ML070030517

Supplement to Application for Technical Specification (TS) Improvement Regarding Steam Generator Tube Integrity

ML070030517
Person / Time
Site:	Cook
Issue date:	12/26/2006
From:	Peifer M Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:6449-01, TAC MD2306, TAC MD2307
Download: ML070030517 (16)

Text

{{#Wiki_filter:Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place Bridgman, MI 49106 POWER" AEP.com A unit ofAmerican ElectricPower December 26, 2006 AEP:NRC:6449-01 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-PI-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 Supplement to Application for Technical Specification (TS) Improvement Regarding Steam Generator Tube Integrity

References:

1. Letter from Joseph N. Jensen, Indiana Michigan Power Company (I&M), to U. S.

Nuclear Regulatory Commission (NRC) Document Control Desk, "Application for Technical Specification (TS) Improvement Regarding Steam Generator Tube Integrity," AEP:NRC:6449, Accession Number ML061570157, dated May 26, 2006.

2. Electronic mail message from Peter Tam, NRC, to Michael Scarpello, I&M, et al.,

                           "D.C. Cook: Proposed RAI on Application for Amendment re. TSTF-449 (TAC MD2306 and MD2307)," Accession Number ML062370043, dated August 23, 2006.

Dear Sir or Madam:

By Reference 1, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant Units 1 and 2, proposed to amend Facility Operating Licenses DPR-58 and DPR-74. I&M proposed to modify Technical Specifications (TS) requirements related to steam generator tube integrity consistent with TS Task Force (TSTF) generic change traveler TSTF-449, "Steam Generator Tube Integrity." By Reference 2, the U. S. Nuclear Regulatory Commission (NRC) staff provided draft questions that were discussed in a conference call with I&M personnel on August 30, 2006. During the conference call I&M stated that it would provide a formal response to the draft questions on the docket. provides an affirmation statement pertaining to this letter. Enclosure 2 provides a response to the NRC questions as discussed during the August 30, 2006, conference call. to this letter provides the marked-up TS page to replace the corresponding page submitted in Attachment IA to Reference 1. Attachment 2 provides the TS page, with the changes incorporated, to replace the corresponding page submitted in Attachment 2A to Reference 1.

U. S. Nuclear Regulatory Commission AEP:NRC:6449-01 Page 2 provides draft TS Bases pages for Unit 1 to replace the corresponding pages submitted in Attachment 3 to Reference 1. Unit 2 TS Bases changes are consistent with Unit 1 draft changes. The proposed TS changes, as supplemented by this letter, remain within the scope of the amendment previously proposed by Reference 1. Therefore, the No Significant Hazards Consideration evaluation and the evaluation of Environmental Considerations provided in Enclosure 2 to Reference 1 continue to bound the proposed changes. Copies of this letter and its attachments are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91. There are no commitments made in this letter. Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428. Site Vice President KAS/rdw

Enclosures:

1. Affirmation

2. Response to Draft Request For Additional Information Attachments: 1. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked To Show Changes

2. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages With the Proposed Changes Incorporated

3. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Bases Pages Marked To Show Changes c: J. L. Caldwell, NRC Region III K. D. Curry, Ft. Wayne AEP, w/o enclosures/attachments J. T. King, MPSC MDEQ - WHMD/RPMWS NRC Resident Inspector P. S. Tam, NRC Washington, DC

Enclosure 1 to AEP:NRC:6449-01 AFFIRMATION I, Mark A. Peifer, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief. Indiana Michigan Power Company MarkA.Pfe Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS c -"DAY OF1DCerYý 1e-_r , 2006 Notary Pbi REGN 0. WENDZEL My Commission Expires otary Public, Berrien Coun.t. K. My Commission Expires Jan. 21,"009

