ML20203L111

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Proposed Tech Specs Pages Re Amend to License NPF-3 Involving Incorporation of New Repair Roll Process for SG Tubes W/Defects in Upper Tube Sheet
ML20203L111
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 02/26/1998
From:
CENTERIOR ENERGY
To:
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ML19317C931 List:
References
NUDOCS 9803050358
Download: ML20203L111 (18)
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_ __

IJtR97 0010 Page 17 m C10R C0 m N1 svs"

STEAll GENERATORS INFORMATION ONLY 4

LIMITING CONDITION FOR OPERATION .__

3.4.5 Each steam Generator shall be OPERABLE with a minimum water level of 18 l inches and the maximum specified below as applicable: j H0 DES I and 2:

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

MODE 3*:

b. 50 inches Startup Range with the SFRCS Low Pressure Trit bypassed and one or both Main feedwater Pump (s) capable c' supplying l Feedwater to any Steam Generator.
c. 96 percent Operate Range with:
1. The SFRCS Low Pressure Trip active.

Or

2. The SFRCd low Pressure Trip bypassed and both Main Feedwater Pumps incapable of supplying Feedwater to the Steam Generators.

l l

H0DE 4:

d. 625 inches Full Range level APPLICABILITY: H0 DES 1, 2, 3, and 4, as above.

ACTION:

a. With one or more steam generators inoperable due to steam generator tube imperfections, restore the inoperable generator (s) to OPERABLE status prior to increasing 1,y above 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 Fours 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 />.

DAVIS-BESSE, UNIT 1 3/4 4-6 Amendment No. 7Y,Y7Y,192 9003050358 900226 PDR ADOCK 05000346 P PDR

0 l0"; " INFORMATION ONLY Floure 3.4 Maximum Allowable Steam Generator Level in H0 DES I and 2

,- 100 (43,96)

/

t$

T> 90 - . t

.5 E

E 80 -

Unacceptable operating '

3 Region .

E, 70 -

O u

2 /

e U 60 - Acceptable u operating S

Region 3 50 - -

m (0,43) 40 - N l l l l l I O 10 20 30 40 50 60 Main steam Superheat ('F)

DAVIS-BESSE, UNIT 1 3/4 4-6a Amendment No.192

LAR97 0016 Pcge 19 REACTOR COOLANT SYSTEM i

STEAM OENERATORS SURVRif 1 ANCE REOUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Ge=rator S= ale Seldon and inanection - Each st:am generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum

, number of steam generstors specified in Table 4-4.1, 4.4.5.2' Steam Generator Tube S=nle Selection and insocction The steam generator tube minimum sample size, inspection result classification, and the correspcmding action required '

shall be as specified in Table 4.4 2. De inservice inspection of steam generator tubes shall be perfonuS at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified accepi:ble per the acceptance criteria of Specification 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:

a. The first sample inspection during each inservice inspection (d x ;;d t: S:

ir:!!= !=p xf =) of each steam generator shall include:

1. All tubes or tuk eeves that previously had detectable wall penetrations

(> 20%)that L tot been plugged or repaired by renale roll or sleeving in the affected ark ITubes renalmd by sleevinn or mnair roll remain available for random selectioni

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

DAVIS-BESSE- UNIT 1 3/4 4-6b Amendment No.192, 9

~

LAlt07-0010 INFORMATION ONLY SURVEILLANCE EQUIREMENTS (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 dees not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall to 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 dr'illed 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 l subjected to less than a full tube inspection provided:
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 inspectson 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 Here 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-BESSE, UNIT 1 3/44-7 Amendment No. 21.112. 3

LAR07-0016 Page21 JG6CTOR GQQldNT SYSTEM SURVEll1ANCE REQUIREMENTS (Continued)

Notes: (1) In all inspections, previously degraded tubes must exhibit significant (> 10%) 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 Jmp_eslien".tsautIKiep The above required inservice inspections of steam generator tubes shall be performed at the followiug frequencies:

a. h kselinenspection+hallle perfwmed 4wAneide-witMhe-first<,eheduled+efueling+utagMmi-no lator-than-Amil40494 Subsequent-linservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the },revious inspection, if the results of two consecutive inspections for a given group' of tubes followirig 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 contir.ued and no additional degradation has occurred, the inspection interval for that group may be extended 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 after 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 accordance with the first sample inspection specified in Table 4.4 2 during the shutdown subsequent to any of the following conditions:
1. Primary to-seccndary tube leaks (not including leaks originating from tube to tube sheet welds) in excess of the limits of Specifiaation 3.4.6.2.

