ET 09-0016, Revision to Technical Specifications 5.5.9, Steam Generator (SG) Program, and TS 5.6.10, Steam Generator Tube Inspection Report, for a Permanent Alternate Repair Criterion

Revision to Technical Specifications 5.5.9, Steam Generator (SG) Program, and TS 5.6.10, Steam Generator Tube Inspection Report, for a Permanent Alternate Repair Criterion

ML091590170
Person / Time
Site:	Wolf Creek
Issue date:	06/02/2009
From:	Garrett T Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ET 09-0016
Download: ML091590170 (37)

Text

W.LF CREEK NUCLEAR OPERATING CORPORATION Terry J. Garrett June 2, 2009 Vice President Engineering ET 09-0016 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

1) Letter dated April 28, 2005, from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station Issuance of Exigent Amendment RE: Steam Generator (SG) Tube Surveillance Program (TAC NO.

MC6757)"

2) Letter dated October 10, 2006, from J. N. Donohew, USNRC, to R. A.

Muench, WCNOC, "Wolf Creek Generating Station Issuance of Amendment RE: Steam Generator Tube Inspections Within the Tubesheet (TAC NO. MD 2467)"

3) Letter dated April 4, 2008, from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station - Issuance of Amendment RE:

Revision to Technical Specification 5.5.9 on the Steam Generator Program (TAC NO. MD8054)"

Subject:

Docket No. 50-482: Revision to Technical Specifications 5.5.9, "Steam Generator (SG) Program," and TS 5.6.10, "Steam Generator Tube Inspection Report," for a Permanent Alternate Repair Criterion Gentlemen:

Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). This amendment request proposes to permanently revise Wolf Creek Generating Station (WCGS) Technical Specification (TS) 5.5.9, "Steam Generator (SG)

Program," to exclude portions of the tube below the top of the steam generator tubesheet from periodic steam generator tube inspections. In addition, this amendment request proposes to revise TS 5.6.10, "Steam Generator Tube Inspection Report," to provide reporting requirements specific to the permanent alternate repair criteria. This permanent change is supported by Westinghouse Electric Company LLC, WCAP-17071-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)."

P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET

ET 09-0016 Page 2 of 4 References 1 and 2 approved one-cycle revisions to TS 5.5.9 to exclude from inspection and repair, portions of the tube below 17 inches from the top of the tubesheet. Reference 3 approved an interim alternate repair criteria that requires full-length inspection of the tubes within the tubesheet but does not require plugging tubes if any circumferential cracking observed in the region greater than 17 inches from the top of the tubesheet is less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads.

WCNOC requests the proposed change be approved by October 1, 2009, to support Refueling Outage 17, which is scheduled to start in October, 2009. Once approved, the amendment will be implemented prior to MODE 4 entry during startup from Refueling Outage 17.

Enclosure I provides the proprietary Westinghouse Electric Company LLC WCAP-17071-P, "H*:

Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)." Enclosure II provides the non- proprietary Westinghouse Electric Company LLC WCAP-17071-NP, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)." As Enclosure I contains information proprietary to Westinghouse Electric Company LLC, it is supported by an affidavit signed by Westinghouse Electric Company LLC, the owner of the information. The affidavit 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) of 10 CFR 2.390 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 2.390 of the Commission's regulations. This affidavit, along with Westinghouse authorization letter, CAW-09-2566, "Application for Withholding Proprietary Information from Public Disclosure," is contained in Enclosure Ill.

Attachment I through IV provide the Evaluation, Markup of TSs, Retyped TS pages, and proposed TS Bases changes, respectively, in support of this amendment request. Attachment IV, proposed changes to the TS Bases, is provided for information only. Final TS Bases changes will be implemented pursuant to TS 5.5.14, "Technical Specification (TS) Bases Control Program," at the time the amendment is implemented. Attachment V provides a List of Regulatory Commitments made by WCNOC in this submittal.

It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

This amendment application was reviewed by the Plant Safety Review Committee. In accordance with 10 CFR 50.91, a copy of this amendment application, with attachments, is being provided to the designated Kansas State official.

ET 09-0016 Page 3 of 4 If you have any questions concerning this matter, please contact me at (620) 364-4084, or Mr.

Richard D. Flannigan at (620) 364-4117.

Sincerely, Terry J. Garrett TJG/rlt Attachments: Evaluation IV Markup of Technical Specification pages II III Markup of Technical Specification Bases pages (for information only)

IV Retyped Technical Specification pages List of Regulatory Commitments Enclosure I - Westinghouse Electric Company LLC WCAP-17071-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)"

11 - Westinghouse Electric Company LLC WCAP-17071-NP, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)"

III - Westinghouse Electric Company LLC LTR-CAW-09-2566, "Application for Withholding Proprietary Information from Public Disclosure" cc: E. E. Collins (NRC), w/a, w/e T. A. Conley (KDHE), w/a, w/e (Enclosure II only)

V. G. Gaddy (NRC), w/a B. K. Singal (NRC), w/a Senior Resident Inspector (NRC), w/a

ET 09-0016 Page 4 of 4 STATE OF KANSAS )) s COUNTY OF COFFEY )

Terry J. Garrett, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

Terr. Garrett Vic resident Engineering SUBSCRIBED and sworn to before me this 2 0 day of JwT- ,2009.

Nota--ublic SL-* GAYLE SHEPHEARDi a&-* Notary Public - State of Kansas.

MyAppt. Expires 171/ o0 /-// _)o I/

Expiration Date _

Attachment I to ET 09-0016 Page 1 of 17 EVALUATION

Subject:

Revision to Technical Specifications 5.5.9, "Steam Generator (SG) Program," and TS 5.6.10, "Steam Generator Tube Inspection Report," for a Permanent Alternate Repair Criterion

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION 4,1 Applicable Regulatory Requirements/Criteria 4.2 Significant Hazards Consideration 4.3 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

Attachment I to ET 09-0016 Page 2 of 17

1.

SUMMARY

DESCRIPTION Wolf Creek Nuclear Operating Corporation (WCNOC) proposes to revise Wolf Creek Generating Station (WCGS) Technical Specification (TS) 5.5.9, "Steam Generator (SG)

Program," to exclude portions of the tube below the top of the steam generator tubesheet from periodic steam generator tube inspections. In addition, this amendment request proposes to revise TS 5.6.10, "Steam Generator Tube Inspection Report," to provide reporting requirements specific to the permanent alternate repair criteria. Application of the structural analysis and leakage evaluation results to exclude portions of the tubes from inspection and repair of tube indications is interpreted to constitute a redefinition of the primary to secondary pressure boundary.