Enclosure 2 to AEP:NRC:6449-01 RESPONSE TO DRAFT REQUEST FOR ADDITIONAL INFORMATION By Reference 1, Indiana Michigan Power Company (I&M) proposed to modify Technical Specification (TS) requirements related to steam generator tube integrity consistent with TS Task Force (TSTF) generic change traveler TSTF-449, "Steam Generator Tube Integrity." By electronic mail message (Reference 2), Nuclear Regulatory Commission (NRC) staff provided I&M with draft questions to discuss by conference call. Following discussions with NRC staff on August 30, 2006, I&M personnel stated that responses discussed during the conference call would be provided formally. The questions are addressed below. NRC Ouestion 1 "On Page 3.4.13-2 of your proposed technical specifications (TS) for Cook Unit 1, Condition C states that if identified LEAKAGE is not within limits, then the required action is to reduce LEAKAGE to within limits within 4 hours. This statement appears to imply that even if there is primary-to-secondary LEAKAGE (which is identified LEAKAGE), you would be allowed to reduce this leakage to within limits. TSTF-449, however, does not allow 4 hours to reduce primary-to-secondary leakage to within limits. Please discuss your plans to modify your proposed TSs in order to be fully consistent with TSTF-449. For example, the TS could be modified as follows: 'C. Identified LEAKAGE not within limits for reasons other than primary-to-secondary LEAKAGE.' Your Bases may also need to be revised to clarify this issue (Page B3.4.13-6)." I&M Response to NRC Ouestion 1 I&M agrees with the NRC staff statement and has modified Unit 1 TS 3.4.13, Condition C to state, "Identified LEAKAGE not within limits for reasons other than primary-to-secondary LEAKAGE." Unit 1 TS Bases 3.4.13 has also been revised to be consistent with the proposed Unit 1 TS 3.4.13. The changes have been included in Attachments 1, 2, and 3. NRC Ouestion 2 "On Page B 3.4.13-4 of your proposed Bases there appears to be a typographical error. The sentence reads: '...performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4).' It appears that Reference 4 should be Reference 6." to AEP:NRC:6449-01 Page 2 I&M Response to NRC Ouestion 2 I&M agrees with the NRC staff statement and has modified TS Bases Page 3.4.13-4 and has included it in Attachment 3 for information. NRC Ouestion 3 "On Page B 3.4.17-2 of your proposed Bases there appears to be a statement missing. The second sentence of the second paragraph states: '...total primary to secondary LEAKAGE from all SGs of 1 gallon per minute (gpm) as a result of accident induced conditions.' Please discuss your plans to modify your proposal to include the following statement: '...or is assumed to increase to 1-gpm...' as part of this sentence which would make your submittal consistent with TSTF-449." I&M Response to NRC Ouestion 3 TS Bases Pages 3.4.13-2 and 3.4.17-2 have been revised for both Unit 1 and Unit 2 TS Bases to be consistent with TSTF-449. The Unit 1 TS Bases pages have been included in Attachment 3 for information. NRC Ouestion 4 "In your proposed Bases, you indicate that the 150 gpd leakage limit is measured at room temperature (e.g., Page B 3.4.13-9). However, in Enclosure 2 you indicate that the primary to secondary leakage is evaluated at hot (615.2-degrees Fahrenheit) conditions. Please discuss your plans to correct the apparent discrepancy." I&M Response to NRC Ouestion 4 I&M radiological accident analyses include both Control Room Habitability and offsite dose analyses. The Control Room Habitability dose analysis assumes the reactor coolant system (RCS) primary-to-secondary (P/S) leakage is at room temperature, while the offsite dose analysis assumes the RCS P/S leakage is at normal operating temperature and pressure. To remain conservative relative to current and proposed TS limiting condition for operation for surveilling P/S leakage, I&M multiplies the Electric Power Research Institute guideline equations for determining leakage by a volume correction factor of 1.52. The correction factor ensures the smaller mass release assumed in the offsite dose analysis bounds the TS measured leak rate. While this is overly conservative relative to Control Room Habitability dose analyses, this is appropriately conservative for offsite dose analyses. NRC staff suggested we clarify this point in CNP TS Bases. Unit 1 and Unit 2 TS Bases Surveillance Requirement 3.4.13.2 have been to AEP:NRC:6449-01 Page 3 revised to clarify CNP calculation of P/S leak rate. The Unit I TS Bases Page 3.4.13-9 has been included in Attachment 3 for information. References for this Enclosure

1. Letter from Joseph N. Jensen, I&M, to NRC Document Control Desk, "Application for Technical Specification (TS) Improvement Regarding Steam Generator Tube Integrity,"

AEP:NRC:6449, Accession Number ML061570157, dated May 26, 2006.

