ML20129G710

Amend 116 to License NPF-30,revising TS 3/4.4 & Associated Bases to Address Installation of Laser Welded Tube Sleeves in Plant Steam Generators

ML20129G710
Person / Time
Site:	Callaway
Issue date:	10/01/1996
From:	Thomas K NRC (Affiliation Not Assigned)
To:
Shared Package
ML20129G714	List: ML20129G710 ML20129G723
References
NUDOCS 9610070272
Download: ML20129G710 (15)

Text

-

j#

%g g'

t UNITED STATES g

,j NUCLEAR REGULATORY COMMISSION l'

WASHINGTON, D.C. 2006HW01

%...../

UNION ELECTRIC COMPANY CALLAWAY PLANT. UNIT 1 DOCKET NO. 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 116 License No. NPF-30 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Union Electric Company (UE, the licensee) dated April 12, 1996, as supplemented by letters dated August 2, 1996, August 19, 1996, and September 5, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I;

.l B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by

)

this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be i

conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; ano E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-30 is hereby amended to read as follows:

l l

9610070272 961001 PDR ADOCK 05000483 p

PDR

- (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 116 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

l 3.

This amendment is eff active as of its date of issuance and will implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COPMISSION j, M u ( t.. T h t v & A Kristine M. Thomas, Project Manager Project "fiectorate IV-2 Divisio,of Reactor Projects III/IV Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 1, 1996

)

4' l

l 1

c r

1 4

REACTOR COOLANT SYSTEM i

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 steam generator (s) to OPERABLE status prior to increasing T, above 200*F.

i SURVEILLANCE RE0VIREMENTS l

4.4.5.0 Each steam generator shall be demonst ated OPERABLE by performance of i

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

4.4.5.1 Steam Generator Samole Selection and Insoection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in 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 Tables 4.4-2 and 4.4-3.

l The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.

When applying the exceptions of 4.4.5.2.a through 4.4.5.2.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring inspection.

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 indicatas critical areas to be inspected, then at least 50% of the l

tubes inspected shall be from these critical areas; i

i b.

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

CALLAWAY - UNIT 1 3/4 4-11 Amendment No. 116 l

i

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (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 i

l 3)

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

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

inspection.

t l

c.

The tubes selected as the second and third samples (if required by Tables 4.4-2 or 4.4-3) during each inservice inspection may be l

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 i

following three categories:

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

t C-2 One or more tubes, but not more f han 1% of the total tubes inspected are defective, or between 5% and 10% of the tota'. 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 1

to be included in the above percentage calculations.

I l

i i

i CALLAWAY - UNIT 1 3/4 4-12 Amendment No.116

REACTOR COOLANT SYSTEM i

i SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.3 Insoection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

j 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 wths after the previous inspection.

If two consecutive inspections, 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; b.

If the results of the inservice inspection of a steam generator conducted in accordance with Tables 4.4-2 and 4.4-3 at 40-month l

i intervals fall in Category C-3, the inspection frequency shall be i

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 i

c.

Addi%nal, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Tables 4.4-2 and 4.4-3 during the shutdown subsequent to l

any of the following conditions:

1 1)

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

i 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.

t CALLAWAY - UNIT 1 3/4 4-13 Amendment No.116 l

REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 4.4.5.4 Accentance Criteria As used in this specification:

a.

1)

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

Dearadation means a service-induced cracking,

wastage, wear or general corrosion occurring on either inside or outside of a tube; 3)

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

% Dearadation means the percentage of the tube wall thickness affected or removed by degradation; 5)

Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube or sleeve containing a g

defect is defective; 6)

Pluaaina or Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving and is equal to 40% of the nominal tube wall thickness. The plugging limit for laser welded sleeves is i

equal to 39% of the nominal sleeve wall thickness; i

7)

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural l

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 Specification 4.4.5.3c., above; i

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.

For a tube repaired by sleeving, the tube inspection shall include the sleeved portion of the tube; 1

CALLAWAY - UNIT 1 3/4 4-14 Amendment No.116

htEACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 9)

Preservice Insoection means an inspection of the full length of each tube in each steam generator performed by addy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspect'ons; and l

10) Tube Repair refers to a process that reestablishes tube i

serviceability. Acceptable tube repairs will be performed by j

the following processes:

a)

Laser welded sleeving as described in Westinghouse t

Technical Report WCAP-14596-P, " Laser Welded Elevated Tube Sheet Sleeves for Westinghouse Model F Steam Generators."

