ML20244D461

Amends 165,163 & 136 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Modifying Tables 3.2.J & 4.2.J to Reflect Vendor Suggested Testing Frequencies for Seismic Monitoring Equipment

ML20244D461
Person / Time
Site:	Browns Ferry
Issue date:	04/13/1989
From:	Black S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20244D464	List: ML20244D461 ML20244D465
References
TAC-00474, TAC-00475, TAC-00476, TAC-474, TAC-475, TAC-476, NUDOCS 8904210357
Download: ML20244D461 (30)

Text

. - _ _ -. _ _ _ _ _ _ _ _ _

L j

i

./

'o,,

UNITF.D STATES

~

.[

g NUCLEAR REGULATORY COMMISSION wasswnow, o. c. roess

%,...../

l l

l TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BRNNS FERRY NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.165 License No. DPR-33 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated September 29, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulat' ions set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliance with the Commission's regulations; D.

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

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

8904210357 090413

~

PDR ADOCK 05000259 p

PDC

l

• 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-33 is hereby amended to read'as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.165, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMISSION

'zenne\\ 4 D

N Rs I Black, Assi m nt Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of 1ssuance: April 13, 1989

ATTACHMENT TO LICENSE AMEN 0 MENT NO.165 FACILITY OPERATING LICENSE NO. OPR-33 DOCKET NO. 50-259 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.

REMOVE INSERT 3.2/4.2-37 3.2/4.2-37 3.2/4.2-58 3.2/4.2-58 3.2/4.2-69 3.2/4.2-69*

3.2/4.2-70 3.2/4.2-70 i

.-____.m_ _ _ _ _ _ _ _

e w

Ew L L L E8 E

• 3I g

5 5 5 2kk d d d g

g 2

1 -Ql

f. f. N2 m

&EE EEE a

d E

5E k5$

k$k kbb g

o o o o o e o o o m

2 u

4 2

5 40 3

5 5

5 s

a m

a s

a 3

3 y

C G

G G

u i

e

~

J EC c

G

.I c c c

y g-m a

g g :1

. g :.

1 1

i 1

y s

wd t C C

g g"~gg;8 g.2 m

3 2

3 Wo o

o o

C 4 4

4 3

E4 4

4 e5 5 :

G s 2

2 0L Rs 2

2 5

W~

.k 2

E 6

6

~

u u o

u d

3

-o o

a U

"U 9 MB R

E t t

2 W

~

t W E.

t c:

t i

~

IE4 E

E o

a u u

6 g

4-3 4-a -

u

~ha5 4sai5y&aA

~

a 4

4 J E4 i J ma;a F

~

n m

a Eo 3.2/4.2-37 Amendment No. 165

_m

___.______m_

no i

tar b

i la C

R R

R R

R R

R R

R lenna h

C t

se T

l S

a T

n N

o EM i t

E c

R n

I u

U A

A A

F Q

A A

A

/ / /

A A

A E

l S

S S

N N N S

S S

R en E

n C

a N

h A

C LL I

EV J.

R k

U c

2 S

e h

y y

y y

y y

4 T

C l

l l

l l

l N

h h

h h

h h

e E

l t

t t

t t

t l

M e

n n n

A A A

n n n b

U n

o o

o

/ / /

o o

o a

R n

M M M N

N N

M M M T

T a

S h

N C

I GN I

R

)

)

O 5

5

)

T 2

0

)

)

)

)

I N

0 5

0 0

0 1

6 4

O 9

2 5

4 9

9 9

M 1

f.

C 5

5 1

1 1

l 5

5 5

I l

E 1

M 1

E

(

E 1

l l

S I

E

(

(

E E

E S

(

b

)

(

(

(

E H

b a

5

)

e 7

0 o

b b

b S

P b

a l

A a

l s

l a a a

R l

s 5

0 a

l l

l G

s e

2 8

s s

s O

r s

6 5

4" R

e o a e

e e

E s

o b S

1 s

s s

L a

l H

1 1

a a a t

E b

f P

E E

b b b

n C

a A

(

(

m e

C d

R e

S

. a e m A a a i G

e e t

a Ed u

d d b O

o o s

d d

r Y i i R

t R

b b

r E

l l

v C

i 1

1 a a s

T b

b b

s O

a L

I v W r r r n

T r

r t

E 0"

6 a

S o o o I

S o o e C

I t

t C

1 1

r t

t t

H c

c A

e C

c c

c a a -

s I

a a a

E e

e K W W M

e e

e M

r r

A C

S e

S r

r r

I E

C Mr r

I T

I I

l P

E o

E I

I I

e R

S L

t t

s L

t t

)

A i i e A 1

1 1

L i i l

I n n i I

A n n t

i b

X U U D N

U U U I

U U I

A A

K n

I A

I

. R R

a I

T a b c

T a

b c

D a

b c

m r

1 2

3 a

~

Sh 53oMa 5.

Il

y-

,s-

' 3.2 R& GEE (Cont'd)

Trip setting of 100 ar/hr for the monitors in the refueling zone are based.upon initiating normal ventilation isolation and SGTS operation so that none of the activity released during the refueling accident leaves I

the Reactor Building via the normal ventilation path but rather all the activity is' processed by-the SGTS.

