ML17331A520
| ML17331A520 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/10/1980 |
| From: | INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | |
| Shared Package | |
| ML17331A518 | List: |
| References | |
| AEP:NRC:00449, AEP:NRC:449, NUDOCS 8012180321 | |
| Download: ML17331A520 (21) | |
Text
ATTACHMENT 3 TO AEP:NRC:00449 Proposed Technical Specifications Changes for Unit 2
PLANT SYSTEMS SURVEILLANCE RE UIREMENTS
'4.
Verifying that each automatic valve in the flow path is in the fully open position whenever the auxiliary feedwater system is placed in automatic control or when above lOX RATED THERMAL POWER.
b.
At least once per 18 months during shutdown by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of the appropriate Engineered Safety Features actuation test signal required by Specification 3/4.3.2.
2.
Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of the appropriate Engineered Safety Features actuation test signal required by Specification 3/4.3.2.
D. C.
COOK - UNIT 2
3/4 7-6
TABLE 3. -3 Cont'd.
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION n
n C)
ED pc FUNCTIONAL UNIT 6.
MOTOR DRIVEN AUXILIARYFEEDWATER PUMPS TOTAL NO.
CHANNELS OF CHANNELS TO TRIP
'MINIMUM CHANNELS APPLICABLE OPERABLE MODES ACTIN a.
Steam Generator Water Level-Low-'w 2/Stm.
Gen.
3/Stm.
Gen.
Any Stm. Gen.
2/Stm.
Gen.
1, 2, 3 14*
b.
4 kv Bus Loss of Voltage c.
Safety Injection d.
Loss of Main Feedwater Pumps 2/Bus 2/Bus 2/Bus 1, 2, 3
1, 2, 3
1I 2,3 14*
I O
Al 7.
TURBINE DRIVEN AUXILIARYFEEDWATER PUMPS a.
Steam Generator Water Level-Low-Low 2/Stm.
Gen.
2/Stm.
Gen.
3/Stm.
Gen.
Any 2 Stm. Gen..
1, 2, 3.
14>>
.b.. Reactor Coolant Pump Bus Undervoltage 4-1/Bus 2.
1, 2, 3 8.
LOSS OF POWER a., 4 kv'Bus Loss of Voltage b.
4 kv Bus Degraded Voltage 3/Bus 3/Bus 2/Bus 2/Bus 2/Bus 2/Bus l~ 2, 3, 4 1, 2, 3, 4 14*
14>>
c
r TABLE 3.
Cont'd.
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS n
CD CD 7C I
FUNCTIONAL UNIT 6.
MOTOR DRIVEN AUXILIARYFEEDWATER PUMPS a.
Steam Generator Water Level-Low-Low b.
4 kv Bus Loss of Voltage c,
Safety Injection d.
Loss of Main Feedwater Pumps TRIP SETPOINT
> lOX of narrow range instrument span each steam generator 3196 volts with a 2-second delay Not Applicable Not Applicable ALLOWABLE VALUES.
> 9X of narrow range instrument span each steam generator 3196 a 18 volts. with a
2 x.2 second delay Not Applicable Not Applicable 7.
TURBINE DRIVEN AUXILIARYFEEDWATER PUMPS.
a.
Steam Generator Water Level--
Low-Low b.
Reactor Coolant Pump Bus Undervoltage
> 10$ of narrow range Tnstrument span each -':
,'; + 2750 Volts-each bus
> 9X of narrow range Tnstrument span each steam generator t >2725 Volts-each bus 8.
LOSS OF POWER a.
4 kv Bus Loss of Voltage b.
4 kv Bus Degraded Voltage 3196 volts with a 2-second delay 3596 volts with a 2.0 min. time delay 3196 a 18 volts with a 2 a.2 second delay 3596 k 18 volts with a 2.0 minute a 6 second time delay
TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES
RESPONSE
TIMES
RESPONSE
TIME IN SECONDS
< 12.0f/24.0H
< 2.0
< 8.0
< 18.08'/28.088 Not Applicable
< 60.0
< 14.08/48.0N c 8.0 INITIATING SIGNAL AND FUNCTION 6.