The NRC previously issued Amendment Number 162 (Reference 1) and Amendment Number 169 (Reference 2) to exclude the portion of the tubes below 17 inches from the top of the tubesheet on a one-time basis. Additionally, the NRC issued Amendment Number 178 (Reference 3) which approved an interim alternate repair criteria that requires full-length inspection of the tubes within the tubesheet but does not require plugging tubes if any circumferential cracking observed in the region greater than 17 inches from the top of the tubesheet is less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads. Amendment Number 178 is only applicable for Refueling Outage 16 and the subsequent operating cycle. This permanent request for amendment would replace the existing interim alternate repair criteria.

This permanent change is supported by Westinghouse Electric Company LLC, WCAP-17071-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," April 2009 (Reference 4) which recommends the 95% probability/50% confidence H* value of 11.'2 inches. WCNOC has chosen to use an H*

value of 13.1 inches for additional conservatism Approval of this amendment application is requested to support Refueling Outage 17 (Fall 2009) and subsequent eddy current inspection intervals as the existing interim alternate repair criteria amendment expires at the end of the current operating cycle.

2. DETAILED DESCRIPTION Changes to TS 5.5.9 are as follows. Deleted text is struckthrough and added text is italicized.

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repair criteria shall be applied as an alternative to the 40% depth-based criteria:

tubes with flaws having a circumferential component less than Or equal to 203 degrees found in the portion of the tube below 17-inches from the top of the tubeshect and above 1 inch from the

Attachment I to ET 09-0016 Page 3 of 17 bottom of the tubesheet do not require Plugging. Tubes with flaws having a circum~ferential component greater than 203 degrees-Io Ino in TriPoGnnn Of !iR tube belowlI: innones roM !nc top OT the tubesheet and above 1 inch fromn the bottom of the tubesheet shall be removed from serVOce.

Tubes with ser~iie induced flaws located within the region fromn the top of the tubesheet to 17 inches below the top of the tubesheet shall be Frnemved from ser~ice. Tubes with seFrce-induced axial cracks found in the portion of the tube below 17-inches from the top of the tubesheet do not require plugging.

When more than one flaw With circumferential componentS i6 found in the portion of the tube below 17 inches from the top of -- L LIIe tuUbeSdIet ajId aboveJ -IiRGH +FOm tmeDOTTO Of t1 nc uocsheet with the total of the cir.umferent*al o*mponents greater than 203 degrees and an axial separation distance of less than 1 inch, then-the tube shall be remoeyd_ fro-m ssewice. When the circumferential components of each of the flaws are added, it is acceptable4to count the overlapped portionS only once in the total of circumferential components.

When one or mo-re flawys With circumferential components are-found in the portion of the tube within 1 in-hfrom the bottom of the tubesheet, and the tetal of the c-rcumferential cOmponents found in the tube exceeds 94 degrees, then the tube shall be removed ftrom serv*ce. When onMe Or mor~e flaws with circumnferential components are found in the por-tion of the tub within 1 inch from the bottom of the tubesheet and within 1 inch-axial separation distance of a flaw above 1 inch from the bottom of the tubesheet, and the total of the circumferential com~ponents fo-und- in the tube exceeds 914 deg-crees,6 then the tube shall be removed fFom ser.v'ie. When the ciFG~rcumtcn;ai components oT CcIn " M CIVVO C21-" C2 ~a ac~eptable to count the 0

overlapped portions Only once in the total Of GFircumferential Gern penent&-Tubes with service-induced flaws located greater than 13.1 inches below the top of the tubesheet do not require plugging. Tubes with service-inducedflaws located in the portion of the tube from the top of the tubesheet to 13.1 inches below the top of the tubesheet shall be plugged upon detection.

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 13.1 inches below the top of the tubesheet on the hot leg side to 13.1 inches below the top of the tubesheet on the cold leg side, the tube to tubesheet weld at the tube inlet to the tube to t"ub'h,*eht;*weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is

Attachment I to ET 09-0016 Page 4 of 17 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 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 from 13.1 inches below the top of the tubesheet on the hot leg side to 13.1 inches below the top of the tubesheet on the cold leg side, 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.

Changes to TS 5.6.10 are as follows:

Steam Generator 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 the 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;

Attachment I to ET 09-0016 Page 5 of 17

e. Number of tubes plugged during the inspection outage for each active degradation mechanism;

f. Total number and percentage of tubes plugged to date; and

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

h. Following completion of an inspection pe~fermed in Refueling Outage 16 (and any inspections pe~fermed in the subsequent operating cycle),

the number of indicatins and lOcation, size, orientai*o*, whethe iniiated on primar o e i for each ser indued Ice withi the thicvknesqs of the tubesheet, and the total of the circumferential components and any circumnferential overlap below 17 inc~hes from the top of the tubesheet as dcteFrnined in accordance with Fo=llowing completion of an inspection pe~fermed in Refueling Outage 16 (and any inspections pe~femed in the subsequent operating cycle),

the prmarFy toecondary a IE-K- rs ,'Fate in'bserd eaeh SG (if it is not practical to assign leakage to an individual SG, the entire primary to secondary LEAKAGE should be eonservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of theepe and

j. Following completion of an inspection pe~.rmed in Refueling Outage 16 (and any inspectonRS pe~formed in the subsequent operating cycle),

the calculated accident leakage rate from the peotien of the tube below 17 inches from the top of the tujaesneez ror the mosr 1miting -aGGideRt4R f

• IB e the moest limiting SG.

h. The primary to secondary LEAKAGE rate observed in each SG (if it is not practicalto assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle precedingthe inspection which is the subject of the report,-

The calculated accidentinduced leakage rate from the portion of the tubes below 13.1 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.03 times the maximum operationalprimary to secondaryleak rate, the report should describe how it was determined; and The results of monitoring for the tube axial displacement (slippage). If slippage is discovered, the implications of discovery and corrective action shall be provided.

