ML20082K371
| ML20082K371 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 04/13/1995 |
| From: | Mark Miller NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20082K374 | List: |
| References | |
| NUDOCS 9504190255 | |
| Download: ML20082K371 (31) | |
Text
_ _ _ _ _ - _ _ _ _ _ _ _ - _ _
jperug 3
4 UNITED STATES y
j NUCLEAR REGULATORY COMMISSION
'2 WASHINGTON, D.C. J0565-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT. UNIT U AMENDMENT TO FACILITY OPERATING Ufjj@i Amendment No.100 License No. DPR-80 1.
The Nuclear Regulatory Commission (the Cor.nission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated August 17, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; 8.
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 (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
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-80 is hereby amended to read as follows:
h D
P
~
, (2)
Technical Specifications i
The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
100, are hereby incorporated in the license.
Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
'Y$ Q h.
Melanie A. Miller, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 13, 1995
p Kti g
is UNITED STATES y
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 206lWH1001 l
"N'
/
PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 l
DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 99 License No. DPR-82 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated Augu.it 17, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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; l
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 l
Part 51 of the Commission's regulations and all applicable j
requirements have been satisfied.
2.
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-82 is hereby amended to read as follows:
l
. (2)
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
99, are hereby incorporated in the license.
Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION k
Melanie A. Miller, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 13,1995 l
l
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 100 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 99 TO FACILITY OPERATING LICENSE N0. DPR-82 DOCKET NOS. 50-275 AND 50-323 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change.
Overleaf pages are also provided to maintain document completeness.
REMOVE INSERT viii viii 3/4 1-8 3/4 1-8 3/4 1-11 3/4 1-11 3/4 4-3 3/4 4-3 3/4 4-5 3/4 4-5 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 3/4 4-35 3/4 4-35 3/4 5-7 3/4 5-7 3/4 5-8 3/4 5-8 B 3/4 1-3 B 3/4 1-3 B 3/4 4-1 B 3/4 4-1 B 3/4 4-7 B 3/4 4-7 B 3/4 4-12 B 3/4 4-12 B 3/4 4-15 B 3/4 4-15 B 3/4 4-16 B 3/4 4-16 B 3/4 5-2 B 3/4 5-2
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1
SECTION EAGI l
3/4.3 INSTRUMENTATION (continued)
Chlorine Detection Systems...............................
3/4 3-54 Explosive Gas Monitoring Instrumentation.................
3/4 3-59 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................
3/4 3-60 j
3/4.4 REACTOR COOLANT SYSTEM l
3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power 0peration..............................
3/4 4-1 Hot Standby..............................................
3/4 4-2 Hot Shutdown.............................................
3/4 4-3 Cold Shutdown - Loops Filled.............................
3/4 4-5 1
Cold Shutdown - Loops Not F t11 ed.........................
3/4 4-6 3/4.4.2 SAFETY VALVES l
l Operating................................................
3/4 4-8 3/4.4.3 PRESSURIZER..............................................
3/4 4-9 3/4.4.4 RELIEF VALVES...........................................
3/4 4-10 3/4.4.5 STEAM GENERATORS........................................
3/4 4-11 DIABLO CANYON - UNITS 1 & 2 vii Amendment Nos. 57 1 OC, 75 a 74, 98 & 97
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION Pf_GE 3/4.4 REACTOR COOLANT SYSTEM (continued)
TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION................
3/4 4-16 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION.......................
3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.............................
3/4 4-18 Ope rati onal Le akage................................... 3/4 4-19 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......
3/4 4-21 3/4.4.8 SPECIFIC ACTIVITY.....................................
3/4 4-25 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
> 1 yCI/ GRAM DOSE EQUIVALENT I-131....................
3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................
3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS..........................
3/4 4-30 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -
APPLICABLE UP TO 12 EFPY.............................
3/4 4-31 l
FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -
APPLICABLE UP T0 12 EFPY.............................
3/4 4-32 l
Overpressure Protection Systems......................
3/4 4-35
)
DIABLO CANYON - UNITS 1 & 2 viii Unit 1 - Amendment No. 54r98,100 Unit 2 - Amendment No. 53147,99
i REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS
^
FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION i
3.1.2.1 As a minimum, one of the following boron injection flow paths shall l
be OPERABLE with motor-operated valves required to change position and pumps j
required to operate for boron injection capable of being powered from an j
OPERABLE emergency power source:
i a.
