ML17017A208
| ML17017A208 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 01/05/2017 |
| From: | Connolly J South Texas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NOC-AE-16003429, TAC ME 4937, TAC ME4936 | |
| Download: ML17017A208 (16) | |
Text
G;;JCC?UCJ Nuclear Operating Company
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South Texas Project Electric Generating Station P.O. Box 285' Wadsworth, Texas 77483 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 and 2 January 5, 2017 NOC-AE-16003429 10 CFR 54 File No. G25 Docket Nos. STN 50-498, STN 50-499 Response to Requests for Additional Information for the Review of the South Texas Project, Units 1 and 2, License Renewal Application - RAI 82.1.8-3 (TAC Nos. ME4936 and ME4937)
Reff1rences:
- 1. Letter; G.T. Powell to the NRC Document Control Desk; "License Renewal Application",
NOC-AE-10002607; dated October 25, 2010. (ML103010257)
- 2. Letter; from the NRC, L. M. James, to STPNOC, G.T. Powell; "Request for Additional Information for the Review of the South Texas Project, Units 1 and 2, License Renewal Application (TAC Nos. ME4936 and, ME4937)", dated December 8, 2016.
By Reference 1, STP Nuclear Operating Company (STPNOC) submitted a License Renewal Application (LRA). In Reference 2, the NRC requested additional information for the review of the South Texas Project LRA. Enclosure 1 to this letter provides STPNOC's response to the requests for additional information, and Enclosure 2 provides line in/line out revisions to the LRA.
STI: 34421468
r NOC-AE-16003429 Page 2 of 3 If there are any questions regarding this submittal, please contact Arden Aldridge, STP License Renewal Project Lead, at (361) 972-8243 or Rafael Gonzales, STP License Renewal Project regulatory point-of-contact, at (361) 972-4779.
Revised regulatory commitments are included in Enclosure 3 to this letter. There are no other regulatory commitments in this letter.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on tf!_~ 6J,J.017 ti te '
amesConnM Site Vice President rjg
Enclosures:
- 1) STPNOC Response to Request for Additional Information
- 3) STPNOC Regulatory Commitments - line in/out
cc:
(paper copy)
Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 G9A) 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U.S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN116
-Wadsworth, TX 77483 Lois James License Renewal Project Manager (Safety)
One White Flint North (MS 011-F1)
U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Tam Tran License Renewal Project Manager (Environmental)
One White Flint North (MS 011 F01)
U. S. Nuclear Regulatory Commission Washi'ngton, DC 20555-0001 (electronic copy)
NOC-AE-16003429 Page 3 of 3 Morgan. Lewis & Beckius LLP
- Steve Frantz U.S. Nuclear Regulatory Commission Lisa M. Regner Lois James Tam Tran NRG South Texas LP Chris O'Hara Jim von Suskil Skip Zahn CPS Energy Kevin Pollo Cris Eugster L. D. Blaylock City of Austin Elaina Ball John Wester Texas Dept. of State Health Services Helen Watkins Robert Free
NOC-AE-16003429 STPNOC Response to Request for Additional Information
RAI 82.1.8-3
Background:
SOUTH TEXAS PROJECT LICENSE RENEWAL APPLICATION (LRA)
REQUEST FOR ADDITIONAL INFORMATION (RAil NOC-AE-16003429 Page 1 of 2 The staff issued License Renewal Interim Staff Guidance (LR-ISG) 2016-01, "Changes to Aging Management Guidance for Various Steam Generator Components" (ADAMS Accession No. ML16237A383). LR ISG 2016 01 provides the following guidance for aging management:
Visual inspections on steam generator head internal areas (head interior surfaces, divider plate assemblies, tubesheets (primary side) and tube-to-tubesheet welds) in order to identify signs of cracking or loss of material (e.g., rust stains and distortion of divider plates).
Frequency of the visual inspections: at least every 72 effective full power months or every third refueling outage whichever results in more frequent inspections Implementation of the latest EPRI steam generator guidelines such as (a) EPRI Report 1022832 (primary-to-secondary leak guidelines); (b) EPRI Report 1025132 (in-situ pressure test guidelines); (c) EPRI Report 3002007571 (integrity assessment guidelines); and (d) EPRI Report 3002007572 (examination guidelines)
Issue:
The staff found a need to confirm whether the applicant's Steam Generator Tube Integrity Program is consistent with the guidance discussed above.
