ML11305A076
| ML11305A076 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 10/25/2011 |
| From: | Rencurrel D South Texas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME4936, TAC ME4937 | |
| Download: ML11305A076 (18) | |
Text
Nuclear Operating Company South Texas Prolect Electric Generatig Station PO Box 289 Wadsworth, Texas 77483 October 25, 2011 NOC-AE-1 1002744 10CFR54 STI: 32991978 File: G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2746 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Response to Requests for Additional Information for the South Texas Proiect License Renewal Anolication (TAC Nos. ME4936 and ME4937'*
References:
- 1. STPNOC Letter dated October 25, 2010, from G. T. Powell to NRC Document Control Desk, "License Renewal Application" (NOC-AE-10002607)
- 2. NRC letter dated September 22, 2011, "Requests for Additional Information for the Review of the South Texas Project, Units 1 and 2 License Renewal Application - Aging Management Review, Set 3 (TAC Nos. ME4936 and ME4937)" (ML11258A161)
By Reference 1, STP Nuclear Operating Company (STPNOC) submitted a License Renewal Application (LRA) for South Texas Project (STP) Units 1 and 2. By Reference 2, the NRC staff requests additional information for review of the STP LRA. STPNOC's response to the request for additional information is provided in Enclosure 1 to this letter. Changes to the LRA described in Enclosure 1 are depicted in line-in/line-out pages provided in Enclosure 2.
There are no regulatory commitments in this letter.
Should you have any questions regarding this letter, please contact either Arden Aldridge, STP License Renewal Project Lead, at (361) 972-8243 or Ken Taplett, STP License Renewal Project regulatory point-of-contact, at (361) 972-8416.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on
/bt_____,,
Date O.W. Rencurrel Senior Vice President, Technical Support & Oversight KJT
Enclosures:
- 1. STPNOC Response to Requests for Additional Information
NOC-AE-l 1002744 Page 2 cc:
(paper copy without enclosures)
Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 612 East Lamar Blvd, Suite 400 Arlington, Texas 76011-4125 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8B1) 11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN1 16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 John W. Daily License Renewal Project Manager (Safety)
U.S. Nuclear Regulatory Commission One White Flint North (MS 011-Fl)
Washington, DC 20555-0001 Tam Tran License Renewal Project Manager (Environmental)
U. S. Nuclear Regulatory Commission One White Flint North (MS O11F01)
Washington, DC 20555-0001 (electronic copy without enclosures)
A. H. Gutterman, Esquire Kathryn M. Sutton, Esquire Morgan, Lewis & Bockius, LLP John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pena City Public Service Peter Nemeth Crain Caton & James, P.C.
C. Mele City of Austin Richard A. Ratliff Alice Rogers Texas Department of State Health Services Balwant K. Singal John W. Daily Tam Tran U. S. Nuclear Regulatory Commission NOC-AE-1 1002744 Enclosure I STPNOC Response to Requests for Additional Information NOC-AE-1 1002744 Page 1 of 7 SOUTH TEXAS PROJECT, UNITS I AND 2 REQUESTS FOR ADDITIONAL INFORMATION -
AGING MANAGEMENT REVIEW SET 3 (TAC NOS. ME4936 AND ME4937)
Selective Leaching (034)
RAI 3.3.2.3.7-1 Backgqround:
In license renewal application (LRA) Table 3.3.2-7, the applicant includes an aging management review (AMR) item for copper alloy greater than 15 percent zinc solenoid valve internally exposed to plant indoor air. The applicant stated that the component will be managed for loss of material using the Selective Leaching of Materials program. The AMR item lists generic note G and plant-specific note 3, indicating that this material exposed to plant indoor air is subject to wetting due to condensation, and thus is subject to loss of material due to selective leaching.
The Generic Aging Lessons Learned (GALL) Report and the Metals Handbook Desk Edition (Second Edition, ASM International, 1998) both state that copper alloy greater than 15 percent zinc may be subject to stress corrosion cracking in solutions containing ammonia or ammonia-like compounds such as amines provided sufficient tensile stresses are present.
Issue:
The staff does not have sufficient information regarding the control or use of ammonia or amines (e.g., cleaning solutions, chemicals, decay of insects) in the vicinity of the instrument air system (air intake) to determine if stress corrosion cracking should be an aging effect requiring aging management.
Request:
Describe what measures are taken to. prevent or limit the presence of ammonia and amines in the instrument air system.
