Text

Amend 151 to License NPF-35,revising TS to Reflect New Setpoints,Operational Parameters & Approved Analysis Methodologies Associated W/Replacement of Unit 1 SGs
	ML20117G846
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Site:	Catawba
Issue date:	08/29/1996
From:	Berkow H; NRC (Affiliation Not Assigned)
To:	;
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	NUDOCS 9609060010
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9k UNITED STATES 0

g NUCLEAR REGULATORY COMMISSION t

WASHINGTON, D.o. 205dW4001 49*****

,o DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE. INC.

DOCKET NO. 50-413 CATAWBA NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 151 License No. NPF-35 1.

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

A.

The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-35 filed by the Duke Power Company, acting for itself, North Carolina Electric Membership Corporation and Saluda River Electric Cooperative, Inc. (licensees), dated September 30, 1994, as supplemented September 18, 1995, January 19, March 15, May 16, and August 27, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I, B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the l

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 condacted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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; and 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.

9609060010 960829 i

PDR ADOCK 05000413 i

P PDR

i

. 2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No.

NPF-35 is hereby amended to read as follows:

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 151, which is attached hereto, is hereby i

incorporated into this license. Duke Power Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION H r ert N. Berkow, Director roject Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Resctor Regulation

Attachment:

Technical Specification Changes Date of Issuai.ce:

August 29, 1996 l

l l

i e

i

1 ATTACHMENT TO LICENSE AMENDMENT NO. 151 FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 Replace the following pages of the Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Paces Insert Paaes VII VII VIII VIII 2-5 2-5 2-8 2-8 2-10 2-10 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 l

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 4-16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-18 3/4 4-18 3/4 4-19 3/4 4-19 3/4 4-20 3/4 4-20 3/4 4-21 3/4 4-21 3/4 4-22 3/4 4-22 l

3/4 4-23 3/4 4-23 3/4 4-24 3/4 4-24 3/4 4-25 3/4 4-25 3/4 4-26 3/4 4-26 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 3/4 4-33 3/4 4-33 3/4 4-34 3/4 4-34 3/4 4-35 3/4 4-35 3/4 4-36 3/4 4-36 3/4 4-37 3/4 4-37 3/4 4-38 3/4 4-38 3/4 4-39 3/4 4-39 3/4 4-40 3/4 4-41 3/4 4-42 3/4 4-43 5-7 5-7 6-22 6-22 B 3/4 4-3 B 3/4 4-3 B 3/4 4-4 B 3/4 4-4

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE 3/4.4.3 PRESSURIZER........................

3/4 4-9 3/4.4.4 RELIEF VALVES.......................

3/4 4-10 3/4.4.5 STEAM GENERATORS 3/4 4-12 i

TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION 3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 3/4 4-18 L

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE

. Leakage Detection Systems.................

3/4 4-19 Operational Leakage....................

3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES....

3/4 4-22 3/4.4.7 CHEMISTRY.........................

3/4 4-24 TABLE 3.4-2 REACTOR. COOLANT SYSTEM CHEMISTRY LIMITS 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS......................

3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY.....................

3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131.................

3/4 4-28 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM..........................

3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS 1

Reactor Coolant System 3/4 4-31 1

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 10 EFPY 3/4 4-32 FIGURE 3.4-3 REACTUR COOLANT SYSTEM C00LD0WN LIMITATIONS -

APPLICABLE UP TO 10 EFPY 3/4 4-33 CATAWBA - UNIT 1 VII Amendment No. 151

a

.C LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE i

TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE 3/4 4-34 t

P re s s u ri z e r...... '..................

3/4 4-35 Overpressure Protection Systems..............

3/4 4-36

]

3/4.4.10 STRUCTURAL INTEGRITY 3/4 4-38 3/4.4.11 REACTOR COOLANT SYSTEM VENTS 3/4 4-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Cold Leg Injection 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350'F 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350*F 3/4 5-8 3/4.5.4 REFUELING WATER STORAGE TANK 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Contai nment Integri ty.................... 3/4 6-1 Containment Leakage....................

3/4 6-2 TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS.......

3/4 6-5 Containment Air Locks...................

3/4 6-8 Internal Pressure.....................

3/4 6-10 Air Temperature......................

3/4 6-11 Containment Vessel Structural Integrity..........

3/4 6-12 Reactor Building Structural Integrity...........

