TMI-12-027, License Amendment Request to Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection

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License Amendment Request to Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection
ML12086A037
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 03/26/2012
From: Jesse M
Exelon Generation Co, Exelon Nuclear
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
TMI-12-027
Download: ML12086A037 (26)
v • d • e


Text

Exelon Nuclear www.exeloncorp.com Exelon.

200 Exelon Way Nuclear Kennett Square, PA 19348 10 CFR 50.90 TMI-12-027 March 26, 2012 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

License Amendment Request to Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection

References:

1. TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
2. Notice of Availability of the "Models for Plant-Specific Adoption of Technical Specifications Task Force Traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,"" dated October 27, 2011.

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the Technical Specifications (TS) for Three Mile Island Nuclear Station, Unit 1 (TMI, Unit 1) to adopt U.S. Nuclear Regulatory Commission (USNRC)-approved Revision 2 to Technical Specifications Task Force (TSTF) Traveler TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (Reference 1). The proposed changes revise Limiting Condition for Operation 3.1.1.2 and Surveillance Requirement 4.19.2, "Steam Generator (SG) Tube Integrity," Specification 6.19, "Steam Generator (SG) Program," and Specification 6.9.6, "Steam Generator Tube Inspection Report," and include TS Bases changes that summarize and clarify the purpose of the TS. The specific changes concern SG inspection periods, and address applicable administrative changes and clarifications.

This change is consistent with the Notice of Availability of TSTF-51 0, Revision 2, (Reference 2) announced in the Federal Register on October 27, 2011 (76 FR 66763) as part of the consolidated line item improvement process (CLIIP).

Attachment 1 of this submittal provides an evaluation of the proposed changes, the requested confirmation of applicability, plant specific verifications, and variations in the proposed TMI, Unit 1 TS changes from the approved TSTF-510 (Reference 1). Attachment 2 provides the existing TS page markups showing the proposed changes. Attachment 3 provides the associated TS Bases markups for information only.

U.S. Nuclear Regulatory Commission March 26, 2012 Page 2 The proposed changes have been reviewed by the TMI, Unit 1 Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed amendment by March 26,2013. Once approved, the amendment shall be implemented within 60 days.

There are no regulatory commitments contained in this letter.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), EGC is notifying the Commonwealth of Pennsylvania of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions conceming this letter, please contact Ms. Wendy E. Croft at (610) 765-5726.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 26th day of March 2012.

Michael D. Je se Director - Lice ng and Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Changes

2. Markup of Technical Specifications Pages
3. Markup of Technical Specifications Bases Pages (For Information Only) cc: USNRC Region I, Regional Administrator USNRC Project Manager, TMI, Unit 1 USNRC Senior Resident Inspector, TMI, Unit 1 Director, Bureau of Radiation Protection, PA Department of Environmental Resources Chairman, Board of County Commissioners, Dauphin County, PA Chairman, Board of Supervisors, Londonderry Township, PA R. R. Janati, Commonwealth of Pennsylvania

ATTACHMENT 1 Evaluation of Proposed Changes Three Mile Island Nuclear Station. Unit 1 Renewed Facility Operating License No. DPR-50

Subject:

Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection Page 1 of 5 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Renewed Facility Operating License DPR-50 for Three Mile Island Nuclear Station, Unit 1 (TMI, Unit 1) to adopt U.S. Nuclear Regulatory Commission (USNRG)-approved Revision 2 to Technical Specifications Task Force (TSTF)

Traveler TSTF-51 0, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (Reference 6.1). The proposed changes revise Limiting Condition for Operation (LCO) 3.1.1.2 and Surveillance Requirement (SR) 4.19.2, "Steam Generator (SG)

Tube Integrity," Specification 6.19, "Steam Generator (SG) Program," and Specification 6.9.6, "Steam Generator Tube Inspection Report," and include TS Bases changes that summarize and clarify the purpose of the TS. The specific changes concern SG inspection periods, and address applicable administrative changes and clarifications.

This change is consistent with the Notice of Availability of TSTF-51 0, Revision 2 (TSTF-51 0),

(Reference 6.2) announced in the Federal Register on October 27, 2011 (76 FR 66763) as part of the consolidated line item improvement process (CLlIP).

