ML20126M291

Proposed Tech Specs Changes 205 & 71 to Allow SG Tube Sleeving,Per Westinghouse & B&W Processes

ML20126M291
Person / Time
Site:	Beaver Valley
Issue date:	12/30/1992
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML19303F172	List: ML20126M264 ML20126M269 ML20126M291 ML20126M296 ML20126M300
References
NUDOCS 9301080261
Download: ML20126M291 (55)

Text

.

ATTACHMENT A-1 Beaver _ Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 205 IGRKED-UP REPLACEMENT PAGES a

Revise the Technical Specification as follows:

Remove Paces Insert Paces 3/4 4-8 3/4 4-8 i

l 3/4 4-9 3/4 4-9 3/4 4-10 3/4 4-10 3/4 4-10a 3/4 4-10a 3/4 4-10d 3/4 4-10d B 3/4 4-2a B 3/4 4-2a p

L l'

9301080261 921230 PDR ADOCK 05000334 i

P PDR I-

CPR-66 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2,

3 and 4.

ACTION:

With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 200*F.

SURVEILLANCE REQUIREMENTS 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 ninimum number of steam generators'specified in Table 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.

Steam generator tubes shall be examined in accordance with Article 8 of Section V

(" Eddy current Examination of Tubular Products") and Appendix IV to Section XI

(" Eddy Current Examination of Nonferromagnetic Steam Generator Heat Exchanger Tubing")

of the applicable year and addenda of the ASME Boiler and Pressure Vessel Code required by 10CFR50, Section 50.55a(g).$ The tubes selected for l

/ each inservice inspection shall include at least 3% of the total-number of tubes in all steam generators; the tubes -elected for these inspections shall be selected on a random basis except:

23& EAT A 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. le a4 kbe.s selechd h eacl, smp b.

The first A inservice inspection (subsequent to the l

preservice inspection) of each steam generator shall include:

g gf 1.

All nonplugged t

es that previously had detectable-Wall penetrations 20 D, and l

@ tt eYt** When 2.

Tubes in those areas where experience has indicated I))SSRT 8 g potentialproblemsy,ad BEAVER VALLEY - UNIT 1 3/4 4-8 Amendment No.

fAdf0SEO x

l Insert A 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 sleeving rare not considered an area requiring' reinspection.

Insert B-3.

At least-3%

of the total number of sleeved tubes in all three steam generators.- A sample size less than 3%

is acceptable provided all the sleeved tubes in the steam generator (s) e alained during' the refueling.

outage are-inspected.

These inspections will include-both the tube and the sleeve, and 4.

A tube-inspection pursuant to Specification

~

4.4.5.4.a.8.

If any se) 3d tube does not permit the passage of the eddy currs probe for.a tube or sleeve inspection, this shall-be-recorded-and an adjacent-tube shall be selected and subjected tto a

tube-inspection.

Insert C c.

The tubes -selected as the second 'and third-samples (if-required by Table 4.' 4 - 2 ) during each inservice inspection 1 may be subjected to a partial tube inspection-provided:

'1.

The tubes selected forethese samples include 1 he_ tubes t

from those' -areas of the tube' sheet array where tubes with imperfections were previously-found, and 2.

'The inspections-include those--portions of'the-tubes.

where-imperfections were previouslyJfound.

~_

~~... _.

DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c.

nd and third inservice inspections may be leee a

full tu e sp on by concentratin ng at least 2}VSEA7 C,i" 50%

of the tubes to e

the inspection on those-areas of s eet array and on tions of the w ere tubes with imperfections were previous y a

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

Catecorv Insoection Results C-1 Less than 5%

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

C-2 One or more

tubes, but not more than 1%

of the total tubes inspected are defective, or between 5%

and 10%

of the total tubes

• inspected are degraded tubes.

C-3 More than 10%

of the total tubes inspected are degraded tubes or more than 1%

of the ' inspected tubes are defective.

g Note:

In all inspections, previously deg$raded tubeSw must exhibit significant 10%)

l Of 5/efW3 [furthbr wall penetrations to be incl.ed in the above percentage calculations.

4.4.5.3 InsDection Frecuencies The above required inservice inspections of steam generator tubes shall-be performed at the fol3owing frequencies:

a.

The-first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.

Subsequent inservice inspections shall be performed at-intervals of not less than 12 nor more than 24 calendar months

after the previous inspection.

If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1

-category or if two consecutive inspections demonstrate that -previously observed degradation has not continued-and no additional degradation has. occurred, the inspection interval may be extended to a maximum of once per 40 months.

l I

BEAVER VALLEY - UNIT 1 3/4 4-9 PA0P03Eb

DPR-66.

REACTOR COOLANT SYSTEM r

SURVEILLANCE REQUIREMENTS--(Continued) b.

If the inservice inspection of a-steam generator conducted in accordance with Table 4.4-2

. requires a third sample-inspection.

whose results-fall--in Category-C-3, the-inspection frequency shall be reduced =to'at least once per 20 months.

The reduction in~ inspection 1 frequency.shall apply until a-subsequent inspection-demonstrates that a third sample inspection is not required.

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 conditionsy -

.l g

tube-to[-tube 1.

Primary-to-secondary tube leaks (not' including leaks-l originating from sheet-welds).in excess of the limits of Specification 3.4.6.2, 2.

A -seismic occurrence greater than the operating Basis Earthquake, 3.

A loss-of-coolant. accident requiring actuation-of the engineered safeguards, or 4.

A main steam line or feedwater line break.

4.4.5.4 Acceotance Criteria a.

As used in this Specification:

of g/ggpq 1.

Imoerfection means an exceptio to ;the dimensions,-

finish.or contour of a

tube-from that required-by I

fabrication drawings or specifications.

Eddy-current:

testing. indications below 20%-of the' nominal' tube; wall-thickness,.

if detectable, '

may be considered-as imperfections.-

2.

Deoradation means a service-induced cracking, wastage, wear.or general corrosion occurring on either inside or outside of a tube, of s hagg.

- l.

3.

Dearaded Tube means-a tube.containing imperfections l

20%

of the nominal wall thickness caused b

degradation.

g,pqS

_,p 4.-

Decradation means the percentage'of the' tube wall l

thickness affected'or removed by degradation.

yttefth bN or uelto p

(

BEAVER VALLEY - UNIT.1.

3/4 4 ;-

PtoAosEb

c-OPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) of repage i

5.

Defect means an imper ection of such severity that it exceeds the plugging limit.

A tube containing a l

defect is defective.

Any tube which does not-permit the passage of the oddy-current inspection probe shall be deemed a defective tube, l

6.

Tdwtgina__ Limit means the imperfection de t beyond whic h Q all be cwe com service INSE:47- )-*because it may become se '

abl+-prin to the next inspect p equal to 40% of the nomi a

ual-1 -tfilckness.

7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis E

5 quake, a

loss-of-coolant

accident, or a stor or feedwater line break as specified in 4.

ve.

8.

Ty

'n means an inspection of the steam g-from the point of entry (hot leg side) ind the U 'aend to the top support to the INSEkr &

l shall be determined OPERABLE after-

.rresponding actions (plug tubes tub :4 all excet

,ging limit)-and all centnining

- thrcq y

.A-reqf red by Table 4.4-N @h ~ -

4.4.5.5 Reports a.

Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged n

each steam generator shall be reported.to the l

Commissi in a

Special Report pursuant to Specification or repa;te) ad Clttvt 6.D.2.

h b.

The complete results of the steam generator tube inservice l

inspection shall be submittert to the Commission in a

Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection.

This Special Report shall include:

g gg 1.

Number and extent of tubes inspected.

l 2.

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

of refa[r1R/

3.

Identification of tubes pluggedf l

BEAVER VALLEY - UNIT 1 3/4 4-10a Amendment No.

MOPOSEb

~..

. Jnsert-D 6..

EJuccina or- ' Repair Limit means-the_ imperfection depth at or _beyond which _the tube:- shall be removed from service by _ plugging or' repaired by sleeving in--the.

affected area-because it~ may-become--unserviceable prior' to: __the next inspection.- The plugging or~ repair 1 limit-imperfection-; depths-are specified-in percentage of nominaliwa11' thickness _as follows:

-a, original tube wall 40%

b.

Babcock & Wilcox kinetic welded sleeve wall 40%

c.

Westinghouse laser welded sleeve wall ~

31%

Insert E 9.

Tube Repair' refers to' sleeving which is: used to maintain a

tube in-service Dor ' return a. tube to service.

This includes the removalLof-plugs that were installed as-a corrective or preventivo measure.,-The following sleeve designs-have been found= acceptable:

a)

Babcock

_Wilcox

-kinetic welded sleeves,.

BAW-2094P, Revision 1

including -kinetic sleeve-

" tooling" and installation process parameter

changes, b)

Westinghouse laser welded

sleeves, WCAP-13483,_

Revision 1.

't

TABLE 4.4-1 g

MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators ~per Unit Two Three Four Two Three Four First Inservice Inspection All One Two Two l

l 2

3 Second & Subsequent Inservice Inspections One One One One Table Notation:

1.

The inservice inspection may be limited to one steam generator on a. rotating schedule encompassing 3 N %.of.the tubes (where H is' the number of steam generators in the plant) if thel results.of the first or previous inspections indicate that all steam generators are performing in a-like. manner.

Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam ' generators.- Under such circumstances the sample sequence shall 1x2 modified to inspect the most severe conditions.

2.

The other-steam generator not inspected during the first inservice inspection shall be inspected.

The third and subsequent inspections should follow the instructions described in 1 above.

3.

Each of the other two ' steam' generators not inspected during. the first inservice inspections shall be inspected during the second and third inspections. 'The: fourth;and subsequent' inspections shall follow the instructions described in 1 above.

3/44-10/c!L l'

BEAVER VALLEY-- UNIT 1.

TAlil.E 4. 4-2 SITR4 GF11F1M'iOR 'ILTIE INSPFUPIG1 g

,o ISP SAMPIE IllSPILTIO!1 2!1D SAMPLE I!1SPirPIGi 3RD SAMPIE INSPfffIG1 7

m*

Sanple size Result Action Rcquired Result Action Required Result Action Rcquired A minimum of C-1 None

!!/A 11/A ti/A N/A S 'Ibl>u per S. G.

C-2 P1 defective tubes C-1 flone N/A N/A and inspect additional fQ defective tubes C-1 None 2S tubes in this C-2 S. G.

and inspect additional j

4S tubes in this S. G.

defective x

/

y goe Of teP8b C-3 Ibrform action for C-3 result-of first sample C-3 lbrionn action for C-3 N/A N/A result of first sanple j

C-3 Inspect all tuber in Al1 other None N/A

!!/A this S. G., plty S.G.s are l

defective tubes and C-1 inspect 2S tubes in each other S. G.

Some S.G.s Ibrform action for 11/A N/A Notification to NRC C-2 1xit no C-2 result of second pirsuant to additio.n1 sample

,h}

Specification 6.6.

S.G.s are p

C-3 Additional Ins [wct all tubes ~

S.G. is each S.G. and pl N/A N/A l

C-3 defective tubes.

Notification to NRC purstunt to Specifi-cation 6.6.

0 S = 3 1. Where H is the nimier of steam generators in.the unit, and n is the number of steam generators inspecthi dor'iseJ.to irq n tion.

i d /.'. I h '. / i J J.' - t IJ!'l 1 1/44-10/6-A!!kuvink'rit flo.

MoPoseb

DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued)

The surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is 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, 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.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter-limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameter

limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation ~would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator).

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

Operating plants have demonstrated that primary-to-secondary leakage of 500 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 and plugged.

N Of /t/ S Wastage-type defects are unlikely with the all volatile reatment (AVT) of secondary coolant.

However, even if a defect o similar type should develop in service, it will be found during theduled inservice steam generator tube examinations.

Pluggingo will be required of all tubes with imperfections exceeding the plugging op pfpelimit. which, by the dcfinitica cf Specificction

'.5,'

c ic ' F --

-ef the tub ncminci 2: 11 thichnccc.

Steam generator tube inspections of operating plants have uemons rated the capability ec reliably detect degradation that has penet ated 20% of the original tube wall thickness.

pggg Whenever the results of any steam generator tubing inservice l

inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 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 Spe tfications, f

necessary.

BEAVER VALLEY - UNIT 1 B 3/4 4-2a Amendment No.

hkQh036b

,,,*-,-,--,.,-m-

.,mm m-.+---.,v-----,-,

,,-em

,w-----m,----,--------.m.--,,---r---,e.e

,-----m,---

m.----n-----c--+=

-.,,,n-->e.---

Insert F Degraded; steam generator-tubes may-be repaired by the installation-

of:. sleeves-which_-span-the'degradedrtube_section.

_.A steam generator:

' tube with a - sleeve--.' installed meets.the_ structural requirements:of'.

~ tubes-which are.not-degraded,-therefore,_the1 sleeve is considered a part-of the' tube.

The surveillance requirements identify,those-sleeving methodologies-approved for use.

If an installed s)eeve is.

found to have through-wall penetration greater than-or_ equal to-the plugging limit,_ the tube must be plugged.

The_: plugging limit for the sleeve is derived from R.G.

1.121 analysis which utilizes-a 20%

allowance for eddy. current uncertainty in determining the depth of u

tube wall. penetration asid additional degra'dation growth.

1 h

i-i l

i.

i-i i

l

{

i

\\

I l

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- - -., _.. -.... -. _,. _. ~ ~... -. _ - -....... _. ~.. - _.. _ _ _ - - _. -.,,..,,. _, _ - - _. -. -., _...,. _. -. ~.... _

ATTACRMENT A-2

.j

. Beaver Valley Power Station,,'Init !!o ~ 2 Proposed Technical Specification Change No. 71:

MARKED-UP REPLACEMENT PAGES t

Revise the Technical Specification =-as follows:

i i

Remove Paces Insert Paces 3/4 4-11 3/4 4 11 i

3/4 4-12

3) 4 4-12 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4 _14 3/4 4-16 3/4.4-16 B 3/4 4-3 B 3/4 4-3 i

7 6

3 e

i i

!. ~

f l

l l

d

NPF-73 f

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:

With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T above 200*F.

avg SURVEILLANCE RE00iREMENTS 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Samole Selection and Inspection - 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 in-service inspection of steam generator tubes shall be performed at the frequen-cies 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.

Steam generator tubes shall be examined in accordance with Article 8 of Section V (" Eddy Current Examination of Tubular Products") and Appendix IV to Section XI (" Eddy Current Examination of Nonferromagnetic Steam Generator Heat Exchanger Tubing") of the applicable year and addenda of the ASME Boiler and Pressure Vessel Code required by 10 CFR 50, Section 50.55a(g).9 The tubes selected for each l

inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these i pections shall be selected on a randos basis except:

.TNS&W Where experience in similar plants with similar water chemistry indicates a.

critical areas to be inspected, then at least 50% of the tubes inspected shall be from these$ critical areas *Mes selecN Int eacN Sumpft. e The first41nservice inspection (subsequent to the preservice inspection)

I b.

of each steam generator t.11 inclu :

c, kt fere Sts If.

All nonplugge as th previously had detectable wall penetrations than 20%, and I

gefahr 2h Tubes in those areas where experience has indicated potential problessy and D SE9" E +

l BEAVER VALLEY - UNIT 2 3/4 4-11 Amendment No.

fRWOS@

Insert-A 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 sleeving-

_are not considered an area requiring reinspection.

Insert B 3.

At least 3%

of the total number of sleeved. tubessin-~

all three steam generators.

A sample. sira less than-3%

is acceptable provided all the sleeved tubes in the steam generator (s) examined -during the refueling _

outage are inspected.

These inspecticas will include-both the tube and the sleeve, and 4.

A tube

. inspection pursuant

.to Specification 4.4.5.4.a.8.

If any selected tuba does not permit?the passage of the eddy current proPe for.a tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected-to a

tube inspection.

Insert c c.

The tubes selected as the second and third. samples;(if required by Table 4.4-2) during each inservice-inspection may be subjected to a partial 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,-and 2.

The -inspections-include those portions of the-tubes-where imperfections were previouslyLfound.

3 r,

,--en e

+, - -

s

-w

l l

NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continuedl

c. h third inservice inspections may oe less tube inspection y (selectin 0% of the DJEWT C '" t"D'S t D' I"59"Ct'd) th' ISSP

"**5 I th' t"D*

sheet arra e portions of the tubes whereTubet-wi(

im ions were previously found.

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

Category Inspection Results c

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 percent of the total tubes inspected are defective, or between 5 percent and 10 percent of the total tubes inspected are degraded tubes.

C-3 Hore than 10 percent of the total tubes inspected are degraded tubes or more than 1 percent of the inspected "g'S & f of SletMeE a

Note:

In all inspe ions, previously degraded tubesYmust exhibit significant (>10 percent) further wall penetrations to be included in the above percentage calculations.

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

a.

The first inservice inspection shall be performed after 6 Effective Full Power Months bat within 24 calendar months of initial critica'ity.

Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.

If two consecutive inspections following service under All Volatile Treatment (AVT) conditions, not inci ding the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

b.

If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspectior. whose results fall in Category C-3, the inspection frequency snall be in-creased at least once per 20 months.

The increase in the inspection frequency shall apply until a suw eguent inspection demonstrates that a third sample inspection is not required.

BEAVER VALLEY - UNIT 2 3/4 4-12 f20 Pond

r-MPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued)

Additional, unscheduled inservice inspections shall be performed on c.

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 conditionsyi

[

1.

Primary-to-secondary tube)( leaks (not including leaks originating l from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, 2.

A seismic occurrence greater than the Operating Basis Earthquake, 3.

A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4.

A main steam line or feedwater line breal 4.4.5.4 Acceptance Criteria a.

As used in this Specification: of sltevt.

1.

Imperfection means n exception to the dimensions, finish or contour of a tube rom that required by fabrication drawings or l

specifications.

Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2.

Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube D,,j g g6 & y Iff l

3.

Degraded Tube means containingimperfectionsf20pecentl

.of the nominal wall thickness caused by degradation.

4.

% Degradation means the percentace of the tube wall thickness I

affected or removed by degrad nion, g

5.

Defect means an imperfection of such severity that it exceeds the plugginA limit.

A tube containing a defect is defective.

I op %r Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.

6.

