ML20072J614

Proposed Tech Specs,Revising TS 3.1.1.1,3.4.5 & Associated Bases

ML20072J614
Person / Time
Site:	Davis Besse
Issue date:	08/22/1994
From:	CENTERIOR ENERGY
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ML20072J536	List: ML20072J534 ML20072J587 ML20072J614
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NUDOCS 9408290109
Download: ML20072J614 (15)

Text

Docket Number 50-346 j License Number NPF-3 l Serial IA.unber 2231 l Attachment ~

Page 12 ADDlil0NAL CHANGf S PREV 100 Sty REACTIVITY CONTROL SYSTEMS PROPOSED BY LETTER 3/4.1 Senal No. o7)/O Date OMM/f't 3/4.1.1 BORATION CONTROL SHUTOOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 1% ak/k.

APPLICABILITY: MODES 1,2*,374and5.

ACTION:

With the SHUTDOWN MARGIN < 1% ak/k, immediately initiate and continue boration at > 18 gpm of 7875 ppm boron or its equivalent, until the required SHUT 00WN MARGIN is restored.

SURVEILLANCE REQUIREMENTS s 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 1% ak/k:

a. Within one hour af ter detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> therbafter while the rod (s) is inoperable. If the inoperable control rod is imovable or untrippable, the above require 4 SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).

b. When in MODES 1 or# 2 , at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, by verifying that regulating rod groups withdrawal is within the limits of Specification 3.1.3.6.

c. When in MODE 2## within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criti-cality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

d. Prior to initial operation above 5% RATED THERMAL POWER af ter each fuel loading by consideration of the factors of e. below, with the regulating rod groups at the maximum insertion limit of Specification 3.1.3.6.

With Kef f > 1.0.

With Keff < l.0.

'n .1 4 e LCO 3Maes Generatog for add;tioul SHL4TDOLJN 11 ARGTtJ re u;re ments.

DA , UNIT 1 1-1 ge29oio,94ogy, p ADOCK 05000346 PDR

Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 13 i REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

e. When in MODES 3, 4 or 5, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position,

3. Reactor coolant system average temperature, '

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1% ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e. above.

The predicted reactivity va' dues shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full power Days af ter each fuel loading. i P

DAVIS-BESSE, UNIT 1 3/4 1-2  ;

.. IN.FO.h.n4ATIC3N ONLY

Docket ihmder 50-346 f i

License Nunder NPF-3 l Serial Nundwr 2231 Attachment f l

P89e H 3/4.1 REACTIVITY CONTROL SYSTEMS BASES m m ,.. mm _

3/4.1.1 BORAT10N CONTROL mrumum IUI4 URLY '

l 314.1.1.I SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2 the reactivity transients associated with postulated accident condition)s are controllable within acceptable and 3) the reactor will be maintained sufficiently subcritical tolimits, prec lude inadvertent criticality in the shutdown condition.

During Modes 1 and 2 the SHUTDOWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insertion limit.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration and RCS T . The most restrictive condition occurs at E0L, with T at no load operWing temperature. The SHUTDOWN MARGIN required is con Ustent with FSAR safety analysis assumptions.

3/4.1.1.2 BORON DILUTION A minimum flow rate of at least 2800 gpm provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual through the Reactor Coolant System in the core during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 2800 gpm will circulate an equivalent Reactor Coolant System volume of 12,110 cubic feet in approximately 30 minutes. The reactivity change rate associated with boron concentration reduction will be within the capability for operator recognition and control.

In MODE 5 or MODE 6, the RCS boron concentration is typically somewhat higher than the boron concentration required by Specification 3.1.1.1 (MODE 5) or Specification of normal sources3.9.1 of (MODE water 6)ddition.,

a and could beinventory At reduced higher than the boroninconcentration conditions the RC$, in order to reduce the possibility of vortexing, the flowrate through the decay heat system may be procedurally restricted to somewhat less than 2800 gpm. In this situation, if water with a boron concentration equal to or greater than the baron concentration associated with the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 (MODE 5) or the boron concentration correspondingtothemorerestrictivereactivltyconditionspecifiedin

, is added to the RCS, the RCS boron concentration Specification is assured to 3.9.1 remain(MODE above6)he t minimum baron concentration associated with the Specification 3.1.1.1 or Specification 3.9.1 requirement, and a flowrate of less than 2800 gpm is not of concern.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC are provided to ensure that the assumptions used in the accident and transien)t analyses remain valid through each fuel cycle. The surveillance requirement for measurement of the MTC each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurance that the coefficient will be maintained within acceptable values throughout each fuel cycle. '

DAVIS-BESSE, UNIT I B 3/4 1-1 Amendment No. U 6,188 INFORMATION ONLY

4 Docket number 50-346 3

License Number NPF-3 Serial Numbar 2231 Attachment Page 15 RE*: TOR COOLANT SYSTEM  !

