Text

Response to Request for Additional Information Re Application for Technical Specification Improvement on Steam Generator Tube Integrity
	ML070460048
Person / Time
Site:	Millstone
Issue date:	02/14/2007
From:	Christian D; Dominion Nuclear Connecticut
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TAC MD2570, TAC MD2571 06-001A
	Download: ML070460048 (38)
	v • d • e

Dominion Nuclear Connecticut, Inc.

5000 I>orninion Boulevard, Glen Allen, Virginia 23060

\\X'ch hddrcas: www.dorn.com February 14, 2007 U. S. Nuclear Regulatory Commission Serial No.

06-001A Attention: Document Control Desk NL&OS/PRW RO One White Flint North Docket Nos.

50-336150-423 11 555 Rockville Pike License Nos.

DPR-65lNPF-49 Rockville, MD 20852-2738 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNITS 2 AND 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY

/TAC NOS. MD2570 AND MD257I 1 In a letter dated May 31, 2006, Dominion Nuclear Connecticut, Inc. (DNC) submitted a request to revise the steam generator integrity requirements in the technical specifications for Millstone Power Station Units 2 and 3 (MPS2 and MPS3), consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity."

In a facsimile dated August 16, 2006, transmitted on August 23, 2006, the NRC forwarded a request for additional information (RAI) in order for the NRC staff to complete its review. The draft RAI was further discussed in a conference call between DNC and the NRC on October 4,2006.

The response to the RAI for MPS2 is provided in Attachment 1 to this letter.

Revised marked up pages for MPS2 are provided in Attachment 2.

The response to the RAI for MPS3 is provided in Attachment 3 to this letter. Revised marked up pages for MPS3 are provided in Attachment 4. The marked up bases pages are provided for information only.

If you have any questions in regard to the responses provided or require additional information, please contact Mr. Paul R. Willoughby at (804) 273-3572.

Very truly yours, David A. Christian Senior Vice President - Nuclear Operations and Chief Nuclear Officer

Serial No. 06-001A Docket Nos. 50-336150-423 Response to Request for Additional lnformation Page 2 of 3 Commitments in this letter: None Attachments: (4)

1. Response to Request for Additional lnformation - Millstone Power Station Unit 2.

2. Revised Marked Up Pages - Millstone Power Station Unit 2.

3. Response to Request for Additional lnformation - Millstone Power Station Unit 3.

4. Revised Marked Up Pages - Millstone Power Station Unit 3.

cc:

U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 1 1555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station

Serial No. 06-001A Docket Nos. 50-336150-423 Response to Request for Additional Information Page 3 of 3 COMMONWEALTH OF VIRGINIA

)

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by David A. Christian, who is Senior Vice President - Nuclear Operations and Chief Nuclear Officer of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this / 4% day of & h c ~ ~. y,2007 My Commission Expires:

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Notary Public (SEAL)

Serial No. 06-001A Docket No.05-336 ATTACHMENT I APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY (TAC NO. MD2570)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No. 06-001A Docket No. 50-336 Response to Request for Additional Information Page 1 of 8 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION MILLSTONE POWER STATION UNIT 2 By letter dated May 31, 2006, Dominion Nuclear Connecticut, Inc. (DNC),

submitted a request to revise the steam generator tube integrity requirements in the Technical Specifications (TSs) for Millstone Power Station, Unit No. 2 (MPS2).

In order for the Nuclear Regulatory Commission (NRC) staff to complete its review, responses to the following questions are needed.

NRC Question No. 1 On page 5 of Attachment 1 of the May 31, 2006, amendment request, DNC indicated that the revision to the TS requirements pertaining to steam generator tube integrity can not be implemented without approval of the alternate source term amendment currently in review by the NRC staff. Please clarify the reason for this (i.e., is it because of the proposed accident induced leakage limit, the proposed increase in the primary-to-secondary leakage limit, and because reference was made to Title 10 of the Code of Federal Regulations Section 50.67 in your proposed Bases.)

