Text

Proposed Tech Specs,Consisting of Reorganization of Apps a & B
	ML18025B502
Person / Time
Site:	Browns Ferry
Issue date:	05/15/1981
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML18025B501	List: ML18025B500; ML18025B502;
References
	NUDOCS 8105210136
	Download: ML18025B502 (193)
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ENCLOSURE PROPOSED CHQNGP$, 2'O TECHNICAL SPECIFICATIONS ERAS FERE NUCLEAR PLANT

UNIT 1

PROPO/ED CHANGES

PRQPOSFD CIIANGLS TO APIiI'.NDIX A TECHNICAL SPECIFICATZONS

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1

1 FUEL CLADDING INTEGRITY 2

1 FUEL CLADDING INTEGRITY" A licabilit A licabilit Applies to the interrelated variables associated with fuel thermal.behavior.

O~b'ective To e'tablish limits which en.,ure the integrity of the fuel cladding.

S ecifications

A-

'Thermal Power Limits

1. Reactor Pressure

> 800 psia and Coze Flow >

10%

of Rated.

When the reactor pressure is greater than 800 psia, the existence of a minimum critical power ratio (MCPH) less than 1.07

. shall constitute violation of the fuel cladding integrity safety limit.

Applies to trip settings of the instruments and devic'es which are provided to prevent the reactor system safety limits from being exceeded.

~Ob ective To define the level of the process variables at which automatic protective action is initiated to prevent the fuel cladding integrity safety limit from being.exceeded.

S ecification The limiting safety system settings Shall be as specified below:

I A-Neutron Flux Trip Settings APRM Flux Scram Trap Setting (Run Mode) ee When the Mode Switch is in the RUN position, the APRM flux scram trip setting shall be:

SS(0.66H

+

54%)

where:

S

= Setting in per-cent of rated thermal power (3293 MWt)

W = Loop recircu-late.on flow

- rate'n'er-

, cent of rated (rated loop recxrculatxon flow rate. equals 34;2x106 lb/hr)

ShFETY LIMIT LIMITING ShFFTY SYSTEM SETTING FN.".r. CLnDDINC'NTEGRIPY

~

I 2.1 FUVI CLADDING INTEGRITY In the event of operation with the core maximum fraction of limiting power density (QtFLPD) greater than fraction ot rated thermal power (FPZ) the setting shall be modified as follovs:

p Sk(0.66M + 5(X)

PIIP CHFLPO O.

c, For no combination of loop recircu-lation flov rate and core thermal

'over shall the APlQf-flux scram trip setting bc allowed to exceed 120M of rated thermal povez.

(Hotc:

These settings assume operation

~'ithin the basic therma) hydrau)ic design criteria.

These criteria arc LllGR C 18.5 kv/ft for 7Ã7 fuel and~

13.4 Lv/ft for 8X8, 8x8R, and P8x8R fUel, MCPR limits of Spec 3.5.k. If it is determined that either o.'hese design criteria is being vio)".ed during operation, action.shal) bc initiated within 15 t;.irutos to rc tore operation within prescribed )i..its.

Surveillance requireraents for APF.::

scram setpnint are given in specification 4.1.B.

I I

l d.

" The APRM Rod block trip setting shall be:

S

< (0. 66M +425)

,RB where:

S ~ = Rod block setting in percent of r'ated thermal power (3293 Mwt)

= Loop recirculation flow rate in percent, of rated

{rated loop recirculation flow rate equals.

~ 34.2 x 10s lb/hr) 9

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1

1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY In the event of operation with the core maximum fraction of limiting power density (CMFLPD) greater than fraction of rated thermal power (FRP) the setting shall be modified as follows:

8

{0.66W +42% ) "~

QIPLPD 2.

Reactor Pressure

~800 PSIA or Core Flow "-10% of rated.

When the reactor pressure is ~800 PSIA or core flow is

=10% of rated, the core thermal power shall not exceed 823 M<t (

25%, of rated thermal power).

2, APRM and IRM Trip Settings (Startup and Hot Standby Modes).

a.

APRM--Nh:.n th'-

reactor mode swatch is in the STARTUP position, the APRM scram shall be set at less than or equal to 15% of rated power.

I

~

b.

IRM The IRM scram shall be set at less than or equal to 120/125 of full scale.

10

SAFETY LXMIT LXMITXNG SAFETY SYSTEM SETTING l.l FUEL CLADDING XNTEGRITX 2.1 FUEL CLADDING INTEGRITY B

Power Transient

-B. Power Transient Tri Settin s

To ensure that the Safety Limi,ts established in Specification

l. l.A are not exceeded, each required scram shall be initiated by its expected scram signal.

The Safety Limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.

Scram and i,sola-tiOn (PCIS groups 2,3,6) reactor low water level

2. Scram--turbine stop valve closure Scram--turbine control valve fast cloaura or turbine trip 538 in above vessel zero S

10 per-cent valve closure

~'-

550 psig 4.

Scramlow con-denser vacuum 5, 'Sc'ram--main steam line isolation 2 23 inches Hg vacuum 5

10 per-cent valve

. alosur~

6, Main steam isola-tion valve closure nuclear system low pressure

~ 825 psig C. Reactor Vessel Hater Level C. Hater Level Tri Settin s

Hhenever there is irradiated fuel in the reactor vessel, the water. level shall. not be less than'7.7 'in. above the top of the normal active fuel zone.

Core spray and LPCI actuation

'reactor 1oi water.

level 2.

HPCI and RCIC actuation--reac-tor low water level

~ 3.

Mairi steam isola-tion valve closure--reactor low water *level 2 378 in.

above vesoei zero a

470 in.

)

above vessel zero 470 in. f above vessel zero U.

2'. 1 bASES fron fuel dmage, assumfng a steady-scarc operation ac the crfp sectfng, over chc encfre recirculation flov range.

The coargfn co chc Safecy Lfoit increases ae the flov decreases for the spec;f ffed c rip sett fng versus flou rc lationshfp 1

therefore, the vorsc case HCPR vhich cou'.d occur during s eady-state operation fs ac, 10RZ of raced cheroal pover because of the APIN rod block trip setting.

The

~ ecual pnver dfscrfbuc fnn fn che core is escabl fshed hy specfffrd control rod sequences and fs wnnicored conc fnuously by che -fn-core LPHtf syscem.

As vith the Apfuf scree cr'fp sect fng, the APR.'t rod block trip seccfnP fs ad jusced dovnvard if che CHFLPD exceeds Fffp thus prvsecvfng che

!>NC rnJ block safety margin.

C.

Reaccor Water Lou Level Scrap and l olac ion (Face c Hain Sceaml fnes)

The sec point for the lcm'~ level scrao fs above the boc coo nf the separator skirt.

This level has been used fn !ransfcnc'nalyses dealing vf ch coolant invencory decrease.

The re ulcs reporccd fo FSAR subsection 14.S shiv that scram and fsolacion of all process lines (excepc wain steao) ac this level adequately procects the fuel and the pressure barrier, because HCPR fs g,rcater than 1.07 in. all cases, and sysceo pressure does noc.reach che safecy valve settings.

Thc scceu setting fs'pproxfoacely 31 inches belov the normal operating range and is thus'adequate ro avofd spurious scr~.

The turbine stop valve closure trip anticipates the pressure, neutron f1ux

'nd heat flux increases that would result from closure of the stop valves.

Mith a

.rip setting oj 10" of valve closure from fu)l open, the resultaht increa.e in heat flux is such that adequate thermal margins are maintained even during the worst case trarsient that assumes the turbine bypass valves remain closed, (Reference 2)

E., Turbine Control Valve Fast Closure. or Turbine Tri Scram

'urbine control valve fast closure or turbine trip scram anticipates the

, pressure, neutron flux, and heat flux increase that, could result from

~ control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.

The reactor protection system initiates a scram in less than 30 milliseconds after

'. the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducinp hydraulic control oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection systeni.

This trip settinq, a nominally 50'; qreater closure time and a different valve characteristic from that of the turbine stop valve, combine tu produce transients very similar to that for the stop valve.

iso signi fi-cant.

change in thCPR occurs.

Relevant transient analyses are discussed in References 2 and 3 of the Final Safety Analysis Report.

This scram is bypassed wlien turbine steam flow is below 30,", of rated, as m asured by'urbine first state pressure.

23

)tain Condenear Lov Vacuum Scrms Yo protect the main condenoer agiinet overpreeeure, a looo of con-denser vacuun initiates outmctic closure of the turbine atop vslveo.

To anticipate the tranoiont and outoaat9c

. oc;aa reoulting frccx the cloaure of the turbine atop>>alvee, low con-deaaor vacuum initiotca a ocracy.

The lov vacuum scrawl eot point Ce

@elected to initiate a ocr'efc.'a the closure of the turbine atop vagvoe ie lnltl648d.

C.

4 H.

Hnin Stnaw Linn Ia~ ation on Lcm Preoeure and Kain Stean Line Iooletion -Scree The leer preeoure isolation of the asin eteeu lines at 825 paid vms provided to protect againet rapid reactor dcpresourixetion and the rec)ulting rapid cooldovn of the vessel..

Advantage is taken of. the ocrara faaturo that occuro uhcn the as'tecui line isolation valvee nro clooed, to provide" for reactor ohutdoMo po that hich poM>>r opera" ticra at lov reactor proooure done not. occur, thus providing protection for the fuel cladding inteOrity safety litsit.

Operation of,the reac" Nor nt preoaurea lovor'han Sp peig requireu that the reactor cede ovitch ba in the. STAR'NP poeition, vhere protection of thc fuel cladding integrity oafoty lirjit 'ie provided by'the IRH end APRH high neutron flux ncrcusa.

Thus, the ceAination of nein oteau'ine lov praoourc isolation tM icolation valva clooure acraa assures tho availability of neutron

- flux acrete protection over the entire range of applicability of the fuel cladding integrity aaiety I&sit.

In addition, the iaolation valve closure ecran anticipates th<<preeouro and flux tranaiente that occur during normal or inedvo tant isolation valve cloeuro.

With the scree eot ot 10 percent of volvo clooure, neutron i'lux dose not increase.

p4

TABLE 3 1

A REACTOR PROTECTIOM SYSTEI1 (SCRAM)

INSTRVMENZATIOM REQVIREMEtiT NN e NOe od:

Operable Inste Channel s Per Trip

~Rute

.I filiTri runetion I

1 Mode Switch in Shutdown 1

Manual Scram IRM (16)

High Flux Inoperative Tr l t--'~ttin 120/125 Indicated on scale Modes in Mhich Function Must Be rable Shut-Startup/Hot dote

~Refuel 7

~Standb X(22)

X (22)

Run

~notion I X

1.A X

1.A (5) 1.A (5) 1 A 2

2 2

2 APRM (16)

High Plux High Flux Inoperative Downscale See Spec.

2.1.A.1 15% rated cower (13)

> 3 Indicated on Scale X (21)

X (21)

(11)

X 1.A or 1.B X(17)

(15) 1.A or 1.B X(17)

~ X 1.A or 1.B (11)

X(12) 1.A or 1

B High Reactor Pressure

< 'l055 psiq X(10)

X 1 A High DrYwell Pressure

{14)

Reactor Low Mater Level (14)

High Mater Level in Scram Discharge Tank 2 '

tf3iz

> 538" above vessel zero

< 50 Gallons X (8)

X (2)

X(8)

X 1 A X

1 A X

1 A

TABLE 3 1

A REACTOR PROTECTION SYSTEM

{SCRAM)

IMSTRIJMEMTATIOM REQUIREMEMT Min Mo of Operable Inst.

Channels Per Trip

~secen i iiiinein euncnion Tri Level Settin Modes in Mhich Function Must Be O

rable Shut Startup/Hot coen

~eeeuei 7

Seendn eun

~ice ion Main Steam Line Isola-tion Valve Closure 5

10% Valve Closure Turbine Cont. Valve Past Closure or k 550 psis Turbine Trip X(3) (6)

X(3) (6)

X(6) 1.A or 1.C X(4) 1Aor 1D Turbine Stop Valve Closure S 10'alve Closure X(4) 1.A or 1.D Turbine First Stage Pressure Permissive Turbine Condenser Low Vacuum Main Steam Line High Radiation (14) not 1154 psi8 I 23 In. Hg, Vacuum 3X 5ormal Full Pover

Background

X(18)

X(3)

X(9)

X (18)

X(18)

{19)

X(3)

X 1.A or 1.C X(9)

X(9) 1.A or 1.C

10.

Not required to bc ope"able when thc reactor prcssure vessel head is not bolted to the vrsscl.

The APRN downscale trap function is only active when the reactor mode switch is in run.

12.

13.

The APRM downscale trip is automatically bypassed when the IRM.instrumentation is operable and not high.

Less than 14 operable LPRM<s will cause a trip system trip'.

14.

channel shared by Reactor protection System and primary Containment and Reactor Vessel Isol t.ion Control System.

channel failure may be a channel failure in each system.

\\

15.

The APRM 15% scram is bypassed in the Run Mode.

16.

17.

18.

,20.

.21.

Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion).

A channel failure may be a channel failure in each system.

If a channel is allowed to be INOPl'.RABLE per Table 3.1.A, thc corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).

Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).

This function must inhibit the automatic bypassing of turbine co:.trol valve fast closure or turbine trip scram and, turbine stop valve closure scram whenever turbine first,stage pressure'is greater than or equal to 154 psig.

Action 1.A or 1.D shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function.

Otherwise, no action is required.

Tne nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively)

'are established based on the normal background at full power.

The allowable setpoints for alarm and reactor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.

The APRM High Flux and Inoperative Trips do not have to be operable in the Refuel Mode if the Source Range Monitors are connected to give a non-coincidence, High Flux scram, at 5 x 10 cps:

The SRH's, 5

chal,l be operablc per Specification 3.10.B.1.

The removal of eight (8) shorting links i., required to provide non-coincidence high-flux scram prot. ction from the Source Range Monitors.

22.

The three required XR'1's per trip channel is not required in the Shutdown or Refuel Modes if at least four IRM's (one in each core quadrant) are connected to give a non-coincidence, High Flux scram'.

The removal of four,(4) shorting links is required to provide "non-coincidence high-flux scram protection from the IRM'0, 23.

A channel may be placed in an inoperable status for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-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.

TABLE Q 1 A REACTOR PROTECTION SYSTEM (SCRAN)

XNSTRUM~.iATION FUNCTIONAL TESTS HINLMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY X)<S R.

AND CONTROL CIRCUITS Group (2)

Functional Test Minimum Frequency (3)

Main Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or Turbine Trip Tr '

Cha lnel and Alarm Tr 'h.I."ne) anc. Alarm Once/Month

( 1)

Once/Month (1)

Turbine First Stage Pressure Permissive Turbine Stop Valve Closure Trip Chan:.-:'nd Alarm Trip channel and Alarm Every 3 Months Once/Month (1)

TABLE 4-1 B

REACTOR PROTECTXOS STSTEM (SCRAM)

INSTRUMENT CALIBRATION MINIMUMCALIBRATION FRIQQSES IZS FOR REACTOR PROTECTION INSTRUMENT CHANNELS.

Instrument Channel IRM High Flux APRM High Flux Output Signal Flow Bias signal LPRM signal High Reactor Pressure High Drywell Pressure Reactor Low Water Level High Mater Level in Scram Uischarge Volume

~

~ Turbine Condenser Low Vacuum Q

Main Steam Line Isolation Valve Closure Main Steam Line High Radiation Turbine First Stage Pressure Permissive Croup (1)

B B

Calibration Comparison to APRM on Control-led stattups (6)-

Heat Balance Calibrate Flow Bias Signal (7)

TIP System Traverse (8)

Standard Pressure Source Standard Pressure Source Pressure standard Note (5)

Standard Vacuum Source Note (5)

Standard Current Source (3)

Standard Pressure source Minimum Frequency (2)

Note (4)

Once every 7 days Once/operating cycle Every 1000 Effective Pull Power Hours Every 3 Months Every 3'Months Every 3 Months Note

{5)

Every 3 Months Note (5)

Every. 3 Months Every 6 Months Turbine Stop Valve Closure Note (5)

Note (5)

NOTES FOR TABLE 4. l. B 1.

A description of three groups is included in the bases of this specification.

2.

Calibrations are not required when the systeras are not required to be operable or are tripped. lf calibrations are

missed, they shall be performed prior to returning the system to an operable status.

3.

The current. source provides an instrument, channel alignment.

Calibration using a radiation source shall be made each refueling outage.

5-f 6.

Required frequency is initial startup following each refueling outage.

Physical. inspection and actuation of these position switches will be performed once pei operating cycle.

On controlled startups

~ overlap between the IRM~ s and APRM<s C

will be verified.

7 ~

The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during'ach operating cycle.

The instrumentation is an analog type with redundant flow signals that, can be compared.

The flow comparator trip and upscale will be functionally tested according to Table 4. 2.C to ensure the p'per operating during the operatin<< cycle.

Refer to 4.1 Bases for

- furtner explanation of ca.<iver;.tion frequency'.

8.

h comp)ctc tip system

.raverse calibratcs the lPR>< signals to the process computer.

Thc indivi<lual LP!8 meter readircs !ril.l be adjus(.c<l as a mini<<>><n at thc beginning of carl< operating cvcle b<.for<

r<.achl>>g 1007. power.

41

Q,f NOISES

(

modes.

In the paver range the AplN system pzovides requ'zed protection.

Reft Section 7.5.7 FSAR.

Thus, the IRH System is ooc required in the Run mode.

The APRM'e and the IK'f's provide adequate coverage ia the startup and incermediate range.

The high reactor pressure, high dzyvell pressure, reactor lov vater level and scram discharge volume high level scrams are requized for Startup an Run modes of plant operation.

They are, therefore, required to be opera<<

. tional for these modes of'eactor operation.

The requirement to haye che scram functions as indicated in Tabl 3.

T ble 3.1.1 o ezable io che Refuel mode is to assure that shifting to the Refuel aude during reactor pbvcr operation does not diminish the need for the zeac't opera e

ea

'toz protection system.

The turbine condenser Iov vacuum scram is only required during povez o erstioo aod must be bypasoed to start up th>> unit.

Belov IS4 psig tur-bine first stage pressure (30X of rated),

the scram signal due to tuzbin p

e stop valve closure, end turbine control valve fast closure (ia bypassed because faux aad pressure sczam sze adequate to protect the reactor.

Because of che APISH downscale limit of

> 3Z vhea in the Run mode and high level limit of

<15X vhen in the Startup

Hode, the transition bstveea the Scartup and Run Modes must ba made vith the APL'nstrumentatioa indicating betveen 3I and 15X of rated povsr or a control rod scram vill occur.

In

addition, che IRN system must be indicating belov the High Plux setting (120/125 of scale) or a scram vill occur vhen ia the Startup Mode.

Poz normal operating conditions, these limits provide assuraaca o

overlap betveen the IBH system and APL'ysrem so that there aze ao "gapa" ia the paver level indications (i.e., ths pover level is continuously aanitorad

!rom beginning of startup to full pover aad from full povar to shutdovn)

~

4'hen pover io befog reducsd, if a transfer to the Startup mode is made sad tha ILt's have ooc been fully iaserced (a malopezatioae'ut aot impossible condition) a control rod block Medisteiy occurs so chat reactivity mssz-tioo by control rod vichdraval csaaoc occur.

TABLE 3a 2 A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOIATION INSTRUMENTATION Minimum No.

Operable Per T~ri S s

'I (ll)

Punct ion 2

Instrument Channel-Reactor Lov Mater Level (6)

Tr Level Settin h 538" above vessel zero Action 1

A or (B and E}

Remarks Belov trip setting does the following:

a.

Initiates Reactor Building Isolation b.

Initiates Primary Containment Isolation c.

Initiates SGTS Instrument Channel-Reactor High Pressure Instrument Channel-Reactor Low Mater Level (LIS-3-56A-D, SM t 1) 100

+

15 psig 470" above vessel zero A

1.

Above trip setting isolates the shutdown cooling suc ion valves of the RHR system.

1.

Below trip setting initiates Hain Steam Line Isolation Instrument Channel-High Dryvell Pressure (6)

(PS-64 56A-D)

Instrument channel-High Radiation Main Steam Line Tunnel (6)

Instrument Channel-Lov Pressure Hain Steam Line S

2.5 Psi8 3 times normal rated full power background 825 Psig (4)

A or (B and E) l.

