Text

Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,Using the Consolidated Line Item Improvement...
	ML13025A163
Person / Time
Site:	Vogtle, Farley
Issue date:	01/23/2013
From:	Ajluni M; Southern Nuclear Operating Co, Southern Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	NL-13-0081
	Download: ML13025A163 (70)
	v • d • e

Mark J. Ajluni, P.E. Southern Nuclear Nuclear Licensing Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birm ingham, Alabama 35201 Tel 205.992.7673 Fax 205 992.7885 January 23, 2013 SOUTHERN'\'

Docket Nos.: 50-348 50-424 COMPANY NL-13-0081 50-364 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Application to Revise Technical Specifications to Adopt TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using The Consolidated Line Item Improvement Process Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is submitting a request for an amendment to the Technical Specifications (TS) for the Joseph M. Farley Nuclear Plant (FNP) - Units 1 and 2 and the Vogtle Electric Generating Plant (VEGP) - Units 1 and 2.

The proposed amendment would modify TS requirements regarding steam generator tube inspections and reporting as described in TSTF-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection. " provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications . Enclosures 2 and 3 provide the existing TS pages marked up to show the proposed changes for FNP and VEGP, respectively. Enclosures 4 and 5 provide revised (clean) TS pages for FNP and VEGP, respectively. Enclosures 6 and 7 provide existing TS Bases pages marked up to show the proposed changes FNP and VEGP, respectively.

SNC requests approval of the proposed license amendments by November 30, 2013. The proposed changes would be implemented within 60 days of issuance of the amendment.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

U. S. Nuclear Regulatory Commission NL-13-00B1 .

Page 2 Mr. Mark J. Ajluni states that he is Nuclear Licensing Director of Southern Nuclear Operating Company (SNC), is authorized to this oath on behalf of SNC, and to the best of knowledge belief, the forth in letter are true.

Sincerely, Mark J. Ajluni Nuclear Licensing Director MJAlRMJ/lac Sworn to subscribed My commission

Enclosures:

1. Enclosure 1, for Change Enclosure Technical Specification - Marked-up Pages (FNP)

Enclosure 3, Technical Specification Change - Marked-up Pages

4. 4, Technical Specification Change - Clean-typed Pages (FNP)

5. 5, Technical Specification Change - Clean-typed (VEGP)

6. 6, Proposed Technical Specification Change (for information only) (FNP)

7. Enclosure 7, Proposed Technical Specification Bases Change (for information only) (VEGP)

U. S. Nuclear Regulatory Commission NL-13-0081 Page 3 cc:

Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. Bost, Executive President & Chief Nuclear Officer Mr. T. A. Lynch, Vice - Farley Mr. Tynan, Vice President - Vogtle Mr. L Ivey, Vice President - Regulatory Affairs RType: CFA04.054; CVC7000 Mr. V. M. McCree, Regional Administrator Mr. Martin, I\lRR Project Manager -Vogtle Mr. L M. Cain, Senior Resident Inspector - Vogtle E. A. Brown, NRR Project Manager - Farley Mr. P. K. Niebaum, Senior Resident Inspector - Farley Mr. J. R. Sowa, Senior Inspector - Farley H. Turner, Environmental Director Protection Division

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Application to Revise Technical Specifications to Adopt TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using The Consolidated line Item Improvement Process Enclosure 1 Basis for Proposed Change

Enclosure 1 Basis Proposed Change Table of Contents

1. Summary Description Assessment

3. Regulatory Analysis

4. Environmental Evaluation

-1 to NL-i3-00Bi Basis for Proposed

1. Summary Description Southern Nuclear Operating Company (SNC) proposes to revise the Technical (TS) for the Joseph M. Farley - Units 1 and 2 (FNP) and the Vogtle Generating Plant - Units 1 and 2 The proposed change revises TS Section "Steam Generator (SG) Program," 0, "Steam Generator Tube Inspection Report," and the Steam Generator Tube Integrity (LCO 3.4.17) for both FNP and The proposed changes are needed to implementation issues associated with periods, and address changes and clarifications.

The proposed amendment is 0, Revision 2, "Revision to Generator Program Sample Selection."

In addition, this oraiOO!~ea indenting for FNP TS Section top of page 5.5-S. This administrative, and has no technical impact on Assessment 2.1 SNC has reviewed TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample " and the model safety evaluation dated October justifications presented in ° 2011 as part of the Federal Register Notice for Comment. SNC has concluded that the model safety evaluation prepared by the NRC are applicable to FNP and VEGP and justify this amendment for the incorporation of the to both the FNP TS and to the P 2.2 Optional Changes and Variations SNC is proposing the following described in the Revision 2, or the applicable model safety evaluation 1:

The FNP TS and the Specifications (STS) - Westinghouse Specifically, the "Steam Generator

° different numbering than the Standard Technical on which TSTF-51 was based.

Inspection is in Section 5.S.7 of the STS, whereas it is in Section 10 for both the FNP and the VEGP TS. In addition, the "Steam Generator (SG) Tube Integrity" is in 3.4.20 of the STS and STS Bases, whereas it is in Section 3.4.17 of the FNP and VEGP TS and TS Bases.

TSTF-510 proposes to combine 5.S.7.1 and, therefore, delete 5.S.7.h.

FNP and VEGP currently do not have ,..rll*..o(~nr.nN,nn requirements from Section 5.S.7.h of the STS. [Note: As stated above, in the STS corresponds to Section 5.S.10 in the FNP TS and the requirements of Section 5.6.7.h of the STS are being added to the requirements in 10f of the FNP TS and VEGP TS.

The proposed change f'r'lr'rt:>f"f~ inconsistency in TSTF-510, Paragraph d.2 of the STS Section Inspection Program". In Section 2.0, "Proposed Change," TSTF-51° ran.rac to "tube repair criteria" in Paragraph d are revised to plugging II However, in the following to NL-13-0081 Basis Proposed Change sentence in Paragraph this change was inadvertently omitted, "If a degradation assessment indicates the potential for a type degradation to occur at a location not previously with a technique capable of detecting this type of degradation location and that may satisfy the applicable the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may prorated" (Emphasis added).

SNC does not have an tube repair criteria. Therefore, sentence is revised to "tube plugging" criteria.

These ron,I"t:>.c are administrative and do affect the applicability of oto the Fi\JP TS.

3. Regulatory Analysis 3.1 The Joseph M. Nuclear Plant Units 1 and 2 (FNP) and the Electric nor",'",,,,, Plant - Units 1 and 2 (VEGP) requests adoption of an approved change to Standard Technical Specifications (STS) into the plant Technical Specifications to revise the Specification "Steam Generator (SG) Program," 5.6.10, Generator Inspection .. and 17, "Steam Generator Tube Integrity," to address inspection periods and other administrative changes and clarHications.

As required by 10 50.91 (a), an of the of no significant hazards consideration is presented below:

1. Does proposed change involve a significant increase in probability or consequences of an accident previously evaluated?

No.

The proposed change the Steam Generator (SG) Program to modify the frequency of verification tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of design basis accidents that are analyzed as part of a licensing The proposed tube inspection frequency sample will continue to ensure that tubes are such that probability a SGTR is not increased. The consequences of a SGTR are bounded by conservative assumptions in the design accident The change will cause the consequences of a SGTR to those assumptions .

.... "r.oTl"".,., it is concluded that this does not a significant in the probability or consequences of an accmelnt previously evaluated.

Does the proposed change the possibility of a new or different kind of accident from any previously evaluated?

Response: No.