DNC Response Current Millstone Power Station Unit 2 (MPS2) Technical Specifications state that Reactor Coolant System leakage shall be limited to: "... c. 0.035 GPM primary-to-secondary leakage through any one steam generator,..." This is equal to approximately 50 gallons per day. The inputs to the current MPS2 accident analyses also assume a primary-to-secondary leak rate of 50 gallons per day.

These leakage levels are extremely low and result in significant limitations in plant operation.

In order to provide additional operational margin, the design basis accidents were reanalyzed under the alternate source term methodology using an assumed primary-to-secondary leak rate of 150 gallons per day.

The offsite dose consequences were found to be acceptable in accordance with 10 CFR 50.67.

(The previous analyses utilized 10 CFR 100.)

The license amendment associated with this change to the accident analysis of record has been submitted to the NRC for approval as a separate license amendment request.

NRC Question No. 2 In the submittal, the definition of LEAKAGE was modified. Please confirm that the modification (i.e., excepting controlled leakage from the definition of identified leakage) is consistent with the currently-approved design and licensing basis. If

Serial No. 06-001A Docket No. 50-336 Response to Request for Additional Information Page 2 of 8 not consistent, please identify all changes and provide the technical justification for them.

DNC Response The modification is consistent with the currently approved design and licensing basis. Excepting controlled leakage from the definition of identified leakage is implicit in the current technical specifications. The change adds clarity by making the exception explicit.

NRC Question No. 3 There are several modifications to the Bases section for reactor coolant system operational leakage that go beyond the changes addressed in TSTF-449.

Please confirm that all changes are consistent with the currently-approved design and licensing basis. The NRC staff notes that DNC has primarily adopted the wording in the standard technical specifications (STS); however, DNC has also incorporated several statements from the existing Bases (e.g., last paragraph in the limiting condition of operation (LCO) section of insert B 314.4.6.2-OL). Please discuss why these latter statements are needed and discuss whether they are consistent with the STS.

DNC Response Modifications to the Bases section are consistent with the current licensing and design basis. DNC has adopted, where possible, the exact wording found in the standard technical specifications (STS). The statements from the existing Bases that have been incorporated are provided for clarity or because the licensing basis is different from the STS. Table 1-1 below lists instances of differences between the STS and the MPS2 proposed TS along with explanations for the differences. The added statements are consistent with the STS.

Definition of PRESSURE BOUNDARY LEAKAGE TABLE 1-1 (RCPB).

Insert Number B314.4.6.2-OL -

the acronym.

Wording that is Different reactor coolant pressure boundary Explanation Wording added to define

Insert Number jefinition of

'rimary to Secondary

-EAKAGE

hrough Any One Steam Generator B314.4.6.2-OL -

definition of IDENTIFIED LEAKAGE B314.4.6.2-OL -

below definition of IDENTIFIED LEAKAGE B314.4.6.2-OL -

ACTION b.

Serial No. 06-001A Docket No. 50-336 Response to Request for Additional Information Page 3 of 8 Wording that is Different The main steam line break (MSLB) accident analysis assumes a primary-to-secondary LEAKAGE of 150 gallons per day per SG. Operational primary-to-secondary LEAKAGE limit is set at 75 gallons per day per SG.

but does not include PRESSURE BOUNDARY LEAKAGE, or CONTROLLED LEAKAGE.

The IDENTIFIED LEAKAGE and UNIDENTIFIED LEAKAGE limits listed in LC0 3.4.6.2 only apply to the reactor coolant system pressure boundary within the containment. Leakage outside of the second isolation valve for containment, which is included in the RCS Leak Rate Calculation, is not considered RCS LEAKAGE and can be subtracted from RCS UNIDENTIFIED LEAKAGE. The definitions for IDENTIFIED LEAKAGE and UNIDENTIFIED LEAKAGE are provided in the technical specifications definitions section, definition 1.14.

It should be noted that LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

Explanation This wording has been added because it is specific to the Millstone Unit 2 MSLB accident analysis which assumes a 150 gallon per day primary-to-secondary leak rate. The Tech Spec leakage limit is set at 75 gallons per day in order to provide margin to the leak rate assumed in the accident analvses.