Above trip setting does the folloving:

a.

Initiates Reactor Building Isolation b.

Initiates Primary Containment

'solation c.

Initiates SGTS 1.

Above trip setting initiates Hain Steam Line Isolation Below trip setting initiates Hai'n Steam Line Isolation 2 (3)

Instrument Channel-High Plow Hain Steam Line S

140% of rated steam flow B

1.

Above trip setting initiates Hain Steam Line Isolation

0 TABLE' 2 A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.

Function Tri Level Settin Action 1

Remarks Instrument Channel-Main Steam Line Tunnel High Temperature Instrument Channel-Reactor Mater Cleanup System Floor Drain High Temperature Instrument Channel Reactor Water Cleanup System Space High Temperature Instrument Channel Reactor Building Venti-lation High Radiation-Reactor Zone Instrument Channel-Reactor Building Venti-lation High Radiation-Refueling Zone 200OF 160 - 180oF 160 - 180oF S t00 mr/hr or downscale G

S 100 mr/hr or downscale P

Above trip setting initiates Main Steam Line Isolation.

1.

Above trip setting initiates Isolation of Reactor Water Cleanup Line from Reactor and Reactor Water Return Line.

1.

Same as a~mve l.

1 upscale or 2 downscale will a.

Initiate SGTS b.

Isolate reactor one and refueling floor.

c.

Close atmosphere control system.

1.

1 upscale or 2 downscale will a.

Initiate SGTS b.

Isolate refueling floor.

c.

Close atmosphere control system 2 (7) (8)

Instrument Channel SGTS Plow - Train A Heaters 2 (7) (8)

Instrument Channel SGTS Plow Train B

Heaters 2 (7) (8)

Instrument Channel SGTS Flow - Train C

Heaters Charcoal Heaters S 2000 cfm R.H. Heaters S 2000 cfm Charcoal Heaters S 2000 cfm R.H. Heaters S 2000 cfm Charcoal Heaters S2000 cfm R.H. Heaters S 2000 cfm H and (A or F)

H and (A or F)

H and (A or P) 1.

Below 2000 cfm, trip setting coal heaters will turn on.

2.

Below 2000 cfm, trip setting heaters will shut off.

l.

Below 2000 cfm, trip setting coal heaters will turn on.

2.

Below 2000 cfm, trip setting heaters will shut off.

1.

Below 2000 cfn, trip setting coal heaters will turn, on.

2.

Below 2000 cfm, trip set ting heaters will shut off.

char-Ro do char-R char-R. H.

TABLE 3 2<<A PRXNARY CONTAINMEN~

AND REACTCR BUILDING ISOLATION INSTRUHENTATZON Hinimum No Operable Per 1

Function Reactor Building Isolation Timer (refueling floor)

Instrument Channel-Static Pressure Control Permissive

{refueling floor)

Tri Level Setti OStS 2secs<<

Action 1

HorF HorF 1.

2<<

Remarks Below trip setting prevents

"",urious trips and system pertur-

'on" from initiating isolation Located in unit 1 only Pe missive for static pressure control

{SGTS A, B, or C on).

Channel shared by permissive on reactor zone static pressure cont.

Static Pressure control Pressure Regulator (Re-fueling Floor) 1/2" H 0 HorF 1.

2 ~

Located in unit 1 only Controls static pressure of refueling floor during reactor building isolation with SGTS running.

Reactor Building Isolation 0

< t <

2 secs.

Timer (reactor zone)

GorA or H Below trip setting prevents spurious trips and system pertur-bations from initiating isolation 1 (9)

Instrument Channel-Static Pressure Control Permissive (reactor xone)

N/A Perm'ssive for static pressure control (SGTS A, B, or C on).

channel shared by permissive on refueling floor static pressure control.

1 (9)

Static Pressure Control Pressure Regulator (reactor zone) 1/2" H 0 controls static pressure of reactor zone during reactor building isolation with SGTS running.

Group "1 (Initiatinq) Logic N/A Group 1 (Actuation) Logic N/A Refer to 'iable 3.7.A for list of valves.

Refer to Table 3.7.A for list of valves.

0 TABLE 3 ~ 2 A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.

Operable Per Functio 2

Group 2 (Initiatin ) Logic Group 2

(RHR Isolation-Actuation) Logic Group 2 (Tip-Actuation)

Logic Group 2 (Drywell Sump Drains-Actuation)

Logic Tri Level Settin N/A N/A Action 1

A or (B and E)

Remarks 1.

Refer to Table 3.7.A for list o" valves.

Group 2 (Reactor Building N/A S Refueling Floor, and Dry-well Vent and Purge-Actuation) Logic Group 3 (Initiating) Logic N/A Group 3 {Actuation) Logic

-N/A Group 6 Logic Group 8 (Initiating) Logic N/A Reactor Building Isolation N/A (refueling floor) Logic Reactor Building Isolation N/A (reactor xone) Logic FandG H or F 8 or G or A 1.

Part of Group 6 Logic.

1.

Refer to Table 3.7.A for list of valves.

1.

Refer to Table 3.7.A for list of valves.

1.

Refer to Table 3.7.A for list of valves.

2.

Same as Group 2 initiating logic.

1.

Logic has.permissive to refueling floor static pressure regulator.

Logic has permissive to reactor zone static pressure regulator.

0 TABLE 3 2 PRIMARY CONTAINMEHP AND REACTOR BUILDIHG ISOLATION INSTRUMENTATION Minimum No.

Operable Per unct on Tri Level Settin Remarks 1(7) (8)

SGTS Train A logic 1(7) (8)

SGTS Train B Logic 1(7) (8)

SGTS Train C Logic Static Pressure Control (refueling floor) Logic NIA L or (A and F)

L or (A and F)

HorF 1.

Located in unit 1 only.

1 (9)

Static Pressure Control (reactor zone) Logic Refe to Table

3. 2.B for RCIC and HPCI functions including G oups 0, 5, and 7 valves.-

0'hannel shared by RPS and Primary Containment 6 Reactor Vessel Isolation Control System.

A channel failure may be a channel failure in each system.

7.

A train is considered a trip system.

Two out of three SGTS trains required.

A -failure of more than one will require action A and F.

10.

There is only one trip system with auto transfer to two power sources.

Refer to Table 3.7.A and its notes for a listing oE Isolation Valve Groups and their initiating signals.

h channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPE1+BLE channel in the same trip system is monitoring that parameter.

rfgs f'R TABLE 3.2.c

~

t

~

1.

For th< st'irtup and run positions of the Reactor Node se)

< ct.or.

swi t;clt, th< r>> sha11 b<'wo o)>or<<hl.<. or tripp<'< trip s

.",t

< mn for

< <<<:h furtct.ion.

The

SRN, IRN ~

an<1 APRN {Sttartup mod"),

bio<:k:; nee<1 not he operable in>>Run>>

mode, and t)te Apl<N {t'Low biased) and RBM rod blocks need not be operable in

>>Startup>>

mode.

If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that. time the operable system is fu'nctionally tested immediately and daily thereafter; if this condition last lonqer than seven days, the system with the inoperable channel shall be tripped. If the fir'st column cannot be met for both trip systems, both trip systems shall be tripped.

2.

W is the recirculation loop flow in percent of design.

Trip level setting is in percent of rated power (3293 MWt).

.. A ratio of FRP/CMFLPD<1.0 is permitted at reduced powaer.

See Specification 2.

1 for APRM control rod block setpoint.

k 3.

IRN Qownscale is bypassed when it is on its lowest range.

(

4 ~

SRM<s A'and C downscale function is bypassed when IRM's A, C, E, and G are above range 2.

SRM's B and D downscale function is by-passed when IRM<a B, D, F, and H are above range 2.

SRM detector not in startup position is bypas: ed when the count rate is 0- 100 CPS or the above condition is satisfied.

5.

One instrument channel; i.e.

trip system may be bypassed be bypassed; Refer to Section core al,terationa.

one APRM or IRM or RBM, per except only one of four SRM may 3.10.B for SRM requirements during XRN charm el s A, E, C, G all

'A 6 C functions.

in range 8 bypasses SRM channels IRN channels B, F, D, H all in range 8 bypasses SRM channels 6

D functions.

7.

Thc Co following operational restraints apply to the HBM only.

Both HBM channels are bypassed when reactor power is ( 30$.

The HBM need not be operable in the >>startup>> position of the re<<ctor mode selector switch.

Two HLN channels are provided and only one of these may bc bypa".;ed from the console.

An HBN channel may bc out, of service for te::t;ing and/or maintenance provided this condition doc'ot lust longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any thirty day period.

If minimum conditiorts for Table 3.2.C are not met, administrative controls, shall be immediately imposed to prevent control rod withdrawal 74

TABLE 3 2 F SURVEILLANCE INSTRUMENTATION M.nimum

% of Operable Instrument Channels Instrument 46 A

LI-3-46 B PI-3-54 PI-3-61 PR-64-50 PI-64-67 TI-64-52 R-64-52 TR-64-52 TI-64-55 TIS-64-55 Lr-64 54 A LI 66 N/A N/A PS"64-'7 TR-64-52 and PS-64-58 B and IS-64-67 LI-84-2A LI-84-13A Instrument Reactor Mater Level Reactor Pressure Drywell Pressure Drywell Temperature Suppression Chamber Air Temperature Suppression Chamber Mater Temperature Suppression Chamber Mater Level Control Rod Position Neutron Monitoring Drywell Pressure Drywell Temperature-and Pressure and Timer CAD Tank "A Level CAD Tank "B" Level Type Indication and Range Indicator - (55" to

+60" Indicator 0-1200 psig Recorder 0-80 psia Indicator 0-80 psia

Recorder, Indicator 0-400~F Recorder 0-400~F Indicator, 0-400~F Indicator -25" to

+25" 6U Indicating

)

Lights

)

SRMg IRM, LPRM

)

0 to 100% power

)

Alarm at 35 psig

)

Alarm if temp.

)

281oF and

)

pressure

> Z.5 plf8) after 30 minute

)

delay

)

Indicator 0 to 100%

Notes (1)

{I)

(3)

(1)

(2)

(3)

(1)

(2)

(3 )

(1)

(2)

(3)

(1)

(2)

(3)

(1)

(2)

(3)

(1)

(2)

(3)

(1)

(2)

(3)

(4)

(1)

(2)

(3)

(4)

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRE) KNTS

. 3. A REACTIVITY CONTR01.5

~

4.3.A REACTIVITy CONTROLS 0

c.

c<1<l t rol rods M Ith sera<

t I>>>>'.<>

Sec<> tcr'hall cbook permit tcd

)>y Spec iflcn-tinn 3.3.C.3 nrc iuoper-

<lblc, but if they can be i<loertcd Mith control rod drive prcssure they need noe bc dioarc>cd electri-cally.

r d,

Control rods <dlth

<1 fal)cd "Ful1-in" or "Pull-out" position Jw I < ch ray be by passed in the Rod Sequence.

Cont ro1 Sy>> t cc>>>nd cons I-dered operable if the actual rod pool tio<l Io knol~.

These rods nu>>t be aoved In oequence to their correct posit!ono (full in on insertion or fully out on uithdraval).

Control rods Mith inopc rob I c accu:luliltofo or tho8a whose position cannot bc positively determined ohall be conoi-d c rcd in op crab le.

Inn pe rab Ic control rod a o ha il

'. be positioned such that Speci-ficatio<l 3.3.A.l io c>et.

In addition, durlnS reactor powa<

operation, no mere than one control rod in any 5 x,5 <>rray may be inoperablc

{<>t leaot opcrablc control rods t>uot aepo ra tc any 2 inoper<>b le ones).

If this Specifica-tion cannot be e>ct, ehe rc<>c-tor oh>ill not bd> otartcd, or if nt pourr, the renctor

<>hnlI bo brought to a shut-do>dn condition Mithin 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s~

Co d.

A second li'censed operator shall verify the conformance eo Specification 3.3.A.2.d before a rod may be bypassed in the Pnd Sequence Contrnl System.

)ihen it is ini.tial.Iy determined that a control rod is incapable of normtal inscrei<ln a teSt shall bc conducted eo demon-straee that the cause of t)lc malfunctinn is not a failure in the control rod drive mechanism.

If thin can l>e demonstrated nn att<'mpt to fully insert the control rod shall be made.

Xf the cnntrol rod cannot be inserted and an investipation has demonstrated that the cause of failure is not. a failed control rod drive mechanism collct hnusinp, a shutdown marpin test shall be m><cle to demon-strate under this condition that the core can be made subcritical for any reactivity condition durinp ehe remainder of the operatinp cycle with the analytically determined h iphes t worth control rod capable of.

withdrawal fully withdrawn, and all other control rods capable of insertion fully inserted.

The cnntrol rod accumulators

, shall be determined operable at least once per 7 days by

verifying that the pressure and level detectors are nnt in

'he alarmed condition.

II.

Control todn I

l.

Each control rod Shall be coupled to it<> driv or coc>pie eely inoar t cd and the 12l B.

CONTROL RODS 1.

The couplinp integrity shall h~

verified fn" each w't""dzn<~l contxol r u as foll<>

4

BRONNS FEts.".Y NUCLEAR PL'ANT F lGURE 3.~.2 K) FACTOR AUTOMATIC FLOW CONTROL MANUAL FLOW CONTROL Scoop-Tuba Sot-Point Calibration position such that Ffovrmax

= 102:S 'A 107.0 '/

112.0%~

117.0%

30 40 50 60 70 CORE FLOW,X 90 IOO

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.A Thermal and Pressurisntion Limitations 4.6.A Thcrral u<<d Prassuri.ratio<<

Limit.

~ t,< o<>a nr as fn<licntc d fn 3.6.A.A, during cool<1own following nuclear shutdown, or during low-level physics

tests, the reactor vessel temperatuie shall be at or above the temperatures of curve 82 of Figure 3.6.1 until removing tension on

. the head stud bolts as specified in 3.6.A.5.

30 Test specimens rcprc.-.c<<tir.. tn rc'<r ton vcs "cl, b<< "c we'<> und'cl.!

haut affected rona mat+i uh>al' j>>atul1cd iri the reactor ves c'd'acant totha vca"cl wall et thc core midplanc level.

Thc number urd type of specnen" wil3. bc in accor<lance " -.". 0:.

rcport <~0-10115.

Thc "pccincns chu13. neet the in-cnt of <STM 105-70.

Samples ahull bc <Cth-drawn at one-fourth and hrce-fourth scr rica life.

Thc reactor vessel shell temperatures during inservice hydrostatic or leak tasting shall be at or above the temperatures shown on curve 81 of figure 3.6-1.

The applicability of this curve to these tests is extended to non-nuclear heatup and ambient loss cooldown associated with these tests only if the heatup and cooldown rates do not exceed 15'F per hour.

4.

Neutron flux wires shall be ir.-

stalled in thc rcncto:

vc""e'dJuccnt to tha roue or vcs"c'all a!, thc core midpl;. level.

Thc wire" ahull bc rc<;ovc',und

'I

'I tc" tcd du."in<> tne.ir" rc'u"-ing outage to expcrincnta11y vari L.,

the culcula cd value" o.

n utror, fluence at one-fourth of hc bcltlinc shcU. thickness the=

ura used to dctcr4nc.thc l<D ':

shift from Fibre 3.6-Z, 6.

7.

The reactor vessel head bolting studs may bc part!ally tensioned (four sequences of tha seating pass) provided the studs and flange materials arc above 70'F.

Before loading the flanges any more, the vessel flange and head flange must be greater than 100'F, and must remain above 100'F while under full tension.

The pump in an idle recircula-cion loop shall not be started unless the temperatures of the coolant within the idle and operating recirculation loops are within 50'F of each other.

The reactor recirculation pumps shall not bc started unless the coolant temperatures between thc dome and the bottom head drain are within 145'F.

6.

7 ~

when thc reactor vcs"c'ead bolt,ing "tud

u. c tc <s< <,-c'

!hc reactor is in u <..olu tion, thc reactor vcs..c.'.".<:U.

temperature immediately c." o" tha head flange shrill bc c"-

mancntly recorded.

Prior to und during ate~Cup o!

an idle rccirculu ion loop, temperature of thc

! cuctcr cno'.-

unt i>> the operating and ii'c loops shaU.

bc perse<<cntly loggc<l.

Prior !.o s anting u rec':..u-tio<<p<n>p>

the reactor cdc.'t tc~>;><:ratu<ca in the do<><c <<<> head drain 'hall bc comp!rcd

<>ad pc=~>ancntly

':.'75

0

LIMITING CONDXTIONB FOR OPERATXON SURVEILLANCE REQUIREMENTS l.6 PRxHARY sYBTEM BoUNDARY

4. 6 PRIMARY SYSTEM BOUNDARY B. 'ool an t Chemistr Prior to startup and at steaming rates less than 100,000 lb/hr, the folloMing limits shall apply.

a.

Conductivity, Umho/cm525~C 2.0

b. 'hloride, ppm

0. 1 B.

Coolant Chemiotr A sample of reactor coolant shall be

.-:,... analyzed:

a.

at least every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months for

~..

conductivity and chloride ion content.

at least every 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months during startups, until the steaming rate is greater than 100,000

.lb/hr, for conductivit:y and

=hloride io:;

content.

c.

at least every 8

hours fox conductivity and chloride ion content uhsn the continuous

conductivity monitor is

'inoperable.

2.

At steaming rates greater than 100,000 lb/hr, the folio>ring limits shall apply a.

Conductivity, umho/cm925~C 2 0 b.

Chloride, ppm 0.2 176

44Pf+V go+

LIHITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

" l. 6 PR'iMARY SYSTEh1 BOUNDARY 4

6 PRIHARY SYSTEH BOUNDARY 0

3 ~

At steaming rates greater than t00,000 lbl'hr, the reactor water quality may exceed specification 3.6.B.2 only for the time limits specified below.

Exceeding these time limits of the following maximum quality limits shall be cause for placing the reactor in the cold shutdown condition.

a.

Conductivity time above umho/cm32'i~C-4 weel:sly'ar.

Hax'um Limit 10 pmholcmPv25nC b.

Chloride concentration time above

0. 2 pprn-4 wee)cs /yea r.

Haxitnum Limit-0.5 ppm.

177

LIMITING CONDITIONS FOR OPERATIO'N SURVEILLANCE RZQUIRHMNTS

3. 6,PRIMARY SYSTEH BOUNDARY 4.6 PRTtlARY SYST!'il BOUNDARY 0

When the reactor is not pressurized, except during

otartup, the reactor water shall be maintained within the following limits.

a.

Conductivity "

10 1!mho/cm9250C b.

Chloride - 0. 5 ppm 2.

During equilibrium power operation an isotopic analysis including guantitative meaourcmento for at least I 131 I

I-133 and I 130 shall be performed monthly on a coolant liguxd sample 178

w vs>~<

0

LIHITING CONDITIONS FOR OPE11LTION SURVEILLANCE REQUIREMENTS pnrrtARY SYSTEM BOUNDARY 4.6 PRIHARY SYSTEH BOUNDARY Mhe'never the reactor is critical, the limits on activity concentra-tions in the reactor coolant shall no t exceed the equilibrium value of 3.2 pc/gm of dose equivalent*

I-131.

This limit may be exceeded following power transients for a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

During ttris activity transient the iodine concentrations shall not exceed 26 pCi/gm whenever the reactor is critical.

The reactor shall not be operated more than 5 percent of its yearly power operation under this exception'for the equilibrium activity limits. If the iodine conceitration in the coolant exceeds 26 pCi/gm, the reactor shall be shut

down, and the steam line isolation valves shall be closed immediately.

That concentration of 1-131 which alone would produce the same tiryroid dose as the quantity of total iodincs actually present.

3.

Additional coolant samples shall be taken whenever the reactor

. activity exceeds one

'ercent of the equili-brium concentration specified in 3.6.8.5 and one of the followirrg conditions are met:

a.

During startup b.

Following a significant

. power change"*

c.

Following an increase in the equilibrium off-gas lev'1 exceeding 10,000 trodi/sec (at the steam jet air ejector) within a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period.

d.

Whenever the equilibri'um iodine limit specified in 3.6.8. 5 is exceeded.

The <<dditional coolant liquid samples shall be taken at 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months intervals for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , qr until a stable iodine concentration below the limiting value (3.2 rrci/

gm) is established.

However, at.

'least 3 consecutive samples ahall be taken in all cases.

