W3F1-2006-0016, To Amendment Request NPF-38-262, Steam Generator Tube Inservice Inspection Program

To Amendment Request NPF-38-262, Steam Generator Tube Inservice Inspection Program

ML061250173
Person / Time
Site:	Waterford
Issue date:	05/03/2006
From:	Tankersley T Entergy Nuclear South, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TSTF-449, Rev 4, W3F1-2006-0016
Download: ML061250173 (40)

Text

asI a=- Entergy Entergy Nuclear s3outh Entergy Operations. Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6780 Fax 504-739-6698 ttanker@entergy.com Tom Tankersley Acting Director, Nuclear Safety Assurance Waterford 3 W3Fl -2006-0016 May 3, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

REFERENCES:

Supplement 2 to Amendment Request NPF-38-262 Steam Generator Tube Inservice Inspection Program Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

1.

Entergy letter dated March 15, 2005, License Amendment Request NPF-38-260 Proposed Technical Specification Change Regarding Tubesheet Inspection Depth for Steam Generator Tube Inspections (W3F1 -2005-0009)

2.

Entergy letter dated July 21, 2005, License Amendment Request NPF-38-262 Proposed Technical Specification Change to Waterford-3 Steam Generator Tube Inservice Inspection Program Using Consolidated. ine Item Improvement Process (W3F1 -2005-0040)

3.

Entergy letter dated February 15, 2006, Supplement to Amendment Request NPF-38-262 Steam Generator Tube Inservice Inspection Program (W3Fl-2006-0007)

4.

Entergy letter dated March 22, 2006, Tubesheet Inspection Depth for Steam Generator Tube Inspections Waterford Steam Electric Staticin, Unit 3 (\\A'3F1-2006-0008)

Dear Sir or Madam:

By letter (Reference 2), Entergy Operations, Inc. (Entergy) proposed a change to the Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical Specifications (TSs) to replace the existing steam generator tube surveillance program with that being proposed by the Technical Specification Task Force in TSTF 449, Revision 4.

On January 3, 2006, Entergy received an NRC Staff Request for Additional Information (RAI) to support the review of the proposed change. On January 19, 2006, Entergy and members of your staff held a call to clarify the additional information requested and discuss an extension to the Entergy RAI response from 30 days to 45 days. On February 15, 2006, Entergy provided a response to the RAI (Reference 3).

1A0F

W3F1I-2006-0016 Page 2 On April 17, 2006, Entergy received a second NRC Staff RAI dated March 31, 2006 to support the review of the proposed change. On April 25, 2006, Entergy discussed with members of your staff our desire to have the proposed TSTF-449 modeled TS change (Reference 2 supplemented by Reference 3) approved prior to the tubesheet inspection depth proposed TS change (C*) to simplify and expedite the review process. This decision will necessitate the removal of references to C* from the proposed specification. RAI questions related to the C*

will be addressed in the proposed C* TS change (Reference 1 supplemented by Reference 4).

Additionally, Entergy had included the already approved welded sleeve alternate repair method from the existing TSs to the proposed TSs in accordance with TSTF-449. However, due to subsequent NRC staff questions related to the use of and inspection techniques for the sleeving repair methodology and with this repair method not being applied at Waterford-3, Entergy will remove this method from this proposed TS change. Entergy's response to this RAI is contained in Attachment 1.

Changes to the TS pages and TS Bases pages, which were originally submitted in Reference 2 and supplemented by References 3 and 4, are proposed. The revised mark-ups are included in Attachments 2 and 3. Note that marked up TS pages in Attachment 2 replace the pages provided in Attachment 2 of Reference 2 and supplemented by Attachment 3 of Reference :3 and Atlachment 6 of Reference 4 in their entirety. Note that marked up TS Bases pages in replace the pages provided in Attachment 4 of Reference 3 in their entirety.

The conclusions of the original no significant hazards consideration included in Reference 2 are not affected by any information contained in this supplemental letter. There are no new commil:ments contained in this letter.

if you have any questions or require additional information, please contact Steve Bennett or Ron Williarms at (479) 858-4626 and (504) 739-6255, respectively.

I declare under penalty of perjury that the foregoing is true and correct. Executed on May 3, 2006.

Sincerely, 1yrs le TET/RlW Attachments:

1. Response to Request for Additional Information

2. Revised Markup of Replacement Pages for All TS Pages

3. Revised Markup of Replacement Pages for All TS Bases Pages

W3F1 -:2003-0016 Page 3 cc:

Dr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U.S. Nuclear Regulatory Commission Attn: Mr. Mel B. Fields MS O-7E1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. 0. Box 4312 Baton Rouge, LA 70821-4312 American Nuclear Insurers Attn: Library Town Center Suite 300S 29th S. Main Street West Hartford, CT 06107-2445 To W3FI-2006-001 6 Response to Request for Additional Information

Attachment I to W3F1 -2006-0016 Page i of 7 Response to Request for Additional Information Question 1:

Currertly, no sleeves are installed in the Waterford Steam Electric Station, Unit 3 (Waterford-3) steam generators (SGs); however, proposed Technical Specification (TS) 6.5.9.f allows the use of sleeving (CENS Report CEN-605-P, "Steam Generator Tube Repair Using Leak Tight Sleeves"). It is the staff's understanding that the tubesheet sleeves, as described in CEN-605-P, have a nickel band in the area of the rolled joint. Based on interactions with other plants, it is not clear whether techniques currently exist to inspect the parent tube located behind (adjacent to) the nickel band for crack-like indications. If this is the case, it is not clear how you will implement proposed TS 6.5.9.d, which requires that the method of inspection should be capable of detecting flaws of any type that may be present along the length of the tube and that may satisfy the applicable tube repair criteria. In light of the above, either (a) discuss your plans for removing this sleeving method from your TSs, (b) provide information supporting the ability of an inspection technique to detect the forms of degradation that could occur in the parent tube adjacent to the nickel band and that may satisfy the applicable tube repair criteria, or (c) prcvide analysis and/or testing results which indicate that inspection of this region (i.e., behind the nickel band) is not needed.

