ML023520628
| ML023520628 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 12/13/2002 |
| From: | Pace P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| WBN-TS-02-16 | |
| Download: ML023520628 (30) | |
Text
Tennessee Valley Authonty, Post Office Box 2000, Spnng City, Tennessee 37381-2000 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.
C.
20555 Gentlemen:
In the Matter of
)
Docket No.
50-390 Tennessee Valley Authority WATTS BAR NUCLEAR PLANT (WBN)
UNIT 1 -
PROPOSED LICENSE AMENDMENT REQUEST CHANGE NO.
WBN-TS-02 STEAM GENERATOR TUBE REPAIR SLEEVE Pursuant to 10 CFR 50.90, TVA is submitting a request for an amendment to WBN's License NPF-90 to change the Technical Specifications (TS) for Unit 1. The proposed TS change will allow the use of Westinghouse leak-limiting Alloy 800 sleeves to repair defective steam generator tubes as an alternative to plugging the tube.
The technique for repairing the degraded tubes is described in the Westinghouse Electric LLC.
Proprietary Class 2 WCAP-15918-P Report Revision 0, (Drafted as CEN-633-P, Revision 05-P),
"Steam Generator Tube Repair for Combustion Engineering and Westinghouse Designed Plant with V4 Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves," dated November 2002.
This report details the analyses and testing performed to verify the adequacy of Alloy 800 sleeves for installation in a steam generator tube and demonstrates sleeving to be an acceptable repair technique. to this letter provides the description and evaluation of the proposed TS change.
This includes TVA's determination that the proposed change does not involve a significant hazards consideration, and is exempt from environmental review. contains copies of the appropriate TS pages from Unit 1 marked-up to show the proposed change. forwards the revised TS pages for Unit 1 which incorporate the proposed changes.
Enclosure 4 provides one copy of the Westinghouse Electric Report Proprietary Class 2 WCAP-15918-P and Enclosure 5 provides Pried ý recycled paper
U.S. Nuclear Regulatory Commission Page 2 DEC 13 2002 a copy of the Non-Proprietary version of the report. provides the Westinghouse Affidavit, Proprietary Information Notice, and Copyright Notice.
NRC approved the use of the leak-limiting Alloy 800 repair sleeves for Calvert Cliffs Nuclear Power Plant Units 1 and 2 on September 1, 1999.
Calvert Cliffs applied the technique in Westinghouse Report CEN-633-P, Revision 3 as the basis for the Alloy 800 repair sleeve.
Revision 3 of the report was for the specific tube size in the Calvert Cliffs steam generators.
Since that time, the report has been revised to include additional testing and analysis and incorporate other industry comments.
Revision 5 of the report is relative to the specific tube size of the Watts Bar steam generators.
TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c) (9).
Additionally, in accordance with 10 CFR 50.91(b) (1),
TVA is sending a copy of this letter and enclosures to the Tennessee State Department of Public Health.
Since Enclosure 4 contains information proprietary to Westinghouse Electric Company, it is accompanied by an affidavit signed by Westinghouse, the owner of the information.
The affidavit (Enclosure 6) sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b) (4)
Section 2.790 of the Commission's regulations.
Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR 2.790 of the Commission's regulation.
Correspondence with the respect to the copyright on proprietary aspects of the Westinghouse information listed above, or the supporting Westinghouse affidavit, should be addressed to H. A. Sepp, Regulatory and Licensing Engineering, Westinghouse Electric Company, P.
- 0. Box 355, Pittsburgh, Pennsylvania 15230-0355.
U.S. Nuclear Regulatory Commission Page 3 DEC 2OO TVA requests approval for the use of this sleeve repair technique before the upcoming Fall 2003 Cycle 5 Refueling Outage to prevent, if appropriate, unnecessary plugging of steam generator tubes.
Approval of the repair technique is needed prior to the outage to prepare the necessary procedures for the scheduled inservice inspection of the steam generator tubes during the outage.
TVA is prepared to meet with the Staff if necessary, to facilitate the Staff's review.
This was discussed with the Staff by teleconference on November 1, 2002.
There are no regulatory commitments associated with this submittal.
If you have any questions about this proposed change, please contact me at (423) 365-1824.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on this 13th day of December, 2002.
Since ly, P. L. Pace Manager of Licensing and Industry Affairs Enclosures
- 1.
TVA Evaluation of Proposed Change.
- 2.
Proposed Technical Specification Changes (mark-up).
- 3.
Proposed Technical Specification Changes (re-typed).
- 4.
Westinghouse Proprietary Class 2, WCAP-15918-P, Revision 0,
(Draft CEN-633-P, Revision 05-P) "Steam Generator Tube Repair for Combustion Engineering and Westinghouse Designed Plant with Y Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves," dated November 2002 (Proprietary).
- 5.
Westinghouse Proprietary Class 3, WCAP-15918-NP, Revision 0, (Draft CEN-633-NP, Revision 05-NP) "Steam Generator Tube Repair for Combustion Engineering and Westinghouse Designed Plant with Y Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves," dated November 2002 (Non-Proprietary).
- 6.
Westinghouse Affidavit, Proprietary Information Notice, and Copyright Notice.
cc:
See page 4
U.S. Nuclear Regulatory Commission Page 4 DEC 13 2002 cc (Enclosures):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr.
L.
Mark Padovan, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St.,
SW, Suite 23T85 Atlanta, Georgia 30303 Mr. Lawrence E.
