ML20009D272

Amend 58 to License DPR-36,revising Tech Specs to Support Operation of Unit at Full Power During Cycle 6

ML20009D272
Person / Time
Site:	Maine Yankee
Issue date:	07/10/1981
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20009D270	List: ML20009D267 ML20009D272 ML20009D276 ML20009D278
References
NUDOCS 8107230323
Download: ML20009D272 (24)

Text

.=

),

is

/

[$.

[h N UC L E AR R E G U L I,i O R Y CO*.* *.*d SSION t

W E

n r.*> $""G T O *J D C 70%5

~#(('

~-

MAINE YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-309 s

MAINE YANKEE ATOMIC POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE i

Amendment No. 58 License No. DPR-36 1.

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

A.

The applications for amendment by Maine Yankee Atomic Power Company, (the licensee) dated February 13, March 25, April 28, and June 30, 1981, as supplemented, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, 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; The issuance of this a[nendment will not besinimical to the common D.

defense and security or to the health and safety of the public; a(d E.

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

D DO O O 03 9 P

PDR 4

w,n.

. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.B(6)(b) of Facility Operating License No. DPR-36 is hereby amended to read as follows:

( b") Technical' Specifications 4

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 58, 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 the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

m. l.. Q' (.. --] 4 GA i

,, ~..

Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 10,1981 s.

l l

l l

1 h

' s ATTACHMENT TO LICENSE AMENDMENT NO. 58 TO FACILITY OPERATING LICENSE NO. OPR-36 DOCKET NO. 50-309

~

s Revise Appendix A as follows:

Remove Pages Insert Pages I

1 2

2 3

3 2.1-2 2.1-2 2.1-4 2.1-4 2.1-5 2.1-5 3.4-8 3.4-8 3.4-9 3.4-9 3.10-2 3.10-2 3.10-10 3.10-10 3.10-11 3.10-11 3.10-12 3.10-12 3.10-13 3.10-13 3.19-1 3.19-1 3.19-2 4.6-1 4.6-1 4.6-2 4.6-2 4.6-3 4.6-3 4.6-4 4.6-4

\\

4.6-5

{

I s

~

I e

/

e

TECHNICAL SPECIFICATICNS DEFINITIONS The following terms are defined for. uniform interpretation of these Technical Specifications:

REACTOR' OPERATING CONDITIONS Rated FoGer A steady-state reactor core output of 2630 Mut.

Reactor Critical i

The reactor is considered critical for purposes of administrative control when the neutron flux logarithmic range channel instrucentation indicates greater than 10 % of rated power.

Power Operation Condition When the reactor is critical and the neutron flux power range instrumentation

-indicates greater than 2% of rated power.

Hot Standby Condition The reactor is considered to be in a hot standby condition if 'the average tempe'rature of the primary coolant (Tavg) is greater than 500 F and any of the control rods are withdrawn and the neutron flux power range instrumentation indicates less than 2% of the rated power.

Hot Shutdown Condition When the reactor is suberitical by 5%dik/k and Tavg is greater than 500 F.

Refueling Shutdown Condition When the primary coolant is at refueling boron concentration and Tavg is l

1ess than 210 F.

Cold Shutdown Condition When the primary coolant is at cold shutdown, boron concentration and Tavg is less than 210 F.

Amendment No. 58

hf; =lig c;r ration Any eperation involving movement of core components when the vessel head is unbolted or removed.

Low Power Physics Testing Testing performed under approved written procedtres to determine control rod worths and other core nuclear properties. Reactor power during these and tests shall not exceed 2% of rated power, not including deca'; heat, primary system temperature and pressure shall be in the rarge of 260 F to Certain deviations from 550*F and 415 psia to 2300 psia, respectively.

normal operating practice which are necessary td enable performing some -

of these tests are permitted in accordance with the specific provisions in these Technical Specifications.

Power kgnge Physics Testing Tests performed under approved written procedures to verify core nuclear Reactor design properties at power and plant response characteristics.

