ML20086C103
| ML20086C103 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/29/1995 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20086C086 | List: |
| References | |
| NUDOCS 9507060277 | |
| Download: ML20086C103 (50) | |
Text
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January 3,1995 CONTAllgfENT SYSTEMS
(
ELECTRIC HYDROGEN RECOMBINERS LIMITING COM ITION FOR OPERATION 3.6.4.2 Two' independent Hydrogen Recombiner Systems shall be OPERABLE.
APPLICABILITY: N0 DES I and 2.
&GIlQll:
With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days 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 />.
SURVEILLANCE REQUIRENENTS f once each@fi~eping intervaj)by:
pl 4.6.4.2 Each dros en nombiner System shall be demonstrated OPERABLE at least
- a. (Verifying during a Hydrogen Recombiner System functional test that the l
minimum reaction chamber gas temperature increases to greater than or G. V \\ equal to 700*F within 90 minutes and is maintained for at least 2 clelh J (hoursandthatthepurgebloweroperatesfor15 minutes.
I b.
Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits, c.
Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e., loose wiring or structural connections, deposits of foreign materials, etc.),
d.
Verifying the integrity of all heater electrical circuits by l
performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than 10,000 ohns, and e.
Verifying during a recombiner system functional test using containment atmospheric air at an acceptable flow rate as determined in Section 4.6.4.2.f that the gas temperature increases to greater than or equal to 1100*F within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and is maintained for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
f.
Verifying during a recombiner system functional test using containment l
atmospheric air that the blower would be capable of delivering at leest 41.52 sc.fc at containment conditions of 12.47 psia and 130*F.
I NILLSTONE - W IT 3 3/46-17 Amendment No. E, S.100 em 9507060277 950629 PDR ADOCK 05000423 P
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eJo chage.
j Pb8 twPo oNW January 3,1995 h
AllI!LIARY FEEDWATER SYSTEM 1
LIMITING CONDITION FOR OPERATION i
3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
a.
Two motor-driven auxiliary feedwater pumps, each' capable of being powered from separate emergency busses, and b.
One steam turbine-driven auxiliary feedwater pump capable of being i
powered from an OPERABLE steam supply system.
i APPLICABILITY: MODES 1, 2, and 3.
ACI1Qti:
a.
With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be 1
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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With two auxiliary feedwater pumps inoperable, be in at least HOT i
STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following a
1 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
W c.
With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. Entry into an OPERATIONAL MODE pursuant to Specification 3.0.4 is not permitted with three auxiliary feedwater pumps inoperable.
SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
a.
At least once per 31 days by:
1)
Verifying that each non-automatic valve in the flow path that I
is not locked, sealed, or otherwise secured in position is in its correct position; and 2)
Verifying that each auxiliary feedwater control and isolation l
valve in the flow path is in the fully open position when above 105 RATED THERMAL POWER.
s 9
NILLSTONE - IBIIT 3 3/47-4 ANDONENT No. E,100 asse
January 3.1995 PLANT SYSTEMS SURVEILLANCEREQUIRENENTS(Continued) b.
At least once per 92 days on a STAGGERED TEST BASIS by:
l 1)
Verifying that on recirculation flow each motor-driven pump l
develops a differential pressure of greater than or equal to 1460 psid when tested pursuant to specification 4.0.5; 2)
Verifying that on recirculation flow the steam turbine-driven l
pump develops a differential pressure of greater than or equal to 1640 psid when the secondary steam supply pressure is greater than 800 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
g a 4 na w l c.
AtleastonceLa-_ M rthe durina shu+dawa) by verifying that each l
auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal.
For the steam turbine-driven auxiliary feedwater pump, the provisions of i
Specification 4.0.4 are not applicable for entry into MODE 3.
4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying flow to each steam generator.
1 I
NILLSTONE - WIT 3 3/47-5 Amendment No. pp.100
% s.1
349-%
PLANT SYSTEMS gg 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent reactor plant component cooling water safety j
loops shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3, and 4.
1 ACTION:
With only one reactor plant component cooling water safety-loop OPERABLE, re-store at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.
SURVEILLANCE REQUIREMENTS
)
4.7.3 At least two reactor plant component cooling water safety loops shall be demonstrated OPERABLE:
At 1 east once per 31 days by verifying that each valve (manual, a.
power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and y rpd A'-
j e
b.
At least once lper 18 months during shutdowrf, by verifying that:
1)
Each automatic valve actuates to its. correct position on its
, associated Engineered Safety Feature actuation signal, and 2)
Each Component Cooling Water System pump starts automatically on an SIS test signal.
i MILLSTONE - UNIT 3 3/4 7-11 AS47
3-39 99 PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM i
LIMITING CONDITION FOR OPERATION r
z 3.7.4 At least two independent service water loops shall be OPERABLE.
iAPPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDB 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 />.
SURVEILLANCE REQUIREMENTS 4.7.4 At least two service water loops shall be demonstrated OPERABLE:
At least once per 31 days by verifying that each valve (manual, a.
power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; an S M T 6 g V A L.
I eam ~REFbE LDJ(r - - - - ~
b.
At least onc per 19 : nth: t-'-- H te ",y by verifying that:
1)
Each automatic valve servicing safety related eouipment actuates to its correct position on its associated Engineered Safety Feature actuation signal, and 2)
Each Service Water System pump starts automatically on an SIS test signal.
f i
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MILLSTONE - UNIT 3 3/4 7-12 6,5Gf g % pggg7 bjo.
I
January 3,1995 PLANT SYSTDis SURVEILLANCE REQUIRDIENTS (Continued) type that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and' detemined OPERABLE per Specification 4.7.10.f. All snubbers found connected to an inoperable comeon hydraulic fluid reservoir shall be counted as unacceptable for detemining the next inspection interval. A review and evaluation shall be performed and documented to justify continued operation with an unacceptable snubber.
If continued operation cannot be justified, the snubber shall be I
declared inoperable and the ACTION requirements shall be met.
i l
d.
Transient twent Insnection l
An inspection shall be performed of all snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients as determined from a review of operational data and a visual inspection of the systems within 6 months following such an event.
In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall i
be verified using at least one of the following: induced snubber movem
- 1) manually piston setting; or (3) stroking the mechanical snubber through its full range of travel.
c, e.
Functional Tests v* h
,,,,,,,r-During thegualino shutdown and Eleast oncedier 18 month [
l thereaft
"" 'Da representative sample of snubbers of ea:h type
'all be tested using one of the followirn sample plans.
The sample plan for each type shall be selected pr1 or to the test period and cannot be changed during the test period.
The NRC Regional Administrator shall be notified in writing of the sample i
plan selected for each snubber type prior to the test period or the sample plan used in the prior test period shall be implemented:
j 1)
At least 105 of the total of each type of snubber shall be functionally tested either in-place or in a bench test.
For i
each snubber of a type that does not meet the functional test i
acceptance criteria of Specification 4.7.10f., an additional 5f, of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested; or i
i MILLSTONE - UNIT 3 3/4 7-13 Amendment g, 77,100 P03%L l
1 m
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.e-1 Docket No. 50-423 B1522jli.
i o
i
.t Millstone Nuclear Power Station, Unit No. 3 i
Proposed Revision to. Technical Specifications l
24-Month Fuel Cycle i
Plant Systems i
Retyped Pages t
P j
1 1
June 1995 i
i
CONTAINNENT SYSTEMS ELECTRIC NYDR0 GEN REC 0NBINERS LIMITING CONDITION FOR OPERATION 3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERABLE.
APPLICABILITY: N0 DES I and 2.
ACTION:
With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days or 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 />.
SURVEILLANCE REQUIRENENTS 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE at least once each REFUELING INTERVAL by:
l a.
Deleted l
b.
Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits, c.
Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e.,
loose wiring or structural connections, deposits of foreign materials, etc.),
d.
Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than 10,000 ohms, and e.
Verifying during a recombiner system functional test using containment atmospheric air at an acceptable flow rate as determined in Section 4.6.4.2.f that the gas temperature increases to greater than or equal to 1100*F within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and is maintained for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
f.
Verifying during a recombiner system functional test using containment atmospheric air that the blower would be capable of delivering at least 41.52 scfm at containment conditions of 12.47 psia and 130*F.
NILLSTONE - UNIT 3 3/4 6-17 Amendment No. (7, JJ. JPP, om
PLANT SYSTEMS SURVEILLANCEREQUIREMENTS(Continued)-
b.
At.least once per.92 days on a STAGGERED TEST BASIS by:
1)
Verifying that on; recirculation flor each motor-driven. pump develops a differential pressure of ' greater than or equal to 1460 psid when tested pursuant to Specification 4.0.5;.
2)
Verifying that.on recirculation flow the steam turbine-driven pump develops a differential pressure of greater than or equal to 1640 psid when the secondary steam supply pressure is greater than 800 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
auxiliary feedwater pump starts as designed automatically upon l At least once each REFUELING INTERVAL by verifying that. each c.
receipt of an Auxiliary Feedwater Actuation test. signal.
For the steam turbine-driven auxiliary feedwater pump, the provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be-demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying flow to each steam generator.
l..
NILLSTONE - UNIT 3 3/4 7-5 Amendment No. pp, Jpp, 0368
.-~
PLANT SYSTEMS 3/4.7.3 REACTOR PUUti COMP 0NENT C00 LIM WATER SYSTEN l
LINITING COMITION FOR OPERATION i
p j
3.7.3 At least two independent reactor plant component cooling water safety.
loops shall be OPERABLE.
i-APPLICABILITY: NODES 1, 2,~3, and 4.