2. Electronic mail message from Peter Tam, NRC, to Michael Scarpello, I&M, et al., "D.C.

Cook: Proposed RAI on Application for Amendment re. TSTF-449 (TAC MD2306 and MD2307),"Accession Number ML062370043, dated August 23, 2006.

Attachment 1 to AEP:NRC:6449-01 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3.4.13-2

RCS Operational LEAKAGE 3.4.13 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Identified LEAKAGE not C.1 Reduce LEAKAGE to within 4 hours within limits tor reas-ons limits. bther th arn primary to secondary LEAKAGE. OR Primary to scconday LEAKAGE not with-i D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time of Condition A, B, AND or C not met. D.2 Be in MODE 5. 36 hours OR Pressure boundary LEAKAGE exists. OR kmirary to secoNdar

    .EAKAGE not within Jirnit.SR 3.4.13.2 not SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.4.13.1                                    NOTE*"

ri: Not required to be performed until 12 hours after establishment of steady state operation. VgoF-.77 _Epec Verify RCS operational AjAEeekage is within 72 hours limits by performance of RCS water inventory balance. Cook Nuclear Plant Unit 1 3.4.13-2 Amendment No. 287

Attachment 2 to AEP:NRC:6449-01 DONALD C. COOK NUCLEAR PLANT UNIT I TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED 3.4.13-2

RCS Operational LEAKAGE 3.4.13 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Identified LEAKAGE not C.A Reduce LEAKAGE to within 4 hours within limits for reasons limits. other than primary to secondary LEAKAGE. D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time of Condition A, B, AND or C not met. D.2 Be in MODE 5. 36 hours OR Pressure boundary LEAKAGE exists. OR Primary to secondary LEAKAGE not within limit. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 NOTES

1. Not required to be performed until 12 hours after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by 72 hours performance of RCS water inventory balance. Cook Nuclear Plant Unit 1 3.4.13-2 Amendment No. 297,

Attachment 3 to AEP:NRC:6449-01 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION BASES PAGES MARKED TO SHOW CHANGES B 3.4.13-2 B 3.4.13-4 B 3.4.13-6 B 3.4.13-9 B 3.4.17-2

RCS Operational LEAKAGE B 3.4.13 BASES APPLICABLE SAFETY ANALYSES (continued) inu~*e. gpm) or is assumed to increase to one gpm as a resultof pccident induced conditions, The LCO requirement to limit primary to pecondary LEAKAGE through any one SG .to less than or equal to 150 ballons per day is significantly less ithan the conditions assumed- in the

                   $afety analysis. at least a" 1 gPm primay to,*eGodary LEAKAGE at initial condition.

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident and other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid. The UFSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released via the steam generator power operated relief valves (and safety valves if their setpoint is reached) if offsite power is not available or if the condenser steam dump system fails to operate. The safety analysis for the SLB accident assumes the amount of primary to secondary LEAKAGE in the three intact SGs is 1 gpm minus a faulted SG tube LEAKAGE of 500 gallons per day as an initial condition. The dose consequences resulting from events resulting in a steam discharge to the atmosphere are within a small fraction of the limits defined in 10 CFR 100 and within GDC-19. The RCS Operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE The 0.8 gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air particulate monitoring equipment can detect within a reasonable time period. The limit is established for the pressurizer surge line in the leak before break methodology. Violation of this LCO could result in Cook Nuclear Plant Unit I B 3.4.13-2 Revision No. 0

RCS Operational LEAKAGE B 3.4.13 BASES LCO (continued)