-r i uyy h g4.f= 4 + means the imperfection depth at or bevnndM the tube s@ be removecFTrv. 5cr@a betmsTTE~m~ay become un-JgjfgT' D --+ serviceable r ta-%e-nen inspection and is equai Lv 00 pe- _

nom al tube wall thickness.

BEAVER VALLEY - UN.T 2 3/4 4-13 McDoSEb

-1 16 sert"D 6.

Pluccina-or Repair Limit-meanslthe:imperfectioncdeptha at or?=beyond--which the tube shall be removedJfrom-.

-service inr plugging or repaired; by sleeving-in~the affected area' because. it-may ~ become -unserviceable-prior to the next inspection.

The: plugging or repair limit imperfection-depths are-.specifiedLin percentage of nominal wall thickness as follows:

a.

Original tube wall

.40%.

b.

Babcock & Wilcox kinetic welded sleeve 1 wall 40%;

c.

Westinghouse laser welded sleeve wall 31%

Insert _E 9.

Tube Reoair refers to sleeving which-is used' tx)-

maintain a

tube in-service or. return a --tube-to service.

This includes the removal of plugs that were installed as a' corrective or preventive measure.' The; following sleeve designs have been'found acceptable:

a)

Babcock Wilcox kinetic-

~ welded

sleeves, BAW-2094P,- Revision 1

including kinetic' sleeve

" tooling" and installation process-parameter changes.

b)

Westinghouse laser welded

sleeves, WCAP-13483, Revision 1; i-L

NPF-73'

-REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS.(Continued) 7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

8.

Tube Inspection 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.

SNSLAT & -->

e npu -

b.

The steam generator shall be de ined OPERABLE after completing the qorresponding actions (plug all tubes exceeding the plugging limitht.d cil tut:: : ntaining thr: gh =ll cr::k:, required by-Table 4.4-2.

'4.4.5.5 Reports

_g9Q Within 15 days following the completion of each insehce inspection a.

of steam generator tubes, the number of tubes plugged'in each steam l

generator _shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2.

g g

Thecompleteresultsofthesteamgeneratortubeknserviceinspection-l b.

shall be included in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection..This report shall include:

or sleeves l

1.

Number and extent of tubesY nspected.--

i 2.

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

Identificationoftubesplugged[ofN)0a8 l

3.

Results of steam gcnerator tube inspections which fall into c.

Category C-3 shall be reported to the Commission pursuant to Specification 6.6 prior to-resumption of plant operation.

The written report shall provide a description of investigations con-ducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

l BEAVER VALLEY - UNIT 2 3/4 4-14 Amendment No.

PbfoSED

TABLE 4.4-2 co i

9 STEAM GENERATOR TUBE INSPECTION E

.1

=J y

let SAMPLE INSPECTION 2nd SAMPLE INSPECTIOE 3rd SAMPLE INSPECTION Sample Size Result Ect ion Required

.kesult Act ion Required Resu lt Act ion Required i A minimum of

'C-1 None N/A

'N/A N/A N/A Q

S Tubes per S.C.

I

'C-2 Mde fect ive C-1 None-N/A N/A' Tubes and inspect l

addit ional 28 tubes C-2 M defect (we tubes and C-1' None i

( in this S.C.

Inspect. additions; 45 N

tubes in this S.C.

9 -2 PluP de fe ct ive tubes-

}

c/ ftpotf g9p

,C-3 Per form act ion : for ' C-3 result of f atet sample C-3 Perform act ion for C-3 N/A result of'first a mple N/A-h.

i C-3 Inspect all tub in All this S.C., plug ot her

.l defective' tubes and S.C.'s-inspect 2S tubee'in, are C-1 None N/A N/A each other~S.C.

some Perform act ion for C-2 S.C.*e result of second ' eample N/A N/A-C-2 but no add'1 Not(fication to NRC S.C. are of repase pursuant t o 5 %.72 -

C-3 (b)(2) of 10 CFR 4

Part M Add'l

! aspect all tubes S.C. is each S.C. sad plug

.l C-3 defective tubes. -

==

Not(fication to NRd N/A N/A pureuset to SM.72 (b)

(2) of 10 CFR Part M 9

S-I

- lihere n ' is the somber df steam generators inspected during an. inspect ion.

4

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) decay heat removal capabilities for RCS temperatures greater than 350'F if one steam generator becomes inoperable due to single failure considerations.

Below 350'F, decay heat is removed by the RHR system.

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 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, 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.

The plant is expected to be operated in a manner such that the secondary coolant will be maintainec' within those parameter limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameter 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 prilary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during opera-tion will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.. Operating plants have demonstrated that primary-to-secondary leakage of 500 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 and plugged.

,4 Wastage-type c efects are unlikely with the all volatile treatment (AVT) of secondary coolant.

However, even if a defect of similar type should develop in service, it will bo found during scheduled inservice steam generator tube exami-nations.

Plugging"will be required of all tubes with imperfections eweeding the et fep pluggingVlimit.wt.ich, by the-definitica of Specif fectica 4.4.5.4.c i 10% of

-the tube a;;inal wcli thickr as.

Steam generator tube inspections of operating plants have demonstrated the apability to reliably detect degradation that has penetrated 20% of the origin 1 tube wall thickness.

INSERT F Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.6 prior to resumption of plant operation.

Such cases will be considered by the Comission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, test, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

BEAVER VALLEY - UNIT 2 8 3/4 4-3 M0POSO

Insert F Degraded. steam generator tubes may be repaired by the installation of sleeves -which span'the degraded tube section.

A steam generator tube with a

sleeve installed meets the structural requirements of tubes which are not degraded, therefore, the sleeve is considered a part of the tube.

The surveillance requirements identify those sleeving methodologies approved for use.

If an installed sleeve is found to have through wall penetration greater than or. equal to the plugging

limit, the tube must be plugged.

The plugging limit for the sleeve is derived from R.G.

1.121 analysis which utilizes a 29%

allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional degradation growth.

h i

menussiiumm i

J

ATTACHMENT B Beaver Valley Power Station, Unit Nos. 1 and 2 Proposed Technical Specification Change No. 205/71 REVISION OF SPECIFICATION 3.4.5 A.

DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would modify the surveillance requirements of Specification 3.4.5,

" Steam Generators ' and Bases 3/4. 4. 5 to allow sleeving at the tube support plate and tubesheet regions in accordance with the processes performed by the vendors Babcock and Wilcox (B&W) and Westinghouse.

The following proposed changes revise the surveillance requirements and bases to identify sleeving as a repair method for defective tubes:

Surveillance Requirement (SR) 4.4.5.2 has been revised by including an exception to the degraded tube inspection requirements for tubes that have been repaired by sleeving.

SR 4.4.5.2.b has been revised to incorporate the Standard Technical Specification (STS) words including the addition of Items 3

and 4.

Items 1 and 2 include editiorial changes, Item 3 specifies the required population of sleeved tubes to be inspected and Item 4 provides alternate inspection requirenients when a

tube or sleeve does not permit passage of an eddy current inspection probe.

SR 4.4.5.2.c has been revised to incorporate the STS words and provides the criteria for partial tube inspection during the second and third inservice inspections.

The note has been revised to address sleeves.

SR 4.4.5.3.c includes aditorial changes.

SR 4.4.5.4.a, Items 1,

2, 3 and 4 have been revised to include sleeves in the steam generator acceptance criteria definitions.

SR 4.4.5.4.a, Item 5 has been revised to include tube repair in the definition of defect.

Item 6 has been changed to include the sleeve imperfection plugging limits.

Item 9 has been added to define " tube repair" and specify the approved sleeving vendors.

SR 4.4.5.4.b has been revised by including tube repair when steam generator operability is determined.

SR 4.4.5.5.a and 4.4.5.5.b have been revised by including tube repair by sleeving in steam generator tube inspection reports.

BV-1 Table 4.4-1 includes a revised page number.

Table 4.4-2 has been revised to include tube repair in the actions required for categories C-2 and C-3.

The BV-1 page number has also been revised.

Base 9 4/4.4.5, Steam Generators, has been revised by including the

']a s is for repalr of degraded steam generator tubes by clee/ing.

-ATTACHMENT B,. continued-Proposed 1 Technical Specification Change No. 205/71L Page 2 B.

BACKGROUND Pressurized water reactor steam generators have experienced tube degradation related to corrosion' phenomena such as

wastage, pitting,-

intergranular

attack, stress corrosion cracking and crevice. corrosion along with other phenomena such as denting-and-vibration wear.

Tubes that experience excessive degradation reduce the integr'ity of the primary-to-secondary pressure.

boundary.

These tubes-are considered defective and nust be repaired or plugged and removed from~ service.

The installation of steam generator tube plugs removes the heat transfer surface of the plugged tube from service and leads to a reduction in the primary coolant flow available for core cooling.

Sleeving is a steam generator tube repair method which secures a-length of-tubing (sleeve) having an outer _ diameter slightly smaller than the inside of the steam generator tube and also spans the degraded region of_ the parent tube.

_ Installation of steam generator sleeves does not greatly affect the heat _ transfer capability or the primary coolant flow rate through the tube-being

sleeved, therefore, a

large number of sleeves can be installed without significantly affecting the-operation of the RCS.

The sleeve spans the degraded section of the tube and maintains the structural integrity of the steam generator tube under normal and accident conditions and limits Hor-prevents leakage if a-through hole in the tube wall should develop.