3/4 6-13 Annulus Ventilation System 3/4 6-14 Containment Purge Systems.................

3/4 6-16 CATAWBA - UNIT 1 VIII Amendment No.

151

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

12. Steam Generator Water Level Low-Low 2 10.7% of narrow range 2 9% of narrow range span span

13. Undervoltage - Reactor 2 77% of bus voltage 2 76% (5016 volts)

Coolant Pumps (5082 volts) with a 0.7s response time

14. Underfrequency - Reactor 2 56.4 Hz with a 2 55.9 Hz Coolant Pumps 0.2s response time

- 15. Turbine Trip a.

Stop Valve EH 2 550 psig 2 500 psig-Pressure Low b.

Turbine Stop Valve Closure 2 1% open 2 1% open

- 16. Safety Injection Input from ESF N.A.

N.A.

CATAWBA - UNIT.1 2-5 Amendment No.

151 O

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TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1:

(Continued)

T' s 585.1*F (Nominal T,yg allowed by Safety Analysis);

K Overtemperature AT reactor trip depressurization setpoint penalty coefficient as presented in the

=

3 Core Operating Limits Report; P

Pressurizer pressure, psig;

=

P' 2235 psig (Nominal RCS operating pressure);

=

Laplace transform operator, s'l; S

=

and f, (AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:

(i)

For qt -9b between the " positive" and " negative" f, (AI) breakpoints as presented in the Core Operating Limits Report; fi (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii)

For each percent AI that the magnitude of q, - gb is more negative than the f, (AI) " negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the fi (AI) " negative" slope presented in the Core Operating Limits Report; and (iii)

For each percent AI that the magnitude of qt -9b is more positive than the fi (AI) " positive" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the fi (AI) " positive" slope presented in the Core Operating Limits Report.

NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.5' of 4

Rated Thennal Power.

~

CATAWBA - UNIT 1 2-8 Amendment No. 151 6

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 3:

(Continued)

K, Overpower AT reactor trip heatup setpoint penalty coefficient as presented in the Core Operating

=

Limits Report for T > T" and K = 0 for T s T",

T As defined in Note 1

=

T" Indicated T,y, at RATED THERMAL POWER (Calibration temperature for AT instrumentation, s 585.1*F),

=

S As defined in Note 1,

=

and f2 (AI) is a function of the indicated differences between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

l (i) for gt between the " positive" and " negative" f2 (AI) breakpoints as presented in the Core 9b i

Operating Limits Report; f2 (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and gt + 9 is total THERMAL POWER in percent of RATED THERMAL b

POWER; 1

(ii) for each ' percent AI that the magnitude of qt -9b is more negative than the /2 (AI) " negative" breakpoint presented in the Core Operating Limits Report, the AT Trip Setpoint shall be automatically reduced by the /2 (AI) " negative" slope presented in the Core Operating Limits Report; and I

r l

(iii) for each percent AI that the magnitude of q, - gb is more positive than the /2 (AI) " Positive" breakpoint presented in the Core Operating Limits Report the AT. Trip Setpoint shall be automatically reduced by the f2 (AI) " positive" slope presented in the Core Operating Limits Report.

NOTE 4:- The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.0% of Rated Thermal Power.

CATAWBA - UNIT 1 2-10 Amendment No.

151 W

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 4

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

4. Steam Line Isolation a.

Manual Initiation N.A.

N.A.

j b.

Automatic Actuation Logic N.A.

N.A.

1 and Actuation Relays c.

Containment Pressure-High-High s 3 psig s 3.2 psig j

j d.

Steam Line Pressure - Low 2: 775 psig 2: 744 psig e.

Steam Line Pressure-s 100 psi s 122.8 psi **

Negative Rate - High 5.

Feedwater Isolation i

i a.

Automatic Actuation Logic N.A.

N.A.

Actuation Relays b.

Steam Generator Water Level-High-High (P-14) s 83.9% of s 85.6% of narrow narrow range range instrument instrument span span i

c.

T

-L w 2: 564*F 2: 561*F avg d.

Doghouse Water Level-High 11 inches 12 inches above 577' above 577' floor level floor level e.

Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.

CATAWBA - UNIT 1 3/4 3-31 Amendment No.

151 a

w-w r~-'

r-TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATIDN TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

6. Turbine Trip a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation N.A.

N.A.

Logic and Actuation Relays c.

Steam Generator Water Level-High-High (P-14) s 83.9% of s 85.6% of narrow narrow range range instrument instrument span span d.

