Text

Proposed Tech Specs 3/4.4 Re Changing TS, RCS & Associated Bases to Address Installation of Electrosleeves in Callaway Plant SGs
	ML20107F079
Person / Time
Site:	Callaway
Issue date:	04/12/1996
From:	; UNION ELECTRIC CO.
To:	;
Shared Package
ML20107F044	List: ML20107F042; ML20107F079;
References
	NUDOCS 9604220213
	Download: ML20107F079 (36)
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1 ULNRC-3358 ATTACHMENT 1 TECHNICAL SPECIFICATION CHANGES (MARKED-UP)

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9604220213 960412 PDR ADOCK 05000403 P PDR

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l4 L A/Al- 3 3 S~8

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}' REACTOR COOLANT SYSTEM .

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3/4.4.5 STEAM GENERATORS l 1

LIMITING CONDITION FOR OPERATION ,

3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3 and 4. '

I ACTION:

With one or more steam generators inoperable, restore the above inoperable 200*F.steam generator (s) to OPERABLE status priorcto increasing T,yg j SURVEILLANCE REQUIREMENTS 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. I l

Tj 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator

,,,gF shall be cetermined OPERABLE ouring shutoown by selecting and inspecting at ;

least the minimum number of steam generators specified in Table 4.4-1. 1 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tune minimum sample size, inspection result classification, The and the corresponding action required shall be as cspe' ified in Table 4,4-2i inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the the inspecte tubes selected for each inservice inspection shall include at least 3% of total number of tubes in all steam generators; the tubes selected for thele inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemist fythe indicates critical areas to be inspected, then at least 50%

tubes inspected shall be from these critical areas; I

inspection

b. The first sample of tubes selected for each inserv steam generator (subsequent to the preservice inspectic shall include

& ag% A earhw 0f 4gsa.a ., u ,A i 4A s . 2. C , pre via s de &ct.c or impa,6cnzmr in l ;. & drea refairtd flce VYG d.v C m f datC;dert'd ~ ~

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33" REVislON 1 f REACTOR CDOLANT' SYSTEM l

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SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria

a. As used in this spe::ification:

1) Imperfection means an exception to the dimensions, finish or 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; ,

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2) Decradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube;

3) Deoraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; tube wall thickness

4) % Dectadation means the er enta degrada oron; Nee Vd-te i or eqw .

5). n imperfectio of such severity that it exceeds Defectmeansflimit.

the plugging A tube ontaining a defect is defective;

h pI oe'rtl@sJr* I i oe Aefsw t or eyo ich OJ 6) Pluccing imit means the imperfection f'

the tube shall be removed from service and is equal t f the nogall thicknessig pg l

7) Unserviceable describes the M nD f a tube if it leaks or l contains a oefect large enough to affect its structural integ- j rity in the event of an Operating Basis Earthquake, a loss-of- ;

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coolant accident, or a steam line or feedwater line break as specified in Specification 4.4.5.3c., above; 8)

Tube Insoection means an inspection of the steam generator tube f rom the point of entry (hot leg side) completely around the U-bend to the top support of the cold legg, sad pk , 4g,,

l refareJ Ly slemy, A % hqu:r>w .rLat l

inlae A daa M fochn .A ofA +4 % ; ad s A

'V Y f"gty I lx/ & Eleeb-o r/eva is

. W 6 20% o/ dnMOW r4n'e %LWen +4ckaerr. A I ~i CALLAWAY - UNIT 1 3/4 4-14 ~l. . .

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- Judet- 33 C8 i REVISION 1 REACTOR COOLANT SYSTEM

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SURVEILLANCE REQUIREMENTS (Continued)

Preservice Inspection means an inspection of the full length of

9) each tube in each steam generator performed by addy current techniques prior to service to establish a baseline condition of t tubing. This inspection shall be performed prior to i RATION using the equipment and techniques tial inspe expected to be use ring subse b.

- rasar 1 The steam generator shall be determined OPERABLE after completing r-the corresponding actions (plu all tubes exceeding the plugging i

limit and all tubes containing hrough-wall c. racks) required by le 4.4-2.

4.4.5.5 Reports inservice inspection

a. Within 15 days following the completion of ea lugge in each steam of steam generator tubes, the number of tubes in a 5 cial Report I generator shall be reported to the Commissio of N pursuant to Specification 6.9.2;

b. The complete resul'ts of the steam generator tube inservice inspection "g shall be submitted to the Commission in a Special Report pursuant to 7

/ Specification 6.9.2 within 12 months following the completion of the 11 inspection. This 5 ecial ort

{ det0 ll46 VCS

1) NumberandextentoftubesMspected, each

2) Location and percent of wall-thickness penetration for indication of an imperfection, and .

Identification of tubes pluggedy oy- r-e. Fad 3) h s which fall into Category

c. Results of steam generator tube i C-3, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption plant operation.

gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

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3/4 4-15 i CALLAWAY - UNIT 1 l

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ULNRC-3358 Attachment 1 l

l INSERT 1 (4.4.5.4.a.9) 1

10) Tube Repair refers to a process that reestablishes tube serviceability. Acceptable tube repairs will be performed by the following processes: i l a) Electrosleeving as described in Framatome ;

Technical Report BAW-10219P, Revision 1, I

"Electrosleeving Qualification for PWR l Recirculating Steam Generator Tube Repair." !

l March 1996 )

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TABLE 4.4-2 S

r- STEAM GENER ATOR TUBE INSPECTION i E I 6

-< 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION 1ST SAMPLE INSPECTION Result Action Required Result Action Required Result Action Required E Sample Sire N. A. N. A.

Z C-1 None N. A. N. A.

~ A minimum of -

S Tubes per S. G.

C-2 Plu e tive bes and nspect addin' A W efeciivgbes VI

_ ~ _N. A.

C-h y %

N. A.

2S ubes in this S. G. C-7 a nspect additional C-2 ' Pluglehtive tubes tubes in this S. G. Pf6tm e tion for dr GfO -

g NN p -C-3 r it of first g-sample is ,

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3 Perform n for ?

N./. N. A. g C-3 C-3 result o ' st 8 sample /

m - .

) C-3 Inspe all tube {in . All other N. A.

N. A. .

> S. G.s are None this S. .. plugWie- .

