Text

Amend 56 to License NPF-12,consisting of Administrative Changes & Functional Definition Clarifications
	ML20207J833
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Site:	Summer
Issue date:	12/16/1986
From:	Rubenstein L; Office of Nuclear Reactor Regulation
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	NUDOCS 8701080615
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o UNITED STATES

. [ ' 3 s ( [,g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 g

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SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. S!!MMER NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 56 License No. NPF-17 1.

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

A.

The application for amendment by South Carolina Electric A Gas Company and South Carolina Public Service Authority (the licensees) dated June 20, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations 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 Comission; C.

There is reasonable assurance (1) 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 conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and j

E.

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

2.

Accordingly, the license is amended by 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-12 is hereby amended to read as follows:

)

8701030615 861216 PDR ADOCK 05000395 P

PDR

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 56

, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION A b., a f. L.

9:

}% ss Lester. S. hubenstein, Director PWR Pro,iect Directorate #2 Division of PWR Licensino-A Office of Nuclear Reactor Reaulation

Attachment:

Changes to the Technical Specifications Date of Issuance: December 16, 1986 1

ATTAC4WFNT TO LICENSE AMEND *ENT AMENDMENT NO. 56 TO FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-395 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 indicatina the areas of change. Corresponding overleaf pages are also provided to maintain document completeness.

Renove Pages Insert Pages 3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-37 3/4 3-37 3/4 3-71 3/4 3-71 3/4 3-72 3/4 3 72 3/4 3-79 3/4 3-79 3/4 5 0 3/4 5 4 R3/4 2-5 B3/4 2-5 B3/4 2-6

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i

.I TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

m.

m i

g TRIP q

ANALOG ACTUATING MODES FOR CHANNEL DEVICE MASTER SLAVE WHICH g

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED 1.

SAFETY INJECTION, REACTOR TRIP t

FEEDWATER ISOLATION, CONTROL ROOM ISOLATION START DIESEL GENERATORS, CONTAINMENT COOLING FANS AND ESSENTIAL SERVICE WATER a.

Manual Initiation N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4 y

b.

Automatic Actuation N.A.

N.A.

M(1)

Q 1, 2, 3, 4 Logic and Actuation Relays

[

c.

Reactor Building S

R M

N.A.

N.A N.A.

N.A.

1, 2, 3 Pressure-High-1 i

]

=

l d.

Pressurizer Pressure--Low S R

M N.A N.A.

N.A.

1,2,3 9

i o.

Differential Pressure S

R M

N.A.

1,2,3 Between Steam Lines--High j

f.

Steam Line Pressure Low S

R M

N.A.

1, 2, 3 i

I2.

REACTOR BUILDING SPRAY a.

Manual Initiation N.A.

N.A.

N. A.

R N.A.

N.A.

1,2,3,4 g

b.

Automatic Actuation N.A.

N.A.

M(1)

Q~

1,2,3,4 l

Logic and Actuation i

j g

Relays l

c.

Reactor Building S

R M

N.A.

1, 2, 3 l

)

Pressure-High-3 j

?

0 i

.n..

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS y,

=

a TRIP gg ANALOG ACTUATING MODES FOR

e

~*

CHANNEL DEVICE MASTER SLAVE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED 3.

CONTAINMENT ISOLATION 4

a.

Phase "A" Isolation

1) Manual N.A.

N.A.

R N.A.

N.A.

1,2,3,4 l

2) Safety Injection See 1 above for all Safety Injection Surveillance Requirements

3) Automatic Actuation N.A.

N.A.

M(1)

Q 1,2,3,4 t[

Logic and Actuation Relays

)l b.

Phase "B" Isolation

1) Automatic Actuation N.A.

N.A.

M(1)

Q 1, 2, 3, 4 Logic and Actuation Relays

2) Reactor Building S

R M

N.A.

1, 2, 3 Pressure-High-3

-c.

Purge and Exhaust Isolation

1) Automatic Actuation N.A.

N.A.

M(1)

Q 1, 2, 3, 4 Logic and Actuation g-Relays

(

2) Containment Radio-S R

M N.A.

