ML20236H011
| ML20236H011 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 10/29/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236H013 | List: |
| References | |
| NUDOCS 8711030389 | |
| Download: ML20236H011 (9) | |
Text
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'4e l' - *t.. A UNITED STATES e
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,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
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j BOSTON EDISON COMPANY i
DOCKET NO. 50-243
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PILGRIM NUCLEAR p0WER STATION b
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AMENDMENT TO FACILITY OPERATING LICENSE 2
fi Amendment No.109
.s License No. DFR-35 2
1.
The Nuciear Regulatory Commission (the Commission) has found that:
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i A.
Thi aqolication for amendment by Boston Edison Company (the licensee) dai/d, June T.g 1987, as supplemented by e letter dated September 1, 198// 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 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 Conrnission's regulations; D.
The issuance of this amendment will not be inimical to the common defense ind 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 Cemission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical
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Specifications as indicated in the attachment to this license amendment, j
and paragraph 3.B of Facility Operating License No. DPR-35 is hereby amended to read as follows:
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2 (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.109, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This'11 cense amendment is effective 30 days after the date of issuance.
FOR THE NUCLEAR REGULATORY COM ISSION'
./
e Acting Direct Project Direc orate I-3 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 8, 1987 1
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ATTACHMENT TO LICENSE AMEN 0 MENT NO.109 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-243 Replace the following pages of the Appendix A Technical Specifications with i
the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pages are provided to maintain document completeness.
Remove Pages Insert Pager 107 107 108 108 l
109 109 113 113 116 116 117 117 I
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~ LIMITING CONDfTION FOR OPERATf0N' SURVEILLANCE REQUIREMENT
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- 3.5.8 Containment coolina subsystem 4.5.B Containment Coolina subsystem (Cont'd)
(Cont'd) 2.
From and after the date that 2.
When one containment cooling one containment cooling subsystem loop hecomes subsystem loop is made or inoperable, the operable found to be inoperable for subsystem loop and its any reason, continued reactor associated diesel generator operation is permissible only shall be demonstrated to be during.the succeeding seven operable immediately and the days unless such subsystem ~
. operable containment cooling loop is sooner made operable, subsystem loop daily thereafter, provided that the other containment cooling subsystem loop, including its associated diesel generator, is operable.
3.
If the requirements of 3.5.B cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C.
HPCI Subsystem C.
,HPCI Subsystem l.
The HPCI Subsystem shall be 1.
HPCI Subsystem testing shall be operable whenever there is performed as follows:
irradiated fuel in the reactor vessel, reactor a.
Simulated Once/ operating pressure is greater than 150 Automatic cycle psig, and reactor coolant Actuation temperature is greater than Test 365'F; except as specified in 3.5 C.2 and 3.5.C.3 below.
b.
Pump Oper-Once/ month and-ability Once/ cycle from the Alternate Shutdown Station c.
Motor Once/ month Operated and j
Valve Once/ cycle s
Operability from the l
Alternate Shutdown Station d.
Flow Rate Once/3 months
- Conditional relief granted from this LCO for the period October 31, 1980 e.
Flow Rate Once/ operating through November 7, 1980.
at 150 psig cycle
. Amendment No. ff, 109 i
107 1
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LIBLUNG CGNDIT10N FOR OPERATION SURVEILLANCE REQUIREMENT
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3.5.C HPCI Subsystem (cont'd) 4.5.C HPCI Subsystem (Cont'd)
The HPCI pump shall deliver at least 4250 gpm for a system head corresponding to a reactor pressure of 1000 to 150 psig.
2.
From and after the date that 2.
l the HPCI Subsystem is made or When it is determined that the HPCI Subsystem is inoperable l
found to be inoperable for any the RCIC, the LPCI subsystem, reason, continued reactor both core spray subsystems, and operation is permissible only the ADS subsystem actuation during the succeeding seven logic shall be demonstrated to days unless such subsystem is be operable immediately.
The sooner made. operable, RCIC syster ?nd ADS subsystem providing that during such logic shall be demonstrated to seven days all active be operable daily thereafter.
components of the ADS subsystem, the RCIC system, j
the'LPCI subsystem and both j
core spray subsystems are t
4 operable.
3.
If the requirements of 3.5.C cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall l
be reduced to below 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.5.D Reactor Core Icolation Coolina 4.5.0 Reactor Core Isolation Coolino (RCIC) Subsystem (RCIC) Subsystem 1.
The RCIC Subsystem shall be 1..
RCIC Subsystem testing shall be i
operable whenever there is performed as follows-1 irradiated fuel in the reactor l
vessel, reactor pressure is a.
Simulated Once/ operating greater than 150 psig, and Automatic cycle reactor coolant temperature is Actuation greater than 365'F; except as Test specified in 3.5.0.2 below.
b.
Pump Once/ month and Operability Once/ cycle from the Alternate Shutdown
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Station c.
Motor Once/ month and Operated Once/ cycle from l
Valve the Alternate Operability Shutdown Station Amendment No. /2, 109
r LfMITING'CONDITf0N FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.0 Reactor Core Isolation Coolina 4.5.D Reactor Core Isolation Coolina (RCIC) Subsystem (Cont'd)
(RCIC) Subsystem (Cont'd) d.
Flow Rate Once/3 months at 1000 psig e.
Flow Rate Once/ operating at 150 psig cycle The RCIC pump shall deliver at least 400 gpm for a system head corresponding to a reactor pressure of 1000 to 150 psig.
2.
From and after the date that 2.
When it is determined that the.RCIC the RCICS is made or found to subsystem is inoperable, the HPCIS be inoperable for any reason, shall be demonstrated to be operable continued reactor power immediately and weekly thereafter.
operation is permissible only during the succeeding seven days provided that during such seven days the HPCIS is operable.