E1

1 to NL-13-0081 for Proposed Change The proposed changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. proposed change not affect the of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

it is that this does not possibility of a new or different kind of accident from any accident previously evaluated.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

SG tubes in water are an of the reactor coolant pressure boundary and, as such, are upon to maintain the primary pressure and inventory. of the reactor coolant pressure boundary, the SG tubes are unique in that they are relied upon as a heat transfer surface the primary and secondary systems such residual heat can removed from the primary system. In addition, the SG tubes also the radioactive products in primary coolant secondary system. In summary, function of a is maintained the integrity of generator integrity is a of the deSign, environment, and the physical condition of the The proposed does not affect deSign or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes that there will not a reduction in margin of safety compared to the current requirements.

Therefore, it is concluded that the proposed change not involve a reduction in a margin of on the above, SNC concludes proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4. Environmental Evaluation proposed would change a requirement with to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would an inspection or surveillance requirement. However, the proposed change 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 offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change the eligibility criterion for categorical exclusion forth in 10 CFR .22{c){9).

10 CFR 51.22(b), no or assessment in connection with the proposed change.

-4

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Application to Revise Technical Specifications to Adopt TSTF-510, II Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, II Using The Consolidated Line Item Improvement Process Enclosure 2 Technical Specification Change - Marked-up Pages (FNP)

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained .

All SG tubes satisfying the tube fej3aif-pJlt9.SiJ l9.",criteria shall be plugged in accordance with the Steam Generator Program .

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

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

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube ~ affected tube(s) is plugging criteria and not maintained until the next plugged in accordance inspection.

with the Steam Generator AND Program. Prior to entering MODE 4 following A.2 Plug the affected tube(s) the next refueling in accordance with the outage or SG tube Steam Generator inspection Program .

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

Farley Units 1 and 2 3.4.17-1 Amendment No. .:uM-(Unit 1)

Amendment No. +ae-(Unit 2)

Tube Integrity 3.4.17 SURVEILLANCE SURVEILLANCE FREQUENCY 3.4.17.1 Verify SG tube integrity in accordance with with Steam Generator Program.

Generator Program 3.4.17.2 tube that satisfies the Prior to entering is in accordance MODE 4 following with the Steam Generator Program. a tube inspection Farley Units 1 and 2 3.4.1 Amendment No. ~(Unit 1)

Amendment No. 2)

Programs and Manuals 5.5 Manuals Reactor Coolant Pump Flywheel Inspection Program (continued)

b. A surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the disassembled flywheel.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program.

program provides controls for Code Class 1 , 2, 3 components. The program shall

a. frequencies specified in selctlc,n the ASME Boiler and Pressure Vessel Code and applicable 1"1""""'1"<> as follows:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At once per 92 days Semiannually or every 6 months At once per 184 days 9 months At least once per 276 days Yearly or annually At least once per 366 days or every 2 years At least once per 731 days provisions of SR are Frequencies for performing

c. provisions of SR 3.0.3 are applicable to testing activities; and Nothing in the ASME Boiler and shall be construed to supersede the requirements of any 5.5.9 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 Condition monitoring 1 2 5.5-5 Amendment No. +00 (Unit 1)

Amendment No. +8& (Unit 2)

Programs and Manuals and Manuals

==':':~:'=":==...J...:::::=.t....:....:..;~~ (continued) assessment means an evaluation of the "as found" condition of the tubing with to the performance criteria for structural integrity and aCClce!nt Al ign indent induced leakage. The found" condition refers to the condition of the with Section tubing during an inspection outage, as determined from the inservice 55.9.b inspection or other prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which 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 L ' - '.... ' ' ' '...C1IL.

1. Structural integrity performance criterion: Ali inservice tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby and cooldownh a-fl4.all anticipated transients included in the design specification,~ and design basis accidents. This includes retaining a safety factor of 3.0 (36P) 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 accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the basis or combination of accidents in accordance with the design and licensing basis, shall also be evaluated determine if the associated loads contribute significantly to burst or collapse. In the assessment tube integrity, those loads that do significantly affect burst or collapse shall be determined and in combination with the loads due pressure with a safety factor of 1 on the combined primary loads and 1.0 on axial secondary

2. Accident induced leakage performance criterion: primary to secondary accident induced leakage rate for any design accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total rate for all SGs and for an individual Accident induced leakage is not to exceed 1 gpm total aI/ SGs.

3. operational LEAKAGE performance criterion is specified in LCO 3.4.1 "RCS Operational LEAKAGE."

Farley Units 1 and 2 5.5-6 Amendment No. .:t-72 (Unit 1)

Amendment No. (Unit 2)

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

c. Provisions for SG tube repair plugging 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.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair plugging 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 assessment 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 replaoementinstallation.

2. 2-:--After the first refueling outage followiog SG installation. inspect each SG at least every 72 effective full power months or at least everv third refueling outage (whichever results in more freIDJ.§.[Jj inspectioos). In addition. the minimum number of tubes inspectsill...a1 each scheduled inspection shaLLbe the number of tubes in all SGs gjyjg~,gJ!,lJh~Jl!JJJJP~[=9L§-G--i~~.ctiQ[) oJJtages sched.uill.d lli~

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 a.t~9J;;~Jl2D.llQiQJ-eyt~?~,lo§R~t~g,~i!b aJ~~bDJ~ly~=~.gR.qJ?~=9f detecting this type of degradation at this location and that may satisfy

~he @..Q.!icable tube plugging criteria, the minimum number of 10catiQQ.S.

jnspecre:d with such a capable jnspecti.on technique during the remainder of the inspection perioQmay be prorated . The fraction of iQcations tQ be inspecte.rLfuLthis potential type of degradation at this IQcation at the end of the inspection period shall be no les.sJbaQJhe.

ratio of the number of times th~i~eduled to be inspected in tile insp~tion p~riod 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 in.s~ctio_n period. E;a"ch inspection period defined below may be extended up to 3 effective full power months to include a SG (continued)

Farley Units 1 and 2 5.5-7 Amendment No. +eJ (Unit 1)

Amendment No. +W (Unit 2)

Programs and Manuals 5.5 In?,,Re_~t12.rJ-~QJJJg£l§ in anin~12..4GtiQo_ period aill1.1h~ s ubse~l-bL~J)1 inspection period beginsgt the conclusion of the included SG ins.Qection outage.

a) After the first refueling outage following SG installation. inspect 100% of the tubes duringJ.he next 144 effective full power months. Th~Q1lS1iiutes the first inspection period; 2J Durillill~e~1120 effective full power montb~J.o.§p.,!iCL~

of th_e tubes. This constitutes th~econd inspection period:

c) During the next 96 effective full power months. inspect 100% of the tubes . This constitutes the third inspection period: and s.:J.J- _D _. a the ceIlli!lnlog~QU!J~G.~~jnsQe_cUOQ_CZ'O-QfJOO tu.b.(;ts.

every 72 effective full power months. This constitutes the fourth and subsequent inspectiQn periods.

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

3. If crack indications are found in any SG tube, then the next inspection for each affected and Dotent1ially affected 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 results in more freque.nlinspect.iQns). 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)

Farley Units 1 and 2 5.5-8 Amendment No. -+eJ (Unit 1)

Amendment No. +W (Unit 2)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.9 Deleted 5.6.10 Steam Generator (SG) Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active dQegradation 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 astWe degradation mechanism,

f. Total number and percentage of tubes plugged to date, andThe number and percentage of tubes plugged to date. and the eflOOive pluggiQ£1 percentage in each steam generator,

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

5.6.11 Alternate AC (AAC) Source Out of Service Report The NRC shall be notified if the AAC source is out of service for greater than 10 days.