Nanny, Director (w/o Enclosures 2-6)
Division of Radiological Health 3rd Floor L & C Annex 401 Church Street Nashville, Tennessee 37243
U.S. Nuclear Regulatory Commission Page 5 DEC 1 3 2002 PLP:RNM Enclosures cc (Enclosures):
R.
J.
K.
Baker, BR 3H-C L.
S.
- Bryant, MOB 2R-WBN M. J.
Burzynski, BR 4X-C M. H.
H.
- Evans, WTC 1G-WBN P.
W. Harris, ADM 1V-WBN J.
C.
- Kammeyer, EQB IA-WBN G.
J.
Laughlin, EQB IA-WBN J.
E.
V. Parscale, ADM IB-WBN J.
E.
Semelsberger, EQB 2W-WBN J.
A.
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN)
UNIT 1 DOCKET NO.
390 PROPOSED LICENSE AMENDMENT REQUEST WBN-TS-02-16 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE
- 1.
DESCRIPTION This letter is a request to amend Operating License NPF-90 for Watts Bar Nuclear Plant Unit 1.
The proposed TS change will allow the use of Westinghouse leak-limiting Alloy 800 sleeves to repair defective steam generator tubes as an alternative to plugging the tube.
The technique for repairing the degraded tubes is described in the Westinghouse Electric LLC. Report, WCAP 15918-P, Revision 0, (Draft CEN-633-P, Revision 05-P),
"Steam Generator Tube Repair for Combustion Engineering and Westinghouse Designed Plant with 3 Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves."
This report details the analyses and testing performed to verify the adequacy of Alloy 800 sleeves for installation in a steam generator tube and demonstrates sleeving to be an acceptable repair technique.
TVA is requesting approval of this amendment before the Fall 2003 Cycle 5 Refueling Outage.
- 2.
PROPOSED CHANGE The proposed change which allows steam generator tube sleeves as a repair criteria affects the following sections of Technical Specifications 5.7.2.12, "Steam Generator (SG)
Tube Surveillance Program,":
Section 5.7.2.12.g.l.f
- revises the definition of Plugging Limit to include the use of tube sleeves.
Section 5.7.2.12.g.l.h - revise the definition of Tube Inspection to exclude 1) the portion of the tube below the F* distance for those tubes without sleeve assemblies and
Section 5.7.2.12.g.l.m -
add a new definition for Tube Repair for defective tubes by installing tube sleeves.
Section 5.7.2.12.g.2 - add the words "or repair" after the word "plug."
E-I
Table 5.7.2.12 Add the words "or repair" any place in the table that it indicates the action is to "plug" the defective tubes.
In summary, the proposed change to Technical Specification Section 5.7.2.12, provides an alternative to plugging defective steam generator tubes.
By repairing the defective tubes with the use of leak limiting sleeves, the tube is allowed to remain in service.
WBN Technical Specification Administrative Section 5.0 does not have a supporting Bases section.
- 3.
BACKGROUND Pressurized water reactor steam generators have experienced tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, stress corrosion cracking, and crevice corrosion, along with other phenomena such as denting and vibration wear.
Tubes that experience excessive degradation reduce the integrity of the primary-to-secondary pressure boundary.
Eddy current examination is used to measure the extent of tube degradation.
When the reduction in tube wall thickness reaches a calculated value commonly know as the plugging criteria, when a crack is found, or when other alternate plugging criteria are exceeded, the tube is considered defective and corrective action is taken.
Currently, the Watts Bar Unit 1 Technical Specifications allow defective tubes to be removed from service by installing plugs at both ends of the tube.
The installation of steam generator tube plugs removes the heat transfer surface of the plugged tube from service and leads to a reduction in the primary coolant flow available for core cooling.
The proposed amendment revises the appropriate technical specification section to permit the use of leak-limiting Alloy 800 repair sleeves developed by Westinghouse to be used at Watts Bar Unit 1. Westinghouse provides two types of leak-limiting Alloy 800 repair sleeves.
The first type of sleeve spans the hard rolled transition zone (TZ) of the steam generator tube at the top of the tubesheet.
The TZ repair sleeve is hydraulically expanded into the tube at the upper end and is hard rolled into the tube within the steam generator tubesheet.
The length of the TZ sleeves permits the sleeve to span the degraded tube section at the top of the tubesheet and places the hydraulic expansions above the sludge pile.
The second type of repair sleeve spans degraded areas of the tube at a tube support (TS) plate elevation or in a free span section and is hydraulically expanded into the degraded tube near each end of the sleeve.
There are two distinct advantages associated with the leak-limiting Alloy 800 repair sleeves.
First, no E1-2
welding, brazing, or heat treatment is required during sleeve installation.
Secondly, the strain within the tube is low, thereby reducing the likelihood of future degradation due to stress-influenced mechanisms.
Although the Alloy 800 repair sleeves allow slight leakage past the sleeve (assuming the parent tube is leaking),
the postulated leakage is well within the WBN Technical Specification limits.
The Westinghouse analysis was performed for steam generator tube repair in Combustion Engineering and Westinghouse designed plants with y inch outer diameter Inconel 600 tubes of varying wall thickness, and addresses a combination of one TZ sleeve and/or up to two TS sleeves that could be installed in a single steam generator tube.
Acceptable sleeve locations covered by the analysis are from the fourth tube support plate elevation down to the top of the tubesheet.