Response 1:

Currently no sleeves are installed in the Waterford-3 SGs and plans are not to install any using this tube repair methodology. Therefore, this tube repair method will be removed from these proposed TS changes.

Question 2:

Proposed TS 6.5.9.d excludes from inspection the portion of each tube from the top support of the cold leg to the cold-leg tube end. This is inconsistent with the corresponding section of the Technical Specification Task Force (TSTF)-449 (5.5.9.d), which states the objective of tube inspection is to detect flaws of any type, 'from the tube-to-tubesheet weld at the tube inlet to the tube-tc-tubesheet weld at the tube outlet." Please discuss your plans to modify the proposed TS to make them consistent with TSTF-449.

Response 2:

As discussed in the cover letter regarding the agreement to have the TSTF-449 format change approved prior to the tubesheet inspection depth change, Entergy agrees to modify the proposed TS to make them consistent with the wording in TSTF-449, 5.5.9.d. The revised 7S pages for this proposed license amendment are contained in Attachment 2.

to W3F1 -2006-0016 Page 2 of 7 Question 3:

Proposed TS 6.5.9.d, states, "in addition to meeting the requirements of d.1 and d.2 below...."

To be consistent with TSTF-449, this should read "... requirements of d.1, d.2, and d.3 below,"

since your February 15, 2006 response to Request for Additional Information (RAI) question 3 added a paragraph that was missing from the original submittal. Please discuss your plans to modify the proposed TSs to make them consistent with TSTF-449. (Emphasis added by the staff.)

Response 3:

This was an editorial oversight in the last RAI response. A corrected page is being provided in Attachmnent 2.

Question 4:

Proposed TS 6.5.9.c addresses SG tube repair criteria. Since a tube is defined as the entire length of the tube, including the tube wall and any repairs to it, it could be construed that the 40% pugging limit is applicable to the sleeves. Please discuss your plans to incorporate the repair criteria for the sleeves into the specification. For example, In the region of a tube repaired in accordance with TS 6.5.9.f, the tube shall be plugged upon detection of any service-induced flaw in (a) the sleeve or (b) the pressure boundary portion of the original tube wall in the sleeve-to-tube joint.

Response 4:

In Entergy's response to RAI 9 received on the SG tubesheet depth submittal (Reference 4,l, Entergy made the following commitment:

If s'eeves are installed, Entergy plans to inspect inservice sleeves over their full length plus 5 inches beyond the sleeve-to-tube rolledjoint in the tube sheet in accordance with the requirements of the EPRI Guidelines using appropriate examination methodology. The tube shall be plugged upon detection of any service induced imperfection, degradation or defect in the sleeve or pressure boundary portion of the original tube wall in the sleeve-to-tube rolled joint. Entergy will periodically inspect sleeves as a minimum in accordance with the existing TS requirements As discussed in the response to RAI question 1 above regarding the removal of the sleeving tube repair method from the TSs, this commitment no longer applies and therefore will be withdrawn.

to W3F1 -2006-0016 Page 3 of 7 Question 5:

Proposed TS Bases Insert B-2 includes only the first sentence of a paragraph from the corresponding TSTF-449 insert (B 3.4.13B). Missing from the proposed Waterford insert is the following:

The Steam Generator Program operational LEAKAGE performance criterion in 14EI [Nuclear Energy Institute] 97-06 states, "The RCS [Reactor Coolant System] operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience wit, SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

In place of this paragraph you have a plant-specific discussion of operational leakage limits.

The staff recognizes that the 75 gallons per day (gpd) operational leakage limit at Waterforcl-3 ensures the radiological consequences will be limited to the appropriate regulatory limits.

However, this limit also reflects operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion (since it is less than 150 gpd through any one SG) in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of SG tube ruptures. Please discuss your plans for modifying your Bases to include the other reason for the operational leakage limit. The staff notes that from the Bases as currently proposed, one may incorrectly conclude 540 gpi is an appropriate operational leakage limit for a "faulted steam generator."

Response 5:

Entergy did not believe that the quote from NEI 97-06 added substantial value in light of the reduced operational leakage limit of 75 gpd. However, for completeness, Entergy will include the remainder of the TSTF-449 proposed TS Bases 3.4-133B into the Waterford-3 Bases insert for 3/4 4.5.2. The proposed Insert B-2 into Waterford-3 Bases will now read:

The primary to secondary leakage limit of 75 gallons per day through any one SG is based on the operational leakage performance criterion in NEI 97-06. The Steam Generator Program operational leakage performance criterion in NEI 97-06 states, 'The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The NEI 97-06 limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Prcgram is an effective measure for minimizing the frequency of steam generator tube ruptures.

Regarding the existing TS Bases 3/4.4.5.2, OPERATIONAL LEAKAGE, Entergy agrees that this discussion could lead one to believe that the assumptions in the accident analysis could be applicable to operational leakage based on the heading of the section. The same accident analysis assumptions are contained in Insert B-1 which is being included in the Bases for T'i 3/4.4.4. Therefore, Entergy will remove the existing discussion in TS Bases 3/4.4.5.2.

A revised TS Bases and Insert B-2 for 3/4.4.5.2 is contained in Attachment 3.