Flow integrators and sump fill rate and pump out rate timers are used to determine. leakage in the drywell. A system whereby the time interval to fill a known volume.will be utilized to provide a backup. An air sampling system is'also provided to detect leakage inside the primary containment (See Table 3.2.E).

For each parameter. monitored, as listed in Table 3.2.F, there are two channels of' instrumentation except as noted. _By comparing readings between the two channels, a near continuous surveillance of instrument performance is available. Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.

Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control room. :An accident signal that isolates primary containment will also automatically' isolate the control room and initiate the emergency pressurization system.. In addition, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressurization system. Activity required to cause automatic actuation is abcut one mRom/hr.

Because of the constant surveillance and control exercised by TVA over the Tennessee Valley, flood levels of large magnitudes can be predicted in advance of their actual occurrence. In all cases, fullLadvantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted; however, the plant flood protection is always in place and does not depend in any way on advanced warning.. Therefore,-during flood conditions, the plant will.be permitted to' operate until' water begins to run across the top of the pumping station at elevation'565. Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition. At that time an orderly shutdown of the plant will be initiated, although surges even to a depth of several feet over the pumping station deck will not cause the loss of the main condenser circulating water ptapa.

BFN 3.2/4.2-69 Unit 1 L

d 3.2 EASES (Cont'd)

The operability of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation dose to the public as.a result of routine'or accidental release of~ radioactive materialssto the atmosphere..This capability is required to; evaluate the need for initiating protective measures to protect the health and safety of the public.

The operability of the seismic instrumentation ensures that sufficient capability is available to promptly determine the seismic response of l

those. features important to safety. This capability is required to permit comparison of the measured response to that used.in the design.

basis for Browns Ferry Nuclear Plant and to determine whether the plant can continue to be operated safely.- The instrumentation provided is consistent with specific portions of the recommendations of-Regulatory Guide 1.12 " Instrumentation for Earthquakes."

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials:in gaseous effluents during actual or potential releases of gaseous effluents.. The alarm / trip setpoints for these instruments will be calculated in accordance with guidance provided in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. _This instrumentation also includes provisions for monitoring the concentration of potentially explosive gas mixturca-in the offgas holdup system. The operability and use of this instrumentation is consistent with-the requirements of General Design Criteria 60, 63, ar.n 64 of Appendix A to 10 CFR Part 50.

Th'e radioactive' liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with guidance provided in the ODCM to. ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR -Part 20 Appendix B, Table II, Column 2.

The OPERABILITY and use of this instrumentation is consistent with the requirements of Genera 1' Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

ATWS/RPT, Anticipated Transients without Screa/ Recirculation Pump Trip system provides a means of limiting the consequences of the.unlikely occurrence of a failure to scram during an ATWS event. The response of the plant to this postulated event (ATWS/RPT) follows.the BWR Owners Group Report by General Electric NEDE-31096-P-A and the accompanying NRC Staff Safety Evaluation Report.

ATWS/RPT utilizes the engineered safety feature (ESF) easter / slave analog trip units (ATU) which consists of four level and four pressure channels total. The initiating logic consists of two independent trip systems each consisting of two reactor done high pressure channels and two reactor vessel low level channels. A coincident trip of either two low levels or two high pressures in the same trip system causes initiation of ATWS/RPT. This signal from either trip system opens one of two B0C BFN 3.2/4.2-70 Amendment No. 164, 165 Unit 1

UNITED STATES NUCLEAR REGULATORY COMMISSION a

I wasHWGTON, D. C. 20056 a

TENNESSEE VALLEY AUTHORITY l

DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.163 License No. DPR-52 1

1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated September 29, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this a:nendment is in accordance with 10 CFR Part 51 of the Cossnission's regulations and all applicable requirements have been satisfied.

~

l

. w

+.

. t i

Accordingly, th'a license is amended by changes to the Technical 2..

Specifications'as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.163, are.hereby incorporated in the license. The licensee f.'111 operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COP 941SSION XT) s Tack,AssishntDirector v

(hv.

zanne for Projects TVA Projects Division Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 13, 1989 4

1

\\

{.*

l ATTACHMENT TO LICENSE AMENDMENT NO.163

]

FACILITY OPERATING LICENSE NO. DPR-52 I

l DOCKET NO. 50-260 l

l

)

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Pages with* are provided to maintain document completeness.

REMOVE INSERT 3.2/4.2-14 3.2/4.2-14*

3.2/4.2-15 3.2/4.2-15*

3.2/4.2-37 3.2/4.2-37 3.2/4.2-58 3.2/4.2-58

~.2/4.2-65 3.2/4.2-65 3.2/4.2-66 3.2/4.2-66 3.2/4.2-67 3.2/4.2-67 3.2/4.2-68 3.2/4.2-68 3.2/4.2-69 3.2/4.2-69 3.2/4.2-70 3.2/4.2-70 3.2/4.2-70a 3.2/4.2-71 3.2/4.2-71 3.2/4.2-72 3.2/4.2-72 3.2/4.2-73 3.2/4.2-73

i a

i a

Sa l re a

a isn n

l e o

ef g eb Cn nl es i ne, r m

von s

t i

t i

i t

i t

g i et rt i re e

r s e

sl i

n i

n os

,yw0 n

,wirim, el rnri el mi wa5oi v,

n i

n i

pa pou g

g ft gd w

,S gl sm n gg gt n.