.'_. fective tu C-1 ,

1 i ,

" N. A.

inspect 2S tu ire g3' .g 5 Perform act. ion for N. A. ,

{-

each cther S. G. C-2 but no C-2 result of second addition l sample Notification to NRC S. G. are /

C-3 j ,

pursuant to $50.72 - I th)(2) of 10 CF R Additional in'spect ' 11 tubes inh

Part 50 S. G. is C-3 each S. and plu defective tu

N. A. N. A.

Notification to N 2 pursuant to $50.72 '

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, (h)(2) of 10 CF R j '

Part 50

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Where N is the number of steam generators in the uriit, and n is the number of steam generators 2 .

ins S=3 % during an inspection n

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$dL awl *- 33s3 REACTOR COOLANT SYSTEM OPERATIONAL"l.EAKAGE LIMITING CON 0! TION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAXAGE, v-

b. 1 gpm UNIDENTIFIED LEAKAGE.

c. [generators ota r actor-to-secondary leakage through all steam not isolated from the Reactor Coolant System and gallons per day through any one steam generator, g/SO d, 10 gpm IDENTIFIED LEAKAGE from tne Reactor Coolant System,'

A M

e. 8 gpm per RC pump CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 + 20 psig, and

f. The leakage from each Reactor Coolant System Pressure Isolation g' valve specified in Table 3.4-1 shall be limited to 0.5 gpm per i nominal inch of valve size up to a mr.ximum of 5 gpm, at a Reactor [

Coolant System pressure of 2235 + 20 psig.*

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APPLICABILITY: MODES 1,'2, 3, and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWrt within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within a hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with an RCS pressure of less than 600 psig.

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Test pressures less than 2235 psig but greater. than 150 psig are allowed.

Observed leakage shall be adjusted for the actual test pressure up to 2235 osig assuming the leakage to be directly proportional to pressure differential to tne one-half power.

f CALLAWAY, UNIT 1 3/4 4-19 AmendmentNo.[

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. . . . . . . . . . Ms/W 3369. - - - - - - . - - - -

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} :=. . rn ,RFACTOR_ COOLANT SYSTEM

. REVISJoy y O

hb 8ASES 3/4.4.5 STEAM GENERATORS 1

The Surveillance Requirements 'for inspection of the steam generator tubes

- ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is 1 based on a modification of Regulatory Guide 1.83, Revision 1. Inservice Inspection of steam generator tubing is essential in order to maintain surveil-lance of the 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 generator tubing also prov, ides a means of characterizing the nature and cause of any tube degradation so that corrective measures can'be taken. ,

Unscheduled inservice inspections are performed on each steam generator following: (1) reactor to secondary tube leaks; (2) a seismic occurrence greater than the Operating Basis Earthquake; and (3) a loss-of-coolant accident

. requiring actuation of the Engineered Safety Features, which for this Specification is defined to be a break greater than that equivalent to the

.. severance of a 1" inside diameter pipe, or, for a main steamline or feedline, a break greater than that equivalent to a steam generator safety valve falling open; to ensure that steam' generator tubes retain sufficient integrity for continued operation. Transients less severe than these do not require inspections because the resulting stresses are well within the stress criteria V ., ' established by Regulatory Guide 1.121, w nplugged steam generator tubes must be ca able of withstandin The plant.is expected to be oper ted in a ma er such that the secondary coolant will be maintained within t se chemistry imits found to result in negligible corrosion of the steam g nerator tubes. If the secondary coolant chemistry is not maintained within these limits, lo lized corrosion may d likely result in stress corrosion racking. The ext ent of cracking during plant operation would be limited ty the limitation o' steam generator tube leakage between the Reactor Coola System and the 50condary Coolant System gallons per day'l er steam generator).

(reactor-to-secondary leakage =

Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loacs imposed during normal operation and by postulated accident / Operating plants have demonstrated that reactor-to-secondary leakage of 0 gallons per day per s team geanerator can readily be detected by radiation monitors of steam generator hinwdown, Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will b'e located 3am!

Id ugge yj gjyyg, ,

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 examinations.

Pluggingwillberegredforalltubeswithimperfectionsexceedingthe

' t' mu na=4=1 ' tM c"::: . Steam generator s pluggin limit.:' _

W (1CyiY' CALLAWAY - UNIT 1 B 3/4 4-3

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/dbA%C-3ns*8f j REVISION _

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REACTOR COOLANT SYSTEM BA50.s ,3 511 AM Cl NTRATORS (Continued)

Lube inspections of operating planth have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thigkness. Results f rom WCAP-10043 have been used to establish plugging limit.

7 Whenever the results of any steam generator tubing inservice inspection f all into Category C-3, these results will be reported to the Commission pursuant to Specification 6.9.2 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 examinations, tests, additional l eddy-current inspection, and revision of the Technical Specifications, if necessary.

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 monitor and detect leakage from the reactor coolant pressure bounda ry. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems,* May 1973.

g 3 /4. 4. 6. 2 OPERATIONAL LEAKAGE g

H

) ' PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may '

j be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

]ndustry experience has shown that while a limited amount of leakage is expected f rom the RCS, the unident'.fied portion of this leakage can be reduced l to a threshold value of less than 1 gpm. This thre- is su. ientiv

e. i low to ensure e r1 cetection >.,1 addit

, . - 6m Bfd

The otal steam generator tube leakage limit of 4-gem Tor all steam '

t nerators not isolated f rom the RC5 ensures that the dosage contribution f rom ie tube leakage will be ,1imited to a small fraction of 10 CFR Part 100 dose ;

guideline values in the event of either a steam generator tube rupture or steam i line break. The ' limit is c m i n . J th the assumptions used in the analysis of these ccidents. The v gpd 1 kage limit per steam generator sures that ste generator tube ntegrity s maintained in the event of a m in steam line upta.? or under DCA condit ons,

' /,tc pp d / S*O wrervaN WW f The 10 ppm IDENTIFIED $EAKAGE limitation provides allowance for a limited amount : %. kno n source hose presence w 1 not interfer the sietection of UNIDE 6 JIED u. 'GE the L::.~ e Dete uion 5ys ems, l '

l i The CONTROLLLO LE AKAGE limitation restricts operation when the total flow l

i trom the reactor coolant pump seals exceeds 8 gpm per RC pump bt a nominal RCS *,;

pre.. ore n1 ??35 p>ig. This limitation ensures adequate performance of the RC i pump w.il t. ,

CAIiAWAY - UN]1 1 B 3/4 4-4 k l f y cy p- g.jy / g ,) $ + / c. ,0 A M N" 5 S

fllaMstanv p u kkmw -4 la M% N~

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I ULNRC-3358 ATTACHMENT 2 TECHNICAL SPECIFICATION CHANGES (RE-TYPED) 1 l

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1 74 L A/M- 33 s 8 REACTOR COOLANT SYSTEM, 3/4.4.5 STEAM GENERATORS .

