3.3.2-3, on page 3.3-148, the 2 nd row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-3 Tube (Itra)One-Time Sheet)
(Internal)
Inspection C
Program
- 83. In Table 3.3.2-3, on page 3.3-148, the 7 th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3Tie1-91 (CS-E-4 (nenl teTm Channel Head)
(Internal)
One-Tiec I Program I
- 84. In Table 3.3.2-3, on page 3.3-149, the 7th row is revised as follows:
Water Heat Heat Treated Chemistry C
Exchanger Transfer TraePhmsryga Stainless Borated Loss of Program VII.El-17 Components Steel Water Material (AP-79) 3.3.1-91 (CS-E-4 Pressure (Internal)
One-Time Tubes)
Boundary Inspection C
Program
- 85. In Table 3.3.2-3, on page 3.3-149, a new row is added after the 7th row as follows:
Water Heat Heat Treated Chemistry A
Exchanger Transfer Stainless Borated Loss of Program VItA4-4 Components Steel Water Material (AP-62) 3.3.1-3 (CS-E-4 Pressure (Interna Me One-Time Tubes)
Boundary Inspection A
Program
- 86. In Table 3.3.2-3, on page 3.3-150, the 3rd row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-4 Tube (Itra)Otte-Time Sheet)
(Internal)
Inspection C
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 24 of 48
- 87. In Table 3.3.2-3, on page 3.3-150, the 8th row is revised as follows:
Heat Water Exchanger Treated Chemistry C
Components Pressure Stainless Borated Loss of Program VII.E1-17 3.3.1-91 (CS-E-5A and Boundary Steel Water Material One-Time (AP-79) 5B Channel (Internal)
Inspection Head)
Program
- 88. In Table 3.3.2-3, on page 3.3-151, the 8th row is revised as follows:
Water Heat Heat Treated Chemistry C
Exchanger Transfer Stainless Borated Loss of Program VIsIt-17 Components Steel Water Material (AP-79) 3.3.1-91 (CS-E-5A and Pressure Se Wter One-Time 5B Tubes)
Boundary (Internal)
Inspection C
_ _Program
- 89. In Table 3.3.2-3, on page 3.3-15 1, a new row is added after the 8th row as follows:
Water Heat Heat Treated Chemistry A
Exchanger Transfer Stainless Borated Reduction Program VII.A4-4 Components Steel Water of Heat (AP-62) 3.3.1-3 (CS-E-SA and Pressure Transfer One-Time 5B Tubes)
Boundary
(
Inspection A
Program
- 90. In Table 3.3.2-3, on page 3.3-152, the 4th row is revised as follows:
Heat Water Exchanger Treated Chemistry C
Components Pressure Stainless Borated Loss of Program VII.El-17 (CS-E-5A and Boundary Steel Water Material One-Time (AP-79) 5B Tube (Internal)
Inspection C
Sheet)
Program
- 91. In Table 3.3.2-3, on page 3.3-153, the 1st row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3.3199 (CS-E-6(Itra)OeTm Channel Head)
(Internal)
Onectiec Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 25 of 48
- 92. In Table 3.3.2-3, on page 3.3-153, the 6 th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.E1-17 3.3.1-91 Components Boundary Steel Water Material One-Tinte (AP-79)
(CS-E-6 Shell)
(Internal)
Inspection C
Program
- 93. In Table 3.3.2-3, on page 3.3-154, the 1st row is revised as follows:
Water Heat Heat Treated Chemistry C
Exchanger Transfer Stainless Borated Loss of Program VII.Ei-17 Components Steel Water Material (AP-79) 3.3.1-91 (CS-E-6 Pressure Se Wter One-Time Tubes)
Boundary (External)
Inspection C
Program
- 94. In Table 3.3.2-3, on page 3.3-154, a new row is added after the 1 st row as follows:
- 95. In Table 3.3.2-3, on page 3.3-154, the 4 th row is revised as follows:
Water Heat Heat Treated Chemistry C
Exchanger Transfer Stainless Borated Loss of Program VII.El-17 Components Steel Water Material (AP-79) 3.3.1-91 (CS-E-6 Pressure (SnterMal)
One-Time Tubes)
Boundary Inspection C
Program
- 96. In Table 3.3.2-3, on page 3.3-154, a new row is added after the 4th row as follows:
Water Heat Heat Treated Chemistry A
Exchanger Transfer Stainless Borated Reduction Program VI.A4-4 Components of Heat 3.3.1-3 (C--6 Pesue Steel Water
!f(AP-62)
(CS-E-6 Pressure (Internal)
Transfer One-Time Tubes)
Boundary Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 26 of 48
- 97. In Table 3.3.2-3, on page 3.3-154, the 7th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-6 Tube o
a eter One-Time Sheet)
(Extemal)
Inspection C
Program I
- 98. In Table 3.3.2-3, on page 3.3-155, the 2 nd row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-6 Tube (Itra)One-Time Sheet)
(Internal)
Inspection C
Program
- 99. In Table 3.3.2-3, on page 3.3-155, the 7th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.E 1-7 Components Boundary Steel Water Material One-Time3(AP179 Sheet)
(Internal)
Inspection C
Program 100.
In Table 3.3.2-3, on page 3.3-156, the 6th row is revised as follows:
Water Heat Heat Treated Chemistry C
Exchanger Transfer Stainless Borated Loss of Program VII.El-17 Components Steel WBrated terial (AP-79) 3.3.1-91 (CS-E-7 Pressure (Internal)
One-Time Tubes)
Boundary Inspection C
Program 101.
In Table 3.3.2-3, on page 3.3-156, the 6'h row is revised as follows:
Water Heat Heat Treated Chemistry A
Exchanger Transfer Stainless Borated Reduction Program VII.A4-4 Components Steel Water of Heat (AP-62) 3.3.1-3 (CS-E-7 Pressure (Snternate Transfer One-Time Tubes)
Boundary Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 27 of 48 102.
In Table 3.3.2-3, on page 3.3-157, the 1st row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material One-Time A-79)
Sheet)
(Internal)
Inspection C
Program 103.
In Table 3.3.2-3, on page 3.3-157, the 6 th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-8 (Inter One-Time Channel Head)
(Internal)
Inspection C
Program 104.
In Table 3.3.2-3, on page 3.3-158, the 3rd row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.El-17 3.1-91 Components Boundary Steel Water Material One-Time (AP-79)
(CS-E-8 Shell)
(Internal)
Inspection C
Program 105.
In Table 3.3.2-3, on page 3.3-158, the 6th row is revised as follows:
Water Heat Heat Treated Chemistry C
Exchanger Transfer Stainless Borated Loss of Program VIiIt-17 Components Stael Bated Loss (AP-79) 3.3.1-91 (CS-E-8 Pressure Steel Water Material One-Time Tubes)
Boundary (External)
Inspection C
I_
I_
I_
I Program I
106.
In Table 3.3.2-3, on page 3.3-158, a new row is added after the 6th row as follows:
Water Heat Heat Treated Chemistry A
Exchanger Transfer Stainless Borated Loss of Program VItA4-4 Components Steel Water Material (AP-62) 3.3.1-3 (CS-E-8 Pressure SeWter One-Time Tubes)
Boundary (External)
Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure I Page 28 of 48 107.
In Table 3.3.2-3, on page 3.3-159, the 1st row is revised as follows:
Water Heat Heat Treated Cryga Exchanger Transfer TreatdrChmisr tCopngenTs Stainless Borated Loss of VII.El-17 3
1-91 Components Steel Water Material (AP-79) 3.1 (CS-E-8 Pressure (Internal)
One-Time, Tubes)
Boundary Inspection C
Program 108.
In Table 3.3.2-3, on page 3.3-159, a new row is added after the I" row as follows:
Water Heat Heat Treated Chemistry A
Exchanger Transfer Stainless Borated Reduction Program VII.A4-4 Components Steel Water of Heat 3.3.1-3 (CS-E-8 Pressure (Internal Transfer One-Time Tubes)
Boundary Inspection A
I_
I_
I_
I_
I_ Program 1 1
109.
In Table 3.3.2-3, on page 3.3-159, the 4th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.E 1-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-8 Tube BEr Materia Otte-Time Sheet)
(External)
Inspection C
Program 110.