$ TEA.'4 GENERATORS LI'CTING CONDITION FOR OPERATION

! 3.-C . 5 Each steam-generater shall be OMRACLC with a watee-level be,tFcen

.%-and-348-incheer . INS E RT /VEW TS. 3. 4. f

): LIMITING coNOITIotJ FOR

' APPL:CA0ll:TY- MODES 1, 2, 2 and-4. OPERATION, MS ATTA c HE.D

ACTION:

With one or more steam generators inoperable due to steam

a. -

generator tube imperfections, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg ab ve 200*F.

b. With one or more steam generators inoperable due to the water level being outside the limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

'SU:.VEILLANCE REQUIREMENTS 24 .4.5.0 Each steam generator shall be demonstrated OPERABLE by performance d of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam

generator shall De determined OPERAdLE' ourina shutdown by selecting and ir.soecting at least the minimum number of steam generators specified in Ta

le 4-4.1.

4. .5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and

tne corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the

frequencies specified in Specification 4.4.5.3 and the inspected tubes shall

be verified acceptable per the acceptance criteria of Specification 4.4.5.4.

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

.I of the total number of tubes in all steam generators; the tubes selected for

' tnese inspections shall be selected on a random basis except; I

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

1. All tubes or tube sleeves that previously had detectable ,

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

2. At least 50E of the tubes inspected shall be in those areas where experience has indicated potential problems.

Amendment No. 2J ,171

?.:5-BESSE - UNIT I 3/44-6

Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 16 New TS 3.4.5 Limiting condition for Operation 3.4.5 Each Steam Generator shall be OPERABLE vith a minimum vater level of 18 inches and the maximum specified belov as applicable:

MODES 1 and 2: ,

a. The acceptable operating region of Figure 3.4-5.

MODE 3: *

b. 30 inches Startup Range with the SFRCS Lov Pressure Trip bypassed and one or both Hain Feedvater Pump (s) capable of supplying Feedvater to any Steam Generator.

c. 96 percent Operate Range with:

1. The SFRCS Lov Pressure Trip active.

Or

2. The SFRCS Low Pressure Trip bypassed and both Main Feedvater Pumps incapable of supplying Feedvater to the Steam Generators.

MODE 4:

d. 625 inches Full Range Level APPLICABILITY: MODES 1,2,3, and 4, as above.

AEstablish adequate SliUTDOWN MA11 GIN to ensure the reactor will stay subcritical during a MODE 3 Ma in 'st eam Line 14 re a k .

~

I Docket Numbar 50-346 License Number NPF-3 I

Serial Number 2231 Attachment Page 17 Figure 3.4-5 Maximum Allowable Steam Generator Level in H0 DES 1 and 2 100 -

(43,96) 2 v

\

/

T 90 -

E a

E

$ 80 - Unacceptable Operating 3 Region E

8, 70 -

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$ 60 -

5 Acceptable U

Operating

@ Region 3 50 -

v2 (0,43) 40 - s 1 I I I i 1, 0 10 20 30 40 50 60 Main Steam Superheat (oF) l NEW figu.re. 3.4-5 To es WSEMED As New Pacie 3/4 4-(o m

- ~

Docket Number 50-346 License Number NPF-3 )

Serial Number 2231 Attachment e- 18 REACTOR COOLANT SYSTEM INFORMATION ONI.Y i l

SURVEILLANCE REQUIREMENTS (Continued)

3. A tube inspection (pursuant to Specification 4.4.5.4.a.9 ) l shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

b. Tubes in the following groups may be excluded from the first random sample if all tubes in a group in both steam generators are inspected. No credit will be taken for these tubes in meeting minimum sample size requirements.

1. Group A-1: Tubes within one, two or three rows of the open inspection lane.

2. Group A-2: Tubes having a drilled opening in the 15th support plate.

3. Group A-3: Tubes included in the rectangle bounded by rows 62 and 90 and by tubes 58 and 76, excluding tubes included in Group A-1.*

c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to less than a full tube inspection provided:

-e t

1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

2. The inspections include those portions of the tubes where imperfections were previously found. ,

The results of each sample inspection shall be classified into one of the following three categories:

Category Inspection Results C-1 Less than 5% of the total tubes .

inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

• Tubes in Group A-3 shall not be excluded after completion of the fifth refueling outage.

DAVIS-BISSE, UNIT 1 INFORMATIO M 0 m-e-

• m.m

• Docket Number 50-346 License IAunber NPF-3 i Serial Number 2231

- ch-Page 19 INFO.RMATION ONLY REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

Notes: (1) In all inspections, 'reviously degraded tubes must exhibit significant s> 107.) further wall penetrations to be included-in the above percentage calculations.