Steam Line Pressure Low a.
Safety Injection (ECCS) b.
Reactor Trip (from SI) c.
Feedwa ter Iso 1 ation d.
Containment Isol ation-Phase "A"
'e.
Containment Purge and Exhaust Isolation f.
Motor Driven Auxiliary Feedwater Pumps Essential Service Mater System h.
Steam Line Isolation 7.
Containment Pressure--High-High a.
Containment Isolation-Phase "8"
c.
Steam Line Isolation d.
Containment Air Recirculation Fan
< 45.0 Not Applicable
< 7.0
< 600.0
, 8.
Steam'Generator Mater Level--High-Hioh a.
Turbine Trip-Reactor Trip b.
Feedwater Isolation Not Applicable Not Applicable 9.
Steam Generator Mater LevelLow-Low a.
Motor Driven Auxiliary Feedwater Pumps
< 60.0 b.
Turbine Driven Auxiliary Feedwater Pumps
< 60.0 10.
4160 volt Emergency 8us Loss of Voltage a.
Motor Driven Auxiliary Feedwater Pumps
< 60.0 ll.
Loss Of Main Feedwater Pum s
a.
Motor Driven Auxiliary Feedwater Pumps 12.
Reactor Coolant Pumo Bus Undervolta';e a.
Turbine Driven Auxiliary Feedwater Puaips c 60.0 D.C.
COOK - UNIT 2 3/4 3-28
n TABLE 4.3-2 Continued EHGIHEEREO SAFETY FEATURE ACTUATIOH SYSTEM INSTRUHENTATIOH
~u~~r. Z mi:arr~
n CD C'C I
FUNCTIONAL UNIT CIIAHHEL CllLCK CIIRHHEL CALIORATIOH CHANNEL NODES IN MHICH'-
FUHCTIOHAL
'URYEILLANCE TEST RE UIRED 4.
STE/N LINE ISOLATION a.
Manual b.
Automatic Actuation Logic.
c.
Containment Pressu're--
Nigh-lligh N.A.
H.A.
H.A.
M(1)
M(2)
M(3) 1,2,3 1,2,3 1,2,3 d.
Steam Floe in Taco Steam;.
S Lines--lligh Coincident with T
-- Low or Steam Line Pf @sure--Low 5.
TURBINE TRIP AHO FEEOWATER ISOLATION a.
Steam Generator Water Level illgh-illgh 6,
MOTOR ORIVEH AUXILIARY FEEDWATER PUMPS, Steam Generator Water S
Level--Low-Low b.
4 kv Bus Loss of Voltage M
1,2,3 1,2,3 1,2,3 1,2,3 c.
Safety Injection I
d.
Loss of Hain Feed Pumps N.A.
N.A.
N.A,
. N.A.
H(2) 1, 2, 3
~
1, 2, 3
TABLE
. -2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE U IREMENTS FUNCTIONAL UNIT 7.
TURBINE DRIVEN AUXILIARYFEEDWATER PUMPS a.
Steam Generator Water Level--Low-Low b..
Reactor Coolant Pump Bus Undervoltage CHANNEL CHECK N.A.
CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE RE UIRED 1,2,3 1,2,3 b.
4 kv Bus Degraded Voltage 8.
LOSS OF POWER a.
4 kv Bus Loss of Voltage 1,2,3,4 1,2,3,4
T
.3-IO POST-ACCIDENT MO ORING INSTRUMENTATION INSTRUMENT MINIMUM CHANNELS OPERABLE l.
Containment Pressure 2.
Reactor Coolant Outlet Temperature - THOT (Wide Range) 3.
Reactor Coolant Inlet Temperature -
TCOLD (Wide Range) 4.
Reactor Coolant Pressure
- Wide Range 5.
Pressurizer Water Level 6.
Steam Line Pressure 7.
Steam Generator Water Level - Narrow Range 8.
Refueling Water Storage Tank Water Level 9.
Boric Acid Tank Solution Level 10.
Auxiliary Feedwater Flow Rate ll.
Reactor Coolant System Subcooling Margin Monit6r 12.