Ilths.lsakit.dttermincitt %nuurcair rollicint rather.tha.selectine a random sampJs.

inspgs1.10.0% of the retsit ,A ints in the affected steam etaggytor. If the results of this insocctiodlin12Aff.L2ecorvmnerform additional insoections in the unnfreeted.sttam -

RCatfALQL

' A group of tubes means:

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

3/4 6 8 Amendment No. 21, DAVIS-DESSE, UNIT I

LAR97 0016 Paga 22 i EMCTOR COOLANT SYSTEM

$URVE1LLANCE nmuinRMENTS (Cantla=A) i

2. A seismic occurrence greater thaa ti. hwating Basis Easthquake.
3. A loss-of coolant accident requiring act alon of the engineered safeguards.
4. A main steam line or feedwater line break.
d. De provisions of Specification 4.0.2 are not applicable.

4.4.5.4 Asssptance Criteria

a. As used in this Specification:
1. Tubinn or Tube means that portion of the tube or tube sleeve which forms the primary system to secondary system boundary.
2. imaarfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or spec lfications. Eddy-current testing indications below 20% of the nominal tube wall thicknen, if detectable, may be considered as imperfections.
3. DegradatiP.n means a service induced cracking, wastage, wear or general ,.

t corrosion occurring on either inside or outside of a tube,

4. Drnraded Tube means a tube containing imperfections it 20% of the nominal wall thickness caused by degradation that has not been repaired by raamir roll or sleeving in the affected area.

^

5.  % Dearadetion means the percentage of the tube wall thickness affected or _

removed by degradation.

6. Difeti means an imperfection of such severity that it exceeds the repair limit. A defective tube is a tube containing a defect that has not been repaired by gspait gallg. sleeving in the affected area or a sleeved tube that has a defect in the -

sleeve.

7. Renair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by ran=le roll or sleeving in the affected area because it may become unserviceable prior to the next inspection and is '

1 equal to 40% of the nominal tube wall thickness. De Babcock and Wilcox process described in Topical Report BAW 212f P will be used for sleeving.

DAVIS-BESSE, UNIT 1 3/44-9 Amendment No. 21,171, l

i I

L._. - , , _ . _ . , . _ _ _ . . . . . . - . . . . ~ . . , ~ ~ = , , _ , . . . . - . - _ . _ _ . - . . . . _ , . _ , _ . . . _ , . _ . ~ . . . _ . - , _ , . .

LAR97-0016 Page 23 KEACIOR COOLANT SYST9d SUP_VRIf I ANCE REOUIRRMENTS (CimeinimaA) , ,

(fsggjgy/) 7. h ranale mil near*aa will andv he a=~4 to raanle hihan with rial. gjg t*m phesheet area. h renair mil nmcess will he nerfanned only on:e per (pgni gaperator tube usina a 1 inch reroll lenath. h new mil area must he free of degradation in order for the renair to be canaldered acceptable. h repgir my pmcess used is desenhed in the Topical Renart BAW 2303P. Revision 3.

8. Unserviceable desenbes the condition of a tube ifit leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a lossef coolant accident, or a steam line or feedwater line break as spec 8.?4 6.14.4.5.3.c. above.
9. Tube insoection means an inspection of the steam generator tube from the point -

of entry completely to the point of exit. h previously existine tube and tubt roll. above the new mil area in the upner tube sheet can he excluded from future periodic Inanection requirements because it is no lanner part of the orennure houndarv once the repair mil is installed.

