ML20031F164
| ML20031F164 | |
| Person / Time | |
|---|---|
| Site: | Surry  (DPR-32-A-072, DPR-32-A-72, DPR-37-A-073, DPR-37-A-73) |
| Issue date: | 09/29/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20031F165 | List: |
| References | |
| NUDOCS 8110190290 | |
| Download: ML20031F164 (50) | |
Text
^
(+;ga j%g,9, UNITED STATES y 9 ),., (, 3 g NUCLEAR REGULATORY COMMISSION
- .lG '-,"/.C WASHINGTON, D. C. 20555 gy j
gv j VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 72 License No. DPR-32 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated November 14, 1980, as supplemented December 23, 1980 and August 21, 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 set fo-th 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 (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 Comission'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 l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements l
have been satisfied.
i l
l 8110190290 810929 PDR ADOCK 05000280 P
, 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-32 is hereby amended to read as follows:
(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 72, 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.
F0p THE NUCpEAR R[{GULATORY COMMISSION L
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U10 arga,Chigf(
t Steven A.
Operating Reactors B nch No. 1 Division of Licensin
Attachment:
Changes to the Technical Specifications Date of Issuance: September 29, 1981 l
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'N UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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WASHING TON, D. C. 20555
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73 License No. DPR-37 1.
The Nuclear Regulatory Coranission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated November 14, 1980, as supplemented December 23, 1980 and August 21, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Corm.ission'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 or the Comission; 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 Comission's regulations; 0.
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.
---.,,,m-,
2-2.
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 DPR-37 is hereby amended to read as follows:
~
(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 73, 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.
FOR THE NUCLEAR REGULATORY COMMISSION I
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$o Y"
Steven A.
arga hie Operating Reactors Qr nch No. 1 Division of Licensi
Attachment:
Changes to the Technical Specifications Udte of Issuance: September 29, 1981 e
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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 72 TO FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 73 TO FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:
Remove Pages Insert Pages 3.1-1 3.1-1 3.1-2 3.1-2 3.1-2a 3.1 - 3 3.1-3 4
3.1-4 3.1-4 3.1-5 3.1-5 3.1-5a 3.1-5b 3.1-5c-3.7-1 3.7-1 3.7-2 3.7-2 3.7-3 3.7-3 3.7-4 3.7-4
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TS 3.1-1 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.
Objectives To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe reactor operation.
These conditions relate to:
operational components, heatup and cooldown, leakage, reactor coolant activity, oxygen and chloride concentrations, minimum temperature for criticality, and reactor coolant system overpres-sure mitigation.
A.
Operational Components Specifications 1.
Reactor Coolant Pumps a.
A reactor shall not be brought critical with less than two pumps, in non-isolated loops, in operation.
Amendment Nos. 72 & 73
TS 3.1-2 b.
If an unscheduled loss of one or more reactor coolant pumps occurs while operating below 10% rated power (P-7) and results in less than two pumps in service, the affected
~
plant shall be shutdown and the reactor made suberitical by inserting all control banks into the coce.
The shutdown rods may remain withdrawn.
c.
When the average reactor coolant loop temperature is greater than 350*F, the following conditions shall be met:
1.
At least two reactor coolant loops shall be operable.
2.
At least one reactor coolant loop shall be in operation.
d.
When the average reactor coo. ant loop temperature is less than or equal to 350*F, the following conditions shall be met:
1.
A uinimum of two non-isolated lesps, consisting of any combination of reactor coolant loops or residual heat removal loops, shall be operable, except as specified in Specification 3.10.A.6.
2.
At least one reactor coolant loop or one residual heat removal loop shall be in operation, except as specified in Specification 3.10.A.6.
Amendment Nos. 72 & 73
TS 3.1-3 l
Reactor power shall not exceed 50% of rated power with only e.
two pumps in operation unless the overtemperature AT trip setpoints have been changed in accordance with Section 2.3, after which power shall not exceed 60% with the inactive loop stop valves open and 65% with the inactive loop stop valves closed.
f.
When all three pumps have been idle for > 15 minutes, the first pump shall not be started unless:
(1) a bubble exists in the pressurizer or (2) the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.
2.
Steam Generator A minimum of two steam generators in non-isolated loop shall be operable when the average reactor coolant temperature is greater than 350*F.
l 3.
Pressurizer Safety Valves a.
One valve shall be operable whenever the head is on the reactor vessel, except during hydrostatic tests.
Amendment Nos. 72 & 73
l TS 3.1-4 l
b.
Three v?1.ves shall be operable when the reactor coolant average temperature is greater than 350'F, the reactor is critical, or the Reactor Coolant System is not connected to
~
the Residual Heat Removal System.
Valve lift settings shall be maintained at 2485 psig i I c.
percent.
4.
Reactor Coolant Loops Loop stop valves shall not be closed in more than one loop unless the Reactor Coolant System is connected to the Residual Heal Removal System and the Residual Heat Removal System is operable.
s 5.
Pressurizer The reactor shall be maintained suberitical by at least 1%
a.
until the steam bubble is established and necessary sprays and at least 125 Kw of heaters are operable.
b.
With the pressurizer inoperable due to inoperable pressurizer heaters, restore the inoperable heaters within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot shutdawn within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the reactor coolant system temperature and pressure less than 350*F and 450 psig, respectively, within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Amendment Nos. 72 & 73 l
J
TS 3.1-5 l
4 c.
With the pressurizer otherwise inoperable, be in at least hot shutdown with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the reactor coolant system temperature and pressure less than
~
350*F and 450 psig, respectively, within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
6.
Relief Valves a.
Two power operated relief valves (PORVs) and their associated block valves shall be operable whenever the reactor keff is 20.99.
1 b.
With one or morr. PORVs inoperable, within I hour either restore the PORV(s) to operable status or close the associattJ block valve (s) and remove power from the block valve (s); otherwise, be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
With one or more block valve (s) inoperable, within I hour either restore the block valve (s) to operable status or close the block valve (s) and raiove power from the block valve (s); otherwise, be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold i
shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
t Basis l
Specification 3.1.A-1 requires that a sufficient number of reactor coolant pumps be operating to provide coastdown core cooling flow in the event of a loss of reactor coolant flow accident. inis provided flow will maintain the t
Amendment Nos. 72 & 73 T
i
,, -. ~....
e TS 3.1-Sa DNBR above 1.30.
Heat transfer analyses also show that reactor hest equiva-lent to approximately 10% of rated power can be removed with natural i
circulation; however, the plant is not designed for critical operation with natural circulation or one loop operation and will not be operated under these conditions.
l When the boron concentration of the Reactor Coolant System is to be reduced the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uni-form concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place.
The residual heat removal pump will circulate the equivalent of the reactor coolant system volume in approximately one half hour.
One steam generator capable of performing its heat transfer function will provide sufficient heat removal capability to remov-core decay heat after a normal reactor shutdown.
The requirement for redundant coolant loops ensures the capability to remove core decay heat when the reactor coolant system average temperature is less than or equal to 350*F. Because of the low-low steam generator water level reactor trip, normal reactor criticality cannot be achieved without water in the steam generators in reactor coolant loops with open loop stop valves. The requirement for two operable rteam generators, combined with the requirements of Specification 3.6, ensure adequate heat removal capabilities for reactor coolant system temperatures of greater than 350*F.
Anendment Nos. 72 & 73 n,
.. ~ - - -,
--.-g
TS 3.1-5b Each of the pressurizer safety valves is designed to relieve 295,000 lbs.
per br. 6f saturated steam at the valve setpoint. Below 350*F and 450 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure. There are no credible accidents which could cccur when the Reactor Coolant System is connected to the Residual Heat Removal System which could give a surge rate exceeding the capacity of one pressurizer safety valve. Also, two safety valves have a capacity greater than the maximum surge rate resulting from complete loss of load.(2)
The limit 3 tion specified in item 4 above on reactor coolant loop isolation will prevent an accidental isolation of all the loops which would eliminate the capability of dissipating core decay heat when the Reactor Coolant System is not connected to the Residual Heat Removal System.
The requirement for steam bubble formati an in the pressurizer when the reactor has passes 1% subcriticality will ensure that the Reactor Coolant System will not be solid when criticality is achieved.