PORV Position Indicator - Limit Switches***
13.
PORV Block Valve Position Indicator - Limit Switches 14.
Safety Valve Position Indicator - Acoustic Monitor 2
2 2/Steam Generator 1/Steam Generator 1/Steam Generator>>
1*>>
1/Valve 1/Valve 1/Valve
- Steam Generator Water Level Channels can be used as a substitute for the corresponding auxiliary feedwater flow rate channel instrument.
- PRODAC 250 subcooling margin readout can be used as a substitute for the subcooling monitor instrument.
~TBBEE 4.3-3 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLAN E RE UIREMENTS INSTRUMENT l.
Containment Pressure 2.
Reactor Coolant. Outlet Temperature - THOT (Wide Range) 3.
Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 4.
Reactor Coolant Pressure - Wide Range 5.
Pressurizer Water Level 6.
Steam Line Pressure 7.
Steam Generator Water Level - Narrow Range 8.
RWST Water Level 9.
Boric Acid Tank Solu
'.on Level 10.
Auxiliary Feedwater Flow Rate ll.
Reactor Coolant System Subcooling Margin Monitor 12.
PORV Position Indicator - Limit Switches 13.
PORV Block Valve Position Indicator - Limit Switches 14.
Safety Valve Position Indicator. - Acoustic Monitor CHANNEL CHECK M
CHANNEL CALIBRATION
RECTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume less than or equal to 62/ of span and at least 150 kW of pressurizer heaters.
APPLICABILITY:
MODES 1, 2, and 3
ACTION:
With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply 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 STANDBY 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 HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With the pressurizer otherwise inoperable, be in at least HOT SHUTDOWN with the reactor trip breakers open within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE RE UIREMENTS 4.4.4.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.4.2 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and energizing the required capacity of heaters.
D.
C.
COOK - UNIT 2 3/4 4-6
'LIEF VALVES - OPERATING cJ g)
LIMITINGCONDITION FOR OPERATION 3.4.11 Three Power Operated Relief Valves (PORVs) and their associated block valves shall be OPERABLE.
APPLICABILITY:
MODES 1, 2 and 3
ACTION:
With one PORV inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, be in at least HOT STANDBY 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 />.
The provisions of Specifications 6.9.1.9, 3.0.3. and 3.0.4 are not applicable.
b.
Ce d.
With two or more PORVs inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore
, the PORVs to OPERABLE status or close the associated block valves and remove power from the block valves; otherwise, be in at least HOT STANDBY 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 COl D SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With one block valve inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either (1) restore the block valve to OPERABLE status or (2) close the block valve and remove power from the block valve or (3) close the associated.
PORV and remove power from its associated Solenoid valve~otherwise, be in at least HOT STANDBY 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 />.
The provisions of Specifications 6.9.1.9, 3.0.3 and 3.0.4 are not applicable.
With two or more block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either (1} restore the block valves to OPERABLE status or (2) close the block valves and remove power from the block valves or (3) close the associated PORVs and remove power from their associated Solenoid valves; otherwise, be in at least HOT STANDBY 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 />.
SURVEILLANCE REQUIREMENTS 4.4.11.1 Each of the three PORVs shall be demonstrated OPERABLE:
a ~
b.
At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, excluding valve operation, and At least once per 18 months by performance of a CHANNEL CALIBRATION.
D. C.
COOK UNIT 2 3g4 4 30
J
~
SURVEILLANCE RE UIREMENTS Cont'd
'4.4.)1.2 Each of the three block valves shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.
The provisions of Specification 4.0.4 are not applicable when Actions 3.4.1l.a or 3.4.1l.c are applied.
4.4.11.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through a complete cycle of full travel while the emergency buses are energized by the on-site diesel generators and on-site plant batteries.
This testing can be performed in conjunction with the requirements of Specifications 4.8.1.1.2.b and 4.8.2.3.2.c.
D-.C.
COOK UNIT2
~
~
l 3/4 4-3I
TABI E 3.6-1 (Continued)
CONTAINMENT ISOLATION VALVES VALVE NUMBER A.