C.

JUSTIFICATION The purpose of sleeving is to repair a degraded tube-in a manner' that maintains the function and integrity of _the tube.

Surveillance Requirement 4.4.5.4 states that a steam generator

~

tube containing a defect is defective.

A defect'is defined as an imperfection of such severity that'it exceeds the plugging 111mit, equal to 40 percent of the_ nominal tube wall thickness.

.All-tubes exceeding the plugging limit are taken out of service by plugging.

Repairs by-means other-than plugging are not currently addressed in the technical specifications.

Tube _ sleeving provides an advantage to: plugging in-that the-tube will remain in' service with the structural integrity of the tube maintained _with only a small reduction in flow and heat: transfer' capability.

.The repaired--tube functions in a

manner similar to the original tube.

The sleeves will be installed in accordance with the processes provided by the vendors and described in the associated reports which address sleeve design, qualification, installation

methods, non-destructive _ examination and ALARA considerations.

The. B&W sleeving process is -described in NRC approved-topical-report BAW-2094P, Revision 1.

The Westinghouse -process is described in attached WCAP-13483, -Revision 1,

" Beaver Valley Units 1

and 2,

Westinghouse _ Series 51 Steam Generator Sleeving Report Laser Welded Sleeves" and is provided for-NRC review and approval.

I' ATTACHMENT B, continued Proposed Technical Specification Change No. 215/71 Page 3 D.

SAFETY ANALYSIS The tube sleeving procedure involves inserting a tube of smaller diameter (the sleeve) inside the tube to be repaired.

Sleeves span a

defective or degraded region of a

tube and are mechanically joined to the parent tube by a roll expansion and weld at the tubes' end

areas, thereby maintaining the steam generator tubing primary-to-secondary pressure boundary under nor.na l and accident conditions.

Thus, sleeving leaves the repaired tube functional.

This is in contrast to plugging, which removes the heat transfer surface of the plugged tube from service and reduces the reactor coolant system (RCS) flow available for reactor core cooling.

Therefore, a large number of sleeves can be installed without significantly affecting either RCS flow rate or plant operating efficiency (as compared to plugging),

and the service life of a steam generator that is experiencing degradation can be extended.

The principal accident associated with this proposed change is the steam generator tube rupture accident.

The environmental effects associated with a

steam generator tube rupture are discussed in BV-1 UFSAR Section 14.2.4 and BV-2 UFSAR Section 15.6.3, Steam Generator Tube Rupture.

For this occurrence, fission products contained in the RCS would be released to the secondary system.

Some of the radioactive noble gases and iodine would la released into the atmosphere through the condenser air removal system and steam line safety valves.

Use of the tube sleeving process will allow the repair of degraded steam generator tubes such that the function and integrity of the tube is maintained, therefore, the steam generator tube rupture accident is not affected by sleeving.

The tube sleeve is specifically designed to repair steam generator tubes which are exhibiting degradation in the tube sheet or at the tube support plate.

The material selected for the-sleeve is thermally treated Alloy 690 Inconel due to its enhanced corrosion resistance properties.

The structural analysis of the sleeve demonstrates that its design meets the ASME Boiler and Pressure Vessel Code (ASME Code)Section III criteria for the steam generator pressure, temperature, and flow design conditions and establishes the minimum reactor coolant pressure boundary wall thickness requirements.

Vibration testing and analysis were performed to demonstrate the adequacy of the sleeved tube for a

40 year design life objective.

Fatigue loadings used during the qualification testing of the sleeve joints - were established in accordance with ASME Code requirements to verify the integrity of the sleeve over the design life of the plant.

Fatigue testing consisted of axial load

cycling, vibration

cycling, pressure cycling and thermal cycling.

The sleeve is designed to accommodate all fatigue that the tubes may experience due to normal plant conditions and all anticipated transients.

ATTACHMENT B, continued Proposed Technical Specification Change No. 205/71 Page 4 The tubesheet and free span joints are mechanical seals produced by roll expanding the sleeve into the tube.

The structural integrity of the joints was proven by subjecting sleeve / tube specimens La a

series of tests representing steam generator service conditions.

These samples were fatigue tested, tensile

tested, thermal cycled, and leak tested to qualify the joints by experimental stress analysi in accordance with ASME Section III.

Corrosion testing has demonstrated the corrosion resistance of the sleeve and the sleeve / tube joints.

Defects which have been spanned by a

sleeve need -not be

)

considered for determination of inspection result categories in accordance with surveillance requirement 4.4.5.2, Steam Generator Tube Sample Selection and Inspection.

For the case in which the degraded tube has been spanned by a sleeve, further tube wall penetrations in the parent tube (from the bottom of the uppermosc rolled joint to the top of the lower-most rolled joint) are considered inconsequential since that portion of the tube no longer constitutes the reactor coolant pressure betndary..Any degradation in the parent tube in the area spanned by the sleeve does not affect the integrity of the pressure boundary and therefore, does not require the same degree of scrutiny as a wall penetration greater than 20 percent in a portion of the tube that does constitute the pressure boundary.

The-inspection requi.rement still applies to a-sleeved tube which has been subjected to a random full length examination and has been'found to have a wall penetration greater than 20 percent in either the portion of the tube which is'not spanned by the sleeve.or_in the sleeve itself.

B&W Kinetic Welded Sleeve Process NRC approved Topical Report BAW-2094P, Revision 1 describes in detail the analytical methods used for the design and qualification of the B&W tube sleeve.

The topical report also contains the results

?

the sleeve design verification which included analysis and-confirmatory testing to demonstrate the acceptability of the steam-generator sleeving technique._ The design and operating conditions-(including transient conditions and cycles) specified for the sleeve in the topical report bound the steam generator design conditions.

i The technical specifications;do not specifically limit the number of steam generator tubes which can be plugged while retaining-L occeptable primary flow. rates.

As discussed in the topical' L

ceport the thermal hydraulic effect of. installing up to 2,000 L

sleeves in each. steam generator has=been analyzed.

Two cases l

were considered, a 2. inch tube roll in the tubesheet with 29 inch sleeves and a

full tube roll. in the tubesheet with 11 inch sleeves.

The 2,000 sleeves consirted of 1,000 tubesheet sleeves l

and 1,000 tube support plate sleeves evenly distributed throughout the' steam generator._

Sleeving 2,000 tubes in each steam generator will reduce RCS-flow up to 0.63 percent and heat transfer performance up to_0.91 percent.

lI

=.

ATTACHMENT B, continued Proposed Technical Specification Change No. 205/71 Page 5 An analysis has been performed in accordance with Regulatory _

Guide' 1.121,. Bases for Plugging Degraded PWR Steam Generator-

Tubes, to establish the sleeve defect plugging criterion.

The plugging limit for the sleeve is calculated to be a 60 percent.

through wall defect.

An additional 20 percent of the wall thickness is deducted as a

combined allowance for postulated degradation due to corrosion and for eddy current testing inaccuracy.

Therefore, a defect plugging limit of 40' percent of the original sleeve wall is established.

A baseline eddy current (ECT) inspection of the installed sleeves is performed prior to operation.

ECT is-used to detect the presence of defects in the steam generator tubes and sleeves.

ECT detects the presence of defect caused variations in the effective electrical conductivity and/or magnetic permeability of the tubes.

The required defect sizes can be detected'and sized in the

sleeve, the parent tube behind the_ sleeve, and the tube above the sleeve.

The B&W-sleeving methodology described in topical report-BAW-2094P,- Revision 1 was accepted by'the_NRC for referencing in licensing applications on _ January

-4,-_1990.

The sleeve installation. procedures described in~BAW-2094P, Revision 1 will' be revised.

to include the kinetic sleeve

" tooling".and installation process parameter changes oescribed in NRC-approved BAW-2045PA, Revision.

1, January. - 1992,

" Recirculating Steam Generators Kinetic Qualification for 3/4 Inch OD Tubes.". These changes were incorporated to resolve field problems or to improve the sleeve instellation rate and will' not. alter the basic installed configuration of the sleeve as described in-BAW-2094P, Revision 1.

We have reviewed the methodology-described-in BAW.

2094P, Revision 1 and determined that.they are applicable to the Beaver Valley units and provide a safe and efficient alternative to plugging.

Westinchouse Laser Welded Sleeve Process section III of the ASME Code was used, during the-development of laser welded

sleeving, for the.

. minimum wall - thickness determination and bounding-stress and fatigue levels for the sleeve.

By showing that the sleeve design meets all facetsLof_'

the applicable subsections of Section-III of the Code, the sleeve design meets the design requirements.of the original tubing.

Regulatory Guide 1.121 is used to develop the plugging _ limit of-the sleeve should sleeve wall degradation occur.. WCAP-13483, Revision 1,.. describes in_ detail the-analytical methods used for the _ design _and qualification of.the Westinghouse tube sleeve.-

The WCAP also contains-the. results of the: sleeve design verification which included analysis and confirmatory testing to demonstrate the acceptabilityr of-the steam generator' sleeving i

technique.

Potentially_ degraded sleeves were shown.(by analysis) a to-retain burst strength ~in-excess'.of'three_ times thefnormal

~

operating _ pressure differential'for the_ Beaver Valley units.