Vi i

i r

sr 1Si sosetg n

id r

n s

n s

n n o

C,

n i i yi ni nayo om y

i y

ye ihI e g

ihmt an i s irai t e i

a t

a ad tt vdn tt r l n.

t g t ppi as t

t t ert ct t

l t

l li ti

.i n i ti eyeu e

e e

e ec ewesan ewpadrS en eosd ey s

s r

s r

rc s

rs n

s l

D si s

n Rs k

a n ui r u nl e.pA t

tt o r

p r

p r

r posmer pa pnnc ed a i e

i e

es ii sr m poednos ii v ii i ee he

.i m

r i

r i

i e rt eeip rt e mpe rt rt mt tt crpte e

t l

t l

lt t

t cl c t

ti t rnn S

R p

p pa np n

e2t a n

ei e ei l yp wurs/t wi wvad l c M

w.

w wu i.

E oI tC o.

tI.

ii i

i i tt ojhea oje 1Rt o

odti io T

lC lI lS 1C li l n g vlR S l nt5 i

l r l anc h s NL Ni e oi e eH D eoa02rn eo enoc ws P

n S

eP uC eS Y

BH MR BC BchldRA Bcw1 1 oi Sf Bica S

EL e G

2 1

2 1

1 At N

1 1

1 Bh I

R L

Ef O

P o O

O C

et T

bi N

l E

oi M

tb N

I d p A

n ea T

o rc N

i i

O t

A A

A A

A A

un C

c qoi A

ei D

o rt N

r A

e ti z

ot E

ni R

o.

o o

o o

l O

r r

r r

r e

si C

e e

e e

e s

i

8. E a

z a

a a

s a

e l

H n

l l

l l

l v

eu

2. T i

e e

e e

e nn 3

t s

s s

s s

e) na S

t s

s s

s s

vt aM EL e

e e

e e

e oh h

LO S

v v

v v

v bg c

BR ai AT l

e e

e e

e e

td T N e

v v

v v

v h

ne 6"

e O

v o

o o

c o

C e

b b

b b

b 1 e m

l a

a a

a a

/r u

R 5o rf S

ip O

0" 0"

8" 8"

4" c

t r 2

se r

7 7

7 7

4 13 I

2 in E

T 4

4 3

3 5

3/

A 1

1 1

1 1

1(

s I

iii I

h T

tb N

I f

oi T

l l

l l

l l

A e

e e

e e

H v

v v

v v4, e

yn v

t I

- e e

- e

- e

- e8

- e ii L

L t

L L1 L

l t N

l l

l l

l l

i O

er er er er er3 er) b I

n ne ne ne ne ne -

ne2 ai T

o nt nt nt nt ntS nt6 pf A

i aa aa aa aa aaI aa ai T

t hW) hW) hW) hW) hWL hWd cd N

c C

D C

D C

D C

0 C

(

C n

E n

w-w-

w-w-

w wa n

M u

t oA t oA t oA t oA t oe t o o

U F

nL8 nL8 nL8 nL8 nl v nL2 i g R

e 5

e 5

e 5

e 5

e i e

5 t

mr -

mr -

mr -

mr -

mrs mr -

ai I

S uo3 uo3 uo3 uo3 uos uo3 it N

rt -

rt -

rt -

rt -

rti rt -

I t cS tcS t cS t cS t cm) t cS it saI saI saI saI sar5 saI ni neL neL neL neL nee 8 neL i

IR(

IR(

IR(

IR(

IRP1 IR(

c i

tlam oe tlua r

.e) e oP1

)

)

h N

(

6 6

T es 1

1 ml y

(

(

ubS 2

2 2

2 1

I ma iro nej ipr MOI og g."2*"-

g,a Qo5 2o* Ea*

e-.$

t t

c0#

t e

s t

n n r t

S l e

c ef g o o a

S a t

a r d,

v o n ct.i C n a

e o. n g e

i cS t g

i n rt e a in mi n

t i

c v r nr n aS i pe u.

i eC n o s i wi

, o e s e n w ai r n d s r

i r

o n t

gt n

g s

f t gl i gl r s m u n e y e n we s n

n n

i i r i r at i ot y y e ih e i h w mt t ert t l a a ad t t r t t o p

t p pi t

il l i t i u t i l e y m.

e o sd eh t e e c e ws e w p a uS s

s n

s t i r r c s

s s

l pD k

t t o i n a

ne nl l e

A r

p n n c p wir r

p o r p o e ed R a

i e e i

e e s i i p ii v v H s r n ei l

e r t rt e e

.H e

mt m

rt e

t r n

t o rl i t t c r t cl l c t

R el e i u p p a n o n

e r a w v ad et si

. l i e ut.

e u r r s oi od t i v c s t I t t v j cI vje e t

l a n c o n el Ci i o n aC o nt t 5S i

e n o c b u r u P n

b o eP b o a a0 S n B i c a A j pMH hi A c rL A c w w1 Ci 1

1 2

1 1

no i

t A

A A

A c

A

)

deun i

tn e

o n

C i

(

t t

B.

e S

g 2

l i

3 e

s v

p g

g g

E e

i i

i 5

s s

s L

L p

p p

B A

o 2

T i

i 5

5 5

r p

T 2

2 2

1 1

1 1

1 e

e e

e

- r

- r

- r

- r u

u u

u l

s l

s l

s l

s e s e s e s e s n

n e n e n e n e o

n r

)

n r n r n r a P H a P

)

a P

)

a P

)

it h

h D

h 0

h D

c C h E C h -

C h -

C h -

F nHi 5 nHigA gA gA n

g u

t 8

t t i 8 t i7 8

nH 5 nH 5 e

e 5

e e

ml 4 ml -

ml 4 ml 4 ul 6 ul 4 ul 6 ul 6 r e -

r e6 r e -

r e -

t wS t w -

t wS t wS s yI s yS s yI s yI n r P n r P n rP n r P I D(

I D(

I D

(

I D (

r

. e) oP 1

)

(

6 N

e s 1

ml v

(

ub S 2

2 2

2 m a i r o n ei ip r MOT

• wb*

E@ y@ +r O.

