Text

Amend 87 to License NPF-43,changing Licensed Thermal Power Level of Reactor from Current Limit of 3,293 Mwt to Increase Limit to 3,430 Mwt
	ML20105D276
Person / Time
Site:	Fermi
Issue date:	09/09/1992
From:	Murley T; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20105D277	List: ML20105D276; ML20105D284;
References
	NUDOCS 9209240503
	Download: ML20105D276 (41)
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DEIROIT EDISON COMPANY FERMI-R DQ[KE1 NO. 50-341 6dL@ DENT 10 FAClll1Y OPERAllNG Ll((MI Amendment No.87 License No. N0f-43 1.

The Nuclear Regulatory Conmission (the Connission) has found that:

A.

The application for amendment by the Detroit Edison Company (the licensee) dated September 24, 1991, and modified January 31, and April 30, 1991, 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 rulet and regulations of the Commission; C.

There is censonable assurante (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 Connission'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 I.

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, paragraph 2.C.(1) of f acility Operating License No. NPF-43 is hereby amended to read as follows:

(1) Maximum Power Level DECO is authorized to operate the facility at reactor core power levels not in excess of 3430 megawatts thermal (100% power) in accordance with the conditions specified herein and in Attachment I to this license.

The items identified in Attachment I to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license.

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Further, the license is amended by changes to the Technical Specifications as indicated in the attachment 'a this ifcense amendment and paragraph 2.C.(2) of f acility Operating License No. NPf-43 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan Ibe Technical Specifications contained in Appendix A, as revised through Amendment No. 87

, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

DECO shall operate the facility in accordanr.e with the Technical I

Specifications and the Environteental Protection Plan.

4.

This license arandment is effective as of beginning of the third refueling outage, currently scheduled for September 12, 1992, with full implementation prior to startup from the third refueling outage.

FOR THE NUCLEAR REGULATORY COMMIS$10N Cr Thomas E. Murley, Director Office of Nuclear Reactor Regulation Attachment

1. Pages 3 and 4 of license *

2. Changes to the Technical Specifications Date of Issuance: Mbn'e-9,19E

Pages 3 and 4 are attached, for convenience, for the composite license to reflect this change.

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, (4) Df.Co, sursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material such as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration and as fission detectors in amounts as required; (5) Deco, pursuant to the Act and 10 CFR Parts 30, 40 c.ad 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) DECO, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess,.but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This license shall be deemed to contain end is subject to the conditions spes fied in the Commission's regulations set forth in 10 f.FR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified of incorporated below:

(1) tiaximum Power level DECO is authorized to operate the facility at the reactor core power levels not in excess of 3430 megawatts thermal (100% power) in l

acccMance with the conditions specified herein and in Atttchment 1 to this license. The itns identified in Attachment I to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No.

87 and the Environmental Protection Pitn contained in Appendix B, are hereby incorporated in the license.

DECO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Antitrust Conditions DECO shall abide by the agreements and interpretations between it and the Department of Justice relating to Article 1, Paragraph 3 of the Electric Power Pool Agreement between Detroit Edison Company and Art tadment No. -r,4 t," B7

4 l Consumers Power Company as specified in a letter from Deco to the Director of Regulation, dated August 13, 1971, and the letter from

~j Richard W. McLaren, Assistant Attorney General, Antitrust Division, U.S. Department of Justice, to Bertram H. Schur, Associate General Counsel,_ Atomic Energy Comission, dated August 16, 1971.

(4) Safetv/ Relief Valve In-Plant _Ig;11no (Section 3.8.1. SSER iSJ!

Prior to completing the startup test prog *am, Deco shall perform a series of in-plant tests of the safety /rel'ef v11ves ($RVs). The acceptance criteria for these tests are containsd in Section 2.13.9,

SRV Load Assessment by In-Plant Tests" of NUREG-0661, 'NRC Acceptance Criteria for the Mark _! Containment Long-Term Program."

The results of these tests shall be reported to the NRC staff-within six months of completing this test series.

(5)

Suopression Pool Temoerature Measurements (Soction 3.8.1. SSER 85)

Deco shall accomplish during the first fuel cycle, all the tasks described in its letter dated March 6, 1985, regarding the series of SRV tests which will confirm its methodology for measuring the suppession pool bulk temperature.

(6)

Environmental Oualification (Section 3.11. SSER 85)

No later than November 30, 1985, Deco shall environmentally qualify all electrical equipment according to the provisions of 10 CFR 50.49.

(7)

Control Room Habitability (31ction'6.4.1. SSER *6) 1 Prior to startup following the first refueling outage, Deco shall provide assurance to the NRC staff that potential contamination pathways through those portions of the control room air-conditioning system which are external to the control room zone will net have a significant adverse impact on control room habitability, or will propose a technical specification which provides for periodic leakage _ testing to assure the integrity of those external portions of the control room air-conditioning systemc i

The parenthetical notation following the title of many license conditions denotes the section of-the_ Safety Evaluation Report (SER).and/or its-i supplements wherein the license condition is discussed..

Amendment No. 87 w

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4 ATTACHMENT TO L! CENSE AMENDHENT NO. 87 TACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341

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Replace the following pages of the Appendix *A' Technical Specifications with tne attached pages.

The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.

Pages marked with an aster'isk (*) indicate overleaf pages and do not contain changes.

l RE! ICY [

INSERT l-5 1-5 2-4 2-4 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 2-1 3/4 2-1 3/4 3-4 3/4 3-4 3/4 3-5 3/4 3-5 3/4 3-15 3/4 3-15 3/4 3-44 3/4 3-44 3/4 4-1 3/4 4-1 3/4 4-2 3/4 4-2 3/4 4-6a 3/4 4-6a 3/4 4-7 3/4 4-7 3/ 4 4-B' 3/4 4-8*

3/4 4-9*

3/4 4-10 3/4 4-10 3/4 4-21 3/4 4-21 3/4 4-23 3/4 4-23 3/4 4-24*

3/4 4-24*

3/4 4-31 3/4 4-31 3/4 5-3*

3/4 5-3*

3/4 5-4 3/4 5-4 3/4 7-13*

3/4 7-13*

3/4 7-14 3/4 7-14 3/4 8-23 3/4 8-23 8 3/4 1-4 8 3/4 1-4 8 3/4 2-la B 3/4 2-la B 3/4 2-4 B 3/4 2-4 B 3/4 4-1 B 3/4 4-1 B 3/4 4-la B 3/4 4-la B 3/4 4-8 B 3/4 4-8 B 3/4 6-1 B 3/4 6-1 B 3/4 6-2 B 3/4 6-2 B 3/d o-3 B 3/4 6-3 P 3/4 6-4 B 3/4 6-4 B 3/4 7-5 B 3/4 7-5 6-2) 6-21

e DEFINITIONS 2.

