ML20140A620

Proposed Tech Specs Deleting References to Steam Generator Tube Sleeving & Repair That Will Not Be Used for Westinghouse Model D5 Steam Generators in Use at Catawba Unit 2

ML20140A620
Person / Time
Site:	Catawba
Issue date:	05/27/1997
From:	DUKE POWER CO.
To:
Shared Package
ML20140A617	List: ML20140A614 ML20140A620
References
NUDOCS 9706040319
Download: ML20140A620 (22)

Text

._

No bhF i

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS 3

7; I

y N

LIMITING CONDITION FOR OPERATION S

3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

t 3

'E ACTION:

'M With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T above 200*F.

og A

SURVEILLANCEREOU,J,3EdENTS E

y iG#

4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of j

Q the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall 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 SamDle Selection and InsDection - 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 fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.

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

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 from these critical areas; b.

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

9706040319 970527 PDR ADOCK 05000414 P

PDR l

CATAWBA - UNIT 2 3/4 4-12 Amendment No.

142

_ _ _ _ _ _ = - _ _ _ _ _ _ -.

y,

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 3 1)

All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),

2)

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

A tube inspection (aursuant to Specification 4.4.5.4a.8) shall be performed on eac1 selected tube.

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

l

(;.

het ;1s@

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.

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

Cateaory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

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

and 10% of the total tubes inspected are degraded tubes.

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

Note:

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

CATAWBA - UNIT 2 3/4 4-13 Amendment No.

142

' )*

l' No CbW R[fM OR COOLANT SYSTEM g

, SURVEILLANCE REQUIREMENTS (Continued)

}l y

h 4.4.5.3 Inspection Freauencies - The above required inservice inspections of l

steam generator tubes shall he performed at the following frequencies:

c.

a.

The first inservice inspection shall be performed after 6 Effective

!I Full Power Months but within 24 calendar months of initial criticality.

Subsequent inservice inspections shall be performed at 9

intervals of not less than 12 nor more than 24 calendar months after

{j the previous inspection.

If two consecutive inspections, not including the preservice inspection, result in all inspection results e

J falling i,,

• he C-1 category or if two consecutive inspections d

demonstra 3at previously observea degradation has not continued l

and no additional degradation has occurred, the inspection interval 1

may be extended to a maximum of once per 40 months; j

b.

If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and c.

Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection

-)

specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1)

Reactor-to-secondary tubes leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or 2)

A seismic occurrence greater than the Operating Basis Earthquake, or 3)

A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or 4)

A main steam line or feedwater line break.

I J

CATAWBA - UNIT 2 3/4 4-14 Amendment No.

142

Il l

\\l

['-

ILEACTOR COOLANT SYSTEM l'

4 SURVEILLANCE REQUIREMENTS (Continued) l t

i 4.4.5.4 Acceptance Criteria a.

As used in this specification:

1)

Imperfection means an_ exception to the dimensions, finish or I

t contour of a tubegr :!::v_)from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tubeGr

!: vowall thickness, if I

detectable, may be considered as imperfections; l

4.

2)

Dearadation means a service-induced cracking,

wastage, wear or l

general corrosion occurring on either inside or' outside of a u

/

tube P eauel; c

f 3)

DecradedTubemeansatubeff:1:M coniiaining imperfections j

D greater than or equal to 20% of the nominal tube (~r

:::ve;wali thickness caused by degradation; i

f 4)

Dearadation means the percentage of the tube (cr :

! :ve} wall

?

thickness affected or removed by degradation; 4

!!y 5)

Defect means an imperfection of such severity that it exceeds limit. A tube (Er :! erg containing a defect is there(~ie defect t

e hoy,n P

i.

1 6) ne;;i (, Limit means the imperfection depth at or beyond which the l

tube shall be removed from service by plugging er repaired by

?

lec;ing.

It ch; ::::: the imperfc:ti n depth at Or beyond 1

9ich : :10 ved tub: chall b; ilu;;;d. The re " r limit is equal to 40% of the nominal tu je( [ :l;;djwalf ic< ness.

l L W43 jl If a tub; i 01;;ved due t; d;;rr.dati;n in the f* di tan:0, th n

ny deft: : f end " the tube bel:ce th ciceve ill net n:::: i j

tete plugging.

The Och:::k f. Wilecx prc:::: d;;;ribcd in Topical R; pert CAW-200(r)-A, R;v. I will b; u d for :lceving.

