ML19352B326
| ML19352B326 | |
| Person / Time | |
|---|---|
| Site: | Surry  (DPR-32-A-071, DPR-32-A-71, DPR-37-A-071, DPR-37-A-71) |
| Issue date: | 06/23/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19352B327 | List: |
| References | |
| NUDOCS 8107020160 | |
| Download: ML19352B326 (30) | |
Text
,
i j
i
\\,)
.. #e a u:n,'o ss UNITED STATES Bky,q h Eh NUCLEAR REGULATORY COMMISSION l
<a ;
WASHINGT ON. D. C. 20555 kk' '$b#,p/ '
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 71 License No. OPR-32 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated May 19, 1981, complies with~the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's, rules and regulations set forth in 10 CFR Chapter I; i
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Co mission; 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 conducted in compliance with the Commission's' regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; 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.
e 6
O lg107020 @
w---,-
e
t 2-
~
2.
Accordingly, Facility Operating License No. DPR-32 is hereby amended by deleting paragraph 3.F. by revising paragraph 3.B to read as follows, and by changing the Technical Specifications as indicated in the attach-ment to this license amendment.
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 71, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
F THE NUC R EGULATORY COMMISSION k Nr ie e
/ Operating _ Reactors ranch #1
- s Division of Licen 19
Attachment:
Changes to the Technical Specifications Date of Issuance: June 23, 1981 f
l l
+
y--
-+p-
,we yg
--eg
-_.-.*-,e.v-
, wa w,--p.-rP-M-
,I
,',o UNITED STATES g
,y c(
g NUCLEAR REGULATORY COMMISSION
-.E WASHINGTON, D. C. 20666 Q
f s
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 71 License No. DPR-37 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licen Se) dated May 19, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations j
set forth in 10 CFR Chapter I; B.
Th-facility will operate in conformity with the ap' plication, tiie provisions of the Act, and the rules and regulations of the Comission; f
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 conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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.
t 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-37 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 71, are hereby incorporated in the licens's. The licensee shall operate the facility in accordane.e with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
Fh(;THE NUCA R GULATORY COMMISSION Axjsat 3:even A. Yarga, Chi.e"
'Operat.i.ng Reactors inch #1 Division 6f Licensi
Attachment:
Changes to th, Technical Specifications Date of Issuance:
June 23,1981 i
i e
J l
I t
l L
1 ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 71 TO FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 71 TO FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:
Remove Pages Insert Pages DPR-32 DPR-37 DPR-32 & DPR-37 i
3.3-1 3.3-1 3.3-1 3.3-2 3.3-2 3.3-2 3.3-3
.3.3-3 3.3-3 3.3-4 3.3-4
- 3. 3-4' 3.3-5 3.3-5 3.3-5 3.3-6 3.3-6 3'.3-6 3.3-7 3.3-7 3.3-7 3.3-8 3.3-8 3.3-8 3.3-9 3.3-9 3.3-9 3.4-1 3.4.2-1 3.4-1 3.4-2 3.4.2-2 3.4-2 3.4-3 3.4.2-3 3.4-3 3.4-4 3.4.2-4 3.4-4
~
3.4-5 3.'4.2-5
~3.4-5 3.4-6 3.6-3 3.6-3 3.6-3 3.8-1 3.8.2-1 3.8-1 3.8-2 3.8.2-2 3.8-2 4
3.8-3 3.8.2-3 3.8-3 3.8-4 3.8.2-4 3.8-4 T.S. Figure 3.8-1 T.S. Figure 3.8.2-1 T.S. Figure 3.8-1 T.S. Figure 3.8.2-1 (cont)
T.S. Figure 3.8-1 (cont) 4.1 -7 4.1 -7 4.1-7 4.5-6 4.5-6 4.11-5 4.11 -5
a f
TS 3.3.-l 3.3 SAFETY INJECTION SYSTEM Applicability Applies to the operating status of the Safety Injection System.