Primary system power may be greater than 2% during these measurements.

average temperature and pressure shall be in the range of 500 F to 580 F Certain deviations from and between 1700 psia to 2300 psia, respectively.

normal operating practices which are necessary to enable the performance of some of 'these tests are permitted in accordance with specific provisions of these Technical Specifications.

Cold Shutdown Boron Concentration

~

The boron concentration shall be suf ficient to mainta'in the reactor at least 5%21k/k subcritical with all control rods in the core.

l Refueling Boron Concentration The boron concentration shall be sufficient to maintain the reactor at least 5%2hk/k suberitical under all refueling conditions.

Quadrant Power Tilt The difference between nuclear power in any core quadrant and the average in all quadrants.

Power in any quad

-1

% TILT = 100 x avg. power of all quad g

C o

e Amendment No. 58 O

?ES..TM F^0TECTIVE SYSTEM Trstrument Channeln One of four independent measurement channels, complete with the sensors, sensor power supply units, amplifiers, and trip modulea provided for each safety parameter.

Reactor Trip The de-energizing of the magnetic jack holding coils which releases the shutdown and regulating control rods and allows them to drop into the core.

a Trip Module A' bistable unit in each nf the instrument channels which is tripped when The relay contact outputs the para, meter signal exceeds a specified limit.

of the trip modules form the reactor protective system logic.

ENGINEERED SAFEGUARDS SYSTEMS Subsystem Each engineered safeguard, system can be initiated by either of two redundant Each, subsystem contains sensor, logic and circuitry fcr engineered subsystemss 4

e I

e e

q.

~

t 9

1 r

l l

4 e

Amendment No. 58

-3_

1

I LIMITING SAFETY SYSTEM SE1 TING - REACTOR FFOIECTION SYSTEM

2.1 Applicability

Applies to reactor trip settings and bypasses for the channels monitoring the process variables which instrument influence the safe operation of the plant.

that Objective:

To provide automatic protective action in the event the process variables approach a safety limit.

Specification:

The reactor protective sistem trip setting limits and bypasses for the required operable instrument channels shall be as fo11cvs:

.'}.1.1 Core Protection a)

Variable Nuclear Overpower f q+10, or 106.5 (whichever is smaller) for 10<Qf100; <20 for Q < 10.

where

/Q = Percent ther=al or nuclear power, whichever is larger.

b)

Thermal Margin / Low Pressure 2.A QDNB + BTc +5 C, or 1835 psig (whichever is larger) where T

. = cold leg tempera,ture, OF c

A

= 2060.7 B

= 17.9 C

= -10053

=Al x QR1 QDNB Al and QR1 are given in Figure 2.1-la and 2.1-lb, respectively.

f b

This trip may be bypassed below 10 percent of rated power.

l c)

The symmetric offset trip and pretrip function shall not exceed the limits shown in Figure 2.1-2, for three loop operation. This trip may be bypassed below 15 percent of rated power.

d)

Low Reactor Coolant Flow

>93 percent of 360,000 GPM (3 pump operation)

This trip may be bypassed below 2 percent of rated power.

1 2.1-1 Amendment No. 29, 38, 49, f$. 58 h

,,,.,,r

--__,yy,yg.yy

,y,,_..e'=

rF"**'

• 4here:

Q,,N B =A X QR c.

1 1

tri and p P = 2060.7 + 17.9 T.

-10053~

e var in T.

= cold leg temperature, OF in 1.30 1

_....._._.._.E.._._.._

.i.

_._. __ _. s.. _. _.

.i__._._._...-

. _.. _. _..._. _.. _. _... _... t. _...... _ _

.. _ _ _ _~-F- - _._...r_._- 3_--- -., _ _.. _. _... __.. _ +.. _..

1. _,0

_.. _..._.w..___.i__.._..__.

..__.4.__

y._

r......