Attachment I to ET 09-0016 Page 6 of 17

3. TECHNICAL EVALUATION 3.1 Background WCGS is a four loop Westinghouse designed plant with Model F steam generators having 5626 tubes in each steam generator. A total of 233 tubes are currently plugged in all four steam generators. The design of the steam generator includes Alloy 600 thermally treated tubing, full depth hydraulically expanded tubesheet joints, and stainless steel tube support plates with broached hole quatrefoils.

The steam generator inspection scope is governed by TS 5.5.9, "Steam Generator (SG)

Program;" Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines,"

(Reference 5); EPRI 1013706, "Pressurized Water Reactor Steam Generator Examination Guidelines," (Reference 6); EPRI 1012987; "Steam Generator. Integrity Assessment Guidelines," (Reference 7); WCGS procedure AP 29A-003, "Steam Generator Management;"

and the results of the degradation assessments required by the Steam Generator Program.

Criterion IX, "Control of Special Processes'" of 10 CFR Part 50, Appendix B, requires in part that nondestructive testing be accomplished by qualified personnel using qualified procedures in accordance with the applicable criteria. The inspection techniques and equipment are capable of reliably detecting the known and potential specific degradation mechanisms applicable to WCGS. The inspection techniques, essential variables and equipment are qualified to Appendix H, "Performance Demonstration for Eddy Current Examination" of the EPRI Steam Generator Examination Guidelines.

Catawba Nuclear Station, Unit 2, (Catawba) reported indication of cracking following nondestructive eddy current examination of the steam generator tubes during their fall 2004 outage. NRC Information Notice (IN) 2005-09, "Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-to-Tubesheet Welds," (Reference 8), provided- industry notification of the Catawba issue. IN 2005-09 noted that Catawba reported crack like indications in the tubes approximately seven inches below the top of the hot leg tubesheet in one tube, and just above the tube-to-tubesheet welds in a region of the tube known as the tack expansion in several other tubes. Indications were also reported in the tube-end welds, also known as tube-to-tubesheet welds, which join the tube to the tubesheet.

WCNOC policies and programs, as well as TS 5.5.9, require the use of applicable industry operating experience in the operation and maintenance of WCGS. The recent experience at Catawba, as noted in IN 2005-09, shows the importance of monitoring all tube locations (such as bulges, dents, dings, and other anomalies from the manufacture of the steam generators) with techniques capable of finding potential forms of degradation that may be occurring at these locations (as discussed in Generic Letter 2004-001, "Requirements for Steam Generator Tube Inspections"). Since the WCGS Westinghouse Model F steam generators were fabricated with Alloy 600 thermally treated tubes similar to the Catawba Unit 2 Westinghouse Model D5 steam generators, a potential exists for WCGS to identify tube indications similar to those reported at Catawba within the hot leg tubesheet region if similar inspections are performed during the fall 2009 refueling outage.

Potential inspection plans for the tubes and tube welds underwent intensive industry discussions in March 2005. The findings in the Catawba steam generator tubes present two distinct issues with regard to the steam generator tubes at WCGS:

Attachment I to ET 09-0016 Page 7 of 17

1) Indications in internal bulges and overexpansions within the hot leg tubesheet; and

2) Indications at the elevation of the tack expansion transition.

Prior to each steam generator tube inspection, a degradation assessment, which includes a review of operating experience, is performed to identify degradation mechanisms that have a potential to be present in the WCGS steam generators. A validation assessment is also performed to verify that the eddy current techniques utilized are capable of detecting those flaw types that are identified in the degradation assessment. Based on operating experience discussed above, WCGS revised the steam generator inspection plan to include sampling of bulges and overexpansions within the tubesheet region on the hot leg side in Refueling Outage 14 (Spring 2005) and Refueling Outage 15 (Fall 2006). The sample is based on the guidance contained in EPRI 1013706, "Pressurized Water Reactor Steam Generator Examination Guidelines," Revision 7, and TS 5.5.9, "Steam Generator (SG) Program." The inspection plan is expanded according to EPRI steam generator examination guidelines if necessary due to confirmed degradation in the region required to be examined (i.e. a tube crack). Degradation was not detected in the tubesheet region in Refueling Outage 14 and Refueling Outage 15 (Fall 2006).

For Refueling Outage 16 (Spring 2008) an Interim Alternate Repair Criteria (IARC) was approved which revised TS 5.5.9. The IARC required an inspection to the full depth of the hot leg tubesheet, but also allowed axial cracks and circumferential cracks less than a specified extent to remain in service at the tube ends. Indications were identified in the tube ends of steam generators B and C which required tube end inspection scope expansions into steam generators A and D. Indications were observed at the hot leg tube ends in steam generators B, C, and D; none were observed in steam generator A. All indications were within about 0.2 inch from the tube end. Indications with circumferential extent greater than 940 were plugged. All axial and circumferentially oriented indications 940 or less in circumferential extent were left in service consistent with the criteria provided in the IARC. Axial indications and indications with circumferential extent of up to, and including, 940 do not challenge the structural and leakage integrity requirements of NEI 97-06.

As a result of these potential issues and to prevent the unnecessarily plugging of additional tubes in the WCGS steam generators, WCNOC is proposing changes to TS 5.5.9 to limit the steam generator tube inspection and repair (plugging) to the portion of tube, from 13.1 inches below the top of the tubesheet on the hot leg side to 13.1 inches below the top of the tubesheet on the coldleg side. In addition, this amendment request proposes to revise TS 5.6.10 to provide reporting requirements specific to the permanent alternate repair criteria.

3.2 Evaluation To preclude unnecessarily plugging tubes in the WCGS steam generators tube inspections will be limited to identifying and plugging degradation in the portion of the tubes within the tubesheet necessary to maintain structural and leakage integrity in both normal and accident conditions. The technical evaluation for the inspection and repair methodology is provided in WCAP-17071-P. The evaluation is based on the use of finite element model structural analysis and a bounding leak rate evaluation based on contact pressure between the tube and the tubesheet during normal and postulated accident conditions. The limited tubesheet inspection

Attachment I to ET 09-0016 Page 8 of 17 criteria were developed for the tubesheet region of the WCGS Model F steam generator considering the most stringent loads associated with plant operation, including transients and postulated accident conditions. The limited tubesheet inspection criteria were selected to prevent tube pullout from the tubesheet due to axial end cap loads acting on the tube and to ensure that the accident induced leakage limits are not exceeded. WCAP-17071-P provides technical justification for limiting the inspection in the tubesheet expansion region to less than the full depth of the tubesheet.