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1 LIMITING CONDITION FOR OPERATION 1

3.4.5 Each steam generator shall be OPERABLE. )

APPLICABILITY: MODES 1,2,3, AND 4. )

l ACTION:

With one or more steam generators inoperable, restore the inoperable steam l Generator (s; io OPERABLE status prior to increasing T.y, above 200 F.

1 SURVEILLANCE REQUIREMENTS 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 Generator Samole Selection and insoection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Tabla 4.4-1.

4.4.5.2 Steam Generator Tube Samole Selection and insoection - The steam generator tube minimum sample size, inspection result classification, and the 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. When applying the exceptions of !

4.4.5.2.a through 4.4.5.2.c, previous defects or imperfections in the area repaired by l sleeving are not considered an area requiring inspection. The tubes selected for each l inservice inspection shallinclude at least 3% of the total number of tubes in all (

steam generators; the tubes selected for these inspections shall be selected on a l random basis except: l

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;

b. Th'e first sample of tubes selected for cach inservice inspection (subsequent to the preservice inspec.tior) o.' each steam generator shall include:

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CALLAWAY - UNIT 1 3/4 4-11

i 74 4 N A & 3 3 S'2P REACTOR COOLANT SYSTEM l SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Accootance Criteria

a. As used in this specification:

1) Imoerfection means an exception to the dimensions, finish or 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;

2)

Dearadation means a service-induced cracking,

wastage, wear or general corrosion occurring on either inside or outside of a tube;

3) Dearaded Tube means a tube containing imperfections greater than l or equal to 20% of the nominal wall thickness caused by i

degradation;

4) % Dearadation means the percentage of the tube wall thickness i affected or removed by degradation; l 5) Oefect means an imperfection of such severity that it exceeds the

! plugging or repair limit. A tube or sleeve containing a defect is defective;

6) Pluaaino or Reoair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving and is equal to 40% of the nominal tube wall thickness.

The plugging limit for electrosfeeves is equal to 20% of the nominal sleeve wall thickness.

7) Unserviceable describes the condition of a tuba if it leaks or contains a defect large enough to affect its structura' integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam l line or feedwater line break as specified in Specification 4.4.5.3c.,

I above;

8) Tube insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. For a tube repaired by sleeving, the t tube inspection shallinclude the sleeved portion of the tube; and i

CALLAWAY - UNIT 1 3/4 4-14

1 74/ A/4& 3 3:5~8 REACTOR COOLANT SYSTEM l SURVEILLANCE REQUIREMENTS (Continued) i

! 9) Preservice Insoection means an inspection of the fulllength of each tube in each steam generator performed by eddy current techniques I prior to service to establish a baseline condition of the tubing. This <

. inspection shall be per_ formed prior to initial POWER OPERATION !

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

10) Tube Reoair refers to a process that reestablishes tube serviceability.

Acceptable tube repairs will be performed by the following processes: ;

I

a. Electrosleeving as described in Framatome Technical Report BAW-10219P, Revision 1, "Electrosfeeving Qualification for PWR Recirculating Steam Generator Tube Repair." March 1996

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by sleeving all tubes exceeding the plugging or repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reoorts

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2:

l b. The complete results of the steam generator tube inservice inspection shall l

be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 montns following the completion of the inspection. This Special Report shallinclude:

1) Number and extent of tubes and sleeves inspected,

2) Location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged or repaired,

c. Results of steam generator tube inspections, which fall into Category C-3,

{ shall be reported in a Special Report to the Commission pursuant to

! Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations

conducted to determine cause of the tube degradation and corrective i measures taken to prevent recurrence.

CALLAWAY - UNIT 1 3/4 4-15

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TABLE 4.4-2 O STEAM GENERATOR TUBE INSPECTION r-si 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION

) Sarnple Sita Result Action Required Result Action Required Result Action Required N.A. N.A. N.A. N.A.

h A rninirnum of S Tubes per S.G.

C-1 None 2

~

Plug or repair defective C-1 None N.A. N.A.

~.4

.a C-2 tubes and inspect additional 2S tubes in this S.G. ~

C-2 Plug or repair $~ective C-1 Nono tubes and inspect additional 4S tubes in this S.G.

C-2 Plug or repair defective tubes G) C-3 Perform action for C-3 g result of first sample b C-3 Perform action for C-3 N.A. N.A.

a result of first sample q

C-3 inspect all tubes in this All other S.G.s None N.A. N.A. h g

S. G., plug or repair are C-1 defective tubes and inspect 2S tubes in each other S.G. g Some S.G.s C-2 Perform action for C-2 N.A. N.A. t Notification to NRC but no additional result of second sample y S.G. are C-3 ba pursuant to 150.72 (b)(2) of 10 CFR Part g

50 g .

Additional S. G. is inspect all tubes in each N.A. N.A.

C-3 S.G. and plug or sepair defective tubes. ,

Notification to NRC pursuant to 150.72 -

(b)(2) of 10 CFR Part 50 S=3N% Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection l

n i

i

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lALJM - 33s~g REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION l

3.4.6.2 Reactor Coolant System leakage shall be limited to: l J

a. No PRESSURE BOUNDARY LEAKAGE, l

b. 1 gpm UNIDENTIFIED LEAKAGE, I

c. 600 gpd total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 150 gallons per day through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 8 gpm per RC pump CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 i 20 psig, and

f. The leakage from each Reactor Coolant System Pressure Isolation Valve

pecified in Table 3.4-1 shall be limited to 0.5 gpm per nominalinch of valve size up to a maximum of 5 gpm, at a Reactor Coolant System pressure of 2235 20 psig.'

APPLICABILITY: MODES 1,2,3, AND 4.

ACTION:

]

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY '

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With any Reactor Coolant System leakage greater than any one of the .

above lir.uts, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the foilowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with an RCS pressure of less than 600 psig.

