Text

Units 1 & 2 - Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification Change
	ML062400349
Person / Time
Site:	Byron, Braidwood
Issue date:	08/18/2006
From:	Benyak D; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-06-109
	Download: ML062400349 (73)
	v • d • e

Exelkn.

Exelon Generation www.exeloncorp.com Nuclear 4300 Winfield Road Warrenville, IL60555 10 CFR 50.90 RS-06-109 August 18, 2006 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. 50-454 and 50-455

Subject:

Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification

References:

(1) Letter from J. A. Bauer (Exelon Generation Company, LLC) to U. S.

NRC, "Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity," dated November 18, 2005 (2) Letter from S. J. Campbell (NRC) to C. M. Crane (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Unit Nos. 1 and 2 - Request for Additional Information Related to Technical Specification Improvement Regarding Steam Generator Tube Integrity," dated August 4, 2006 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Appendix A Technical Specifications (TS), of Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively. The proposed changes were requested to revise the TS requirements related to steam generator tube integrity. The change was consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity."

In the course of their review the NRC determined that additional information is required as documented in Reference 2. The attachments to this letter provide the response to the requested information. The response is subdivided as shown below.

Attachment 1 provides the EGC responses to the Request for Additional Information.

August 17, 2006 U.S. Nuclear Regulatory Commission Page 2 Attachments 2-A and 2-B include the marked-up TS pages with the proposed changes indicated for Braidwood Station and Byron Station, respectively.

Attachments 3-A and 3-B include the associated typed TS pages with the proposed changes incorporated for Braidwood Station and Byron Station, respectively.

Attachments 4-A and 4-B include the associated revised TS Bases for Braidwood Station and Byron Station, respectively, for information only. provides an affidavit for withholding signed by Westinghouse Electric Company, LLC, the owner of proprietary information provided in Attachment 6. Also enclosed are a Westinghouse authorization letter CAW-06-2181, Proprietary Information Notice and Copyright Notice. provides a proprietary version of Westinghouse document SG-SGDA-06-20-P-Attachment, Revision 1, "Exelon: Byron Unit 2 and Braidwood Unit 2, Response to Request for Additional Information on FP&L Seabrook License Amendment Request 05 Limited Inspection of the Steam Generator Tube Portion Within the Tubesheet." provides a non-proprietary version of Westinghouse document SG-SGDA-06-20-NP-Attachment, Revision 1. contains information proprietary to Westinghouse Electric Company, LLC; it is supported by an affidavit signed by Westinghouse, the owner of the information.

The affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR 2.390.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-06-2181 and should be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

August 18, 2006 U.S. Nuclear Regulatory Commission Page 3 The information provided in this letter does not affect the supporting analysis for the original license amendment request as described in Reference 1. No other information contained in the Reference 1 letter is affected by this additional information.

The No Significant Hazards Consideration and the Environmental Consideration provided in Attachment 1 of Reference 1 are not affected by this additional information.

In accordance with 10 CFR 50.91(b), "State consultation," EGC is providing the State of Illinois with a copy of this letter and its attachments to the designated State Official.

If you have any questions about this letter, please contact David Chrzanowski at (630) 657-2816.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 18th day of August 2006.

Respectfully, Darin M. Benyak Manager - Licensing : Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification Change A: Revised Markup of Technical Specifications Pages for Braidwood Station B: Revised Markup of Technical Specifications Pages for Byron Station A: Revised Typed Technical Specifications Pages for Braidwood Station B: Revised Typed Technical Specifications Pages for Byron Station A: Revised Typed Technical Specifications Bases Page for Braidwood Station B: Revised Typed Technical Specifications Bases Page for Byron Station : Application for Withholding and Affidavit : Proprietary Version - Westinghouse SG-SGDA-06-20-P-Attachment, Revision 1 : Non-proprietary Version - Westinghouse SG-SGDA-06-20-NP-Attachment, Revision 1

Attachment 1 BRAIDWOOD STATION UNITS 1 AND 2 Docket Nos. STN 50-456 and STN 50-457 License Nos. NPF-72 and NPF-77 and BYRON STATION UNITS 1 AND 2 Docket Nos. STN 50-454 and STN 50-455 License Nos. NPF-37 and NPF-66 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification Change

Attachment 1 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification Note that the SG-SGDA-06-20-P-Attachment, Revision 1 (Attachment 6 to this submittal) and SG-SGDA-06-20-NP-Attachment, Revision 1 (Attachment 7 to this submittal) reports provided by Westinghouse Electric Company, LLC (Westinghouse) were intended to provide responses to NRC draft questions posed to Seabrook Station, Braidwood and Byron Stations as well as providing information in support of other utilities proposing similar steam generator tube inspection license amendment requests. The responses to these draft questions in the Westinghouse reports are formatted in the Seabrook Station draft question sequence.