ACTION:
l-i With only one reactor plant component cooling water safety loop OPERABLE, re-store at least two loops to OPERABLE. status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 8Y 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 />..
i SURVEILLANCE REQUIRENENTS I
i 4.7.3 At least two reactor plant component cooling water safety loops shall be demonstrated OPERABLE:
a.
At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its I
correct position; and j
b.
At least once each _ REFUELING INTERVAL by verifying that:
l I)
Each automatic valve actuates to its correct position on its associated Engineered Safety Feature ' actuation signal, and l
2)
Each Component Cooling Water System pump starts automatically I
on an SIS test signal.
)
NILLSTONE - UNIT 3 3/4 7-11 Amendment No.
caos
PLANT SYSTEMS 3/4.7.4 SERVICE MATER SYSTEN LIMITIM COMITION FOR OPERATION 3.7.4 At.least two independent service water loops shall be OPERABLE.
APPLICABILITY: N0 DES 1, 2, 3, and 4.
ACTION:
With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.
$URVEILLANCE REQUIREMENTS 4
4.7.4 At least two service water loops shall be demonstrated OPERABLE:
a.
At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and b.
At least once each REFUELING INTERVAL by verifying that:
1)
Each automatic valve servicing safety-related equipment actuates to its correct position on its associated Engineered Safety Feature actuation signal, and 2)
Each Service Water System pump starts automatically on an SIS test signal.
MILLSTONE - UNIT 3 3/4 7-12 Amendment No.
ONO
U~
PLANT SYSTEMS SURVEILLANCE REQUIRENENTS (Continued) type that may be generically susceptible; and (2) the affected l
snubber is functionally tested in the as-found condition and determined OPERABLE per Specification 4.7.10.f. All snubbers found connected to an inoperable common hydraulic fluid reservoir shall be counted as unacceptable for determining the next. inspection interval. A review and evaluation shall be performed and documented to justify continued operation with an unacceptable snubber.
If continued operation cannot be justified, the snubber shall be declared inoperable and the ACTION requirements shall be met.
d.
Transient Event Insnection-An inspection shall be erformed of all. snubbers attached to sections of systems that ave experienced unexpected, potentially damaging transients as determined from a review of operational data and a visual inspection of the systems within 6 months.following such an event.
In addition to satisfying the visual. inspection acceptance criteria,' freedom-of-motion of mechanical snubbers shall be verified using at least one of the following:
(1) manually-induced snubber movement; or (2) evaluation of in-place snubber piston setting; or_ (3) stroking the mechanical snubber through its full range of travel.
e.
Functional Tests During the first refueling shutdown and at least once each REFUELING INTERVAL thereafter, a representative sample of snubbers of each type shall be tested using one of the following sample plans. The sample plan for each type shall be selected prior to the test period and cannot be changed during the test period.
The NRC Regional.
Administrator shall be notified in writing of the sample plan selected for each. snubber type prior to the test. period or the sample plan used in.the prior test period shall be implemented:
1)
At least 10% of the total of each type of snubber shall. be functionally tested either in-place or in a bench test.
For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.10f.. an additional 5%
of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested; or NILLSTONE - UNIT 3 3/4 7-23 AmendmentJJ,77,JPP, 0372
}
s
' Docket No. 50-423 B15225 l
L 1
i s
Millstone Nuclear Power Station, Unit No. 3 Description of the Proposed Technical Specification Changes i
i t
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June 1995 i
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U.S. Nuclear Regulatory Commission B15225/ Attachment 3/Page 1 June 29, 1995 Millstone Nuclear Power Station, Unit No. 3 Description of the Proposed Technical specification Changes l
Introduction Millstone Unit No. 3 has begun Cycle 6 which utilizes a 24-month fuel cycle.
To take advantage of this longer fuel cycle, NNECO is proposing to modify the frequency of a
number of the surveillance requirements existing in the Millstone Unit No. 3 l
Technical Specifications.
The proposed changes are described below:
Description of the Pronosed Chances Sections 4.6.4.2.b throuah 4.6.4.2.f.
Electric Hydrocen Recombiners. Surveillance Recuirement 1.
Surveillance Requirement 4.6.4.2.b through 4.6.4.2.f verifies the operability of the hydrogen recombiners by performing a
channel calibration of applicable instrumentation and control
- circuits, by verifying integrity of all electrical heaters, and by performing a functional test of the hydrogen recombiner system.
NNECO proposes to extend the frequency of Surveillance Requirement 4.6.6.2 from at least once per 18 months to once each refueling interval.
No changes to the Bases section are proposed.
The proposed changes are consistent with the guidance contained in Generic Letter (GL) 94-04.
In addition, NNECO proposes to delete SR 4.6.4.2.a, since this surveillance is addressed in the performance of 4.6.4.2.e and 4.6.4.2.f.
2.
Section 4.7.1.2.1.
Auxiliary Feedwater System.
Surveillance Recuirements The proposed change modifies the frequency of Surveillance Requirement 4.7.1.2.1.c of the Millstone Unit No. 3 Technical Specifications.
This surveillance requirement verifies automatic starting capability of each auxiliary feedwater system pump once per 18 months during shutdown.
NNECO proposes to extend the frequency of Surveillance Requirement 4.7.1.2.1.c from at least once per 18 months to at least once each refueling interval (i.e., nominal 24 months).
In addition, the phrase "during shutdown" is Surveillance Requirement 4.7.1.2.1.c is being deleted.
Because the terms, " Shutdown" and " Cold Shutdown" are defined in the Millstone Unit No. 3 Technical Specifications as operating modes or conditions, the added restriction to u
n
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./
- U.S.. Nuclear Regulatory Commission
~
^ B15225/ Attachment 3/Page 2
' June-29,11995 E
-perform certain surveillances. may be misinterpreted.-
1,
'This change (1.e.',
deletion of phrase "during shutdown")
i
-is consistent with the recommendations -of GL 91-04.
No changes are proposed to Bases Section 3/4.7.1.2.
3.
Section 4.7.3.
Reactor Plant Connonent Coolina Water System. Surveillance Reauirements l
The proposed change modifies the frequency of l
Surveillance Requirement 4.7.3.b of the Millstone Unit I
No. 3 Technical Specifications.
This-surveillance
(
requirement verifies the operability of the reactor plant l-component cooling water system pumps.and automatic valves in the flow path.
NNECO proposes to extend the frequency l
of Surveillance Requirement 4.7.3.b from at least once-per 18 months to at least once each ' refueling. interval (i.e., nominal 24 months).
In addition, the phrase "during shutdown" in.Surveillanco Requirement 4.7.3.b is being deleted.
Because the terms.
l
" Hot Shutdown" and " Cold Shutdown" are defined in the l
Millstone Unit No. 3 Technical Specifications
'as operating modes or conditions, the added restriction to perform certain surveillances may be misinterpreted.
j This change (i.e.,
deletion of the-phrase "during shutdown")
is - consistent with the recommendations of-GL 91-04.
No ' changes are proposed to Bases section 3/4.7.3.
4.
Section 4.7.4.
Service Water System.
Surveillance Requirements The proposed change modifies the frequency of Surveillance Requirement 4.7.4.b of the Millstone Unit No. 3 Technical Specifications.
This surveillance requirement verifies the operability of the service water system pumps and automatic valves which serve safety-related equipment.
In addition,- the phrase "during shutdown" in Surveillance Requirement 4.7.4.b is being deleted.
Because the terms " Hot Shutdown" and " Cold Shutdown" are defined in the Millstone Unit No. 3 Technical Specifications as' operating modes or conditions, the added restriction' to perform certain surveillances may be misinterpreted.
This change is consistent with the recommendations of GL 91-04.
I 5.
Section 4.7.10. Snubbers. Surveillance Reauirements L-The proposed change modifies the frequency of-I' Surveillance Requirement 4.7.10.e of-the Millstone Unit l
No. 3 Technical Specifications.
This surveillance requirement requires the functional testing of a
)
.f' U.S. Nuclear. Regulatory Commission-B15225/ Attachment 3/Page 3 June.29, 1995.
representative sample of snubbers of each type.
NNECO proposes to extend the frequency of surveillance Requirement 4.7.10.e from at.least once per 18 months to at least once each refueling interval (i.e.,
nominal 24 months).
In addition, the phrase "during shutdown" in surveillance l
Requirement 4.7.10.e is being deleted.
Because the terms
" Hot Shutdown" and " Cold Shutdown" are defined in the l
Millstone Unit No. 3 Technical specifications as operating modes or conditions, the added restriction to i
perform certain surveillances may be. misinterpreted.
}
This change (i.e.,
deletion of the phrase "during shutdown")
is consistent with the recommendations of GL 91-04.
No changes to Bases section 3/4.7.10 are proposed.
i l
i I
l l
l l
si Docket No. 50-423
+
B15225 l
Millstone Nuclear Power Station, Unit No. 3
(
Proposed Revision to Technical Specifications 24-Month Refuel Cycle l
l Safety Assessment and Significant Hazards Consideration for Changes to:
l l
' Electric Hydrogen Recombiners j
Auxiliary Feedwater System Reactor Plant Component Cooling Water System Service Water System Snubbers i
l.
l l
June 1995
1 U.S. Nuclear' Regulatory Commission B15225/ Attachment 4/Page 1 June 29, 1995 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications 24-Month Refuel Cycle Safety Assessment and Significant Hazards Consideration safety Assessment and significant Essards ' consideration for
)
Changes to Electrical Mydrogen Recombiners, Auxiliary Feedwater j
system, Reactor Plant Component Cooling Water system, service Water system and snubbers
Background
on June 7, 1995, Millstone Unit No. 3 began operating on a 24-month (nominal) fuel cycle instead of the previous 18-month cycles.