Wording which defines CONTROLLED LEAKAGE deleted as it is a defined term.

Millstone Unit 2 specific terminology. Wording added for clarity to explicitly identify the physical limits of PRESSURE BOUNDARY LEAKAGE.

Insert Number 6314.4.6.2-OL -

SURVEILLANCE REQUIREMENTS B314.4.6.2-OL -

SURVEILLANCE REQUIREMENTS 6314.4.6.2-OL -

SURVEILLANCE REQUIREMENTS

& TS 4.4.6.2.2 6314.4.6.2-OL -

APPLICABLE SAFETY ANALYSES -

OPERATIONAL LEAKAGE B314.4.6.2-OL -

APPLICABLE SAFETY ANALYSES -

OPERATIONAL LEAKAGE B314.4.6.2-OL -

APPLICABLE SAFETY ANALYSES -

OPERATIONAL LEAKAGE Serial No. 06-001A Docket No. 50-336 Response to Request for Additional Information Page 4 of 8 Wording that is Different seal "leakoff' flows (2 instances)

LC0 3.4.6.1, "Leakage Detection Systems."

LEAKAGE of "75" gallons per day (several instances) secondary LEAKAGE "from any one steam generator (SG) of 150 gpd or" "main" steam line break ("MVSLB) accident (2 instances)

The FSAR (Reference 3) analysis for SGTR assumes the contaminated secondary fluid is only briefly released via safety valves or atmospheric dump valves.

Explanation Millstone Unit 2 specific terminology. There is no "injection" flow at Millstone Unit 2. "Return" flow is called "leakoff' at Millstone Unit 2.

This is the proper Millstone Unit 2 reference.

The Tech Spec leakage limit is set at 75 gallons per day in order to provide margin to the leak rate assumed in the accident analyses.

Words added for clarity.

Millstone Unit 2 specific terminology.

Deleted reference to condenser and reference to "inconsequential." Neither are supported by the Millstone Unit 2 accident analysis.

Serial No. 06-001A Docket No. 50-336 Response to Request for Additional Information Page 5 of 8 Insert Number B314.4.6.2-OL -

Explanation Millstone Unit 2 specific -

for consistency with accident analysis.

Wording that is Different the MSLB accident assumes "1 50 APPLICABLE SAFETY ANALYSES -

OPERATIONAL LEAKAGE Note: The quotes used in the "Wording that is Different" column denote the subject of the explanation provided. Other minor differences are related to numbering of the subsections of the referenced technical specifications.

Differences in these instances are to maintain consistency with the numbering scheme of the MPS2 Technical Specifications.

gpd primary-to-secondary LEAKAGE is through the affected generator and 150 gpd from the intact SG as an initial condition."

NRC Question No. 4 As currently written, there are several references to tube sleeving in the proposed TSs (e.g., in TSs 3.4.5, 4.4.5, 6.9.1.9, and 6.28 and the proposed Bases).

Please confirm that the analysis previously submitted to support sleeving at MPS2 specifically addressed the replacement steam generators, and provide the reference for this analysis. If the analysis did not address sleeving for the replacement steam generators, please discuss plans to either remove the sleeving option from the TSs (in all applicable sections) or submit the relevant testing and analysis for NRC staff review and approval (or basis for why the NRC staff review and approval is not required).

In the event DNC can demonstrate that sleeving is authorized for the replacement steam generators (per the information provided in response to the above questions), discuss the plans to incorporate the inspection, repair, and installation requirements into proposed TS 6.26. For example, discuss the plans to incorporate the repair criteria for the authorized sleeving methods in proposed TS 6.26.c. and to remove the "Note" regarding tube repair from proposed TS 3.4.5. With respect to this "Note", the NRC staff observes that "defective" is no longer defined in the proposed TS. In addition, discuss plans to modify the proposed Bases (third paragraph in proposed insert B 314.4.5) to indicate that a steam generator tube includes any repairs made to it (which would make it consistent with TSTF-449).