Attachment I to W3Fl -2006-0016 Page e of 7 Question 6:

In the Limiting Condition for Operation section of your BASES Section 3/4.4.4, "STEAM GENERATOR TUBE INTEGRITY", the reference to Regulatory Guide 1.121 is omitted frorr, the bullet dealing with the structural integrity performance criterion (i.e., where Subsection NB cf the Americ:an Society of Mechanical Engineers Boiler and Pressure Vessel Code is referenced).

Since Regulatory Guide 1.121 was used in the development of the structural integrity performance criterion, it is not clear why it is not referenced. Please discuss your plans to modify your proposal to address this comment.

Response 6:

The new structural integrity analysis that is being performed for Waterford-3 supersedes the!

typical analysis performed per draft RG 1.121 and therefore was not initially included. However, since tie structural integrity analysis incorporates approaches and methodologies from RG 1.121, the reference to draft RG 1.121 will be added to the Bases. A revised TS Bases Insert B-1 is contained in Attachment 3.

Question 7:

You included a commitment in Attachment 4 indicating all loads that can significantly affect burst or collapse will be determined and assessed. In this commitment, there is a statement that indicates: "These loads, as well as the other analyses to support a 40% plugging limit, will be analyzed for the Waterford-3 SG licensing basis. These analyses will be performed and documented under the requirements of 10 CFR 50.59."

The NRC staff is aware of the industry's efforts to assess the effects of non-pressure loads on tube integrity (structural and leakage integrity). These efforts include an assessment of whether changes are needed to the industry guidelines to ensure these loads are appropriately accounted for in tube integrity evaluations (i.e., in the methods used to determine whether the performance criteria have been exceeded).

However, your statements seem to imply that the on-going industry efforts may affect the 40%

tube plugging limit. The reason for this is not clear since the 40% plugging limit was developed with consideration of non-pressure loads (consistent with the guidance in Regulatory Guide 1.121). Please clarify the meaning of your commitment which should include a determinaticin of whether it is needed.

Response 7:

Entergy believes the intent of this commitment has been misinterpreted by the NRC. At the time of the submittal, Entergy had not performed the new structural integrity analysis to comply with NEI 97-06. The only intent of this commitment is to state that the structural integrity and plugging limit calculation would be completed prior to implementation of the TS amendment. It is believed that the analysis results can be incorporated into the Waterford-3 licensing basis under 10CFR.50.59 and should not require NRC review and approval.

to W3Fl -2006-0016 Page I5 of 7 Question 8:

A separate license amendment request to apply the C* inspection criterion at Waterford-3 was submitted on March 15, 2005 and is still under NRC staff review. This would require tube inspection to a depth of 10.4 inches below the top of the hot-leg tubesheet or hot-leg expansion transition, whichever is lower. If your C* amendment is approved before the TSTF amendment, it may be necessary to amend the specifications in your TSTF amendment. Similarly, if you desire approval of the TSTF amendment before approval of the C* amendment, it will be necessary to remove references to C* from the specifications.

The following question was included in RAI question 9 about your C* amendment proposal. The staff notes that this will need to be addressed before the C* criterion can be incorporated into your proposed TSs modeled after TSTF-449.

The Waterford[-31 technical specifications (4.4.4.4.b) currently allow installation of leak-tight sleeves according to CENS Report CEN-605-P. Since sleeves could extend into the tubesheet below the C* distance, the proposed technical specifications would not require an insp:ection of this portion of the sleeve (including the lower sleeve joint.) Sleeves were rot addressed in the testing and analysis used to justify excluding part of the tube from inspection (WCAP-16208-P, Rev. 1). What plans do you have to modify the technical specifications to ensure the lower ends of sleeves (i.e., those within the tubesheet below the C* distance) will be inspected?

Response 8:

As discussed in the cover letter and responses to RAI questions I and 4 above, the sleeving tube repair method will be removed from these proposed TS changes. Therefore, the need for inspection of these sleeves no longer applies.

Question 9:

In your proposed TS 3.4.5.2.c under OPERATIONAL LEAKAGE for the RCS, the primary-tc-secondary limit is 75 gpd per SG. The wording in TSTF-449 and in your proposed accident-inducei leakage performance criterion (TS 6.5.9.b.2) is "through any one" SG. Please discuss your plans for modifying your proposed TS to make the wording of your leakage limits fully consistent with your performance criteria and the TSTF. (Emphasis added by the staff.)

Response 9:

The term 'per SG' was used in several locations in the existing TSs and its usage was carried over to the proposed TSs. However, for consistency Entergy has made changes, where appropriate, to use the term "through any one SG". The appropriately revised TS pages are contained in Attachment 2 and the appropriately revised TS Bases pages are contained in.

to W3F -:2006-0016 Page 6 of 7 Question 10:

Proposed TS 6.5.9.b.3, the operational leakage performance criterion, refers to Limiting Condition for Operation 3.4.5.2 as "Operational Leakage." The wording used in your proposed TS 3.4.5.2 is "Reactor Coolant System Operational Leakage," and the TSTF-449 wording is "RCS Operational Leakage." Please discuss how you will modify your proposed TS to make them consistent with either your existing wording or the TSTF wording. (Emphasis added by the staff.)

Response 10:

"Reactor Coolant System" is used in Waterford-3 TS LCO 3.4.5.2 when referring to operational leakagD. Therefore, Entergy will correct references of "RCS operational leakage" or "operational leakage" to "Reactor Coolant System operational leakage". The appropriately revised TS pages are contained in Attachment 2 and the appropriately revised TS Bases pages are contained in Attachment 3.

Question 11:

In your February 15, 2006, response to RAI questionl, you proposed changes to the ACTION section of TS 3/4.4.4, "Steam Generator (SG) Tube Integrity." Paragraph a.1 of the proposed insert Estates:

Wilhin 7 days verify tube integrity of the affected tube(s) is maintained until the next inspection, (Emphasis added by the staff.)