Request:
Clarify whether the Steam Generator Tube Integrity Program is consistent with the guidance discussed above (i.e., conduct of visual inspections, visual inspection frequency, and implementation or plans for implementation of the latest EPRI steam generator guidelines by the implementation dates provided by the industry). If not, provide justification of why the applicant's Steam Generator Tube Integrity Program is adequate for aging management. In addition, provide updated UFSAR supplement for this program as necessary.
STPNOC Response:
NOC-AE-16003429 Page 2 of 2 LRA Table 3.1.2-4, AppendixA1.8, Appendix 82.1.8, and LRA Basis Document AMP Xl.M19,
' Steam Generator Tube Integrity, have been revised consistent with the guidance provided in LR ISG 2016-01, "Changes to Aging Management Guidance for Various Steam Generator Components":
Perform visual inspections on steam generator head internal areas (head interior
. surfaces, divider plate assemblies, tubesheets (primary side) and tube-to-tubesheet welds) in order to identify signs of cracking or loss of material.
Frequency of the visual inspections: at least every 72 effective full power months or every third refueling outage whichever results in more frequent inspections.
Implementation of the latest (i.e. implemented within the approved implementation period) EPRI steam generator guidelines such as (a) EPRI Report PWR Primary-To-Secondary Leak Guidelines; (b) EPRI Report Steam Generator In-Situ Pressure Test Guidelines; (c) EPRI Report Steam Generator Program Integrity Assessment Guidelines; and (d) EPRI Report PWR Steam Generator Examination Guidelines. provides the line-in/line-out revision to LRA Appendix A 1.8, LRA Appendix 82.1.8, and Table 3.1 ~2-4. provides the line-in/line-out revision to LRA Table A4-1 adding LRA Commitment item #48.
NOC-AE-16003429 STPNOC LRA Appendix A and B Line in/out Sections Affected LRA Section Table 3.1.2-4 A1.8 82.1.8
T, bl 3 1 2 4 a e.. -
Component Type SG Divider Plate SG Divider Plate SG Tube-to-tubesheet Welds SG Tube-to-tubesheet Welds SG Primary Head SG Prima!Y Head R
t V:
I I t eacor esse, n erna s, an dR t
C I tS t eacor ooan ;ysem-s Intended Material Environment Aging Effect Function Requiring Management OF Nickel Alloys Reactor Coolant Cracking (Ext)
OF Nickel Alloys Reactor Coolant Loss of Material (Ext}
NOC-AE-16003429 Page 1 of 7 ummaryo fA
~gmg M anagemen tE I f
va ua ion-St eam G enera ors Aging Management NU REG-Table 1 Item Notes Program 1801 Vol.
2 Item Steam Generator IV.01-6 3.1.1.81
£A,4 Tubing Integrity: (B2.1.8) and Water Chemistry (B2.1.2)
Steam Generator None None H.4 Tubing Integrity: (B2.1.8) and Water Chemist!Y (82.1.2)
Space between changes - For clarity, only the line items with changes haven provided in this submittal PB Nickel Alloys Reactor Coolant Cracking Steam Generator None None
.ti (Ext}
Tubing Integrity (B2.1.8) and Water Chemist!Y (82.1.2)
PB Nickel Alloys Reactor Coolant Loss of Material Steam Generator None None
.ti (Ext)
Tubing Integrity: (B2.1.8) and Water Chemist!Y (B2.1.2)
Space between changes - For clarity, only the line items with changes haven provided in this submittal PB Carbon Steel Reactor Coolant Cracking ASME Section XI IV.01-1 3.1.1.68
&A with (Int) lnservice Inspection, Stainless Subsections IWB, IWC, Steel and IWO for Class 1 Cladding components (82.1.1 ).,_
aAti-Water Chemistry (B2.1.2) and Steam Generator Tubing lntearitv (82.1.8)
PB Carbon Steel Reactor Coolant Loss of Material Steam Generator None None
.ti with (Ext)
Tubing Integrity: (B2.1.8)
Stainless and Water Chemist!Y Steel (B2.1.2)
Claddina
Component Intended Material Environment Aging Effect Type Function Requiring Management Aging Management NU REG-Program 1801 Vol.
2 Item NOC-AE-16003429 Page 2of7 Table 1 Item Notes Space between changes - For clarity, only the line items with changes haven provided in this submittal SG Tubesheet PB Carbon Steel Reactor Coolant Cracking Steam Generator IV.01-18 3.1.1.73 C, 2 Gladded with (Ext)
Tubing Integrity (82.1.8)
Nickel Alloys and Water Chemistry (82.1.2)
SG Tubesheet PB Carbon Steel Reactor Coolant Loss of Material Steam Generator None None
.ti.