A flow path from the boric acid tanks via a boric acid transfer pump j
and charging pump to the Reactor Coolant System if the boric acid j
l storage tank in Specification 3.1.2.5a. is OPERABLE,.or i
b.
The flow path from the refueling water storage tank via a charging j
pump to the Reactor Coolant System if the refueling water storage j
tank in Specification 3.1.2.5b. is OPERABLE.
APPLICABILITY:
MODES 5 and 6.
ACTION:
With none of the above flow paths OPERABLE or capable of being powered from an i
OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
d l
SURVEILLANCE REQUIREMENTS 4
4.1.2.1 At least one of the above required flow paths shall be demonstrated j
OPERABLE:
I At least once per 7 days by verifying that the temperature of the flow a.
i path is greater than or equal to 65 F when a flow path from the boric j
acid tanks is used, and
]
l b.
At least once per 31 days by verifying that each valve (manual, power-operated or automatic) in the flow path that is not locked, sealed or otherwise securied in position, is in its correct position.
4 I
I l
4 3
a DIABLD CANYDN - UNITS I & 2 3/4 1-7 Amendment Nos.
53 & 52, 72 8 71 e
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REACTIVITY CONTROL SYSTEMS i
j FLOW PATHS - OPERATING i
t j
LIMITING CONDITION FOR OPERATION 3.1.2.2 Each of the following boron injection flow paths shall be OPERABLE:
a.
The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS), and b.
The flow path from the refueling water storage tank via a charging i
pump to the RCS.
APPLICABILITY: MODES 1, 2, 3 and 4#.
1 i
ACTION:
i a.
With the flow path from the boric acid tanks inoperable, restore the inoperable flow path 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 and borated to a SHUTDOWN MARGIN equivalent to at 4
least 1% ok/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With the flow path from the refueling water storage tank inoperable, restore the flow path to OPERABLE status within I hour 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 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 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:
a.
At least once per 7 days by verifying that the temperature of the flow path from the boric acid tanks is greater than or equal to 65'F, b.
At least once per 31 days by verifying that each valve (manual, power-operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position, c.
At least once per 18 months by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal, and d.
At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a delivers at least 30 gpm to the RCS.
- 0nly one boron injection flow path is required to be OPERABLE whdnever the temperature of one or more of the RCS cold legs is less than or equal to 270*F.
g DIABLO CANYON - UNITS 1 & 2 3/4 1-8 Unit 1 - Amendment No. 63rM,100 Unit 2 - Amendment No. 6BrM,99
4 g
REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4#.
ACTION:
With only one charging pump OPERABLE, restore at least two charging pumps 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 and borated to a SHUTOOWN MARGIN equivalent to at least 1% Ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE0VIREMENTS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE when tested pursuant to Specification 4.0.5.
In addition, when the above required charging pumps include a centrifugal charging pump (s), verify that, on recirculation flow, each required centrifugal charging pump (s) develops a differential pressure of greater than or equal to 2400 psid.
4.1.2.4.2 All centrifugal charging pumps, except the above required OPERABLE pump, shall be demonstrated inoperable
- 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 Reactor Coolant System (RCS) cold legs is less than or equal to 270*F by verifying that the motor breaker D.C.
l control power is de-energized.
- A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 270*F.
l
- An inoperable pump may be made OPERABLE for testing per Specification 4.0.5 provided the discharge of the pump has been isolated from the Reactor Coolant System by an isolation valve with power removed from the valve operator, or by a sealed closed manual isolation valve.
DIABLO CANYON - UNITS 1 & 2 3/4 1-11 Unit 1 - Amendment No. 100 i
Unit 2 - Amendment No. 99
4 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN j
LIMITING CONDITIM FOR OPERATION l
3.1.2.5 As a minimue, one of the following borated water sources shall be OPERABLE:
l a.
A Boric Acid Storage System with:
- 1) A minimum contained borated water volume of 2,499 gallons,
- 2) A boron concentration between 7,000 and 7,700 ppe, and 1
- 3) A minimum solation temperature of 65'F.
b.
The Refueling Water Storage Tank (RWST) with:
j i
- 1) A minimum contained borated water volume of 50,000 gallons, i
- 2) A minimum boron concentration of 2300 ppm, and
- 3) A minimum solution temperature of 35'F.
APPLICABILITY: MODES 5 and 6.
ACTION:
With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
i I
SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:
e i
a.