The requirement that 125 Kw of pressurizer heaters and their associated controls be capable of being s7pplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at hot shutdown.
Amendment fios. 72 & 73
TS 3.1-Se The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.
References:
(1) FSAR Section 14.2.9 (2) FSAR Section 14.2.10 I
l l
l Amendment Nos. 72 & 73
TS 3.7-1 3.7 INSTRUMENTATION SYSTEMS Operational Safety Instrumentation Applicability:
Applies to reactor and safety features instrumentation systems.
Objectives:
To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable limits are exceeded, and to delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety.
Specification:
A.
For on-line testng or in the event of a sub-system instrumentation channel failure, plant operation at rated power shall be permitted to continue in accordance with TS Tables 3.7-1 through 3.7-3.
B.
In the event the number of channels of a particular sub-system in service falls below the limits given in the column entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in Column 4 of TS tables 3.7-1 through 3.7-3.
l Amendment Nos. 72 & 73 l
.= -
TS 3.7-2 C.
In the event of sub-system instrumentation channel failure permitted by Specification 3.7-B, Tables 3.7-1 through 3.7-3 need not be observed during the short period of time and operable sub-system channel
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are tested n
.e failed channel must be blocked to prevent unnecessary reactor trip.
D.
The Engineered Safety Features initiation instrumentation setting limits shall be as stated in TS Table 3.7-4.
E.
Automatic functions operated from radiation monitor alarm shall be as stated in TS Table 3.7-5.
The requirements of Specification 3.0.1 are not applicable.
F.
The accident monitoring instrumentation for its associated operable components listed in TS Table 3.7-6 shall be operable in accordance with the following:
1.
With the number of operable accident monitoring instrumentation channels less than the total number of channels shown in TS Table 3.7-6, either restore the inoperable channel (s) to operable status within 7 days or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
With the number of operable accident monitoring instrumentaton channels less than the minimum channels operable requirement of TS Table 3.7-6, either restore the inoperable channel (s) to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Amendment Nos. 72 & 73
TS 3.7-3 Basis i
Instrument Operating Conditions During plant operations, the complete instrumentation system will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits.
jafety is not compromised, however, by continuing operation with certain instru-mentation channels out of service since provisions were made for this in the plant design. This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.
Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for channel calibration and test at power. Exceptions are backup channels such as reactor coolant pump breakers. The removal of one trip channel en process control equipment is accomplished by piscing that d
channel bistable in a tripped mode; e.g., a two-out-of-three circuit becomes a one oat-of-two circuit. The nuclear instrumentation system channels are not intentionally placed in a tripped mode since the test signal is superimposed on the normal dete: tor signal to test at power. Testing of the NIS power range channel requires:
(a) bypassing the Dropped Rod protection from NIS, for the channel being tested: and (b) placing the AT/T,y8 protection channel set that is being fed from the NIS channel in the trip mode and (c) defeating the power mismatch section of T,y control channels when the appropriate NIS channel is Amendment Nos. 72 & 73
___, _.. _, _. _, _ ~
_. _ _ _.- _.. _ _... _ _ _ _ _,. ~. -., _
TS 3.7-4 being tested.
However, the Rod Position System and remaining NIS channels still provide the dropped-rod protection. Testing does not trip the system unless a trip condition exists in a concurrent channel.
Instrumentation has been provided to sense accident conditions and to initiate (I) operation of the Engineered Safety Features.
Safety Injection System Actuation Protection ag. inst a Loss of Coolant or Steam Break Accident is brought about by automatic actuation of the Safety Injection System which provides emergency cooling and reduction of reactivity.
The Loss of Coolant Accident is characterized by depressurization of the Reactor Coolant System and rapid loss of reactor coolant to the containment.
The Engineered Safeguards Instrumentaton has been designed to sense these effects of the Loss of Coo). ant accident by detecting low pressurizer pressure to generator signals actuating the SIS active phase. The SIS active phase is also actuated by a high containment pressure signal brought about by loss of high e'hthalpy coolant to the containment.
This actuation signal acts as a backup to the low pressurizer pressure actuation of the SIS and also adds diversity to protection against loss of coolant.
Signals are also provided to actuate the SIS upon sensing the effects of a steam line break accident.
Therefore, SIS actuation following a steam line break is designed to occur upon sensing high differential steam pressure
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Amendment Nos. 72 & 73
TS 3.7-5 between the steam header and steam generator line or upon sensing high steam line flow in coincidence with low reactor coolant average temperature or low steam line pressure.
The increase in the extraction of RCS heat following a steam line break results in reactor coolant temperature and pressure reduction. For this reason pro-tection against a steam line brea accident is also provided by low pressurizer pressure actuating safety injection.
Protection is also provided for a steam line break in the containment by actuation of SIS upon sensing high containment pressure.
SIS actuation injects highly borated fluid into the Reactor Coolant System in order to counter the reactivity insertion brough about by cooldown of the reactor coolant which occurs during a steam line break accident.
The Engineered Safety Features also initiate containment spray upon sensing a high-high containment pressure signal. The containment spray acts to reduce 1
containment pressure in the event of a loss of coolant or steam line break accident inside the containment. The containment spray cools the containment directly and limits the release of fission products by absorbing iodine should it be released to the containment.
Amendment Nos. 72 & 73 1
TS 3.7-6 Containment spray is designed to be actuated at a higher containment pressure (approximately 50% of design containment pressure) than the SIS (10% of design).
Since spurious actuation of containment spray is to be avoided, it is initiated only on coincidence of high-high containment pressure sensed by 3 out of the 4 containment pressure signals provided for its actuation.
Steam Line Isolation Steam line isolation signals are initiated by the Engineered Safety Features closing all steam line trip valves.
In the event of a steam line break, this action prevents continuous, uncontrolled steam release from more than one steam generator by isolating the steam lines on high-high containment pressure or high steam line flow with coincident low steam line pressure or low reactor coolant average temperature. Protection is afforded for breaks inside or outside the containment even when it is assumed that there is a single failure in the steam 1.ne isolation system.
Feedwater Line Isolation The feedwater lines are isolated upon actuation of the Safety Injection System in order to prevent excessive cooldown of the reactor coolan, sjstem. This sitigates the effects of an accident such as steam break which in itself causes excessive coolant temperature cooldown.
Feedwater line isolation also reduces the consequences of a steam line break inside the containment, by stopping the entry of feedwater.
Amendment Nos. 72 & 73
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TS 3.7-7 e
Auxiliary Feedwater System Actuation The automatic, initiation of auxiliary feedwater flow to the steam generators by instruments identified in Table 3.7-2 ensures that the Reactor Coolant System Decay Heat can be removed following loss of main feedwater flow.
This is consistent with the requirements of the "THI-2 Lesson Learned Task Force Status Report", NUREG-0578, item 2.1.7.b.
Setting Limits 1.
The high containment pressure limit is set at about 10% of design containment Initiation of Safety Injection protects against loss of coolant (2) pressure.
or steam line breat (3) accidents as discussed in the safety analysis.
2.
The high-high containment pressure limit is set at about 50% of design containment pressure.
Initiation of Containment Spray and Steam Line Isolation protects against large loss of coolant (2) or steam line break l
accidents (3) as discussed in the safety analysis.
(
3.
The pressurizer low pressure setpoint for safety injection acutation is 1
set substantially below system operating pressure limits. However, it is sufficiently high to protect against a loss-of-coolant accident as shown in the safety analysis. (2) l Amendment Nos. 72 & 73 l
-,e, a-mw ev-g yy w
mm%
TS 3.7-8 t
4.
The steam line high differential pressure limit is set well below the differential pressure expected in the event of a large steam line break accident as shown in the safety analysis. (3) 5.
The high steam line flow differential pressure setpoint is constant at 40% full flow between no ? ad and 20% load and increasing linearly to 110% of full flow at full load in order to protect against large steam line break accidents. The coincident low T,yg setting limit for SIS and steam line isolation initiation is set below its hot shutdown value.