Y CiV h$u I

l

r E

ML B

RP A I R 1"

I P

'1" NE I

1 1

1 1

1 1

MO TN I

3 g

g O

g g

g A A A 1

1 1

P l

1 1

/ / /

n T

0 0 0

N N N 0 0 0

o E

i S

ta tne T

m N

g g

g u

E 5

5 5

r PE g

g g

g g

g 2

2 2

t EG 0

0 0

0 0

0 J.

s t N N

n LA 1

1 1

5 5

5 5 5 5 2

I SR 2

2 2

A 0 0 0

0 0 0 0

0 0

3 a

E l

i M

n e

r

)

b o

a 0

t

)

T i

0 2

4 0

0 0

1 5

4

)

)

)

n

)

o 5

5 M

9 1

5 I

1 15 5

9 9

9 1

2 1

c 5

6 6

E 5

5 i

S 1

(

ms H

1 E

1

)

e 1

1 1

P E

(

E 5

)

n E

E E

ie A

(

(

7 0

ie

(

(

(

l R

b G

b a

b 5

0 b

b b

O a

l a

2 8

4" a

a a

R l

s l

6 5

l l

l n

E s

s S

1 s

s s

o L

r H

1 1

i E

e o

e P

E E

e e

e t

n s

s s

a C

s o

s A

(

(

C a

l a

R e

S a

a a

c A

b f

b G

e e t

E b

b b

i Y

. O o n r

H d

R a a o E

o i v

C a

i R

l n

e s

T o a O

d d

d L

I d

d d

w W m

T i

1 i

E 0"

6" i

1 i

S b

b b

C a

S b

b b

o I

C 1

1 r

o H

r r

. A W.

m I

o o o

o C

r r

r r

e e n

E t

t a K W M

t t

t l

c c

e A

C S

e S

c c

c o

I a a E

R r

I a

a a

r K H o

E e

e e

t T

T e e l

P N

r r

e R

R c

S r

r r

n L

s L

o INA I

ie A

1 1

1 L

1 1

1 c

1 1

I A

R X

U U D

X U U U I

U U U h

T A

A X

t S

I I

A i

N R

R I

W I

T a

b c

T a

b.

c B

a b

c 2

1 2

3 t

l in U

-N FS h

53&3n O

j

~

no i

tar b

i la C

R R

R R

R R R

R R

lenna hC tse T

l S

a T

n N

a E

l M

t E

c R

n I

u U

F A A A Q

A A A

/ / /

A A A E

R S

S S

N N N S

S S

I E

C NA r'

L L

I E

V

'J R

k U

c 2

S e

h y

y y

y y

y 4

TN C

l l

l l

l l

h h

h h

h h

e E

l t

t t

t t

t M

e n n n A A A n n n l

b t

n o

o o

/ / /

o o

o t

a T

Ti n

M M M N

N N

M M M a

S h

N C

I GN I

R

)

)

O 5

5

)

T 2

0

)

)

)

)

I N

0 5

0 0

0 1

6 4

O 9

2 5

4 9

9 9

M 1

6 5

1 1

C l

5 51 5

5 I

1 E

1 M

1 E

(

E 1

1 1

S E

(

(

E E

E I

S

(

b d1 E

(

(

(

H b

a e

S P

b a

l 7

4 R

l g

5 J

ic b

b b

A a

l s

a a

a n

l l

l G

s r_

'5 e

2 s

s s

0 s

6 4"

R e

o a

e e

e E

s o

b S

1 s

s s

L a

l H

l 1

a a

a t

E b

f A

(

(

m P

E E

b b

b n

C e

C e

m A

e e.

d t

e S

. e e G e e t

E e

i u

d d b O

n o s

H d d d r

Y i i R i l

w C

i i i t

R b

b r

E e a s

T b

b b

s O

o L

I n

T r

r t

E v W r

r r

I S

e o a C 0"

6" a

S o o e I

t t

r C

1 1

r t

t t

H c

c e A a

C c

c c

a a n s

I a a a E

e e e MC W M

e e e W

M r

r e

S e

S r

r r

I E

C E

r I

T i

I l

P S

o E

i I

I e

R W

c S

L t

t s

L t

t t

A i

e A

1 1

1 L

I n

it i

I A

i l

i X

U h

n i

n 2

h D

X U U U I

U U

I t

A A

X t

I I

A i

R R

I n

T a b c

T a b c

B a.

b c

U

-N F

1 2

3 B

l u'uN ~

E a$a &.

u s

3.2. AAAEA In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates astion to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary systes isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the Specifications are (1) to. assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such' systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.