The corresponding section of the TSTF states:

Wilhin 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection. (Emphasis added by the staff.)

The TSTF wording could eliminate the need to shut down the facility in the event that tube integrity is only maintained until a refueling outage and not until the next SG tube inspection.

Please discuss your plans to revise your proposed TS to make them consistent with the TSTF Response 11:

Action a.1 of Insert 2 for TS 3/4.4.4 will be revised to:

Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or 'SG tube inspection.

The appropriately revised TS pages are contained in Attachment 2.

to W3Fl-:2006-0016 Page 7 of 7 Question 12:

On page 6 of 8 of Attachment 4 in your February 15, 2006 RAI response, the final bullet under "Limiting Condition for Operation" discusses operational leakage. The staff notes there appears to be an unnecessary bracket in the next-to-last sentence between "SGTR" [steam generator tube rupture] and "under." Please delete this bracket, or provide the missing information and closing bracket you intended to include.

Response 12:

The bracket has been removed and the revised Insert B-1 for TS Bases 3/4.4.4 is contained in.

To W3F1 -2006-0016 Revised Markup of Replacement Pages for All TS Pages to W3F1 -2006-0016 Page 1 of 19 [Replacement Page)

DEFINITIONS IDENTIFIED LEAKAGE (Continued)

b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c.

Reactor Coolant System leakage through a steam generator to the secondary system.

MEMBER(S) OF THE PUBLIC 1.15 MEMBER(S) OF THE PUBLIC means any individual except when that individual is receiving an occupational dose.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.16 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be Included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specification 6.9.1.7 and 6.9.1.8.

OPERABLE -

OPERABILITY 1.17 A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.18 An OPERATIONAL MODE (i.e. MODE) shall correspond to any cne inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.2.

WATERFORD - UNIT 3 1 -4 Amendment No. 68-84, 116 to W3Fl -2006-0016 Page 2 of 19 [Replacement Page]

D)EFINITIONS PHYSICS TESTS

1.19 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and i1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Comission.

PLANAR RADIAL PEAKING FACTOR - F

1.20 The PLANAR RADIAL PEAKING FACTOR is the ratio of-the peak to plane average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.

11RESSURE BOUNDARY LEAKAGE 0-tF0V

L.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (exce t5

%ta34b leakage) through a non isolable fault in a Reactor Coolant System component lxdy, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM (PCP)

L.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that process-ling and packaging of solid radioactive wastes based 6n demonstrated processing (if actual or simulated wtt solid wastes will be accomplished in such a way as 1.o assure compliance with 10 CFR Parts 20, 61, and 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

!,URGE -

PURGING 1.23 PURGE or PURGING shall be the controlled process of discharging air or Sias from a confinement to maintain temperature, pressure, humidity, concentra-1.ion or other operating condition, in such a manner that replacement air or gias is required to purify the confinement.

WATERFORD -

UNIT 3 1-5 Amendment No. b8 to W3F1 -2006-0016 Page 3 of 19 [Replacement Page]

REiCTOR COOLAT STE iQ[-4 4 STEAM GFNFRATORR70S5?

-J-E I N rey-6 y LI14ITING CONDITION FOR OPERATION 4.3.

1 E5ah steam gemerate-shel! he orE*eEAD.&dT APPLICABILITY:

MODES 1,.2, 3, and 4.

ACTrION:

XL'h An"

^r more stae g'paorr irgoperabl.

rstri 4hp iroporublo r

1 ZR "AeFrtor(6) to OPE__

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-at p

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lon F.

SURVEILLANCE REQUIREMENTS 1-A A

.^

1

_+

-,,1o A%

e-sAftnX EDDR 2^-f,,

I. 4. It.u V

c st m generatur shall be demo 1iLraI.U urLMnMLL u VvI5I-r1G16il-C up the following gmented inservice inspec n program.

4.4.4.1 ear Generator Each steam generator shall determined OPERABLE dur shutdown by sele ing and inspecting at lea the minimum number of st m generators spec ied in Table 4.4-1.

,4 4.4.2 Seam npnpr X

2 P

Salqplpeti K and Tnrrtinn -

The st generator tube minimu ample size, inspee,2 on result classificatio and the corresponding actiog required shall be specified in Table 4.4-.

The inservice inspec? n of steam genera r tubes shall be perform at the frequencies sp i fied in Specific on 4.4.4.3 and the inspe ed tubes shall,

be verified ceptable per the ceptance criteria of Spee ication 4.4.4.4 The tubes elected for each ervice inpection shall i ude at least 3of the to number of tubes all steam generators; t tubes selected r

thte inspections shall selected on a random b s except:

a.

Where ex ience in similar plants th similar water emistry indica s critical areas to be i pected, then at 1 St 50% of the tub inspected shall be fro ese critical are

b.

he first sample of tube selected for each

• ervice inspection

/

(subsequent to the pre rvice inspection) each steam generator shall include:

WATERFORD - UNIT 3 3/4 4-10 A v*%"rvv0+

to W3F1 -2006-0016 Page 4 of 19 [Replacement Page]

SUVLLANCE REQUIREMENTS (C D4f/<ued)

I.

All onplugged tubes th previously had detecta e wall p etrations (greate an 20%).

2.

Tubes in those ar s where experience has ndlcated potential prob s.

3.