WCAP-15918-P Revision 0, (Draft Report CEN-633-P Revision 05-P) (Enclosure 4) provides a detailed description of the design, installation, and testing associated with the Alloy 800 leak-limiting repair sleeves.
NRC approved the use of the leak-limiting Alloy 800 repair sleeves for Calvert Cliffs Nuclear Power Plant Units 1 and 2 on September 1, 1999.
Calvert Cliffs applied the technique in Westinghouse Report CEN-633-P, Revision 3 as the basis for the Alloy 800 repair sleeve.
Revision 3 of the report was for the specific tube size in the Calvert Cliffs steam generators.
Since that time, the report has been revised to include additional testing and analysis and incorporate other industry comments.
Revision 5 of the CEN-633-P report (WCAP-15918-P, Revision 0) is relative to the specific tube size of the Watts Bar steam generators.
- 4.
TECHNICAL ANALYSIS The principal accident associated with this proposed change is the steam generator tube rupture (SGTR) event.
The consequences associated with a SGTR event are discussed in Watts Bar Unit 1 Updated Final Safety Analysis Report Section 15.4.3, "Steam Generator Tube Rupture."
The SGTR event is a breach of the barrier between the reactor coolant system and the main steam system.
The integrity of this barrier is significant from the standpoint of radiological safety in that a leaking steam generator tube allows the transfer of reactor coolant into the main steam system.
In the event of a SGTR, radioactivity contained in the reactor coolant mixes with water in the shell side of the affected steam generator.
This radioactivity is transported by steam to the turbine and then to the condenser, or directly to the condenser via the turbine bypass valves, or directly to the atmosphere via the atmospheric dump valves, main steam safety valves, or the auxiliary feedwater pump turbine E1-3
exhaust.
Noncondensible radioactive gases in the condenser are removed by the condenser air removal system and discharged to the plant vent.
The use of Westinghouse leak-limiting Alloy 800 sleeves allows the repair of degraded steam generator tubes such that the function and integrity of the tube is maintained; therefore, the SGTR accident is not affected.
The consequences of a hypothetical failure of a leak limiting Alloy 800 repair sleeve and/or associated steam generator tube would be bounded by the current SGTR analysis described above.
Due to the slight reduction in diameter caused by the sleeve wall thickness, primary coolant release rates would be slightly less than assumed for the SGTR analysis and, therefore, would result in lower total primary fluid mass release to the secondary system.
A main steam line break (MSLB) or feedwater line break (FLB) will not cause a SGTR since the sleeves are analyzed for a maximum accident differential pressure greater than that predicted in the Watts Bar Unit 1 safety analysis.
The impact of repair sleeving on steam generator performance, heat transfer, and flow restriction is minimal and insignificant compared to plugging.
The proposed WBN Technical Specification change to allow the use of leak-limiting Alloy 800 repair sleeves does not adversely impact any other previously evaluated design basis accident.
Evaluation of the proposed leak-limiting Alloy 800 repair sleeves indicates no detrimental effects on the sleeve or sleeved tube assembly from reactor system flow, primary or secondary coolant chemistries, thermal conditions or transients, or other pressure conditions that may be experienced at Watts Bar Unit 1. The minimal leakage, which is assumed but not expected, experienced during normal operation is well within the established leakage limits when combined with calculated leakage for other alternate plugging criteria.
Data and calculation methodology concerning the reduction in primary coolant flow rate and sleeve-to-plug equivalency ratios is contained in Section 10 of Enclosure 4.
Table 1 below provides a comparison of loading conditions assumed in with respect to Watts Bar Unit 1 corresponding operating and accident values.
The values assumed in are either equivalent or more conservative than Watts Bar Unit 1 plant specific values.
E1-4
TABLE I LOADING CONDITION COMPARISON Watts Bar Unit 1 WCAP-15918-P T-Hot (Primary) Inlet Actual 619.1 0 F 620°F Design 650oF 650OF T-Steam (Secondary)
Actual 538 0 F 526.5 0F Design 600oF 570OF Prim-to-Sec AT Actual 81.1 0 F 93.5 0 F Pressure Primary Actual 2250 psia 2250 psia Design 2500 psia 2500 psia Pressure Secondary Actual 947 psia 877 psia Design 1185 psia 1200 psia Prim-to-Sec AP Actual 1303 psid 1373 psid MSLB/FLB 2405 psi*
2850 psi LOCA 1185 psig 1198 psi Based on pressurizer relief valve setting including instrument error.
The detailed report describing the specific qualifications of Alloy 800 leak-limiting repair sleeves is contained in the proprietary Westinghouse WCAP-15918-P (Draft Report CEN-633-P Revision 05-P),
(Enclosure 4).
The summary of the results from the report are discussed below.
General Structural Assessment The Alloy 800 tubing, from which the repair sleeves are fabricated, is procured to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section II, Part B, SB-163, NiFeCr Alloy UNS N08800, and Section III, Subsection NB-2000.
Additionally, supplemental requirements more tightly controlling parameters within the limits allowed by the ASME specification are imposed.
Fatigue and stress analysis of the sleeved tube assemblies have been completed in accordance with the requirements of Section III ASME Boiler and Pressure Vessel Code and NRC's Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes."
Steam generator tubes with installed Alloy 800 repair sleeves meet the structural integrity requirements of tubes that are not degraded.
Even in the event of the severance of the steam generator tube in the region behind the sleeve, the repaired sleeve will provide the required structural support and acceptable leakage between the primary and secondary systems for normal operating and accident conditions.