When necessary, one channel may be made inoperable for brief intervals to

.l conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances arplicitly stated where the high and low values are both critical and may have a substantial effect on j

safety. The setpoints of other instrumentation, where only the high or low and of the setting has a direct bearing on safety, are chosen at a level away fr n the normal operating range to prevent inadvertent actuation of Iao safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 538 inches above vessel zero closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor water level instrumentation that is set to trip when reactor water level is 470 inches above vessel zero (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems. The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain vs17es (Group 7).

The low water level instrumentation set to trip at 378 inches above vessel zero (Table 3.2.B) closes the Main Steam Isolation Valves, the l

l Main Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1).

Details of valve grouping and required closing times are given in Specification 3.7.

These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.5) i BFN 3.2/4.2-65 Amendment No. 157, 163 i

Unit 2 1

l.

3.2 R&gg (Crnt'd)'

initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators.. These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operatica so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to

'the complete circianferential break of a 28-inch recirculation line and with the trip. setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water-level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves.. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also.

ADS provides for automatic nuclear stieam system depressurization, if needed, for small breaks in the nuclear system so that the LPCI and the CSS can operate to protect the fuel from overheating. ADS uses six of the 13 MSRVs to relieve the high pressure steam to the suppression pool.

ADS initiates when the following conditions exist: low reactor water level permissive (level 3), low reactor water level (level 1), high drywell pressure or the high drywell pressure bypass timer timed out (12 1/2 min.), and a 105 second time delay. In addftion, at least one RER pump or two core spray pumps must be running.

The high pressure bypass timer is added to meet the requirements of NUREG 0737, Item II.K.3.18.

This timer will bypass the.high drywell pressure permissive after a sustained low water level. The worst case condition is a main steam line break outside primary containment with HPCI inoperable. With the bypass timer set at 15 minutes, a Peak Cladding Temperature (PCT) of 1424' F is reached for the vorst case event. This temperature is well below the limiting PCT of 2200* F.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass' inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity to the environs is well below 10 CFA 100 guidelines. Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided.on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200*F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 spe; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steaa

]

BFN 3.2/4.2-66 Amendment No. 162, 163 Unit 2

~

e

,e 3.2 548E1 (Cont'd) flow instrumentation is a backup to the temperature instrumentation. In the event of a loss of the reactor building ventilation system, radiant heating in ths vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases.can cause an unnecessary main steam line isolation and reactor seras. Permission is provided to bypass the temperature trip for four hours to avoid an

. unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

Hia;h radiation monitors in the main steam line tunnel have been provided to detect. gross fuel failure as in the control rod drop accident. With the established nominal setting of three times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference Section 14.6.2 FSAR. An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also.

Pressure instriasentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 pais.

The EPCI high flow and temperature instriasentation are provided to detecc a break in the EPCI steam piping. Tripping of this instrinnentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE.

High temperature in the vicinity of the HPCI equipment is sensed by four sets of four binetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system.

The HPCI trip settings of 90 psi for high flow and 200*F for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" H O for high flow and 2

200*F for temperature are based on the same criteria as the HPCI.

High temperature at the Reactor Water Cleanup (RWCU) System floor drain in the space near the RWCU system or in the space near the pipe trench containing RWCU piping could indicate a break in the cleanup system.

When high temperature occurs, the cleanup system is isolated.

The instrtamentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system sven during periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed.

BFN 3.2/4.2-67 Amendment No. 156, 163 Unit 2 e

-l v

l l

3.2 A&AEA (Cont'd)

The contro1~ rod block functions are provided to prevent ascessive control rod withdrawal so that MCPR does not decrease to 1.07.

The trip logic for this function is 1-out-of-n e.g., any trip on one of six APRMs, i

'eight IBMs, or four SENs will result in a rod block.

'l The minis m instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is set. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This does not significantly increase,the risk of an inadvertent control rod withdrawal, as the other i

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 reduction in MCPR, especially during operation at reduced flow. The APRM provides gross core protection; i.e.,

limits the gross core power increase from. withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than 1.07.

4 The RBM rod block function provides local protection of the core; i.e.,

the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.

If the IBM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IBM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.

A downscale indication is an indication the instrument has failed or the I

instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

i For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary r,nd minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, i

testing, or calibration, and also minimizes the risk of inadvertent

{

operation; i.e., only one instrument channel out of service.

Two post treatment off-gas radiation monitors are provided and, when their trip point is reached, cause an isolation of the off-gas line.

Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip or both have a downscale trip.

BTN 3.2/4.2-68 Amendment No. 163 Unit 2

E-w 3.2 34331 (Crat'd)

Both instruments are required for, trip but the instruents are set so that the instantaneous stack release rate limit given in Specification 3.s is not exceeded.

Four radiagien monitors are provided for each unit which initiate Primary Containaams Isolation (Group 6 isolation valves) Reactor Building Isolation and operation of the Standby Gas Treatment System. These l

instrument channels monitor the radiation in the reactor zone ventilation l

exhaust ducts and in the refueling sone.

Trip setting of 100 ar/hr for the monitors'in the refueltag zone are based upon initiating normal ventilation isolation and SGTS operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the SGTS.

Flow integrators and sump fill rate and ptmp out rate timers are used to determine leakage in the drywell. A system whereby the time interval to fill a known voltme will be utilized to provide a backup..An air sampling system is also provided to detect leakage inside the primary containment (See Table 3.2.3).

