ML063000024

Tech Spec Pages for Amendments 250 and 249 Regarding Implementation of Generic Safety Issue 191

ML063000024
Person / Time
Site:	Surry
Issue date:	10/12/2006
From:	NRC/NRR/ADRO/DORL/LPLII-1
To:
Martin R, NRR/DORL, 415-1493
References
GSI-191, TAC MC9724, TAC MC9725
Download: ML063000024 (10)

Text

3. This renewed license shall be deemed to contain and Is subject to the conditions specified Inthe following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 7.0.32 of 10 CFR Part 70; and is subject to all.applicabie provisions of the Act and the..

rules, regulations, and orders of the Commission now or hereafter Ineffect; and Is subject to the additional conditions specified below:

A. Maximum Power Level The licensee Is authorized to operate the facility at steady slate reactor core power levels not in excess of 2546 miegawatts (thermal).

B. Technical Specifications The Technical Specifications.contalned InAppendix A, as revised through Amendment No. 250 are hereby Incorporated Inthe renewed license.

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

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facility operating records In accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 I. Fire Protection The licensee shall implemeht and maintain In effect the provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report and as approved Inthe SER dated September 19, 1979, (and Supplements dated May 29, 1980, October 9, 1980, December 18, 1980, February 13, 1981, December 4, 1981, April 27, 1982, November 18, 1982, January 17, 1984, February 25, 1988, and Renewed License No. DPR-32 Amendment No. 250

• E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

. This renewed license shall be deemed to contain and Is subject to.the conditions.

specified In the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 OFR Part 30, Section 40.41 of IOCFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and Is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter In effect; and Is subject to the additional conditions specified below:.

A. Maximum Power Level The licensee Is authorized to operate the faclilty at steady stale reactor core power levels not In excess of 2548 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 249'are hereby Incorporated in this renewed license.

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

0. Reports The licensee shall make certain reports In accordance with the requirements of the Technical Specifications.

D. -Records The licensee shall keep facility operating records Inaccordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment 65 G. Deleted by Amendment 227 H. Deleted by Amendment 227 Renewed License Amendment No, No, DPR-37 249

TS 3.4-3 Basis The spray systems in each reactor unit consist of two separate parallel Containment Spray Subsystems, each of 100 percent capacity, and four separate parallel Recirculation Spray Subsystems, each of 50 percent capacity.

Each Containment Spray Subsystem draws water independently from the refueling water storage tank (RWST). The water in the tank is cooled to 45TF or below by circulating the water through one of the two RWST coolers with one of the two recirculating pumps. The water temperature is maintained by two mechanical refrigerating units as required. In each Containment Spray Subsystem, the water flows from the tank through an electric motor driven containment spray pump and is sprayed into the containment atmosphere through two separate sets of spray nozzles. The capacity of the spray systems to depressurize the containment in the event of a Design Basis Accident is a function of the pressure and temperature of the containment atmosphere, the service water temperature, and the

• temperature in the refueling water storage tank as discussed in the Basis of Specification 3.8.

Each Recirculation Spray Subsystem draws water from the common containment sump. In each subsystem the water flows through a recirculation spray pump and recirculation spray cooler, and is sprayed into the containment atmosphere through a separate set of spray nozzles. Two of the recirculation spray pumps are located inside the containment and two outside the containment in the containment auxiliary structure.

With one Containment Spray Subsystem and two Recirculation Spray Subsystems operating together, the spray systems are capable of cooling and depressurizing the containment to 1.0 psig in less than 60 minutes and to subatmospheric pressure within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following the Design Basis Accident. The Recirculation Spray Subsystems are capable of maintaining subatmospheric pressure in the containment indefinitely following the Design Basis Accident when used in conjunction with the Containment Vacuum System to remove any long term air inleakage. The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 1.0 psig (from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) and is maintained less than 0.0 psig (after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).