7)

Unserviceable describes the condition of a tube if it leaks or L

contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above; 8)

Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg Eide) completely around the U-bend to the top support of the cold leg; CATAWBA - UNIT 2 3/4 4-15 Amendment No.

142 A

~

(

,.3' y

k 4-e i

j REACTOR COOLANT SYSTEM l) f-SURVEILLANCE REOUIREMENTS (Continued)

{'

(;

9)

Preservice Inspection means an inspection of the full length of i:

each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of J

the tubing.

This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to J,

be used during subsequent inservice inspections.

10) Tube Roll ExfIansion is hat porti of a tube w ch has been

(

increas in diamet by a roll g process su that no cre ice 1:

exis etween th outside di eter of the be and the tu sheet.

g i

1 F* Distanc is the mini m lengtn of e roll expand portion l

of the t e which can t contain an defects in or r to ensure i

the t e does not p out of the ubesheet.

Th

• distance s 1.6 inches and is easured from the bottom of e roll ansion transi on or the to of the tubesh t if the b tom of the roll ex e)nsion is abo the top of t tubesheet.

2 d-Included in is distance a safety fac r of 3 plus 0.5 inch eddy rrent vertic measurement certainty.

12) bltern e tube pluaa' a criteria do not requir the tube to be remov d from servi or repaired w n the tube egradation exc ds the repai limit so long s the degra ation is in that p tion of the be from F* to e bottom o the tubesheet.

his definiti does not appl to tubes w h degradation (i.e.,

4 i

ations cracking) in he F* dista e.

h Et URO plug'q b.

The steam generator shcil be detemined OPERABLE after comple ing the corresponding actions (Flug er rep & all tubes exceeding the,cpair

>i limit and all tubes containing through-wall cracks) required by Table i

4.4-2.

t J

0l 4.4.5.5 Reports a.

Within 15 days following the completion of each inservice inssection of steam generator tubes, the number of tubes rO ;ird in eac1 steam generator shall be reported to De Commission in a Special Report pursuant to Specification 6.9.2; l

b.

The complete results of the steam generator tube 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 Feport shall include:

CATAWBA - UNIT 2 3/4 4-16 Amendment No.

142 i

i l

f' '

REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) l 1)

Number and extent of tubes inspected, 2)

Location and percent of wall-thickness penetration for each indication of an imperfection, and Pl%3e Identification of tubes repaire f.l 3)

Results of steam generator tube inspections, which fall into Category j

c.

C-3, shall be reported in a Special Report to the Commission pursuant i

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

• .j investigations conducted to determine cause of the tube degradation 1

d and corrective measures taken to prevent recurrence.

i ib.

&L NbC

.I 51U L Ubcu l l j !

Y i

3 u

.)

'?

r i

Ii 5 =

i '

l CATAWBA - UNIT 2 3/4 4-17 Amendment No.

142

A%%%iMHus%<.

m.wae49 mugsenkm,

Table 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION T

Preservice Inspection No Yes No. of Steam Generators per Unit Four Four First Inservice Inspection All Two l

2

' Second & Subsequent Inservice Inspections One One TABLE NOTATIONS i

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 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 be modified to inspect the most severe conditions.

2 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.

k O

I CATAWBA - UNIT 2 3/4 4-18 Amendment No.

142 N

-..Y

pe= N w. spin yEtvse-.-% >.,

aw.

N

'N TABLE 4.4-2 Ssm STEAM GENERATOR TUBE INSPECTION g

1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTIO 3RD SAMPLE INSPECTION Action Re[uired Result Action Required Sample Size Result Action Required Result 1

A minimum of C-1 None N.A.

N.A.

N.A.

N.A.

I S Tubes per C-2 Plug defective tubes C-1 Noqe N.A.

N.A.

S.G.

and inspect additional C-2 P!

deflective tubes and C-1 None 2S tubes in this S.G.

inspect additional 4S C-2 Plug defective tubes tubes in this S.G.

C-3 Perform action for C-3 l

result of first sample C-3 Perform action for C-3 N.A.

N.A.

result of first sample C4 Inspect all tubes in this All other None N.A.

N.A.

S.G., plug defective S.G.s are C-1 tubes and inspect 2S tubes in each other S.G.

Some S.G.s Perform action for C-2 N.A.

N.A.