Objective To define those limiting conditions for cperation that are necessary to provide sufficient borated cooling water to remove decay heat from the i
core in emergency situations.
Specifications A.
A reactor shall not be made critical un1.ess,the following conditions are met:
The refueling water storage tank contains not less than 387,100 gal of borated water. The baron concentration shall be at least 2000 ppm and not greater than 2200 ppa.
2.
Eaca accumulator system is pressurized to at leist 600 psia and 3
3 contains a minimum of 975 ft and a maximum of 389 ft of borated water with a boron concentration of at least 1950 ppe.
~
3.
The boron injection tank and isolated portion of the inlet and outlet piping contains no less than 900 gallons of water with a boron concentration equivalent to at least 11.5%rto 13% weight boric acid solution at a temperature of at least 145'F.
Addi-t
. tionally, recirculation between a unit's Boron Injection Tank l
and the Boric Acid Tank (s) assigned to the unit shall be main-l
{
tained.
Amendment Nos. 71 & 71
=
l i
I TS 3.3-2 4.
Two channels of heat tracing shall be available for the flow paths.
5.
Two charging pumps are operable.
6.
Two low head safety injection pumps are operable.
7.
All valves, piping, and interlocks associated with the above components which are required to operate under accident condi-tions are operable.
8.
The Charging Pump Cooling Water Subsystem shall be operating as follows:
a.
Make-up water from the Component Cooling Water Subsystem shall be available.
b.
Two charging pump component cooling water pumps and two
., charging pump service water pumps shall be operable.
c.
Two charging pump intermediate seal coolers shall be operable.
9.
During power operation the A.C. power shall be removed from the following motor operated valves with the valve in the open position:
Unit No. 1
' Unit No. 2
~
10.
During power operation the A.C. power shall be removed from the following motor operated valves with the valve in the closed position:
Unit No. 1 Unit No. 2 l
MOV 1869B MOV 2869B MOV 1890A MOV 2890A MOV 1890B MOV 2890B
TS 3.3-3 11.
The accumulator discharge valves listed below in non-isolated loops shall be blocked open by de-energiziag the valve motor operator when the reactor coolant system pressure is greater than 1000 psig.
Unit No. 1 Unit No. 2 MOV 1865A MOV 2865A MOV 1865B MOV 2865B MOV 1865C MOV 2865C 12.
Power operation with less than three loops in service is pro-hibited. The following loop isolation valves shall have AC
~
power rer~.3 and be locked in open position daring power operation.
Unit No. 1 Unit No. 2 s
MOV 1590 MOV 2590 MOV 1591 MOV 2591 l
MOV 1594 MOV 2594 MOV 1595 MOV 2595 1
13.
The total system uncollected leakage from valves, flanges, and pumps located outside containment shall not exceed the limit shown in Table 4.11-1 as verified by inspection during system testing.
Individual component leakage may exceed the design value given in Table 4.11-1 provided that the total allowable
~
system uncollected leakage is not exceeded.
Amendment Nos. 71 & 71
s TS 3.3-4 B.
The requirements of Specification 3.3-A may be modified to allow one of the following components to be inoperable at any one time.
If the system is ot restored te meet tae requirements of Specification 3.3-A within the time period specified, the reactor shall initirlly be placed in the shutdown condition.
If the requirements of Specification 3.3-A are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the reacto-shall be placed in cold shutdown condition.
1.
Que accumulator may be isolated for a period not to exceed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
2.
Two charging pumps per unit may be out service, provided immediate attention is directed to making repairs and one pump is rescored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
One,loy head safety injection pump per unit may be out f service, provided immediate attention is directed to making repairs and the pump is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The other low head safety injection pump shall be terted to demonstrate operability prior to initiating repair of the inoperable pump and shall be tested once every eight (8) hours thereafter, until both pumps are in an operable status or' the reactor is shutdown.
~
4.
Any one valve in the Safety Injection System may be inoperable
~
provided repairs are initiated immediately and are completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Prior to initiating repairs, all automatic valves in the redunda'.t system shall be tested to demonstra'.e operability.