Farley Units 1 and 2 5.6-6 Amendment No.-+74 (Unit 1)

Amendment No.--1--e+ (Unit 2)

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Application to Revise Technical Specifications to Adopt TSTF-510, II Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using The Consolidated Line Item Improvement Process Enclosure 3 Technical Specification Change - Marked-up Pages (VEGP)

Tube Integrity 17 3.4 COOLANT SYSTEM (RCS) 17 (SG) Integrity LCO 17 tube integrity shall be maintained.

All tubes satisfying the tube Fef:liCtf-*RlYWm.c1*,for,<:> shall plugged in accordance with the Steam APPLICABILITY: 1,2, and ACTIONS Separate Condition entry is allowed for each CONDITION REQUIRED ACTION COMPLETION TIME One or more tubes A.1 Verify tube integrity the 7 days satisfying the tube affected tube(s) is maintained QJJJgg.L~icrltCana and not until the next refueling outage plugged in <:>"f'r.nri<:>rl"t> or SG tube inspection.

with the Steam Program.

Plug the tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection Required Action and B.1 in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months tube integrity not maintained.

Vogtle Units 1 2 3.4.17-1 Amendment (Unit 1)

Amendment No.424 (Unit 2)

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance with Steam Generator Program. the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies Prior to entering the tube ~plugging criteria is plugged in MODE 4 following a accordance with the Steam Generator Program. SG tube inspection Vogtle Units 1 and 2 3.4.17-2 Amendment No.-444 (Unit 1)

Amendment No.4-24 (Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 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 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 downl aREi-ali anticipated transients included in the design specification+/-1 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.

(continued)

Vogtle Units 1 and 2 5.5-7 Amendment NO.-t44 (Unit 1)

Amendment No.~ (Unit 2)

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

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG . Leakage is not to exceed 1 gpm per SG.

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

c. Provisions for SG tube ~~91D9: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 plugging criteria shall be applied as an alternative to the 40% depth based criteria:

Tubes with service-induced flaws located greater than 15.2 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 15.2 inches below the top of the tubesheet shall be plugged upon detection.

(continued)

Vogtle Units 1 and 2 5.5-8 Amendment No.-4-e7-(Unit 1)

Amendment No. +4Q.-(Unit 2)

Programs and Manuals and Manuals

==:"':"":::=:'=":=.!....I,.;::::"=I-!-'-""'l"-='-'-' (continued )

d. Provisions for tube SG tube inspections performed. The number and portions of the tubes inspected and methods of inspection shall performed with objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld the tube outlet, and may satisfy the applicable tube Portions of the tube below 1 below top of the tubesheet are excluded from requirement.

Vogtle Units 1 and 2 Amendment No. .::t-e+-(Unit 1)

Amendment No. +4Q-(Unit 2)

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

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. Aft assessment of degradation ~essment 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 replacomentinstallation.