At least once per 7 days by:
1)
Verifying the boron concentration of the water, t
l 2)
Verifying the contained borated water volume, and 3)
Verifying the boric acid storage tank solution temperature when it is the source of borated water.
i b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when it i
is the source of borated water and the outside ambient air temperature is less than 35'F.
l 4
J DIABLO CANYON - UNITS 1 & 2 3/4 1-12 Amendment Nos.
53 & 52, 72 & 7
REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3
- At least two of the loops / trains listed below shall be OPERABLE and at least one of these loops / trains shall be in operation:*
a.
Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,**
b.
Reactor Coolant Loop 2 and its associated steam generator and reactor i
coolant pump,**
l c.
Reactor Coolant Loop 3 and its associated steam generator and reactor
'i coolant pump,**
d.
Reactor Coolant Loop 4 and its ar,sociated steam generator and reactor coolant pump,**
e.
Residual Heat Removal (RHR) Train 1, and j
f.
Residual Heat Removal (RHR) Train 2.
APPLICABILITY: MODE 4.
I ACTION:
a.
With less than the above required loops / trains OPERABLE, immediately l
initiate corrective action to return the required loop / train to
)
OPERABLE status as soon as possible; if the remaining OPERABLE i
loop / train is an RHR train, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With no loop / train 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 / train to operation.
- All reactor coolant pumps and RHR pumps may be deenergized for up to I hour 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 one or more of the Reactor Coolant System cold leg temperatures less than or equal to 270*F unless:
l (1) the pressurizer water level is less than 50%, or (2) secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.
DIABLO CANYON - UNITS 1 & 2 3/4 4-3 Unit 1 - Amendment No.100 Unit 2 - Amendment No. 99
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pumps, if not in operation, shall be determined 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 secondary side water level to be greater than or equal to 15% 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 loop or RHR train 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 />.
l DIABLO CANYON - UNITS 1 & 2 3/4 4-4 A
COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) train shall be OPERABLE and 1
in operation *, and either:
a.
One additional RHR train shall be OPERABLE #, or l
b.
The secondary side water level of at least two steam generators shall be greater than 15%.
APPLICABILITY: MODE 5 with reactor coolant loops filled ##.
ACTION:
j a.
With one of the RHR trains inoperable and with less than the required steam generator water level, immediately initiate corrective action to return the inoperable RHR train to OPERABLE status or restore the required steam generator water level as soon as possible.
b.
With no RHR train 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 RHR train to operation.
SURVEILLANCE RE0VIREMENTS 4.4.1.4.1.1 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 />.
i 4.4.1.4.1.2 At least one RHR train 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 RHR pump may be deenergized for up to I hour 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.
- 0ne RHR train 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 RHR train is OPERABLE and in operation.
- A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 270*F unless:
l (1) the pressurizer water level is less than 50%, or (2) the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.
DIABLO CANYON - UNITS 1 & 2 3/4 4-5 Unit 1 - Amendment No.100 Unit 2 - Amendment No. 99
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l Controlling Material:
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1/4T: Unit 1 Lower shell Loneltudinet Weld 3 442 c. RT., a 1/4T = 161'F.
3/4T: Unit 2 Intermodlate Shett Plate 05454 2. RT., a 3/4T = 131'F l
FIGURE 3.4-2 i
REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 12 EFPY 4
i DIABLO CANYON - UNITS 1 & 2 3/4 4-31 Unit 1 - Amendment No. 64.100 Unit 2 - Amendment No. 63,gg i
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1/4T: Unit 1 Lower Shett Loneltudinst Weld 3 442 C. RT., a 1/47 = 161'F.
i 3/4T: Unit 2 Intermediate Shett Piste 35454 2. RT., a 3/4T = 131'F i
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FIGURE 3.4-3 l
REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 12 EFPi i
i 4
i DIABLO CANYON - UNITS 1 & 2 3/44-32 Unit 1 - Amendment No. 64,100 l
Unit 2 - Amendment No. E,99 i
i
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS
)
LIMITING CONDITION FOR OPERATION I
'). 4. 9. 3 The following Overpressure Protection Systems shall be OPERABLE:
1 Two Class 1 power-operated relief valves (PORVs) with a lift setting a.
of less than or equal to 435 psig, or l
b.
The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.07 square inches.
APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 270*F, MODE 5 and MODE 6 with the reactor vessel head on and the l
vessel head closure bolts not fully de-tensioned.
