ML20206G827

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Application for Amends to Licenses NPF-4 & NPF-7,deleting &/Or Relocating Addl primary-to-secondary Leak Rate Limits & Enhanced Leakage Monitoring Requirements Imposed Following 1987 SG Tube Rupture Event
ML20206G827
Person / Time
Site: North Anna  Dominion icon.png
Issue date: 05/03/1999
From: Christian D
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20206G830 List:
References
99-263, NUDOCS 9905100156
Download: ML20206G827 (11)
v • d • e


Text

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Vincisir IsarcTunc ANi> I'owEn C<nieANv Rn:ssuoNas, VIRGINI A 2326:

May 3, 1999 U.S. Nuclear Regulatory Commission Serial No.99-263 Attention: Document Control Desk NL&OS/GSS/ETS R1

. Washington, D.C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF-7 Gentlemen:

~ VIRGINlA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CHANGES PRIMARY-TO-SECONDARY LEAKAGE RATE PROVISIONS AND DETECTION SYSTEM OPERABILITY REQUIREMENTS Pursuant to 10 CFR 50.90, Virginia Electric and Power Company requests amendments, in the form of changes to the Technical Specifications and to Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed changes will delete and/or relocate the additional primary-to-secondary leak rate limits and enhanced leakage monitoring requirements imposed following the 1987 steam generator tube rupture event. With the completion of steam generator replacement, these additional rettrictions are no longer necessary as part of the North Anna Technical Specifications. A discussion of the proposed Technical Specifications changes is provided in Attachment 1.

f The proposed Technical Specifications changes have been reviewed and approved by s the Station Nuclear Safety and Operating Committee and the Management Safety oI Review Committee. It has been determined that the proposed Technical Specifications changes do not involve an unreviewed safety question as defined in 10 CFR 50.59 or a

\o significant hazards consideration as defined in 10 CFR 50.92. The proposed Technical Specifications changes are provided as a mark-up in Attachment 2 and a typed version in Attachment 3. The basis for our determination that the changes do not involve a significant hazards consideration is provided in Attachment 4.

9905100156 990503 i PDR ADOCK 05000338 I

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if you have any further questions, please contact us. l l

Very truly yours, D. A. Christian Vice President- Nuclear Operations Attachments

1. Discussion of Changes
2. Mark-up of Technical Specifications Changes
3. Proposed Technical Specifications Changes
4. Significant Hazards Consideration Determination Commitments made in this letter: l
1. None cc: U.S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. J. Morgan NRC Senior Resident Inspector North Anna Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street Richmond, VA 23219 Mr. J. E. Reasor Old Dominion Electric Cooperative innsbrook Corporate Center 4201 Dominion Blvd.

Glen Allen, Virginia 23060

COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by D. A. Christian, who is Vice President - ,

Nuclear Operations, of Virginia Electric and Power Company. He has affirmed l 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 day of ALV ,19 b.

My Commission Expires: March 31,2000.

b]O0flkb) (YhLtd

/ Notary Public (SEAL)

i Attachment 1 Discussion of Changes l

North Anna Power Station Units 1 and 2 Virginia Electric and Power Company N-

Discussion of Changes introduction Pursuant to 10 CFR 50.90, Virginia Electric and Power Company requests changes to Technical Specifications Section 3.4.6, " Reactor Coolant System" Limiting Conditions for Operation (LCO), associated Action Statements, Surveillance Requirements (SR),

and Bases. The proposed changes will delete and/or relocate the additional stringent prirnary-to-secondary leak rate limits and enhanced leakage monitoring requirements that were imposed by Technical Specifications 3.4.6,3 and 3.4.6.4 when operating in Mode 1 above 50% power for the original steam generators. These requirements were established following the Unit 1 steam generator tube rupture (SGTR) event on July 15, 1987 to ensure timely detection of eiereasing primary-to-secondary leakage rates.