To take advantage of this longer fuel cycle, NNECO is proposing to modify the frequency of a number of the Surveillance Requirements existing in the Millstone Unit No.
3 Technical Specifications.
This request deletes Surveillance Requirement 4.6.4.2.a and also modifies the frequency of Surveillance Requirements 4.6.4.2.b through f,
4.7.1.2.1.c, 4.7.3.b.1, 4.7.3.b.2, 4.7.4.b.1, 4.7.4.b.2, and 4.7.10.e of the Millstone' Unit No. 3 Technical Specifications.
These Surveillance Requirements deal with the i
verification of the operability of the electric hydrogen recombiners, the operability - of the automatic start feature for the auxiliary feedwater (AFW)
- pump, the operability of the reactor plant component cooling water system automatic isolution valves and pumps, the operability of the service water system (SWS) isolation valves and pumps, and the operability of the snubbers.
1 In the near future, NNECO will be proposing additional changes to the Millstone Unit No. 3 Technical Specifications to prepare for the conversion to 24-month fuel cycles.
Each of these submittals will contain evaluations that are independent and which stand-i alone.
i A.
Electric Hydrocen Recombiners Safety Assessment The electric hydrogen recombiner system (HRS) is a
safety
- related, seismic Category I
system used to. control the concentration of hydrogen inside containme.nt to below the flammability limit of four percent after a loss of coolant accident (LOCA).
The HRS has two 100 percent capacity trains which are used to control the hydrogen in the containment
{
atmosphere at a
safe concentration following a LOCA.
The i
recombiners, which are manually initiated, recombine hydrogen
.._....c a
j
'U.S. Nuclear. Regulatory Commission
- 815225/ Attachment 4/Page 2 June 29, 1995 with oxygen from the containment atmosphere.to form water.
This is. accomplished by the recirculation of air from the containment through the recombiner which consists of a blower,
- heater, catalytic recombiner, and cooler.
The components required ' to perform this function are located on the Hydrogen Recombiner skid j
-(3HCS*RBNRIA/B) and include the piping and valves leading to and from them.
While the plant is in Modes 1 or 2,
a minimum of two electric hydrogen recombiners are required to be operable.
This ensures I
that at least one recombiner is available, given a single failure.
The HRS is not required to be operable in Modes 3 through 6.
l The Millstone Unit No. 3 Technical Specifications demonstrates the operability of the HRS.through the performance of Surveillance Requirements 4.6.4.2.a, 4.6.4.2.b, 4.6.4.2.c, 4.6.4.2.d, 4.6.4.2.e, and 4.6.4.2.f.
Surveillance Requirement 4.6.4.2.a is proposed to be deleted since this surveillance is a reduced scope of the testing and is redundant to Surveillance Requirement 4.6.4.2.e and 4.6.4.2.f.
Surveillance Requirement 4.6.4.2.a uses the same test procedure as 4.6.4.2.e and 4.6.4.2.f, while varying the acceptance criteria.
Since Surveillance Requirements 4.6.4.2.e and 4.6.4.2.f perform the same' functional test as Surveillance Requirement 4.6.4.2.a, Surveillance Requirement 4.6.4.2.a can be deleted.
l NNECO also proposes to redefine the basis of the term refueling interval.
Currently, Millstone Unit No. 3 operates on an 18-month fuel cycle.
Beginning with Cycle 6, Millstone Unit No. 3 will be utilizing fuel capable of operating for a nominal 24-month fuel cycle
- and, therefore, is proposing to conduct Surveillance Requirements 4.6.4.2.b, 4.6.4.2.c, 4.6.4.2.d, 4.6.4.2.e, and
- 4. 6.4. 2. f ' at least once each refueling ' interval (i.e., every 24 months i 25 percent).
Surveillance Requirement 4.6.4.2.b requires that a
channel calibration of all recombiner instrumentation 'and control circuits be performed each refueling interval, currently, the refueling interval is assumed to be 18 months.
The proposed modification will increase the frequency of Surveillance Requirement 4.6.4.2.b to at least once each 24 months.
A review was performed of the equipment performance over the last four operating cycles to determine the impact of extending the frequency of Surveillance Requirement 4.6.4.2.b.
This evaluation included a
review of surveillance
- results, preventative maintenance records, and the frequency and types of corrective maintenance.
A review of these records has shown no operationally significant failures over the past four operating cycles.
Since the proposed
i
]
y..
- U.S.' Nuclear Regulatory Commission
'B15225/ Attachment 4/Page 3 June 29, 1995 modification does not impact the operation of-the hydrogen recombiners instrumentation, a decrease 'in the frequency of.
testing is considered to have a. minimal impact on safety.
Therefore, surveillance Requirement'
- 4. 6.4.2.b - can be extended from its current value of once every 18 months (f. 25 percent) to
]
the new value of once every 24 months (* 25 percent).
]
Surveillance Requirement 4.6.4.2.c requires that_ a verification i
be performed that there has been no degradation to the internals of the recombiners since the last examination..
Surveillance Requirement 4.6.4.2.d requires the verification of the integrity i
of all electrical circuits of the recombiner by performing a resistance to ground test following the above' functional tests.
A review of the resistance to ground tests for ' the hydrogen recombiner heaters shows that the resistances are at least 8000 times greater than the minimum resistance of 10,000 ohms. _The r
previous visual examinations, conducted over the previous j
refueling cycles have all been satisfactory.
1 A review of the preventative maintenance records revealed that
-j there are no electrical preventative maintenance-activities' performed on an 18-month basis.
The preventative maintenance required to be performed are conducted at quarterly intervals or j
at 3 or 10 year intervals and could'be performed while the plant is in operation.
l i
corrective maintenance work performed on the hydrogen recombiners
'during the last four cycles were all minor activities.
In all-
)
cases, the repairs were able to be made with no adverse impact to plant operation.
Extension of the performance of Surveillance Requirements 4.6.4.2.c and 4.6.4.2.d from 18 months (i 25 percent) to 24 months
(*
25 percent) is ~ justified by considering the surveillance history, the frequency of other surveillances, and j
the preventative maintenance activities that are performed on the j
hydrogen recombiner.
Surveillance Requirements 4.6.4.2.e and 4.6.4.2.f verifies the operability of the hydrogen recombiner blowers, heaters and-recombiners.
A review of the tests of tho'~recombiners identified two failures.
One failure was caused by a loose termination to a breaker which caused the
'A' recombiner to_ trip.
There-have been no repetitive failures of the breaker connection since that time.
The second iesue was caused by NNECO's imposition on itself of overly restrictive acceptance criteria, which caused the recombiners to fail the surveillance.
The improper acceptance criteria was modified by NNECO and accepted by the NRC Staff via License Amendment No. 63.
NNECO's.
investigation into this event determined that no component degradation occurred.
m
+ _. - - -, -
l j
.i;
}
U.S. Nuclear Regulatory Commission J
B15225/ Attachment 4/Page 4 1
- June 29, 1995 A' review of the corrective maintenance - records have shown only i
one significant mechanical' equipment failure over.the past four.
j operating - cycles.
The blower 3HCS*C1B suffered an. oil leak - in 1
1993.which required the replacement of tho' blower.
There was no impact on the plant operation as the unit was in a refueling j
outage at the time.
No other failures of this -type have occurred.
l t
Quarterly valve surveillance tests (these tests are not required.
per technical specification) provide some assurance of the proper operation of the hydrogen recombiner system.
During these tests the maintenance valves are opened and proper flow rates verified I
during system operation.
This test is intended to confirm proper operation of-the valves in the system.
The proper operation of i
the heater and recombiners are not verified by this test.
However,.the blowers and control systems must function properly in order to meet the acceptance criteria of this quarterly test.
A review of the preventative maintenance records revealed-that there is no preventative maintenance required on an 18-month refueling outage basis.
As such, the extension of the operating-
-cycle from a nominal 18 months (i 25 percent) to 24 months (i 25 l
percent) will not adversely impact the maintenance schedule of j
the equipment.
A Probabilistic Risk - Assessment (PRA) review concluded that the proposed change is not risk significant.
\\
Sionificant Basards Consideration i
i In accordance with 10CFR50.92, NNECO has reviewed the proposed changes and has concluded that they do not involve a significant.
hazards consideration (SHC).-
The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised.
l The proposed changes do not involve an SHC because the changes would not:
1.
Involve a
significant increase in the probability or consequences of an accident previously evaluated.
l The proposed deletion of Surveillance Requirement 4.6.4.2.a I
will not involve a significant increase in the probability or.
consequences of an accident. previously evaluated since the surveillance required by 4.6.4.2.a is addressed by the performance of Surveillance Requirements 4.6.4.2.e.
and 4.6.4.2.f.
The proposed changes to Surveillance Requirements 4.6.4.2.b, 4.6.4.2.c, 4.6.4.2.d, 4.6.4.2.e, and 4.6.4.2.f extend the frequency of demonstrating operability of the hydrogen recombiners.
The proposed changes would extend the frequency from at least once per refueling (presently the refueling is
U.S.. Nuclear Regulatory Commission l
B15225/ Attachment 4/Page 5 i
June 29, 1995 i
defined as once per 18 months) to once each refueling l
interval (i.e., 24 months).
The proposed changes do not alter the intent or method by l
which the surveillances are conducted.
In addition, the acceptance criterion for the ' extended surveillances are unchanged.
As such, the proposed changes will not degrade the ability of the hydrogen recombiners to perform their i
intended function.
1 Since the proposed changes eliminate a duplication of. a Surveillance Requirement and for the remaining, only affects
)
the surveillance frequency of a safety related system that is
+
manually initiated 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> subsequent to a
- LOCA, the proposed changes cannot affect the probability of any i
previously analyzed accident.
