ML17223A867
| ML17223A867 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/01/1990 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17223A868 | List: |
| References | |
| NUDOCS 9008070383 | |
| Download: ML17223A867 (29) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 FLORIDA POWER-5 LIGHT COMPANY DOCKET NO. 50-335 ST.
LUCIE PLANT UNIT NO. I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 46 License No. DPR-67 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A..
The application for amendment by Florida Power 8 Light Company, (the licensee) dated February 7, 1990, as supplemented June 19, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations.set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, 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 aIIerIdment 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.
E.
The issuance of this amendment wi 11 riot be inimical to the common oeferIse and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 5I of'he CorrmIission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly, Facility Operating License No.
DPR-67 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.(2) to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A
and B, as revised through Amendment No.
g6, 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.
Attachment:
Changes to the Technical Specifications erbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor'Regulation Date of Issuance:
August 1, 1990
ATTACHMENT TO LICENSE AMENDMENT NO.
46 TO FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace the following pages of'the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document com-pleteness.
Remove Pa es 1-4 3/4 4-3, 3/4 4-5 3/4 4-10 3/4 4-29 3/4 4-31a 3/4 4-31b 3/4 4-32 3/4 4-35 3/4 4-37a B 3/4 4-1 '
3/4 4-3 B 3/4 4-8 B 3/4 4-11 Insert Pa es 1-4 3/4 4-3 3/4 4-5 3/4 4-10 3/4 4-29 3/4 4-31a 3/4 4-31b 3/4 4-32 3/4 4-35 3/4 4-37a B 3/4 4-1 B 3/4 4<<3 B 3/4 4-8 B 3/4 4-11
DEFINITIONS DOSE E UIVALENT I"131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/
gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually. present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
E - AVERAGE DISINTEGRATION ENERGY
- l. 11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for
- isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95K of the total non-iodine activity in the coolant.
ENGINEERED SAFETY FEATURES
RESPONSE
TIME 1.12 The ENGINEERED SAFETY FEATURES
RESPONSE
TIME shall be that time interval from when the monitored parameter. exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).
Times shall include diesel generator starting and sequence loading delays where applicable.
FRE UENCY NOTATION 1.13 The FRE(UENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table l. l.
GASEOUS RADWASTE TREATMENT. SYSTEM
- l. 14 A GASEOUS RADWASTE TREATMENT SYSTEM is. any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:
a.
Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or b.
Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c.
Reactor Coolant System leakage through a steam generator to the secondary system.
ST.
LUCIE - UNIT 2 1-3
DEFINITIONS LOW TEMPERATURE QVERPRESSURE PROTECTION RANGE -
RCS 1.16 The LOW TEMPERATURE OVERPRESSURE PROTECTION RANGE is that operating condition when
$1) the RCS cold leg temperature is less than or equal to that specified in Table 3.4-3, and (2) the Reactor Coolant System is not vented to containment by an opening of at least 3.58 square inches.
MEMBER S OF THE PUBLIC 1.17 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.
This category does not include employees of the licensee, its contractors or vendors.
Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.
OFFSITE DOSE CALCULATION MANUAL ODCM 1.18 The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and shall include the Radiological Environmental Monitoring Sample point locations.
- 1. 19 A system, subsystem,
- train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s),
and when all necessary attendant instrumentation, controls, electrical
- power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem,
- train, component or device to perform its function(s) are also capable of performing their related support function(s).
OPERATIONAL MODE -
MODE 1.20 An OPERATIONAL MODE (i.e.
MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.2.
PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.
ST.
LUCIE - UNIT 2 Amendment No. ~6~ 88'6>
REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR QPERATIQN 3.4. 1.3 At least two,of the loop(s)/train(s) listed below shall be GPERABLE and at least one Reactor Coolant and/or shutdown cooling loops shall be in operation."
a.
Reactor Coolant Loop 2A and its associated steam generator and at least one associated Reactor Coolant pump,""
b.
Reactor Coolant Loop 28 and its associated steam generator and at least one associated Reactor Coolant pump,""
c.
Shutdown Cooling Train 2A, d.
Shutdown Cooling Train 2B.
APPLICABILITY:
MQDE 4.
ACTION:
b.
With less than the above required Reactor Coolant and/or shutdown cooling loops OPERABLE, immediately initiate corective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling loop, be in COLD SHUTDQWN within 3Q hours.