2. After the first refueling outage following SG installation. i~spect each

~~~.,alle_90~t",EZ¥'~Y_4a~ffJ}J.;..t l¥elYlLpower mopths or at least eVeLY=Qlb.er refueling outage (whichever results in more frequent inspections). In ad.d.iooILJb.-eJIllnlO1um nUJIlber of tubes inspected at eaclLs.ffieJ1.u.l..e.d inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheJiuled in each inspection period as_defined in__~b. and c below. If a~.a_d~assessment indkille.s.

the. potential for a type of degradation to os;cJ.!L...aLa 10catioD.JlQJ previous ly inspected with a technique capable of de.Le....cting this type of

~ad~n-.aLthis location and that may satisfy the applicatlliLtub..e.

plugging criteria . the minimum number of locations inspected with such acapable inspection technique during the remainder of the inspection Q§riod may ge prorated . The fraction of locations to be inspected for thiS---f!otential type of de..9La..d~tion at this location at the end Qf...1he inspection period shall be no less than the ratio of the numbeL.QUin:le..s.

the SGj~s.c heduled to be imQ.e.cted in th ej~Q~G.ti.QrJ period af1~L1tv~

determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is

~cglH~£i1JJed ..tQJ?,gj!l§JJ,~f.i~dJ.IJ thtils.@,ctL~.Re

o E.am~-Q period defined below may be extended up to 3 effective full power illQilths to include a SG inSD_ectiQn..QU!agVn.jmJo.sD~ction peQQ.d_.a.nd the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation. inspect 1.Qir'lQ QfJtN..1u.b.es during th§..Jle~O effective full pOWeL1IlQDibs...

This coos..titute._s the first inspection period :

,9-1 During the Qext 96 effe.ctiv.e..full power months. inspect 100% of the tubes. This cons1iliJ.tes the second inspection period : and (continued)

Vogtle Units 1 and 2 5.5-10 Amendment No. +a7-(Unit 1)

Amendment No. ~(Unit 2)

Programs and Manuals 5.5 c) During the remaining life of the SGs. inspect 100% of the tubes

~ery 72~ctive full power months. This cons_titutes the thir:d..and s.J.Lbsequent inspection. periods.

2. Inspect 100% of the tubes at sequential periods of 120, QO, 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) 'Nithout being inspected.

3. If crack indications are found in portions of the SG tube not excluded above, then the next inspection for each a.ffe.cted and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is lessresults in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic nondestructive 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 .

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:

(continued)

Vogtle Units 1 and 2 5.5-11 Amendment No. +a+--(Unit 1)

Amendment No. ~(Unit 2)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.9 Deleted.

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 dQegradation 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 dateI.tliul.u.m.b.er and percent~ge oj tubes plugged to-.d~d the effective plugging percentage in each steam generator,

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; and

i. The calculated accident induced leakage rate from the portion of the tubes below 15.2 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 2.48 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined.

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

Vogtle Units 1 and 2 5.6-6 Amendment No. ~(Unit 1)

Amendment No. +49-(Unit 2)

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant* Units 1 and 2 Application to Revise Technical Specifications to Adopt TSTF-S1 0, II Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using The Consolidated line Item Improvement Enclosure 4 Technical Specification Change - Clean-typed Pages (FNP)

SG Tube Integrity 17 3.4 COOLANT SYSTEM (RCS) 17 Generator (SG) Tube Integrity LCO 3.4.17 tube integrity shall maintained.

All tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, and ACTIONS

""""<::Ir<::lto Condition entry is allowed for each CONDITION REQUIRED ACTION COMPLETION TIME One or more Verify integrity of satisfying the tube affected tube(s) is plugging criteria and not maintained until the next plugged in accordance inspection.

with Steam Generator Program.

A.2 Plug the affected tube(s) Prior to in accordance with the MODE 4 following Generator the next refueling Program. outage or SG tube inspection Required Action and Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Condition A not met.

in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SG tube integrity not maintained.

Farley Units 1 2 3.4.1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance with Steam Generator Program. the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube plugging criteria is plugged in accordance with the MODE 4 following Steam Generator Program . a SG tube inspection Farley Units 1 and 2 3.4.17-2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals and Manuals Reactor Coolant Pump Flywheel Inspection Program (continued)

b. A surface examination (magnetic particle and/or liquid penetrant) of ext,OSEW surfaces of the disassembled flywheel.

provisions of SR 3.0.2 and SR 3.0.3 are applicable to the ..... a':::u'T/"l.r Coolant Pump Flywheel Inspection Program.

5.5.8 program provides controls for inservice testing ASME Code 1, and 3 components. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Code and applicable Addenda terminology Required Frequencies inservice testing for performing inservice activities testing activities Weekly At once per 7 days Monthly At once per 31 days Quarterly or every 3 months At once per days Semiannually or every 6 months At least once 184 days 9 months At least once per 276 days Yearly or annually At least once 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of 3.0.3 are applicable to and

d. Nothing in the ASME Boiler and Vessel shall be construed to supersede requirements of any TS.

5.5.9 A Steam Generator Program shall be established and implemented to ensure that tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

a. Provisions for condition monitoring as~,>es:srrleniS Condition monitoring Farley Units 1 and 2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals 5.5.9 Steam Generator (SG) Program (continued) assessment means an evaluation 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 tubes.

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

b. Performance criteria for SG integrity. SG integrity be maintained by the performance criteria for tube structural integrity.

accident induced and operational LEAKAGE.

1. Structural integrity performance criterion: All SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup. operation in the power hot standby and cooldown), all anticipated transients included in design specification, and basis accidents. includes retaining a safety factor of (3L1P) against burst under normal steady full power operation primary to pressure differential and a safety factor of 1.4 against burst applied to the design accident primary to differentials. Apart from the above requirements. additional loading conditions associated with the design basis accidents, or combination accidents in accordance with the design licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the integrity. those loads that significantly burst or collapse shall be determined and in combination with the due to pressure with a safety factor of 1 on the combined primary loads and 1.0 on axial secondary loads.

Accident induced leakage performance criterion: The primary to induced leakage rate for any design accident, other than a tube rupture, shall not exceed rate in the accident in terms of total all SGs and leakage rate for an individual Accident induced leakage is not to 1 gpm total for three

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

Units 1 and 2 Amendment No. (Unit 1)

Amendment No. (Unit

Programs and Manuals 5.5 Manuals 5.5.9 (continued)

c. for plugging Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal wall thickness be plugged.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall performed. The and the tubes and ..... "'.t... l'\rI'"

of inspection shall performed with the objective detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) may be present along the length of tube, from tube-to-tubesheet weld at tube inlet to the tube-to-tubesheet weld tube outlet, that may the applicable plugging The tube-to-tubesheet weld is part of the In addition to meeting the requirements of 1, and below, the scope, methods, and inspection intervals shall such as to ensure that SG tube integrity is maintained until the next inspection. A degradation performed to the type of flaws which the may be susceptible and, based on this assessment, to determine which inspection

.....c.t ... ,'\rI'" need employed what 10ciatICIn

1. Inspect 100% the tubes in each SG during first refueling following installation.

2. After the first outage following SG installation, inspect SG at least every effective full power months or at least every third refueling (whichever results in more inspections). In addition, the minimum number of tubes inspected at scheduled shall number of in all SGs by the number of SG inspection outages scheduled in inspection period as defined in a, b, c and d below. If a degradation d;:,~)t:::;:';::'l indicates potential a of degradation to occur at a location not previously inspected with a technique capable of detecting this type degradation location and may satisfy applicable plugging the minimum number of locations inspected with such a inspection technique during remainder of inspection period may be prorated. The fraction locations to inspected for this potential type of degradation at this location the end of inspection shall be no than the ratio number times the is scheduled to inspected in the inspection after the that a new form of could potentially be occurring at this divided by total number times the is scheduled to inspected in inspection period.

Each inspection period defined below may extended up to 3 effective full power months to include a inspection in an period and the subsequent inspection period begins at the Units 1 2 5.5-7 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals 5.5.9 =="--"'=':":'==~=::....I..":'==~ (continued) conclusion of the included inspection outage.

a) After the refueling outage following installation, inspect 100% of the tubes during the next 144 effective full power months.

This constitutes first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. constitutes the period; c) During the next 96 effective full power months, 100% of the This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 1 of the tubes 72 full power months. This constitutes fourth inspection

3. If indications are found in SG then next for each affected and potentially affected for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, testing, or engineering evaluation indicates that a crack-like indication is not with a crack(s), indication not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

5.5.10 This n ......n .. ",,..,, controls for monitoring secondary to 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 to measure the of the critical variables;

