ML20114A060

Amend 171 to License NPF-3,revising TS 3/4.4.5 & Bases to Allow Use of B&W Steam Generator Tube Sleeving Process to Effect Repairs of Defective Steam Generator Tubes

ML20114A060
Person / Time
Site:	Davis Besse
Issue date:	07/28/1992
From:	Hopkins J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20114A061	List: ML20114A060 ML20114A063
References
NUDOCS 9208100093
Download: ML20114A060 (12)

Text

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i UNITED STATES

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j NUCLEAR REGULATORY COMMISSION g

WASHINGTON. D.C. 20066 4

,,,,e TOLEDO EDISON COMPANY p

i CENTER 10R SERVICE COMPANY RQ-l THE CLEVELAND ELECTRIC llLUMINATING COMPANY i:

DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. 1 i

i AMENDMENT TO FACILITY-0PERATING LICENSE Amendmant' No.171 License No NPF-3 j

1.

The Nuclear Regulatory Commission (the Commission) has found that:-

A.

The application for amendment-by the Toledo Edison Company,:Centerior j

Service Company, and the Cleveland Electric 111uminating Company.

l (the licensees). dated August 16, 1991, supplemented February 3,-1992, 1

complies with the standards a'nd requirements of the' Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules a.nd regulations j

set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with_the application, the-provisions of-the Act, and the rules and regulations of the-

. Commission; C.

There is-reasonable assurance (i) that the:acti'ities authorized by v

l this amendment can-be conducted without endangering the health;and l

safety.of the public, and (ii)' that such activities will be-conducted:

in compliance with the Commission's regulations; D. 'The issuance of..this amendment will:not be inimical to the common

[

defense and security or to the. health and safety of the public; and i

E.

The issuance of this amendment ~~is in accordance with 10 CFR Part 51 af the Commission's regulations'and all applicable requirements have buen satisfied.

l 2.

Accordingly, the license is amended by. changes to the-Technical

. Specifications as indicated-in the attachment'to'this; license amendment, and paragraph 2.C.(2) of facility Operating License No. NPF-3-is hereby i

amended to read as follows:

9208100093 920728 PDR-ADOCK 05000346 P

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2 (a) Technical Specifications l

The Technical Specifications contained in Appendix A, as revised through Amendment No.171, are hereby incorporated in the license.

The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance.

FOR THE NUCLEAR REGULATORY COMMISSION D."

c Jon B. Hopkins, Sr.-Project Manager Project Directorate 111-3 Division of Reactor Projects Ill/IV/V Of fice of Nuclear Reactor Regulation.

Attachment:

Changes to the Technical Specifications Date of issuance: July 28,1992 l

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ATTACHMENT TO LICENSE AMENDMENT NO,171 FAClllTY OPERATINO IICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages, The revised pages are identified by amendment number and cor.tain vertical lines indicating the area of change, The corresponding overleaf pages arc-also provided to maintain document completeness.

Remove Insert 3/4 4-6 3/4 4-6 3/4 4-9 3/4 4-9 3/4 4-10 3/4 4-10 3/4 4-12 3/4 4-12 B 3/4 4-2 B 3/4 4-2 B 3/4 4-3 B-3/4 4-3 I

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I REACTOR COOLANT SYSTEM PRES 5URIZER 3

LIMITING CONDITION FOR OPERATION k

3.a.4 The pressurizer shall be OPERABLE with:

i a.

A steam bubble.

b.

A water level between 45 and 305 inches.

APPLICABILITY:

MODES 1 and 2.

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ACTION:

4 With the pressurizer inoperable, be in at.least HOT STANDBY with the control rod drive trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVE!LLANCE REQUIREMENTS f;

I a.a.

The pressurizer shall be demonstrated OPERABLE by verifying j

ressarizer level to be within limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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l REACTOR COOLANT SYSTEM

! STEAM GENERATORS LIMITING C0NDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE with a water level between 18 and 348 inches.

liAPPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

i, ll a.