STPNOC Response:
The instrument air intakes are located inside the Turbine Generator Building (TGB). The TGB contains the instrument air compressor intake area which is a designated housekeeping area with no standing water or nearby enclosures in which insects could build-up. No ammonia-based chemicals are used in the instrument air compressor intake area. Some cleaning solutions may contain a very low concentration of ammonia but the cleaning solutions would be used at such a low percentage of time that it is considered insignificant. Therefore, instrument air intake piping is not exposed to airborne amines or ammonia-based compounds in sufficient quantities to be of significance. A review of South Texas Project (STP) operating experience did not find any evidence of stress corrosion cracking associated with copper alloy greater than 15 percent zinc.
NOC-AE-1 1002744 Page 2 of 7 Buried Piping (035)
RAI 3.3.2.3.4-1 Backgiround:
In LRA Table 3.3.2-4, the applicant includes an AMR item for copper alloy piping (greater than 8 percent aluminum) externally exposed to soil. The applicant stated that the component will be managed for loss of material using the Buried Piping and Tanks Inspection program. The AMR item lists generic note G, indicating that the environment is not in the GALL Report for the material and environment. The Metals Handbook states that copper alloy (greater than 8 percent aluminum) may be subject to stress corrosion cracking in solutions containing ammonia or ammonia-like compounds such as amines, provided sufficient tensile stresses are present.
Issue:
Table 2.2-1 of the Updated Final Safety Analysis Report states a chemical plant that previously used anhydrous ammonia is located 4.8 miles NNE of the South Texas Project (STP) site. The Environmental Report (ER) states that some of the STP site east of the main cooling reservoir is leased for cattle grazing. The ER also describes the land surrounding the plant as fairly flat and used for ranchland and farmland.
The plant procedure for implementing the Buried Piping and Tanks Inspection Program states that soil analysis data (i.e., pH, resistivity, redox potential, sulfide and sulfate ion concentration, chloride concentration, conductivity, and moisture content) should be collected during excavations to help assess the likelihood of pipe outside diameter corrosion. However, the procedure does not indicate if the soil analysis tests for the presence of ammonia or ammonia-like compounds.
The staff does not have sufficient information regarding the presence or absence of ammonia or ammonia-like compounds in the soil in and around the buried copper alloy (greater than 8 percent aluminum) piping such that stress corrosion cracking could be ruled out as a possible aging effect requiring aging management.
Request:
Describe what, if any, measures are taken to detect the presence or absence of ammonia in the soil near the buried piping of interest. If it is determined that there is a potential for ammonia or ammonia-like compounds to be present in the soil in the vicinity of the piping of interest, describe what measures will be taken to manage stress corrosion cracking of the buried copper alloy (greater than 8 percent aluminum) piping.
STPNOC Response:
The outdoor environment at STP is not subject to industrial pollution. The closest industrial facility is 4.8 miles away. There have been no industrial ammonia events detected at the STP plant site. The land adjacent to the plant site is open range for cattle and farming. There are no large concentrations of cattle within five miles of the plant site that could generate excessive detrimental gases or concentrated solid waste. There is no runoff from adjacent land onto the plant site. A search of the STP corrective action database did not identify any ammonia or ammonia-like compound spills or contaminations that affected on-site soil conditions. There is no evidence to expect the soil at STP to have elevated levels of ammonia or ammonia-like NOC-AE-1 1002744 Page 3 of 7 compounds. Therefore, it is not necessary to consider ammonia-related aging effects on buried copper alloy (>8% aluminum) components.
Diesel Exhaust Piping (078)
RAI 3.3.2.2.3.3-1
Background:
LRA Table 3.3.2-21 includes stainless steel expansion joints exposed to diesel exhaust (internal) for the nonsafety-related diesel generator that are being managed for loss of material. For the corresponding material and environment, the GALL Report recommends managing for both loss of material and cracking due to stress corrosion cracking, and recommends using a plant-specific AMP.
Issue:
The stainless steel expansion joint exposed to diesel exhaust in LRA Table 3.3.2-21 is not being managed for stress corrosion cracking as recommended by the GALL Report.
Request:
Provide the basis for not managing the stainless steel expansion joint exposed to diesel exhaust in Table 3.3.2-21 for stress corrosion cracking or provide a suitable AMP that will manage this aging effect for this material and environment combination.
