Text

Proposed TS Table 3.2.C Re Instrumentation That Initiates Rod blocks,4.2.C Re Surveillance Requirements for Instrumentation That Initiate Rod Blocks & LCO 4.3.A.2 Re Reactivity Margin - Inoperable Control Rods
	ML20102A369
Person / Time
Site:	Browns Ferry
Issue date:	07/20/1992
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML18058B967	List: ML20102A369;
References
	NUDOCS 9207270010
	Download: ML20102A369 (65)
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ENCil)SURE1 PROPOSED TECIINICAL SPECIFICATION CilANGE BROWNS FERRY NUCLEAR PLANT UNITS 1,2 AND 3 (TVA BlHP TS 310) 2 Ohhh0-PD$

l'ROl'OSED TECilNICAL Sl'ECIFICATION CllANGE IIROWNS FEltitY NUCl.EAll I'LANT UNIT 1 (TVA lith'i' TS 310)

/

i i

UNIT 1 EFFECTIVE PACE LIST REMOVE INSERT i

3.2/4.2-25 3.2/4.2-25 3.2/4.2-50 3.2/4.2-50 3.3/4.3-2 3.3/4.3-2 3.3/4.3-3 3.3/4.3-3 3.3/4.3-4.

3.3/4.3-4 3.3/4.3-5 3.3/4.3-5 3.3/4.3-6 3.3/4.3-6 3.3/4.3-7 3.3/4.3-7 3.3/4.3-8 3.3/4.3-8 3.3/4.3-9

-3.3/4.3-9 v

3.3/4.3-10 3.3/4.3-10 3.3/4.3-14 3.3/4.3-14 3.3/4.3-15 3.3/4.3-15 3.3/4.3-16 3.3/4.3-16 3.3/4.3-19 3.3/4.3-19 t

1 4

...m

.o

.=

IAhkE OF CONTEHIl Section Pate Noi 1.0-1 1.0 Definitions........

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.1/2.1-1 1.1/2.1 Fuel Cladding Integrity..

e 1.2/2.2 Reactor Coolant System Intagrity..........

1.2/2.2-1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMEHI]

3.1/4.1-1 3.1/4.1 Reactor Protection System..

3.2/4.2 Protective Instrumentation.............

3.2/4.2-1 A.

Primary Containment and Reactor Building Isolation functions.

3.2/4.2-1 B.

Core and Containment Cooling Systems -

Initiation and Control...........

3.2/4.2-1 C.

Control Rod Block Actuation.

3.2/4.2-2 D.

Radioactive Liquid Effluent Monitoring Instrumentation...............

3.2/4.2-3 E.

Drywell Leak Detection......

3.2/4.2-4 F.

Surveillance Instrumentation.....'....

3.2/4.2-4 C.

Control Room Isolation...

3.2/4.2-4 H.

Flood Protection...

3.2/4.2-4 I.

Meteorological Monitoring Instrumentation...

3.2/4.2-4 J.

Seismic Monitoring Instrumentation..

3.2/4.2-5 K.

Radioactive Gaseous Effluent Monitoring Instrumentation 3.2/4.2-6 L.

ATWS Recirculation Pump Trip.

3.2/4.2-6a 3.3/4.3 Reactivity Control.................

3.3/4.3-1 A..

Reactivity-Limitations...

3.3/4.3-1 B.

Control' Rods.................

3.3/4.3-5 C.

Scram Insertion Times...

3.3/4.3-9 BFN

'l Unit 1

~

I TABLE 3.2.C INSTRtpcENTATION THAT INITIATES ROD BLDCKS 5#

-n 5*

Miniatum Operable

. Channels Per Trio Function (5)

Function Trio level Settins 4(1)

ADRM Upscale (Flow Blas) 10.66W + 421 (2) 4(1)

APRM Upscale (Startup Mode) (8) 1121 4(1)

APRM Downscale (9) 13%

4(1)

APRM Inoperative (10b) 2(7)

RBM Upscale (Flow Bias) 10.66W + 401 (2)(13) 2(7)

RBM Downscale (9) 131 2(7)

RBM Ineperative (10c) 6(1)

IRM Upscale (8) 1103/125 of full scale y

6(1)

IRM Downscale (3)(8) 15/125 of full scale 6(1)

IRM Detector not in Startup Position (8)

(11) n 6(1)

IRM Inoperative (8)

(10a)

E 5

3(1) (6)

SRM Upscale (8) i 1X10 counts /sec.

3(1) (6)

SRM Downscale (4)(B) 13 counts /sec.

3(1) (6)

SRM Detector not in Startup Position (4)(8) (11)

,3(1) (6)

=SRM Inoperative (8)

(10a) 2(1)

Flow Bias Comparator 110% difference in recirculation flows 2(1)

Flow Bias Upscale 11151 recirculation flow 1

Rod Block Logic N/A" 1(12)-

High Vater Level in West 125 gal.

Scram Discharge Tank (LS-85-45L) 1(12)

High Vater Level in East 125 gal.

Scram Discharge Tank (LS-85-45M)

TABLE 1.2.C SURVEILLANCE REQUIREMENTS FOR INSTRUMENTATION THAT INITIATE R00 BLOCKS Function functional Test Calibratien (171 Instrument Check

2:

. APRM Upscale (Flow Blas)

(1)

(13) once/3 month once/ day (8)

APRM Upscale (Startup Mode)

(1)

(13) once/3 months once/ day (8)

APRM Downscale (1)

(13) once/3 monthe once/ day (8)

APRM Inoperative (1)

(13)

N/A once/ day (8)

RBM Upscale (Flow Blas)

(1)

(13) once/6 months once/ day (S)

RBM Downscale (1)

(13) once/6 months once/ day (8)

RBM Inoperative _

(1)

(13)

N/A once/ day (8)

(-

IRM Upscale (1)(2)

(13) once/3 months once/ day (8) i IRM Downscale (1)(2)

(13) once/3 months once/ day (8)

IRM Detector Not in Startup Position (2) (once operating cycle) once/3perating cycle (12)

N/A IRM Inoperative (1)(2)

(13)

N/A N/A

.. w I

SRM Upscale (1)(2)

(13) ence/3 months once/ day (8)

.SRM Downscale (11(2)

(13) once/3 months once/ day (8)

SRM Detector Not in Startup Position (2) (once/ operating cycle) once/ operating cycle (12)

N/A SRM Inoperative (1)(2)

(13)

N/A N/A Flow Bias Cowarator (1)(15) once/ operating cycle (20)

N/A Flow Blas Upscale.

(1)(13) once/3 months N/A Rod Block Logic-(16)

N/A N/A l'

West Scram Discharge once/ quarter once/ operating cycle N/A Tank Water Level High (LS & 5L)

East Scram Discharge once/ quarter once/ operating cycle N/A Tank Water Level High (LS-85-45M)

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPIFATION SUlyEILI.ANCE REOUIREMENTS 3.3.A.2 Reactivity marrin - inoperable 4.3.A.2 Reactivity _marr.in - in-control rods operable control roja a.

Control rod drives which can-a.

Each partially or i

not be moved with control fully withdrawn rod drive pressure shall be OPERABLE control considered inoperable.

If rod shall be a partially or fully with-exercised one notch drawn control rod drive can-at 1 cast once each not be moved with drive or week when operating scram pressure the reactor above the power shall be brought to the COLD level cutoff of the SIIUTDOWN CONDITION within 24 RWM.

In the event hours and shall not be power operation is started unless (1) investi-continuing with gation han demonstrated that three or more the cause of the failure is inoperable control not a failed control rod rods, this ttat drive mechanism collet shall be performed housing and (2) adequate at least once each chutdown margin has been day, when operating demonstrated as required above the power level by Specification 4.3.A.2.c.

cutoff of the RWM.

r BFN 3.3/4.3-2 Unit 1

- ~.

3.3/4.3 REACTIVITLf0KIEQL LIMITING CORDITIONS FQR OrrRATION SURVEILLANCE REOUIEMiENTS 3.3.A.2 Reactivity martin - inoperable 4.3.A.2 Reactivity mars.in - in-ggntrol roda (Cont'd)

Eperable control roda (Cont'd) b.

The control rod direc-b.

DELETED tional control valves for inoperable control rods shall be disarmed electrically.

c.

Control rods with scram c.

When it is initially times greater than those determined that a control permitted by Specification rod is incapable of 3.3.C.3 are inoperable, normal insertion a test but if they can be shall be conducted to inserted with control demonstrate that the rod drive pressure they cause of the malfunction need not be disarmed is not a failure in the electrically.

control rod drive mechanism.

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 control rod drive mechanism collet housing, a shutdown l

margin test shall be made to demonstrate under this condition that the core'can be made suberitical for any reactivity condition during the remainder of the operating cycle with the analytically determined highest worth i

control rod capable of

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

BTN 3.3/4.3-3 Unit 1 1

-. - = -.. _.

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.A.2 Reactivity agIgin - inoperable 4.3.A.2 Reactivity marain - in-control rods (Cont'd) operable control rods (Cont'd) d.

DELETED d.

The control rod accumulators shall be e.

Control rods with inoperable determined OPERABLE at accumulators or those whose least once per 7 days by

+

position cannot be verifying that the positively determined shall pressure and level be considered inoperable, detectors are not in the alarmed condition.

f.

Inoperable control rods shall be positioned such that $pecification 3.3.A.1 is met.

In addition, during reactor power operation, no more than one control rod in any 5x5 array may be inoperable (at least 4 OPERABLE control rods must separate any 2 inoperable ones).

If thin specification cannot be met the reactor shall not be started, or if at power, the reactor shall be brought to a shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , l.

BFN 3.3/t 4-4 Unit 1

~.

'I 3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERAllpN SURVEll1ANCI: IlLOUIRI:MENTS 3.3.B.

Control Rode 4.3.D.

Control Rods 1.

Each contral rod shall be 1.

The coupling integrity coupled to its drive or shall be verified for completely inserted and the each withdrawn control i

control rod directional rod as follows:

control valves disarmed l

clectrically. This a.

Verify that the j

requirement does not apply control rod is j

in the SHUTDOWN CONDITION following the drive when the reactor is vented, by observing any l

Two control rod drives may response in the be removed as long as nuclear instru-Specification 3.3.A.1 mentation each time is met.

a rod la moved when the reactor la operating above i

the preset power level cutoff of the RWM.

b.

When the rod is fully withdrawn the first time after each refueling outage or after 1

maintenance, observe that the drive does not go to the overtravel position.

2.

The control rod drive 2.

The control rod drive housing support system shall housing support system be in place during REACTOR shall be inspected POWER OPERATION or when the after reassembly and reactor coolant system is the results of the pressurized above atmospheric inspection recorded.

pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.3.A.1 is met.

BF1 3.3/4.3 5 Unit 1

3.3/4.3 arACTIVITY CONTROL LIMITING. CONDITIONS FOR OPERATION SURVEft1.14CE REQUIREMENTS' 3.3.B.

Control Rods 4.3.B.

Control Rods

't I

3.a DELETED 3.a DELETED 1

3.b Whenever the reactor is 3.b.1 The Rod Worth in the,startup or run modes Minimiter (RWM) shall l

below 10% rated power, the be demonstrated to be Rod Worth Minimizer (RWM)

OPERABLE for a shall be OPERABLE.

reactor startup by the following checks:

1.

Should the RWM become a.

By demonstrating inoperable after the that the control first twelve rods have rod patterns and been withdrawn, the Banked position.

t start-up may continue Withdrawal provided that a second Sequence (or.

licensed operator or equivalent) input other technically to the RWM qualified member of the computer are i

plant staff is present correctly loaded.

at the console verifying following any compliance with the loading of the prescribed control rod program.into the l

program, computer.

2.

Should the RWM be b.

Within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s-l inoperable before the prior to withdrawal l

first twelve rods are of control rods for withdrawn, start-up may the purpose of continue provided a making the reacto; l

second licensed operator critical verify-l or other technically

-proper annunciation qualified member of the of the selection plant staff is present error-of at least at the console verifying one out-of-sequence compliance with the control rod, prescribed control rod program. Use of this

c. 'Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior l

provision is limited to.

to withdrawal.of one plant startup per control. rods-for the calendar. year.

purpose of making the reactor critical ~, the rod block function of l

the-RWM shall'be verified by moving an out-of-sequence-control rod.

y BFN..

3.3/4.3-6 Unit 1 L

1 p

3.3/4.3 REACTIVfTY CONTROL LitilIlliQ QQNDITIONS EOR OPERATION SURVE1LLANCE REOUIREtiENTS 3.3.B.

Control Rode 4.3.B.

Control Rode 3 b (Cont'd) 3.b.2 The Rod Worth Minimizer (RWM) 1 Should the RWM become shall be inoperable on a shutdown, demonstrated to be shutdown may continue OPERABLE for a reactor provided that a second shutdown by the licensed operator or other following checket technically qualified member of the plant staff is present a.

By demonstrating at the console verifying that the control compliance with the rod patterns and prescribed control rod Banked Position program.

Withdrawal Sequence (or equivalent) input to the RWM computer are correctly loaded following any loading of the program into the computer.

b.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to RWM automatic initiation when reducing thermal power, verify proper annunciation of the selection error of at least j

one out-of-sequence I

control rod.

l c.

Within one hour after RWM automatic-initation when-i reducing thermal-power, the rod l

block function of the RWM shall be-verified by moving an out-of-sequence l

-control rod.

l BrN 3.3/4.3-7 Unit 1-l

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS POR OPERATION SURVEILLANCE REOUIRI:MENTS 3.3.B.

Control Rods 4.3.B. Gp.n11gl_Epfg 3.c.

If Specifications 3.3.D.3.b.1 3.b.3 When the RWM is not through 3.3.B.3.b.3 cannot OPERABLE a second be met the reactor shall.

1' tensed operator not be started, or if the other technically reactor la in the run or qualified member of startup modes at less than the plant staff shall l

10% rated power, control rod verify that the correct movement may be only by rod program is followed.

actuating the manual scram or placing the reactor mode switch in the shutdown position.

i 4.

Control rods shall not be 4.

Prior to control rod j

withdrawn for startup or withdrawal for startup refueling unless at 1 cast or during refueling, two source range channels verify that at least two have an observed count rate source range channels equal to or greater than have an observed count three counts per second.

rate of at least three counts.per second.

5.

During operation with 5.

Wnen a limiting limiting control rod control rod pattern i

patterns, as determined by exists, an instrument the designated qualified functional test of the personnel, either:

RBM shall be performed prior to withdrawal of a.

Both RBM channels shall the designated rod (s) be OPERABLE:

and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.-

or b.

Control rod withdrawal shall be blocked, l

l l

l Bf3 3.3/4.3-8 Unit 1

,a

3.3/4.3 REAGTIVITY CONTROL LJMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIRQENTS 3.3.C.

Scram Insertion Times 4.3.C.

Erram Insertion Times 1.

The average scram 1.

After each refueling insertion time, based on outage, all OPERABLE the deenergiration of the rods shall be scram pilot valve sole-scram-time tested from t

noids as time zero, of the fully withdrawn all OPERABLE control rods position with the in the reactor power nuclear system operation condition shall pressure above 800 be no greater thans pais. This testing shall be completed prior to exceeding 40%

power.

Below 10%

l power, only rods in those sequences which i

% Inserted Fro.3 Ava. Scram Inser-were fully withdrawn Fu1IV. Withdrawn tion Times (sec) in the region from 100% rod density to 5

0.375 50% tod density shall 20 0.90 be scram-time tested.

50 2.0 90 3.500 f

l l

BFN 3.3/4.3-9 Unit 1

d a

d j

s J

i 1

THIS PACE INTENTIONALLY LEFT BLANK 1

f l

l BrN 3.3/4.3-10 l

Unit 1

3.3/4.3 DAEld (C:nt'd) i 2.

Renetivity margin - inoperable control rods - Specification 3.3.A.2 requires that a rod be taken out of service if it cannot be moved with drive pressure.

If the rod is fully inserted and disarmed electrically *, it is in a safe position of maximum contribution to shutdown reactivity.

If it is disarmed electrically in a nonfully !nserted position, that position shall be consistent with the shutdown reactivity limitations stated in Specification 3.3.A.I.

This assures that the core can be shut down at all times with the remaininc i

control rods assuming the strongest OPERABLE control rod does not insert. Also if damage within the cortrol rod drive mechanism and in particular, cracks in drive internal housingh, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out.

Circumferential cracks resulting from stress-assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs.

This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods.

Limiting the period of operation with a potentially severed rod after detecting one stuck rod will assure that the reactor vill not be operated with a large number of rods with failed collet housings. The Rod Worth Minimiser is not automatically bypassed until reactor power is above the preset power level cutoff. Therefore, control rod movement is restricted and the singic notch exercise surveillance test is only performed above this power level. The Rod Worth Minimizer prevents movement of out-of-sequence rods unless power is above the preset power level cutoff.

B.

Control Rods 1.

Control rod dropout accidents an discussed in the FSAR can lead to significant core damage.

If coupling integrity is maintained, the possibility of a rod dropout accident ic eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive.

Absence of such response to drive movement could indicate an uncoupled condition. Rod position indication is required for proper function of the Rod Worth Minimizer.

To disarm the drive electrically, four amphenol type plug contactors are removed from the drive insert and withdrawal solenoids rendering the rod incapable of withdrawal. This procedure is equivalent to valving out the-drive and is preferred because, in this condition, drive water cools and minimizes crud accumulation in the drive. Electrical disarming does not.

eliminate position indication.

BFN 3.3/4.3-14 Unit 1

3.3/4.3 BASES (C:nt'd) p 2.

The control rod housing support restricth the outward movement of a control rod to less than 3 inches in the extren'ely remote F

event of a housing failure. The amount of reactivity which h

could be added by this small amount of rod withdrawal, which i.

4' less than a normal single withdrawai increment: will not contribute to any damage to the primary coolant system. The design basis is given in subsection 3.5.2 of the FSAR and the safety evaluation is given in subsection 3.5.4 This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing. Additionally, the support is not required if all control rods are fully inserted and it an adequate shutdown margin with one control rod withdrawn has been demonstrated, since the reactor would remain subcritical even in the event of complete ejection of the strongest control rod.

3.

The Rot ' rth Minimizer (RWM) restricts withdrawals and q

inserth ' d control rods to prespecified sequences. All patterna.usociated with these sequences have the characteristic that, assuming the worst single deviation from the sequence, the drop of any control rod from the fully-inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in any pellet average enthalpy in excess of 260 calories per gram. An enthalpy of 280 calories per gram is well below the level at which rapid fuel dispersal could occur (i.e., 425 calories per gram).

Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Reference Sections 3.6.6, 7.16.5.3, and 14.6.2 of the FSAR, and NEDE-24011-P-A, Amendment 17.

In performing the function described above, the RWM is not d

required to impose any restrictions at core power levels in excess of 10 percent of rated. Material in the cited reference-l shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occurring at power greater than 10 percent, regardless of the rod pattern. This is true for l

all normal and abnormal patterns including those which maximize individual control rod worth.

BFN 3.3/4.3-15 Unit i

3.3/4.3 EASES (Ccnt'd)

At power levels below 10 percent of rated, abnormal control rod l

patterns could produce rod worths high enough to be of concern l

relative to the 280 calorie per gram rod drop limit.

In this rangetheRWMconstrainsthecontrolrodsequencesandpatternsq to those which involve only acceptable rod worths.

The Rod Worth Minimiter provides automatic supervision to d

assure that out of sequence control rods will not be withdrawn or inserted; i.e.,

it limits operator deviations from planned withdrawal sequences. Reference Section 7.16.5.3 of the FSAR.

The RWM functions as a backup to procedural control of control rod sequences, which limit the maximum reactivity worth of control rods. When the Rod Worth Minimizer is out of service, special criteria allow a second licensed operator or other technically qualified member of the plant staff to manually fulfill the control rod pattern conformance functions of this system. The requirement that the RWH be OPERABLE for the withdrawal of the first twelve rods on a startup is to ensure that a high degree of RWM availability is maintained.

The functions of the RWM make it unnecessary to specify a d

license limit on rod worth to preclude unacceptable consequences in the event of a control rod drop. At low powers, below 10 percent, the RWM forces adherence to acceptable (Banked Position Withdrawal Sequence or equivalent) rod patterns. Above 10 percent of rated power, no constraint on rod pattern is required to assure that rod drop accident consequences are acceptable.

Control rod pattern constraints above 10 percent of rated power are irposed by power l

distributionrequirements,asdefinedinSections3.5.I,3.5.J.d 4.5.I, and 4.5.J of these techr.ical specifications, 4.

The Source Range Monitor (SRM) system performa no automatic safety system function; i.e., it has no scrom function.

It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second assures that any transient, should it occur, begins at or above the initini value of 10-8 of rated power used in the analyses of trannients from cold conditions.

One OPERABLE SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal.. A minimum of two OPERABLE SRMs are provided as an added conservatism.

BFN 3.3/4.3-16 Unit 1 1

~_--_.-_.-_-__m--___

-.m.--

m

_m-----

3.3/4.3 BASES (C:nt'd)

The surveillance requirement for scram testing of all the control rods after each refueling outage and 10 percent of the control rods at 16-week intervals is adequate for determining the OPERABILITY of the control rod system yet is not so frequent as to cause excessive wear on the control rod system components.

The numerical values assigned to the predicted scram performance are based on the analysis of data from other BWRs with control rod drives the same as those on Browns Ferry Nuclear Plant.

The occurrence of scram times within the limits, but significantly longer than the average, should be viewed as en indication of systematic problem with control rod drives especially if the number of drives exhibiting such scram times exceeds eight, the allowable number of inoperable rods.

l In the analytical treatment of the transienta which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a neutron l

sensor reaching the scram point and the start of control rod motion.

This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results.

Approximately the first 90 milliseconds of each of these time intervals result from sensor and circuit delays after which the pilot scram solenoid deenergizes to 120 milliseconds later, the control rod motion is estimated to actually begin. However, 200 milliseconds, rather than 120 milliseconds, are conservatively assumed for this time interval in the transient analyses and are.also included in the allowable scram insertion times of Specification 3.3.C.

t 5

l l

BFN 3.3/4.3-19 Unit 1

PROPOSED TECIINICAL SPECIFICATION CIIANGE BROWNS FERRY NUCLEAR PLANT UNIl'2 (TVA BFNP TS 310) 1

1 1

UNIT 2 ETTECTIVE PAGE LIST f

l REMOVE INSERT i

i i

3.2/4.2-25 3.2/4.2-25 1

3.2/4.2-50 3.2/4.2-50 3.3/4.3-2 3.3/4.3-2 3.3/4.3-3 3.3/4.3-3 l

3.3/4.3-4 3.3/4.3-4 3.3/4.3-5 3.3/4.3-5 j

4 3.3/4.3-0 3.3/4.3-6 3.3/4.3-7 3.3/4.3-7 2

3.3/4.3-8 3.3/4.3-8 3.3/4.3-9 3.3/4.3-9' 3.3/4.3-10 3.3/4.3-10 i

4 3.3/4.3-14 3.3/4.3-14 3.3/4.3-15 3.3/4.3-15 3.3/4.3-16 3.3/4.3-16 3.3/4.3-19 3.3/4.3.19 i

3

.I i

l.-

i R:

u m

om, w

l IABLE OF CONTENIA 1

l Section Pane No.

1 1.0 Definitions.

1.0-1 i

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l

1.1/2.1 Fuel Cladding Integrity..............

1.1/2.1-1 1.2/2.2 Reactor Coolant System Integrity.

1.2/2.2-1 LIMITING C0 EDITIONS FOR OPERATION AEll SURVEILLANCE REOUIEDEEIE i

3.1/4.1-1 3.1/4.1 Reactor Protection System..

)

3.2/4.2 Protective Instrumentation............

3.2/4.2-1 A.

Primary Containment and Reactor Building l

Isolation Functions.

3.2/4.2-1 B.

Core and Containment Cooling Systems -

Initiation and Control...........

3.2/4.2-1 C.

Control Rod Block Actuation..........

3.2/4.2-2 D.

Radioactive Liquid Effluent Monitoring Instrumentation..

3.2/4.2-3 i

E.

Drywell Leak Detection............

3.2/4.2-4 F.

Surveillance Instrumentation..

3.2/4.2-4 G.

Control Room Isolation...........

3.2/4.2-4 H.

Flood Protection...............

3.2/4.2-4 I.

Meteorological Monitoring Instrumentation...

3.2/4.2-4 J.

Seismic Monitoring Instrumentation......

3.2/4.2-5 K.

Radioactive Gaseous Effluent Monitoring Instrumentation' 3.2/4.2-6 L.

ATWS Recirculation Pump Trip.

3.2/4.2-6a 3.3/4.3 Reactivity Control...-..............

3.3/4.3-1 A.

Reactivity Limitations..

3,3/4.3-1 l

B.

Control Rods.

3.3/4.3...........-.....

l l

L.

Scram Insertion Times..

3.3/4.3-9 l

BFN i

Unit 2

1

-.>w

-p 4 Ar a

e

+1-W TABLE 3.2.C INSTRUMENTATION THAT INITIATES RCD BLOCKS l

hN

z Minimum Operable N

Channels Per Trio Function (5)

Function Trio tevel Setties 4(1)

APRM Upscale (Flow Bias) 10.58W + 50% (2) 4(1)

APRM Upscale (Startup Mode) (8) 112%

4(1)

APRM Downscale (9) 13%

4(1)

APRM Inoperative (10b) 2(7)

RBM Upscale (Flow Blas) 10.66W + 40% (2)(13) 2(7)

RBM Downscale (9) 13%

2(7)

RSM Inoperative (10c) 6(1)

IRM Upscale (B) 1108/125 of full scale w

6(4)

IRM Downscale (3)(8) 15/125 of full scale 6(1)

IRM Detector not in Startup Fos' tion (8)

(11) 6(1)

IRM Incperative (8)

(10a) w 5

3(1) (6)

SRM Upscale (8) i 1X10 counts /sec.

3(1) (6)

SRM Downscale i,4)(8) 13 counts /sec.

3(1) (6)

SRM Detector not in Startep Position (4)(8) (11) 3(1) (6)

SRM Inoperative (8)

(ICa) 2(1)

Flow Bias Comparator 110% difference in recircubtion flows 2(1)

Flow Bias Upscale 1115% recirculation flow 1.

Rod Block Logic N/A 1(12).

High Water Level in West 125 gal.

Scram Discharge Tank (LS-85-45L) 1(12)

High Water Level in East 125 gal.

Scram Discharge Tank (LS-85-45M)

TABLE 4.2.C

$URVEILLAfCE REQUIREMENTS FDR INSTRUMENTATION THAT INIT1WiE ROD BLOCKS h*

' Function Functional Test Calibration (171 Instrument Check

&N APRM Upscale (Flew Blas)

(1)

(13) once/3 months once/ day (8)

APRM Upscale (Startup Mode)

(1)

(13) once/3 months cace/ day (S)

APRM Downscale (1)

(13) once/3 months once/ day (B)

APRH Inoperative' (1)

(13)

N/A once/ day (8)

RBM Upscale (Flow Blas)

(1)

(13) once/6 months once/ day (8)

RBM Downscale (1)

(13) once/6 months once/ day (8)

RBM Inoperative (1)

(13)

N/A once/ day (8)

IRM Upscale' (1)(2)

(13) once/3 months once/ day (8)

IRM Downscale (1)(2)

(13) once/3 months once/ cay (8)

F IRM Detector Not in Startup Position (2) (once operating cycle) once/ operating de (12)

N/A IRM Inoperative (1)(2l-(13)

K/A N/A SRH Upscale (1)(2)

(13) once/3 months once/ day f3)

SRM Downscale (1)(2)

(13) once/3 months once/ day (8)

SRM Detector Not in Startup Position.

(2) (once/ operating cycle) once/ operating cy.le (12)

N/A SRM Inoperative (1)(2)

(13).

N/A N/A s

Flow Bias Comparator (1)(15) once/ operating cyc?e (23)

N/A Flow Bias Upscals (1)(15) once/3 months N/A Rod Block Logic.

(16)

N/A N/A West Scram Discharge.

once/ quarter once/18 months N/A Tark Water Level High (LS-85-45L)

East Scram Discharge once/ quarter-once/19 months N/A Tank Water Level High (LS-85-45M)

3.3/4.3 ' REACTIVITY CONTROL' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.A.2 Reactivity Mart a - Inonerable 4.3.A.2 Reactivity Martin - In-t

[ontrol Reda pagrable Control Rods a.

Control-' rod drives which can-a.--Each partially or not be moved with control fully withdrawn-rod drive-pressure shall be-OPERABLE control-considered in?perable.-- If-rod shall be a partially or fully with-exercised one notch drawn =contro1~ rod drive can-at least once each not be moved with drive or week when operating scram pressure the reactor-above the power shall be brought to the COLD level cutoff of the SHUTDOWN CONDITION within 24 RWM. -In the event hmars and shall not be-power operation is started unless (1)_investi--

continuing with-gation has demonstrated that three or more the cause of the failure is-inoperable control not a failed control' rod rods,-this test drive mechanism collet shall be performed.

d housing and (2) adequate at'least once each' shutdown margin has been day, when operating demonstrated as required above the power level by Specification 4.3.A.2.c.

cutoff of the'RWM.

i i

t-t 2

l BFN 3.3/4.3-2 Unit 2

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEYLLANCE REQUIREMENTS 3.3.A.2 Reactivity Martin - Inoperable 4.3.A.2 Reactivity Martin - In-Control Rods (Cont'd) operable Control Rods (Cont'd) h.

The control rod direc-b.

DELETED tional control valves for inoperable control rods shall be disarmed electrically.

c.

Control rods with scram c.

When it is initially times greater than those determined that a control permitted by Specification rod is incapable of 3.3.C.3 are inoper6ble, normal insertion a test but if they can be shall be conducted to inserted with control demonstrate that the rod drive pressure they cause of the malfunction

d not be disarmed is not a failure in the

% *rically, control rod drive mechanism.- 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 control rod drive mechanism collet housing, a shutdown margin test shall be made to demonstrate under this condition that the core can be made suberitical 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.

BFN 3.3/4.3-3 Unit 2 i

l l

J,. 3/4. 3 REACTIVITY CONTROL LIMITING CONDITIONS-POR OPERATION SURVEILLANCE REQUIREMENTS'~

3.3.A.2 Reactivity Martin - Inonerable 4.3.A.2 -Reactivity Martin --In-Control Rods (Cont'd)-

.Roerable Control Rods.

(Cont'd) d.

DELETED d.--The control rod accumulatorsTsha11:be.

e.

Control rods with inoperable

' determined OPERABLE at.

accumulators or those whose.

least once per_7 days _by position cannot be:

verifying that the--

positively determined shall pressure - and : level -

be-considered inoperable.:

detectors are not.in the-alarmed condition.

f.

Inoperable control rods shall be positioned such_that Specification 3.3.A.1 is met.

In addition, during reactor-

-power operation, no_more than'one control-rod-in-any 5x5-array may be inoperable-(at least 4 OPERABLE control: rods-must separate any 2' inoperable ones).? If this. specification cannot be met the reactor shall not be started, or if at power, the reactor shall-be brought to a shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

=i 1

.i BFN 3.3/4.3-4

'i Unit 2

3.3/4.3 REACTIVITY CONTRE LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.B.

Control Rods 4.3.B.

Control Rods 1.

Each control rod shall be 1.

The coupling integrity coupled to its drive or shall be verified for completely inserted and.the each withdrawn control control rod directional rod as follows:

control valves disarmed electrically. This.

a.

Verify that the requirement does not apply control rod is in the SHUTDOWN CONDITION following the drive when the reactor is vented, by observing any l

Two control rod drives may response in the be removed as long as nuclear instru-Specification 3.3.A.1 mentation each. time is met, a rod is moved when the reactor is operating above the preset power level cutoff of-the RWM.

b.

When the rod is fully withdrawn the first time after each refueling outage or after maintenance, observe that the drive does not go to the overtravel positior.

2.

The control rod drive 2.

The control rod drive housing support system shall nousing support system be in place during REACTOR shall be inspected POWER 02ERATION or when the after reassembly and reactor coolant system is the results of the.

pressurized above atmospheric inspection recorded.

with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.3.A.1 is met.-

BFN 3.3/4.3-5 Unit 2

L)/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.B.

Control Rods 4.3.B.

Control-Rods 3.a DELETED 3.a DEhETED-3.b Whenever-the reactor is 3.b.1 The Rod Vorth in the startup or run modes Minimizer (RWM) shall below 10% rated pow 6r, the

-be demonstrated to to Rod Worth Minimizer (RWM)

OPERABLE for a shall be OPERABLE.

reactor startup by the following checkst 1.

Should the RWM become a.

By e*.amanstrating inoperable after the that the control first twelve rods have rod patterns and been withdrawn, the Banked Position start-up may continue Withdrawal provided that t. second-Sequence (or-licensed operator or equivalent) input other. technically to the RWM.

qualified member of the computer are plant staff is present correctly loaded at the console verifying following any compliance with thel loadirig of the prescribed control rod-

-program into the' program.

computer.

2.

Should the RWM be b.; Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable before the prior _to withdrawal first twelve rods are of control rods for withdrawn, start-up may the purpose of l

continue provided a-making;the' reactor second licensed operator critica1Lverify.

or other technically proper-annunciation qualif'^t member of-the of the selection-plant..aff is present-error of at.least-at the console verifying one out-of-sequence compliance with the; control' rod..

prescribed control rod program. Use of this c.- - Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior -

provision is limited-to to withdrawal of one plant startup'per control rods-for the calendar-year.

-purpose of making the reactor-critical, the rod

block function of-the RWM'shall be verified byl moving an out-of-sequence-

' cont rol.; rod.

BFN 3.3/4.3-6 Unit 2

~1

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS 5'OR OPERATION SURVEILLANCE REOUIREMENTS-3.3.B.

Control Rods-4.3.B.

G.pntrol Roh 3.b (Cont'd) 3.b.2. The-Rod Worth-Minimizer (RWM)-

3.

Should the RWt1 become shall be inoperable on a-shutdown, demonstrated to be shutdown may continue-OPERABLE for a reactor provided that a second shutdown by the licensed operator or other following checks:1 technically qualified member of the plant staff is-present a.

.By demonstrating at the console' verifying that-tb-control compliance with the roi ernseand-preheribed control rod Bankeu Position-

-program.

Withdrawal. Sequence

-(or equivalent) input-to the RWM computer are correctly loaded-following_any-

-loading of the

-program into the computer.-

b.--Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to RWM-automatic =

initiation when 1

reducing: thermal power, verify l

proper annunciation

.ofLthe-salection

-error of at<1 east---

one out-of-sequence

control rod.

.: c. Within one hour'.-

after RWM automatic initation wht.n redueing thermal.

. power, the rod

-: block. function of-

-the RWM shall.be-verified by: moving-L an out-of--sequence

-contro11 rod.

a

,.g BFN 3.3/4.3-7:

. Unit 2

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.B.

Control Rods 4.3.B..QgIttrol Rods 3.c.

If Specifications 3.3.B.3.b.1 3.b.3 When the RWM is not through 3.3.B.3.b.3 cannot OPERABLE a second-be met the reactor shall licensed operator not be started, or if the or other technically reactor is in the run or qualified member of startup modes at less than the plant staff shall 10% rated power, control rod verify that the correct movement may be only by rod program is followed, actuating the manual scram

~

or placing the reactor mode switch in the shutdown position.

4.

Control rods shall not be 4.

Prior to control rod withdrawn for startup or withdrawal for startup refueling unless at least

.or during refueling, two source range channels verify that at least two have an observed count rate source range channels equal to or greater than have an observed count three counts per second, rate of at least three counts per second.

5.

During operation with 5.

When a limiting limiting control rod control rod r ttern a

patterns, as determined by exists, an instrument the designated qualified functional test of the personnel, either:

RBM shall be performed prior to withdrawal of a.

Both RBM channels shall the designated rod (s) be OPERABLE:

and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

or b.

Control rod withdrawal shall be blocked.

6 4

BFN 3.3/4.3-8 Unit 2

~ ~.. ~.-

l 3.3/4.3 REACTIVITY CONTROL-l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.C.

Scram Insertion Times 4.3.C.

Scram Insertion Times' 1.

The average scram 1.

After each refueling'

~

insertion time, based on outage, all OPERABLE the deenergitation of the rods shall be scram pilot valve sole-scram-time-tested from noids as time zero, of the fully withdrawn-all OPERABLE control rods position with the in the reactor power-

. nuclear system operation condition shall pressure above 800.

l

' be no greater thant psig. This testing l

shall.be-completed prior to exceeding:40%.

j pcVer. Below 10%

]

power, only rods in

. those sequences which -

)

% Inserted From-Ava. Scram Inser-were' fully withdrawn i

Pully Withdrawn. tion Times (sec) in the region _from:

100%-rod density to-i 5

0.375 50% rod density shall 20

- 0.90 be scram-time tested.

50 2.0 90 3.500 I

~

d t

i t,.

4.

4 4

1 BFN

3. 3/4. 3-9 --

j

' Unit 2-1 f

- -,- b,

,,,a J ~

-y y....,,,..,,,,

m -.4,s

,,..,44

,._m,e-.

-THIS PAGE INTENTIONALLY LEFT BLANK i

l i

BFN 3*3/4.3-10.

Unit 2 n.

3.3/4.3 BASES (Cent'd) 2.

Reactivity Martin - Inoocrable Control Rqda - Specification 3.3.A.2 requires that a tod be taken out of service if it cannot be moved with drive pressure.

If the rod is fully inserted and disarmed electrically *, it is in a safe position of maximum contribution to shutdown reactivity.

If it is disarmed electrically in a nonfully inserted position that position shall be consistent with-the shutdown reactivity limitations stated in Specification 3.3.A.l.

This assures that the core can be shut down at all times with the remaining control rods assuming the strongest OPERABLE control rod does not insert. Also if damage within the control rod drive mechanism and in particular, cracks in drive internal housings,.

cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out.

Circumferential etacks resulting from stress-assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs. This type of cracking could occur-in a number of drives and if the cracks propagated until severance of the' collet housing occurred, scram could be prevented in the affected rods.

Limiting the period of operation-with a potentially severed rod after detecting one stuck rod will asnure that the reactor will not be operated with a large number of rods with failed collet housings. The Rod Worth Minimizer is not automatically bypassed until reactor power is above the preset power level cutoff. Therefore, control rod movement is restricted and the single notch exercise surveillance test is only performed above this power level. The Rod Worth Minimizer prevents movement of out-of-sequence rods unless power is above the preset power level cutoff.

B.

Control Rods 1.

Control rod dropout accidents as discussed in the FSAR can lead to significant core damage.

If coupling integrity is maintained, the possibility.of a rod dropout accident is eliminated. The overtravel position feature provides-a positive check as only uncoupled drives may reach this position. Neutron instrumentation respor.se to rod movement provides a' verification that the rod is following its drive.

Absence of sudh response to drive movement could indicate an uncoupled condition. Rod position indication'is required for proper function of the Rod Worth Minimizer.

To disarm the drive electrically, four amphenol type plug. connectors are removed from the drive insert and withdrawal solenoids rendering the' rod incapable of withdrawal. This procedure is equivalent to valving out the drive and-is preferred because, in this condition, drive water cools and minimizes crud accumulation in the drive. Electrical disarming does not eliminate position indication.

BFN 3.3/4.3-14 Unit 2

3.3/4.3 BASES (C:nt'd)-

2.

The control tod housing support restricts'the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any dmnage to the primary coolant system. The design basis is given in subsection 3.5.2 of the FSAR and the safety evaluation is given in subsection 3.5.4.

This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing. Additionally, the support is not-required if all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has been demonstrated, since the reactor would remain suberitical even in the event of complete ejection of the strongest centrol rod 3.

The Red Worth Minimizer (RWM) restricts withdrawals and j

insertions of control _ rods to prespecified sequences. All patterns associated with these sequences have the-characteristic that assuming the worst single deviation from the sequence, the drop of any control _ rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in any pellet average enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories per gram is well below the level at which rapid fuel dispersal.could occur (i.e., 425 calories per gram). Primary system damage in this accident is not possible valess a significant amount of fuel is rapidly dispersed. Reference Sections 3.6.6, 7.16.5.3, and 14.6.2 of the FSAR, and NEDE-240ll-P-A, Amendment 17.

In performing the function described ebove,.the RWM is not

-l required to impose any restrictions at core power levels in-excess of 10 percent of rated. Materialinthecitedreferencel shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occurring at power greater than-all normal and abnormal patterns including thoss which maximiz 10 percent, regardless of the rod pattern. This is true for individual control rod worth.

i BFN 3.3/4.3-15 Unit 2

3.3/4.3 M EIA (C:nt'd)

At power levels below 10 percent of rated, abnormal control rod I patterns could produce rod worths high enough to be of concern relative to the 280 calorie per gram rod drop limit.

In this range the RWM constrains the control rod sequences and patterns i-to those which involve only acceptable rod worths.

The Rod Worth Minimizer provides automatic supervision to d

assure that out of sequence control rods vill not be withdrawn or inserted; i.e.,

i.t limits operator deviations from planned withdrawal sequences. Reference Section 7.16.5.3 of the FSAR.

The RWM functions as a backup to procedural centrol of control rod sequences, which limit the maximum reactivity worth of control rods. When the Rod Worth Minimizer is out of service, special criteria allow a second licensed operator or other technically qualified member of the plant staff to manually fulfill the control rod pattern confnrmance functions of this system. The requirement that the RWM be OPERABLE for the withdrawal of the first tvcive rods on a startup is to ensure that a high degree of RWM availability is maintained.

The functions of the RWM make it unnecessary to specify a d

license limit on rod worth to preclude unacceptable.

consequences in the event of a control rod drop. At low powers, below 10 percent, the RWM forces adherence to acceptable (Banked Position Withdrawal Sequence or equivalent) rod patterns. Above 10 percent of rated power, no constraint on rod pattern is required to assure that rod drop accident consequences are acceptable. Control rod pattern constraints above 10 percent of rated power are imposed by power l

distributionrequirements,asdefinedinSections3.5.I,3.5.J,d 4.5.I, and 4.5.J of these technical specifications.

4.

The Source Range Monitor (SRM) system performs no automatic safety system function; i.e., it has =no scram function.

It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per.second assures that any~ transient, should it occur, beginn at or above the initial value of 10-8 of rated power ur i in the analyses of transients from cold conditions.

One OPERABLE SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns.of scattered -

control rod withdrawal. A minimum of two.0PERABLE SRMs are provided as an added conservatism.

J BFN 3.3/4.3-16 Unit 2

3.3/4.3 BME1 (C:nt'd)

The surveillance requirement for scram testing of all the control' rods after each refueling outagc and 10 percent of the control rods at 16-week intervals is adequate for determining the OPERABILITY of the control rod system yet.is not so frequent as to cause excessive wear on the control rod system components.

he numerical values assigned-to the predicted scram performance are based on the analysis of data from other BWRs with control rod drives the same as those on Browns Ferry Nuclear Plant.

The occurrence of scram times within the limits, but significantly longer.

than the average, should be viewed as an indication of systematic problem with control rod drives especially if the number of drives exhibiting such scram times exceeds eight, the allowable number of inoperable rods.

In the analytical treatment of the transients which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of control rod motion.-

This is adequate and conservative when compared _to the typical time delay-of about 210 milliseconds _ estimated from scram _ test results.

Approximately the first 90 milliseconds,of each of these time-intervals-result from sensor and circuit delays after which the pilot scram-solenoid deenergizes to 120 milliseconds later, the control _ rod motion is estimated to actually begin. However, 200 milliseconds, rather than 120 milliseconds, are conservatively assumed for this time interval-in the transient analyses and are also included in the allowable scram insertion times of Specification 3.3.C.

l i

i l

i fFN 3.3/4.3-19 l

Unit 2 l

l

[

PROPOSED TECIINICAL SPECIFICATION CilANGE BROWNS FERRY NUCLEAR PLANT UNIT 3 (TVA BFNP TS 310) d v

y-

UNIT 3 EFFECTIVE PACE LIST 4

REMOVE INSERT-1 i

l 3.2/4.2-24 3.2/4.2-24

- 3.2/4.2-49 3.2/4.2-49 3.3/4.3-2 3.3/4.3-2

-3.3/4.3-4 3.3/4.3-4 3.3/4.3-5 3.3/4.3-5 4

3.3/4.3-6 3.3/4.3-6 l

3.3/4.3-7 3.3/4.3-7 3.3/4.3-8 3.3/4.3-8 3.3/4.3-9 3.3/4.3-9 3.3/4.3-10 3.3/4.3-10 3.3/4.3-14 3.3/4.3-14 3.3/4.3-15 3.3/4.3-15 3.3/4.3-16 3.3/4.3-16 3.3/4.3-19 3.3/4.3-19.

T r-4 4

p a

p

TAELE OF CONTEELS Section Page No.

1.0 Definitions.

1.0-1 SAFETY LIMITS MP LIMITING SAFETY SYSTEM

$JTTINGS 1.1/2.1-1 1.1/2.1 Fuel Cladding Integrity..

1.2/2.2 Reactor Coolant System Integrity.

1.2/2.2-1 LIMITING CONDITIONS FOR OPERATION AND SURVEILIANCE REOUIREMENTS 3.1/4.1-1 3.1/4.1 Reactor Protection System..

3.2/4.2 Protective Instrumentation..

3.2/4.2-1 A.

Primary Containment and Reactor Building Isolation Functions.

3.2/4.2-1 B.

Core and Containment Cooling ?ystems -

Initiation and Control.

3.2/4.2-1 C.

Control Rod Block Actuation....

3.2/4.2-2 D.

Radioactive Liquid Effluent Monitoring Instrumentation....

3.2/4.2-3 E.

Drywell Leak Detection...

3.2/4.2-4 F.

Surveillance Instrumentation.........

3.2/4'.2-4 G.

Control Room Isolation..

3.2/4.2-4 H.

Flood Protection.....

3.2/4.2-4 I.

Meteorological Monitoring Instrumentation.

3.2/4.2-4 J.

Seismic Monitoring Instrumentation......

.3.2/4.2-5 K.

Radioactive Gaseous Effluent Monitoring Instrumentation 3.2/4.2-6 L.

ATWS Recirculat$on Pump Trip.

3.2/4.2-6a 3.3/4.3-1 3.3/4.3 Reactivity Control..

A.

Reactivity Limitations.

3.3/4.3-1 B.

Control Rods........

3.3/4.3-5 C.

Scram Insertion T.imes.

3.3/4.3-9 BFN i

Unit 3

..7..._-...

.m.

4 i

l 2

TASLE 3.2.C INSTRUMENTATION THAT INITIATES R00 BLOCKS i -.

c: exs l

iE Minimum Operable

.. Channels Per 42 Trio Function (M Function Trio Level Settino 1,

.4(1)

' APRM Upscale (Flow Blas) 10.66W + 421 (2).

\\

, 4(1) '.

APRM Upscale (Startup Mode) (8)

. 112%

4(1).

1 APRM Downscale' (9) -

13% l

. i-

'4(1)

_APRM Inoperative:

(10b)

-t 7

2(7).

RBM Upscale (Flow' Bias)

,10.66W + 401 (2)(13)

L13%

2(7)'

' RBM Downscale (9).

.Y

[

2(7)E

- RBM Inocerative (10c) 6(1)

' IRM Upscale (e) 1108/125 of full scale

-6(1)n LIRM Downscale (3)(8)-

15/125 of full scale 5

'~

i 6(1)

. IRM Detector not' in' Startup Position (8). T(11) w

'k.-

~6(1)

IRM Inoperative (8)

(10a) i(

y 3(1)'(6)

SRM, Upscale (8)

I'1X105 counts /sec.

.y

{

'Z' 3(1).(6).

J SRM Downscale (4)(8)-

13 counts /sec.

4

-3(1) (6)i

,SM Detector not.in Startup Position (4)(8) (11) -

}'

3(1) (6).

Sc4 Inoperative (8)-

(10a) 2(1).

Flow Blas Comparator 110% difference in recirculation flows 2(1)

Flow Bias Upscale 711151 recirculation flow a

1 Rod Block Logic-N/A

-High Water Level in West

-125 gal.-

.- l 1(12)

Scram Discharpt Tank-

~

(LS-85-45L)-

M

[ -:,

=1(12)

' High Water Level in East 125 gal.

h

' Scram Discharge Tank.

c

.(LS-85-45M)-

.y.

}

.c,.

~

s #.. ~

... ~

. m.

-.7...~..

. ~.....

mc.

- ~

..4

(

r.:

4 TABLE 4.2.C im ^

. SURVEILLANCE REQUIREMENTS FOR INSTRUMENTATION THAT INITIATE ROD BLOCKS l

i

'@ 08 Function' Functional Test Calibration (17)

Instrument Check r-E

- i l APRM Upscale (Flow Blas)

(1)

(13) ence/3 months once/ day (8) w APRM Upscale (Startup Mode)

-(1)

(13) once/3 months once/ day (8) l 4

'APRM Downscale (1)

(13) once/3 months once/ day (8) t.

APRM Inoperative ^

,(1)

(13)

N/A once/ day (8)

.. RBM Upscale (Flow Blas)

'(1)

(13) once/6 months once/ day (8)

RBM Downscale (1)

(13) once/6 months once/ day (8)

(1)

(13)

N/A ence/ day (8) t L RDM Inoperative -

4

IRM Upscale

-(1)(2)

(13)

.once/3 months-coce/ day (8)

L b

IRM Downscale;

'(1)(2)

(13)'

once/3 months-once/ day (8)

's 1

!:IRM Detector Not.in Startup Position (2) (once operating cycle) once/ operating cycle (12)

N/A w.

-(1)(2)

(13)-

N/A N/A U

RD

?IRM Inoperative:

1 (1)(2)

(15) once/3 months' once/ day (8)

L L SRM Upscale; 1

.i, t '

J

{

j-eSRM Downscale

.(1)(2)

(13) once/3 months once/ day (8) l SRM Detector NotLin Startup Position '

~ (2) -(once'/ operating cycle) J once/ operating cycle (12)

N/A' LSRM Inoperative 5 y (1)(2)

(13)-

L N/A lN/A

~ I t

Flow Bia's Comparator L(1)(15) once/ operating cycle (20)

N/A

~

L.

F

Flow Blas: Upscale ?

_. (I)(15).

!once/3 months N/A e

. Rod Block Logic;

(16)J

.N/A N/A sWest Scram Discharge...

once/ quarter' once/ operating cycle N/A 4

ETank Water Level High' (LS-85-45L)g

East Scram Discharge once/ quarter once/ operating cycle -

.N/A-1 (Tank Waterele~el High :

(LS-85-45M)1 a

.g

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m

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, a.

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l 3.3/4;3 REACTIVITY CONTROL.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS~

3.3.A.2 Reactivity marrin - inoperable 4.3.A.2 ' Reactivity martin - in-control rods ooerable control rodg' a.

Control rod drives which can-

a. -Each partially or not be moved vm a control

-fully withdrawn rod drive pressure shall be OPERABLE control considered inoperable.-

If..

rod shall be-a partially or. fully with-exercised one notch drawn control rc2 drive can-at least'once each not be moved with drive or-

. week when operating scram pressure the reactor above the power shall:be brought to the COLD level cutoff of the

+

SHUTDOWN CONDITION within 24 RWM.

In the event l

hours and shall not be power operation is started unless (1) investi-continuing with gation has demonstrated that three or more the cause of the failure is inoperable control not a failed control rod rods, this test.

-drive mechanism collet shall be performedL housing and (2) adequate

-at leastLonce each shutdown margin has been day, when operatiIg demonstrated as required above the power level by Specification 4.3.A.2.c.

cutoff of the RWM.

b.

The control rod direc-

b. --DELETED l

ticnal control valves for inoperable control rods shall be disarmed' electrically.

l i-1 i

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1

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4 BFN 3.3/4.3-2 Unit 3 1

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3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.A.2 Reactivity margin - inonerable 4.3.A.2 Reactivity margin - in-control rods (Cont'd) operable centrol L2dA (Cont'd) d.

DELETED d.

The control rod accumulators shall be e.

Control rods wich inoperable determined OPERABLE at accumulators o; those whose least once per 7 days by position cannet be verifying that the positively determined shall pressure and level be considered inoperable.-

detectors are not in the alarmed condition, f.

Inoperable control-rods shall be positioned such that Specification 3.3.A.1 is met.

In addition, during reactor power operation, no more than one control rod in any 5x5 array may be inoperable (at least 4 OPERABLE control rods must separate any 2 inoperable ones).

If this specification cannot be met the reactor shall not be started, or if at power, the reactor shall be brought to a shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

BFN 3.3/4.3-4 Unit 3

3.3/4.3 REACTIVITY CONTROL LlHITING CONDITIONS POR OPERATION SURVEILLANCE REOUIREMENTS 3.3.B.

G2ntrol Rods 4.3.B.

Control Rods 1.

Each control-rod shs11 be 1.

The coupling integrity coupled to its drive or shall be verified for completely inserted and the each withdrawn control control rod directional rod as fo11ovas control valves disarmed electrically. This a.

Verify that the requirement does_not apply control rod is in the SHUTDOWN CONDITION following the drive when the reactor is vented.

by observing any l

Two control rod drives may response in the be removed as long as nuclear instru-Specification 3.3.A.1 mentation each time is met.

a rod is moved when the reactor is operating above the preset power level cutoff-of the RWM.

b.

When the rod is fully withdrawn the first time c

after each refueling outage er after maintenance, observe that the drive does not go to the overtravel position.

2.

The control rod drive 2.

The coutrol rod drive housing support system shall housing support system be in place during REACTOR shall be inspected.

POWER OPERATION or when the after reassembly and reactor coolant system is the results of the prescurized above atmospheric inspection recorded, pressure with fuel in the reacter vessel, unless all control rods are fully inserted and Specification 3.3.A.1 is met.

BFN 3.3/4.3-5 Unit 3 l

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.B.

Control Roda 4.3.B.

Control Rods f

3.a DELETED 3.a DELETED 3.b Whenever the reac'or is 3.b.1 The Rod Worth in the startup or run modes Minimiter (RWM) shall below 10% rated power, the be demonstrated to be Rod Worth !iinimizer (RWM)

OPERABLE for a shall be OP"RABLE.

reactor startup by the following checks:

1.

Sho"lo the RWM become a.

By demonstrating inoperable after the that the control first twelve rods have rod patterna and been withdrawn, the Banked Position start-up may continue Withdrawal provided thut a second Sequence (or licensed operator or equivalent) input other technically to the RWM qualified member of the computer are plant staff is present correctly loaded at the console verifying following any compliance with the loading of the prescribed control rod program.into the

program, computer 2.

Should the RWM be b.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable before the prior to withdrawal first twelve rods are of control rods for withdrawn, start-up may the purpose of continue provided a making the reactor second licensed operator critical verify or other technically proper annunciation qualified member of the of the selection plant staff is present error of at least at the console verifying one out-of-sequence compliance with the

' control rod.

prescribed control rod program. Use of this c.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior provision is limited to to withdrawal of one plant startup per control rods for the calendar year, purpose of making the reactor critical, the rod block function of the RWM shall be verified by moving an out-of-sequence control rod.

BFN 3.1/4.3-6 Unit 3

e 3.3/4.3 REACTIVITY CONTROL-LIMITING CONDITIONS POR OPERATION SURVEILLANCE REOUIREMENTS 3.3.B.

Control Rods 4.3.B.

Control-Rods 3.b (Cont'd) 3.b.2 The Rod Worth Minimizer (RWM) 3.

Should the RWM become shall be inoperable on a shutdown, demonstrated to be shutdown may continue OPERABLE for'a reactor provided that a second shutdown by the licensed operator or otber following checks:

technically qualified member of the plant staff is present a.

By demonstrating at the console verifying that the control compliance with the rod-patterns and preacribed control rod Banked Position program.

Withdrawal Sequence (or equivalent) input to the RWM computer are correctly-loaded following any loading of the program into the computer.

b.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to RWM automatic initiation when reducing thermal power, verify proper annunciation of the selection-error of at least-one out-of-sequence control rod.

c.

Within-one hour after RWM automatic initation when reducing thermal power, the rod block function of' the RWM shall be verified by moving an out-of-sequence

-control rod.

BFN 3.3/4.3-7 Unit 3

3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.B.

Control Rode 4.3.B. Control Rode 3.c.

If Specifications 3.3.B.3.b.1 3.b.3 When the RWM is not through 3.3.B.3.b.3 cannot OPERABLE a second be met the reactor shall licensed operator not be started, or if the or other technically-reactor is in the run or-qualified member of startup modes at less than the plant staff shall l

10% rated power, control rod verify that the correct movement may be only by rod program is followed.

actuating the manual scram

~

or placing the reactor mode switch in the shutdown position.

4.

Control rods shall not be 4.

Prior to control rod withdrawn for startup or withdrawal for'startup refueling unless at least or during refueling, two source range channels verify that at least two have an observed count rate source range channels equal to or greater than have an observed count three counts per second.

rate of et least three counts per second.

5.

During operation with 5.

Who S limiting limiting control rod cont. 1 rod pattern patterns, as determined by exists., an instrument the designated qualified funct ional test of the personnel, either:

RBM shall be performed prior to withdrawal of a.

Both RBM channels shall the designated rod (s)-

be OPERABLE:

and at least once per-24 hours thereafter.

or b.

Control rod withdrawal shall be blocked.

BFN 3.3/4.3-8 Unit 3

3.3/4.3 REACTIVITY CONTRQL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.C.

Scram Insertion Times 4.3.C.

Scram Insertion Times 1.

The average scram 1.

After each refueling insertion time, based on outage, all OPERABLE the deenergizaticn of the rods shall be scram pilot valve sole-scram-time t.eated from noids as time zero, of the fully withdrewn al? OPERABLE control rods position with the

.in the reactor power nuclear system operation condition shall pressure above 800 be no greater than:

psig. This testing shall be completed prior to exceeding 40%

l power. Below 10%

power, only rods in thosesequenceswhichd

% Inserted From Ave. Scram Insn-were fully withdrawn Fully Withdrawn tion Times (secl in the region from 100% rod density to 5

0.375 50% rod density shall 20 0.90 be scram-time tested, 50 2.0 90 3.5 BFN 3.3/4.3-9 Unit 3

e a

.a-a m-w

.Aa d

1 THIS PAGE INTENTIONALLY LEIT BLANK 4

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3.3/4.3 BASES (C:nt'd) 2.

Etactivity margin - inonerable-control rods - Specification 3.3.A.2 requires that a rod be taken out of service if it cannot be moved with drive pressure.

If the rod is fully inserted and disarmed electrically *, it is in a safe position of maximum contribution to shutdown reactivity.

If it is disarmed electrically in a nonfully inserted position, that position shall be consistent with the shutdown reactivity limitations stated in Specification 3.3.A.I.

This assures that the core can be shut down at all times with the remaining control rods assuming the strongest OPERABLE control rod does not insert. Also if damage within the control rod drive mechanism and in particular, cracks i-drive internal housings, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out.

Circumferential cracks resulting from stress-assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods.

Limiting the period of operation with a potentially severed rod after detecting one stuck rod will assure that the reactor will not be operated with a ir ge number of rods with failed collet housings. The Rod Wortt Ninimizer is not automatically bypassed until reactor power is above the preset power level cutoff. Therefore, control rod movement is restricted and the i

single notch exercise surveillance test is only performed above this power level. The Rod Worth Minimizer prevents movement of out-of-sequence rods unless power is above the preset power level cutoff.

a B.

Control Rods 1.

Control rod dropout accidents as discussed in the FSAR can lead to significant core damage.

If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives ray reach this position. Neutron instrumentation response to rod movemeat provides a verification that the rod is following its drive.

Absence of such response to drive movement could' indicate an uncoupled condition. Rod positiou indication is required for proper function of the Rod Wortn Minimizer.

To disarm the drive electrically, four amphenol type plug connectors are removed from the drive insert and withdrawal solenoids rcndering the rod incapal le of withdrawal. This procedure is, equivalent to valving out the drive and is preferred because, in this condition,-drive water cools and minimizes crud accumulation in the drive. Electrical disarming does not i

eliminate position indication.

BFN 3.3/4.3-14 Unit 3

~ --

'f j

3.3/4.3 EAEEE (C:nt'd) i 2.

The control rod housing support restricts the outward movement of a cor. trol rod to less than three inches in the extremely l

remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system.

The derian basic is given in subsection 3.5.2 of the FSAR and the safety evaluation is given in subsection 3.5.4.

This support is not required it th5 reactor coolant system is at atmospheric pressure since there vottid then be no driving force to rapidly eject e drive housing. Additionally, the support is i

not required if all control redo are fully inserted and if an i

adequate shutdown margin with one control rod withdrawn has 4

been demonstrated, since the reactor would remain subcritical even in the event of complete ejection of the otrongest control rod.

3.

The Rod Worth Minimiter (RWM) restricts withdrawals and insertions of control rods to prespecified sequences. All patterna associated with these sequences have the characteristic that, assumina the worst single deviation from the dequence, the drop of any control rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in any pellet average enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories per gram is well below the level at which rapid fuel dispersal could occur (i.e., 425 calories per gram). primary system dainage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Reference Sections 3.6.6, 7.16.5.3, and 14.6.2 of the FSAR, and NEDE-24011-p-A, Amendment 17.

In performing the function described above, the RMM is not d

required to impose any restrictions at core power levels in excess of 10 percent of rated. Material in the cited reference l

l shows that it is impossible to reach 280 calories per gram in the event of a contrcl rod drop occurring at power greater than 10 percent regardices of the rod pattern. This is true for l

all normal and abnoscal patterns including those which maximize individual control rod worth.

4 BIH 3.3/1 1-35 Unit 3

3.3/4.3 BA$rd (C:nt'd)

Atpowerlevelsbelow10percentofrated,abnormalcontrolrodl patterns could produce rod worths high enough to be of concern relative to the 280 calorie per gram rod drop limit.

In this rangetheRWMconstrainsthecontrolrodsequencesandpatternsj to those which involve only acceptable rod worths.

The Rod Worth Minimizer provides automatic supervision to i

assure that out of sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. Reference Secticn 7.16.5.3 of the FSAR.

The RWM functions as a backup to procedural control of tontrol rod sequer.;es, which limit the maximum reactivity worth of control rods. When the Rod Worth Minimir.er is out of service, special critula allow a second licensed operator or other technically qualified member of the plant staff to manually fulfill the control rod pattern conformance functions of this system. The requirement that the RWM be OPERABLE for the withdr

al-of the first twelve rods on a startup is to ensure that a L h degree of RWM availability is maintained.

The functione of the RWM make it unnecessary to specify a d

license limit on rod worth to preclude unacceptabis consequences in the event of a control tod drop. At low powers, below 10 percent, the RWM forces adherence to acceptable (Banked Position Withdrawal Sequence or equivalent) rod patterns. Above 10 percent of rated power, no constraint on rod pattern is required to a.ssure that rod drop accident consequences are acceptable. Control rod pattern constraints above 10 percent of rated power are imposed by power l

distribution requirements, as defined in Sections 3.5.I, 3.5.J d

4.5.1, and 4.5.J of these technical specifications, 4.

The Source Range Monitor (SRM) system performa no automatic safety system function; i.e., it has no scram function.

It doe 9 provide the operator with r visual indication-of neutron

uel. The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least three counts per second assures that any transient should it occur, begins at or above the initial value of 10-g of rated power b. sed in the analyses of transients from cold conditione.

One OPERABLE SRM channel would be adequate to monitor the approach to criticality uaing homogeneous patterns of scattered control rod withdrawal. A minimum of two OPERABLE ST.Ms are provided as an added conservatism.

l l

BFN 3.3/4.3-16 l

Unit 3 l

3.3/4.3 EASES (C:nt'd)

The surveillance requirement for scram testing of all the control rods after each refueling outage and 10 percent of the control rods at 16-week intervals is adequate for determining the OPERABILITY of the control rod system yet is not so frequent as to cause excessive wear on the control j

rod system components.

'the numerical values assigned to the predicted scram performance are based on the analysis of data from other BWRs with control rod drives the same as those on Browns Ferry Nucicar Plant.

The occurrence of scram times within the limits, but significant1*r longer than the average, should be viewed as an indication of systematic problem with control rod drives especially if the number of drives exhibiting such scram times exceeds eight, the allowable number of inoperable rods.

In the analytical treatment of the transients which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of control rod motion.

This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results.

Approximately the first. 90 milliseconds of coch of these time intervals result from sensor and circuit delays after which the pilot scram solenoid deenergizes to 120 milliseconds later, the control rod motion is estimated to actually begin. However, 200 milliseconds, rather than 120 milliseconds, are conservatively assumed for this time interval in the transient analyses and are also included in the allowable scram insertion times of Specificatien 3.3.C.

6 BFN 3.3/4.3-19 Unit 3

ENCLOSURE 2 REASON TIIE FOR CIIANGE, Df3CRIITION AND JUSTIFICATION BROWNS FERRY NUCLEAR PLANT (HFN)

UNITS 1,2, AND 3 (TVA BFNP TS 310)

REASON IVR TIIE CII ANGE The Rod Sequence Control System (RSCS) and Rod Worth Minimizer (RWM) are designed to mitigate the consequences of a control rod drop accident (RDA) by placing restrictions on the sequence in which control rods are withdrawn from or inserted into the core and the control rod patterns achieved during plant startup. The RSCS was requirul for BWR reactors at a time when the RDA consequences were believed to be more severe than current analyses now demonstrate. Current analyses show that the consequences of a RDA are effectively mitigated by conformance with control rod patterns equivalent to the Banked Position Withdrawal Sequence (BPWS) as enforced by the RWM. These analyses also demonstrate that the power level at which the RDA is a concern is much lower than that considered in the oiiginal analysis.

DESCRIPTION OF TIIE PROPOSED Cl!ANGE

1. The existir.g Units 1, 2, and 3 Technical Specification (TS) Tables 3.2.C, Instrumentation That Initiates Rod P, locks, and 4.2.C, Surveillance Requirements For Instrumcatation That Initiates Rod Blocks, contain an entry, "RSCS Restraint (PS85-61A,B)."

l The proposed change deletes this entry for all three units.

2. The proposed change deletes the following Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs) in their entirety for all three units:

LCDs SRs 3.3. A.2.d 4.3. A.2.b 3.3.B.3.a 4.3.B.3.a l

l

3. The existing SR 4.3.A.2.a currently reads:

Each partially or fully withdrawn OPERABLE control rod shall be exercised one notch at least once each week when operating above 30% power, in the event power operation is continuing with three or more inoperable control rods, this test shall be performed at least once each day, when operating above 30% power.

Page 2 of 8 The proposed change reads for all three units:

Each partially or fully withdrawn OPERABLE control rod shall be exercised one notch at least once each week when operating above the power level cutoff of the RWM. In the event power operation is continuing with three or more inoperable control rods, this test shall be performed at least once each day, when operating above the power !cvel cutoff of the RWM.

4. SR 4.3.B.I.a currently reads:

Verify that the control rod is following the drive by observing a resp;.ise in the nuclean instrumentation each time a rod is moved when the reactor is operating above the preset power level of the RSCS.

The following change to SR 4.3.B.1.a is proposed for all three units:

Verify that the control rod is following the drive by oburving any response in the nuclear instrumentation each time a rod is moved when the reactor is operhting above the preset power icvel cutoff of the RWM.

5. The proposed change deletes the existing text for LCO 3.3.B.3.b in its entirety and is replaces it with the followirg tml for all three units:

Whenever the reactor is in the startup or run modes below 10% rated power, the Rod Worth Minimizer (RWM) shall be operable.

1. Should the RWM become inoperable after the first twelve rods have been withdrawn, the start up may continue provided that a second lleensed operator or other technically qualified member of the plant staff is present at the console verifying compliance with the prescribed control rod program.

2. Should the RWM be inoperable before the first twelve rods are withdrawn, start-up may continue provided a second licensed operator or other technically qualified member of the plant staff is present at the console verifying compliance with the prescribed control rod program. Use of this provision is limited to one plant startup per calendar year.

3. Should the RWM become inoperable on a shutdown, shutdown may continue provided that a second licensed operator or other technically qualified member of the plant staff is present at the console verifying compliance with the prescribed control rod program.

Page 3 of 8 4

6. SR 4.3.B.3.b,1.a currently reads for all three units:

The Rod Worth Minimizer (RWM) shall be demonstrated OPERABLE for a reactor start-up by the following checks:

By demonstrating that the control rod patterns and sequence input to the RWM computer are correctly loaded followl'ig any loading of the program into the computer.

The propowd change revises this text as follows for all three units:

The Rod Worth Minimizer (RWM) shall be demonstrated to be OPERABLE for a reactor start up by the following checks:

4 By demonstrating that the control rod patterns and Banked Position Withdrawal Sequence (or equivalent) input to the RWM computer are correctly loaded following any k ading of the program into the computer.

7. The existing text for SR 4.3.B.3.b.2.a currently reads for all three units:

The Rod Worth Minimizer (RWM) shall be demonstrated OPERABLE for a reactor start up by the following checks:

By demonstrating that the control rod patterns and sequence input to the RWM computer are correctly loaded following any loading of the program into the computer.

The proposed change revises this text as follows for all three units:

The Rod Worth Minimizer (RWM) shall be demonstrated to be OPERABLE for a reactor shutdown by the following checks:

By demonstrating that the control rod patterns and Banked Position Withdrawal Sequence (or equivalent) input to the RWM computer are correctly loaded following any loading of the program into the computer.

1 l

j Page 4 of 8

8. LCO 3.3.B.3.c currently reads for all three units:

I If Speci0 cation 3.3.B.3.A through.b cannot be rnet the reactor shall not be started.

or if the reactor is in the run modes at less than 20% rated power, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the shutdown position.

The proposed change revises this text as follows for all three units:

If Specifications 3.3.B.3.b.1 through 3.3.B.3.b.3 cannot be met the reactor shall not be started, or if the reactor is in the run or startup modes at less than 10% rated power, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the shutdown position.

9. The existing text for SR 4.3.B.3.b.3 currently reads as follows for all three units:

l When the RWM is not OPERABLE a second licensed operator or other technically qualified member of the plant staff shall verify that the correct rod program is followed except as specified in 3.3.B.3.a.

The proposed revision reads as follows for all three units:

When the RWM is not OPERABLE a second licensed operator or other technically qualified member of the plant staff shall verify that the correct rod program is followed.

10. SR 4.3.C.1 currently reads for all three units:

After each refueling outage, all OPERABLE rods shall be scram time tested from the l

fully withdrswn position with the nuclear system pressure above 800 psig. This testing shall be completed prior to exceeding 40% power. Below 20% power, only rods in those sequences (An and Au or Bn ana Bu) which were fully withdrawn in the in the region from 100% rod density to 50% rod density sliall be scram-tested.

The sequence restraints imposed upon the control rods in the 100-50 percent rod density groups to the preset power level may be removed by use of the individual bypass switches associated with those control rods which are fully or partially withdrawn and are not within the 100-50 percent rod density groups. In order to bypass a rod, the actual rod axial position must be known; and the rod must be in the l

correct in sequence position.

l I

IT/

Page 5 of 8 The proposed revision to SR 4.3.C.1 reads:

After each refueling outage, all OPERABLE rods shall be scram time tested from the fully withdrawn position with the nuclear systern pressure above 800 paig. This testing shall be completed prior to exceeding 40% power. Iklow 10% power, only rods in those sequences which were fully withdrawa in the it gion from 100% rod density to 50% rod density shall be scram tested,

11. BASES 3.3/4.3.A.2 for all three units presently reads in part:

... The Rod Sequence Control System is not automatically bypassed until reactor power is above 20 percent power. Therefore, control rod movement is restricted and the single notch exercise survel lance test is only performed above this power !cvel.

The Rod Sequence Control System prevents movement of out-of sequence rods unless power is above 20 percent.

The proposed revision revises this text as follows for all three units.

... The Rod Worth Minimizer is n'>t automatically bypassed until reactor power is above the preset power level cutoff. Therefore, control rod movement is restricted and the single notch exercise surveillance test is only performed above this power level. The Rod Worth Minimirer prevents inovement of out of sequence rods unless power is above the preset power level cutoff.

12. BASES 3.3/4.3.B.1 for all three units presently reads in part:

... Rod position indication is required for proper function of the Rod Sequence Control System and the rod worth minimizer.

The proposed revision to BASES 3.3/4,3.B.1 for all three units reads ns follows:

... Rod position indication is required for proper funcdon of the Rod Worth Minimizer.

13. A proposed revision to BASES 3.3/4.3.B.3 reads as folbws for all three units:

The Rod Worth Minimizer (RWM) restricts withdrawals and lar,crtions of control rods to prespecified sequences. All patterns... Reference Sections 3.6.6,7.16.5.3, and 14.6.2 of the FSAR, and NEDE-240ll-P A, Arrendment 17.

[

Page 6 of 8 In performing the funcdon described above, the RWM is not required to impose any restrictions at core power levels in excess of 10 percent of rated. Material in the cited reference shows that it is imoossible to reach 280 calo.*ies per gram in the event of a control rod drop occurring at power greater than 10 percent, regardless of the rod pattern. This is true for all r.ormal and abnormal patterns including those v:hich maximize individual control rod worth.

At power levels below 10 perwnt of rated, abnormal control red patterns could produce rod worths high enougn to be of concern relative to the 280 calorie per fram rod drop limit. In this range the RWM constrains the control rod sequences and patterns to those which involve only acceptable rod worths.

The Rod Worth Minimizer provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. Reference Section 7.16.5.3 of the FSAR. The RWM functions as a backup to procedural control of control rod requences, which limit the maximum reactivity worth of control rods. When the Rod Worth Minimizer is out of service, special criteria allow a second licent.ed operator or other technically qualified member of the plant staff to manaally fulfill the contrul rod pattern conformance functions of this system. The requirement that the RWM be operable for the withdrawal of the first twelve rods on a startup is to ensure that a high degree of RWM availability is maintained.

The functions of the RWM make it unnecessary to specify a license limit on rod worth to preclude unacceptable consequences in the event of a control rod drop. At low powers, below 10 percent, the RWM forces adherence to acceptable (Banked Position Withdrawal Sequence or equivalent) rod patterns. Above 10 percent of rated power, no constraint on rod pattern is required to assure that rod drop accident consequences are ecceptable. Control rod pattern constraints above 10 percent of rated power are imposed by power distribution requirements, as defined in Sections 3.5.1,3.5.J,4.5.1, and 4.5.J of the!- technical specifications.

14. The last two paragraphs of BASES 3.3.C/4.3.C presently read:

In order to perfoim scram testing...

... In addition, RSCS will, prevent movement of rods in the 50 percent density to preset power level range until the scrammed rod has been withdrawn.

The proposed revision deletes these two paragraphs for all three units.

Page 7 of 8 JUSTIFICATION FOR TIIE PROPOSED CIIANGE The purpose of this proposed technical specification change is to eliminate the requirement

for use of the Rod Sequence Control System Control System and to decrease the power level setpoint above which the Rod Worth hiinimirer (RWM) would no longer be required to be used from the existing 20 percent rated power setpoint to a new setpoint of 10 percent rated power. This change is applicable to DFN Units 1, 2, and 3. These proposed technical specincation amendments are based on and are consistent with the NRC Safety Evaluation issued on December 27,1987 which approved Amendment 17 of General Electric Topical report NEDE-24011-P-A, ' General Electric Standard Application for Reactor Fuel."

The RSCS restricts control rod movement to minimire the individual rod worth of control rods to lessen the consequences of a Rod Drop Accident (RDA). Control rod movement is restricted through the use of rod select, insert, and rod withdrawal blocks. The RSCS is a hardwired (as opposed to a computer controlled), redundant system to the RWM. It is independent of the RWM in terms of inputs and outputs, but the two systems are compatible.

The RSCS is designed to monitor and block, when necessary, operator control rod selection, withdrawal and insertion actions, and thus assists la preventing significant control rod pattern errors which could lead to high reactivity worth (if dropped). A significant pattern error is one of several abnormal events which must occur to lave a RDA which might exceed the fuel enthalpy criteria for the event. The RSCS was designed only for possible mitigation of the RDA an<! is active only during low power (curready less than 20 percent rated pcwer) when a RDA might be significant. It does not prevent a RDA. A similar pattern control function is provided by the RWM, a computer-controlled system.

In response to NEDE-240111 P A submitted by the BWR Owner's Group, the NRC staff issued a safety evaluation (A. C. Thadani to J. S. Charnley, Acceptance for keferencing of Licensing Topical Report NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel," Revision 8, Amendment 17) approving the methodology for 1) elimination of the RSCS while retaining the RWM to provide backup to the operator for control rod pattern control and 2) lowering the setpoint for cutoff of the RWM to 10 percent rmed power from its current 20 percent level. This safety evaluation concluded that the proposed changes were acceptable provided:

1. The TSs should require provisions for minimizing operations without the RWM operable.

2. The occasional necessary use of a second operator replacement should be strengthened by a utility review of re!cvant procedures, related forms, and quality control to assure that the second operator provides an effective ar.d truly independeat monitoring process. A discussion of this review should accon pany the request for RSCS removal.

Eaclosure 2 Page 8 of 8

3. The rod patterns used sould be at least equivalent to Banked Position Withdrawal Sequence (BPWS) patterns.

With respect to item 1. above, the proposed TSs allow only one reactor startup per calendar year with the RWM inoperable prior to or during the withdrawal of the first twelve control rods. This will ensure that operations with the RWM inoperable are minimized.

These provisions are modeled after provisions previously found to be acceptable by the NRC staff for the application of the results of the topical report. These provisions address the need to promote effective maintenance. ' tl.c RWM by severely limiting operation with the system bypassed. Commencement of a reactor startup with an inoperable RWM is generally not allowed, with a once per calendar year exemption to allow for unusual or abnormal situations. However, once a reactor startup has commenced and significantly progressed, specincally after twelve rods are withdrawn, the evolution may be completed using the verification provisions. BFN believes that these provisions provide strong incentive for RWM maintenance without engendering excessive operationa4 restrictions and that, therefore, item

1. is adequately addressed.

Regarding item 2. above, the requirements for rod selection and rod motion verification aloi.g with the specific actions expected of the verifier are in place at BFN. 2 0185, Control Rod Drive Control System Operating Instruction and Surveillance Instruction 2-SI-4.3.B.3.b.3, RWM Progmm Verification address the administrative requirements for rod motion verification when the RWM is bypassed or inoperable for any other reason. The following controls are included in these instructions:

n Bypass of the RWM may only be performed at the direction of the Shift Operations o

Supervisor.

Whenever the RWM is bypassed or inoperable, proper rod motion is verified as each o

cor. trol rod movement is accomplished, Controls to ensure that the proper control rod movement data sheet is utilized.

o Identification of the technically qualified members of the plant staff which may be o

utillral for rod program verification (currently limited to a nuclear engineer or STA).

With respect to item 3), BpWS patterns are in use at BFN. The proposed changes to TS surveillance requirements 4.3.B.3.b.1.a and 4.3.B.3.b.2.a require that the BPWS pattern or 3

equivalent be correctly loaded into the computer as a condition for RWM operability.

TVA believes that the requirements of the NRC safety evaluation of December 27,1987 have been addressed and the proposed changes are accepta' ole.

ENCLOSURE 3 PROPOSED NO SIGNIFICANT liAZARDS CONSIDERATIONS DETERMINATION BROWNS FERRY NUCLEAR PLANT UNITS 1,2, AND 3 (TVA BFNP TS 310)

DESCRIPTION OF TIIE PROPOSED TECIINICAL SPECIFICATION CIIANGE The purpose of this proposed technical specification change is to climinate the requirement for use of the Rod Sequence Coatrol System and to decrease the power level setpoint above which the Rod Worth Minimizer (RWM) would no longer be required to be used from the existing 20 percent rated power setpoint to a new setpoint of 10 percent rated power. This change is applicable to BFN Units 1,2, and 3. These proposed technical specification amendments are based on and are consistent with the NRC Safety Evaluation issued on December 27,1987 which approved Amendment l7 of General Electric Topical report NEDE-240ll P A, ' General Electric Standard Application for Reactor Fuel."

BASES FOR PROPOSED NO SIGNIFICANT llAZARDS CONSIDERATIONS DETERMINATION NRC has provided scandards for determining ivhether a significant hazards consideration exists as stated in 10 CFR 50.91(c). A proposed amendment to an operating licenw involves no significant hazards considerations if operation of the facility in accordance with the propos:d amendment would not (1) in, olve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed TS change is judged to involve no significant hazards considerations based on the following:

l. The proposed amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated.

Eliminating the RSCS and decreasing the RWM setpolut have no effect on the probability of any previously evaluated accident because these systems play no role in any accident initiating mechanism. These systems act to mitigate the consequences of the rod drop accident (RDA). The probability of an RDA is dependent only on the control rod drive system and mechanisms themselves, and not in any way on the RSCS or RWM. Therefore the proposed changes do not involve a significant increase in the.

probability of any accident previously evaluated.

f

Page 2 of 3 A study of the RDA sponsored by the BWR Owner's Group (NEDE-240ll-A-P) has concluded that the RSCS is unnecessary. This study was approved by the NRC in a safety evaluation dated December 27,1987. The RSCS functions as a redundant system to the RWM. As long as the RWM is operable, the RSCS is not needed since the RWM presents control rod pattern errors, in the event the RWM is unavailable, the proposed technical specifications require that control rod movement and compliance with the prescribed control rod pattern be verified by a second licensed operator or other technically qualified member of the plant staff. In addition, to further minimite control rod movement at low power with the RWM out of service, the proposed technical specifications permit only one plant startup per year with the RWM out of service prior to or during the withdrawal of the 11rst twelve control rods. Therefore, the consequences of an RDA as previously evaluated will not be increased as a result of the climination of the RSCS.

The effects of a RDA are more severe at low power levels and are less severe as power level increases. Although the original calculations showed that no significant RDA could occur above 10% power, the NRC required that the generic BWR technical specifications be written to require operation of the RWM below 20 percent power t.o account for uncertainties in thu analysis. Recently, more refined calculations conducted for the NRC (NUREG 28109, " Thermal Hydraulic Effects on Control Rod Drop Accident in a BWR') have shown that even with the maximum single control rod position error, and most multiple control rod pattern errors, the peak fuel rod enthalpy reached during an RDA from these control rod patternt would not exceed the NRC limi'. of 280 caloriet per gram for RDAs above 10 percent power. These more recent calculations corroborate the original GE analyses. Therefore, the proposed decreased setpoint for the RWM will not result in a significant increase in the consequences of any accideut previously evaluated.

2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Operation of the RSCS.md RWM cannot cause or prevent an accident. These systems function to minimite the consequences of a RDA. The RDA is evaluated in the FSAR, and the effect of the proposed changes are discussed in item 1) above..

Elimination of the RSCS and decreasing the RWM setpoint will have no impact on the operation of any other systems, and therefore would not contribute to a malfunction in any other equipment nor create the possibility for an accident to occur which has not previously been evaluated.

Page 3 of 3

3. The proposed amendment does not involve a significant reduction in the margin of safety.

Elimination of the RSCS will not result in a significant reduction in the margin of safety for the reasons discussed in Item 1. above and summarized below:

a. NRC and industry studies have demonstrated that the possibility of a RDA resulting in unacceptable consequences is so low as to negate the requirement for the RSCS,

b. Current calculations have shown that the consequences of an RDA are acceptable above 10 percent power.

c. The RSCS is redundant in function to the RWM. Eliminating the RSCS does not eliminate the control rod pattem monitoring function performed by the RWM.

d. To <nsure that RWM unavailability will be minimized, the proposed technical specification changes allow only one startup per calendar year with the RWM out of service prior to or during the withdrawal of the first twelve control rods. If the RWM is out of service below 10 percent power, control rod movement and compliance with prescribed control rod patterns will be verified by a second licensed operator or other technically qualified member of the plant staff.

No significant reduction in the margin of safety will result from decreasing the RWM setpcint from 20 percent power to 10 percent power because calculations have shown that even with the maximum single control rod position error, and most multiple control rod pattern errors, the peak fuel rod enthalpy reached during an RDA from these control rod patterns would not exceed the NRC limit of 280 calories per gram for RDAs above 10 l

percent power.

CONCLUSION TVA has evaluated the proposed amendment described above against the criteria given in 10 CFR 50.92(c) in accordance with the requirements of 10 CFR 50.91(a)(1). This evaluation has determined that the proposed amendment will nel (1) involve a significant increase in the probability or conseauences of an accident previously evalested, (2) create the oossibility for a new cr different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a inargin of safety. Thus, TVA has concluded that the proposed rmendment does not involve a significant hazards consideration.

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ML20102A369

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