Trip of All Main N.A.

N.A.

Feedwater Pumps e.

Reactor Trip (P-4)

N.A.

f.

Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.

7. Containment Pressure Control System a.

Start Permissive s 0.4 psid s 0.45 psid b.

Termination 2 0.3 psid 2: 0.25 psid

8. Auxiliary Feedwater a.

Manual Initiation N.A N.A.

b.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays CATAWBA - UNIT 1 3/4 3-32 Amendment No. 151 e

l i

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

8. Auxiliary Feedwater (Continued) c.

Steam Generator Water Level - Low-Low 2: 10.7% of 2: 9% of narrow range narrow range span span d.

Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.

e.

Loss-of-Offsite Power 2: 3500 V 2: 3242 V f.

Trip of All Main Feedwater Pumps N.A.

N.A.

g.

Auxiliary Feedwater Suction Pressure-Low

1) CAPS 5220, 5221, 5222 2: 10.5 psig 2: 9.5 psig

- 2) CAPS 5230, 5231, 5232 2: 6.2 psig 2: 5.2 psig

9. Containment Sump Recirculation a.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays b.

Refueling Water Storage 2: 177.15 inches 2: 162.4 inches Tank Level-Low Coincident With Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.

CATAWBA - UNIT 1 3/4 3-33 Amendment No.

151

~

REACTOR COOLANT SYSTEM l

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 performed 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.

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

The results of each sample inspection shall be classified into one of the following three ca,tegories:

Cateaorv inspection Results l

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 i

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 l

degraded tubes or more than 1% of the inspected l

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.

l l

CATAWBA - UNIT 1 3/4 4-13 Amendment No. 151

i REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.3 Inspection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

The first inservice inspection after steam generator replacement a.

shall be performed after at least 6 Effective Full Power Months but within 24 calendar months of initial criticality after steam generator replacement. 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, 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 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 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 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.

CATAWBA - UNIT 1 3/4 4-14 Amendment No. 151

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

As used in this specification:

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

considered as imperfections; 2)

Dearadation means a service-induced cracking,

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

3)

Dearaded 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 affected or removed by degradation; j

5)

Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective;

)

6)

Pluaaina Limit means the imperfection depth at or beyond which 1

the tube shall be removed from service by plugging.

The plugging limit is equal to 40% of the nominal tube wall thickness.

7)

Unserviceable describes the condition of a tube if it leaks or 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

9)

Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy 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 inspections.

CATAWBA - UNIT 1 3/4 4-15 Amendment No.151

REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) b.

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

4.4.5.5 Reports 1

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; 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 i

inspection. This Special Report shall include:

1)

Number and extent of tubes inspected, f

2)

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

3)

Identification of tubes plugged.

g c.

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

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

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.

CATAWBA - UNIT 1 3/4 4-16 Amendment No. 151

Table 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection' No Yes No. of Steam Generators per Unit Four Four First Inservice Inspection after Steam All Two Generator Replacement l

2 Second & Subsequent Inservice Inspections-One One TABLE NOTATIONS 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 after steam generator replacement 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.

CATAWBA - UNIT 1 3/4 4-17 Amendment No. 151 Al

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 minimum of C-1 None N.A.

N.A.

S Tubes per S.G.

C-2 Plug defective tubes C-1 None N.A.

N.A.

and inspect additional 2S tubes in this S.G.

C-2 Plug deflective tubes and C-1 None inspect additional 4S tubes in this S.G.

C-2 Plug defective tubes C-3 Perform 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 this All other None N.A.

N.A.

S.G., plug defective S.G.s are C-1 tubes and inspect 2S l

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-3 150.72(b)(2) of 10 CFR Part 50.

Additional Inspect all tubes in each N.A.

N.A.

S.G.is C-3 S.G. and plug defective g

tubes. Notification to NRC, pursuant to 150.72 (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.

I CATAWBA - UNIT 1 3/4 4-18 Amendment No.

1 51 e

4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3

LEAKAGE DETECTION SYSTEMS i

LIMITING CONDITION FOR OPERATION j

3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

The Containment Atmosphere Gaseous Radioactivity Monitoring System, a.

b.

The Containment Floor and Equipment Sump Level and Flow Monitoring Subsystem, and c.

Either the Containment Ventilation Unit Condensate Drain Tank Level Monitoring Subsystem or the Containment Atmosphere Particulate Radioactivity Monitoring System.