While the proposed changes lengthen the duration between surveillances, the extensions have no significant impact on the reliability or availability j
of the systems.
Consequently, the recombiners will continue to operate as required, and there is no impact on the consequences of any accidents analyzed previously.
.i 2.
Create the possibility of a new or different' kind of accident i
from any accident previously evaluated.
The deletion of Surveillance Requirement 4.6.4.2.a removes a
[
duplicate surveillance requirement which is adequately addressed by the performance of Surveillance Requirements t
4.6.4.2.e and 4.6.4.2.f.
The remaining changes do not alter the intent' or method by which the surveillances are conducted, do not involve any physical changes to the plant, do not alter the way any structure, system, or component functions, and do not modify the manner in.which the plant is operated.
As such, the proposed changes do not introduce a new failure mode.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed.
3.
Involve a significant reduction in a margin of safety.
The deletion of Surveillance Requirement 4.6.4.2.a does not decrease the margin of safety since Surveillance Requirement 4.6.4.2.a is captured in Surveillance Requirements 4.6.4.2.e and 4.6.4.2.f.
Changing the frequency of Surveillance Requirements 4.6.4.2.b through
- 4. 6. 4. 2. f from at least once per 18 months to once each refueling interval does not change the basis for the frequency.
The proposed change to the surveillance frequency does not change the intent or method for performing the surveillance.
- Further, the current inservice testing
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 6 June 29, 1995 requirements, and the previous history of reliability of the system provides assurance that the changes will not affect the reliability of the HRS.
Thus, it is concluded that there is no impact on the margin of safety B.
Auxiliary Feedwater systen Safety Assessment The AFW system (AFWS) provides a supply of high-pressure feedwater to the secondary side of the steam generators for removing heat from the reactor coolant system following a loss of normal feedwater.
It also provides a cooling source in the event of a small break loss-of-coolant accident (LOCA).
Furthermore, the system is used in the event of a main steam line break, loss of power (LOP), or low-low steam generator water level conditions.
Under LOP conditions, the AFWS maintains the plant at a standby condition.
Under station blackout conditions, the turbine-driven AFW pump will be responsible for maintaining AFWS flow using only Class 1E DC power.
Additionally, the AFWS operates during startup and hot standby to maintain water level in the steam generator.
The AFWS is comprised of two motor-driven pumps, one turbine-driven pump, and the associated piping and valves necessary to connect the demineralized water storage tank (DWST) to the pump suctions and the pump discharges to the feedwater system.
The pumps and valves are supplied by redundant electrical power.
The AFWS possesses redundant piping flow paths from the pumps to the steam generators to ensure that i
the required flow is provided to at least two steam i
generators, assuming a single failure.
The AFWS is capable of supplying sufficient heat-removal capacity to remove the sensible and decay heat from the reactor core in the event of a single failure.
During normal plant operation, the AFWS pumps are on standby.
The motor-driven AFWS pumps start automatically whenever any of the following conditions occur:
1)
3) containment depressurization actuation signal (CDA),
and 4) two-of-four low-low water level signals in any one steam generator.
]
During startup and hot standby, the motor-driven AFWS pumps take sucti.on from the non-safety-related condensate storage tank (CST) to maintain steam generator water levels.
In the event of an SIS, LOP, CDA signal, or two-of-four low-low water level in any one steam generator, the suction source is
U.S. Nuclear Regulatory Commission B15225/ Attachment'4/Page 7 June 29, 1995.
automatically switched to the DWST, and the suction lines from.the CST are isolated.
l t
The turbine-driven AFWS pump starts automatically on two-of-i four low-low water level signals in any two-of-four steam i
generators.
The Millstone Unit No. 3 Technical Specifications contain
[
requirements for demonstrating the operability of the AFWS.
They are Surveillance Requirements 4.7.1.2.1.a, b, and c, and 4.7.1.2.2.
Currently, Surveillance Requirement 4.7.1.2.1.c requires that each AFWS pump starts automatically upon l
receipt of an AFW actuation test signal at least once per 18 months during shutdown.
NNECO is proposing to change the l
frequency of Surveillance Requirement 4.7.1.2.1.c from at l
least once per 18 months during shutdown to at least once i
each refueling interval.
In addition, the phrase "during shutdown" in Surveillance l
Requirement 4.7.1.2.1.c is being deleted.
Because the term
" Hot Shutdown" and
" Cold Shutdown" are defined in the Millstone Unit No. 3 Technical Specifications as operating modes or conditions, the added restriction to perform certain surveillances may be misinterpreted.
The proposed deletion of the term "during shutdown" is consistent with the i
recommendation of Generic Letter (GL) 91-04.
In GL 91-04, the NRC has concluded that the technical specifications need not restrict surveillances as only being performed during shutdown.
However, the NRC indicated that if the performance of a refueling interval surveillance during plant operation would adversely affect safety, the 1
licensee should postpone the surveillance until the' plant is shut down for refueling or in a' condition or mode consistent with safe conduct of that surveillance.
NNECO believes that the deletion of the words "during shutdown" has no safety-impact as long as tha surveillances are conducted in any mode 3
or condition without impacting the plant safety.
i Surveillance Requirements 4.7.1.2.1.a and b and 4.7.1.2.2 provide additional assurance of the operability of the AFWS.
l Surveillance Requirement 4.7.1.2.1.b requires that each AFWS pump be demonstrated operable at least once per 92 days on a j
staggered test basis by: 1) verifying that each pump develops the appropriate differential pressure at the appropriate conditions.
Surveillance Requirement 4.7.1.2.1.a provides l
additional assurance of the operability of the AFWS by i
verifying that each non-automatic valve in the flow path that i
is not locked, sealed, or otherwise secured in position is in i
its correct position; and 2) verifying that each AFW control and isolation valve in the flow path is in the fully open j
i t
U.S. Nuclear Regulatory Commission i
l B15225/ Attachment 4/Page 8 June 29, 1995 position when above 10% rated thermal power.
Surveillance Requirement 4.7.1.2.2 requires that an AFW flow path to each steam generator be demonstrated operable following each cold shutdown of greater than 30 days prior to entering. Mode 2 by verifying flow to each steam generator.
The Class 2 and 3 portions of the AFWS are tested, as
- required, by ASME Section XI.
This ' inservice testing provides additional assurance of the mechanical operability of the Class 2 and 3 AFWS components.
Additionally, the automatic actuation logic and relays
)
responsible for starting the AFWS pumps are required to be l
tested by Surveillance Requirement 4.3.2.1 of the Millstone Unit Nc.
3 Technical Specifications.
Surveillance Requiremont 4.3.2.1 requires the conduct of monthly actuation logic tests and master relay tests and quarterly slave relay tests for the automatic actuation logics and actuation relays.
Equipment performance over the last four operating cycles was evaluated to determine the impact of extending the frequency.
of Surveillance Requirement 4.7.1.2.1.c.
This evaluation included a
review of surveillance
- results, preventive maintenance records, and the frequency and type of corrective maintenance.
The AFWS motor-driven pumps have started as required when tested in accordance with surveillance Requirements 4.7.1.2.1.c.
The turbine driven pump is tested at least quarterly because of overlapping technical specification surveillances.
With respect to the turbine-driven AFW pumps, surveillance procedures (SP 3622.3 and SP 3646A.8 (A.9)), which are used i
to comply with Technical Specification 4.7.1.2.1.b's 18-month surveillance requirement, are performed at least quarterly because of overlap with other technical specification surveillance requirements (e.g.,
Surveillance 4.7.1.2.1.b, AFW pump hydraulic performance surveillance requirements and Table 4.3-2, Engineered Safety Features Actuation System Instrumentation surveillance requirements) and S.P. 3622.3, Section 4.4, turbine-driven AFW pump cold start test requirements.
There are no preventive maintenance activities for the AFW pump auto-start feature (and which would not be surveilled by other more frequent testing) that are performed on an 18-month interval.
Also, no corrective maintenance activities associated with the motor-driven AFW pump auto-start feature were identified.
The turbine driven AFW pump has had two
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 9 June 29, 1995 recent auto-start failures and corrective maintenance activities have been performed to address these failures.
However, at least quarterly, a cold start turbine driven AFW pump surveillance is conducted.
Therefore, these corrective maintenance activities have no impact on the proposed 18 month surveillance frequency change.
Based on this review, the auto-start feature for the AFWS pumps has been determined to be reliable, except for two recent turbine driven AFW pump auto-start failures.
- However, other overlapping turbine driven AFW pump (quarterly) plant surveillances comprehensively test the auto-start feature.
Therefore, the proposed technical specification change poses no safety reduction since the pump is tested quarterly by overlapping surveillance testing.
- Also, there is no indication that the proposed extension could cause deterioration in the condition or performance of any of the subject AFWS components.
A PRA review concluded that the proposed change is not risk significant.
Significant Easards Consideration NNECO has reviewed the proposed change in accordance with 10CFR50.92 and concluded that the change does not involve a SHC.
The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised.
The proposed change does not involve a SHC because the change would not:
1.
Involve a
significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to surveillance Requirement 4.7.1.2.1.c of the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying that each AFWS pump automatically starts upon receipt of an AFW actuation test signal.
The proposal would extend the frequency from at least once per 18 months to at least once each refueling interval (i.e.,
nominal 24-months).
i Changing the frequency of Surveillance Requirement 4.7.1.2.1.c from at least once per 18 months during shutdown to at least once each refueling interval does not change the basis for the fraquency.
The frequency was chosen because of the need to perform this verification under the conditions that apply during a plant outage, and to avoid the potential of an unplanned transient if the surveillance were conducted with the plant at power.
The proposed change does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any
_y
~
)
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 10 June 29, 1995 structure, system,. or component. functions, and does not modify the manner in which the plant is operated.
