ML17321A813
| ML17321A813 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/09/1985 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17321A816 | List: |
| References | |
| NUDOCS 8508190485 | |
| Download: ML17321A813 (24) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-315 DONALD C.
COOK NUCLEAR PLANT UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.88 License No.
DPR-58 2.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
.The applications for amendments by Indiana and Michigan Electric Company (the licensee) dated July 18, 1985 and July 19,
- 1985, as supplemented by letter dated July 3, 1985, compTy 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; B.
The. facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be 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.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-58 is hereby amended to read as follows:
.. 850819'0485 850809'g PDR ADDCH, 050003i5
., 'P 'DB
J'
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 88
, are hereby incorporated in-the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The change to the Technical Specifications is to be effective within 30 days of issuance.
4.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
August g, )gB5 S even A. Varga, Chief Op rating Reactors Branch ¹I Division of Licensing
ATTACHMENT TQ LICENSE AMENDMENT AMENDMENT NO.
'88 FACILITY OPERATING LICENSE NO.
DPR-58 DOCKET NO. 50-315 Revise Appendix A as follows:
Remove Pa es Insert Pa es 3/4 l-ll 3/4 4-3 3/4 4-3d 3/4 4-27 3/4 4-28 3/4 4-31 3/4 5-7 3/4 5-8 B 3/4 1-3 B 3/4 4-1 B 3/4 4-5*
B 3/4 4-6 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-10 B 3/4 4-11 B 3/4 5-2 3/4 1-11 3/4 4-3 3/4 4-3d 3/4 4-27 3/4 4-28 3/4 4-31 3/4 5-7 3/4 5-8 B 3/4 1-3 B 3/4 4-1 B 3/4 4-5*
B 3/4 4-6 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-10 B 3/4 4-11 B 3/4 5-2
- Included for convenience.
No changes.
REACTIVITY CONTROL SYSTEMS CHARGING PUMP -
SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.
APPLICABILITY; MODES 5 and 6.
ACTION:
a.
b.
C.
With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity.
With more than one charging pump OPERABLE or with a safety injection pump(s)
OPERABLE when the temperature of any RCS cold leg is less than or equal to 17O~F, unless the reactor vessel head is removed, remove the additional charging pump(s) and the safety injection pump(s) motor circuit breakers from the electrical power circuit within one hour.
The provisions of Specification 3.0.3 ar not applicable.
SURVEILLANCE REQUIREMENTS
- 4. 1.2.3.1 The above required charging pump shall be demonstrated OPERABLE at least once per 31 days by:
a.
Starting (unless already operating) the pump from the control
- room, b.
Verifying, that on recirculation flow, the pump develops a discharge pressure of > 2390 psig, c.
Ver'.fying pump operation for at least 15 minutes, and d.
Verifying that the pump is aligned to receive electrical power from an OPERABLE emergency bus.
4.1.2.3.2 All charging pumps and safety iniection pumps, excluding the above required OPERABL charging pump, shall be demonstrated inoperable by verifying that the motor circuit breakers have been removed from their electrical power supply circuits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, except when:
a.
The reactor vessel head is removed, or b.
The temperature of all RCS cold legs is greater than 17O F.
D.
C.
COOK - UNIT 1 3/4 1-11 Amendment No. 88
RE".C":3R COOLANT SYSTEM SHUTDOWN LIMI '.NG CONDITION FOR OPERATION 3.4.1.3 a.
At least two of the coolant loops listed below shall be OPERABLE:
1.
Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,"
2.
Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,*
3.
Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,>>
~ e
~
4.
Reactor Coolant Loop 4 and its associated steam generator and..
reactor coolant pump,*
5'.
Residual Heat Removal - East,~
6.
Residual Meat Removal - West,~
b.
At least one of the above coolant loops shall be in operation.*~
APPLICABILITY:
MODES 4 and 5
ACTION:
a.
With less than the above required loops OPERABLE, imediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN with-in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With no coolant loop in. operation, suspend all operations in-volving a reduction in boron concentration of the Reactor Cool-ant System and immediately initiate corrective a tion to return the required coolant loop to operation.
A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 176 ~F unless
- 1) the pressurizer water volume is less than ~2.00" of span or 2) the secondary water temperature of each steam generator is less than 50 oF above each of the RCS cold leg t mper-atures.
Operability of a reactor coolant loop(s) does not require an OPERABLE auxiliary feedwater system.