STPNOC Response:
LRA Table 3.3.2-21 and Section 3.3.2.1.21 will be revised to add an aging effect of cracking for stainless steel expansion joints exposed to an internal environment of diesel exhaust using NUREG 1801 AMR line VII.H2-1. Aging management program B2.1.22, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components, will be used to manage the cracking of stainless steel expansion joints exposed to an internal environment of diesel. provides the line-in/line-out sections of the License Renewal Application.
Boric Acid Corrosion (010)
RAI 3.1.1.58-1
Background:
In LRA Tables 3.1.2-1, 3.1.2-2, 3.1.2-3, and 3.1.2-4, the applicant stated that several steel component external surfaces exposed to borated water leakage are managed for loss of material by the Boric Acid Corrosion Program (LRA Section B2.1.4). These items are associated with LRA Table 3.1-1, item 3.1.1.58.
The updated staff guidance in SRP-LR, Revision 2, Table 3.1-1, item 48, states that steel external surfaces, including reactor vessel top head, bottom head, and reactor coolant pressure boundary piping or components adjacent to dissimilar metal welds exposed to air with borated water leakage, should be managed for loss of material due to boric acid corrosion by GALL NOC-AE-1 1002744 Page 4 of 7 AMPs XI.M10, "Boric Acid Corrosion" and XI.M1 1 B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components."
The GALL AMP XI.M1 1 B "scope of program" program element states that this program manages loss of material due to boric acid corrosion in steel components in the vicinity of nickel-alloy components, including, but not limited to, reactor vessel components, steam generator components, pressurizer components, and reactor coolant system piping. The program description states that inspection activities should be in accordance with 10 CFR 50.55a, including ASME Code Cases N-722-1 and N-729-1, and industry guidelines for inspection of primary system butt welds (e.g. MRP-1 39).
Issue:
The program description in LRA Section B2.1.4, "Boric Acid Corrosion," refers to inservice inspections in accordance to ASME Code Section Xl; however, it is not clear to the staff whether the requirements in 10 CFR 50.55a, including Code Cases N-722-1 and N-729-1, and MRP-1 39, are incorporated in those inspections.
Request:
Clarify whether the inservice inspections in the Boric Acid Corrosion Program are in accordance with 10 CFR 50.55a, including ASME Code Cases N-722-1 and N-72901, and MRP-139. If not, provide information on what equivalent inspection activities will be used to manage loss of material due to boric acid corrosion of steel components in the vicinity of nickel-alloy reactor coolant pressure boundary components.
STPNOC Response:
The inspection activities identified by 10 CFR 50.55a, including ASME Code Cases N-722-1 and N-729-1, and MRP-1 39, pertain to the detection of cracking of nickel-alloy components (including welds) and do not pertain to the loss of material of carbon steel components. At STP the aging effect of loss of material due to boric acid corrosion for steel components in the vicinity of nickel-alloy components is managed by the Boric Acid Corrosion program (B2.1.4). The Boric Acid Corrosion program includes provisions to identify leakage, inspect and examine for evidence of leakage, evaluate the effects of leakage, and initiate corrective actions, as appropriate.
The aging effect of cracking of susceptible nickel-alloy components (including welds) of the reactor pressure boundary is managed by Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors program (B2.1.5), and plant-specific program Nickel Alloy Aging Management program (B2.1.34). Aging management program B2.1.5 requires implementation of ASME Code Case N-729-1, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(D)(2) through (6), in the ASME Section Xl, In-Service Inspections.
Aging management program B2.1.34 requires implementation of ASME code case N-722, subject to the conditions listed in 10 CFR 50.55a(g)(6)(ii)(E)(2) through (4), into the ASME Section Xl, In-Service Inspections. Aging management program B2.1.34 also implements examinations consistent with MRP-1 39.
As discussed above, ASME code cases N-729-1 and N-722 (subject to the conditions of 10CFR50.55a) and MRP-139 are applied to management of aging of Nickel Alloy components, similar to GALL Rev 2 XI.M1 1 B requirements.
NOC-AE-1 1002744 Page 5 of 7 RAI 3.3.1.88-1
Background:
SRP-LR, Revision 1, Table 3.3-1, item 88 states that aluminum and copper alloy greater than 15 percent Zn piping, piping components, and piping elements exposed to air with borated water leakage should be managed for loss of material due to borated water leakage by GALL AMP XI.M10, "Boric Acid Corrosion." LRA Table 3.3.1, item 3.3.1.88 states that this item is not applicable because there is no in-scope aluminum or copper alloy greater than 15 percent Zn piping, piping components, or piping elements exposed to air with borated water leakage in the auxiliary systems.