3.
If the requirements of 3.5.0 cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to or below 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l 3.5.E Automatic Depressurization 4.5.E Automatic Depressurization System System (ADS)
(ADS) 1.
The Automatic Depressurization 1.
During each operating cycle the Subsystem shall be operable following tests shall be performed whenever there is irradiated on the ADS:
fuel in the reactor vessel and the reactor pressure is a.
A simulated automatic actuation greater than 104 psig and test shall be performed prior to prior to a startup from a Cold startup after each refueling 4
Condition, except as specified outage.
in 3.5.E.2 below.
b.
With the reactor at pressure, each relief valve shall be manually opened until a corresponding change in reactor pressure or main turbine bypass valve positions ihdicate that steam is flowing from the valve.
c.
Perform a test from the i
alternate shutdow'n panel to verify that the relief valve solenoids actuate.
Test shall be performed after each refueling outage prior to Amendment 47, E7,109, 109
BASES:
3.5.A Core Spray and LPCI Subsystem This specification assures that adequate emergency cooling capability is available whenever irradiated fuel is in the reactor vessel.
Based on the loss of coolant analysis performed by General Electric in accordance with Section 50.46 and Appendix K of 10CFR50, the Pi'igrim I Emergency Core Cooling Systems are adequate to provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident, to limit calculated fuel clad temperature to less than 2200*F, to limit calculated local metal water reaction to less than or equal to 17%, and to limit calculated core wide metal water reaction to less than or equal to 1%.
The limiting conditions of operation in Specifications 3.5. A.1 through 3.5. A.6 specify the combinations of operable subsystems to assure the availability of the minimum cooling systems noted above.
No single failure of CSCS equipment occurring during a loss-of-coolant accident under these limiting conditions of operation will result in inadequate cooling of the reactor core.
Core spray distribution has been shown, in full-scale tests of systems similar in design to that of Pilgrim, to exceed the minimum requirements by at least 257..
In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irrediated fuel.
The accident analysis takes credit for core spray flow into the core at vessel pressure below 205 psig.
However, the analysis is conservative in that no credit is taken for spray cooling heat transfer in the hottest fuel bundle until the pressure at rated flow for the core spray (104 psig vessel pressure) is reached.
The LPCI subsystem is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident.
This system functions in combination with the core spray system to prevent excessive fuel clad temperature.
The LPCI subsystem and the core spray subsystem provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double-ended recirculation line break without assistance from the high pressure emergency core cooling subsystems.
The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate.
The method and concept are described in reference (1). Using the results developed in Amendment No, JB,109, 113 o
MSIS: -
3.5.C HP_C1 The. limiting conditions for operating the HPCI System are derived from the Station Nuclear Safety Operational Analysis (Appendix G) and a detailed functional analysis of the HPCI System (Section 6).
The HPCIS is provided to assure that the reactor core is. adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear r
system and loss-of-coolant which does not result in rapid depressurization of the reactor vessel.
The HPCIS permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.
The HPCIS continues to operate until reactor vessel pressure is below the pressure at which LPCI operation or Core Spray System operation maintains core cooling.
The capacity of the system is selected to provide this required core cooling.
The HPCI pump is designed to pump 4250 gpm at reactor pressures between 1100 and 150 psig.
Two sources of water are available.
Initially, demineralized water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor.
When the HPCI System begins operation, the reactor depressurizes more rapidly I
than would occur if HPCI was not initiated due to the condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI System. As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through the b?eak. Continued depressurization causes the break flow to decrease below the HPCI flow and the
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liquid inventory begins to rise.
This type of response is typical of the small breaks.
The core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the capacity range of the HPCI.
The analysis in the FSAR, Appendix G, shows that the ADS provides a single failure proof path for depressurization for postulated transients and accidents.
The RCIC is required as an alternate source of makeup to the HPCI only in the case of loss of all offsite A-C power.
Considering the HPCI and the ADS plus RCIC as redundant paths, reference (1) methods would give an estimated allowable repair time of 10 days based on the one month testing frequency. Considering this and the judgments of the reliability of the ADS and RCIC systems, a 7-day period is specified.
The requirement that HPCI be operable when reactor coolant temperature is greater than 365'F is included in Specification 3.5.C.1 to clarify that HPCI need not be operable during certain testing (e.g., reactor vessel hydro testing at high reactor pressure and low reactor coolant temperature).
365'F is approximately equal to the saturation steam temperature at 150 psig.
Amendment No. 109, M6 j
MSIS:
3.5.0 RCIC System The RCIC is designed to provide makeup to the nuclear system as part of the planned operation for periods when the normal heat sink is unavailable.
The nuclear safety analysis, FSAR Appendix G, shows that RCIC also serves as redundant makeup system on total loss of all offsite power in the event that HPCI is unavailable.
In all other postuiated accidents and transients, the ADS provides redundancy for the HPCI.
Based on this and judgments on the reliability of the HPCI system, an allowable repair time of seven days is specified.
Immediate and weekly demonstrations of the HPCI operability during RCIC outage is considered adequate based on judgment and practicality. More frequent testing would cause undesirable steam flow interruption and thermal cycling transients.
The requirement that RCIC be operable when reactor coolant temperature is greater than 365'F is included in Specification 3.5.0.1 to clarify that RCIC need not be operable during certain testing (e.g., reactor vessel hydro testing at high reactor pressure and low reactor coolant temperature). 365'F is approximately equal to the saturation steam temperature at 150 psig.
Amendment No, 109, II7
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