In Table 3.3.2-3, on page 3.3-159, the 7th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.E-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-8 Tube B
a eter One-Time Sheet)
(Internal)
Inspection C
Program 111.
In Table 3.3.2-3, on page 3.3-160, the 4th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Leakage Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-63 (Spatial)
(Internal)
One-Time Channel Head)
Inspection C
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 29 of 48 112.
In Table 3.3.2-3, on page 3.3-161, the 4th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Leakage Stainless Borated Loss of Program VII.El-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-64 (Spatial)
(Internal)
One-cTime Channel Head)
Inspection C
Program 113.
In Table 3.3.2-3, on page 3.3-162, the 1st row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Leakage Stainless Borated Loss of Program VIsEI-I7 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-64 (Spatial)
Se Wter One-Time Shell)
Inspection C
Program 114.
In Table 3.3.2-3, on page 3.3-162, the 6th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Leakage Stainless Borated Loss of Program VII.EI-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-65 (Spatial)
(Internal)
Onte-Time Channel Head)
(InternanC Program I
115.
In Table 3.3.2-3, on page 3.3-164, the 4th row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Leakage Stainless Borated Loss of Program VII.E1-17 Components Boundary Steel Water Material (AP-79) 3.3.1-91 (CS-E-139 (Spatial)
(Internal)
One-Time Shell)
Inspection C
Program 116.
In Table 3.3.2-3, on page 3.3-170, the 6th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Instrumentation (Spatial)
Stainless Borated Loss of Program VII.E-17 331-91 Element Steel Water Material One-Time (AP-79)
Pressure (Internal)
Boundary Inspection A
________Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 30 of 48 117.
In Table 3.3.2-3, on page 3.3-172, the 1st row is revised as follows:
Leakage Water Boundary Chemistry A
(Spatial)
Treated Chem Orifice Stainless Borated Loss of Program VII.EI-17 331-91 Pressure Steel Water Material One-Time (AP-79)
Boundary (Internal)
Inspection A
Throttle Program 118.
In Table 3.3.2-3, on page 3.3-174, the 1st row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping And (Spatial)
Stainless Borated Loss of Program VII.E1-17 3.3.1-91 Fittings Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program 119.
In Table 3.3.2-3, on page 3.3-174, the 2nd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping And (Spatial)
Stainless Borated Program VII.El-20 Fittings Steel Water Cracking (AP-82) 3.3.1-90 Pressure
> 140OF One-Time Boundary (Internal)
Inspection A
________Program 120.
In Table 3.3.2-3, on page 3.3-177, the 1st row is revised as follows:
Water Treated Chemistry A
Pump Casing Pressure CASS Borated Loss of Program VII.E 1-17 3.3.1-91 Boundary Water Material One-Time (AP-79)
(iternal)
Inspection A
Program 121.
In Table 3.3.2-3, on page 3.3-178, the 2"d row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Pump Casing (Spatial)
Stainless Borated Loss of Program VII.E1-17 3.3.1-91 Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 31 of 48 122.
In Table 3.3.2-3, on page 3.3-178, the 3rd row is revised as follows:
Water Leakage Treated Chemistry A
Boundary Stainless Borated Program VII.E-20
-90 Pump Casing (Spatial)
Steel Water Cracking (AP-82) 3.3.1
>140 OF One-Time (Internal)
Inspection A
Program 123.
In Table 3.3.2-3, on page 3.3-179, the 4h row is revised as follows:
Water Pump Casing Treated Chemistry A
(High Head Pressure Stainless Borated Loss of Program VII.EI-17 Centrifugal Boundary Steel Water Material One-Timi7e9)
Charging (Internal)
On-iispeoi Pump)
Inspection A
Program 124.
In Table 3.3.2-3, on page 3.3-180, the 1st row is revised as follows:
Leakage Water Boundary Treated Chemistry C
Tank (Spatial)
Stainless Borated Loss of Program VII.EI-17 3.3.1-91 Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspectioni I Boundary Program I
125.
In Table 3.3.2-3, on page 3.3-180, the 2"d row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Program VII.El-20 Tank Boundary Steel Water Cracking (AP-82) 3.3.1-90 (Spatial)
> 140 OF One-Time (Internal)
Inspection A
Program 126.
In Table 3.3.2-3, on page 3.3-181, the 5th row is revised as follows:
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 32 of 48 127.
In Table 3.3.2-3, on page 3.3-181, the 6th row is revised as follows:
Water Leakage WA Boundary Treated Chemistry Thermowell (Spatial)
Stainless Water Program VII.El-20 ThroelSteel Water Cracking (AP-82) 3.3.1-90
>140°F One-Time Pressure Boundary (Internal)
Inspection A
I_ Boundary Program 128.
In Table 3.3.2-3, on page 3.3-182, the 51h row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Valve Body (Spatial)
Borated Loss of Program VII.E1-17 CASS Water Material (AP-79) e.3.1-91 Pressure (Internal)
Inspection A
Boundary Program 129.
In Table 3.3.2-3, on page 3.3-182, the 6th row is revised as follows:
Leakage Water Luary Treated Chemistry A
Valaveaod Borated Program VIlE] -20 Valve Body (pta)
CASS Water Cracking (AP-82) 3.3.1-90 Pressure
> 140 OF One-Tine Boundary (Internal)
Inspection A
Program 130.
In Table 3.3.2-3, on page 3.3-184, the 2nd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Valve Body (Spatial)
Stainless Borated Loss of Program VII.E-17 Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program 131.
In Table 3.3.2-3, on page 3.3-184, the 3rd row is revised as follows:
Water Leakage Treated Chemistry A
Boundary Borated Program VII-20 Valve Body (Spatial)
Stainless Water Cracking (AP-82) 3.3.1-90 Steel
>140 OF Otte-Time Pressure Boundary (Internal)
Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 33 of 48 132.
In Table 3.3.2-24, on page 3.3-380, the 7th row is revised as follows:
Water Treated Chemistry A
Leakage Treatdrahmis F H eBnar Stainless Borated Loss of Program VII.El-17 3
Flexible Hose Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program 133.
In Table 3.3.2-24, on page 3.3-381, the 4th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Loss of Program VII.E1-17 Fittings Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program 134.
In Table 3.3.2-24, on page 3.3-384, the Ist row is revised as follows:
Leakage Boundary Water (Spatial)
Chemistry A
TreatedChmsrA Valve Body Pressure Stainless Borated Loss of Program VII.EI-17 3.3.1-91 Boundary Steel Water Material One-Time (AP-79)
Structural Inspection A
Integrity Program (Attached)_
135.
In Table 3.3.2-3 1, on page 3.3-429, the 5th row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.E1-17 Orifice Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
Onte-Time Inspection A
I Program
United States Nuclear Regulatory Commission Page 34 of 48 SBK-L-12123 / Enclosure 1 136.
In Table 3.3.2-31, on page 3.3-430, the 2nd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Loss of Program VII.EI-17 3.3.1-91 Fittings Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program 137.
In Table 3.3.2-3 1, on page 3.3-432, the 3rd row is revised as follows:
138.
In Table 3.3.2-3 1, on page 3.3-433, the 2nd row is revised as follows:
139.
In Table 3.3.2-32, on page 3.3-435, the 7th row is revised as follows:
Water Treated Chemistry A
Piping and Leakage Stainless Borated Loss of Program VII.El-17 Fittings Boundary Steel Water Material One-Time (AP-79)
(Internal)
Inspection A
Program 140.
In Table 3.3.2-32, on page 3.3-437, the 4 th row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.El-17 Valve Body Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 35 of 48 141.
In Table 3.3.2-33, on page 3.3-439, the 4th row is revised as follows:
142.
In Table 3.3.2-33, on page 3.3-440, the 1st row is revised as follows:
Leakage Water Boundary Treated Chemistry Valve Body (Spatial)
CASS Borated Loss of Program VII.El-17 3.3.1-91 Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program 143.
In Table 3.3.2-35, on page 3.3-446, the 2 d row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Program VII.El-20 Flexible Hose Boundary Steel Water Cracking (AP-82) 31-90 (Spatial)
>140 OF One-Time (Internal)
Inspection A
Program 144.
In Table 3.3.2-35, on page 3.3-446, the 3 row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.A3-8 Flexible Hose Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program 145.