Also, these Westinghouse reports were prepared prior to the formal issuance of the finalized version of the Braidwood Station and Byron Station NRC questions, as a result, the documents make reference to two questions that no longer require clarification.

Therefore, the responses provided for "Exelon Byron/Braidwood Draft RAI Questions 5 and 7" contained in the Westinghouse reports are no longer applicable.

Wherever necessary, the specific location in these Westinghouse reports of the relevant Braidwood Station and Byron Station information is identified in the response to the question.

NRC Question 1:

Proposed Specification 5.5.9.f discusses provisions for Unit 2 Steam Generator(SG) tube repairmethods. The proposedtechnical specifications (TS) for BraidwoodUnits 1 and 2, and Byron Units 1 and2, including LCO 3.4.19, refer to "pluggedor repaired."

However, proposed TS 5.5.9.f addressesprovisions for SG tube repairmethods for BraidwoodStation Unit 2 and Byron Station Unit 2 only. The staff is concerned that the proposed TS could be misconstruedto mean that there are no restrictionswith respect to repairsfor BraidwoodStation Unit 1 and Byron Station Unit 1. Discussyour plans for revising specification 5.5.9.f such as to clarify that tube repairsmay not be performed for Unit 1.

Exelon Generation Comoanv. LLC (EGC) response to Question 1:

In order to clarify that there are currently no approved SG tube repair methods for Braidwood Station Unit 1 and Byron Station Unit 1, the proposed TS 5.5.9.f wording is being revised. The statement: 'There are no approved tube repair methods for the Unit 1 SGs," is being added to Section 5.5.9.f. See Attachment 2-A (for Braidwood Station) and 2-B (for Byron Station) for the revised wording of Insert 5.5-9 and Insert 5.5-10. To aid the review, these inserts are highlighted to indicate the most recent changes. The typed TS pages incorporating these changes are provided in Attachment 3-A (for Braidwood Station) and Attachment 3-B (for Byron Station).

Page 1 of 6

Attachment 1 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification NRC Question 2:

The NRC staff notes that the existing TS reportingrequirementsdo not addressissues associatedwith implementation of the tubesheet inspection and alternaterepaircriterion.

Discuss Exelon's plans to revise proposed TS 5.6.9, "Steam Generator(SG) Inspection Report," to include reportingrequirements applicable to the implementation of the tubesheet inspection and alternaterepaircriterion,similarto what the staff has requestedother licensees submitting applicationsfor tubesheet inspection and alternate repaircriteria.

EGC response to Question 2:

The proposed TS 5.6.9 wording is being revised to include three additional items to be included in the 180-day SG report for Braidwood Station Unit 2 and Byron Station Unit 2.

The additional items address the reporting details regarding indications detected in the upper 17-inches of the tubesheet thickness, the operational primary to secondary leakage rate observed during the previous Unit 2 cycle, and the calculated accident leakage rate from the lowermost 4-inches of tubing.

See Attachment 2-A (for Braidwood Station) and 2-B (for Byron Station) for the revised wording of Insert 5.6-6. To aid the review, this insert is highlighted to indicate the most recent change. The typed TS pages incorporating these TS 5.6.9 changes are provided in Attachment 3-A (for Braidwood Station) and Attachment 3-B (for Byron Station).

NRC Question 3:

On page 7 of 17 of Attachment 1, "Evaluationof ProposedChanges,"of the November 18, 2005 submittal, Exelon indicatedthat the roll transitionzone sleeve lowerjoint is located nearthe neutralaxis of the tubesheet (i.e., within the portion of the tube that will be inspected). Confirm that the Licensing Report CEN-621-P, Revision 00, "Commonwealth Edison Byron and Braidwood Unit 1 and Unit 2 Steam GeneratorsTube Repair Using Leak Tight Sleeves, FinalReport," dated April 1995, precludes the establishmentof the joint in the lower 4 inches of the tubesheet. If it does not, provide technicaljustification why a joint in the lower 4 inches of the tubesheet is allowable.

Alternatively, rewrite the proposed TSs to preclude the establishmentof joints in the lower 4 inches of the tubesheet.