....... s.-....._..

g 1,15

1.,_

1

. _ _. _ ___.._L..__..__...

_..._....__.._.r_..... _..

t-

..__._.t.__._______. _ _.... _......

u_..._=

____.m

__......_.a_..

_ _. _.. _. +

2.

.._..g.....

.... _...._....,4 p _.__.. _ __

_..y.

1.10

. _ ___4

_._..a _ _

1

.04. 70

_ _. _...... _. _......._.. 4.

=

___.i___..4.___.__-

.__.__..J...___.

.i

__=..... _. _... _.. _. _.. _..

1_.. /.._

__...., ~

_.. g 3 71 +__, 990 _. _ __...

6 1,05 l

._..s._.

i

. _. _ _, _ -. _...__.s........

.... -j.

i _......

.__.t.._..

...._.__.J.t__......

s...... _.....

.... s,..

4

( _.._...... _.... ___

....s,...

g.00 l

i

.5

.4

.3

.2

.1

.0

.1

.2

.3

.4

.5 A U-L

• Symmetric Offset Y

=

g U+L MAINE YANKEE Therma 14(argin/ Low Pressure Trip Setpoint Figure Technical versus 2 1-la l

Specification (Yr versus A )

2.1-4 1

\\

n Amendsen,t No. 29,. 38, Ag, Ag, 58'

• 'h e ra : A x QR =

O_.. B I

1 u..

trip - 2060.T Q

+ 17.9 T

- 10053 and p DNB in var T

= cold leg temperature g

1.2l

.i a

I f

I 3...

I l

h6 5

1.0 I

e

g..... i.

g,

. j..

i t

h

'~

I s

+

.l l.

.. -".l, T) -

. L...

i..

.l.

jg2 @ '.... '.

.8

.......i......;l.

...jg :. -- -

.t..:..

g:.-. : n.]......

i?l.3 3.j. '

i: !

~~ '.j

i;inq. iijti 33 iii j

.. W.

q:.:7 I,7"iF l..: i,.:

3....

• 6 QRI

.....1:..

.....l':....

li:

.g.

.;j j.; -

.?T ' '

i:.:

.t.:

.l..

I

-). -

4-.

. i.:

r-4

.l

.l.

.4

...3,

_ l.:....

.l.

..:. : [.

....s

,3 s

g

\\\\.

..h:

..f *

. l l."

~! '

..j.:

- l:.

-=

e. n.

• =l.. -:-l. -

i-2 i

c g..

-l 2_ w.:..:+.

...p)...;._... ;. l._.;;;.

..l;;

- e _. i...._.._ i

.i 1.i i

y 0-

~

O

.2

.4

.6

.8 1.0 1.2 s

Fraction of Rated Thermal Power MAINE YANKEE n ermal Margin / Low Pressure Technical Trip Setpoint Part 2 Figure

/

$pecifications (Fraction of Rated n ermal Towe versus QR )

2.1-lb!

1 2.1-5 l

Amendment No. 29, 38, #9, 48, 58

m.

.w...,. n.,.

Es.

FLUT.NCE VS. 33"JP FOR 1/4T AND 3/4T POSITIONS FOR ?>._NE Y12;KIE s

2 i

+

l E

19

\\ f 0-~ ' ;

10

.-.---=_.====1

, y. _ -- r = -

- -. ~=== wg5g. --

f__' _. _....

.-=

_s - -

.=_._/

J

/

t s

1 E

s' g>

1

=a E.

S A

8

/

o w

n. 5

.r a i g,a >

f i

a

-v a.

a y

r s

e-s f

o s

8 t

/

I gg; I

i t

hr t

y a

q l

l f

=t 1

y i

/

5'

/

i 18

._ f 10 f

i i

1 s.

I I

g s

r i

i I

.t

~

t 1

\\

I I

i

{

f l

E.

o I

l' l

i e

I 1017 i

f i

1 2

3 4

5 6

7 8

WHch x 10 3.4-8

% 58

?-

=""T'"

I r

O !U[id id',U}h l. "

. 6 tjl i

q!

t l

I'

l.!'