- The normal or emergency power source may be inoperable in NODE 5.
- All reactor coolant pumps and residual heat removal pumps may be de-energizea or up
~o 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> orovided 1) no operatisons
=re pe!iitted that wou;d ":-use dilution of the reac.or coolant system boron concentration, and 2j cor cut-let temperature i" maintained at least li0 oF below satura.
on t~i;perature.
0.
C.
Cook - Unit 1
3/4 4-3 Amendment No. 88
REACTOR COOLANT SYSTEM ACTION. Continued L
Below P-7:0 a.-
b.
C.
Startup and Power operation below P-7 may proceed provided at least two reactor coolant loops and associated pumps are in operation.
Hot standby, hot shutdown, and cold shutdown operation may proceed provided at least one reactor coolant loop in operation with an associated reactor coolant or residual heat r'emoval pump; however, operation for up to 15 minutes with no pump in operation is permissible to accommodate transition between residual heat removal pump and reactor coolant puntp operation.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.1.4.1 With one reactor coolant loop and associated pump not in operation, at least once per 7 days determine that:
a ~
b.
The applicable reactor trip system and/or ESF actuation system instrume'ntation channels specified in the ACTION statements above have been placed in their tripped conditions, and If P-8 interlock setpoint has been reset for 3 loop operation, its setpoint is
< 76X of RATED THERMAL POWER.
4.4.1.4.2 Within 30 minutes prior to the start of a reactor coolant pump when any RCS cold leg temperature is
< 170 'F, verify that:
I a.
The <emperature of the secondary water of each steam generator is
< 504F above the temperature of each of the RCS cold legs, or b.
The pressurizer water volume is less than 1116 cubic feet, equivalent to less than 62 indicated on the wide range. level indicator.
A reactor coolant pump shall not be started with one or more.of the RCS cold leg temperatures less than'r equal to 170 OF unless
- 1) the pressurizer water volume is less than 1116 cubic feet j62>> of span or 2) the secondary water temperature of each steam generator is less than 50'F above each of the'CS cold leg temperatures.
D.C.
COOK - UNIT 1
3/4 4-3d If I
A;"..endment No. 88
a n
n00 7C 2800 2600 2400 2200 REACTOR COOLANT SYSTEM HEATUP LIMITATIONSAP-PLICABLE FOR FIRST 12 EFFECTIVE FULL POWER YEARS (MARGINS OF 60 PSIG AND 104F ARE IN-CLUDED FOR POSSIBLE INSTRUMENT ERROR.)
LEAK TEST LIMIT 4l Ihl CO De ls lt C4I
~J CO
~J C
0 O
CP lo O
V Oo 2000 1800 1600 1400 1200 1000 800 600 400 MATERIAL PROPERTY BASIS WELD METAL CU ~ 0.31%,
P~ 0.017%
INITIALRT
~ 234oF NDT (3/4T)
= 117oF PRESSURE-TEMPERATURE LIMIT FOR HEATUP RATES UP TO 60 F/HR UNACCEPTABLE OPERATION ACCEPTABLE OPERATION CRITICAL1TY LIMIT 200 60
~
100 150 200 250 300 350 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE
( F)
Figure 3.4-2
- 400, 450 I
I' CO 00 REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE LIMITS VERSUS 60oF/HOUR RATE CRITICALITY AND HYDROSTATIC TEST LiKMIT
O00 2800 2600 2400 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS AP-PLICABLE FOR FIRST 12 EFFECTIVE FULL POWER "IEARS.
(MARGINS OF 60 PSIG AND 10oF ARE IN-CLUDED FOR POSSIBLE INSTRUMENT ERROR.)
2200 CP CO Oe e
Clle De 8V el CO
~J C
0 O
CP 404fV Q
04 2000 1800 1600 1400 1200 1000 800 600 400 MATERIAL PROPERTY BASIS WELD METAL CU ~ 0.31%,
P = 0.017%
INITIALRT
~ 234 F
NDT (3/4T)
= 117 F
PRESSURE TEMPERATURE LIMITS COOLDOWN RATE oF/HR 0 F/Hr UNACCEPTABLE OPERATION ACCEPTABLE OPERATION 200 60 F/
I>
Hr 60 100 150 200 100 F/H 250 300 350 400
'.450 0
AVERAGE REACTOR COOLANT SYSTEM TEHPERATURE
( F)
Figure 3.4-3 e
REACTOR COOZdWT SYSTEM PRESSURE - TEMPERATURE LIMITS VERSUS COOLDOWN RATES
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONOITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:
a.