Amendment Nos. 250 249

TABLE 3.7-2 (Continued)

ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDITIONS Minimum Total Number OPERABLE Chi inels Permissible Operator Functional Unit Of Channels Channels To' Trip Bypass Conditions Actions

3. AUXILIARY FEEDWATER (continued)

e. Trip of main feedwater pumps - start motor 2/MFW 1/MFW pump 2-1 each 24 driven pumps pump MFW pump

f. Automatic actuation logic 2 2 1 22

4. LOSS OF POWER

a. 4.16 kv emergency bus undervoltage (loss of 3/bus 2/bus 2/bus 26 voltage)

b. 4.16 kv emergency bus undervoltage (degraded 3/bus 2/bus 2/bus 26 voltage)

5. NON-ESSENTIAL SERVICE WATER ISOLATION

a. Low intake canal level 4 3 3 20

b. Automatic actuation logic 2 2 1 14

6. ENGINEERED SAFEGAURDS ACTUATION INTERLOCKS - Note A

a. Pressurizer pressure, P- 11 3 2 2 23

b. Low-low Tavg, P-12. 3 2 2 23

c. Reactor trip, P-4 2 2 1 24

7. RECIRCULATION MODE TRANSFER

a. RWST Level - Low-Low 4 3 2 25 I

b. Automatic Actuation Logic and Actuation 2 2 1 14 Relays 0 8. RECIRCULATION SPRAY

a. RWST Level - Low Coincident with High High 4 3 2 20 Fs Containment Pressure -3 C4

b. Automatic Actuation Logic and Actuation 2 2 1 14 Relays CN)

Note A - Engineered Safeguards Actuation Interlocks are described in Table 4.1-A

TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING No. Functional Unit Channel Action Setting Limit 6 AUXILIARY FEEDWATER

a. Steam Generator Water Level Low-Low Aux. Feedwater Initiation > 14.5% narrow range S/G Blowdown Isolation

b. RCP Undervoltage Aux. Feedwater Initiation > 70% nominal

c. Safety Injection Aux. Feedwater Initiation All S.I. setpoints

d. Station Blackout Aux. Feedwater Initiation > 46.7% nominal

e. Main Feedwater Pump Trip Aux. Feedwater Initiation N.A.

7 LOSS OF POWER

a. 4.16 KV Emergency Bus Undervoltage Emergency Bus Separation > 2975 volts and <3265 volts with a 2 (Loss of Voltage) and Diesel start (+5, -0.1) second time delay

b. 4.16 KV Emergency Bus Undervoltage Emergency Bus Separation > 3830 volts and < 3881 volts with a 60 (Degraded Voltage) and Diesel start (+/-3.0) second time delay (Non CLS, Non SI) 7 (+/-0.35) second time delay (CLS or SI Conditions) 8 NON-ESSENTIAL SERVICE WATER ISOLATION

a. Low Intake Canal Level Isolation of Service Water 23 feet-6 inches flow to non-essential loads 9 RECIRCULATION MODE TRANSFER z a. RWST Level-Low-Low Initiation of Recirculation _ 11.25% I 0 Mode Transfer System <15.75%

10 TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water Level High-High Turbine Trip < 80% narrow range 0-o.

Feedwater Isolation tJ.q 11 RWST Level Low (coincident with High High Containment Pressure)

Recirculation Spray Pump Start

> 59%

<61% I t-En

TS 3.8-4 (3) assuring that environmental conditions will not preclude access to close the valves and

4) that this administrative or manual action will prevent the release of radioactivity outside the containment.

The Reactor Coolant System temperature and pressure being below 350'F and 450 psig, respectively, ensures that no significant amount of flashing steam will be formed and hence that there would be no significant pressure buildup in the containment if there is a loss-of-coolant accident. Therefore, the containment internal pressure is not required to be subatmospheric prior to exceeding 350'F and 450 psig.

The allowable value for the containment air partial pressure is presented in TS Figure 3.8-1 for service water temperatures from 25 to 95'F. The RWST water shall have a maximum temperature of 450 F.

