Text

Proposed Tech Specs,Standardizing Both Unit Formats for Testing Reactor Coolant Pressure Isolation Valves & Defining Allowable Leakage Criteria Per NRC Guidance
	ML20083M919
Person / Time
Site:	Farley
Issue date:	04/10/1984
From:	; ALABAMA POWER CO.
To:	;
Shared Package
ML20083M917	List: ML20083M914; ML20083M919;
References
	NUDOCS 8404180391
	Download: ML20083M919 (9)
	v • d • e

. .

Attachment Proposed Changes to T.S. Pages Unit 1 4

Remo'.e Page V Replace Page V

, Remove Page 3/4.4-17 Replace Page 3/4.4-17 Remove Page 3/4.4-18 Replace Page 3/4.4-18 4

Remove Page 3/4.4-19 Replace Page 3/4.4-19 Delete Page 3/4.4-19a Delete Page 3/4.4-19b Remove Page B 3/4.4-4 Replace Page B 3/4.4-4

' Unit 2 Remove Page 3/4.4-17 Replace Page 3/4.4-17 Delete Page 3/4.4-17a r Remove Page 3/4.4-18 Replace Page 3/4.4-10 Remove Page 3/4.4-19 Replace Page 3/4.4-19 i

3 l

h 1

8404180391 840410 PDR ADOCK 05000348 P PDR . , _

.l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE JFQUIREMENTS SECTION PAGE R

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation . . . . . . . . . . . 3/4 4-1 Hot Standby . . . . . . . . . . . . . . . . . . . 3/4 4-2 Hot Shutdown . . . . . . . . . . . . . . . . . . 3/4 4-3 Col d Shu tdown . . . . . . . . . . . . . . . . . . 3/4 4-4a

, 3/4.4.2 SAFETY VALVES - SHUTDOWN . . . . . . . . . . . . 3/4 4-5 3/4.4.3 SAFETY VALVES - OPERATING . . . . . . . . . . . . 3/4 4-6 3/4.4.4 PRESSURIZER . . . . . . . . . . . . . . . . . . . 3/4 4-7 3/4.4.5 RELIEF VALVES . . . . . . . . . . . . . . . . . . 3/4 4-8 3/4.4.6 STEAM GENERATORS . . . . . . . . . . . . . . . . 3/4 4-9 3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems . . . . . . . - . . . . . 3/4 4-16 i Operational Leakage . . . . . . . . . . . . . . . 3/4 4-17 l

3/4.4.8 CH EM I STRY . . . . . . . . . . . . . . . . . . . . 3 /4 4-2 0 3/4.4.9 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . 3/4 4-23 3/4.4.10 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System . . . . . . . . . . . . . 3/4 4-27 Pressuri zer . . . . . . . . . . . . . . . . . . . 3/4 4-31 Overpressure Protection Systems . . . . . . . . . 3/4 4-32

3/4.4.11 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components . . . . . . 3/4 4-34 FARLEY - UNIT 1 V AMENDMENT NO.

'~"

' REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235120 psig.

f. The maximum allowable leakage of any Reactor Coolant System Pressure Isolation Valve shall be as specified in Table 3.4-1 at a pressure of 2235 1 20 psig.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit specified in Table 3.4-1, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves,-

or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . l SURVEILLANCE REQUIREMENTS 4.4.7.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by;

a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the containment air cooler condensate level system or containment atmosphere gaseous radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

FARLEY - UNIT 1 3/4 4-17 AMENDMENT NO.

L

4 REACTOR COOLANT SYSTEM I SURVEILLANCE REQUIREMENTS (Continued) l c. Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals at least once per 31 days when the Reactor Coolant System pressure is 2235 + 20 psig with the modulating valve fully open. The provisions of !

Specification 4.0.4 are not applicable for entry into l j' MODE 3 or 4. ;

i L d. Performance of a REACTOR COOLANT SYSTEM water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

l e. Monitoring the reactor head flange leakoff system at l 1 east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i 4.4.7.2.2 Each Reactor Coolant System Pressure Isolation Valve specifled in Table 3.4-1 shall be demonstrated OPERABLE pitrsuant to Specification 4.0.5 except that in lieu of any leakage testing required by Specification 4.0.5, each valve should be demonstrated OPERABLE by verifying leakage to be '

within the allowable leakage criteria of 0.5 gpm per inch of nominal valve size with an upper limit of 5 gpm, and the 1 measured leak rate for any given test cannot reduce the difference between the results of the previous test and 5 gpm by more than 50%:

l a. Every refueling outage during startup.

j b. Prior to returning the valve to service following i maintenance,. repair or replacement work on the valve affecting the seating capability of the valve.