The selected design criteria for the repaired sleeves ensure that all design and licensing requirements are considered.
El-5
Extensive testing and analysis have been performed on the repair sleeve and sleeve-to-tube joints to demonstrate that these design criteria are met.
Mechanical testing has been performed to support the analyses prepared using ASME Code stress allowables.
Corrosion testing of sleeve/tube assemblies has been performed in Belgium (Laborelec) and the U.S.
(Westinghouse) with satisfactory results.
These results, when analyzed in conjunction with corrosion test results from the tungsten inert gas-welded sleeve program, confirm the adequacy of the sleeve joint design.
The Alloy 800 sleeve material showed no signs of degradation under high temperature and pressure conditions in a caustic environment, while the sleeve/tube specimens maintained primary side pressure and exhibited no leakage throughout the duration of the test program.
Earlier design variations of this sleeve/tube assembly (larger diametrical hydraulic expansion or varying number of expansions/configurations) were used at KORI 1 (South Korea) and Tihange 3 (Belgium) steam generators.
The current design configuration is in service at Angra 1 (Brazil),
KRSKO (Slovenia),
Ringhals 4 (Sweden),
Tihange 2 (Belgium),
Ulchin 1 & 2 (South Korea),
and Calvert Cliffs 1 and 2 (U.S.) steam generators.
Regulatory Guide 1.121 along with Electrical Power Research Institute's (EPRI's) Steam Generator Degradation Specific Management Flaw Handbook, which adds margin to account for configuration of a long axial crack, are used to develop the structural limit of the repair sleeve should sleeve wall degradation occur as described in Section 8.2 of Enclosure 4.
Alloy 800 leak-limiting repair sleeves are shown (by test and analysis) to retain burst strength in excess of three times the normal operating pressure differential at end of cycle conditions.
No credit for the presence of the parent tube behind the sleeve is assumed when performing the minimum wall burst evaluation for the Alloy 800 repair sleeve.
For sleeves with minimum wall thickness, the structural limit imperfection depth is determined conservatively to be 48 percent (%) and bounds both normal and accident conditions.
Appendix H of the EPRI Pressurized Water Reactor (PWR)
Steam Generator Examination Guideline specify that adequate flaw detection capability in the parent tube be demonstrated for flaws greater than or equal to 60% throughwall.
For the purpose of this sleeve inspection qualification, these values were conservatively reduced to greater than or equal to 50% throughwall for the parent tube and greater than or equal to 45% for the sleeve in order to provide an operational margin between the detection limit and the structural limit for defect growth.
A sufficient number of flaw samples has been used to demonstrate that the statistical requirements for probability of detection are met.
The proposed Technical Specification changes require that a sleeved tube be plugged upon the detection of a defect in the pressure E1-6
boundary portion of the sleeve/tube assembly.
Corrosion Assessment Historically, Alloy 800 has been used successfully for steam generator tubes, tube plugs, and sleeves primarily in Western Europe.
Over 200,000 Alloy 800 tubes have been used for up to 23 years with only minimal tube failures (thinning/wastage, wear).
No evidence of primary or secondary side stress corrosion cracking has been identified in any Alloy 800 tube.
Over 5,300 Alloy 800 repair sleeves of the leak-limiting type design have been used in 10 units worldwide of which none have identified any service induced stress corrosion cracking in the sleeved tube assembly to-date.
Accelerated corrosion testing of Alloy 800 repair sleeve/tube assemblies has been performed in simulated primary and secondary side steam generator environments.
The specific details of Alloy 800 repair sleeve corrosion performance are contained in Section 6 of Enclosure 4.
Mechanical Integrity Assessment Mechanical testing of Alloy 800 repair sleeve/tube assemblies was performed using mock-up steam generator tubes.
The tests determined axial load, pressure load, collapse, burst, leak rates, wear, load cycling, and thermal cycling capability.
The test results correlated well with applicable structural analysis results (analyses always conservative).
The loading conditions developed in Section 8 of Enclosure 4 were used to develop the conditions that were tested in Section 7 of Enclosure 4.
The temperature and pressure differentials described in Section 8 of Enclosure 4 are conservative with respect to Watts Bar Unit 1 operating and accident conditions.
Leakage Rate Assessment The Alloy 800 TZ and TS repair sleeve leakage characteristics were evaluated at room and operating temperatures so that all possible plant conditions would be enveloped by the test results.
Based on the worst-case leakage and excluding calculated leakage from alternate plugging criteria in effect, over 11,000 TZ sleeves could be installed and still meet the Technical Specification leakage limit of 150 gallons per day primary to secondary leakage for a single steam generator during normal operation and 1 gallon per minute for all steam generators during MSLB accident conditions.
Details of the leakage assessment are contained in Section 7 of Enclosure 4.
Sleeve Examination A post-installation sleeve/tube assembly examination will be performed using eddy current testing techniques qualified per EPRI Appendix H criteria.
This examination will establish inservice inspection baseline data and EI-7
initial installation acceptance data on the primary pressure boundary of the sleeve/tube assembly repair.
Subsequent examinations will be consistent with plant Technical Specifications and EPRI Steam Generator Examination Guideline Revision 6 inspection requirements.
Section 5 of Enclosure 4 describes repair sleeve/tube assembly examination methodology.
Conclusion Based on past usage, extensive testing, and analysis the Westinghouse Alloy 800 leak-limiting repair sleeves provide satisfactory repair of defective steam generator tubes.