LRA Section 3.3.2.1.19 states that the chemical and volume control system (CVCS), an auxiliary system, contains an environment of borated water leakage. The staff noted that in LRA Table 3.3.2-19, the AMR results for the CVCS include an item for aluminum insulation; however, the only environment cited is plant indoor air (external).
Issue:
Given that borated water leakage is a recognized environment in the CVCS, it is not clear to the staff why the aluminum insulation in this system is not managed for loss of material due to boric acid corrosion.
Request:
Clarify whether the aluminum insulation in the chemical and volume control system may be exposed to borated water leakage. If so, state how loss of material due to boric acid corrosion will be managed.
STPNOC Response:
Aging management evaluation for a treated borated water leakage environment is considered applicable only for components that contain treated borated water, and is not applicable for adjacent system components or insulation on the piping that contains the treated borated water.
It is possible that the aluminum sheathing could be exposed to treated borated water leakage.
The loss of material due to boric acid corrosion caused by treated borated water leakage from the system to the aluminum sheathing is managed by the Boric Acid Corrosion program as described in LRA Appendix B2.1.4. The Boric Acid Corrosion program covers mechanical, electrical, and structural components made of materials susceptible to boric acid corrosion on which borated water leakage may occur and evaluates any components with evidence of boric acid exposure, including insulation aluminum sheathing.
Reactor Head Closure Studs Program (003)
RAI B2.1.3-4
Background:
SRP-LR, Revision 2, Table 3.0-1 addresses aging management programs used to manage the aging effects associated with various systems and the descriptions of the programs, which are acceptable for the UFSAR supplement. Specifically, SRP-LR, Revision 2, Table 3.0-1 addresses NOC-AE-1 1002744 Page 6 of 7 the UFSAR supplement description of GALL AMP XI.M3, "Reactor Head Closure Studs," by referring to the inservice inspections in conformance with the requirements of the ASME Code, Section Xl, Subsection IWB, Table IWB-2500-1 and preventive measures to mitigate cracking.
SRP-LR, Revision 2, Table 3.0-1 further states that the program also relies on recommendations to address reactor head stud bolting degradation as delineated in NUREG-1339 and NRC Regulatory Guide (RG) 1.65. NUREG-1339 and RG 1.65 indicate that molybdenum sulfide is a potential contributor to stress corrosion cracking (SCC). NUREG-1 339 and RG 1.65 (Revision 1, April 2010) also include guidance for the yield strength levels of the bolting material resistant to SCC.
In comparison, LRA Section A1.3 provides the UFSAR supplement description for LRA Section B2.1.3, "Reactor Head Closure Studs Program." This LRA Section states that the applicant's program follows the preventive measures in RG 1.65. However, the UFSAR supplement described in LRA Section A1.3 does not include the statement that the applicant's program relies on recommendations to address reactor head stud bolting degradation as delineated in NUREG-1339 and NRC RG 1.65.
Issue:
In contrast with SRP-LR, Revision 2, Table 3.0-1, the applicant's UFSAR supplement for the Reactor Head Closure Studs Program (described in LRA Section A1.3) does not include the statement that the applicant's program relies on recommendations to address reactor head stud bolting degradation as delineated in NUREG-1339 and NRC RG 1.65. The licensing basis for this program for the period of extended operation may not be adequate if the applicant does not incorporate this information in its UFSAR supplement.
Request:
Revise the applicant's UFSAR supplement description for the Reactor Head Closure Studs Program to be consistent with the UFSAR supplement described in SRP-LR, Revision 2, Table 3.0-1, which incorporates recommendations in NUREG-1339 and NRC RG 1.65.
If the applicant has determined that a revision to the UFSAR supplement description is not necessary, justify why the omission of the information from the UFSAR supplement, regarding NUREG-1339 and NRC RG 1.65, is acceptable to provide an adequate licensing basis for this program for the period of extended operation.