The coincident steam line pressure setting limit is set below the full load operating pressure. The safety analysis shows that these settings provide protection in the event of a large steam line break.(3)
Automatic Function Operated from Radiation Monitors The Process Radiation Monitoring System continuously monitors selected lines containing or possibly containing, radioactive effluent.
Certain channels in this system actuate control valves on a high-activity alarm signal. Additional information on the Process Radiation Monitoring System is available in the FSAR. b)
Accident Monitoring Instrumentation The operability of the accident monitoring instrumentation is Table 3.7-6 ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. On the pressurizer PORV's, the pertinent channels consist of limit switch indication and acaustic Amendment Nos. 72 & 73
TS 3.7-9 monitor indication. The pressurizer safety valves utilize an acoustic monitor channel and a downstream high temperature indication channel.
This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident", December 1975, and NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short Term Recommendations".
References (1) FSAR - Section 7.5 (2) FSAR - Section 14.5 (3) FSAR - Section 14.3.2 (4) FSAR - Section 11.3.3 Amendment Nos. 72 & 73
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TABL.E 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 4
1 2
3 4
OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN.
OF EXCEPT AS CONDI-OPERABLE REDUN-PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET 1.
Manual 1
Maintain hot shutdown 2.
Nuclear Flux Power Range 3
2 Low trip setting when 2 Maintain hot of 4 power channels greater shutdown than 10% of full power 3.
Nuclear Flux Intermediate 1
2 of 4 power channels greater Maintain hot inge than 101 full power shutdown 4.
Nuclear Flux Source Range 1
1 of 2 intermediate range lisintain hot 10 channels greater than 10 shutdown amps 4
5.
Overtemperature AT 2
1 Maintain hot 3g shutdown E.g 6.
Overpower AT 2
1 Maintain hot 3
shutdown k
7.
Low Pressurizer Pressure 2
1 3 of 4 nuclear power channels Maintain hot and 2 of 2 turbine load shutdown M
chacnels less than 10% of y
rated power m
uw 8.
Hi Pressurizer Pressure 2
1 Same as Item 7 above Maintain hot y
shutdown g
e TABLE 3.7-1 REACTOR TRIP t
INSTRUMENT OPERATING CONDITIONS I
~
1 2
3 4
OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN.
OF EXCEPT AS CONDI-OPERABLE REDUN-PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET l
9.
Pressurtzer-Hi Water Level 2
1 3 of 4 nuclear power Maintain hot l
channels and 2 of 2 shutdown l
turbine load channels l
less than 10% of rated power 1
10.
Low Flow 2/ operable If inoperable loop Maintain hot l
loop channels are not in service shutdown they must be placed in the tripped mode 11.
1 Maintain less than 10% rated power 12.
Lo-Lo Steam Generator 2/non-iso-1/non-Maintain hot Water Level lated loop isolated loop shutdown p
- 13. Underfrequency 4KV Bus 2
1 Maintain hot shutdown o.
5
- 14. Undervoltage 4KV Bus 2
1 Maintain hot 5
shutdown E
Y M
M wL E:
0
TABLE 3.7-1 REACTOR TRIP
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INSTR!HENT OPERATING CONDITIONS 1
2 3
4
~
OPERATOR ACTION IF CONDITIONS OF DEGREE COI.UMN 1 OR 2 MIN.
OF EXCEPT AS CONDI-OPERAS,LE REDUN-PERMISSIBII BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET 15.
Control rod misalignment Monitor **
a) rod position deviation 1
Log individual rod positions once/ hour, and after a load change
> 10% or after > 30 inches of control rod motion.
b) quadrant power tilt 1
Log individual upper monitor (upper and upper and lower ion lower excore neutron chamber currents once/
detectors) hour and after a load rht ge > 10% or after
>- 30 inches of control rod motion.
- 16. Safety Injection See Item 1 of TS Table 3.7-2 8n Gt
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us D*
i
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ne l
l
TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1
2 3
4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN.
OF EXCEPT AS CONDI-OPERABLE REDUN-PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS C_ANNOT BE MET
- 17. Low steam generator 1/non-iso-Maintain hot water level with lated loop shutdown steam /feedwater 1/non-iso-mismatch flow lated loop 1
- If both rod misalignment monitors (a and b) inoperable for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or more, the nuclear overpower trip shall be reset to 93 percent of rated power in addition to the 1
increased surveillance noted.
I d
a 8e r
M o.
C
TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION I
1 2
3 4
I OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN.
OF EXCEPT AS CONDI-OPERABLE REDUN-PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET 1.
SAFETY INJECTION a.
Manual 1
0 1
Cold shutdown b.
High Containment Press.
3 1
Cold shutdown c.
High Differential Press.
2/non-iso-1/non-Primary Pressure Cold shutdown between any Steam Line and lated loop isolated less than 2000 psig the Steam Line Header loop except when reactor is critical d.
Pressurizer Low-Low Press.
2 1
Primary Pressure Cold shutdown less than 2000 psig except when reactor is critical e.
High Steam Flow in 2/3 1/steamline Reactor Coolant aver-Cold shutdown k
Steam Lines with' Low T 2T signals 1 age temperature less i
or Low Steam Line Press!8 2 SleIm Press.
I than 543*F (nominal) l Signals during heatup and 5
cooldown E,
v
- With the specified minimum operable channels the 2/3 high steam flow is already in the trip mode.
N 1
tn u
i i
4
TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDITIONS 1
2 3
4 OPERATOR ACTION IF CONDITIONS OF DEGREE CDLUMN 1 OR 2 MIN.
OF EXCEPT AS CONDI-OPERABLE REDUN-PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET 2.
CONTAINMENT dPRAY a.
Manual 2
Cold shutdown b.
High Conta!nment Press.
3 1
Cold shutdown (Hi-Hi Setpoint) 3.
Steam Generator Water Level Low-Low
- i. Start Motor 2/Stm. Gen.
1 Loop Stop Valve in res-Place inoperable Driven Pumps pective loop closed channel in Tripped II. Start Turbine 2/Stm. Gen.
I condition within Driven Pumps one hout Eg b.
1 Place inos,erable g-Start 'Ibrbine Driven Pump channel in Tripped g
condition within r+
one hour 5
T c.
Safety Injection (All safety injection initiating functions and requirements)
Start Motor Driven Pumps y
0
- Must actuate 2 switches simultaneously.
.?
G:
TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDIT0NS 1
2 3
4 OPERATOR ACTION i
IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN.
OF EXCEPT AS CONDI-l OPERABLE REDUN-PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET 3
d.
Station Blackout 2
0 Restore inoperable Start Motor Driven Pump channel within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shatdown i
within the follow-ing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
e.
Trip of Main Feedwater Pumps 1/ Pump 1/ Pump Restore inope>'able Start Motor Pumps channel withi 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the follow-27 ing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, m
n
?
Ri D1 i
y>
a.
U T
5;'
TABLE 3.7-3 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS INSTRUMENT OPERATING CONDITIONS 1
2 3
4 i
OPERATOR ACTION IF CONDITIONS OF DEGREE C0f,UMN 1 OR 2 MIN.
OF EXCEPT AS CONDI-OPERABLE REDUN-PERMISSIBLE BYPASS TIONED BY COLUMN 3 4
FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET I
1.
CONTAINMENT ISOLATION a.
Safety Injection See Item No. 1 of Table 3.7-2 Cold shutdown b.
Manual 1
Hot shutdown c.
High Conta$nment Press.
3 1
Cold shutdown (Hi Setpcinti d.
High Containment Press.
3 1
Cold shutdown 2.
STEAM LINE ISOLATION a.
High Steam Flow in 2/3 lines 1/steamline Cold shutdown and 2/3 Low T or 2/3 2/T 1
Low Steam Pre $ sere sig$aEs k
2 Stm. Press.
I 1
g, signals 5
b.
High Containment Press.
3 1
Cold shutdown 2
(Hi-Hi Level)
O c.
Manual 1/line Hot shutdown M
H,
{
3.
FEEDWATER LINE ISOLATION o.
a.