. DAVIS-BESSE, UNIT I 3/44-9a Amendment No. 21,171,

LAR97-0016 Pag 3 24 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

10. Preacrvice I==~ tion means an inspection of the full length of each tube in each steam generator periwmed by eddy current techniques prior to service to establish a l baseline condition of the tubing. His inspection shall be performed prior to initial J POWER OPERATION using the equipment and techniques expected to be used  ;

during subsequent inservice inspections.

b. %c steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by ra=le roll or 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 generator tube inservice inspection shall be submitted on an annual basis in a report for the period in which this inspection was completed. His report shallinclude:
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,er41eeved or ra=Ir rallad.
c. Results of steam generator tube inspections which fall into Category C 3 and require notification of the Commission shall be reported prior to sesumption of plant operation.

His report shall provide a description of investigations conducted to determine cause of the tube degradation a.xl corrective measures taken to prevent recurrence.

I 4.4.5.6 ne steam generator shall be demonstrated OPERABLE by verifying 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 perfo ed as per Section 4.4.5.2, an additional but totally ^

separate inspection shall be performed on special intert. . peripheral tubes in the vicinity of the secured internal auxiliary feedwater header his testing shall only be required on the steam generator selected for inspection, and the test shall require inspection only between I

c l

DAVIS BESSE, UNIT 1 3/4410 Amendment No. 8,27,62, 171,184 i

I

'.__;_-. .____._____ ..__-_.__.....,_..____,_x-.,__;_._ _ _ , _ . . . _ , _ _ _ _ _ , _ . . , _ , _ _ , , _ . _ _ _ _ _ . . . , ,~ - .- _

IAR97-0016 Pngo 25 REACTOR COOLANT SYSEM SURVRIIIANCE REOUIREMENTS (Continued) the upper tube sheet and the 15th tube support plate. The tuks select al for inspection shall re 3 resent the entito circumference of the steam generator and shall tc al at least 150 peripheral tu xs.

1 4.4.5.8 Visual inspections of the secured internal auxiliary feedwater header, header to shroud attachment welds, and the external header thermal sleeves shall be performed on each steam generator through the auxiliary feedwater injection penetrations.

These inspections shall be performed during the third and four.h refueling outages and at the ten year ISI. l l

4.4.5.9 Wh'en steam cenentorjube insocction is nerfonned_as oer Section A4.5.7. ari additlanal aut totally seoarate insocct on shall be 1:,crformed on soccial interest tubes tl isit have xenicoaired ave tv the rena r roll orocess. This insnecdgn shall be oerformed on 100% of the tubggjhat H xen reoaired by the reoair roll orocess. The insocction shall be limited to the reoair to1 qpj god the roll trannidons of the reoair roll. Defective or denraded tubes found in the rena r_rgli renton as a result of the insocction need not be included in determ ninghlgggggtion Restats Catenorv for the ceneral steam cencrator insocction.

ADDilM*AL OGS PMV:00$LY Ph0PO$lD tv LOMR l

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Date 97 ' t 23 DAVIS-DESSE, UNIT 1 3/4 4-10a Amendment No.62

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m. TABLE 4.4-2 

                                                                                                                                                                                                            *' G tC 4 STEAM GENERATORTUBEINSPECTION                                                          - "6 y

g . c- - , 1 IST3AMPIEINSPECITON 2ND SAMPIRINSPECITON 3RDSAMPLEIN3FECITOIJ E Action Requir.ai Result Action Required Result l ActionRequma" Sampir Size Result , P A nununum 'v*i None N/A N/A y N/A N/A l l C" of S Tubes C2 Plug or repeirbyM C-1 None N/A N/A 5- per S G. (1) '~ i 

                                    -                                              254eevingdefenvc tubes and                                                            g C2               Plu                                C1                    None impect additimal25 tubes in                          %gxrepairbyM ee E.kfective tubes I

this S.G. and inspecirdditional 45 C-2 Plur or re by M tubes in thir S.G. d M@eringdefective [ 3 mbes P ta 2' Peixm actica fer C-3 rescrt A C-3 l of first sample h . j C-3 Perform ac:ior for C-3 result N/A N/A of frst sample O> O9 g C3 Inspect all tubes in this S.G., All .xher t pluror repair by S.G.s are C-I None N/A N/A g g, gg Esteevmg defective sabes and Sorne S.G.s , y Perform action for C-2 rest I

' @ mspect 25 tubes in each other C-2 but no of second samp'e N/A N/A S.G. Report to the NRC prior additional C~h
o. , to iesumption of plant S.G. are C-3 Eg ,A
8. .- Additional Inspect alltubes in each S.G.