STPNOC Response:
The Reactor Head Closure Studs program manages cracking and loss of material by conducting ASME Section Xl inspections of reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, and bushings. The program includes periodic visual, surface, and volumetric examinations of reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, and bushings and the performance of visual inspections of the reactor vessel flange closure during primary system leakage tests. The program implements recommendations in NUREG-1 339 and NRC Regulatory Guide 1.65 to address reactor head stud bolting degradation except for yield strength of existing bolting materials. The program implements ASME Section X1 Code, Subsection IWB, and detects reactor vessel stud, nut, washer, and bushing cracking, loss of material due to wear and corrosion, and reactor coolant leakage from the reactor pressure vessel flange. STP will use the ASME Code edition consistent with the provisions of 10 CFR 50.55a during the period of extended operation.
NOC-AE-1 1002744 Page 7 of 7 LRA Appendices A1.3 and B2.1.3 will be revised to include reference to NUREG-1 339 and NRC Regulatory Guide 1.65. provides the line-in/line-out sections of the License Renewal Application.
NOC-AE-1 1002744 STPNOC LRA Changes with Line-inlLine-out Annotations NOC-AE-1 1002744 Page 1 of 7 List of Revised LRA Sections
-Affected LRA Section RAI Section 3.3.2.1.21 3.3.2.2.3.3-1 Table 3.3.2-21 3.3.2.2.3.3-1 Appendix A1.3 B2.1.3-4 Appendix B2.1.3 B2.1.3-4 South Texas Project License Renewal Application Amendment 6 NOC-AE-1 1002744 Page 2 of 7 3.3.2.1.21 Nonsafety-related Diesel Generators and Auxiliary Fuel Oil System Aging Effects Requiring Management The following nonsafety-related diesel generators and auxiliary fuel oil system aging effects require management:
Crackinq Loss of material Loss of preload South Texas Project License Renewal Application Amendment 6 NOC-AE-1 1002744 Page 3 of 7 Table 3.3.2-21 Auxiliary Systems - Summary of Aging Management Evaluation - Nonsafety-related Diesel Generators and Al i tvili;n~
t:"m n1 Qi/olfnm
~ '.~~L
~~"
Component Type 1Intended Material Environment Aging Effect Aging Management, NUREG-Table.1 Notes Function Requiring Program f1801 Vol.
Item Management Item Closure Bolting Expansion Joint PB tp._BB Carbon Steel Stainless Steel Plant Indoor Air
(.Ext)
Diesel Exhaust Diesel Exhaust (Int)
Li C
Expansion Joint PB oss of preload
'Bolting Integrity 1(B2.1.7)_
racking
.Inspection of Internal
\\
Surfaces in Miscellaneous Piping
,and Ducting
_ Components (B2.1.22) oss of material Inspection of Internal
\\
'Surfaces in
'Miscellaneous Piping
.and Ducting
'Components (B2.1.22)
/11.1-5
/Il. H2-*
/11.H2-:
3.3.1.45 Stainless Steel Lc B
1 3.3.1.06 2
3.3.1.18 E
South Texas Project License Renewal Application Amendment 6 NOC-AE-1 1002744 Page 4 of 7 A1.3 REACTOR HEAD CLOSURE STUDS The Reactor Head Closure Studs program manages cracking and loss of material by conducting ASME Section Xl inspections of reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, and bushings. The program includes periodic visual, surface, and volumetric examinations of reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, and bushings and performs visual inspections of the reactor vessel flange closure during primary system leakage tests. The pro-gram implements recommendations in NUREG-1339 and NRC Regulatory Guide 1.65 to address reactor head stud bolting degradation except for yield strength of existing bolting materials. The program implements ASME Section Xl code, Subsection IWB, and detects reactor vessel stud, nut, washer, and bushing cracking, loss of material due to wear and corrosion, and reactor coolant leakage from the reactor vessel flange. STP will use the ASME Code edition consistent with the provisions of 10 CFR 50.55a during the period of extended operation.
South Texas Project License Renewal Application Amendment 6 NOC-AE-1 1002744 Page 5 of 7 B2.1.3 Reactor Head Closure Studs Program Description The Reactor Head Closure Studs program manages cracking and loss of material by conducting ASME Section Xl inspections of reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, and bushings. The program includes periodic visual, surface, and volumetric examinations of reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, and bushings and performs visual inspections of the reactor vessel flange closure during primary system leakage tests. The STP program implements ASME Section Xl code, Subsection IWB, 2004 Edition. Reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, and bushings are identified in ASME Section Xl Tables IWB-2500-1 and are within the scope of license renewal. ST.P follows the preventive m.easures The program implements recommendations in NUREG-1339 and iR-NRC Regulatory Guide 1.65, Material and Inspection for Reactor Vessel Closure Studs, to address reactor head stud boltinq degradation except for yield strength of existing bolting materials. STP uses lubricants on reactor head closure stud threads after reactor head closure stud, nut, and washer cleaning and examinations are complete. The lubricants are compatible with the stud material and operating environment and do not include MoS 2 which is a potential contributor to stress corrosion cracking.
