ML20101G664
| ML20101G664 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 06/05/1992 |
| From: | Chambers J GENERAL ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 9206260225 | |
| Download: ML20101G664 (54) | |
Text
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REACTIVITY CONTROL SYSTEM 3.1.5 DRAFT 3.1 PIACTIVITY CONTROL SYSTEM 3.1.5 centrol m d Scre Mcu-M 9ters -
o LCO 3.1.5 Each control rod scrar?. accumulator shall be OPEPABLE.
APPLICABILITY: FDDES 1 and 2.
t ACTIONS NOTE------- g Separat4 Condition entry is allowed for et.h contqcl rod scram accumulator.
3 a.---.
t CONDITION REQUIPID ACTICN COMPLETION TIFE A.
One or more control rod A.1 Verify all control rods Imed_ately -
scram accumulator (s) associated with inoperab]e inoperable accumulators are fully inserted.
E 4
A.2 Declare the associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> control rods inoperable.
)
B.
Required Action and B.1 Be in FDDE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Associated Conpletion Time not met.
SURVEI CE PEQUIREMENTS j
SURVEILLANCE
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FREQUENCY SR 3.1.5.1 verifyeachcontrolrodscramaccumula\\or 7 days I
pressura is 2 (18501 psig.
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Control Rod Scram Accumulators B 3.1.5 B 3.1.5 cent ml Red Scre Ar m 31aters P.AL.S BACKGROLUD The control rod scram accumulators are part of the Control Rod Drive (CRD) system and are provided to ensure the control rods scram under varying reactor conditions.
The control rod scram accumulators store sufficient energy to fully insert two control rods at any reactor vessel pressure. The accumulator is a hydraulic cylinder with a free-floating piston. The piston separates the water used to scram the control rods from the nitrogen which provides the required energy. The scram accumulators are necessary to scram the control rods within the required insertion tims of LCO 3.1.4," Control Rod Scram Times."
APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the control rod scram function are presented in References 1, 2, 3 and 4.
The Design Basis Accident (DBA) and tl.e transient analyses assume that all of the control rods scram at a specified insertion rate. OPEPABILITY of each individual-control rod scram accumulator, along with ICO 3.1,3,
" Control Rod OPERABILITY," and LCO 3.1.4, ensures that the scram reactivity-assumed in the design basis transient and accident analyses can be met..The existence of an inoperable accumulator may invalidate prior. scram time measurements and may result in the-complete loss of scram capability for the associated control reds.
The scram function of the CRD system, and therefore the OPERABILITY of the accumulators, protects the FCPR Safety Limit (see Bases for LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (FCPR)") and the 1% cladding plastic strain fuel design limit (tee Bases for ICO.
3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION PATE (APLHGR)"), which ensure that-no fuel damage will occur if these limits are not exceeded (see Bases for LCO 3.1.4).
Also, the scram function at low' reactor vessel pressure (i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accident. (see Bases for LCO 3.1.6, " Rod Pattern Control").
ABWR STS B 3.1.5-1 6/4/92 '12 :39 PM l
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Control Rod Scram Accumulators 3
B 3. l'. 5 l
DRAFT The Control Rod Scram Accumulators satisfy the reqairements of criterion 3 of the imC Policy 3
Statement.
i LCO l
t The OPERABILITY of the control rod scram accumulators is regaired to ensure that adegaate scram insertion capability exists when needed over the entire range of 2
reactor pressures. The OPERAB7LITY of the scram accu ulators is based on maintaining adequate j
accumulator pressure.
1 i
APPLICABILITY In FCDES 1 and 2, the scram function is reqaired for mitigation of DBAs and transients and, therefore, the j
scram _ accumulators must be OPERABLE to support the scram function.
In FODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations l
LCO 3.10.3 " Single Control Rod withdrawal-Hot Shutdown," and-LCO 3.10.4 " Single-Control Rod i
Withdrawa1-Cold-Shutdown," which provide adequate requirements for control rod scram accumulator l
OPERABILITY under these conditions. Requirements for scram accumulators in FODE 5 are contained in i
LCO 3.9.5, " Control Rod OPERABILITY-Refueling."
ACTIONS l
The Actions table is modified by a Note indicating l-that a separate Condition. entry is allowed for each control rod. This is acceptable since the Required Actions for each Condition provides appropriate _. _
compensatory action for each inoperable control-rod.
Complying with the Required Actions may allow-for continued operation, and subsequent inoperable control rods governed by subsequent Condition entry and application of associated Required Actions.
(
A.1. A.2 With one or more control rod' scram accumulators inoperable the scram function could be severely degraded, because the accumulators are the primary source of scram force for the control rods at all reactor pressures. Therefore,'it must be verified
-immediately that all control rods associated with inoperable _ scram accumulators are fully inserted. The associated control rods must also'be declared' ABWR STS-B 3.1.5-2
- 6/4/92 12:39 PM
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c Control Rod Scram Accumulators B 3.1.5 inoperaole within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The allowed'Ccopletion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable for Required Action A,2-considering the low probability of a DBA or transient occurring during the time the accumulator is inoperable. Additionally, an autom tic reactor scram function is provided on sensed low pressure in the CRD charging water header (see LCO 3.3.1.1, "RPS Instrumentation"). This anticipatory reactor trip protects against the possibility of significant-pressure degradation (and thus reduced scram force) concurrently in multiple control rod scram accumulators due to a transient in the CRD hydraulic system.