Closed by at least one manual valve, blank flange, or deactivtted automatic valve secured in its closed position, except as provided in Table 3.6.31 or Specificatien 3.6.3.

b.

All primary containment equipment hatches are closed and sealed, c.

Each primary containment air lock is in compliance with the requirements cf Specification 3.6.1.3.

d.

The primary containment leakage rates are within the limits of Specification 3.6.1.2.

e.

The suppression chamber is in compliance with the requirement of Specification 3.6.2.1.

f.

The sealing mechenism associated with each primary containment penetration, e.g., welds, bellows, or 0 rings, is OPEPABLE.

g.

The suppression chamber to reactor building vacuum breakers are in compliance with Specification 3.6.4.2.

THE PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 State regulat ons, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.31 PURGE or PURGING is the contro11eG process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

MTED THERMAL POWER 3.32 RAT [0 THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3430 MWT.

l FERMI - UNIT 2 1-5 Amendment No. ft, 82, 87

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TARIE 2.2.1-1 REBCTOR PROTICTION sisiiH InsiRUMENTRTICR SFTPolNTS ALLOWABLE TRIP SETPOINT VALUES _

FUNCTIONAL UNIT 93 1.

Intermediate Range Monitor, Neutrcn Flux - liigh s 120/125 divisions of s 122/125 divisions of full scale of full scale 2C 2.

Average Pnwer Range Monitor:

Neutron Flux-Upscale, Setdown s 15% of RATED

< 20% of RATED THERMAL POWER THERMAL POWER a.

-4 b.

Flow Blased Simulated Therma Power-Upscale

1) During two recirculation loop operation:

< 0.63 W+61.4%, with 5 0.63 W+64.3%, with a.

Flow Biased a maximum of a maximum of s 113.5% of RATED s 115.5% of RATED b.

High Flow Clamped THERMAL POWER THERMAL POWER

2) During single recirculation loop operation:

s 0.63W+56.3%,**

s 0.63W+59.2%,**

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a.

Flow Biased NA NA b.

High Flow Clamped Fixed Neutron Flux-Upscale s 118% of RATED s 120% of RATED THERMAL POWER THERMAL POWER c.

NA NA d.

Inoperative 3.

Reactor Vessel Steam Dome Pressnre - High s 1093 psig s III3 psig E

2 173.4 inches

2 171.9 inches 2

4.

Reactor Vessel low Water level - Level 3 E

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See Bases Figure B 3/4 3-1.

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- During single recirculation loop operation, rather than adjusting the APRM Flew Blased Setpoints to comply with the single loop values, the gain of the APRMs may be adjusted for a period not to exceed f

w.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months such that the final APRM readings are at least 5.1% of rated power greater than 100% times FRTP, provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL POWER and a notice v.

2" 93 of adjustment is posted on the reactor control panel.

A REACTIVITY CONTROL Sy,ilLH1 SVRVEltlANCE RE0VIREMENTS (Continuedl b.

At l u st once per 31 days by:

1.

Verifying the continuity of the explosive charge.

2.

Determining that the concentration of boron in solution is within the limits of Figure 3.1.51 by chemical analysis.'

3.

Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise securei in position, is in its correct position.

c.

Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm at' a pressure of greater than or equal to 1215 psig is met, j

d.

At least once per 18 months during shutdown by:

1.

Initiating one of the standby liquid control system loops, including an explosive valve, and ver'.fying that a flow pain from the pumps to the reactor pressure vessel is available by pumping demineralized water into tie reactor vessel.

The replacement charge for the explosive valve shall be from the same manufactured t.atch as the one fired or from another batch which has been certified by having one charge of that batch successfully fired.

Both injection loops shall be tested in 36 months.

2.

Demonstrating that the pump relief valve setpoint is less than or equal to 1400 psig and verifying that the relief valve does not actuate during recirculation to the test tank.

3.

Demonstrating that all piping between the storage tank and the explosive vaives is unblocked by pumping from the storage tank to the test tank and then draining and flushing the piping with demineralized water.**

4.

Demonstrating that the storage tank heaters are OPERABLE for mixing by verifying the expected temperature rise of the sodium pentaborate solution in the storage tank after the heaters are energized, e.

At least once per 18 months sample and analyze the sodium pentaborate solution to verify that the Boron 10 Isotope enrichment exceeds 65 atom percent.

This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below the 48'F limit.

- This test shall also be performed whenever the solution temperature drops below the 40*f limit and may be performed by any series of sequential, overlapping or total flow path steps such that the entire flow path is included, l

FERM1 - UNIT;2 3/4 1-20 Amendment No. 3,87 t_____

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FERMI - UNIT 2 3/4 1 21 Amendment No. #,87

_-- _ -. - _ - - - - _ - - _ _ _ _ - _ - _ - _ _ _ ~. _ _..

r 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE L

llMITING.4,JITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall not exceed:

a.

The MAPLHGR limit which has been approved for the respective fuel and lattice type as a function of tie average planar exposure (as determined by the NRC approved methodology described in.

GESTARll),or b.

When hand calculations are required, the most limiting lattice type MAPLHGR limit as a function of the average planar exposure-shown in the CORE OPERATING LIMITS REPORT (COLR) for the applicable fuel type, l