c. Identification of sampling which include monitoring the condenser hotwells evidence condenser in leakage;

d. Procedures the recording and management of data; Units 1 and 2 Amendment No. (Unit 1)

Amendment No. (Unit

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Secondary Water Chemistry Program (continued)

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 3, and in accordance with ASME N510-1989. The FNP Final Safety Analysis Report identifies the relevant surveillance testing requirements.

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 < 0.5% when tested in accordance with ASME N510-1989 at the system flowrate specified below.

ESF Ventilation System Flowrate (CFM)

CREFS Recirculation 2,000 ~ 10%

CREFS Filtration 1,000~ 10%

CREFS Pressurization 300 + 25% to - 10%

PRF Post LOCA Mode 5,000 ~ 10%

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.5% when tested in accordance with ASME N510-1989 at the system flowrate specified below.

ESF Ventilation System Flowrate (CFM)

CREFS Recirculation 2,000 ~ 10%

CREFS Filtration 1,000 ~ 10%

CREFS Pressurization 300 + 25% to - 10%

PRF Post LOCA Mode 5,000 ~ 10%

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in ASME N51 0-1989, shows the methyl iodide penetration less than the value (continued)

Farley Units 1 and 2 5.5-9 Amendment No. (Unit 1)

Amendment 1\10. (Unit 2)

Programs and Manuals 5.5 and Manuals 5.5.11 specified when tested in accordance with ASTM D3803-1989 at a temperature of::; 30°C and greater or equal to relative humidity specified below.

CREFS Recirculation 70%

Filtration 70%

CREFS Pressurization 0.5% 70%

Post LOCA Mode 5% 95%

Pressurization methyl iodide penetration limit is on depth.

d. Demonstrate for each of ESF systems that the pressure drop across filters and the charcoal is less than the value below when tested in accordance with ASME 1\1510-1 at the system flow rate specified below.

Delta P Flowrate Recirculation 2,000.::!:. 10%

Filtration 2.9 1,000.::!:. 10%

CREFS Pressurization 2.2 300 + 25% to - 10%

PRF Post LOCA Mode 2.6 5,OOO.::!:. 10%

e. Demonstrate that the heaters for the Pressurization System dissipate the value specified below when tested in accordance with ASME 1\1510-1989.

The provisions of SR 3.0.3 are applicable to the test frequencies.

12 program provides controls for potentially explosive mixtures contained in the Gas System, quantity of radioactivity contained in tanks, the quantity of radioactivity contained in unprotected outdoor liquid C"T"r".L'\ tanks.

Units 1 and 2 o Amendment 1\10. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas System and a surveillance program to ensure the limits are maintained.

Such limits shall be to

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure 2 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled of tanks' contents; and

c. A surveillance to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste System is than 10 curies.

The provisions of SR and 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance 13 A fuel oil testing program implement required testing of both new fuel oil and stored fuel oil be established. The program shall include sampling and testing acceptance all in accordance with ASTM Standards. The purpose of program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to the emergency diesel generator tanks by determining that the oil

1. an API gravity or an absolute specific gravity within limits,

2. a point kinematic within limits for ASTM 20 oil, and

3. a clear and bright appearance with proper color; or a water and sediment content within limits.

b. Other properties for ASTM 20 fuel oil are within limits within 31 days following sampling and addition to tanks; and Farley Units 1 and 2 5.5-11 Amendment No. (Unit 1)

Amendment No. (Unit

Programs Manuals and Manuals 5.5.13

c. Total particulate concentration of oil is s 10 mgtl when tested 31 days.

d. provIsions SR 3.0.2 and 3.0.3 are applicable to the Fuel Oil Testing Program surveillance test frequencies.

14 This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of made under appropriate administrative

b. may make changes to Bases without prior NRC approval provided the changes do not require of the following:

1. a in the incorporated in the or

2. a change to the updated or that NRC approval pursuant to 10 CFR

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet criteria of Specification 14b above be and approved by the prior to implementation.

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

15 Safety Function Determination Program (SFDP)

This program ensures of safety function is detected and appropriate taken. Upon entry into LCO 3.0.6, an evaluation shall be to determine if of safety exists. Additionally, appropriate actions as a result of the support system inoperability and corresponding exception to entering supported system Condition Required This program implements requirements of LCO The SFDP the following:

a. Provisions for cross checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go Units 1 and 2 2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

b. Provisions for ensuring the plant is maintained in a condition if a loss function condition

c. to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

of safety function exists assuming no concurrent failure, a function assumed in the analysis cannot be performed. the purpose this program, a safety function may exist a support system is inoperable, and:

a. A required system redundant to the system(s) supported by inoperable support system is inoperable; or

b. A required redundant to the system(s) in turn supported by the inoperable supported system is inoperable; or

c. A required system redundant to the support system(s) for supported (a) and (b) above is also inoperable.

The SFDP identifies where a of function If a loss of function is determined to exist by this program, the appropriate Conditions and Required Actions the in which loss of safety function are required to entered.

5.5.16 The main from the rigid anchor points of containment penetrations downstream and including the main steam header shall be The extent of the inservice examinations completed during each inspection interval (IWA 2400, Code, 1974 Edition, Section XI) shall provide 100 percent volumetric examination of circumferential and longitudinal pipe welds to the The areas subject to examination are defined in accordance with examination category for 2 piping welds in IWC-2520.

5.5.17 A program shall to implement the rate of containment as required by 10 CFR (0) and 10 CFR Appendix J, Units 1 and 2 3 Amendment No. (Unit 1)

Amendment No. (Unit

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 Containment Leakage Rate Testing Program (continued)

Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J":

Section 9.2.3: The next Type A test, after the March 1994 test for Unit 1 and the March 1995 test for Unit 2, shall be performed during refueling outage R22 (Spring 2009) for Unit 1 and during refueling outage R20 (Spring 2010) for Unit 2. This is a one time exception.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 43.8 psig.

The maximum allowable containment leakage rate, La, at Pa, is 0.15% of containment air weight per day.

Leakage rate acceptance criteria are:

a. Containment overall leakage rate acceptance criterion is ::; 1.0 La. During plant startup following testing in accordance with this program, the leakage rate acceptance criteria are ::; 0.60 La for the combined Type Band C tests, and ::; 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1. Overall air lock leakage rate is ::; 0.05 La when tested at ~ Pa .

2. For each door, leakage rate is ::; 0.01 La when pressurized to ~ 10 psig.

c. During plant startup following testing in accordance with this program, the leakage rate acceptance criterion for each containment purge penetration flowpath is ::; 0.05 La.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

(continued)

Farley Units 1 and 2 5.5-14 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals 5.5.18 Control Room Integrity Program (CRIP)

A Control Room Integrity (CRIP) shall established and implemented to ensure that the control room integrity is maintained such that a radiological or a (e.g., byproducts, halon, will not prevent the control room operators from controlling the reactor during normal or conditions. program shall require testing as outlined below. should be performed when changes are made to structures, and components which could impact Control Room Impact integrity. systems components may internal or external to the CRE. Testing should also conducted following a modification or a that could affect Testing should also be if the conditions associated with a particular challenge result in a change in operating mode, alignment or system response that could result in a new limiting condition. Testing should commensurate with the type modification or Testing should be conducted in alignment that results in to the operators.

A CRIP shall established to implement the following:

a. Demonstrate, using Regulatory Guide (RG) 1.197 and ASTM , that CRE is than the below The below do not include 10 cfm assumed in accident analysis for ingress I egress.

i) cfm when the control room ventilation systems are aligned in the emergency recirculation mode operation, ii) 600 when the control room ventilation systems are aligned in the isolation of operation, and iii) cfm when the control room ventilation systems are aligned in the normal mode of operation;

b. Demonstrate that leakage characteristics of the will not result in simultaneous loss reactor control from the control room the hot shutdown panels;

c. Maintain a configuration control and a design and licensing control program and a preventative maintenance program. As a minimum, the CRE configuration control program will determine whether i) CRE differential pressure relative to adjacent areas and ii) the control room ventilation system flow as determined in with ASME N51 989 or ASTM E2029-99, are consistent with the values measured at time the ASTM E741 test was performed. If item i or ii has changed, how this has affected of If there has degradation in inleakage characteristics of Units 1 and 2 5 Amendment No. (Unit 1)

Amendment No. (Unit

Programs and Manuals 5,5 and Manuals 5.5.18 CRE since E741 test, a determination should be made whether licensing basis analyses remain valid. If the licensing basis analyses remain valid, the CRE

d. the CRE in accordance with the testing methods and at the frequencies specified in RG 1.197, Revision 0, May 2003.

The provisions of SR are applicable to control room inleakage testing frequencies.

5.5.19 program provides controls Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed intervals sufficient to assure the Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of Surveillance Requirements which the Frequency is controlled by the program.

