ML20236E167
| ML20236E167 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 07/17/1987 |
| From: | Virgilio M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236E169 | List: |
| References | |
| NUDOCS 8707310258 | |
| Download: ML20236E167 (9) | |
Text
.
e arco UNITED STATES o.!
',g NUCLEAR REGULATORY COMMISSION
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j WASHINGTON, D. C. 20555 g
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i DETROIT EDISON COMPANY l
WOLVERINE POWER SUPPLY COOPERATIVE, INCORPORATED l
DOCKET NO. 50-341 FERMI-2 s
AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 8 License No. NPF-43 1.
The Nuclear Regulatory Commission (the Commission) has found that:
1 A.
The application for amendment by the Detroit Edison Company (the l
licensee) dated March 9, 1987, complies with the standards and i
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; 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
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C.
There is reasonable assurance (i) that the activities authorized by I
this amendment can be conducted without endangering the health and j
l safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; l
D.
The issuance of this amendment will not be inimical to the common l
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.
l 2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-43 is hereby amended to read as follows:
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.08, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license.
DECO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
B7073102go 97o,3y DR ADOCK 05000341 PDR
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.g-I 3.
This amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION W
a Marti J. Virgilio, Acting Director Project Directorate III-1 Division of Reactor Projects - III, IV, V and Special Projects
Enclosure:
Changes to the Technical Specifications l
Date of Issuance: July 17,1987 l
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ATTACHMENT TO LICENSE AMENDMENT N0.-8 l
FACILITY OPERATING LICENSE N0. NPF-43 DOCKET NO. 50-341 l
1 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
The corresponding i
overleaf pages are also provided to maintain document completeness.
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REMOVE INSERT 3/4 5-2 3/4 5 3/4 8-24 3/4 8-24 83/4 6-1 B3/4 6-1 I
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3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with:
The core spray system (CSS) consisting of two subsystems with each a.
subsystem comprised of:
1.
Two OPERABLE CSS pumps, and 2.
An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.
b.
The low pressure coolant injection (LPCI) system of the residual heat removal system consisting of two subsystems with each l
subsystem comprised of 1.
Two OPERABLE LPCI'(RHR) pumps, and 2.
An OPERABLE flow path capable of taking suction from the l
suppression chamber and transferring the water to the reactor-vessel.***
c.
The high pressure cooling injection (HPCI) system consisting of:
1.
2.
An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
d.
The automatic depressurization system (ADS) with at least five OPERABLE ADS valves.
APPLICABILITY:
OPERATIONAL CONDITION 1, 2* ** # and 3* **.
- The HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 150 psig.
- Upon receipt of an LPCI initiation signal, operator action is required to manually open the torus suction valves to facilitate LPCI operation if the LPCI system is in the RHR shutdown cooling mode of operation per Specifica-tion 3 4.9.1.
- See Special Test Exception 3.10.6.
FERMI - UNIT 2 3/4 5-1 Amendment No. 8 L------------
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J EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
)
l ACTION:
a.
For the core spray system:
1.
With one CSS subsystem inoperable, provided that at least one j
LPCI pump in each LPCI subsystem is OPERABLE, restore the i
inoperable CSS subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within tLe following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
With both CSS subsystems inoperable, be in at least HOT i
SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next i
l 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
For the LPCI system:
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1.
With one LPCI pump in either or both LPCI subsystems inoperable, provided that at least one CSS subsystem is OPERABLE, restore the inoperable LPCI pump (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in i
COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
With one LPCI subsystem otherwise inoperable, provided that both CSS subsystems are OPERABLE, restore the inoperable LPCI subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
With a LPCI system cross-tie valve closed, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i 4.
With both LPCI subsystems otherwise inoperable, be in at least'
)
HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.*
5.
The provisions of Specification 3.0.4 are not applicable for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the purpose of establishing the RHR system i
in the LPCI mode once the reactor vessel pressure is greater than the RHR cut-in permissive setpoint.
c.
For the HPCI system, provided the CSS, the LPCI system, the ADS and the RCIC system are OPERABLE:
1.
With the HPCI system inoperable, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to 1 150 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
FERMI - UNIT 2 3/4 5-2 Amendment No. 8
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3/4.6 ' CONTAINMENT SYSTEMS l
BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive mate-I rials.from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation-doses to within the limits'of 10 CFR Part 100.during accident conditions.
3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure of 56.5 psig, P.
As an added conserva-a tism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L during performance of the periodic tests to account for a
possible degradation of the containment leakage barriers between leakage tests.
Operating experience with the main steam line isolation valves has
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indicated that degradation has occasionally occurred in the leak tightness of j
the valves; therefore the special requirement for testing these valves.
The surveillance testing for measuring leakage rates is consistent with i
the requirements of Appendix J of 10 CFR Part 50 with the exception of exemptions granted for main steam isolation valve leak testing and testing the airlocks after each opening and analyzing the Type A test data.
Appendix J to 10 CFR Part 50, Paragraph III.A.3, requires that all Type A tests be conducted in accordance with the provisions of N45.4-1972, " Leakage.