The basis for determining the portion of the tube which requires eddy current inspection within the tubesheet is based upon evaluation and testing programs that quantified the tube-to-tubesheet radial contact pressure for bounding plant conditions as described in WCAP-17071-P. The tube-to-tubesheet radial contact pressure provides resistance to tube pullout.

Primary to secondary leakage from tube degradation in the tubesheet area is accounted for in several design basis accidents: feedwater line break (FLB), steam line break (SLB), locked rotor, and control rod ejection. The radiological dose consequences associated with this assumed leakage are evaluated to ensure that they remain within regulatory limits (e.g. 10 CFR Part 100, 10 CFR 50.67, GDC 19). The accident induced leakage performance criteria are intended to ensure the primary to secondary leak rate during any accident does not exceed the primary to secondary leak rate assumed in the accident analysis. The limiting leakage ratio of 2.03 is independent of the H* distance defined in WCAP-17071-P.

The constraint that is provided by the tubesheet precludes tube burst for cracks within the tubesheet. The criteria for tube burst described in NEI 97-06 and NRC Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," (Reference 9) are satisfied due to the constraint provided by the tubesheet. Through application of the limited tubesheet inspection scope as described below, the existing operating leakage limit provides assurance that excessive leakage (i.e., greater than accident analysis assumptions) will not occur. The accident induced leak rate limit for WCGS is 1.0 gpm. The TS 3.4.13, "RCS Operational LEAKAGE," operational leak rate limit is 150 gpd (0.1 gpm) through any one steam generator. Consequently, accident leakage is approximately 10 times the allowable leakage, if only one steam generator is leaking. Using a SLB/FLB overall leakage factor of 2.03, accident induced leakage is approximately 0.5 gpm, if all 4 steam generators are leaking at 150 gpd at the beginning of the accident. Therefore, significant margin exists between the conservatively estimated accident induced leakage and the allowable accident leakage (1.0 gpm).

Plant-specific operating conditions are used to generate the overall leakage factor ratios that are to be used in the condition monitoring and operational assessments. The plant-specific data provide the initial conditions for application of the transient input data. The results of the analysis of the plant-specific inputs to determine the bounding plant for each model of steam generator and to assure that the design basis accident contact pressures are greater than the normal operating pressure contact pressure are contained in Section 6 of Reference 4.

The leak rate ratio (accident induced leak rate to operational leak rate) is directly proportional to the change in differential pressure and inversely proportional to the dynamic viscosity. Since dynamic viscosity decreases with an increase in temperature, an increase in temperature results in an increase in leak rate. However, for both the postulated SLB and FLB events, a plant cool down event would occur and the subsequent temperatures in the reactor coolant

Attachment I to ET 09-0016 Page 9 of 17 system (RCS) would not be expected to exceed the temperatures at plant no load conditions.

Thus, an increase in leakage would not be expected to occur as a result of the temperature change. The increase in leakage would only be a function of the increase in primary to secondary pressure differential. The resulting leak rate ratio for the SLB and FLB events is 2.03 for WCGS, which is the bounding value for all steam generator designs.

The other design basis accidents, such as the postulated locked rotor event and the control rod ejection event, are conservatively modeled using the design specification transients to result in increased temperatures in the steam generator hot and cold legs for a period of time. As previously noted, dynamic viscosity decreases with increasing temperature. Therefore, leakage would be expected to increase due to decreasing viscosity and increasing differential pressure for the duration of time that there is a rise in RCS temperature. For transients other than a SLB and a FLB, the length of time that a plant with Model F steam generators will exceed the normal operating differential pressure across the tubesheet is less than 30 seconds. As the accident induced leakage performance criteria is defined in gallons per minute, the leak rate for a locked rotor ejection event can be integrated over a minute to compare to the limit. Time integration permits an increase in acceptable leakage during the time of peak pressure differential by approximately a factor of two because of the short duration (less than 30 seconds) of the elevated pressure differential. This translates into an effective reduction in leakage factor by the same factor of two for the locked rotor event. Therefore, for the locked rotor event, the leakage factor of 1.77 (Table 9-7, Reference 4) for WCGS is adjusted downward to a factor of, 0.89. Similarly, for the control rod ejection event, the duration of the elevated pressure differential is less than 10 seconds. Thus, the peak leakage factor may be reduced by a factor of six from 2.65 to 0.44. Due to the short duration of the transients above normal operating pressure differential, no leakage factor is required for the locked rotor and control rod ejection events (i.e., the leakage factor is under 1.0 for both transients).

The plant transient response following a full power double-ended main feedwater line rupture corresponding to "best estimate" initial conditions and operating characteristics as generally presented in the USAR Chapter 15.0 safety analysis, indicates that the transient for a Model F steam generator exhibits a cooldown characteristic instead of a heatup transient. The use of either the component design specification transient or the Chapter 15.0 safety transient for leakage analysis for FLB is overly conservative because:

The assumptions on which the FLB design transient is based are specifically intended to establish a conservative structural (fatigue) design basis for RCS components; however, H* does not involve component structural and fatigue issues. The best estimate transient is considered more appropriate for use in the H* leakage calculations.

For the Model F steam generator, the FLB transient curve (Figure 9-5, Reference 4) represents a double-ended rupture of the main feedwater line concurrent with both station blackout (loss of main feedwater and reactor coolant pump coast down) and turbine trip.

The assumptions on which the FLB safety analysis is based are specifically intended to establish a conservative basis for minimum auxiliary feedwater (AFW) capacity requirements and combines worst case assumptions which are exceptionally more severe when the FLB occurs inside containment. For example, environmental errors that are applied to reactor trip and engineered safety feature actuation would no longer

Attachment I to ET 09-0016 Page 10 of 17 be applicable. This would result in much earlier reactor trip and greatly increase the steam generator liquid mass available to provide cooling to the RCS.

A SLB event would have similarities to a FLB except that the break flow path would include the secondary separators, which could only result in an increased initial cooldown (because of retained liquid inventory available for cooling) when compared to the FLB transient. A SLB could not result in more limiting temperature conditions than a FLB.