The horizontal upper limit line in TS Figure 3.8-1 is based on MSLB peak calculated

,pressure criteria, and the sloped line from 70'F to 95°F service water temperatures is based on LOCA depressurization criteria.

Amendment Nos. 250, 249

TS 3.8-5 If the containment air partial pressure rises to a point above the allowable value thereactor shall be brought to the HOT SHUTDOWN condition. If a LOCA occurs at the time the containment air partial pressure is at the maximum allowable value, the maximum containment pressure will be less than design pressure (45 psig), the containment will depressurize to 1.0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0.0 psig within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 1.0 psig for the interval from 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following the Design Basis Accident.

If the containment air partial pressure cannot be maintained greater than or equal to the minimum pressure in Figure 3.8-1, the reactor shall be brought to the HOT SHUTDOWN condition. The shell and dome plate liner of the containment are capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia.

References UFSAR Section 4.2.2.4 Reactor Coolant Pump UFSAR Section 5.2 Containment Isolation UFSAR Section 5.2.1 Design Bases UFSAR Section 5.2.2 Isolation Design UFSAR Section 5.3.4 Containment Vacuum System Amendment Nos. 250, 249

TS Figure 3.8-1 SURRY TECHNICAL SPECIFICATION CURVE FOR CONTAINMENT ALLOWABLE AIR PARTIAL PRESSURE INDICATION VS. SERVICE WATER TEMPERATURE 11.6 11.4

-(70, 11.3) 11.2 ACCEPTABLE OPERATION INSIDE THE LINES

0. 11.0

• 10.8 CONTAINMENT TEMPERATURE BETWEEN 75 F AND 125 F a- 10.6 I ....... _____ ____

N R.

10.4 .1 _________ II I .1 1 4- I- 4 4> 4 -I- 4 (95, 10.3)

N 10.2 -- (25, 10.3) 4 1 (70, 10.1) z0

~L I .1

- F ~ - b- i - -- (95,101) 10.0 .. ..1 1 1 .1 ..... .I ...... . ...... ..... I 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Service Water Temperature, deg-F

TS 3.19-2 If the requirements of Specification 3.19.B.1, 3.19.B.2, or 3.19.B.3 are not met within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after achieving HOT SHUTDOWN, both units shall be placed in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Basis Following a design basis accident, the containment will be depressurized to 1.0 psig in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and to subatmospheric pressure within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 1.0 psig for the interval from 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following the Design Basis Accident. Beyond 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, containment pressure is assumed to be less than 0.0 psig, terminating leakage from containment. The main control room is maintained at a positive differential pressure using bottled air during the first hour, when the containment leakrate is greatest.

The main control room is contained in the control room pressure boundary or envelope, which is defined in the Technical Specification 3.23 Basis.

The control room pressure boundary is permitted tobe opened intermittently under administrative control without declaring the boundary inoperable. The administrative control must provide the capability to re-establish the control room pressure boundary. For normal ingress into and egress from the pressure boundary, the individual entering or exiting the area has control of the door.

Amendment Nos.250, 249

TABLE 4.1-1 (Continued)

MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description Check Calibrate Test Remarks

10. Rod Position Bank Counters S(1,2) N.A. N.A. 1) Each six inches of rod motion when data logger is out Q(3) of service

2) With analog rod position

3) For the control banks, the benchboard indicators shall be checked against the output of the bank overlap unit.

11. Steam Generator Level S R Q

12. Deleted

13. Deleted

14. Deleted

15. Recirculation Mode Transfer

a. Refueling Water Storage Tank S R Q Level-Low-Low I

b. Automatic Actuation Logic and N.A. N.A. M Actuation Relays

16. Recirculation Spray Pump Start II

a. RWST Level-Low S R Q

17. Reactor Containment Pressure-CLS *D R Q(1) 1) Isolation valve signal and spray signal

18. Deleted z 19. Deleted

20. Deleted 4_

21. Deleted

22. Steam Line Pressure S R Q -H