An isotopic analysis shall be perforaert for each

sample, and quant.itat.ive measurements made to determine the dose equivalent I-131 concentration.

If the total iodine activity of tire sample is below 0.32 <<ci/gm, an isotopic analysis to determin equivalr'nt I-131 is not required.

For the purpose of this sect.i>>>>

on s-~rd r>>rr frequency, a significant power exchar~.~

defined as a charrgi exceeding 16 of power in less th.rr 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

\\

179

r LIMITINC CONDITIONS FOR OP RATION SURVEILLANCE REQVZREMi:NTS 3.6 PRIMARY SYSTEM BOUNDARY 6

PRIMARY SYSTEM BOUNDARY C.

Coolant Leaka e

Any time'irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212oF, reactor coolant leakage into the primary containment from unidentifi'ed sources-shall not exceed 5'prn.

Xn addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 g pm.

C.

Coolant Leaka e

1.

Reactor coolant system leakage shall be checked by the sump and air sampling system and recorded at least once per day>>

gith the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ..

Both the sump and air sampling systems shall be operable during reactcr power operation.

From and after the date that one of these systems is made or found to be inoperable for 'any

reason, reactor power operation is permissible only during the succeeding seven days.

Tha air sampling system may be removed from service for a period of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for calibration, functional testing, and maintenance ~ithout pro-viding a temporary monitor.

180

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS a.6.L

~J ~ t F

~

3.6.F Jet r).nr'ismatch b,

The indicated vs)ue uf ence flov rate varies fro~ thr value derived frere lnop flov uersurccrents by oorr:

than 1GX.

The di ffu'cr to lover pl< rr~LQ dlffcrcnticl prsssucc rend-ing on an individucl )et pucp varies free ct resrr of all 5ct purzp d'/rrrrri>>

ticl pressures hy more thar, 10X.

The reactor shc11 not be operated with one recircu'ation

. loop out of scrrJce fvr more than P!: hours.

With the reactor opcratiri-, if one recirculation loop is out of service, the plant shall be placed in a hot shutdown conditior, cri thin 24 hortrs uru.ess the loop is sooner ret!trued to service.

Z.

uhencvcc'here is rccirculsc'on flov with che reactor in chc Startup or Run Hodr and cnc -c-circuls cion punp ia operas!n i vith chc'qucli.czr vr lvc c)ried, the diffuoer to lover p)cnu~

diffcrcncial pccssrrre sholl checked hrrily'and the differ:n-tinl prcssure of an indivlr'.v~1

)et pump in c lrrr p shall not vary frcyI rhz rocsn of all pucrp dif".Crenti.al pccscures lu that loop by rssre, than 10i..

2,

)n).lo!4nS one pump oper.". ion, the discharge valve of.the low spr cd prr..p rray not bc opened u~'css the sp ed of thc foster puce is less than

~0,r of its rated speed.

3.

Steady state operation with both retirculatinn purrps out of ser-vice, for up to 12 hrs is per-mitted.

Ourihg such interval restart, of the recirculation, urrps is permitted, provided the sop discharge temperature is within 75oF of the saturation temperature of the reactor vessel water as determined by dome pressure,

.The total elapsed time in natural circula-tion and orre pump operation must he no greater than 24 nrs.

F.

Jet P~s~

Flov Hisrsstch 1.

Rccirculscion pucp speeds shill be checked and logged ac icssc once

'pcc day ~

2.

No additional surveillance required.

3.

Before starting either recirculation pump during steady state operation, check and log the loop discharge temperature and dome saturation temperature.

C.

Structural Inde.rit~

1.

The structural integrity of the primary system shall be G.

Structural Integritv 1;

Table 4.G.A together with sup-plementary notes, speci th 182

0'

E1H 1 7 1HC Cr...~ s T 1 ON5 FOR OP ERAT 1 ON SURVEIf.LANCE RE UXREHENTS 7.A Prfmnr Cones fnmenc

< ~ 7.A Prftnar Conrafnranc C.

Tuo dryucll-supprcssfon chamber vacuum breakers

'ay be decennfned to be inoperable Eor opening.

C ~

vslvon shall be exercised ieaoedistely and every 15 days th resf ccr unt i 1 the.

indpecsble valve has been

~ rc:turned to normal. service.

Once each operating cycle each vacuun breaker valve ehs11 be inspected for proper operacfua of the valve and limit suitchao.

d.

1 f speciEf cat fons

3. 7.A.4.a,

~.b, or.c cannot be mec, the unft nhnl1 bc placed in a cold ahucdoun condition in sn orderly manner uithin 24'ours.

5.

Or.

en Concentration d 0 5.

0 K

A leak test of the dryuell to suppreosion chsaber structure shall be con-ducted during each operating cyc)e.

Accept-able. leak rate io 0.14 ib/

sec of pritasry containment atoosphere uith 1 psi differential.

co Concentration a.

After completion of chc fire-related startup retescin

program, containment nteoop> ere shall be reduced to less than 4X oxygen vith nitro-gen gss during rcaccor pouer operation uith reac-tor coolant pressure above 100 psig, except as speci-

'fied in 3.7.A.5.b.

b, Mithin the 24-hour period subsequent to placing the reactor in the Run caode foilouing n nhucdoun, thc contafnaent atmosphere oxygen concentration shall bc rrhuccd co Lena than 4Z by vo].umo nnd enfntnfncd in chfa condition.

Dc-fncrc-.

ing easy commence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to s shutdoun b.

The primary containment oxygen concentration shall be measures and recorded daily.

The oxygen measure ient shall be adjusted to account for the uncertainty of the method used by adding a

predetermined error function.

The methods used to measure the primary containment oxygen con-centration shall be calibrated once every refueling cycle.

c.

1f npccfffcacfon 3.7.A.5.a and 3.7.A.5. b cannot bc ncc, an order?y 235 nhutdnvn ah 11 bc fnftfated and chc rnaccor aha if bc, in a Cold Shucdoun condition ufthfn 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

~

LIHITING CONDITIONS FOR OPERATION 3.9.h Auxiliar Electrical E ui ment SURVFILLANCE RE UXREHENTS 4.9.A Auxiliar Electrical E ui ment common trans-former capable of sup-plying power to the shutdown boards.

the specified time sequence.

c.

Once a month the quantity of diesel fuel available shall be logged.

b.

A fourth operable units 1 and 2 diesel generator.

4.

Buses and Boards Available a.

Start buses 1A and 1B are energized.

b.

The units 1 and 2 4-kV shutdown boards are energized.

c.

The 480-kV shutdown boards associated with the unit arnergized.

d.

Each diesel generator shall be given an annual inspection in accordance with instructions based on the manufacturer's recommendations.

e.

Once a month a sample, of diesel fuel shall be checked for quality.

The quality shall be within acceptable limits specified in Table 1

of the latest revision to ASTH D975 and logged.

d.

The Units 1 6 2 Diesel Aux Boards are energized e.

Undervoltage relays operable on start buses 1A and lB and 4-kV shut-down boards, A, B, C>

and D.

f.

Shutdown Busses 1

& 2 energized 5.

The 250-Volt unit and shut-down board batteries and a

battery charger for each battery boards are operable.

6.

Logic Systems a.

Common accident signal logic system is operable.

b.

480-V load shedding logic system is operable.

2.

D.

C Power System Unit Batteries (250-Vol.t) Diese'1 Generator Batteries (125-Volt) and Shutdown Board Batteries (250-Volt) a.

Every week the specific gravity and the voltage of the pilot cell, and temperature of an adjacent cell 'and overall battery, voltage shall be measured and logged, b.

Every three months the N

measurements shall be

'ade of voltage of each cell'o nearest 0.1 vol:~

specific gravity of e'a "h cell, and temperature of every fifth cell.

Th~se measurements shall be logged.

7.

" There shall be a minimum of 103,300 gallons of diesel fuel in the standby diesel generator fuel tanks, 293 A battery rated disch r'rapacity) test shal; t.

performed and the vol:.'ag~",

time, and output current measurement" shal.

b~

3o~<ed at

~nt~". a'"

n.

exec.d 8,

1 t

0

L1HITIHH cnHHITfoHs FoR opYRATIoN SURVEILLANCE RE VIREHENTS

3. 10 CARY, ALTERATIOHS igc10 CORE ALTERATIONS A

lie<<bilit A

1icabi lit Applies to thc fuel handling and core reactivity ligfgitationa.

Applicgg to the periodic testing of those interlocks and instru-mentation used during refueling and core alterations.

O~bac t'ive To ensure that core'reactivity is within the capability of the control rods and to prevent criticality during refueling.

~Ob ective To verify the operability of instrumentation and interlocks used'n refueling,and

core, alterationse S ccification A.

Rcfuclin Interlocks Refuelin Inter locks 1.

The reactor vade switch shall be locked in the "Refuel" position during core alterations and the refuel ing interlocks shall be operable except as specified in 3clO.A,5

<<nd 3.10.A.6, below.

1.

Prior to any fuel hand-ling vith the head off the'eactor

vessel, the refueling interlocks shall be functionally tested.

They shall be tested at weekly inter-vals thereafter uncil no longer required.

Ihey shall alar) be tested fol-loving any repair work associated with the inter-locks.

2

~

Fuel shall not be loaded into thc reactor core unless all control rods are fully inserted.

2.

Prior to" performing con-trol rod or control rod drive maintenance on con-trol cells without fl rcngoving fuel asseffgblics, it shall be demonfgtrfgted that the core can be

<<fade subcrit ical by a eergie ef 0.38 Cerce c eft/k at any tiege during the'aintenance with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.

Alternave tivcly it the remaining 302

I >~II Isg CP;<<iITIpv< Fpa OPFita.!O'I J. lO.A RcIi/eII/iz. In'.<rip

~s 4, lp.h Re(uc I in Inc e. locks I

F control rode are fully inserted and have had their direccional con-trol valves electrically disabled I it i.s auffi

cient, co d erron>> t ra t e that the core is aub-cr$ tical arith a eiarrfn o/

t le et 0.38 tlerc.ent dk/k

.'any tine dur JnR the eiaintenance.

h control rod on Mhich ciaintenance i>> bcinII perforated shall be considered inopcrabl,e.

).

The fuel Rrapplc hoist load sMitch shall be >>et at,

< 1,000 lbs.

3.

No additional surveillance required.

I ( the (rane-iiiounted auxi-liary hoist, thc eanorsii-aounted curlllary hoist, or thc service 'plat(o~ hoist is to be used (or handIInF, (uel vith thc head c(f the reactor vessel, thc load i(nit suitch on thc hoist to be used sha.'1 be sat at

< 400 lbs.

4.

"No additional surveillance I

required'.

h a.axInuw o( tvo nor.-

ad]occnt control rods nay bc uithdrsi~ tron the core for thc purpose o( pcr(or-w(nt cont rol roC an4/or control rod.drlv>> nainten-

ance, provided thc (ollov-inR conditions are satio-(ied:

5.

No additional surveillance required.

I a ~

The reactor node ai/Itch

~ ha I I be locked In the "re(uel" pcs I t ion.

The rc(ueIIng Interlock vhich prevents aiore than one contro'I rod (con beinR uith4rcm may be bypassed (or one of the control r'oC>>

on.Mhich ma in tenancy i s bc In F, performed.

All other 303

C}

0

f.'I>IIIll:I; rounlTloas vr>n or utATIoa SURVKII.DNCK RF. UIRDtPHTS 3.1D.A Nefuellnli In>mrlnrk>>

rrfunl I ng I nt nr I or Its

~ hnll be operable.

b.

A euf f Iclen>

number of control rods shall be npernble so that the core can ba madr cub-crit I c nl vil.h the atrongrst oprrnblo con>>

trol rod fully vith-dravn nnd

~ 11 o the t'perable control rods fully inserted

~ or al)

,. directional control valve> for rem>>lninF, cont rol ro>ln nhn ll bc dian>>sed electrical)y nnd nuff lclcnt margin" lo critlcnllty shall ba da>nonstrsted.

c.

If mnintenar>re is to be pcrfor>>>ed on l,vn control rod drives they must be acparnl.ed by mire 'ti>an tvo control cells in any direction.

d.

An approprinte number of Sibyl' nrr available ns def incd in spec ifI-cation 3.10.8.

'G.

Any numb> r nf control rods mny be vt(h>lrn>s> or remnvcd, from the rcnctr>r cora, pro-

'iding the folloving condi-tiono are satisfied:

a.

Tha reactor

>>>ode svitch ia loc'aed in the "re-fuel" position.

The refueling int.er lock vhic'h prevents e>orc than ona control rod from 6;

Pith the mode seleL.tor switch in the refuel mode, only one control rod may be withdrawn until two licensed operators have confirmed that all fuel has been removed from around the next rod to be withdrawn.

i

LIMITING CONDITIONS FOR OPEPAT1ON SURVEILLANCE REQUIREMENT 1.

1 1

FI l>.":

l>1<OT ECTIOH

! > YSTF~MS il. 11 FIRE PROTECTION SYSTEMS Th-,

CO, Fire Protection SyStem shall be operable.

with a minimum of 8-1/2 tons

{0. 5 Tank)

CO.

in sto age unit=

1 and 2 ~

with a mxnxmum of 3 tons

{0.5 Tank)

COg storage unit 3.

ci' I 1 1 O IAa I J. (:

initia"ion logic operas'e.

I f speci 1:%cation 3.11. 8.1.a or 3.11.0.'1 b or

3. 11. B. 1. c cannot be

met, a pa trolling f-ire watch with portable fire equipment shall be establis'ned to ensure that each area where protection is lost is checked hnu r ly.

CO Fi. re Prntection System B

CO

~ Fire Protection System CO~ Fare Protect@on Testing:

I tern a.

Simulated automa tic and manual actuation F>.au-n, One&yea

-b. Storage

. Checked tank daily pressure and level c.,

CO q Spray Once/ 3 header an'd

'yea'r.s nozzle inspection for blockage 2.

When the cable spreading room CO 1 Fire Protection is inoperable, one 125 pound (or larger) portable fire e:(tinguisher shall be placed at each entrance.

3.

If sp(ciiicati.ons

3. 11. H. 1..<,

3.11.B. l.b, or

3. 11. D. l. e a cot met wi t,h.i.n f day >,

the al, l.ected unit(s) shall l.'.e in cold shutdown within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months

'19

~,

6.0 A--:~INISTRATrrjE CONTROLS 0

B-Sou rce Tes'ts Results of required leak tests performed on sources if the tests reveal the presence of 0.005 microcrrrie or more of removable contamination.

C ~

Sof ci r 1 R~~>sets (irr writing to the t)ir<<ctor of R{.<gional Office of Irrspection

<and Enforcement).;

Reports on the following ar<<<is shall be submitted as noted:

b.

Fatigue Usage Evaluation 6.6 a.

Secondary Containment 4.7.C Leak iRate Testing (5) within 90 days of cornpLetion of each test.

'nnual Operating RepoM.

c.

Seismic.1nstrumentation 3.P.J.3 Within 10 dai s Inoperabi lity a f'trr 30 days 0f i no(tr.r ab i

'1 i ty 0

D.

d.

.Relief Valve Tailpipe Instrum{*.ntation 3.2.F Within 30 days after inoperability

,of

.hermocouple an acoustic monitor on one valve.

d 1

3.2.I,2 Within 10 days after / day~ of inoperability e ~

'Heteorological l<onitoring Instrumentatiorr Inoperability E

~Seconal R~eort {in oritict to the Director oi Retionht Office of Inspection and t'.nforcerncnt)

Data shall be retrieved from all seismic instrurtrents actuated during a seismic event and analyzed to determine, the magnitude of the vibratory ground motion.

A Special Report shall be submitted within 10 'days after the event describing the magnitude, frequency. spectrum, and resultant effect upon plant features important to safety.

356

PROPOSED CHAhGES TO APPENDIX B TFCHNICAI SPECIFICATIONS

Hctcoro1ogkcal data shall be sununarixed ard reported consistent vith the recce;.endations of Regulatory Guide 1.21 (June 1974) and'egulatory Cuidq 1.23 (February 19t2),

and meteorological ybservatiors shall be recorded in a fore consistent vith National Heather Service procedures.

If'he outage of any meteorological instrtraent(s) required by Regulatory Guide 1.23 (Fcbruar) 1972) exceeds seven consecutive days, the total outage time, the dates of outage, the cause + the outage, nnd-the instru>>.

ment(") involved shall be reported vithin10'Rays of the initiation of the outage to the VShRC, Office of Inspection and Enforcement

~ vith a copy to the Office of Nuclear Reactor Regulation, Division of Operating peactors.

Zlc~=nts of this progran may be modified or terminated in accordance vith Subsection 5.6.3(c).

".".Te collection of'eteorological data at the plant site, provides information for usc in developing atmospheric diffusion parameter" f'r estimate.icg potential radiation doses'o the public resulting from actual routine or,'bnormal releases of radioactive materials to the y.tmo pherc, and for a" sessing the actual impact of the plant cooling system on the atmospheric cnvironncnt of'he ite area.

A n.cteorological data collection prcgraa as,. dcscrib d above is accessary to meet the requirements of. subparagraph 50.36a(a)(C) of'0 CFR Part 50, Appendix D to 10 CFR Part 50, and Appendix E to 10 CFB Part 50.

(3)

Ref. Section 6.7.3.D Appendix A Technical Specifications.

I 5-5-3

'Mrittcn procedures described in Section 5.5. l shall bc r<<vl<<ued l>y PORC

>llul approved by thc Plant Manager prior to implr>>><<station.

Temporary rh>>ng<><

a procedure which do not chango the intent of thc pla>>t st.nf f k>><>ul<<lgat>l<>

in tho nrcn affected by the proccdurc and rh<. additional nl>prov.>l

<>f a semi>> r of thc plant staff who holds a Senior Reactor Operator license.

Su<'h

<'.l>>>>>j;v>>

shall be documented and subsequently reviewed by PORC and <<pproved by tl>>>

Plant Manager.

e 5.6 Re ortin Re uirements 5.6.1 A report shall be prepared by Fnvironmentnl Compliance and submitted to DNP following the end of each 12-month period of operation, which shall su>>m>arise the results of the nonradio logical environmental monitoring program.

5.6.2 Routin Re ortin a.

A nummary rcport shall be prepared for both the inplant monitoring program and the nonradiologicnl monitoring programs and submitted to thc Director of Division of Operating Reactors, NR(, as part of the Annual Operating Report within 120 days after December 31 of each year.

aad>olo leal anvixonne'otal H~on>tox>n Routine Re ortin Reporting Requirements:

1.

TVA shall prepare a report entitled "Environmental Radio-activity Levels - Drowns Perry Nuclear Plant - Annual Report."

The report shall cover the previous 12 months of operation and shall be submitted to the Di.rector of the NRC Region II Office (with a copy to the Director, Office of.Nuclear Reactor RHgulation) within 120 days after January 1 of each year.

The report format shown in Regulatory Guide 4.S Title 1 shell be used.

,The report shall include summaries, interpretations, nnd evaluations of the results of the radiologi cnl environmental surveillance activities for the report period, including n comparison with preoperationnl studies and/or operational controls (ns noprn-priate),

and an assessment of the observed impacts of the plan'0 operation on the environments.

If harmful effects or evidence of irreversible damage nre detected by thc mo<<in< 'ing-the licensee shall provide an analysis of the'problem r>d a proposed course of action to nllcvinte the problem.

UNIT 2 PROPOSED CHANGES

PROPOSED CfODGES TO APPENDIX h TECHNICAL SPECIFICATIONS

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1-1 FUEL CLADDING INTEGRITY F 1 FUEL CLADDING INTEGRITY A

> licabilit A licabilit Applies to the interrelated variables associated with fuel thermal behavior.

~cb 'ective To establish limits which ensure the j.ntegrity of the fuel cladding.

S ecifications A-Thermal Power Limits

1. Reactor Pressure

> 800 psia and Core Flow >

10%

of Rated.

When the reactor pressure is greater than 800 psia, the existence of a minimum critical power ratio (MCPR) less than 1

~ 07 shall constitute violation of the fuel cladding integrity safety limit.

Applies to trip settings of the instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.

~Ob ective To define the level of the process variables at, which automatic protective action is initiated to prevent the fuel cladding integrity safety limit from being exceeded.

S ecification The limiting safety system settings shall be as specified below:

A-Neutron Flux Trip Settings ApRM Flux Scram Trap Setting (Run Node) a.