Safety Injection See Item No. 1 of Table 3.7-2 Cold shutdown s
- With the specified minimum operable channels the 2/3 high steam flow is already in the trip mode
4 TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING No.
FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT 1
High Containment Pressure (High Contain-a) Safety Injection 55 psig
~
ment Pressure Signal) b) Containment Vacuum Pump Trip c) High Press. Containment Iso.
d) Safety Injection Contain. Iso.
e) F.W. Line Isolation 2
High High containment Pressure (High High a) Containment Spray
$25 psig Containment Pressure Signals) o) Recirculation Spray c) Steam Line Isolation d) High High Press. Contain. Iso.
3 Pressurizer Low Low Pressure a) Safety Injection 21,700 psig b) Safety Injection Cont. Iso.
c) Feedwater Line Isolation 4
High Differential Pressure Between a) Safety Injection
$150 psi Steam Line and the Steam Line Header b) Safety Injection Contain. Iso.
c) F.W. Line Isolation 5
High Steam Flow in 2/3 Steam Lines a) Safety Injection 540% (at zero load) of full steam flow 540% (at 20% load) of full steam flow b) Steam Line Isolation 5110% (at full load) of Ef c) Safety Injection Contain. Iso.
full steam flow d) F.W. Line Isolation Coincident with Low T,y8 or Low Steam 2541*F T,yg EF Line Pressure 2500 psig steam line pressure U
D a-Ei u>
L h
~
TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING No.
FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT 6
AITXILIARY FEEDWATER a.
Steam Generator Water Level Low-Low Aux. Feedwater Initiation 25% narrow range S/G Blowdown Isolation b.
RCP Undervoltage Aux. Feedwater Initiation 270% nominal c.
Safety Injection Aux. Feedwater Initiation All S.I. setpoints d.
Station Blackout Aux. Feedwater Initiation 246.7% nominal e.
Main Feedwater Pump Trip Aux. Feedwater Initiation N.A.
a a
8e M
o.
\\
4 l
TABLE 3.7-5 l.
AUTOMATIC FUNCTIONS OPERATED FROM RADIATION MONITORS ALARM i
1 AUTOMATIC FUNCTION MONITORING ALARM $ETPOINT MONITOR CHANNEL AT ALARM CONDITIONS REQUIREMENTS pCI/cc i
1.
Process vent particulate Stops discharge from contain.
See Specifications Particulatg $4x10' and gas monitors vacuum systems and waste 3.11 and 4.9 Gas 59x10 (RM-GW-101 & RM-GW-102) gas decay tanks (shuts Valve Nos. RCV-GW-160, FCV-GW-260, FCV-GW-101) l' 2.
Component cooling water Shuts surge tank vent valve See Specifications STwice Background j
radiation monitors HCV-CC-100 3.13 and 4.9 l
(RM-CC-105 & RM-CC-106)
[
3.
Liquid waste disposal Shuts effluent discharge See Specifications 51.5x10'3 radiation monitors valves FCV-LW-104A and 3.11 and 4.9 (RM-LW-108)
FCV-LW-104B 4.
Condenser air ejector Diverts flow to the contain-See Specification 51.3 I
radiation monitors ment of the affected unit 3.11 and 4.9 i
l (RM-SV-111 & RM-SV-211)
(0 pens TV-SV-102 and shuts i
TV-SV-103 or opens TV-SV-202 and shuts TV-SV-203) 5.
Containment particulte Taips affected unit's purge See Specifications Particulatg59x10' j
y and gas monitors supply and exhaust fans, 3.10 and 4.0 Cas 51x10 (RM-RMS-159 & RM-RMS-160, closes affected unit's 8-RM-RMS-259 & RM-RMS-260) purge air butterfly valves j
(MOV-VS-100A, B, C & D or MOV-VS-200A, B, C & D)
E i
6.
Manipulator crane area Trips affected unit's purge See Specifications 550 ares /hr L
monitors (RM-RMS-162 &
supply and exhaust fans, 3.10 and 4.9 y
RM-RMS-262) closes affected unit's N
{
purge air butterfly valves d
(MOV-VS-100A, B, C & D or MOV-VS-200A, B, C & D 1
k E
TABLE 3.7-6 ACCIDENT MONITORING INSTRUMENTATION TOTAL NO.
MINIMUM CHANNELS INSTRUMENT OF CHANNELS OPERABLE 1.
Auxiliary Feedwater Flow Rate 1 per S/G 1 per S/,G 2.
Reactor Coolant System Subcooling Margin Monitor 2
1 3.
PORV Position Indicator (Primsry Detector) 1/ valve 1/ valve 4.
PORV Position Indicator (Backup Detector) 1/ valve 0
5.
PORV Block Valve Position Indicator 1/ valve 1/ valve 6.
Safety Valve Position Indicator (Primary Detector) 1/ valve 1/ valve 7.
Safety Valve Position Indicator (Backup Detector) 1/ valve 0
i 8
lit j
2" O*
M m
N
?,
Ct t
1
I 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 containment 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 condition.
2.
The reactor containment shall not be purged while the reactor is operating, except as stated in Specification 3.8.A.3.
3.
Durinr, 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.
Amendment Nos. 72 & 73
miry.vs 5.
Positive reactivity changes shall not be made by rod drive motion or boron dilution unless the containment integrity is intact.
6.
The containment isolation valves shall be listed in Tables 3.8-1 and 3.8-2.
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 p essure (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 procedores.
3.
If the internal pressure falls below 8.25 psia the reactor shall be placed in the cold shutdown condition.
4.
If the air partial pressure cannot be maintained greater than or equal to 9.0 psia, the reactor shall be brought to the hot shutdown condition.
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.
Amendment Nos. 72 & 73 e
u 0
e e
TS 3.8-3 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 under any circumstance, even though fuel and control rod assemblies are being moved.
The allowable value for the containment air partial pressure is presented in TS Fig. 3.8-1 for service water temperatures from 25 tc 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 temperature 6f 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 subatnospheric peak pressure criteria.
The curve shall be interpreted as follows:
The horizuntal limit lin,t designates the allowable air partial
\\
pressure value for the given everage containment temperature.
The horizontal limit line applie.s 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 112*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 116*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. 72 & 73
l TS 3.8-4
/
i If the containment air r tial 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 oc, curs 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 bour, and the maximum subatmospheric peak pressure will be less than 0.0 psig.
1 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 ccadition. 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 FSAR Section 4.3.2-Rezetor Coolant Pump FSAR Section 5.2 Containment Isolation FSAR Section 5.2.1 Design Bases FSAR Section 5.5.2 Isolation Design l
l t
1 I
i.
Amendment Nos. 72 & 73 e
4 TS 3.8-5 6
TABLE 3.8-1 a
UNIT NO. 1 CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION A.
PHASE I CONTAINMENT ISOLATION (SAFETY INJECTION SIGNAL) 1.
MOV-1867C Boron Injection Tank Outlet 2.
MOV-1867D Boron Injection Tank Outlet 3.
MOV-1285A Charging Line 4.
MOV-1381 Reactor Coolant Pump Seal Water Return 5.
HCV-12004 Letdown Orifice Isolation 6.
HCV-1200B Letdown Orifice Isolation 7.
HCV-1200C Letdown Orifice Isolation 8.
TV-SI-101A Accumulator N helief Line 2
9.
TV-SI-101B Accumulator N Relief Line 2
10.
TV-SI-100 Accumulator N Relief Line 2
11.
TV-VG-109A Primary Drain Transfer Tank Vent 12.
TV-VG-109B Primary Drain Transfer Tank Vent 13.
TV-DG-108A Primary Drain Transfer Pump Discharge l
14.
TV-DG-108B Primary Drain Transfer Pump Discharge 15.
TV-CC-109A*
Component Cooling from RHR's 16.
TV-CC-109B*
Component Cooling from RHR's 17.
TV-SS-100A Pressurizer Liquid Sample 18.
TV-SS-100B Pressurizer Liquid Sample 19.
TV-SS-101A Pressurizer Vapor Sample 20.
TV-SS-101B Pressurizer Vapor Sample Amendment Nos. 72 & 73 mm
,.,eowne--
e
,a vow e *v w,
TS 3.8-6 TABLE 3.8-1 UNIT NO. 1 CONTAINMENT ISOLATION VALV } (Continued)
VALVE NUMBER FUNCTION 21.