In accordance with plant operating procedures, the operator would take action following a high energy secondary line break to stabilize the RCS conditions. The expectation for a SLB or FLB with credited operator action is to stop the system cooldown through isolation of the faulted steam generator and control of temperature by the AFW System. Steam pressure control would be established by either the steam generator safety valves or control system (atmospheric relief valves). For any of the steam pressure control operations, the maximum temperature would be approximately the no load temperature and would be well below normal operating temperature.

Since the best estimate FLB transient temperature would not be expected to exceed the normal operating temperature, the viscosity ratio for the FLB transient is set to 1.0.

The leakage factor of 2.03 for WCGS, for a postulated SLB/FLB, has been calculated as shown in Table 9-7 of Reference 4. The leakage factor of 2.03 is a bounding value for all steam generators, both hot and cold legs, in Table 9-7 of Reference 4. Specifically, for the condition monitoring (CM) assessment, the component of leakage from the prior cycle from below the H*

distance will be multiplied by a factor of 2.03 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the operational assessment (OA), the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.03 and compared to the observed operational leakage.

Reference 4 redefines the primary pressure boundary. The tube-to-tubesheet weld no longer functions as a portion of this boundary. The hydraulic expansion of the tube into the tubesheet over the H* distance now functions as the primary pressure boundary in the area of the tube and tubesheet, maintaining the structural and leakage integrity over the full range of steam generator operating conditions, including the most limiting accident conditions. The evaluation in Reference 4 determined that degradation in tubing below 11.2 inches from the top of the tubesheet does not require inspection or repair (plugging). The inspection of the portion of the tubes above 11.2 inches from the top of the tubesheet for tubes that have been hydraulically expanded in the tubesheet provides a high level of confidence that the structural and leakage performance criteria are maintained during normal operating and accident conditions.

Reference 4 (Section 8.0) recommended a final value of H* of 11.2 inches below the top of the tubesheet for the entire bundle of tubes. However, WCNOC has chosen to use a more conservative value of 13.1 inches. This more conservative value was discussed between the NRC staff and industry representatives on April 24, 2009 and May 1, 2009.

Attachment I to ET 09-0016 Page 11 of 17 Reference 4, Section 9.8, provides a review of leak rate susceptibility to tube slippage and concluded that the tubes are fully restrained against motion under very conservative design and analysis assumptions such that tube slippage is not a credible event for any tube in the bundle.

However, in response to a NRC staff request, WCNOC commits to monitor for tube slippage as part of the steam generator tube inspection program. A proposed change to TS 5.6.10 adds a new reporting requirement for slippage monitoring. If no tube slippage is identified as a result of the monitoring, the Steam Generator Tube Report would indicate no tube slippage was detected.

In addition the NRC staff has requested that licensees determine if there are any significant deviations in the location of the bottom of the expansion transition (BET) relative to the top of tubesheet that would invalidate assumptions in Reference 4. Therefore, WCNOC commits to perform a one-time verification of tube expansion locations to determine if any significant deviations exist from the top of tubesheet to the BET. If any significant deviations are found, the condition will be entered into the plants corrective action program.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria General Design Criteria (GDC) 1, 2, 4, 14, 30, 31, and 32 of 10 CFR 50, Appendix A, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity.

GDC 19 of 10 CFR 50, Appendix A, defines requirements for the control room and for the radiation protection of the operators working within it. Accidents involving the leakage or burst of steam generator tubing comprise a challenge to the habitability of the control room.

10 CFR 50, Appendix B, establishes quality assurance requirements for the design, construction, and operation of safety related components. The pertinent requirements of this appendix apply to all activities affecting the safety related functions of these components.

These requirements are described in Criteria IX, XI, and XVI of Appendix B and include control of special processes, inspection, testing, and corrective action.

10 CFR 100, Reactor Site Criteria, established reactor siting criteria, with respect to the risk of public exposure to the release of radioactive fission products. Accidents involving leakage or tube burst of steam generator tubing may comprise a challenge to containment and therefore involve an increased risk of radioactive release.

Under 10 CFR 50.65, the Maintenance Rule, licensees classify steam generators as risk-significant components because they are relied upon to remain functional during and after design basis events. Steam generators are to be monitored under 10 CFR 50.65(a)(2) against industry established performance criteria. Meeting the performance criteria of Nuclear Energy Institute (NEI) 97-06, Revision 2, "Steam Generator Program Guidelines," provides reasonable assurance that the steam generator tubing remains capable of fulfilling its specific safety function of maintaining the reactor coolant pressure boundary. The NEI 97-06, Revision 2, steam generator performance criteria are:

Attachment I to ET 09-0016 Page 12 of 17

• All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and-a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial loads.

• The primary to secondary accident induced leakage rate for any design basis accident, other than a steam generator tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all steam generators and leakage rate for an individual steam generator. Leakage is not to exceed 1 gpm per steam generator, except for specific types of degradation at specific locations when implementing alternate repair criteria as documented in the Steam Generator Program technical specifications.

• The Reactor Coolant System (RCS) operational primary to secondary leakage through any one steam generator shall be limited to 150 gallons per day.

The proposed change defines the portion of the tube as the length of tube that is engaged in the tubesheet from the secondary face that is required to maintain structural and leakage integrity over the full range of steam generator operating conditions, including the most limiting accident conditions. The evaluation in Enclosure I (WCAP-17071-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)") determined that degradation in tubing below 11.2 inches from the top of the tubesheet does not require plugging and serves as the bases for the steam generator tubesheet inspection program. As such, the WCGS inspection program provides a high level of confidence that the structural and leakage criteria are maintained during normal operating and accident conditions.

4.2 Significant Hazards Consideration This amendment application proposes to revise Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," to exclude portions of the tubes within the tubesheet from periodic steam generator inspections. In addition, the amendment application proposes to revise TS 5.6.10, "Steam Generator Tube Inspection Report," to provide reporting requirements specific to the permanent alternate repair criteria. Application of the structural analysis and leak rate evaluation results, to exclude portions of the tubes from inspection and repair is interpreted to constitute a redefinition of the primary to secondary pressure boundary.