A tube ins ctlon (pursuant to Spe Icatlon 4.4.4.4a.

shall b erformed on each selec d tube. If any se cted tube dees not permit the passa of the eddy curr probe forX.tube inspection, this all be recorded a an acent tube shall

/

selected and subject to a tube inspect n.

c.

e tubes selected as t econd and third s les (if required by

-L A

A A

ut

--bjected to tubes r i

th

'N ts where The resul following I/

Categ C-3 f the

/

rv /

Iys~ction Results

/

Less than of the total tube nspected are degr ded tubes and none the ins p ed tubes are defect e.

or more tubes, but ot more than 1% cf he total tubes insp ted are defective, or between 5% and of the total tubes lnspected are d aded tubes.

More than 1 of the total tubes in ectid are degra d tubes or more than 1 of the inspect tubes are defective.

In all ilnspecti

,s previously degrade ubes must exhibit signi cant (greater than 10 further wall penetratio Vto be included in the

ove percentage calculat a

ns.

WATERFORD - UNIT 3 3/4 4-11 N s Po, e. !

Vq q-1-7 Amendment No. j**~1 to W3F1 -2006-0016 Page 5 of 19 [Replacement Page]

SURVEIL CNCE REoUIREMENTS (C.,. nued

4. 43 nsectio-The abo required inservice nspections of eam generator tube all be performe at the following fre encles:

a. The firs inservice inspec on shall be perforce after 6 Effect1 Full Pa er Months but wi in 24 calender month of initial crit-ical i.

Subsequent in rvice inspections s 1 be performed tnt vals of not less1 han 12 nor mare than 4 calendar Pont after th previous Inspec on.

If two consecut e inspections fo owing rvlce under AVT ndltions, not inclu ng the preservic lnspecticn, esult in all in ection results falli into the C-1 Ca gory or if two consecutive nspections demonstr e that previousl observed

/

degradation h not continued and n additional d gr atton has occurred, t inspection interval ay be extended t maximum of once per 4 months.

b.

If the sults of the inserv e inspection of steam generator condu ed in accordance wit Table 4.4-2 at 4 month intervals fal into ategory C-3, the in ection frequency hall be increased to/at.

le t once per 20 month The increase i inspection frequency all apply until the *bsequent inspect ns satisfy the crit Ia of pecification 4.4.4. 6.; the interval y then be extended tpa

/ aximum of once per 0 Oonths.

/

Additional, uns eduled inservice spections shall be rformed on each steau gen ator in accordan with the first sam e inspection specified in able 4.4-2 durin the shutdown subsequ t to any of the follovi g conditions:

1. Pr ary-to-secondary bus leaks (not i Ing leaks a iginating from tu -to-tube sheet welds in excess of th imits of Specit Lion 3.4.5.2.

2 A seismic occur nce greater than th Operating Basis Earthquake.

3. A loss-of-olant accident requ ing actuation o the engineer safeguards.

4. A ma steam line or main dwater line bre TERFORD - UNIT 3 3/ 4-12 to W3Fl -2006-0016 Page 6 of 19 [Replacement Page]

RACTOR COOM T SYSTEM EURVE REQUIREMENTS (Con ned

4..4 4

ccetnc t

a.

As used I his Specificat n I.

in or tube s that portion of t ube or sleeve which forms the rimary system to sac dary system press e

/

/ ~~boundary.

means an exceptlon a

the dimensions, nish or I contour a tube from that re med by fabricati drawings or spe fications. Eddy-cur nt testing indica ons below 20%

the nominal tube wa thickness, if det table, may be nsidered as imperfe ons.

3, means a s ice-Induced crac ng, wastage, wear, or genera corrosio occurring on sthe inside or outside

/

/ ~of a tube.

//

/4 grad Tube ans a tube contal ng imperfections gre or than or equa to 20% of the nomi al wall thickness ca ed by

/

~degradatiof means the per ntage of the tube w 1 thic ss af ected or re od by degradation.

6.

means an i ier ction of such sever that it Scees the pluggi or repair limit. A be containing a defect is defecti P!Eiasn sRCe Iinlitit means the perfection depth or beyond which he tube shall be re ed from service b plugging op repaired by sleeving ecause it may bec unservic ble prior to the nex inspection and is qual to

/4nx of >e nominal tube wall a nS$./

describes t condition of a ie f it leaks o contains a defect la e enough to affec lts structural ntegrity in the even an Operating B is Earthquake, a loss-of-coolant acce t, or a steam 1e or feedwater line

/ ~break as speciflie n 4.4.4.3c., ao mtet Inspec ans an inspectl of the steam gener r tube from th point of entry h 9 leg side) completel around the -bend to the top pport of the cold 1.

/IO. Prer nsecon mea ian nspection of th ull length of ea tube in each ste generator perfore y eddy cur nt techniques pri to service to estab sh a baseline c dition of the tub g. This inspection s performed

/or to field hyd static test and prio to initial POVER OPERATION using t equipment and tech ques expected to be used during sub quent Inservice insp tions.

314 4-13 to W3F1 -2006-0016 Page 7 of 19 [Replacement Page]

to W3F1 -2006-0016 Page 8 of 19 [Replacement Page]

ins /tins I that I steam erators 1

perfor in a

/

m e. :Not that une ome circ tances, t operati conditio In OnO more st generator may be f nct to be e

an those n other steam ge rators. U er such ci instances he sampeesequence all be oodf to 1nspec t r

/ on

^T~e~~u

-U";;3 V4 4-15 to W3F1 -2006-0016 Page 9 of 19 [Replacement Page]

I i Ud~1

IRa WNW1 U'i 2

2i IIJ

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3Jul1

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,iS to W3F', -2006-0016 Page 10 of 19 [Replacement Page]

Insert 1 (TS 314.4.4) 3.4.4

a. SG tube integrity shall be maintained.