With no Reactor Coolant or shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
All Reactor Coolant pumps and shutdown cooling pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are, permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.
""A Reactor Coolant pump, shall not be started with two idle loops and one or more of the Reactor Coolart System cold l.eg temperatures less than or equal to that specified in Table 3.4-3 unless the 'secondary watei temper'ature.of each steam generator is less than 40'.F above each of the Reactor Coolant System cold leg temperatures.
ST.
LUCIE - UNIT 2 3/4 4"3 Amendment No.l~~
3E/~ 4o~
0 0 M
~
REACTOR COOLANT SYSTEM HOT SHUTDOWN SURVEILLANCE RE UIREMENTS 4.4. 1.3. 1 The required Reactor Coolant, pump(s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4. 1. 3.2 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be > 10K indicated narrow range level 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. 1.3.3 At least one Reactor Coolant or shutdown cooling loop shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ST.
LUCIE - UNIT 2 3/4 4-4
REACTOR COOLANT SYSTEM COLD SHUTDOWN -
LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4. 1.4. 1 At least, one shutdown cooling loop shall be OPERABLE and in operation",
and either:-
a.
One additional shutdown cooling loop shall be OPERABLE
, or b.
The secondary side water level of at least two steam generators shall be greater than lOX indicated narrow range level.
APPLICABILITY:
MODE 5 with Reactor Coolant loops filled ACTION:
With one of the shutdown cooling loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the inoperable shutdown cooling loop to OPERABLE status or to restore the required steam generator level as soon as possible.
With no shutdown cooling loop in operation, suspend all operations involving a reducti.on in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling loop to operation.
SURVEILLANCE RE UIREMENTS
'.4.
1.4. l. 1 The secondary side water level of at least two steam generators when required shall be determined to be within limits 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. 1,4. 1.2 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The shutdown cooling pump may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature.is maintained at least 10'F below saturation temperature.
One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other shutdown cooling loop is OPERABLE and in. operation.
¹¹A Reactor Coolant pump shall 'not be started with two idle loops unless the secondary water temperature. of each steam generator fs less than 40'F
'bove each of the. Reactor Coolant System coTd'eg temperatures.
ST.
LUCIE " UNIT 2 3/4 4-5 Amendment No. Ni
~1'~ 46,
REACTOR COOLANT SYSTEM COLD SHUTDOWN -
LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4. 1.4.2 Two shutdown cooling loops shall be OPERABLE and at least one shutdown cooling loop shall be in operation."'PPLICABILITY:
MODE 5 with reactor coolant loops not filled.
ACTION:
aO With less than the above required loops OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate corrective action to return the required loops to OPERABLE status as soon as possible.
With no shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate corrective action to return the'equired shutdown cooling loop to operation.
SURVEILLANCE RE UIREMENTS 4.4.1.4.2 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other shutdown cooling loop is OPERABLE and in operation.
The shutdown cooling pump may be deenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are perrhitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 104F below saturation temperature.
ST.
LUCIE - UNIT 2 3/4 4-6
REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a minimum water level of greater than or equal to 27% indicated level and a maximum water level of less than or equal to 68K indicated level and at least two groups of pressurizer heaters capable of being powered from 1E buses each having a nominal capacity of at least 150 kW.
APPLICABILITY:
MODES 1, 2, and 3.
ACTION:
a4 With one group of the above required pressurizer heaters inoperable, restore at least two groups to OPERABLE status 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
With the pressurizer otherwise inoperable, be in at least HOT STANDBY 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 in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE RE UIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limits 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. 3. 2 The capacity of each of the above required groups of pressurizer heaters shall be verified to be at least 150 kW at least once per 92 days.
4.4.3.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by verifying that on an Engineered Safety Features Actuation test signal concurrent with a loss of offsite power:
a.
the pressurizer heaters are automatically shed from the emergency power sources, and b.
the pressurizer heaters can be reconnected to their respective buses manually from the control room after resetting of the ESFAS test signal.
ST.
LUCIE - UNIT 2 3/4 4"9 Amendment No.
A'~ ll
REACTOR COOLANT SYSTEM 3/4.4;4 PORV BLOCK VALVES LIMITING CONDITION FOR OPERATION 3.4.4 Each Power Operated Relief Valve (PORV) Block valve shall be OPERABLE.