5.
One channel of heat tracing may be inoperable for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, provided immediate attention is directed to making repairs.
Amendment Nas. 71 & 71
TS 3.3-5 6.
One charging puan component cooling water pump or one charging pump service water pump may be out of service provided,the pump is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
7.
One charging pump intermediate seal cooler or other passivc component may be out of service provided the system may still operate at 100 percent capacity and sepairs are completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
8.
Power may be restored to any valve referenced in Specifications 3.3.A.9 and 3.3.A.10 for the purpose of valve testing or maintenance provided that no more than one valve has power restored and provided that testing and maintenance is completed and power removed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
9.
Power may be restored to any valve referenced in Specification 3.3.A.11 l for Uh6 purpose bf ' valve testing or maintenance provided that no more than one valve has power restored and provided that testing or maintenance is completed and power removed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
10.
Recirculation between a unit's Boron Injection Tank and the Boric Acid Tank (s) assigned to the unit may be terminated for a period not to exceed two hours, provided all other parameters (temperatures, boron concentration, volume) of the Boron Injee-tion Tank are within Specification 3.3.A.3 and immediate l'
attention is directed to making repairs.
11.
The total uncollected syster leakage for valves, flanges, and pumps located outside containment can exceed the limit shown in Table 4.11-1 provided immediate attention is directed to making l
repairs and system leakage is returned to within limits within 7 days.
Amendment Nos. 71 & 71
TS 3.3-6
_ Basis The normal procedure for starting the reactor is, first, to heat the reactor coolant te near operating temperature by running the reactor coolant pumps. The reactor is then make critical by withdrawing control rods and/or diluting boron in the coolant. With this mode of startup the Safety In.iection System is required to be operable as specified. During low power physics tests there is a negligible ambunt of energy stored in the system; therefore an accident comparable in severity to the Design Basis Accident is not possible, and the full capacity of the Safety Injection System is not required.
The operabl,e status of the various systems and, components is to be,
demonstrated by periodic tests, detailed in TS Section 4.1.
A large I
fraction of these tests are performed while the reactor is operating in the power range.
If a component is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short time. A single component being inoperable does not negate the ability of the system to perform its function, but it reduces the redundancy provided in the re actor design and thereby limits the ability to tolerate additional equipment failures.
To provide maximum assurance that the redundant conponent(s) will operate if required to do so, the redundant component (s) are to be tested prior to-initiating repair of the inoperable component and, in some. cases are to be retested at intervals during the repair period.
In some cases, i.e.
charging pumps, additional components are installed to allow a component to be inoperable without affecting system redundancy. For those cases Amenduent Nos. 71 & 71
e s
TS 3.3-7 which are not so designed, if it deve! ops that (a) the inoperable component is not repaired within the specified allowable. time period, or (b) a second component in the same or related system is found to be inoperable, the reactor will initially be put in the hot shutdown condition to provide for reduction of the decay heat from the fuel, and consequent reduction of cooling requirements after a postulated loss-of-coolant accident. After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in the hot shutdown condition, if the malfunction (s) are not corrected the reactor will be placed in cold shutdown condition, following normal shutdown and cooldown procedures.
The Specification requires prompt action to effect repairs of an inoperable componeat, and therefore in most cases repairs will be completed in less than the s,pecified allowable repair times. Furthermore, the specified repair times do not apply to regularly scheduled maintenance of the Safety Injection System, which is normally to be performed during refueling shut-downs. The limiting times for repair are based on: estimates of the time required to diagnose and correct various postulated malfunctions using sa'fe and proper procedures, the availability of tools, materials and
~
~
equipment; health physics requirements and the extent to which other systems provide functions 1 redundancy to the system under repair.
Assuming the reactor has been operating at full rated-power for at least 100 days, the magnitude of the decay heat production decreases as follows after initiating hot shutdown.