TMI, Unit 1 installed replacement SGs during the fall 2009 refueling outage (T1 R18). The TMI, Unit 1 TSs were previously revised to be consistent with TSTF-449, "Steam Generator Tube Integrity," Revision 4 (Reference 6.3) for its replacement SGs (References 6.4 and 6.5).

The proposed changes in TSTF-51 0 reflect the industry's early implementation experience with respect to TSTF-449. TSTF-510 characterizes the changes as editorial corrections, changes, and clarifications intended to improve internal consistency, consistency with implementing industry documents, and usability without changing the intent of the requirements. The proposed changes are an improvement to the existing SG inspection requirements and continue to provide assurance that the plant licensing basis will be maintained between SG inspections.

Furthermore, the first SG inservice inspection following replacement was completed during the fall 2011 refueling outage (T1 R19) in accordance with Technical Specification 6.19.d.1. The first 144 effective full power month (EFPM) Inspection Period was entered following start-up from T1R19 as defined in the current and proposed Technical Specification 6.19.d.2.

There are no significant variations from the USNRC-approved TSTF-510 in this submittal.

2.0 DETAILED DESCRIPTION Exelon Generation Company, LLC, (EGG) has reviewed the Model Safety Evaluation provided in the Federal Register Notice of Availability for TSTF-51 0 dated October 27, 2011 and its associated referenced information provided in TSTF-510. EGC has concluded that the technical justifications presented in Section 3.0, Technical Evaluation, of the Model Safety Evaluation and its associated referenced information are applicable to TMI, Unit 1 and justify this amendment for incorporation of the changes to the corresponding TS.

Consistent with TSTF-510 the proposed changes are as follows:

NOTE: Proposed revisions to the TS Bases are also included in this application for information only. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

  • Clarifications are made to TS 3.1.1.2.a.2, 3.1.1.2.a.3, 4.19.2, 6.19.c, 6.19.d, and their applicable TS Bases Sections. References to "tube repair criteria" are revised to "tube

Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection Page 2 of 5 plugging criteria" to be consistent with the treatment of SG tube repair throughout Specification 6.19.

  • TS 6.19, "Steam Generator (SG) Program," is revised to make an editorial correction to the introductory paragraph. The last sentence is revised from "In addition, the Steam Generator Program shall include the following provisions:" to "In addition, the Steam Generator Program shall include the following:" The subsequent paragraph starts with "Provisions for

... " and stating "provisions" in the introductory paragraph is duplicative.

  • An editorial correction is made to TS 6.19.b.1. The closing parenthesis is misplaced. It currently states "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 inappropriately includes anticipated transients in the description of normal operating conditions. The sentence is revised to, "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), all anticipated transients included in the design specification, and design basis accidents. "
  • Clarifications are made to TS 6.19.d. The term "An assessment of degradation" is replaced with "A degradation assessment" to be consistent with the terminology used in the industry program documents.
  • TS 6.19.d.1 is revised from "Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement" to "Inspect 100% of the tubes in each SG during the first refueling outage following SG installation." This wording change will allow the SG Program to apply to both existing plants and new plants.
  • The proposed change revises TS 6.19.d.2 within the SG Program to modify the frequency of verification of SG tube integrity and SG tube sample selection to reduce implementation issues experienced with the current specification. The revised specification is consistent with the existing specification in that it continues to be based on SG tube material type, age, condition and cycle length, and continues to address the time dependence of degradation and prevent front end or back end loading of inspections. In addition, the maximum interval allowed between inspections is the same as in the current TS.
  • TS 6.19.d.3 refers to "next inspection for each SG ... shall not exceed 24 effective full power months or one refueling outage (whichever is less)." An editorial change is made to the parenthetical statement in order to clarify the intent. It is revised to "(whichever results in more frequent inspections)." TS 6.19.d.3 is also revised to clarify the SG inspection requirements when crack indications are found.
  • TS 6.9.6, "Steam Generator Tube Inspection Report," is revised to change the reporting requirements. TS 6.9.6.f is revised to require reporting the effective plugging percentage.