EGC response to Question 3:

As indicated in topical report (CEN-621-P) Figures 4-3A, "Roll Transition Zone Sleeve (Short) Installation," and 4-3B, "Roll Transition Zone Sleeve (Long) Installation," the bottom edge of the ABB-CE sleeve design is 11.25 inches to 17.0 inches from the hot leg tube end, thus installation of the sleeve within the bottom four inches of the tubesheet is prohibited.

Page 2 of 6

Attachment 1 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification NRC Question 4:

In the proposed TS (and TSTF-449), a SG tube is defined as the entire length of the tube, including the tube wall (andany repairsmade to it), between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. Given this definition, the proposed repaircriteriain TS 5.5.9.c could be misinterpreted.Discuss Exelon's plans to modify the TSs to more clearly define the repaircriteria for the sleeved portion of a tube.

EGC response to Question 4:

TS 5.5.9.c, "Provisions for SG tube repair criteria," is being revised. See Attachment 2-A (for Braidwood Station) and 2-B (for Byron Station) for the revised wording of Insert 5.5-

8. To aid the review, this insert is highlighted to indicate the most recent change. The typed TS pages incorporating these TS 5.5.9 changes are provided in Attachment 3-A (for Braidwood Station) and Attachment 3-B (for Byron Station).

NRC Question 5:

Regarding the revised BASES,

a. ProposedPage B 3.4.13 - 3 states, "The dose consequences resulting from the Locked Rotor with a ConcurrentSG PORV[power-operatedrelief valve] FailureAccident are well within the limits defined in 10 CFR [Part] 100." This statement replaces the current statement, "The dose consequences resulting from the SLB accidentare well within the limits defined in 10 CFR [Part]100." Why has this revision been made?

Aren't the consequences of all analyzed events within the limits defined in 10 CFR Part 100? If so, why not simply state that?

b. On page B 3.4.19-4, there is a statement that the accidentinduced leakage criteriais 1 gpm for all steam generatorsexcept for specific types of degradationat specific locations. This wording (i.e., the exception) is not in the proposed TS. Describe Exelon's plans for resolving this discrepancy.

EGC response to Question 5:

The current wording in the Braidwood Station and Byron Station Bases, page 3.4.13-3 states: 'The dose consequences resulting from the Locked Rotor with a Concurrent SG PORV Failure accident are well within the limits defined in 10 CFR 100." There were no changes proposed to the current wording. The wording identified in Question 5 that states: 'The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100," is the wording proposed in TSTF-449. This generic statement in the TSTF-449 should have been identified as a bracketed change since the limiting design basis accident (DBA) from a dose perspective is site-specific.

For Braidwood and Byron Stations the Locked Rotor with a concurrent SG Power Operated Relief Valve (PORV) Failure presents the design basis challenge for primary to Page 3 of 6

Attachment 1 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification secondary SG leakage only. As stated earlier in Section B 3.4.13, the Braidwood Station and Byron Station:

"Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a Locked Rotor with a Concurrent Steam Generator (SG) Power Operated Relief Valve (PORV) Failure accident because such leakage contaminates the secondary fluid. Other accidents or transients involve secondary steam release to the atmosphere, such as a Steam Generator Tube Rupture (SGTR). The SGTR is more limiting than the Locked Rotor with a Concurrent SG PORV Failure for site radiation releases."

While this specific Braidwood Station and Byron Station Bases 3.4.13 wording accurately describes the limiting DBA from a primary to secondary SG tube leakage perspective and no changes are being considered for this TS amendment submittal, the Braidwood Station and Byron Station license amendment request to incorporate Alternative Source Term, currently under NRC review, is proposing to modify the current Base 3.4.13 words to encompass all relevant DBAs by stating:

'The dose consequences resulting from the MSLB, SGTR, Control Rod Ejection and Locked Rotor accidents are within the limits defined in 10 CFR 50.67."

No additional actions are required to address this part of Question 5.

The original proposed wording on page 4 of new Bases Section B 3.4.19 states:

"The accident induced leakage requirement of 1 gpm for all SGs, except for specific types of degradation at specific locations where the NRC has approved greater accident induced leakage, bounds the accident analysis assumptions for primary to secondary LEAKAGE."

This wording was included TSTF-449 to address a licensee's alternate repair criteria allowing for greater than one gpm leakage. Braidwood Station and Byron Station are not invoking this type of alternate repair criteria, therefore, the words will be revised to state:

"The accident induced leakage requirement of 1 gpm for all SGs bounds the accident analysis assumptions for primary to secondary LEAKAGE."