When the Mode Switch is in the RUN position, the APRM flux scram trip setting shall be:

S< (0 66W

+

54%)

where:

S = Setting in per-cent of rated thermal power (3293 MHt)

W = Loop recircu-lati:on flow rate'n per-cent of rated (rated loop recirculation flow rate equals 34.2x10+ lb/hr)

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.

1 FUEL CLADDING INTEGRITY l

2.1 FUEL CLADDING INTEGRITY b

In thc event of operation with the core maximum fraction of limiting

~

power density (CK?LPD) greater than fraction of rated thermal pover (PVJ')

the setting shall be modified aa follows:

S+ (0.66W + 54X)

FR'Bl.P D t

F r no combination of loop recircu-c, or n lation flow rate and core thermal power s a

ver shall the APRM flux scram trip

~F setting be allovcd to exceed 120~

, of rated thermal pover.

O (Note: These settings assume operation within the basic thermal hydraulic design criteria. These criteria are LHGR<18.5 kw/ft for 7x7 fuel and ~13.4 kw/ft.for 8x8 8xgR.

and P8xgR>

and MCPR within limits of Spccificatinn 3.5.k. If

.(t is determined that r (ther of these design critcr.'a is being v'olc. cd during opcrat ion, action shall bc initia ed within 15 minutes to rc tore operation within prescribed limits Surveillance requirements foz APRM scram setpoint are given in specif ication 4.1.B.

d, 'he APRM Rod block trip setting shall be:

s (0.66H

+02%)

BB where:

Rod block sett'.ng in percent of rated thermal power (3293 Mwt)

Loop recirculation flow rate in percen of rated (rated loop recirculation flow rate e luals,

34. 2 x 10'.b/hr)

SAFETY LIMIT LXMITING SAFFTY SYSTEM SETTING I

1.

1 FUEL GLADDING INTEGRITY 2.1 FUEL CLADDXNG XNTEGRITY Xn the event of operation with the core maximum fraction of limiting power density (CHFLPD)'reater than fraction of rated thermal power (FR~P

~ the setting shall be modified as follows:

0~6

{0.66W +02%

) FDF CHFLFD 2.

Reactor Pressure

<<800 PSIA or Core Flow 10% of rated.

When the reactor pressure is "800 PSIA or core flow is "-10% of rated, the core thermal power shall not exceed 823 MWt (~5/ of rated thermal power).

2.

APRM and IRM Trip Settings (Startup and Hot Standby Modes).

a..APRM-When the reactor mode switch is, in the STARTUP position, the APRM scram shall be set at less than or equal to 15% of rated power.

b.

IRM The IRM scram shall be set at less than'or equal to

'l20/125 of full scale.

10

SAFETY LIMIT LXMITXNG SAFETY SYSTEM SETTING 1.1 FUEL CLADDING XNTEGRXTY 2.1 FUEL CLADDXNG XNTEGRITY B

Power Transient;

.S.

Power Transient Tri 'Settin s

To ensure that the Safety Limits established in Specification

>.l.n are not exceeded,

'ach required scram shall be initiated by its expected scram signal.

The Safety Limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.

Scram and isola-(PCIS groups 2

3 6) reactor low

~ater level 2.

Scramturbine stop valve closure I

Scram--turbine control valve fast clo"ura or turbine trip.

Scram--low con-denser vacuum 5, 'Scram--main steam line isolation 2 538 in.

above vessel zero S

>0 per-

'ent, valve closure 550 psig 2 23 inches Hg vacuum 5

t.0 per-cent valve

.. closur~

6.

Main steam. isola-tion valve closure

--nuclear system low pressure

~825 psig C. Reactor Vessel Water Level C. Water Level Tri Settin s

Whenever there is irradiated fuel.in, the reactor vessel,

~ the water level shall not be less than'7.7 'in. above the top of the normal active fuel zone.

Core spray and LPCI actuation

'reactor lofti 'water.

level 2.

HPCX and RCXC actuation reac-tor low water level Hain steam isola-tion valve closure--r eactor low water level 2 378 in.

above vessel zero 4'70

'bove vessc

)

zero 470 Xn.

(

above ves ~el zero

2'. 1 SASES tron (uel damage, assuming a steady-state operation ac the crip setting', over the entire recirculation I foM range.

The margin to the Safety Licit*fncreases as the flou decreases (or the spi cf(fcd r rip setting versus f lou rc lationship; therefore, the uorsr case HCPR Mhich could occur during s eady-state operatfon fs

~t 108X ol rated the roal pover because of the APRH rod block trfp setting.

The

~ccual poi'er dfscrfbuc loci In the core fs cstabl fshed by specf. fed control rod sequences and fs Nionfrorrd continuously by the in-core LPRH system.

As Mith the APRH scram c rip ss c t fng, the ApRH rod block trip set t fng fs adjusted dovni ard lf the CMFI PO exceeds Flap thus preserving che

!PRN ref block safety iiargfn.

C.

Reaccor Mater Loi>> Level Scrim and Isolacfon (F>>ice t thin Steamlfnes)

The set point for the lo>> level scram fs above the bottom of the separator skirt.

Thfs level has been used in transient analyses dealing vich coolant fnvenrory decrease.

The results reported in FSAR subsection 14.5 shcu thar scram and fs'olatfon of all process lines (except vain stean) at this level adequately protects the fuel and the pressure barrier, because HCPR is greater than 1.07 in all cases, and systeo pressure does not, reach the satcry valve settings.

Tiic scram setting is approximately 31 inches belou the normal operating range and fs thus'adequate ro devoid spurfous scr~.

The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves.

With a trip setting of 10">> of valve closure fromm full open, the resultant increase in heat flux is such that adequate thermal margins are maintained

'even during the worst case transient that assumes the turbine bypass valves remain closed.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Tri "Scram Turbine control valve fast closure or turbine trip scram anticipates the

pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.

The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraplic control oil pressure at the main turbine control va'lve actuator disc dump valves.

This loss of pressure ic sensed by pressure switches whose contacts form the one-out-ot'-two-twice logic input to the reactor protection system.

This trip setting, a nominally 50': greater closure tlnle.and a different valve characteiisti'c from that of the turbine sto'p valve.

combine tv produce transients very similar to that for the stop valve.

iso signifi-cant change in t<CPR occurs.

Relevant transient analyses are discussed

=in References 2 and 3 of the Final Safety Analysis Report.

This scram is bypassed when turbine steam flow is below 30.". of rated, as measured by turbine first state pressure.

i 23

0

t.

Hain Condenser Lou Yacuuo Scram To protect the aein.condenser ag~inst overprcseurc, a loss of con-denser vacuun initiates autoactlc closure af the turbine stop valves.

To anticipate the transient and autoaatic ocran resulting fron the closure of the turbine stop>>alves.

lou con-denser vacuvcs inihiatea a acres.

The lov vacuua scram oat point ie oeiected to initiate 4 ocraa befc 'e the closure of the turbine stop valves is initiated.

C. C'R.

Hain Stoaa Line Ia~ stion on Lov prcssure and Hain Steam Line Isolation Scrsn

'Ae lou pressure isolation of the csin stean lines at 825 psig ass

, provided to protect 'a6ainst rapid reactor depressurixation and the resulting rapid cooidovn of the vessel.

Advantage is taken of.the ocraa feature that occurs shen the aain etecua line isolation va?vee are closed, to provide'for reactor shutdovn oo that high povar opera-

,, tion at lcw reactor prooaura does not occur, thus providing protection for the fuel cia'dding 'integrity safety liait.

Operation of the reac-tor at pressures 1ovei than 82> paig requires thit the recctor code evitch ba in.thi.STAR'NP position, vheze protection of thc fueL cladding integrity safety lioit 'ia provided by'the IRH and APRN high neutron flux

~ ocraas.

Thus, the coabination of caaln ataxia )ine lov pressure isolation end isolation valve closure scree assures the availability o! neutron

~

dlux scram protection over ths entire range of applicab'lity of the fu I cladding integrity safety Limit.

In addition, the isolation valve closure scree anticipates the pressure.and flux transients that occur during norzeal or inadvertent isolation valve closure.

Mith the sera~

set ot 10 percent of valve closure, neutron flux does not increase.

TABLE 3 1 A REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMEN1ATIOR REQUIREMENT Min. No.

of Operable Inst Channels Per Trip 1

Mode Switch in Shutdown 1

Manual Scram IRM (16)'igh Flux Inoperative Tri Level Settin S 120/125 Indicated on scale Modes in Which Function Must Be erable Shut Startup/Hot uceeu

~Refuel 7

~eteudb X(ZZ)

X (22) uuu

~tuttut e

X 1.A X

1eA (5) 1.A (5) 1 A 2

tet 2

2 2

APRM (16)

High Flux High Flux Inoperative Downscale See Spec.

2.1.A.1 S

15% rated power (13)

I 3 Indicated on Scale X (21)

X (21)

(11)

X1Aor1B X(17)

(15) 1.A or 1.B X(17)

- X 1.A or 1.B (11)

X(12) 1.A or 1.B High Reactor Pressure

< 1055 psig X (10)

X X

1,A High Drywell Pressure (10)

Reactor Low Water Level (10)

High Water Level in Scram Discharge Tank 2.5 psis h 538" above vessel zero S 50 Gallons X (8)

X (2)

X(8)

X 1eA X

1eA X

1,A

TABLE 3 1 A REACTOR PROTECTIOH SYSTEM (SCRAM)

XHSTRUMEHTATIOH REQUIREMEHT Min. Ho.

of Operable Inst.

Channels Per Trip Tri Level Settin Modes in Mhich Function Must Be O erable Shut-Startup/Hot deem

~Ref ne1 7

~stands ann

~antann 1

I" 2

Main Steam Line Isola-tion Valve Closure e

Turbine Cont Valve Past Closure <<

Turbine Trip 5

10% Valve Closure k 550 psig 1.A or 1.C X(4) 1.A or 1.D X(3) (6) -

X(3) {6)

X (6)

Tu bine Stop Valve closure 5

10% Valve Closure X(4) 1.A or 1.D Turbine First Stage Pressure Permissive.

Turbine Condenser Low Vacuum Main Steam Line High Radiation (14) not h 154 psig I 23 In. Hg, Vacuum 3X Hormal Full Power

Background

X(18)

X(3)

X{9)

X (18)

(19)

X(3)

X lAorl C X{9)

X(9) 1.A or 1.C

0

l0.

Not required to bc operable when the reactor pressure vessel head is not bolted to the vessel.

11.

The ARRL downscale trip function is only active when the reactor mode switch is in run.

12.

The APRM downscale trip is automatically bypassed when the IRM.instrumentation is operable and not high.

13.

Less than 14 operable LPRM's vill cause a trip system trip-1q.

C anne s

r

.h 1 shared by Reactor Protection System and Primary S stem.

A Containmen an t 'nt and Reactor Vessel Xsol-t.ion Control y

channel failure may be a channel failure in each y

s stem.

15.

The APRM 15$ scram is bypassed in the Run Mode.

Reactor Protection System and Reactor Hanual Contxol 1 f il b

h 1 f il o ).

A channel a

ure m

y (o

s stem.

If a channel is allowed to be hat same channel may be inoperable in thc corresponding function, in that same c ann the Reactor Manual Control System, 17.

Hot required while performing low power p y h si cs tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).

18 Tli fu ction must inhibit the automatic yp b

assin of turbine control 1 s Unc tri scram and tuxbine stop valve c osure 1

valve fast closure or turbine tr p scram an

~ nor eual to scram whenever tur ne irs

~

bi f'

stage pressure is greater than or equa 15') psig,

'I

19. Action 1.A or 1.0 shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function.

Otherwise, no action is required.

'i a set oints for alarm and reactor trip (1.5 and 3.0 times

, blished based on the normal background background, respectively) are csta s

e ase tri are at full power.

T e a owa h

11 bid setpoints for alarm and reactor tr p are 1,2-1'.8 and 2.4-3.6 times background, respectively.

do not have to be operable 21.

The APRH 1iigh Flux and Inoperative Trips do not ha in theRcue oc f

1 M d if the Source Range Monitors are connected to give a non-coincidence, ig e

Hi h Flux scram, at 5 x 10 cps; The a

1 c operable per Specifica shorting links is required to provide non-coincidcncc g -

ux protection from the Source Range Monitors.

anneI. is not required in the 22'he three required IRM's pex trip channeI. is S

o o

Re uel o

e a

quadrant) are connected to give a non-coinci ence, g

non-coincidence high-flux scram protection from the IRH's.

xn an inoperable status for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for 23.

A channel may be placed in an no required surve anc illance without placing the tr p sys PERABLE channel in the same trip condition provided at least one OPERABL c ann system is monitoring that parameter.

TABLE

- REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENTATION FUNCTIONAL TESTS

!iZNIMUM FUNCTIONAL TEST FREQUENCIES FOR SA=-TY IiNS'R. AND CONTROL CIRCUITS Group (2)

Functional Test Minimum Frequency (3)

Main steam Line Isola=ion valve closure

'iirbine Control Valve Fast Closure or Turbine Trip Trip Channel and Alarm Trip Channel

,.nd Alarm Once/Month (1)

Once/Month (I)

Turbine First Stage Pressure Permissive Turbine Stop Valve Closure Trip Channel and Alarm Trip Channel and Alarm Every 3 Months Once/Month (1)

TABLE 4-1-B REACTOR PROTECTION SISTEM (SCRAM)

INSTRUMENT CALIBRATION MINIMQM CALIBRATION PREQQENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS, Instrument Channel IRM High Flux APRM High Flux Output Signal Flow Bias Signal LPRM Signal High Reactor Pressure

'High Drywell Pressure Reactor Low Rater Level High Rater Level in Sczam Discharge Volume c Turbine Condenser Low Vacuum Main Steam Line Isolation Valve Closure Main Steam Line High Radiation Turbine First Stage Pressure Permissive Group (1 )

B B

Calibration Comparison to APRM on Control-led stirrups (6)

Heat Balance Cal brate Flow Bias Signal (7)

TIP System Traverse (8)

Standard Pressure Source Standard Pressure Source Pressure Standard No e (5)

Standard Vacuum Source Note (5)

Standard Current Source (3)

Standard Pzessuze Source Minimum.Frequency (2)

Note (4)

Once every 7 days Once/operating cycle Evezy 1000 Effective Full Power Hours Every 3 Months Every 3 Months Every 3 Months Note (5)

Every 3 Months Note (5}

Every 3 Months Every 6 Months Turbine Stop Valve Closure Note (5}

Note (5)

NQTEs FoR TABLE 4. 1. B 1.

A description of three groups is included in the bases of t:hi.t "p<".cification.

2.

Calibration are not required when the systems are not, required to be operable or are tripped.

If calibrations are

missed, they shall be performed prior to returning the system to an operable status.

3.

The current source provides an instrument channel alignment.

Calibration using a radiation source shall be made each refueling outage.

4-Requited frequency is-initial startup following each refueling outage.

6.

On controlled startups

~< overlap between th will be verified.

5.

Physical inspection and actuation of these position switches'ill be performed once pei operating cycle.

e IRM's and APRM's

-7.

The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compaxed.

The flow comparator trip and upscale will be functionally tested according to Table 4. 2.C to ensure the proper operating during the operating cycle.

Refer to 4.1 Bases for

- further explanation of calibration frequency.

h complete tip system ".r'averse calibratcs the LPR'. signals to the process computer.

The individual LPR."! meter readires will be adjusted as a minimum at the beginning of each operatinp cvcle before reaching 100/ power.

3,f oases modes.

In the pover range the APRM Ref. Section 7.5.7 PSAR.

Thus, the Run mode.

The APRM's aad the IR l's startup and intermediate range.

system provides requ'red prateccion.

IRM System is not required in the provide adequate coverage in the The high reactor pressure, high dryvell pressure, reactor lov vater level and scram discharge volume high level scrams are required for Startup and Run mades of plant operation.

They are, therefore, required to be opera-tional for these modes of'eactor operation.

The requirement to have the scram functions as indicated in Table 3.1.

1 operable ia the RefueL mode is to assure that shifting to the Refuel mode during resctor pnvcr opcratioa does not dimiaish the need for the reactor protection system.

The turbine condenser Lov vacuum scram is only required during paver operation and must be bypsaoed to start up the unit.

Belov 154 psig tur-bine first etage pressure (30Z of rated),

the scram signal due to turbine op "

"e osu e ~

and turbine control valve fast closure is bypassed because flux and pressure scram are adequate to protect the reactor.

Because of the APRM dovnscale limit of 3X vhea in the Run mode and high level limit of

< 15Z vhen in the Startup

Mode, the transition batveen the Startup and Run Modes must be made vith the APRN instrumentation indicatiag betveen 3Z and 15Z of rated paver or a control rod scram vill occur.

In

addition, the IRN system must be indicating belov the High Flux setting

{120/125 of scale) or a scram vill occur vhen in the Startup Mode.

For normal operating conditions, these limits provide assurance of over'lap II lt betveen the IRM system and APRt system so that there are no 'apa in the pover level indicatione (i.e.,

the pover level is continuously monitored

'.rom beginning of startup to full paver

<<nd from full pover to shutdovn).

v'hen pover is being reduced, if a transfer to the Startup mode is made snd the IRM's have not been fully inserted

{a msloperstional but not impossible condition) a controL rod block immediately occurs so that reactivity ~sar" tion by control rod vithdraval cannot occur.

I,IXITINCP CONOITIONS FOR OPERATION 3.r.lt Ylo c prot ctl*

The unit shall be shutdown and placed fn the cold condition when Mheelor Reservoir lake stage rises to a levol such that water fro>>> the reservoir begins to run across.the pu>>>ping station deck at elevation 565.

SlfRVEILLA.'fCZ RE VIREXEHTS 4 ~ 2.H Flood Protection Surveillance shall be perfomet!

on the instrumentation that e>onitors the reservoir level as

.indicated in Table. 4.2.H.

Require>aents for instru>3>entation that e>onltozs the reservoir level is given in Table 3.2.H.

3.2.I Hctcorolocl col. I~tonitorin Instrumentation 4.2.I Hetcoro'>>:;ical Monitorin Instrumentation The meteorological monad oring instru-mentation listed in table 3.2.I shall be operable at all times.

1.

With the number of operable met.eo1n! o!3,'. cal monitor in,v, rhannel>>

less than acquired by tabl 3.2.I, restore the inoperable channel{")

to operable status within 7 days.

Each meteorologic& moni orir~ inst'=eat channel sha33 be demonstrated ooerable by the perfo~rc o" the CMI>H CH:-CK at least once per 24 hou"s and th" C!'.VP>c.

CALIBMTIOifat leas once each 6 =oaths.

2.

With onc or more of the meteoro-logical monitoring channels inoperable for morc than 7 days, 1>rcparc

<<nd submit a Special Bcport ti tf>e Commission, f>>1rsuant to Speci f'Icntf.on f'>.7.3.C within the next.

10 days outlining the cause of the malfunction and the plans for restoring the system to operable st 1 1'>is ~

53

TABLE 3 ~ 2 A PRIMARY CONTAINMENT AND REACTOR BUILDING XSOLATION INSTRUMENTATXON Minimum No.

Operable Per Tri S s 1 (ll) 2 Function Instrument Channel-Reactor High Pressure Instrument Channel-Reactor Low water Level (6}

Tri Level settin

.Action 1

100

+

15 psig

> 538" above vessel zero A or (B and E)

Remarks 1.

Below trip setting does the following:

a Initiates Reactor Building

. Isolation b.

Initiates Primary Containment Isolation c.

Initiates SGTS 1.

Above trip setting isolates the shutdown cooling suction valves of the RHR system.

Instrument Channel-Reactor Low Water Level (LXS-3-56A-D~

SW 41) 470" above vessel zero A

1 ~

Below trip setting initiates Main Steam Line Xsolation Instrument Channel-High Drywell Pressure (6)

(PS-6c}-56A-D)

S 2.5 Psis A or (B and E)

Above trip setting does the following:

a.

Initiates Reactor Building Isolation b.

Initiates Primary Containment Isolation c.

Initiates SGTS 2 (3)

Xnstrument Channel-High Radiation Main Steam Line Tunnel (6)

Instrument Channel-Low Pressure Main Steam Line Instrument Channel-High Flow Hain Steam Line 3 times normal rated full power background 2

825 Psig (u}

5 140% of rated steam flow

- B Above trip setting initiates Hain Steam Line Isolation 1.