ACTION:
a.
With one Class 1 PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through an RCS vent of greater than or equal to 2.07 square inches vent 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 one Class 1 PORY inoperable in MODES 5 or 6 with the reactor vessel head on and the vessel head closure bolts not fully detensioned, restore the inoperable PORV to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or depressurize and vent the RCS through an RCS vent of greater than or equal to 2.07 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 />.
c.
With both PORVs inoperable, depressurize and vent the RCS through an RCS vent of greater than or equal to 2.07 square inches vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
d.
In the event either the PORVs or the RCS vent are used to mitigate an 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 on the transient, and any corrective action necessary to prevent recurrence.
66 81,100 DIABLO CANYON - UNITS 1 & 2 3/4 4-35 Unit I - Amendment No.
7 Unit 2 - Amendment No. 5460,99
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SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each Class 1 PORV shall be demonstrated OPERABLE by:
a.
Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV i
actur. tion channel, but excluding valve operation, at least once per
{
31 ' Jays; j
b.
Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and c.
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 /> i
when the PORV is being used for overpressure protection.
1 i
4.4.9.3.2 The RCS vent shall be verified to be open when the vent is being i
used for overpressure protection at least once per 31 days when the pathway is i
provided by a valve (s) that is locked, sealed, or otherwise secured in the open position; otherwise verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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DIABLO CANYON - UNITS 1 & 2 3/4 4-36 Amendment Nos.81 & 80
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j EMERGENCY CORE COOLING SYSTEMS f
3/4.5.3 ECCS SUBSYSTEMS - T;; LESS THAN 350*F a
1 LIMITING CONDITION FOR OPERATION t
3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be j
OPERABLE:
t
[
a.
One OPERABLE centrifugal charging pump,*
b.
One OPERABLE Residual Heat Removal heat exchanger, c.
One OPERABLE Residual Heat Removal pump, and d.
An OPERABLE flow path capable of taking suction from the Refueling Water Storage Tank upon being manually realigned and transferring 1
j sucticn to the containment sump during the recirculation phase of operation.
4 APPLICABILITY: MODE 4.
i ACTION:
i With no ECCS subsystem OPERABLE because of the inoperability of a.
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 I hour or be in COLD SHUTDOWN within the next j
20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
I b.
With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat j
removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T, less than 350*F by use of alternate heat removal methods.
1 c.
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 j
describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
i 1
SA maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 270*F.
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DIABLO CANYON - UNITS 1 & 2 3/4 5-7 Unit 1 - Amendment No.100 Unit 2 - Amendment No. 99
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0VIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of Specification 4.5.2.
4.5.3.2 All centrifugal charging pumps and Safety Injection pumps, except the above allowed OPERABLE pumps, shall be demonstrated inoperable
- 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 270*F by verifying that the motor circuit breakers D.C.
I control power is de-energized.
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- An inoperable pump may be made OPERABLE for testing or for filling accumulators provided the discharge of the pump has been isoitted from the Reactor Coolant System by an isolation valve with power removed from the valve operator, or by a sealed closed manual isolation valve.
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DIABLO CANYON - UNITS 1 & 2 3/4 5-8 Unit 1 - Amendment No.100 Unit 2 - Amendment No. 99
9
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REACTIVITY CONTROL S) STEMS BASES 1
1 B0 RATION SYSTEMS (Continued) l The contained water volume limits include allowance for water not available because of discharge line location and other physical j
characteristics.
\\
The OPERABILITY of one Boron Injection System during REFUELING ensures i
i that this system is available for reactivity control while in MODE 6.
l The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all centrifugal charging pumps except the required OPERABLE pump to be inoperable below 270*F provides l
assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
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l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES
)
The specifications of this section ensure that:
(1) acceptable power j
distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is i
maintained, and (3) the potential effects of rod misalignment on associated i
accident an'alyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure j
compliance with the control rod alignment and insertion limits. Group demand position can be determined from:
(1) the group step counters, or (2) the plant computer, or (3) for control rods, the P to A converter at the rod control cabinet.
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The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the
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original design criteria are met. Continued operation of the Rod Control system is allowed with multiple immovable rods, that are still trippable and within alignment, for periods up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to allow maintenance and/or i
testing of the Rod Control system (additional information is included in j
Attachment C of the Westinghouse letter to the NRC on Movable Assemblies, December 21,1984.) Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation.