Design criteria were established based on the requirements of the ASME Code and Regulatory Guide 1.121.
Qualified nondestructive examination techniques will be used to perform necessary repair sleeve and tube inspections for defect detection, and to verify proper installation of the repair sleeve.
In conclusion, based on the considerations discussed
- above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
- 5.
REGULATORY SAFETY ANALYSIS The proposed change is a request to amend Operating License NPF-90 for Watts Bar Nuclear Plant Unit 1. This proposed change revises the Technical Specification for the Steam Generator Tube Surveillance Program to allow the use of Westinghouse leak-limiting Alloy 800 repair sleeves, to repair defective steam generator tubes as an alternative to plugging the tubes.
5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendments(s) by focusing on the three standards set forth in 210 CFR 50.92, "Issuance of Amendment,"
as discussed below:
- 1.
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No.
The Westinghouse Alloy 800 leak-limiting repair sleeves are designed using the applicable American Society of Mechanical Engineers (ASME) Boiler and E1-8
Pressure Vessel Code and, therefore, meet the design objectives of the original steam generator tubing.
The applied stresses and fatigue usage for the repair sleeves are bounded by the limits established in the ASME Code.
Mechanical testing has shown that the structural strength of repair sleeves under normal, upset, emergency, and faulted conditions provides margin to the acceptance limits.
These acceptance limits bound the most limiting (three times normal operating pressure differential) burst margin recommended by NRC's Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes."
Burst testing of sleeve/tube assemblies has demonstrated that no unacceptable levels of primary-to secondary leakage are expected during any plant condition.
The Alloy 800 repair sleeve depth-based structural limit is determined using the NRC guidance and the pressure stress equation of ASME Code,Section III with additional margin added to account for configuration of long axial cracks.
A bounding detection threshold value has been conservatively identified and statistically established to account for growth and determine the repair sleeve/tube assembly plugging limit.
A sleeved tube is plugged on detection of degradation in the sleeve/tube assembly.
Evaluation of the repaired steam generator tube testing and analysis indicates no detrimental effects on the sleeve or sleeved tube assembly from reactor system flow, primary or secondary coolant chemistries, thermal conditions or transients, or pressure conditions as may be experienced at Watts Bar Unit 1. Corrosion testing and historical performance of sleeve/tube assemblies indicates no evidence of sleeve or tube corrosion considered detrimental under anticipated service conditions.
The implementation of the proposed amendment has no significant effect on either the configuration of the plant or the manner in which it is operated.
The consequences of a hypothetical failure of the sleeve/tube assembly is bounded by the current steam generator tube rupture (SGTR) analysis described in Watts Bar Unit 1 Updated Final Safety Analysis Report.
Due to the slight reduction in diameter caused by the sleeve wall thickness, primary coolant release rates would be slightly less than assumed for the steam generator tube rupture analysis and; therefore, would result in lower total primary fluid mass release to the secondary system.
A main steam line break or feedwater line break will not cause a SGTR since E1-9
the sleeves are analyzed for a maximum accident differential pressure greater that that predicted in the Watts Bar Unit 1 safety analysis.
The minimal repair sleeve/tube assembly leakage that could occur during plant operation is well within the Technical Specification leakage limits when grouped with current alternate plugging criteria calculated leakage values.
Therefore, TVA has concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
No.
The Alloy 800 leak-limiting repair sleeves are designed using the applicable ASME Code as guidance; therefore, it meets the objectives of the original steam generator tubing.
As a result, the functions of the steam generators will not be significantly affected by the installation of the proposed sleeve.
The proposed repair sleeves do not interact with any other plant systems.
Any accident as a result of potential tube or sleeve degradation in the repaired portion of the tube is bounded by the existing SGTR accident analysis.
The continued integrity of the installed sleeve/tube assembly is periodically verified by the Technical Specification requirements and the sleeved tube plugged on detection of degradation.
The implementation of the proposed amendment has no significant effect on either the configuration of the plant, or the manner in which it is operated.
Therefore, TVA concludes that this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3.
Does the proposed change involve a significant reduction in a margin of safety?
Response
No.
The repair of degraded steam generator tubes with Alloy 800 leak-limiting repair sleeves restores the structural integrity of the degraded tube under normal operating and postulated accident conditions and thereby maintains current core cooling margin as opposed to plugging the tube and taking it out of service.
The design safety factors utilized for the repair sleeves are El-10
consistent with the safety factors in the ASME Boiler and Pressure Vessel Code used in the original steam generator design.
The portions of the installed sleeve/tube assembly that represent the reactor coolant pressure boundary can be monitored for the initiation of sleeve/tube wall degradation and affected tube plugged on detection.
Use of the previously identified design criteria and design verification testing assures that the margin to safety is not significantly different from the original steam generator tubes.
Therefore, TVA concludes that the proposed change does not involve a significant reduction in a
margin of safety.
Based on the above, TVA concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c),
and, accordingly, a finding of "no significant hazards consideration" is justified.
5.2 Applicable ReQulatory Recquirements/Criteria Based on past usage, extensive testing and analysis, the Westinghouse Alloy 800 leak-limiting repair sleeves provide satisfactory repair of defective steam generator tubes.
Design criteria were established based on the requirements of the ASME Code and Regulatory Guide 1.121.
Qualified nondestructive examination techniques will be used to perform necessary repair sleeve and tube inspections for defect detection, and to verify proper installation of the repair sleeve.