The:-

requirements ~ of Regulatory Guide.l.83, " Inservice Inspection of-i

ATTACHMENT B, continued Proposed Technical Specification Change No. 205/71 Page 6 PWR Steam Generator Tubes" are implemented, and a baseline eddy current inspection of the installed sleeves is performed prior to operation.

An ultrasonic inspection of a sample of the free span weld joints is also performed prior to operation.

The ultrasonic inspection is used to verify that the minimum acceptable fusion zone thickness of the weld is achieved.

This minimum weld fusion zone thickness has been shown by analysis to satisfy the requirements of the ASME Code with regard to acceptable stress during operating and accident conditions.

The standard eddy current inspection procedure involves the use of a

bobbin eddy current probe, with two circumferential1y wound coils which are displaced axially along the probe body.. The coils are connected in the differential mode where the system responds only when there is a difference in the properties of the material surrounding the two coils.

The coils are excited by using an eddy current instrument that displays changes in the material surrounding the coils by measuring. the electrical impedance of the coils.

The outputs of the various frequencies are combined and recorded.

The combined data yields an output in which signals resulting from conditions that do not affect the integrity of the tube are reduced.

By reducing unwanted signals, improved inspectability of the tubing results.

Regions in the-steam generatur such as the tube support plate, tubesheet weld area and sleeve transition zones are examples of areas where multifrequency processing has proven valuable in.providing improved inspectability.

After sleeve installation, all. sleeved' tubes are subjected to an eddy current inspection which includes a

verification of correct sleeve installation for process

control, degradation inspection and establishing a baseline for all subsequent inspection comparisons..

Leakage testing under conditions considered to be more severe

-than 2xpected during all' operating plant conditions has shown that the laser welded sleeve does not introduce additional primary to secondary leakage during a postulated steam line break.

event.

Leakage testing has also shown that the seal wold of the lower joint in the tubesheet sleeve is.not required in order to preclude leakage during normal operation or accident conditions.

S]eeve/ tube leakage test specimens were subjected to both fatigue.

and thermal cycling tests prior to final leak rate evaluation testing.

The load level applied.during the-fatigue testing exceeded the maximum' axial load applied to.the sleeve during the most severe pressure loading condition.

Thermal cycling tests simulated a standard plant heatup/cooldown cycle.

~

Westinghouse has evaluated the laser welded sleeving-process in WCAP-13483, Revision 1.

A copy of the WCAP is provided in-4 Attachment D for NRC review and acceptance.

We have reviewed the methodology described in the WCAP and determined that.it provides a safe and efficient alternative to plugging.

ATTACHt4E!1T B, continued Proposed Technical Specification Change 110, 205/71 Page 7 Conclusion Based on the Regulatory Guide 1.121 guidelines for tube degradation

limits, appropriate plugging limits have been established.

Eddy current techniques are available to perform necessary sleeve and tube inspections for defect detection and to verify proper installation of the sleeve.

Available techniques are capable of providing adequate defect sensitivity in the required areas of the tube and sleeve pressure boundary.

Proprietary methods described in the vendor reports with supporting qualification data demonstr ate the inspectability of the sleeve and underlying tube.

In addition, we are committing to qualify the adequacy of any system that is used for periodic inservice inspection and to evaluate and, if practical, implement they are developed and qualified for better testing methods as use.

E.

[10 SIGt1IFICAtJT HAZARDS EVALUI.TIO!1 The no significant hazard considerations involved with the proposed

.mendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:

The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a

facility licensed under paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:

(1)

Involve a

significant increase in the probability or

~

consequences of an accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The following evaluation is provided for the no significant hazards consideration standards.

1.

Does the change involve a

significant increase in the probability or consequences of an accident previously evaluated?

Some steam generator tubes have been found to have a varying amount of wall degradation.

When the degradation is extensive, the normal practice of plugging defective tubes reduces the effectiveness of the steam generators and eventually will reduce the performance of the nuclear steam supply system.

An alternative to plugging tubes is installing a

sleeve as a new pressure boundary inside the original tube to bridge the degraded area, thus permitting

- ATTACHMENT B, continued Proposed Technical Specification Change No. 205/71 Page 8 the tubes to remain in service.

The integrity of the

)

repaired steam generator tubes will be equivalent to that or the original tube and will allow the tube to continue performing its heat transfer function.

The proposed change allows the installation of steam generator tube sleeves in accordance with the vendor methodologies provided by the B&W Kinetic welded sleeving

)

process described in NRC approved topical report BAW-2094P, Revision 1

and the Westinghouse laser welded sleeving process described in WCAP-13483, Revision 1, provided for NRC review and acceptance.

The sleeve installation 1

procedures described in BAW-2094P, Revision 1

will be revised to include the kinetic sleeve

" tooling" and installation process parameter changes described in ENRC approved BAW-2045PA, Revision 1,

January

.1992,

" Recirculating Steam Generators Kinetic Qualification for 3/4 Inch OD Tubes."

These changes were incorporated to resolve field problems or to improve the sleeve installation rate and will not alter the basic installed configuration of the sleeve as described in BAW-2094P, Revision 1.

We-have reviewed the methodology described in BAW 2094P, Revision 1 and determined that they are applicable to the Beaver Valley units and provide a

safe and efficient alternative to plugging.

We have reviewed the methodologies described in these vendor reports and determined that they provide a safe and efficient alternative to plugging.

Eddy current techniques are available to perform sleeve and tube inspections for defect detection and to. verify proper installation of the sleeve.

Available techniques are capable of providing adequate defect sensitivity in the required areas of the tube and sleeve pressure boundary.

Proprietary methods described

.i n.

the vendor reports with supporting qualification data demonstrate the-inspectability of the sleeve and underlying tube.

In addition, we are committing to qualify the adequacy of any system that is used for periodic inservice inspection and to evaluate and, if practical, implement testing methods as better methods are developed and qualified for use.

The structural integrity of the repaired tube is' restored to that of an undegraded tube and the tube and sleeves will be inspected periodically in accordance with the technical specification surveillance requirements.

Sleeving does not affect 'the UFSAR steam _ generator tube rupture accident,-

-therefore, the proposed change does not involve a

significant.-increase-in the probability or consequences of an accident previously evaluated.

2.

Does -the change create the possibility of a new or different kind of accident from many accident previously evaluated?-

Both the structural integrity and the. heat transfer capability of the steam generators will not be.significantly affected by the installation of sleeves..

In addition, the

-m

.+-

.ce aw -

+

g 3.h rr v

~

3

' TTACll!4E!1T B, continued j

}

oposed Technical Specification Change llo. 205/71 eage 9

.?/

sleeves are attached to the inside of the tubes and cannot ere '

interact with any of the other plant systems.

The sleeves have been analyzed and tested and the repair methods have been evaluated to ensure they satisfy the required design conditions.

Sleeving returns the degraded tube to a

serviceable condition and the sleeved tube functions in essentially the same manner as the original tube.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the change involve a significant reduction in a margin of safety?

The heat transfer capabilities of the steam generators will be improved by utilizing the sleeving process rather than the currently required plugging.

Installing sleeves slightly reduces the RCS flow and heat tral.cfor capabilities,

however, this reduction is significantly less than that of tubes that have been plugged.

Sleeving maintains the structural integrity of the steam generators to ensure the RCS pressure boundary is adequate for the expected design conditions, therefore, the proposed change does not involve a

significant reduction in the margin of safety.

F.

110 SIG!!IFICAliT llAZARDS C011SIDERATIOli DETERMI!1ATIOli Based on the considerations expressed above, it is concluded that the activities associated with this licence amendment request satisfies the no significant hazards consideration standards of 10 CFR 50.92(c)

and, accordingly, a

no significant hazards consideration finding is justified.

ATTACllMEllT C-1 Beaver Valley Power Station, Unit flo. 1 Proposed Technical Specification Change 11 o. 205 TYPED ItEPLACEME!1T PAGES Typed Pages:

3/4 4-8 3/4 4-9 3/4 4-10 3/4 4-10a 3/4 4-10b 3/4 4-10c shifted backward 3/4 4-10d shifted backward 3/4 4-100 shifted backward B 3/4 4-1 shifted forward B 3/4 4-la shifted forward B 3/4 4-2 shifted forward B 3/4 4-2a i

[

'k 1,

1

1 DPR-66 REACTOR COOLANT SYSTTE JL4&4. 5 STEA!4 GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2,

3 and 4.

Ap1LQU:

With one or more steam generators inoperable, rostore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 200'F.

i SURVEILLANCE REQUIREMENTS e

4.4.5.1 E. team Generator Samnlo Selection _and Insnection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and Inspectiom - The steam generator tube minimum samplo

sizo, inspection result classification, and the corresponding action required shall be as specified in Tablo 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 veritied acceptable por the acceptanco critoria of Specification 4.4.5.4.

Steam generator tubes shall be examined in accordance with Articlo-8 of Section V

(" Eddy current Examination of Tubular Products") and Appendix IV to Section XI

(" Eddy Current Examination of Nonferromagnetic Steam Generator Heat Exchanger Tubing")

of

'the applicable year and addenda of the ASME Boilor and Pressure Vossel Codo required by

10CFR50, Section 50.55a(g).

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 slooving are not considered an area requiring reinspection.

The tubos selected for each inservice inspection shall include at least 3% of the total number of tubos in all steam generators; the tubos selected for those inspections shall be selected on a random basis except a.