,i?;

,l!

I I

I 5

e : '

,I

~

1I l"

.F ii

.l 4

ili ij b.

,j ;. i. -.

l nl

I 9

n l3 1

3 0

1 9

9 l

1 N

E 2

T D17 I

F T

7

~

J N9 m

GC 1 1

i Y1 I l

C Rr I M

OWH l i1 MnA I

I l

M I

U i

1 d.

i l

G E

i'

. M9 R

i\\l 8

iI l

7 i1 6

i1

)

l, y;

5

,.g A

E

\\.

~

1 C

N i

4 y

N E

l g

ll n.

g g

I 8

y F

j g

3 1 2 l

0 n 1

T

/a R

N H

n T

C I

2 E

.F C

a, f

F c

+

m I

D I

f u

1 S

R I

1 r

/

/

1 I

'tI

,I I.

9 o

l 8

'b 7

6' J5 Ib 7

j 1

~

4 0

l 1

l l.

3

'd l

]

2 0

0 0

0000 0 0 0

0 0

0987 6 5 5

4 2

1 s

p~

Ng' t-s i $ i.s,r f+

x y.o P7*

>i!,i1' 1:!

]jjilj!

jji 1!

ij;j l

i;

,il1 1

l'GU u

14 kw/ft E > 0.50 and C.O > 792 s0/gu L 16 kw/ft E < 0.50 L-Exposed Fuel:

14.0 kw/ft 1 > 0.50 L

16.0 kw/ft E < 0.50 L-where E is fraction of' core height and CAB is c'ycle averaghburnup.

Should any of these limits be exceeded, immediate action will be taken to restore the linear heat rate to within the appropriate limit specified above.

P The total radial peaking f actor, defined as Ff = F 2.

(1 + Tq), shall be evalueted at least once a month during power operation above 50% of rated full power.

/

2.1Ffisthelatest available unrodded radial peak determined from the incore monitoring system for a condition where all CEAs are at or above the 100% power insertio'n limit. T9 is given by the following expression:

'(Pa-Pc)2 + (Pb-Pd)2 T

=

Pc'Pd)2 9

(Pa+Pb +

Pi = relative quadrant power determined from incore system for quadrant i, when the incore system is operable and by Specification 3.10.B.4 othervi,e.

l 2.2 If the measured valu6 of Ff exceeds the value given in Figure 3.10-4, perform one of the following within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

a) Reduce symmetric offset pre-trip alarm and trip band (Figure 2.1-2), thermal margin / low pressure trip limit (Figure 2.1-1 and Tech.

Spec. 2.1), and excore IDCA monitoring limit (Figure 3.10-3) by a factor:

T

~

1 FR measured i

Ff(Figure 3.10-4) or Amendment No.fg, Jg, gg, pg, SR 3.10-2

l Cy c ! e A e. a t;c E.u.7..:p <,.'9 2 ". D/.' 70

~~~-~~'r-

~

i i-i i.I i

100'

..f I

i i

l

-l s

......l

{.

i j.

j 1

' l

'l 90 l

i l

-l ---

i C.-03,87.-C)

I h

3 l\\

i S 0.

~

i

\\

.c.

,..:.u,

~

'j.

l

- / l

..l-i i

!/i l

l

~

i I

/

i 1

\\

l l

i

- /t o.

j p

t.. M,69.1.)

l l

i if.

i-l

.i-A l

.i

\\

l I

i

. i'..

--( -4fh61.))

I J

60 l

L,'

• W M. * ) ()

.1 j-d l

J.

t I

t 5&

l.

.=

-.r.

1.

]

l l

l.

j.

~

i.-.

I 40-i

Al l

i l

l I

8 30 I -

j I

i 20[

.5

.4

.3

.2

.1 0

+.1

+.2

+.3

+.4

+.5 Excore Monitor S.O. Alarm Band for LOCA Limiting Conditions of Operation When Incore gf0 2 MAINE YANKEE 3.

-Technical Monitors are Inoperable Specifications 3.10 10 Amendment No. 58

Cyc:e u.:..

wp > 73: ' 'O, :.TU 1

l l.

l' 100 i

l I

l 1:

l i

j I.

l E~2 I

l 90

(.08.38.5)

,(.04,88'.5)!

l t

/

I l

\\./

l l

/

l 80

/

i l

i l(.19.j79.0) j

\\

l l/

i

\\

70

/

lk

,/

l N

t i

A '

t

- : /:

t

-J-t

\\

s

. Q.(.40,64.,1.)

t,

+.aV37 T),-

-6.-

l.

60 i

. l, ;

jt.

l l

l i

1

...t -

_.. r.

?

50 l

n.:..

.. }. :-

- ;; ~

.i::~,. -

t

=.L -

40

l.....

~

.i

.s

. s i.

l l.i

j. j l

4 j

30 i

[.

j

.j:

.l'

{

1

t..

t j

i I

'1 1

20

.5

.4

.3

.2

.1 0

+.1

+.2

+.3

+.4

+.5 i

J l

Excore Monitor 5.0. Alarm Band for LOCA

]

MAINE YANKEE

,, Lihiting Conditions of Operation When Incore Technical Monitors a g goperable Specifications z

M b. N o ',58 w--w.

';o t e :

1.

This curve ine'

?.-s 10*. calcuIstic -al u cert ain'.y 2.

F{=F[*1.03 3.

Measured F{ should be augmented by measure:r.ent

-uncertainty (8%) before ccciparison to this curve.

a i

I i

I i

l l

i 1' - ( Q-i 1

7-I I

1-

.(

0, 1.6203 )

(.05,'1.6084 )

1.65 (0.5, 1.6082 )

(1.0, 1.6015 )

j (2.0, 1.6070 )

(4.0, 1.6297 )

(6.0, 1.6227-)

(8.0, 1.6320 )

1.67 l

(10., 1.6538 )

(12., 1.6643 )

N l

1.66 I

v I

l s

1.65

.j 1.6%

I l

I

/

l

.d'-

i fsf' 1~6' 1

i i/N

./

i l

lY i

I

[

bA 1.6.,

.:-i. f.

i l.i r

l l

.t.

g

I

l l

r ixq s.;.

26 i

i i[

l' j

j j

j:

l' 0

2 4

6 8

10 12 Cycle' Average Exposure (K.WD/MT)

Allowable Unrodded Radial Peak Versus Fi gure MAINE YANKEE Technical Cycle Average Burnup 3.10-4 3.10-12 Specification 58 Amanonent No. 38, 48, $.$,

CEA's are to be maintained at or above the 100%

Note:

power insertion limit when applying 3.10.B.2.2b 4

110 i I

i l

i i

i s

I I

I j

. A 2.7.100' I

i-

-l.

l y

l l

I

~

VI-5 N

i 7.95,90 s

0 90 i

I 3

l V 'h' 2

i j

i w

I o

I 14.4,80 i

l h

~ !._ZIN i

s

. l....

l I

.i.

l 70 I

l 0

2 4

6 8

10 12 4 14 16 18 i

l l

T I * '

Allowable % Increase in FR l

l IO Increase in Total Radial Figure Allowable Power Level vs.

3 10-5 Technical Peak Specification 3.10-13 Amendment No. 38, #93 fS, 58 1

3.19 SAFETY INJECTION SYSTEM Applicability: Applies to the condition of safety injection system.

Objective:

To define the condition of the safety injection system required during reactor operation.

Specification: a)

None of the following valves may be closed unless the reactor is suberitical.

~

1.

Any safety injection tank isolation valve (SIA-M-11, 21, 31) 2.

Any safety injection header isolation valve (HSI-16, 26, 36) 3.

Any loop isolation valve (RC-M-11, 12, 21, 22, 31, 32).

b)

The reactor shall not be critical unless the following conditions are met:

/

1.

The saf ety injection tank isolation valves shall be disabled in the open position.

This shall require the following:

a.

The breakers shall be locked and tagged open.

b.

The disconnect switches for each valve power operator, shall be locked and tagged open.

2.

The loop isolation valves shall be disabled in the open position.

This shall require the following:

a.

The breakers shall be locked and tagged open.

b.

The disconnect switchd> for each valve power operator shall be locked and tagged open.

3.

The following ECCS check valve barriers shall have been determined to be intact in accordance with Technical Specification 4.6.A.2.f.

Barrier Loop 1 a

HSI-17 and HSI-61 b

LSI-12 Loop 2 a

HSI-27 and HSI-62 b

LSI-22 Loop 3 a

HSI-37 and HSI-63 b

LSI-32 Amendment No. 19, if, 56, 58 3.19-1