Test pressures less than 2235 psig but greater than 150 psig are allowed.

Observed leakage shall be adjusted for the actual test pressure up to 2235 psig 7

l assuming tha leakage to be directly proportional to pressure differential to the

one-half power.

l CALLAWAY - UNIT 1 3/4 4-19 Amendment No. 66

4 7A. L su'e.C - 3359 REACTOR COOLANT SYSTEM BASES ;

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes l ensure that the structuralintegrity 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 essentialin 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, manufacturing 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.

Unscheduled inservice inspections are performed on each steam generator l following: (1) reactor to secondary tube leaks; (2) a seismic occurrence greater than the Operating Basis Earthquake; and (3) a loss-of-coolant accident requiring actuation ;

of the Engineered Safety Features, which for this Specification is defined to be a l break greater than that equivalent to the severance of a 1" inside diameter pipe, or, for a main steamline or feedline, a break greater than that equivalent to a steam generator safety valve failing open; to ensure that steam generator tubes retain l sufficient integrity for continued operation. Transients less severe than these do not require inspections because the resulting stresses are well within the stress criteria established by Regulatory Guide 1.121, which unplugged steam generator tubes must be capable of withstanding.

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 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 Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 150 gallons per day j per steam generator). Cracks having a reactor-to-secondary leakage less than this l limit 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 reactor-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located, plugged or repaired.

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 examinations. Plugging or repair will be required for all tubes with imperfections exceeding the plugging or d repair limit. Steam generator tube inspections of operating plants have demonstrated CALLAWAY - UNIT 1 B 3/4 4-3

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74LA4ec- 334~3

-l REACTOR COOLANT SYSTEM

_ BASES STEAM GENERATORS (Continued) the capability to reliably detect degradation that has penetrated 20% of the original

. tube wall thickness. Results from WCAP-10043 have been used to establish plugging limit.

The plugging or repair limit for the pressure boundary portion of electrosleeves is determined to be 20% through-wall (by NDE).

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.9.2 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 examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. 1 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 monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may l be indicative of an impending gross failure of the pressure boundary. Therefore, the l presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly 1 placed in COLD SHUTDOWN.

Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additionalleakage.

The total steam generator tube leakage limit of 600 gpd for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 600 gpd limit is conservative compared to the assumptions used in the analysis of these accidents. The 150 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.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the i detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow from the reactor coolant pump seals exceeds 8 gpm per RC pump at a nominal RCS pressure of 2235 psig. This limitation ensures adequate performance of the RC pump seals.

CALLAWAY - UNIT 1 8 3/4 4-4 l

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ULNRC-3358 ATTACHMENT 3 l

l SAFETY EVALUATION 1 1 l l l

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ULNRC-3358 Attachment 3 Page 1 of 8 SAFETY EVALUATION INTRODUCTION This proposed amendment revises the Surveillance Requirements of Technical Specification (TS) 3/4.4.5 " Steam Generators" and 3.4.6.2 " Operational Leakage" and associated Bases as appropriate, to address the installation of electrosleeves in the Callaway Plant steam generators.

This license amendment request revises TS 3/4.4.5, 3.4.6.2 and associated Bases to include:

a. Electrosleeving (or sleeving) per Framatome Technical Report BAW-10219P, Revision 1, as an approved tube repair method,

b. The associated sleeve wall depth-based plugging limit value and inspection requirements,

c. Reduction of the tube plugging limit from 48% to 40%

through wall (of the nominal tube wall thickness) to be consistent with NUREG-1431, " Standard Technical Specifications for Westinghouse Plants", and,

d. The reduction of the primary to secondary normal operational leakage limit from 500 to 150 gpd per steam generator.

Currently, tubes with indications of degradation in excess of the plugging criteria are removed from service by plugging.

Removal of a tube from service results in a reduction of reactor coolant flow through the steam generator. This small reduction in flow can impact the margin in the reactor coolant flow through the steam generator in LOCA analyses and I on the heat transfer efficiency of the steam generator.

l Repair of a tube via electrochemical deposition of material maintains the tube heat transfer area and results in a much l smaller RCS flow reduction. Therefore, the use of sleeving l in lieu of plugging helps to assure that minimum flow rates l are maintained in excess of that required for operation at

! full power. Any combination of sleeving and plugging, up to l a level such that the effect will not reduce the minimum reactor coolant flow rate to below the current TS limit or below the plugging limits analyzed in the Callaway Safety Analysis Report is acceptable. The sleeve / plug equivalency results are contained in BAW-10219P.

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ULNRC-3358 Attachment 3 Page 2 of 8 BACKGROUND Callaway has Westinghouse Model F steam generators which utilize 11/16" OD x 0.040" nominal wall thickness tubes. The first ten rows of tubes at Callaway are thermally treated ;

Alloy 600 (1326 tubes), while the remainder of the tubes I (4300 tubes) are mill annealed Alloy 600. The Callaway tubes are hydraulically expanded within the tubesheet region. The pressure utilized for the expansion process is designed to provide a radial preload between the tube and tubesheet such 1 that the tube to tubesheet gap is completely reduced during all conditions.

The current Callaway TS require steam generator tubes with I eddy current indications of 48% through wall or greater to be removed from service. This amendment proposes to permit the repair of degraded steam generator tubes by the installation l of electrosleeves. Electrosleeving is the structural repair of a degraded tube by electrodeposition of ultra-fine-grained high purity nickel on the inner surface of a tube.

1 EVALUATION j 1

Generic Structural Assessment Electroformed sleeves have been designed to Section III, Subsection NB-3300 and applicable code cases, of the 1989 Edition of the ASME Code. Fatigue and stress analyses of the sleeved tube assemblies have been completed in accordance with the requirements of Section III, Subsection NB-3200 and applicable code cases, of the 1989 Edition of the ASME Code.

The results of the primary stress intensity evaluation, primary plus secondary stress intensity range evaluation and fatigue evaluation indicate that the ASME Code allowable limits are not exceeded. That is, stress intensities are bounded by the minimum limits for the electrosleeve material and cumulative fatigue usage is less than 1.0. Therefore, the design of the sleeve pressure boundary meets the design objectives of the original tubing.

Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes" and the ASME Code are used to develop the plugging limit of the sleeve should sleeve wall degradation occur. Potentially degraded sleeves are shown (by test and analysis) to retain burst strength in excess of three times the normal operating pressure differential at end of cycle conditions. No credit for the presence of the parent tube behind the sleeve is assumed when performing the minimum wall / burst evaluation.

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ULNRC-3358 Attachment 3 Page 3 of 8 The sleeve structural analysis utilizes a generic set of design and transient loading inputs which are intended to bound all Westinghouse Model F steam generators. The temperature and pressure variances used in the assumed operating conditions and genc-ic transients are bounding.

An ultrasonic inspection of L lectrosleeve is performed prior to placing the sleeve il 'vice to verify correct electrosleeve position, propF alieve to tube bonding, that the minimum acceptable sleeve th*ckness is achieved and to provide a baseline jnspection of the new pressure boundary.

The loading cycles that were applied to the electrosleeve analysis and testing were those for a 40 year plant life cycle. Therefore, the fatigue analysis is bounding for an operating plant. The results of the fatigue analysis indicate acceptable usage factors for the entire range of permitted sleeve thickness.

Leakage Assessment Leakage testing of 5/8", 3/4" and 7/8" electrosleeve assemblies under conditions considered to be more severe than expected during all operating plant conditions has shown that electrosleeving does not introduce additional primary to secondary leakage during a postulated steam line break event.

Electrosleeves were subjected to thermal and fatigue cycling and then leak tested at pressure differentials of greater than 3110 psi, which far exceeds the expected maximum feed line break or steam line break pressure differential.

Leakage testing has also shown that the electrosleeve is essentially leaktight during all plant conditions. Due to time limitations, leakage testing of electrosleeve specimens for 11/16" tubes has not been performed. These tests are scheduled to be completed before the Refuel 8 outage at Callaway, which is currently scheduled for Fall 1996. The leakage test results will be provided to NRC upon completion.

The leak testing program for 11/16" tubes will be consistent with the previously performed electrosleeve leakage tests.

Corrosion Assessment Nickel has performed well historically with regard to corrosion. Accelerated corrosion tests also show that fine-l grained nickel exhibits resistance to stress corrosion i

cracking equal to or greater than rolled tube transitions.

Any sleeve degradation can be detected by nondestructive examination (NDE).

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ULNRC-3358 Attachment 3 Page 4 of 8 Mechanical Integrity Assessment Mechanical testing of 5/8", 3/4" and 7/8" electrosleeves i indicates that.the axial load bearing capability exceeds the l most limiting pressure end cap loading established by RG 1.121. The sleeve structural integrity requirements. include safety factors inherent to the requirements of the ASME Code.

Installation of electrosleeves restores the integrity of the primary pressure boundary and the tube is leaktight. The structural analysis and mechanical performance of the sleeves are based on installation in the hot leg of the steam generators.

Statistical evaluation was performed to show that the electrosleeve material is not tube size dependent. All structural analysis for the 11/16" electrosleeve has been completed. Mechanical integrity tests for the 11/16" electrosleeve will be provided upon completion. It is expected that the 11/16" electrosleeve design will perform similar to the 5/8", 3/4" and 7/8" designs.

Sleeving of Previously Pluqqed Indications -i

)

The electrosleeve installation requirements applicable to I active tubes which have been identified as containing degradation indications which exceed the repair limit are no different for the sleeving of previously plugged tubes. A new " baseline" inspection of the entire tube length must be performed prior to sleeve installation in a previously plugged tube. Historically at Callaway, only the top of tubesheet region has experienced stress corrosion cracking.

The analysis also supports' sleeve' installation in an OD circumferentially cracked tube, therefore, the extent of the originally identified degradation indication should not affect electrosleeve installation.

EVALUATION The proposed changes to the TS do not involve an Unreviewed Safety Question because operation of Callaway Plant with this change would not:

1. Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.

The electrosleeve configuration has been designed and analyzed in accordance with the requirements of the ASME

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ULNRC-3358 Attachment 3 Page 5 of 8 Code. The applied stresses and fatigue usage for the sleeve are bounded by the limits established in the ASME Code. ASME Code minimum material property values are used for the structural and plugging limit analysis. Mechanical testing has shown that the structural strength of nickel electrosleeves under normal, upset and faulted conditions provides margin to the acceptance limits. These acceptance limits bound the most limiting (3 times normal operating pressure differential) burst margin recommended by RG 1.121.

Leakage testing for 5/8", 7/8" and 3/4" tube sleeves has demonstrated that no unacceptable levels of primary to secondary leakage are expected during any plant. condition.

Similar tests of 11/16" tube electrosleeves will be completed prior to Refuel 8.

The sleeve nominal wall thickness (used for developing the depth-based plugging limit for the sleeve) is determined using the guidance of Regulatory Guide 1.121 and the pressure stress equation of Section III of the ASME Code. The limiting requirement of Regulatory Guide 1.121, which applies to part throughwall degradation, is that the minimum acceptable wall must maintain a factor of safety of three against tube failure under normal operating (design) conditions. A bounding set of design and transient ~ loading input conditions was used for the minimum wall thickness l evaluation in the generic evaluation. Evaluation of the minimum acceptable wall thickness for normal, upset and postulated accident condition loading per the ASME Code indicates these conditions are bounded by the design condition requirement minimum wall thickness.

A bounding tube wall degradation growth rate per cycle and an i NDE uncertainty has been assumed for determining the sleeve TS plugging limit. The sleeve wall degradation extent determined by NDE, which would require plugging sleeved tubes, is developed using the guidance of RG 1.121 and is defined in BAW-10219P to be 20% throughwall.

l The consequences of failure of the sleeve are bounded by the current steam generator tube rupture analysis included in the Callaway FSAR. Due to the slight reduction in diameter caused by the sleeve wall thickness, primary coolant release rates would be slightly less than assumed for the steam generator tube rupture analysis (depending on the break location), and therefore, would result in lower total primary fluid mass release to the secondary system.

The proposed change does not adversely impact any other previously evaluated design basis accident or the results of I

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I ULNRC-3358 Attachment 3 Page 6 of 8 LOCA and non-LOCA accident analyses for the current TS minimum reactor coolant system flow rate. The results of the analyses and testing demonstrate that the electrosleeve is an l ccceptable means of maintaining tube integrity. Furthermore, l per Regulatory Guide 1.83 recommendations, the sleeved tube can be monitored through periodic inspections with present

! NDE techniques. These measures demonstrate that installation of sleeves spanning degraded areas of the tube will restore the tube to a condition consistent with its original design basis.