DNC Response DNC has reviewed the licensing basis related to sleeving of steam generator tubes as an alternate repair criteria for MPS2 and concludes that the scope of the associated analysis did not extend to the replacement steam generators.

Therefore, DNC is removing reference to sleeving of steam generator tubes as a

Serial No. 06-001A Docket No. 50-336 Response to Request for Additional Information Page 6 of 8 repair option for MPS2. See Attachment 2 for replacement marked up pages.

The marked up bases pages are provided for information only.

NRC Question No. 5 The NRC staff notes that DNC is proposing to increase the primary-to-secondary leakage limit from 0.035 gallons per minute (approximately 50 (gallons per day) gpd) to 75 gpd. Please provide the technical basis for this change in terms of demonstrating that 75 gpd is less than the leakage rate associated with the largest crack that could exist and still satisfy the performance criteria.

Alternatively, discuss plans to revise the proposal to leave the limit at 50 gpd.

In addition, please discuss whether "leakage" in TS 3.4.6.2.c should be capitalized (consistent with TSTF-449).

DNC Response The Regulatory Guide 1.121 analysis of record confirmed that 150 gpd is less than the leakage rate associated with the largest crack that could exist and still satisfy the performance criteria. The leakage rate associated with the largest crack was reviewed in conjunction with the installation of replacement steam generators and was found to be bounding. Furthermore, 150 gpd is also consistent with the TSTF value. The 75 gpd limit proposed in the Technical Specifications is likewise bounded by the existing RG 1.I21 analysis.

"Leakage" in TS 3.4.6.2.c will be capitalized (consistent with TSTF-449). See for replacement marked up pages. The marked up bases pages are provided for information only.

NRC Question No. 6 In proposed surveillance requirement (SR) 4.4.6.2.2.c., DNC has a note that "the provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or MODE 4." A similar note exists in the current TSs, although it does not reference MODE 3. Please clarify the purpose of this note and the reason for adding MODE 3 to the "Note1'. Alternatively, discuss plans for making the proposal consistent with TSTF-449 and removing this "Note" in its entirety.

DNC Response The last sentence of the "Note1' which states, "The provisions of specification 4.0.4 are not applicable for entry into MODE 3 or MODE 4", will be removed.

See Attachment 2 for replacement marked up pages. The marked up bases pages are provided for information only.

Serial No. 06-001A Docket No. 50-336 Response to Request for Additional Information Page 7 of 8 NRC Question No. 7 With respect to the accident-induced leakage limit in proposed TS 6.26.b.2, the NRC staff notes that the second sentence is inconsistent with TSTF-449. As currently written, the proposed wording for the second sentence appears to have been written to be consistent with the current design and licensing basis (or the proposed design and licensing basis - see question 1). However, the second sentence in the accident-induced leakage performance criteria in TSTF-449 is not intended to be an interpretation of the current design and licensing basis.

Rather, the second sentence (in the accident-induced leakage performance criteria in TSTF-449) is intended to ensure that the potential for induced leakage during severe accidents will be maintained at a level that will not increase risk.

This is discussed in the NRC staffs generic safety evaluation of TSTF-449. As a result, please discuss plans to modify the second sentence to be consistent with TSTF-449.

DNC Response The sentence will be modified to read, "Leakage is not to exceed 150 gpd per SG." See Attachment 2 for replacement marked up pages. The marked up bases pages are provided for information only.

NRC Question No. 8 In insert B 314.4.6.2-0L under LC0 "c", there are two extra sentences included in the proposal that are not in TSTF-449. With respect to the last sentence, it appears to indicate that limiting the operational primary-to-secondary leak rate to 75 gpd will ensure that the accident-induced leakage limit will be met. The basis for this statement is not clear since there is operating experience that indicates that operational leakage can go from low levels (e.g., less than 75 gpd) to several hundred gallons per minute (i.e., a tube rupture) during normal operation. As a result, please discuss plans for removing this statement.

DNC Response The last sentence will be removed. See Attachment 2 for replacement marked up pages. The marked up bases pages are provided for information only.

