ML20058L842

Amend 53 to License NPF-43,revising Tech Specs to Allow Extended Operation of Plant at Reduced Power W/Single Recirculation Loop in Operation

ML20058L842
Person / Time
Site:	Fermi
Issue date:	07/27/1990
From:	Pierson R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20058L846	List: ML20058L842 ML20058L847
References
NUDOCS 9008080223
Download: ML20058L842 (40)

Text

_

?

ff o.,

UNITED STATES

{

3 g

NUCLEAR REGULATORY COMMISSION E

WASHINGTON, D. C. 20555 0,

%,..... /

DETROIT EDISON COMPANY DOCKET NO. 50-341 FERMI-2 AMENDMENT TO FACILITY OPERATlHG LICENSE Amendment No. 53 License No. NPF-43 1.

The Nuclear Regulatory Comission (the Comission) has found that:

I A.

The application for amendment by the Detroit Edison Company (the licensee)datedAugust4,1988assupplementedAugust 18, 1989, complies with the standards and requirements of the Atomic Energy i

Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

Thereisreasonableassurance(1)thattheactivitiesauthorizedby 5

l this aner.dn.ent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted L

in compliance with the Comission's regulations; 1

D.

The issuance of this amendment will not be inimical to the comon defense ar.d security or to the health and safety of the public; and i

E.

The issuance of this emendaent is in accordance with 10 CFR Part 51 of I

the Cottission's regulations and all applicable requirer (nts have been satisfieo.

t 2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendtent and paragraph 2.C.(2) of 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. 54, andtheEnvironmentalProtectionPiancontainedin Appendix B, are hereby incorporated in the license.

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

9008080223 900727 PDR ADOCK 05000341 P

PDC

1 2-3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 125( C-Robert C. Pierson, Acting Director Project Directorate 111-1 Division of Reactor Projects - III,

!Y, V & Special Projects Office of Nuclear Reactor Regulation Attachnent:

Changes to the Technical l

Specifications Date of Issuat cc: July 27, 1990 l

l i

ATTACWENT T0,1.ltgSE, AMgDMENT,N0.53 T

2 FACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341 Replace the following pases of the Appendix "A" Technical Specifications with the attached pages. Tie revised pages are ider.tified by Arnendrnent nurnber and j

contain a vertical line iridicating the area of change.

REMOVE INSERT v*

v*

1 vi vi

)

xiii xiii xxi xx1 2-1 2-1 2-3 2-3 1

2-4 2-4 2-Aa B 2-7 B 2-7 B 2-7a B 2-8*

B 2-8*

3/4 2-1 3/4 2-1 3/4 2-5 3/4 2-5 3/4 2-Sa 3/4 3-41 3/4 3-41 3/4 3-42*

3/4 3-42*

3/4 3-44 3/4 3-44 3/4 3-44a

)

3/4 4-1 3/4 4-1 3/4 4-2 3/4 4-2 3/4 4-3 3/4 4-3 3/4 4-4 3/4 4-4 3/4 4-5 3/4 4-5 3/4 4-6 3/4 4-6 3/4 4-6a 3/4 4-30 3/4 4-31 2/4 10-3*

3/4 10-3*

3/4 10-4 3/4 10-4 B 3/4 2-1 B 3/4 2-1 B 3/4 2-la B 3/4 2-3 8 3/4 2-3 B 3/4 4-1 B 3/4 4-1 B 3/4 4-la B 3/4 4-2*

B 3/4 4-2*

B 3/4 4-8 B 3/4 4-9 5M'e'rleat page pr ovided to trairitain docuo ent corapleter4 ss.

!!c cher.ges contair.ed in these pages.

INDEX LIMITil4GCONDIT10h5FOROPERATIONANDSURVEILLAhCEREpVIREMENTS

.............................._...............nur"-""-

3/4.3 INSTRUMEji,T A,T,1,0) 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...........

3/4 3-1 3/4.3.2 ISOLATION AC10AT 1011 I NSTRUMENTATION.................

3 /4 3-9 3/4.3.3 EMERGEllCY CORE C00LikG SYSTEM ACTUA110N INSTRUMENTATION............................-..........

3/4 3-23 3/4.3.4 ATWS RECIRCULATION PUMP TRIP SYSTEM ACTUATION lilSTRUMENTATION.....................................

3/4 3-32 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION lit S1 RUMENT ATION...................................... 3 /4 3 -3 6 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTAT10h...................

3/4 3-41 3/4.3.7 N0lilTORING INSTRUMENTATION Radiation Fonitoring Inst rurnentation.................

3/4 3-47 Sei sn.ic Moni t oring I nstrunient a t ion...................

3/4 3-51 l'eteorological Monitoring Instrunientation............

3/4 3-54 1

Rentote Shutdown Systen Instrunentation and Controls..

3/4 3-57 Accident Vonitoring Instrurientation..................

3/4 3-60 Source Range Monitors...............................

3/4 3-64 Traver sing I n-Core Probe Systerr......................

3/4 3-65 C h l o r i n e De t e c t i o n Sy s t ern............................

3/4 3-66 Fire Detection Instronientat ion.......................

3/4 3-67 Lcese-Pa rt Det ec tion Systern..........................

3/4 3-70 Radioactive Liquid Effluent Monitoring I nst runient a t i on.....................................

3 /4 3 71 Radicactive Gasecus Effluent l'cnitoring I n s t runie n t a t i o n.....................................

3 /4 3 -7 6 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM...................

3/4 3-85 3/4.3.9 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTAT10h......................................

3/4 3-06 FERl'1 - UNIT 2 v

An endirent No. 53

INDEX LIMIT 1kG CCADITIONS FOR OPERATION AhD SURVEILLANCE,_REQUIREME3TS

= -.

SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation loops..................................

3/4 4-1 Jet Pumps............................................

3/4 4-4 Recirculation Pumps..................................

3/4 4-5 Idle Recirculation Loop Startup......................

3/4 4-6 3/4.4.2 SAFETY / RELIEF VALVES Sefety/ Relief Va1vis.................................

3/4 4-7 Safety /Felief Valves Lcw. Low Set Tunction............

3/4 4-8 3/4 4.3 FEACTOR C00LAtT SYSTEh LEAKAGE Le a kage Dete ction Sy sters............................

3/4 4-9 O Pe r6 t i ona l L ea ka ge.................................. 3/4 4-10 3/4.4.4 CHEMISTRY.......................s...,................

3/4 4-13 5/4.4.5 SPECIFIC ACTIVITY....................................

3/4 4-16 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Feactor Coolant System...............................

3/4 4-19 l

Reector Steam 00me...................................

3/4 4-23 l

3/4.4.7 NAIN STEAM LINE 150LAT10l! YALVES.....................

3/4 4-24 3/4.4.8 STRUCTURAL INTEGRITY.................................

3/4 4-25 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown.........................................

3/4 4-26 Cold Shutdown........................................

3/4 4-28 3/4.4.10 CORE THERIML HYDRAULIC STABILITY.....................

3/4 4-30 3./.4..5 EME,RG,Eli,C,Y,,C,0RE, COOL 1NG SYSTEMS l

3/4.5.1 FCCS - OPERATING.....................................

3/45-1 3/4.5.2 ECCS - SHUTD0WN......................................