Additionally, portions of the enhanced leakage monitoring requirements for the N-16 radiation monitoring systems will be relocated to the Technical Requirements Manual (TRM).

Since the SGTR event, new steam generators have been installed for Unit i during the 1993 outage and Unit 2 during the 1995 outage with excellent performance. As a result, the more stringent leak rate limits and enhanced leakage monitoring are no longer deemed necessary. The existing Technical Specification requirements to limit Reactor Coolant System (RCS) operating leakage as specified in LCO 3.4.6.2, associated Action Statements, and Surveillance Requirements 4.4.6.2.1 and 4.4.6.2.2 are bounded by the existing accidents analyzed and will be retained.

The proposed changes do not create an unreviewed safety question and are consistent with the improved Technical Specification RCS operational leakage limits specified in NUREG-1431.

Licensing and Design Bases 10CFR50, Appendix A, General Design Criteria 30 states that means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage. Additionally, Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems" describes the acceptable methods for selecting leakage detection systems.

The RCS leakage detection systems are designed to monitor and detect leakage from the Reactor Coolant Pressure Boundary during normal operations to provide prompt and quantitative information to the operators to permit immediate corrective actions should a leak be detrimental to the safety of the plant.

Page 1 of 7

Discussion On July 15,1987, North Anna Unit 1 experienced a steam generator tube rupture event due to fatigue caused by flow-induced vibration. As part of the corrective actions, very conservative corrective measures were implemented to reduce the probability of fatigue-induced tube rupture. These measures included the installation of downcomer flow resistance plates to reduce the source of loads associated with the fatigue mechanism in the U-bend area and the preventive plugging of potentially susceptible tubes. Additionally, enhanced leakage monitoring instrumentation was installed and more stringent leak rates were established for operation in Mode 1 above 50% power in the unlikely event that the downcomer modification and preventive plugging were unsuccessful in preventing occurrence of a similar fatigue failure.

By letter dated December 4,1987, Virginia Electric and Power Company requested Amendments to the Technical Specifications that would incorporate the more stringent primary-to-secondary coolant system leakage limits, and establish operability and surveillance requirements to assure the operability of the existing and new N-16 instrumentation used to verify compliance with the revised leakage limits. These stringent leakage limits and increased monitoring requirements were incorporated into Technical Specifications by Amendments 109 for Unit 1 and 95 for Unit 2 issued on December 12,1988.

Since the SGTR event, new steam generators have been installed in Units 1 and 2, and the more stringent leakage rate limits and enhanced leakage monitoring are no longer deemed necessary. Therefore, the proposed changes retain the existing Technical Specification 3.4.6.2 requirements to limit Reactor Coolant System (RCS) leakage and provide a clarification footnote for Surveillance Requirement 4.4.6.2.1.d. The purpose of the RCS Operational Leakage LCO is to limit system operation in the presence of leakage from the identified sources to amounts that do not compromise safety. The total steam generator tube leakage limit of 1 gpm for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break, i The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

Therefore, pursuant to 10 CFR 50.90, Virginia Electric and Power Company requests changes to the Technical Specifications Section 3.4.6, " Reactor Coolant System" Limiting Conditions for Operation, associated Action Statements, Surveillance Requirements, and Bases. The proposed changes retain and/or relocate the original requirements and delete the more stringent primary-to-secondary leakage limits and increased monitoring requirements of Technical Specifications 3.4.6.3 and 3.4.6.4 that were imposed when operating in Mode 1 above 50% power following the steam generator tube rupture (SGTR) event at Unit 1 on July 15, 1987. Portions of the Page 2 of 7

= ,

enhanced leakage monitoring requirements for the N-16 radiation monitoring systems will be relocated to the TRM. The proposed changes do not create an unreviewed

. safety question and are consistent with the improved Technical Specification requirements specified in NUREG-1431. l The changes remove and/or relocate the more stringent primary-to-secondary leakage rates and surveillance requirements imposed as a result of the SGTR in 1987, when operating greater than 50% power, which are no longer deemed necessary with the replacement of the steam generators at North Anna Units 1 and 2. These added very conservative requirements ensured that if in the event of a fatigue induced crack were to occur in one or more degraded steam generators then the resulting leak would be detected in sufficient time to conduct an orderly shutdown prior to catastrophic tube failure.