Time After Shutdown Decay Heat, % of Rated Power 1 min.
3.7 30 min.
1.6
TS 3.3-8 l
Time After Shutdown Decay Heat, % of Rated Power I hour 1.3 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.75 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.48 Thus, the requirement for core cooling in case of a postulated loss-of-coolant accident while in the hot shutdown condition is reduced by orders of magnitude below the requirements for handling a postulated loss-of-coolant accident occurring during power operation. Placing and maintain-ing the reactor in the hot shutdown condition significantly reduces the potential consequences of a loss-of-coolant accident, allows access to some of the Safety Injection System components in order to effect repairs, and minim zes the exposure to thermal cycling.
Failure to complete repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to hot shutdown condition is considered indicative of unforeseen problens, i.e., possibly the need of major maintenance.
In such a case the reactor is to be put it.o the cold shutdown condition.
The accumulators are able to accept leakage from the Reactor Coolant System without any effect on their availability. Allowable inleakage is based on the volume of water that can be added to the initial amount without exceed-ing the volume given it Specification 3.3.A.2.
The maximum acceptable inleakage is 14 cubic feet per tank.
Amendment Nos. 71 & 71
1 0
TS 3.3-9 1
The accumula: ors (one for each loop). discharge into the cold leg of the j
reactor coolant piping when Reactor Coolant System pressure decreases j
below accumulator pressure, thus assuring rapid core cooling for large breaks. The line from each accumulator is provided with a motorized valve to isolate the accumulator during reactor start-up and shutdown to preclude the discharge of the contents of the accumulator when not required. These valves receive a signal to open when safety injection is initiated.
To assure that the accumulator valves satisfy the single failure criterion, they will be blocked open by de-energizing the valve motor operators when the reactor coolant press _ure, exceeds 1000 psig. The operating pressure of the Reactor Coolant System is 2235 psig and safety injection is initiated when this pre.ssure drops to 600 psia. De-energiz-ing the motor operator when the pressure exceeds 1000 psig allows sufficient time during normal startup operation to perform the actions required to de-energize the valve. This procedure will assure that there is an operable flow path from each accumulator to the Reactor Coolant System duriag power operation and that safety injection can be accom-plished.
The removal of power from the valves listed in the specification will assure that the systems of which they are a part satisfy the single failure criterion.
l Continuous recirculation between the Boron Injection Tank and the Boric Acid Tank (s) ensures that a unit's Boron Injection Tank is full of con-centrated boric acid at all times.
o TS 3.4-1 l
J l
3.4 SPRAY SYSTEMS Applicability Applies to the operational status of the Spray Systems.
Objective To define those conait.ons of the Spray Systems necessary to assure safe i
unit operation.
Specification A.
A unit's Reactor Coolant System temperature or pressure shall not be made to exceed 350'F or 450 psig, respectively, or the reactor shall not be made critic'al unless the following Spray System conditions in the unit are met:
1.
Two Containment Spray Subsystems, including containment spray pumps and motor drives, piping, and valves shall be operable.
2.
Four Recirculation Spray Subsystems, including recirculation spray pucys, coolers, piping, and valves shall be operable.
3.
The refueling water storage tank shall 'ce.r.ain not less than 387,10e gal and not greater than 398,000 gal of borated water at a maximum temperature as shown in TS Fig. 3.8-1 l
If this volume of water catnot be maintained by makeup, or the temperature maintained below that specified in TS Fig. 3.8-1, the reactor shall be shutdown until repairs can be made. The water shall be borated to a boron concentration not less than
/hmilfiirgi1RDft 71 &.71_
~.
S l
TS 3.4-2 2,000 ppa and not greater than 2,200 ppm which will assure that l
the reactor is in the refueling shutdown condition when all control rod assemblies are inserted.
4.
The refueling water chemical addition tank shall contain not less than 4,200 gal of r.olution with a sodium hydroxide concen-tretion of not less than 17 percent by weight and not greater than 18 percent by weight.