3/4 5-6 l

3/4.5.3 SUPPRESSION CHAMEER..................................

3/4 5-8 FERMI - UNIT 2 vi An.endment No. 53

INDEX BASES g 9 _........................................................ g...

INSTRUMENTA_TIO,h(Continued)

MONITORlhG INSTRUMENTATION (Continued)

Meteorologic61 Mor itoring Instrurnentation.......

B 3/4 3-4 Remote Shutdown Systera Instrurr.entation and Controls........................................

B 3/4 3-4 Accider.t Monitoring Instrumenta tion.............

B 3/4 3-4 Source Range Konitors...........................

B 3/4 3-4 Traversing In-Core Probe System.................

B 3/4 3-4 Chlorine Detection Sy stera.......................

B 3/4 3-5 Fi re Detection Instruinentation................... B 3/4 3-5 Loose-Part Detection Sy stem.....................

B 3/4 3-5 l

Fedioactive Lic;uid Effluent Monitoring Instrumentation.................................

B 3/4 3-5 Fedicettive Gastous Effluent Monitoring I n s t r uine n t a t i o n.................................

B 3/4 3-6 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM.............

B 3/4 3-6 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEMS ACTUATION I ftST RUMENTATION.................................

B 3/4 3-6 l

1/4.4 REACTOR COOLANT SYSTEM 1

l 3/4.4.1 RECIRCULATION SYSTEM............................

B 3/4 4-1 l

5/4.4.2 SAFETY / RELIEF VALVES............................

B 3/4 4-la l L

3/4.4.3 REACTOR COOLANT SYSTEM LE.'KAGE Leakage Detection Systems.......................

D 3/4 4-2 l

l Ope. ration 61 Leakage.............................

B 3/4 4-2 1

3/4.4.4 CHEMISTRY.......................................

B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY...............................

B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.....................

B 3/4 4-4 1

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES................

B 3/4 4-5 1

3/4.4.8 STRUCTURAL INTEGRITY............................

B 3/4 4 5 3/4.4.9 RESIDUAL HEAT RLM0 VAL...........................

B 3/4 4-5 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY................

B 3/4 4-8 l FERMI - UNIT 2 xiii An.endment No. (9, 53

INDEX LIST OF FIGURES

+

FIGURE PAGr 3.1.5-1 SODIUM PENTABORATE VOLUME / CONCENTRATION REQUIREMENTS...................................

3/4 1-21 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAFLHGR) VS. AVERAGE PLANAR EXPOSURE INITIAL CORE FUEL TYPE 8CR183..................

3/4 2-?

3.2.1-2 MAXIMUM AVERAGE PLAMAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR233..................

3/4 2-3 3.2.1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPuiGR) VS. AVERAGE PLANAR EXPOSURE, i

RELOAD FUEL TYPE BC3180........................

3/4 2-4 3.2.1-4 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, i

t RELOAD FUEL TYPE BC318E).........................

3/4 2-4A 3.2.3-1 BOC TO 12,700 MWD /ST, MINIMUM CRITICAL POWER j

RATIO (MCPR) VS. TAU AT RATED FLOW............

3/4 2-8 l

l 3.2.3-1A 12,700 MWD /ST TO 13,700 MWD /ST, MINIMUM CRITICAL POWER RATIO (MCPR) VS. TAU AT RATED FLOW......

3/4 2-8A l

3.2.3-1B 13,700 MWD /ST TO EOC, MINIMUM CRITICAL POWER RATIO (MCFR) VS. TAU AT RATED FLOW............

3/4 2-8B 3.2.3-2 FLOW CORRECTION (K ) FACTOR.....................

3/4 2-9 f

3.4.1.4-1 THERMAL POWER VS CORE FLOW.....................

3/4 4-6a l

3.4.6.1-1.

MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE....................

3/4 4-21 l

3.4.10-1 THERMAL POWER VS. CORE FL0W.....................

3/4 4-31 l

4.7.5-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST.....

3/4 7-21 B 3/4 3-1 REACTOR VESSEL WATER LEVEL...................... B 3/4 37 B 3/4.4.6-1 FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A l

FUNCTION OF SERVICE LIFE........................ B 3/4 4 '/

8 3/4.6.2-1 LOCAL POOL TEMPERATURE LIMIT.................... B 3/4 6-5 B 3/4.7.3-1 ARRANGEMENT OF SHORE BARRIER SURVEY POINTS...... B 3/4 7-6 5.1.1-1 EXCLUSION AREA.................................

5-2 5.1.2-1 LOW POPULATION ZONE............................

5-3 5.1.3-1 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS......................................

5-4 FERMI - UNIT 2 xxi Amendment No. 30, 42, 53 l

y v

2.0 SAFETY LIMITS AND LIMITING SAFE 1' SYSTEM SETTINGS 2.1~ SAFETY LIMITS-THERMAL P0gj Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

])

ALTION:

With THERMAL POWER exceedin 25% of RATED THERMAL POWER and the reactor vessel steam dome pre m re 3ss than 785 psig or core flow less than 10%

of rated flow, be in 9t least HOT SPJTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirernents of Specificatio s.7.1.

?

=

THERMAL POWER,,H,ig,h, Pre,ssure,an,d,Hjph F, low

?.1.2 TheMINIMLMCRITICALPOWERRATIO(MCPR)shallnotbelessthanthe

- Safety Limit MCPR of 1.07 for two recirculation loop-operation er.d shall not be less than the Safety Limit MCPR of 1.08 for single loop operation with.the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABIL,1T,Y : -OPERATIONAL CONDITIONS 1 and 2.

A_CTION:

With MCPR less than the Safety Limit-KCPR of 1.07 for two recirculation loop operaticn or less than the Safety Limit MCPR of 1.08 for single loop operation and with the reactor vessel steam dome pressure greater than 785 psig end core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and cwply with the requirements ;;f Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

B ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTD0KH with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with

=

the requirements of Specification 6.7.1.

FERMI - UNIT 2 2-1 Amendment No. H, 53

I SAFETY LIMITS AND LIMITINC SAFETY SYSTEM SETTINGS t

2.2 LINITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SE1 POINTS f

2.2.1 The reactor protection system instrun.entation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

APPLICABILITY: As shown in Table 3.3.1 1.

ACTION:

'With a reactor protection system instrur.tentation setpoint* less conservative

[

.than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable ar.d apply the applicable ACTION staten:ent requirement of Specification 3.3.1 ontil the chatr.el is restored to OPERABLE status with its setpoint adjusted consister,t with the Trip Setpoint value.

7 1

L

• The APhifTlbhFblased instrur.ieritation need not be declared inoperable upon enterir.g sir;gle recirculation loop operation provided the setpoints are adjusted

't within 4 hcurs per Specification 3.4.1.1.

1 5

i t

FERMI - UNIT 2 2-3 Amendment No. 53

.. ~...

s TABLE 2.2.1-1 j

E REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS

~

-4 ALLOWABLE.

rv FUNCTIONAL UNIT TRIP SETPOINT VALUES 1.

Intermediate Range Monitor, Neutron Flux-High 1 120/125 divisions of i 122/125 divisions' full scale of full scale 2.

Average Power Range Monitor:

a.