Below trip setting initiates Hain Steam Line Isolation Above trip setting initiates Hain Steam Line Isolation

TABLE 3. 2.A PRIMARY CONTAINMEMP AND R=ACTOR BUILDING ISOIATION INSTRUMENTATION Minimum No.

Operable Per 2

Instrument Channel-Main Steam Line Tunnel High Temperature 2000F Leve t

n Act on 1

rks Above trap setting snit?ates Main Steam Line Isolation.

Instrument Channel-Reactor Watez Cleanup System Plooz Drain High Temperature Instrument channel-Reactor Water Cleanup System Space High Temperature Instrument Channel-Reactor Building Venti-lation High Radiation-Reactor Zone Instrument Channel.-

Reactor Building Venti-lation High Radiation-Refueling Zone 160 180oF 160 - 180oF S

100 mr/hr oz downscale G

100 mr/hr or dovnscale F

1.

Above trip setting initiates Isolation of Reactor Water Cleanup Line from Reactor and Reactor Water Return Line.

1.

Same as above 1 upscale or 2 downscale will a.

Initiate SGTS b.

Isolate reactor zone and refueling floor.

c.

Close atmosphere control system.

1.

1 upscale or 2-downscale will a.

Initiate SGTS b.

Isolate refueling floor.

c.

Close atmosphere control system 2 (7) (8)

Instrument Channel SGTS Flow - Train A Heaters charcoal Heaters S 2000 cfm H and 1.

R.H. Heaters

< 2000 cfm

{A or P) 2 ~

Below. 2000 cfm, trio setting coal heaters will turn on.

Below 200C cfm, trip setting heaters will shut off.

char-R. H.

2(7) (8) 2(7) (8)

Instrument Channel SGTS Plow - Train B

H'eaters Instrument Channel SGTS Flow - Train C

Heaters Charcoal Heaters S2000 cfm H and R.H. Heatezs S 2000 cfm (A or F) 1 ~

2.

Charcoal Heaters S 2000 cfm H and 1.

R.H. Heaters S 2000 cfm (A or F) 2 ~

Below 2000 cfm, trip setting coal heaters will urn on.

Below 2000 cfm, trip setting heaters will shut off.

Belo~ 2000 cfm, trip setting

'coal heaters will turn, on.

Below 2000 cfm, trip setting heatezs will shut off.

char-Ro H.

char-R. H.

TABLE 3 ~ 2 A PRIMARy CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.

Operable Per Function Tri Level Settin Action 1

Remarks 2

Group 2 (Initiatin ) Logic N/A A. or

{B and E) 1.

Refer to Table 3.7.A for list of valves.

Group 2

(RHR Isolation-Actuation) Logic Gxoup 2 (Tip-Actuarion)

Logic Group 2 (Dryvell Sump Drains-Actuation)

=ogic Group 2 (Reactor Building S Refueling Floor, and Dry-vell Vent and Purge-Actuation) Logic N/A N/A N/A D

F and G

1.

Part of Group 6 Logic.

Group 3 (Initiating) Logic N/A 1.

Refer to Table 3.7.A for list of valves.

Group 3 (Actuation) Logic Group 6 Logic N/A N/A F andG 1.

Refer to Table 3.7.A for list of valves.

Group 8 (Initiating) Logic N/A Reactor Building Isolation N/A (refueling floor) Logic Reactor Building Isolation N/A (reactor zone)

Logic HorF H or G

oz' t.

Refer to Table 3.7.A for list of valves.

2.

Same as Group 2 initiating logic'.

1.

Logic has,.permissive to refueling floor static pressure.xegulator.

1.

Logic has permissive to reactor zone static pressure regulator.

TABLE 3i2 A PRIMARy CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.

Operable Per Punction Reactor Building Isolation Timer (refueling floor)

Txi Level Setti

,0<t52secs HorP Remarks

1. 'elow trip setting prevents purious trips and system pertux-bations from initiating isolation Instrument, Channel-Static Pressure Contxol Permissive (refueling floor)

Stat'ic Pressure Contxol Pressure Regulator (Re-fueling Ploor)

N/A

< 1/2" H 0 H or F Hor F 20 1.

20 Located in unit 1 only Permissive for static pressure control (SGTS A, B, or C on).

Channel shared by permissive on reactor zone static pressure cont.

Located in unit' only Controls static pressure of refueling floor during reactor building isolation with SGTS running.

Reactor Building Isolation 0 < t 5 2 secs.

Timer (reactor zone)

GorA or 8 Below trip setting prevents spurious trips and syste~ pertur-bations from initiating isolation 1 (9)

Instrument channel-Static Pressure Control Permissive

{reactor zone)

N/A Pexmissive for static pressure control (SGTS A, B, or C on).

Channel shared by permissive on refueling floor static pressure control.

" 1(9)

Static Pressure Control Pressure Regulator (reactor zone)

S 1/2" H 0 Controls static pressure of reactor zone during reactor building isolation with SGTS running.

Group 1 (Initiating) Logic N/A 1 ~

Refer to Table 3.7.A for list of valves.

Group 1 (Actuation) Logic N/A Refer to Table 3.7.A for list of valves.

0

TABLE 3 2 A PRIMARY CONTAIAKÃF AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimun No.

Operable Per unct on Tri Level Settin Remarks 1(7) (8)

SGTS Train A.ingle 1(7) (8)

SGTS Train B Logic 1(7) (8)

SGTS Train C Logic Static Pressure Control (refueling floor) 'Logic N/A N/A L or (A and F)

L or (A and F)

HorF 1.

Located in unit 1 only.

1(~)

Static Pressure Control (reactor cone)

Logic N/A Refer to Table 3.2.B for RCIC and HPCI functions including G oups 0, 5, and 7 valves.

6.

Channel shared by RPS and Primary Containment 6 Reactor Vessel Isolation Control System.

A channel failure may be a channel failure in each system.

7.

A train is considered a trip system 8.

Two out of three SGTS trains required.

A -failure of more than one will require action A and F.

9.

Tliere, is only one trip system with auto transfer to two power sources.

10.

Refer to Table 3.7.A an ts no e

7 A d it tes for a listing of Isolation Valve Groups and their initiating signals.

. A channel may be placed in an inoperable status 'for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPEIQBLH channel in the same trip system is monitoring that parameter.

0, S

One instrument channel; i.e.,

one-APRH 'or IRM or RBM, per trip.system may be bypassed except only one of four SRH may be bypassed-Rcfcr to Section 3.10.B for SRH rcqiiircmcnts during

.core alterations.

6.

XRN Channelo A, E, C,

G all in range 8 bypaSSeS SIN Channelo A

C C functions.

t<OTES FofC TABLE 3.2.C 1.

For the

. tartup and run positions of the Reactor Node Selector Switch, there shall be two operable or tripped trip systems for each function.

The

SRH, IRM, and APRM (Startup mode),

blocks need not be operable in <<Run<<

mode, and the APRN (Flow biased) and RBM rod blocks need not be operable in

<<Startup>>

mode.

If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition last-lonqer than seven days, the system with

'the inoperable channel shall be tripped. If the first column cannot be met for both trip systems, both trip systems shall

. be tripped.

2. I is the recirculation loop flow in percent 'of design.

Trip level setting is in percent. of rated power (3293 MWt).

~

... A ratio of FPZ/CHFLPD<1.0 is permitted at reduced "power.

See Speci,fication

2. 1 for APRM control rod block setpoint.

3.

IRN downscale is bypassed when it is on its lowest range.

<4.

SRH's A and C downscale function is bypassed when IRH s A, C, E,

'and G are above range 2.

SRH's B and D downscale function is by-passed when IRH's B, D, F, and H are above range 2.

SRH detector not in startup position is bypas: cd 'when the count

'rata is > 100 CPS or the above condition is satisfied.

XRM channels B, F, D,

H all in range 8 bypasses SRM" channel" 0

C D functions.

7.

Thc b

P following operation& restraints apply to thc RBM only.

Both RBH channels are bypassed when reactor power is g 30$.

Thc RBN need not be operable in the <<startup<< position'of thc reactor mode selector switch.

Two RBM channels are provided and only onc of the c may be bypaa.cd from the console.

An RBl4 channel may bc out of service.

for tc ting and/or maintenance provided this condition doc not last longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any thirty day period.

If minimus conditions for Table 3.2.C are not mct, administrative couLrola, 'hall be imniediat.cly imposed to prcvcnt control rod 0

withdrawal.

74

0 TABLE 3 2 F SURVEILLANCE INSTRUMENTATION Minimum S of Operable Instrument Channels Instrument, 0

LI 3-46 A

LI 3-46 B

PZ-3-54 PZ-3-61 PR 64 50 PI 64 67 TI-64-52 TR-64-52 TR-64-52 TI-64-55 TIS 64-55 LX-64 54 A LI-64-66 N/A PS-64-67 TR-64-52 and PS-64-58 B and IS-60-67 LI-84-2A LI-84-13A Instrument Reactor Water Level Reactor Pressure Drywell Pressure Drywell Temperature Suppression Chamber Air Temperature Suppression Chamber Water Temperature Suppression Chamber Water Level Control Rod Position Neutron Monitoring Drywell Pressure Drywell Temperature and Pressure and Timer CAD Tank "A" Level CAD Tank "B" Level Type Indication and Range Indicator - 155" to

+60" Indicator 0-1200 psig Recorder 0-80 psia Indicator 0-80 psia Recorder, Indicator p 4ppop Recorder 0-400DP Indicator, 0-4004F Indicator -25" to

+25" 6V Indicating

)

Lights

)

SRMr IRMA LPRM

)

0 to 100% power

)

Alarm at 35 psig

)

Alarm if temp.

)

2814F and

)

pressure

> 2 5

@gas

)

after 30 minute

)

delay

)

Indicator 0 to 100%

Indicator 0 to l00%

Nates (1)

(2)

(3)

(1)

(2)

(3)

(1)

(2)

(3)

(1)

(2) '(3)

(1)

(2)

(3) f 1)

(2)

(3)

(1)

(2)

(3)

(1)

(2)

(3)

(4)

(1)

(2)

(3)

(4)

LIMITING CONDITIONS FOR OPERATION

'VRVEILLANCE REqlJIREMENTS

3. 3. A REACTIVITY COi'ITROLS

~

l.3.A REACTIVITV CONTROLS 0

c.

Control rods vtth ucrnu times gicatci than those p xmittcd by Specif ica-tion 3. 3.C. 3 a rc iiio pe r-able, but if they can be inserted wf.th control rod drive pressure they need not be 'disarmed electri.-

cally.

d.

Control rods wi.th a fail d "Full-in" or "t"ul1-out" position sMitch may be'y-passed in the Rod Sequence Control System and consi-dered oprrable if thc actusl rod position is knoiw.

These rods civet be moved in sequence to their correct poaitioiis (full in on insertion or full out on withdroval).

e.

Control rods Mith inoperable accv sulati ro o" those uhosc position canriot, bc positively determined shall be consi-dered inoperable.

f.

Inoperable control rods shall

'.be positioned such that Speci.-

fication 3.3.A.l is met.

In addition, during reactor power opcratioti, no toore than one control rod in any 5 x 5 array may be inoperable (ot least 4 operable control rods riust separate any 2 inoperable ones).

If this Speci.fica-tion cannot be act thc reac-tor ahdBll not bo started, or if at poBdd r, the reactor shall be brought to a shut-doBdn condition Mithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months h.

A second 1l,censed operator shall verify the conformance to Specification 3.3.A.2.d before a rod may be bypassed in the Rod Sequence Control System.'ien it is initially determined that a control rod is incapable of normal insertion a test sliall be conducted to demon-strate that the cause of the malfunction is not a failure in the control rod drive meclianism.

'If this can be demonstrated an attempt to fully insert the control rod shall be made.

If the control rod cannot be inserted and an investigation has demonstrated that the cause of. failure is not" a failed cont To 1 rod drive mechanism collet housing, a shutdown margin test shall be made to demon-strate under this condition that the core can be made subcritical for any reactivity condition during the remainder of, the operating cycle with the analytically determined highest worth control rod capable of withdrawal fully.withdrawn, and all other control rods capable of insertion fully inserted.

d.

The control rod accumulators

. shall he determined operable at least once per 7 days by verifying that the pressure and.level detectors are not in

~'he alarmed, condition.

B.

Control Rodo l.

Each control rod shn11 be coupled to its driv or completely inserted and the 121 B.

CDNTRDL. RODS 1;

The coupling integrity shall be verified for each withdrawn control rod as follows:

BRONNS FERRY NUCLEAR PLANT F lGURE 3.5.2 Kf FACTOR AUTOMATIC FLOW CONTROL MANUAL FLOW CONTROL Scoon-Tube Set-Point Calibrotion poiition such thot Flowmox

"- 102;S%

107 0'/

112 0%

117.0 %

30 50 60 70 CORE FLOW,X 80 90 i00

LIMI'ZING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.A Thermal and Pressurization Limitations 4.6.h Thcn~<i and Prcssurizatto<<

Limit,at,ion.".

3.

4, D<<ring heatup by non-nuclear mean.

nr as indicated in 3.G.A.4, during cnoldown following n<<cl ear.

shutdown, or during low-level physics tests.

the reactor vessel tempernt<<re shall he at or above the temperatures of curve I2 of Figure 3.6.I until. removing tension on the head st<<d bolts as spcri fted 3n 3.6.h.5.

The reactor vessel shell temperatures during inservice hydrostatic or leak testing shall be at or above the temperatures shown on curve Ill of figure 3.6-1.

The applicability of this curve to these tests is extended to non-nuclear heatup and ambient loss cooldown associated with these tests only if the

. heatup and cooldown rates do not exceed 15'F pcr hour.

3.

Test spccfmcns reprc"cnt.'n~ tn reactor vcs-cl, base wci<> and Meld heat affcctcd zone metal @hall be installed iii'hc reactor vessel adgaccnt t.o t.hc vessel "all at the core midp1~nc level.

Thc number and type of "pcc'::.;,"

wf.13 hc 1n accordance;,h rcport H)2)0-10115.

%lie "pecim~ns shaU. mcct t,hc in-cnt of ~ST<

="

3.05-70.

Samples "hail bc with-drawn at onc-fou=th and threc-fourt,h" service life.

4.

Neutron flu< wires shel'e in-stalled in tnc rcacto:

vc "c'dJacent to thc reactor vcssc'all at thc core midpla;. lc:ci.

The wires shall be rcmove'nd

. tc"ted du"irg the first rc"uc'ing outage to experimentally verify thc calculated values of scut,ron flucncc at, onc-fourth of hc bclt&nc shell t)o.ckress the=

are used to dctbrnknc thc NDT':

shift from Figurc 3.6-Z 5.

6.

7 ~

The reactor vessel head bolting studs may be partially tensioned (four sequences of the seating pass) provided the studs and flange materials are above 70'F. Before loading the flanges any more, the vessel flange and head flange must be greater than 100'F, and must remain above 100'F while under full tension.

The pump in an idle recircula-tion loop shall not be started unless the temperatures of the coolant within the idle and operating recirculation loops are within 50'F of each other.

Thc reactor recirculation pumps shall not be started unless the coolant temperatures between the dome and the bottom head drain are within 145'F.

Mhcn t,hc reactor vc: "c'ead bolting stud-are t,cn"io.-.c'nd thc reactor is in a co' con:U.-

tion, the reactor vessel

";.cll tcmpcraturc immediately cclow thc head flange shall bc pc=-

mancntly recorded.

6.

Prior to and during tittup of an. idle recirculation loop, the tcmpcraLurc of thc rc<<ctcr coo.-

ant; in the operating and M.c loops shall be pcraencntly logged.

7.

Prior to starting a rcc'r..'.a-tion pump, thc reactor coc.'-'tt, tcmpcrat,urcs in thc dome

<<::d in thc bot.tom head drain ha'1 bc compared and pc~cntly 1.oggcd.

175

LIMITINGCONDITIONS FOR OPERATION SURYEILLANCE REQUIREMENTS

1. 6 PRIMARY SYSTEM BOUNDARY I~ 6

., P IMARY SYSTEM BOUNDARY B.

oolant chem str 1.

Prior to startup and at steaming rates less than 100,000 lb/hr, the following limits shall apply.

a.

Conductivity, umho/cmB25~C 2.0 b.

Chloride, ppm 0.1 I

B.

Coolant Chemistr

~ 1 A sample of reactor coolant shall be'::

... analyzed:

7 a.

at least every

~

96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months for conductivity and chloride ion content.

at, least every 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months during startups, until the steaming rate is greater than 100,000 lb/hr, for conductivity and chloride ion content.

Ce at least every 8

hours for conductivity,and chloride ion content

~hen

<he continuous

~

conductxvity monitor is inoperable.

2..

At steaming rates greater than 100,000 lb/hr, the follcwing limits shall 'apply.

a ~

b.

Conductivity, ymho/cm925~C 2

0

Chloride, ppm 0.2 176

0,

LZHITIHG CONDITIONS FOR OPERATION SURVEILLANCE RFQUIRElMNTS

~ ~ '0 ~ at ht

l. ri PRIMARY SYSTEM BOUNDARY 4

6 PRIMARY SYSTEM BOUNDARY i 3 4 At steaming rates greater than f00,000 lb/hr, the reactor water quality may exceed specification 3.6.B.2 only for the time limits specified below.

Exceeding these time limits of the following maximum quality limits shall be cause for placing the reactor in the cold shutdown condition.

a.

Conductivity time above 2 rrmho/crn925oC 4 weeks/year.

Max mum Limit 10 pmho/cmib25<C b.

Chloride concentration time above 0.2 ppm-4 weeks/year.

Haximum Limit-0.5 pprn.

177

I

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRE1&NTS

3. 6, PRIMARY SYSTEM BOUNDARY 6

PRIMARY SYSTEM BOUNDARY 4

When the reactor is not pressurized, except during

startup, the reactor water shall be maintained 'within the following limits.

a.

Conductivity "

pmho/cm$ 25 C

During equilibrium power operation an

, isotopic analysis-,

including quantitative measurements for at least I-131, I-132, I-133, and I-134 shall be performed monthly on a coolant liquid sample.

b.

Chloride " 0.5 ppm 178

0 LIMITINC CONDITIONS FOR OPEI ATION SURVEILLANCE REQUIREMENTS

3. 6 PRDUlRY SYSTFM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Whenever the reactor is critical, the limits ori activity concentra-tions in the reactor coolant shall not exceed the equilibrium value of 3.2 uc/gm of dose equivalent*

1-131.

This limit may be exceeded following power transients for a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

During this activity transient the iodine concentrations shall not exceed 26 uCi/gm whenever the

'reactor is critical.

The reactor shall not be operated more than 5 percent of its yearly power operation under this exception for the equilibrium activity limits. If the iodine concentration in the coolant exce'eds 26 >Ci/gm, the reactor shall be shut

down, and the steam line. isolation valves shall be closed immediately.

That concentration of 1-131 which alone would produce the same thyroid dose as the quantity of total iodines actually present.

3.

Additional coolant samples shall be taken whenever the reactor

. activity exceeds one percent of the equili-brium concentration specified in 3.6.8.5 and one of the following conditions are met:

a.

During startup b.

Following a significant

. power change"*

c.

Following an increase in the equilibrium

off-gas lev 1 exceeding 10,000 uCi/sec (at the steam jet air ejector')

within a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period.

d.

Whenever the equilibri'um iodine limit specified in 3.6'. 5 is exceeded.

The additional coolant liquid samples shall be taken at 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months intervals for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , qr until a stable iodine concentration below the limiting value (3.2 upi/

go/ is established.

However, at least 3 consecutive samples ahall be taken in all cases.

An isotopic analysis shall be performed for each

sample, and quantitative measurements made to determine the dose equivalent I-131 concentration.

If the total iodine activity of the sample is below 0.32 uci/gm, an isotopic analysis to determine equival<<nt I-131 is not required.

For the purpose of this section on sampling frequency,

a significant power exchange is defined as a change exceeding 15" of rated power in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

\\

179

0

LIMITING CONDITIOHS FOR OPERATION SURVEILLANCE RE@1'IREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 6

PRIMARY SYSTEM BOUNDARY C

Coolant Leaka e

Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212oF, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5

gpm.

In addition, the total reactor coolant system leakage into the pri ma ry con ta inment shall not exceed 25 gpmo Both the sump and air sampling systems shall be operable during reactor power operation.