(2) Where special inspections are performed pursuant to 4.4.5.2.b, defective or degraded tubes found as a result of the inspection shall be included in determining the Inspection Results Category for that special inspection but need not be included in determining the Inspection Results Category for the general steam generator inspection.

4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. The baseline inspection shall be performed to coincide with the first scheduled refueling outage but no later.than April 30, 1980.

Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If the results of two consecutive inspec-tions for a given group

• of tubes following service under all volatile treatment (AVT) conditions fall into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occu-red, the inspection interval for that group may be extendeo to a maximum of 40 months.

b. If the results of the inservice inspection of a steam generator performed in accordance with Table 4.4-2 at 40 month intervals for a given group

• of tubes fall in Category C-3, subsequent inservice inspections shall be performed at intervals of not less than 10 nor more than 20 calendar months af ter the previous inspection. The increase in inspection frequency shall apply until a subsequent inspection meets the conditions specified in 4.4.5.3a and the interval can be extended to 40 months.

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordaace with the first sample inspection specified ir .ra ble 4.4-2 during the shutdown subsequent to any of the fol -.ap conditions:

1. Primary-to-secondary tut". m ks (not in 12 ding leaks originating from tube-to tube sheet weid.) in excess of the limits of Specification 3.4.6.2.

• A group of tubes means:

(a) All tubes inspected pursuant to 4.4.5.2.b or (b) All tubes in a steam generator less those inspected pursuant to 4.4.5.2.b.

INFORYATION C NLY DAVIS-BESSE. UNIT 1 3/ -s Amendment No. 2:

Docket Number 50-346 License Number NPF-3 Serial Number 2231

T"' INFO RMATION ONLY 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 applicable.

4.4.5.4 Acceptance Criteria

a. As used in this Specification:

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

2. Imperfection means an exception to the dimensions, finish or l contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

3. Degradation means c service-induced cracking, wastage, wear I or general corrosion occurring on either inside or outside of a tube.

4. Degraded Tube treans a tube containing imperfections > 20% of the nominal wall thickness caused by degradation that has not been repaired by sleeving in the affected area.

5.  % Degrada *.fon means the percentage of the tube wall thickness affected or removed by degradation.

6. Defect means an . imperfection of such severity that it exceecs

~

the repair limiti A defective tube is a tube containing a defect that has not been repaired by sleeving in 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 the tube shall be removed from service by plugging or repaired by sleeving in the affected area because it may beco.te unservice-able prior to the next inspection and is equal to 40% of the nominal tube wall thickness. The Babcock and Wilcox process described in Topical Report BAW-2120P will be used for sleeving.

8. Unserviceable dascribes the condition of a tube if it leaks or l contains a defect large enough to af.fect its structural integ-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 inspection of the steam generator j tube from the point of entry completely to the point of e ri-1 l

DAVIS-BESSE, UNIT 1 Ement No. 21.1 1

}

I

Docket tuber 50-346 l 1.icense IM ber NPF-3 )

Serial Number 2231 Attachment

' REACTOR COOLANT SYSTEM INFORMATION ONLY SURVEILLANCE REQLHREMENTS (Continued)

. 10. preservice Inspection means an inspection of the full length of each tube in each steam generator performed {

by eddy current techniques prior to s' ervice to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections. i

^

b. 1 The steam generator shall be determined 0PERABLE after completing '

the corresponding actions (plug or repair by sleeving in the affected areas all tubes exceeding the repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports

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 ge'n erator 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.

3. Identification of tubes plugged or sleeved.

c. Results of steam generator tube inspections which fall into Category i C-3 and require notification of the Commission shall be reported prior to resumption' of plant operation. 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 i steam generator 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 />.

4.4.5.7 When steam generator tube inspection is performed as per Section 4.4.5.2 an additional but totally separate inspection  :

shall be performed on special interest peripheral tubes in the vicinity of the secured internal auxiliary feedwater' header. This testing shall only be required on the steam generator selected for inspectian, and the test shall require inspection only between DAVIS-BESSE. UNIT 1 3/4 4-10 Amendment No. 8,27,62,172. *  ;

i INFORMATION ONLY

Docket thimber 50-346 License tAlmber t@F-3 s- u1 * * - 2231 INFORMATION ONI.Y Attachment '

REACTOR COOLANT SYSTEM "SURVEILEAMCE REQUIREMENTS'(Continued) the upper tube sheet and the 15th tube support plate. The tubes-selected for inspection shall represent the entire circumference of the steam generator and shall total at least 150 peripheral tubes.