With one or more steam generators inoperable due to steam-

't generator tube imperfections, restore the inoperable generator (s) j, to OPERABLE status prior to increasing T above 200 F.

l avg b.

With one or.more steam generators inoperable due to the water l

4 level being outside the limits,- be in at least HOT STANDBY l

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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SURVEILLANCE REQUIREMENTS i

,14.4.5.0 Each steam generator shall be demonstrated OPERABLE by -performance-

,:of the following augnented inservice inspection program and the requirements l

! of Specification 4.0.5.

i

!-4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam l

': generator shall oc determined OPERAdLE ourina shutdown by selecting and I

inspecting at least the minimum number of steam generators specified in l

Table 4-4.1.

[4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and j the corresponding action required shall be as specified in Table 4.4-2.

The iinservice inspection of steam generator tubes shall be performed at the _.

a j ! frequencies specified in S.necification 4.4.5.3 and the inspected tubes =shall l

i be verified acceptable per the acceptance criteria of Spacification 4.4.5.4'.

The tubes selected for each inservice inspection shall include at _least 3%

of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

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a.

The first sample inspection during each inservice inspection (subsequent to the baseline inspection) of each steam generator shall include:

1 1.

All tubes or tube sleeves that previously had detectable j

wall penetrations (> 20%) that have not been -plugged or ll repaired by sleeving in the affected area.

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2.

At least 50% of the tubes inspected snall be in those areas where experience has indicated potential problems.

DAVIS-BESSE - UNIT 1 3/44-6 Amendment No. 21,171

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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2.

A seismic occurrence greater than the Operating Basis Earthquake.

3.

A loss-of-coolant accident requiring actuation of -the engineered safeguards.

4.

A main steam line or feedwater line break, d.

The provisions of Specification 4.0.2 are not applicab'e.

4.4.5.4 Acceptance Criteria a.

As used in this Specification:

il Ji 1.

Tubing or Tube means that portion of the tube or tube sleeve which forms the primary system to secondary system boundary.

l 2.

Imperfection means an exception to the dir'ensions, finish or l-ll contour of a tube from that _ required by fabrication drawings or specifications.

Eddy-current testing indications below

i 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

3.

Degradation means a service-induced cracking, wastage, wear i

or general corrosion occurring on-either inside or outside l

of a tube.

4.

Degraded Tube means a tube containing imperfections > 20% of l

i the nominal wall thickness caused by-degradation that has l

not been repaired by sleeving in the affected area.

5.

t Degradation means the percentage of the tube wall thickness affected or removed by degradation, ji 6.

Efect means an imperfection of such severity that it exceeds

'l the repair limit. A defective tube is a tube containing a defect that has not been repaired by sleeving i_n the affected area or a sleeved tube that has a defect in the sleeve.

7.

Repair Limit means the imperfection depth at or beyond which l

the tube shall be removed from service by plugging or repaired by sleeving in the affected area because it may become unservice-able prior to the next inspection and is equal to 40% of the I

nominal tube wall thickness.

The Babcock and Wilcox process described in Topical Report BAW-2120P will be used for sleeving.

8.

Unserviceable describes the condition of a tube if it leaks or' l_

contains a defect _ large enough to affect its structural integ-E rity in the event of an Operating-Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above, 9.

Tube Inspection means an i_nspection of the steam generator l

tube from the point of entry completely to the point of exit.

DAVIS-BESSE, UNIT l-3/4 4-9 Amendment No. 2J,171

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I REACTOR COOLANT SYSTEM jSURVElLLANCEREQUIREMENTS(Continued)

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10.

Preservice Inspection means an inspection of the full j

length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection je shall be performed prior to initial POWER OPERATION using i!

the equipment and techniques expected to be used during ij subsequent inservice inspections.

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b.

The steam generator shall be determined OPERABLE af ter completing jj the corresponding actions (plug or repair by sleeving in' the

-l affected areas all tubes exceeding the repair limit and 'all tubes containing through-wall cracks) required by Table 4.4-2.

i4.4.5.5 Reports i

a.

Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

b, The complete results of the steam generator tube inservice inspection shall be submitted on an annual basis in a report-for the period in which this inspection was completed. This-report shall include:

1.

Number and extent of tubes inspected.

2.

Location and percent of wall-thickness penetration for each indication of an imperfection.

1 3.

Identification of tubes plugged or sleeved.

.l t t i

c.

Results of steam generator tube inspections which fall into Category C-3 and require-prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant. operation.

The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

4.4.5.6 The steam generator shall be demonstrated OPERABLE by verifying steam generator level to be within limits at least or.ce per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I 4.4.5.7 When steam generator tube inspection is performed as per Section 4.4.5.2, an additi_onal but totally separate inspection

!I shall be performed on special interest peripheral tubes in the I

vicinity of the secured internal auxiliary feedwater header.

This j

testing shall only be' required on the steam generator selected

'i for inspection, and the test shall require inspection only between DAVIS-BESSE, UNIT 1 3/4 4-10 Amendment No. E,27,62,171

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-_ q ThBLE 4.4-2 STEAM GFNERATOR Ttf9E INSPECTION

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W t*5 tra in 3RD SAMPLE IRSFECTION M

IST SWe!.E IRSPECTION 2ND.5AMPj,E J 3FECTIDM Sample size Result Action Regyfred

_ Result

, _ Action Reguired~

_~ Desult Action poquired g

+4 A einleum of C-1

' None N/A N/A N/A N/A H

S To p per p

s.a.

C Flug or repair try C-I Wone'

.N/A N/A eleeving defective tubes and inspect Plug or repair by

.C-1 mone additional 25 tubes C-2 eleaving defective C-2 Flag or repair try in this 9.O.

totes and Inesnt miseving defoetive adrtit ional 43 tubes tubes in this'5.C.

C-3 Perfore action for C-3 re-suit of first a

sample Perfare action for C-3 C-1 result of first N/A N/A to sample

~f C-3 Inspect all tubes in All other this S.G.,

plug or S.C.s are pone N/A N/A e

N repair try sleeving C-1 defective tubes and.

Some S.G.s Perfore action for.

Inspect 25 tubes in C-2 Irut no C-2 result of second N/A N/A each other 3,C.

additional sample S.C.

are Prompt notification-C-3

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to NRC pursuant to.

Additional Inspect all' tubes in l

leppelfication6.9.1 S.C.

is eacti S.C. and plog or

'N/A N/A C-1 repair by sleeving def eet t ve t abee.

Frompt notification to NRC pornuant to

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spectficatton 6.9.1

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D 1; "b where # is the number of steam gamerators in the unit, and n is the nereber of stems generators inspected (1) S =

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(2) For tubes tropected pursuant to 4.4.%2.br No action is required for C-1 results.' For C-2 results in one or ttrth tubes.

For C-3 results to one or both steen generators,

f steam generators plog or repair ty sleeving def er*tive plug or repair by s!*.eving defetive totes and provide prevet notifiestion of MPC persuant to Specification 6.6.

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! REACTOR COOLANT SYSTEM BASES 3/4.4.4 PRESSURI2ER A steam bubble in the Dressurizer ensures that the RCS is nut a hydraulically solid system and is capable of accommodating pressure surges during operation.

,' The steam bubble '1150 protects the pressurizer code safety valves and pilot j!operatedreliefvalveagainstwaterrelief.

4 The low level limit is based on providing enough water volume to prevent a reactor cooiant system low pressure condition that would actuate the Reactor i

Protection System or the Safety Feature Actuation System.

The high level limit' l

is based on providing enough steam volume,to prevent a pressurizer high level

as a result of any transient.

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The pilot operated relief valve and steam. bubble function to relieve RCS i

j

,, pressure during all design transients.

Operation of the pilot operated relief i

vaive minimizes the undesirable opening of the spring-loaded pressurizer code l

safety valves.