Conformance of the electrosleeve design with the applicable sections of the ASME Code and results of the leakage and mechanical tests, support the conclusion that installation of electrosleeves will not increase the probability or consequences of an accident previously evaluated.

2. Create the possibility for an accident or malfunction of equipment of a different type than any previously evaluated in the Safety Analysis Report.

Electrosleeving will not advers<,ty affect any plant component. Stress and fatigue snalysis of the repair has shown that the ASME Code and Regulatory Guide 1.121 criteria are not exceeded. Implementation of electrosleeving maintains overall tube bundle structural and leakage integrity at a level consistent with that of the original tubing during all plant conditions. Leak and mechanical l

testing of electrosleeves support the conclusions of the calculations that each sleeve retains both structural and leakage integrity during all conditions. Sleeving of tubes does not provide a mechanism resulting in an accident outside l

, of the area affected by the sleeves. Any accident as a l result of potential tube or sleev- legradation in the i I

repaired portion of the tube is t .ded by the existing tube rupture accident analysis.

Implementation of sleeving will reduce the potential for primary to secondary leakage during a postulated steam line break while not significantly impacting available primary coolant flow area in the event of a LOCA. By effectively isolating degraded areas of the tube through repair, the l potential for steam line break leakage is reduced. These

! degraded intersections now are returned to a condition

! consistent with the Design Basis. While the installation of a sleeve reduces primary coolant flow, the reduction is far below that caused by plugging. Therefore, far greater primary coolant flow area is maintained through sleeving versus plugging.

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, ULNRC-3358 Attachment 3 i Page 7 of 8 I

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3. Reduce the margin of safety as defined in the basis for any Technical Specification.

The electrosleeve repair of degraded steam generator tubes has been shown by analysis to restore the integrity of the

tube bundle consistent with its original design basis I condition, i.e., tube / sleeve operational and faulted condition stresses are bounded by the ASME Code requirements -

l and the repaired tubes are leaktight. The safety factors used in the design of sleeves for the repair of degraded !

tubes are consistent with the safety factors in the ASME Code

used in steam generator design. The portions of the installed sleeve assembly which represent the reactor coolant pressure boundary can be monitored for the initiation and progression of sleeve / tube wall degradation, thus satisfying the requirements of Regulatory Guide 1.83. 'The portion of l 1 the tube bridged by the sleeve is effectively removed from the pressure boundary, and the sleeve then forms the new pressure boundary. The areas of the sleeved tube assembly which require inspection are defined in BAW-10219P.

In addition, since the installed sleeve represents a portion of the pressure boundary, a baseline inspection of these j areas is required prior to operation with sleeves installed. !

The effect of sleeving on the design transients and accident l analyses has been reviewed based on the installation of

? sleeves up to the level of steam generator tube plugging coincident with the minimum reactor flow rate and the i Callaway Safety Analysis.

( Provisional requirements cited in other NRC Safety Evaluation Reports addressing the implementation of sleeving have

! required the reduction of the individual steam generator f normal operation primary to secondary leakage limit from 500 to 150 gpd. Consistent with these evaluations, Union Electric will reduce the per steam generator leak rate limit of 500 gpd in TS 3.4.6.2.c to 150 gpd. The establishment of this leakage limit at 150 gpd provides additional safety margin.

Finally, Union Electric will reduce the tube plugging limit from 48% through wall to 40% through wall to be consistent with NUREG-1431. The establishment of the plugging limit at 40% through wall provides additional safety margin.

l ULNRC-3358 Attachment 3 Page 8 of 8 CONCLUSION Given the above discussions as well as'those presented in the Significant Hazards Consideration, the proposed change.does not adversely affect or endanger the health or safety of.the.

general-public or involve an Unreviewed Safety Question.

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I l j ULNRC-3358 ATTACHMENT 4 SIGNIFICANT HAZARDS EVALUATION l

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ULNRC-3358 Attachment 4

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SIGNIFICANT HAZARDS EVALUATION ,

1 INTRODUCTION i

This proposed amendment revises the Surveillance Requirements I of Technical Specification (TS) 3/4.4.5 " Steam Generators" I and 3.4.6.2 " Operational Leakage" and associated Bases as l I

appropriate, to address the installation of electrosleeves in l

l the Callaway Plant steam generators.

This license amendment request revises TS 3/4.4.5, 3.4.6.2 and associated Bases to include-l l

a. Electrosleeving (or sleeving) per Framatome i f Technical Report BAW-10219P, Revision 1, as an approved tube repair method, l b. The associated sleeve wall depth-based plugging

! limit value and inspection requirements,

c. Reduction of the tube plugging limit from 48% to 40% l I

through wall (of the nominal tube wall thickness) to be consistent with NUREG-1431, " Standard Technical l Specifications for Westinghouse Plants", and, i

d. The reduction of the primary to secondary normal operational leakage limit from 500 to 150 gpd per steam generator.

Currently, tubes with indications of degradation in excess of the plugging criteria are removed from service by plugging. l l Removal of a tube from service results in a reduction of l reactor coolant flow through the steam generator. This small reduction in flow can impact the margin in the reactor l coolant flow through the steam generator in LOCA analyses and l

on the heat transfer efficiency of the steam generator.

Repair of a tube via electrochemical deposition of material j maintains the tube heat transfer area and results in a much I smaller RCS flow reduction. Therefore, the use of sleeving in lieu of plugging helps to assure that minimum flow rates I are maintained in excess of that required for operation at full power. Any combination of sleeving and plugging, up to a level such that the effect will not reduce the minimum reactor coolant flow rate to below the current TS limit or d

below the plugging limits analyzed in the Callaway Safety Analysis Report is acceptable. The sleeve / plug equivalency results are contained in BAW-10219P.

ULNRC-3358 Attachment 4 i Page 2 of 8 l

BACKGROUND Callaway has Westinghouse Model F steam generators which utilize 11/16" OD x 0.040" nominal wall thickness tubes. The I first ten rows of tubes at Callaway are thermally treated Alloy 600 (1326 tubes), while the remainder of the tubes (4300 tubes) are mill annealed Alloy 600. The Callaway tubes are hydraulically expanded within the tubesheet region. The pressure utilized for the expansion process is designed to provide a radial preload between the tube and tubesheet such that the tube to tubesheet gap is completely reduced during all conditions.