Neutron Flux-Upscale, Setdown i 15% of RATED THERMAL POWER 1 20% of RATED-THERMAL POWER b.

Flow Biased Simulated Thermal Power-Upscale

1) During two recirculation loop operation:

sui a.

Flow Biased' 1 0.58 W+59%, with 1 0.58 W+62%, with a maxime.2 of a maximum of b.

High Flow Clamped i 113.5% of RATED 1 115.5% of RATED THERMAL POWER THERMAL POWER'

2) During single recirculation loop operation:

a.

Flow-Biased' 10.58W+54.4%,**

10.58W+57.4%,**

b.

High Flow Clamped NA NA c.

Fixed Neutron Flux-Upecaie i 118% of RATED 1 120% of RATED THERMAL POWER THERMAL POWER 2,

d.

Inoperative N.A.

N.A.

3.

Reactor Vessel Steen Dome Pressure 'High 5 1068 psig i 1088 psig A

4.

Reactor Vessel Low Water-Level - Level 3 1 173.4 inches

• 1 171.9 inches 5

• See Bases Figure B 3/4 3-1.

m

• • During single recirculation loop operation, rather than adjusting the APRM Flow Biased Setpoints to comply with the single loop values, the gain of the APRMs may be adjusted for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such that the final APRM readings are at least 5.3% of rated power graater than 100% times FRTP, provided that the adjusted APRM readings do not exceed 100% of' RATED THERMAL. POWER and a notice of adjustment is posted on the reactor control panel.

m..

m x.

1

.m TABLE' 2.2.1-1

ox REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued);

c ALLOWABLE FUNCTIONAL UNI!

TRIP SETPOINT

' VALUES

-4 5.

Main Steam Line Isolation Valve - Closure 1 8% closed i 12% closed y

6.

Main Steam Line Radiation - High 5 3.0 x full power background 5 3.6 x full power backts aund

~7.

Drywell Pressure - High 1 1.68 psig

.i 1.88 psig 8.

Scram Discharge Volume Water Level'- High a.

Float Switch

< 594'8"

< 596'0" b.'

Level Tiansmitter

< 592'6"

< 596'0" 9.

Turbine Stop Valve - Closure 1 5% closed 5 7% closed 4

to 10.

Turbine Control Valve Fast Closure Initiation of fast closure M.A.

.p 11.

Reactor Mode Switch Shutdown Position N.A.

N.A.

12.

Manual Scram M.A.

N.A.

13.

Deleted i

I l

l I

9 l

@a CL Bm 3r+

Z O

f

.m

.y...

,w w

" i *

'N

~

^N'

m m

t LIlllTIh6 SAFETY SYSTEM SETTINGS BASES REACTORPROTECTIONSYSTEMINSTRUMENTATIONSETPOINTS(Continued)

A,v,e,r,agePowerRang,eMonitor(Continued)

Because the flux distribution associated with uniform rod withdrawals does not I

involve high local peaks and because several rods cust be moved to change power by a significant amount, the' rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate.

In an assumed uniform l

rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more theti adequate to assure shutdown before the pcwer could exceed the Safety Limit.

The 15% neutron flux trip remains active until the mode switch is placed in i

the Run position.

l l

The APRM trip system is calibrated using heat balance data taken during steady state conditions.

Fission chambers provide the basic input to the systera and therefore the nonitors respond directly and quickly to changes due to transient operation for the case of the Fixed Neutron-Flux-Upscale setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel.

For the Flow Biased Neutron Flux-High setpoint, e time constant of 6 t I seconds is introduced into the ficw biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximun value is used for the flow biased setpoint as shown in Table 2.2.1-1.

l The APRM setpoints were selected to provide adequate margin for the Safety l

Liraits and yet allow operating n,argin that reduces the possibility of unneces.

sary shutdown. The flow referenced trip setpoint n:ust be adjusted by the specified formula in Specification 3.2.2 in order to naintain these neargins L

when MFLPD is greater thanJor equal to FRTP, For single recirculation loop oper6 tion, the reduced APRM setpoints are based on a tkW value of 8%. The

~

46W value corrects for the difference in indicated drive flow (in percentage of i

drive flow which produces rated core flow) between two loop and single loop operation of the same core flow. The decrease in setpoint is derived by multiplying the slope of the setpoint curve by 8%. The High Flow Clamped Flow Biased Neutron Flux-High setpoint is'not applicable to single loop operation as core power levels which would require this limit are not achievable in a single locp configuration.

3.

Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear systen. process-barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reettivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious l

FERMI - UNIT 2 B 2-7 Amendment No. 53

LIMITING SAFETY SYSTEM SETTINGS

?

BASES REACTOR PROTE_CTION SYSTEM INSTRUMENTATION _ SETPOINTS (Continued) 3.

Reactor Vesse_1 Steam Dome Pressure-High (bontinued) trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared' to the highest pressure tF

'urs in the system during a transient. This trip setpoint is effective ~at Iflow conditions when the turbine stop valve closure trip is bypassed, ibine trip under these conditions, the transient analysis indico sate margin to the thermal hydraulic limit.

l i

II b

l FERMI - UNIT 2 B 2-7a Amendment No. 53 j'

Lu

T LIMITING SAFETY SYSTEM SETTINGS BASES

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

REACTOR _ PROTECTION SYSTEM lHSTRUMENTATION SETPOINTS (Continued) 4 Reactor Vessel Low i. ter Level-Level 3 The reactor vessel water level trip setpoint was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel tc assure that there is adequate protection for the fuel and pressure limits.

E.

Pain Steam Line Isolation Yalve-Closur.

The n.ain steam line isolation valve closure trip was provided to limit the ar.; cur;t of fission product release for certain pcstulated events. The PSIY't at e closed autoratictlly from reasured par 6teters such as high steam flow, high steam line radiation, low reactor water level, high steam tur.nel ter.;perature, and low steam !ine pressure. The MSIV's closure scram antici3ates the ptessure and flux transients which could follow MSIV closure and there)y protects teactor vessel pressure and fuel thermal / hydraulic Safety Limits.

6.

Pain Steam Line Radiatier..High The main steam line radiation detectors are provided to detect a gross feilure of the fuel cladding. When the high radiation is detected, a trip is initiated to reduce the continued failure cf fuel cledding. At the same time the n.ain steam line isolatict.-valves are closed to limit the release of fission products.

The trip setting is'high enough above background radiation levels

.tc prevent spuricus-trips yet low enough to prorptly detect gross f ailures in the futi cladding.

No credit ucs taken for operation.cf this trip in the eccident ar.tlyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability-cf the. F.eactor Protection System.

7.

Dryv.cil Prgs,sure,-H,ich e

High pressure in the drywell could indicate a break in the primary pressure boundary systems or a complete loss of drywell cooling. The reactor is tripped in order to mininize the possibility cf fuel damage ar.d reduce the an.ount of energy being aoded to the coolant. _The trip setting was selected as 1cw as possible without causing spurious trips.

FERMI - UNIT 2 B 2-8

.-um--miim-mm------

~.

~3/4.2 P OW E _R D_I S,T,R I,B U,TJ,0,N,,L I MJ T S

,3.f 4. 2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE J

LIMITING CONDITION FOR OPERA 110N 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall not exceed:

a.