IV-SS-103 Residual Heat Removal System Sample
^
22.
TV-SS-106A Reactor Loolant Hot Leg Sample 23.
TV-SS-106B Reactor Coolant Hot Leg Sample 24.
TV-SS-102A Reactor Coolant Cold Leg Sample 7.5.
TV-SS-102B Reactor Coolant Cold Leg Sample 26.
TV-SS-104A Pressurizer Relief Tank Vapor Sample 27.
TV-SS-104B Pressurizer Relief Tank Vapor Sample 28.
TV-CH-1204 Letdown Isolation Valve 29.
TV-PG-1519A Primary Grade Water to Pressurizer Relief Tank 30.
TV-BD-100A*
Steam Generator Blowdown Valve 31.
TV-BD-100B*
Steam Generator Blowdown Valve 32.
TV-BD-100C*
Steam Generator Blowdown Valve 33.
TV-BD-100D*
Steam Generator Blowdown Valve 34.
TV-BD-1.00E*
Steam Generator Blowdown Valve 35.
TV-BD-100F*
Steam Generator Blowdown Valve 36.
TV-DA-100A Containment Sump Pump Isolation 37.
TV-DA-100B Containment Sump Pump Isolation 38.
TV-MS -109*
Main Steam Drain Trip Valve 39.
TV-MS-110*
Main Steam Drain Trip Valve 40.
TV-LM-100A Containment Isolation Monitoring 41.
TV-LM-100B Containment Isolation Monitoring 42.
TV-LM-100C Con.ainment Isolation Monitoring e
Amendment Nos. 72 & 73
TS 3.8-7 i
TABLE 3.8-1 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)
VALVE NUMBER FUNCTION 43.
TV-LM-100D Containment Isolation Monitoring 44.
TV-LM-100E Containment Isolation Monitoring 45.
TV-LM-100F Containment Isolation Monitoring 46.
TV-LM-100G Containment Isolation Monitoring 47.
TV-LM-100H Containment Isolation Monitoring 48.
TV-CV-150A Containment Vacuum Suction Valve 49.
TV-CV-150B Containment Vacuum Suction Valve 50.
TV-LM-101A Leakage Monitoring Sealed Reference 51.
TV-LM-101B Leakage Monitoring Sealed Reference 52.
TV-CV-150C Containment Vacuum Suction Valve 53.
TV-CV-150D Containment Vacuum Suction Valve 54.
TV-SV-102A Condenser Air Ejector Vent Trip Valve B.
PHASE II CONTAINMENT ISOLATION (HI CLS SIGNAL) 1.
TV-RM-100A Containment Air & Particulate Rad. Mon. TV's 2.
TV-RM-100B Containment Air & Particulate Rad. Hon. TV's 3.
TV-RM-100C Containment Air & Particulate Rad. Mon. TV's 4.
TV-IA-101A Containment Instr
'r Compressor Suction 5.
TV-IA-101B Containment Instr. Air Compressor Suction 1
1 Amendment Nos. 72 & 73 l
is 3.8-6 i
TABLE 3.8-1 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)
VALVEliUMBER-FUNCTION l
C.
PHASE III CONTAINMENT ISOLATION (HI-HI CLS SIGNAL) 1.
TV-MS-101A*
Main Steam Trip Valve 2.
TV-MS-101B*
Main Steam Trip Valve 3.
TV-IA-100 Containment Instr. Air Compressor Disch. Vlv.
4.
TV-MS-101C*
Main Steam Trip Valve 5.
TV-CC-107*
CC from RCP Thermal Barriers 6.
TV-CC-101A*
CC from A Air Recire.
l 7.
TV-CC-101B*
CC from B Air Recire.
8.
TV-CC-101C*
CC from C Air Recire.
9.
TV-CC-105A*
TV-CC-105B*
11.
TV-CC-105C*
CONTAINMENT PURGE & EXHAUST 1.
MOV-VS-100C R.C. Purge Exhaust M0V's 2.
J0V-VS-100D R.C. Purge Exhaust MOV's 3.
MOV-VS-101 R.C. Purge Exhaust Bypass MOV l
4.
HOV-VS-100A R.C. Purge Supply MOV's 1
5.
MOV-VS-100B R.C. Purge Supply MOV's 6.
MOV-VS-102 contain. Vacuum Breaker 5tmos. Supply MOV 1
I Amendment iios. 72 & 73
TS 3.8-9 r
TABLE 3.8-1 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)
VALVE NUMBER FUNCTION E.
REMOTE MANUAL VALVES 1.
MOV-CS-101A Containment Spray Discharge Valve 2.
MOV-CS-101B Containment Spray Discharge Valve 3.
MOV-CS-101C Containment Spray Discharge Valve 4.
MOV-CS-101D Containment Spray Discharge Valve 5.
MOV-RS-155A Outside Recire. Spray Suction Valve 6.
MOV-RS-155B Outside Recire. Spray Suction Valve 7.
MOV-RS-156A Outside Recire. Discharge Valve 8.
MOV-RS-156B Outside Recire. Discharge Valve 9.
MOV-1842 Bypasses Boron Injec. Tank to Cold Leg Injec.
10.
MOV-RH-100 Resi. Heat Remov. to RWST 11.
FCV-1160 Loop Fill Header Flow Valve 12.
MOV-1890A Lo Header S. I. Pump Disch. from Hot Leg 13.
MOV-1890B Lo Header S. I. Pump Disch. from Hot Leg 14 MOV-1890C Lo Header S. I. Pump Disch. from Cold Leg 15.
MOV-1869A Iso. from Hot Leg to Hi Header S. I. Line A 16.
MOV-1869B Iso. from Hot Leg to Hi Header S. I. Line B 17.
MOV-1860A Iso. from Sump to Lo Header S. I.
18.
MOV-1860B Iso. Valve from Sump to Lo Header S. I.
19.
MOV-SW-104A*
MOV-SW-104B*
MOV-SW-104C*
SW to "C" HX's Amendment Nos. 72 & 73
. - - _ _ - -, - - - - ~. -.. - - - _ -.
L K 118-30 i
TABLE 3.8-1 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)
VALVE NUMBER FUNCTION 1
- 22. MOV-SW-104D*
- 23. MOV-SW-105A*
- 24. MOV-SW-105B*
- 25. MOV-SW-105C*
- 26. MOV-SW-105D*
- 27. HCV-CV-100 Cont. Vacuum Isolation i
F.
MANUAL VALVES 1.
1-SI-150 Boron Injection Tank 1" line 2.
1-SI-32 Accumulator Fill Valve 3.
1-GW-182 Discharge from Hydrogen Analyzer 4.
1-GW-183 Discharge from Hydrogen Analyzer i
l 5.
1-SA-60 Service Air to Containment 6.
1-SA-62 Service Air to Containment 7.
1-IA-446 Instrument Air to Containment 8.
1-VA-1 Outside Isolation from Primary Vent Pot 9.
1-VA-6 Inside Isolation from Primary Vent Pot 10.
2-IA-446 Cross Tie from #2 Instrument Air Header 11.
1-GW-175 Suction from Containment to H Analyzer 2
12.
1-GW-166 Suction from Containment to H Analyzer 2
13.
1-GW-174 Inlet to Cont. from H Analyzer Outside Cont.
2 14.
1-FP-151 Outside Iso. Vlv for Cont. Fire Protection l
15.
1-FP-152 Outside Iso. Viv for Cont. Fire Protection l
Amendment Nos. 72 & 73
i $~T.~b-i l 1
TABLE 3.8-1 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)
VALVENifMBER FUNCTION 16.
1-RL-3 Inlet V1v to Cavity from RCS Outside Cont.
17.
1-RL-5 Inlet Vlv to Cavity from RCS Inside Cont.
18.
1-RL-13 Suction Viv to 1-RL-P-1A Inside Containment 19.
1-RL-15 Suction V1v to 1-RL-P-1A Outside Containment 20.
1-SI-73 Accumulator N Fill Vlv Outside Containment 2
21.
1-SI-174 Bypasses MOV-1869A 22.
1-CV-2 Cont. Vacuum Isolation G.
CONTAINMENT CHECK VALVES 1.