In addition, i
those accident analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.
The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or i
equal to 541*F and with all reactor coolant pumps operatin,g ensures that the measured drop times will be representative of insertion times experienced 3
during a Reactor trip at operating conditions.
3 Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is 4
i inoperable. These verification frequencies are adequate for assuring that the i
applicable LCO's are satisfied.
j DIABLO CANYON - UNITS 1 & 2 B 3/4 1-3 Unit 1 - Amendment No. 44,100 j
Unit 2 - Amendment No. &3,99 i
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3/4.4 REACTOR COOLANT SYSTEM 1
l BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in l
operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients.
1 In MODE 3, two reactor coolant loops provida sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop rrovides sufficient heat d
removal if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. Single failure considerations require that two 4
loops be OPERABLE at all times.
j In MODE 4, and MODE 5 with reactor coolant loops filled, a s*.ngle reactor coolant loop or RHR train provides sufficient heat removal capabslity for removing decay heat; but single failure considerations require that at least i
two loops (either RHR or RCS) be OPERABLE.
In MODE 5, with reactor coolant loops not filled, a single RHR train provides sufficient heat removal capability for removing decay heat; but r, ingle failure considerations anJ the unavailability of the steam generator as a heat removing component require that at least two RHR trains 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 with one or more RCS cold legs less than or equal to 270*F are provided to prevent RCS pressure l
transients, caused by energy additions from the Secondary Coolant System, j
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 i
Appendix G by:
(1) restricting the water volume in the pressurizer and thereby providing a volume for the reactor coolant to expand into, or (2) restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.
i 3/4.4.2 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 110% of the i
valve's Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
1 i
DIABLO CANYON - UNITS 1 & 2 B 3/4 4-1 Unit 1 - Amendment No. 100 Unit 2 - Amendment No. 99
REACTOR COOLANT SYSTEM BASES I
SAFETY VALVES (Continued) l In the event that no safety valves are OPERABLE, an operating RHR loop, l
connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
In addition, the Overpressure Protection System (r S ief valves) provides a diverse means of protection against RCS overpressurization at low temperatures.
During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
The combined relief capacity of all of these valv6s is greater than the maximum surge rate resulting from a complete loss of load essuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss of-load) and also assuming no operation of the power-operated relief valves or sttam dump valves.
Demonstration of the safety valves' lift 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.
t 3/4.4.3 PRESSURIZER i
The limit on the maximum water volume in the pressurizer assures that the pa ameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation.
The ruaximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.
The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolcnt System pressure and establish natural circulation.
3/4.4.4 RELIEF VALVES r
In MODES 1, 2, and 3 the power-operated relief valves (PORVs) provide an RCS pressure boundary, manual RCS pressure control for mitigation of accidents, and automatic RCS pressure relief to minimize challenges to the safety valves.
The functions of providing an RCS pressure boundary and manual RCS pressure control for mitigation of accidents such as steam generator tube rupture are the safety related function of the PORVs in MODES 1, 2, and 3.
The capability of the PORV to perform its function of providing an RCS pressure boundary requires that the PORV or its associated block valve is closed. The capability of the PORVs to perform manual RCS pressure control for mitigation of accidents is based on manual actuation and does not require the automatic RCS pressure control function. The automatic RCS pressure control function of the PORys is not a safety-related function in MODES 1, 2, and 3.
The automatic pressure control function limits the number of challenges to the safety valves, but the safety valves perform the safety function of RCS overpressure DIABLO CANYON - UNITS 1 & 2 B 3/4 4 2 Amendment Nos.-33.& 36-81 & 80
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~ --;......
BASES i
3/4.4.9 PRES $8E/ TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Presure Vessel Code,Section III, Appendix G:
a l
1.
The reactor coolant temperature and pressure and system heatup and cool-down rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
i a.
Allowable combinations of pressure and temperature for specific j
i temperature change rates are below and to the right of the limit
{
lines shown.
Limit lines for cooldown rates between those presented may be obtained by interpolation; and b.
Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only.
For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater i
capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
1 2.
These limit lines shall be calculated periodically using methods provided i
- below, I'
3.
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F, i
4.
Deleted 4
5.
System preservice hydrotests and inservice leak and hydrotests shall be i
performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI. Allowable pressures and tempera-tures for inservice leak and hydrostatic tests are given in Figure 3.4-2.
i 6.