The MAPLHGR limit which has bean approved for the respective fuel andlatticetypeasafunctionoftheaverageplanarexpe%re(as determined by the NRC approved inethodology described in GIMAR-II),

or b.

When hand calculations are required, the inost limiting lattice type HAPLHGR limit as a function of the average planar exposure shown in the Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, and 3.2.1-4 for the applicable bundle type.

The above limits shall be ir.ultiplied by a factor of 0.90 during single loop operation, APPLICABIL),TY: OPERATIONAL CONDIT10N 1, when THERMAL POWER is greater than or equal to I'Si of RATED THERMAL POWER.

ACTION:

l-With an APLHGR exceeding the above limits, initiate corrective action within 1E nir.utes ar.d restore APLHGR te within the required lin4its within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or i

t ieduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 tours.

SURVEILLANCE RE,QUIREMENTS j

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

a.

At 1 cast once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after con:pletion of a THERMAL POWER increase of at l

least 15% of RATED THERMAL POWER, and l

c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is l

operating with a LIMITING CONTROL ROD PAT 1ERN for APLHGR.

)

l d.

The provisions of Specification 4.0.4 are not applicable.

l l

L FERMI - UNIT 2 3/4 2-1 Aniendment No. 42, 53 l

L

.I POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased neutron flux-high scram trip setpoint (S) and flow biased neutron flux-high control rod blod tMp setpoint (SRB) shall be established according to the following relationships:

i TRIP SETPOINT ALLOWABLE VALUE

1. During two recirculation loop operation:

l S 5 (0.58W + 59%)T S 5 (0.58W + 62%)T Sgg 5 (0.58W + 50%)T Sgg 5 (0.58W + 53%)T

2.

During single recirculation loop operation:

V 5 5 (0.58W + 54.4%)T S < (0.58W + 57.4%)T l

S 1 (0.58W + 45.4%)T S

1 (0.58W + 48.4%)T RB RB L

where:

S and S are in percent cf RATED THERMAL POWER, W=LoohBrecirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million 1bs/hr, at 100% of RATED THERMAL POWER T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER l-divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY.

T is i

applied only if less than or equal to 1.0 H

p APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

-equal to 25% of RATED THERMAL POWER.

ACTION:

1 With the APRM flow biased neutron flux-high scram trip setpoint.and/or the flow biased neutron flux-high control rod block trip setpoint less conservative than the value shown'in the Allowable Value column for S or-S as above determined, I

initiatecorrectiveactionwithin15minutesandadjust$B,nd/ ors to be a

consistentwiththeTripSetpointvalue*#within6~hoursorreduceYHERMALPOWER I

R to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• With MFLPD greater than the FRTP during power ascension up to 90% of RATED

' THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal-to 100% times MFLPDj j

provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL

-POWER and a notice of adjustment is posted on the reactor control panel.

With MFLPD greater than FRTP and a single recirculation loop in operation, if the.

APRM flow biased setpoints have not been adjusted to their single loop values, then the minimum -required APRM reading must be increased by an additional 5.3%

)

of rated power.

1. During single recirculation loop operation with FRTP greater than or equal to l

MFLPD, rather than adjusting the APRM setpoints to comply with the single loop values, the APRM gain may be adjusted for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such that tnu final APRM readings are at least 5.3% of rated power greater than 100%

times FRTP, provided that the adjusted APRM readings do not exceed 100% RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

l FERMI - UNIT 2 3/4 2-5 Amendment No. 9, 42, 53 l

l

. POLLER DIST,RJ),U, TION LIMITS e

SURVELL,L,ANC,E, REQUIREMENTS 4.2.2 The FRTP and:the NFLPD for each class of fuel shall be determined, the value of T calculated, and the rnost recent actual APRM flow biased neutron flux-high scram and flow biased neutron flux-high control rod block trip setpoints verified to be witnin the above limits or adjusted, or the APRM 9t.in readings shall be verified as indicated below*f, as required:

a.

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

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL P0llER increase of at

-least 15% of RATED THERMAL POWER, and.

1 c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with MFLPD greater than or equal to FRTP.

d.

- The provisions of Specification 4.0.4 are not applicable.

• With MFLPD greater than the FRTP during pover ascension us to 90% of. RATED.

THERMAL POWER, rather than adjusting the APRM setpoints, tie APRM gain may be adjusted such that APRM readings are greater than or equal to 100% times MFLPD, provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL POWER and a. notice of adjustment is posted on the reactor control panel. With-MFLPD greater than FRTP arid a single recirculation loop in operation, if the l

APRM f1w b' hed setpoints have not been adjusted to their single loop values.

then the timum required APRM reading trust be increased by an additional 5.3%

of rated, ser.

1. During single-recirculation loop operation with FRTP. greater than or equal to-MFLPD, rather than adjusting the APRM setpoints to comply with the single loop values, the APRM. gain may be adjusted for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such that the final APRM readings are at least 5.3r of rated power greater than 100%

times FETP, provided that the adjusted APRM readings do not exceed 100% RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel..

l l

l l

t FERMI - UNIT 2 3/4 2-Sa Amendment flo. 7, 53 m-t

~

~

INSTRUMENTATION' 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.

The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.

L APPLICABILITY: As shown in Table 3.3.6-1.

l-ACTION:'

a.

With a control rod block instrumentation channel trip setpoint* 1ess conservative than the value shown in the Allowable Values column of L

Table 3.3.6-2, declare the channel inoperable until the channel is l

restored to OPERABLE status with its trip setpoint adjusted consistent-L with the Trip Setpoint value.

L L

b.

With the number of OPERABLE channels less than required by the Minimum p

OPERABLE Channels per Trip Function requirement, take the ACTION L

required by Table 3.3.6-1.

L l

l SURVEILLANCE REQUI,REMENTS 4.3.6 Each of the above required control rod block trip systems and E

instrumentation channels shall be demonstrated OPERABLE by the performance of b

the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OFERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.

3.

I '

• The APRM Flow Biased Neutron Flux-High and Rod Block Monitor instrumentation need not be declared inoperable upon entering single reactor recirculation loop operation provided the setpoints are adjusted within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per Specification 3.4.1.1.

l.

li FERMI - UNIT 2 3/4 3-41 Amendment No. 53

].L, -

TABLE 3.3.6 ' CONTROL ROD BLOCK INSTRUMENTATION

m.

g MINIMUM APPLICABLE 3

OPERA 8LE CHANNELS OPERATIONAL Ii TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION

.g 1.

ROD BLOCK MONITOR *)

I q

a.

Upscale 2

1*.

60 b.

Inoperative 2

1*

60 y

c. ~Downscale 2

1*

60 2.

APRM a.

Flow Biased Neutron Flux -

High.

4 1

61 b.

Inoperative 4

1, 2, 5 61 c.

Downscale 4

1 61 d.

Neutron Flux - Upscale, Setdown 4

2, 5 61 3.

SOURCE RANGE MONITORS Detector not full in(b) 3 2

61 a.

II) w 2

5 61 b.

Upscale (c) 3 2

61 I) s 2

5 61 Inoperative (c) l c.

3 2

61 I) 2 5

61 d.

Downscale(d) 3 2

61 II) 2 5

61 4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in 6

2, 5 61 b.