The current Callaway TS require steam generator tubes with eddy current indications of 48% through wall or greater to be removed from service. This amendment proposes to permit the repair of degraded steam generator tubes by the installation of electrosleeves. Electrosleeving is the structural repair of a degraded tube by electrodeposition of ultra-fine-grained high purity nickel on the inner surface of a tube.

EVALUATION I

Generic Structural Assessment l Electroformed sleeves have been designed to Section III, Subsection NB-3300 and applicable code cases, of the 1989 Edition of the ASME Code. Fatigue and stress analyses of the I sleeved tube assemblies have been completed in accordance with the requirements of Section III, Subsection NB-3200 and applicable code cases, of the 1989 Edition of the ASME Code.

The results of the primary stress intensity evaluation, primary plus secondary stress intensity range evaluation and l fatigue evaluation indicate that the ASME Code allowable limits are not exceeded. That is, stress intensities are bounded by the minimum limits for the electrosleeve material and cumulative fatigue usage is less than 1.0. Therefore, the design of the sleeve pressure boundary meets the design objectives of the original tubing.

Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes" and the ASME Code are used to develop the plugging limit of the sleeve should sleeve wall degradation occur. Potentially degraded sleeves are shown (by test and analysis) to retain burst st rength in excess of three times the normal operating pressures differential at end of cycle conditions. No credit for the presence of the parent tube behind the sleeve is assumed when performing the minimum wall / burst evaluation.

ULNRC-3358 Attachment 4 Page 3 of 8 The sleeve structural analysis utilizes a generic set of design and transient loading inputs which are intended to bound all Westinghouse Model F steam generators. The temperature and pressure variances used in the assumed operating conditions and generic transients are bounding.

An ultrasonic inspection of the electrosleeve is performed prior to placing the sleeve in service to verify correct

! electrosleeve position, proper sleeve to tube bonding, that j the minimum acceptable sleeve thickness is achieved and to l provide a baseline inspection of the new pressure boundary.

The loading cycles that were applied to the electrosleeve analysis and testing were those for a 40 year plant life l

cycle. Therefore, the fatigue analysis is bounding for an l operating plant. The results of the fatigue analysis l indicate acceptable usage factors for the entire range of j permitted sleeve thickness.

l Leakage Assessment

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' l Leakage testing of 5/8", 3/4" and 7/8" electrosleeves under conditions considered to be more severe than expected during l all operating plant conditions has shown that electrosleeving does not introduce additional primary to secondary leakage i I

during a postulated steam line break event. Electrosleeves were subjected to thermal and fatigue cycling and then leak tested at pressure differentials of greater than 3110 psi, t

which far exceeds the expected maximum feed line break or steam line break pressure differential. Leakage testing has also shown that the electrosleeve is essentially leaktight during all plant conditions. Due to time limitations, leakage testing of electrosleeve specimens for 11/16" tubes has not been performed. These tests are scheduled to be completed before the Refuel 8 outage at Callaway, which is I currently scheduled for Fall 1996. The leakage test results will be provided to NRC upon completion. The leak testing program for 11/16" tubes will be consistent with the previously performed electrosleeve leakage tests.

Corrosion Assessment Nickel has performed well historically with regard to corrosion. Accelerated corrosion tests also show that fine-grained nickel exhibits resistance to stress corrosion cracking equal to or greater than rolled tube transitions.

Any sleeve degradation can be detected by nondestructive examination (NDE).

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ULNRC-3358 l Attachment 4 Page 4 of 8 Mechanical Integrity Assessment Mechanical testing of 5/8", 3/4" and 7/8" electrosleeves l indicates that the axial load bearing capability exceeds the most limiting pressure end cap loading established by RG

! 1.121. The sleeve structural integrity requirements include safety factors inherent to the requirements of the ASME Code.

Installation of electrosleeves restores the integrity of the primary pressure boundary and the tube is leaktight. The structural analysis and mechanical performance of the sleeves are based on installation in the hot leg of the steam generators.

Statistical evaluation was performed to show that the electrosleeve material is not tube size dependent. All structural analysis for the 11/16" electrosleeve has been completed. Mechanical integrity tests for the 11/16" electrosleeve will be provided upon completion. It is expected that the 11/16" electrosleeve design will perform similar to the 5/8", 3/4" and 7/8" designs.

Sleeving of Previously Plugged Indications The electrosleeve installation requirements applicable to active tubes which have been identified as containing degradation indications which exceed the repair limit are no different for the sleeving of previously plugged tubes. A new " baseline" inspection of the entire tube length must be performed prior to sleeve installation in a previously plugged tube. Historically at Callaway, only the top of tubesheet region has experienced stress corrosion cracking.

The analysis also supports sleeve installation in a OD circumferentially cracked tube, therefore, the extent of the originally identified degradation indication should not affect electrosleeve installation.

EVALUATION The proposed changes to the TS do not involve an Unreviewed Safety Question because operation of Callaway Plant with this change would not:

1. Involve a significant increase in the probability of occurrence or the consequences of an accident or i malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.

l The electrosleeve configuration has been designed and analyzed in accordance with the requirements of the ASME

i ULNRC-3358 Attachment 4 Page 5 of 8 Code. The applied stresses and fatigue usage for the sleeve are bounded by the limits established in the ASME Code. ASME l Code minimum material property values are used for the structural and plugging limit analysis. Mechanical testing has shown that the structural strength of nickel electrosleeves under normal, upset and faulted conditions provides margin to the acceptance limits. These acceptance limits bound the most limiting (3 times normal operating I pressure differential) burst margin recommended by RG 1.121.

Leakage testing for 5/8", 7/8" and 3/4" tube sleeves has demonstrated that no unacceptable levels of primary to ,

secondary leakage are expected during any plant condition. I Similar tests of 11/16" tube electrosleeves will be completed )

prior to Refuel 8.

l The sleeve nominal wall thickness (used for developing the depth-based plugging limit for the sleeve) is determined using the guidance of Regulatory Guide 1.121 and the pressure stress equation of Section III of the ASME Code. The limiting requirement of Regulatory Guide 1.121, which applies to part throughwall degradation, is that the minimum acceptable wall must maintain a factor of safety of three against tube failure under normal operating (design) conditions. A bounding set of design and transient loading input conditions was used for the minimum wall thickness evaluation in the generic evaluation. Evaluation of the i minimum acceptable wall thickness for normal, upset and postulated accident condition loading per the ASME Code indicates these conditions are bounded by the design condition requirement minimum wall thickness.