In Table 3.3.2-35, on page 3.3-449, the 6th row is revised as follows:
Water Treated Chemistry A
Instrumentation Leakage Stainless Borated Loss of Program VII.A3-8 Element Boundary Steel Water Material One(TP-)e (A.79)
(Internal)
Inspection A
Program
United States Nuclear Regulatory Commission Page 36 of 48 SBK-L-12123 / Enclosure 1 146.
In Table 3.3.2-35, on page 3.3-450, the 2 nd row is revised as follows:
Water Leakage Treated Chemistry A
Boundary Piping and (Spatial)
Stainless Borated Program VILE 1-20 Water Cracking 1-8) 3.3.1-90 Fittings Steel
>140 OF One-Time (A-82)
Pressure Boundary (Internal)
Inspection A
Program 147.
In Table 3.3.2-35, on page 3.3-450, the 3 d row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Loss of Program VII.A3-8 3.31-91 Fittings Steel Water Material One-Time (AP-79)
Pressure (Internal)
InecTioA BonayInspection A
Boundary Program 148.
In Table 3.3.2-35, on page 3.3-45 1, the 9 th row is revised as follows:
Water Treated Chemistry' C
Tannary Stainless Borated Loss of Program VIL.A3-8 3
Tank Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection C
Program 149.
In Table 3.3.2-35, on page 3.3-452, the 4th row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.A3-8 Thermowell Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program 150.
In Table 3.3.2-35, on page 3.3-453, the 4th row is revised as follows:
Water Treated Chemistry A
Leakage Borated Loss of Program V11.A3-8 Valve Body Boundary CASS Bated ProgrmalI(AP-7 3.3.1-91 (Spatial)
Water Material Ote-Time (Internal)
Inspection A
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 37 of 48 151.
In Table 3.3.2-35, on page 3.3-454, the 1st row is revised as follows:
Water Leakage Aae Boundary Treated Chemistry A
(Spatial)
Stailess Borated Program Valve Body St Water Cracking (AP-2) 3.3.1-90 Steel
>140 "F One-Time (AP-82)
Pressure Boundary (Internal)
Inspection A
Program 152.
In Table 3.3.2-35, on page 3.3-454, the 2nd row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Valve Body (Spatial)
Stainless Borated Loss of Program VII.A3-8 Steel Water Material OneTine (AP-79)
Pressure (Internal)
Inspection Boundary PronaA
________Programn 153.
In Table 3.3.2-39, on page 3.3-482, the 9th row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VJI.A3-8 Filter Housing Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program I
154.
In Table 3.3.2-39, on page 3.3-483, the 6th row is revised as follows:
Heat Water Exchanger Treated Chemistry C
Components Pressure Stainless Borated Loss of Program VII.A3-8 3.3.1-91 (SF-E-15A &
Boundary Steel Water Material One-Tine (AP-79)
B Channel (Internal)
Inspection Head)
Program 155.
In Table 3.3.2-39, on page 3.3-484, the 1st row is revised as follows:
Heat Water Exchanger Steel Treated Chemistry Components Pressure With Borated Loss of Program VIIA3-8 (SFE-5A Bundry Stainless Woater Matria of Pone-T VIIA3e (SF-E-15A &
BoundaryWater Material (AP-79) 3.3.1-91 B Channel Steel (Internal)
One-Time Head Cover)
Cladding Inspection C
Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 38 of 48 156.
In Table 3.3.2-39, on page 3.3-484, the 7th row is revised as follows:
Water Heat Heat Treated Chemistry C
Exchanger Transfer Stainless Borated Loss of Program VII.A3-8 Components Stness ated ProgrmalI(AP-7 3.3.1-91 (SF-E-I 5A &
Pressure Steel Water Material One-Time (AP-79)
B Tubes)
Boundary (Internal)
Inspection C
Program 157.
In Table 3.3.2-39, on page 3.3-484, the 7th row is revised as follows:
Water Heat Heat Chemistry A
Exchanger Transfer Stiles Treated Reduction Program IA-Components Stainless Borated Reuto n
Progra VII.A4-4 3.-
(SF-E-15A &
Pressure Steel Water (AP-62) 3.3.1-3 B Tubes)
Boundary (Internal)
Transfer One-Time Inspection A
Program 158.
In Table 3.3.2-39, on page 3.3-485, the 1st row is revised as follows:
Water Heat Treated Chemistry C
Exchanger Pressure Stainless Borated Loss of Program VII.A3-8 Components dary Steel Water Material (AP-79) 3.3.1-91 (SF-E-15A &
Boun (Internal)
One-Time P
B Tubesheet)
Inspection C
Program 159.
In Table 3.3.2-39, on page 3.3-485, the 4 th row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Instrumentation (Spatial)
Stainless Borated Loss of Program VII.A3-8 3.31-91 Element Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program 160.
In Table 3.3.2-39, on page 3.3-486, the 1st row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Piping and (Spatial)
Stainless Borated Loss of Program VII.A3-8 3.3.1-91 Fittings Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program
United States Nuclear Regulatory Commission Page 39 of 48 SBK-L-12123 / Enclosure 1 161.
In Table 3.3.2-39, on page 3.3-487, the 4th row is revised as follows:
162.
In Table 3.3.2-39, on page 3.3-487, the 7th row is revised as follows:
Water Treated Chemistry C
Leakage Stainless Borated Loss of Program VII.A3-8 Tank Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-cTime Inspection C
Program 163.
In Table 3.3.2-39, on page 3.3-488, the 3 d row is revised as follows:
Leakage Water Boundary Treated Chemistry A
Thermowell (Spatial)
Stainless Borated Loss of Program VII.A3-8 3.3.1-91 Steel Water Material One-Time (AP-79)
Pressure (internal)
Inspection A
Boundary Program 164.
In Table 3.3.2-39, on page 3.3-489, the Ist row is revised as follows:
165.
In Table 3.3.2-39, on page 3.3-489, the 5 th row is revised as follows:
Water Leakage WA Boundary Treated ChemistrA Valve Body (Spatial)
Stainless Borated Loss of Program V1I.A3-8 3.3.1-91 Steel Water Material One-Time (AP-79)
Pressure (Internal)
Inspection A
Boundary Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 40 of 48 166.
In Table 3.3.2-41, on page 3.3-497, the 3'd row is revised as follows:
Water Treated Chemistry A
Piping and Lk Stainless Borated Loss of Program VII.A3-8 Fittings Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time hIspection A
Program 167.
In Table 3.3.2-41, on page 3.3-497, the 4 th row is revised as follows:
Water Treated Chemistry A
Piping and Leakage Stainless Borated Program VLI.E1-20 Boundary Water Cracking (AP-82) 3.3.1-90 Fittings (Spatial)
Steel
>140 OF One-Time A
(Internal)
Inspection A
Program 168.
In Table 3.3.2-42, on page 12 of Enclosure 2 of SBK-L-1 1069, dated 4-22-2011, item "e" is revised as follows:
Water Treated Chemistry A
Piping and Leakage Stainless Borated Loss of Program VII.EI-17 Fittings Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program 169.
In Table 3.3.2-42, on page 12 of Enclosure 2 of SBK-L-11069, dated 4-22-2011, item "f' is revised as follows:
Water Treated Chemistry A
Leakage Borated Loss of Program V1tErI-17 Valve Body Boundary CASS Bated ProgrialI(AP-7 3.3.1-91 (Spatial)
Water Material One-Time (Internal)
Inspection A
Program 170.
In Table 3.3.2-43, on page 3.3-503, the 71h row is revised as follows:
Water Treated Chemistry A
Piping and Leakage Stainless Borated Loss of Program VII.EI-17 Boundary 3.3.1-91 Fittings (Spatial)
Steel Water Material One-Time (AP-79)
S i(Internal)
Inspection A
Program
United States Nuclear Regulatory Commission Page 41 of 48 SBK-L-12123 / Enclosure I 171.
In Table 3.3.2-43, on page 3.3-505, the 1st row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.El-17 Valve Body Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-cTioe Inspection A
Program 172.
In Table 3.3.2-45, on page 3.3-515, the 5th row is revised as follows:
Water Heat Treated Chemistry C
D Stainless Borated Loss of Program VII.El-I 7 Components Boundary Steel Water Material (i3.3.1-91 (WLD-E-43 (Spatial)
(APe7)tern One-Time Channel Head)
(Internal)
Inspection C
Program 173.