Eveapy M :

If an;. of the ECCS c'.eck valve barriers specified above do not r.eet the accaptance criteria of Technical Specification 4.6.A.2.f, then the reactor may be made or remain critical in accordance with the provisions of Specification 4.6.A.2.f.

Basis:- The position restrictions on the loop isolation valves, safety injection header isolation valves, and the safety injection tank isolation valves are necessary to assure that plant operation is

- restricted to conditions considered in the loss-of-coolant accident analysis.

The three check valves in the ECCS line to each loop provides assurance

. that a valve failure will not result in unrestricted flow of pressurized

, reactor coolant into lower pressure connecting piping outside the containment.

The valve integrity testing required by Technical Specification 4.6.A.2.f assures that the rate of flow under a valve failure condition will not exceed the pressure relief capacity of the line.

It further provides periodic assurance that the check valves are intact.

The two check valves closest to the loop are grouped together as a single check valve barrier for test purposes.

The first valve provides a thermal barrier preventing thermal distortion from af fecting the tightness of the second valve.

The third valve alone constitutes a check valve barrier.

In addition to the check valves the ECCS line to each loop contains a Motor Operated Valve (MOV) which is closed except for periodic monthly testing.

The MOV and reactor side piping is designed for full system precsure and also capable of preventing an overpressure condition of-connecting piping.

The exception permits time to schedule an orderly shutdown and maintenance of a def ective valve while providing assurance that two separate intact barriers always exist.

l l

l l

i l

Amendment No. 56, 58 3.19-2

/

e b

f, 4.6 P!3 IODIC TESTI"G SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS STEAM GENERATOR AUXILIARY FEED PUMPS MAIN STEAM EXCESS FLOW CHECK VALVES Applicability:

Applies to the saf ety injection system, the containment spray system, chemical injection system, the containment cooling system, the auxiliary feedwater system, and the main steam excess flow check valves.

~

Objective:

To verify that the subject systems will respond promptly and perform their intended functions, if required.

Specifieation:

A.

SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS 1.

The following tests will be performed monthly whenever plant conditions are as defined in Section 3.6.A of these Specifications.

a., Emergency Core Cooling System (ECCS) pumps:

Both operable high pressure saf ety injection (HPSI) pumps shall be tested by operating in the charging mode.

Both operable low pressure safety Lijection (LPSI) pumps and both operable containment spray (CS) pumps shall be tested by operating

'in the recirculation mode.

Acceptable performance shall be that pumps attain rated heads, operate for at least 15 minutes, and that the associated instrumenta-tion and controls function properly.

b.

ECCS Valves:

All automatically operated valves that.are required to operate to assure core flooding, or containment spray shall be exercised. The volume control tank (VCT) outlet to charging pump suction valves shall be exercised through part travel and all other valves shall be visually checked to verify proper operating position.

Exception:

LSI-M-11, 21 or 31 shall not be tested when the associated ECCS check valve barrier leakage f alls into Condition 2 or 3, as defined in Specification 4.6.A.2.f.

2.

The following tests will be performed at each refueling interval:

a.

ECCS Pumps:

One HPSI pump shall be flow tested at 1000 psig discharge head.

Amendment No.

45, 58 4.6-1

One LPSI pump and one CS pump shall be fle. tested at 100 poi discharge head.

During these tests flow fistribution thru the HPSI and LPSI flow orifices will be checked.

Acceptance performance shall be that the pumps and orifices attain flow values used in the safety analysis.

Alternate pumps will be tested at each refueling interval, so that all pumps will be tested within any five year period.

b.

ECCS Valves:

i All automatically operated valves and the notor operated fill header root valves shall be exercised through their full travel forth in in conjunction with the actuation signal testing set Table 4.1-2 of Technical Specifications.

c.

Safety Injection Tanks:

' Each safety injection tank will be flow tested by opening the tank isolation valve suf ficient to verify check valve operation.

d.

The correct position of each electrical and mechanical position stop for the following throttle valves shall be verified:

1) Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of maintenance on the valve when the HPSI system is required to be operable.

2) At least once per 4 months Valve Numbers HSI-M-ll HSl-M-12 HSI-M-21 HSI-M-22 HSI-M-31 HSI-M-32 e.

A flow balance test, as described in 4.6.A.2 above, shall be performed during shutdown to confirm the injection flow rates assumed in the Safety Analysis following completion of HPSI or LPSI system modifica-tions that alter system flow characteristics.

f.

ECCS Check Valves The check valve barriers defined in Technical Specification 3.19.b.3 shall be determined to be intact by leak terting.

Amendment No. If, 5, 55s,58 4.6-2

t i

Check scive barriers shall be determined to be intact tl. rough satisfaction of the following acceptance criteria.

Acceptance Criteria Condition 1 - Barrier (a) less than 15 gpm and barrier (b),

in the same loop, less than 5 gpm.

~

No additional action required.

Condition 2 - Barrier (a) less than 15 gpm and barrier (b),

in the same loop, greater than 5 spm.

The reactor may be made or remain critical for up to 30 days provided the affected ECCS line Motor Operated Valve remains closed.

Condition 3 - Barrier (a) greater than 15 gpm and barrier (b), in the same loop, less than 5 gpm.

I The reactor may be made or remain critical for up to 30 days provided the affected ECCS line Motor Operated Valve remains closed.

Condition 4 - Barrier (a) greater than 15 gpm und barrier (b),

]

in the same loop, greater than 5 gpm.

The reactor shall not be made or remain critical for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

Containment Spray Headers:

The containment spray flow.iozzles will be tested every five years.

The test will consist of pressurizing the headers with air and verifying <~iat the nozzles are free of obstruction.

j s-4.

Containment Isolation Valves:

Where practicable, each containment isolation valve shull be stroked to the position required to fulfill its safety vunction every three months.

Those valves that cannot be tested without possible adverse effects during plant operation shall be tested during each cold shut-down if not teeted during the previous three months.

B.

STEAM GENERATOR AUXILIARY FEED PUMPS Prior to plant startup following an extended cold shutdown, a flow test will be performed to verify the normal flow path from the demineralized water storage tank to the steam generators.

The flow test will be conducted with the AFW system valves in their normal. alignment.

)

Amendment-No. 45,55, 58 4.6-3 l

e

Monthly in:,pections shall be perforr.ad to verify that all manual va.!ves in the AFW system necessary to assure flow f rom the prin.ary water source to the steam generators are locked in tne proper position.

During normal plant operation, each auxiliary feed pump shall be tested at quarterly intervals to demonstrate operability of pumps, system valves and instrumentation.

C.

MAIN STEAM EXCESS FLOW CHECK VALVES The main steam excess flow check valves shall be tested once every 6 weeks for movement of the valve disc through ardistance of approxii.ately cne and ong-half inches.

These valves will be tested through full travel distance dur4ng each refueling ir.terval.

_ Basis:

The safety injection system and the containment spray system are principal

~

plant, safeguards sf,; ems that are normally operable during reactor operation.

Complete system tests cannot be performed when the reactor is operating because'of their inter-relation with operating systems.

The method of assuring operability of these systems is a combination of complete system tests performed during refueling shutdowns and monthly tests of active system components (pumps and valves) which can be performed during reactor operation.