TS 6.9.6.h, which required reporting the effective plugging percentage, is deleted. The word "active" was removed from paragraph 6.9.6.b and 6.9.6.e to be consistent with TS 6.19.

Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection Page 3 of 5

3.0 TECHNICAL EVALUATION

EGC has reviewed the Model Safety Evaluation provided in the Federal Register Notice of Availability for TSTF-510 dated October 27,2011 and its associated referenced information provided in TSTF-51 O. EGC has concluded that the history and system descriptions presented in Section 2.0, Regulatory Evaluation, and the technical justifications presented in Section 3.0, Technical Evaluation, of the Model Safety Evaluation and its associated referenced information are applicable to TMI, Unit 1 and justify this amendment for incorporation of the changes to the corresponding TS.

There are no significant variations or deviations in EGC's proposal from the TS changes described in the Federal Register Notice of Availability of TSTF-51 0 dated October 27, 2011 and its associated referenced information provided in TSTF-510. The three minor variations to the approved TSTF-51 0 are described below:

1. TMI, Unit 1 is a custom technical specification PWR plant and, therefore, the applicable TSs and associated Bases section numbers are different from the Babcock and Wilcox Owners Group (BWOG) Standard TSs provided in TSTF-510.
2. TSTF-510, Section 2.0, paragraph four, states "An editorial correction is made to add a missing closing bracket to the end of Paragraph 5.5.9.b.2." This change affects a bracketed (plant-specific) sentence in the BWOG Standard TS. TMI, Unit 1 did not incorporate the bracketed wording in corresponding TS 6.19.b.2 and therefore this change does not apply.
3. One of the editorial improvements in TSTF-51 0 was revising references to "tube repair criteria" to "tube plugging [or repair] criteria." The BWOG Standard TS mark-up of TS Section 5.5, "Steam Generator Tube Inspection Program" provided in TSTF-510, contained three versions of paragraph 5.5.9.d.2 (one each for 600MA tubing, 600TT tubing, and 690TT tubing). All three versions contain the following statement:

"If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated." (emphasis added)

The TSTF determined that this paragraph contains an administrative error. The italicized phrase in paragraph 5.5.9.d.2 (above) should state "tube plugging [or repair] criteria,"

consistent with the other changes made in TSTF-51 O. TMI, Unit 1 has corrected this administrative error in the submitted mark-up for TS Section 6.19.d.2, Insert 1. This change meets the original intent of TSTF 510.

Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection Page 4 of 5

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements I Criteria EGC has reviewed the Model Safety Evaluation provided in the Federal Register Notice of Availability for TSTF-51 a dated October 27, 2011 and its associated referenced information provided in TSTF-51 O. EGC has concluded that the regulatory history presented in Section 2.0, Regulatory Evaluation, of the Model Safety Evaluation and its associated referenced information are applicable to TMI, Unit 1 and justify this amendment for incorporation of the changes to the corresponding TS.

Additional reviews have determined that the proposed changes do not require any exemption or relief from regulatory requirements other than the TS, and do not affect conformance to any General Design Criteria differently than described in the Updated Final Safety Analysis Report (UFSAR).

4.2 Precedent This change is consistent with the Notice of Availability of TSTF-51 0, Revision 2, (Reference 2) announced in the Federal Register on October 27, 2011 (76 FR 66763) as part of the consolidated line item improvement process.

4.3 No Significant Hazards Consideration Determination Exelon Generation Company, LLC, (EGG) has reviewed the proposed no significant hazards consideration determination included in the model safety evaluation provided in the Federal Register Notice of Availability of TSTF-51 a dated October 27, 2011 and its associated referenced information provided in TSTF-51 0, Revision 2. EGC has concluded that the proposed determination presented in the notice is applicable to Three Mile Island Nuclear Station, Unit 1 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a).

Based on the above, EGC concludes that the proposed changes do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Revise Steam Generator Program Inspection Frequencies and Tube Sample Selection Page 5 of 5

5.0 ENVIRONMENTAL CONSIDERATION

EGC has reviewed the environmental evaluation included in the model safety evaluation provided in the Federal Register Notice of Availability of TSTF-510 dated October 27,2011 and its associated referenced information provided in TSTF-51 0, Revision 2. ECG has concluded that the staffs findings presented in that evaluation are applicable to Three Mile Island Nuclear Station, Unit 1 and the evaluation is hereby incorporated by reference for this application.