APPL 1 CAB 1t1TY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or j

equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the above limits, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or-4 reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within.the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVE1LLANCE RE0V1REMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the-limits' required by Specification 3.2.1:

f; La.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion ofz a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor-is-operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.-

f a

FERMI --UNIT 2 3/4 2 1 Amendment No, ff, J),l%/,87 y

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l TABLE 3.3.1 1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATfMENTS Be in at least HOT SHUTDOWN within 12 hout t.
ACTION 1 Verify all insertable control rods to be it.e<rted in the core ACTION 2 and lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
Suspend all operations involving CORE ALTERATIONS and insert ACTION 3 all insertable control rods within I hour.
Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
ACTION 4 Be in STARTUP with the main steam line isolation valves closed ACTION 5 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Initiate a reduction ir, THERMAL POWER within 15 minutes and ACTION 6 redute turbine first stage pressure to s 161.9 psig, equivalent j
to THERMAL POWER less than 30% of RATED THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
Verify all insertable control rods to be inserted within I Atl10N 7 hour8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .
Lock the reactor mode switch in the Shutdown position within 1 Ati10N 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
Suspend all operations involving CORE ALTERATIONS, and insert ACTION 9 all insertable control rods and lock the reactor mode switch in the Shutdown position within I hour, i
FERMI - UNIT 2 3/4 3-4 Amendment No.67
TABLE 3.LJ;l (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status fer up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.
(C) Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, the " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn.'
(d) When the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRM',, 6 IRiis and per Specification 3.9.2, 2 SRMs.
(0) An APRM channel is inoperable if there are let,s than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.
(f) This function is not requireo to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(C) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
(h) This function is not required to be OPERABLE when PRIMARY CONTAINMEfD INTEGRili is not required.
(i)
With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
U) This function shall be automatically bypassed when turbine first stage pressure is s 161.9 psig, equivalent to THERMAL POWER less than 30% of l
RATED THERMAL POWER.
Not required f or control rods removed per Specification 3.9.10.1 or 3.9.10.2.
FERM) - UNIT 2 3/4 3 5 Amendment No. 7E. 87
TABLE 3.3.2-2 "x3 1501ATION ACTUATION INSTRUMENTATION SEiPOINTS 9
All0WABLE c
TRIP FUNCTION TRIP SETPOINT VALUE 1.
PRIMARY CONTAINMENT ISOLATION a.
Reactor Yessel low Water level 1) level 3
> 173.4 inches *
> 171.9 inches 2)
Level 2
> 110.8 inches
z 103.8 inches 3) levei 1 2 31.8 inches *
> 24.8 inches b.
Drywell Pressere - High s 1.68 psig s 1.88 psig
,s
[
c.
Main Steam tir.e G
1)
Radiation - High s 3.0 x full power background s 3.6 x full power background 2)
Pressure - Low 2 756 psig 2 736 psig 3)
Flow - High 5 115.4 psid 5 118.4 psid i
E
=
d.
Main Steam Line Tunnel Temperature - High s 200'F s 206'F e.
Condenser Pressure - High s 6.85 psia s 7.05 psia A
z?
f.
Turbine 81dg. Area g
Temperature - High s 200*F s 206'F h
g.
Deleted
~
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h.
Manual Initiation NA NA l
MMM M TABLE 3.3.6-2 CONTROL ROD BlogK THSTRUMENTATION SETPOINTS ALLWABLE VAll)r TRIP SETPOINT
-,s y
TRIP FUNCTION As specified in the CORE OPERAilNG 1.
R00 BLOCK MONITOR As specified in the a.
Upscale EORE OPERATING LIMITS REPORT LIMITS RLPORT c
NA NA b.
Inoperative 2 92.3% of Reference level
> 94% of Referenr3 Level c.
Dornscale I
2.
APRM s 0.63 W + 58.59 Flow Blased Neutron Flux - liigh s 0.63 W + 55.6%*
with a maximum of a.
1) During two recirculation with a maximum of 1I0%
loop operatior 108%
I 8
s 0.63 W + 53.4%
s 0.63 W + 50.5% *
2) During single recirculation NA loop operation NA 2 3% of RATED THERMAL POWER yg b.
Inoperative 5% of RATED THERMAL POWER Y
c.
Downscale s 12% of RATED THERMAL POWER s 14% of RATED TilERMAL 2
A d.
Neutron Flux - Upscale, Setdown K4 3.
SOURCE RANGE MONITORS NA s 1.6 x 105 cp3 Detector not full in s 1.0 x 105 cps a.
g NA b.
Upscale NA 2 2 cps **
g Inoperative c.
2 3 cps **
o.
S d.
Downscale
. r+
~'the APRM rod block function is veled as a function of recirculation loop drive flow (W).
20.
0.7 cps provided the signal-to-noise ratto 3 1y with the single loop values, the gain of l
May be reduced to 3
uring single rectreulation loop operation. rather than adjusting the APRM Flow Blased setpoints to cry,f hal APsM resdings are THERMAL POWER and a notice of adjusteret is posted o.e the reactor the APRMs may tie adjusted for a period not to oceed f2 tours such that thetimes TR1". provided that th 0
g
.g control panel.
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3/4.4 REh_ TOR COOL ANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM REC.lRCELAIJON LOOPS LIMITING COND1110N FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.
APPLICAEll.Ul: OPERATIONAL CONDITIONS 1 and 2*.
ACTION:
With one reactor coolant system recirculation loop not in operation:
a.
1.
Witoin 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:
a) Place the individual recirculation pump flow controller for the operating recirculation pump in the Manual mode.
b) Reduce THERMAL POWER to less than or equal to 67.2% of RATED l
THERMAL POWER.
c) Limit the speed of the operating recirculation pump to less than or equal to 75% of rated pump speed.
d)
Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2.
i l
e) Reduce the Average Power Range Monitor (APRM) Scram and Ro/ Block I
Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6.
f)
Perform Surveillance Requirement 4.4.1.1.4 if THERMAL POWER is l
less than or equal to 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is less than or equal to 50% of rated loop flow.
2.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , With no '.eactor coolant system recirculation loop in operation while in b.
OPERATIONAL CONDITION 1, immediately place the Reactor Mode Gwitch in the.
SHUTDOWN position, With no reactor coolant system recirculation loops in operation, while in c.
OPERATIONAL CONDITION 2, initiate measures to place the unit in at least HOT $HUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
See Special Test Exception 3.10.4.
APRM gain adjustments may be made in lieu of adjusting the APRM Flow Biased Setpoints to comply with the single loop values for a period of up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .
l 1
FERMI - UNIT 2 3/4 4-1 Amendment No. EJ,Ef,%9,EJ,67
REACTOR COOLANT SYST(d SVRVElllANCE RE0VIREMENTS 4.4.1.1.1 Each pump discharge valve shall. be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each STARTilP* prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER.-
4,4.1.1.2 Each pump MG set scoop tube mechanical and electrical stop shall be-demonstrated OPERABLE with overspeed setpoints less than or equal to 110%-
and 107%, respectively, of rated core flow, at least once per 18 months.
4.4.1.1.3 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that:
1 a.
THERMAL POWER is less than or equal to 67.2% of RATED' l
THERMAL POWER, and b.