5.
All valves, piping, and interlocks associated with the above components which are required to operate under accident conditions shall be operable.
6.
The total uncollected system leakage from valves, flanges, and pumps located outside containment shall not exceed the limit shownin) Table 45-1asverifiedby.inspectionduringsystem testing.
Individual component leakege may exceed the design valuegiveninTable4.5-1providedthatthetotIlallowed system uncollected leckage is not Exceeded.
B.
During power operation the requirements of Specification 3.4-A may I,
~
be modified to allow the following components to be inoperable.
If the components are not restored to meet the requirements of Specifi-cation 3.4-A within the time period specified below, the reactor shall be placed in the hot shutdown condition.
If the requirements l
of Specification 3.4-A are not satisfied within an additional 48 l
~
hou'rs the reactor shall be placed in the cold stetdown condition using normal operating procedures.
i Amendment Nos. 71 & 'l l
TS 3.4-3 l
1.
One Containment Spray Subsystem may be out of service, provided Lunediate attention is directed to making repairs and the sub-system can be restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The other Containment Sprey Subsystem shall be tested as specified in Specification 4.5-A to demonstrate operability prior to initiating repair of the inoperable system.
2.
One outside Recirculation Spray Subsystem may be out of service provided immediate attention is directed to making repairs and the subsystem can be restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The other Recirculation Spray subsystem shall be tested as specified in Specification 4.5-A.to_ demonstrate operability prior to initiating repair of the inoperable system.
3.
One inside Recirculation Spray Subsystem may be out of service provided immediate attention is directed to making repairs and the subsystem can be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The other Recirculation Spray subsystems shall be tested as specified in Specification 4.5-A to demonstrate operability prior to initiating repair of the inope'rable subsystems.
4.
The total uncollected system leakage from valves, flanges, and pumps located outside containment can exceed the limit shown in Table 4.5-1 provided immediate attention is directed to making repairs and system leakage is returned to within limits within 7 days.
Amendment Nos. 71 & 71
~
..o TS 3.4-4 C.
Should the refueling water storage tank temperature fail to be main-tained at or below 45'F, the containment pressure and temperature shall be maintained in accordance with TS Fig. 3.8-1 to vaintain the cap-ability of the Spray System with the higher refueling water temperature.
If the containment temperature and. pressure cannot be maintained within the limits of TS Fig. 3.8-1, the reactor shall be placed in the cold shutdown condition.
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 I
capacity.,.
Each Containment Spray Subsystem draws water independently from the 398,000 gal. capacity refueling water storage tank. The water in the tank is cooled to 45'F or below by circulating the tank water through one of the two refueling water storage tank coolers through the use of one of
~
the two refueling vater 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 tanks as discussed in Specifi-cation 3.8-B.
TS 3.4-5 l
Each Recirculation Spray Subsystem draws water from the common containment pump.
In each subsystem the water flows through a recirculation spray pump and recirculation spray cooler, and is sprayed into the containment atmos-phere through a separate set of spray nozzles. Two of the recirculation spray pumps are located inside the containeent and two outside the contain-ment in the containment auxiliary structure.
With one Contsinment Spray Subsystem and two Recirculation Spray Sub-systems operating together, the Spray Systems are capable of cooling and depressurizing the containment to subatmospheric pressure in less than 60 minutes following the Design Basis Accident.
The Recirculation Spray Subsystems are capable of maintaining subatmospheric pressure in the con-tainment indefinitely following the Design Basis Accident when used in
,4 conjunction with the Containment Vacuum System to remove any long term air in leakage.
In addition.to supplying water tr the Containment Spray System, the refuel-ing water storage tank is aise a source of water for safety injection following an accident. This sater is borated to a concentration which assures reactor shutdown by apyroximately 10 percent ak/k when all control rod assemblies are inserted and then the reactor is cooled down for refueling.