6.0 REFERENCES

6.1. TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."

6.2. Notice of Availability of the "Models for Plant-Specific Adoption of Technical Specifications Task Force Traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,"" dated October 27, 2011.

6.3. TSTF-449, Revision 4, "Steam Generator Tube Integrity."

6.4. Letter from P. Bamford (NRC) to C. Pardee (Exelon Corporation, LLC), "Three Mile Island Nuclear Station, Unit 1 - Issuance of Amendment RE: Technical Specification Changes to Reflect Steam Generator Replacement (TAC NO. MD9923)," dated September 15, 2009.

6.5. Letter from P. Bamford (NRC) to C. Crane (Exelon Corporation, LLC), "Three Mile Island Nuclear Station, Unit 1 -Issuance of Amendment 261 Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process and Generic Letter 2006-01 (TAC Nos. MD1807 and MD0115)," dated September 27,2007.

ATTACHMENT 2 Markup of Technical Specifications Pages Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR*50 REVISED TECHNICAL SPECIFICATIONS PAGES 3-1a 3-1b 4-77 6-19 6-20 6-26 6-27 6-28

3.1 REACTOR COOLANT SYSTEM 3.1.1 OPERATIONAL COMPONENTS Applicability Applies to the operating status of reactor coolant system components.

Objective To specify those limiting conditions for operation of reactor coolant system components which must be met to ensure safe reactor operations.

Specification 3.1.1.1 Reactor Coolant Pumps

a. Pump combinations permissible for given power levels shall be as shown in Specification Table 2.3.1.
b. Power operation with one idle reactor coolant pump in each loop shall be restricted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the reactor is not returned to an acceptable RC pump operating combination at the end of the 24-hour period, the reactor shall be in a hot shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
c. The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.

3.1.1.2 Steam Generator (SG) Tube Integrity

a. Whenever the reactor coolant average temperature is above 200°F, the following conditions are required:

(1.) SG tube integrity shall be maintained.

AND ~PIUgging I (2.) All SG tubes satisfying the tube ~ criteria shall be plugged in accordance with the Steam Generator Program. (The Steam Generator Program is described in Section 6.19.)

ACTIONS:

NOTE--------------------------------------------------

Entry into Sections 3.1.1.2.a.(3.) and (4.), below, is allowed for each SG tube.

(3.) If the requirements of Section 3.1.1.2.a.(2.) are not met for one or more tubes then perform the following:

3-1a Amendment No. 12, 17,28,47, 98, ~

~PIUgging I With one or more SG tubes satisfying the tube Fepa+f criteria and not plugged in accordance with the Steam Generator Program:

a. Verify within 7 days that tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, AND
b. Plug the affected tube(s) in accordance with the Steam Generator Program prior to exceeding a reactor coolant average temperature of 200°F following the next refueling outage or SG tube inspection.

(4.) If Action 3., above, is not completed within the specified completion times, or SG tube integrity is not maintained, be in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

3.1.1.3 Pressurizer Safety Valves

a. The reactor shall not remain critical unless both pressurizer code safety valves are operable with a lift setting of 2500 psig +/- 1%.
b. When the reactor is subcritical, at least one pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III.

3-1 b Amendment No. ~

4.19 STEAM GENERATOR (SG) TUBE INTEGRITY Applicability: Whenever the reactor coolant average temperature is above 200°F Surveillance Requirements (SR):

Each steam generator shall be determined to have tube integrity by performance of the following:

4.19.1 Verify SG tube integrity in accordance with the Steam Generator Pro ram.

plugging 4.19.2 Verify that each inspected SG tube that satisfies the tube . cntena IS pugged in accordance with the Steam Generator Program prior to exceeding an average reactor coolant temperature of 200°F following an SG tube inspection.

BASES:

BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by TS Section 3.4.

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively.

The SG performance criteria are used to manage SG tube degradation.

Specification 6.19, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.19, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and 4-77 Amendment No. 47,294 (12-22-78)

6.9.5 CORE OPERATING LIMITS REPORT 6.9.5.1 The core operating limits addressed by the individual Technical Specifications shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle.