The individual recirculation pump flow controller for the operating recirculation pump is in the Manual mode, and c.
The speed of the operating recirculation pump is less than or equal to 75% of rated pump speed.
4.4.1.1.4 With one reactor coolant system loop not in operation with THERMAL POWER less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of rated. loop flow, verify the following differential temperature requirements are met within no more than 15 minutes prior to either THERMAL POWER increase or recirculation flow increase:
a.
Less than or equal to 145'F bctween reactor vessel steam space coolant and bottom head drain line coolant, and b.
Less than or equal to 50*F between the reactor coolant within the loop not in operation and the coolant.in the reactor pressure vessel **, and c.
Less than or equal to 50*F between the reactor coolant within the loop not in operation and the operating loop.**
1f not performed within the previous 31 days.
Requirement does-not apply when the recirculation loop not in operation is isolated from the reactor pressur] vessel.
FERMI - UNIT 2 3/4 4-2 Amendment No. EJ, ES, 87-
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3/4.4.2 SAFETY / RELIEF VALVES SATETY/ RELIEF YALVES L1HlTING CONDITION FOR OPERATION 3.4.2.1 The safety valve function of at least 11 of the following reactor coolant syste7 safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*
5 safety / relief valves 01135 psig il%
l 5 safety / relief valves 0 1145 psig al%
5 safety / relief valves 0 1155 psig al%
APPLIC ABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
a.
With the safety valve function of less than 11 of the above safety / relief valves OPERABLE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b.
With one or more safety / relief valves stuck open, provided that suppression pool average water tempert.ture 4 less than 95'F, close the stuck open safety / relief valve (s),.f v' ale to close the stuck open valve (s) within 2 minutes or if supprenton pool average water temperature is 95'F or greater, place the reactor mode switch in the Shutdown position.
c.
With one or more safety / relief valve position indicators inoperable, restore the inoperable indicator (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
SURVEILLANCE REQUIREMENTS 4.4.2.1.1 The valve position indicator for each safety / relief valve shall be demonstrated OPERABLE with the pressure setpoint of each of the tail-pipe pressure switches verified to be 30 2 5 psig by performance of a CHANNEL CAllBRATION at least once per 18 months.
4.4.2.1.2 At least 1/2 of the safety relief valves shall be set pressure tested at least once per 18 months, such that all 15 safety relief valves are set pressure tested at least once per 40 months.
The lift setting pressure shall correspond to ambient conditions of the v&lves at nominal operating temperatures and pressures.
I FERMI - UNIT 2 3/4 4-7 Amendment No.67
PE ACTOR COOL ANT SYSTEM SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION tlMITING CONDITION FOR OPERATION 3.4.2.2 The low-low set function of the following reactor coolant system safety / relief valves shall be OPERABLE with the following settings:
Low-Low Set Function Low-Low Set Function Setooint (esial Allowable Value (osial Valve No, M
[lgig Dagn (lgig F013A 1017 905 1037 F013G 1047 935 1067 APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With the low-low set function of one of the above required reactor a.
coolant system safety / relief valves inoperable, restore the inoperable low low set functien to OPERABLE status within 14 days or be ir at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.
With the low-low set function of both of the above required reactor coolant system safety / relief valves inoperable, be in at least HDT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
SURVElLLANCE REQUIREMENTS 4.4.2.2 The low-low set function pressure attuation instrumentation shall be demonstrated OPERABLE by performar 9 of a:
a.
CHANNEL FUNCTIONAL TEST at least once per 31 days, b.
CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic operation of the entire system at least once per 18 months.
Closing pressure must be at least 100 psi less than actual opening pressure.
FERMI - UNIT 2 3/4 4-8
REACTOR COOLANT-SYSTEM 3/4-4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITINr CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE:
a.
The primary containment atmosphere gaseous radioactivity monitoring.
system channel.
b.
The primary containment sump flow monitoring system consisting of:-
1.
The drywell floor drain sump level, flow and pump-run time system, and 1
2.
The drywell equipment drain sump level, flow and' pump-run-time system.
l c.
The drywell floor drain sump level monitoring system.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
i ACTION:
With only two of the above-required leakage detection systems OPERABLE,-
restore the inoperable detection system to OPERABLE status. within 30 days;-
when the. required gaseous radioactive monitoring system is inoperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least on.e per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,.
otherwise, be~in at least HOT SHUTOOWN within the:next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
SURVEILLANCE RE001REMFNTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by:
a.
Primary containment-atmosphere gaseous monitoring systems-performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a CHANNEL-FUNCTIONAL TEST at'least once per 31 days and a-CHANNEL CALIBRATION at least once per 18 months.
b.
Primary containment sump flow and drywell floor drain sump level monitoring systems-performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days:and a CHANNEL CALIBRATION TEST at least once-per 18 months.
FERMI - UNIT 2.
3/4 4-9
t REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE.
b.
5 gpm UNIDENTIFIED LEAKAGE.
c.
25 gpm total leakage averaged over any 24-hour period, d.
1 gpm leakage at a reactor coolant system pressure of 1045
10 psig from j
any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1, e.
2 gpm increase in UNIDENTIFIED LEAKAGE within any 4-hour period.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b.
With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
c.
With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one other closed manual, deactivated automatic, or check
valve, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
d.
With one or more of the high/ low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-2-inoperable, restore the inoperable monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; restore the inoperable monitor (s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN vithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
e.
With any reactor coolant system UNIDENTIFIED LEAKAGE increase greater than 2 gpm within any 4-hour period, identify the source of leakage increase as not service sensitive Type 304 or 316 austenitic stainless steel within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
Which has been verified not to exceed the allowable leakage limit at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.
FERMI - UNIT 2 3/4 4-10 Amendment No. 87
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i FERMI - UNIT 2 3/4 4-21 Amendment No. 77,67
s REACTOR COOLANT SYSTEM-REACTOR STEAM DOME LIMITING-CONDITION FOR OPERATION 3,4.6.2 The pressure in the reactor steam dome shall be less Lthan-1045 psig.
l APPLICABILITY:
OPERATIONAL CONDITIONS 1* and 2*.
ACTION:
With the reactor steam dome pressure exceedingfl045 psig, reduce the pressure-to less than 1045 psig within 15 minutes or be in at least HOT SHUTDOWN with.in - !
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