Amendment Nos. 71 & 71 l
y
,.,.. - = -, -, -,, -..,. -
n
+,
TS 3.4-6 References FSAR Section 4 Reactor Coc'nnt System FSAR Section 6.3.1 Containment Spray Subsystem FSAR Section 6.3.1 Recirculation Spray Pumps and Coolers FSAR Section 6.3.1 Refueling Water Chemical Addition Tank FSAR Section 6.3.1 Refueling Water Storage Tank FSAR Section 14.5.2 Design Basis Accident FSAR Section 14.5.5 Cont,inment Transient Analysis t
L Amendment Nos. 71 & 71 l
f
Amendment Nos. 71 & 71 T.S. 3.6-3 450 psig, respectively, residual heat removal requirements are normally
' satisfied by steam bypass to the condenser.
If the condenser is unavail-able, steam can be released to the atmosphere through the safety valves, power operated relief valves, or the 4 inch decay beat release line.
The capability to supply feedwater to the generators is normally provided In the event of by the operation of the Condensate and Feedwater Systems.
complete loss of electrical power to the station, residual heat removal would continue to be assured by the availability of either the steam driven auxiliary feedwater pump or one of the motor driven auxiliary feedwater pumps and the 110,000 ga11on condensate storage tank.
A minimus of 92,000 gallobs of water in the 110,000 gallon condensate tank
~
.~
is sufficient for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of residuhl heat removal following a rea>; tor rip and loss of all off-site electrical power.
If the protected condensate storage tank level is reduced to 60,000 gallons, the im=ediately available gallon cond' nsate tank can be gravity-replenishment water in the 300,000 e
feed to the protected tank if required for residual heat removal. An alter-nate supply of feedwater to the auxiliary feedwater pump suction is also available from the Fire Protection System Main in the auxiliary feedwater pump cubicle.
l The five main steam code safety valves associated with each steam generator have a total combined capacity of 3,725,575 pounds per hour at their individual set pressure; the total combined capacity of all fifteen main steam ode safety valves is 11,176,725 pounds per hour.
f The ultimate pc r rating steam flow is 11,167,913 pounds per hour. The ombined capacity of the safety valves ccquired by Specification 3.6 always exceeds the total steam flow corresponding to the maximus. steady-
- w. aht sina4_dyrine _one. two or three reactor
..---..--.-s--
-, o
IS 3.8-1 l
3.8 CONTAINMENT Applicability Applies to the integrity and operating pressure of the reactor containment.
Objective To define the limiting operating status of the reactor containment for unit operation.
Specification A.
Containment Integrity and Operating Pressure 1.
The qontainme'nt, integrity, as* defined in TS Section 1.0, shall not be violated, except as specified in Specification 3.8.A.2 below, unless the reactor is in the cold shutdown confit 'an.
2.
The reactor containment shall not be purged while the reactor i operating, except as stated in Specification 3.8.A.3.
3.
During the plant startup, the remote manual valve on the steam jet
~
air ejector suction line may be open, if under administrative control, while containment vacuum is being established. The Reactor Coolant System temperature and pressure must not exceed 350*F and 450 psig, respectively, until the air partial pressure in the containment has been reduced to a value equal to, or below, that specified in TS Fig. 3.8-1.
~
4.
The containment integrity shall not be violated when the reactor vessel head is unbolted unless a shutdown margin greater than 10 percent Ak/k is maintained.
Ann @:Gn% NE.71 @.71
TS 3.8-2 5.
Positive reactivity changes shall not be made by rod drive motion or boron dilution unless the containment integrity is intact.
B.
Internal Pressure 1.
If the internal air partial pressure rises to a point 0.25 psi above the allowable value of the air partial pressure (TS Fig. 3.8-1),
the reactor shall be brought to the hot shutdown condition.
2.
If the leakage condition cannot be corrected without violating the containment integrity or if the internal partial pressure continues to rise, the reactor shall be brought to'the cold shutdown condition utilizing normal operating procedures.
3.
If the internal pressure falls below 8.25 psia the reactor shall be placed in the cold shutdown condition.