N.A.

1, 2, 3, 4 g

activity-High 5

3) Safety Injection See 1 above for all Safety Injection Surveillance Requirements.

I.

l

i TABLE 4.3-2 (Centinued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENIS

o e

TRIP g

ANALOG ACTUATING MODES FOR

~

CHANNEL DEVICE MASTER SLAVE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED i

4.

STEAM LINE ISOLATION

c. ' Manual N.A.

N.A.

NA.

R N.A.

N.A.

1,2,3 b.

Automatic Actuation Logic N.A.

N.A.

M(1)

Q 1, 2, 3 and Actuation Relays c.

Reactor Building Pres-S R

H N.A.

N.A.

1,2,3 y

sure-High-2

[

d.

Steam F1ow in Two Steam S

R M

N.A.

1,2,3 g,

Lines--High Coincident S

R M

N.A.

1,2,3 With T,,g,_ Low-Low w

5.

TURBINE TRIP AND FEE 0 WATER ISOLATION c.

Steam Generator Water S

R M

N.A.

1, 2 Level--High-High b.

Automatic Actuation N.A.

N.A.

M(1)

Q 1, 2 Logic and Actuation Relay 6.

EMERGENCY FEEDWATER c.

Manual N.A.

N.A.

R N.A.

N.A.

1, 2, 3 3

b.

Automatic Actuation N.A.

N.A.

M(1)

Q 1, 2, 3

{

Logic and Actuation Relays Po c.

Steam Generator Water S

R M

N.A.

1,2,3 5

Level--Low-Low 5

i

TABLE 4.3-2 (Continued) 9 ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRENENTS E

TRIP

~

ANALOG ACTUATING N00ES FOR CHANNEL DEVICE NASTER SLAVE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED AUXILIARY FEEDWATER (Continued) d.

Undervoltage - ESF N.A.

R N.A.

N.A.

1, 2, 3 e.

Safety Injection See 1 above for all Safety Injection Surveillance Requirements f.

Trip of Main Feedwater N.A.

N.A.

R N.A.

N.A.

1, 2 Pumps w

s 1

[g.

Suction transfer on S

R N

N.A.

1,2,3 low pressure cn 7.

LOSS OF POWER a.

7.2 kV Emergency Bus N.A.

R N.A.

N.A.

1,2,3,4 Undervoltage (Loss of Voltage) b.

7.2 kV Emergency Bus N.A.

R N.A.

N.A.

1, 2, 3, 4 Undervoltage (Degraded Voltage) 8.

AUT0NATIC SWITCNOVER TO CONTAlff9ENT SUNP a.

RWST level low-low S

R N

N.A.

1, 2, 3 b.

Automatic Actuation N.A.

N.A.

N(1)

Q.

1, 2, 3 Logic and Actuation Relays

TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

=

ANALOG CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL gy INSTRUMENT CHECK CHECK CALIBRATION TEST l

1.

GROSS BETA OR Gale 9A RADIOACTIVITY MONITORS l

PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE l

a. Liquid Radwaste Effluent Line - RM-L5, RM-L9 D

P R(2)

Q(1)

b. Nuclear Blowdown Effluent Line - RM-L7 D

P R(2)

Q(1)

c. Steam Generator Blowdown Effluent Line -

D M

R(2)

Q(1)

RM-L3, RM-LIO

,s*

d. Turbine Building Sump Effluent Line - RM-L8 0

M R(2)

Q(1) 4 w

e. Condensate Demineralizer Backwash Line RM-L11 D M

R(2)

Q(4) 2.

FLOW RATE MEASUREilENT DEVICES i

a. Liquid Radwaste Effluent Line D(3)

N.A.

R Q

b. Penstocks Minimum Flow Interlock D(3)

N.A.

R Q

c. Nuclear Blowdown Effluent Line D(3)

N.A.

R Q

g I

y

d. Steam Generator Blowdown

'O(3)

N.A.

R Q'

Effluent Line f+

y 3.

TANK LEVEL INDICATING DEVICES

a. Condensate Storage Tanks D

N.A.