.i Rate Testing of Containment Structures for Nuclear Reactors." N45.4-1972 1
requires that Type A test data be analyzed using point-to point or total time analytical techniques.
Specification 4.6.1.2a requires use of the mass plot analytical technique.
The mass plot method is considered the better analytical technique, since it yields a confidence interval which is a small fraction of the calculated leak rate; and the interval decreases as more data sets are added to the calculation.
The total time and point-to point techniques nay give con-fidence intervals, which are large fractions of the calculated leak rate, and the intervals may increase as more data sets are added.
The mass plot method is endorsed by ANSI /ANS 56.8-1981 (Containment System Leakage Requirements) which superseded N45.4-1972.
3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The. limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6.1.1 and 1
1 1
FERMI - UNIT 2 B 3/4 6-1 Amendment No. 8 i
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CONTAINMENT SYSTEMS BASES PRIMARY CONTAINMENT AIR LOCKS (Continued) 3.6.1.2.
The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation.
Only one closed door in each air lock is required to maintain the integrity of the containment.
1 3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM i
Calculated doses resulting from the maximum leakt ge allowance for the main steamline isolation valves in the postuleted LOCA situations would be a small fraction of the 10 CFR Part 100 guidelines, provided the main steam line system l
from the isolation valves up to and including the turbine condenser remains i
intact.
Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not always been maintained continuously.
The requirement l
for the leakage control system will reduce the untreated leakage from the MSIVs when isolation of the primary system and containment is required.
3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 56.5 psig in the event of a LOCA.
A visual inspection in conjunction with Type A leakage tests is sufficient to I
demonstrate this capability.
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3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE i
l The limitations on drywell and suppression chamber internal pressure ensure that the containment peak pressure of 56.5 psig does not exceed the maximum allowable pressure of 62 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 2 psid.
3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 340 F during LOCA conditions and is consistent with the safety analysis.
i 3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM Leakage integrity tests with a maximum allowable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage failure develops.
The 0.60 L leakage limit shall not be a
exceeded when the leakage rates determined by the leakage integrity tests of these vclves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.
FERMI - UNIT 2 B 3/4 6-2 Amendment No. 8
l TABLE 3.8.4.3-1 (Continued).
i MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION j
SYSTEM (S)
VALVE NUMBER AFFECTED E41-F022 HPCI E41-F041 HPCI E41-F042 HPCI E41-F059 HPCI E41-F075 HPCI E41-F079 HPCI E41-F600 HPCI j
(
7.
E51-F001 Reactor Core Isolation Cooling
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System (RCIC)
I E51-F002 RCIC.
j E51-F007 RCIC
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E51-F012 RCIC E51-F013 RCIC E51-F019 RCIC-E51-F022 RCIC E51-F029 RCIC E51-F031 RCIC E51-F045 RCIC E51-F046 RCIC E51-F059 RCIC E51-F062 RCIC-E51-F084 RCIC-8.
G1154-F018 Drywell Floor Drain System G1154-F600 Drywell Floor Drain System j
j 9.
G33-F001 Reactor Water Clean-Up System (RWCU) i G33-F004 RWCU 10.
G51-F600 Torus Water Management System (TWMS)
G51-F601 TWMS G51-F602 TWMS G51-F603 TWMS G51-F604 TWMS G51-F605 TWMS G51-F606 TWMS G51-F607 TWMS
- 11. N11-F607 Main Steam System N11-F608 Main Steam System N11-F609 Main Steam Systes N11-F610 Main Steam System f
FERMI - UNIT 2 3/4 8-23 Amendment No. 8 I
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cy M'
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W TABLE 3.8.4.3-1 (Continued) f f
MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION
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SYSTEM (S)
VALVE NUMBER AFFECTED 12.
DELETED l
13.
P44-F601A Emergency Equipment Cooling Water-(EECW)
P44-F601B EECW P44-F602A EECW P44-F602B EECW P44-F603A EECW P44-F6038 EECW P44-F604 EECW P44-F605A EECW P44-F605B EECW P44-F606A EECW P44-F606B EECW P44-F607A EECW.
P44-F607B EECW P44-F608 EECW P44-F613 EECW P44-F614 EECW P44-F615 EECW-P44-F616 EECW 14.
P50-F603 Compressed Air Systems P50-F604 Compressed Air Systems 15.
T48-F601A Containment Atmosphere Control-System T48-F601B Containment Atmosphere Control System T48-F602A Containment Atmosphere Control System I
T48-F602B Containment Atmosphere Control System T48-F603A Containment Atmosphere Control System T48-F603B Containment Atmosphere Control System T48-F604A Containment Atmosphere Control System T48-F604B Containment Atmosphere Control System T48-F605A Containment Atmosphere Control System T48-F605B Containment Atmosphere Control System T48-F606A Containment Atmosphere Control System T48-F606B Containment Atmosphere Control System
.T4803-F601 Containment Atmosphere Purging System T4803-F602 Containment Atmosphere Purging System
- 16. -T49-F601 Primary Containment Pneumatic Supply System T49-F602 Primary Containment Pneumatic Supply System FERMI - UNIT 2 3/4 8-24' Amendment No. 8