In Table 3.3.2-45, on page 3.3-516, the 2nd row is revised as follows:
Water Treated Chemistry A
Instrumentation Leakage Stainless Borated Loss of Program VII.El-17 Element Boundary Steel Water Material (AP-79) 3.3.199 (Spatial)
(Internal)
One-cTime Inspection A
Program 174.
In Table 3.3.2-45, on page 3.3-517, the 3rd row is revised as follows:
175.
In Table 3.3.2-45, on page 3.3-518, the 5th row is revised as follows:
Water Piping and Treated Chemistry A
Fittings Pressure Stainless Borated Loss of Program VII.El-17 3.3.1-91 (Containment Boundary Steel Water Material One-Time (AP-79)
Isolation)
(Internal)
Inspection A
Program
United States Nuclear Regulatory Commission Page 42 of 48 SBK-L-12123 / Enclosure 1 176.
In Table 3.3.2-45, on page 3.3-519, the 5th row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.El-17 Pump Casing Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program 177.
In Table 3.3.2-45, on page 3.3-520, the 1st row is revised as follows:
Water Treated Chemistry C
Leakage Stainless Borated Loss of Program VII.E1-17 Tank Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Tione Inspection C
Program 178.
In Table 3.3.2-45, on page 3.3-520, the 7th row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.EI-17 Thermowell Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time Inspection A
Program 179.
In Table 3.3.2-45, on page 3.3-521, the 4th row is revised as follows:
Water Treated Chemistry A
Leakage Borated Loss of Program VII.E 1-17 Valve Body Boundary CASS Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
Oae-Time
)
Inspection A
Program 180.
In Table 3.3.2-45, on page 3.3-522, the 2nd row is revised as follows:
Water Treated Chemistry A
Leakage Stainless Borated Loss of Program VII.E1-17 Valve Body Boundary Steel Water Material (AP-79) 3.3.1-91 (Spatial)
(Internal)
One-Time A
Inspection Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 43 of 48 181.
In Table 3.3.2-45, on page 3.3-523, the 1st row is revised as follows:
182.
In Table 3.3.2-45, on page 3.3-523, the 5th row is revised as follows:
Water Treated Chemistry A
Valve Body Pressure Stainless Borated Loss of Program VII.El-17 (Containment Boundary Steel Water Material (AP-79) 3.3.1-91 Isolation)
(Internal)
One-Time Inspection A
I__
I_
_Program I
I 183.
In Table 3.4.2-1, on page 3.4-40, the 6h row is revised as follows:
Water Treated Chemistry A
Thermowell Pressure Stainless Borated Loss of Program VII.EI-17 3.3.1-91 Boundary Steel Water Material One-Time (AP-79)
(Internal)
Inspection A
Program 184.
In Section 3.5.2.1.5, on page 3.5.8, the following program has been added to the Section "Aging Management Programs" One-Time Inspection Program (B.2.1.20) 185.
In Section 3.5.2.1.6, on page 3.5.9, the following program has been added to the Section "Aging Management Programs" 0
One-Time Inspection Program (B.2.1.20)
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 44 of 48 186.
In Table 3.5.1, on page 3.5-46, line item 3.5.1-46 is revised as follows:
3.5.1-46 Group 5:
fuel pool liners Cracking due to stress corrosion cracking; loss of material due to pitting and crevice corrosion Water Chemistry and Monitoring of spent fuel pool water level and level of fluid in the leak chase channel No The spent fuel pool is normally maintained less than 140 0F, therefore Stress Corrosion Cracking is not an aging effect that requires management. Crevice and pitting corrosion are managed by the Water Chemistry Program, B.2.1.2. The One-Time Inspection Program, B.2.1.20, will be used to verify the effectiveness of the Water Chemistry Program.
187.
In Table 3.5.2-5, on page 3.5-223, the 4 th line is revised as follows:
Water PST - Stainless Treated Chemistry Steel -FSB-
- Shelter, Stainless Borated Program III.A5-13 Exposed to Protection Steel Water Cracking (T-14) 3.5.1-46 A, 507 Treated (Etra)One-Time Borated Water (External)
Inspection Program 188.
In Table 3.5.2-5, on page 3.5-223, the 5th line is revised as follows:
Water PST - Stainless Treated Chemistry Steel -FSB-
- Shelter, Stainless Borated Loss of Program III.A5-13 Exposed to Protection Steel Water Material (T-14) 3.5.1-46 A
Treated (Etra)One-Time Borated Water (External)Inspection Program 189.
In Table 3.5.2-6, on page 3.5-233, the 5 h line is revised as follows:
Water ASME Class Treated Chemistry 2/3 - Stainless Structural Stainless Borated Loss of Program III.A5-13 3.5.1-46 A
Steel - in Support Steel Water Material One-Time (T-14)
Treated Water (External)
Inspection Program
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 1 Page 45 of 48 190.
In Table 3.5.2-6, on page 3.5-238, the 6 th line is revised as follows:
Water Spent Fuel Treated Chemistry Rack Support -
Structural Stainless Borated Prograa VIkAg2-7 Stainless Steel Support Steel Water Cracking (A-97) 3.3.1-90 A, 507
- in Treated (Etra)One-Time Water (External)
Inspection Program 191.
In Table 3.5.2-6, on page 3.5-238, a new line is added after the 6th line as follows:
Water Spent Fuel Treated Chemistry Rack Support Structural Stainless Borated Loss of Program III.A5-13
- Stainless Support Steel Water AMaterial (T 14)
Steel - in Spt eWer Maei One-Time Treated Water (External Inspection Program
United States Nuclear Regulatory Commission Page 46 of 48 SBK-L-12123 / Enclosure 1 RAI B.2.1.9-2
Background:
GALL Report AMP XI.M18, "Bolting Integrity," manages aging of closure bolting for pressure retaining components. The program includes periodic inspection of closure bolting for indication of loss of preload, cracking, and loss of material due to corrosion, rust, etc. GALL Report AMP XI.M36, "External Surfaces Monitoring of Mechanical Components," manages loss of material, cracking, and change in material properties of component external surfaces during system inspections and walkdowns.
In recent reviews of license renewal applications and operating experience, the NRC staff noted that Seabrook Station may have used, or currently uses, seal cap enclosures to contain water leakage. The staff also noted that the use of such enclosures may not be accounted for in their license renewal application. For example, the environment within seal cap enclosures may be submerged, rather than the air environment of the original component design. Also, enclosures may prevent the direct inspections of bolting and component external surfaces within the Bolting Integrity and External Surfaces Monitoring Programs, respectively.
Issue:
It is unclear to the staff whether Seabrook Station is using seal cap enclosures to contain water leakage, and if so, how bolting and component external surfaces within seal cap enclosures will be age managed, since direct inspection is not possible.
Request:
- 1. For all instances where seal cap enclosures surround pressure-retaining bolting and the external surfaces of in-scope components:
- a. Describe the leaking water environment (e.g., reactor coolant, secondary water, borated water) and the materials of construction of the bolting and component external surfaces that are exposed to that environment.
- b. Describe how the bolting and component external surfaces will be managed for loss of material, loss of preload, cracking, and change in material properties, as appropriate, in the submerged environment. Add associated AMR line items, if necessary.
- c. If the use of seal cap enclosures prevents the direct inspections within the Bolting Integrity and External Surfaces Monitoring Programs, provide technical justification for how the aging effects will be effectively managed during the period of extended operation.
- 2. Describe how the use of seal cap enclosures is controlled such that aging is managed as described in L.b and 1.c.
United States Nuclear Regulatory Commission Page 47 of 48 SBK-L-12123 / Enclosure 1 NextEra Energy Seabrook Response:
1 a.
Seabrook Station has one seal cap enclosure that surrounds the pressure-retaining bolts of valve 1-SI-V-82. The valve is an ASME Code Class 1, 6-inch, Westinghouse swing check valve. The leakage water environment is treated borated water. The valve body is SA182, F304, bonnet is SA240, TP304, stud SA453, GR660, nuts SA194, GR6 and seal cap enclosure is SA479, TP 316/316L.
1 b.
NextEra installed a seal cap enclosure on SI-V-82 during the current operating cycle, (2011 Forced Outage) to allow continued operation of the unit until such time that the valve could be repaired. The installation of a seal cap enclosure creates a submerged environment that prevents the aging management of the bolting and component external surfaces for loss of material, loss of preload, cracking, and change in material properties.