Serial No. 06-001A Docket No. 50-336 Response to Request for Additional Information Page 8 of 8 NRC Question No. 9 In insert B 314.4.6.2-OL, some of the information in the "Applicable Safety Analyses - Operational Leakage" section duplicates some of the material under LC0 "c". Please discuss plans to remove this redundancy (i.e., it does not appear that the information is needed in the "Applicable Safety Analyses -

Operational Leakage" section).

DNC Response The last four sentences in first paragraph of the "APPLICABLE SAFETY ANALYSES - OPERATIONAL LEAKAGE" section of insert B314 4.6.2-OL will be deleted. See Attachment 2 for replacement marked up pages. The marked up bases pages are provided for information only.

Serial No. 06-001A Docket No.05-336 ATTACHMENT 2 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY (TAC NO. MD25701 REVISED MARKED UP PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No. 06-001A Docket No. 50-336 INSERT 3.4.5 STEAM GENERATOR TUBE INTEGRITY

\\

LIMITING CONDITION FOR OPERATION 3.4.5 Steam Generator (SG) tube integrity shall be maintained.

AND -

the tube repair criteria shall be ccordance with the Steam Generator APPLICABILITY:

MODES 1, 2, 3, and 4.

ore SG tubes satisfying the tube repair criteria and not plugged ccordance with the Steam Generator Program:

1.

Verify tube integrity of the affected tube@) is maintained until the next refueling outage o inspection within 7 days, and P

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

b. With Required ACTION and associated Completion Time of ACTION a. not met or SG tube integrity not maintained:

1.

Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

2. Be in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Serial No. 06-001A Docket No. 50-336 SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam Generator at satisfies the tube repair criteria rdance with the Steam Generator Program OWN following a SG tube inspection.

Serial No. 06-001A Docket No. 50-336 REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System akage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, 1 GPM UNIDENTIFIED 10 GPM IDENTIFIED LEAKAG

\\APPLICABILITY MODES I, 2,3 and 4.

With any PRESSURE BOUNDARY LEAKAGE, be in COLD System leakage hours.

SURVEILLANCE REQUIREMENTS to secondary leakage shall be demonstrated to be within the ary to secondary leak rate determination at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . T on 4.0.4 are not applicable for entry into MODE 4.

MILLSTONE - UNIT 2 Amendment No. 25,3;r, 82,85, W,

=,-l%.urs,a

Serial No. 06-001A Docket No. 50-336 INSERT 3.4.6.2 ACTION:

a. With any RCS operational LEAKAGE not within limits for reasons other than PRESSURE BOUNDARY LEAKAGE or primary to secondary LEAKAGE, reduce LEAKAGE to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. With ACTION and associated Completion Time of ACTION a. not met, or PRESSURE BOUNDARY LEAKAGE exists, or primary to secondary LEAKAGE not within limits, be in HOT STANDBY Lvithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2.1...................................

NOTES -----------------------------------------------

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

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Verify primary to secondary LEAKAGE is I 75 gallons per day through any one SG at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Serial No. 06-001A Docket No. 50-336 INSERT 6.9.1.9 STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.9 A report shall be submitted within 180 days after initial entry into MODE 4 following completion of an inspection performed in accordance with TS 6.26, 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, induced indications,

i.

Repair method utilized and the number of tubes repaired by each repair

Serial No. 06-001A Docket No. 50-336 INSERT 6.26 6.26 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 foundJ' 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 or repair of tubes. Condition monitoring assessments shall be outage during which the SG tubes are inspected, confirm that the performance criteria are being met.

b.

Provision for 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.

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.

Serial No. 06-001A Docket No. 50-336

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube ruDture, shall not exceed the leakaae 3.4.6.2, "Reactor Coolant System Operational LEAKAGE."

c.

Provisions for SG tube repair criteria: Tubes fou d by inservice inspection to contain flaws with a depth equal to or ex ng 40% of the nominal tube wall thickness shall be plugged d r repaire.