1-FP-153 Inside Cont. - Fire Protection Header 2.
1-VP-12 Inside Cont. - Air Eject Disch to Cont.
3.
1-RS-17 Inside Cont. - RS Disch to Cont. A 4.
1-RS-11 Inside Cont. - RS Disch to Cont. B 5.
1-CS-13 Inside Cont. - Discharge of 1-CS-P-1A 6.
1-CS-24 Inside Cont. - Discharge of 1-CS-P-1B l
l 7.
1-IA-938 Inside Cont. - Disch of Cont. IA Component 8.
2-IA-446 Manual Valve - Disch. of IA Component Unit #2 9.
1-SI-234 Check Inside Cont. - N to Accumulator 2
10.
1-IA-939 Chech Inside Cont. - Disch. of Cont. lA Component Unit #1 11.
1-IA-446 Manual Vlv - Disch. of Unit 1 Instr. Air Comp.
l Amendment Nos. 72 & 73 l
L'
^
TS 3.8-12 TABLE 3.8-1 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)
VALVE NUMBER FUNCTION 12.
1-RC-160 Check Valve Inside Contain. from PG Supply 13.
1-RM-3 Check Valve Inside Contain. - Rad. Monitoring Suc.
14.
1-IA-939 Instr. Air Check Valve to Containment 15.
1-SA-446 Service Air Check Valve to Containment 16.
1-CC-177*
1-CC-176*
1-SI-225 HHSI from BIT 19.
1-CC-242*
CC to "A" Air Recire.
20.
1-CC-233*
CC to "B" Air Recire.
21.
1-CC-224*
CC to "C" Air Recire.
22.
1-CH-309 Normal Chg. Hdr 23.
1-CC-1*
1-CC-58*
1-CC-59*
1-SI-224 HHSI BIT Bypass 27.
1-SI-226 HHS1 to Hot Legs 28.
1-SI-228 LHSI Pp Discharge 29.
1-SI-229 LHSI Pp Discharge 30.
1-SI-227 LHSI to Hot Leg I
{
- - Not subject to Type "C" Testing.
- - Modifications to this table should be submitted to the NRC as part of the next license amendment.
l Amendment Nos. 72 & 73
TS 3.8 33 i
TABLE 3.8-2 UNIT NO. 2 CONTAINMENT ISOLATION VALVES 4
VALVE NUMBLR FUNCTION i
A.
PHASE I CONTAINMENT ISOLATION (SAFETY INJECTION SIGNAL) 1.
MOV-2867C Boron Injection Tank Outlet 2.
MOV-2867D Boron Injection Tank Outlet 3.
MOV-2289A Charging Line 4.
MOV-2381 Reactor Coolant Pump Seal Water Return 5.
HCV-2200A Letdown Orifice Isolation 6.
HCV-2200B Letdown Orifice Isolation 7.
HCV-2200C Letdown Orifice Isolation 8.
TV-SI-201A Accumulator N Relief Line 2
9.
TV-SI-201B Accumulator N Relief Line 2
10.
TV-SI-200 Accumulator N Relief Line 2
11.
TV-VG-209A Primary Drain Transfer Tank Vent 12.
TV-VG-209B Primary Drain Transfer Tank Vent 13.
TV-DG-208A Primary Drain Transfer Pump Discharge 14.
TV-DG-208B Primary Drain Transfer Pump Discharge 15.
TV-CC-209A*
Component Cooling from RHR's 16.
TV-CC-209B*
Component Cooling frcm RHR's 17.
TV-SS-200A Pressurizer Liquid Sample 18.
TV-SS-200B Pressurizer Liquid Sample 19.
TV-SS-201A Pressurizer Vapor Sample 20.
TV-SS-201B Pressurizer Vapor Sample Amendment Nos. 72 & 73 l
~ ~
TS 3.8-14 i
TABLE 3.8-2 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)
VALVE NUMBER FUNCTION 21.
TV-SS-203 Residual Heat Removal System Sample 22.
TV-SS-206A Reactor bolant Hot Leg Sample 23.
TV-SS-206B Reactor Coolant Hot Leg Sample 24.
TV-SS-202A Reactor Coolant Cold Leg Sample 25.
TV-SS-202B Reactor Coolant Cold Leg Sample 26.
TV-SS-204A Pressurizer Relief Tank Vapor Sample 27.
TV-SS-204B Pressurizer Relief Tank Vapor Sample 28.
TV-CH-2204 Letdown Isolation Valve 29.
TV-PG-2519A
, Primary Grade blater to Pressurizer Relief Tank 30.
TV-BD-200A*
Steam Generator Blowdown Valve 31.
TV-BD-200B*
Steam Generator Blowdown Valve 32.
TV-BD-200C*
Steam Generator Blowdown Valve 33.
TV-BD-200D*
Steam Generator Blowdown Valve 34.
TV-BD-200E*
Steam Generator Blowdown Valve 35.
TV-BD-200F*
Ster.m Generator Blowdown Valve 36.
TV-DA-200A Containment Sump Pump Isolation 37.
TV-DA-200B Containment Sump Pump Isolation 38.
TV-MS-209*
Main Steam Drain Trip Valve 39.
TV-MS-210*
Main Steam Drain Trip Valve 40.
TV-LM-200A Containment Isolation Monitoring 41.
TV-LM-200B Containment Isolation Monitoring 42.
TV-LM-200C Containment Isolation Monitoring Amendment Nos. 72 & 73
.. = -..- :-
TS 3.8-15 TABLE 3.8-2 UNIT NO. 2 CDNTAINMENT ISOLATION VALVES (Continued)
VALVE NUMBER FUNCTION 43.
TV-LM-200D Containment Isolation Monitorin, 44.
TV-LM-200E Containment Isolation Monitoring 45.
TV-LM-200F Containment Isolation Monitoring 46.
TV-LM-200G Containment Isolation Monitoring 47.
TV-LM-200H Containment Isohtion Monitoring 48.
TV-CV-250A Containment Vacuum Suction Valve 49.
TV-CV-250B Containment Vacuum Suction Valve 50.
TV-In-201A Leakage Monitoring Sealed Reference 51.
TV-LM-201B Leakage Monitoring Sealed Reference 52.
TV-CV-250C Containment Vacuum Suction Valve 53.
TV-CV-250D Containment Vacuum Suction Valve 54.
TV-SV-202A Condenser Air Ejector Vent Trip Valve B.
PHASE II CONTAINMENT ISOLATION (HI CLS SIGNAL) l l
1.
TV-RM-200A Containment Air & Particulate Rad. Mon. TV's 2.
TV-RM-200B Containment Air & Particulate Rad. Mon. TV's 3.
TV-RM-200C Containment Air & Particulate Rad. Mon. TV's 4.
TV-IA-201A Containment Instr. Air Compressor Suction I
5.
TV-IA-201B Containment Instr. Air Compressor Suction Amendment Nos. 72 & 73
~ ~'.
.J 1 :
TS 3.8-16 s
TABLE 3.8-2 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)
VALVE NUMBER FUNCTION C.
PHASE III CONTAINMENT ISOLATION (HI-HI CLS SIGNAL) 1.
TV-MS-201A*
Main Steam Trip Valve 2.
TV-MS-201B*
Main Steam Trip Valve 3.
TV-IA-200 Containment Instr. Air Compressor Disch. V1v.
4.
TV-MS-201C*
Main Steam Trip Valve 5.
TV-CC-207*
CC from RCP Thermal Parriers 6.
TV-CC-201A*
CC from A Air Recire.
7.
TV-CC-201B*
CC from B Air Recirc.
8.
TV-CC-201C*
CC from C Air Recire.
9.
TV-CC-205A*
TV-CC-205B*
TV-CC-205C*
CONTAINMENT PURGE & EXHAUST l
1.
MOV-VS-200C R.C. Purge Exhaust M0V's I
2.
MOV-VS-200D R.C. Purge Exhaust MOV's l
l 3.
MOV-VS-201 R.C. Purge Exhaust Bypass MOV l
4.
MOV-VS-200A R.C. Purge Supply MOV's 5.
Mr VS-200B R.C. Purge Supply MOV's 6.