6.9.5.2 The analytical methods used to determine the core operating limits addressed by the individual Technical Specifications shall be those previously reviewed and approved by the NRC for use at TMI-1, specifically:

(1) BAW-10179 P-A, "Safety and Methodology for Acceptable Cycle Reload Analyses." The current revision level shall be specified in the COLR.

(2) TR-078-A, "TMI-1 Transient Analyses Using the RETRAN Computer Code", Revision O. NRC SER dated 2/10/97.

(3) TR-087-A, "TMI-1 Core Thermal-Hydraulic Methodology Using the VIPRE-01 Computer Code", Revision O. NRC SER dated 12/19/96.

(4) TR-091-A, "Steady State Reactor Physics Methodology for TMI-1",

Revision O. NRC SER dated 2/21/96.

(5) TR-092P-A, "TMI-1 Reload Design and Setpoint Methodology",

Revision O. NRC SER dated 4/22/97.

(6) BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel", NRC SER dated February 4, 2000.

6.9.5.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient/accident analysis limits) of the safety analysis are met.

6.9.5.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

6.9.6 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the average reactor coolant temperature exceeds 200°F following completion of an inspection performed in accordance with Section 6.19, Steam Generator (SG) Program. The report shall include:

The scope of inspections performed on each SG, 1.---lY.-;~~e4e§f:a4atiE~ mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism, 6-19 Amendment No. 72,77,129,137,141,149,150,168,173,178,202, 233, 261, ~
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 aeti¥e degradation mechanism, f.
g. The results of condition monitoring, inclu ing the results of tube pulls and in-situ testing, "f::: 11.

'\~ The number and

~ percentage of tubes plugged to date, and 1.-----------1 the effective plugging percentage in each steam generator, 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

a. Records of normal station operation including power levels and periods of operation at each power level.
b. Records of principal maintenance activities, including inspection, repairs, substitution, or replacement of principal items of equipment related to nuclear safety.
c. All REPORTABLE EVENTS.
d. Records of periodic checks, tests and calibrations.
e. Records of reactor physics tests and other special tests related to nuclear safety.
f. Changes to procedures required by Specification 6.8.1.
g. Deleted
h. Test results, in units of microcuries, for leak tests performed on licensed sealed sources.
i. Results of annual physical inventory verifying accountability of licensed sources on record.
j. Control Room Log Book.
k. Control Room Supervisor Log Book.

6-20 Amendment No.72, 77,129,137,141,150,173,180,219, 261, ~

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
1. A change in the TS incorporated in the license or
2. A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59.
c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
d. Proposed changes that meet the criteria of Specification 6.18. b.1 or 6.18.b.2 above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).

6.19 STEAM GENERATOR (SG) PROGRAM 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 pFO'Jisions:

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 lias 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 dow aM all anticipated transients included in the design specification, nd design basL..,.i-s---+---i),

accidents. This includes retaining a safety factor of 3.0 agai .'i!.lt...;tr&ld.l..........ld.LJ.lo4.:to<.l. +-!

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.

6-26 Amendment No. 2W, 29+

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.

3. The operational leakage performance criterion is specified in TS 3.1.6, "LEAKAGE."
c. Provisions for SG tubit:feeaif criteria.

~----------.plugglng I . I

1. 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.

6-27 Amendment No. ~, 2+4-

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 satis the a lica ~ criteria. The tube-to-tubesheet weld is no pa 0 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 ef." . hall be performed to determine the type and location of flaws to which A

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

assessment

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG r:e:Pla:e:e:rn:e::.: linstallation I I ~""""-Ins-e-rt---'1
2. Insp t tubes at sequential periods of 144, 108,72, and, t~fter, I 60 effeetive full power months. The first sequential period shall be Gonsidered to be§!in after the first inserviee inspeotion of the 8Gs. In addition, inspeet 50% of tRe tl:lBes By tRe refl:lelin§! ol:lta§!e nearest the midpoint of the period and the remainin§! eO% By the refuelin§! outa§!e nearest the end of the period. No 8G shall operate for R10re than 72 effeetive full pO\'Jor months or three refuel in§!

ol:lta§!es (whichever is less) withol:lt bein§! inspeeted.