As such, the proposed change in the frequency of Surveillance Requirement 4.7.1.2.1.c will not degrade the ability of the AFWS to perform its safety function.
- Also, the AFWS is designed to perform its intended safety function even in the event of a single failure.
Additional assurance of the operability of the AFWS is I
provided by additional surveillance requirements.
.Also, the class 2 and 3 portions of the AFWS is tested, as required, by ASME Section XI.
This inservice testing provides additional assurance of the mechanical operability of the Class 2 and 3 AFWS components.
Also, the automatic actuation logic and relays responsible for starting the AFWS pumps are required to be tested by Surveillance Requirement 4.3.2.1 of the Millstone Unit No. 3 Technical Specifications.
Equipr.ent performance over the last four operating cycles was i
evaluated to determine the impact of extending the frequency of Surveillance Requirement 4.7.1.2.1.c.
This evaluation included a
review of surveillance
- results, preventive maintenance records, and the frequency and type of corrective maintenance.
It concluded that the auto-start feature for th0 AFWS pumps is reliable, and that there is no indication i
that the proposed extension could cause deterioration in the condition or performance of any of the s'abj ect AFWS components.
Since the proposed change only affects the surveillance frequency for safety systems that are used to mitigate accidents, the change cannot affect the probability of any previously analyzed accident.
While the proposed change can lengthen the intervals between surveillances, the increases in intervals has been evaluated and it is concluded that there is no significant impact on the reliability or availability of the safety system and consequently, there is no impact on the consequences on any analyzed accident.
2.
Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change to surveillance Requirement 4.7.1.2.1.b of the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying that each AFWS pump automatically starts upon receipt of an AFW actuation test signal.
The proposal would extend the frequency from at least once per 18 months to at least once each refueling interval (nominal 4
24 months).
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 11 June 29, 1995 Changing the frequency of Surveillance Requirement 4.7.1.2.1.c from at least once per 18 months during shutdown i
to at least once each refueling interval does not change the basis for the frequency.
The frequency was chosen because of l
the need to perform this verification under the conditions i
'that apply during a plant outage, and to avoid the potential of an unplanned transient if the surveillance were conducted with the plant at power.
The proposed change does not alter the intent or method by l
which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure, -system, or component functions, and does not modify the manner in.which the plant is operated.
As such, l
the proposed change cannot create the possibility of a new or i
different kind of accident from any previously evaluated.
l 3.
Involve a significant reduction in a margin of safety.
The proposed change to surveillance Requirement 4.7.1.2.1.c of the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying that each AFWS pump automatically l
starts upon receipt of an AFW actuation test signal.
The proposal would extend the frequency from at least per i
18-months to at least once each refueling interval (24-months).
l The proposed change to surveillance frequency is still consistent with the basis for.the frequency, and the intent or method of performing the surveillance _ is unchanged.
i Further, the current inservice testing requirements and the previous history of reliability of the system provides assurance that the changes will not affect the reliability of the auxiliary feedwater system.
Thus, it is concluded that there is no impact on the margin of safety.
C.
Reactor Plant Connonent Cooling Water Systen l
safety Assessment The reactor plant component cooling water (RPCCW) system is a l
closed loop cooling system that transfers heat from reactor auxiliaries to the service water system during plant l
operations and during normal and emergency cooldown/ shutdown.
l Additionally, the RPCCW provides makeup water to several cooling subsystems.
The RPCCW system consists of three half-capacity motor-driven cooling water
- pumps, three half capacity heat exchangers, a surge tank, a chemical addition tank, associated piping, valves, instrumentation, controls j
and auxiliary electrical equipment.
r u.
L-U.S. Nuclear Regulatory Commission
)
B15225/ Attachment 4/Page 12 June 29, 1995 During normal operation, two reactor plant component cooling pumps and two reactor plant component cooling heat exchangers accommodate the heat removal load.
A' spara pump and heat exchanger is provided to allow pump or heat exchanger I
maintenance.
One reactor plant component cooling pump is fed by one emergency bus, a second reactor plant component cooling pump is fed by the second redundant emergency bus, while the spare reactor plant component cooling pump may be manually connected to either emergency bus.
During accident conditions which do not cause a CDA signal, one reactor plant component cooling pump and one reactor plant component cooling heat exchanger accommodate the heat removal load.
During accident conditions which cause a CDA signal, the I
RPCCW heat exchanger service water (SW) flow is automatically isolated to provide adequate flow to the containment i
recirculation coolers.
Service water flow is re-established to the RPCCW heat exchangers once a second service water pump has started.
The RPCCW system is then available for' transferring heat to the ultimate heat sink.
A failure of l
one power supply train or any reactor plant component in one train does not prevent the system from performing its safety function.
i Surveillance Requirement 4.7.3.b of the Millstone Unit No. 3 Technical Specifications verifies that each automatic valve in the RPCCW that services safety-related equipment actuates to its. correct position on its associated engineered safety l
feature actuation signal, and that each RPCCW pump starts automatically on a safety injection test signal.
Currently, this verification is required to be performed at least once per 18 months during shutdown.
NNECO is proposing to change this frequency to at least once each refueling interval (i.e., at least once each 24 months (nominal)).
Changing the frequency of Surveillance Requirement 4.7.3.b from at least once per 18 months during shutdown to at least once each 24 months does not change the basis for the i
frequency.
The frequency was chosen because of the need to 1
perform this verification under the conditions that apply i
during a plant outage, and to avoid the potential of an unplanned transient if the surveillance were conducted with t
the plant at power.
In addition, the phrase "during shutdown" in Surveillance Requirement 4.7.3.b is being deleted.
Because the term " Hot Shutdown" and " Cold Shutdown" are defined in the Millstone Unit No. 3 Technical Specifications as operating modes or l
conditions, the added restriction to perform certain surveillances may be misinterpreted.
The proposed deletion j
of the term "during shutdown" is consistent with the recommendation of GL 91-04.
l t
I i
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 13 June 29, 1995 In GL 91-04, the NRC has concluded that the technical specifications need not restrict surveillances as only being performed during shutdown.
However, the NRC indicated that if the performance of a refueling interval surveillance during plant operation would adversely affect safety, the licensee should postpone the surveillance until the plant is shut down for refueling or in a condition or mode consistent with safe conduct of that surveillance.
NNECO believes that the deletion of the words "during shutdown" has no safety impact as long as the surveillances are conducted in any mode or condition without impacting the plant safety.
The Class 3 portion of the RPCCW is tested, as required, by ASME Section XI.
This inservice testing provides additional assurance of the mechanical operability of the class 3 RPCCW components.
Also, the automatic actuation logic and relays responsible for starting the RPCCW pump and for actuating the RPCCW isolation valves are required to be tested by surveillance Requirement 4.3.2.1 of the Millstone Unit No. 3 Technical Specifications.
Surveillance Requirement 4.3.2.1 requires the conduct of monthly actuation logic tests and master relay tests and quarterly salve relay tests for the automatic actuation logics and actuation relays.
Equipment performance over the last four operating cycles was evaluated to determine the impact of extending the frequency of Surveillance Requirement 4.7.3.b.
This evaluation included a
review of surveillance
- results, preventive maintenance records, and the frequency and type of corrective maintenance.
Over the past four cycles, there have been 12 tests (six Train A tests and six Train B tests) to verify that the RPCCW valves actuate to their required position, and 12 tests (six on each Train) to verify that the RPCCW pumps automatically start on a safety injection test signal.
The RPCCW pumps passed each of the tests, while only one valve failed to actuate to its required position.
This was due to a breaker trip on control power.
The repairs were made and the valve was successfully retested.
On two separate instances, two different valves provided dual indication.
In both instances, the valves were inspected and were found to be in the correct position.
The two isolated problems were repaired and the valves were retested.
These test results demonstrate that the reliability of the RPCCW isolation valves and pumps is high.
Preventive maintenance activities for the RPCCW isolation valves that are scheduled on a refueling or an 18-month
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 14 June 29, 1995 interval include:
(1) the removal, cleaning, inspection, and testing of motor-operated valve (MOV) breakers; and (2) the reinstallation of MOV
- breakers, and performance of maintenance on Limitorque motor operators, including the checking of grease levels and motor operator condition.
l Review of the RPCCW system regularly scheduled preventative (PM) act
?.i ws associated with the RPCCW pump auto-start feature (and which would ng be surveyed by other more frequent testing required by other technical specifications) with a refueling or 18 month interval, has concluded that there are no PMs with this interval.
Therefore, there are no refueling or 18 month PMs which, if changed to 24 months, would adversely affect the conclusion that the technical specifications surveillance test interval can be increased.
Corrective maintenance records for both the RPCCW automatic valves and the automatic start feature of the RPCCW pumps have been reviewed to see if any maintenance activities would indicate that a 24 month surveillance interval for Technical Specification 4.7.3.b would be imprudent.
The RPCCW system maintenance activity review did not identify any maintenance activities which would indicate that extending the surveillance interval would be inappropriate.
A PRA review concluded that the proposed change is not risk significant.
Sionificant Hazards Consideration NNECO has reviewed the proposed changes in accordance with 10CFR50.92 and concluded that the changes do not involve an SHC.
The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised.
The proposed changes do not involve an SHC because the change would not:
1.
Involve a
significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to Surveillance Requirement 4.7.3.b of the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying that each RPCCW valve actuates to its required position in response to their associated engineered safety features actuation signal and for verifying that each RPCCW pump starts automatically on a safety injection test signal.
The proposal would extend the frequency from at least once per 18 months to at least once each refueling interval (nominal 24-months).
Changing the frequency of Surveillance Requirement 4.7.3.b from at least once per 18 months during shutdown to at least once each refueling interval does not change the basis for the frequency.