SURVE1LtANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to-be less-than J045 psig at least'once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
_l_
Not applicable during anticipated transients.
jFERMI - UNIT 2 3/4 4-23 Amendment.No 87
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-O REACTOR COOLANT SYSTEM 3/4.4.7 - MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 4
3.4.7 Two rain steam line isolation valves (MSIVs) per main steam line shall
~
be 0PERABLE with closing times greater than or equal to 3 seconds and less than or equal to 5 seconds.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, end 3.
ACTION:
a.
With one or more MSIVs inoperable:
1.
Maintain at least one MSlY OPERABLE in each-affected main steam line that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:
a)-
Restore the inoperable valve (s) to OPERABLE status, or b)
Isolate the affected main steam line by use of a deactivated MSly in the closed position.
2.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
l-SURVE1LLANCE RE0VIREMENTS
~ 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 5 seconds when tested pursuant to Specification 4.0.5.
FERM] --UNIT 2 3/4 4-24 Amendment No. 83
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EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION:
(Continued) d.
For the ADS:
1.
With one of the above required ADS valves inoperable, provided the HPCI system, the CSS and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to s 150 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
2.
With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to s 150 psig wi. thin the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
e.
With a CSS header AP instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or determine the CSS header AP locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; otherwise, declare the associated CSS subsystem inoperable, f.
With cn LPCI or CSS system discharge line " keep filled" alarm instrumentation inoperable, perform Surveillance Requirement 4.5.1.a.l.a.
9 In the event an ECCS system 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 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.
SURVE1LLANCE REQUIREMENTS 4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:
a.
At least once per 31 days:
l.
For the CSS, the LPCI system, and the HPCI system:
a) VerifyinD by venting at the high point vents tL t the system piping from the pump discharge valve to the system isolation valve is filled with water.
b) 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.
2.
For the_LPCI system, verifying that the cross-tit valve is open.
Except tnat an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may oe in position for another mode of operation.
FERMI - UNIT 2 3/4 5-3
4 8
l EMERGENCY CORE COOLING SYSTEMS SURVElllANCE RE0VIREMENTS (Continued) 3.
For the HPCI system, verifying that the HPCI pump flow controller is in the correct position, b.
Verifying that. when pursuant to Specification 4.0.5:
1.
The two CSS pumps in each subsystem together develop a flow of at least 6350 gpm against a test line pressure of greater than or equal to 270 psig, corresponding to a reactor vessel pressure of a 100 psig.
2.
Each LPC1 pump in each subsystem develops a flow of at least 10,000 gpm against a test line pressure of a 230 psig, corresponding to a reactor vessel to primary containment differential pressure of a 20 psig.
3.
The HPCI pump develops a flow of at least 5000 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure including injection line losses when steam is being supplied to the turbine at 1025 +20, -80 psig.*
l c-At least once per 18 months:
1.
For the CSS, the LPCI system, and the HPCI system, performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.
2.
for the HPCI system, verifying that:
a) The system develops a flow of at least 5000 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure including injection line losses when steam is being supplied to the turbine at 165 + 50, -0 psig.*
b) The suction for the '.iPCI system is automatically transferred fron the condensate storage tank to the suppression chamber on a condensate storage trnk water level - low signal and on a suppression chamber - water level higt, signal.
3.
Performing a CHANNEL CALIBRATION of the CSS and the LPCI system discharge line " keep filled" alarm instrumentation.
4.
Performing a CHANNEL CAllBRATION of the CSS header AP instrumentation and verifying the setpoint to be s the allowable value of 1.0 psid.
The provisions of Specification 4.0.4 are not applicable provided th-surveillance is performed within 12 houts after reactor steam pres-is adequate to perform the test.
l FERM1 '- UNIT 2 3/4 5-4 Amendment No. 87
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.u TABLE 3.7.31-(Continued).
SURVEY POINTS FOR SHORE BARRIER
SURVEY LOCATION **
DECEMBER 1984 POINT NORTH SOUTH EAST WEST CONTROL ELEVATION 9A N7529 E5948 583.04.
9B N7531 E5961 582.10
)
9C N7531
.E5965 579.91=
9D N75?6 E5973 575.13.
j 10A N7612 E5937 583.85 10B N7610 E5950 582.21.
10C N7618 E5961
-582.56 10D N7616 E5972
-576.58-11A N7721 E5940 583.15 i
llB N7721 E5956
.582.03.
11C N7718 E5963 579.82 11D N7722 E5971 576.43..
12A N7814 E5949 581.86 12B N7809 E5955-581.11 12C N7814 E5965 578 88 12D N7815 E5975 577.81 Measuring reference points are anchored into the capstones using.
conter notched >self-drilling bolts.
See Figure B 3/4.7.3-1 for location sketch.-
-FERMI:- UNIT 2 3/4-7-13
PLANT SYSTEMS 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM llM1 TING CONDITION FOR OPERATION The reactor core isolation cooling (RCIC) system shall be OPERABLE with 3.7.4 an OPERABLE flow path capable of taking suction from the suppression pool and transferring the water to the reactor pressure vessel.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig.
ACTION.
With the RCIC system inoperable, operation may continue provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days, otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or egn1 to 150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
SURVEltLANCE REQUIREMENTS 4.7.4 The RCit system shall be demonstrated OPERABLE:
At least once per 31 days by:
a.
1.
Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.
2.
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.
3.
Verifying that the pump flow controller is in the correct
position, b.
At least once per 92 days by verifying that the RCIC p"mp develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure including injection line losses when steam is being supplied to the turbine at 1025 + 20, - 80 psig.*
j
The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor stcam pressure is adequate to perform the test.
\\
FERM1 - UNIT 2 3/4 7-14 Amendment No. 87
TABLE 3.8.4.3-1 (Continued) tiQTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION SYSTEM (S)
-VALVE NUMBER AFFECT [Q E41-F022 HPCI E41-F041 HPCI E41-F042 HPCI E41-F059 HPCI E41-F075 HPCI E41-F079 HPCI E41 F600 HPCI
7. E51-F001 Reactor Core Isolation Cooling System (RCIC)
E51-F002 RCIC E51-F007 RCIC E51-F008 RCIC E51-F010 RCIC 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 E51-F095 RCIC
.l l
8. Gll54 F018 Drywell Floor Drain System G1154-F600 Drywell Floor Drain System
9. G33-F001 Reactor Water Clean-Up System.(RWCV) 4 G33-F004 RWCU
30. G51-F600 Torus Water Management System (TWMS)
G51-F601 TW:15 G51-F602 TWHS G51-F603 TWMS G51-F604 TWM5 G51-F605 TWMS G51-F606 TWMS G51-F607 TWMS
31. N11-F607 Main Steam System Nil-F608 Main Steam System N11-F609 Main Steam System Nll-F610 Main Steam System FERMI - UNIT 2 3/4 8-23 Amendment No.8,87
l PEACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANfBY L10VID CONTROL SUl@
1he design objective of the Standby Liquid Control (SLC) System is two fold.