Upscale 6

2, 5

'61~

Inoperatig) 6 2, 5 61 c.

d.

Downscale 5

2, 5 61 5.

SCRAM DISCHARGE VOLUME t

a.

-Water Level-High 2

1, 2, 5**

62 b.

' Scram Trip Bypass 2

2, 5**

62 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale 2

1 62 b.

Inoperative 2

1 62 c.

Comparator 2

1 62 7.

REACTOR M00E SWITCH SHUTDOWN POSITION 2

3, 4_

63

=. -

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRilMENTATION SETPOINTS -

TRIP FUNCTION TRIP SETPOINT

-ALLOWABLE VALUE-g 1.

ROD BLOCK MONITOR

a. ' Upscale

1) During two recirculation

< 0.66 W + 40%

< 0.66 W + 43%

c loop operation.

2 U

2) During single recirculation

< 0.66 W + 34.7%

< 0.66 W + 37.7%

N loop operation b.

Inoperative NA NA c.

Downscale 1 5% of RATED THERMAL POWER

> 3% of RATED THERMAL POWER 2.

APRM a.

Flow Biased Neutron Flux - High

1) During two recirculation 1 0.58 W + 50%*

s'O.58 W + 53%*

loop operation

2) During single recirculation 1 0.58 W + 45.4%#

1 0.58 W + 48.4%#

loop operation b.

Inoperative NA NA c.

Dcwnscale

> 5% of RATED THERMAL POWER

> 3% of RATED THERMAL. POWER R

d.

Neutron Flux - Upscale, Setdown 512%ofRATEDTHERMALPOWER 514%ofRATEDTHERMALPOWER 3.

SOURCE RANGE MONITORS g

a.

Detector not full in NA NA b.

Upscale

. < 1.0 x 105 cps 1 1.6 x 10s cps c.

Inoperative NA NA d.

Downscale 1 3 cps **

1 2 cps **

• The APRM rod block function is, varied as a function of recirculation loop drive flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.

• • May be reduced to 1 0.7 cps provided the signal-to-noise ratio.1 20.

m*..s M

1. During single recirculation loop operation, rather than adjusting the APRM and RBM Flow Biased Setpoints to comply with the single loop values, the gain of the APRMs may be adjusted for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such that the final APRM readings are at least 5.3% of rated power greater than 100% times FRTP, provided that g

the adjusted APRM readings do not exceed 100% of. RATED THERMAL' POWER and a notice of adjustment is posted on the reactor control panel.

p

2..

1

- TABLE 3.3.6-2 l

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS (Continued) l

'{

TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 4.

INTERMEDIATE RANGE MONITORS ru a.

Detector not full in NA NA b.

Upscale 5 108/125 divisions of 5 110/125 divisions of full scale full scale c.

Inoperative NA NA-d.

Downscale

> 5/125 divisions of

> 3/125 divisions of S.

SCRAM DISCHARGE VOLUME-I"II' scale full scale a.

Water; Level-High 1 589'11 "

-1 591'0" b.

Scram Trip Bypass NA NA 6.

REACTOR COOLANT-SYSTEM RECIRCULATION FLOW M

l a.

Upscale

< 108/125% of rated flow

< 111/125% of rated flow Y

b.

Inoperative EA 5A l

,S c.

Comparator 1 10% flow deviation 5 11% flow deviation l

7.

REACTOR MODE SWITCH SHUTDOWN POSITION NA NA Y

8-O.

.~.

. ~.

s.#.

._w..

3 /A. 4 - QAC,T,0,R,C,0,0j., ANT, S,Y,5 TEM 3/4.4.1-RECIRCULATION SYSTEM

-RECIRCULAi!3N LOOPS LIMITING C_0NDli 0N,F03,,0,P,E,RA,T10N _ _,,,,,,,,,,,, _,,,,,,,, _,,,,,,,,,,,_

3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*.

ACTION:

a.

With one reactor coolant system recirculation loop not in operation:

L 1.

Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a) Place the individual recirculation pump flow controller for the operating recirculation pump in the Manual mode.

l-b) Reduce' THERMAL POWER to less than or equal to 70% of RATED THERMAL POWER.

c) Limit the speed of the operating recirculation pump to less than i

or equal to 70% 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.

e) Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value of 0.90 times the two recirculation loop operation limit per Specification-3.2.1.

f) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable for single recirculation loop operationf per l-Specifications 2.2.~1, 3.2.2 and 3.3.6.-

9) Perform Surveillance Requirement 4.4.1.1.4 if THERMAL POWER is 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.

The provisions of Specification 3.0.4 are not applicable.

3.

Otherwise, be in at least HOT SHUTDOWN within the next 12. hours.

b.

With no reactor coolant system recirculation loop in operation while in 0FERAT10hAL CONDITION ~1, irrmediately place the Reactor Mode Switch in the SHUTDOWN position, i

c.

With no reactor coolant system recirculation loops in operation, while in 0FERATIONAL CONDITION 2, initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

• S'e'e"Sp'eWaTText Exception 3.10.4

~

1. APRM gain adjustm'.nts may be made in lieu of adjusting the APRM and RBM Flow Biased Setpoints to comply with the single loop values f';r a period of up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

FERMI - UNIT 2 3/4 4-1 Amendment No. 53 o

't REACTORCOOLANTSLSTjM e

SURVEILLANCL Q UIREl6ENTS' 4.4.1.1.1 Each pump discharge valve shall be demonstrated OPERABLE by cycling eact ve.1ve through at least one complete cycle of full travel during each STARTUP* prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER.

4.a.1.1.2 Each pump MG set scoop tube mechanical aM eiectrical stop shall be derronstrated OPERABLE with overspeed setpoints iess than or equal to 105% and 102.5%, 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:

a.

THERMAL POWER is less than or equal to 70% of RATED 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 Pith 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 locp less than or equal to 50% of rated loop flow, verify the following differential temperature requirements are met within no mcre than 15 minutes prior to either THERMAL POWER increase or recirculation flw increase:

a.

Less than or equal to 145'F between 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 cr eation and the coolant in the reactor presn re vessel *e, and c.

Less than or eg al to 50'F between the reactor coolant within-the loop not in operation and the operating loop.**'

• If not performed within the previous 31 days.

• • Requirement does not apply when the recirculation loop not in operation is isolated from the reactor pressure vessel.

FERMI - UNIT 2 3/4 4-2 Amendment No. 53

4.:

5 8

L:,

i l-THERHAL POWER VERSUS CORE FLOW l

l '>

FIGURE 3.4.1.1 DELETED.

t l

l i

l l

1e

\\

l a

l

'l

\\

I l

i i

FERMI - UNIT 2 3/4 4-3 Amendment No. 53

c., 7 r

u.

REACTOR COOLANT SY_ST_EM_

JET PUMPS

- LIMITillG CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.

A,P,P,LICABillTY: OPERATIONAL CONDITIONS I and 2.

. ACTION:

With one or more jet pumps 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 />, r

SURVEILL,A_N,CE REQUIREMENTS 4.4.1.2 With THERMAL POWER greater than 25% of. RATED THERMAL POWER, each of i

l the above reeuired jet pumps shall be-denonstrated OPERABLE at least once per 1

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

• by deterr,ining cperating rtcirculation loop flow (s), total core flow, and' diffuser-to-lower pler:um differential pressure for each operating jet purr.p and verifying that nu two of the following conditions occur:

i a.