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. results in more frequent inspections affected and potentially affected
e. Provisions for monitoring operational primary to secondary leakage.

6-28 Amendment Ne,. ~, 2-74

INSERT 1

2. After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated.

The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

ATTACHMENT 3 Markup of Technical Specifications Bases Pages (For Information Only)

Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 REVISED TECHNICAL SPECIFICATIONS BASES PAGES 4-78 4-80 4-81 4-82 4-83

BACKGROUND (continued) operational leakage. The SG performance criteria are described in Specification 6.19. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

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 a SGTR event assumes a bounding primary to secondary leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (Le., 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 conservatively assumed to be equal to, or greater than, the TS 3.1.4, "Reactor Coolant System 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 TS 3.1.1.2.a The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the . criteria be plugged in accordance with the S . tor Pro ra. plugging plugging During a SG inspe 'on, any inspected tube that satisfies the Steam Generator Program fef*!H: criteria is removed from service by plugging. If a tube was determined to satisfy the ~ criteria but was not plugged, the tube may still have tube integriM' Iplugging I 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.

The tube-to-tubesheet weld is not considered part of the tube.

4-78 Amendment No. 47, 153, 237, 261, 27-+

(12-22-78)

LCO (continued) The accident induced leakage rate includes any primary to secondary leakage existing prior to the accident in addition to primary to secondary leakage induced during the accident.

The operational leakage performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational leakage is contained in TS 3.1.6.3, "LEAKAGE," and limits primary to secondary leakage through anyone SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABI L1TY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced when the reactor coolant system average temperature is above 200°F.

RCS conditions are far less challenging when average temperature is at or below 200°F; primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for leakage.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

3.1.1.2.a.(3.)a. and 3.1.1.2.a.(3.)b.

.--------1 piugg ing 3.1.1.2.a.(3.) applies if it is discovered that one or re SG tubes examined in an inservice inspection satisfy the tube fepatf criteria but were not plugged in accordance with the Steam Generator Program as required by Surveillance Requirement 4.19.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG .... plugging Steam Generator Program. The S . criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, 3.1.1.2.a.(4.) applies.

4-80 Amendment No 116,149, 153,206,237,261,2+4

ACTIONS (continued)

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action 3.1.1.2.a.(3.)b. allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to exceeding a reactor coolant average temperature of 200°F following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

3.1.1.2.a.(4.)

If the Required Actions and associated Completion Times of Condition 3.1.1.2.a.(3.) are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENT SR 4.19.1:

During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, "Steam Generator Program Guidelines" (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the prevo . period.

plugging The Steam Gene tor Program determines the scope of the inspection and the methods ed to determine whether the tubes contain flaws satisfying the tube fepaif criteria. Inspection scope (Le., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also 4-81 Amendment No. 47, 83, 91, 103, 129, 14Q, 153, 157, 206, 209, 237, 2&t

SURVEILLANCE REQUIREMENTS (continued) specifies the inspection methods to be used to find potential degradation.

Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.19.1. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.19 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. ~Insert 21 SURVEILLANCE REQUIREMENT SR 4.19.2: I .

...-----lP uggmg During an SG ins tion, any inspected tube that satisfies the Steam Generator Progra FepaH: criteria is removed from service by plugging. plugging The tub . criteria delineate in peci Ication 6.19 are intended to ensure that tubes accepted for continued service satisfy the SG I .

performance criteria with allowance for error in the flaw size r plugging measurement and for future flaw growth. In addition, the tubEWFepaH:

criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

4-82 Amendment No. 47, 86, 116, 14Q, 153, 206, 20Q, 237, 261, ~

INSERT 2 If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 6.19 until subsequent inspections support extending the inspection interval.

The frequency of "prior to exceeding an average reactor coolant temperature of 200°F following an SG tube inspection" en~re5 that the [plugging Surveillance has been completed and all tubes meeting th fePaif criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

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".

4-83 (Pages 4-84 through 4-85 deleted)

Amendment No. 47, 12Q, 206, 20Q, 237, 261, 2+4

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