One objective is to provide backup capability for bringing the reactor from full power to 1 cold, Xenon-free shutdown, assuming that the withdrawn control rods retiain fixed in the rated power pattern. The second objective of the SLC System l
!s to meet the requirement of the ATWS Rule, specifically 10 CFR 50.62 paragraph (c)(4) which states that, in part:
"Each boiling watei reattor must have standby liquid control system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution."
The SLC System uses enriched Boron-10 (contained in the Sodium pentaborate solution) to comply with 10 CFR 50.62 paragraph (c)(4). The met _ hods used to determine compliance with the ATWS Rule are 'in accordance with Reference 2.
To meet both objectives, it is necessary to inject a minimum quantity of 2560 l
net gallons of 65 atom percent Boron-10 enriched sodium pentaborate in a solution having a concentration of no less than 9.0 weight percent (see Figure 3.1.5-1 for equivalent volumes and concentration ranges). The equivalent concentration of natural boron required to shutdown the reactor is 720 parts per million (ppm) in the 70*F moderator, including the Recirculation loops and with the RHR Shutdown Cooling Subsystems in operation.
In addition to this, a 25 percent margin is provided to allow for leakage and imperfect mixing (900 ppm).
The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible sodium pentaborate solution volume range, which adequately compensates for the positive reactivity effects due to moderator temperature reduction and xenon decay during shutdown.
The temperature requirement is necessary to ensure that the sodium pentaborate remains in solution.
With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
The SLC tank heaters are only required when mixing sodium pentaborate and/or water to establish the required solution operating parameters during additions to the SLC tank. Normal operation of the SLCS does not depend on these tank heaters to maintain the solution above its saturation temperature.
Technical requirements have been placed on the tank heater circuit breakers to ensure that their failure will not degrade other SLC components (see Specification 3/4.8.4.5).
Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.
Analysis of Boron-10 enrichment each 18 months provides sufficient assurance that the minimum enrichment of Boron-10 will be maintained.
FERMI - UNIT 2 B 3/4 1-4 Amendment No. #, A2, 87
-3/4.2' POWER DISTR'180 TION LIMITS-BASES J
3/4.2.1 AVERAGE PLANAR ' LINEAR HEAT GENERATION RATE (Continuedl l.
Power-and flow dependent adjustments are provided in'the COLR-to assure that.the i
fuel thermal' mechanical-design criteria are preserved during abnormei transients initiated from off-rated conditions, l
-FERP.I - UNIT 2
.B 3/4 2-la Amendment No. ff U,#,6 67.
i POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO (Continued)
Details on how evaluations are performed, on the methods used, and how the MCPR limit is adjusted for eparation at less than rated power and flow conditions are given in References 1 ans 3 and the CORE OPERATING LlHITS REPORT.
At THERMAL POWER levels less than or equal to 25 percent of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.
During initial startup testing of the plant, a MCPR evaluation will be made at 25 percent of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.
The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25 percent of RATED THERNAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement fer calculating MCPR when a limiting controi rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.
3.4.2.4 LINEAR HEAT GENERATION RATE The thermal expansion rate of U0, pellets and Zircalloy cladding are different in that, during heatup, the fuel pellet could come into contact with the cladding and create stress, if the stress exceeds the yield stress of the cladding material, the cladding will crack. The LHGR limit assures that at any exposure,1% plastic strain on the clad is not exceeded.
This limit is a function of fuel type and is presented in the CORE OPERATING LIMITS REPORT.
References:
1.
General Electric Standard Application for Reactor Fuel", NEDE-240ll-P A (the approved version at the time the reload analyses are performed shall be identified in the COLR).
2.
"The GESTR-LOCA and SAFER Models int the Evaluation of the Loss-of-I Coolant Accident - SAFER /GESTR Application Methodology" NEDE 23785-1-PA (the approved version at the time the reload analyses are performed shall l
be identified in the COLR).
i 3.
" Fermi 2 Maximum Extended Operating Domain Analysis", NEDC-3184E July 1990.
I I
FiRMI - UNIT 2 B 3/4 2-4 Amendment No JS,fg,ff,Ef,ES,p
}/4 4 REACTOR COOLANT SYSTEM a
[ASFS I
3/4.4.1 RECIRCULATION SYS.IQj The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operatfen is permitted at power level is up i
to 67.2% of RATED THERML POWER if the MCPR fuel cladding safety limit is j
increased as noted by Specification 2.1.2.
APRM scram and control rod block setpoints (or APRM gains) are adjusted as noted in Tables 2.2.1 1 and 3.3.6 2, respectively, A time period of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is allowed to make these adjustments l
l following the establishment of single loop operation since the need for single loop operation often cannot ce anticipated. MCPR operating limits adjustments in Specification 3.2.3 for different plant operating situations are applicable to both single and two recirculation loop operation.
To prevent potential control system oscillations from occurring in the recirculation flow control system, the operating mode of the recirculation flow control system must be restricted to the manual control mode for single-loop operation.
Additionally, surveillance on the pump speed of operating recirculation loop is' imposed to exclude the pessibility of excessive. core internals vibration.
The surveillance on differential temperatures below 30% THERMAL POWER or 50% rated recirculation loop flow is to prevent undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during a power or flow increase following extended operation in the single recirculation loop mode.
An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding.the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.
Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.
Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation.
The limits will ensure an adequate core flow coastdown from e..her recirculation loop following a LOCA.
In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.
1 In order to prevent undue stress on the vessel nozzles and bottom head-region, the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop. The loop temperature must also be within 50'F of the reactor pressure vessel coolant temperature to prevent thermal shock to tne recirculation pump and recirculation nozzles.
FERMI - UNIT 2 B 3/4 4-1 Amendnient No. JJ,87
L/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)
Sudden equalization of a temperature difference greater than 145'F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause unO stress in the reactor vessel bottom head.
Requirements are imposed to prohibit idle loop startup above the 777. -od j
line to minimize the potential for initiating core the.nal-hydraulic instability.
3/a.4.2 SAFETY / RELIEF VALVES 9
The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 11 OPERABLE safety / relief valves is required to limit reactor pressure to within ASME Ill allowable values for the worst case upset transient.
Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.
The low-low set system ensures that a potentially high thrust load (designated as lead case C.3.3) on the SRV discharge lines is eliminated during subsequent actuations. This is achieved by automatically lowering the closing setpoint of two valves and lowering the opening setpoint of two valves following the initial opening.
Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced setpoint does not violate the design basis.
FERMI - UNIT 2 8 3/4 4-la Amendment No. EJ B
REACTOR COOLANT SYSTEM BASIS 3/4.4.10 CORE THERMAL HYDRAUllt STABILITY BWR cores typically operate with the presence of global flux noise in a stable mode which is due to random boiling and flow noise. As the power / flow conditions are changed, along with other syste'n parameters (pressure, subcooling, pow distribution, etc.) the thermal hydraulic / reactor kinetic feedback mechanism can be enhanced such that random perturbations may result in sustained limit cycle or divergent oscillations in power and flow.
Two major modes of oscillations have been observed in BWRs. The first mode is
{.
the fundamental or core-wide oscillation mode in which the entire core
}
oscillates in phase in a given axial plane. The second mode involves regional oscillation ir which one half of the core oscillates 180 degrees out of phase with the other half.
Studies have indicated that adequate margin to the Safety Limit Minimum Critical Power Ratio (SLMCPR) may not exist during regional oscillations.
Region A and B of Figure 3.4.10-1 represent the least stable conditions of the plant (high power / low flow).
Region A and B are usually entered as the result of a plant transient (for example, retirculation pump trips) and therefore are generally not considered part of the normal operating domain.
Since all stability events (including tr.
aperience) have occurred in either Region A or B, these regions are avoides a minimize the possibility of encountering oscillations and potentially challenging the SLMCPR. Therefore, intentional operation in Regions A or B is not allowed.
It is recognized that during certain abnormal conditions within the plant, it may becone necessary to enter Region A or B for the purpose of protecting equipment which, were it to fail, could impact plant safety or for the purpose of protecting a safety or fuel e
operating limit.
In these cases, the appropriate actions for the region entered would be performed as required.
Most oscillations that have occurred during testing and operation have occurred at or above the 96% rod line with core flow near natural circulation.
l This behavior is consistent with analysis which predict reduced stability margin with increasing power or decreasing flow.
As core flow is increased or power decreased, the probability of oscillations occurring will decrease.
Region A of Figure 3.4.10-1 bounds the majority of the stability events and tests observed in GE BWRs.
Since Region A represents the least stable region of the power / flow operating domain, the potential to rapidly encounter large 3
magnitude core thermal hydraulic oscillations is increased. During transients, the operator may not have sufficient time to manually insert control rods to mitigate the oscillations before they reach an unacceptable magnitude.
Therefore, the prompt action of manually scramming the plant when Region A is entered is required to ensure protection of the SLMCPR.
FERMI - LINIT 2 B 3/4 4-8 Amendment No. U,87
4 9
3/4.6 -CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINYtNT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated let: rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE B0UNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
PRIMARY CONTAINMENT INTEGRITY is demonstrated by leak rate testi.ng and by verifying that all primary containment penetrations not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during acciuent conditions are closed by locked valves, blank flanges or deactivated automatic valves secured in the closed position.
For test, vent and drain connections which are part of the containment boundary, a threaded pipe cap with acceptable sealant in addition to the containment isolation valve (s) provides protection equivalent to a blank flange.
3/4.6,1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primaiy 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.
Updated analysis a
demonstrates maximum expected pressure is less than 56.5 psig. As an added l
conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the :pecial requirement for testing these valves.
The surveillance testing for measuring leakage rates is consistent with 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-Rate Testing of Containment Structures for Nuclear Reactors." ~ N45.4-1972 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 calculaticr.
The total time and point-to-point techniques may give confidence intervals, which are large frcctions of the calculated leak rate, and the intervals may increase as mor,e data sets are added.
FERMI - UNIT 2 B 3/4 6-1 Amendment No. E, A9,s7
=
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.
3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR Pat; 100 guidelines, provided the main steam line system from the isolation valves up to and including the-turbine condenser remains 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 for the leakage control system vill reduce the untreated leakage from the MSIVs when isolation of the primary system and containment is required.
3/4,6.1,5' PR1HARY 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 demonstrate this capability.
3/4.6,1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the containment peak pressure of less than 56.5 psig does not exceed the maximum l
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.E,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 temperatura of 340*F during LOCA conditions and is consistent with the safety analysis.
3/4.6.1,8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The drywell and suppression chamber purge supply and exhaust isolation valves are maintained closed during a majority of the plant operating time.
Maintaining these valves closed (even though they have been qualified to close against the buildup of pressure-in primary containment in the event of DBA/LOCA) reduces the potential for release of excessive quantities of radioactive material.
FERMI - UNIT 2 B 3/4 6-2 Amendment No. M,0.87
\\
l CONTAINMENT SYSTEMS BASES DRYWELL AND SVPPRESSION CHAMBER PURGE SYSTEM (Continued)
Purging or venting through the Standby Gas Treatment System (SGTS) imposes a vtlnerability factor on the integrity of the SGTS.
Should a LOCA occur while the purge pathway is through the SGTS the associated pressure surge, before the purge valves close, may adversely affect the integrity of the SGTS charcoal filters. Therefore, PURGING or VENTING through the SGTS is limited to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 days. This time limit is not imposed when venting through the SCTS with the 1-inch valves or when PURGING or VENTING through the Reactor Building Ventilation System with any of the purge valves.
Leakage integrity tests with a maximum allowable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient caterial seal degradation and will allow the opportunity for repair before gross leakage failure develops. The 0.60 La leakage limit shall not be ex.eeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.
The 6,10, 20, and 24 inch purge valves are generally configured in a three (3) valve arrangement at each of the associated purge penetrations.
The valves are leak tested by pressurizing between the three valves and a total leakage is determined as opposed to a single valve leakage.
Verifying that the measured leakage rate is less than 0.5 L for this multi-valve arrangement a
is more conservative than a limit of 0.5 La for a single valve.