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

                   *': Pri~mary toSecondary LEAKAGE Tlirouah Any One SG The imit of 150 gallons per day per SG is based on the operational EAKAGE performance criterion in NEI 97-06, Steam Generator rogram Guidelines (Ref. 6). The Steam Generator Program perational LEAKAGE performance criterion in NEI 97-06 statesi
                             'The RCS operational primary to secondary leakage through any one 6G shall be limited to 150 gallons per day." The limit is based on
                               )perating experience with SG tube degradation mechanisms that vesult in tube leakage. The operational leakage rate criterion In
                                =onjunction with the implementation of the Steam Generator Program s an effective measure for minimizing the frequency.of steam oenerator tube ruptures.

d ,. -,,',,., +, tR - I AVAKAGE t,* ua- AlSl,,-{,no.-m " Total prima*ly to secondary LEAKAGE amounting I to I gpm through tho*intct SGs produces acceptable ef...te and contrGo room doses in the SGTR accident analysis. For the SLB accident, the amo.unt o primary to seconda-'y LEAKAGE in the three intact SG ir, asAumed to be 1 gpmn minus a faulted SG tube LE-=AK-A.GE of 500 gallons per day. The LCO limit of 600 gallons per day is mor~e cOnsew.ative than the 1 gpmA value assumed- in t-he offesite dose calculations. This limit is imposd to hel m.inimizo the potontial for excessive leakage or tube burst inthe event of a MSLB or LOCA consistent with the LCOQ limit on primary to secondary LEAKAGE through any one SG. In addition, the coenseryative limit is appropriate due to the inrGeased steamn reease as a result of the replacemnent S~s. Violation of this LCO c-ould eXceed the offsite dose limits for these accidents,. Primary to secoendary LEAKAGE must be included in the toa allowable limit for identified LEAKAGE. e..,.,.Pgma te R--ida LEAK(AGE thmug..h Any. G(e S Cook Nuclear Plant Unit I B 3.4.13-4 Revision No. 0

RCS Operational LEAKAGE B 3.4.13 BASES APPLICABILITY (continued) LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE. ACTIONS A.1 and A.2 With unidentified LEAKAGE > 0.8 gpm, the pressurizer surge line must be verified not to be the source of unidentified LEAKAGE or the unidentified LEAKAGE must be reduced to within limit within 4 hours. These Required Actions are necessary to satisfy the requirements for the application of Leak-Before-Break methodology to the pressurizer surge line as documented in Reference 4 and approved by the NRC as documented in Reference 5, and are necessary to prevent further deterioration of the RCPB associated with the pressurizer surge line. The Completion Time allows time to verify leakage rates and either identify the unidentified LEAKAGE or reduce LEAKAGE to within limit before the reactor must be shut down. B.1 Unidentified LEAKAGE > 1.0 gpm must be reduced to < 1.0 gpm within 4 hours. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB. C.1 Identified LEAKAGE: GF f[tv* ej.tion of primary to secondary LEAKAGEI in excess of the LCO limits must be reduced to within limits within 4 hours. This Completion Time allows time to verify leakage rates and reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB. D.1 and D.2 If any Required Action and associated Completion Time of Condition A, B, or C is not met, i pressure boundary LEAKAGE exists, or if

• S*Ro3Le-*

3. is not n met, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary Cook Nuclear Plant Unit 1 B 3.4.13-6 Revision No. 0

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.4.13.2 This-S verifies thait primary to 'secondary LEAKAGE is less thin or equal o 150 gallons per day through any one SG. Satisfying the primary to Wecondary LEAKAGE limit ensures that the operational LEAKAGE erformance criterion in the Steam Generator Program is met. If this SR snot met, compliance with LCO 3.4.17, "Steam Generator Tube integrity," should be evaluated. The primary to secondary LEAKAGE is rneasured at room temperature as described in Reference 7. Prior to 0ompadson with the 150 gallons per day TS limit, the measured primary osecondary LEAKAGE is multiplied by a volume correction factor of