MOV-VS-262 Contain. Vacuum Breaker Atmos. Supply MOV l
Amendment Nos. 72 & 73
15 3.8-17 TABLE 3.8-2 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Cantinued) i VALVE NUMBER FUNCTION E.
REMOTE MANUAL VALVES 1.
MOV-CS-201A Containment Spray Discharge Valve 2.
M0V-CS-201B Containment Spray Discharge Valve 3.
MOV-CS-201C Containment Spray Discharge Valve 4.
HOV-CS-201D Containment Spray Discharge Valve 5.
MOV-RS-255A Outside Recirculation Spray Suction Valve 6.
MOV-RS-255B Outside Recire. Spray Suction Valve 7.
MOV-RS-256A Outside Recirc. Discharge Valve 8.
MOV-RS-256B Outside Recire. Discharge Valve 9.
MOV-2842 Bypasses Boron Injec. Tank to Cold Leg Injec.
10.
MOV-RH-200 Resi. Heat Remov. to RWST 11.
FCV-2160 Loop Fill Header Flow Valve 12.
MOV-2890A Lo Header S.I. Pump Disch. from Hot Leg 13.
MOV-2890B Lo Header S.I. Pump Disch. from Hot Leg 14.
MOV-2890C Lo Header S.I. Pump Disch. from Cold Leg 15.
MOV-286?A Iso. from Hot Leg to Hi Header S. I. Line A 16.
MOV-2869B Iso. from Hot Leg to Hi Header S. I. Line B 17.
MOV-2860A Iso. from Sump to Lo Header S. I.
18.
MOV-2860B I,so. Valve from Sump to Lo Header S. I.
19.
MOV-SW-204A*
MOV-SW-204B*
MOV-SW-204C*
SW to "C" HX's Amendment Nos. 72 & 73
i TABLE 3.8-2 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)
VALVE NIJMBER FUNCTION
- 22. MOV-SW-204D*
- 23. MOV-SW-205A*
- 24. MOV-SW-205B*
- 25. MOV-SW-205C*
SW from "C" EX's
- 26. MOV-SW-205b-SW from "D" HX's
- 27. HCV-CV-200 Cont. Vacuum Isolation F.
MANUAL VALVES 1.
2-SI-150 Boron Injection Tank 1" line 2.
2-SI-32 Accumulator Fill Valve s
3.
2-GW-182 Discharge from Hydrogen Analyzer 4.
2-GW-183 Discharge from Hydrogen Analyzer 5.
2-SA-60 Service Air 6.
2-SA-62 Service Air 7.
2-IA-446 Instrument Air to Containment 8.
2-VA-1 Outside Isolation from Primary Vent Pot 9.
2-VA-6 Inside Isolation from Primery Vent Pot 10.
2-IA-446 Cross Tie from #1 Instrument Air Header 11.
2-GW-175 Suction from Cont. to H Analyzer 2
12.
2-GW-166 Suction from Cont. to H Aralyzer 2
13.
2-GW-174 Inlet to cont. from H Analyzer Outside Cont.
2 14.
2-FP-151 Outside Iso. Vlv for Cont. Fire Protection -
15.
2-FP-152 Outside Iso. Vlv for Cont. Fire Protection Amendment Nos. 72 & 73
g TABLE 3.8-2 UNIT NO. 2 CONTAINMENI ISOLATION VALVES (Continued)
VALVE NUMBER FUNCTION 16.
2-RL-3 Inlet V1v to Cavity from RCS Outside Cont.
17.
2-RL-5 Inlet Vlv to Cavity from RCS Inside Cont.
18.
2-RL-13 Suction Vlv to 2-RL-P-1A Inside Containment 19.
2-RL-15 Suction V1v to 2-RL-P-1A Outside Containment 20.
2-SI-73 Accumulator N Fill V1v Outside Containment 2
21.
2-SI-174 Bypasses MOV-1869A 22.
2-CV-2 Cont. Vacuum Isolation G.
CONTAINMENT CHECK VALVES 1.
2-FP-153 Inside Cont. - Fire Protection Header 2.
2-VP-12 Inside Cont. - Air Eject Disch to Cont.
3.
2-RS-17 Inside Cont. - RS Disch to Cont. A 4.
2-RS-11 Inside Cont. - RS Disch to Cont. B 5.
2-CS-13 Inside Cont. - Discharge of 2-CS-P-1A 6.
2-CS-24 Inside Cont. - Discharge of 2-CS-P-1B 7.
2-IA-938 Inside Cont. - Disch of Cont. IA Component 8.
2-IA-446 Manual Valve - Disch. of IA Component Unit #2 9,
2-SI-234 Check Inside Cont. - N l* A**"*"I**
2 10.
2-IA-939 -
Check Inside Cont. - Disch. of Cont. IA Component Unit #2 11.
2-IA-446 Manual Vlv - Disch. of Unit 2 Instr. Air Comp.
Amendment Nos. 72 & 73
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.c TABLE 3.8-2 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)
VALVENbMBER FUNCTION 12.
2-RC-160 Check Valve Inside Contain, from PG Supply 13.
2-RM-3 Check Valve Inside Contain. - Rad. Monitoring Suc.
14.
2-IA-939 Instr. Air Check Valve to Containment 15.
2-SA-446 Service Air Check Valve to Containment 16.
2-CC-177*
2-CC-176*
2-SI-225 HHSI from BIT 19.
2-CC-242*
CC to "A" Air Recire.
20.
2-CC-233*
CC to "B" Air Recire.
21.
2-CC-224*
CC to "C" Air Recire.
22.
2-CH-309 Normal Chg. Edr 23.
2-CC-1*
2-CC-58*
2-CC-59*
2-SI-224 HHSI BIT Bypass 27.
2 ",I-226 HHSI to Hot Legs 28.
2-SI-228 LHSI Pp Discharge 29.
2-SI-229 LHSI Pp Discharge 30.
2-SI-227 LHSI to Hot Leg Not subject to Type "C" Testing.
Modifications to this table should be submitted to the NRC as part of the next license amendment.
Amendment Nos. 72 & 73 i
i
TS 4.1-1 s
4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiticg conditions for operation.
Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.
Specification A.
Calibration, testing, and checking of instrumentation channels shall l
be performed as detailed in Table 4.1-1.
B.
Equipment tests shall be conducted as detailed below and in Table 4.1-2A.
i 1.
Each Pressurizer PORV shall be demonstrated operable:
a.
At least once per 31 days by performance of a channel functional test, excluding valve operation, and b.
At least once per 18 months by performance of a channel calibration.
l l
Amendment flos. 72 & 73
TS 4.1-la 2.
Each Pressurizer PORV block valve shall be demonstrated operable at least once per 92 days by operating the valve through one complete cycle of full travel.
3.
The pressurizer water volume shall be determined to be within its limit as defined in Specification 2.3.A.3.a at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the reactor is not suberitical by at least 1% ak/k.
C.
Sampling tests shall be conducted as detailed in Table 4.1-2B.
D.
Whenever containment integrity is not required, only the asterisked items in Table 4.1-1.and 4.1-2A and 4.1-2B are applicable.
E.
Flushing of sensitized stainless steel pipe sections shall be conducted as detailed in TS Table 4.1-3A and 4.1-3B.
O I
l l
Amendment Nos. 72 & 73
TABLE 4.1-1 MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS i
1.
Nuclear Power Range S
D (1)
BW(2)
- 1) Against a heat balance standard J
.M(3)
Q (3)
- 2) Signal of AT; bistable action (permissive, rod stop, strips)
- 3) Upper and lower chambers for symetrig offset by means of the moveable incor detector system.
2.
Nuclear Intermediate Range
- S(1)
N.A.
P(2)
- 1) Once/ shift when in service
- 2) Log level; bistable action (permissive, rod stop, trip)
]
3.
Nuclear Source Range
- S(1)
N.A.
P(2)
- 1) Once/ Shift when in service
- 2) Bistable action (alarm, trip) i 4.
Reactor Coolant Temperature
- S R
BW(1)
- 1) Overtemperature - aT BW(2)
- 2) Overpower - AT 5.
Reactor Coolant Flow S
R H
6.
Pressurizer Water Level S
R H
E 7.