Therefore, removal of the seal cap enclosure and restoration of the original configuration is planned to be completed no later than December 31, 2014. With the removal of the seal cap enclosure the existing aging management programs are sufficient to age manage the bolting and component external surfaces for loss of material, loss of preload, cracking, and change in material properties during the period of extended operation.
I c.
See 1 b.
- 2.
NextEra will remove the existing seal cap enclosure by December 31, 2014 and has no current plans to install any new seal cap enclosures. Therefore, the need to age manage the bolting and component external surfaces internal to the seal cap enclosure and inclusion of an associated AMR line items is not necessary.
License Renewal Application Appendix A, Section A.2.1.9, page A-9, is changed by adding a new second paragraph as follows:
Seabrook Station has one seal cap enclosure that surrounds the pressure-retaining bolts of valve 1-SI-V-82. The seal cap enclosures on SI-V-82 was installed during the 2011 Forced Outage to allow continued operation of the unit until such time that the valve could be repaired. The installation of a seal cap enclosures creates a submerged environment that prevents the aging management of the bolting and component external surfaces for loss of material, loss of preload, cracking, and change in material properties. Therefore, removal of the seal cap enclosures and restoration of the original configuration is planned to be completed no later than December 31, 2014. With the removal of the seal cap enclosures the existing aging management programs will remain sufficient to age manage the bolting and component external surfaces for loss of material, loss of preload, cracking, and change in material properties during the period of extended operation.
United States Nuclear Regulatory Commission Page 48 of 48 SBK-L-12123 / Enclosure 1 License Renewal Application Appendix B, Section B.2.1.9, page B-57 is changed by adding a new second paragraph as follows:
Seabrook Station has one seal cap enclosures that surrounds the pressure-retaining bolts of valve 1-SI-V-82. The seal cap enclosures on SI-V-82 was installed during the 2011 Forced Outage to allow continued operation of the unit until such time that the valve could be repaired. The installation of a seal cap enclosures creates a submerged environment that prevents the aging management of the bolting and component external surfaces for loss of material, loss of preload, cracking, and change in material properties. Therefore, removal of the seal cap enclosures and restoration of the original configuration is planned to be completed no later than December 31, 2014. With the removal of the seal cap enclosures the existing aging management programs will remain sufficient to age manage the bolting and component external surfaces for loss of material, loss of preload, cracking, and change in material properties during the period of extended operation.
to SBK-L-12123 Changes to the Seabrook Station License Renewal Application Associated with LR-ISG-2011-02, "Aging Management Program for Steam Generators"
United States Nuclear Regulatory Commission Page 1 of 1 SBK-L-12123 / Enclosure 2 LR-ISG-2011-02: Aging Management Program for Steam Generators LR-ISG-2011-02 recommends that applicants for license renewal follow the guidance provided in Revision 3 of NEI 97-06 when implementing their steam generator aging management program, including using Revision 3 of the Steam Generator Integrity Assessment Guidelines.
The following chfanges have been made to the NextEra Seabrook License Renewal Application in accordance with the recommendations made in LR-ISG-2011-02.
- 1. B.2.1.10, Steam GeneratorTube Integrity Program, is revised as follows:
- a. On page B-61, the 1st sentence of the 3 rd paragraph is revised as follows:
The Seabrook Station Steam Generator Tube Integrity Program is based on NEI 97-06 Rev. 2 3, "Steam Generator Program Guidelines ", the response and cormnitment to Generic Letter 97-06, "Degradation of Steam Generator Internals" and Seabrook Station Technical Specification 3/4.4.5, "Steam Generators," which ensure that the performance criteria for structural integrity, accident-induced leakage, and operational leakage are not exceeded.
- b. On page B-61, "NUREG-1801 Consistency" section is revised as follows:
This program, with the ex.epti.n noted below
, is consistent with NUREG-1801-XI.M19 as modified by LR-ISG-2011-02.
- c. On page B-63, the "Exceptions to NUREG-1801" section is revised as follows:
There are no exceptions to NUREG-1801-X1.MJ9 as modified by LR-ISG-2011-02.
NUREG 1.801 XI.UU 19 states "... the licensee
.s emmitient to implement the SG degrawdation management programn deseribed in NE! 97 06, are adequate to manage the effeets of aging on the SG tubes-, plufgs-, sleeves, and tube supports."
+iec Keter-ences section fer P4UR 1801 -XIA419 ientities NELI 9/ 06, &team Generator Program Guidelines " as Revision 1, dated Januar-y 200 1.
The Seabrook Station Steam Generator Tube integrit' Program is based on NEI 97 06, "Steam Generator Pro gram Guidelines ", Revision 2, dated May 205 Revision 2 of NEI 97 06 did not r-educe the ffinetional requirements of Revisionl 1.
in NEI correspondence with the NRC (Alex Mar-ion to Dr. Br-ian W Sheron) dated September 9, 2005, "Steam Generator Progr-am Guidelines, Revision 2 ", NEI states that Revision 2 of NEI 97 06 is eonsistent with Tcfical Specification Task For-ec Traveler TSTF 4149 Revision 4, "Steam Generator Tube Integrity'i The NRC staff review and approval of TST-F 1149, Revision 4, was documnented in Gener-ic Letter 2006 01, "Steam Generator Tube hitegrit' and 4s-seeiated Tcehnieal Spqeeoefieains ".
Seabrook Station implemented T-STF 449 with Lieense Amnendmnent 115 to Technical Specifications in June of 2007.
The approval of TST-F 1149 Revision 4 justifies the use of Revision 2 of NEI 97 06.
Program Element-sAffetefd:. 0le MP W 1 6360pe ef praffl-ano.
to SBK-L-12123 Clarification to Responses to RAI B.2.1.10-1 Associated with Steam Generator Tube-to Tubesheet Weld Inspection Plan, RAI B.2.1.26-1 Associated with Flash Point Testing, and to Steam Generator Divider Plate Inspection Plan
United States Nuclear Regulatory Commission Page 1 of 5 SBK-L-12123 / Enclosure 3 Clarification to Response to RAI B.2.1.10-1 Steam Generator Tube-to Tubesheet Weld Inspection Plans On December 14, 2010, the NRC issued Request for Additional Information (RAI)
B.2.1.10-1 related to Steam Generator Tube Integrity. Discussion with the NRC staff on May 29, 2012 identified a need for further clarification of the NextEra Seabrook response to that RAI.
To provide clarity regarding the intent of the commitments relative to addressing the potential failure of the steam generator reactor coolant pressure boundary due to PWSCC cracking of tube-to-tubesheet welds, the response has been revised as indicated below.
NextEra Energy Seabrook Revised Response to RAI B.2.1.10-1:
- 1. Based on the currently approved alternate repair criteria, the Seabrook Station steam generator tube-to-tubesheet welds are not included in the reactor coolant pressure boundary. This alternate repair criteria has not yet been permanently approved.
- 2. Unless a pcnantent alternate r-epair critria changing the ASME code boundary is approved by the NRC, or the Seabrook Station steam generatcrs are haniged to eliminate PAWSC-C susceptible tube to tubeshect welds, Seabrook Station will submit craeking due to PWISCC at least twenty fourf months prior to entering the Period of Extended Oprtin This plant speeifie programn will either
- 2. Seabrook Station will to address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two options:
- 1) Perform a one-time inspection of a representative sample of tube to tubesheet welds in all steam generators to determine if PWSCC cracking is present and, if cracking is identified, resolve the condition through engineering evaluation justifying continued operation or repair the condition, as appropriate, and establish an ongoing monitoring program to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators, or
- 2) Perform an analytical evaluation showing that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence of tube-to-tubesheet weld cracking, and or by redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for that the tube to tub..heet..
eld are not required to perform a reactor coolant pressure boundary function. The redefinition of the reactor coolant pressure boundary must be approved by the NRC as part of a license amendment request.
Based on the above discussion, the following changes have been made to the Seabrook License Renewal Application.