                   }1.52. The correction factor ensures the offsite dose analyses, which assume primary to secondary leakage is at normal operating temperature and pressure, remain bounding. The operational LEAKAGE rate limit
                   ýpplies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, then all of the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

rs ITK6e Sr-v~eil lanhce I's-- mod ifie'd-b-a-W ~e--wh~ich st'a-te's thfa-t-the -_Srve6ilIla'nc'e not required to be performed until 12 hours after establishment of teady state operation. For RCS primary to secondary LEAKAGE

                    *etermination, steady state is defined as stable RCS pressure, emperature, power level, pressurizer and makeup tank levels, makeup
                     .nd letdown,,and RCP seal injection and return flows.

to trend interval of early rimary to secondary LEAKAGE and recognizes the importance is a reasonable Frequency of 7hours, akage SuLrveillan~ce detection in the prevention of accidents. The primary to [e econdary LEAKAGE is determined using continuous process radiiation onitors or radiochemical grazb.sampling in accordance with the EPRI

                     .uAidpline~s_(Ref. 7)Z, This SR pric: the means ncccssa~y to determnine SG OPEP.AB!LITY in no      erainal MODE. The rcguircen~et t9 demonstate SG tb ingrt in ccodance with the Steam Generator Programn emphasizes the importan-e-of SG tube integrity, oene; though this SuF~eillance cannot be pe ome at nrGmal opercating conditions REFERENCES          1. UFSAR, Section 1.4.3.

2. Regulatory Guide 1.45, May 1973.

3. UFSAR, Section 14.2.4.

4. Letter from Indiana Michigan Power Company (M. W. Rencheck) to the NRC dated October 26, 2000 (Letter C1000-20).

Cook Nuclear Plant Unit I B 3.4.13-9 Revision No. 0

SG Tube Integrity B 3.4.17 hpPilCPiL .The6steam generatortube rupture (SGTR) accident is the limiting design 6AFETY- basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of an SGTR event assumes a bounding Orimary to secondary LEAKAGE rate equal to the operational LEAKAGE ( ate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage ate associated with a double-ended rupture of a single tube. The Pccident analysis for an SGTR assumes the contaminated secondary fluid is released to the atmosphere via the SG power operated relief Valves. The analysis for design basis acc6dents and transients other than an SGTR assumes the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the

                    ýtmosphere is based on the total primary to secondary LEAKAGE from all SGs of one gallon per minute (gpm) or is assumed to increase to one Upm 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 Ppecific Activity," limits. For accidents that assume fuel damage, the prdmary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC
                    *pproved licensing basis.(e.g., a small fraction, of these limits).

teimgerlr~toi: fulbe1ntegr-ty satisfies Criteron

                                                ... ..."...........   .      . 2. of 10. CFR 500.36(c)(2)(ii), .

I.CO 17ThieLC~f~f'reqjurir tb intgrj6ity be rainane iit Id. T.f6LCO also

                   . ..equires that all SG tubes that satisfy the repair criteria be plugged in
                    ýccordance with the Steam Generator Program,.

uring an SG inspection_,any inspected tube that satisfies the Stearr Senerator Program repair criteria is removed from service by plugging. 1, tube was determined to satisfy the repair criteria but was no.t ,plugged, he tube may, still have tube integrjty... te-co-nt-ext o&f this SeiiaonanGtueisdfined as the enire [ength of the tube, including the tube wall, between the tube4to-tubesheet

                     *,eld at the tube inlet and the tuba.to-tubesheet weld at the tube outlet.

re[e tube-to-tubesh.bet weld is not con.sidered park of.the tube..

                            ~~ibie'hastu1ieyt                        aife h Gpifrai rida.. The SG performa Tmce cijte 'p re.defined in Specificationfib.57 Ste   ,Gbhenator Progem

• iam,"d'. ýsibe acceptable SG tube

[errormance. The Steam Generator Program also provides the eefrEt.sfor g Rev. X.X B 3.4.17-2 Cook Nuclear Unit II Plant Unit Nuclear Plant B 3.4.17-2 Rev. X.X}}