Gladded with (Ext)
Tubing lntegri~ (82.1.8)
Nickel Alloys and Water Chemistry (82.1.2)
Notes for Table 3.1.2-4:
Standard Notes:
A Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.
B Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP takes some exceptions to NUREG-1801 AMP.
C Component is different, but consistent with NUREG-1801 item for material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.
E Consistent with NUREG-1801 for material, environment, and aging effect, but a different aging management program is credited or NUREG-1801 identifies a plant-specific aging management program.
G Environment not in NUREG-1801 for this component and material.
H Aging effect not in NUREG-1801 for this component, material and environment combination.
Plant Specific Notes:
-1 The inspection requirements of ASME Code Case N-722 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/182 that have been mitigated by weld overlay.
2 The tubesheets in the replacement steam generators installed at STP are made of carbon steel cladded with the Alloy 690.The tube-tubesheet welds are flush fusion welds with cladding.
3 The replacement steam generator primary manway is equipped with stainless steel gasket insert.
4 The divider plates in the replacement steam generators installed at STP are made of Alloy 690 with the associated Alloy 52/152 weld materials.
5 The replacement steam generator secondary manway, handhole and inspection ports are equipped with nickel alloy gasket inserts.
A1.8 STEAM GENERATOR TUBE INTEGRITY NOC-AE-16003429 Page 3 of 7 The Steam Generator Tube Integrity program is applicable to managing the aging of steam generator tubes. plugs, sleeves. divider plate assemblies. tube-to-tubesheet welds. primary head (interior surfaces), tubesheet(s) (primary side), and secondary side components that are contained within the steam generator.The Steam Generator Tube Integrity program manages cracking and loss of material of the follmving component types: steam generator tubes, tube support plates, secondary side access covers, secondary nozzles, moisture separators, primary head and divider plates, internal structures, flow distribution baffles, feedwater rings, and auxiliary feedwater (AFVV) spray pipes. The program manages the cracking of plugs. The program also manages the wall thinning of the following component types: upper steam drum internals, moisture separators, feedwater rings, and AFW spray pipes.
The program ensures the integrity of the primary to secondary pressure boundary through preventive measures, inspections, degradation assessments, condition monitoring, operational assessments, tube plugging, and leakage monitoring activities necessary to manage potential steam generator tube degradation, including mechanically induced phenomena, such as wear and impingement damage.
The aging management measures employed includes nondestructive examinations, visual inspection, sludge removal, tube plugging, in-situ pressure testing and maintaining the chemistry environment by removal of impurities and addition of chemicals to control pH and oxygen. NOE inspection and primary to secondary leak rate monitoring are conducted consistent with the requirements of STP Units 1 and 2 Technical Specifications, NEI 97-06, Steam Generator Program Guidelines and the latest (i.e. implemented within the approved implementation period) EPRI Steam Generator Integrity Assessment Guidelines, EPRI PWR Primary-to-Secondary Leak Guidelines, EPRI Steam Generator In-Situ Pressure Test Guidelines. and EPRI PWR Steam Generator Examination Guidelines.
Performance criteria are maintained for operational leakage, accident induced leakage and structural integrity as prescribed in the Technical Specifications. Tube structural integrity limits consistent with Regulatory Guide 1.121 are applied.
Visual inspection are performed of the divider plates assemblies. tubesheets (primary side),
tube to tubesheet welds. and primary head (interior surfaces) for signs of cracking and loss of material are performed at least every 72 effective full power months or every third refueling outage, whichever results in more frequent inspections.
82.1.8 Steam Generator Tube Integrity Program Description NOC-AE-16003429 Page 4 of 7 The Steam Generator Tube Integrity program is applicable to managing the aging of steam generator tubes. plugs. sleeves. divider plate assemblies. tube-to-tubesheet welds. primary head (interior surfaces). tubesheet(s) (primary side). and secondary side components that are contained within the steam generator. The Steam Generator Tube Integrity program manages the loss of material of the following component types: steam generator tubes, tube support plates, secondary side assess savers, secondary nozzles, moisture separators, internal structures, flmv distribution baffles, feedvJater rings, auxiliary feed*.vater V\\FW) spray pipes, primary head and divider plates. The program man.ages the cracking of the follo1Ning component types: steam generator tubes, plugs, tube support plates, secondary side access savers, secondary nozzles, primary head and divider plates, internal structures, flm*J distribution baffles, feedwater rings, and AFW spray pipes. The program also manages wall thinning of the following component types: moisture separators, feedwater rings, and AFW spray pipes.