The indicated o>erating recirculation loop flow (s) differs by.more l

.than-10% from tie established pump speed-loop flow characteristics, i

b.~

The indicated total core ficw differs by more than 10% from the established _ total core flow value derived from recirculation loop l

flow measurements.-

L c.

The indicated diffuser-to-lower plenum differential pressure of any individual jet pump in an operating 1 cop differs from the mean cf all I

jet pump differential pressures in the same loop by more than 20%

deviation from its normal deviation.

I l-l W e provisions ct Specification 4.0.4 are not applicable provided that this j

surveillance is perforn.ed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter exceeding 25% of RATED THERMAL POWER.

FERMI - UNIT 2 3/4 4-4 Amendment No. 53

[

l REACTOR.00,0,LANT SYSTEM f

I

.R.E.C.I RC ULAT.I.0.li. PUMPS l

Lil!1 TING CONDITION FOR OPERATION i

3.4.1.3 Recirculation punp speed shall be n.aintained within:

-j i

a.

5% of each other with core flow greater than or equal to 70% of i

rated core flow.

1 I

b.

10% of each other with core flow less than 70% of rated core flow.

AP,P,LICABILITY: OPERATIONAL CONDITIONS I and 2*, during two recirculation loop operation.

j ACTION:

I i

With the recirculation punp speeds different by more than the specified l

lin.its, either:

l l

a.

Restore the recirculation pun:p speeds to within the specified limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or' i

b.

Shutdown one of the recirculation loops and take the ACTION required

(

by Specification 3.4.1.1.

Otherwise, 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 />.

j i

SLf,VEILLANCEREQUIREMENTS 4.a.1.3 Recirculation pump speed shall be verified to be within the limits l-at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Te'e'3hecialTestException3.10.4.

l-1 FERMI - UNIT 2 3/4 4 5 Amendment No. 53 r

4 REACTORC0_0LANT_SYSTEf.

S IDLE RECIM ULATION LOOP STARTUP i

Litt1T1h6 CCliDITION FOR OPERATION

.3.4.1.4 An idle recirculation loop shall not be started unless THERMAL POWER is within the unrestricted zone of Figure 3.4.1.4-1 and the temperature differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than or equal to 145'F, and:

a.

When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is,less than or equal i

to 50*F, or l

b.

When only one locp has been. idle, unless the temperature differential l

tatteen the reactor coolant within the idle and operating recircula-l tion loops is less than or equal to 50'F-and the operating loop 1

flow rate is less than or equal to 50'l of rated loop flow, j

l l

APPLICABILITY:

OPERATIONAL C0liDIT10h5 1, 2, 3 and 4.

{

j ACTION:

i With THERl'AL POWER, temperature differences and/or flow rates exceeding the abcse limits, suspend startup of any idle recirculation loop.

i l

I I

i SURVE1L L ANCE,,R,E,QUIREMENTS

-l 1

4.4.1.4 The THERMAL POWER, temperature differentials ar.d flow rate shall be determined to be within the limits within 15 minutes prior to startup of an idli u circulation loop.

l.

9 t.

l l

l-FERMI - UNIT 2 3/4 4-6 Amendment No. 53

a t

e e

=.

7 3

.E.,

b W

g @-

7 u

E n,.

w a

w g

...........o.

m i

g O

d

-u m

a a

g s

wcr Z=

l 1

8......

g

.......=

i l

?

J E'

S S

S R

2 2

l.

(031VW %) W340d 1VMW3H13W00

(

FERMI - UNIT 2 3/4 4-6a Amendment No. 53 L

I

/

REACTOR COOLANT SYSTEM l

- 3/4;.4.10 ' CORE-THERMAL HYDRAULIC STl.31LITY LIMITING CONDITION FOR OPERATION 3.4.10 The Reactor core shall not be operated in Region A or Region B of

. Figure 3.4.10-1.

1 APPLICABILITY:

OPERATIONAL CONDITION 1-ACTION:

a.

With the Reactor operating in Region A of Figure 3.4.10-1, immediately place the Reactor Mode Switch in the SHUTDOWN position.

b.

With the Reactor operating in Region B of Figure 3.4.10-1, immediately initiate action to exit Region B by inserting control rods.

c.

If, while exiting Region B, core thermal hydraulic instability occurs as evidence by APRM readings oscillating by greater than j

or equal to 10% of RATED THERMAL POWER peak to peak or LPRM i

readings oscillating greater than or equal to 30 watts /cm2 peak to peak, immediately place the Reactor Mode Switch in the SHUTDOWN position.

q SURVEILLANCE REQUIREMENTS 4.4.10.1 The provisions of Specification 4.0.4 are not applicable.

4.4.10.2 With THERMAL POWER greater than 30% of RATED THERMAL POWER and Core Flow less than 50% of Rated Core Flow, verify that the reactor core is not operatir.g in Region A or Region B of Figure 3.4.10-1 at least once every'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

7

. FERMI - UNIT 2 3/4 4-30 Amendment No. 53

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

o-(

l.

l l

l' R

=

6 l

.I a'

E 8

7 m

o

[

meu see3 %SP 8

3 9

i n

5

~

5 n.

t U

meu saco %DF c

S

.9a 3

Q C

a l

E h

y

\\

5, I

\\

I S

q s

g 2

j 5

E

-e i

)

\\.

I L

e a

a a

a

=

l (GRAVW %) W3AAOd 1TWW3HA 1

FERMI - UNIT 2 3/4 4-31 Amendment No. 53 i

SPECIAL TEST EXCEPTIONS 3]4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS L1HITING CONDITION FOR OPERATION 3.10.3 The provisions of Specification 3.9.1, Specification 3.9.3, and Table 1.2 nay be suspended to permit the reactor n. ode switch to be in the Startup position and to allow nore than one control rod to be withdrawn for shutdown nargin demonstration, provided that at least the following requirements are satisfied, a.

The source range monitors are OPERABLE per Specification 3.9.2 with the RPS circuitry "shortir.g links" removed.

b.

The rod worth mininiizer is OPERABLE per Specification 3.1.4.1, or con-forn.ance with the shutdown margin dentonstration procedure is verified by a second licensed operator or other technically qualified n.en.ber of the unit technical staff, c.

The " rod-out-notch-override" ccntrol shall not be used during cut-of-sequence movement of the contro' rods, d.

No other CORE ALTERATIONS are in progress.

EllCABILITY: CFERATIONAL CONDITION S, during shutdown margin demonstrations.

ACTION:

With the requirements of the abcVe specification not satisfied, imniediately place the reactor node switch in the Shutdown or Refuel positien.

- SL'PVEllL ANC,E,,R,EpulREMENTS 4.10.3 Within '30 niinutes prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the perforn.ance of a shutdown raargin denionstration, verify that; a.

The source range monitors are OPERABLE per Specification 3.9.2, b.

The rod worth n.inimizer is OPERABLE per Specification 3.1.4.1 or a ncond licensed operator or other technically qualified n;en.ber of the unit technical staff is present and verifies conipliance with the shut-dcwn demonstration procedures, and c.

No other CORE ALTERATIONS are in progress.