Changes will also add a footnote to Surveillance Requirement 4.4.6.2.1.d stating that primary-to-secondary leakage surveillance is not required below 50% power. Primary- j to-secondary leakage is measured by the performance of RCS water inventory balance j in conjunction wlth radioisotope monitoring. The RCS water inventory balance is

{

performed at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with the reactor at steady state operating conditions. A steady state condition above 50% power comparison between the secondary radioisotope concentrations and the primary radioisotope concentrations is necessary for an accurate primary-to-secondary leak rate determination.

The relocated requirements for the N-16 radiation monitoring system to the TRM will provide an additional method that establishes operability and surveillance requirements l to verify compliance with the Technical Specifications primary-to-secondary leakage limits l Specific Changes - Technical Specifications LCO 3.4.6.2.c Delete reference to "not isolated from the Reactor Coolant System" and footnote "". LCO will now read:

Existing LCO (Units 1 and 2)

c. 1 GPM total primary-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System, * (Unit 1)
c. 1 GPM total primary-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System, ** (Unit 2)

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1 Change to: f I

c. -1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator, LCO 3.4.6.2.c Delete following footnotes at bottom of page for Units 1 and 2:  ;
  • When in Mode 1 above 50% power, provisions of Specification 3.4.6.3 apply. (Unit 1)
    • When in Mode 1 above 50% power, provisions of Specification 3.4.6.3 apply. (Unit 2) l Surveillance Requirement 4.4.6.2.1.d Add footnote in surveillance requirement and at bottom of page as noted:

4.4.6.2.1.d -

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 /> during steady state operation **, and (Unit 1)

    • Primary-to-secondary leakage not required below 50% power l

I 4.4.6.2.1.d - 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 /> during steady state operation *, and (Unit 2)

TS Bases 3/4.4.5 Delete following 3"',4* and 5* paragraphs as follow:

l "It has been determined, however, that certain conditions within the steam generator i may produce limited displacement fluidelastic instability in the tube bundle that may ,

result in fatigue failure of a tube. Modifications have been accomplished in all steam 1 generators consisting of installation 'of downcomer resistance plates and preventive I plugging of potentially susceptible tubes. Even though these measures are considered to have been very conservative and highly effective in reducing the probability of fatigue induced tube rupture, enhanced leakage monitoring and more stringent leak rate limits have been established. Leakage is now limited to 100 gpd (rather than 500 gpd) per steam generator when operating at greater than 50% power. Cyclic life analysis of fatigue induced tube cracks has shown that, assuming a post-modification maximum stress amplitude of 7 ksi, a leak rate of up to 500 gpd would be reached some 90 minutes prior to tube rupture. Therefore, the 100 gpd leak rate limit is bounding since

a. the 100 gpd limit would be detected well in advance of reaching 500 gpd, b, the time required for leak rate detection and power reduction to less than 50% is expected to be less than 90 minutes, and Page 4 of 7
c. the maximum stress amplitude is anticipated to lie in the 5 ksi range which would allow for much earlier leak before break warning than would occur in the assumed 7 ksi case.

These assumptions also include an appropriate allowance for measurement uncertainty. (

References:

Virginia Electric and Power Co., " North Anna Unit 1 July 15, 1987 Steam Generator Tube Rupture Event Report, Revision 1, September 15,1987, and Westinghouse WCAP-11601, " North Anna Unit 1 Steam Generator Tube Rupture and Remedial Actions Technical Evaluation, September 1987").