.c. Following valve activation due to automatic or manual action or flow through the valve for valves identified

in Table 3.4-1 by an asterisk.

d. The provisions of Specification 4.0.4 are not applicable for entry into M00E:3 or 4.

F # To satisfy ALARA requirements, leakage may be measured indirectly

! (as from performance of pressure indicators) if accomplished in

accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve

[ compliance with the leakage criteria. ,

i b*

. FARLEY - UNIT 1 3/4 4-18 AMENOMENT NO.

l

-- = - - ..

w , y -

[ ,, .,-

} .s.

s

, y

\

-; x 5

TABLE 3.4-1

(

REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES

! t s ,

VALVE LMUMBER -

J-s.

DESCRIPTION MAXIMUM ALLOWABLE LEAKAGE **

~h1E!1V001A, l'2* GA7 E " 5.000 GPM 12" GATE 5.000 GPM

,_ Q1E11V0018 '

Q1E11V016A' 12" GATE 5.000 GPM i~

Q1E11V016B

\

k; - 12" GATE 4 5.000 GPM 01E11V021A \ 5" CHECK '\ 3.000 GPM Q1E11V0218 '6" CHECK 3.000 GPM {

Q1E11V021C , 6" CHECK 3.000 GPM l s

Q1E21V032A 12" CHECK 5.000 GPM

Q1E21V032B s

\ o 12'; CHECK 5.000 GPM 12" CHECK s 5.000 GPM

Q1E21V032C 2

Q1E21V037A '

li" CHECK 5.000 GPM

Q1E21V0376 ' , ' 12", CHECK 5.000 GPM

Q1E21V037C _ -12." CHECK 5.000 GPM Q1E11V042A 10' CHECK 5.000 GPM l Q1E11V042B ' l'1" CHECK. 5.000 GPM i

Q1E21V076A 6" CHECK 3.000 GPM

Q1E21V0768 , 6" CHECK 3.000 GPM

01E21V077A 6" CHECK 3.000 GPM

QlE21V0778% i s 6"' CHECK ,

3.000 GPM Q1E21V077C 6" CHECK 3.000 GPM

~.. ,

% % s g- ,

\ , s .

.\ .

15

'p I%

w

,_ h i 3

,\. ,

s, s 's' L

,c. ,f4 ~ s-

Innii;ates the r'equirementa,of Section'4.4 7 2.2

Item (ff are applicably C

\ ** The measured leak rate for any given test cannot. reduce the difference between 'the results of' the '

previous test and 5 gpm by more than 50%. -

7

'h.

S

~

s -FARLEY - UNIT 1 4 3/4.4-19' AMEN 0 MENT NO.

z c ..4 5 6 $,

w:

l I

REACTOR COOLANT SYSTEM BASES 3/4.4.7 REACTOR COOLANT-SYSTEM LEAKAGE 3/4.4.7.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These. detection systems,are consistent with the reconstendations of Regulatory Guide 1.45, " Reactor Coolant Pressure

. Boundary Leakage Detection Systems," May 1973.

3/4.4.7.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be _ reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE-limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. - '

The, CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to.the reactor coolant pump seals exceeds 31 GPM with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity, thereby reducing the

_ probability of gross valve failure and consequent intersystem LOCA.

Leakage from.the RCS Pressure Isolation valves is IDENTIFIED LEAKAGE and will be considered a portion of the allowed limit.

The total steam generator tube _ leakage limit of 1 GPM .for 'all

- steam generators ensures'that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break.

The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity _is maintained in the event ~

of a main steam line. rupture or'under LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since.

it may be indicative of an impending gross failure of:the pressure boundary. 'Therefore, the presence of any PRESSURE' BOUNDARY _ LEAKAGE requires the unit to _ be promptly placed'in. COLD SHUTDOWN.

FARLEY-UNIT'1 :B 3/4 4.4 - AMENDENT NO. .