For each parameter monitored, as listed in Table 3.2.F, there are two channels of instrumentation except as noted. By comparing readings between the two channels, a near continuous surveillance of instrument performance is available. Any deviation in readings will initiate en early recalibration, thereby maintaining the quality of the instriment readings.

Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to l

the control room. An accident signal that isolates primary containment will also automatically isolate the control room and initiate the emergency pressurization system. In addition, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressurization system. Activity required to cause automatic actuation is about one mrem /hr.

Because of the constant surveillance and control exercised by TVA over the Tennessee Valley, flood levels of large magnitudes can be predicted in advance of their actual occurrence. In all cases, full advantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted; however, the plant flood protection is always in place and does not depend in any way on advanced warning. Therefore, during flood conditions, the plant will be permitted to operate until water begins to run across the top of the pumping station at elevation 565. Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition. At that time an orderly shutdown of the plant will be initiated, although surges even to a depth of several feet over the ptaping statica deck will not cause the loss of the main condenser circulating water pumps.

BFN 3.2/4.2-69 Amendment No. 163 Unit 2

l o

3.2 R&g31 (Cont'd)-

The operability of the meteorological instrumentation ensures that sufficient meteorological data is'available for estimating potential radiation dose to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.

The operability of the seismic instrumentation ensures that sufficient i

capability is available to promptly determine the seismic response of

(

those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for Browns Ferry Nuclear Plant and to determine whether the plant can continue to be operated safely. The instrumentation provided is consistent with specific portions of the recommendations of Regulatory Guide 1.12 " Instrumentation for Earthquakes."

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments will be calculated in accordance with guidance provided in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes' provisions for monitoring the concentration of potentially explosive gas mixtures in the offgas holdup system. The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in

)

liquid effluents during actual or potential releases of liquid l

effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with guidance provided in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20 Appendix B, Table II, Column 2.

The OPERABILITY and use of this instrumentation is consistent with the requirements.of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

l ATWS/RPT, Anticipated Transients without Scram / Recirculation Pump Trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an ATWS event. The response of the plant to this postulated event (ATWS/RPT) follows the BWR Owners Group Report by General Electric NEDE-31096-P-A an( the accompanying NRC Staff Safety Evaluation Report.

ATWS/RPT utilizes the engineered safety feature (EST) master / slave analog trip units (ATU) which consists of four level and four pressure channels total. The initiating logic consists of two independent trip systems each consisting of two reactor done high pressure channels and two reactor vessel lov level channels. A coincident trip of either two low levels or two high pressures in the same trip system causes initiation of BFN 3.2/4.2-70 Amendment No. 161, 163 Unit 2

3.2 AAAEA (Cont'd).

ATWS/RPT. This signal from either trip system opens one of two BOC (and-of-cycle) breakers in series /(the other system opens the other breaker) between the pump motor and the Motor Generator set driving each recirculation pump. Both systems are completely redundant such that only one trip system is necessary to perform the ATWS/RPT function. Power comes from the 250 VDC shutdown boards.

Setpoints for reactor done high pressure and reactor vessel low level are such that a normal Reactor Protection System scram and accompanying recirculation pump trip would occur before or coincident with the trip by ATWS/RPT.

- 4.2 R&II,1 The instrumentation listed in Tables 4.2.A through 4.2.F will be functionally tested and calibrated at regularly scheduled intervals. The same design reliability goal as the Reactor Protection System of 0.99999 generally applies for all applications of (1-out-of-2) X (2) logic.

Therefore, on-off sensors are tested once/3 months, and bistable trips-l associated with analog sensors and amplifiers are tested once/ week.

Those instruments which, when tripped, result in a rod block have their contacts arranged in a.1-out-of-n logic, and all are. capable of being bypassed. For such a tripping arrarigament with bypass capability provided, there is an'optistan test interval that should be maintained in order to maximize the reliability of a given channel (7). This takes account of the fact that testing degrades reliability and the optistui interval between tests is approximately given by:

i=

Where:

i= the optimum interval between tests.

l-t=

the time the trip contacts are disabled 1

from performing their function while l

the test is in progress.

I r=

the expected failure rate of the relays.

To test'the trip relays requires that the channel be bypassed, the test t

made, and the system returned to its initial state.

It is assumed this task requires an estimated 30 minutes to complete in a thorough and worirmanlike manner and that the relays have a failure rate of 10-6 failures per hour. Using this data and the above operation, the optimum test interval is:

2M.D '

3 i=

= 1 x 10

-6 h

10

= 40 days BFN 3.2/4.2-71 Amendraent No. 161, 163 Unit 2 1

.4.2 BASES (Crat'd)

For additional marain a' tant interval of anaa nar month will be used initially.

The sensors and electronic apparatus have not been included here'as these are analog devices with readeuts in the control room and the sensors and electronie apparatus can be checked by comparison with other like instruments. The checks which ore made on a daily basis are adequate to assure operability of the sensors and electronic apparatus, and the test interval given above provides for optimum testing of the relay circuits.

The above calculated test interval optimizes each individual channel, considering it to be independent of all others. As an example, assume that there are two channels with an individual technician assigned to each. Each technician tests his channel at the optimum frequency, but the two technicians are not allowed to communicate so that one can advise the other that his channel is under test. Under these conditions, it is possible for both channels to be under test simultaneously. Now, assume that the technicians are required to communicate and that two channels are never tested at the same time.