1 g g [-' S.G. is C-3 and ein orrepar by M N/A G- sleeving defective N/A ~ 

                                 #$                                                                                                      tubes. Report to the NRC prior to resumption of plet operanon.                                                                       l
                                                        -                                                                                                                                                                Y (1) S=3H% Where N 'is the number of steam generators in the unit, and n is the number of stea n genattors inspected daing an inspection.                          l t

n i i l

O LAR97 0010 Pago28 Rl! ACTOR COOL, ANT SYSTEM OPERN110NAL LEAKAGE LIMI11NG CONDITION FOR OPERATION 3,4.6.2 Reactor Coolant System leakage sha'l be limited to:

a. No PRESSURE BOUNDARY LEARAGF,
b. 1 GPM UNIDENTIFIED LEAKAGE,
c. 150 GPDI-GPM total-primary to secondarji leakage through.theJubes of any ong steam generators,
d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,
c. 10 GPM CONTROLLED LEAKAGE, and
f. 5 GPM leakage from any Reactor Coolant System Pressure Isolation Valve as specified in Table 3.4 2.

APPLICADILITY: MODES 1,2,3 and 4 AC110N:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least IIOT STANDBY within 6 hours and in COLD SilUTDOWN within the following 30 hours,
b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours or be in at least 110T STANDBY within the next 6 hours and in COLD SilUTDOWN within the following 30 hours except as pennitted by paragraph c below,
c. In the event that integrity of any pressure isolation valve specified in Table 3.4 2 cannot be demonstrated, POWER OPERATION may continue, provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition.(a)
d. The provisions of Section 3.0.4 are not applicable fc: mtry into MODES 3 and 4 for the purpose of testing the isolation valves in Table 3.4-2.

W Motor operated valves shall be placed in the closed position and power supplies deenergized. { DAVIS-BESSE, UNIT 1 3/4415 Order dtd. 4/20/81 Amendment No. 135,180 3.-

O 4 LAR974016 Page 29 l REACTOR COOI).NT SYSTEM 1 SURVRif LANCE REOUIREMENTS ! 4.4.6.2.1 Reactor Coolant System leakage s shall be demonstrated to be witMn each of the above

limits by:

I a. - Monitoring the containmeat atmosphere gaseous or particulate radioactivity at j- least once per 12 hours. o ! b. - - Monitoring the containment sump level and flow indication at least once per 12 I bM. . c. Measurement of the CONTROLLED LEAKAGE from the reactor coolari pump seals to the makeup sys'em when the Reactor Coolant System pressure is 2185 i 20 psig at least once pet 31 days.

d. Perfonnance of a Reat tor Coolant System water inventory balance at least once i_ per 72 hours during stcady state operation.
e. An evaluattar of necnadary watar radinchemletev for determination of nrim arv to amtsn triary .ma emne t),ronah the meamen omnarminra 31 lamat nnce ner 72 hours t urin;
                            =e==<

l v ='ae- oneratird L 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve spt cified in Table 3.4 2 shall be 

        ' individually demonstrated OPER ABLE by verifying leakage testing (or the equivalent) to be within its hmit prior to entering MODE 2:
a. A'ter each refueling outage, ,.
b. Whenever the plant has been in COLD SHUTDOWN for 7 days, or more, and if -

leakage testing las not been performed in the previous 9 months, and

c. Prior to retuming the valve to service following maintenance, repair or replacement work on the valve. -
d. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.