United States Nuclear Regulatory Commission Page 2 of 5 SBK-L-12123 / Enclosure 3
- 1. In Section A.2.1.10, on page A-10, the following new second paragraph has been added:
Seabrook Station will address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two options:
- 1) Perform a one-time inspection of a representative sample of tube to tubesheet welds in all steam generators to determine if PWSCC cracking is present and, if cracking is identified, resolve the condition through engineering evaluation justifying continued operation or repair the condition, as appropriate, and establish an ongoing monitoring program to perform routine tube-to-tubesheet weld inspections for tihe remaining life of the steam generators, or
- 2) Perform an analytical evaluation showing that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence of tube-to-tubesheet weld cracking, or by redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure boundary function. The redefinition of the rector coolant pressure boundary must be approved by the NRC as part of a license amendment request Option 1 or 2 will be completed at least 24 months prior to entering the period of extended operation.
- 2. In Section B.2. 1.10, on page B-64, the following Enhancement has been added:
Seabrook Station will address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of thefollowing two options:
- 1) Perform a one-time inspection of a representative sample of tube to tubesheet welds in all steam generators to determine if PWSCC cracking is present and, if cracking is identified, resolve the condition through engineering evaluation justifying continued operation or repair the condition, as appropriate, and establish an ongoing monitoring program to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators, or
- 2) Perform an analytical evaluation showing that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence of tube-to-tubesheet weld cracking, or by redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure
United States Nuclear Regulatory Commission Page 3 of 5 SBK-L-12123 / Enclosure 3 boundary function. The redefinition of the rector coolant pressure boundary must be approved by the NRC as part of a license amendment request.
Option 1 or 2 will be completed at least 24 months prior to entering the period of extended operation.
Program Elements Affected: Element 4 (Detection of Aging Effects).
In Section A.3, License Renewal Commitment #54 has been revised as follows:
No.
PROGRAM COMMITMENT UFSAR SCHEDULE or TOPIC I LOCATION I 54 Steam Generator Tube Integrity I Tý~ov
ý
ý
"+ ýf
ý
+
ýý
- ý A2 9
1 0 changing the ASMEf code boundary' is approeved by the NRC, or the Seabrook Station steam generator-s are changed to elifn-*.nate PWSCC--
susceptible tube to tubesheet welds, submit-a plant speeific agin managtent progr-am to manage the potential aging effect of er-acking due toPWSGC at least twenty four-months priort d L.*.£.
IV Pr-egFam to he submitted to NRG a At least 24 months prior to entering the period of extended operation.
entermig the PerioJ of Extendled Operation.
NextEra will address the potentialfor cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one ofthefollowing two options:
- 1) Perform a one-time inspection of a representative sample of tube-to-tubesheet welds in all steam generators to determine if PWSCC cracking is present and, if cracking is identified, resolve the condition through engineering evaluation justifying continued operation or repair the condition, as appropriate, and establish an ongoing monitoring program to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators, or
- 2) Perform an analytical evaluation showing that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence of tube-to-tubesheet weld cracking, or redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure boundary function.
The redefinition of the rector coolant pressure boundary must be approved by the NRC as part of a license amendment request.
United States Nuclear Regulatory Commission Page 4 of 5 SBK-L-12123 / Enclosure 3 II.
Steam Generator Divider Plate Inspection Plans On June 4, 2012, the NRC requested that the NextEra Seabrook commitment to inspect the Steam Generator divider plates for evidence of primary water stress corrosion cracking (Commitment #55) also be reflected in Appendices A and B. On June 18, 2012, the NRC staff indicated that current operating experience indicates that the divider plate inspection should be performed no sooner than five years prior to entering the period of extended operation.
Based on the discussion above the following License Renewal Application changes are provided.
In Section A.3, License Renewal Commitment #55 has been revised as follows:
Steam Generator Seabrook will perform an inspection of each Within five years tPprior to 55 Tube Integrity steam generator to assess the condition of the A.2.1.10 entering the period of extended divider plate assembly.
operation Based on the above discussion, the following changes have been made to the Seabrook License Renewal Application.
- 1. In Section A.2.1.10, on page A-10, the following new 3 rd paragraph has been added:
Seabrook will perform an inspection of each steam generator to assess the condition of the divider plate assembly within five years prior to entering the period of extended operation.
- 2. In Section B.2.1.10, on page B-63, the following has been added after the last paragraph of the Program Description :
Seabrook will perform an inspection of each steam generator to assess the condition of the divider plate assembly within five years prior to entering the period of extended operation. The inspection techniques used will be capable of detecting primary water stress corrosion cracking in the steam generator divider plate assemblies and their associated welds. Any evidence of cracking will be documented and evaluated under the corrective action program.
United States Nuclear Regulatory Commission Page 5 of 5 SBK-L-12123 / Enclosure 3 IlL.
Clarification to Response to RAI B.2.1.26-1 Associated with Flash Point Testing In response to RAI B.2.1.26-1 (Seabrook letter SBK-L-1 1002, Enclosure 1) Seabrook revised Section B.2.1.26 to state that "lube oil analysis required will include "Flash Point" when there is a potential for contamination of the lubrication oil by fuel."
Consistent with that response, Appendix A, Section A.2.1.26, Lubricating Oil Analysis has been changed as follows:
A.2.1.26 Lubricating Oil Analysis The Lubricating Oil Analysis Program obtains and analyzes lubricating oil samples from plant equipment to ensure that the oil quality is maintained within established limits. The program provides an early indication of adverse equipment condition in lubricating oil environments.
The Seabrook Station Lubricating Oil Analysis Program includes sampling and analysis of lubricating oil for components within the scope of license renewal and subject to aging management review, that are exposed to lubricating oil and for which pressure boundary integrity or heat transfer is required for the component to perform its intended function. The lube oil analysis required will include "Flash Point" when there is a potentialfor contamination of the lubrication oil by fueL to SBK-L-12123 LRA Appendix A - Final Safety Report Supplement Table A.3 License Renewal Commitment List
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 1 of 14 A.3 LICENSE RENEWAL COMMITMENT LIST UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Program to be implemented prior to the period of extended operation. Inspection plan to be An inspection plan for Reactor Vessel Internals will be submitted submitted to NRC not later than PWR Vessel Internals for NRC review and approval.
A.2.1.7 2 years after receipt of the renewed license or not less than 24 months prior to the period of extended operation, whichever comes first.
Closed-Cycle Cooling Enhance the program to include visual inspection for cracking, Prior to the period of extended
- 2.
Water loss of material and fouling when the in-scope systems are A.2.1.12 operation opened for maintenance.
Inspection of Overhead Heavy Load and Light Enhance the program to monitor general corrosion on the crane Prior to the period of extended
- 3.
Load (Related to and trolley structural components and the effects of wear on the A.2.1.13 operation Refueling) Handling rails in the rail system.
Systems Inspection of Overhead Heavy Load and Light Prior to the period of extended
- 4.
Load (Related to Enhance the program to list additional cranes for monitoring.
A.2.1.13 operation Refueling) Handling Systems
- 5.
Compressed Air Enhance the program to include an annual air quality test A.2.1.14 Prior to the period of extended Monitoring requirement for the Diesel Generator compressed air sub system.
operation
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 2 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to perform visual inspection of penetration Prior to the period of extended seals by a fire protection qualified inspector.
operation.
Enhance the program to add inspection requirements such as Prior to the period of extended
- 7.
Fire Protection spalling, and loss of material caused by freeze-thaw, chemical A.2.1.15 operation.
attack, and reaction with aggregates by qualified inspector.
Enhance the program to include the performance of visual Prior to the period of extended
- 8.
Fire Protection inspection of fire-rated doors by a fire protection qualified A.2.1.15 operation.
inspector.
Enhance the program to include NFPA 25 guidance for "where sprinklers have been in place for 50 years, they shall be replaced Prior to the period of extended
- 9.
Fire Water System or representative samples from one or more sample areas shall be A.2.1.16 operation.
submitted to a recognized testing laboratory for field service testing".
Enhance the program to include the performance of periodic flow
- 10.
Fire Water System testing of the fire water system in accordance with the guidance A.2.1.16 operaton.
of NFPA 25.
operation.
Enhance the program to include the performance of periodic visual or volumetric inspection of the internal surface of the fire protection system upon each entry to the system for routine or corrective maintenance. These inspections will be documented and trended to determine if a representative number of Within ten years prior to the
- 11.
Fire Water System inspections have been performed prior to the period of extended A.2.1.16 period of extended operation.
operation. If a representative number of inspections have not been performed prior to the period of extended operation, focused inspections will be conducted. These inspections will be performed within ten years prior to the period of extended operation.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 3 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to include components and aging effects
- 12.