The proposed change defines the portion of the tube that must be inspected and repaired. A justification has been developed by Westinghouse Electric Company, LLC to identify the specific inspection depth below which any type of axial or circumferential primary water stress

Attachment I to ET 09-0016 Page 13 of 17 corrosion cracking can be shown to have no impact on NEI 97-06, Revision 2, performance criteria.

WCAP-17071-P, Section 8.0, recommended a final value of H* of 11.2 inches from the top of the tubesheet for the entire bundle of tubes. However, WCNOC has chosen to use a more conservative value of 13.1 inches. This more conservative value was discussed between the NRC staff and industry representatives on April 24, 2009 and May 1, 2009.

WCNOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 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 previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The proposed change that alters the steam generator inspection criteria does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident.

Of the applicable accidents previously evaluated, the limiting transients with consideration to the proposed change to the steam generator tube inspection and repair criteria are the steam generator tube rupture (SGTR) event and the feedline break (FLB) postulated accidents.

During the SGTR event, the required structural integrity margins of the steam generator tubes and the tube-to-tubesheet joint over the H* distance will be maintained. Tube rupture in tubes with cracks within the tubesheet is precluded by the presence of the tubesheet and constraint provided by the tube-to-tubesheet joint. Tube burst cannot occur within the thickness of the tubesheet. The tube-to-tubesheet joint constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tubesheet, from the differential pressure between the primary and secondary side, and tubesheet deflection. Based on this design, the structural margins against burst, as discussed in Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," and TS 5.5.9 are maintained for both normal and postulated accident conditions.

The proposed change has no impact on the structural or leakage integrity of the portion of the tube outside of the tubesheet. The proposed change maintains structural and leakage integrity of the steam generator tubes consistent with the performance criteria in TS 5.5.9.

Therefore, the proposed change results in no significant increase in the probability of the occurrence of a SGTR accident.

Attachment I to ET 09-0016 Page 14 of 17 At normal operating pressures, leakage from tube degradation below the proposed limited inspection depth is limited by the tube-to-tubesheet joint. Consequently, negligible normal operating leakage is expected from degradation below the inspected depth within the tubesheet region. The consequences of an SGTR event are not affected by the primary to secondary leakage flow during the event as primary to secondary leakage flow through a postulated tube that has been pulled out of the tubesheet is essentially equivalent to a severed tube. Therefore, the proposed changes do not result in a significant increase in the consequences of a SGTR.

The probability of a SLB is unaffected by the potential failure of a steam generator tube as the failure of the tube is not an initiator for a SLB event.

The leakage factor of 2.03 for WCGS, for a postulated SLB/FLB, has been calculated as shown in Table 9-7 of WCAP-1 7071-P and will be applied to the normal operating leakage associated with the tubesheet expansion region in the condition monitoring (CM) and operational assessment (OA). The leakage factor of 2.03 is a bounding value for all steam generators, both hot and cold legs, in Table 9-7 of Reference 2. Through application of the limited tubesheet inspection scope, the existing operating leakage limit provides assurance that excessive leakage (i.e., greater than accident analysis assumptions) will not occur. The accident induced leak rate limit for WCGS is 1.0 gpm.

The TS 3.4.13, "RCS Operational LEAKAGE," operational leak rate limit is 150 gpd (0.1 gpm) through any one steam generator. Consequently, accident leakage is approximately 10 times the allowable leakage, if only one steam generator is leaking. Using a SLB/FLB overall leakage factor of 2.03, accident induced leakage is approximately 0.5 gpm, if all 4 steam generators are leaking at 150 gpd at the beginning of the accident. Therefore, significant margin exists between the conservatively estimated accident induced leakage and the allowable accident leakage (1.0 gpm).

No leakage factor will be applied to the locked rotor or control rod ejection transients due to their short duration.

For the CM assessment, the component of leakage from the prior cycle from below the H*

distance will be multiplied by a factor of 2.03 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the OA, the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.03 and compared to the observed operational leakage.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Attachment I to ET 09-0016 Page 15 of 17 (2) Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change that alters the steam generator inspection and reporting criteria does not introduce any new equipment, create new failure modes for existing equiphrent, or create any new limiting single failures. Plant operation will not be altered, and safety functions will continue to perform as previously assumed in accident analyses.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Does the change involve a significant reduction in a margin of safety?

Response: No The proposed change defines the portion of the tube that must be inspected and repaired.

WCAP-17071-P identifies the specific inspection depth below which any type tube degradation shown to have no impact on the performance criteria in NEI 97-06, Revision 2.

The proposed change that alters the steam generator inspection and reporting criteria maintains the required structural margins of the steam generator tubes for both normal and accident conditions. NEI 97-06, Revision 2, and RG 1.121, are used as the bases in the development of the limited tubesheet inspection depth methodology for determining that steam generator tube integrity considerations are maintained within acceptable limits.

RG 1.121 describes a method acceptable to the NRC for meeting GDC 14, "Reactor Coolant Pressure Boundary," GDC 15, "Reactor Coolant System Design," GDC 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," and GDC 32, "Inspection of Reactor Coolant Pressure Boundary," by reducing the probability and consequences of a SGTR. RG 1.121 concludes that by determining the limiting safe conditions for tube wall degradation the probability and consequences of a SGTR are reduced. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the American Society of Mechanical Engineers (ASME) Code.

For axially-oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially-oriented cracking, WCAP-17071-P, defines a length of degradation-free expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces, with applicable safety factors applied. Application of the limited hot and cold leg tubesheet inspection criteria will preclude unacceptable primary to secondary leakage during all plant conditions. Using the methodology for determining leakage as described in WCAP-17071-P, it is shown that significant margin exists between conservatively estimated accident induced leakage and the allowable accident leakage (1.0 gpm) if all four steam generators are assumed to be leaking at the TS leakage limit at the beginning of the design basis accident.

Therefore, the proposed change does not involve a significant reduction in any margin of safety.

Attachment I to ET 09-0016 Page 16 of 17 4.3 Conclusion The hydraulically expanded portion of the tube from the top of the tubesheet to 11.2 inches below the top of the tubesheet is the length of tube that is engaged within the tubesheet (secondary face) that is required to maintain structural and leakage integrity over the full range of steam generating operating conditions, including the most limiting accident conditions.