4.4.6.2.3 wenever the integrity of a pressure isolation valve listed in Table 3.4 2 cannot be demonstrated, d dennir.e and record the integrity of the high pressure flowpath on a daily basis. , Integrity shrdi be determined by performing either a leakage test of the remaining pressure isolation valve, or a combined leakage test of the temnining pressure isolation valve in a series with the closed motor operated containment isolation valve. In addition, record the position of tl.e closed motor-operated containment isolation valve located in the high pressure piping on a daily basis. Order dated 4/20/81 DAVIS BEFSE, UNIT 1 3/4416 Amendment No. 54,135, 180,1 % ,

O e l:g;;o= = u > *- 2 INFORMATION ONLY REACTOR COOLANT SYSTEM PRESSURE ISOLA, TION VALVES c ,, 4 VALVE NUMBERS (b)_  !%XIMtf4 ALLOWABLE L'EAKAGE (a)(c) 

           . SYSTEM CF-30                           1       5 0 gpm
1. Decay Heat Removal 1 5 0 gpm Decay Heat Removat CF-31 2.

1 S 0 gpa

3. Decay Heat Removal @ 76 1 5 0 gpm 4 Decay Hea; Aemoval OH-77 Notes:

(a) 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpa are ansidered acceptable if the latest measured rate has not exceedew the rate determined by the previous test by an amou1t that reduces the surgin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
3. Leakage rates greater than 1,0 gpm out less than or equal to 5.0 gpu art considered unacceple if the latest measured rate e.veeded the rate determined by the previour test by an amount g

that reduces the margin between measured leakage rate and the caximum permissible rate of S.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gem are considered unaccept'able.

(b) Valves CF-30 and CF-31 will be tested with the Reactor Coolant system pressure >1200 psig. Yalves 576 and DH 77 will be Mini- tcsted with normal Core Floeding Tank pressure which is >b75 psig. mue differential test pressure across each valve shall not be less  ! ' than 130 psid. To satisfy ALARA requirements, leakage may be measured indirectly (c) (as from the performance of pressurs indicators) if acc:mplished in accordance with apprzved procedurus and supported by coc@utations showing that the method is capable of demonstrating valve compliance 

                           .ith the lukage criteria.

3/4 a-16a Order dtd. 4/20/81 DAVIS-BESSE, UNIT 1

LAR97-0016 Page 31 REACTOR COOLANT SYSTEM 

    . BASES 3/4.4.4 PRESSURI7RR l A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is
     -capable of accommodating 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 'Ihe high level limit is based on providing enough steam volume to prevent a pressurizer high twel as a result of any transient. 

     . The pilot operated relief valve and steam bubble function to relieve RCS pressure during all design transients. Operation of the pilot operated relief valve mhlmizes the undesirable opening of the spring-loaded pressurizer code safety valves.

3/4.4.5 STEAM OENERATORS 

      ' The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this poition 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 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 due to design, manufactr:ing errors, or inservice conditions
       . that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. A process equivalent to the inspection method described in Topical Report BAW-2120P
      - will be used for inservice inspection of steam generator tube sleeves. This inspection will .

pm.'ide ensurance of RCS integrity. Tiie plant is expected to be operated in a manner such that the seconda:y 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 chemissy.

1 limits, le Jized corrosion may likely result in stress corrosion cracking. -The extent of cracking during plaut 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 = 150 GPD thmnoh any ane s'eMD.Esaggg4-GPM). Cracks having a primary-to-secomlary leakage less than this limit during operation will have an ad:quate margin of safety to withstand the loads imposed during normal DAVIS-BESSE, UNIT 1 B 3/4 4-2 Amendment No. 135,171, l 