C-2 but no result of second sample Notification to NRC additional pursuant to S.G. are C-3 550.72(b)(2) of 10 CFR Part 50.

Additiona!

Inspect all tubes in each N.A.

N.A.

S.G.is C-3 S.G. and plug defective tubes. Notification to NRC pursuant to 150.72 (b)(2) of 10 CFR 50.

S = 3 (N/n)% Where N is the number of steem generators in the unit, and n is the number of steam generators inspected during an inspection.

l l

CATAWBA - UNIT 2 3/4 4-19 Amendment No.

142

i REACTOR COOLANT SYSTEM BASES RELIEF VALVES (Continued) of PORVs to control reactor coolant system pressure except for limited periods where the PORY has been isolated due to excessive seat leakage and except for i

limited periods where the PORV and/or block valve is closed because of testing i

and is fully capable of being returned to its normal alignment at any time,

l I

provided that this evolution is covered by an approved procedure.

This is a function that reduces challenges to the code safety valves for overpressurization events..5) Hanual control of a block valve to isolate a stuck-open PORV. Testing of the PORVs includes the emergency Ny supply from 1 :

1 the Cold Leg Accumulators.

This test demonstrates that the valves in the supply line operate satisfactorily and that the nonsafety portion of the i

instrument air system is not necessary for proper PORV operation.

~

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the Reactor Coolant System will be maintained. The program for inservice inspection of steam gen-erator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

j Inservice inspection of steam generator tubing is essential in order to main-tain surveillance of the. conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manu-facturing 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.

c i

The B&W pro s (or me d equivalen to the inspection method de ibed i 4

t Topical ort BAW-5(P)-A,Rev

, will be used Inservice '

pection of stea enerator eeves is al required to ens RCS integ

• y.

Because the s eyes intr uce changes

• the wall thick s and diame

, they reduce t e sensitivi of eddy cur t testing, ther ore, special nspection metho must b used. A met is described

• opical Repo BAW-2045(P)-A,

v. I wi~t supporting y dation data th demonstrates.. e inspectabili of the s eve and unde ying tube. As quired by NRC or licensees a orized to se this rep process, Cata a commits to v idate the ad acy of any sys-tem that i used for perio '

inservice in etions of t sleeves, and will evaluate nd, as deemed propriate by D e Power Com y, implement testing tmethods as better met. ds are develope and validat for commercial use.

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

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 150 gallons per day per steam generator).

CATAWBA - UNIT 2 B 3/4 4-3

\\

fl REACTOR COOLANT SYSTEM

, BASES

]

STEAM GENERATORS (Continued)

Cracks having a reactor-to-secondary leakage less than this limit during l

operation will have an adequate margin of safety to withstand the loads imposed during nonnal operation and by postulated accidents. Operating plants have demonstrated thr.t reactor-to-secondary leakage of 150 gallons per day per f

i steam generator car. readily be detected.

Leakage in excess of this limit will i

require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and repaired.

l Wastage-type defects are unlikely with proper chemistry treatment of the

$ econdary coolant. However, even if a defect should develop in service, it will be found during scheduled intervice steam generator tube examinations. A

"~

Nc air will be required for all tubes with imperfections exceeding the repur i

liInit of 40% of the tube nominal wall thickness. Defective :te generater 4'he i stallation of :lceve: which span the are: cf-

-tubes c:r be repa4*:

n i

i degradation, ani-serve cs e rcplacc;cnt pressurc boundary for the degraded

)

il9 portion of the ?&v-al-%w4tg the tube te rem:!" i" service. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage t,'.pe degradation that has penetrated 20% of the original tube wall thickness.

j 1

-Tubc; experiencing cuter diancter stres; cerrction cracking within the thickne:: cf the tube cupport plate; cre plugged or repaired by the-criter.e

.)

If 4.4.5.4..13.

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 6.9.2 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, addi-

'~

tional eddy-current inspection, and revision of the Technical Specifications, if necessary.

If e tube is slecved duc to dcgradatica in th: F* distance, then eny defects in th tubc bcicw the :1ceve will rc :in in service withcut-repair.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS Ine Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

CATAWBA - UNIT 2 B 3/4 4-4

1 J

t 1

Typed Clean Pages s

s b

a 4

4 i

s 4

l 1

l i

i a

~,

REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 1)

All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),

2)

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

A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be perfonned on each selected tube.

If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be l

selected and subjected to a tube inspection.