R Q

INSTRUMENTATION TABLE 4.3-8 (Continued)

TABLE NOTATION (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Loss of Power (alarm only).

3.

Low flow (alarm only).

4.

Instrument indicates a downscale failure (alarm only).

5.

Normal /8ypass switch set in Bypass (alarm only).

6.

Other instrument controls not set in operate mode.

I (2) The initial CHANNEL CALIBRATION shall be performed using one or more of l

the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods I

of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

(4) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic I

isolation of this pathway and local panel alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Loss of Power (alarm only).

3.

Low flow (alarm only).

4.

Instrument indicates a downscale failure (alarm only).

5.

Normal / Bypass switch set in Bypass (alarm only).

6.

Other instrument controls not set in operate mode.

SUMMER - UNIT 1 3/4 3-72 Amendment No.20.56

INSTRUMENTATION TABLE 4.3-9 (Continued)

TABLE NOTATION At all times.

During waste gas holdup system operation (treatment for primary system offgases).

(1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Loss of Power (alarm only).

3.

Low flow (alarm only).

4.

Instrument indicates a downscale failure (alarm only).

5.

Normal / Bypass switch set in Bypass (alarm only).

6.

Other instrument controls not set in operate mode.

(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm setpoint.

2.

Loss of Power.

3.

Low flow.

4.

Instrument indicates a downscale failure.

5.

Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.

These standards shall permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1.

1500 1 30 ppm hydrogen, balance nitrogen, for the outlet hydrogen monitor and 2.

4 1 0.1 volume percent hydrogen, balance nitrogen, for the inlet hydrogen monitor.

(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1.

75 i 1.5 ppm oxygen, baic,ce nitrogen, for the outlet oxygen monitor and 2.

3.5 1 0.1 volume percent oxygen, balance nitrogen, for the inlet oxygen monitor.

SUMMER - UNIT 1 3/4 3-79 Amendment No. 7,56

=

INSTRUMENTATION LOOSE-PART DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The loose par't detection system sha'11 be OPERABLE.

APPLICABILITY: MODES 1 and 2 ACTION:

a.

With one or more loose part detection system channels inoperable for more than 30 days, prepare and submit a Special Report to the Commis-sion pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the charinel(s) to OPERABLE status, b.

,The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each channel of the loose part detection system shall be demonstrated OPERABLE by performance of:

a.

A CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , bl An ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and c.

A CHANNEL CALIBRATION at least once per 18 months.

SUMMER - UNIT 1 3/4 3-80

,-n-----.--

t EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

l a.

A minimum contained borated water volume of 453,800 gallons, b.

A boron concentration of between 2000 and 2100 ppm of beron, and c.

A minimum water temperature of 40*F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or 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 .

SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

l At least once per 7 days by:

a.

1.

Verifying the contained borated water volume in the tank, and 2.

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is less than 40*F.

SUMMER - UNIT 1 3/4 5-9 Amendment No.##.56

POWER DISTRIBUTION LIMIT BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

F limit for Rated Thermal Power (FRTP) as provided in the Radial Peaking xy x

Factor Limit Report per specification 6.9.1.11 was determined from expected power control maneuvers over the full range of burnup conditions in the core.

When RCS flow rate and F are measured, no additional allowances are q

necessary prior to comparison with the limits of Figures 3.2-3.

Measurement N

errors of 3.5% for RCS total flow rate and 4% for F have been allowed for AN in determining the limits of Figure 3.2-3.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3.

3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup l

testing and periodically during power operation.

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts.

A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted. The limit of 1.02 was selected to provide an allowance for q

the uncertainty associated with the indicated power tilt.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by g

reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of 4 symmetric thimbles. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.

The limits are consistent i

SUMMER - UNIT 1 B 3/4 2-5 Amendment No. H,56

}

POWFR DISTRIBUTION LIMIT BASES l

HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTAALPY RISE HOT CHANNEL FACTOR (Continued) with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

l i

i t

i i

I SUMMER - UNIT 1 8 3/4 2-6 Amendment No.56

.___

ML20207J833

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