March 1996 (W Proprietary) l.

b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by sleeving all tubes exceeding the plugging or repair limit and all tubes containing i

through-wall cracks) required by Tables 4.4-2 and 4.4-3.

4.4.5.5 Reporti Within 15 days following the completion of each inservice inspection a.

of steam generator tubes, the number of tubes plugged or repaired in l

each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; 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:

1)

Number and extent of tubes and sleeves inspected,

[

2)

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

Identification of tubes plugged or repaired.

l 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 i

and corrective measures taken to prevent recurrence,

!+

i l

CALLAWAY - UNIT 1 3/4 4-15 Amendment No. 116 l

n2 N

5 TABLE 4 4-1 MINIMUM NUMBER OF STE AM GENER ATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspecten No Yes No. of Steam Generators per Unit Two Three Four Two Three Four Fast inservice inspection All One Two Two 3

I l

OM One Saond & Subsecuent Inservice Inspections One One TABLE NOTATIONS M

1. The mservice mopection may be timeted to one steam generator on a rotating schedule encompassing 3 N % of the tubes 6

twhere N es the number of steam generators m the plant) of the results of t e twst or previous inspectens endecate that h

A all steam generators are performeng m a tihe manner Note that under some cwcumstances, the operatmg conditions m one or more steam generators may be found to be more severe than those m other steam generators. Under such circum cn stances the sampfe sequence shall be modified to inspect the most severe conditions.

2. The other steam generator not inspected durirq the first mservice inspection shall be inspected _ The third and subsequent mspections should follow the instructions described in 1 above.

3. Each of the other two steam generators not mspected durmg the first inservice inspections shall be inspected during the second and therd inspectens. The fourth and subsequent mspections shall follow the instructions described in 1 above.

t b

e

=

w

+

ve =

v.

-rv-.-

,-rw-r-

---r

.s

- =

A

n TABLE 4.4-2 r-r=

-D STEAM GENERATOR TUBE INSPECTION D-<

t Cb 1ST SAMPLE INSPECTION l

2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Sies Resut Action Required l

Result Action Regsired Result Action Required A rnitumum of S C-1 None N.A.

N.A.

N.A.

N.A.

Tubes per S.0.

C-2 Plug or repair defective C-1 None N.A.

N.A.

l tthee end Inspect odditional 2S tubee in tNo S.G.

C-2 Plug or repair defective C-1 None l

tubes and inspect additional 45 tubes in thie S.G.

C-2 Plug or repair

[

(

defective tubee a

C-3 Perform act6on for C-3 result of first eensle L

C-3 Perform action for C-3 N.A.

N.A.

N L

result of first eergle C-3 Inspect oli tubes in this All other S.G.e None N.A.

N.A.

I S. G., plug or repair see C-1 defective tubee and inspect 2S tubes in each other S.G.

Borne S.O.e C-2 Perform setteet tot C 2 N.A.

N.A.

Notificetten to NRC but no odditional result of eeeend compte pursuant to 150.72 8.G.ero C-3 tbH21 of to crR l' art 50 Additionel S. G. le inepect oli tubee in each N.A.

N.A.

C-3 S.G. end plug er repair l

defective tuhes.

Notification to NRC i

3 purovent to 150.72 o.

(b)(2) of to CFR Port 50 S=3N % where N is the m,mber of steem venerefore in th. unit. and n is the number of steem eeneretore in.pected durine en inop.cison R

.o

.4 O

r e

TABLE 4.4-3 STEAM GENERATOR REPAIRED TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION Sample Size Result Act on Required Result Action Required A minimum of 20% of C-1 None N.A.

N A.

repaired tubes (1)(2)

C2 Plug defective repaired tubes C1 None and inspect 100% of the repaired tubes in this S G.

C-2 Plug defective repaired tubes C3 Perform acton for C-3 result of first sample C3 inspect all repaired tubes in this All other S G.s are None S G., plug defective tubes and C1 inspect 20% of the repaired tubes in each other S.G.

Notification to NRC pursuant to Some S.G s C-2 Perform action for C 2 result of 150 72 (b)(2) of 10 CFR Par 150 but no adddional first sample j

$ G. are C-3 l

i Additional S.G. is inspect all repsired tubes in C3 each S G. and plug defective tubes Notification to NRC pursuant to $50.72 (b)(2) of 10 l

I CFR Part 50 (1) Each repair method is consdered a separate population for determination of scope expansson.

(2) The inpection of repaired tubes may be performed on tubes from 1 to 4 steam generators based on outage plans.

CALLAWAY - UNIT I 3/4 4-17a Amendment No.116

RE' ACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING JOITION FOR OPERATION l

3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall l

be OPERABLE:

l a.