PHASE "A" ISOLATION 61.
NCR-252 62.
QCM-250 63.
QCM-350 64.
QCR-300 65.
QCR-301 66.
RCR-100 67.
RCR-101 68.
VCR-10 69.
VCR-11 70.
VCR-20 7 1.
VCR-21 7 2.
XCR-100 7 3.
XCR-101 FUNCTION (Continued)
Primary Water to Pressurizer Relief Tank RCP Seal Water Discharge RCP Seal Water Discharge Letdown to Letdown Hx.
Letdown to Letdown Hx.
PRZ Relief Tank to Gas Anal.
PR2 Relief Tank to Gas Anal.
Glycol Supply to Fan Cooler Glycol Supply to Fan Cooler Glycol Supply from Fan Cooler Glycol Supply from Fan Cooler Control Air to Containment Control Air to Containment Isolation ISOLATION TIME IN SECONDS
< 10
< 15
< 15
< 10
< 10
< 10
< 10
< 10 l I
< 10
< 10
< 10
< 10
< 10
v
REACTOR COOLANT SYSTEM BASES Oemonstration of the safety valves'ift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME(Boiler and Pressure Code.
3 4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a
hydraulically solid system and is capable of accormodating pressure surges during operation.
The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief.
The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.
Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
The requirement that 150 KN of pressurizer heaters and their associated controls be capable of being sappl'/ed electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of-offsite power conditi on to maintain natural circulation at HOT STANOBY.
3 4.4.5 STEAM GENERATORS e
urve>
ance
. eau>rements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the condi tions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice condi tions that lead to cor-rosion.
Inservice inspection of steam generator tubing also provides a
means of char acterizing the nature and cause of any tube degradation so
~
that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secon-dary coolant will be maintained within those chemistry limits.found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage
= 500 gallons.per day per steam generator ).
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Operating plants nave demons-rated that primary-to-secondary leakage of 500 gallons per day per s-.earn gene. ator can readily
- 0. C.
COOK - UNIT 2 B 3/4 4-2
TABLE 3.6-1 (Continued)
CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION A.
PHASE "A" ISOLATION (Continued)
ISOLATION TIME IN SECONDS 74.
XCR-102 75.
XCR-103 B.
Phase "B" ISOLATION Control Air to Containment Isolation Control Air to Containment
< 10
< 10 1.
CCM-451 CCM-452 3.
CCM-453 4.
CCM-454 5.'CM-458 6.
CCM-459 7.
ECR-31 8.
ECR-32 CCW from RCP Oil Coolers CCW from RCP Oil Coolers CCW from RCP Thermal Barrier CCW from RCP Thermal Barrier CCW to RCP Oil Coolers 5 Thermal Barrier CCW to RCP Oil Coolers 8 Thermal Barrier
< 60
< 60
< 30
< 30
< 60
< 60 Containment Air Particle Radio Gas Detector
< 10 Containment Air Particle Radio Gas Detector
< 10
TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITIONP LICENSE CATEGORY SOL OL Non-Licensed 1, 2, 3
& 4 5
& 6 APPLICABLE MODES Shift Technical Advisor None Required
- Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS.
$Shift crew composition may be less than the minimum requirements for a period of time n'ot to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on duty shift crew members provided immediate action is 'taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.
- Shared With O.C.
COOK - UNIT 1.
Ql ~
t
- 0. C.
COOK - UNIT 2 6-4
ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF UALIFICATIONS 6.3.1 Each member o
the facility staff shall meet or exceed the minimum quali.
cations of ANSI N18.1-1971 for comparable positions, e"cept for (1) the Radiati<
Protection Supervisor who shall meet or exceed the qualifications of Regulatory uide 1.8, September 1975 and (2) the Shift Technical Advisor who shall have a
bachelor's degree or equivalent in a scientific or engineering discipline with pecific training in plant design, and response and analysis of the plant for ansients an~ accidents.*
6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Plant Manager and shall meet or exceed the require-ents of Section 27 of the NFPA Code-1976.