4.
If the, air parti,al pressure cannot be maintained greater than or equal to 9.0 psia, the reactor shall be brought to the hot shutdown condition.
I Basis 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 build-up in the containment if there is a loss-of-coolant accident.
The shutdown margins are selected based on the type of activities that are being carried out. The 10 percent ak/k shutdown margin during refueling precludes criticality ander any circumstance, even though fuel and control rod assemblics are being moved.
l l
l l
l Amendment Nos. 71 & 71
l l
TS 3.8-3 l
The allowable value for the containment air partial pressure is presented in TS Fig. 3.8-l'^for service water temperatures from 25 to 90*F.
The allowable value varies as shown in TS Fig. 3.8-1 for a given containment average temperature. The RWST water shall have a maximum terperature of 45'F.
The horizontal limit lines in TS Fig. 3.8-1 are based on LOCA peak calcu-lated pressure criteria, and the sloped line is based on LOCA subatmospheric peak pressure criteria.
The curve shall be interpreted as follows:
The horizontal limit line designates the allowable air partial pressure value for the given average containment temperature.
The, horizontal limit line applies for service water temperatures from 25*F to the sloped line intersection value (maximum service water temperature).
From TS Fig. 3.8-1, if the containment average temperature is ll2*F and the service water temperature is less than or equal to 83*F, the allow-able air partial pressure value shall be less than or equal to 9.65 psia.
If the average containment temperature is ll6*F and the service water temperature is less than or equal to 88*F, the allowable air partial pressure value shall be less than or equal to 9.35 psia. These horizontal limit lines are a result of the higher allowable initial containment average temperatures and the analysis of the pump suction break.
Amendment Nos. 71 & 71
~ - - -.
-n.-,,-
.,,.n,-
-e s.,y
-,,---, -e-,-
, n
,w--
-e
TS 3.8-4 e
i l
4 If the containment air partial pressure rises to a point 0.25 psi above the allowable value, the reactor shall be brought to the hot shutdown condition.
If a LOCA occurs at the time the containment air partial pressure is 0.25 psi above the allowable value, the maximum containment pressure will be less than 45 psig, the containment will depressurize 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 the maximum subatmospheric peak pressure will be less than 0.0 psig.
If the containment air partial pressure cannot be maintained greater than or equal to 9.0 psia, 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.
p-Refe'rences FSAR Section 4.3.2 Reactor Coolcnt Pump FSAR Section 5.2 Containment isolation FSAR Section 5.2.1 Design Bases FSAR Section 5.5.2 Isolation' Design Amendment Nos. 71 & 71 i
T.S. Figure 3.8-1 ALLOWABLE AIR PARTIAL PRESSURE SURRY POWER STATION
... 10.4 _ _ _ _ _ _
'.N_".100'F MINIMUM,,,_,' _.,,
, _,,,, )
TC= W F j 77eE
~l ' ~ i TC = 104'F
.- 4 0.2 Tc = toe *F 750F.
...__...p, 10.0 -
-. -.._~.._..-...
.. -. -.. 7,5 Tc = tos F 7gop g s
- c. _.
g _... 7,
.g m
810F.$,.
9.8 S,
c Tc = 112*F a
830F #
b 9.6-
-~
a-----
l
=
r -
-2._.
_._:.._...._.1__.
0 4
i TC - 114 F 0
86 F w
d i
8 ga 9.'
-j TC = 116'F ggop
}.
. i... I TC*118 F 900F 4.2 I
. __..].
i
- TC = 120 F '
0
/ 920F e
l--TC = 120*F M AXIMUM
- _--9.0 25 35 45
_._a -. 55 65 75 85 95 t
l
\\
I
?
.nRVICE WATER TEMPERATURE (*F)
I Amendment Nos. 71 & 71
TS Figure 3.8-1 FIGURE 3.8-1 (Continued)
FIGURE NOTATION TC - Containment average temperature.
FIGURE NOTES 1.