No more than one block valve shall be open at any one time.
APPLICABILITY:
MODES 1, 2, and 3.
'CTION:
. b.
C.
With one or more block valve(s) 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 block valve(s) to OPERABLE status or close the block valve(s) and remove power from the block valve(s}; 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 />.
With both block valves
- open, close one block valve within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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 COLO 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 specification 3.0.4 are not applicable.
SURVEILLANCE RE UIREMENTS 4.4.4 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet':the requirements of Action a. or b. above.
ST.
LUCIE - UNIT 2 3/4 4-Zo Amendment No. ~;g6
'I REACTOR COOLANT SYSTEM 3/4. 4. 9
. PRESSURE/TEMPERATURE LIMITS REACTOR COOL'ANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant'System (except the'ressurizer) temperature and 'pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-3 and '3.4-4 during heatup, cooldown, criticality, and inservice leak and. hydrostatic testing.
APPLICABILITY: At al 1 times.
ACTION:
l With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations 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 reduce the RCS T to less than 200'F within'the'-next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> in accordance with Figurek 3.4-3 and 3.4~4.
SURVEILLANCE RE UIREHENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup,
- cooldown, and inservice leak and hydrostatic testing operations.
ST.
LUCIE - UNIT 2 3/4 4-29 Amendment No. gg, Q,, 46,
REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (Continued 4.4. 9. 1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50 Appendix I{ in accordance with the schedule in Table 4.4-5.
The results of these examinations shall be used to update Figures 3.4-2, 3.'4-3 and 3.4-4.
ST.
LUCIE - UNIT 2 3/4 4-30 Amendment No. ides 31
FIGURE 3.4-2 ST. LUCIE-2 P/T LIMITS, 15 EFPY HEATUP'AND CORE CRITICAL 2500 50 F/HR ISOTHERMAL 2000 1000 500 CJJ CL Ch CJJ LJJ 1500 LJJN Ch CJJ JJJK 00 JJJ J
0 Ox I
Vl O
L, LOWEST SERVICE TEMPERATURE 160 F ISOTHERMAL 551 PBA CORE CRITICAL ALLOWABLEHEATUP RATES RATE, ~F/HR TEMP. LIMIT~F 50 AT ALL TEMPERATURES 5&F/HR MIN. BOLTUP TKMP SIPP 100 200 300 400 500 TC INDICATED REACTOR COOLANT SYSTEM TEMPERATURE'OF ST.
LUCIE - UNIT 2 3/4 4-3la Amendment No. 87, 46P
FIGURE 3.4 3 ST. LUCIK-2 P/T LNNTS, 15 KFPY
. COOLDOVtN AHO NSERVCK-TEST NSERVICE HYORQSTATIC TEST 2000 LOWKST SGlVlCK TEMPERATURK 1QPF TO ISOTl%cdNNAL, oooo ISOTl%c3NNAL 1~/l%l
~/l%
0 7' /t%
1QPF/l%
0 TC MXCATKOREACTOR COOLAI4T SYSTEM TEMPERATURE, <<F ST.
LUGIE - UNIT 2 3/4 4-31b Amendment No. g7, 46,
RGURK 3.+4 Sf. LUC%.2 P/T LIITSi 15 KFPY MAXtMUMALLOWAbLECOOLOOWN RATES 100
$0 e0 I
C io
'A~ F/t%
......... 30
'130 140 To ~ NQCATEO REACTOR COOLANT TQS%RATURE, F
NOTE: A MAXNNMCOCNJX)WN RATE Of 100 F/HR 8 ALLOWKOAT ANY TE%N%RATURE AbOVK 1NPP ST.
LUGIE - UNIT 2 3/4 4-32 Amendment No.
$7,:46,
4 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS
'IMITING CONDITION FOR OPERATION 3.4.9.3 Unless the RCS is depressurized and vented by atlhas't 3.58 square
- inches, at least one of the following overpressure protection systems shall be OPERABLE:
a.
Two power-operated relief valves (PORVs) with a liftsetting of less than or equal to 470 psia and with their associated block valves open.
These valves may only be used to satisfy low temperature overpressure protection (LTOP) when the RCS cold leg temperature is greater than the temperature listed in Table 3.4-4.
b.