The program ensures the integrity of the primary to secondary pressure boundary through assessments of potential degradation mechanisms, inspections, tube integrity assessments, maintenance plugging and repairs, primary to secondary leakage monitoring, maintenance of the secondary-side integrity, primary side and secondary side water chemistry, and foreign material exclusion.. STP procedural guidance ensures performance criteria for tube structural integrity, operational leakage and accident induced leakage.
Reporting criteria, inspection scope and frequency, assessments, plugging criteria, and primary to secondary leak rate monitoring, monitoring and controlling primary and secondary side water chemistry are consistent with the requirements of STP Units 1 and 2 Technical Specifications, the Maintenance Rule (10 CFR 50.65), NEI 97-06, Steam Generator Program Guidelines, Revision 3 and the latest (i.e. implemented within the approved implementation period) EPRI 1019038,Steam Generator Integrity Assessment Guidelines, EPRI PWR Primarv-to-Secondary Leak Guidelines. EPRI Steam Generator In-Situ Pressure Test Guidelines, and EPRI PWR Steam Generator Examination Guidelines. Tube structural integrity limits consistent with Regulatory Guide 1.121, Bases for Plugging Degraded PWR Steam Generator Tubes are applied as detailed in U FSAR Section 3.12.1.
Visual inspection of the divider plates assemblies. tubesheets (primary side). tube to tubesheet welds. and primary head (interior surfaces) for signs of cracking and loss of material are performed at least every 72 effective full power months or every third refueling outage.
whichever results in more frequent inspections.
The training and qualification standards for personnel engaged in the acquisition and/or evaluation of steam generator non-destructive examination (NOE) activities are specified in a station administrative procedure, and inspection practices are consistent with the EPRI PWR Steam Generator Examination Guidelines. STP programmatic guidance also requires that each inspection be based on a degradation assessment created for each refueling outage that considers existing and potential degradation mechanisms.
The STP steam generators were replaced with Westinghouse Delta 94 Steam Generators in 2000 and 2002 for Units 1 and 2, respectively. The STP replacement steam generators are equipped with Alloy 690TT tubes. The tube support plates are fabricated from type 405 stainless steel and the tube holes are trefoil-broached. Due to the advanced design and
NOC-AE-16003429 Page 5of7 features, and material seleotion of the replaoement steam generators, the previously signifioant degradation meohanisms of tube support plate erosion/oorrosion and oorrosion induoed denting, divider plate oraoking, and wrapper drop have insignifioant potential of ooourring. Sinoe the STP replaoement steam generators are not susceptible to the modes of degradation defined in Generio Letter 97 06, the STP response to NRG Generio Letter 97 06 is no longer applicable.
STP has observed minor tube to support plate wear on the replacement steam generators. To date STP has observed six tube to support plate wear indications. three in each unit. with the largest being 10% thru wall. STP has also observed one potential under deposit pitting corrosion indication in SG 10 during 1RE19. Anti-vibration bar (AVB) wear in the STP steam generators is unlikely, since the STP replacement steam generators have a U-bend region designed to prevent the potential mechanical wear seen in the first generation steam generators.
NUREG-1801 Consistency The Steam Generator Tube Integrity program is an existing program that. following enhancement. will be ffi-consistent with NUREG-1801, Section Xl.M19, Steam Generator Tube Integrity.
Exceptions to NUREG-1801 None Enhancements Prior to the period of extended operation. the following enhancement will be implemented in the following program elements:
Scoping (Element 1)
Procedures will be revised to specify visual inspection of the divider plates assemblies.
tubesheets (primary side). tube to tubesheet welds. and primary head (interior surfaces) for signs of cracking and loss of material.
Parameters of Monitored or Inspected (Element 3). Detection of Aging Effects (Element 4)
Procedures will be revised to specify visual inspection of the divider plates assemblies.
tubesheets (primary side). tube to tubesheet welds. and primary head (interior surfaces) for signs of cracking and loss of material are performed at least every 72 effective full power months or every third refueling outage. whichever results in more frequent inspections.
Acceptance Criteria (Element 6)
Procedures will be revised to evaluate the acceptability of any degraded conditions of the divider plate assemblies. tubesheets (primary side). tube to tubesheet welds. and primary head (interior surfaces) on a case-by-case basis.