The typed Bases page incorporating this revised wording is provided, for information only, in Attachment 4-A (for Braidwood Station) and Attachment 4-B (for Byron Station).

Page 4 of 6

Attachment 1 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification NRC Question 6:

Under the proposed 17-inch tubesheet inspection zone, it is the Exelon's contention that the accident leakage integrity of the tubing below the 17-inch inspection zone is ensured by the bell-weatherprinciple. The NRC staff requests that Exelon submit a leakage sensitivity study to support the conservatism of the bellwether approach. That is, leakage during accidents will not exceed two times that observed during normal operatingconditions. The NRC staff requests that this study consideraxial and circumferentialflaws locatedat the bottom of the tubesheet at three tubesheet radial locations;i.e., at the zero radius, mid-radius,and peripherallocations. For each type of crack at each location, leakage under normal operatingand accident conditionsshould be evaluated consideringonly the crack leakage resistance,consideringonly the tube to tubesheet annulus resistanceand, lastly, consideringthe total resistance of the crack and annulus to leakage.

EGC response to Question 6:

The complete response to this question is provided on pages 8 through 16 of the proprietary version of SG-SGDA-06-20-P-Attachment, Revision 1 (i.e., Attachment 6 of this submittal) and is identified as "Response 5." Because the response to this question contains information proprietary to Westinghouse, only partial information is provided on pages 8 through 16 of the non-proprietary version (i.e., Attachment 7).

NRC Question 7:

Section 8.2 of Attachment 7provides a justification for why ligament tearing of circumferentialcracks is not a significant concern. Provide a justificationfor why ligament tearingof axialcracks at the bottom of the tubesheet at the peripheryis similarlynot a significantconcern.

EGC response to Question 7:

The complete response to this question is provided on pages 20 through 23 of the proprietary version of SG-SGDA-06-20-P-Attachment, Revision 1 (i.e., Attachment 6) and is identified as "Response 9." Because the response to this question contains information proprietary to Westinghouse, only partial information is provided on pages 20 through 23 of the non-proprietary version (i.e., Attachment 7).

Page 5 of 6

Attachment 1 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification NRC Question 8:

Are there any tubes in the Byron and Braidwood, Unit 2 SGs which were not fully expanded (pernominal)within the tubesheet? If so, please describe the extent of this condition andjustify why the amendment request is sufficient to ensure the structural and leakage integrity of the affected tube joints.

EGC response to Question 8:

A review of eddy current data for Byron Station Unit 2 was performed during the 12 th refueling outage in the Fall of 2005 (B2R112). One SG tube was identified as not being fully expanded through the tubesheet. This tube was rerolled and expanded an appropriate length to accommodate a mechanical plug. The tube was then preventively plugged prior to restart from the B2R12 refueling outage.

As described on page 25 of both the Attachment 6 (proprietary) and the Attachment 7 (non-proprietary) versions of SG-SGDA-06-20-P-Attachment, Revision 1, identified as "Response 10," the eddy current reports for Braidwood Station Unit 2 were reviewed to determine if there were indications of partially or fully unexpanded tubes within the tubesheet. Based on this data all the tubes currently inservice, within the Braidwood Station Unit 2 tubesheet, are fully expanded.

Page 6 of 6

Attachment 2A BRAIDWOOD STATION UNITS 1 AND 2 Docket Nos. STN 50-456 and STN 50-457 License Nos. NPF-72 and NPF-77 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification Change Revised Markup of Technical Specifications Pages

Steam Generator Program (Braidwood)

INSERT 5.5-8

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed a total of 1 gpm for all SGs.

3. The operational LEAKAGE performance criteria is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria.

1. ITubes found by inservice inspection to contain flaws in a non-sleeved region with a depth equal to or exceeding 40% of the nominal wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in TS,5.5.9.c.4.

2. Sl eeVessfound by inservice inspection to contain flaws with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness shall be plugged:

i. ... TIG welded sleeves (per TS 5.5.9.f.2.i): 32%

3. -Tubes with a flaw in a sleeve to tube joint that occurs in the sleeve or in the original tube wall of the joint shall be plugged:

4.' The' following tube repair criteria may be applied as an alternate to the 40% depth-based criteria of Technical Specification 5.5.9.c.1:

i.. For Unit 2 only,' degradationh fouird in the portion of the tube below.17 inches from the top of the hot leg tubesheet does not require plugging or repair.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Unit 2, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. The tube-to-tubesheet weld is not

Steam Generator Program (Braidwood) part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to

Steam Generator Program (Braidwood)

INSERT 5.5-9 determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the Unit 1 tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

Inspect 100% of the Unit 2 tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of

Steam Generator Program (Braidwood)

INSERT 5.5-10 these Specifications, tube plugging is not a repair.