 ,0 d                                                                                                         1 LAlm.
  • 16 i Pago32 REACTOR COOLANT SYSTEM BASES (Continued) operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 150 GPD 4GPM 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 locate i and plugged or repaired by reoair rolline or sleeving in the affected areas. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolent. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. 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 40% through wah. Tubes with imperfections exceeding the repair limit of 40% of the nominal wall thickness will be plugged or repaired by reoair rolline or sleeving the affected areas. Davis-Besse will evaluate, and as approprice implement, better testing methods which are developed and validat ,d for commercial use so as to enable detection of degradation as small as 20% through w~all 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 cPained during the baseline sleeved tube inspection. An additional reoair method for decraded steam cenerator tubes consists of rerollino the tubeliD the unoer tubesheet to create a new roll area _and oressure boundarv for the tube. The reoair roll orocess will ensure that the area of decradnion will not serve as a cressure boundarv. thus oermittine the +2he to remain in service. The decraded area of the tube can be excluded from future oeriodic insocction reauirements because it is no loncer oart of the cressure bcundary once gg,Igpeir roll is installed in the unoer tubesheet. All tubes which have been reoaired usine the renair roll orocess will have the new roll area jginggfed durine the inservice insucction. Defective ordecraded tube indications found in the new roll area as a rgsult of the insocction of the recair roll and any indications found in the originally rolled recion of the rerolled tube need not be included in determinine the Insoection Results Catecorv for the ceneral steam cenerator insocction. Thqlgonir ro!1 orocess will be oerformed only once oer steam cenerator tube usine a 1 inch reroll lencth as described in the Tonical Renort BAW-2303P. Revision 3. Thus. multiole acolications of the rerolline orocess to any individual tube is not accentable. The new roll areunust be free of decradation in order for the reoair to be considered accentable. After the new roll a ea is initially deemed accentable. future decradation in the new roll area will be analyzed to determine jf the tube is defective and needs to be removed from service. The rerolline orocess is described ig the Toolcal Reoort BAW-2303P. Revision 3. DAVIS-BESSE, UNIT I B 3/4 4-3 Amendment No. 171,184,192, 

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LAR97-0016 Page 33 REACTOR COOLANT SYSTEM a BASES (Continued) Whenever the results of any steam generator tubing inscrylce inspection fall into Category C-3, these results shall be reported to the Commission prior to resumption of plant operation. - Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory exanunations, tests, additional eddy-cunent inspection, and a revision of tl'c Technical Specifications, if ne,:essary. 4 The steam generator waar level limits are consistent with the initial assumptions in the USAR.

While in MODE 3, examples of Main Feedwater Pumps that are incapable of supplying feedwater to the Steam Generators are tripped pumps or a manual valve closed in the discharge j flowpath. The reactivity requirements to ensure adequate SIIUTDOWN MARGIN are provided in plant operating procedures.

1 i i i 4 ( DAVIS-BESSE, UNIT I B 3/4 4-33 Amendment No. 171,184,192, J 

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LAR974)016 - Page 34 REACTOR CQQLANT SYSTEM BASES m 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to , detect and monitor leakage from the Reactor Coolant Pressure Boundary, nese detection systems are consistent with the recommendation of Regulatory Guide 1.45, 

         " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973, 3/4.4.6.2 OPERATIONAL LEAKAGE i

l PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary herefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the Unit to be promptly placed in COLD SHUTDOWN. Industry experience has shown that, while a limited amount ofleakage is expected from the RCS, the UNIDENTIFIED LEAKAGE portion of this can be reduced to a threshold value ofless that 1 GPM. His threshold value is sufTiciently low to ensure early detection of additional leakage. The total-steam generator tube icakage limit of 150 GPD_thropoh any one 4-GPM fer an steam generators ensures that the dosage contribution from tube leakage will be limited to a small fraction of 10 CFR Part 100 lindts in the event of either a steam generator tube rupture or steam line break. ,A-ne-1 GPM total nrimary to secondary leakaoe limit is :: it:: cid i: x =nption: used in the analysis of these accidents. The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount ofleakage from known sources whose presence will not interfere with tiie detection of UNIDENTIFIED LEAKAGE by the leakage detection systems, ne CONTROLLED LEAKAGE limit of 10 GPM restricts operation with a total RCS leakage from all RC pump seals in excess of 10 GPM. The sarveillance requirements for RCS Pressure Isolation Valves provice added assurance of valve integrity th:reby reducing the probability of gross valve failure and consequent intersystem LOCA Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. DAVIS-BESSE, UNIT 1 B 3/4 44 Amendment 180 _j}}

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