The frequency was chosen because of the need
.m U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 15 June 29, 1995 to perform this verification under the conditions that apply during a plant outage, and to avoid the potential of an unplanned transient if the surveillance'were conducted with the plant at power.
)
The proposed change does not alter the intent or method ' by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter ' the way any structure,
- system, or component functions, and does not modify the manner in which the plant is operated.
As such, the proposed change in the frequency of Surveillance Requirement 4.7.3.b will not degrade the ability of the RPCCW to perform its safety function.
Also, the RPCCW is designed to perform its intended safety function even in the event of a single failure.
Equipment performance over the last four operating cycles was evaluated to determine the impact of extending the frequency of Surveillance Requirement 4.7.3.b.
This evaluation included a
review of surveillance
- results, preventive maintenance records, and the frequency and type of corrective maintenance.
It concludes that the RPCCW isolation valves and. pumps are highly
- reliable, and that there is no l
indication that the proposed
-extension could
.cause deterioration in the condition or performance of any of the subject RPCCW components.
The Class 3 portion of the RPCCW is tested, as required, by ASME Section XI.
This inservice testing provides additional l
assurance of the mechanical operability of the Class 3 RPCCW components.
Also, the automatic actuation logic and relays responsible for starting the RPCCW pumps and for' actuating the RPCCW isolation valves are required to be tested by Surveillance Requirement 4.3.2.1 of the Millstone Unit No. 3 Technical Specifications.
Since the proposed change only affects the surveillance frequency for safety systems that are used to mitigate accidents, the change cannot affect the probability of any previously analyzed accident.
While the proposed change can lengthen the interval between surveillances, the increase in interval has been evaluated and it is concluded that there is no significant impact on the reliability or availability of these safety systems and consequently, there is no impact on the consequences on any analyzed accident.
2.
Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change to Surveillance Requirement 4.7.3.b of the Millstone Unit No. 3 Technical Specifications extends the i
l U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 16 June 29, 1995 l
frequency for verifying that each RPCCW isolation valve actuates to its required position in response to their associated engineered safety features actuation signal, and for verifying that each RPCCW pump starts automatically on a safety injection test signal.
The proposal would extend the frequency from at least once per 18 months during shutdown to at least once each refueling interval (nominal 24 months).
The proposed change does not alter-the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure,
- system, or component functions, and does not modify the manner in which the plant is operated.
As such, the proposed change cannot create the possibility of a new or different kind of accident from any previously evaluated.
3.
Involve a significant reduction in a margin of safety.
The proposed change to Surveillance Requirement 4.7.3.b of I
the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying that each RPCCW isolation valve actuates to its required position in response to their associated engineered safety features actuation signal, and for verifying that each RPCCW pump starts automatically on a 1
safety injection test signal.
The proposal would extend the frequency from at least once per 18 months during shutdown to at least once each refueling interval (nominal 24 months).
The proposed change to the surveillance frequency are still consistent with the basis for the frequency and the intent or method of performing the surveillance is unchanged.
- Further, the current inservice testing requirement and previous history of reliability of the service water system provide assurance that the changes will not affect the reliability of the service water system.
Thus, it is concluded that there is no impact on the margin of safety.
D.
service Water system Safety Assessment The Millstone Unit No.
3 SWS transfers heat from safety-related structures, systems, and components to a heat sink, during both normal and accident conditions.
Additionally, the SWS supplies cooling water for various non-safety related structures,
- systems, and components, during normal operations, and provides an emergency source of makeup water to the fuel pool and an emergency backup source of water to the AFWS and control building chilled water system.
.~ -
~U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 17 June 29, 1995 The SWS consists of two redundant flow paths, each consisting i
of two 100% capacity pumps, two self-cleaning strainers, two booster pumps, piping, and valves.
Power is supplied to the redundant SWS pumps from separate emergency buses.
Each pump can supply the minimum cooling water flow requirements during a design basis accident (DBA) coincident with a LOP and l
during cold shutdown coincident.with a LOP.
Redundant MOVs are used for isolation and diversion functions; the redundant
[
MOVs receive power from separate emergency. buses.
i During normal operations, the SWS pumps in each header are I
operated in an alternating manner to achieve uniform wear.
One SWS pump on each redundant header is in operation, while the remaining pump on each header remains in standby.
The majority of the SWS is continually in use and is monitored and observed by personnel to assure continued safe operation of the plant.
Additionally, SWS flow and temperature data for the reactor plant component cooling water heat exchangers i
is taken periodically to indicate possible biological fouling problems.
f f
Surveillance Requirement 4.7.4.a of the Millstone Unit'No. 3 Technical Specifications requires that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position be verified to be in its correct position at least once per 31 days.
This Surveillance Requirement provides assurance that the proper flow paths exist for SWS operation.
This Surveillance Requirement does not require any testing or valve manipulation; it only involves verification that those valves are in the correct position.
i In the event of a LOCA or high energy line break within the containment, a
containment depressurization
- signal, automatically closes the MOVs in the supply lines to the t
nonsafety-related equipment, and automatically opens motor-operated isolation valves in the supply lines to the containment recirculation coolers.
In such an accident, coincident with a LOP, two SW pumps, each supplied by a separate emergency bus, start automatically.
Surveillance Requirement 4.7.4.b of the Millstone Unit No. 3 Technical Specifications verifies that each automatic valve in the SWS that services safety-related equipment actuates to i
its correct position on its associated engineered safety feature actuation signal, and that each SW pump starts automatically on a safety injection test signal.
Currently, this verification is required to be performed at least once per 18 months during shutdown.
NNECO is proposing to change this frequency to at least once each refueling interval i
(i.e., at least once each 24 months (nominal)).
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U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 18 June 29, 1995 Changing the frequency of Surveillance Requirement 4.7.4.b from at least once per 18 months during shutdown to at least once each 24 months does not change the basis for the frequency.
The frequency was chosen because of the need to perform this verification under the conditions that apply during a plant outage, and to avoid the potential of an unplanned transient if the surveillance were conducted with the plant at power.
In addition, the phrase "during shutdown" in Surveillance i
Requirement 4.7.4.b is being deleted.
.Because of the terms 1
" Hot Shutdown" and
" Cold Shutdown" are defined in -the Millstone Unit No. 3 Technical Specifications as operating modes or conditions, the added restriction to perform certain surveillances may be misinterpreted.
The proposed deletion of the term "during shutdown" is consistent with the recommendation of GL 91-04.
In GL 91-04, the NRC has concluded that the technical specifications need not restrict surveillances as only being performed during shutdown.
However, the NRC indicated that if the performance of a refueling interval surveillance during plant operation would adversely affect safety, the licensee should postpone the surveillance until the plant is shut down for refueling or in a condition or mode consistent with safe conduct of that surveillance.
NNECO believes that the deletion of the words "during shutdown" has no safety impact as long as the surveillances are conducted in any mode or condition without impacting the plant safety.
The Class 3 portion of the SWS is tested, as required, by ASME Section XI.
This inservice testing provides additional assurance of the mechanical operability of the class 3 SWS components.
Also, the automatic actuation logic and relays responsible for starting the SW pumps (i.e., safety injection signal) and for actuating the SWS isolation valves (i.e., CDA signal) are required to be tested by Surveillance Requirement 4.3.2.1 of the Millstone Unit No.
3 Technical Specifications.
Surveillance Requirement 4.3.2.1 requires the conduct of monthly actuation logic tests and master relay tests and quarterly slave relay tests for the safety injection and containment spray (i.e.,
CDA signal) automatic actuation logics and actuation relays.
Equipment performance over the last four operating cycles was evaluated to determine the impact of extending the frequency of Surveillance Requirement 4.7.4.b.
This evaluation included a
review of surveillance
- results, preventive
2 U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 19 June 29, 1995 maintenance records, and the frequency and type of corrective maintenance.
Over the past four cycles, there have been 11 tests. (five Train A tests and 6 Train B tests) to verify that the SWS isolation valves actuate to their required position, and 12 tests (six' for each Train) to verify that the SW. pumps automatically start on a safety injection test signal.
The SW. pumps passed each of the tests, while only one valve failed to actuate to its required position.
These test results demonstrate that the reliability of the SWS isolation valves and pumps is high.
Valve 3SWP*MOV50D failed to actuate to its required position during the test conducted on December 1,
1987.
This valve failed to actuate to its required position due to dirty contacts found on an auxiliary relay.
This condition was rectified, and the valve returned to service.
Preventive maintenance activities for the SWS isolation valves that are scheduled on a refueling or an 18-month interval' include: (1) the removal, cleaning, inspection, and testing of MOV breakers; and (2) the reinstallation of MOV breakers, and performance of maintenance on Limitorque motor operators, including the checking of grease levels and motor operator condition.
No preventive maintenance activities associated with the SWS pump automatic start feature were identified that are conducted on either.a refueling or an 18-month interval.
Extending the 18-month intervals for breaker maintenance and lubrication of the SWS MOVs are acceptable because of the low frequency of use of the
- valves, the past operating experiences associated with the valves, and the moderate normal ambient conditions that the valves experience.
Corrective maintenance records for both the SWS isolation valves and the automatic start feature of the SWS pumps were reviewed to determine if any maintenance activities would indicate that extension of the surveillance frequency would be imprudent.
This review did not identify any maintenance activities that would
. indicate that extending the surveillance frequency would be inappropriate.
A PRA review concluded that the proposed change is not risk significant.
Sionificant Easards consideration NNECC has reviewed the proposed change in accordance with 10CFR50.92 and concluded that the change does not involve a SHC.
The basis for this conclusion is that the three criteria of
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 20
' June 29, 1995 j
i 10CFR50.92(c) are not compromised.