Two shutdown cooling relief valves (SDCRVs) with a liftsetting of less than or equal to 350 psia.
c.
One PORV with a liftsetting of less than or equhl to 470 psia and with its associated block valve open in conjunction with the use of one SDCRV with a, liftsetting of less than or equal to 350 psia.
This combination may only be used to satisfy LTOP when the RCS cold leg temperature is greater than the temperature listed in Table 3.4-4.
APPLICABILITY:
MODES 4, 5 and 6.
ACTION:
a.
With either a PORV'r an SDCRV being used for LTOP inoperable, restore'at least two overpressure protection devices to OPERABLE status within 7 days or:
1.
Depressurize and vent the RCS with a minimum vent area of 3.58 square inches within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; OR'.
Be at a temperature above the LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3 within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b.
With none of the overpressure protection devices being used for ESTOP OPERABLE, within the next eight hours either:
I 1.
Restore at least one::overpressure'rotection device to OPERABLE status or vent the RCS; OR 2.
Be at a temperature above the'OW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3.
With cold leg temperature within the LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3.
ST.
LUCIE - UNIT 2 3/4 4-35 Amendment No. 75, gg, 45'
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued):
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, SDCRVs 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.
SURVEILLANCE RE UIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
a.
In addition to the requirements of Specification 4.0.5, operating the PORV through one complete cycle of full travel at least once per 18 months.
ST.
LUCIE - UNIT 2 3/4 4-36 Amendment No. 1g, 31
'URVEILLANCE RE UIREMENTS Continued b.
Performance of a CHANNEL FUNCTIONAL'EST on the PORV actuation
- channel, but excluding valve operation, within 31 days prior to enter ing a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.
C.
d.
Performance of a CHANNEL CALIBRATION on the PORV actuation channel, at least once per 18 months.
Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.
4.4.9.3.2 The RCS vent(s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />" when the vent(s) is being used for overpressure protection.
Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
ST..LUCIE - UNIT 2 3/4 4-37
I TABLE 3.4-3 I
LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE Operating Period,EFPY'uring
~Heatu During Cooldown Cold Le Tem erature
< operating pit'i6d < l5
< 247
< 230 TABLE 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP Operating Period EFPY 6 < operating period
< 15
- cold, F'ur ing
~Heatu 165
- cold, F'uring Cooldown 165 ST.
LUCIE - UNIT 2 3/4 4-37a Amendment No.
8f. 46.
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.20 during all normal operations and anticipated transients.
In MODES 1 and 2
with one reactor coolant loop not -in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
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 MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least =two loops (either shutdown cooling or RCS) be OPERABLE.
In MODE 5 w'th reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat. removing component, require that at least two shutdown cooling loops b'e OPERABLE.
The operation of one reactor coolant pump or one shutdown cooling 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 reductions will, therefore, be within the capability of operator recognition and control.
The restriction on starting a reactor coolant pump in MODES 4 and 5, with two idle loops and one or more RCS cold leg temperatures less than or equal to that specified in Table 3.4-3 is provided to prevent RCS pressure transients, caused by energy additions from the secondary system from exceeding the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients by (1) sizing each PORV to mitigate the pressure transient of an inadvertent safety injection actuation in a water-solid RCS with pressurizer heaters energized, (2) restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 40'F above each of the RCS cold leg temperatures, (3) using SDCRVs to mitigate RCP start transients and the transients caused by inadvertent SIAS actuation and charging water, and (4) rendering one HPSI pump inoperable when the RCS is at low temperatures.
3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.
Each safety valve is designed to relieve 212,182 lbs per hour of saturated steam at the valve setpoint.
The relief capacity of a single safety valve is adequate to relieve any overpres-sure condition which could occur during shutdown.
In the event that no safety valves are
- OPERABLE, an operating shutdown cooling loop, connected to the
In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.
ST.
LUCIE - UNIT 2 B 3/4 4-1 Amendment No. N, 8f. 4
~
REACTOR COOLANT SYSTEM BASES.
SAFETY VALVES Continued During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.
The combined relief capacity of these valves is sufficient to limit the system pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER 'and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pres-surizer Pressure-High) is reached (i.e.,
no credit is taken for a direct reactor trip on the loss of turbine} and also assuming no operation of the pressurizer power-operated relief valve or steam dump valves.