The criticality limit on Figure 3.4-2 is based on the minimum allowable temperature of 295'F for an inservice hydrostatic test of 110% of l
operating pressure.
i The fracture toughness testing of the ferritic materials in the reactor j
vessel was performed in accordance with the 1966 Edition for Unit I and the 1968 Edition for Unit 2 of the ASME Boiler and Pressure Vessel Code,Section III. These properties are then evaluated in accordance with the NRC 1
Standard Review Plan.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil ductility reference temperature, RT at the end of 12 effective full power years (EFPY) of service life. The N,EFPY service life period is chosen such that the limiting RT, at the 1/4T location in the core region i
i DIABLO CANYON - UNITS 1 & 2 B 3/4 4-7 Unit 1 - Amendment No. 6h98,100 Unit 2 - Amendment No. G h97,99
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DIABLO CANYON - UNITS 1 & 2 8 3/4 4-8 Amendment Nos'. 54 and 53 i
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i DIABLO CANYON - UNITS 1 & 2 8 3/4 4-11 Anendment Nos. 54 and 53 i
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BASES j
PRESSURE / TEMPERATURE LIMITS (Continued)
(
l 1s greater than the RT of the limiting unirradiated material. The selection of such' a liEting RT Coolant System will be operated., assures that all components in the Reactor conservatively in accordance with applicable i
l Code requirements.
I The reactor vessel materials have been tested to determine their initial j
the results of these tests are shown in the FSAR Update.
Reactor RT.,; tion and resultant fast neutron (E greater than 1 MeV) irradiation can opera I
cause an increase in the RT Therefore, an adjusted reference temperature, or.
]
based upon the fluence, copper content and nickel content of the material in computed by Regulatory Guide question, can be predicted using value of ART ' on Predicted Radiation Damage i
1.99, Revision 2, " Effects of Residual ElemenIs j
i to Reactor Vessel Materials," for the maximum neutron fluence at the locations j
of interest. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT., at the end of 12 EFPY.
1 i
Values of ART determined in this manner will be used until the results 7
]
from N material, surveillance program, evaluated according to ASTM E185-82, can be. _.,ed.
Capsules will be removed in accordance with the requirements of ASTM E185 and 10 CFR Part 50, Appendix H.
The surveillance specimen l
withdrawal schedule will be maintained in the FSAR Update. The lead factor 1
represents the relationship between the fast neutron flux density at the i
location of the capsule and the inner wall of the reactor vessel. The heatup determined from the and cooldown curves must be recalculated when the ART
- he equivalent capsule i
)
surveillance capsule exceeds the calculated ART,7 for t j
radiation exposure.
l 4
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Allowable pressure-temperature relationships for various heatup and i
j cooldown rates are calculated using methods derived from Appendix G in j
Section III of the ASME Boiler and Pressure Vessel Code as required by l
Appendix G to 10 CFR Part 50 and these methods are discussed in detail in the j
following paragraphs.
1 1
The general method for calculating heatup 7.nd cooldown limit curves is i
based upon the principles of the linear elastic fracture mechanics (LEFM) j technology.
In the calculation procedures a semi-elliptical surface defect j
with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is 1
assumed to exist at the inside of the vessel wall as well as at the outside of j
the vessel wall. The dimensions of this postulated crack, referred to in j
Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor i
operation limit curves developed for this reference crack are conservative and
{-
provide sufficient safety margins for protection against non-ductile failure.
}
To assure that the radiation embrittlement effects are accounted for in the l
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j DIABLO CANYON - UNITS 1 & 2 B 3/4 4-12 Unit 1 - Amendment No. 64,100 Unit 2 - Amendment No. 63S9 l
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REACTOR COOLANT SYSIEM i
BASES PRESSURE / TEMPERATURE LIMITS (Continued) heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value 4
of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, i
the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure 4
stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
]
Rather, each heatup rate of interest must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any j
given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.