b. Changes to the Frequencies in the Surveillance Frequency Control Program shall be made in accordance with NEt 04-10, "Risk-Informed Method for Control Surveillance Frequencies," Revision 1.

c. provIsions Surveillance Requirements 3,0.2 and 3.0.3 are applicable the Frequencies established in the Frequency Control Program.

Farley Units 1 and 2 5.5-16 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6.9 Deleted 10 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 shall include:

a. The of inspections on SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation

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

e. Number of plugged during inspection outage for each degradation mechanism,

f. The number and percentage of tubes plugged to date, and the effective plugging in

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

11 The NRC shall notified if the MC source is out of service for greater than 10 days.

Farley Units 1 2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Application to Revise Technical Specifications to Adopt TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using The Consolidated Line Item Improvement Process 5

Technical Specification Change - Clean-typed Pages (VEGP)

SG Tube Integrity 17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube "":;,,,.-.nl 3.4.17 tube integrity shall be maintained.

AND All satisfying the tube plugging criteria shall be plugged in accordance with Steam Generator Program.

APPLICABILITY: MODES 1, and 4.

ACTIONS Condition entry is allowed CONDITION ACTION COMPLETION TI or more tubes A.1 Verify tube integrity of 7 days satisfying tube affected tube(s) is maintained plugging criteria and not the refueling outage plugged in accordance or SG tube inspection.

with the Steam Generator Program.

Plug the tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator next refueling outage or SG tube inspection B. Required Action and in MODE 3. 6 associated Completion Time Condition A not met.

Be in MODE 5. hours tube integrity not maintained.

Vogtle Units 1 and 2 3.4.17-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Tube Integrity 3.4.17 SURVEILLANCE SURVEILLANCE FREQUENCY SR 17.1 SG tube integrity in accordance with the In accordance with Program. Steam Program 3.4.17.2 Verify that inspected tube that Prior to entering the tube plugging criteria is plugged in MODE 4 following a with the Generator Program. inspection Vogtle Units 1 2 17-2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals A Steam Generator Program shall be established and implemented to ensure tube integrity is maintained. In addition, the Steam Program shall include the following:

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

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

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

1. Structural integrity performance criterion: All in-service steam tubes shall retain structural integrity over full 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. This includes retaining a safety factor of 3.0 against burst under normal steady 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 Apart from the requirements, additional loading conditions associated with the design basis accidents, or combination of in accordance with the design and licensing shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment tube integrity, those loads that do significantly affect burst or collapse shall 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.

Vogtle Units 1 2 5.5-7 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals

=::"::::':"';";""';:::::"::::";":":::":"':='--l,..:::::..=.......,...:...;:;..;;o",-=-,-, (continued)

Accident induced leakage performance The primary to secondary accident induced leakage rate for any design basis accident, other a rupture, shall not the assumed in the accident analysis in terms of total leakage rate for and rate for an individual is not to exceed 1 gpm per SG.

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

c. Provisions for tube plugging 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 plugging criteria shall be applied as an alternative to the 40% depth based criteria:

Tubes with service-induced flaws located greater than 15.2 inches below top of the tubesheet do not plugging. Tubes with service-induced flaws located in the portion of the tube from the top of tubesheet to 15.2 below the top of the shall be plugged upon detection.

Vogtle Units 1 and 2 5.5-8 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals and Manuals 5.5.9 ==,-,..:.....:::=.:...:.=,-",=....t==-.t...:....:..;== (continued)

d. Provisions for Periodic 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 volumetric flaws, axial and circumferential cracks) that may be present along the length of from tube-to-tubesheet weld the tube inlet to the tube-to-tubesheet weld at the tube outlet, and may satisfy the applicable tube plugging criteria. Portions of the tube below 15.2 inches below the top ofthe tubesheet are from this requirement.

Vogtle Units 1 and 2 Amendment No. (Unit 1)

Amendment No. (Unit

Programs Manuals 5.5 and Manuals

=::"=:"':":""';;;:=':";:";::;";"':=:""'>"'="-=l-!-:...::..;Jc:.=..:...: (co nti nued)

The tube-to-tubesheet weld is not part of the tube. In addition to the requirements of d.1, d.2, and below, the inspection scope, inspection methods, inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation shall be performed to determine the type and location flaws to which the tubes may be susceptible and, based on this to determine which inspection methods need to employed and at what locations.

1. Inspect 100% of in during the first refueling outage following installation.

2. refueling outage following SG installation, inspect at 48 full power months or least other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall the number of tubes in all divided by number of inspection outages scheduled in inspection period as in a, b, and c below. If a the potential for a type of degradation to occur at a location not previously inspected with a technique of detecting type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to inspected for this potential type of degradation at this location at the end of the inspection period shall be no than the of the number of times SG is to inspected in inspection period the determination that a new form of degradation could potentially be occurring this location divided by total of the SG is scheduled to be inspected in inspection period. inspection period defined below may extended up to 3 effective full power months to include a inspection outage in an inspection period and the subsequent inspection period begins at the conclusion the included a) After first refueling outage following 100% of the tubes during the next 120 full power months. This constitutes the first inspection period.

b) During the next 96 effective full power months, 100% of the tubes. This constitutes the second inspection period; and Vogtle Units 1 and 2 o Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 U,.,..,...,.-:>,...,,,,, and Manuals 5.5.9 (continued) c) During the remaining of the inspect 100% of the tubes every effective full power months. This constitutes third and subsequent inspection periods.

3. If crack indications are found in portions of then the next inspection for and potentially affected for the degradation mechanism that caused the crack indication shall not 24 full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic nondestructive testing, or engineering evaluation indicates that a crack-like indication is not with a crack(s), then the indication need not treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

10 This program provides controls for monitoring secondary water chemistry to inhibit tube degradation. The program shall include:

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

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

c. Identification of nrnf't:>co sampling points;

d. for recording management of

e. Procedures defining corrective actions for off control point chemistry

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

Vogtle Units 1 and 2 1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals 11 A program shall established to implement the following required of Engineered Safety Feature 'filter ventilation systems at the frequencies in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1980:

a. Demonstrate for of the that an inplace of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass S 0.05% when in with Regulatory Guide 1.52, Revision 2, and ASME N510-1980 at the system flow rate specified below

+/- 10%.

ESF Ventilation System Flow Control Room Emergency Filtration System (CREFS) 19,000 CFM Uonai',."'t ...... n Area Filtration (PPAFES) 15,500 CFM

b. Demonstrate for each of the systems that an inplace test of the charcoal adsorber shows a penetration and system bypass S 0.05% when with Regulatory Guide 1.52, Revision 2, and the system flow rate specified below +/- 10%.

Ventilation System Flow CREFS 19,000 CFM 15,500 CFM

c. Demonstrate for each of the systems that a laboratory of a of charcoal adsorber, when as described in Regulatory Guide 1 Revision 2, shows the methyl iodide penetration than or equal the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30"C than or equal to the relative humidity specified below.

Ventilation System Penetration RH

.2% 70%

10% 95%

Vogtle Units 1 and 2 5.5-12 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 5.5 Manuals 11 (continued)

d. Demonstrate for the that the combined HEPA filters, the charcoal adsorbers, and cooling coils is less than specified below when tested in accordance with Regulatory Guide 1 Revision and ASME N510-1989 at the system flow rate specified below +/- 10%.

Ventilation System Delta P Flow Rate 7.1 19,000 CFM water gauge 6 in. 15,500 CFM gauge

e. Demonstrate that the heaters for the dissipate 95 kW when corrected to 460 V in with ASME N510-1989.

provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.5.12 This program provides controls for potentially explosive gas mixtures contained in the Waste the quantity of radioactivity in each Gas Decay Tank, and the quantity of radioactivity contained in unprotected outdoor liquid The radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated to Waste System Leak or Failure." The liquid radwaste quantities shall be limited to 10 curies per outdoor tank in accordance with Standard Review Section 15.7.3, "Postulated Radioactive Release due to Tank Failures."

program shall include:

a. The limits for concentrations hydrogen oxygen in the Gaseous Waste System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria whether or not system is to withstand a hydrogen explosion);

Vogtle Units 1 and 2 3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 5.5 and Manuals 12

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas tank is less than amount that would result in a whole body exposure of:2: 0.5 rem to any individual in an area, in event of an uncontrolled release of the tanks' contents; and