In conformance with 10 CFR 50.55a(g)(4)(ii), the STP ISI Program is updated during each successive 120-month inspection interval to comply with the requirements of the latest edition of the Code specified twelve months before the start of the inspection interval. STP will use the ASME Code Edition consistent with the provisions of 10 CFR 50.55a during the period of extended operation.
Potential cracking and loss of material conditions in reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, and bushings are detected through visual, surface, or volumetric examinations in accordance with ASME Section Xl requirements in STP procedures every ten years. These inspections are conducted during refueling outages. Reactor vessel studs are removed from the reactor vessel flange each refueling outage. Studs, nuts, washers, and bushings are stored in protective racks after removal. Reactor vessel flange holes are plugged with water tight plugs during cavity flooding. These methods assure the holes, studs, nuts, washers, and bushings are protected from borated water during cavity flooding. Reactor vessel flange leakage is detected prior to reactor startup during reactor coolant system pressure testing each refueling outage. The STP program has proven to be effective in preventing and detecting potential aging effects of reactor vessel flange stud hole threads, closure studs, nuts, washers, and bushings.
NUREG-1801 Consistency The Reactor Head Closure Studs program is an existing program that is consistent, with exception to NUREG-1801,Section XI.M3, Reactor Head Closure Studs.
South Texas Project License Renewal Application Amendment 6 NOC-AE-1 1002744 Page 6 of 7 Exceptions to NUREG-1801 Proqram Elements Affected:
Scope of Program (Element 1)
Regulatory Guide 1.65 states that the ultimate tensile strength of stud bolting material should not exceed 170 ksi. One closure head insert has a tensile strength of 174.5 ksi. STP credits inservice inspections that are within the scope of this AMP, which are implemented in accordance with the STP Inservice Inspection Program, Examination Category B-G-1 requirements, as the basis for managing cracking in these components. This is in accordance with the "parameters monitored or inspected" and "detection of aging effects" program elements in NUREG 1801,Section XI.M3. In addition, the studs, nuts and washers are coated with a lubricant which is compatible with the stud materials, and the studs, nuts, and washers are protected from exposure to boric acid by removing them and plugging the reactor vessel flange holes during cavity flooding.
Corrective Actions (Element 7)
NUREG-1801,Section XI.M3 specifies the use of Regulatory Guide 1.65 requirements for closure stud and nut material. STP uses SA-540, Grade B-24 (as modified by Code Case 1605) stud material. The use of this material has been found acceptable to the NRC for this application within the limitations discussed in Regulatory Guide 1.85, Materials Code Case Acceptability.
Enhancements None Operating Experience Review of plant-specific operating experience has not revealed any program adequacy issues with the Reactor Head Closure Studs program for reactor vessel closure studs, nuts, washers, bushings, and flange thread holes. No cases of cracking due to SCC or IGSCC have been identified with STP reactor vessel studs, nuts, washers, bushings, and flange stud holes.
Review of the Refueling Outage Inservice Inspection Summary Reports for Interval 2 indicates there were no repair/replacement items identified with reactor vessel closure studs, nuts, washers, bushings, or flange thread holes. None of the repair/replacement items indicate any implementation issues with the STP ASME Section Xl Program for reactor closure studs, nuts, washers, bushings, or flange thread holes.
The ISI Program at STP is updated to account for industry operating experience. ASME Section Xl is also revised every three years and addenda issued in the interim, which allows the code to be updated to reflect operating experience. The requirement to update the ISI Program to reference more recent editions of ASME Section Xl at the end of each inspection interval ensures the ISI Program reflects enhancements due to operating experience that have been incorporated into ASME Section Xl.
South Texas Project License Renewal Application Amendment 6 NOC-AE-1 1002744 Page 7 of 7 Conclusion The continued implementation of the Reactor Head Closure Studs 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.
South Texas Project License Renewal Application Amendment 6