(7) UCRL-50451, Improving Availability and Readiness of Field Equipment Through Periodic Inspection, Benjamin Epstein, Albert Shiff, July 16, 1968, page 10, Equation (24), Lawrence Radiation Laboratory.

Forbidding simultaneous testing improves the availability of the system over that which would be achieved by testing each channel independently.

These one-out-of-n trip systems will be tested one at a time in order to h

take advantage of this inherent improvement in availability.

Optimizing each channel independently may not truly optimize the system considering the overall rules of system operation. However, true system optimization is a complex problem. The optimums are broad, not sharp, and optimizing the individual channels is generally adequate for the system.

The formula given above minimizes the unavailability of a single channel which must be bypassed during testing. The minimization of the unavailability is illustrated by Curve No. 1 of Figure 4.2-1 which assumes that a channel has a failure rate of 0.1 x 10-6/ hour and 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is required to test it.

The unavailability is a minimum at a test interval 1, of 3.16 x 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />.

If two similar channels are used in a 1-out-of-2 configuration, the test interval for minis m unavailability changes as a function of the rules for testing. The simplest case is to test each one independent of the other.

In this case, there is assumed to be a finite probability that both may be bypassed at one time. This case is shown by Curve No. 2.

Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test interval. Thus, if the two channels are tested independently, the equation above yields the test interval for minis m unavailability.

BFN 3.2/4.2-72 Amendment No. 163 Unit 2 i

3 4.2 BASER (Cont'd)

A more' usual case is that the testing.is not done independently. If both channels are bypassed and tested at the same time, the result is shown in Curve No. 3.

Essa that the minimum occurs at about 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, much

. longer than for asses 1 and 2.

Also, the minimum is not nearly as low as Case 2 which indicates that this method of testing does not take full advantage of the redundant channel. Bypassing both channels for simultaneous testing should be avoided.

The most likely case would.be to stipulate that one channel be bypassed, tested, and restored, and then immediately following, the second channel be bypassed, tested, and restored. This is shown by Curve No. 4.

Note that there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel..

The best test procedure of all those examined is to perfectly stagger the tests. That is, if the test interval is four months, test one or the j

i other channel every two months. This is shown in Curve No. 5.

The difference between Cases 4 and 5 is negligible. There may be other argissants, however, that more strongly support the perfectly staggered tests, including reductions in human error.

-The conclusions to be drawn are these:

1. 'A 1-out-of-n system may be treated the same as a single channel in terms of choosing a test interval; and

'2.

more than one channel should not be bypassed for testing at any one time.

The radiation monitors in the refueling area ventilation duct which initiate building isolation and standby gas treatment operation are arranged in two 1-out-of-2 logic systems. The bases given for the rod blocks apply here also and were used to arrive at the functional testing frequency. The off-gas post treatment monitors are connected in a 2-out-of-2 1 mic arrangement. Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate.

The automatic pressure relief instrumentation can be considered to be a 1-out-of-2 logic system and the discussion above applies also.

)

l The criteria for ensuring the reliability and accuracy of the radioactive

]

gaseous effluent instrumentation is listed in Table 4.2.K.

The criteria for ensuring the reliability and accuracy of the radioactive liquid effluent instrtusentaiton is listed in Table 4.2.D.

j i

BFN 3.2/4.2-73 Amendment No. 163 Unit 2 J

/

k, UNITED STATES NUCLEAR REGULATORY COMMISSION a

.{

j wAsmwoTow, o. c. nossa j

\\,...../

)

1

~

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.136 License No. DPR-68 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated September 29, 1988, compl.'es with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

I

s.

2.

'Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-68 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.136. are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 4

N pys, S anne ack, Assistant Director or Projects TVA Projects Division Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 13,1989 i

~

' ATTACHMENT TO LICENSE AMENDMENT NO.136 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 1

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.

REMOVE INSERT 3.2/4.2-37 3.2/4.2-37 3.2/4.2-57 3.2/4.2-57 3.2/4.2-68 3.2/4.2-68*

3.2/4.2-69 3.2/4.2.

l 4

I L L L g

r

• 3 o o =

g QQ5 kkk d d d 2

1 5

32S 5

888 888 N N "

d 5

5 4ii 444 kbN m

o o o o o o o o o

• c 2.c 2

o i

5 N

3 5

5 5

2 5

n 5

5 o

G 3 y C

5 5

G i

m

~

Ec c

G I

c c c

a g

9 o

2 1

1 y

1 1

1 w ;

I g

d c

c g

3 3

E yo o 3 3

3 3

~

o 3

g go o

o Qf f

f 1

1 5

a i

a ag g ; R3 i

i g

g2 2

2

=j k

W a W t t

t w

l a

a 5 :

2 C

t c:

t

-a u

e u 6

6 E

4-5 g-a -

u yaa5 gaaa5 saa o

E 5 4 i J 5 4 i J E 4 i J P

m n

m a

3.2/4.2-36 i

}

?

no i

tar b

i la C

R R

R R

R R

R R

R lenna h

C t

se T

l S

a T

n N

o E

i M

t E

c R

n I

u U

F A

A A

Q A

A A

/ / /

A A

A E

l S

S S

N N

N S

S S

R en E

n C

a N

h A

C L

L I

E VR J.

