ML20062G527
| ML20062G527 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/10/1982 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20062G525 | List: |
| References | |
| NUDOCS 8208130062 | |
| Download: ML20062G527 (11) | |
Text
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I ATTACHENT 1 Proposed Unit 2 Technical Specifications 3/4.4.7.3 New Page 3/4 4 19 New Page 3/4 4-19a New Page 3/4 4-19b Revised Unit 1 Technical Specification 4.4.7.3.2 sf g Revised Page 3/4 4-19a Revised Unit 2 Technical Specifications 3/4.4.7.2 Revised Page 3/4 4-17 Revised Page 3/4 4-18 Revised Page 3/4 4-19 8208130062 820810 DRADOCK05000g
. PRESSURE ISOLATION VALVES-LIMITING CONDITION FOR OPERATION 3.4.7.3 Reactor Coolant System pressure isolation valves shall be operational.
APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:
- 1. All pressure isolation valves listed in Table 3.4-1 shall be functional as a pressure isolation device, except as specified in 2. Valve leakage shall not exceed the amounts indicated in Table 3.4-1.
12 . In the event that integrity of_ any pressure isolation valve specified in-Table 3.4-1 cannot be demonstrated, reactor operation may continue, provided that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at least two valves in each high pressure line having a non-functional valve are in, and remain in, the mode corresponding to the isolated condition.(a)
- 3. If ACTION 1 and 2 cannot be met, an orderly shutdown 'shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the reactor shall be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
(a) Motor operated. valves shall be placed in the closed position and power supplies de-energized.
FARLEY - UNIT 2 3/4 4-19
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENT
___.y- --
4.4.7.3.1 Each pressure isolation valve listed in Table 3.4-1 shall be
. demonstrated OPERABLE pursuant 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 limit of Table 3.4-1:(a)
- a. Every refueling outage during startup.
- b. Prior to returning the valve to service following maintenance, repair or replacement work on the valve affecting the seating capability of the val ve.
- c. Following valve actuation due to flow through the valve (s) identified in Table 3.4-1 by an asterisk.
- d. The provision of Specification 4.0.4 is not applicable for entry into MODE 3 or 4.
(a) To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and suppor'.ed by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
FARLEY - UNIT 2 3/4 4-19a l
TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Q2E11V016A Q2E21V062A Q2E11V001A Q2E21V0628 Q2E11V016B Q2E21V062C Q2E11V001B Q2E21V066A Q2E21V066B Q2E11V051A Q2E21V066C Q2E11V051B Q2E21V077C Q2E11V051C Q2E21V078A-Q2E11V021A Q2E21V078B Q2E11V021B- Q2E21V078C Q2E11V021C Q2E21V079A Q2E11V042A Q2E21V079B Q2E11V042B. Q2E21V079C Q2E21V077A* -Q2E21V032A*
Q2E21V077B* Q2E21V032B*
Q2E21V076A* Q2E21V032C*
Q2E21V076B* Q2E21V037A*
Q2E21V037B*
Q2E21V037C*
ALLOWABLE LEAKAGE RATES:
- 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.
However, for initial tests, or tests following valve repair or replacement, leakage rates less than -or equal to 5.0 gpm are considered -
acceptable.
- 2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of. 5.0 gpm by 50% or greater.
l 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm i are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
- 4. Leakage rates greater than 5.0 gpm are considered unacceptable.
i Minimum differential test pressure shall not be less than 150 psid.
l l
l i
FARLEY - UNIT 2 3/4 4-19b L
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENT 4.4.7.3.1 Each pressure isolation valve listed in Table 4.4-2a shall be demonstrated OPERA 8LE pursuant to Specification 4.0.5, except that in lieu of any leakage testing required by Specification 4.0.5,eachvalveshouldbedemonstratedOPERABLQyverifying leakage to be within the limit of Table 4.4-2a:
- a. Every refueling outage during startup. ;
- b. Prior to returning the valve to service following maintenance, repair or replacement work on the valve affecting the seating capability of the valve.
- c. Following the plant being placed in a cold shutdown condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> duration if leakage testing has not been accomplished within 12 months.
- d. The provision of Specification 4.0.4 is not applicable for entry into MODE 3 or 4.
4.4.7.3.2I WheTiever' integrity,[f a pressur[ isolation Salve / listed in []-
Tablar4.4-26cannotbe,demonstratedIthe,integrityof'the ,
j/ .
remaining'valverin each higVpress'ure 1)ne hayfng a/ leaking/valve .
spaill be' determined,and recorded / daily, In additic'n, th6
/ [positJ6n of ,the pipi,ng shall be re' corded' daily'. other
/
j closkd
/
h valve
/' r/e/1 loc'
~
pgfC C l
l
(*)To satisfy ALARA requirements, leakage may be measured indirectly (as .
from the performance of pressure indicators) if accomplished in accord-ance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
'l FARLEY-UNIT 1 3/4 4-19a
' REACTOR COOLANT SYSTEM . . _
LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to:
.a. No PRESSURE BOUNDARY LEAXAGE,
- 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 LEAXAGE from the Reactor Coolant System, and
- e. 31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235
- 20 psig.