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

Insert 2 (TS 3/4.4.4)

Separate Action entry is allowed for each SG tube.

a. With one or more SG tubes satisfying the tube repair criteria and are not plugged in accordance with the Steam Generator Program,

1. Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and

2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.

b. If the required Action and Allowed Outage Time of Action a. above cannot be met or the SG tube integrity cannot be maintained, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN with the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Insert 3 (TS 3/4.4.4) 4.4.4.1 Verify SG tube integrity in accordance with the Steam Generator Program.

4.4.4.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

to W3F1 -2006-0016 Page 11 of 19 [Replacement Page]

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.5.2 Reactor Coolant Systema e limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKA)E,

c. 75 gallons per day primary oEe'condary leakag

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 1 gpm leakage at a Reactor Coolant System pressure of 2250 +/- 20 psia from any Reactor Coolant System pressure isolation valve specified in Table 3.4-1.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, e in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWNl-Tthitn-the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any Reactor Coolant Systemfleaiag reaer than any one of thy--

limits, excluding PRESSURE BOUNDARY LEAKAGE d leakage from Reactor\\

¢ Coolant System pressure isolation valves, reduce the~leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be In at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any Reactor Coolant System pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valve, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS NOTE:

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

4.4.5.2.1 Reactor Coolant System leakagesshall be demonstrated to be within each of the above limits by performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

)c~lS/

WATERFORD - UNIT 3 3/4 4-18 AMENDMENT NO. 497, i -9 to W3Fl -2006-0016 Page 12 of 19 [Replacement Page]

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4 2 h Reactor Coolant System pressure isolation valve specified in Tab Section A and Section B, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve,

d. Following valve actuation for valves in Section B due to automatic or manual action or flow through the valve:

1.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying valve closure, and

2.

Within 31 days by verifying leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.

ch Reactor Coolant System pressure isolation valve power-operated valve ecified in Table 3.4-1, Section C. shall be demonstrated OPERABLE by verifying leakage to be within Its limit:

a. At least once per 18 rnonths, and

b. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve.

The provisions of Specification 4.0,4 are not applicable for entry into MODE 3 or 4.

WATERFORD - UNIT 3 314 4-1 9 AMENDMENT NO. 964P to W3F1 -2006-0016 Page 13 of 19 [Replacement Page]

Insert 4 (TS 3.4.5.2) or any primary to secondary leakage not within limit, Insert 5 (Note to SR 4.4.5.2) except for primary to secondary leakage, Insert 6 (TS 4.4.5.2.2) 4.4.5.2.2 Primary to secondary leakage shall be verified to be < 75 gallons per day through any one SG at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

to W3Fl -2006-0016 Page 14 of 19 [Replacement Page]

ADMINISTRATIVE CONTROLS 6.5.8 INSERVICE TESTING PROGRAM This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

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

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

b. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice testing activities.

c. The provisions of Specification 4.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

7 WATERFORD - UNIT 3 6-7a AMENDMENT.NO. 1I*

to W3F1 -2006-0016 Page 15 of 19 [Replacement Page]

Pages 6-9 through page 6-13 not used I

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,'a F-K WATERFORD - UNIT 3 6-8 Next Page is 6-14 AMENDMENT NO. 16, 63, -,

100,109, 488, I

to W3F1 -2006-0016 Page 16 of 19 [Replacement Page]

Insert 7 (New SG Program) 6.5.9, 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 nspected or plugged to confirm that the performance criteria are being met.

b.

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

1.

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

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Primary to secondary leakage is not to exceed 540 gpd through any one SG.

3.

The operational leakage performance criterion is specified in LCO 3.4.5.2, "Reactor Coolant System operational leakage."

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

to W3F1 -2006-0016 Page 17 of 19 [Replacement Page]

d.

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

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

2.

Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.

3.

If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 eiTective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or encineering 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.

to W3F1 -2006-0016 Page 18 of 19 [Replacement Page]

A0MINIjTaTIVECONQI R ANNUAL REPORTS (Continued)

(1)

Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2)

Results of the last Isotopic analysis for radlolodine pertfonned prior to exoseading the limit, recuftc of analyic while limit was exceeded and results of one analysis after the radlolodine activity was reduced to less than limit. Each result should include date and time of samplinn and the radloladine concentrations:

(3)

Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the Vest sample In which the limit was exceeded; (4)

Graph of the 1-131 concentration and one other radloodcine isotope concentration In microcules per gram as a function of time for the duration of the specific activity above steady-state level; and (5)

The time duration when the specific activity of the primary coolant exceeded the radlolodine limit.

6.9.1.5 BELEETI I

WATERFORD - UNIT 3 6-17a AMENDMENT NO. *446.

}M

)

to W3F1 -2006-0016 Page 19 of 19 [Replacement Page]

Insert 8 STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.5 A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.5.9, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG,

b.

Active degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

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

e.

Number of tubes plugged during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

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

h.

The effective plugging percentage for all plugging in each SG.

To W3FI -2006-0016 Revised Markup of Replacement Pages for All Technical Specification Bases Pages For Information Only to W3F1 -:2006-0016 Page 1 of 8 [Replacement Page]

fMCTOR COOLANT SYSTEM MA ~S SAFE;TY VALVES (Continued) valves are OPERABLE. an operating shutdown cooling loop, connected to the RCS.

provides overpressure relief capability and will prevent RCS overpressurization.

In addition, the overpressure protection system provides a diverse means of protection against RCS overpressurization at low temperatures.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.

The combined relief capacity of these valves Is sufficient to limit the system pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pre! sure-High) is reached and also assuming no operation of the steam dump va!ves.

Demonstration of the safety valves' lift settings will occur only during reactor shutdown and will be performed in accordance with the provisions of Section Xl of the ASME Boiler and Pressure Vessel Code.