The proposed change does not involve a SHC because the change would not-1.
Involve a
significant-increase in the probability or consequences of an accident previously evaluated.
The proposed change to Surveillance Requirement 4.7.4.b of l
the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying that each SWS isolation valve i
actuates to its required position in response to their associated engineered safety features actuation signal, and for verifying that each SWS pump starts automatically on a i
safety injection test signal.
The proposal would extend the l
frequency from at least once per la months to at least once each refueling interval (nominal 24 months).
l l
Changing the frequency of Surveillance Requirement 4.7.4.b from at.least once per 18 months during shutdown to at least once each refueling interval does not change the basis for l
the frequency.
The frequency was chosen because of the.need j
to perform this verification under the conditions that apply i
during a plant outage, and to avoid the potential of an unplanned transient if the surveillance were conducted with the plant at power.
The proposed change does not alter the intent or method by which the surveillances are conducted, does not involve any I
physical changes to the plant, does not alter the way any structure,
- system, or component functions, and does not.
modify the manner in which the plant is operated.
As such, the proposed change in the frequency of Surveillance Requirement 4.7.4.b will not degrade the ability of the SWS to perform its safety function.
Also, the SWS is designed to perform its intended safety function even in the event of a single failure.
The majority of the SWS is continually in use and is monitored and observed by personnel to assure continued safe operation of the plant.
Additionally, SWS flow and temperature data for the RPCCW heat exchangers is taken periodically to indicate possible biological fouling problems.
Equipment performance over the last four operating cycles was evaluated to determine the impact of extending the frequency of Surveil?ance Requirement 4.7.4.b.
This evaluation i
included a
review of surveillance
- results, preventive i
maintenance records, and the frequency and type of corrective 1
maintenance.
It concluded that the SWS isolation valves and pumps are highly reliable, and that there is no indication l
that the proposed extension could cause deterioration in the I
r
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 21 June 29, 1995-condition or performance of any of the subject SWS components.
The class 3 portion of the SWS is tested, as required, by ASME Section XI.
This inservice testing provides additional assurance of the mechanical operability of the class 3 SWS components.
Also, the automatic actuation logic and relays responsible for starting the SW pumps (i.e., safety injection signal) and for actuating the SWS isolation valves are required to be tested by Surveillance Requirement 4.3.2.1 of the Millstone Unit No. 3 Technical Specifications.
Since the proposed change only affects the surveillance j
frequency for safety systems that are used to mitigate i
accidents, the change cannot affect the probability of any previously analyzed accident.
While the proposed change can lengthen the intervals between surveillances, the increases in intervals has been evaluated and it is concluded that there is no significant impact on the reliability or availability of these safety systems and consequently, there is no impact on.the consequences on any analyzed accident.
I 2.
Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change to Surveillance Requirement 4.7.4.b of
?
the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying that each SWS isolation valve l
actuates to its required position in response = to their j
associated engineered safety features actuation signal, and for verifying that each SW pump starts automatically on a safety injection test signal.
The proposal would extend the-frequency from at least once per 18 months to at least once i
each refueling interval (nominal 24 months).
i The proposed change does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any l
structure,
- system, or component functions, and does not modify the manner in which the plant is operated.
As such, the proposed change cannot create the possibility of a new or different kind of accident from any previously evaluated.
l 3.
Involve a significant reduction in a margin of safety.
}
The proposed change to Surveillance Requirement 4.7.4.b of the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying that each SWS isolation valve actuates to its required position in response to their associated engineered safety features actuation signal, and for verifying that each SW pump starts automatically on a
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 22 June 29, 1995 safety injection test signal.
The proposal would extend the frequency from at least once per 18 months to at least once each refueling interval (nominal 24 months).
The proposed change to surveillance frequency is still consistent with the basis for the frequency ind the intent or method of performing the surveillance is unchanged.
- Further, the current inservice testing requirements and previous j
history of reliability of the SWS provide assurance that the 1
change will not affect the reliability of the SWS.
Thus, it is concluded that there is no impact on the margin of safety.
E.
Snubbers Safety Assessment Surveillance Requirement 4.7.10.e of the Millstone Unit No. 3 Technical Specifications requires that a
representative sample of each snubber type be tested every 18 months via one of three defined test plans.
NNECO is proposing to change the frequency of this surveillance requirement from at least once per 18 months to at least once each refueling interval.
The proposed change to surveillance Requirement 4.7.10 e does not change the manner in which snubbers are tested, nor does it change the actions that are required in response to the discovery of failed snubbers.
In addition, the phrase "during shutdown" in Surveillance Requirement 4.7.10.e is being deleted.
Because of the terms
" Hot Shutdown" and
" Cold Shutdown" are defined in the Millstone Unit No. 3 Technical Specifications as operating modes or conditions, the added restriction to perform certain surveillances may be misinterpreted.
The proposed deletion of the term "during shutdown" is consistent with the recommendation of GL 91-04.
In GL 91-04, the NRC has concluded that the technical specifications need not restrict surveillances as only being performed during shutdown.
However, the NRC indicated that if the performance of a refueling interval surveillance during plant operation would adversely affect safety, the licensee should postpone the surveillance until the plant is shut down for refueling or in a condition or mode consistent with safe conduct of that surveillance.
NNECO believes that the deletion of the words "during shutdown" has no safety impact as long as the surveillances are conducted in any mode or condition without impacting the plant safety.
At Millstone Unit No.
3, two sample plans are used to test the four types of snubbers currently installed in the plant.
Type A (small mechanical), Type C (large mechanical), and
^
^
i_
'U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 23 June 29, 1995 Type D (large hydraulic) snubbers are tested to the "10 percent plan," while the Type B (medium mechanical) are tested to the "37 plan."
The 10 percent plan requires an initial test sample size of 10 percent of that type of snubber be tested.
Under the 37 plan, an initial test sample size of 37 anubbers is tested.
Each plan requires an additional sample equal to approximately one half the initial sample size be tested for.each identified failure.
Testing normally continues until no more failures are found or until all of the snubbers of that type are tested.
Because both of these test plans are self correcting in
- nature, each requiring additional testing when functional failures are identified, the increased test interval will have a
negligible effect upon the overall reliability of the snubber population.
Snubber testing experience at Millstone Unit No. 3 has shown that failure rates are not necessarily a direct function of the length of the test interval or snubber age.
During the i
third refueling
- outage, after an operating cycle of approximately 22
- months, the function testing program identified multiple small acchanical snubber (Type A) failures.
Following this refunling outage, extended mid-cycle maintenance shutdowns required the plant to request a test interval extension for snubbers.
NRC approval was ultimately obtained in early 1993 for an amendment to the Millstone Unit No. 3 Technical Specifications.
The amendment extended the functional test interval to 30 months by allowing a one-time extension to the 18 month snubber test i
interval.
An expanded sample of snubbers was tested during the fourth refueling outage.
Only one Type A snubber failed, j
even though the interval between the third and fourth refueling outages was approximately 30 months.
The fact that I
the snubber rate did not rise due to the extended interval supports the contention that failure rates are not a direct function of the snubber test interval.
The most recent industry guidelines concerning snubbers are contained in the ASME OM Code (1990),
Subsection ISTD, entitled, " Inservice Testing of Dynamic Restraints (Snubbers) in Light Water Reactor Power Plants."
This document requires that snubber testing be performed at refueling outages, rather than at a set monthly interval as required by the Millstone Unit No.
3 Technical Specifications.
The functional test program in this standard is designed to provide a 954 confidence level that 90% to 100% of the snubber population is operable.
It is essentially the same program that is contained in the Millstone Unit No.
3
. Technical Specifications.
Although the OM Code is somewhat more complex with respect to failure mode grouping and corrective
- actions, it is less restrictive as far as
~
i U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 24 June 29, 1995 additional testing which could result from test failures.
Because both programs are basically equivalent, it can be concluded that an increase in the Millstone Unit No.
3's snubber test interval will not significantly impact our confidence level in the reliability of the snubber population.
As such, the effect upon overall plant safety will be negligible.
A PRA review concluded that the proposed change is not risk significant.
This determination is reinforced by the results of piping stress analyses which have been performed to assess the impact,of snubbers which have failed to meet functional test acceptance criteria.
The results to date have shown that neither piping system functionality nor structural integrity have ever been compromised.
The Millstone Unit No.
3 Technical Specifications also require that the service life of snubbers be monitored in order to ensure the service life is not exceeded prior to the next surveillance interval.
Therefore, snubber maintenance records will be reviewed on a time frame which is consistent with the 24 month operating cycle.
These reviews will ensure that snubber service life will not be exceeded prior to the next scheduled review.
Bionificant Hazards Consideration NNECO has reviewed the proposed change in accordance with 10CFR50.92 and concluded that the change does not involve a SHC.
The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised.
The proposed change does not involve a SHC because the change would not:
1.
Involve a
significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to Surveillance Requirement 4.7.10.e of the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying snubber operability.
The frequency will be changed from at least once per 18 months during shutdown to at least once each refueling interval (i.e.,
nominal 24 months).
This proposed change will not have a significant impact on the overall reliability of the snubber population.
This is due, in part, to the fact that the snubber test plans are self correcting.
As functional test failures tre identified, additional snubbers are required to be tested.
Thus, the reliability of the snubber population is maintained.
The proposed change does not alter the intent or method by which the surveillances are conducted, does not involve any
U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 25 June 29, 1995 physical changes to the plant, does not alter the way any structure,
- system, or component functions, and does not modify the manner in which the plant is operated.
As such, the proposed change in the frequency of Surveillance Requirement 4.7.10.e will not degrade the ability of the snubbers to perform their safety function.