From and after the date that one of these systems is made or found to be inoperable for 'any

reason, reactor power operation is permissible only during the succeeding seven days.

C Coolant Leaka e

1.

Reactor coolant system leakage shall

'e checked by the sump and air sampling system and recorded at least once per day.

2.

With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 2Q hours.

The air sampling system may he removed from service for a period of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for calibration, functional testing,"

and maintenance without pro<<

viding a temporary monitor.

180

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.6.L

~J ~ t P

~

b, The indicated va]ue of core flolI rate varies fro thr value dcri ved frtut lnop flov urr ureocnte by ooru than 102.

c.

The diffo cr to lover pliuun differential praaaure read-ing on an individual )et pump varico free th (leau of all 5ct purr!frrnn-tial preaaurea by nore than 10X.

The reactor shall, not bc operated Mith one recircu'ation lo.>p out of scrv9ce for more (has 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the reactor opcral,itt", lf one rccirct(l.ation loop is out of service, thc plant shaD be placed in a hot shutdown condi tion<<within 24 hotu;s unless the loop is sooner ret!tmed to service, 2.

Mhenever there is rccircula('el flow with the reactor in the Startup or Run Hodt and cne "e-circulut ion punp io operut'.n t vith the'equal(e'er vrlva c)ried, the dif!voe r to levee plena~

diffcrcntial ptcaaore shall checked hui12" and the diffnrtn-tinl prccaurc of an fndividusl

]et pump in a lonp shall not vary frcri the +can of all,'et pump dif ercntiai prcsuyres in that lonp by uvre than 10>..

2.

I'nllo:4ng onc purp ope. n ion, th di charge valve of'.the lou e

speed pump may not be opene unless the speed of the faster, pua'p ugly is less than

<<O,i of ita

= rated cpeed.

3, Stcudy stat. operation with both recirculation pumps out of ser-vice for up to 12 hrs is per-mitted.

Durihg such interval restart of the recirculation,

~umps is permitted, provided the oop'discharge temperature is within 75oF of the saturation temperature of the reacto~

vessel water as determined by dome pressure.

The total elapsed time in natural circula-tion and one pump operation must be no greater than 24 nrs,.

P.

Jet Pu~

Fino Miaaotch 1.

Recirculation punp apeeds shall be cheeked and logged at least once par day.

2.

No additional surveillance required.

3.

Before starting either recirculation pump during steady state operation, check and log the loop discharge temperature and dome saturation temperature.

1.

The structural integrity of the primary system shall be G.

Structural Integrit 1.

Table 4.6.A together with sup-pleme'nt'ary notes, specifies the 182

4selA

> W ~ j 0

LIHIYIMGCi.:~a'CfnHS FOR OPCRATIOH SURVEILLAHCE RY UIREHEHTS 3.7.A Prie nr Containment

<<.'7.A Primer Conta incur ent valves shall be exercised iuuoediately and every 15 days thereafter until the inoperable valve has been

~ returned to normal service.

c.

Mo dryucl 1-suppression chamber vacuum breakers "easy be determined to be inoperable for opening.

Once each operating cycle each vacuua breaker valve shal]

be inspected for proper operant fun of the valve and lait switches.

d.

If specifications 3.7.A.4.a,

.b, or.c cannot be met, the unit nhall Lc placed in 'a cold ehutdoun condition in an orderly manner within 24'ours.

d.

A leak test of the drywell to suppression chamber structure shall be con-ducted during each operating cycle.

Accept-

. able leak rate in 0.14 lb/

sec of primary containaent atuosphere with 1 psi differential.

S.

Ox en Concentration S.

0 eo Concentration a.

After completion of thc fire-r<<lated stnrtup retestin

program, containment atnonpl.ere shall be reduced co lean than 4Z oxygen with nitro-gen gas during reactor pouer operation with reac-tor coolant pressure above 100 psig, except as speci-fied in 3.7.A.5,b.

b.

Withfn the 24-hour period

~ ubnequent to placing the reactor in the Run mode following a shutdovn f the containment stuoaphere oxygen concentration shall be reduced to lena than 42 by volume and raa inta ined in

'hio condition.

De-inert-ing may commence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a shutdown.

s.

The primary containment oxygen concentration shall be measures and recorded daily.

The oxygen measurement shall be adjusted to account for the uncertainty of the method used by adding a

predetermined error function.

b.

The methods used to measure the primary containment oxygen con-centrat'ion shall be calibrated once every refueling cycle.

c.

If 'pecification 3.7.A.S.o and 3.7.A.5. b cannot Le net, an orderly

$35 nhutdn~

oh ll be initiated and the rractor <<he'll br in a Cold Shutdoun condition uithfn 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ~.

0

LIHITING CONDITTONS FOR OPERATION UEREHFNTS

<r 3.9.A Auxiliar Electrical E ui ment 4.9.A Auxiliar

~

Electrical E ui ment common trans-former capable of sup-plying power to the shutdown boards.

c the specified time sequence.

Once a month the'uantity of diesel fuel available shall be logged.

b.

A fourth operable units 1 and 2 diesel generator.

a.

Start buses 1A and 1B are energized.

b.

The units 1 and 2 4-kV shutdown boards are energized.

c.

"The -480-kV shutdown boards associated with the unit're energized.

4.

Buses and Boards Available d.

e.

Each diesel generator shall be given an annual inspection in accordance with instructions based on the manufacturer's recommendations.

Once a month a sample of diesel fuel shall be checked for quality.

The quality shall be wi.thin acceptable limits specified in Table 1

of the latest revision to ASTM D975 and logged.

d.

The Units 1

& 2 Diesel Aux Boards are energi.zed e,

Undervoltage relays operable on start buses 1A and 1B and 4-kV shut-down boards, A, B, C, and D.

f.

Shutdown Busses 1

& 2 energized 2.

D.

C Power System Unit Batteries (250-Voit) Diesel Generator Batteries (125-Volt) and Shutdown Board Batteries (250-Volt)

Every week the specific gravity and the voltage of the pilot cell, and temperature of an adjacent cell and overall battery voltage shall be measured and logged.

5 ~

The 250-Volt unit and shut-down board batteries and a

battery charger for each battery boards are operable.

6.

Logic Systems a.

Common accident signal logic system is operable.

b.

480-V load shedding logic system is operable.

b.

Every three months the measurements shall be made of voltage of each cell to nearest 0.1 vol:>

specific gravity of each

cell, and temperature oi every fifth cell.

These measurements shall be logged.

~'.

There shall be a minimum of 103,300 gallons of diesel fuel ln the standby diesel generator fuel tanks.

293 A battery rated discharge (capacity) test shall be performed.and the voltage,

time, and output current measurements shall be logged at intervals not to exceed 24 months'

I.INITINr. rnNiiIT(ONS FOR OPYRATIOH

3. lO CORI'. AI.TERATIOHS SURVEZLLAHCE RE UIRFHEHTS 4 10 CORE ALTERATIOHS A licabilit A licabilit Applies to the fuel handling and core reactivity limitations.

Applies to the periodic testing of those interlocks and instru-mentation used during refueling and core alterations.

O~becttvc To ensure that core'reactivity is uichin the capability of the control rods and to prevent criticality during refueling.

~OO ecttve To verify the operability of instrumentation and inter locks used in refueling snd core alterations.

S ccification A.

Rcfuclin Interlocks Refuelin interlocks l.

The reactor mode suicch shall be locked in che "Refuel" position during core alterations and the refuclinF interlocks shall be operable except as specified in 3elO.A.5 and 3.IO.AO6 belou.

1.

Prior to any fuel hand-ling uith the head off the reactor vessel, the re fu cling inter locks shall be functionally tested.

They shall be tested st ueekly inter-vale chereafter until no longer required.

They shall, ala

> be tested fol-loving any repel,r uork associated vith the inter-locks.

2.

Fuel shall noc be loaded into the reactor core unless all control rods are fully inserted.

2.

Prior to performing con-trol rod or concrol rod drive maintenance on con-trol cells uithout removing tuel assemblies, it shall be demonstrated chat the core can be made subcritical by a margin of 0.38 percent ak/k at any time during the maintenance uith the strongest operable control rod fully uithdravn and

~ll other operable rods fully inserted.

Alterna-tively i! the remaining 302

I.INITINC CONOIT\\Ou FC'" O"r<~

).10.'A Rcfu>>1!n). fn:>>rip ks Sl>RvF'.I.laHCr. Icr&IIRI:HK!ITS 4i)O.A Re!uclfn fnc>>clocks control rods are fully inserted and have hsd their dfrectfonal con-trol valves electrically

disarned, ic is suffi-cient to de~on>>crace chat the core is sub-crftfcal vfth s margin of at least 0.38 nercent 4k/k

'any circe during the euLfncenance.

h control rod on vhich mafncenance i>> being performed shall bc consider>>d inoperable.

The fuel grapple hoist load svftch shall be sec ac

<<1,000 lbs.

'3.

No add.f.tionnl surveillance

~

required.

0:

4. lf the franc-~ounted auxi-liary hoisc, thc ranorafl-aounted ouxf1 facy boise, or thc scrvfcc 'platform hoist is co be used for handlfng if'uel. vfch thc head off che reactor vessel, che load lfnit svitch on thc hoist to be used shall be >>st sc

<< 400 lbs 4.

No edditional surveillance required.

h e,sxfnu~ of tuo non-adjocenc control rods may be vichdrauw fron the core foc thc purpos>> of pcrfor-mfnr, control cod snd/or control rod drive nafncen-

ance, provfded the foliov-ing condicfons arc sacis-1 led:

5.

No additional surveillance required.

a.

The r>>secor node svicch

~ haf I be locked fn the "rc(ucl" position.

Thc refuelfng fnr er lock vlifch pc>>v>>nc s nor>>

chan one control cod fr'm bein"vfthdrcvn aoy be bypass>>d for one of the control rods on.uhfch eLafncensnct i>> being performed.

All ocher 303

I,IHI'lI<if: rORI) IT IAI<S Y(n< OPI;MAT IAII SIIRVV.I I.LAHCV, RV. I!IRD1VIITS

)10<

~ll

< ~

refunlfnR lnCorlocks

~hall be operabls.

4.10.A Refuel in interlocks b.

A suf f fcfen<

number of control rods shall be operable so that the core can ba

<.ade sub-critfca1 uith thc scronseat opcrsblo con-trol rod fully vfth-drs<<<<

and sll ocher'perable control rods ful.ly fnscrccd, or a)l directional control vstvcu for rematnfnV, concrol rods shall bc dfsa<<ocd electrically and nuff fcfcnt marZin Io criticality shall ba demonstrated.

c.

If m<<fntcnsr<ce is to be performed on tvo control rod drives they <<<us'e separated by <<<>rc 'ti<sn tMo control cells in any direction.

d.

An sppropr!atc nu<<<bcr of S!IH' err avs f table as dcftnc<l fn spcctff-cacfon 3.10.8.

f<.

Any n<<<<<l<c r <<I control rods may be ufth<Irn<<<< or removed fro<o the react<<r core pro-vfdfng the I'olio<<fng condi-ctons arc sacfof fed:

i.

Ths reactor mode svttch fs locked fn che "re-fuel" posfcfon

~

Thc refueling int,crlock vhfch prevents o<orc than ono control rod fro<a 6.

Pith the mode selector switch in the refuel mode, only one control rod may be withdrawn until two licensed operators have confirmed that all fuel has been removed from around the next rod to be withdrawn.

304

0 LIttITEHG CO'tDITEONS FOR 0?EPAT1ON SURVEILLANCE REQUIREMENTS t.

1 1

Yt l!4: t'I!<)TiCT I Atl.'lYli'rF~4lS tt.

1 1 FIRE PRM'FCTIO'.4 SYSTEMS CO Fir'e Protection System Th>>

COz Fire Protection System.

shall b.

operable:

B CO Fire Protection S stem 2

1.

COz Fire Protection Testing:

With a minimum of 8-1/2 tons (0. 5 Tank)

CO.

in storage unit=

"1 and 2.

I tern a.

Simui a ted automa tic and manual actuation Freauenc.

Once/yea.

2.

3.

ttith a minimutn of 3 t:ons (0. 5 Ttlnk )

CO g storage unit 3.

C.'att'om't tJ.c it t.itiation 2 oyic operable.

If specification 3.11.8.1.a or

3. 11. B. 1. b or

3. 11. B. 1. c cannot be

met, a

pa trolling fire watch wi th portable fire equipment hall be established to ensure that each area where protection is lost is checked hou r iy.

If specifications

3. 11. H. 1.,c,

3. 11

~ B. 1.b, o>>

3. 11. B. 1.c.!r".not nte t wit.hin 7 da ys, the affected unit(s) shall 1;e in cold shutdown <<ithin 2!t hours

~

2.

Checked dai ly Mhen the cable spreading room CO >

Fire Protection is inoperable, one 125 pound (or larger) portable fire extingui her shall be placed at each entrance.

4 ~ Storaqe tank pressure and level CO< Spray Once/3 header and

years, nozzle inspection for blockage 319

0

6 ~ 0 ihB< tINIST itATIVE CONTROLS B.

Source Tests

'Results of required leak tests performed on source if the tests reveal the presence of 0.005 microcurie or more of removal:le contat<>inatjon.

C.

Special Reports (in writing t:o the Director of Regional Office of Inspection and Enforcement)..

Reports on the following areas "hall be submitted as noted:

a.

Secondary Containment tt.7.c Leak Rate Testing(5) within 90'lays of completion of each test.

b.

Fatigue Usage Evaluation 6.6 Annual Ope rat:ing Repo t 0

,C.

Seismic Instrumentation Inoperabi 1 i ty d ~

.Relief Valve Tailpipe Instruttentation e ~

'Heteorolol,ical ldonitorinc Instrumentation In>>petal>lilt:y 3.?.J.3 3.2,P 3.2.I.2 Wi tlt in 10 da<'s after 30 days of inopet abi 1 i ty Within 30 days after inoperability of t:her>,.>>couple and acoustic monitor on one valve.

Within-10 days after 7 days of in<>per<tb 5 1 ity

'..~Sl< rial ReI>>>rt (in wri.tittl, to the l)irector of ltrl,iot>hi Off'

< of In71>acti<>n anforces>ent)

Data shall be retrieved from all seismic instruf<lcnts actuated duri",p a sei. mic event and analyzed to deters>ine, the mapnitude of the vibratory ground motion.

A Special Report shall be submitted within 10 'days after the event describing the magnitude, frequency, spectrum, and resultant effect upon plant feature<a important to safety.

356

PROPOSED CHANGES TO APPENDIX B I

TECHNICAL SPECIFICATIONS

tletcoro ogical data shall be surumazized and reported consistent vith thc rcco~~nd<<ions of Regulatory Guide 1.21 (Junc 1974) and Regle.atory 23 (February -1972)

~ and sEetcorological ybservati'ons shall be recorded in a form consistent vith ?lational Mcather Service Procedurf;s..

Xf the outage of any meteorological instraaent(s) requi ed by ReyQatory Guide 1.23 (February 1972) exceeds seven consecutive days, the total outage time, the dates of outage, the cause + the outage, and the instru-.

men s

nvo ve ment(s) involved shall be reported vi hin l0'Fogs of thc initiation of the outvc to the USliRC, Office of Xnspection and ".nforccment, vith a copy to the Ofiice of Nuclear Reactor Regulation, Division of Operating Reactors.

Flerr=nts of this program may be modified or tend.nated in accordance vith Subsection 5.6.3(c)..

".'he collection of meteorological data at the plant site provides inforaatkon for use in developing atmospheric diffusion parameters for cstimctIog potential radiation doses to thc public resulting from actual routine or abnormal releases of radioactive materials to the atmosphere, and for a" ses" ing thc actu" 1 impact of thc plant cooling system on the atmospheric environment of the ite area.

A>>eteoro'ogical data collection program as describ d above is necessary to meet the requirements of subpercgraph 50.36a{a){E".) of 10 ~s Part 50, Appendix D to 10 CFR Part 50, and Appendix Z to 10 CFR Part 50.

(3)

Ref. Section 6.7.3.D Appendix A Technical Specifications.

0,

5-5.3

'<<ittcn proccdu<<es dc>>cribcd in Sect fon 5.5.1>>hall bc riv1>>Med by PORC and approved by the Plant Hnnagcr prior to i>><~>l<><>cntatfon.

Tc<>>por<<ry. <<h<><<y<

>< tu a procedure which do not chnngc thc intent of the plant>>tnff kn<>Ml<dg<>l>l<>

in the area affected by tho procedure und the additional approval ol n numb<<r of thc plant staff who holds a Senior Reactor Oporator liccns<:.

Such

<J>>>ny<>><

shall be documented and subsequently reviewed by PORC and

<<pprov<.d by th>>

Plant Manager; 5.6 Re ortin Re uircments 5.6.1 A report shall be prepared by Fnvironmcntal Compliance and submitted to DNP following the end of each 12~onth period of operation, which shall summarize the results of the nonradiological environmental monitoring program.

5.6.2 Routin Re ortin a.

A summary report shall be prepared for both the inplant monitoring program and the nonradiological monitoring, programs and submitted to the Director of Division of Ope. ating Reactors, NRC, us p"rt of the Annual Operating Report within 120 days after December 31 of each year.

b.

Radiolo ical Environmental Honitorin Routine Re ortin Reporting Requirements:

1.

TVA shall prepare a report entitled "Environmental Radio-activity Levels -'rowns Ferry Nuclear Plant

>> Annual Report."

The report shall cover the previous 12 months of operation and shall be submitted to the Director of the NRC Region,II Office {with a copy to the Director, Office of Nuclear Reactor R6gulation) within 120 days after January 1 of each year.

The rcport format shown in Regulatory Guide 4.8 Title 1 shall be used.

The report shall include summaries, interpretations, and evaluations of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies and/or operational controls

{as appro-priate),

and an assessment of the observed impacts of the plans operation on the environments.

If harmful effects or evidence of irreversible damage are detected by the monitoring, the licensee shall provide an analysis of the problem and a

'roposed,. course of action to alleviate thc problem.

UNIT 3 PROPOSED CUANGES

PROPOSED CHANGES TO APPENDIX h TECHNICAT, SPECIFICATIONS

SAFETY LIMIT LIMITING SAFFTY SYSTEM SETTING 1

1 FUEL CLADDING INTEGRITY 2

1 FUEL CLADDING INTEGRITY A licabilit Applies to the interrelated variables associated with fuel thermal behavior.

~cb 'ective To establish limits which ensure the integrity of the fuel cladding.

S ecifications A-Thermal Power Limits J. Reactor Pressure

> 800 psia and Core Flow >

10%

of Rated.

When the reactor pressure is greater than 800 psia, the existence of a minimum critical power ratio (MCPR) less than 1.07 shall constitute violation of the fuel cladding integrity safety limit.

A licabilit Applies to trip settings of the instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.

~cb ective To define the level of the process variables at which automatic protective action is initiated to prevent the fuel cladding integrity safety limit from being exceeded.

S ecification The limiting safety system settings shall be as specified below:

A Neutron Flux Trip Settinps APRM Flux Scram Trip Setting (Run Mode) a.

When the Mode Switch is in the RUN position, the APRM flux scram trip setting shall be:

S<(0.66W

+

54%)

where".

S

= Setting in per-cent of rated thermal power (3293 MWt)

0

,SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING l.l FUEL CLADDING INTEGRITY 2 ~ 1 FUEL CLADDING INTEGRITY M = Loop recircu-lation flow rate in per-cent of rated (rated loop recirculation flow rate equals 34;2xl0~ lb/hr)

In the event of operation with the core maximum fraction of, limiting power density (CMPLPD) greater than fraction of rated thermal power (FRP) the setting shall be modified as follows<

S<(0.66M

+

54%) FRP c.

For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

(NOTE:

These settings assume operation within the basic thermal hydraulic design criteria.

These criteria are LHGR < 13.4k'/ft and MCPR within the limits of 3.5.K.

10

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1

FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY If +t ~s determine that either of these design criteria is being violated during operation, action shall be xnitiated within 15 minutes to restore operation within the prescribed limits.