Pressurizer Pressure (High &
S R
M l
Low) e 8.
4 Kv Voltage and Frequency S
R H
Reactor protection circuit only 9.
Analog Rod Position
- S(1,2)
R M(3)
- 1) With step counters DI E$
(4)
- 2) Each six inches of rod motion e=
n.
when data logger is out of ja
-a service es
- 3) Rod bottom bistable action
- 4) NA when reactor is in cold chut-down
TABLE 4.1-1 (Continued)
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 loRger is out of service
- 2) With analog rod position
)
11.
Steam Generator Level S
R H
12.
Chargir.g Flow N.A.
R N.A.
- 13. Residual Heat Removal Pump Flow N.A R
N.A.
- 14. Boric Acid Tank Level
- D R
N.A.
- 15. Refueling Water Storage S
R M
Tank Level
- 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
M(1)
- 1) Isolation Valve signal and spray signal 19.
Processing and Area Radiation
- D R
M Monitoring Systems o.
20.
Boric Acid Control N.A.
R N.A.
Eo 21. Containment Sump Level N.A.
R N.A.
- 22. Accumulator Level and Pressure S
R N.A.
U 23.
Containment Pressure-Vacuum Pump S
R N.A.
E' System i
o.
w 24.
Steam Line Pressure S
R H
TABLE 4.1-1 CHANNEL DESCRIPTION CIECK CALIBRATE TEST REMARKS
- 25. Turbine First Stage Pressure S
R M
- 26. Emergency Plan Radiation Instr.
- M R
M
- 27. Environmental Radiation Monitors
- M N.A.
N.A.
TLD Dosimeters
- 28. Logic Channel Testing N.A.
N.A.
M
- 29. Turbine Overspeed Protection N.A.
R R
l Trip Channel (Electrical)
- 30. Turbine Trip Setpoint N.A.
R R
Stop valve closure or low EH fluid pressure
- 31. Seismic Instrumentation M
SA M
- 32. Reactor Trip Breaker N.A.
N.A.
M
- 33. Reactor Coolant Pressure (Low)
N.A.
R N.A.
- 34. Auxiliary Feedwater a.
Steam Generator Water S
R M
Level Low-Low b*
+
b.
R H
co E.
Po c.
S.I.
(All Safety Injection surveillance requiremeets) 5
==
d.
Station Blackout N.A.
R N.A.
8 e.
Main Feedwater Pump Trip N.A.
N.A.
R S - Each shift M - Monthly o.
D - Daily P - Prior to each startup if not done previous week L
u" W - Weekly R - Each Refueling Shutdown
~
r NA - Not applicable BW - Every two weeks
(
SA - Semiannually AP - After each startup if not done previous week c.
Q - Every 90 effective full power days
- See Specification 4.1D
TABLE 4.1-2 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION I.
Auxiliary Feedwater Flow Rate P
R 2.
Reactor Coolant System Subcooling Margin Monitor M
R 3.
PORV Position Indicator (Primary Detector)
M R
4.
PORV Position Indicator (Backup Detector)
M R
5.
PORV Block Valve Position Indicator M
R 6.
Safety Val. Position Indicator M
R 7.
Safety Valve Position Indicator (Backup Detector)
M R
l.
a n
~
M*
w E
N (a
g i
1
TABLE 4.1-2A MINIMUM FREQUENCY FOR EQUIPMENT TESTS i
FSAR SECTION l
DESCRIPTION TEST FREQUENCY REFERENCE 5
1.
Control Rod Assemblies Rod drop times of all full length Each refueling shutdown or after 7
l rods at hot and cold conditions disassembly or maintenance re-l quiring the breech of the Reactor l
Coolant System integrity I
2.
Control Room Assemblies Partial movement of all rods Every 2 weeks 7
i l
3.
Refueling Water Chemical Functional Each refueling shutdown 6
i Addition Tank l
4.
Pressurizer Safety Valves Setpoint Each refueling shutdown 4
l 5.
Main Steam Safety Valves Setpoint Each refueling shutdown 10 i
6.
Containment Isolation Trip
- Functional Each refueling shutdown 5
f 7.
Refueling System Interlocks
- Functional Prior to refueling 9.12 8.
Service Water System
- Functional Each refueling shutdown 9.9 4
9.
Fire Protection Pump and Functional Monthly 9.10 Power Supply 10.
Primary System Leakage
- Evaluate Daily 4
g 11.
Diesel Fuel Supply
- Fuel Ir.ventory 5 days / week 8.5
@g-12. Boric Acid Piping Heat
- 0perational Monthly 9.1 g
Tracing "ircuits a
g 13. Main Steam Line Trip Functional 10 F
(1) Full closure (1) Each cold shutdown (2) Partial closure (2) Before each startup en
)
Jh U
L h
A s
TABLE 4.1-2A (CONTINUED)
MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION DESCRIPTION TEST FREQUENCY REFERENCE
- 14. Service Water System Valves Functional Each refueling 9.9 in Line Supplying Recircu-lation Spray Heat Exchangers 15.
Control Room Ventilation
- Ability to maintain positive pres-Each refueling interval 9.13 System sure for I hour using a volume of (approx. every 12-18 months) air equivalent co or less than stored in the bottled air supply
- 16. Reactor Vessel Overpressure Functional & Setpoint Prior to decreasing RCS None l
Mitigating System (except temperature below 350 F backup air supply) and monthly while the RCS is <350 F and the Reactor Vessel Head is bolted
- 17. Reactor Vessel Overpressure Setpoint Refueling Nons Mitigating System Backup Air Supply F
e" Ut 2
f*
m
TABLE 4.1-2A (CONTINUED)
MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION DESCRIPTION TEST FREQUENCY REFERENCE -
- 1. Periodic leakage testing (a)(b) on each valvQ 18.
Primary Coolant System Functional Pressure Isolation Valves listed in Specification 3.1.C.7a shall be accomplished prior to entering power operation condition after every time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomp-lished in the preceeding 9 months, and prior to returning the valve to service after maintenance, repair or replace-ment work is performed.
(a) To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators)
J if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
I 1
l (b)
Minimum differential test pressure shall not be below 150 psid.
l
- See Specification 4.1.D.
r B
d
+
=
M
- 4
'_N l
o.
j Es I
~6 TS 6.4-6 H.
Practice of site evacuation exercises shall be conducted annually, following emergency procedures and including a check of communications with off-site report groups. An annual review of the Emergency Plan will be performed.
I.
The industrial recurity program which has been established for the station shall be implemented, and appropriate investigation and/or corrective action shall be taken if the provisions of the program.re violated. An annual review of the program shall be performed.
J.
The facility fire protection program and implementing procedures wb2ch have been established for the station shall be implemented. The program shall be reviewed at least once every two years.
K Systems Integrity 1
The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.
This program shall include the following:
1.
Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2.
Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
Amendment Nos. 72 & 73
TS 6.4-7 L.
Iodine Monitoring The licensee shall implement a program which will ensure the capability to accurately determine the airborne. iodine concentration in vital area under accident conditions. This program shall include the following:
1.
Training of personnel,
/
e 2.
Procedures for monitoring, and 4
3.
Provisions for maintenance of sampling and analysis equipment..
Amendment Nos. 72 & 73
-,m,
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- * ~'"
. <~
g TS 6.6-4 r.
(1) A tabulation on an annual basis of the number
, i.
of station, utility and other personnel (includ-ing contractors) receiving exposures greater thau 100 mrem /yr and their associated man rem exposure according to work and job functions, e
~
e.g., reactor operations and surveillance, inser-vice inspection, routine maintenance, special maintenance (describe maintenance), waste pro-cessing, and refueling. The dose assignment to 1
s various duty functions may be estimates based
/
on pocket dosimeter, TLD, or film badge measure-ments. Small exposures totalling less than 20%
of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work func-tions.
i i
C.
Monthly Operating Report: Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the Reactor Coolant System PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Management and Program
[
I l
Analysis, U. S. Nuclear Regulatory Commission, Washington, D. C.
20555, with a copy to the Regional Office of Inspection and Enforce-ment, no later than the 15th of each month following the calendar month covered by the report.
I s_.
l l
A Amendment Nos. 72 & 73 1
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