'!3/4.4.5 STEAM GENERATORS i

ll The Surveillance Requirements for inspection of the steam generator tubes ensure ll that the structural integrity of tHs purtion of the RCS will be maintained.

y The program for inservice inspection of steam generator tubes is based.on a

mocification of Regulatory Guide 1.83, Revision 1.. Inservice inspection of i

l steam generator tubing is essential in order to maintain surveillance of the j

conditions of the tubes in the event that there is evidence of mechanical damage

' or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generatur tubing also provides a means of characterizing the nature and cause of any tube t

j degradation so that corrective measures can be taken. A process. equivalent to l' the inspection method described in Topical Report BAW-2120P will be used for l

l' inservice inspection of steam generator tube sleeves. This inspection will provide ensurance of RCS integrity.

i

! The plant is expected to be operated in a manner ~such that the secondary j

coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. _If'the secondary coolant chemistry is not maintained within these chemistry limits, localized i

corrosion may likely result in stress corrosion cracking..'The extent'of.

4 cracking during plant operation would be limited by the limitation-of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 GPM).

Cracks having a primary-to-secondary leakage less than this limit.during operation will have an adequate margin of safety to withstand the loads imposed during normal i

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DAVIS-BESSE, UNIT 1 B 3/4 4-2 knendment No. DS,171 4

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REACTOR COOLANT SYSTEM BASES (Continued) operation and by postulated accidents.

Operating plants have demonstrated i

that primary-to-secondary leakage of 1 GPM can be detected by monitoring the secondary coolant.

Leakage in excess of this limit will_ require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired by sleeving in the affected areas.

l Wastage-type defects are unlikely with proper chemistry treatment.of the l secondary coolant. However, even if a defect should develop in service,

' it will be found during scheduled inservice steam generator tube examina-tions. As described in Topical Report BAW-2120P, degradation as smell as-l 20L through wall car, be detected in all areas of a tube sleeve except for i the roll expanded areas and the sleeve end, where the limit of detectability is 405 through wall.

Tubes with imperfections exceeding the repair limit.of

!; 405 of the nominal wall thickness will be plugged or repaired by sleeving

the affected areas.

Davis-Besse will evaluate, and as appropriate implement,

, better testing methods whica are developed and validated for commercia) use

so as to enable detcction of degradation _ as small as 20% through wall,without exception.

Until wch time as 20% penetration ccn be detected in the~ roll l; expanded areas and the sleeve end, inspection results will be compared to those

, obtained during the baseline slee';ed tube-inspection.

Wnenever the results of any steam generator tubing inservice inspection fall

, into Category C-3, these results will be promptly reported to the Commission pursuart to Specification 6.9.1 prior to-resumption of plant operation.

Such lI cases will be considered by tne Ce, mission on a case-by-case basis and may

,, rrstit in a requirement for analjas, laboratory examinations, tests,

,i aoditional eddy-current inspection, and revision of the Technical Specifica-

' tions, if necessary.

l; The steam generator water level limits are consistent with the initial

!assumptionsintheFSAR.

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REACTOR COOLANT SYSTEM i

BASES i

l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 1

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems reouired by this specification are provided to detect and manit.or leakage trom ' e Reactor Coolent Pressure Boundary. These detection s/ stems are consisteilt with the recommendations of Regulatcry Guide 1.45, "he?.ctor C'olant Pressure Boundary Leakage o

Detection Systems,r May.1973 3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

4 1

Therefore, the presence of-any PRESSURE BOUNDARY _ LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

1

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Industry experience has shown that, while a limited amount of leakage is expected from the RCS, the UNIDENTIFIED LEAKAGE portion of this can be reduced to a threshold value of less than 1 GPM. 'This threshold value is i

sufficiently low to ensure early detection of additional leakage, i

The total steam generator tuce leakage limit of 1 GPM for-all steam generators ensures that the dosage-contribution from-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. : The<1 GPM limit is consistent with the assumptions used in the analysis of-these accidents.

i The 10 GPM IDENTIFIED-LEAKAGE limitation provides allowance for a i

limited amount of leakate from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE ~ by the leakage j

detection systems.

i The CONTROLLED LEAKAGE limit of 10 GPMl restricts operation with a total RCS leakage from all RC pump seals in excess of 10 GPM.

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