Aboveground Steel Tanks required by the Aboveground Steel Tanks.
A.2.1.17 Prior to the period of extended
- 12.
boveroud Stel Tnksoperation.
Enhance the program to include an ultrasonic inspection and
- 13.
Aboveground Steel Tanks evaluation of the internal bottom surface of the two Fire A.2.1.17 Witin tenyea prior t
Protection Water Storage Tanks.
period of extended operation.
Enhance program to add requirements to 1) sample and analyze
- 14.
Fuel Oil Chemistry new fuel deliveries for biodiesel prior to offloading to the A.2.1.18 Prior to the period of extended Auxiliary Boiler fuel oil storage tank and 2) periodically sample operation.
stored fuel in the Auxiliary Boiler fuel oil storage tank.
Enhance the program to add requirements to check for the Prior to the period of extended
- 15.
Fuel Oil Chemistry presence of water in the Auxiliary Boiler fuel oil storage tank at A.2.1.18 operation.
least once per quarter and to remove water as necessary.
Enhance the program to require draining, cleaning and inspection Prior to the period of extended
- 16.
Fuel Oil Chemistry of the diesel fire pump fuel oil day tanks on a frequency of at A.2.1.18 least once every ten years.
operation.
Enhance the program to require ultrasonic thickness measurement of the tank bottom during the 10-year draining, Prior to the period of extended
- 17.
Fuel Oil Chemistry cleaning and inspection of the Diesel Generator fuel oil storage A.2.1.18 operation.
tanks, Diesel Generator fuel oil day tanks, diesel fire pump fuel oil day tanks and auxiliary boiler fuel oil storage tank.
Reactor Vessel Enhance the program to specify that all pulled and tested Prior to the period of extended
- 18.
Surveillance capsules, unless discarded before August 31, 2000, are placed in A.2.1.19 operation.
storage.
- 19.
Reactor Vessel Enhance the program to specify that if plant operations exceed A.2.1.19 Prior to the period of extended Surveillance the limitations or bounds defined by the Reactor Vessel operation.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 4 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Surveillance Program, such as operating at a lower cold leg temperature or higher fluence, the impact of plant operation changes on the extent of Reactor Vessel embrittlement will be evaluated and the NRC will be notified.
Enhance the program as necessary to ensure the appropriate withdrawal schedule for capsules remaining in the vessel such that one capsule will be withdrawn at an outage in which the
- 20.
Reactor Vessel capsule receives a neutron fluence that meets the schedule A.2.1.19 Prior to the period of extended Surveillance requirements of 10 CFR 50 Appendix H and ASTM E185-82 and operation.
that bounds the 60-year fluence, and the remaining capsule(s) will be removed from the vessel unless determined to provide meaningful metallurgical data.
Enhance the program to ensure that any capsule removed,
- 21.
Reactor Vessel without the intent to test it, is stored in a manner which maintains A.2.1.19 Prior to the period of extended Surveillance it in a condition which would permit its future use, including operation.
during the period of extended operation.
Within ten years prior to the
- 22.
One-Time Inspection Implement the One Time Inspection Program.
A.2.1.20 perio tended prio n.
period of extended operation.
Implement the Selective Leaching of Materials Program. The Selective Leaching of program will include a one-time inspection of selected Within five years prior to the
- 23.
Materials components where selective leaching has not been identified and A.2.1.21 period of extended operation.
periodic inspections of selected components where selective leaching has been identified.
Buried Piping And Tanks Implement the Buried Piping And Tanks Inspection Program.
Within ten years prior to
- 24.
Inspection A.2.1.22 entering the period of extended operation One-Time Inspection of Implement the One-Time Inspection of ASME Code Class 1 Within ten years prior to the
- 25.
ASME Code Class I Small Small Bore-Piping Program.
A.2.1.23 period of extended operation.
Bore-Piping
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 5 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and effects of
- 26.
External Surfaces interest, the refueling outage inspection frequency, the A.2.1.24 Prior to the period of extended Monitoring inspections of opportunity for possible corrosion under operation.
insulation, the training requirements for inspectors and the required periodic reviews to determine program effectiveness.
Inspection of Internal
- 27.
Surfaces in Miscellaneous Implement the Inspection of Internal Surfaces in Miscellaneous A.2.1.25 Prior to the period of extended Piping and Ducting Piping and Ducting Components Program.
operation.
Components Enhance the program to add required equipment, lube oil analysis A.2.1.26 Prior to the period of extended
- 28.
Lubricating Oil Analysis required, sampling frequency, and periodic oil changes.
operation.
Enhance the program to sample the oil for the Reactor Coolant A.2.1.26 Prior to the period of extended
- 29.
Lubricating Oil Analysis pump oil collection tanks.
operation.
Enhance the program to require the performance of a one-time
- 30.
Lubricating Oil Analysis ultrasonic thickness measurement of the lower portion of the A.2.1.26 Prior to the period of extended Reactor Coolant pump oil collection tanks prior to the period of operation.
extended operation.
ASME Section XI, Enhance procedure to include the definition of "Responsible A.2.1.28 Prior to the period of extended Subsection IWL Engineer".
operation.
- 32.
Structures Monitoring Enhance procedure to add the aging effects, additional locations, A.2.1.31 Prior to the period of extended Program inspection frequency and ultrasonic test requirements.
operation.
Structures Monitoring Enhance procedure to include inspection of opportunity when Prior to the period of extended
- 33.
Program planning excavation work that would expose inaccessible A.2.1.31 operation.
concrete.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 6 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Electrical Cables and Connections Not Subject Implement the Electrical Cables and Connections Not Subject to
- 34.
to 10 CFR 50.49 10 CFR 50.49 Environmental Qualification Requirements A.2.1.32 Prior to the period of extended Environmental operation.
Qualification program.
Requirements Electrical Cables and Connections Not Subject to 10 CFR 50.49 Implement the Electrical Cables and Connections Not Subject to Prior to the period of extended
- 35.
Environmental 10 CFR 50.49 Environmental Qualification Requirements Used A.2.1.33 operation.
Qualification in Instrumentation Circuits program.
Requirements Used in Instrumentation Circuits Inaccessible Power Cables Not Subject to 10 CFR Implement the Inaccessible Power Cables Not Subject to 10 CFR Prior to the period of extended
- 36.
50.49 Environmental IpeetteIacsilPoeCalsNtSbcto10FR A.2.1.34 oeain Qualification 50.49 Environmental Qualification Requirements program.
operation.
Qualification Requirements Prior to the period of extended
- 37.
Metal Enclosed Bus Implement the Metal Enclosed Bus program.
A.2.1.35 operaton.
operation.
Prior to the period of extended
- 38.
Fuse Holders Implement the Fuse Holders program.
A.2.1.36 operaton.
operation.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 7 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Electrical Cable Connections Not Subject to
- 39.
10 CFR 50.49 Implement the Electrical Cable Connections Not Subject to 10 A.2.1.37 Prior to the period of extended Environmental CFR 50.49 Environmental Qualification Requirements program.
operation.
Qualification Requirements Prior to the period of extended
- 40.
345 KV SF6 Bus Implement the 345 KV SF6 Bus program.
A.2.2.1 oraton.
operation.
- 41.
Metal Fatigue of Reactor Enhance the program to include additional transients beyond A.2.3.1 Prior to the period of extended Coolant Pressure Boundary those defined in the Technical Specifications and UFSAR.
operation.
- 42.
Metal Fatigue of Reactor Enhance the program to implement a software program, to count A.2.3.1 Prior to the period of extended Coolant Pressure Boundary transients to monitor cumulative usage on selected components.
operation.
The updated analyses will be Pressure -Temperature Seabrook Station will submit updates to the P-T curves and submitted at the appropriate
- 43.
Limits, including Low LTOP limits to the NRC at the appropriate time to comply with A.2.4.1.4 time to comply with 10 CFR 50 Temperature Overpressure 10 CFR 50 Appendix G.
Appendix G, Fracture Protection Limits Toughness Requirements.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 8 of 14 UFSAR PROGRAM or TOPIC COMMITMENT I
SCHEDULE 44.