This limit, along with the enhanced monitoring system, should provide sufficient notification to permit orderly shutdown prior to a potential tube rupture event. Leakage in excess of any of these limits will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged."

TS Bases 3/4.4.6.2 Delete following introduction phrase in the third sentence of the 4*

paragraph and the entire 5* paragraph as noted below:

' ge ere!, fer ? crt

! eperet!c et er be!ce' 50% perter, The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

l When operating at greater than 50% power, more stringent primary-to-secondary leakage limits of 300 gallons per day (GPD) total from all three steam generators and 100 gpd from an individual steam generator have been imposed. These limits ensure that in the event that a fatigue induced crack were to occur in one or more generators, the resulting leak would be detected in sufficient time to conduct an orderly shutdown prior to catastrophic tube failure. The limits on an increase in leakage of 60 gpd between surveillance intervals and for an increasing trend indicating that 100 gpd would be exceeded within 90 minutes ensure that, in the event of fatigue crack initiation, power can be reduced to a level below which propagation will not occur. In the latter case, the limit also provides for orderly shutdown since the 100 gpd limit is being approached. These leakage rates are conservative with regard to dosage contribution in that they are less than the previously analyzed total amount of 1 GPM and 500 GPD for any single steam generator.

S_afety Significance The inservice inspection program of the steam generator tubes is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation that leads to corrosion which may result in stress corrosion cracking.

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The extent of cracking during plant operation would be limited to the limitation of steam generator tube leakage between the primary and secondary coolant systems .

(primary-to-secondary leakage = 500 gpd per steam generator). Cracks having a  !

primary-to-secondary leakage rate less than 500 gpd during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gpd per steam generator can readily be detected by  !

radiation monitors of steam generator blowdown.

Virginia Electric and Power Company has reviewed the proposed Technical Specification changes and has determined that the proposed changes would not pose an unreviewed safety question. These proposed changes eliminate overly conservative primary-to-secondary leakage rate restrictions above 50% power, and i the operability and surveillance requirements for the leakage monitoring instrumentation. Specifically, operation of the North Anna Power Station in accordance with the proposed Technical Specification changes will not:

1. Involve a significant increase in the probability or consequences of an I accident previovely evaluated. )

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Eliminating the conservative primary-to-secondary leakage limits associated I I

with the replaced steam generators and the operability requirements for the leakage monitoring instrumentation does not change the operation of the plant. The steam generators will be operated, inspected, and maintained in the same manner. No new accident initiators are established as a result of the proposed changes. Therefore, the probability of occurrence is not increased for any accident previously evaluated.

Removing the conservative primary-to-secondary leakage limits associated l l

with the replaced steam generators and the operability requirements for the leakage monitoring instrumentation does not change the operation of the plant.. Although the conservative leakage limits are being deleted, the remaining leakage limits will maintain the dose rate, in the event of a tube rupture, within the analyzed limits. Therefore, there is no increase in the consequences of any accident previously analyzed.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated. 1 The proposed changes do not affect the operation of the plant. The steam generators will be operated, inspected, and maintained in the same manner.

There are no modifications to the plant or steam generators as a result of the change. No new accident or event initiators are created by the removal of the conservative primary-to-secondary leakage limits associated with the Page 6 of 7

4 replaced steam generators and the operability requirements for the leakage monitoring instrumentation. Therefore, the proposed changes do not create the possibility of any accident or malfunction of a different type.

3. Involve a significant reduction in the margin of safety as defined in the bases I on any Technical Specifications. j 1

The proposed changes have no effect on any safety analyses assumptions. t The remaining limits maintain primary-to-secondary leakage within the )

accident analysis assumptions. The proposed changes only eliminate overly conservative primary-to-secondary leakage requirements and the operability 1 and surveillance requirements for the leakage monitoring system associated with the replaced steam generators. Therefore, the proposed changes do not result in a reduction in a margin of safety.

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