Demonstration 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 Vessel Code.
3/4.4. 3 PRESSURIZER An OPERABLE pressurizer provides pressure control for the Reactor Coolant System during operations with both forced reactor coolant flow and with natural circulation flow.
The minimum water level in the pressurizer assures the pressurizer
- heaters, which are required to achieve and maintain pressure
- control, remain covered with water to prevent failure, which could occur if the heaters were energized uncovered.
The maximum water level in the pres-surizer ensures that this parameter is maintained within the envelope of operation assumed in the safety analysis.
The maximum water level also ensures that the RCS is not a hydraulically solid system and that a steam bubble will be provided to accommodate pressure surges during operation.
The steam bubble also protects the pressurizer code safety valves against water relief.
The requirement to verify that on an Engineered Safety Features Actuation test signal concurrent with a loss of offsite power the pressurizer heaters are automatically shed from the emergency power sources is to ensure that the non-Class 1E heaters do not reduce the reliability of or overload the emergency, power source.
The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability to control Reactor Coolant System pressure and establish and maintain natural circulation.
ST.
LUCIE " UNIT 2 B 3/4 4-2
REACTOR COOLANT SYSTEM BASES 3/4.4.4 PORV BLOCK VALVES The power operated relief valves (PORVs) 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 PORVs in conjunc-tion with a reactor trip on a Pressurizer Pressure-High signal minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
The opening of the PORVs fulfills no safety-related function and no credit is taken for their operation in the safety analysis for MODE 1, 2, or 3.
Each
.PORV has a remotely operated block valve to provide a.'.positive shutoff capability should a relief valve become inoperable.
Since it is impractical and undesirable to actually open the PORVs.,to demonstate their reclosing, it becomes necessary to verify OPERABILITY of the PORV block valves to ensure the capability to isolate a malfunctioning PORV.
As the PORVs are pilot operated and require some system pressure to operate, it is impractical to test them with the block valve closed.
The PORYs are sized to provide low temperature overpressure protection (LTOP).
Since both PORVs must be OPERABLE when used for LTOP, both blot:k valves will be open during operation within the LTOP range.
As the PORV capacity required to perform the LTOP function is excessive for operation in MODE 1, 2, or 3, it is necessary that the operation of more than one PORY be precluded during these NODES.
- Thus, one block valve must be shut during MODES 1, 2, and 3.
3/4.4.5 STEAN GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integt ity of this portion of the RCS will be maintained.
The program for inseryice 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 conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
3 ST.
LUCIE - UNIT 2 B 3/4 4-3 Amendment No. Q, 46~
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS Continued Inservice inspection of steam generator tubing also provides a meany of characterizing 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 secondary coolant, will be maintained within tho'se 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
= 1.0 gpm from both steam generators.
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 have demon-strated that primary-to-secondary leakage of 0.5 gpm per steam generator can readily be detected by radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
- However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40K of the tube nominal wall thickness.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20K of the original tube wall thickness.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4. 6. 1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by thi.s specification are provided to monitor and detect leakage frbm the Reactor Coolant Pressure Boundary.
These detection systems are consistent with the recommendations of ST.
LUCIE - UNIT 2 B 3/4 4"4 Amendment No. I3
REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY Continued DELETED ST.
LUCIE - UNIT 2 B 3/4 4-7 Amendment No. 44
REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Section,5.2 of the FSAR.. During startup and shutdown,'the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients through the reactor vessel wall produce thermal stresses which are compressive at the reactor vessel inside surface and which are tensile at the reactor vessel outside surface.
Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the outside surface location.
However, since neutron irradiation damage is larger at the inside surface location when compared to the outside surface, the inside surface flaw may be more limiting.
Consequently. for the heatup analysis both the inside and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting.
During cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and which are compressive at the reactor vessel outside sur'face.
Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the i>>ide surface location.
Since the neutron indication damage is also greatest at the inside surface location the inside surface flaw is the limiting location.
<<>>equently, only the inside surface flaw must'e evaluated for the c'ooldown analysis.