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

ACTION:

With only two of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

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, b.

Containment Floor and Equipment Sump Level and Flow Monitoring Subsystem-performance of CHANNEL CALIBRATION at least once per 18 months, and c.

Containment Ventilation Unit Condensate Drain Tank Level Monitoring Subsystem-performance of CHANNEL CALIBRATION at least once per 18 months.

CATAWBA - UNIT 1 3/4 4-19 Amendment No.

151 l

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 gpm UNIDENTIFIED LEAKAGE, c.

0.4 gpm total reactor-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generatur, d.

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

40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig, and f.

1 gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

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

ACTION:

a.

With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and. leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage l

rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 CATAWBA - UNIT 1 3/4 4-20 Amendment No.

1 51 l

REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstra he within each of the above limits by:

Monitoring the containment atmosphere gaseous or particulate radio-a.

activity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; b.

Monitoring the containment floor and equipment sumps inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump c.

seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; d.

Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and e.

Monitoring the reactor head flange leakoff at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a.

At least once per 18 months, b.

Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, Prior to returning the valve to service following maintenance, repair c.

or replacement work on the valve, and d.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

CATAWBA - UNIT 1 3/4 4-21 Amendment No.

151 l

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION NI59 Accumulator Discharge NI60 Accumulator Discharge NI70 Accumulator Discharge i

NI71 Accumulator Discharge NI81 Accumualtor Discharge NI82 Accumulator Discharge NI93 Accumulator Discharge NI94 Accumulator Discharge l

NI124 Safety Injection (Hot Leg)

NI125 Residual Heat Removal (Hot Leg) i NI126 Safety Injection (Hot Leg) i NI128 Safety Injection (Hot Leg)

NI129 Residual Heat Reinval (Hot Leg)

NI134 Safety Injection (Hot Leg)

NI156 Safety Injection (Hot Leg)

NI157 Safety Injection (Hot Leg)

NI159 Safety Injection (Hot Leg)

NI160 Safety Injection (Hot Leg)

NI165 Safety Injection / Residual Heat Removal (Cold Leg NII67 Safety Injection / Residual Heat Removal (Cold Leg NI169 Safety Injection / Residual Heat Removal Cold Leg NI171 Safety Injection / Residual Heat Removal Cold Leg NI175 Safety Injection / Residual Heat Removal Cold Leg NI176 Safety Injection / Residual Heat Removal (Cold Leg NI180 Safety Injection / Residual Heat Removal (Cold Leg NI181 Safety Injection / Residual Heat Removal (Cold Leg l

l CATAWBA - UNIT 1 3/4 4-22 Amendment No. 151 l

l

1 TABLE 3.4-1 (Continued)

REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION ND1B*#

Residual Heat Removal ND2A*#

Residual Heat Removal ND36B*#

Residual Heat Removal ND37A*#

Residual Heat Removal i

Testing per Specification 4.4.6.2.2d. not applicable due to positive indication of valve position in Control Room.

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

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

CATAWBA - UNIT 1 3/4 4-23 Amendment No. 1 51 l

i REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2.

APPLICABILITY: At all times.

ACTION:

MODES 1, 2, 3, and 4:

With any one or more chemistry parameter in excess of its Steady-a.

State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; and b.

With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

At All Other Times:

l With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects.of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4.

SURVEILLANCE RE0VIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3.

CATAWBA - UNIT 1 3/4 4-24 Amendment No. 151 l

_ _ ~

I i

TABLE 3.4-2 j

REACTOR COOLANT SYSTEM i

i CHEMISTRY LIMITS l;

]

STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT 1

l

}

Dissolved Oxygen

s 0.10 ppm s 1.00 ppm i

l Chloride s 0.15 ppm s 1.50 ppm l

Fluoride s 0.15 ppm s 1.50 ppm i

i i

1 i

i i

t i

4 i

i 1

1 1

l l

I i

l i-5 4

d 1

4 i

i i

~

l

Limit not applicable with T,yg less than or equal to 250*F.

i CATAWBA - UNIT 1 3/4 4-25 Amendment No.

151 l

(

1

TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REOUIREMENTS SAMPLE AND PARAMETER ANALYSIS FRE00ENCY Dissolved Oxygen

At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Chloride At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Fluoride At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 9

Not required with T,yg less than or equal to 250'F CATAWBA - UNIT 1 3/4 4-26 Amendment No. 151 l

c

___-~

REACTOR COOLANT SYSTEM j

3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a.

Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and b.

Less than or equal to 100/E microCuries per gram of gross specific activity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2 and 3*:

a.

With the specific activity of the reactor coolant greater than 1 i

microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months J

during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T,yg less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; j

b.

With_the gross specific activity of the reactor coolant greater than 100/E microcuries per gram of gross radioactivity, be in at least HOT STANDBY with T,yg less than 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and c.

The provisions of Specification 3.0.4 are not applicable.

MODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E micro curies per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

With T,yg greater than or equal to 500'F.

CATAWBA - UNIT 1 3/4 4-27 Amendment No.

151 l

l

i, i- -

..l L

..).;.r I. 4..' 1-

..r-p -

J., A a,.

E

.21-L.

e.-

.i.

e

.g U 250 3

2 s

g

. ". ~.. '

~

"'I.

~

3.,,

H 200 O

. UNACCEPTABLE i

o

...O...P E R A.TIO N g

g gg m

g z 150 4

.\\

a

..T c) c)

o m

O

&c 4 100' w

m E

I i

z ACCEPTABLE OPERATION 50 2

Dc w

w m

O l

I l

0 20 30 40 50 60 70 80 90 100 l

PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 Ci/ gram DOSE EQUIVALENT I-131 CATAWBA - UNIT 1 3/4 4-28 Amendment No.

151 l

O

TABLE 4.4-4 l

REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM i

TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED 1.

Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1, 2, 3, 4 Determination **

2.

Isotopic Analysis for DOSE EQUIVA-1 per 14 days 1

LENT I-131 Concentration 3.

Radiochemical for I Determination ***

1 per 6 months

1 4.

Isotopic Analysis for Iodine a)

Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 1,2,3,4,5 Including I-131, I-133, and I-135 whenever the specific activity exceeds 1 pCi/ gram DOSE EQUIVALENT I-131 or 100/E Ci/ gram of gross radioactivity, and b) One sample between 2 1,2,3 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15%

of the RATED THERMAL POWER within a 1-hour period.

CATAWBA - UNIT 1 3/4 4-29 Amendment No. 1 51 l

TABLE 4.4-4 (Continued)

TABLE NOTATIONS Until the specific act:vity of the Reactor Coolant System is restored within its limits.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radioiodines. The total specific activity shall be the sum of the beta-gamma activity in the sample within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level.

The latest available data may be used for pure beta-emitting radionuclides.

- - A radiochemical analysis for E shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines, which is identified in the reactor coolant.

The specific activities.for these individual radionuclides shall be used in the determination of E, for the reactor coolant sample. Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95% confidence level.

CATAWBA - UNIT 1 3/4 4-30 Amendment No. 151 l

4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 60*F in any 1-hour period, b.

A maximum cooldown of 100'F in any 1-hour period, and A maximum temperature change of less than or equal to 10*F in any c.

1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural l

integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDRY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the T and pressure to less than 200*F and 500 psig, respectively, within the fo1 Ding 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE0VIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

j 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in j

Table 4.4-5.

The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

CATAWBA - UNIT 1 3/4 4-31 Amendment No. 151 l

2500

^

2250 l

LEAK TEST LIMIT i

l i

2000 I

l l

1750 l

6 UNACC PTABLEl l

[

Ui OPE R AT ION n.g 1500 i

l

[

ea E

E 1250 a.

O l

Y i

5 1 o*

Q r

z

/

i l

750

/

i i

500 i

CR.lTICALITY L IT BASED i

ON INSE RVICE YOROSTATIC i

TEST TEMP (24 GF) FOR THE i

250 SERVICE PERIOb UP TO l10 EFPY ACCEPTABLE' l

ORERATION I

j j

'i 0

O 50 100 150 200 250 300

.350 400 450' 500 INDICATED TEMPERATURE (DEG. F) curve APetN:AsLE FOR NEATUP MATERIAL B ASIS RATES UP 70 00*P/H4 POR THE SERVICE PERICO UP TO 10 EPPY CONTROLLING MATERIAL-UNIT 2 INTERMEDIATE SHELL CONTAINS MARGIN OF 10 7 PLATE 88005-2 8

COPPER CONTENT C.07wr%

ANO 80 PS8G FOR POS$1SLE NICKEL CONTENT- 0.81 we %

INSTRUMENT ERRORS.