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.l., 6.2, 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.

lnspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

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

Serial No. 06-001A Docket No. 50-336

3.

If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not asso6iated with a crack(s), then the indication need not be treated as a crack.

e.

Provisio

Serial No. 06-001A Docket No. 50-336 INSERT B 314.4.5 314.4.5 STEAM GENERATOR TUBE INTEGRITY LC0 The LC0 requires th be maintained. The satisfy the repair criteria b accordance with the Stea During a SG inspection, any inspected tu Steam Generator Program repair criteria i removed from service by plugging. If a tu to satisfy the repair criteria but was not plugged, or repaired, 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 6.26, "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 tubes 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

Serial No. 06-001A Docket No. 50-336 cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced during MODES 1,2,3, and 4.

RCS conditions are far less challenging during MODES 5 and 6 than during MODES 1,2,3, and 4. During MODES 5 and 6, primary to secondary differential is low, resulting in lower stresses and reduced potential for LEAKAGE.

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 operation, and subsequent affected SG tubes are governed by subsequent ACTION entry and application of associated ACTIONS.

a.1 and a.2 ACTION a. applies if it is discovered that one or ore SG dhe tube tubes examined in an inservice insp repair criteria but were not plugge or re aire accordance with the Steam Genera o rog am as required of SG tube integrity of the made. Steam generator tube e SG performance criteria rator Program. The SG repair degradation that allow for ns while still providing orrnance criteria will continue to ne if a SG tube that should have s tube integrity, an evaluation onstrates 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

Serial No. 06-001A Docket No. 50-336 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. lnspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of TS 4.4.5.1. 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 and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.26 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

During an SG inspection, any inspected tube the Steam Generator Program repair criteria i removed from service by plugging. The tube delineated in Specification 6.26 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube 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.

Serial No. 06-001A 1

Docket No. 50-336 Q/'

'steam generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program.

BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LC0 3.4.1.I, "STARTUP and POWER OPERATION," LC0 3.4.1 -2, "HOT STANDBY,"

LC0 3.4.1.3, "HOT SHUTDOWN," and LC0 3.4.1.4, "COLD SHUTDOWN, 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 6.26, "Steam Generator (SG) Program,"

requires that a program be established and implemented to

Serial No. 06-001A Docket No. 50-336 INSERT B 314.4.6.2-01.

LC0 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 LC0 could result in continued degradation of the reactor coolant pressure boundary (RCPB). LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b.

UNIDENTIFIED LEAKAGE One gallon per minute (gpm) of UNIDENTIFIED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LC0 could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c.

Primarv to Secondarv LEAKAGE throuah Anv One Steam Generator:

The limit of 75 gallons per day per Steam Generator (SG) is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Reference 4) and the Accident Analyses described in the FSAR (Reference 3). 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 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 Program is an effective measure for minimizing the frequency of steam generator tube ruptures. The main

Serial No. 06-001A Docket No. 50-336 The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the IDENTIFIED LEAKAGE from the UNIDENTIFIED LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS LEAKAGE detection.

This LC0 deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analysis radiation release assumptions from being exceeded. The consequences of violating this LC0 include the possibility of a loss of coolant accident (LOCA).

APPLICABLE SAFETY ANALYSES - OPERATIONAL LEAKAGE Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from any one steam generator (SG) of 150 gpd or from all SGs of 300 gpd as a result of accident induced conditions. The LC0 requirement to limit primary to secondary LEAKAGE through any one SG to less than or Guidelines (Reference 4) with conservatism added to account for the MPS2 radiological analysis. 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 limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage.

The operational leakage rate criterion, as modified to account for the MPS2 radiological analysis, in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

e dose releases outside containment resulting from a main steam line break (MSLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to

Serial No. 06-001A Docket No.05-423 ATTACHMENT 3 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY (TAC NO. MD25711 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No. 06-001A Docket No. 50-423 Response to Request for Additional Information Page 1 of 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION MILLSTONE POWER STATION UNIT 3 By letter dated May 31, 2006, Dominion Nuclear Connecticut, Inc. (DNC),

submitted a request to revise the steam generator tube integrity requirements in the Technical Specifications (TSs) for Millstone Power Station, Unit No. 3 (MPS3).