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tran;ltlcn t~>< '
e c t.~ Cry lato y ~[~ "" shift e
z da 1.99
"(actor ta R
safgt ty t
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3 ~ 4 3
$.EFby g, l,'L~.i'l
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'4%1T 0
>>e heatup and cooldown l,imit.curves Figures 3.4-2, 3.4-3 and 3.4-4 are
~>><<te curves which were prepared by determining the most conservative
- case, the inside or outside wall controlling, for any heatup rate of up to 9"e>> F.per hour or cooldown rate of up" to 100 degr ees F per hour.
The nd cooldown curves were prepared based upon the most limiting value p" d<<ted adjusted reference temperature at.15'EFPY, and.they include
"~"ts <<r pressure differences between the reactor vessel. beltline and press"r>ter knstrurient taps.
r vessel materials have been tested to determine their initial
>>e reacto t " t~ of these tests are shown in Table B 3/4.4-1.
Reactor operation ihgl'gpss
)n th n<<<<on (E greater than 1 MeY) irradiation will cause an a) t~~ <n<t~,l <> NOT An ad]usted reference temperature can be predicted using n the RT a'.)ustaents for NG(~ b) Che'fluence (E greater than 1 MeV), including appropriate t>e ~all tht
" r4n attenuation and neutron energy spectrum variations through opper and nickel contents of the material, and d) the from the curve shown in Figure B 3/4.4-1 as recommended evision 2, "Effects 'of Residual Elements on Predicted
>easel i)0terQlg.."
,The..heatup apd cooldown limQ;.
a"" 3.4-.~i'nclude predicted adjustments for this B 3/4 4-B Amendment No. J$, g], 46,
~ ~
REACTOR COOLANT SYSTEM BASES The actual shift in RT T of the vessel material will be established periodically dur ing operatiI by removing and evaluating, in accordance, with ASTM E185-73 and 10 CFR Appendix H, reactor vessel material irradiation surveil-lance specimens installed near the inside wall of the reactor vessel in the core area.
The surveillance specimen withdrawal schedule is shown in Table 4.4-'5.
Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.
The heatup and cooldown curves must be recalculated when the delta RT determined from the surveillance capsule is different from the calculated IINta RTN for the equivalent capsule radiation exposure.
The lead factors shown% Table 4.4-5 are the ratio of neutron flux at the surveillance capsule to that at the reactor inside surface.
The pressure-temperature limit lines shown on Figures 3.4-2, 3.4-3 and 3.4-4 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.
The maximum RTN T for all Reactor Coolant System pressure-retaining materials, with the 3xception of the reactor pres'sure
- vessel, has been determined to be 60'F.
The Lowest Service Temperature limit line shown on Figures 3.4-2, 3.4-3 and 3.4-4 is based upon this RT T since Article NB-2332 (Summer Addenda of 1972) of Section III of the ASME IIII>ler and Pressure Vessel Code requires the Lowest Service Temperature to be RT DT + 100 F for piping,
- pumps, and valves.
Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia.
The limitations imposed on the pressurizer heatup and cooldown rates 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 ASHE Code requirements.
The OPERABILITY of two PORVs, two SDCRVs or an RCS vent opening of greater than 3.58 square inches ensures that the RCS will 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 leg temperatures aro less than or equal to the LTOP temperatures.'he Low Temperature Ove'rpressure Protection System has adequate relieving capability.to protect the RCS from ovei pressurization when the transient is limited to either (1) a safety injection actuation in a water-solid RCS with 'the pressurizer heaters energized or (2) the start of an idle RCP with the secondary water. temperature of the steam generator less than or equal to 40'F above the RCS cold leg temperatures with the pressurtker water-solid.
ST.
LUCIE - UNIT 2 B 3/4 4-11 Amendment No. 1$. X/s 46,
REACTOR COOLANT SYSTEM BASES 3/4.4. 10 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling.
The OPERABILITY of at least one reactor Coolant System vent path from the reactor vessel head and the pressQrizer steam space ensures the capa-bility exists to perform this function.
The redundancy design of the Reactor Coolant System vent systems serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System vent system are consistent with the requirements of Item II.b.l of NUREG-0737, "Clarification of TMI Action Plan Requirements,"
November '1980.
3/4.4. 11 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.
These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50. 55a(g) except where specific wr itten relief has been granted by the Commission pursuant to 10 CFR Part 50. 55a (g) (6) (i).
Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through'Summer 1973.
ST.
LUCIE " UNIT 2 B 3/4 4-12