6.5 REVIEW ANO AUOIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE PNSRC FUNCTION 6.5.1.1 The PNSRC shall function to advise the Plant Manager on all matters related to nuclear safety.
COMPOSITION 6.5.1.2 The PNSRC shall be composed of the:
Chairman:
Member:
Hember:
Member:
Member:
Member:
Member:
Member:
Member:
Member:
Plant Manager or designated alternate Asst. Plant Hanager Operations Supervisor Technical Supervisor Maintenance Supervisor-Instrument and Control Engineer Nuclear Engineer Chemical Supervisor Performance Supervising Engineer Radiation Protection Supervisor
- Full compliance by January 1,1981 D. C.
COOK - UNIT 2
INSTRUMENTATION BASES 3/4.3.3. 6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and folIowing an accident.
3/4.3
~ 3.7 AXIAL POWER DISTRIBUTION MONITIORING SYSTEM APDMS OPERABILITY of the APDMS ensures that sufficient capability is available for the measurement of the neutron flux spatial distribution within the reactor core.
This capability is required to 1) monitor the core f1ux patterns that are representative of the peak core power density and 2) limit the core average axial power profile such that the total power peaking factor F~ is maintained within acceptable limits.
3 4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.
This capability is required in order to detect and locate fires in their early stages.
Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.
In the event that a portion o, the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is returned to service.
3/4.3. 4 TURBINE OVERS PEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive.overspeed.
Protec-tion from turbine excessive overspeed is required since excessive over-speed of the turbine could generat potentially damaging missiles which could impact and damage safety related components, equipment or structures.
0.
C.
COOK - UNIT 2 B 3/4 3-3
REACTOR COOLANT SYSTEM/
The actual shirt in RTi T of the vessel material will be established periodically durina ooeratioII by removina and evaluatina, in accordance with AS'Pt E185-73, reactor vessel material irradiation surveillance speci-mens installed near the inside wall of the reactor vessel in the core area.
Since the neutron soectra at the irradiation samoles and vessel inside radius are essentially identical, the measur ed transition shift for a sample can be aoolied with confidence to the adjacent section of the reactor vessel.
The heatuo and cooldown curves must be recalculated when the 4RT determined rrom the surveillance caosule is different from the ca1cula
<RTNOT for the equivalent capsule radiation exposure.
The pressure-temoerature limit lines shown on Fiaure 3.4-2 for reactor criticality and ror inservice leak and hydrostatic testinq have be n
provided to assuro c"moliance with the minimum temperature requirements of Appendix G to 10 CFR 50.
The number of reactor vesseI irradiation surveillance specimens and the frequencies
-or removina and testina these specimens are provided in Table 4.4-5 to assure complianc with the r equirements of Appendix H to IO CFR Part 50.
The limitations imoosed on pressurizer heatuo and cooldown and spray water temperature dif-,erential are orovided to assure that the pressurizer is operated within the design criteria assumed for <<he fatique analysis performed in accordance with the ASIDE Code requirements.
3/4.4.10 STRtlCTURAL INTEGRITY The inspection and testina proarams for ASIDE Code Class I, 2 and 3
components ensure that the structural intearity of these comoonents wiII be maintained at an acceotabIe level throuahout the life of the plant.
To the extent aoplicable, the inspection program for these components is in compliance with Section XI of the ASIDE Hoiler and Pressure Vessel Code.
3 4.4.11 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves.
These relic. 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 supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.
D.
C.
COOK - UNIT 2 1 3/4 4-In
ATTACHMENT 4 TO AEP:NRC:00449 DONALD C.
COOK NUCLEAR PLANT UNIT NOS.
1 AND 2 DOCKET NOS.
50-315 AND 50-316 LICENSE NOS.
DPR-58 AND DPR-74 PROPOSED LICENSE CONDITIONS FOR SYSTEt1S INTEGRITY AND IODIt<i YiOt<ITOR ING S stem Inte rit 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.
Iodine Monitorin The licensee shall implem nt a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.
This program shall include the following:
1.
Training of personnel, 2;
Procedures for monitoring, and 3.
Provisions for maintenance of sampling and analysis equipment.