In order for the Nuclear Regulatory Commission (NRC) staff to complete its review, responses to the following questions are needed.

NRC Question No. 1 Proposed TS 1.16.2.d indicates that reactor coolant system pressure isolation valve leakage shall be considered identified leakage. This definition does not appear to currently be in the TSs nor is it directly specified in the STS (although the Bases Section for pressure isolation valve leakage in the STS indicates that leakage through both pressure isolation valves in series is identified leakage).

Please discuss the reason for adding this item to the definition of "identified leakage" since it deviates from the STS.

DNC Response Upon review, DNC agrees that inclusion of proposed TS 1.16.2.d is inconsistent with the STS. Attachment 4 provides marked up pages reflecting removal of TS 1.16.2.d, respectively. The marked up bases pages are provided for information only.

NRC Question No. 2 In proposed surveillance requirement (SR) 4.4.6.2.1.d., DNC has a note that "the provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or MODE 4." A similar note exists in the current TS. Please clarify the purpose of this note. Alternatively, discuss plans for making the proposal consistent with TSTF-449 and removing this "Note1' in its entirety.

DNC Response The last sentence of the "Note" which states, "The provisions of specification 4.0.4 are not applicable for entry into MODE 3 or MODE 4", will be removed.

See Attachment 4 for replacement marked up pages.

Serial No. 06-001A Docket No. 50-423 Response to Request for Additional Information Page 2 of 3 NRC Question No. 3 With respect to the accident-induced leakage limit in proposed TS 6.8.4.g.b.2, the NRC staff notes that the second sentence is inconsistent with TSTF-449. As currently written, the proposed wording for the second sentence appears to have been written to be consistent with the current design and licensing basis (or the proposed design and licensing basis - see question 1 for MPS2). However, the second sentence in the accident-induced leakage performance criteria in TSTF-449 is not intended to be an interpretation of the current design and licensing basis.

Rather, the second sentence (in the accident-induced leakage performance criteria in TSTF-449) is intended to ensure that the potential for induced leakage during severe accidents will be maintained at a level that will not increase risk. This is discussed in the NRC staff's generic safety evaluation of TSTF-449. As a result, please discuss plans to modify the second sentence to be consistent with TSTF-449.

DNC Response The sentence will be modified to read, "Leakage is not to exceed 500 gpd per SG." See Attachment 4 for replacement marked up pages.

NRC Question No. 4 In proposed TS 6.8.4.g.d, there appears to be two typographical errors. In the second sentence, it appears that "tube" has been deleted before the word "outlet1,. In the fourth sentence, there is an extra period following "d.1". Please discuss plans to correct these apparent typographical errors.

DNC Response The typographical errors have been corrected.

See Attachment 4 for replacement marked up pages. The marked up bases pages are provided for information only.

NRC Question No. 5 There are several modifications to the Bases section for reactor coolant system operational leakage that go beyond the changes addressed in TSTF-449.

Please confirm that all changes are consistent with the currently-approved design and licensing basis.

The NRC staff notes that DNC has primarily adopted the wording in the STS; however, DNC has not done this in all cases.

For example, DNC did not incorporate into the "Applicability" section a discussion on pressure isolation

Serial No. 06-001A Docket No. 50-423 Response to Request for Additional Information Page 3 of 3 valve leakage which is present in the STS. In addition, DNC did not incorporate some of the Bases information from the pressure isolation valve section in the STS (refer to question 1). Please discuss the reason for these apparent omissions.

DNC Response DNC will add the Standard Technical Specifications (STS) discussion of pressure isolation valve leakage as the third paragraph under APPLICABILITY in BASES section 314.4.6.2. Bases changes are consistent with currently-approved design and licensing basis. See Attachment 4 for replacement marked up pages. The marked up bases pages are provided for information only.