c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment is limited to s; 10 curies tank, excluding tritium and dissolved or entrained noble This surveillance program provides assurance that in the event of an uncontrolled of tank's contents, the resulting concentrations would less than the limits of 10 CFR 20, Appendix Table 2, Column 2, at the potable water supply and the nearest surface water supply in an unrestricted area.

The provisions of 3.0.2 and SR 3.0.3 are applicable to and Storage Tank Radioactivity Monitoring Program surveillance 13 A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining fuel oil

1. an API gravity or an absolute specific gravity within limits, or an API gravity or specific gravity within limits when compared to the supplier's certificate;

2. a flash point within limits for ASTM 2D fuel oil, and, if gravity was not determined by comparison with supplier's certification, a kinematic viscosity within limits for ASTM 2D fuel oil; and

3. a clear and bright appearance with proper color.

b. Other properties for ASTM 2D fuel oil are within limits within 30 days following sampling and addition to storage tanks; and

c. Total particulate concentration of the fuel oil is s; 10 mgtl when tested 31 Vogtle Units 1 and 2 5.5-14 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 5.5 Manuals 5.5.13 provisions of SR 3.0.2 and SR 3.0.3 are applicable to Oil Testing Program surveillance frequencies.

5.5.14 program provides a means for processing changes to the of these Technical Specifications.

a. Changes to the of the shall be made under appropriate administrative controls and reviews

b. may make to without prior NRC approval provided the do not require either of the following:

1. a change in the incorporated in the license; or

2. a change to the updated FSAR or that requires NRC approval pursuant to 10 CFR 50.59.

a. Control Program contain provisions to ensure that Bases are maintained consistent with the

b. Proposed changes that meet the criteria of (b) above shall be reviewed and approved the NRC prior to implementation. to the implemented without prior NRC approval shall be provided to NRC on a frequency consistent with 10 CFR 50.71 (e).

15 This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into 3.0.6, an evaluation made to if of function exists. Additionally. other appropriate actions may taken as a of the support system inoperability and corresponding exception to entering supported system Condition Required Actions. program implements the requirements of LCO 3.0.6. SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss the capability to perform the safety function assumed in the accident analysis does not go undetected;

b. Provisions for ensuring the plant is maintained in a condition if a loss of function condition Vogtle Units 1 and 2 5.5-15 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals and Manuals 15 Safety Function Determination Program (SFDP) (continued)

c. Provisions to ensure that an inoperable supported system's Completion is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

A loss safety function when, no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP where a of safety function If a of function is determined to exist by this program, the appropriate Conditions and Required Actions of the in which the of safety function are required to entered.

16 MS and FW Piping Inspection Program program shall provide for the inspection of four Main Steam and Feedwater lines from the containment penetration flued head outboard welds, up to the first five-way restraint. extent of the inservice examinations completed during each inspection interval (ASME Code Section XI) shall provide 100%

volumetric examination of circumferential longitudinal welds to extent practical. This augmented inservice inspection is consistent with the requirements NRC Branch Technical Position MEB 3-1, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment," November 1975 and Section 6.6 the FSAR.

17 Containment Leakage Rate Testing Program A program shall established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 50, Appendix J, Vogtle Units 1 and 2 6 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5 .17 Containment Leakage Rate Testing Program (continued)

Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Testing Program," dated September 1995, as modified by the following exceptions:

1. Leakage rate testing for containment purge valves with resilient seals is performed once per 18 months in accordance with LCO 3.6.3, SR 3.6.3.6 and SR 3.0.2.

2. Containment personnel air lock door seals will be tested prior to reestablishing containment integrity when the air lock has been used for containment entry. When containment integrity is required and the air lock has been used for containment entry, door seals will be tested at least once per 30 days during the period that containment entry(ies) is (are) being made.

3. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief or alternative has been authorized by the NRC. At the discretion of the licensee, the containment concrete visual examinations may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenance/refueling outage.

4. A one time exception to NEI 94-01, Rev. 0, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J":

Section 9.2.3: The next Type A test, after the March 2002 test for Unit 1 and the March 1995 test for Unit 2, shall be performed within 15 years.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, Pa, is 37 psig.

The maximum allowable containment leakage rate, La, at Pa, is 0.2% of primary containment air weight per day.

Leakage rate acceptance criteria are:

(continued)

Vogtle Units 1 and 2 5.5-17 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and 5.5 and Manuals 5.5.17 =.!..!.!.:l:<==~==~:""::'::::'=-:"'=~::L-!...":"'::::'::1!..:::::.!..!..!. (continued)

a. Containment overall leakage acceptance criteria are ~ 1.0 During the first unit startup following testing in with this program, the leakage rate acceptance criteria are' ~ 0.60 for the combined Type B Type C tests, ~ 0.75 for Type A tests;

b. Air lock testing are:

1) Overall air lock lealKaale rate is ~ when ? Pa ,

2) For each the leakage is ~ 0.01 when pressurized to

? Pa .

3.0.2 do not apply to the frequencies specified in the L..O:;::.::lI'l.CIUO:;:: Rate Program.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Testing Program.

18 The Configuration Risk Management Program (CRMP) provides a risk-informed assessment to manage associated with equipment inoperability. The program applies to technical specification structures, or components for which a risk-informed allowed outage time has been granted. The program shall include the following elements:

a. Provisions for control and implementation of a 1 at power internal events PRA-informed methodology. The assessment shall be capable of evaluating applicable plant configuration.

b. Provisions for performing an assessment prior to entering LCO Condition for preplanned activities.

c. Provisions for performing an as~seSiS after entering the LCO Condition for unplanned entry into the Condition.

d. Provisions for assessing for additional actions of additional equipment out of service conditions while in

e. Provisions for considering other applicable risk significant contributors such as Level 2 and events, qualitatively or quantitatively.

Vogtle Units 1 and 2 8 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 and Manuals 19 This program provides for restoration and maintenance, based on recommendations of Standard 995, Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," of following:

a. Actions to restore battery cells with float voltage < 2.13 V, and

b. Actions to had been discovered with electrolyte 5.5.20 A Control Room Envelope Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System occupants can control the reactor safely under normal conditions maintain it in a condition following a radiological event, chemical or a smoke challenge. program shall ensure that adequate radiation protection is provided permit access occupancy under accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of body for the duration of the accident. program shall include the following elements:

a. The definition and CRE boundary.

b. Requirements for the boundary in its condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into CRE in accordance with the methods at Frequencies specified in Sections C.1 and C.2 Regulatory 1.1 "Demonstrating Control Room Envelope Integrity at Nuclear Power " Revision 0, May 2003, and (ii) assessing habitability at the Frequencies specified in Sections 1 and of Regulatory Guide 1.1 Revision O.

d. Measurement, at designated of the relative to all external areas adjacent to the CRE boundary during pressurization mode of operation by one train of the CREFS, operating at flow by the VFTP, at a Frequency of 18 months on a STAGGERED BASIS. The results shall trended and as part of the 18 month of the boundary.

Vogtle Units 1 and 2 9 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals and Manuals 5.5.20 "::::'=:"':"::':"'=-:;;';;::;:::::::='::-=-==:::::===..L....:....':::::::'=:..:..! (contin ued)

e. The limits on unfiltered air inleakage into the CRE.

limits shall be stated in a manner to allow direct comparison to the

.",t.lt",..,,:,,',! air inleakage measured the testing described in paragraph c.

The unfiltered air inleakage limit for radiological is the inleakage flow assumed in the licensing basis analyses of DBA Unfiltered inleakage for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the basis.

f. provisions of 3.0.2 are applicable to the Frequencies for habitability, determining unfiltered and measuring pressure CRE boundary as required by paragraphs c and 5.5.21 This program controls for The program ensure that Surveillance Requirements cn£>f'.t.ol'l Specifications are performed intervals to assure as~;oclatE~a Limiting Conditions for Operation are met.

a. Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the is controlled by the program.

b. to in the Surveillance Frequency Control Program shall made in with I 04-10, "Risk-Informed Method for of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.3 are applicable to the Frequencies established in Surveillance Frequency Control Program.

Vogtle Units 1 and 2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6 6.9 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 in with the Specification Generator (SG) Program. report shall include:

a. The scope of inspections performed on

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

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

e. Number of tubes plugged during inspection outage for each degradation mechanism,

f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator,

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

h. The primary to secondary LEAKAGE rate observed in each (if it is not practical to LL ......",...,U', to an individual SG, primary to secondary LEAKAGE should be conservatively assumed to from one SG) during the cycle preceding the inspection which is the report; and