3/4.6.2 DEPRESSUR17AT10N SYSTEMS The specifications M this section ensure that the primary containment pressure will not exceea t' a maximum allowable pressure of 62 psig during primary system blowdown f rom full operating pressure.
The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.
The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1045 psig.
Since all of the gases in the drywell are purged into the j
suppression chamber air space during a loss-of-coolant accident, the pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure.
T5e design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
TERMI - UNIT 2 B 3/4 6 3 Amendment No. "E,87 l
l
m 2
CONTAINMENT SYSTEMS BASES DEPRESSUR17AT10N SYSTEMS (Continued)
Using the minimum or maximt' water volumes given in this specification, containment pressure during the Design basis accident is less than 56.5 psig j
which is below the maximum allowable pressure of 62 psig. Maximum water volume of 124,220 ft* results in a downcomer submergence of 3'4" and the minimum volume of 121,080 ft8 results in a submergance of 3'0".
The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170'F.
Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.
Under full power operation conditions, a design basis accident blowdown from an initial suppression chamber water temperature of 95'F results in a water temperature of approximately 135'F in the short term following the blowdown.
At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase.
If both RHR loops are used for containment cooling, there is no dependence on containment overpressure for post-LOCA operations.
The large thermal capacitance of the suppression pool is also utilized during piant transients requiring safety / relief valve (SRV) actuation.
Steam is discharged from the main steam lines through the SRVs and their accompanying discharge lines into the suppression pool where it is condensed, resulting in an increase in the temperature of the suppression pcol water.
Although stable steam condensation is expected at all pool temperatures.
NUREG 0783 imposes a local temperature limit shown in Figure B 3/4.0.2 4 in the vicinity of the T-type quencher discharge device.
The limiting plant transients with respect to heat input to the suppression pool have been analyzed. The conservative analysis showed that limiting the average water temperature to less than or equal to 170'F will result in local pool temperatures below the condensation stability limit of Bases Figure B 3/4.6.2-1.
Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below 200'F during any period of relief valve operation.
Specifications have bee placed on +he envelope of reactor operating conditions so that the reacter can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings, Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally change very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.
By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual FERM1 - UNIT 2 B 3/4 6-4 Amendment No.87
- ~ _.
PLANT SYSTEMS BASES 3/4.7.9 MAIN TURBINE BYPASS SYSTEM AND MOISTURE SEPARATOR REHEATER The main turbine bypass system is an active bypass system designed to open the bypass valves in the event of a turbine trip to decrease the severity of the pressure transient.
Each valve is sized to pass approximately 12% percent j
reactor steam flow in the full-open position for a controlled total bypass of approximately 25 percent reactor steam flow. The main turbine bypass system l
is required to be OPERABLE consistent with the assumptions of the Feedwater Controller failure analysis.
The primary purpose of the moisture separator reheater is to improve cycle efficiency by using primary system steam to heat the high pressure turbine exhaust before it enters the low-pressure turbines.
In doing so, it also provides a passive steam bypass flow of about 10 percent that mitigates the early effects of over-pressure transients. The moisture separator reheater is required to be OPERABLE consistent with the assumptions of the Main Turbine Trip with Turbine Bypass failure analysis and the Feedwater Controller failure analysis.
The operation with one or both of the main turbine bypasses inoperable or the moisture separator reheater inoperable to perform preventive or corrective maintenance above 25 percent RATED THERMAL POWER, requires, after one hour, the evaluation of the MCPR in accordance with Specification 3.2.3.
If the MCPR is within the bounds established by Specification 3.2.3, power increases to or operation above 25 percent RATED THERMAL POWER is allowed.
3/4.7.11 APPENDlX R ALTERNATIVE SHUTDOWN AUXILIARY SYSTEMS The systems identified in this section are those utilized for Appendix R Alternative shutdown but not included in other sections of the Technical Specifications. The ACTION statements assure that the auxiliary systems will be OPERABLE or that acceptable alternative means are established to achieve the same objective.
There-are_four independent Combustion Turbine Generator units onsite.
JTG 11 Unit I has a diesel engine starter and thus can be started independently from offsite power.
CTG 11 Units 2, 3, and 4 have AC-motor starters and rely on a 480 volt AC feed.
The phrase
alternative source of power", as used in Specification 3.7.11, ACTION b.2, is defined as a tource of power that is not reliant on offsite power for starting (if required) or operating (if already running) and capable of supplying the required loads on the 4160-volt busses associated with the Alternative Shutdown System.
i One of the two installed Standby Feedwater Pumps and one of the two listed Drywell Cooling Units are necessary for Appencix.R ' Alternative shutdown.
l Therefore unlimited operation with one of the two components inoperable is justified provided increased surveillance is performed on the components which remain OPERABLE.
L FERMI - UNIT 2 B
'5 Amendment No. 79, 59,57 L
i
ADMINISTRATIVE CONTROL 5 SPECIAL REPORTS 6.9.2 Special reports shi.11 be submitted to the Regional Administrator of the Regional Office of the NRC within the time period spsicified for each report, f&R1 0PERATING LIMITS REPORT 6.9.3 Selected cycle specific core operating limits shall be established and documented in the CORE OPERATING L1 HITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in General Electric Company re) orts NEDE 240ll P A and NEDE 23785 1-PA.
The core operating limits shall se determined so that all j
applicable limits (e.g., fuel thermal rnechanical limits, core thermal-hydraulic limits, ECC$ limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The LOLR, including any mid cycle revisions or supplement thereto, shall be submitted upon issuance to the NRC Document Control Desk, with copies to the Regional Administrator and Resident inspector prior to use.
6.10 REC 000 RETENT12!j 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimt.m period indicated.
6.10.2 The following records shall be retained for at least 5 years:
a.
Records and logs of unit operatW covering time interval at each power level.
b.
Recordt and logs of principal Aintestee a".ivitier
'pections, repair. and raplacement of principal items cf aquipms related to nuclear safety.
c.
ALL REPORTABLE EVENTS, d.
Records of surveillance activities, inspections, and calibrations required by these lechnical Spt ;ifications.
e, Records of changes made to the procedures required by Specification 6.8.1.
f.
Records of radioactive shipments.
g.
Reprds of sealed tource and fission detector lea tests and resdts.
h.
Records of annual physical inventory of all sealed sour.: material of record, fERM1 - UNIT 2 6-21 Amendment No. JJ, ff,57
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ML20105D276

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References:

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