Allowable operating air partial pressure in the containment as a function of service water temperature.
2.
Refueling Water Storage Tank temperature 5 45'F.
3.
Horizontal lines designate allowable air partial pressure setpoint per given containment average temperature.
4.
Each,, containment, temperature line is a maximum for the given air partial pressure.
5.
Hot shutdown is required for containment air partial pressure increase greater than 0.25 psi above the allowable value or less than 9.0 psir..
6:
Cold shutdown is required for containment air partial pressure less than 8.25 psia.
l l
1 l-Amendment Nos. 71 & 71 l
l l
l i
i l
-,m-.
-~,.
1 TABLE 4.1-1 (Continued) i l
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 of service f
- 2) With analog rod position 11.
Steam Generator Level S
R H
i 12.
Charging Flow N.A.
R N.A.
13.
Residual Heal Removal Pump Flow N.A.
R N.A.
14.
Boric Acid Tank Level
- D R
N.A.
15.
Refueling Water Storage Tank Level S
R H
16.
Boron Injection Tank Level W
N.A.
N.A.
17.
Volume Control Tank Level N.A.
R N.A.
18.
Reactor Containment Pressure-CLS
- D R
H (1)
- 1) Isolation Valve signal and spray signal 4
19.
Process and Area Radiation Monitoring System
- D R
H s
20.
Boric Acid Control N.A.
n N.A.
21.
Containment Pump Level N.A.
R N.A.
s 22.
Accumulator Level and Pressure S
R N.A.
I.
?
d 23.
Containment Pressure-Vacuum Pump 4
i System S
R N.A.
j 24.
Steam Line Pressure S
R H
8 e
i Amendment Nos. 71 & 71 i
l 4
e 1
TABLE 4.5-1 RECIRCULATION SUBSYSTEM LEAKAGE
- Design Leakage to No.
Uncollected Vent and of Type of Leakage Control and Unit Leakage Drain System, ce per hr**
cc per hr Item Units Leakage Rate i
Recirculation spray 2
No leak of spray water due to tanden 0
0 pumps seal arrangement i
40 drops per min per flange
- a. pump 4
480 0
- b. Valves -
4 460 0
bonnet to body (larger than 2 in.)
Valves - Stem 4
Backseated, double packing with 0
16 leakoffs leakoff - 4 cc per hr per in, stem diameter Miscellaneous 2
Flanges body, packed sten.- 4 drop 24 0
small valves per min Total 964 16
- Based on two subsystems in operation under DBA conditions.
H" Total Allowed System Uncollected Leakage is 964.cc/hr.
- Individual component uncollected leakage may exceed the design value provided that the total j"
allowable system uncollected leakage is not exceeded.
' Amendment Nos. 71 & 71
1 TABLE 4.11-1 EXTERNAL RECIRCULATION LOOP LEAKAGE (Safety Injection Systes Only) e Design Design Leakage to Leakage to No. of Type of Leakage Control and Unit Atmosphere Waste Disposal cc per hr**
Tank, cc per br Item Units Leakage Rate i
Low Head Safety 2
Hechanical Seal with. leakoff -
0 24 Injection Pumps 4 drop per min Safety Injection 3
Mechanical Seal with leakoff -
0 36 Charging 4 drop per min Flanges:
- a. Pump 10 Gasket - adjusted to zero leakage 1,200 0
following any test - 40 drop per ein, per flange
- b. Valves Bonnet to 54 2,240 0
Body (larger than 2 in.)
i Valves - Stem Leakoffs 27
.Backseated, double packing with 0
108 leakoff - 4 cc per hr per in stem diameter Misc. Valves 33 Flanges body packed stems - 4 drop per min 396 0
s Totals 3,836 168 E!
C Total Allowed System Uncollected Leakage is 3,836 cc/hr 3,
- Individual component uncollected leakage may exceed the design value provided that the total allowable system uncollected leakage is not exceeded.
l Amendment Nos. 71 & 71 l
-