A bounding tube wall degradation growth rate per cycle and an NDE uncertainty has been assumed for determining the sleeve TS plugging limit. The sleeve wall degradation extent determined by NDE, which would require plugging sleeved tubes, is developed using the guidance of RG 1.121 and is defined in BAW-10219P to be 20% throughwall.

The consequences of failure of the sleeve are bounded by the current steam generator tube rupture analysis included in the Callaway FSAR. Due to the slight reduction in diameter caused by the sleeve wall thickness, primary coolant release rates would be slightly less than assumed for the steam generator tube rupture analysis (depending on the break l

location), and therefore, would result in lower total primary l fluid mass release to the secondary system.

l The proposed change does not adversely impact any other i previously evaluated design basis accident or the results of f

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I ULNRC-3358 Attachment 4 Page 6 of 8 LOCA and non-LOCA accident analyses for the current TS minimum reactor coolant system flow rate. The results of the analyses and testing demonstrate that the electrosleeve is an acceptable means of maintaining tube integrity. Furthermore, per Regulatory Guide 1.83 recommendations, the sleeved tube can be monitored through periodic inspections with present NDE techniques. These measures demonstrate that installation of sleeves spanning degraded areas of the tube will restore the tube to a condition consistent with its original design basis.

Conformance of the electrosleeve design with the applicable sections of the ASME Code and results of the leakage and mechanical tests, support the conclusion that installation of l electrosleeves will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of i accident from any previously evaluated in the Safety '

Analysis Report, l

Electrosleeving does not represent a potential to adversely l affect any plant component. Stress and fatigue analysis of l the repair has shown that the ASME Code and Regulatory Guide l 1.121 criteria are not exceeded. Implementation of l electrosleeving maintains overall tube bundle structural and I leakage integrity at a level consistent to that of the originally supplied tubing during all plant conditions. Leak and mechanical testing of electrosleeves support the conclusions of the calculations that each sleeve retains both structural and leakage integrity during all conditions.

Sleeving of tubes does not provide a mechanism resulting in an accident outside of the area affected by the sleeves. Any accident as a result of potential tube or sleeve degradation in the repaired portion of the tube is bounded by the existing tube rupture accident analysis.

Implementation of sleeving will reduce the potential for primary to secondary leakage during a postulated steam line break while not significantly impacting available primary coolant flow area in the event of a LOCA. By effectively isolating degraded areas of the tube through repair, the potential for steam line break leakage is reduced. These degraded intersections now are returned to a condition consistent with the Design Basis. While the installation of a sleeve reduces primary coolant flow, the reduction is far below that caused by plugging. Therefore, far greater

1 ULNRC-3358 Attachment 4 Page 7 of 8 primary coolant flow area is maintained through sleeving i versus plugging. l

3. Involve a significant reduction in a margin of safety.. l The electrosleeve repair of degraded steam generator tubes I has been shown by analysis to restore the integrity of the l tube bundle consistent with its original design basis i condition, i.e., tube / sleeve operational and faulted ;

I condition stresses'are bounded by the ASME Code requirements and the repaired tubes are leaktight. The safety factors used in the design of sleeves for the repair of degraded tubes are consistent with the safety factors in the ASME Code :

L used in steam generator design.- The portions of the installed sleeve assembly which represent the reactor coolant l pressure boundary can be monitored for the initiation and ;

I progression of sleeve / tube wall degradation, thus satisfying I i the requirements of Regulatory Guide 1.83. The portion of l the tube bridged by the sleeve is effectively removed from

! the pressure boundary, and the sleeve then forms the new 1 l pressure boundary. The areas of the sleeved tube assembly -i which require inspection are defined in BAW-10219P.

In addition, since the installed sleeve represents a portion l of the pressure boundary, a baseline inspection of these l areas is required prior to operation with sleeves installed.

The effect of sleeving on the design transients and accident i analyses has been reviewed based on the installation of sleeves up to the level of steam generator tube plugging coincident with the minimum reactor flow rate and the Callaway Safety Analysis.

Provisional requirements cited in other NRC Safety Evaluation Reports addressing the implementation.of sleeving have required the reduction of the individual steam generator normal operation primary to secondary leakage limit from 500 to 150 gpd. Consistent with these evaluations, Union Electric will reduce the per steam generator leak rate limit of 500 gpd in TS 3.4.6.2.c to 150 gpd. The establishment of this leakage limit at 150 gpd provides additional safety margin.

Finally, Union Electric will reduce the tube plugging limit from 48% through wall to 40% through wall to be consistent with NUREG-1431. The' establishment of the plugging limit at 40% through wall provides additional safety margin.

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. . . u ULNRC-3358 l Attachment 4 Page 8 of 8 l

CONCLUSION Given the above discussions, the proposed change does not )

adversely affect or endanger the health or safety of the

{

general public or involve a significant hazards '

consideration.

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ULNRC-3358 ATTACHMENT 5 l ENVIRONMENTAL CONSIDERATION l

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ULNRC-3358 Attachment 5 Page 1 of 1 ,

ENVIRONMENTAL CONSIDERATION This proposed amendment revises the Surveillance Requirements of Technical Specification (TS) 3/4.4.5 " Steam Generators" and :

3.4.6.2 " Operational Leakage" and associated Bases as appropriate, to address the installation of electrosleeves in the Callaway Plant steam generators.

The proposed amendment involves changes with respect to the use of facility components located within the restricted area, !

as defined in 10 CFR 20, and changes surveillance I requirements. Union Electric has determined that the proposed I amendment does not involve:

(1) A significant hazard consideration, as discussed in Attachment 4 of this amendment application; l

(?) A significant change in the types or significant increase in the amounts of any effluents that may be released offsite; I

(3) A significant increase in individual or cumulative occupational radiation exposure.

l Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22 (c) (9) . Pursuant to 10 CFR 51.22(b), no environmental impact statement or enviromental assessment need be prepared L in connection with the issuance of this amendment.

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ML20107F079

Text

Dearadation means a service-induced cracking,

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