1. There are no approved tube repair methods for the Unit 1 SGs.

2., All acceptable repair methods for the Unit 2 SGs are listed below.

i. TIG welded sleeving as described in ABB Combustion Engineering Inc., Technical Reports:

Licensing Report CEN-621-P, Revision 00, "Cormmonwealth Edison Byron and Braidwood Unit 1 and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, FINAL REPORT," April 1995; and Licensing Report CEN-627-P, "Operating Performance of the ABB CENO Steam Generator Tube Sleeve for Use at Commonwealth Edison Byron and Braidwood Units 1 and 2," January 1996; subject to the limitations and restrictions as noted by the NRC Staff.

Steam Generator Program (Braidwood)

Insert TS 5.6-6 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged or repaired to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,

h. The effective plugging percentage for all plugging and tube repairs in each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

j. .For Unit, 2,- the number ofihndications and location, size, orientation, and whether initiated on primary or secondary side for each indication detected in the upper 17-inches of the tubesheet thickness.

k. or Unit 2, *the boperational -primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report,' and

1. For Unit 2, the calculated accident leakage rate from the lowermost 4-inches of tubing for the most limiting accident in the most limiting steam generator. In addition, if the calculated accident leakage rate from the most limiting accident is less than 2 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined.

Attachment 2B BYRON STATION UNITS 1 AND 2 Docket Nos. STN 50-454 and STN 50-455 License Nos. NPF-37 and NPF-66 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification Change Revised Markup of Technical Specifications Pages

Steam Generator Program (Byron)

INSERT 5.5-8

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed a total of 1 gpm for all SGs.

3. The operational LEAKAGE performance criteria is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria.

1. Tubes found by inservice inspection' to contain flaws in a non-sleeved region with a depth equal to or exceeding 40% of the nominal wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in TS 5.5.9.6.4.

2. ... Sleeves 'found by inserVice inspection to contain flaws with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness shall be plugged:

i. TIG welded sleeves (per TS 5.5.9.f.2.i): 32%

3.: Tubes 'with a-flaw in-a sleeve'to tube joint that occurs in the sleeve or in the original tube wall of the joint shall be plugged.

4.1' The followihg tube repair-criteria may be applied as an alternate to the 40% depth-based criteria of Technical Specification 5.5.9.c.1:

i.',- For Unit 2 only degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging or repair.,

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Unit 2, the portion of e tube below 17 inches from the top of the hot leg tubesheet is excluded. The tube-to-tubesheet weld is not

Steam Generator Program (Byron) part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to

Steam Generator Program (Byron)

INSERT 5.5-9 determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the Unit 1 tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

Inspect 100% of the Unit 2 tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of

Steam Generator Program (Byron)

INSERT 5.5-10 these Specifications, tube plugging is not a repair.

1. There are no approved tube repair methods for the Unit 1 SGs.

2. Alliacceptable repair methods for the Unit 2 SGs are

.listed below.

i. TIG welded sleeving as described in ABB Combustion Engineering Inc., Technical Reports:

Licensing Report CEN-621-P, Revision 00, "Commonwealth Edison Byron and Braidwood Unit 1 and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, FINAL REPORT," April 1995; and Licensing Report CEN-627-P, "Operating Performance of the ABB CENO Steam Generator Tube Sleeve for Use at Conmnonwealth Edison Byron and Braidwood Units 1 and 2," January 1996; subject to the limitations and restrictions as noted by the NRC Staff.

Steam Generator Program (Byron)

Insert TS 5.6-6 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged or repaired to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,

h. The effective plugging percentage for all plugging and tube repairs in each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

For Unit 2, th*number. of indications and location, size, orientation, and whether initiated: on primary or secondary side for each indication detected in the upper 17-inches of the tubesheet thickness.

k, For Unit 2, the O-eratiOnal primary tosecOndary leakage

,rate observed (greater than'three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report, and

1. For Unit 2,-the 'calculated accident leakage rate from the lowermost 4-inches of tubing for generator.

the most limiting In accident in the most limiting steam addition, if the calculated accident leakage rate from the most limiting accident is less than 2 times the maximum operational primary to secondary leakage rate,'

the report should describe how it was determined.