Where experience in similar plants with similar water chemistry indicatos critical areas to be inspected, then at

-loast 50%

of-the tubes inspected shall be from those critical areas, b.

The first samplo of tubes selected for each inservice l

inspection (subsequent to the preservico inspection) of_each steam generator shall includo:

1.

All nonplugged tubes that previously had detectable l

wall penetrations greator than 20%, and l

BEAVER VALLEY - UNIT 1 3/4 4-8 Amendment No.

PROPOSED i

I i

DPR-66 REACTQR COOLANT SYSTEfj

{

l SURVEILLANCE REQUIREMENTS (Continued) 2.

Tubos in those areas whero experienco has indicated potential problems, and 3.

At least 3%

of the total number of slooved tubos in all throo steam generators.

A sample sizo loss than 3%

is acceptable provided all the slooved tubos in the steam generator (s) oxamined during the refueling outago are inspected.

Those inspections will include both the tubo and the sloovo, and r

4.

A tube inspection pursuant to Specification 4.4.5.4.a.8.

If any selected tubo does not permit tho passago of the oddy current probe for a tube or alcovo inspection, this shall be recorded and an adjacent tube shall be solocted and subjected to-a tubo inspection.

c.

The tubos selected as the second and third samples (if required by -Tablo 4.4-2) during each inservico inspection may be subjected to a partial tube inspection providod:

1.

The tubos selected for those samplos include the tubos from thoso areas of the tube sheet array where tubos with imperfections woro previously found,-and 2.

The inspectio.1s include those portions of the tubes where imperfootions were previously found.

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

Catecorv Ingoection Results-C-1 Loss than 5%

of the total tubos inspected are. degraded tubes and none of the inspected --tubes are defective.

C-2 One-or.more. tubes,.but.not moro-than 1%

of the

. total tubes inspected

. are defectivo, or-betwoon-5% :and 10%: of the total tubes inspected.

are

.dograded tubes.

BEAVER VALLEY - UNIT 1 3/4 4 Amendment No.-

PROPOSED

t DPR-66 i

REACTOR COOLANT SYSTEM t

SURVE1LLANCE REQUIREMENTS (Continued)

-f

=

C-3 Moro than 10%

of the total tubos inspected are degraded tubos or-more than 1% _of the inspected tubes are defectivo.

Note:

In all inspoctions, previously_ degradod 1

tubos or sloovos must exhibit significant (greator than 10%) further wall penetrations to be included in the above percentago calculations.

4.4.5.3 Insnection Frecuencies The above required inservice inspections of steam generator tubos shall be performed at-the following frequencios:

a.

The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.

Subsequent inservice inspections shall be performed at intervals of not less-than_12_nor.

more than 24 calendar months attor-the previous inspection.

If two consecutive inspections following.

service under AVT conditions, not_ including.the preservice

' 3 inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections a

demonstrato that previously observed degradation has not continued and no additional degradation has occurrod, the inspection interval may be--extended to a maximum of once per 40 months.

b.

.If the inservice inspection of a steam generator conducted In accordanco-with Table 4.4-2 requires a third sample inspection whose results

. fall in-Category C-3,

.the inspection frequency shall be reduced to at least once per 20 months.

The reduction in inspection frequency shall-apply until a

subsequent inspection demonstratos that a third sample inspection is not required.

c.

Additional, unscheduled inservico _ 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:

l.

1.

Primary-to-secondary ; tube leaks (not including leaks I

originating from-tube-to-tube sheet welds) _in' excess.

of the limits of ' Specification 3.4.6.2,_

2.

A seismic occurrence greater than the operating Basis Earthquake,

' BEAVER l VALLEY - UNIT.1-3/4 4-10 Amendment No.

PROPOSED p

6

DPR-66 REAcTon COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) s 3.

A loss-of-coolant accident requiring actuation of the engineered safeguards, or

+

4.

A main steam line or feedwater line break.

4.4.5.4 Acceptance Criteria a.

As used in this Specificatient 1.

Jmoerfection means an exception to the dimensions, finish or contour of a

tube or sleeve from that I

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 or sleeve.

l 3.

Decraded Tub.g means a

tube or sleeve -containing I

imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.

4.

1 Dearadation means the percentage of the tube or sleeve wall thickness affected or removed by degradation.

5.

Defect means an i.mperfection of such severity that it exceeds the plugging or repair limit.

A tube l

containing a defect is defective.

Any tube which does not permit the passage of the eddy-current inspection probe shall.be deemed a defective tube.

6.

Plucaina-or Repair Limit means the imperfection depth I

at or beyond which the tube-shall be removed from service by plugging or repaired by sleeving in the affected area because it may become unserviceable prior to the next inspection..The plugging or repair'

. limit imperfection depths are specified in percentage of nominal wall thickness as follows:

a.

Original tube wall 40%'

b.

Babcock & Wilcox kinetic welded sleeve wall 40%-

c.

Westinghouse laser welded sleeve wall

-31%

BEAVER VALLEY - UNIT 1 3/4 4-10a Amendment No.

PROPOSED

DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

==

7.

Unserviceable describen the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a

loss-of-coolant

accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

8.

Tube Inspection 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 to the cold leg.

9.

Tube Repair refers to sleeving which is used to maintain a

tube in-service or return a

tube to service.

This includes the removal of plugs that were installed as a corrective or preventive measure._ The following sleeve designs have been found acceptable:

a)

Babcock Wilcox kinetic welded

sleeves, BAW-2094P, Revision 1

including kinetic sleeve

" tooling" and installation process parameter.

changes.

b)

Westinghouse laser welded

sleeves, WCAP-13483, Revision 1.

b.

The steam generator shall be determined OPERABLE after completing the corresponding actions -(plug or' repair all tubes exceeding the plugging or repair limit) required by.

Table 4.4-2.

4.4.5.5 Beoorts 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 l

reported to the Commission in a Special Report pursuant to Specification 6.9.2.

b.

The complete results of the steam generator tube and sleeve-l inservice inspection shall be submitted to the Commission in a

Special Report pursuant to Specification'6.9.2 within 12 months following the completion of the inspection. ~This Special Report shall include:

1.

Number and extent of tobes and sleeves inspected.

- l-2.

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

3.

Identification of tubes plugged or repaired.

_l BEAVER VALLEY - UNIT 1 3/4 4-10b.

Amendment No.

PROPOSED

_m

__ - = -. _ _

DPR-66 BEFTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c.

Results of steam generator tube insp'ections which fall into-Category C-3 shall be reported to the Commission pursuant to _ Specification 6.6 prior to resumption of plant.

operation.

The written report shall_ provide a description of investigations conducted' to determine the cause-_of-the tube degradation and corrective measures taken to prevent recurrence.

i BEAVER VALLEY - UNIT _1 3l 4-10c

'A'mendment No.

l

-PROPOSED 2

.. -.... ~. -. -.. -....

a -

TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection ~

No Yes No. of1 Steam Generators per Unit Two Three Four Two Three Four First. Inservice Inspection All One Two Two l

l 2

3 Second & Subsequent Inservice Inspections One One One One Table Notation:

1.

The~ inservice., inspection may. be' limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators'in the plant) if the results of the first or previous-inspections indicate that all steam generators are performing in a 'like manner.

Note-that under some circumstances, the operating conditions Ein 'one or more steam. generators may be found to be more severe than those in other steamL generators.

Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.

2.

The-other.. steam generatorinot inspected during the'first inservice inspection shall be inspected.

The third and subsequent inspections should follow the instructions described in 1 above.

3.

Each off'the. 'other two steam generators. not inspected during the.first' inservice inspections shall

.be' inspected during the second and third inspections.

The fourth and subsequent inspections shall~followithe instructions described in 1 above.

-BEAVER VALLEY -JUNIT 1 3/4L4-10d l

PROPOSED

. - =

TABLE 4.4-2 STEAM GBJERATOR TUBE INSPECPION

.o:

y'O 1ST SAMPLE INSPECTION 2ND SAMPE INSPECTION 3RD SAMPLE INSPECTION m

Sample Size Result Action Required Result Action Required Result Action Ikquired A mininum of C-1 None N/A N/A N/A N/A S Tubes'per

..G.

C Plug or repair C-1 None N/A N/A l

defective tubes and inspect additional C-2 Plug or repair defective C-1 lbne j

2S tubes in this S.G.

tubes and inspect additional 4S tubes in C-2 Plug or repair

.l this S.G.

defective tubes C-3 Perfom action for C-3 result of first sample C-3 Perform action for C-3 N/A N/A result of first sample C-3 Inspect all tubes in All other None N/A N/A this S.G., plug or S.G.s are repair defective C-1 tubes and inspect 2S tubes in each other S.G.

Same S.G.s Perform action for N/A N/A C-2 but no C-2 result of second Notification to NRC additional sample punsuant to S.G.s are Specification 6.6 C-3 Additional

' Inspect all tubes in S.G.-is each S.G. and plug or N/A N/A C-3

. repair defective tubes.

Ibtification to NRC pursuant'to Specifi-

' cation 6.6.

N S== 3 % Where N is the number of steam generators in the unit, and_ n is the number of steam generators inspected U

during an inspection.

bE/ NIB VAllr/ -. UNIT 1 3/4 4-10e Amendment No.