In conclusion, based on the considerations discussed
- above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
6.0.
ENVIRONMENTAL IMPACT CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, El-11
the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c) (9).
Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
El-12
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN)
UNIT 1 PROPOSED LICENSE AMENDMENT CHANGE WBN-TS-02-16 MARKED PAGES I.
AFFECTED PAGE LIST 5.0-18 5.0-19 5.0-19b 5.0-20 INSERT II.
MARKED PAGES See attached.
E2-1
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG)
Tube Surveillance Program (continued) c)
A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or d)
A main steam line or feedwater line break.
- g.
Acceptance Criteria
- 1.
Terms as used in this specification will be defined as follows:
a)
Degradation - A service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube; b)
Degraded Tube -
A tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; c)
% Degradation - The percentage of the tube wall thickness affected or removed by degradation; d)
Defect - An imperfection of such severity that it exceeds the plugging limit.
A tube containing a defect is defective; e)
Imperfection - An exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections; f)
Plugqing Limit means the imperfection depth at ewith INSERT A or beyond which the tube shall be removed from Replacewtservice and is equal to 40% of the nominal tube wall thickness.
This definition does not apply to the portion of the tube in the tubesheet below the F* distance provided the tube is not degraded within the F* distance for F* tubes.
For tubes to wlich tne F* criteria is appLied, a minimum of 1.5 inches of the tube into the tubesheet from the top of the tubesheet or from the bottom of the roll transition, whichever is lower in elevation, shall be inspected using rotating pancake coil eddy current technique or an inspection method shown to give equivalent or better information on the orientation and length of cracking.
A minimum of 1.40 inches (which includes NDE uncertainty) of continuous, sound expanded tube must be established, extending from either the bottom of the roll transition or the top of the tubesheet, whichever is lower in elevation, to the uppermost extent of the indication.
(Continued)
Watts Bar-Unit 1 5.0-18 Amendment 27, 38 Watts Bar-Unit 1 5.0-18 Amendment 27, 38
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG) Tube Surveillance Program (continued)
This definition does not apply to flow distribution baffles and tube support plate intersections for which the voltage-based repair criteria are being applied.
Refer to Specification 5.7.2.12.g.1.1 for repair limit applicable to these intersections.
g)
Preservice Inspection - An inspection of the full length of each tube in each SG performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection shall be performed prior to initial MODE 1 operation using the equipment and techniques expected to be used during subsequent inservice inspections.
h)
Tube Inspection - An inspection of the SG tube Replace with INSERT B from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and i)
Unserviceable - The condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operational Basis Earthquake, a loss-of coolant accident, or a steam line or feedwater line break accident as specified in Specification 5.7.2.12.f.
j)
F* Distance is the distance into the tubesheet from the bottom of the steam generator tube roll transition or the top of the tubesheet, whichever is lower in elevation (further into the tubesheet), that has been conservatively chosen to be 1.40 inches (which includes NDE uncertainty).
k)
F* Tube is the tube with degradation equal to or greater than 40%,
below the F* distance and not degraded (i.e., no indications of degradation) within the F* distance.
- 1)
The Tube Support Plate Repair Limit The Tube Support Plate Repair Limit is used for the disposition of Alloy 600 steam generator tubes for continued service that are experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates and flow distribution baffle (FDB).
At tube support plate intersections (and FDB),
the repair limit is based on maintaining steam generator tube serviceability as described below:
(continued)
Watts Bar-Unit 1 5.0-19 Amendment 27, 38
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG) Tube Surveillance Program (continued)
The mid-cycle repair limits are determined from the following equations:
VwURL
=
- VsT, 1.0 + NDE + Gr [CL - At]
L CL J VMIL =
VMUR-(VUL VLRL)
[CL - At]
L CL j where:
V
=
upper voltage repair limit VLRL
=
lower voltage repair limit V*
=
mid-cycle upper voltage repair limit based on time into cycle VMLRL
=
mid-cycle lower voltage repair limit based on VMRL and time into cycle At
=
length of time since last scheduled inspection during which VUL and VLRL were implemented CL
=
cycle length (the time between two scheduled steam generator inspections)
VSL
=
structural limit voltage Gr
=
average growth rate per cycle length NDE
=
95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e. a value of 20-percent has been approved by the NRC)
Implementation of these mid-cycle repair limits should follow the same approach used in specifications 5.7.2.12.g.1.1.1, 5.7.2.12.g.1.1.2, and 5.7.2.12.g.1.1.3.
The upper voltage repair limit is calculated according to the methodology in GL 95-05 as supplemented.
VoR will differ at the tube support plates and flow SINSERT C 0
distribution baffle.
- 2.
The SG shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 5.7.2.12-1.
- h.
Reports - The content and frequency of written reports shall be in accordance with Specification 5.9.9.
(continued)
- 5. 0-19b Amendment 38 Watts Bar-Unit 1
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)
TABLE 5.7.2.12-1 STEAM GENERATOR TUBE INSPECTION SUPPLEMENTAL SAMPLING REQUIREMENTS ist Sample Inspection 2nd Sample Inspection 3rd Sample Inspection Sample Size Result Action Result Action Result Action I
I Required I
Required I
Required N/A N/A N/A None C-I A minimum of S tubes per SG I
I t
1 N/A N/A None C-I Plug or Repair defective tubes and inspect an additional 2S tubes in this SG C-2 Plug or Repair C-I N/A defective tubes and inspect an additional 4S tubes in this SG.