RTNOT "'IIAL"33*E 8

RTgyAPTER 10 EPPY 1/4T.10e*P 3/ 4T, 92* P FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 10 EFPY CATAWBA - UNIT 1 3/4 4-32 Amendment No.

151 l

i 2500 2250 -

2000

)

i 1750 I

I i

6 UNACCEPTABLE i

OPERATION G

15M m

3 M

l 1250 a.c N

1000 5

CMLDOWh ACCEPTABLE g

RATES OPERA"lON OF.'H R 750 g;W l

20 k m i

500

u y 60 i

100-250 j

l 1

0 O

50 100 150 200 250 300 350 400 450 500 INDICATED TEMPER ATURE (DEG. F)

CURVE APPLMllASLE FOR COOLDOWN MATE REAL Basis RATES UP TO ige *P/MR FOR THE CONTROLLING MATERIAL UNif 2 INTERMEOfATE SHELL SERVICE PERICO UP TO 10 EPPY PLATE Stele 2 CONTAING MARGIN OF 10*P COPPER CONTENT-0.07mt%

ANO to Psl0 POR POSSISLE NICK E L CONTENT-0.81 wt%

INSTRUMENT ERRORS.

RTNOTINITIALMP RTNOTAPTER 10 EPPY 1/47,104*P j

3/47 SI*P i

l FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -

APPLICABLE FOR THE FIRST 10 EFPY CATAWBA - UNIT 1 3/4 4-33 Amendment No. 151 l

o

TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE CAPSULE VESSEL LEAD NUMBER LOCATION FACTOR WITHDRAWN TIME (EFPY)

U 58.5 3.85 Standby V

61*

3.65 9

W 121.5*

3.85 Standby X

238.5*

3.85 Standby Y

241*

3.65 5

Z 301.5 3.85 First Refueling CATAWBA - UNIT 1 3/4 4-34 Amendment No. 151 1

m

.m..

m

1 REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

a.

A maximum heatup of 100*F in any 1-hour period, and b.

A maximum cooldown of 200'F in any 1-hour period.

APPLICABILITY: At all times.

i i

ACTION:

i With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE0VIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.

CATAWBA - UNIT 1 3/4 4-35 Amendment No. 151 l

1

l REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION l

3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:

a.

Two power operated relief valves (PORVs) with a lift setting of less than or equal to 450 psig, or b.

The Reactor Coolant System depressurized with a Reactor Coolant System vent of greater than or equal to 4.5 square inches.

APPLICABILITY: MODE 4 when the temperature of any Reactor Coolant System cold leg is less than or equal to 285*F, MODE 5 and MODE 6 when the head is on the reactor vessel.

ACTION:

a.

With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days or complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

With one PORV inoperable in MODES 5 or 6, restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c.

With both PORVs inoperable, complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d.

In the event either the PORVs or the Reactor Coolant System vent (s) are used to mitigate a Reactor Coolant System pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence.

e.

The provisions of Specification 3.0.4 are not applicable.

4 4

+

t

. CATAWBA - UNIT 1 3/4 4-36 Amendment No. 151 l

1

4 REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per 31 days; b.

Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months c.

when the PORV is being used for overpressure protection.

4.4.9.3.2 The Reactor Coolant System vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent (s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

CATAWBA - UNIT 1 3/4 4-37 Amendment No.

151 l

REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.10.

APPLICABILITY: All MODES.

ACTION:

a.

With the. structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.

b.

With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.

With the structural integrity of any ASME Code Class 3 component (s) c.

not conforming to the above requirements, restore the structural

~

integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.

SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

CATAWBA - UNIT 1 3/4 4-38 Amendment No.

151

[

j

p REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 At least one Reactor Coolant System vent path consisting of at least two valves in series powered from emergency buses shall be OPERABLE and closed

at each of the following locations:

a.

Reactor Vessel Head b.

Pressurizer steam space APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one of the above Reactor Coolant System vent paths inoper-a.

able, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 1

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

b.

With both of the above Reactor Coolunt System vent paths inoper-able, maintain the inoperable vent paths closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.11 Each Reactor Coolant System vent path shall be demonstrated OPERABLE i

at least once per 18 months by:

a.

Verifying all manual isolation valves in each vent path are locked in the open position, and b.