The tubes selected as the second and third samples (if required by l

c.

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.

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

Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

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

and 10% of the total tubes inspected are degraded tubes.

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

Note:

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

CATAWBA - UNIT 2 3/44-13 Amendment No.

b REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.

As used in this specification:

1)

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

specifications. Eddy-current testing indications below 20% of 4~

the nominal tube wall thickness, if detectable, may be l

considered as imperfections; 1

2)

Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube; 1

3)

Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal tube wall thickness caused by degradation; 4)

% Dearadation means the percentage of the tube wall thickness l

affected or removed by degradation; 5)

Defect means an imperfection of such severity that it exceeds j

the plugging limit. A tube containing a defect is defective; l

6)

Pluaaina Limit means the imperfection depth at or beyond which the tube shan be removed from service by plugging. The plugging limit is equal to 40% of'the nominal tube wall thickness.

7)

Unserviceable ilescribes the condition of a tube if it leaks or j

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

1 1

I l

CATAWBA - UNIT 2 3/44-15 Amendment No.

REACTOR COOLANT SYSTEM j

SURVEILLANCE RE0VIREMENTS (Continued) 9)

Preservice inspection means an inspection of the full length of each tube in each steam generator perfomed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be perfomed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

t I

j b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit l

and all tubes containing through-wall cracks) required by Table 4.4-2.

)

]

4.4.5.5 ReDorts J

a.

Within 15 days following the completion of each inservice inspection 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; 4

b.

The complete results of the steam generator tube 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:

l 4

l CATAWBA - UNIT 2 3/44-16 Amendment No.

REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 1)

Number and extent of tubes inspected, 2)

Location and percent of wall-thickness penetration for each indication of an imperfection, and 3)

Identification of tubes plugged.

j c.

Results of steam generator tube inspections, which fall into Category C-3, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

I i

i CATAWBA - UNIT 2 3/44-17 Amendment No.

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimua of S C-1 None N.A.

N.A.

N.A.

N.A.

C-2

• Plug defective tubes C-1 None N.A.

N.A.

and inspect additional 25 tubes in this S.G.

C-2 Plug defective tubes C-1 Mone I

and inspect additional 45 tubes in this S.G.

C-2 Plug defective tubes C-3 Perform action for C-3 result of first sample C-3 Perfom 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 S.G.s are C-1 defective tubes and 5 me S.G.s C-Perform action for C-2 N.A.

N.A.

oth r 2 but no result of second sample tiona Notification to NRC pursuant to

$50.72(b)(2) of 10 CFR Additional Inspect all tubes in N.A.

N.A.

Part 50.

S.G.is C-3 each S.G. and plug defective tubes.

Notification to NRC pursuant to $50.72 (b)(2) of 10 CFR 50.

S = 3 (N/n)% Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection.

CATAWBA - UNIT 2 3/44-19 Amendment No.

REACTOR COOLANT SYSTEM BASES RELIEF VALVES (Continued) of PORVs to control reactor coolant system pressure except for limited periods I

where the PORV has been isolated due to excessive seat leakage and except for limited periods where the PORV and/or block valve is closed because of testing and is fully capable of being returned to its nonnal alignment at any time, provided that this evolution is covered by an approved procedure. This is a function that reduces challenges to the code safety valves for overpressurization events.

5) Manual control of a block valve to isolate a j

stuck-open PORV. Testing of the PORVs includes the emergency N2 supply from the Cold Leg Accumulators.

This test demonstrates that the valves in the supply line operate satisfactorily and that the nonsafety portion of the i

instrument air system is not necessary for proper PORV operation.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the Reactor Coolant System will be maintained. The program for inservice inspection of steam gen-erator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to main-i tain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manu-4 facturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing i

the nature and cause of any tube degradation so that corrective measures can i

be taken.

~

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

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System 4

(reactor-to-secondary leakage = 150 gallons per day per steam generator).

1 CATAWBA - UNIT 2 B3/44-3

REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued)

Cracks having a reactor-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 reactor-to-secondary leakage of 150 gallons per day per steam generatcr ce readily be detected. Leakage in excess of this limit will require plar.c shutdcwn and an unscheduled inspection, during which the leaking tubes will be located and repaired.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. Hodever, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for ail tubes with imperfections exceeding the plugging limit of 40% of the tube cominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20% of the original tube wall thickness.

l 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 6.9.2 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, addi-tional eddy-current inspection, and revision of the Technical Specifications, if necessary.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The Leakage Detection Systems required by this specification are provided l

to monitor and detect leakage from the reactor coolant pressure boundary.