Two power operated relief valves (PORVs) with a lift setting of less than or equal to 400 psig, or b.
C.
One power operated relief valve (PORV) with a lift setting of less than or equal to 400 psig and the RHR safety va1ve with a lift setting of less than or equal to 450 psig, or A reactor coolant system vent of greater than or equal to 2 square inches.
APPLICABILITY:
When the temperature of one or more of the RCS cold legs is less than or equal to 170 F, except when the reactor vessel head is removed.
ACTION:
a.
With two PORV's inoperable or with one PORV inoperable and the RHR safety valve inoperable, either restore the inoperable PORV(s) or RHR safety valve to OPERABLE status within 7 days or depressurize and vent the RCS through an at least 2 square inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until the inoperable PORV or RHR safety valve has been restored to OPERABLE status.
b.
With both PORVs inoperable, depressurize and vent the RCS through an at least 2 square inch vent(s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both PORVs or one PORV and the RHR safety valve have been restored to OPERABLE status.
C.
In the event either the
RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.
The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent(s) on the transient and any corrective action necessary to prevent recurrence.
d.
The provisions of Specification 3.0.4 are not applicable.
O.C.
COOK - UNIT 1 3/4 4-31 Amendment No. 88
EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEHS - T
< 350 F
av LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
a.
One OPERABLE centrifugal charging pump;0 b.
One OPERABLE residual heat removal heat exchanger, C.
d.
One OPERABLE residual heat removal
- pump, and J
An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually reali'gned and transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY:
MODE 4.
ACTION:
a.
With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging
'pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or main-tain the Reactor Coolant System T
less than 350 F by use of avg alternate heat removal methods.
C.
d.
With more than one charging pump OPERABLE or with a safety injection pump(s)
OPERABLE when the temperature of any RCS cold leg is less than or equal to I70 F,
remove the additional charging pump(s) and the safety injection pump(s) motor circuit breakers from the electrical power circuit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date.
0A maximum of one centrifugal charging pump shall be OPERABLE and both safety injection pumps shall be inoperable whenever the temperature of one or more of the RCS cold legs is less than or equal to I7Q F.
D.
C.
COOK -.UNIT 1 3/4 5"7 Amendment No.
88
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.
4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE charging pump, shall be demonstrated inoperable, by verifying that the motor circuit breakers have been removed from their electrical 'power supply circuits, 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 temperature of one or more of the RCS cold legs is less than or equal to 170~F as determined at least once
- .per.>our when any RCS cold leg temperature is between I7O~F and 200'F.
D.
C.
COOK-- 4NIT 1 3/4 5"8 Amendment No.
REACTIVITY CONTROL SYSTEMS BASES L
BORATION SYSTEMS Continued
~With the RCS average temperature above 200~F, one injection system is acceptable without single failure consideration on the basis, of,the stable reactivity condition of the reactor and the additional restr,ictions prohibiting CORE ALTERATIONS and positive reactivity change in the event, the single injection system becomes inoperable.
""The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injec" tion pumps, except the required OPERABLE charging pump, to be inoperable below
>70'F, unless the reactor vessel head is removed, provides assurance that a
mass addition pressure transient can be relieved by the operation of a single PORV.
The boration capability required below 200 F is sufficient to provide a
SHUTDOWN MARGIN of 1X hk/k after xenon decay and cooldown fr'om 2004F to 140'F.
This condition requires either 835 gallons of 20,000 ppm borated water from the boric acid storage tanks of 9690 gallons of 1950 ppm borated water from the refueling water storage tank.
3/4. 1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are mainta'ined, (2) the minimum SHUTDOWN MARGIN is main-
- tained, and (3) limit the potential effects of rod ejection accident.
OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.
The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met.
Misalignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restric-tions provide assurance of fuel rod integrity during continued operation.
The reactivity worth of a misaligned rod is limited for the remainder of the fuel cycle to prevent exceeding the assumptions used in the accident analysis for a rod ejection accident.
The maximum rod drop time restriction is consistent 'with the assumed rod drop time used in the accident analyses.
Measurement with T
> 5414F and avg-with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
/,
I D.
C.
COOK - UNIT 1 B 3/4 1-3 Amendment No.