WCAP-17071-P determined that tube degradation below 11.2 inches below the top of the tubesheet does not require plugging and serves as the basis for the limited tubesheet inspection criteria. WCAP-17071-P also shows that, upon implementation of the H* criterion, that the TS 3.4.13, "RCS Operational LEAKAGE," operational leak rate limit of 150 gpd precludes unacceptable leakage during any postulated accident that models primary to secondary leakage.

Based on the considerations above, 1) there is a 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. ENVIRONMENTAL CONSIDERATION WCNOC has evaluated the proposed amendment for environmental considerations. The 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, and 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, the proposed amendments meet the eligibility criterion for categorical exclusion set for in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6. REFERENCES

1. Letter dated April 28, 2005, from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station Issuance of Exigent Amendment RE: Steam Generator (SG) Tube Surveillance Program (TAC NO. MC6757)."

2. Letter dated October 10, 2006, from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station Issuance of Amendment RE: Steam Generator Tube Inspections Within the Tubesheet (TAC NO. MD 2467).

3. Letter dated April 4, 2008, from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station - Issuance of Amendment RE: Revision to Technical Specification 5.5.9 on the Steam Generator Program (TAC NO. MD8054).

Attachment I to ET 09-0016 Page 17 of 17

4. Westinghouse Electric Company LLC, WCAP-17071-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," April 2009.

5. NEI 97-06, Rev. 2, "Steam Generator Program Guidelines," May 2005.

6. EPRI 1013706, "Pressurized Water Reactor Steam Generator Examination Guidelines."

7. EPRI 1012987; "Steam Generator Integrity Assessment Guidelines," July 2006.

8. NRC Information Notice 2005-09, "Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-to-Tubesheet Welds," April 7, 2005.

9. NRC Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," August 1976.

Attachment II to ET 09-0016 Page 1 of 5 ATTACHMENT II Markup of Technical Specification pages

Attachment II to ET 09-0016 Page 2 of 5 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

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

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repair criteria shall be applied as an alternative to the 40% depth-based criteria:

1. For fuelin utage 1 and the ubsequent erating cycl

-- *..t_ ,av , -s uircu wit s a frential co ponent less tlan or

-Tt6e~AQti sevtcje-na1AC-fA ual to 3 degr s,found the portion f the tube b ow 17 4

• • i=Us 16*,cAte S*. 4v--teiv inche rom thetp of the besheet a above 1 in from the 13.1 c6s +Our-*e.

of bott of the beshee o not requi plugging. bes with flaw

-+'e.+Aloasi-i & v.ot* reiqvLr h ing a cir umferen I compone greater tha 203 degrees rTues Let% Serv*cet- ound in e portio of the tube low 17 inch s from the to f the nc,,.c._eA 4 1.*&u,.s;i / tubes et and a ve 1 inch f m the botto of the tubes et shall 14 o ?ov,~~i pe. &F+c-6 co be r oved f m service.

+hr-jep O -14Af,~lcct +0 3.

I fn 6cA.. ba*.,, .* r-- +bF ubes w service-in ced flaws o ated within t region from

&f- tL%6.r ,*..- s.ll. 1, the to f the tube e to 17 inc es below the p of the ui~e/1tt!," tub eet shall b removed fr service. T_tufe4 es with service-in uced axial c cks found i the portion ofhe tube below 1

.nches from e top of the besheet do ot require pluggig.

When re than on aw with circ ferential comp ents is found the portio of the tube ow 7inches the top of th tub sheet and ove 1 inch fr m the bottom of e tubesheet w t e total of th circuferen i components ater than 203 degrees a an axial se ration distance less than 1 in , then the tube all be rem ed from service hen the circ ferential comp ents of eac of the flaws are ded, it is acc table to co the over~la ed portions onl nce in the tot of ci cumferentiaomponents.

When on or more flaws circumferen,* I componen are found i the portion oft tube within 1 ch from the ottom oft tube eet, and the t I of the circu erential com nents fou in e tube excee 94 degrees, t n the tube s all be rem ed rom service. en one or mo flaws with cumferent/

components e found in th ortion of th ube within inch from the botto f the tubeshe and within inch axial paration Saf flaw abov 1 inch from e bo(om theotubesinuet, (continued)

Wolf Creek - Unit 1 5.0-12 Amendment No. 123, 153, 172, 178

Attachment II to ET 09-0016 Page 3 of 5 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) and th otal of the rcumferen/ compon ts found*'the tube ex ds 94 degrres; then tho be shal e remove from servie.

en the circ ferential imponent of each of e flaws ar added, it isscceptable count th overlappe portions o once in the tot of circumf ential co ponents.

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 131 in&v, I,, 4 4V* ,0t any type (e.g., volumetric flaws, axial and circumferential cracks) that may hrules'ý"ert 6, t S- be resent along the length of the tube, from t*"te-tu'J ,tee4_e

.k6 13. k at tu inlo o tu-to-

- e e dat. outee1,t ýd be-..owi", +oC, +-% may satisfy the applicable tube repair criteria. The tube-to-tubesheet

- *[ l 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 6,e ) ,. 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 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 tubev6lhn the next o'f -One_ + icdt Om inspection for each SG for the degradation mechanism that Se. 5 to, 13.1 Olc.Vie.

'61e caused the crack indication shall not exceed 24 effective full bebW -oFl _ WAe. *Le~eC OK-'E power months or one refueling outage (whichever is less). If ay, -Fne coldefinitive 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.

(continued)

Wolf Creek - Unit 1 5.0-13 Amendment No. 23**163* -172, 178

Attachment II to ET 09-0016 Page 4 of 5 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator 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 the 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 during the inspection outage for each active degradation mechanism;

f. Total number and percentage of tubes plugged to date;g

g. The results of condition monitoring, including the results of tube pulls and in-situ testin*c -_C

h. Fol ing co etion an ins pction p ormed in efueling 0 ge (an (and annspe,,otnsape7m annspe n pe rmed in e subseqent operati cycle),

the num r of I cations nd locati , size, or ntation, wh er initiat on prI ary or secondary ide for ea service-ind ced flaw wifn the t kness the tube eet,and ebtotalof th umfererial mpone s and icumf p below 17fnches fromhe top of e tubes et as det mined n cordance wi TS 5.5.9 F owing c pletion an inspeci n performe n Refuelin Outage 6 (and y inspecti s perfor in the sub quent ope ting cycle),

the pri ry to sec dary LEA GE rate ob rved in e SG (if it is/

not actical to sign leak to an indivi al SG, th entire prima qto s onda L KAGE sh Id be conse tively ass ed to be fron he cyc preceding t i tio the bject of there rt; and

j. Fol 1,adan ing compl n ion of an ins mdint cinsp ction pe med in Refue ating bsqen ng Outage cycl, calula acietla ert rte portion to the tube b eow S17"ice m h o o uern o temsiiting acc'* ent in (

" the mo imiting SG,. / ____ .