Additionally, Surveillance Requirement 4.7.10.1 requires that snubber service life be monitored to ensure that the service life is not exceeded between surveillance inspections.
Therefore, snubber maintenance records will be reviewed on a time frame which is consistent with the nominal 24 month operating cycle.
2.
Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change to Surveillance Requirement 4.7.10.e of the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying snubber operability.
The frequency will be changed from at least once per 18 months during shutdown to at least once each refueling interval (i.e.,
nominal 24 months).
The proposed change does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure,
- system, or component functions, and does not modify the manner in which the plant is operated.
As such, the proposed change cannot create the possibility of a new or different kind of accident from any previously evaluated.
3.
Involve a significant reduction in a margin of safety.
The proposed change to Surveillance Requirement 4.7.10.e of the Millstone Unit No. 3 Technical Specifications extends the frequency for verifying snubber operability.
The frequency will be changed from at least once per 18 months during shutdown to at least once each refueling interval (i.e.,
nominal 24 months).
This proposed change will not have a significant impact on the overall reliability of the snubber population.
This is due, in part, to the fact that the snubber test plants are self correcting.
As functional test failures are identified, additional snubbers are required to be tested.
Thus, the reliability of the snubber population is maintained.
The proposed change does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure,
- system, or component functions, and does not
l U.S. Nuclear Regulatory Commission B15225/ Attachment 4/Page 26 June 29, 1995 modify the manner in which the plant is operated.
As such, the proposed change in the frequency of Surveillance Requirement 4.7.10.e will not degrade the ability of the snubbers to perform their safety function.
Additionally, Surveillance Requirement 4.7.10.1 requires that snubber service life be monitored to ensure that the service life. is not exceeded between surveillance inspections.
Therefore, snubber maintenance records will be reviewed on a time frame which is consistent with the 24 month operating cycle.
0 Paga 1 of 7 m
A.
Components covered by Technical specification'4.6.4.2.b component Description 3HCS-PT1A Hydrogen Recombiner Inlet Gas Pressure Transmitter 3HCS-PI1A Hydrogen..
Recombiner Inlet Gas Pressure-
[
Indicator 3HCS-PT2A Hydrogen Recombiner. Return Gas Pressure
-Transmitter 3HCS-PI2A Hydrogen Recombiner Return Gas.
Pressure Indicator 3HCS-FT1A Radiant Heater Differential Pressure and Low l
Flow Alarm Pressure Transmitter 3HCS-FI1A-Radiant Heater Differential Pressure and Low Flow Alarm Pressure Indicator 3HCS-FSIA Radiant Heater Differential Pressure and Low Flow Alarm Pressure Switch 3HCS*TS1A Recombiner Temperature and Air Blast Heat Exchanger Control Temperature Switch 3HCS*TS3A Radiant
- Heater, Recombiner and Reaction Chamber Temperature Indication, control and Alarm. Temp Switch 3HCS* TIS 6A Radiant
- Heater, Recombiner and Reaction Chamber Temperature Indication, Control and Alarm Temp Ind. Switch 3HCS* TIS 7A Radiant
- Heater, Recombiner and Reaction Chamber Temperature Indication, Control and Alarm Temp Ind. Switch 3HCS*TS4A1A Radiant Heater and Reaction
-Chamber Temperature Control and ' Alarm Temperature Switch 3HCS-TS4A1B Radiant-Heater and Reaction
' Chamber Temperature Control and Alarm Temperature Switch 3HCS-TSSA1A Radiant Heater and Reaction Chamber Temperature Control and Alarm Temperature Switch 3HCS-TSSA1B Radiant Heater and Reaction Chamber Temperature Control and Alarm Temperature Switch
p'-;,,.
- Pag 3_2 cf 7
+-
s. o.-
Description 3HCS*TICSA1 Radiant.
Heater and Reaction-Chamber Temperature control and Alarm Temperature Indicating Controller 3HCS-PT1B Hydro Recombiner Inlet Gas Pressure-Transmitter 3HCS-PI1B Hydro Recombiner Inlet Gas Pressure Indicator 3HCS-PT2B Hydrogan Recombiner Return Gas Pressure l
Transmitter 3HCS-PI2B Hydrogen Recombiner Return Gas Pressure l
Indicator l
1 3HCS-FT1B Radiant Heater Differential Pressure and Low Flow Alarm Pressure Transmitter 3HCS-FI1B Radiant Heater Differential Pressure and Low-Flow Alarm Pressure Indicator
)
3HCS-PS1B Radiant Heater Differential Pressure and Low Flow Alarm Pressure Switch 3HCS*TS1B Recombiner Temperature and Air ' Blast Heat Exchanger Control Temperature Switch 3HCS*TS3B Radiant
- Heater, Recombiner and Reaction Chamber Temperature Indication, Control and Alarm Temp switch 3HCS* TIS 6B Radiant
- Heater, Recombiner and Reaction Chamber Temperature Indication, control and Alarm Temp Ind Switch 3HCS* TIS 7B Radiant
- Heater, Recombiner and Reaction Chamber Temperature Indication, Control and Alarm Temp Ind Switch-3HCS*TS4B1A Radiant Heater and.
Reaction Chamber Temperature Control and Alarm Temperature Switch 3HCS-TS4B1B Radiant Heater and Reaction Chamber Temperature Control and Alarm Temperature Switch 3HCS-TSSB1A Radiant Heater and Reaction Chamber Temperature Control and Alarm Temperature Switch
- -i Prgs 3 of 7 Connonent Description 3HCS-TS5BlB Radiant Heater and Reaction Chamber Temperature Control and Alarm Temperature Switch 3HCS*5IC5B1 Radiant Heater and Reaction Chamber Temperature control-and Alarm Temperature Indicating Controller 1
1 i
i I
J r
.. L*
2*
Paga 4 of 7 B.
Componente Covered by Technical specification 4.6.4.2.o & 4
- .c.
Description 3HCS*RBNRIA
- Hydrogen Recombiner Unit 3HCS*E1A Radiant Heat 3HCS*RBNR1B Hydrogen Recombiner Unit 3HCS*ElB Radiant Heat C.
Components Covered by Technical Specification 4.6.4.2.e & f.
Description 2-3HCS*RBNRIA Train A DBA Hydrogen Recombiner 3HCS*RBNR1B Train B DBA Hydrogen Recombiner 3HCS*C1A Train A Positive Displacement Blower 3HCS*C1B Train B Positive Displacement Blower D.
Components Covered by Technical Specification 4.7.1.2.1.c Component Description 3FWA*P1A Motor-Driven Auxiliary Feedwater Pump 3FWA*P1B Motor-Driven Auxiliary Feedwater Pump 3FWA*P2 Turbine-Driven Auxiliary Feedwater Pump
P ga 5 of 7 E.
Components Covered By Technical Specification 4.7.3.b.1 Component Descrintion 3CCP*AOV10A Non-QA Header Supply Isolation Valve 3CCP*AOV10B Non-QA Header Supply Isolation Valve 3CCP*AOV19A Non-QA Header Return Isolation Valve 3CCP*AOV19B Non-QA Header Return Isolation Valve 3CCP*MOV45A Containment Isolation Valve 3CCP*MOV45B Containment Isolation Valve j
3CCP*MOV48A Containment Isolation Valve 3CCP*MOV48B Containment Isolation Valve 3CCP*MOV49A Containment Isolation Valve 3CCP*MOV49B Containment Isolation Valve 3CCP*AOV179A Component Cooling Cross Connection Valve 3CCP*AOV179B Component Cooling Cross Connection Valve 3CCP*AOV180A Component Cooling Cross Connection Valve 3CCP*AOV180B Component Cooling Cross Connection Valve 3CCP*AOV194A Non-QA Header Return Isolation Valve 3CCP*AOV194B Non-QA Header Return Isolation Valve 3CCP*AOV197A Non-QA Header Supply Isolation Valve 3CCP*AOV197B Non-QA Header Supply Isolation Valve 3CCP*MOV222 Train A CAR Fan Isolation Valve 3CCP*MOV223 Train A CAR Fan Isolation Valve 3CCP*MOV224 Train A CAR Fan Isolation Valve 3CCP*MOV225 Train A CAR Fan Isolation Valve 3CCP*MOV226 Train B CAR Fan Isolation Valve 3CCP*MOV227 Train B CAR Fan Isolation Valve
Paga 6 cf 7 Component Descrintion 3CCP*MOV228 Train B CAR Fan Isolation Valve 3CCP*MOV229 Train B CAR Fan Isolation Valve F.
Components Covered by Technical specification 4.7.3.b.2 Component Description 3CCP*P1A - Note 1 RPCCW Pump A 3CCP*P1B - Note 1 RFCCW Pump B 3CCP*P1C - Note 1 RPCCW Pump C Note 1 includes all electrical and control system components associated with autostart features.
l
'o Pngs 7 of 7 G.
Service Water Automatic Valves Covered Under Technical specification 4.7.4.b.1 Component Descrintion 3SWP*AOV39A/B Diesel Generator Heat Exchanger Valves 3SWP*MOV50A/B Reactor Plant Component Cooling Water Heat Exchanger Valves 3SWP*MOV54A/B/C/D Recirculation Spray System Heat Exchanger Valves 3SWP*MOV115A/B Circulating Water Pumps 3SWP*MOV71AB Turbine Plant Component Cooling Water System Heat Exchanger Valves 3WTC*AOV25A/B Chlorination l
H.
Components Covered by Technical Specification 4.7.4.b.2 Component paperietion 3SWP*P1A Service Water Pump A 3SWP*PIB Service Water Pump B 3SWP*P1C Service Water Pump C 3SWP*P1D Service Water Pump D 1
1
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