FERMI - UNIT 2 3/4 10-3

1 SPECI AL T,E,ST,,E,X,C,EP,TJO,NS 3],4.10.4 RECIRCULATION LOOPS 1

-ffb........

- _ _ :: r -- -- r- - -- ~- ~~ ~ --

3.10.4 The requirements of Specific 6tions 3.4.1.1 and 3.4.1.3 that 1

. recirculation loops be in operaticn with matched put.p speed may be suspended for up to P4 hours for the performance of PHYSICS TESTS, provided that THERMAL P0l!ER dc.es not exceed 5% of RATED THERMAL POWER.

APPLICABI,LJ,TY: OPERATIONAL CONDITION 2, during PHYSICS TESTS.

ACTION:

.I a.

With the above specified titre limit exceeded, insert all cor. trol rods.

b.

b!ith the above specified THERitAL POKER lir.!it exceeded during PHYSICS TESTS, imediately place the reactor mode switch in the Shutdown pcs h ion.

SURVEILj.,AN,CE, REQUIREMENTS i

4.10.4.1 The tirne during which the abcve specified requirement has been suspended tra11 be verified to bc less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at least once per hour during PHYSICS TESTS.

I 4.10.4.2 THERMAL POWER shall te determined to be less than 5% of RATED THERMAL POWER 6t least once per hour during PHYSICS TESTS.

l

. FERMI - UNIT 2 3/4 10-4 Amendment flo. 53 i

T 3/4.2 POVER DISTRIBUTION LIMITS

-............ n n : n n n n n :n C:n :n n n : n: n n n n n n n :------~~ ~ ---

The specifications of this section assure that the peak cladding tempera-ture following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENEPATION RATE 1he peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad tempera-ture is calculated assuming a LHGR for the highest powered rod which is equal to.

or less than the design LHGR corrected for densification. This LHGR times 1.02 i

is used in the heatup code along with the exposure dependent steady state gap-

)

conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RA1E (APLHGR) is this LHGR of the highest j

powered rod divided by its local peaking factor. The limitir.g value for APLHGR i

is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3 and 3.2.1-4 l

The Technical Specificotion MAPLHGR value is the ncst limiting composite of the fue.1 mechanical design analysis MAPLHGR and the ECCS MAPLHGR, Fuel Mechanical Design Analysis:

l'RC approved methods (specified in l

Reference 1) are used to demonstrate that all fuel rods in a lattice, L

cperating at the bounding power history, meet the fuel design limits i

specified in Reference 1.

This bcunding power history is used as the L

tasis.for the fuel d" sign analysis MAPLHGR value.

LOCA Analysis: A LOCA analysis is performed -in accordance with 10 CFR 50 Appendix K to denonstrate that the MAPLHGR values con. ply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most l

limiting break size, break location, and single failure conibination for l

tte plant.

Only the n.ost limiting MAPLHGR values are shown in the Technical Specification figures for multiple lattice fuel. When hand calculations are required, these Technical Specifications MAPLHGR figure values for that fuel type are used for all lattices in that bundle.

l For some fuel bundle designs MAPLHGR depends only on bundle type and burnup. Other fuel bundles have MAPLHGRs that vary axially depending upon the specific con.bination of enriched uranium and gadulinia that comprises a fuel bundle cross section at a particular axial node.

Each particular combination I

of enriched uranium and gadolinia, for these fuel bundle types, is called a lattice type. These particular fuel bundle types have MAPLHGRs that vary by L

lattice (axially) as well as with fuel burnup.

J-1 l

l FERMI - UNIT 2 B 3/4 2-1 Amendment No. U, 53

i 3/4.2 POWE,R,DISTRJBUTJON, LIMITS-

- - -BASES-3/4.2.1'AVERA_G_EPLANARLlHEARHEATGENERATIONRATE,(Co,n,tinued)

For plant operation with a single operating recirculation loop, the above MAPLHGR limits are multiplied by 0.90. The constant factor of 0.90 is derived from LOCA analysis initiated from single loop operation to account for earlier L

boiling transition at the limiting fuel node compared to the standard LOCA j.

analysis.

Reference 1.

" General Electric Standard Application for Reactor Fuel," NE0E-24011-P-A L

(letestapprovedrevision).

l l

t 1?

1.

l

(-

l l

l l

I l'

l l.

l l..

l FERMI - UNIT.2 B 3/4 2.la AmendmentNo.97.53 l

1

p-

-t BASESTABLEB_3.f.j-j i

SIGNIFICANT INPUT PARAMETERS TO THE LOS_S.0F-COOLALT, A,CCJD,ENT, AtlALYSJS 1

Plant Parameters:

L Core THERMAL P0WER...............

3430 MWt* which corresponds to 105% of rated steam flow 6

L Vessel Steam Output..............

14.86 x 10 lbm/hr which corresponds to 105% of rated H

steam flow 1

Vessel Steam Dome Pressure.......

1055 psia Design-Basis Recirculation Line l

-Break Area for:

2

'a.

Large Breaks 4.1 ft 2

b.

Small Breaks 0.1 ft Fuel Paranieters:

PEAK TECHNICAL INITIAL' SPECIFICATION DESIGN MINIMUM LlHEAR HEAT AXIAL CRITICAL l

FUEL BUNDLE GENERATION RATE PEAKING POWER l

F

TYPE GEONETRY

( kW/f t)

FACTOR RATIO

)

Initial Core 8x8 13.4 1.4 1.18**

F1rst Reload 8x8 14.4 1.4 1.18**

i

-A more detailed listing of input of each model and its source is presented in

'j Section 11 of Reference 1 and subsection 6.3 of the UFSAR.

• This power level meets the Appendix K requirement of 102%. The core-heatup calculation assumes a bundle power consistent with operation of the highest powered rod at'102% of its Technical Specification LINEAR HEAT GEhERATION RATE linit.

l-

• • For single recirculation loop operation, loss of nucleate boiling is assun:ed l

at 0.1 sectnd after LOCA regardless of initial MCPR.

l l

FERMI - UNIT 2

:l' 2-3 Amendment No. #2. ##, 53

7 3/A.4 REACTOR CUDLAfiT SYSTEM r

BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted at power level is up to 70% of RATED THERMAL POWER if the MCPR Cuel cladding safety limit is ircreased as noted by Specification 2.1.2.

APRM scram and control rod b1Mk setpoints (or APRM gains) are adjusted as noted in Tables 2.2.1-1 and 3.3.6 2, respectively. MAPLPCR limits are decrened by the factor given in Specification 3.2.1.

A time period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is aliowed to make these adjustnents following the est6blishment of single loop operation since the need for single loop operation often cannot be anticipated., MCpR operating lin:its adjustutents in Specification 3.2.3 for different plant operating situetions are applicable to both single and two recirculation loop operation.

To prevent potential control systen. oscillations from occurring in the recirculation flow control systen., the operating mode of the tecirculation flow centrol system n'ust be restricted to the nanval control node for single-loop operation.

Additionally, surveillence on the pun:p speed of operating recirculation 1ccp is irposed to exclude the possibility of excessive core internals vibration.