. f.' 'l G3iM Je'akage f[om any Re' acto'r Co'olan# tSystem/ressureIsola(icn/
P
/ V tS
'y 'y"<e RtryTab)6
&alve/speciftee L L / 3,4-1 it/a/ Reactor / J Coolin/ ystem /jressureJ
/
s06ES 1, 3 and 4 L 7g{ c.,
APPLICABILITY: . .
ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least NOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,
- 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Ii c 'Wittian Re'actofCoolant/SystemiPrelsure/Isolafion/alve'laakige V gr'eate'r th'at the ab'ove ,11mit,/isojate the high pplessure' portion of the/sffected/syste's from the' low, pressure poi tion with'in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,by '
use of,at least t'wo c1'osed,inanual or/ d eactpat 6 auto'matic/ valves, oy'be/in ati leas't HOT STANDBY withir} the pextj5 hour's and'in CDCD SHUTDOWNp'ithidthe/fo11g4fng/30hoors.2 / / / / /
0 SURVEILLANCE REOUIREMENTS ._
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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. - '
- b. Monitoring the containment air cooler condensate level system or containment atmosphere gaseous radioactivity monitor at least once er 12 houn. '
FAney-Uu tr2.
2 /( Ll- M _
(*
\
SURVEILLANCE RE0UIREMENTS (Contin'ued) '
- c. Measurement of the CONTROLLED LEAKAGE frein 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. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- e. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b(7.I2 5ac deac.t r Coolant $ yste'm P e re',I'solatio'n Valve /speciffbd in i
J ble' 3.,4-1 s_ hall b4 demdnstra'ted 0PERAB E pur,s'uant to'Speci f,ication
'except f that/n li,e'u of/any j l'eakag,e test'ing required,by Speci.fication/4.0
- 4. 0. 5,<
each v'a lve/shouJd be/ demonstrated 0,P,ERABLE,by verifying leakage to'be within its Aim t( - '
,/ / ,/ / / / .
/* ,
~
- a. E~very'refM1ing/ outage durintj startdp. ,
,/b./ l / / /
Priorf to returnin,g/
/
/ / / -
the val've to , service fo'11owing maintenance /
lepair or/replagement wdrk on the valve /affecting the seating ~
~
cap'abi }ity of the va,1v'e. j/
/
j/ /
/ / / / f' , ,
/Following yalve aciuationsdue to autcmatic or nianual action or /
/c./ fidw throiJgh the' valve for valves / identified in Table,/3.4-14y an/~
f / /
/. a'steris k{ / / j/ / /
I ,tlThe,p / !
ns of Specification 4.0f4
/rovisio/
/ are not applicable /
/
for /
- d.
e,ntry 1 t$ MODE 3'or4 /
j / _/ /
9Cf6 m
l ,
1 rwnre w u -sua
- \
[
\, ,
,\ TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLITION VALVES
. S.
\ Q2E11V016A
\ Q2E11V001A i Q2E11V016A s Q2E11V0018
\
\ Q2E11V051A sQ2E11V051B, Q2E11V051C Q2E11V021A -
Q2E11V0218 / I Q2E11V021C tx !
Q2E11V042A Q2E11V0428
/'M46% '}
Q2E21V077A*
Q2E21V077B*
Q2E21V076A*
Q2E21V076B*
M P[0 *
/
j y '_ ^ ~ ~
,/
Q2E21V062A Q2E21V062B Q2E21V062C, Q2E21V066A\
Q2E21V066Bi Q2E21V066C '
Q2E21V077C Q2E21V078A iQ2E21V0788 \
/ Q2E21V078C \
f Q2E21V079A Q2E21V0798
\g
/ Q2E21V079C 1
\
/ Q2E21V032A* \
/
Q2E21V0328*
Q2E21V032C*
\g
/ Q2E21V037A* \
/ Q2E21V037B*
\
,/ Q2E21V037C*
/
1 1
1 FARLEY-UNIT 2 .y4 4-19 4
e
, - . . ~ -- ,
1
-o- ATTACHMENT 2 I
Safety Evaluation for Proposed Changes to the FNP-1 and-2 Technical Specifications Sections 3/4.4.7.2 and 3/4.4.7.3 f l
I
- 1.