The test interval is based on the judgment that more frequent testing would not significantly increase the reliability (i.e., the proba-4 bility that the component would oprate when required), yet more f requent tests would result in increased wear over a long period of time.

The monthly part travel exercising of the VCT outlet to charging pump suction valves, in lieu of the full travel exercise, is conducted to preclude an interruption of normal plant operat.'.ons.

Redundant valves have been used to assure proper lineup in the event of ECCS actuation.

Other ECCS valves whose operation is not req'uired to assure core flooding or containment spray shall be tested during each thfueling shutdown period in accordance with 2.b.

The three check valves in the ECCS line to each loop provides assurance that a valve f ailure will not result in unrestricted flow of pressurized reactor coolant into lower pressure connecting piping outside the containment.

The valve integrity testing required by Technical Specification 4.6.A.2.f assures that the rate of flow under a valve failure condition will not exceed the pressure relief capacity of the line.

It further provides periodic assurance that the check valves are intact.

The two check valves closest to the loop are grouped together as a single check valve barrier for test purposes.

The first valve provides a thermal barrier preventing thermal distortion from affecting the tightness of the second valve.

The third valve alone constitutes a check valve barrier.

Amendment No. 45, 55, 58 '

4.6-4

~... -.

. _. _ _.. ~., _.,, _ - __

The check valves are hard seated swing checks designe' to withstand the rigors of long term RHR operation without damage and the greatest assurance of integrity and dependability.

In addition to the check valves the ECCS line to each loop contains a Motor Operated Valve (MOV) which is closed except for periodic monthly testing.

The NOV and reactor side piping is designed for full system pressure and is also capable of preventing'an overpressure condition of connecting piping.

The leakage criteria provide an acceptable balance betvcen the need to maiptain a degree of tightness as a criterion of integrity on one hand and ALARA and power dependability considerations en the other giving due credit to the unique design feature of and protection provided by the four valves in series.

Verification that the spray piping and nozzles are open will be made initially by a suitably sensitive method, and at least every five years thereafter.

Since all piping material is all stainless steel, normally in a dty condition, and with no plugging mechanism available, the retest every five years is considered to be more than adequate.

Other systems that are important to the emergency cooling function are the SI tanks, the component cooling system and the service water system.

The Si tanks are a passive safety feature.

In accordance with Specifica-tion 4.1 (Table 4.1-2, Item 11), the water volume and pressure in the SI tanks are checked periodically.

The component cooling and service water systems operate when the reactor,is in operation and are continuously monitored for satisfactory performance.

The three month testing interval of the steam generator auxiliary feed pumps verifles their operability by recirculating water to the demineralized water tank.

Prior to plant startup following an ext ~oded cold shutdown, a flow test is performed on the Auxiliary Feedwater System to functionally verify the system alignment from the' demineralized water storage tank to the steam generators.

Monthly inspections are performed to verify that all manual valves in the Auxiliary Feedwater System from the primary water source to the steam generators are locked in the proper position.

Proper functioning of the steam turbine admission valve and starting of the auxiliary feed pump will demonstrate the operability of the steam i

driven pump. Verification of correct operation will be made both i em t

instrumentation with the main control room and direct visual obse s u

of the pumps.

l The main steam, excess flaw check valves serve to limit an excessive reactor coolant system cooldown rate and resultant reactivity insertion l

following a main steam break incident.

Their freedom to move will be verified period,ical19 Amendment No. 45, Jg, 58 4.6-5

._ - _