These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

j CATAWBA - UNIT 2 B 3/4 4-4

.. =..

3 4

f i

e a

L

l

~

• Catawba Nuclear Station, Unit 2 Technical Specification Change i

l Page 1 of 3 l

Proposed Revision to Technical Specification 4.4.5.

The surveillance Requirements are changed to delete repair methods that are not applicable to the Westinghouse Model D5 l

l Steam Generators used in Unit 2 and will not be used as repair methods. References to F*,

sleeving and alternate tube i

plugging criteria are deleted.

Unused paragraph numbers and one miss-spelled word in Table 4.4-2 will also be corrected.

i Technical Justification h

This proposed change to the Technical Specifications deletes

~

repair repair criteria (F* and Alternate Tube Plugging) and repair methods (sleeving) that are not applicable to the Catawba Unit 2 steam generators.

These criteria have not been used and are not applicable to the Westinghouse Model D5 steam generators in use in Catawba Unit 2.

The repair methods were applicable to Catawba Unit 1 only prior to steam generator replacement.

At the time of the Technical Specification split (Amendment No. 142) it was not recognized that these repair methods were applicable to Catawba Unit 1 only and should not have been included in the Catawba Unit 2 Technical Specifications.

These changes will make the Unit 2 steam generator surveillance requirements consistent with the Standard Technical Specifications, NUREG -0452, Rev.

4.

Deletion of unused paragraph numbers and correction of a typographical error are administrative and not technical changes.

The proposed changes are consistent with Catawba's planned implementation of the new Improved Standard Technical Specifications (ISTS).

Duke is pursing this amendment request separately and prior to Catawba's ISTS submittal date in order to correct identified errors in the Catawba Technical Specifications in a more timely manner.

NO SIGNIFICANT HAZARDS EVALUATION Pursuant to 10CFR50.92, Duke Power Company has determined that this license amendment request involves No Significant Hazards l

Considerations.

The changes proposed in this amendment apply to the Technical Specification surveillance requirements for the steam generators. The surveillance requirements are changed to delete repair criteria (F' and Alternate Tube Plugging) and repair methods (sleeving) that are not applicable to the Catawba Unit 2 steam generators.

These 1

__._-_______._-_~._._.__-.m i

L Catawba Nuclear Station, Unit 2 Technical Specification Change Page 2 of'3 criteria have not been used and are not applicable to the Westinghouse Model D5 steam. generators in use.in Catawba Unit i

2. The proposed changes also delete unused paragraph numbers and corrects one typographical error (these are considered administrative changes).

L The determination of no significant hazards was made by l

applying the NRC established standards contained in regulation j

10CER50.92.

These standards assure that any changes to the l

operation of Catawba Nuclear Station in accordance with this I

amendment consider the following:

l 1)

Will the change involve a significant increase in the L

probability or consequences of an accident previously-evaluated?

L No. This amendment to the Catawba Unit 2 Technical l

Specifications-will have no impact on operation of the facility since the change will delete steam generator repair methods that are not applicable to the Catawba Unit 2 steam generators and have not been used to repair the Catawba Unit 2 steam generators.

2)

Will the change create the possibility of a new cn:

different type of. accident from any accident previously evaluated?

No.

This amendment will delete steam generator repair methods that are not applicable and have not been used.

Therefore, j

the proposed changes will not create the possibility of a new 1

or different accident.

{

i I

3)

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

No.

This amendment will delete steam generator repair methods that are not applicable and have not been used.

There will be no impact on safety margins as a result of these changes.

Catawba Nuclear Station, Unit 2 Technical Specification Change I

Page 3 of 3 Environmental Impact Assessment J

i This change to the Technical Specifications wi.'.1 delete stean generator surveillance requirements that are not applicable and not being used.

It has been determined that this amendment will not involve a significant hazards consideration, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and that there is no significant increase in the individual or cumulative occupational radiation exposure.

This amendment request therefore meets the criteria 4

of 10 CFR 51.22. (c) (9) for categorical exclusion from an environmental impact statement.

i Committee Reviews This proposed change to the Technical Specifications has been reviewed and approved by the Catawba Plrnt Operations Review Committee and the Nuclear Safety Review ;

'rd.

I