Background:
J. M. Farley Nuclear Plant - Units 1 and 2_are required to test reactor coolant system pressure isolation valves per Technical Specifications 4.4.7.3 and 4.4.7.2.2, respectively. - The testing requirements and acceptance criteria for these valves contained in the Unit 2 Technical Specifications have proven to be too re-strictive, while the Unit 1 Technical Specification testing re-quirements and acceptance criteria for these same valves have been found to be adequate in determining valve operability and at the same time have not resulted in excessive personnel radiation exposure. The Unit 2 test acceptance criteria of 1 gpm versus the Unit 1 criteria of 1 to 5 gpm has proven to be excessively re-strictive without concomitant safety benefit. Therefore, Alabama Power Company respectfully requests that the Unit 1 Technical Specification testing requirements and acceptance criteria be incorporated into the Unit 2 Technical Specifications.
A revision to the Unit 1 Technical Specifications is also proposed which deletes paragraph 4.4.7.3.2.. This paragraph is in conflict with the existing ACTION statement and is unnecessary since the ACTION statement adequately addresses valve integrity.
II.
References:
(1) FNP Unit 1 Technical Specifications 3/4.4.7.3
_(2) FNP Unit 2 Technical Specifications 3/4.4.7.2 III. Bases:
It is proposed to add Specifications 3/4.4.7.3 (which are equiv-alent to the Unit I requirements) and delete Specifications 3.4.7.2.f, ACTION statement C, and 4.4.7.2.2 from the Unit 2 Technical Specifications. The Unit i version contains necessary clarifications in testing requirements and represents an improvement over those in the Unit 2 Technical Specifications.
In summary, the current Unit i leakage criteria is that leakage rates greater than 1 gpm but less than or equal to 5 gpm are con-sidered acceptable if the latest measured rate _has not exceeded the rate determined by the previous test by an amount that reduces the margin between leakage rate and the maximum permissible rate of 5 gpm by 50% or greater. This Unit 1 criteria is herein referred to as the 1 to 5 gpm limit.
- ' ATTACHMENT 2 Page 2 l
The Unit 2 test acceptance criteria of 1 gpm versus the :1 to 5 gpm limit on Unit 1 for the . reactor coolant. system pressure isolation valve leak test has proven to be excessively restrictive.with no corresponding. safety benefit. Valves that did not _ pass either the -
1 gpm or the 1 to 5.gpm acceptance criteria.have been found to-contain the ~same type irregularities causing the valves not to seat completely. No evidence of impending valve failure has been found. In addition, six valves repaired out of a total'of sixteen tested to the 1 gpm criteria resulted in plant personnel ~ receiving.
approximately 10 times the radiation exposure of that associated with the 1 to 5 gpm limit, i.e., 25.0 rem versus 2.5 rem.
Therefore, the 1 to 5_ gpm limit will provide adequate assurance of valve integrity, and at the same time will'not compromise the health and safety of plant personnel.
The purpose of the valve surveillance testing'is to reduce the pro-bability of a 1.0CA resulting from valve failure between the reactor coolant system and interconnecting low pressure systems (i.e.,
Event V in WASli-1400). The probability of such an occurrence is extremely low as stated in WASH-1400, on the order of 10-6/ year.
Also as stated in WASH-1400, yearly testing of the valves will re-duce the failure probability to approximately 10-7/ year, an order of magnitude. decrease. Therefore, the proper stroking of the valves during test perform =d on a yearly basis is the primary indication of acceptable valve. integrity, while the leakage criteria is only a secondary indication. The 1 to 5 gpm leakage criteria is a more reasonable test which does not represent an-increase in probability of valve failure over that represented by the 1 gpm leakage criteria. Thus there is not a significant affect-on the safe operation of Unit 2 as a result of this proposed change in valve leakage test criteria.
In addition, it is proposed to delete Paragraph 4.4.7.3.2 from the Unit 1~ version of the Technical Specifications. This.is because the existing-ACTION statement of Section 3/4.4.7.3 adequately ad-dresses valve integrity and is in conflict with Paragraph 4.4.7.3.2. Since the- ACTION statement requires valve and/or system integrity to be maintained / isolated or the plant is to initiate an orderly shutdown within one hour, the l requirement of Paragraph '
4.4.7.3.2 to daily monitor leakage of downstream valves in the system is inappropriate. Furthermore, such valve leakage moni -
toring cannot'be performed while the reactor is in operation or while the- reactor coolant' system is pressurized due to. personnel
. safety considerations. Therefore, the deletion of Paragraph 4.4.7.3.2 does not af fect the safe operation of Unit 1.
ATTACHMENT 2 Page 3
'IV..
Conclusion:
The proposed changes to the Units 1 and 2 Technical Specifications represent improvements in the testing required for the reactor .)
coolant ' system pressure isolation valves and do not involve an un-reviewed. safety question as defined by 10CFR50.59.
. . . - - . _ . - . . - . . .