0 t ~
I 4
t lt
3/4.4 REACTOR COOLANT SYSTEM B<SES 3!4.'.1 R
ACTOR C"OLAN.
LOOPS The plant is designed to operate with all reac."r coolant loocs in ooeration, and main ain DNBR above 1.30 during all normal operations and anticipated transients.
With one reactor coolant loop not in operation, THERMAL POWER is restricted to
< 51 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset.
Either action ensures that the O'NBR wi 11 be maintained above 1.30.
A loss of flow in. wo loops will cause a reactor trip if ope, ating aoove P-7 {ll percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a'eac or trip if operating above P-B {51 percent of RAT=D THtRMA'CWER)
In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.
In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump below P-7 with one or more RCS cold legs less than or equal to 170 F are provided to prevent RCS pressure transients, caused by energy additions from the secondary
- system, which could exceed the limits of Appendix G to 10 CFR part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either
( 1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCp's to when the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.
3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point.
The relief capaci,ty of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop
, connected to the
0.
C.
COOK - UNIT 1
S 3/i i-1 Amendaent No.
88
REACTOR COOLANT SYSTEM BASES The surveillance requirements provide adequate assurance that con-centrations in excess. of the limits will be detected in sufficient time to take corrective action.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.
The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site para-meters of the D. C.
Cook site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site.
This reevaluation may result in higher limits.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/gram DOSE E(UIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
Operation with specific activity levels exceeding 1.0 pCi/gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be restricted to no more than 10 percent of the unit's yearly operating time since the activity levels allowed by Figure 3.4-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.
Reducing T
to (500'F prevents the release of activity should a
steam generator Nbe rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.
A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
D. C.
COOK-UNIT 1
B 3/4 4-5
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REACTOR COOLANT SYSTEM BASES The 12 EFPY heatup and cooldown curves were developed based on the following:
1.
The core beltline weld material being the limiting material with a copper and phosphorus content of.31$ and
.017K.
2.
The projected fluence values contained in Table XII of the Southwest Research Institute report, "Reactor Vessel Material Surveillance Program for Donald C.
Cook Unit No. 1, Analysis of Capsule Y," dated January 1984.
3.
Figure 1, NRC Regulatory Guide 1.99, Revision 1
The shift in RT of the vessel material will be established periodically during IIPration by removing and evaluating reactor vessel material irradiation surveillance specimen dosimetry installed near the inside wall of the reactor vessel.
The projected fluence values obtained will be used to calculate the change in RTNDT in accordance with Regulatory Guide 1.99, Revision l.
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure complianc@ with the minimum temperature requirements of Appendix G
to 10 CFR 50.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements o'f Appendix H to 10 CFR Part 50.
The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in
, accordance with the ASME Code requirements.
The OPERABILITY of two PORVs, one PORV and the RHR safety valve, or an RCS vent opening of greater than or equal to 2 square inches ensures that the RCS wi11 be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are:less than or aqua'I to 17g'F.
Either PQRV or RHR safety valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water tempera-ture of the steam generator less than or equal to 50'F above the RCS mold leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS.
D.
C.. COOK - UNIT 1 B 3/4 4-7 Amendment No. 88
INTENTIONALLY LEFT BLANK D. C.
COOK UNIT I B 3/4 4-8 Amendment No.
88
INTENTIONALLY LEFT BLANK D. C.
COOK UNIT 1 B 3/4 4-9 Amendment No 88
4 0
I
'c"
~
INTENTIONALLY LEFT BLANK D. C.
COOK - UNIT 1 B 3/4 4-10 Amendment No. 88
INTENTIONALLY LEFT BLANK D. C.
COOK - UNIT 1 B 3/4 4-11 Amendment No 88
EMERGENCY CORE COOLING SYSTEMS BASES With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.
The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injec" tion pumps, except the required OPERABLE charging pump, to be inoperable below
'7O~F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the'afety analyses are met and that subsystem OPERABILITY is maintained.
Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.
Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:
(1) prevent total pump flow from e/ceeding runout condi-tions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assump-tions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.
3/4.5.4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown.
RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture.
The limits on injection tank minimum contained volume and boron concentration ensure that the assumptions used in the steam line break analysis are met.
The OPERABILITY of the redundant heat tracing channels associated with the boron injection system ensure that the solubility of the boron solution will be maintained above the solubility limit of 135'F at 21000 ppm boron.
D.
C.
COOK " UNIT 1 B 3/4 5"2 Amendment No.
88