Environmentally-Assisted Fatigue Analyses (TLAA)
NextEra Seabrook will perform a review of design basis ASME Class 1 component fatigue evaluations to determine whether the NUREG/CR-6260-based components that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting components for the Seabrook plant configuration. If more limiting components are identified, the most limiting component will be evaluated for the effects of the reactor coolant environment on fatigue usage. If the limiting location identified consists of nickel alloy, the enviromnentally-assisted fatigue calculation for nickel alloy will be performed using the rules of NUREG/CR-6909.
(1) Consistent with the Metal Fatigue of Reactor Coolant Pressure Boundary Program Seabrook Station will update the fatigue usage calculations using refined fatigue analyses, if necessary, to determine acceptable CUFs (i.e., less than 1.0) when accounting for the effects of the reactor water environment. This includes applying the appropriate Fen factors to valid CUFs determined from an existing fatigue analysis valid for the period of extended operation or from an analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case).
(2) If acceptable CUFs cannot be demonstrated for all the selected locations, then additional plant-specific locations will be evaluated. For the additional plant-specific locations, if CUF, including environmental effects is greater than 1.0, then Corrective Actions will be initiated, in accordance with the Metal Fatigue of Reactor Coolant Pressure Boundary Program, B.2.3.1.
Corrective Actions will include inspection, repair, or replacement of the affected locations before exceeding a CUF of 1.0 or the effects of fatigue will be managed by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by the NRC).
A.2.4.2.3 At least two years prior to entering the period of extended operation.
United States Nuclear Regulatoty Commission SBK-L-12123 / Enclosure 4 Page 9 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION
- 45.
Number Not Used Protective Coating Enhance the program by designating and qualifying an Inspector Prior to the period of extended
- 46.
Monitoring and Coordinator and an Inspection Results Evaluator.
A.2.1.38 operation Maintenance Enhance the program by including, "Instruments and Equipment Protective Coating needed for inspection may include, but not be limited to, Prior to the period of extended
- 47.
Monitoring and flashlight, spotlights, marker pen, mirror, measuring tape, A.2.1.38 operation Maintenance magnifier, binoculars, camera with or without wide angle lens, and self sealing polyethylene sample bags."
Protective Coating Enhance the program to include a review of the previous two Prior to the period of extended
- 48.
Monitoring and monitoring reports.
A.2.1.38 operation Maintenance Enhance the program to require that the inspection report is to be Protective Coating evaluated by the responsible evaluation personnel, who is to Prior to the period of extended
- 49.
Monitoring and A213 Maintenance prepare a summary of findings and recommendations for future operation surveillance or repair.
Within the next two refueling outages, OR15 or OR16, and
- 50.
ASME Section XI, Perform UT testing of the containment liner plate in the vicinity A.2.1.27 repeated at intervals of no more Subsection IWE of the moisture barrier for loss of material.
than five refueling outages.
Number Not Used 51.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 10 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION
.ASME Section XI, Implement measures to maintain the exterior surface of the
- 52.
Subsection IWL Containment Structure, from elevation -30 feet to +20 feet, in a A.2.1.28 By December 31, 2012 dewatered state.
Reactor Head Closure Replace the spare reactor head closure stud(s) manufactured from Prior to the period of extended
- 53.
Studs the bar that has a yield strength > 150 ksi with ones that do not A.2.1.3 operation.
exceed 150 ksi.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 11 of 14 UFSAR PROGRAM or TOPIC COMMITMENT j
SCHEDULE 54.
Steam Generator Tube Integrity Unless a permanent alternate repair-eriteria ehanging the ASMEF
.. d. bundar.y is approved by the NRC, or-the Scabrook Station steam generators are changed to eliminate PIASCC susceptiblc
.be to tubehet*
welds, submit a plant specific agin management progr-am to manage the potential aging effcct ot efaeking due to PWSCC at leas tI nt fou months prior-to entefing the Period of Extended Operation.
NextEra will address the potentialfor cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two options:
- 1) Perform a one-time inspection of a representative sample of tube-to-tubesheet welds in all steam generators to determine if PWSCC cracking is present and, if cracking is identified, resolve the condition through engineering evaluation justifying continued operation or repair the condition, as appropriate, and establish an ongoing monitoring program to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators, or
- 2) Perform an analytical evaluation showing that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence of tube-to-tubesheet weld cracking, or redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure boundary function. The redefinition of the rector coolant pressure boundary must be approved by the NRC as part of a license amendment request.
A.2. 1.10 Programn to be submitted to NRG-a At least 24 months prior to entering the period of extended operation.
Within five years tPprior to Steam Generator Tube Seabrook will perform an inspection of each steam generator to A.2.1.Wi entering the period of extended Integrity assess the condition of the divider plate assembly.
operation.
United States Nuclear Regulatory SBK-L-12123 / Enclosure 4 Commission Page 12 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the period of
- 56.
Water-Syste Guideline operating ranges and Action Level values for A.2.1.12 Water System hyrznCndslae, extended operation.
hydrazine and sulfates.
Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the period of
- 57.
Water System Guideline operating ranges and Action Level values for Diesel A.2.1.12 extended operation.
Generator Cooling Water Jacket pH.
Update Technical Requirement Program 5.1, (Diesel Fuel Oil Prior to the period of extended
- 58.
Fuel Oil Chemistry Testing Program) ASTM standards to ASTM D2709-96 and A.2.1.18 operation.
ASTM D4057-95 required by the GALL XI.M30 Rev 1 Nickel Alloy Nozzles and The Nickel Alloy Aging Nozzles and Penetrations program will Prior to the period of extended
- 59.
Penetrations implement applicable Bulletins, Generic Letters, and staff A.2.2.3 operation.
accepted industry guidelines.
Buried Piping and Tanks Implement the design change replacing the buried Auxiliary Prior to entering the period of
- 60.
Inspection Boiler supply piping with a pipe-within-pipe configuration with A.2.1.22 extended operation.
leak detection capability.
Within ten years prior to
- 61.
Compressed Air Replace the flexible hoses associated with the Diesel Generator A.2.1.14 entering the period of extended Monitoring Program air compressors on a frequency of every 10 years.
operathon.
Enhance the, program to include a statement that sampling Prior to the period of extended
- 62.
Water Chemistry frequencies are increased when chemistry action levels are A.2.1.2 operation.
exceeded.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 13 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Ensure that the quarterly CVCS Charging Pump testing is continued during the PEO. Additionally, add a precaution to the
- 63.
Flow Induced Erosion test procedure to state that an increase in the CVCS Charging Prior to the period of extended Pump mini flow above the acceptance criteria may be indicative operation of erosion of the mini flow orifice as described in LER 50-275/94-023.
Soil analysis shall be performed prior to entering the period of extended operation to determine the corrosivity of the soil in the vicinity of non-cathodically protected steel pipe within the scope A.2.1.22 Prior to entering the period of
- 64.
Buried Piping and Tanks of this program. If the initial analysis shows the soil to be non-extended operation.
Inspection corrosive, this analysis will be re-performed every ten years thereafter.
Implement measures to ensure that the movable incore detectors Prior to entering the period of
- 65.
are not returned to service during the period of extended Flux Thimble Tube operation.
N/A extended operation Number Not Used 66.
Perform one shallow core bore in an area that was continuously
- 67.
Structures Monitoring wetted from borated water to be examined for concrete A.2.1.31 No later than December 3 1, Program degradation and also expose rebar to detect any degradation such 2015 as loss of material.
Perform sampling at the leakoff collection points for chlorides,
- 68.
Structures Monitoring sulfates, pH and iron once every three months.
A.2.1.31 Starting January 2014 Program
- 69.
Open-Cycle Cooling Water Replace the Diesel Generator Heat Exchanger Plastisol PVC Prior to the period of
- 69.
Open-Cylem Cooling Water lined Service Water piping with piping fabricated from AL6XN A.2. 1.11 Prirntotheperiodo System
- material, extended operation.
United States Nuclear Regulatory Commission SBK-L-12123 / Enclosure 4 Page 14 of 14 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Inspect the piping downstream of CC-V-444 and CC-V-446 to
- 70.
Closed-Cycle Cooling determine whether the loss of material due to cavitation A.2.1.12 Within ten years prior to the Water System induced erosion has been eliminated or whether this remains an period of extended operation.
issue in the primary component cooling water system.
Implement the Alkali-Silica Reaction (ASR) Monitoring
- 71.
Alkali-Silica Reaction Program Prior to entering the period of (ASR) Monitoring Program extended operation.