k U

c 2

S e

h y

y y

y y

y 4

TN C

l l

l l

l l

h h

h h

h h

e E

l t

t t

t t

t l

M e

n n n

A A A n

n n b

U n

o o

o

/ / /

o o

o a

R n

M M M N N N

M M M T

T a

S h

N C

I GN I

R

)

)

O 5

5

)

T 2

0

)

)

)

)

I 5

0 0

0 N

0 1

6 4

O 9

2 5

4 9

9 9

M 5

5 1

6 1

1 1

C l

5 5

5 I

1 E

1 M

l E

(

E l

1 1

S E

(

(

E E

E I

S

(

b

)

(

(

(

E H

b a

5

)

e S

7 0

c b

b b

P b

a l

i a

a a

A a

l s

t l

s 5

0 n

l l

l iG s

e 2

8 s

s s

O r

s 6

5 4"

e e

e R

e o a E

s o

b S

1 s

s s

L a

l H

l 1

a a a t

E b

f P

E E

b b

b n

C a A

(

(

m m

A a

d R

e S

e C

. a i G

e e t

E a a a u

d d b O

o o s

H d

d d

r Y

1 i

R l

l y

C 1

1 i

t R

b b

r E

a a s

T b

b b

s O

a L

I v W r

r r

n T

r r

t E

0" 6"

a S

o o o I

S o o e C

I t

t C

1 1

r t

t t

H c

c -

A a

C c

c c

a a s

I a a a d e e -

K W W M

e e e e

S r

r r

f r

r -

A C

I E

C o

r I

T I

I l

P A

p a

E I

I I

e R

E c

S L

t t

s L

t t

t 1

L i i i A i i e A 1

1 i

I A

n n h 3

I n h W4 WI U U A

t D

X t

X U

A X

t i

I I

A I

n R

R T

a b c

T a b c

B a

b c

U N

F 1

2 3

B l

$aRo" N.

o Nh

~

7 t

h 3.2 3A331 (Cont'd)

Trip setting of 100 ar/hr for the monitors in the refueling zone'are based upon initiating normal. ventilation isolation and SGTS e.peration so that none of the activity released during the refueling accident leaves

-the Reactor Building via the normal ventilation' path but rt.ther all the activity is processed by the SGTS.

Flow integrators and sump fill rate and pump out rate timees are used to determine leakage in the dryvell.. A system whereby the time interval to fill a known volume will be utilized to provide a backup. An air sampling system is also provided to detect leakage inside the primary containment (See Table 3.2.E).

For each parameter monitored, as listed in Table 3.2.F, there are two channels of instrumentation except as noted. By comparing readings between the two channels, a near continuous surveillance of instrument performance is available. Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.

Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control. room. An accident signal that isolates primary containment

. will also automatically isolate the control room and initiate the emergency pressurization system. In addition, there are radiation monitors.in the normal ventilation system that will isolate the control room and initiate the emergency pressurization system. Activity required to cause automatic actuation is about one mrem /hr.

Because of the constant surveillance and control exercised by TVA over the Tennessee Valley, flood levels of large magnitudes can be predicted-in advance of'their actual occurrence. In all cases, full advantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted; however, the plant flood protection is always in place and does not depend in any way on advanced warning. Therefore, during flood conditions, the plant will be permitted to operate until water begins to run across the top of the pumping station'at elevation 565. Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition. At that time an orderly shutdown of the plant will be initiated, although surges even to a depth of several feet over the pumping station deck will not cause the loss of the main condenser circulating water pumps.

BFN 3.2/4.2-64 Unit 3

3.2 RA111 (Cent'd)

S 4

The operability of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation dose. to the public as a result of routine or accidental release of radiomative materials to the atmosphers. This capability is required to evaluate the need.for initiating protective measures to protect the health and safety of the public.

The operability of the seismic instrumentation ensures that sufficient capability.is available to promptly determine the seismic response of l

those features.important to safety. Tais capability is required to permit cor.parison of the measured response to that used in the design basis for Browns Ferry Nuclear Plant and to determine whether the plant can continue to be operated safely. The instrumentation provided is consistent with specific portions of the recommendations of Regulatory Guide 1.12 " Instrumentation for Earthquakes."

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments will be calculated in accordance with guidance provided in the ODCM to ensure that the alare/ trip will occur prior to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring the concentration of potentially explosive' ass mixtures in the offgas holdup system. The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with guidance provided in the ODCM to ensure l

that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20 Appendix B, Table II, Column 2.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

j i

ATWS/RPT, Anticipated Transients without Scram / Recirculation Pump Trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an ATWS event. The response of the plant to this postulated event (ATWS/RPT) follows the BWR Owners i

Group Report by General Electric NEDE-31096-P-A and the accompanying NRC Staff Safety Evaluation Report.

l ATWS/RFT utilizes the engineered safety feature (E1J) master / slave analog trip units (ATU) which consists of four level and four pressure channels i

total. The initiating logic consists of two independent trip systems sach consisting of two reactor done high pressure channels and two reactor vessel low level channels.- A coincident trip of either two low levels or two high pressures in the same trip system causes initiation of ATWS/RPT. This signal from either trip system opens one of two E0C BFN 3.2/4.2-69 Amendment No. 135, 136 Unit 3 A

_ _ _ _. _ _ _ _ _ _. _ _