Operating Experience NOC-AE-16003429 Page 6 of 7 The Degradation Assessment for STP examines industry experience for Westinghouse advanced-design steam generators to determine the potential degradation mechanisms for STP steam generators. To date, the dominant degradation mechanism detected in U.S. replacement steam generators equipped with Alloy 690TT tubing has been mechanical wear, primarily from foreign objects and a few cases of anti-vibration bar (AVB) wear as well as support plate wear.
Based on industry operating experience and the loose parts wear experienced in STP steam generator 1 D, tube wear at AVB intersections and loose parts wear are considered potential degradation mechanisms. Minor support plate wear and potential under deposit pitting corrosion have been observed at STP. Other degradation mechanisms have a very low likelihood of occurrence. STP has experienced chemistry events with chloride, hydrazine and sodium, where inspected parameters have been found at concentrations outside the specified operating range. All conditions were evaluated and corrective actions were instituted, when appropriate, to prevent reoccurrence.
Pre-service Non-Destructive Examination inspections of the new STP steam generators were performed at the manufacturing site. These included 100 percent full length bobbin coil examinations, and additional tests on select areas of interest. As a result of this preservice inspection, a total of six tubes in the Unit 2 steam generators and 108 tubes in the Unit 1 steam generators were plugged. There are 7585 tubes in each of the four steam generators at STP.
Unit 1 SG B had the most tubes plugged (40) as a result of this pre-service inspection. This pre-service inspection also established the baseline for future eddy current testing of the STP steam generators.
Unit 1 steam generators were replaced in 1 RE09 (March 2000). Unit 1 refueling outage 1 RE11 occurred in April of 2003. In Unit 1, during operating cycle 11, a feed water heater event released foreign materials into steam generator 1 D. Four tubes in the steam generator 1 D cold leg were identified with wear due to wire from the feedwater heater event of operating cycle 11.
One tube required plugging due to a wear depth of 44 percent and the remaining tubes had wear depths of less than 20 percent and remained in service. Two tubes with volumetric indications greater than 20 percent were plugged. The Condition Monitoring limits were met and an Operation Assessment showed no challenge to tube integrity for the next cycle for steam generator 1 D. Steam generators 1A/B/C were approved for the next 3 cycles of operation.
During 1 RE14 (April 2008), 220 foreign objects from the feedwater event of operating cycle 11 were identified during the secondary side visual inspections. No wear was found due to the stabilizer wire that was the most common of the foreign objects. Foreign object retrieval removed 150 of the foreign objects during the 1 RE14 (April 2008) outage. For the items remaining, an engineering evaluation supported by detailed wear calculations provided the conclusion that no tube in steam generator 1 D would experience wear exceeding the plugging limit during the next two cycles of operation. Additional detailed wear calculations will be required before operation beyond the next two cycles and are included in the SG Degradation Assessment.
Based on a review of operating experience, degradation has been consistent with industry experience, including the operating experience identified in NUREG-1801. STP has effectively monitored and trended abnormal conditions. Appropriate corrective actions have been taken, including increasing sampling frequencies, physical inspections, and repair. As additional industry and plant-specific applicable operating experience becomes available, it will be
NOC-AE-16003429 Page 7of7 evaluated and incorporated into the program through the Condition Reporting Process or the Operating Experience program.
Conclusion The continued implementation of the Steam Generator Tube Integrity program provides reasonable assurance that aging effects will be managed such that the systems and components within the scope of this program will continue to perform their intended functions consistent with the current.licensing basis for the period of extended operation.
NOC-AE-16003429 STPNOC Regulatory Commitments - line in/out
Table A4-1 License Renewal Commitments Item#
Commitment 48 Enhance the Steam Generator Tube Integrity grogram grocedures to:
- Sgecify gerform visual insgections of the steam generator head internal areas (head interior surfaces, divider glate assemblies, tubesheets (Qrima!Y side} and tube-to-tubesheet welds} for signs of cracking or loss of material.
Sgecify the freguency of the visual insgections be at least eve!Y 72 effective full gower months or eve!Y third refueling outage whichever results in more freguent insgectior:ts.
Procedures will be revised to evaluate the accegtability of any degraded conditions of the divider glate assemblies, tubesheets (Qrima!Y side}, tube to tubesheet welds, and grima!Y head (interior surfaces} on a case-by-case basis.
LRA Section 82.1.8 NOC-AE-16003429 Page 1 of 1 Implementation Schedule Comglete no later than six months grior to the geriod of extended ogeration CR 16-15866