Cycling each valve in the vent path through at least one i

complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING.

For the plants using power operated relief valve (PORV) as a vent path, PORV block is not required to be closed if the PORV is operable.

CATAWBA - UNIT 1 3/4 4-39 Amendment No. 1 51 l

DESIGN FEATURES CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material of which 102 inches shall be 100% boron carbide and remaining 40-inch tip shall be 80% silver,15% indium, and 5% cadmium.

All control rods shall be clad with stainless steel tubing.

l 5.4 REACTOR COOLANT SYSTEM i

DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a.

In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the l

applicable Surveillance Requirements, l

b.

For a pressure of 2485 psig, and c.

For a temperature of 650*F, except for the pressurizer which is

)

680*F.

i VOLUME i

5.4.2 The total water and steam volume of the Reactor Coolant System is 1

13,050

100 cubic feet at a nominal T,y, of 525*F.

5.5 METEOROLOGICAL TOWER LOCATION' 5.5.1 The meteorological tower shall be located as shown in Figure 5.1-1.

5.6 FUEL STORAGE CRITICALITY 5.6.1

a. The spent fuel storage racks are designed and shall be maintained with:

1) k,,5 0.95 if fully flooded with unborated water as d,escribed in Section 9.1 of the FSAR; and 2)

A nominal 13.5" center to center distance between fuel assemblies placed in the spent fuel storage racks.

b. The new fuel storage racks are designed and shall be maintained with:

1) k,,5 0.95 if fully flooded with unborated water as d,escribed in Section 9.1 of the FSAR; and 2) k,,5 0.98 if moderated by aqueous foam as described in S,ection 9.1 of the FSAR; and i

CATAWBA - UNIT 1 5-7 Amendment No.

151

1 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

7.

DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 4.7.13.3 - Standby Makeup Pump Water Supply Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3.9.12 - Spent Fuel Pool Boron Concentration.)

8.

DPC-NE-3002, Rev. 1, "FSAR Chapter 15 System Transient Analysis Methodology," SER dated December, 1995.

(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits) 9.

DPC-NE-3000P, Rev. 1, " Thermal-Hydraulic Transient Analysis Methodology,"

SER dated December,1995.

(Modeling used in the system thermal-hydraulic analyses)

10. DPC-NE-1004A, " Design Methodology Using CASMO-3/ Simulate-3P," November 1992.

~(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)

11. DPC-NE-2004P-A, " Duke Power Company McGuire and Catatbe Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC Proprietary).

(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3

- Nuclear Enthalpy Rise Hot Channel Factor Fag (X,Y).)

12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel," October 1990 (DPC Proprietary).

(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints.)

13. DPC-NE-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,"

February 1995 (DPC Proprietary).

(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor)

CATAWBA - UNIT 1 6-22 Amendment No.151

?

REACTOR COOLANT SYSTEM BASES j

RELIEF VALVES (Continued) l of PORVs to control reactor coolant system pressure except for limited periods 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 i

and is fully capable of being returned to its normal alignment at any time, provided that this evolution is covered by an approved procedure. This is a i

function that reduces challenges to the code safety valves for I

overpressurization events. 5) Manual control of a block valve to isolate a 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 1

supply line operate satisfactorily and that the nonsafety portion of the j

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-l 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 i

inspection of steam generator tubing also provides a means of characterizing j

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 chemistry limits found to result in 4

negligible corrosion of the steam generator tubes.

If the secondary coolant 4

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).

i CATAWBA - UNIT 1 B 3/4 4-3 Amendment No.151 l

il, REACTOR COOLANT SYSTEM l

BASES t

~ 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 i

imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 150 gallons per day per steam generator can readily be detected.

Leakage in excess of this limit will require plant shutdown 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. However, even if a defect should develo) in service, it will be found during scheduled inservice steam generator tu)e examinations.

Plugging will be required for all tubes with imperfections exceeding the

)

plugging limit of 40% of the tube nominal 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 i

original tube wall thickness.

l Whenever the results of any steam generator tubing inservice inspection l

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 i

result in a requirement for analysis, laboratory examinations, tests, addi-tional eddy-current inspection, and revision of the Techni~ cal Specifications, if necessary.

l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS l

The 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 i

1973.

i l

l l

i i

CATAWBA - UNIT 1 B 3/4 4-4 Amendment No. 151 l

_

ML20117G846

Text

Attachment:

Dearadation means a service-induced cracking,

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