i. calculated accident induced from the portion of tubes below 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 rate from most limiting accident is less than 2.48 times the maximum operational primary to secondary leakage the report should how it was determined.

j. The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective shall Vogtle Units 1 and 2 5.6-6 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant* Units 1 and 2 Application to Revise Specifications to Adopt TSTF-S1 "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using The Consolidated II Item Improvement jJro'ce!)S Enclosure 6 Proposed Technical Specification Bases Change (for information only) (fNP)

Tube Integrity B 17 BACKGROUND processes used to the SG performance are defined (continued) by the Generator Program Guidelines (Ref. 1).

APPLICABLE tube rupture (SGTR) limiting SAFETY ANALYSES design event for tubes and avoiding an is the basis for this Specification. analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate to the operational LEAKAGE rate limits in LCO 3, "RCS LEAKAGE,"

plus the leakage associated with a double-ended rupture of a tube. accident analysis for a SGTR assumes the contaminated secondary fluid is released via the main steam safety valves. majority of the activity to the atmosphere from the tube rupture.

analysis for design basis accidents and transients other than a assume the SG retain their structural integrity (Le., they are assumed not to rupture.) In these analyses, the discharge to atmosphere is total primary to secondary LEAKAGE from all of 1 as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of EQUIVALENT 1-131 is to be equal to 3.4.16, Specific Activity," limits. accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The consequences of events are within the limits of GOC 19 10 CFR 100 3) or the NRC approved licenSing (e.g., a small fraction these limits).

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

LCO maintained.

the fefla+F-j~~~CI Generator Program.

During an SG inspection, any inspected tube satisfies the Steam Generator Program is removed from service by plugging. If a tube was determined to satisfy the feFlaIf-ILJJ,L9JJ~

criteria but was not plugged, the may still Units 1 and 2 B 3.4.1 Revision 24 I

SG Tube Integrity B 17 BASES LCO induced leakage includes any primary to secondary (continued) existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

operational LEAKAGE performance criterion provides an indication SG tube conditions during plant operation.

limit on operational LEAKAGE is contained in LCO 3.4.13, Operational " and limits primary to secondary LEAKAGE through anyone SG to 1 gpd. This limit is based on the assumption that a crack leaking this amount would not propagate to a SGTR under stress conditions of a LOCA or a steam line break. If this amount LEAKAGE is to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across tubes is large. differential pressures across tubes can only experienced in MODE 1,2,3, or RCS conditions are far challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential is low, resulting in lower and reduced potential for LEAKAGE.

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

Condition A applies if it is discovered that one or more examined in an inservice inspection satisfy the tube ~ttttf--J,W"lW!!JU, criteria but were not plugged in accordance with the Steam Generator Program as required by 3.4.1 An evaluation of SG tube integrity of the Farley Units 1 2 B 17-4 Revision 24 I

Tube Integrity B 3.4.17 BASES ACTIONS :....::.:..:"'-=.:.::::....::...:.:.:::::: (continued) tube(s) must on meeting the performance criteria Generator Program. The define limits on tube degradation that allow for flaw growth between while still providing assurance that the performance f'rIT,::;rl~ 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 that demonstrates that the SG performance criteria will continue to met until the next tube inspection. tube 1'>,.0',.. ..,1'\ determination is on the estimated of the the time the situation is discovered and growth the degradation prior to the next SG tube inspection. If it is that tube integrity is not being maintained, Condition B A Completion Time of 7 is sufficient to complete evaluation while minimizing the risk of plant operation with a tube that may not integrity.

If the determines the affected tube(s) tube integrity, Required Action allows plant operation to continue until the next refueling outage or inspection provided the inspection interval continues to be supported by an operational that affected tubes. However, the affected tube(s) must be plugged prior to entering 4 following the next refueling outage inspection. This Completion Time is acceptable until the next is supported by If the Actions and Completion of Condition A are not met or if integrity is not maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion are reasonable, based on f'\n~lr",tinn experience, to reach the desired plant conditions from full power conditions in an orderly manner without challenging plant systems.

Farley 1 and 2 B 3.4.17-5

SG Tube Integrity B 3.4.17 BASES SURVEILLANCE SR 3.4.17.1 REQUIREMENTS 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 previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair Q.\.uqqing 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 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 3.4.17.1. 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 5.5.9 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indjcations are found in any SG tube. the maximumjas.pection interval for all affected and potentially affected SGs is restrictedJ2y Specification 5.5.9 until subsequent inspections support extending the inspection interval.

SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair Qluqqing criteria is removed from service by plugging. The tube Dlugging~ criteria delineated in Specification 5.5.9 (continued)

Farley Units 1 and 2 B 3.4.17-6 Revision ~

Tube Integrity B 7 SURVEILLANCE SR 3.4.17.2 (continued)

REQUIREMENTS are intended to ensure that accepted for continued satisfy the SG performance criteria with allowance for error in the flaw measurement and for future flaw growth. addition, the tube in conjunction with other elements of the Steam Generator Program, ensure that the performance criteria will continue to be until the next inspection of subject tube(s).

Reference 1 and Reference 6 provide guidance for performing operational assessments to verify that tubes remaining in service will continue to meet the performance I'I',t.c>,.,,,,

Frequency of "Prior to entering MODE 4 following a SG inspection" ensures that the Surveillance has been completed and all tubes the are plugged prior to subjecting the tubes to significant primary to secondary pressure differential.

1. 97-06, Generator Program Guidelines."

2. 10 CFR Appendix A, GDC 19.

3. 10 100.

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

5. Draft Regulatory Guide 1.1 , "Basis for Plugging Degraded Steam Generator Tubes," August 1

6. RI TR-1 "Pressurized Water Reactor Steam Generator Examination Guidelines."

Farley Units 1 and 2 B 3.4.17-7 Revision

Joseph M. farley Nuclear Plant - Units 1 and 2 VogUe Electric Generating Plant - Units 1 and 2 Application to Technical Specifications to Adopt 0, "Revision to Steam Generator Program Inspection frequencies and Tube Sample Selection," Using The Consolidated Une Item Improvement Process Enclosure 7 Proposed Technical Specification aal:;es Change (for information only) (VEGP)

SG Tube Integrity B 3.4.17 BASES (continued)

APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting SAFETY ANALYSES design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a 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 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 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 !'JRC 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 pluaallli;bcriteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair Qjugging criteria is removed from service by plugging. If a tube was determined to satisfy the repair plugging 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. Portions of the tube below 15.2 inches below the top of the (continued)

Vogtle Units 1 and 2 B3.4.17-2 REVISION2d

SG Tube Integrity 83.4.17 8ASES ACTIONS A.1 and A.2 (continued)

Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair pluggil}-f}

criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair pluggimLcriteria 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, Condition 8 applies.

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 A.2 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 entering MODE 4 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.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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.

(continued)

Vogtle Units 1 and 2 83.4.17-5 Rev. 09/06

SG Tube Integrity B 3.4.17 BASES (continued)

SURVEILLANCE SR 3.4.17.1 REQUIREMENTS 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 previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair plugging criteria. Inspection scope (i.e., 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 specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.17.1.

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 5.5.9 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, llie..JIlaximLJminspection interval for all afutcted and potentially affecJad.

SGs is restricted by Specificatio..o...5.,.5.9 until subsequent inspecti~

support extending the inspection interval SR 3.4.17 .2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program Fej3aff-plugging criteria is removed from service by plugging. The tube repair fllu9.9lmLcriteria delineated in Specification 5.5.9 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube ropair Rlu£Lqlng

( continued)

Vogtle Units 1 and 2 B 3.4.17-6 Rev. 0 9106

SG Tube Integrity B3.4.17 BASES SURVEILLANCE SR 3.4.17.2 (continued)

REQUIREMENTS 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.

The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair plugging 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."

7. License Amendment Nos. 167 and 149, "Vogtle Electric Generating Plant, Units 1 and 2, Issuance of Amendments Regarding Revision to Technical Specifications 5.5.9, "Steam Generator (SG) Program," and 5.6.10, "Steam Generator Tube Inspection Report," (TAC Nos. ME8313 and ME8314),"

September 10, 2012.

Vogtle Units 1 and 2 B 3.4.17-7 REVISION~ I

v t e Letter Sequence Request
TAC:MF0562, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions Answers Results Approval Other:
MONTHYEAR2013-01-30 [Table View]

NL-13-0081, Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,Using the Consolidated Line Item Improvement...

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NL-13-0081, Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,Using the Consolidated Line Item Improvement...

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