The surveillance on differential tenperatures below 307 THERMAL POWER or 50% rated recirculation loop flow is to prevent undue thermal stress on vessel nczzles recirculation pump and vessel bottom head during a power or flowincreasefollcwingextendedoperationinthesinglerecirculationloop n,cde.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in care of a design-basis-accident, inctease the blowdown area and reduce the capability of reflooding the core; thus, tbc requirtrent for shutdown of the facility with a jet pump inoperable.

Jet purp foilure can be detected by n.cnitoring jet pump performance en a prescribed schedule for significant degradation.

Recirculation punip speed n.ismetch limits are in compliance with the.~CCS LOCA analysis design criteria for two recirculation loop operation. Ttt limits will er.sure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during two locp operation, continued operation is permitted in a single recirculation loop node.

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 locp. The loop ternperature must also be within 50*F of the teactor pressure vessel coolant temperature to prevent thermal shock to the recirculation put.:p and recirculation nozzles.

FERMI - UNIT 2 B 3/4 4-1 Amendment No. 53 l

em,

~

e 3/4_.4 REAR. TOR C,0,0,LfNJT,SYSTg sA$5,$,,,n,,,,==__,,_===,

3/4.4.1_ RJCJ,R,C,U,L,AJJON, SJ,$J,EM,J, Con,tj,nuL l d

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 undue stre'ss in the reactor vessel bottern head.

Requireri.ents are imposed to prohibit idle loop startup above the 80f rod line to n.inimize the potential for initiating core t; ermal-hydraulic instability.

l 3 /4. 4. 2,,S,A[,E,TJ/f;,E,LJ((, VALy,ES, The safety vaht. function of the safety / relief valves operate to prevent th reactor ccolant systen frcir, being pressurized above the Safety Limit of 13P5 rsig in accordance with the ASME Code. A total of 11 OPERABLE safety /

relief valves is required to lir.it reactor pressurt to within ASME III allowable values for the worst case upset transient.

Derrenstration of the safety / relief valve lif t settings will occur only curing shutdcwn and will to performed in accordance with the provisicr.s of l

Specifiu tion 4.0.5.

The low-low set system ensures that a potentially tigh thrust load (desig-

[

natcd as load case 0.3.3) on the SRV discharge lines is eliminated during sub-sequent actuations. This is achieved by autornatically lowering the closing set-l point of twc valves arid lowering the opening setpoint of two valves following the ir.itial ciening.

Sufficier.t redundancy is provided for the low-low set systen such that it,ilure of any ore valve to cpen or cicst et its reduced setpuir.t c'oes not violate the design batis.

L l

l FERMI - UNIT 2 B 3/4 4-la Arnendtrent No. 53

~

l

'l REACTOR COOLANT SYSTEM BASES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection I

Systems", May 1973.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from tht. reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due. to equipment design and the detection i

capability of the instrumentation for Mtermining system leakage was also con-sidered.

The eviden+.e obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would g m rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.

Service sensitive reactor coolant system Type 304 and 316 austenitic stainless steel piping; i.e., those that are subject to high stress or that (

.tain relatively stagnant, intermittent, or low flow fluids, requires additional surveillance and leakage limits.

The purpose of the RCS interface valves leakage pressure monitors (LPMs) is to provide assurance of the integrity of the Reactor Coolt.nt System pres-sure isolation valves which form a high/ low pressure boundary.

The LPM is designed to alarm on increasing pressure on the low pressure side of the high/

low pressure interface to provide indication to the operator of abnormal interface valve leakage.

t The Surveillance Requirements for RCS pressure isolation valves provide added assarance of valve integrity thereby reducing the probability of gross valve fai?ure and consequent intersystem LOCA.

Leakage from the RCS pressur6 isolatien values is IDENTIFIED LEAKAGE and will be considered as a portion of the sMowed litait.

3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

FERhl - UNIT--2 B 3/4 4-2 Amendment No.14

i RFACTOR C00 Laid SYS,TJM

]

BASES

)

3/4.4.10 CORE THERMAL HYDRAULIC STABILITY BWR cores typically operete with the presence of global flux noise in a stable mode which is due to randem boiling and flow noise.conditicos are changed, alcrig w As the i

pressure, subcooling, power distribution, etc.) the thermal hydraulic / reactor kinetic feedback nectanistr can be enhanced such that random pet turbations may result in sustair:ed limit cycle or divergent oscillations in power and flow.

Two major enodes of oscillations have been observed in BWRs. The first mode is the fundenent61 or core-wide oscillation node in which the entire core cscillates in phase in a given axial plar,e. The second mode involves regior.a1 oscillation in which one half of the core oscillates 180 degrees out of phase with the other half.

Studics have iridicated that adtquate rnergin to the Safety Limit Minimum Critic 61 Pcker Ratio (SLHCPR) may not exist during repicnal oscillations, i

Region A and D of figure 3.4.10-1 represent the least stable conditions of the plant (high power / low flow). Region A and B are uscelly entered as the result cf a plant transient (for example, recirculation punp trips) and therefore are generally not considered pert of the norn.a1 operating domain.

Since all stability tvents (including test experience) have occurred in either Region A or B these regions are avoided to n.inimize the possibility of encountering oscihaticns and potentially challenging the SLMCPR. Therefore, intentional cperation in Regions A cr B is not allcwed.

It is recognized that during certain tboormal conditions within the plant, it may betonie necessary to enter Regico A or B for the purpose of protecting equipinent which, were.it to fail, could impact plant safety or for purpose of protecting a safety or fuel operatir g limit.

In these cases, the appropriate actions for the region entered would be perforrned as required, t'ost oscillattens that have occurred during testing and operation have cccurred at or above the 100% rod line with core flow r. ear natural circulation. This bt-havicr is consistent with analysis which predict reduced stability nargin i

with increasing power or decreasing flow. As core flew is increased or power decreased, the probability of oscillttions occurring will & crease.

Region A of Figure 3.4.10-1 bounds the n.ajority of the stability (vents and tests l

observed in CE EWRs Since Region A represents the least steble region of the pcwer/ flew operatin domain, the potential to rapidly encounter large magnitude cort. thermal hycrau ic oscillations is increased. During transients, the operator may not have sufficient time to ner.vally insert control rods to mitigate the oscillations before they reach an unaccepteble magnitude.

Therefore, the prornpt action of trarvally scran' ming the plant when Region A is j

entered is required to ensure protect Sn of the SLMCPR.

FERMI - UNIT 2 B 3/4 4-8 Amendntent No. S3

PEACTOR COOLANT SYSTEM

% C C C:! C ::~;;; M Z ::I n Z :::: Z C C::C :2 n l:::::::: CII!!::: C*

3/4.4.10CORETHERMALHYDRAULICSTABILITY(Continued)

Based on tect and operating experience, the frequency of core thermal hyd:aulic oscillations is less in Region B than in Region A.

Decay ratios are expected and predicted to be lower in this region since Region B covers a lower power and higher flow range than Region A.

Also, the margin to the SLMCPR will typically be larger in Region B than in Region A.

With more margin to SLMCPR and n lower probability of oscillations, exiting Region B by control rod insertion is justified.

However, if oscillations are observed while exiting

)

Region B, the reactor will be manually scramed.

The poter.tial for core thermal hydraulic oscillations to occur outside of Regiuns A and B is very small and therefore special requirements are not necessary outside of these regions.

FERMI - UNIT 2 B 3/4 4-9 Amendment No. 53