4.4.5.8 Visual inspections of the secured internal auxiliary feedwater TStR header, header to shroud attachment welds, and the external header M M)oca thermal sleeves shall be perfonned on each steam generator through the auxiliary feedwater injection penetrations.

These inspections shall be perfonned during the third and fourth refueling outages and at the ten-year ISI.

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U n n I IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION to ' (n E Uk Sample Size Reesdt Action Required Result Action Required Hesult Action Required A mlnitivn of C-1 None N/A N/A N/A N/A S Tubes per S.G. (t) arawn. C-2 Plug or repair by C-1 None N/A N/A

== sleeving defective 86=g tLees and inspect C-2 Plug or repanr by C-1 l None Witional 2S tubes in this S U. sleevWective C .

ftbes and inspect additional 4S tubes C-2 Plug or repair by sleeving defective N

in this S.G. , tubes

@4

• == w C-3 l Perform action for C-3 result of first Am 2: samp'e E Paform action for

" """ C-3 N/A N/A r"a C-3 result of first sanWe

==ms C-3 inspect att tubes in All ottw aase M this S.G., plug or S.G s are None l

N/A N/A *mJll n repair by sleevirg C-1 b Q Q

, *2i oefective tubes and Inspect 2S tubes in "'Tl 2 @S each otner S.G. Some S G.s Perform action for O y CM Repon to the NRC C-2 but no C-2 result of second gS prior to resumption it I sa m e N/A N/A  %

c :. d plant operatm C-3 b Additiona all tidus in O" '

S.G. is C. mi pite or d

,g b C-3 rrpair ty simvirs uD defective tidus. N/A N/A ui _ Hquxt to tic 150 Q M prior to rmagtim Z

$ of plant or* rat _im.

R N

Q (t> S - 3 % Where N is the number of steam generators in the untt, and n is the number of steam generators inspected during an inspection.

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Docket Number 50-346 - - -

l License Number NPF-3 - - ~ . 1 Serial Ntunber 2231

"t INFORMATION ONLY i  ;

REACTOR COOLANT SYSTEM ' L ,_ J

~ ' ' ' ~ ~ ~ ~ - - - - - -

BASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accorrvnodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and pilot operated relief valve against water relief.

The low level limit is based on providing enough water volume to prevent a reactor coolant system low pressure condition that would actuate the Reactor Protection System or the Safety Feature Actuation System. The high level limit is based on providing enough steam volume to prevent a pressurizer high level as ,a result of any transient.

The pilot operated relief valve and steam bubble f, unction to relieve RCS pressure during all design transients. Operation of the pilot operated relief valve minimizes the undesirable opening of the spring-loaded pressurizer code ,

safety valves.

3/4.4.5 STEA!4 GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure tnat the structural integrity of this portion of the RCS will be maintained.  :

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

modification of Regulatory Guide 1.83. Revision 1. Inservice inspection of steam generator tubing is esser.tial in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical da age or progressive degradation due to design, manufacturing errors, or inservice -

conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a ceans of characterizing the nature and cause of any tuce degradation so that corrective measures can be taken. A process equivalent to tne inspection method described in Topical Report BAW-2120P will be used for -

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

The plant is expected to be operated in a manner such that the secondary 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 corrosion may likely result in stress corrosion cracking. The extent of 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 i primary-to-secondary leakage less than this limit during operation will have en adequate margin]f-safety-toyithstand the loads imposed during normal INFORMATION ONLY 2.'iS-BESSE, UNIT 1 i : : ~: - f Amencment No. 725,1M

Doc!tet IAunber 50-346 License Number NPF-3 Serial Number 2231 4 Attachment Page 26 REACTOR COOLANT SYSTEM BASES (Continued) operation and by postulated accidents. Operating plants have demonstrated 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.

Wastage-type defects are unlikely with proper chemistry treatment of the 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 small as '

20% through wall can be detected in all areas of a tube sleeve except for 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 40% of the nominal well thickness will be plugged or repaired by sleeving the affected areas. Davis-Besse will evaluate, and as appropriate implement, better testing methods which are developed and validated for comercial use so as to enable detection of degradation as small. as 20% through wall without exception. Until such time as 20% penetration can be detected in the roll expanded areas and the sleeve end. inspection results will be compared to those obtained during the baseline sleeved tube inspection.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results shallSuch be reported to the Comission cases will be considered by prior to resumption of plant operation.the Comission on a case-by-cas analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

usAR u airt F <d ater l eeoluu er to the .Stea rn amps that a,re incapable of .suff yings ora. rnanual valve clased in the d4chvee-Generator.s are trippsd pu SHUTDOWN flow atk. The reactiviey re usrements to ensare adepate (1AR IN are provided in pl nt operatingroccolares I

t B 3/4 4-3 Amendment No. 171,184 DAVIS-BESSE. UNIT 1

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