C-2 Plug or Repair defective tubes C-3 Perform action for C-3 result of first sample N/A N/A Perform action for C-3 result of first sample.
4-
+
I I
t N/A N/A None All other SGs C-1 Some SGs C-Perform action N/A N/A 2 but no for C-2 result other is C-of second 3
sample N/A Inspect all tubes in each SG and plug or repair defective tubes.
Notification to NRC pursuant to 10CFR50 72.
S = 3 N/n %
Where N is the number of SGs in the unit and n is the number of S.G S inspected during an inspection.
Inspect all tubes in this SG, plug or repair defective tubes and inspect 2S tubes in each other SG.
Notification to NRC pursuant to 10CFR50 72 C-3 Additional SG is C-3 C-2 C-3 (continued)
I
( cont inued)
I N/A I
N/A I
N/A 5.0-20 Watts Bar-Unit I
INSERT A:
Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging, or repaired by sleeving in the affected area.
The plugging limit imperfection depths are specified as follows:
- 1.
Original tube wall at greater than or equal to 40% nominal wall.
- 2.
Westinghouse Alloy 800 leak-limiting repair sleeve/tube assembly at detection of degradation as described in the proprietary Westinghouse WCAP-15918-P (Draft CEN-633-P, Revision 05-P),
"Steam Generator Tube Repair For Combustion Engineering and Westinghouse Designed Plants with Y Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves."
This definition does not apply to the portion of the original tube in the tubesheet below the F* distance provided the tube is not degraded within the F* distance for F* tubes.
INSERT B:
Tube Inspection - An inspection of the SG tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg excluding the portion of tube within the tubesheet below the F* distance for a tube with no tubesheet sleeve and excluding the portion of tube within the tubesheet below the sleeve for a tube with a tubesheet sleeve.
INSERT C:
m)
Tube Repair refers to a process that reestablishes tube serviceability.
Tube repair of defective tubes will be performed where applicable by installation of the Westinghouse Alloy 800 leak-limiting repair sleeve as described in the proprietary Westinghouse Report WCAP-15918-P (Draft CEN-633-P, Revision 05-P),
"Steam Generator Tube Repair For Combustion Engineering and Westinghouse Designed Plants with Y, Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves".
ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN)
UNIT 1 PROPOSED TECHNICAL SPECIFICATION (TS)
CHANGE WBN-TS-02-16 REVISED PAGES I.
AFFECTED PAGE LIST 5.0-18 5.0-19 5.0-19a 5.0-19b 5.0-19c 5.0-20 II.
REVISED PAGES See attached.
E3-1
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG) Tube Surveillance Program (continued) c)
A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or d)
A main steam line or feedwater line break.
- g.
Acceptance Criteria
- 1.
Terms as used in this specification will be defined as follows:
a)
Degradation - A service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube; b)
Degraded Tube - A tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; c)
% Degradation - The percentage of the tube wall thickness affected or removed by degradation; d)
Defect - An imperfection of such severity that it exceeds the plugging limit.
A tube containing a defect is defective; e)
Imperfection - An exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections; f)
Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging, or repaired by sleeving in the affected area.
The plugging limit imperfection depths are specified as follows:
- 1. original tube wall at greater than or equal to 40% of the nominal wall.
- 2. Westinghouse Alloy 800 leak-limiting repair sleeve/tube assembly at detection of degradation as described in the proprietary Westinghouse WCAP-15918-P (Draft CEN-633-P, Revision 05-P),
"Steam Generator Tube Repair For Combustion Engineering and Westinghouse Designed Plants with Y4 Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves."
This definition does not apply to the portion of the original tube in the tubesheet below the F*
distance provided the tube is not degraded within the F* distance for F* tubes.
For tubes to which the F* criteria is applied, a minimum of 1.5 inches of the tube into the tubesheet from the top of the tubesheet or from the bottom of the roll transition, whichever is lower in elevation, shall be inspected using rotating pancake coil eddy current technique or (Continued)
Watts Bar-Unit 1 5.0-18 Amendment 27, 38 Amendment 27, 38 Watts Bar-Unit 1 5.0-18
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG)
Tube Surveillance Program (continued) an inspection method shown to give equivalent or better information on the orientation and length of cracking.
A minimum of 1.40 inches (which includes NDSE uncertainty) of continuous, sound expanded tube must be established, extending from either the bottom of the roll transition or the top of the tubesheet, whichever is lower in elevation, to the uppermost extent of the indication.
This definition does not apply to flow distribution baffles and tube support plate intersections for which the voltage-based repair criteria are being applied.
Refer to Specification 5.7.2.12.g.1.1 for repair limit applicable to these intersections.
g)
Preservice Inspection - An inspection of the full length of each tube in each SG performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection shall be performed prior to initial MODE 1 operation using the equipment and techniques expected to be used during subsequent inservice inspections.
h)
Tube Inspection - An inspection of the SG tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg excluding the portion of tube within the tubesheet below the F* distance for a tube with no tubesheet sleeve and excluding the portion of tube within the tubesheet below the sleeve for a tube with a tubesheet sleeve.
i)
Unserviceable - The condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operational Basis Earthquake, a loss-of coolant accident, or a steam line or feedwater line break accident as specified in Specification 5.7.2.12.f.
j)
F* Distance is the distance into the tubesheet from the bottom of the steam generator tube roll transition or the top of the tubesheet, whichever is lower in elevation (further into the tubesheet), that has been conservatively chosen to be 1.40 inches (which includes NDE uncertainty).
k)
F* Tube is the tube with degradation equal to or greater than 40%,
below the F* distance and not degraded (i.e., no indications of degradation) within the F* distance.