.3/f.3 PRESSURIZER An OPERABLE pressurizer provides pressure control for the Reactor Cootant System during operations with both forced reactor coolant now and with natural circulation tow. The minimum water level in the pressurizer assures the pressurizer heaters, which are required to achieve and maintain pressure con-Lrol, remain covered with water to prevent failure, which could occur if the heaters were energized while uncovered. The maximum water level In the pres-surizer ensures that this parameter is maintained within the envelope of operation assumed In the safety analysis. The maximum water lcvel also ensures that the RCS is not a hydraulically safid system and that a steam bubble will be provided to accommodate pressure surges during operation. The steatr bubble also protects the pressurizer code safety valves against water relief. The reqt Irement to verify that on an SIAS test signal the pressurizer heaters are automatically shed from the emergency power sources is to ensure that the non-Clas 1 E heaters do not reduce the reliability of or overload the emergency power source. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability to control Reactor Coolant System pressure and establish and maintain natural circulation.

The auxiliary pressurizer spray Is used to depressurize the RCS by cooling the pressurizer steam space. The auxiliary pressurizer spray is used during those periods when normal pressurizer spray is not available, such as the later stages of a normal RCS cooldown. The auxiliary pressurizer spray also distri-butes boron to the pressurizer when normal pressurizer spray is not available.

The auxi iary pressurizer spray is used, In conjunction with the throttling of the HPSI pumps, during the recovery from a steam generator tube rupture acci-denl. The auxiliary pressurizer spray is also used during a natural circulation cooldown as a safety related means of RCS depressurization to achieve shutdown cooling system initiation conditions and subsequent COLD SHUTDOWN per the require.

ments of Branch Technical Position (RSB) 5-1.

3L42l.4 STEAM GENERATOR rA I ) erLl R-(TV I/hsure that Mstructural inter of this paon of th CS will am-Ktained. T program for ins ice inspe n of sI m generat ubes is WxrERFORD - UNIT 3 B 3/4 4-2 Amendment No.;a2F to W3F1 -:2006-0016 Page 2 of 8 [Replacement Page]

UA-CTOSR C;Q_ 6NT.Y$TEM PA STEAM GENERATM Continued based on a mT ification of Regulatory Gu 1.83, Revision Inservice Inspection gfsteam generator tubing is sential In order maintain surv gillat of the conditions of the I es In the event t there Is evidenpd of mechanical damage orgrogressive degr ation due to design, man acturing errors, or Inservic conditions that I d to corrosion, Invice inspection of steam nerator tubing a provides a means of aracterizing the nature a cause of any tu degradation so that corr lye measures can be taken.

' RU 04-1243 Ch 391 The plant is e ected to be operedi a rnanner such thate scondary

/

cool3nt will be mai ined within thos hemistry limits found to r ult In negl gible corros of the steam ge orator tubes. If the secon ry coolant chemistry is n malntained withi hese limits, localized corr Ion may likely result' stress corrosionytacking. The extent of cra ing during plan" ape tion would be limW by the limitation of stea generator tube leak; g tween the prim coolant system and the condary coolant sy m

(prin

-to-secondary I age -75 gallons per da er steam generator Cracks h

ig a primary-to-scondary leakage less than is lImit during opera n will have adequate' safety to withstand the ds imposed during rmal operation and byxstulated accidents. Ope ting plants have de nstrated that primary-to-condary leakage of 75 p ons per day per ste generator can readily be detected radiatin monitors of ste generator tilowdo

. Leakage in excess oft 75 gallon per day limit I ypecification 3.4,5.2 ill require plant shuticw and an unscheduled

'nsp ion, during which I leakage tubes will be lo ed and p1 ged or repaired.

• PMKz4 $243. Ct 38)

/ Wastage-typ dectir unieywt rp hmistry treatment of t ccndary coolant. Howev, even if a defect sh Id (levelop in service, it

/will e found during sche led inservice steam enerotor tube examinatio Plugging or sleeving w' be required for all tu s with imperfections exceeding the pluggig or repair limit as d ned in Surveillance Requ mont 4.4.41.4. Defective tbes may be repair y sleeving In accordan ith CENS Rep~rt CEN 0

. 'Waterford 3 Stea Generator Tube Repair ing Leak Tigh Sleeves, Rev' ion 00-P. dated Dec her 1992. Steam gener or tube inspections operating plants ha demonstrated the capa ty to reliably detect de adation that has pen rated 20% of the original be wa!l thickne

. Sleeved tubes wile included in the periodic be inspection for tl.inservice Inspection rogram.

/

Whenever th etsof any steam gener r tubing !inseyt Inspection/

fali into Category C-these results will be pro tly reported to e Commission purs nt to Specification 6.9.1 ppr the resump n of plant operation, Suc ses will be considered Ithe Commis I n on a case-by-case basis and ma esult in a requirement for nalysis, laborory examinations.

tests, additi al eddy-current Inspectio and revision the Technical Speifics ns, if necessary, AMENDMENT NO. 2-i47, WATERFORD - UNIT 3 B 3/4 4-3 CHANGE NO.

to W3F1 -2006-0016 Page 3 of 8 (Replacement Page]

REACTOR COOLANT SYSTEM.

BASES (continued)

Monitoring Containment Sump In-Leakage Flow During automatic operation of the containment sump pumps (after a containment sump pump has operated), the flow calculation performed by the plant monitoring computer based on a level change will no longer be accurate since the level in the sump will be lowering. A 20 minute timE period has been conservatively determined based on engineering calculations for this equipment operation. In addition, upon reboot of the plant monitoring computer, a period of 10 minutes is required for the leak rate calculation to become available. It has been determined these time periods (independent or combined) of calculation sump in-leakage flow inaccuracies, the instrumentation remains adequate to detect a leakage rate, or its equivalent, of one gpm in less than one hour; therefore, the containment sump level instrumentation and the corresponding flow calculation is considered to remain operable.