The Containment Atmosphere Particulate Radioactivity Monitoring

System, j

l

[

b.

The Containment Normal Sump Level Measurement System, and 1

i c.

Either the Containment Air Cooler Condensate Flow Rate or the Containment Atmosphere Gaseous Radioactivity Monitoring System.

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

ACTION:

With only two of the above required Leakage Detection Systems OPERABLE, l

operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed for gaseous and particulate radio-activity or a gamma isotopic analysis of the containment atmosphere is performed using the Post Accident Sampling System at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

j a.

Containment Atmosphere Gaseous and Particulate Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, h.

Containment Normal Sump Level Measurement System performance of l

- CHANNEL CALIBRATION at least once per 18 months, and j

c.

Containment Air Cooler Condensate Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.

CALLAWAV - UNIT 1 3/4 4-18

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LINITING CONDITION FOR OPERATION l

3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

I gpa UNIDENTIFIED LEAKAGE,

)

I c.

600 gpd total reactor-to-secondary leakage through all steam l

l generators not isolated from the Reactor Coolant System and 150 gallons per day through any one steam generator, l

d.

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

8 gpm per RC pump CONTROLLED LEAKAGE at a Reactor Coolant System I

pressure of 2235 i 20 psig, and 1

f.

The leakage from each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be limited to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gps, at a Reactor Coolant System pressure of 2235 't 20 psig.*

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

ACTION:

)

a.

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

b.

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

c.

With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, reduce the leakage rate to within i

limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with an RCS pressure of less than 600 psig.

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

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

I i

i CALLAWAY - UNIT 1 3/4 4-19 Amendment No. 66,116 I

l

REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surve111ance 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.

Unscheduled inservice inspections are pe'rformed on each steam generator following:

(1) reactor to secondary tube leaks; (2) a seismic occurrence greater than the Operating Basis Earthquake; and (3) a loss-of-coolant accident requiring actuation of the Engineered Safety Features, which for this Specification is defined to be a break greater than that equivalent to the severance of a 1" inside diameter pipe, or, for a main steamline or feedline, a break greater than that equivalent to a steam generator safety valve failing open; to ensure that steam generator tubes retain sufficient integrity for continued operation.

Transients less severe than these do not require inspections because the resulting stresses are well within the stress criteria established by Regulatory Guide 1.121, which unplugged steam generator tubes must be capable of withstanding.

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

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 150 gallons per day per steam generator).

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 l

steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located, plugged or repaired.

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

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

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.

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

CALLAWAY - UNIT 1 B 3/4 4-3 Amendment No.116

i REACTOR COOLANT SYSTEM BASES i

i STEAM GENERATORS (Continued)

}

j The plueging or repair limit for the pressure boundary portion of laser welded sleeves is determined to be 39% through-wall (by NDE). The laser j

welded sleeve repair limit applicable to the pressure boundary portion of the l

I sleeve is established in WCAP-14596. Appropriate NDE techniques are also j

j discussed in WCAP-14596.

i 1

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results wiL1 be reported to the Commission pursuant to Specification 6.9.2 prior to resumption of plant operation. Such I

cases will be considered by the Commission on a case-by-case basis and may

{

result in a requirement for analysis, laboratory examinations, tests, l

additional eddy-current inspection, and revision of the Technical i

Specifications, if necessary, i

j 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE i

j 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressura boundary. These Detection Systems are consistent with the recommendati)ns of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE i

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to i

1 be promptly placed in COLD SHUTDOWN.

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

}

The total steam generator tube leakage limit of 600 gpd for all steam l

generators not isolated from the RCS ensures that the dosage contribution from the tuue leakage will be limited to a small. fraction of 10 CFR Part 100 dose guideline va'ues in the event of either a steam enerator tube rupture or steam line break.

The 600 gpd limit is conservative compared to the assumptions used in the analysis of these accidents. The 150 gpd leakage limit per steam generator ensures that steam generator tube integrity is j

maintained in the event of a main steam line rupture or under LOCA conditions.

j The 10 gpa IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere j

with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.

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

i

~

CALLAWAY - UNIT 1 B 3/4 4-4 Amendment No.116 1

~,

o ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.116 TO FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE INSERT 3/4 4-11 3/4 4-11 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 4-15 3/4 4-15 3/4 4-17 3/4 4-17 3/4 4-17a 3/4 4-19 3/4 4-19 B 3/4 4-3 8 3/4 4-3 B 3/4 4-4 8 3/4 4-4 l

<