,' e N

y

ff ' REACTOR COOLANT SYSTEM

, s - OPERATIONAL LEAKAGE s' LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE, L b. 1 GPM UNIDENTIFIED LEAKAGE, i

c. 1 GPM total primary-to-secondary leakage through all steam generators and 50.1 gallons per day through any one steam generator,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235120 psig.

f. The maximum allowable leakage of any Reactor Coolant System Pressure Isolation Valve shall be as specified in Table 3.4-1 at a pressure of 2235 1 20 psig.

APPLICABILITY: MODES'1, 2, 3 and 4 ACTION:

a. With any PRESSURE B0UNDARY LEAKAGE, be in at least _ HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD. SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

-c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit specified'in Table 3.4-1, isolate the high pressure portion of the affected system -

from the low pressure portion within _4 hours by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD. SHUTDOWN within the following _30 hours.

SURVEILLANCE REQUIREMENTS

4.4.7.2.1' Reactor Coolant System leakages _ shall be demonstrated to be within each of the above limits by;-

a. Monitoring the' containment atmosphere particulate

. . radioactivity ~ monitor at-least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the containment. air cooler condensate:1evel l system or containment atmosphere gaseous -radioactivity

- monitor at,least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . '

1 FARLEY:-; UNIT ~2 ~3/4 4-17l : AMENDMENT NO.-

w -

n

I REACTOR COOLANT-SYSTEM f

. SURVEILLANCE REQUIREMENTS (Continued)

c. Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals at least once per 31 days when the Reactor Coolant System pressure is 2235 + 20 psig with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

d. Perfonnance of a REACTOR COOLANT SYSTEM water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . <

e. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.7.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE i pursuant to Specification 4.0.5 except that in lieu of any 1-leakage testing required by Specification 4.0.5,- each valve should be demonstrated OPERABLE by verifying leakage to be l within the allowable leakage criteria of 0.5 gpm per inch of nominal valve size with an upper limit of 5 gpm, and the

, measured leak rate for any given test cannot reduce the difference between the results of the previous test and 5 gpm by more than 50%:

L

a. Every refueling outage during startup.

j' b. Prior to returning the valve to service following'-

maintenance, repair or replacement work on the valve

affecting the seating capability of the valve.

t

c. .Following valve activation due to automatic or manual action or flow through the valve for valves identified ;

'in Table 3.4-1 by an asterisk..

4

d. The provisions of Specification _4.0.4 are not applicable

~ for. entry into MODE 3 or 4.

l 4

To satisfy ALARA requirements, leakage may be' measured indirectly ,

-(as from performance of pressure indicators) if accomplished in, accordance with approved procedures and supported by computations i= showing that the method is capable of demonstrating valve

compliance with the leakage criteria.

l l

f .FARLEY.- UNIT 2-3/4:4-18 AMENOMENT NO.

-. , , , .. , , . . . . . .2 . :.s .. .- . .

_ J;

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE MAXIMUM l NUMBER DESCRIPTION ALLOWABLE LEAKAGE **

Q2E11VU01A 12" GATE 5.000 GPM Q2E11V001B 12" GATE 5.000 GPM Q2E11V016A 12" GATE 5.000 GPM Q2E11V016B 12" GATE 5.000 GPM Q2E11V021A 6" CHECK 3.000 GPM Q2E11V021B 6" CHECK 3.000 GPM Q2E11V021C 6" CHECK 3.000 GPM

Q2E21V032A 12" CHECK 5.000 GPM

Q2E21V032B 12" CHECK 5.000 GPM

Q2E21V032C 12" CHECK 5.000 GPM

Q2E21V037A 12" CHECK 5.000 GPM

Q2E21V037B 12" CHECK 5.000 GPM

Q2E21V037C 12" CHECK 5.000 GPM Q2E11V042A 10" CHECK 5.000 GPM '

Q2E11V042B 10" CHECK 5.000 GPM

Q2E21V076A 6" CHECK 3.000 GPM

Q2E21V076B 6" CHECK 3.000 GPM

Q2E21V077A 6" CHECK 3.000 GPM

Q2E21V077B 6" CHECK 3.000 GPM Q2E21V077C 6" CHECK 3.000 GPM

Indicates the requirements of Section 4.4.7.2.2 Item (c) are applicable.

- The measured leak rate for any given test cannot- reduce the difference between the results of the previous test and 5 gpm by more

-than 50%.

FARLEYT- UNIT 2 3/4 4-19 AMENDMENT NO.

ML20083M919

Text

Navigation menu

Search