Surveillance requirements for APRN scram setpoints

. are given in specification 4

1 B).

The APRM Rod block trip setting shall be:

S (0.66W +42%)

where:

S~

= Rod block setting in percent of rated thermal power (3293 MWt)

W

= Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.2 x 10~ 1b/hr)

In the event of operation with the core maximum fraction~

of limiting power density (CMFLPD) greater'han fraction of rated thermal power (FR~P the setting shall be modified as follows:

0~< f0.660 +42%

) P~

CNPLPD

5

SAFETY LIMIT LXMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRXTX 2

1 FUEL CLADDING INTEGRITY 2.

I(carter I'ressure

- 800 PSIA or Core Flow =10% of rated.

2, nPRM and IBM TrIp Settlnps (Startup and Hot Standby Modes).

When the reactor pressure is ~800 PSIA or core flow is ~10/ of rated, the core thermal power shall not exceed 823 MWt { 25% of rated thermal power).

a.

APRM Mhen the-reactor mode switch is in the STARTUP position, the APRM scram shall be set at less than or equal to 15% of rated power.

b.

IRM The IRM scram shall be set. at less than or equal to 120/125 of full scale.

12

0

SAFETY LIMIT LXMITXNG SAFETY SYSTEM SETTXNG 1

1 FUEL CLADDING INTEGRXTY 2.1 FUEL CLADDING XNTEGRITY B.

'ower Transient B. Power Transient Tri 'Settin s

To ensure that the Safety Limits established in Specification

1. 1.A are not exceeded,
'ach required scram shall be initiated by its expected scram signal.
The Safety Limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.
Scram and isola-(PCZS groups 2,3,6) reactor low water level 2.
Scramturbine stop valve closure Scram--turbine control valve fast clo"ura or turbine trip.
2 538 in.
above vessel zero S
10 per-cent valve closure 550 psig 4.
Scram--1ow con-denser vacuum 5,
Scram--main steam line isolation 2 23 inches Hg vacu~a 5
10 per-cent valve
.. "losuro 6,
Main steam isola-tion valve closure nuclear system low pressure
~825 psic C. Reactor Vessel Mater Level C. Mater Level Tri Settin s
Mheneuer there is irradiated fuel in thc reactor vessel,
~ the water level shal1 not be less than'7.7 'in. above the top of the normal active fuel zone.
Core spray and LpCI actuation
'reactor low water level HPCI and RCIC actuation reac-tor low water level 3,
Main steam isola-tion valve closure--reactor low water level 2 378 in.
above voboel zero 2
470 in. I above vessel zero h
47O Xn.
(
above vessel zero
-= 13
C ~
a given point at constant recirculation flow rate, and thus to protect aqainst the condition of a MCPR less than 1.05.
This rod block trip setting, which is automatically 'varied with recirculation loop flow rate, prevents an inciease in the reactor power level to excess values due to control rod withdrawal.
The flow variable trip setting provides substantial margin from fuel da sz.ge, assuming a steady-state operation at the trip setting, over the entire recirculation flow range.
The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during the steady-state operation is at 108$ of rated thermal power because of the APRM rod block trip setting.
The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.
As with the APRM scram trip
settinq, the APRM rod block trip setting is adjusted downward if the CHFLPD exceads FRP thus preservine t he APRM rod block safety margin.
Reactor Water Low Level Scram and Isolation Exce t Main Steamlines
~
The set point for the low level scram is above the bottom of the separator skirt.
This level has been used in transient analyses deal'ing with coolant inventory decrease.
The results reported in FSAR subsection N14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.05 in all cases, and system pressure does not reach the safety valve settings.
The scram setting is approximately 31 inches below the normal operating ranqe and is thus adequate to avoid spurious scrams.
D Turbine Sto Valve Closure Scram The turbine stop valve closure trip anticipates the pressure-neutron flux and heat flux increases that would result from closure of the stop valves.
With a trip setting of 10K of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst. case transient that assumes the turbine bypass valves remain closed.
(Reference
2).
Turbine Control Valve Fast Closure or Turbine Tri Scram Turbine control valve fast closure or turbine trip scram anticipates the
pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.
The reactor protection system initiates a scram in less than 30 milliseconds aEter the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.
This scram is achieved by rapidly reducing hydraulic control 22
oil pressure at the main turbine control valve actuator disc dump valves."
This loss of pressure is sensed by pressure switches whose contacts form
.the one-out-of-two-twice logic input to the reactor protection system.
This trip setting, a nominally 50X greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve.
No signifi-cant. change in MCPR occurs.
Relevant transient analyses are discussed in References 2 and 3 of the Final Safety Analysis Report.
This scram is bypassed when turbine steam flow is below 30X of rated, as measured bv the turbine first staae nressure.
F.
Main Condenser Low Vacuum Scram To p'rotect the main concenser against overpressure, a loss of condenser vacuum initiates automatic closureof the turbine stop valves and turbine bypass valves.
To anticipate the transient and automatic scram resulting from the closure of the" turbine stop valves, low condenser vacuum initiates a
scram.
The low vaccum scram set point is selected to initiate a scram before the closure of the turbine stop valves is initiated.
G.
6 H.
Main Steam Line Isolation on Low Pressure and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 850 psiq was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel.
Advantaqe is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at.
low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit.
Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the STARTUP 23
~n TABLE 3 n 1 e A REACTOR PROTECTION SYSTEM (SCRAM)
INSTRUMENI'ATION REQUIREMENT Min. No.
of Operable Inst.
Channels Per Trip I
1 Mode S~itch in shutdown 1
Manual Scram IRM (1 6)
High Plux Inoperative Tri Level Settin 120/125 Indicated on scale Modes in Which Function Mist Be erable Shut-Startup/Hot datn
~Refuel 1
R~tendn X (22)
X (22)
X Run
~notion 1
X 1.A X
1.A (5) 1 CA (5) 1 A 2
2 2
2 APRM (16)
High Flux High Flux
~
Inoperative Downscale See Spec.
2.1.A. 1 S
15% rated power (13)
? 3 Indicated on Scale X (21)
X {21)
(11)
X 1.A or 1.B X(17)
{15) 1.A or 1.B X(17)
X 1.A or 1eB (11)
X(12) 1.A or 1.B High Reactor Pressure 5 1055 psig X (10)
X 1 A High Drywell Pressure (la)
S 25Dsis X (8)
X(8)
X 1 A Reactor Low Water Level (10)
> 538" above vessel zero X
1 A High Water Level in Scram Discharge Tank Main Steam Line Isola-tion Valve Closure Turbine Cont. Valve Past, Closure
<r Turbine Trip 5 50 Gallons 10){ Valve Closure k 550 psis X
X(2)
X leA X{0) 1.A or 1.D X(3) (6)
X(3) {6)
X(6) 1.A or 1.C
TABLE 3 1 A REACTOR PROTECTION SYSTEM (SCRAM)
INSTRUMENTATION REQUIREMENT e
Min. No of Operable Inst.
Channels Per Trip 4
Turbine Stop Valve Closure Tri Level Settin N 10% Valve Closure Modes in Which Function Must Be O erable Shut-Startup/Hot Ooen
~ReEuel 7
S~eendb nnn
~Action
'I X(4) 1 A or 1
D 2
Turbine, First Stage Pressure Permissive Turbine Condenser Low Vacuum Main Steam Line High Radiation (14) not 5154 Psis I 23 In. Hg, Vacuum 3X.Normal Full Power
Background
X(18)
X(3)
X{9)
X (18)
X(18)
(19)
X{3)
X 1.A or 1.C X{9) '(9) 1.A or 1.C
12.
The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not, high.
13.
Less than,14 operable LPRM>s will cause a trip system trip.
14.
Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Xsol +ion Control System.
A channel failure may.be a channel failure in each system.
15., The APRM 15% scram is bypassed in the Run Mode.
16. Channel shared by Reactor Protection ys S stem and Reactor Manual Control annel failure S stem (Rod Block Portion).
A channel failure may be a channel a
ure f
l 1 i llowed to be INOPERABLE per Table 3.1.A, r~hle in the corresponding function in that same channel may be inoperable n
the Reactor Manual Control System (Rod Block).
17.
Not required while performing.low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).
l.'l. 'I'l ls l
i ti~>n must inhibit the automatic l>vpassing of e<<rbine c'ontrol valve fast closure or turbine trip scram and eurbinc stop valve cl scram whenever turbine first stage pressure 4s greater than or equal to 1'j4 psig.
19. Action 1.A or 1.D shall be taken only if the permissive fails
~
in such a manner to prevent the affected RPS logic from performing its intended function.
Otherwise, no action is
'required.
20.
The nominal setpoints for alarm and reactor trip (1.5. and 3.0 times
. background, respectively) are established based on the normal backgroun at full power.
The allowable setpoints for alarm and reactor trip are 1.2-"1.8 and 2.4-3.6 times background, respectively.
21. The APRM High Flux and Inoperative Trips do not have to be operable in the Refuel Mode if the Source Range Monitors are connected to give a non-coincidence, High Flux scram, at 5 x 10 cps; The SRM's shall be operable per Specification 3.10.B.l.
The removal of eight
( )
shorting links's required to provide non-coincidence high-flux scram protection from the Source Range Monitors.
22. The three required JRM's per trip channel is not required in the Shutdown or Refuel Modes if at least four IRM's (one in each core quadrant) are connected to give a non<<coincidence, High Flux scram.
The removal of four (4) shorting links is required to provide non-coincidence high-flux scram protection from the IRM's.
nel ma be laced in an inoperable status for p
u to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for r
l cin the trip system 5n the tripped required surveillance without p ac ng e
condl.tion prov e
a a.
id d t le st one OPERABLF. channel'in the same er p system is monitoring that parameter.
TABIE 0 1
A REACTOR PROTECTION SYSTEM (SCRAM)
INSTRUMENTATION FUNCTIONAL TESTS MINIMUMFUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR AND CONTROL CIRCUITS Main Steam Line High Radiation Main Steam Line Isolation Valve Closure Turbine Control Valve Past Closure Group - (2)
B Functional Test Trip Channel and Alarm (4)
Trip Channel and Alarm Trip Channel and Alarm Minimum Frequency (3)
Once/Meek Once/Month
{1)
Once/Month
{1)
Turbine First Stage Pressure Permissive Turbine Stop Valve Closure Trip Channel and Alarm Trip Channel and Alarm Every 3 Months Once/Month (1)
e, ~
< ~
a Calibration Comparison to APRM on Control-6cat'ctlps (6)
Instrument Channel Group (1)'RM High Flux Note (a)
C TABLE 4 1 ~ B REACTOR PROTECTION SYSTEM (SCRAM)
INSTRUMENT CALIBRATION MINIMUMCALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS
=-Minimum Frequency (2)
APRM High Flux Output Signal Flow Bias Signal LPRM Signal High Reactor Pressure High Drywell Pressure Reactor Low 'Rater 'Level High Rater Level in Scram Discharge Volume
~ Turbine Condenser Low Vacuum Main Steam Line Isolation Valve Closure Main Steam Line High Radiation Turbine First, Stage Pressure Permissive B
Heat Balance calibrate Flow.Bias Signal (7)
TIP System Traverse (8)
Standard Pressure Source Standard Pressure Source A
Pressure Standard A
Note (5)
A Standard Vacuum Source Note (5)
Standard Current Source (3)
Standard Pressure Source Once every 7 days Once/operating cycle Every 1000 Effective Full Power Hours 1
I Every 3 Months Every 3 Months Every 3 Months Note (5)
Every 3 Months Note (5)
Every 3 Months Every 6 Months
~
~
Turbine Stop Valve Closure Note (5)
Note (5)
NOTES FOR TABLE 4 1. B A description of three groups is included in the bases of this specification.
2 ~
Calibrations are not required when the systems are not required to be operable or are tripped. If calibrations are
missed, they shall be performed prior to returning the system to an operable status.
3 ~
The current source provides an instrument channel alignment.
Calibration using a radiation source shall be made each refueling outage.
5.
Required frequency is initial startup following each refueling outage.
Physical inspection and actuation of these position switches will be performed once per operating cycle.
6.
On controlled startups will'e verified.
overlap between the IRM's and APRM's 7 ~
The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks
'uring each operating cycle.
The instrumentation is an analog type with redundant flow signals that, can be compared.
The flow comparator trip and upscale will be functionally tested.according to Table 4. 2.C to ensure the proper operating during the operating cycle.
Refer to 4.1 Bases for
'urther explanation of calibration frequency.
8.
A complete tip system
".xavexse calibx'ates the LPRH signals to the process computer.:
The individual LPRN meter readincs wi.ll be adjusted as a minimum at the beginning of each operatinp cvcle before reaching 100/ power.
0 40
'I which a scram would be required but not b'e ablo to perform its function adequately.
A source range monitor (SRN) system is also provided to supply additional neutron level information during startup but has no scram functions Ref Section 7 5.0 TSAR Thus, the IRM'is required in the Refuel and Startup modes.
In the power range the APRM system provides required protection.
Ref. Section 7.5.7 FGAR ~
Thus g the IRM System is not required in the Run mode.
The APRNes'and the IRM~s provide adequate coverage in the startup and intermediate range>>
fl The high reactor pressure, high drywell pressure, reactor low water level and scram discharge volume.high level scrams are required for Startup and Run modes of plant operation.
They are, therefore, required to be operational for these modes of reactor operation.
The requirement to have the scram functions as indicated in Table 3.1 ~ 1 operable in the Refuel mode is to assixre,that shifting to the Re&xel mode during reactor power operation does not diminish the need for the reactor protection system.
The turbine condenser low vacuum scram is only required during power operation and must be bypassed to start up the unit Below 150 psig turbine first stage pressure (30% of rated), the scram signal due to turbine stop valve closure, and turbine control valve fast closure, is bypassed because flux and pressure scram are adequate to protect. the reactor.,
Because of the ApRM downscale limit of 2 3% when in the Run mode and high level limit of 5 15% when in the Startup Mode, the transition between the Startup and Run Nodes must be made with the', APRM instmxmentation indicating betwee'n 3% and 15% of rated power or a control rod scram will occur.
In addition, the IRM system must be indicating below the High Flux setting (120/125 of scale) or a scram will occur when in the Startup Mode.
For normal operating conditions, these limits provide assurance of overlap between the ZRM system and APRM system so that there are no "gaps" in the power level indications (i.e., the power level is continuously monitored from beginning of startup. to full power and from full power to shutdown).
When power is'being reduced, if a transfer to the Startup mode is made and the IRM~s have not been fully inserted (a maloperational but not impossible condition) a control rod block immediately occurs so that reactivity insertion by control rod withdrawal cannot occur.
I 43
LIMITXNG'CONDITXONS FOR OPERATION SURVEILLANCE REQUIREMENTS
3. 3 PROTECTED INST~RON ST+~ON 4.2 PROTHCTXVF. XNSTRUNHNTATION 2 0 With one or more of the meteorological monitoring channels inoperable for more than 7 days, prepare and'ubmit a Special Report to the Commission, pursuant to specification 6.7. 3. C within the next 10 days outlining the cause of the malfunction and the plans for restoring the system to operable status.
55
TABLE 3 2 A PRIMARY CONTAINMENT AND REACTOR'UILDING ISOIATION INSTRUMENTATION Minimum No.
Operable Per Tri S s 1 (ll) l 2
Tzi Level Settin Function Instrument Channel -
?
538> above vessel zero Reactor Low Water Level (6).
Action A or
{B and E)
Remarks Below trip setting does the following:
a.
Initiates Reactor Building Isolation b.
Initiates Primary Containment.
Zsolation c.
Initiates SGTS Instzument Channel-Reactor High Pressure Instrument, Channel-Reactor Low Water Level (LIS-3 56A-Di SW 01) 100
+ 15 psig 2 470",above vessel zero 1.
Above trip setting isolates the shutdown cooling suction valves of the RHR system.
1.
Below trip setting initiates Main Steam Line Isolation Instrument Channel-High Drywell Pressure (6)
(PS-64-56A-D)
S 2,5 Psis A or (B and E) l.
Above trip setting does the following:
a.
Initiates Reactor Building Isolation b.
Initiates Primary Containment Isolation c.
Initiates SGTS Instrument, Channel High Radiation Main Steam Line Tunnel (6)
Instrument Channel-Low Pressure Main Steam Line S
3 times normal rated full power background 825 Psig
(<)
B 1.
Above trip setting initiates Main "Steam Line Isolation 1.
Below trip setting initiates Main Steam Zine Isolation 2 (3)
Instrument Channel-High Flow Main Steam Line Instzument Channel-Main Steam Line Tunnel, High Temperatuze S
100$ of rated steam flow B
200DF 1.
Above trip setting initiates Main Steam Line Isolation 1.
Above trip setting initiates Main Steam Line Zsolation.
TABLE 3 ~ 2~A PRIMARY CONTAINMENT AHD REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.
Operable Per Tzi Level Setti Action Remarks Instrument Channel-Reactor Water Cleanup System Floor Drain High Temperature Instrument Channel-Reactor Water Cleanup System Space High Temperature Instrument Channel -
Reactor Building Venti-lation High Radiation-Reactor Zone 160 1800 F 160 - 180oF S
100 mrlhr or downscale l.
Above trip setting initiates Isolation of Reactor Water Cleanup Line from Reactor and Reactor Water Return Line.
Same as above 1.
1 upscale or 2 downscale will a.
Initiate SGTS b.
Isolate reactor zone and refueling floor.
c.
Close atmosphere control system.
Instrument Channel-Reactor Building Venti-lation High Radiation-Refueling Zone S
100 mr/hr or downscale 1.
1 upscale or 2 downscale will a.
Initiate SGTS b.
Isolate refueling floor.
c.
Close atmosphere control system 2 (7) (8)
Instrument Channel SGTS Plow - Train A Heaters Charcoal Heaters S 2000 cfm H and R.H. Heaters S 2000 cfm (A or F) 1.
Below 2000 cfm, trip setting coal heaters will turn on.
2.
Below 2000 cfm, trip setting heaters will shut off.
char-R. H.
2(7) (8) 2(7) (8)
Instrument Channel SGTS Plow - Train B
Heaters Instrument Channel SGTS Flow - Train C
Heaters Charcoal Heaters S 2000 cfm R.H. Heaters S 2000 cfm H and (A or F) charcoal Heaters S2000 cfm H and R.H.'eaters S 2000 cfm (A or F) 1.
Below 2000 cfm, trip setting coal heaters will turn on..
2.
Below 2000 cfm, trip setting heaters will shut off.
1.
Below 2000 cfm, trip setting coal heaters will turn on.
2.
Below 2000 cfm, trip setting heaters will shut off.
char-R. H.
char-R. H.
0 TABLE 3 2.A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.
Function Reactor Building Isolation Timer (refueling floor)
Tri Level Setti 05t<2secs.
Action 1
HorF Remarks 1
.Below trip setting prevents spurious trips and system pertur-bations from initiating isolation Instrument Channel-Static Pressure Control Permissive (refueling floor)
Static Pzessuze Control Pressure Regulator (Re-fueling Floor)
Reactor Building Isolation Timer (reactor xone)
Instrument Channel-Static Pressure Control Permissive (reactor zone)
N/A 1/2" H 0 0 < t 8 2 secs.
N/A HorF HorF G or A or H 1 ~
2 ~
1 2 ~
Located in unit 1 only Pezmissive for static pressure control (SGTS A, B, or C on).
Channel shared by permissive on reactor xone static pressure cont.
Located in unit 1 only Controls static pressure of refueling floor during reactor building isolation with SGTS running.
Below trip setting prevents spurious trips and system pertur-bations from initiating isolation Permissive for static pressure control (SGTS A, B, or C on).
Channel shared by permissive on
= refueling floor static pressure control.
Static Pressure Control Pressure Regulator (reactor zone)
S 1/2n H O
- Contzols static pressure of reactor xone during reactor building isolation with SGTS running.
Group 1 (Initiating) Logic N/A 1.. Refer to Table 3.7.A for list of valves.
Group 1
(Actuation) Logic N/A Refer to Table 3.7.A for list of valves.
TABLE 3' A
~
~
+
+~
+
4 PRIMARy OONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.
Operable Per I
2 Group 2 (Initiating) Logic Group 2
(RHR Isolation-Actuation) Logic Group 2 (Tip-Actuation)
Logic Group 2 (Drywell Sump Drains-Actuation) Logic Tri Level Settin N/A N/A N/A N/A Action 1
A or (B and E)
Remarks 1.
Refer to Table 3.7.A for list of valves.
1 S.
Group 2 (Reactor Building N/A
. 6 Refueling Floor, and Dry-well Vent and Purge-Actuation) Logic Group 3 (Initiating) Logic N/A Group 3 (Actuation) Logic N/A F andG 1.
Part of Group 6 Logic.
1.
Refer to Table 3.7.A for list of valves.
Group 6 Logic N/A F andG 1.
Refer to Table 3.7.A for list of valves.
Group 8 (Initiating) Logic N/A 1.
Refer to Table 3.7.A for list of valves.
Reactor Building Isolation N/A (refueling floor) Logic Reactor Building Isolation N/A (reactor zone)
Logic H or F HorG or A 2.
Same as Group 2 initiating logic.
1.
Logic has permissive to refueling floor static pressure regulator.
1.
Logic has permissive to reactor zone static pressure regulator.
TABLE 3 2 A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.
Operable Per l(7) (8)
SGTS Train A Logic 1(7) (8)
SGTS Train B Logic 1(7) (8)
SGTS Train C Logic Static Pressure Control (refueling floor) Logic Tri Level Settin N/A Action 1
L or (A and F)
L or (A and FJ L or (A and F)
H or F Remarks 1.
Located in unit 1 only.
1 (9)
Static Pressure Control (reactor zone) Logic N/A Refer to Table 3.2.B for RCIC and HPCI functions including Groups 0, 5, and 7 valves.
3.
There are four channels per steam line of which two must be t
operable.
Only required in Run Mode (interlocked with Mode Switch).
5.
Not required in Run Mode (bypassed
iy mode switch).
6.
Channel shared by RPS and Primary Containment 8 Reactor Vessel Isolation Control System.
A channel failure may be a channel failure in each system.
7.
A train i" considered a trip system.
8.
9 ~
Two out of three SGTS trains required.
A.-failure of more than one will require action A and F.
There is only one trip system with auto transfer to two power sources.
10.
Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals.
11.
A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLH channel in the same trip system is monitoring that parameter.
0
.63
XCBS Fof< TABLII 3.2.C For th~ staztup and run positions of the Reactor Mode selector s~itch, there shall be two operable or tripped trip systems for each function.
The
SRN, IRM+ and APRN (Startup mode),
blocks need not be operable in >>Run>>
mode, and the APRM (Flow biased) and RBM rod blocks need not be operable in
>>Startup>>
mode.
If the first, column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition last longer than seven
days, the system with the inoperable channel shall be tripped. If the first column cannot be met for both trip systems, both trip systems shall be tripped.
2.
3.
W i's the recirculation loop flow in percent of design.
Trip level "setting is in percent -of rated power (3293 MWt).
A ratio of PRP/CMFLPD<1.0 is permitted at reduced
'ower.
See Specification
2. 1 for APRM control rod block setpoint.
IRN downscale is bypassed when it is on its lowest range.
4, 5.
SRM's A and C downscale function is bypassed when XRM's A, C, E, and G are above range 2.
SRM's B and D downscale function is by-passed when IRM's B, D, P, and H are above range 2.
~
SRM detector not in startup position is bypassed when the count rate is 0- 100 CPS or the above condition is satisfied.
I One instrument channel; i.e.,
one APRM or IRM or RBM>'- per trip system may be bypassed except only one of four SRM may be bypassed.
Refer to Section 3.10.8 for SRM requirements during core alterations.
6.
XRM channels A, E, C, G all in range 8 bypasses SRM channels A 6 C functions.
IRM channels B, F, D,
H all in range 8 bypasses SRM channels B 6 D functions.
0 7.
The
'b.
ce following operational restraints apply to the RBM only.
Both RBII channels are bypassed when reactor power is ( 30$.
The RBM need not be operable in the >>startup>> position of the reactor mode selector switch.
Two RIM channels are provided and only one of these may be bypassed from the console, An RIN channel may be out of service'or testing and/or maintenance provided this condition does not last longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any thirty day period.
Xf minimum conditions for Table 3.2.C are not met, administrative controls, shall be immediately imposed to prevent control rod withdrawal.
77
0
TABLE 3 2 F
'URVEILLANCE INSTR(tMENTATZON Minimum ¹ of Operable Instrument Channels Instrument ¹ LX-3 46 A LI-3-46 B Instrument.
Reactor water Level Type Indication and Range Indicator - 155" to y 60" Notes (1)
(2)
(3)
PZ-3-54 PZ-3-61 Reactor Pressure Indicator 0-1200 psig (1)
(2)
(3)
PR-64-50 PI-64-67 TI-64-52 TR-64-52 TR-64-52 TI-64-55 TZS-64-55 LZ-64-54 A LI 66 N/A N/A Ps-64-67 TR-64-52 and PS-64-58 B and ZS-64 67 LI, 84-2A LZ-84-13A Drywell Pressure Drywell Temperature Suppression Chamber Air Temperature Suppression Chamber Water Temperature Suppression Chamber Rater Level Control Rod Position Neutron Monitoring Drywell Pressure Drywell Temperature and Pressure and Timer CAD Tank "A" Level CAD Tank "B" Level Recorder 0-80 psia Indicator 0-80 psia
Recorder, Zndicator Q 4000P Recorder 0-400op Indicator, 0-400~F Indicator -25" to
+25" 6V Indicating
)
Lights
)
SRM, IRM, LPRM
)
0 to 100% power
)
Alarm at 35 psig
)
)
Alarm if temp.
)
> 281op and
)
pressure
> 2 5 pzfg) after 30 minute
)
delay
)
Indicator 0 to 1004 Indicator 0 to 100%
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(4)
(1)
(2)
(3)
(4)
LINITINC CONDITIONS FOR OPERATION SURVEILLANCE REOL IREHENTS 3.3 REACTIVITY CONTR01 4.3 REACTIVITY CONTROL c ~
d.
Control rods with scram times greater than those permitted by Specification 3
3 ~ C 3 axe inoperable, but if they can be ins'erted with control rod drive pressure they need not be disarmed electrically.
Contxol rods with a failed
<<Full-in<< or
<<Full-Out<<
position switch may be bypassed in the Rod Sequence Control System and considered operable if the actual rod position is known.
These rods must be moved in sequence to their correct positions (full in on insertion or full out on withdrawal).
C ~
When it is initially determined that a
control rod is incap-able of normal insertion a test shall be con-ducted to demonstrate that the cause of the malfunction is not a failure in the control rod drive mechanism.
If this can be demon-strated an attempt to fully insert the control rod cannot be inserted and an investigation has demonstrated that the cause of failure is not a failed control rod drive mechanism collet housing, a shut-,
down margin test shall be made to demonstrate under this condit'ion that the core can be made subcritical for any reactivity condition during the remainder of the operating cycle with the analytically determined, highest worth control rod capable of withdrawal>
fully withdrawn, and all other control rods capable of insertion fully inserted.
The control rod accumulators shall be determined operable. at least once per 7
days by verifying that the pressure and level detectors
.are not in the alarmed
'condition.
120
'ROGANS FERRY NUCLEAR PLANT FIGURE 3.5.2 K) FACTOR.
AUTOMATIC FLOW CONTROL MANUAL FLOW CONTROL Scoop-Tube Set-Point Calibration po4ition such that Flowinox = 10%5%
i07 0 '/
ill.0%
>>r.0 I 40 50 60 j0 CORE FLOW,X 80 90
NG CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
3. 6 PRIMARY SYSTEM BOUNDARY 4 '
PRIMARY SYSTEM BOUNDARY e.
Reactor vessel bottom head temperature Reactor vessel shell adjacent to shell flange 2 ~
Durinq all operations with a critical core,,
other'han for low level physics tests, the reactor vessel shell and fluid temperatures shall be at or above the temp'erature of curve Number 3 of figure.
3. 6-1.
2 ~
Reactor vessel metal temperature at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell. adjacent to shell flange, shall be recorded at least every 15 minutes durinq inservice hydrostatic or leak testing when the vessl pressure is 312 psiq.
3.
During heatup by non-nuclear means, except when the vessel is vented or as indicated in 3. 6.A.4, during cooldown following nuclear shutdown, or during low-'level physics
tests, the reactor vessel temperature shall be at or above the tem-peratures of curve 82 of Figure 3.6.1 until removing tension on the head stud
.ol.ts as specified in 3.6.A.5.
I 3.
Test specimens representinq the reactor vessel, base
weld, and weld heat affected zone metal shall be installed in the reactor vessel adjacent to the vessel wall.at the core midplane level.
The number and type of specimens will be in accordance with GE report NED0-10115.
The specimens shall meet the intent of ASTH E 185-70.
Samples shall be withdrawn at one-fourth and three-fourths service life.
185
LIHITINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
.3. 6 PRIMARY SYSTEM BOUNDARY 4
6 PRIMARY SYSTEM BOUNDARY C.
Coolant Leaka e
Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212oF, reactor coolant leakage into the primary containment.
from unidentified sources shall not exceed 5
gpm.
In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm.
C Coolant Leaka e
Reactor coolant system leakage shall be checked by the sump and air sampling system and recorded at least once per daye 2.
With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
0 2.
Both the sump and air sampling systems shall be operable during reactor power operation.
From and after the date that one of these systems is made or found to be inoperable for 'any
reason, reactor power operation is permissible only during the succeeding seven days.
'3.
The air samplin8 system may be removed from service for a period of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for calibration, functional testing, and maintenance without pro-viding a temporary monitor.
If the condition in 1
or 2 above cannot be
met, an orderly shutdown shall be inititated and the reactor shall be shutdown in the Cold Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
191
I,IMITING CONDITIONS FOR OPERATION SURVEILLANCE REOVIREMENTS 3
6 PRIMARY SYSTEM BOUNDARY 4
6 PRIMARY SYSTEM BOUNDARY Fo Jet Pum Plow Mismatch The reactor shall not be operated with one recirculation loop out of service for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
With the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a hot shut".down condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the loop is sooner returned to
.service.
F Pet Pum Flow Mismatch 1.
Recirculation pump speeds shall be checked and logged at least once per day.
2.
Following one-pump operation, the discharge valve of the l'ow speed pump may not be opened unless the speed of the faster pump is less than 50% of its rated speed.
2.
No additional surveillance required.
3.
Steady state operation with both recirculation pumps out of service for up to 12 hrs is permitted.
During such interval restart of the recirculation pumps is permitted, provided the loop discharge temperature is within 75 F of the saturation temperature of the reactor vessel water as determined by dome pressure.
The total elapsed time in natural circulation and,one pump operation must be no greater than 24 hrs.
3 ~
Before starting either pump during steady state operation check and log the loop dis-charge temperature and the dome saturation temperature.
195
~IMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS g~R R
M 4
6 PR MARY SYSTEM 8 NO Y
G., Structural Inte rit G.
Structural Inte rit, The structural integrity of the primary system shall be maintained at the level required by the original acceptance standards throughout the life of the plant.
The reactor shall be maintained in a cold shutdown condi.tion until each, indication of 'a defect has been investigated and evaluated.
Table 4
6 A together with supplementary
notes, specifies the inservice inspection surveillance require'ments of the reactor coolant system as follows:
a.
areas to be inspected b.
percent of areas to be inspected during the inspection interval c.
inspection frequency d.
methods used for inspection 2
Evaluation of inservice inspections will be made to the acceptance standards specified for the original equipment.
3'he inspection interval shall be 10 years.
4 ~
Additional inspections shall be performed on certain circumferential pipe welds as listed to provide additional protection against pipe whip, which could damage auxiliary and control systems.
196 Fee dwater - GFw-9 ~
KFW-13, GFW-12~
GFW-26, KFW-31~ GFW-29 ~
KFW 39 GFW" s KFV >A, an l
.'a~
LrHITINr. CONDITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS 4 '
CORI'AINMENT SYST S
5.
zf speci fications 3
7 A 4 a~
b~
or.c, cannot be met, the unit shall be placed in a cold shutdown condition in an orderly manner within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
a 0 After completion of the 300-hour warranty runr conta inme nt atmosphere shall be reduced to less than 4%
oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psigi except as specified in 37 ~A5b Oxygen Concentration 5.
A leak test. of the drywell to suppression chamber structure shall be conducted during each operating cycle.
Acceptable leak rate is 0.14 lb/'sec of.
primary containment atmosphere with 1 psi differential.
Oxygen Concentration The primary containment oxygen concentration shall be measured and recorded daily.
The oxygen measurement shall be ad)usted to accourit for the uncertainty of the method used by adding a predetermined error function.
b.
Within the 24-hour period subsequent to placing the reactor in the Run mode following a shutdown, the
'ontainment atmosphere oxygen concentration shall be reduced to less than 4%
by volume and maintained in this condition.
De-inerting may commence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a shutdown.
245 b.
The methods used to measure the primary con-tainment oxygen concen-tration shall be cali-brated once every refueling cycle.
I.ttllrlt:<: COttl)lTI(t'iS IY)R OPI'.RAT(Ott SURVEILI.AtICE RFPU IRE)IEbTS
'I.')
Al:y.:LIARY ELt;(:1 R ICAL SYSTI'.t'.
4.9 AIIXILIARYELECTRlC,'L SYSTE.'I l!Il<<ra( Il<<n In (.litt( Situ(down
(.und I t ion Vh<<n<<vnr thc react, or is in thc cold sin>tdown cnndl( ion witt: Irradiated tuel In th<< r<<actor, thc a<<a()ah(llty of electric (niw<<r si<<all h<<as npeclf lcd (n Section 3.9.A cxccpt as spccificd herein.
l.
At !cast two unit 3
diesel generators and their associated 4-I V shutdown boards shall be opcrablc.
2.
An additional source of pov<<r consistinR of wibc of the foI lowInt,!
Onc lbl-kV transmission lInc and Its associated coollnn tow<<r transformer capable of supply(.np power to thc unit 3
sin! tdown boards.
A third operable d(<<! cl Rcnera(or.
At I<<as(. <<nc unit 3
4IIO V slllltdown boar(1 rust b<<operable.
Onc 480 V RHOV board motor-generator
{II-0) set is requir-ed for each RI".OV board
{D or. E) required to support operation oi the RIIR system in acco'rdance with 3.5.8.9.
326
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
3. 10 CORE ALTERATIONS 4 ~ 1 0 COR E ALTERATIONS 2 ~
Fuel shall not be loaded into the reactor core unless all contxol rods are fully inserted.
2 0 Prior to performing control rod or control rod drive maintenance on control cells without removing fuel assemblies, it shall be demonstrated that the core can be made subcritical by a margin of 0.38 percent dk/k at any time during the maintenance with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.
Alternatively if the remaining control rods are fully inserted and have had their directional control valves electr ically disarmed, it is sufficient to demonstrate that the.
core is subcritical with a margin of at
, least 0.38 percent hk/k at any time during the maintenance.
A control rod on which maintenance is being performed shall'" be considered inoperable.
3.
The fuel grapple hoist load switch shall be set at 1,000 lbs.
3.
No additional sur-veillance required.
- 332
0
LIMITINGCONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.10 CORE ALTERATIONS 4
10 CORE ALTERATIONS 4 ~
If the frame-mounted auxiliary hoist, the monorail-mounted auxiliary hoist, or the service platform hoist is to be used for handling fuel with the head off the reactor vessel, the load limit switch on the hoist to be used shall be set at ( 400 lbs.
4.
No additional surveillance re'quired.
0 5
A maximum of two nonad jacent control rods may be withdrawn from the core for the purpose of performing control rod and/or control rod drive maintenance, provided the following conditions are satisfied:
5 No additional surveillance required.
a ~
The reactor mode switch shall he locked in the nrefue position.
The refueling interlock which prevents more than one control rod from heing withdrawn may be bypassed for one of the control rods on which maintenance is being performed.
All other refueling interlocks shall be operable.
333
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
3. 10 CORE ALTERATIONS 4
10 CORE ALTERATIONS 0
6-a ~
The reactor mode switch is locked in the nxefueln position.
The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed on a withdrawn control rod after the fuel
'ssemblies in the cell containing (controlled by) that contxol rod have been removed from the reactor core.
All other refueling interlocks shall be operable.
Any number of control rods may be withdrawn ox removed from the reactor core providing the following conditions are satisfied:
6.
With the mode selector switch in the refuel mode, only one control rod may be withdrawn until two licensed operators have confirmed that all fuel has been removed from around the next rod to be withdrawn.
0 335
I T,IMITING CONDITIONS FOR OPSRhTXCN 8URVSILLhNCE REQUIREMENTS
3. 11 FIRE PPOTECTION SYSTEMS 11 RE PROl'BC%ION S STEMS B.
CO~ Fire Protection S stem 1
1.
The CO, Fire.
Protection System shall be operable:
B>>
CO Fire Protect on S stem 2
1 CO~ Fire Protection Testing:
a ~
With a minimum of 8-1/2 tons (0
5 Tank)
COg in storage units 1 and 2.
~eg~uency
a. Simulated, Once/year autowa tie.
and manual actuation b.
With a minimum of 3 tons (0.5 Tank)
CO>
storage unit 3 J
b. Storage tank pressure and level Checked daily 2 ~
c.
Automatic initiation logic operable.
I. specification
3. 11. B. 1. a or 3.11.B.1.b or 3.11.B.1.c cannot be
met, a patrolling fire watch with portable fire equipment shall be established to ensure that each area where protection is lost is checked hourly.
If specifications
3. 11. B. 1. a, 3
11 B
1 bi or 3.11.B.1.c are not met within 7 days, the affected unit(s) shall be in cold shutdown within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .
F ~
c, CO~ Spray Once/3 header and years nmxle inspection for. blockage When the cable spreading room ~z
~ire Protection i~
inoperable, one 125-pound for larger) portable fire extinguisher s) all b.
nlaced at each entrance.
351
Wl,.>I'
6.0 ADMINISTRATIWR CONTROLS B
Source Tests Results of required leak tests performed on sources if the tests reveal the presence of'0.005 microcurie or more of removable contamination.
C.
S ecial Re orts (in writing to the Director of Regional Office of Inspection and Enforcement).
1.
Reports on the following areas shall be submitted as noted:
a.
Secondary Containment 4.7.C Leak Rate Testing(5)
Within 90 days of completion
'f each test.
b.
Fatigue Usage Evaluation 6.6 Annual Operating.
Report c,
.-.Relief Valve Tailpipe Instrumentation 3.2.F Seismic Instrumentation 3.2.J.3 Inoperabi 1 ity Within 30 days after inoperability of thermocouple and acoustic monitor
,on one valve.
Within 10 days after 30 days of inoperabi litv Within 10 days after 7 days of Inoperability Meteorological Monitoring Instrumentation Inoperability
~geofel R~eort (in writing to the Dit'eetor of Regional Office of Inspection and Enforcement)
Data shall be retrieved from all seismic instruments actuated during a seismic event and analyzed to determine the magnitude of the vibratory ground motion.
A Special Report shall be submitted within 10 'days after the event describing the magnitude, frequency. spectrum, and resultant effect upon plant features important to safety.
I 386
PROPOSED CHANGES TO APPENDIX B TFCHNICAL SPFCIFICATIONS
Slctcorologkca1 data shall be summarised and reposed consistent with the rcconm ndations of Regulatory Cuide 1.2l (June 1974) and R<<ulatorj Ciddc 1.23 (February -1972),
and meteorological pbscrvati'ons egal> be recorded in a form consistent vith ltational Heather Ger vice p.ocedurcs
~
Xf the outage of any retcorological instrument(s) required by Regulatory Cuidc 1.23 (February 1972) exceeds, seven consecutiv= day", the total outcgc time, the dates of outage, the cause g the outage, and the instru-.
I vent(
) involved sha11 be reported within lO'Rays of the initiation of she ou ~e to the UShRC, Office of Inspection and inforccrent, uiih a copy to the Office of Ãuclear Peactor ReyQ.ation, Division of Operating Reactors.
ilcrr nts of this progran may be modified or terminated in accordance vith Subsection 5 6.3(c).
"'he collection of meteorological data at the plant site provides information for usc in dcve)oping atmospheric diffusion parameters for cstimct.bing potential radiation dose" to the public resulting from actual routine o."
abnorral releases of radioactive anterials to the y,trosphcre, anc for c"sessing the actual impact of the plant cooling system on the ataosph ric cnvironnent of the ite area.
A meteorological date, collection program as describ d above is necessary to neet the requirement-of subparagraph 50.36a(a)(2) of 10
~s Part 50, Appendix D to IO CFR Part 50, and Appendix F. to 10 CFB Part O.
(1)
Ref. Section 6.7.3.D Appendix A Technical Specifications.
0

ML18025B502

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