I PROPOSFD.

DPR-66 3/4.4 REACTOR COOLANT SYSTFE BASES 3/4.4..

REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design DNBR limit during all normal operations and anticipated transients.

In Modes 1 and 2, with one reactor coolant loop not in operation, THERMAL POWER _is restricted 'to less than or equal to 31 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset.

Either action ensures that the DNBR will be maintained above the design-DNBR limit.

A loss of flow in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a-loss of flow in one loop will cause a reactor trip if operating above P-8 (31 percent of RATED THERMAL POWER).

In MODE 3,

a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, due-to tr.e initial conditions assumed in the analysis for the control rod bank withdrawal from a subcritical condition, two operating coolant loops are required to meet the DNB design: basis for this Condition II event.

In MODES 4

and 5,

a single reactor coolant loop or RHR subsystem provides sufficient heat removal capability for removing. decay-heat; but single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

l The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to -ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction -will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more-L.

RCS cold legs less-than or equal to 275'F a.~e provided to prevent l

RCS pressure transients, caused by energy additions _from_ the secondary

system, which could exceed the limits of Appendix G to 10 CPR Part 50.

The RCS-will be -protected against! overpressure transients and will not exceed the limits of Appendix-G by either (1) restricting the water level in the pressurizer and thereby l

providing a

volume for the primary coolant to expand into or (2) by restricting starting of the -RCPs to when.the secondary-water temperature-of each steam generator is less_than 25'F above each of the RCS cold leg temperatures.

Power is removed-from the isolated loop stop valves (hot leg and-cold leg) to ensure that no reactivity addition to the core can l

BEAVER VALLEY - UNIT 1 B 3/4 4-1 Amendment No.

PROPOSED

-g,-<

E -- e,- --

-y

-m---,

y--

y,v-

.,--e

--*w

-r--w--e i--ne

--v-,

w w-w,---~w

-o--

-r-w--

~DPR 3/4.4 REACTOR COOLANT SYSTEM.

BASES 3/4.4.1 -REACTOR COOLANT LOOPS, (continued) occur while the loop is isolated due to inadvertent opening of the isolated loop stop valves.

Isolated loop startup is limited to' Modes 5

and 6 in accordance with the NRC SER on N-1 loop operation.

Verification of the isolated loop boron concentration -prior to opening the isolated loop stop valves provides a reassurance of the-adequacy of the shutdcWn margin in the remainder of the system..

Restoration of power to the hot leg stop-valve allows opening this valve to complete the recirculation flowpath in conjunction with the relief line bypassing the cold leg stop valve.and ensures adequate nixing in the isolated loop.

This enables the temperature and boron concentration of the isolated loop to be brought to equilibrium with the remainder of the system.

Limiting the temperature differential between the isolated loop and the remainder-of the system prior to opening-the cold leg stop valve prevents any significant reactivity effects due to cool water addition to the core.

Startup of an idle loop will inject cool water from the loop into the core.

The reactivity transient resulting from this cool water-injection is minimized by delaying isolated loop'startup_until its temperature is within 20aF of the operating loops.

Making the reactor suberitical prior to-loop startup_ prevents any power _ spike which could result from this cool water induced reactivity transient.

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves' operate to prevent _the RCS from being pressurized above its Safety Limit of 2735 psig.:

Each~ safety valve is. designed to -relieve 345,000 :lbs. per hour of - saturated -

steam at the. valve-set point.

The relief capacity of a' single i

safety valve is adequate _to relieve any. overpressure condition which j

could occur during shutdown.

In the event that no safety valves are l

OPERABLE, an operating RHR

loop, connected -to the RCS, provides overpressure reli.ef capability and.

Will prevent RCS i

overpressurization.

During operation, all pressurizer code safetyLvalves mustibe-OPERABLE to prevent the RCS from being pressurized above its safety limit oof 2735 _psig. _ The combined relief capacity of all_of these valves.is greater than the maximum surge rate resulting from a complete loss-of-load-assuming-.no reactor. trip _until the-first-

-Reactor. Protective-System: trip set point-is reached (i.e., no_ credit-is taken.for a

direct. reactor trip on the-loss of load) and also-assuming no-operation-of the power operated relief ivalves 'c r steam-

. dump valves.

BEAVER VALLEY - UNIT 1-B 3/4 4-la.

Amendment No.

PROPOSED 1

~

~. -

DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES (Continued)

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.4 PRESSURIZER The requirement that (150)kw of pressurizer heaters ~and their associated controls be capable of being supplied eJectrical power from an emergency bus provides assurance that these heaters can be energized during a

loss of offsite power condition to maintain natural circulation at HOT STANDBY.

3/4.4.5 STEAM GENERATORS One OPERABLE steam generator in a non-isolated reactor coolant-loop provides sufficient heat removal capability to remove decay heat after. a reactor shutdown.

The requirement for two OPERABLE steam generators, combined with other requirements of the Limiting Conditions for Operation ensures adequate decay heat removal capabilities for RCS temperatures. greater than 350*F if one steam generator becomes inoperable due to single failure considerations.

Below 350*F, decay heat is removed by the RHR system.

The ' Surveillance Requirements for inspection of the steam generator tubes ensure -that the structural integrity of'this portion of the RCS-will be maintained.

The program for inservice inspection of steam generator tubes is based'on a modification of Regulatory Guide 1.83, Revision 1.

Inservice. inspection of steam generator-tubing is essential =in order to maintain surveillance of the conditions of the-tubes in the event that there is evidence of mechanical damage cut 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.

The plant is expected: to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes.

If-the secondary coolant chemistry is not maintained within these, parameter

limits, localized corrosion may likely result in-stress corrosion cracking.

The extent of cracking during plant

BEAVER VALLEY - UNIT 1 B 3/4 4-2 Amendment No.

PROPOSED

DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continuedl 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 = 500 gallons per day per steam generator).

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

Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of eteam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant.

However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging or repair l

will be required of all tubes with imperfections exceeding the plugging or repair limit.

Degraded steam generator tubes may be repaired by the installation of sleeves which span the degraded tube section.

A steam generator tube with a sleeve installed meets the structural requirements of tubes which are not degraded, therefore, the sleeve is considered a

part of the tube.

The surveillance requirements identify those sleeving methodologies approved-for use.

If an installed sleeve is found to have through wall penetration greater than or equal to the plugging limit, the tube must be plugged.

The plugging limit for the sleeve is derived from R.G.

1.121 analysis which utilizes a 20% allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional degradation growth.

Steam. generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20%

of the original tube wall-thickness.

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'6.6 prior to resumption of plant operation.

Such cases will'be considered by the Commission on a

case-b*j-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.

BEAVER VALLEY - UNIT 1

-B 3/4 4-2c Amendment No.

PROPOSED l

ATTACl!MF11T C-2 I

Beaver Valley Power Station, Unit 11o. 2 I

a, Proposed Technical Specification Change llo. 71 l

TYPED REPLACEMEliT PAGES r.

l Typed Pages:

3/4 4-11 3/4 4-12 3/4 4-13 i

3/4 4-14 i.

3/4 4-14a added 3/4 4-14b added 3/4 4/16 t

i B 3/4 4-3 B 3/4 4-3a added t

I i

l L

)

I

NPF-73 REACTOR COOLANT' SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2,

3 and 4.

ACTION:

With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing

_T above avg 200*F.

SURVEILLANCE REQUIREMENTS 4.4.5.3 Steam Generator Samole Selection and Inspection - Each._ steam generator shall be determined OPERABLE during shutdown by' selecting and inspecting at. least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Samole Selection and Inspection - 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 specifiedfin Specification 4.4.5.3 and the inspected tubes-shall be verified-acceptable per the-acceptance. criteria of Specification 4.4.5.4.

Steam generator tubes shall be examined in accordance with-Article a of Section V

(" Eddy -Current Excmination_of1 Tubular Products") and-

' Appendix IV.

to Section

-)C[

(" Eddy _ Current Examination of~

Nonferromagnetic Steam Generator Heat Exchanger Tubing")

of Lthe applicable year and addenda of the ASME Boiler and Pressure Vessel-Code required by

10CFR50, Section -50.55a(g).

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 sleeving are not considered an-area requiring reinspection. ~ 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.

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 fromi these critical areas.

b.

b.-

The first sample of tubes. selected for each--inservice inspection' (subsequent to the'preservice inspection) of each steam-generator:shall include:

1 BEAVER VALLEY UNIT 2 3/4 4-11 Amendment No.

.)

PROPOSED q

t M

M c

e v+-m'rW t'

Y W-

.%'-4 m

NPF-73 REACTOR COOLANT SYSTEM

~ SURVEILLANCE REQUIREMENTS (Continued) 1.

All nonplugged tubes that previously had detectable wall penetrations greater than 20%, and j

2.

Tubes in those areas where experience has indicated potential problems, and 3.

At least 3%

of the total number of sleeved tubes,in all three steam generators.

A sample size less than 3%

is acceptable provided all the sleeved tubes in the steam generator (s) examined during the refueling-outage are inspected.

These inspections will include both the tube and the sleeve, and 4.

A tube inspection pursuant to Specification-4.4.5.4.a.8.

If any selected tube'does not permit the passage of the eddy current probe for a tube or sleeve inspection, this shall be recorded and an-adjacent tube shall be selected and subjected to a

tube inspection.

c.