4 l
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the j
course of the heatup ramp the controlling condition switches from the inside l
to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of both Class 1 PORVs or an RCS vent opening of at least 2.07 square inches ensures that the RCS will be protected from pressure transients that could exceed the limits of Appendix G to 10 CFR Part 50 when operating at low temperatures. Low temperature is defined as less than or s
i equal to the reactor coolant temperature corresponding to a reactor vessel wall temperature of RT 90*F, where RT is evaluated at the beltline location (1/4T), whichis+ controlling intTe Appendix G Pressure-Temperature (60*F/hr heatup) limits. This definition is consistent with Branch Technical Position RSB 5-2, and defines the LTOP enable temperature of 270*F, applicable through 12 EFPY.
i DIABLO CANYON - UNITS 1 & 2 B 3/4 4-15 Unit 1 - Amendment No. Bh98,100 Unit 2 - Amendment No. 8b97,99
REACTOR COOLANT SYSIEH BASES LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued)
OPERABILITY of the PORVs for LTOP use requires a lift setting of less than or equal to 435-psig. This setpoint ensures that either Class 1 PORV has adequate relieving capability to protect the RCS from overpressurization for all anticipated transients, concurrent with any single active failure. The limiting transient for LTOP is a mass injection event based on the combined ECCS injection line flow from one centrifugal charging pump and the positive displacement pump, into a water-sold RCS, with letdown isolated.
The 435 psig setpoint was determined for this event based on a PORV stroke time less than or equal to 3.5 seconds, reactor service less than or equal to 12 EFPY, and administrative controls on RCP operation, charging pump operability, and the ECCS injection flow path.
The Maximum Allowed PORY Setpoint for the LTOPs will be modified, if required, based on the results of examinations of the reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H.
The surveillance specimen withdrawal schedule is maintained in l
the FSAR Update.
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i DIABLO CANYON - UNITS 1 & 2 B 3/4 4-16 Unit 1 - Amendment No. 98,100 Unit 2 - Amendment No. 97,99
3/4.5 EMERGENCY CORE COOLING SYSTEMS 8ASES 1
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1
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3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the
}
core through each of the cold legs in the event the RCS pressure falls below 1
the pressure of the accumulators.
This initial surge of water into the core provides the initial cooling mechaniss during large RCS pipe ruptures.
(
The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.
i The accumulator
" operating bypasses" power operated isolation valves are considered to be i
in the context of IEEE Std. 179-1971, which requires that j
bypasses of a protective function be removed automatically whenever permissive i
conditions are not met.
In addition as these accumulator isolation valves fail to meet single failure criteria, remo, val of power to the valves is required.
i The limits for operation with an accumulator inoperable for any reason
{
except an isolation valve closed etnisizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator i
which may result in unacceptable peak cladding temperatures.
i If a closed isolation valve cannot be famediately opened, the full capability of one i
accumulator is not available and prompt action is required to place the reactor in a MODE where this capability is not required.
i i
3/4.5.2 and 3/4. 5. 3 ECCS SUBSYSTEMS i
.1' The OPERABILITV of two ECCS subsystems ensures that sufficient emergency 1
core cooling capability will be available in the event of a LOCA assuming the j
loss of one subsystem through any single failure consideration.
operating in conjunction with the accumulators is capable of supplying sufficientEithe 4
core cooling to Ifait the peak cladding temperatures within acceptable limits
{
for all postulated break sites ranging from the double ended break of the largest RCS cold leg pipe downward.
In addition, each ECCS subsysten provides j
{
long ters core cooling capability in the recirculation mode during the accident j
recovery perled.
3' able without single failure consideration on the basis of the condition of the reactor and the Italted core cooling requirements.
j; DIABLO CANYON - UNITS 1 & 2 8 3/4 5-1 I
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EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSIDL1 (Continued)
The requirement to maintain the RHR Suction Valves 8701 and 8702 in the locked closed condition in MODES 1, 2 and 3 provides assurance that a fire could not cause inadvertent opening of these valves when the RCS is pressurized to near operating pressure. These valves are not part of an ECCS subsystem.
The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all centrifugal charging pumps and Safety Injection pumps except the required OPERABLE charging pump to be inoperable below 270*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 ensure that, at a minimum, the assumptions.used in the safety analyses are met and that subsystem OPERABILITY is maintained. The safety analyses make assumptions with respect to minimum total system resistance, minimum and maximum total injection line resistance, and minimum individual injection line resistance. These resistances in conjunction with the ranges of potential pump performance are used to calculate the minimum and maximum ECCS flows assumed in the safety analyses..
The minimum flow Surveillance Requirement ensures that the maximum injection line resistance assumptions are met. These assumptions are used to calculate minimum flows to the RCS for safety analyses which are limited by minimum ECCS flow to the RCS.
DIABLO CANYON - UNITS 1 & 2 B 3/4 5-2 Unit 1 - Amendment No. 66,100 Unit 2 - Amendment No. 64#9
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