References 1:

10 CFR 50, Appendix A, Section IV, GDC 30.

2.

Regulatory Guide 1.45, Revision 0, dated May 1973.

3.

UFSAR, Sections 5.2.5 and 12.3.

• *(DRN 04-122n. Ch. 33) 3/4.4.5.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than I gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowances for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowable limit.

The 75 gallon per daygpd) per steam gen at tube leakage limi'nstkres that the 7 radi ogical conseque es, inc ding that from e leak e, will be lip d to the\\1 CFR50.6

/

limits br offsite do and within tye limits of eneral Desi Criterio 19 for contro oom se.

For thos analy d events that do0ot re t in faulted stea e rators, greater tha o equal to 75 gpd pri

-to-secondary leakag er steam generator is sumed in the analysi or those analyzed ts that result in a fa e steam generator (H.,

LB), 540 gpd priary-to secon ry lea ge is assume rough e faulted ste generatr while grea r than or e al to 7 gpd prima to-secon ry leakage i assume

rough the in Ct steagenerator.-

RN 04-1243. Ch. 38)

WATERFORD - UNIT 3 B 3/4 4-4e CHANGE NO. -a3, ".,

to W3F1-2006-0016 Page 4 of 8 [Replacement Page]

Insert B-1

Background

Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and seconcary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.1.1, "RCS Loops - MODES 1 and 2," LCO 3.4.1.2, "RCS Loops - MODE 3**," LCO 3.4.1.3, "RCS Loops - MODE 4," and LCO 3.4.1.4, "RCS Loops - MODE 5 with reactor coolant loops filled**."

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

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

Specification 6.5.9, Steam Generator Program, requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.5.9, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage.

The SG performance criteria are described in Specification 6.5.9. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions. The processes used to meet the SG performance criteria are defined by NEI 97-06, Steam Generator Program Guidelines (Reference 1).

Safety Analysis The Steam Generator Tube Rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding a SGTR is the basis for this Specification. The analysis of a SGTR event is based on the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes a Loss of Offsite Power with subsequent releases to the atmosphere via Main Steam Safety Valves and Atmospheric Dump Valves.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) For those analyzed events that do not result in faulted steam generators, greater than or equal to 75 gpd primary to seconcary leakage per steam generator is assumed in the analysis. For those analyzed events that result in a faulted steam generator (e.g., MSLB), 540 gpd primary to secondary leakage is assumed through the faulted steam generator while greater than or equal to 75 gpd primary to seconcary leakage is assumed through the intact steam generator.

to W3F1-:2006-0016 Page 5 of 8 [Replacement Page]

For accidents that do not involve fuel damage, the primary coolant activity level is assumed,:o be equal to the LCO 3.4.7 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 and 10 CFR 50.67. Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Limitingi Condition for Operation The LC:O requires that SG tube integrity be maintained. The LCO also requires that all SG tubes :hat satisfy the repair criteria be plugged in accordance with the Steam Generator Program. During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity. In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.5.9, Steam Generator Program, and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubas under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that significantly affect burst or collapse. In that context, the term "sic nificantly" is defined as "An accident loading condition other than differential pressure is corsidered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse cordition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the to W3F1 -.2006-0016 Page 6 of 8 [Replacement Page]

design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB and Draft Regulatory Guide 1.121.

• The accident induced leakage performance criterion ensures that the primary to secondary leakage caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 540 gpd through any one SG. The accident induced leakage rate includes any primary to secondary leakage existing prior to the accident in addition to primary to secondary leakage induced during the accident.

The operational leakage performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational leakage is contained in LCO 3.4.5.2, Reactor Coolant System operational leakage, and limits primary to secondary leakage through any one SG to < 75 gallons per day. This limit is based on assumptions in radiological analyses. This limit is less than the 150 gallons per day through any one SG limit of NEI 97-06, which assumes that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a Main Steam Line Brea'<. If this amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

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

Action 'a." applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 4.4.4.2. 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 criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Action "b" applies.

An allowed outage 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, 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.

Howev3r, the affected tube(s) must be plugged prior to entering HOT SHUTDOWN to W3F1 -:2006-0016 Page 7 of 8 [Replacement Page]

following the next refueling outage or SG inspection. This time period is acceptable since operation until the next inspection is supported by the operational assessment.

Action "b" applies if the actions and associated allowed outage time of Action "a" are not mel: or if SG tube integrity is not being maintained, the reactor must be brought to HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The allowed outage time 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.

Surveil ance Requirements During shutdown periods the SGs are inspected as required by SR 4.4.4.1 and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Reference 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 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, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.4.1. The frequency is determined by the operational assessment and other limits in the SG examination guidelines.

(Reference 6). The Steam Generator Program uses information on existing degradations arid growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.5.9 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

As required by SR 4.4.4.2 any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specifica:ion 6.5.9 a e 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 flaxv growth. In addition, the tube repair 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.

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The frequency of prior to entering HOT SHUTDOWN following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair 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 C FR 50 Appendix A, GDC 19.

3. 10 CFR 50.67.

4. ASME Boiler and Pressure Vessel Code, Section 1I1, Subsection NB.

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

6. EPFRI, Pressurized Water Reactor Steam Generator Examination Guidelines.

Insert 13-2 The primary to secondary leakage limit of 75 gallons per day through any one SG is based con the operational leakage performance criterion in NEI 97-06. The Steam Generator Program operational leakage performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The NEI 97-06 limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion (since it is less than 150 gpd through any one SG) in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures