ML20063N164

Proposed Tech Spec Changes to Amend Licenses DPR-33,DPR-52 & DPR-68,making Administrative Changes & Updating Requirements

ML20063N164
Person / Time
Site:	Browns Ferry
Issue date:	09/14/1982
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20063N162	List: ML20063N161 ML20063N164
References
NUDOCS 8209200041
Download: ML20063N164 (32)

Text

.

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS (TVA BFNP TS 176)

BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 i

4 i

i i

i l

i i

l l

l l

i i

i B209200041 820914 j PDR ADOCK 05000259 P PDR l

r

-,c..

NOTES FOR TABLE 3.2.C

1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function. The SRM, IRM, and APRM (Startup Mode), blocks need not be operable in "Run" mode, and the APRM (flow biased) and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter.

If this condition lasta longer than seven days, the system with the inoperable channel shall be tripped. If the first column cannot be met for both trip systems, both trip systems shall be tripped.

Trip

2. W is the recirculation loop flow in percent of design.

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

A ratio of.FRP/CMFLPD<l.0 is permitted at reduced powcr. See specification 2.1 for APRM control rod block cetpoint.

3. IRM downscale is bypassed when it is on its lowest range.

k. SRM's A and C downscale function is bypassed when IRM's A, C, E, and C are above range 2. SRM's B and D downscale function is by-passed when IRM's B, D. F, and H are above range 2.

SRM detector not in startup position is bypasned when the count ,

rate is k 100 CPS cc the above condition is satisfied.

one APRM or IRM or RBM, per One instrument channel; i.e.,

5. /

trip . system may be bypassed except only one of 'four SRM may (-

be bypassed. Refer to Section 3.10.D for SRM requirements during

. core alterations. .

6. IRM channe s A, E, C, G, all in range 8 bypasses SRM channels A & C functions.

IRM channels B, F, D, H, all in range 8 bypasses SRM channels B & D functions.

7. The following operational restraints apply to the RBM only.

a. Both RBM channels are bypassed when reactor power is 6.30%.

b. The RBM need not be operable in the "startup" position of the reactor mode selector switch. . - -

c. Two RBM channels are provided and only one of these any be bypassed from the conele. if the inoperable channel cannot be restored within, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel'shall be placed in t'he tripped.cond.ition within one hour.

If minimual conditions for Table 3 2.C are not met, administrative lt d.

controls shall be immediately imposed to prevent onntrol rod withdrawal.

74

e

~

/ ?"

MOTt$ FoR TABLES 4.2.A THROUCH 4.2.H (Continue $)

l f 14 Upscale trip is functionally tested during functional test time as required by section 4.7.8.1.a and 4.7.C.1.c.

15. The flow bias comparator will be tested by putting one flow unit in "Te s t (producing 1/2 scram) and adjusting the test input to obtain comparator rod block. The flow bise upscale will be verified by ober.rving a local upscale trip light during operation and verified that it will produce a rod block during the operating cycle. .

16. Performed during operating cycle. Fortions of the logic is checked more f requently during functional teste of the functions that produce

. a rod block.

l

17. This calibration consists of removing the function from service and performing an electronic calibration of the channel.

18. Functional test is limited to the condition where secondary containment integrity is not required as specified in sections 3.7.C.2 and 3.7.C.3.

19. Tunctional tcat is limited to the time where the SCT5 is required to meet the requirements of section 4.7.C.1.e,.

20. Calibration of the comparator requires the inputs from both recirculation loops to be interrupted, thereby removing the flow bias signal to the APRM and ret and scramming the reactor. This calibration can only be performed durtag an outage.

21. Logic test is limited to the time where actual operatien of the equipment ('-

is permissible.

22. One channel of either the reacter zone or refueling zone Reactor Building Ventilation Radiation Monitoring System may be administrative 1y bypassed for a period cot to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for functional testing and calibration.

23. The Reactor Cleanup System Space Temperature eenitors are RTD's that feed a temperature switch in the control room. The temperature switch may be tested conthly by using a simulated signal. The RTD itself is a highly

, reliable instrument and lese f requent testing is necessary.

24. This instrument check consists of comparing the thermocouple readings for all valves for consistence and for noninal expected values (not required during refueling outages).

25. During each refueling outage, all acoustic monitoring channels shall be calibrated. This calibration includes verification of accelerometer response due to mechanical excitation in the vicinity of the sensor.

26. This instrument check consists of comparing the background signal levels for all valves for consistency and for nominal expected values (not required during refueling outages) .

110 L

i A

~ nasts J.2 provided to detect and temperature inst rumentation art-last rut.entation re-The HPCI hip,h (;ow steam piping. Tripping ofTripping tbts logic for the high a break in the HPCI suits in actuattun of HPC1 isolation valves.and all sensors are required to be nperable.

flow is a 1 out of 2 logic, is sensed by 4 Illgh t emperature in the vicinity of the HPCI h equipnentThe 16 traperature evitches witcher. In each trip sets of 4 biectallic temperature switc es.

are arranged in 2 trip systems with 8 temperature s system. 20C*P fo: high ten-The HFCI trip settings of 90 psi for high flow er.dcore uncovery is prevented a perature are such that release

• La within 11M ts.

and temperature instrumentation are arranged the samt The RCIC high flow trip setetng of 450" 620 ( r high flow and C as that for the HPCI. The 200*P for tenacrature are based on the same criteria as the HP 1 ld indicate lingh temperature at the Reactor Cleanuo Syst,ee floor drain coukhen high temperat a break in the c leanup syst em, system is isolated.

dual The bus systen.

instrunentattnn which initiatesh C5CSevenaction t during as arranctd in the Spectitcation preserves the ef fectiveness of t e cys d. en An exception to periods this is when when natntenance or testing is betnr. perforneloCic functio excessive centrol funcLAons are provided to to prevent 1.07. The trip logic The control rod block decrease rod withdrawal so that1MCPR out ofdoes n: not e . g . , any t r ip on one o f s ix APitzt 's ,

for this function is or f our SRtt's will re sult in a rod block.

eight 1R.'1's in s t rurien t a-channel requirements assure sufficient The etntmur instr, ment The ninteum in s t rustent the s tregle f ailure criteria is ces.

tiun to assuie cleannel requirement s for the RM1 cay be reduced 'oy one for salnt testing, or cellbration. other channel is available. l of "an inadvertent control' rod withdrawal, as the'and the R control rods. is flow biased and prevents a 41 Anificant reddc-reduced flow. The APM pro-The APRM rod block function tion in MCPR , cepecially during operation i.e., atlisitte the gross core power increase vides gross core protection. The from withdrawal of control rods in the normal is esintained witheraval greater sequence.

than 1.07.

trips are set so (54t MCPR local protection of the core; i.e.,

The RSH rod block function provides in a local rcF, ton of the core, for a the prevention of critical power single rod withdrawal error from e limiting control rod pattern.

113

?

i l

l TA3LE 3.7.A FANARY CO!(TAIETT 1501ATION VALVES Nu:2ber of Power Ma xt:nua Action en

_ Operated Valves Operating Normal toitiattog Croup Valve Identt'icaties Inbcard_ Outboard Tine (sec.) Fosition $12aal 1  % 10 steamitne isolatton valves 4 4 3eT<5 0 GC (FCV-1-14,26,37,&51 jl-15, 27, 38, & 52) 1 Main steamitae drain isolation 1 1 15 C SC s valves PCV-1-55 & 1-56

]

1 Reactor Vater sample line isola- 1 1 5 ~~ C SC f i

tion valves 3 -

2 R1tAS shutdown cooling supply isolatico valves FCV-74-48 & 47 1 1 40 C SC 2 RJDtS - LPCI to reactor FCV-74-53, 67 2 30 C SC 2 Reactor vessel head spray 1361a-tion valves FC'i-74-77, 78 1 1 10 C SC i

! 2 RitRS flush and drain vent to supprecolon charaber 4 20 C SC l

FCV-74-102, 103, 119, & 120 2 Suppression Chaeber Drain 2 15 C 3C FCV 75-57,58 .

2 Drywatt equipment drain discharge tulation valves FCV-77-15A, & ISE 2 15 0 cc 2 Dry.rell floor drata discharge isolatlun valves FCV-77-2A & 28 2 15 0 GC l

l l

[ _ _ _

TABLE 3.7.A (Continued)

Action os .

hxiana hcber o' Fover Operating Normal Initiating Operated Valves Position _

Signal Time (sec.)_ _

inboard Outboard Grous Talve Identification 0 CC 3

Reactor water cleanup system supply 1 1 30 isolation valves FCV-69-1, & 2 0 GC 1 ,

10 0 CC 4

FCV 73-81 (Eypass around FCV 73-3) 20 1 1 4 HFCIS steamline isolation valves PCV-73-2 & 3 0 CC ,

1 15 1

5 RCICS steamline isolation valves

, FCV-71-2 & 3 3 C

)

sc 6 Dryvell nitrogen purge inlet isola- 1 5

' tion valves (FCV-76-18)

SC

! 5 C

6 Suppression chamber nitrogen purge 1 talet isolation valves (FCV-76-19)

C SC 6 Dryvell h in Exhaust isolation 2 2,5 valves (FCV-66-29 and 30)

SC 2.5 C 6 Suppression chamber main exhaust 2 .

isolatina valves (FCV-64-32 and 33)

SC C

6 Dryv411/ Suppression Chamber purga 1 2.5 inlet OcV-64-17) 8C C

6 Dryw11 Atuadphere purge inlet '

1 2.5 DCV-64-14) l O G .

l l

I. .

3 9-TABLE 3.7.A(Continued) .

Il i

4 Nunber of Power Maximum Actten en y . Operated Valves Operating Nonnal Initiating .

Group Valve Identification Inboard Outboard Time (sec.) Position Signal y

/ 6 Suppression Chanter purge inlet (FCV-64-19) I 2.5 C SC 6 Drywell/ Suppression Chanber nitro-

• genpurgeinlet(FCV-76-17) 1 5 C SC I 6. Drywell Exhaust Valve Bypass to Standby Gas Treatment System

( FCV-64-31) 1 C SC 5

6 Suppression Chaaber Exhaust Valve i

" Bypass to Standby) Gas Treatment System (FCV-64-34 1 5 C SC ;

, 6 Drywell/ Suppression Chamber Nitrogen Purge Inlet (FCV-76-24) 1 5 C SC -

l i

i 6 Systen Suction Isolation Valves to Air Compressors "A" and "B" 2 15 0 GC (FCV-32-62, 63) 0 GC 7 RCICSteamlineDrain(FCV-71-6A,68) 2 5 7 RCIC Condensate Pusp Drain (FCV-71-7A, 78) 2 5 C SC l 7 HPCI Hotwell pung discharge isola-l: 2 5 C SC tion valves (FCV-73-17A 178) 5 0 GC 7 HPCI steamline drain (FCV 73-6A, 6B) 2 ,

t 4

3 I per guide .NA C GC 8 TIP Suide Tubes (5) tube 2

I

TAB:.E 3.7.A (Continued) ,

Action en hxinu:3 ' Initiating Nu.ber of Pow r Operating Normal

• Operated Valves _ Signal _

~

Tire frec.) Position _

Valve Identificatio Ink ard, Ou%ocrd Group.

C Process Standby 11guld control system 1 1 NA check valves CV 63-526 & 525 0 Frocess 2 NA 2

Feednter check valves

'CV-3-558, 572, 554, & 568 Process 0

Control rod hydraulic. return 1 1 KA check valves CV-85-576 & S73 C Process RH2S - 1.PCI to reactor check NA 2

v.1ves CV-74-54 & 68 o '

10 C SC 2

6 CAD System Torus /Drywell Exhaust i to Standby Cas Treatment (FCV-84-19,20)

NA C Process Core Spray Discharge to Reactor 2 Check Valves FCV-75-26,54 4

[ .

8,

~

TABLE 3.7.A (Continued) ,

Number of Power Maximum Action on Operated Valves Operating Normal Initiating Group Valve Identification Inboard Outboard Time (sec.) Position Signal __

6 Drywell AP air compressor suction 1 10 C SC valve (FCV-64-139) 6 Drywell AP air compressor discharge 1 10 C SC valve (FCV-64-140)

D 6 Drywell CAM suction valves 2 10 0 GC y (FCV-90-254A and 254B) 6 Drywell CAM discharge valves 2 10 0 GC (FCV-90-257A and 257B) 6 Drywell CAM suction valve 1 10 0

~

GC (FCV-90-255)

O

. o 6.0 ADMINISTRATIVE CONTROLS -

C.

Drills on actions to be taken under emergency conditions involving release of radioactivity are specified in the radiological emergency Annual drills sna11 also be plan and shall be conducted annually. conducted on the actions to be taken related systems or components.

D. Radiation Control Procedures Radiation Control Procedures shall be shall These procedures maintained and made available show permissible to all station personnel.

radiation exposure and shall Le consistent with the requirements of 10 CFR 20.

This radiation protection program shall be organized to meet the requirements of 10 CFR 20 except in lieu of the " control device" or " alarm signal" required by paragraph 20.203 (c) of 10 CFR 20:

1. Each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and co posted as a high radiation area and entrance thereto shall spicuously Any be controlled by requiring issuance of a Special Work Permit'.

individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area.

A radiation monitoring device which continuously integrates the (

b.

radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made af ter the dose rate level in the area has been established and personnel have been made knowledgeable of them, c.

An individual qualified in radiation protection procedures who This is' equipped with a radiation dose rate monitoring device.

individual shall be responsible for providing posicive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the f acility ,

HedbthPhysicist in the Special Work Permit.

2. Each high radiation area in which the intensity of radiation is greater than 1,000 mrem /hr shall be subject to the provisions of (1) above; and, in addition, access to the source and/or area shall be secured by lock (s).

The key (s) shall be under the administrative control of the shift engineer.

In the case of a high radiation area established for a period of 30 days or less, direct surveillance to prevent unauthorized entry may be sub-stituted for permanent access control, s.Health Physics personnel, or personnel escorted by Health Physics personnel, in accordance with approved emergency procedures, shall be exempt from the SWP issuance requirement during the performance of their assigned radiation [

pratection duties, provided they comply with approved radiation protection L' procedures for entry into high radiation areas. l 1

339 i

i

6.0 ADriINISTRATIVE CONTROLS B. Source Tests j Results of required leak tests performed on sources if the tests reveal the presence of'O.005 microcurie or more of removable contamination.

C. Special Reports (in writing to the Director of Regional Of fice of Inspection and Enforcement) .

1. Reports on the following areas shall be submitted ae noted: ~

a. Secondary Containment 4.7.C Within 90 Leak Rate Testing (5) days of completion of each test.

b. Fatigue Usage 6.6 Annual Evaluation Operating Report

c. . Relief Valve Tailpipe 3.2.F Within 30 days Instrumentation after inoperability

, , I of thernocouple and acoustic monitor on one valve.

d. Seismic Instrumentation 3.2.J.3 Within 10 days Inoperability af ter 30 days of inoperability

e. Meteorological, Monitoring 3.2.1.2 Within 10 days Instrumentation after 7 days of Ir. operability inoperability

f. Primary Containment 4.7.A.2 Witnia 90 <tays Integrated Leak Rate of completion of Testing each test.

g D. Special Report (in writing to the Director of Region'ai Office of Inspection and Enforcement)

Data sha'll be retrieved from all seismic instruments actuated during a seismic event and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be submitted within 10 days after the event describing the magnitude, frequency spectrum, a'nd resultant effect upon plant features important to safety.

356

" -= .

j m

e 4

,. 1 UNIT 2

's-l 4

I l

l l

l l

2 f

f i

i l

f l

f l -

S . ,

b

. L89IEE F0fLT^u t.E 3.2.c

1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip ,

syntoms for each function. The SRM, IRM, and APIU4 (St.a r tu p mode) , blocks need not,be operable in "Run" mode, and the ApRM (Flow biased) and RBM rod blocks need not be operable in "Startupa mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to r,even days provided that during that time the operable system

• is functionally tected immediately and daily thercafter; if this condition last longer than seven days, the system with the inoperable channel shall be tripped. If the first column cannot be met for both trip systems, both trip systems shall .

be tripped. ..

W is the recirculation loop flow in percent of design. Trip 2.

level setting is in percent of rated power (329 3 MWt) .

. A ratio of FRP/CMFLPD<1.0 is permitted at reduced power. See specification 2.1 for APRM control rod block i setpoint.

3. II4M downscale is bypassed when it is on its lowest range.

t

4. SRM's A and C downscale function is bypassed when IRM's A, C, E, and G are above range 2. SRM's B and D downscale function is by-passed when IRM's B, D F, and H are above range 2 2-SRM detc.:tos not in startup position is bypasued when the count

' rate is t 100 CPS or the above condition is satisfied.

5. One instrument channel; i.e., one APRM or IRM or RBM,' per trip system may be bypassed except only one of four URM may be bypassed. Rcfer to Section 3.10.B for SRM requirements during,

. core alterations.

6. IItM channels A, E, C, G all in range 8 bypasses SRM channels A & C functions.

IHM channel s B, F, D, 11 all in range 8 bypauses SF<M ch.innels DG D functions.

. 7 The following operational restraints apply to the IIIW only.

a. Both REM channels are bypassed when reactor power is f 30%.

b. The RI11 need not be operable in 'the "startup" position of the renacter modo selector switch.

x c. Two R124 channels are provided and only one of these :nay be bypassed from the console. If the inoperable channel cannot be' restored within 24. hours, the inoperable channel shall be placed in the tripped condition within one hour.

d. If minim;n conditions for Tabic 3.2.C are not, met, adminintrative controin, chall bo imnediately imposed.to prevent control rod wi tM rawa l .

74

• e

't Motts roR TA8tEs 6.2.A THROUCH 4.2.H (Continued)_

time se

14. Upecale trip is functionally tested during functional test required by section 4.7.8.1.a and 4.7.C.1.S.

15. The flow bias comparator will be tested by putting one flow input valt in to obtain

" Test" (producing 1/2 scram) and adjusting the testThe flow bias upscale will be verified comparator rod block. observing a local upscale trip light during operation and verified that it will produce a rod block during the operating cycle. .

16. Perforned during operating cycle. Fortions of the logic is checked more frequently during functions 1 tests of the functions that produce a rod block.

function from service and

17. This calibration consists of removing the performing an electronic calibration of the channel.

is limited to the condition where secon ad ry containment

18. Functional teet required se specified in sections 3,7 C.2 and 3.7.C.3.

integrity is not

19. Functional tent is limited to the time where the SCTS is required to meet the requirements of section 4.7.C.1.3L

20. Calibration of the comparator requires the inputs from both recirculation loops to be interrupted. thereby removingThis the celibration flow bLas signal can onlyto the be AFRM and RBM and scramming the reactor.

performed duri.sg an outage.

te limited to the time where actual operation of the equipment (

21. Logic test (

is permissible.

22. One channet of either the reactor zone or refueling zona Reactor Building Ventilation Radiation Monitoring System may be admin?.atratively bypassed for a period rot to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for functional testing and es11bration.

are RTD's that feed

23. The Reactor Cleanup System Space Temperature monitorsThe temperature evitch may be a temperature switch in the control room.

i l

tested monthly by using a simulated signal. The RTD itself is a highly reliable instrument and less frequent testing is neceesary.

24 This instrument check consists of comparing the thermocouple readings for all valves for consistence and for nominal expected values (not required during refueling outages).

25. During each refueling outage, all acoustic monitoring channels shall be calibrated. This calibration includes verification of accelerometer I response due to mechanical excitation in the vicinity of the sensor.

26. This instrument check consists of comparing the background signal levels fnr all valves for con'istency and for nominal expected values (not required during refueling outages).

I 110 l

w --

3.2 .BASF5

\.

The HPCI htr.h i;ow and temperature instrueentation arr provided to detect a break in the HPCI steam P! ping. Tripping of this 1.sst rasientation re-eults in actuation of HPCI 1sulation valves. Tripping logic for the hggh flov is a 1 out of 2 logic, and all sensors are required to be operable.

411r.h teaperature in the vicinity of the HF'*! equipnent is sensed by 4 sets of 4 btrecallic temperature switches. The 16 teaperature evitches are arranRed in 7 trir synterer with 8 temperature switche: in each trip system.

The HrCI trip settinar c.' % ps! f *s i t a r,b 81cv et.d 70C*T for high tere-perature' arc suc h ths t cota tn'ovety is prevented ar.d fission product release is within IIM t s.

The RCIC high finw a n.1 rearerJtute IF*=tru*entat(Od 3rr arrangCd the safst as that for the llPCI. The trip setting c! 450" h,fl fo; high flov and 200*T for t er:pe ra tu s e srt baard on the same c riteita as t!.e HPC1.

liigh tenperatore at the Fractor Cicanuo Syst,ce floor drain could Indicate a break in c l.e cleanup systea. Lhen high tempercture occurs, the cleanup system is isolated.

The instrumentation which initiates CSCS actlen is arranced in a dual bus system. As fer other vital instrurentation Arrant,cd in this fashion, the Specification preserves the effectiveness of the systen even during periode when nalntenance or testing is betnr. perforned. An esecption to this is when lectc funettona; testinr. to being perforne*

The control sort blact fune.t. inns are provided to prevent excessive control sod withdrawal so t h.* t !!CpR doe s nor. decrease to 1.07. The trip log (c for this function is I out of n: c.g. . .sny t r ip on one of six APM's,

~

eight I P# s . or f our SM's will resu1t in a red block.

The nint xs turba+ent channei re t aire ent s ase are suf fic ient in s t rurie n t a -

tion to assuit t oe c ir.r. l e (M1ure criters: 15 eier . Two RBM channels are pro-vided and only one of these may be bypassed from the console, for mainte-nnnee and/or testing, provided that this condition does not last lotiger thrui 2h hours in any thirty day period. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

Tne APRN rad bloc'- funce!nn is flow b!ased and prevents a s!Anificant reduc-tion in t'CI"t , es.m t.elly durtng operation at tcJaced flow. The APM pro-vide

• Cross core prot e c t i< n. i.e.. II,1ta the g ess cere power increase froc withdrawal of centrol rods in the normal wit 5craval sequence. The trips are set an (54t K r.e is maintained greater than 1.07.

1he RDH tod block f um:t toa provide.e b cal protect 1rn of the core; 1.e.,

the prevention of critical re.cra in 3 local region of the core, for a single t'od withdrawal error from 11rnitin5 control rod pattern.

M%

113

J straverLL.ANCE REQUIREMENTS l .f.I.'.LT big _ CON'D1T10.W_E0)Lnff o aviost 4.6.C Structural Integrity 1,r,.r.  ?;s s isc t ieral Intelt,ity inservice inspection surveil-maintained at the level re- lance requi,rements of the reac-quired lay the original accep- tor coolant system as follove:

\-

% tance standards throughout the life of the plant. The reactor chall be maintained a. areas to be inspected in a cold shutdown enndition until each indication nf a b. percent of arena to be defect hna been investigated inspected during the and evaluated. inspection interval

! c. inspection frequency

d. methods used for inspection i

2. . Evaluation of inservice inspec-l tions will be made to the accep-tance standards specified for the original equipment.

l l

3. The inspection interval shall be 10 years.

4. Additional inspections shall be performed on certain circumferen-tial pipe welds as listed to pro-vide additional protection aAatnet pipe whip, which could damage auxiliary and control sys-

\%., tems.

Feedwater - CFW-9, KFW-13 CFW-12, CrW-26, KYW-31, CrW-29, l KFW-39, CrW-15

. KFW-38, and CFW-32 Main steam - CMS-6, EKS-24, CMS-32. KMS-104 CMS-15, and CKS-24 RHR -

~ DSRHR-4, DSRHR-7, DS.M -6 Core Spray - TCS-407 TCS-423 TSCS-408 TSCS-424-18J

! I

e w 6.0 QMINISTRATIVE CONTROLS C. Drills on actions to be taken under emergency conditions involving release of radioactivity are specified in the radiological emergency plan and shall be conducted annually. Annual drills shall also be conducted on the actions to be taken following f ailures of safety related systems or components.

D. Radiation Control Procedures Radiation Control Procedures shall be maintained and made available to all station personnel. These procedures shall show permissible radiation exposure and shall be consistent with the requirements of 10 CFR 20. This radiation protection program shall be organized to meet the requirements of 10 CFR 20 except in lieu of the " control device" or " alarm signal" required by paragraph 20.203 (c) of 10 CFR 20

1. Each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and con-spicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Special Work Permi$'. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made af ter the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activitics within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Special Work Permit. ,

2. Each high radiation area in which the intensity of radiation is greater than 1,000 mrem /hr shall be subject to the provisions of (1) above; and, in addition, access to the source and/or area shall be secured by lock (s).

The key (s) shall be under the administrative control of the shif t engineer.

In the case of a high radiation area established for a period of 30 days or less, direct surveillance to prevent unauthorized entry may be sub-stituted for permanent access control.

a.

Health Physics personnel, or personnel escorted by Health Physics personnel, l

in accordance with approved emergency procedures, shall be exempt from the SWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.

139

6.0 ADMINISTRATIVE CONTROLS B. Source Tests  ;

Results of required leak tests performed on sources if the tests reveal the presence of'O.005 microcurie or more of removable contamination.

C. Special Reports (in writing to the Director of Regional Office of Inspection and Enforcement) . .

1. Reports on the following areas shall be

• submitted as noted:

a. Secondary Containment 4.7.C Within 90

,, Leak Rate Testing (5) days of completion of each test.

b. Fatigue Usage 6.6 Annual Evaluation Operating Report C. . Relief Valve Tailpipe 3.2.F Within 30 days Instrumentation after inoperability

. . t of thernocoupic anc' acoustic monitor on one valve.

~

d, Seismic l'nstrumentation 3. 2.J . 3 Within 10 days inoperability after 30 days of inoperability

> e. Meteorological, Monitoring 3.2.I.2 Within 10 days Instrumentation af ter 7 days of Inoperability - inoperability

f. Primary Containment 4.7.A.2 Within 90 days Integrated Leak Rate of copietion of Testing each test.

D. Special Report (in writing to the Director of Regional Office of Inspection and. Enforcement)

Data 'sha'11 be retrieved from all seismic instruments' actuated during a seismic event and analyzed to determine the magnitude of the vibratory ground motion. A Special Report s' hall be submitted within 10 days after the event describing the magnitude, frequency spectrum, a'nd resultant

. effect upon plant features important to safety. ,

356 ..

t

4 - e .--- .a -

1

) UNIT 3 f

I i

t i

j i

I 7

l 4

1 n

I 4

i

I

!!'JIES POD TABLE 3.2.C startup and run positions of the Reactor Mode 1 For thr-Selector Switch, there shall be two operable or tripped trip systemsblocks for each function. The SRM, IRM, and APRM (Startup

• mode) , need not be operable in "Run" mode, and the APRM (Flow biased) and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time che operable system is functionally tested immediately and daily thereafter; if this condition Jast longer than seven days,Ifthe system with the first column the inoperable. :2annel shall be tripped.

cannot be met for both trip systems, both trip systems shall be tripped.

Trip

2. W is the recirculation loop flow in percent of design.

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

A ratio of FRP/CMFLPD41.0 is permitted at reduced power. See Specification 2.1 for APRM control rod block setpoint.

3. IRM downscale is bypassed when it is on its lowest range.

4. SRM's A and C downscale function is bypassed when IRM's A, C, E, and C are above range 2. SRM's B and D downscale function is by-passed when IRM's B, D, F, and H'are above range 2.

SRM detector not in startup position is bypassed when the count , ,

rate is g 100 CPS or the above condition is satisfied.

5. One instrument channel; i.e., one APRM or IRM or RBM, per trip system may be bypassed except only one of four URM may be bypansed. Nfer to Sect ion 3.10.D for Slut requirementn during, core alterations.

IHM channels A, E, C, G all in range 8 bypasses SRM channels 6.

A & C f unctions.

IRM channels B, F, D, H all in range 8 bypasses SRM channels B C D functions.

7. The following operational restraints apply to the RP14 only,

a. Both RP44 channels are bypassed when reactor power is f 30%.

b. The RH4 need not be operable in the "startup" position of the s

reacter mode selector switch,

c. Two Blid channele are prov'ided and only one of these may be bypaaued from the console.

If the inoperable channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall be placed in the tripped condition within one hour,

d. If minimun conditions for Tabic 3 2.C are not met, administrative controls, chall be immediately imposed to prevent control rod withd rawal . .

77

[

^

e o

JUN 181981

{

MOTES FoR TABLES 4.2.A THROUCTI 4.2.H (Continued)

14. Upecale trip is functionally tested during functional test time as required by section 4.7.8.1.s and 4.7.C.2.c.

13. The flow bias comparator will be tested by putting one flow uait in

" Test" (producing 1/2 acram) and adjusting the test input to obtain comparator rod block. The flow bias upscale will be verified by observing a local upscale trip light during operation and verified that it will produce a rod block during the operating cycle. ,

16 Performed during operating cycle. Fortions of the logic is e.hecked more frequently during functional teste of the functions that produce a rod block.

17. This calibration consists of removing the function f rom service and performing an electronic calibration of the channel.

18. Functional test is limited to the condition where secondary containment integrity is not required se specified in sec tions 3.7.C.2 and 3. 7.C.3.

19. Functional tent is limited to the time where the SCTS is required to meet the requirements of section 4.7.C.1.3L

20. Calibration of the comparator requires the inputs from both recirculation loops to be interrupted, thereby removing the flow bLas signal to the APRM and RBM and scramming the reactor. This calibration can only be performed durias an outage.

21. Logic test te limited to the time where actual operation of the equipment (,

is permissible. -

22. One channci of either the reactor zone ur refueling zone Reactor 8uilding Ventilation Radiation Monitoring System may be administratively bypassed for a period cot to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for functional testing and calibration.

23. The Reactor Cleanup System Space Temperature monitors are RTD's that feed a t empe ra ture evitch in the control room. The temperature switch may be tested monthly by using a simulated signal. The RTD itself is a highly reliable instrument and less frequent testing te necessary.

24. This instrument check consists of comparing the thermocouple readings for all valves for consistence and for nominal expected values (not required during refueling outages).

25. Durina each refueling outage, all acoustic monitoring channels shall be calibrated. This calibratioa includes verification of accelerometer response due to mechanical excitation in the vicinity of the sensor.

26. This instrument check consists of comparing the background signal levels for all valves for consistency and for nominal expected values (not required during refueling outanes).

107 a

Pressure instrument at ion is provided to close the main steam isolation valves in Run Mode when the main steam line pressure drops below 825 psig.

The HPCI high flow and temperature instrumentation are provided a to detect a hreak in the HPCI steam piping. Tripping of this instrumeritation results in actuation of HPCI isolation valves.

Tri ppi nq logic for the high flow is a 1 out of 2 logic, and all sensors are reguired to be operable.

High temperature in the vicinity of the HPCI equipment is sensed

, by 4 sets of 4 binnetallic temperature switches. The 16 temperature switches are arranged in 2 trip systems with 8 temperature switches in each trip system.

The HPCI trip set tings of 90 psi for high flow and 2000r for high temperature are such that core uncovery is prevented and fission product release is within limits.

The FCIC high flow arid temocrature instrumentation are arranged the same as that for the HPCI. Tne trip setting of 450" water for high flow and 2000F for temperature are based on the same criteria as the HFCI.

High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated.

The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the specification preserves the ef fectiveness of the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing i s being perf ormed.

The control rcd block functions are provided to prevent excessive

control rod withdrawol so that MCPR does not decrease to 1.07 The trip logic for this f unction is 1 out of n: e.g., any trip on one of six APRM's, eight IRM's, or four SRM's will result in a rod block.

The minimum instrument channel requirements assure sufficient inst rumentation to assure the single f ailure criteria is met. Two RBM channels are provided ar.d only one of these ztay be bypassed fror, the console, for maintenance and/or testing provided that this condition does not last longer than

21. hours in any thirty day period, This does not significantly increase the risk of an inadvertent control rod iithdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

The APRM rod block function is flow biased and prevents a .

significant reduct ion in MCPR, especially dur inq operation at reduced flow. The APRM provides gross core protection; i.e.,

limit s the gross core power increase f rom withdrawal of control l -

. 110 -

,j 1

1

SURVEILI.ANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION 3.b PRIM ARY SYSTD4 BOUNDARY 4.6 ERJP8ARY SYSTEM BOUNDARY Main steam-GMS-6, KMS-24, GMS-32, KMS-104,..

GMS-15, and GMS-24 RER -DSRHR-6, DSRHR-7, and DSRHR-4 Core Spray TCS-407 TCS-423 TSCS-408 TSCS-424 Reactor -DSRWC-4, DSRWC-3, Cleanup DSRWC-6, and DSRWC-5 HPCI -THPCI-70 THPCI-70A THPCI-71, and THPCI-7 2

5. System hydrostatic tests in accordance with Article IS-500 of Section XI of the ASME Code at or near the end of each

- - . inspection interval and prior to startup following each l

refueling outage.

I The pressure-temperature limits for these tests will te in accordance with specification 3.6.A.3.

~

R E'EREMCE

1. Plant Safety Analysis (BFNP FSAR subsection 4.12)

.r 197

\

! {

1

e t TABLE 3.7. A PRIJoutY ColfrADetDt? IsotATIost VALVES Nuaber of Power Maxir.ua Action on operated valves opera ting Normal Initiating Group Valve Identification Inboard outtoard Time (sec.) Position signal Standby liquid control system check valves (cv 63-526 s 525) 1 1 NA c Process Feedwater check valves 2 2 N4 0 Process (cv-3-554, 572, 554 8 568) control rod bydraulic return check valves (cv-45-576 s 573) 1 1 NA 0 Process RERS 'LPCI to reactor check valves (cV-74-54 5 68) 2 MA c Process Core Spray discharge to reactor check valves (FCV-75-26 and 54) 2 NA C Process 6 Drywell AP air compressor suction valve (FCV 64-139) 1 10 C SC 6 Drywell dP air compressor discharge ,

valve (FCV 64-140) 1 10 C SC 6 Drywell CAM discharge valves .

(FCV 90-257A and 257B) 2 10 0 CC 6 Dryvell CAM suction valves (FCV 90-254A and 254B) 2 10 0 CC -

6 Dryvell CAM suction valve (FCV 90-255) 1 10 0 CC r

'6 CAD System Torus /Drywell Exhaust 2 . 10 C SC

' to Standby Gas Treatment -

(FCV 84-19, 20) l l

i

(

l l

• s O

, o 6.0 ADMINISTRATIVE CONTROLS C. Drills on actions to be taken under emergency conditions involving release of radioactivity are specified in the radiological emergency plan and shall be conducted annually. Annual drills shall also be conducted on the actions to be taken following f ailures of safety related systems or components.

D. Radiation Control Procedures Radiation Control Procedures shall be maintained and made availsble to all station personnel. These procedures shall show permissible radiation exposure and shall be consistent with the requirements of

' 10 CFR 20. This radiation protection program shall be organized to meet the requirements of 10 CFR 20 except in lieu of the " control device" or " alarm signal" required by paragraph 20.203 (c) of 10 CFR 20:

1. Each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and con-spicuously posted as a high radiation area and entrance thegeto shall be controlled by requiring issuance of a Special Work Permit . Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiction dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them,

c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responaible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Special Work Permit. ,

2. Each high radiation area in which the intensity of radiation is greater than 1,000 mrem /hr shall be subject to the provisions of (1) above; and, in addition, access to the source and/or area shall be secured by lock (s).

The key (s) shall be under the administrative control of the shift engineer.

In the case of a high radiation area established for a period of 30 days or less, direct surveillance to prevent unauthorized entry may be sub-stituted for permanent access control.

' Health Physics personnel, or personnel escorted by Fealth Physics personnel, in accordance with approved emergency procedures, shall be exempt from the SWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.

369_ __.

6.0 ADMINISTRATIVE CONTROLS B. Source Tests j l

Results of required leak tests performed on sources if the tests reveal the presence of'O.005 microcurie or more of removable containination.

c. Special 3eports (in writing to the Director of Regional Of fice of Inspection and Enforcement) . .

1. Reports on the following areas shall be

• submitted as noted:

a. Secondary containment 4.7.c Within 90 Leak Rate Testing (5) days of 4 completion of each test.

b. Fatigue Usage 6.6 Annual Evaluation Operating-l Report

c. . Relief Valve Tailpipe 3.2.F Within 30 days Instrumentation after inoperability

. . I of thernocouple and acoustic monitor on one valve.

d. Seismic l'nstrumentation 3. 2.J . 3 Within 10 days Inoperability . af ter 30 days of inoperability

e. Meteorological, Monitoring 3.2.I.2 Within 10 days Instrumentation after 7 days of Inoperability -

Inoperebil,ity

. f. Primary Contalcunent 4.7.A.2 Within 90 days Integrated Leak Rate of completion of Testing each test.

D. Special Report (in writing to the Director of Rey.ionhi Office of Inspection and Enforcement)

Data sha'11 be retrieved from all seismic instruments actuated during a seismic event and analyzed to determine the magn,itude of the vibratory ground motion. A Special g Report shall be submitted within 10 days after the event describing the magnitude, frequency spectrum, a'nd resultant

. effect upon plant features important to safety.

8 386 I

if

ENCLOSURE 2 DESCRIPTION AND JUSTIFICATION (TVA BFNP TS 176)

BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 Units 1 and 2 - pages 74 and 113 Unit 3 - pages 77 and 110 Description It is proposed to remove the requirement for cumulative out of service time for Rod Block Monitor.

Justification By previous Amendment Nos. 74, 71, and 46 to the operating licenses for Browns Ferry units 1, 2, and 3, issued August 6, 1981, the definition of

" Cumulative Downtime" was removed from the technical specifications. The change proposed by this request is for clarity only. This change removes a requirement that has no action other than generation of an LER.

Present technical specifications require that a cumulative downtime of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30-day period not be exceeded. However, no action is required after such downtime has elapsed resulting in lack of clarity. Since cumulative downtime does not appear in the BWR standard technical specifications, we have changed this requirement to a simple 24-hour period allowed for downtime with specific action to be taken in the event that the'24-hour period is exceeded. The effect of this change on plant operation will be to increase the margin of safety related to the rod block monitor by adding the requirement that the RBM be tripped after a 24-hour downtime period.

Clarification is needed based on Licensee Event Report BFR0 260/82009.

There is no required action to be taken when the RBM is out of service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30-day period. Standard technical specifications contain no such requirement and NRC has previously recognized that such cumulative downtime recording is not required by deleting this from Definitions (Item Z Cumulative Downtime) by the license amendments referenced above. This change is also supported by the RBM being a backup to the written sequence for withdrawal of control rods.

This change to the technical specifications does not adversely affect operation, safety margins, accident analysis, or overall plant safety.

s Units 1 and 2 - page 110 Unit 3 - page 107 Description Change Note 19 for tables 4.2. A through 4.2.H reference from "4.7.C.1.c" to "4.7.C.1.a."

Justification There is no section 4.7.C.1.c. The correct reference is 4.7.C.1.a.

This is an administrative change that has no safety significance.

Therefore, a safety analysis or additional justification is not appropriate.

Unit 2 - page 183 Unit 3 - Page 197 Description Change the core spray weld numbers under 4.6.G.4 to:

Core Spray - TCS-407 TCS-423 TSCS-408 TSCS-424 Justification Part of the core spray piping has been changed out from stainless steel to carbon steel. This change to the technical specifications is being made to incorporate the new weld numbers that correspond to the location of the original weld numbers.

There is no impact on plant safety. The core spray weld numbers originally in section 4.6.G.4 are being replaced by weld numbers resulting from a system piping modification. The replacement weld numbers correspond in location to those originally in 4.6.G.4.

NOTE: There are no plans to perform the same piping modifications to unit 1. Therefore, similar revisions to unit 1 technical specifications are not planned.

Unit 1 - page 250 Description and Justification Change "FCV 74-57,58" to "FCV 75-57,58." This corrects a typographical error.

Unit 1 - page 251 Description and Justification Delete valve FCV 69-12. This is not a containment isolation valve.

Isolation is provided by check valves69-579 and 3-572.

Unit 1 - page 252 Description and Justification Change the normal operation of FCV 71-7A,7B from "0" to "C" and change action on Initiating signal from "GC" to "SC." The normal position of these valves is closed.

Change "FCV 75-57,58" to "FCV 73-6A,6B." These valve numbers were incorrectly placed in the table.

Unit 1 - page 253 Unit 3 - page 265 4

Description Add FCV 84-19 and FCV 84-20 to table 3 7.A.

Justification As a result of the routine NRC inspection on November 26 through December 25, 1981 (inspection reports 50-259/81-37, -260/81-37,

-296/81-37), FCV 84-19 and -20 were found to be missing from table 3.7. A.

These valves are primary containment isolation valves and should be included in this table. Technical specification 4.7.D.1.a requires automatically initiated power operated isolation valves to be closure time tested at least once per operating cycle. FCV 84-19 and FCV 84-20 are being added to table 3 7.A to require closure time testing at least once per operating cycle. These valves were inadvertently omitted from the table. This change, therefore, only serves to increase the safety of plant operation since it adds testing requirements for valves which should have been included in earlier revisions to the technical specifications.

I l

i i

1 s

Unit 1 - page 253a Description Add "FCV 64-139," and "FCV 64-140" to table 3.7.A. These valves were added to unit 1 as part of the addition of the drywell pressurization system. These valves were inadvertently not included in table 3 7.A when the valves were installed in the plant. Addition of valves FCV 90-254A, FCV 90-254B, FCV 90-257A, FCV 90-257B, and FCV 90-255 to table 3 7. A.

Justification 1

These valves were advertently omitted from the table. These valves are being added to reflect plant configuration and to require closure time testing at least once per operating cycle. This change serves to increase safety of plant operation by adding test requirements which should have been included in earlier revisions to the technical specifications.

Units 1 and 2 - page 339 Unit 3 - page 369 Description This change is a revision to the administrative procedures for access to high radiation areas.

Justification The reason for the proposed change is to allow more flexibility in the control of access to high radiation areas. The proposed change would allow use of direct surveillance as a substitute for permanent access I

control to preveitt unauthorized entry to a high radiation area if the high radiation area will only be established for 30 days or less.

The proposed revision will not affect operations, safety margins, and accident analysis, however, it should enhance overall plant safety by permitting broader administrative latitudes to control personnel access to high radiation areas and sources.

i 1

Units 1 and 2 - page 356 Unit 3 - page 386 Description Include instructions in section 6.7.3.C (page 356) of the Appendix A technical specifications to specify the requirements to submit primary containment integrated leak rate test reports within 90 days of completion of each test.

.fustification Primary containment integrated leak rate tests are performed in accordance with ASME code for Boilers and Pressure Vessels,Section III, Subsection B as specified in section 4.7. A.2 of the technical specifications. As required in 10 CFR 50 Appendix J, the report for this test must be submitted to NRC within 90 days from the completion of the test. Provisions for reporting primary containment integrated leak rate tests are not included in the reporting requirements of the technical specifications. This change is being submitted in order to clarify the reporting requirements specified above, and therefore, does not affect the overall safety of the plant.