The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected.to a partial 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, and 2.

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

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

Catecorv Inspection Results C-1 Less than 5%

of the total tubes inspected are degraded tubes and-none of the inspected tubestare defective.

C-2 One or more' tubes, -but not more than 1%

of the total tubes L

inspected are defective, _ o r.

l-between 5%

and 10%

of the total tubes inspected are degraded-l tubes.

BEAVER VALLEY - UNIT 2 3/4 4-12 Amendment 1No.

PROPOSED

-NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

-=,

C-3 More than 10%

of the total tubes inspected are degraded -tubes or more than 1%

of the -inspected tubes ere defective.

Note:

In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10%) further wall penetrations to be included -in the above percentage calculations.

The above ' required inservice 4.4.5.3 Insocction Freauencing inspections of steam generator tubes shall be performed at the following frequencies:

a.

The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.

Subsequent inservice inspectionsL shall be performed at intervals-of-not less than 12Vnor more than 24 calendar months after the previous inspection.

If-two consecutive. inspections. following service under All Volatile Treatment (AVT) conditions,-not including the preservice' _ inspection, result in all inspection results _ falling into the C-1 category or-if two consecutive.

inspections demonstrate that' '1previously.

-observed degradation.has -not continued :and no 4 additional-degradation has

occurred, the inspection 11nterval may be extended-to a maximum of once perf40 months.

b.

If the inservice inspection of;a steam generator _ conducted in accordance with Table 14.4-2 requires a_ third sample-inspection whose results fall in-' Category _-C-3,

.the inspection frequency shall be reduced toLat least once per-20 months.-

'The reduction in inspection frequencynshall-apply until a

subsequent linspection demonstrates _that a.

third sample inspection is not' required.

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 theffollowing conditions:

l _'

1.

Primary-to-secondaryl tube leaks: (not including:leaksh E

originating-from tube-to-tube. sheet welds)-in excess-

of-the limits of Specification'3.4.6.2, 2.

A seismic occurrence greater than the Operating._ Basis Earthquake, BEAVER VALLEY - UNIT 2-3/4 4-13)

-Amendment No.

PROPOSED-s.

NPF REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 3.,

A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4.

A main steam line or feedwater line break.

4.4.5.4 Acceptance Criteria As used in this Specification:

a.

1.

Imperfection means-an exception to the_ dimensions, finish or contour of a

tube.or sleeve from that I

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.

Q_earadation hieans a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of-a tube or sleeve.

l 3.

Dearaded Tube means a

tube or. sleeve containing imperfections greater-than or. equal to 20% of the nominal wall thickness caused by degradation.

4.

Decradation means the percentage of-the tube or sleeve wall thickness affected

-or.

removed by degradation.

5.

Defect means an_ imperfection of such severity that it I

exceeds the plugging or repair limit.-.

A-tube containing a defect is defective.- Any tube which does not permit the passage of-the eddy-current inspection probe shall be deemed a defective tube.

6..

Pluccinq or Repair Limit means.the imperfection _ depth l

at or beyond which the-tube' shall-be removed from service by. plugging or repaired by' sleeving in the affected-area because 'it may become. unserviceable prior to.the next-inspection.

The plugging.or repair.

- limi t' imperfection depths are specified in percentage of nomina 11 wall thickness as follows:

a.

-Original. tube. wall 40%

b.

Babcock & Wilcox kinetic. welded. sleeve. wall 40%

c.

' Westinghouse laser welded sleeve wall-f31%.

BEAVER VALLEY - UNIT 2 3/4 4-14 Amendment:No.

PROPOSED

NPP-73 RE6CTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l 7.

Unserviceable -describes the condition of a tube if it leaks-or contains a' defect large_enough to affect its structural integrity in the event of an Operating Basis Earthquake, a

loss-of-coolant

accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

8.

Tube Inspection 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 tc the cold leg.

9.

Tube Ronalt refers to sleeving _ which is used-to maintain a

tube in-service or return a

tube to-service.

This includes the removal of plugs that were installed as a corrective or preventive measure.

The following sleeve designs have been found acceptable:

a)

Babcock Wilcox kinetic welded.

sleeves, BAW-2094P, Revision 1

including kinetic sleeve

" tooling" and installation process-parameter changes.

b)

Westinghouse laser welded

sleeves, WCAP-13483, Revision 1.

i b.

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

4.4.5.5 Reports a.

With'in 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 I

reported' to the Commission in a Special Report' pursuant to-Specification 6.9.2.

b.

The-complete results of the steam generator tube and sleeve-

_l~' D{

inservice inspection shall be submitted to the Commission in a

Special Report pursuant to Spec'ification 6.9.2 within 12 months following the completion--of the inspection.- This Special Report shall include:

1.

Number and' extent of tubes and sleeves inspected.

2.

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

-3.

Identification of tubes plugged or-repaired.

l-BEAVER _ VALLEY -. UNIT 2 3/4 4-14a Amendment No.

PROPOSED-

= -

NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c.

Results of steam-generator tube inspections _which. fall intoL Category C-3 shall be reported to the Commission pursuant to specification 6.6 prior to resumption of plant operation.

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

recurrence.

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BEAVER VALLEY -? UNIT-2 3/4 4-14b Amendmcnt'No.

_ PROPOSED.

g-u x.....

,,_; a.

-..2

~ -

TABLE 4.4-2 SPEAM GFNERA'IOR 'IUBE INSPTTION

o IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPTTION 7

Sa gle Size Result Action Rcquired Result Action Rrquired Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per S.G.

C-2 Plug or repair C-1 None N/A N/A l

defective tubes and inspect additional C-2 Plug or repair defective C-1 None

{

2S tubes in this S.G.

tubes and inspect additional 4S tubes in C-2 Plto or repair l

this S.G.

defective tubes C-3 Perfom action for C-3 result of first sa::ple C-3 Perfom action for C-3 N/A N/A f

result of first sarple C-3 Inspect all tubes in All other None N/A N/A this S.G., plug or S.G.s are repair defective C-1 tubes and inspect 2S tubes in each other S.G.

Some S.G.s Perform action for N/A N/A C-2 but no C-2 result of second Notification to NRC additional sample pursuant to 50.72 S.G.s are (b)(2) of 10 CFR C-3 Part 50 Additional Inspect all tubes in S.G. is each S.G. arrl plug or N/A N/A C-3 repair defective tubes.

Notification to IEC pursuant to 50.72 (b)

(2) of 10 CFR Part 50 S=2%

hhere n is the number of steam generators inspected during an inspection.

during an irrpection.

BEAVER VAILEY - UNIT 2 3/4 4-16 Amendment No.

PROPOSED

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM-GENERATORS (Continued) decay heat removal capabilities for RCS temperatures greater than 350*F if one steam generator becomes inoperable due to single failure considerations.

Below 350*F, decay heat is removed b/ the RHR system.

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is 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, 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.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those. parameter limits found to result in negligible corrosion of the. steam generator tubes.

If the secondary' coolant chemistry is not maintained witain these parameter

limits, localized corrosion may likely result in stress corrosion cracking.

The e.Mtent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coola'nt system (primary-to-secondary leakage = 500 gallons per day.per steam generator).

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

Operating' plants _have demonstrated that primary-to-secondary leakage of 500 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 and plugged.

Wastage-type defects are unlikely with the all volatile treatment ~(ANT) of secondary coolant.

However, even if a defect of similar type.should develop -in -service, it will be found during scheduled' inservice steam generator tubo examinations.

Plugging or repair will be required of all tubes with imperfections exceeding the plugging or repair limit.

Degraded steam generator tubes'may be BEAVER VALLEY.- UNIT 2 B 3/4 4-3 Amendment No.

PROPOSED a

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) repaired by the installation of sleeves which span thefdegraded tube section.

A steam generator tube with a-sleeve installed meets the structural requirements of tubes which are not degraded, therefore,-

the sleeve is considered a

part of the tube.

The surveillance requirements identify those sleeving methodologies approved for use.

If an installed sleeve is found to have-through wall penetration greater than or equal to the plugging limit, the tube.

must be plugged.

The plugging limit for the sleeve is~ derived from R.G.

1.121 analysis which utilizes a 20% allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional degradation growth.

Steam generator tube inspections of p

operating plants have demonstrated the capability to r aliably detect degradation that has penetrated 20%

of the original-tube wall thickness.

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 6.6 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.

BEAVER VALLEY - UNIT 2 B 3/4 4-3a Amendment No.

PROPOSED

ATTACHMENT D

)

Beaver Valley Power-Station,-Unit Nos. 1 and 2 Proposed ~ Technical Specification Change-Nos. - 205 and 71 WESTINGHOUSE: SLEEVING REPORT i

WCAP-13483, Revision 1 Beaver Valley Units 1 and 2, Westinghouse Series 51 Steam Generator Sleeving Report 1

Laser Welded Sleeves d

)

I i

ATTACHMENT E Beaver Valley Power Station, Unit.Nos. -1 and -2 Proposed Technical Specification Change Nos. 205 and;71 H

I WESTINGHOUSE APPLICATION FOR WITHHOLDING The following are included:

Westinghouse Letter,-

Application For Withholding.

Proprietary Information From Public Disclosure Accompanying Affidavit Proprietary Information Notice Copyright Notice i

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