- 1)
The Tube Support Plate Repair Limit - The Tube Support Plate Repair Limit is used for the disposition of Alloy 600 steam generator tubes for continued service that are experiencing (Continued)
Watts Bar-Unit 1 5.0-19 Amendment 27, 38
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG)
Tube Surveillance Program (continued) predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates and flow distribution baffle (FDB).
At tube support plate intersections (and FDB),
the repair limit is based on maintaining steam generator tube serviceability as described below:
- 1.
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the flow distribution baffles and tube support plates with bobbin voltages less than or equal to the lower voltage repair limit of 1.0 volt will be allowed to remain in service.
- 2.
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the flow distribution baffles and tube support plates with the bobbin voltage greater than the lower voltage repair limit of 1.0 volt, will be repaired, except as noted in Specification 5.7.2.12.g.1.1.3 below.
- 3.
Steam generator tubes with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the flow distribution baffles and tube support plates with a bobbin voltage greater than the lower voltage repair limit of 1.0 volt but less than or equal to the upper voltage limit*,
may remain inservice if a rotating pancake coil inspection does not detect degradation.
Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit* will be plugged or repaired.
- 4.
Certain intersections as identified in of WAT-D-10709 ("Tennessee Valley Authority, Watts Bar Nuclear Power Plant Unit 1, Application for Implementation of Voltage Based Repair Criteria, Westinghouse Steam Generator Tubes Affected by ODSCC at TSPs,"
J.
W.
Irons, Revision 0, 1/12/00) will be excluded from application of the voltage based repair criteria as it is determined that these intersection may collapse or (continued)
Watts Bar-Unit 1 5.0-19a Amendment 38
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG) Tube Surveillance Program (continued) deform following a postulated LOCA + SSE event.
- 5.
If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 5.7.2.12.g.1.1.1, 5.7.2.12.g.1.1.2, and 5.7.2.12.g.1.1.3.
The mid-cycle repair limits are determined from the following equations:
V M
=
V L 1.0 + NDE + Gr [CL At]
L CL J VMuR
=
VMUL (VUL VLRL)
[CL -
At]
L CL I where:
V
=
upper voltage repair limit V
=
lower voltage repair limit VML
=
mid-cycle upper voltage repair limit based on time into cycle VMuR
=
mid-cycle lower voltage repair limit based on V*
and time into cycle At
=
length of time since last scheduled inspection during which VRL and Vu* were implemented CL
=
cycle length (the time between two scheduled steam generator inspections)
VSL
=
structural limit voltage Gr
=
average growth rate per cycle length NDE
=
95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e. a value of 20-percent has been approved by the NRC)
Implementation of these mid-cycle repair limits should follow the same approach used in specifications 5.7.2.12.g.1.1.1, 5.7.2.12.g.1.1.2, and 5.7.2.12.g.1.1.3.
The upper voltage repair limit is calculated according to the methodology in GL 95-05 as supplemented.
VuR will differ at the tube support plates and flow distribution baffle.
(continued)
Amendment 38 Watts Bar-Unit 1
- 5. 0-19b
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Steam Generator (SG)
Tube Surveillance Program (continued) m)
Tube Repair refers to a process that reestablishes tube serviceability. Tube repair of defective tubes will be performed where applicable by installation of the Westinghouse Alloy 800 leak-limiting repair sleeve as described in the proprietary Westinghouse Report WCAP-15918-P (Draft CEN-633-P, Revision 05-P),
"Steam Generator Tube Repair For Combustion Engineering and Westinghouse Designed Plants with Y4 Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves".
- 2.
The SG shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 5.7.2.12-1.
- h.
Reports - The content and frequency of written reports shall be in accordance with Specification 5.9.9.
(continued)
Amendment 38 Watts Bar-Unit 1 5.0-19c
Procedures,
- Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)
TABLE 5.7.2.12-1 STEAM GENERATOR TUBE INSPECTION SUPPLEMENTAL SAMPLING REQUIREMENTS ist Sample Inspection 2nd Sample Inspection 3rd Sample Inspection Sample Size Result Action Result Action Result Action Required Required Required A minimum of C-1 None N/A N/A N/A N/A S tubes per SG C-2 Plug or Repair C-i None N/A N/A defective tubes and inspect an additional 2S tubes in this SG.
C-2 Plug or Repair C-i N/A defective tubes and inspect an additional 4S tubes in this SG C-2 Plug or Repair defective tubes.
C-3 Perform action for C-3 result of first sample.
C-3 Perform action N/A N/A for C-3 result of first sample C-3 Inspect all All other None N/A N/A tubes in this SGs C-i SG, plug or repair defective tubes and inspect 2S tubes in each other SG.
Notification to NRC pursuant to 10CFR50 72 Some SGs C-Perform action N/A N/A 2 but no for C-2 result other is C-of second 3
sample Additional Inspect all N/A N/A SG is C-3 tubes in each SG and plug or repair defective tubes.
Notification to NRC pursuant to 10CFR50.72 S = 3 N/n t Where N is the number of SGs in the unit and n is the number of S G.s inspected during an inspection (continued)
Amendment Watts Bar-Unit 1 5.0-20