NRC Question No. 6 In the "Actions" section of the Bases section for reactor coolant system operational leakage, DNC indicated that cold shutdown must be achieved within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This statement is inconsistent with the actual requirement which indicates that the reactor should be in cold shutdown "within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Please discuss plans to make the Bases and the Action requirements consistent.

DNC Response DNC has changed the "Actions" section of the bases to be consistent with the actual requirement. Note that the language selected, while different from the TSTF, is consistent with other MPS3 technical specification sections and the MPS3 licensing basis. See Attachment 4 for replacement marked up pages.

The marked up bases pages are provided for information only.

Serial No. 06-001A Docket No. 50-423 ATTACHMENT 4 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY (TAC NO. MD2571)

REVISED MARKED UP PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No. 06-001A Docket No.05-423 INSERT 1.16 LEAKAGE 1.I 6 LEAKAGE shall be:

1.1 6.1 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals, and 1.16.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

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 1.16.3 PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in a RCS component body, pipe wall, or vessel wall, and 1.I 6.4 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all LEAKAGE which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

Serial No. 06-001A Docket No.05-423 INSERT 4.4.6.2.1.d INSERT 4.4.6.2.1.e em...........................................

NOTE..................... -..........................

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

Verification that primary to secondary LEAKAGE is I 150 gallons per day through any one Steam Generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and;

Serial No. 06-001A Docket No.05-423

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 rom a y one SG not isolated from the RCS.

3. The operational LEAKAGE performance criterion is specified in RCS LC0 3.4.6.2, "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.

any type (e.g., volumetric flaws, axial be present along the length of the the tube inlet to the satisfy the intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

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

2.

lnspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

Serial No. 06-001A Docket No.05-423 The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2250 psia. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the safety analyses.

A limit of 40 gpm is placed on CONTROLLED LEAKAGE.

f.

RCS Pressure Isolation Valve LEAKAGE The specified allowable leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure.

It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced

~otentials for LEAKAGE.

b., c. UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE or RCS pressure isolation valve LEAKAGE in excess of the LC0 limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 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.

a., b. c. If any PRESSURE BOUNDARY LEAKAGE exists, or primary to secondary LEAKAGE is not within limits, or if UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or RCS pressure isolation valve LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 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

Serial No. 06-001A Docket No.05-423 Insert B 314 4.6.2 Applicability LC0 3.4.6.2.f, RCS Pressure Isolation Valve (PIV) Leakage, measures leakage through each individual PIV and can impact this LCO. Of the two PlVs 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.

Serial No. 06-001A Docket No.05-423 LEAKAGE. Th boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In COLD SHUTDOWN, the pressure stresses acting on the reactor coolant pressure boundary are much lower, and further deterioration is much less likely.

SURVEILLANCE REQUIREMENTS CONTROLLED LEAKAGE is determined under a set of reference conditions, listed below:

a.

One Charging Pump in operation.

b.

RCS pressure at 2250 +I-20 psia.

By limiting CONTROLLED LEAKAGE to 40 gprn during normal operation, it can be assured that during an Sl with only one charging pump injecting, RCP seal injection flow will continue to remain less than 80 gpm as assumed in the accident analysis. When the seal injection throttle valves are set with a normal charging lineup, the throttle valve position bounds conditions where higher charging header pressures could exist. Therefore, conditions which create higher charging header pressures such as an isolated charging line, or two pumps in service are bounded by the single pump-normal system lineup surveillance configuration. Basic accident analysis assumptions are that 80 gpm flow is provided to the seals by a single pump in a runout condition.

Verifying RCS LEAKAGE to be within the LC0 limits ensures the integrity of the reactor coolant pressure boundary is maintained. PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance of an RCS water inventory balance.

v t e Letter Sequence Response to RAI
TAC:MD2570, Steam Generator Tube Integrity (Approved, Closed) TAC:MD2571, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request Acceptance... Administration Questions 20 March 2007 Answers 14 February 2007 26 April 2007 Results Approval Other: ML071560531
MONTHYEAR2007-03-20 [Table View]

ML070460048

Text

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