Wolf Creek - Unit 1 5.0-28 Amendment No. 123, 142, 158 ,161 4-78, 179

Attachment II to ET 09-0016 Page 5 of 5 INSERT 5.0-28

h. The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report; The calculated accident induced leakage rate from the portion of the tubes below 13.1 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.03 times the maximum operational primary to secondary leak rate, the report should describe how it was determined; and

j. The results of monitoring for the tube axial displacement (slippage). If slippage is discovered, the implications of discovery and corrective action shall be provided.

Attachment III to ET 09-0016 Page 1 of 3 j

ATTACHMENT III Markup of Technical Specification Bases pages (for information only)

Attachment III to ET 09-0016 Page 2 of 3 SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of an SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for an SGTR assumes the contaminated secondary fluid is released to the atmosphere via SG atmospheric relief valves and safety valves.

The analysis for design basis accidents and transients other than an SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 1 gallon per minute or is assumed to increase to 1 gallon per minute as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet F efuel' g OSutag6 and th subseq nt operatig cycl, an interim ternat repair crirrion for t portion the tub elow inches fro) r tetet ified TS 5.5. .1. (R . 7) The tube-to-u sheet w_ w iss not o sidered art of the be. /

Wolf Creek - Unit B 3.4.17-2 .RRevision 37 A pwr*wv'neyit alternxt*~r_ mAsY ¶ appeo'kveA"~c trley;d

. C.$

t-t4 c.fAL -jvwcs6iv P 1nuiw. .The. Wfdrtchsc

-t% Mb5iovz -th 'toAe-0ý ovl.(W~

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pr~tMp Y"4.q.Lt LiL"&A~ 6Ze.-A a.-u~ jA Wjprvssam~%

,61-v' .vo

-Ml riml of S& &peeoAv;%j si~cMAt;% 1h. *%os* tm4

Attachment III to ET 09-0016 Page 3 of 3 SG Tube Integrity B 3.4.17 BASES REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10 CFR 50 Appendix A, GDC 19.

3. 10 CFR 100.

4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6. EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

7. License Amendment No.

VýM7-ý Wolf Creek - Unit 1 B 3.4.17-7 Revision 37

Attachment IV to ET 09-0016 Page 1 of 4 ATTACHMENT IV Retyped Technical Specification pages

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Proqram A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

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 1 gpm per SG.

(continued)

Wolf Creek - Unit 1 5.0-11 Amendment No. 123,1-5, 164

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

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

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repair criteria shall be applied as an alternative to the 40% depth-based criteria:

1. Tubes with service-induced flaws located greater than 13.1' inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 13.1 inches below the top of the tubesheet shall be plugged upon detection.

(continued)

Wolf Creek - Unit 1 5.0-12 Amendment No. 123, 163, 172, 178,

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 13.1 inches below the top of the the tubesheet on the hot leg side to 13.1 inches below the top of the tubesheet on the cold leg side, and that may satisfy the applicable tube repair criteria. 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 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 from 13.1 inches below the top of the tubesheet on the hot leg side to 13.1 inches below the top of the tubesheet on the cold leg side, 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.

(continued)

Wolf Creek - Unit 1 5.0-13 Amendment No. 123, 153, 172,178,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.11 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, and in accordance with the guidance specified below.

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1% when tested in accordance with Regulatory Guide 1.52, Revision 2 at the system flowrate specified below +/- 10%.

ESF Ventilation System Flowrate Control Room Emergency Ventilation System-Filtration 2000 cfm Control Room Emergency Ventilation System-Pressurization 750 cfm Auxiliary/Fuel Building Emergency Exhaust 6500 cfm (continued)

Wolf Creek - Unit 1 5.0-14 Amendment No. 423,,159, 164  :

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits,"

and

2. Specification 3.4.12, "Low Temperature Overpressure Protection System."

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

1. WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

7 5.6.7 Not Used.

5.6.8 PAM Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.9 Not Used.

(continued)

Wolf Creek - Unit 1 5.0-27 Amendment No. 123, 130, 142, 157, 158-,164,-7-9, 180

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator 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 the 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 during the inspection outage for each active degradation mechanism;

f. Total number and percentage of tubes plugged to date;

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

h. The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report; The calculated accident induced leakage rate from the portion of the tubes below 13.1 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.03 times the maximum operational primary to secondary leak rate, the report should describe how it was determined; and

j. The results of monitoring for the tube axial displacement (slippage). If slippage is discovered, the implications of discovery and corrective action shall be provided.

Wolf Creek - Unit 1 5.0-28 Amendment No. 123, 142,158, 161, 1-7-8, 79,

Attachment V to ET 09-0016 Page 1 of 1 ATTACHMENT V REGULATORY COMMITMENTS The following table identifies those actions committed to by WCNOC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr.

Richard Flannigan at (620) 364-4117.

Regulatory Commitments Due Date / Event Once approved, the amendment will be implemented prior to Prior to MODE 4 entry MODE 4 entry during startup from Refueling Outage 17. Final TS during startup from Bases changes will be implemented pursuant to TS 5.5.14, Refueling Outage 17 "Technical Specification (TS) Bases Control Program," at the time the amendment is implemented.

In response to a NRC staff request, WCNOC commits to monitor Prior to MODE 4 entry for tube slippage as part of the steam generator tube inspection during startup from program. Refueling Outage 17 In addition the NRC staff has requested that licensees determine if Prior to MODE 4 entry there are any significant deviations in the location of the bottom of during startup from the expansion transition (BET) relative to the top of tubesheet that Refueling Outage 17 would invalidate assumptions in Reference 4. Therefore, WCNOC commits to perform a one-time verification of tube expansion locations to determine if any significant deviations exist from the top of tubesheet to the BET. If any significant deviations are found, the condition will be entered into the plants corrective action program.