Attachment 3A BRAIDWOOD STATION UNITS 1 AND 2 Docket Nos. STN 50-456 and STN 50-457 License Nos. NPF-72 and NPF-77 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification Change Revised Typed Technical Specifications Pages 5.5-8 5.5-9 5.5-10 5.6-6 5.6-7

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed a total of 1 gpm for all SGs.

3. The operational LEAKAGE performance criteria is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria.

1. Tubes found by inservice inspection to contain flaws in a non-sleeved region with a depth equal to or exceeding 40% of the nominal wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in TS 5.5.9.c.4.

2. Sleeves found by inservice inspection to contain flaws with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness shall e plugged:

i. TIG welded sleeves (per TS 5.5.9.f.2.i): 32%

3. Tubes with a flaw in a sleeve to tube joint that occurs in the sleeve or in the original tube wall of the joint shall be plugged.

4. The following tube repair criteria may be applied as an alternate to the 40% depth-based criteria of Technical Specification 5.5.9.c.1:

i. For Unit 2 only, degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging or repair.

BRAIDWOOD - UNITS 1 & 2 5.5- 8 Amendment

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Proqram (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Unit 2 the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the.

tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the Unit 1 tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

Inspect 100% of the Unit 2 tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

BRAIDWOOD - UNITS 1 & 2 5.5- 9 Amendment

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair.

1. There are no approved tube repair methods for the Unit 1 SGs.

2. All acceptable repair methods for the Unit 2 SGs are listed below.

i. TIG welded sleeving as described in ABB Combustion Engineering Inc., Technical Reports:

Licensing Report CEN-621-P, Revision 00, "Commonwealth Edison Byron and Braidwood Unit 1 and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, FINAL REPORT," April 1995; and Licensing Report CEN-627-P, "Operating Performance of the ABB CENO Steam Generator Tube Sleeve for Use at Commonwealth Edison Byron and Braidwood Units 1 and 2," January 1996; subject to the limitations and restrictions as noted by the NRC Staff.

BRAIDWOOD - UNITS 1 & 2 5.5 - 10 Amendment

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.8 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported in the Inservice Inspection Summary Report in accordance with 10 CFR 50.55a and ASME Section XI, 1992 Edition with the 1992 Addenda.

5.6.9 Steam Generator (SG) Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program.

The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged or repaired to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,

h. The effective plugging percentage for all plugging and tube repairs in each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

j. For Unit 2, the number of indications and location, size, orientation, and whether initiated on primary or secondary side for each indication detected in the upper 17-inches of the tubesheet thickness.

BRAIDWOOD - UNITS 1 & 2 5.6- 6 Amendment

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.9 Steam Generator (SG) Tube Inspection Report (continued)

k. For Unit 2, the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report, and

1. For Unit 2, the calculated accident leakage rate from the lowermost 4-inches of tubing for the most limiting accident ii the most limiting steam generator. In addition, if the calculated accident leakage rate from the most limiting accident is less than 2 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined.

BRAIDWOOD - UNITS 1 & 2 5.6 - 7 Amendment

Attachment 3B BYRON STATION UNITS 1 AND 2 Docket Nos. STN 50-454 and STN 50-455 License Nos. NPF-37 and NPF-66 Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification Change Revised Typed Technical Specifications Pages 5.5-8 5.5-9 5.5-10 5.6-6 5.6-7

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) ProQram (continued)

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed a total of 1 gpm for all SGs.

3. The operational LEAKAGE performance criteria is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria.

1. Tubes found by inservice inspection to contain flaws in a non-sleeved region with a depth equal to or exceeding 40% of the nominal wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in TS 5.5.9.c.4.

2. Sleeves found by inservice inspection to contain flaws with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness shall be pl ugged:

i. TIG welded sleeves (per TS 5.5.9.f.2.i): 32%

3. Tubes with a flaw in a sleeve to tube joint that occurs in the sleeve or in the original tube wall of the joint shall be plugged.

4. The following tube repair criteria may be applied as an alternate to the 40% depth-based criteria of Technical Specification 5.5.9.c.1:

i. For Unit 2 only, degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging or repair.

BYRON- UNITS 1 & 2 5.5- 8 Amendment

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weldat the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Unit 2 the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the Unit 1 tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

Inspect 100% of the Unit 2 tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

BYRON - UNITS 1 & 2 5.5- 9 Amendment