M If Required Actions and associated Ccapletion Times of Condition A are not met, the plant must be brought to a FODE in which the ICO does not apply. To achieve this status, the plant must be brought to at least FDDE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion' Time is reasonable, based on operating experience,-to reach the required plant conditions from full power conditions in an orderly m nner and without challenging plant systems.-
SURVEII17dCE REQUIRE E US 19 3.1.5.1 SR 3.1.5.1 requires that the accumulator pressure be checked every 7. days to ensure that adeq.: ate accumulator pressure exists to_ provide sufficient:
scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is-considered inoperable. The minimum accumulator pressure-of 1850 psig is well below the expected pressure of approximately 2150 psig (Ref. 2).
Declaring the accumulator inoperable when the minimum pressure _is not maintained ensures that significant-degradation.-in scram times does_not occur. 'The 7-day-Frequency.has been shown_to be acceptable through operating experience.and takes into account other indications available in the control rocm.
ABWR STS B 3.1.5-3 6/4/92 12:39 PM
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i; Control Rod Scram Accumulators 4-B 3.1.5 DRAFT 3
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ABWR SSAR, Section (5.2.2.2.2.2].
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ASWR SSAR, Section 15.4.1.
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RIPS Operating 1
DRAFT 3.4 PEACTOR CCOIR,'T SYSTEM 9
3.4.1 Rea m r Interna' P-s (PIPri Cperat ing LCO 3.4.1 At least nine RIPS shall be in operation,
)
2 With eight RIPS in operation THEPHAL POWER shall be s 95% RTP, 2
With seven RIPS in operation THER.% POWER a all be 5 90% RTP, 4
APPLICABILITY:
tODES 1 and 2.
AC"'!ONS CCtCITION REQUIRED ACTION CCMPLETION TIME i
A.
Eight RIPS in opera *. ion A.1
- tUIE --
with reactor power Provisions of ILO 3.0.4 i
are not applicable.
Restore to at least nine 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> RIPS in operation.
3 A.2 Reduce THEA % POWER to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 5 95% RTP.
B.
Seven RIPS in operation B.1 tUTE with reactor power Provisions of ILO 3.0.4
> 90% RTP.
are not applicable.
i Restore to at least nine 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> RIPS in operation.
2 B.2' Reduce THERMAL PCWER to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 5 90% RTP.
C.
Five or six RIPS in C.1 Reduce THEPMAL POWER to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> operation.
$ 25% RTP.
Hi2 C.2 Restore to at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from seven RIPS in operation.
discovery of less then seven RIPS in operation (continued)-
6/5/92 5:07 FM
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RIPS Operating 3.4.1 DRAFT
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CONDITION REQUIRED ACTION CCt@LETION TIFE i
D.
Four or less RIPS in D.1 Be in FDDE 2.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
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D.2 Restore to at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from j
seven RIPS in operation, discovery of less then seven 4
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RIPS in i
operation 1
E.
Required Action and E.1 Be in FODE 3, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, B, C or D not met.
4 i
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SURVEILLANCE REQUIPDENTS SURVEILIRICE FREQUENCY SR 3.4.1.1 Verify at least nine RIPS are in operation at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i
any TFEEGL TOWER level, O
With only eight RIPS in operation, verify THEMIAL POWER is 6 95% RTP, I
With only seven RIPS in operation, verify THERMAL POWER is 6 90% RTP.
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ABWR STS 3.4,1-2 6/5/92 5:07 FM
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I RIPS Operating a
B 3 4.1 DRAFT BASES I
B 3.4 FIAC7CR COO!M SYSTEM B 3.4.1 Reamr Internal Pms Cm ating l
BASES i
BACKGROUtO The Reactor Coolant Recirculation System is designed to provide a forced coolant flow through the core to remove heat from the fuel. The forced coolant flow remves more heat from the fuel than would be possible with just natural circulation. The forcad' flow, therefore, allows operaeion at significantly higher-power than would otherwise be possible The recirculation system also controls reactivity over a wide span of reactor power by varying the recirculation flow rate to cor. trol the void content of the moderator. The reactor coolant recirculation system consists of ten recirculation pumps internal to the reactor vessel.
Each reactor internal pump (RIP) includes a wet motor, an adjustable speed drive (ASD) to control pu.p speed, an external heat exchanger to -
cool the purrp motor, and associated instrumentation.
The pump motors protrude from the bottom of the reactor vessel into the lower drywell area and the motor casings are part of the reactor coolant pressure boundary. The pumps themselves are considered reactor-vessel internals.
The recirculated coolant consists of-saturated water -
from the steam separators and dryers that has been subcooled by incoming feedwater. This water passes-down the annulus between the reactor vessel, wall and the core shroud.
It then flows to the inlet of the t
'Ps that are located equidistant.around the plate (c
>un deck) for.Tling the bottom of the annulus area.
s The cotat core' flow passes through the RIPS into the area below-the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core.
l-The subcooled water enters the bottom of the fuel channels and contacts the fuel cladding, where. heat.is transferred to the coolant. As:it rises, the coolant
.begins to boil, creating steam voids within the fuel channel that continue until the coolant exits the core.
Because of reduced mderation, the steam ABWR STS B 3.4.1-1 6/5/92' 5:21 PM-
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RIPS Operating B 3.4.1 DRAFT BASES voiding introduces negative reactivity that must be compensated for to mintain or to increase reactor power. The recirculation flow control alicas operators to increase recirculation flow and sweep some of the voids train the fuel channel,. overcoming 3
the negative reactiv':/ void effect.
Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation 1.e.,
70 to 3 l0% RTP) without having to cove control-rods and distub desirable _ flux patterns.
Each RIP is mnually started from the mio control room. The AF1. provide recfalation cf individual RIP speed, and therefore flow. The flow through each RIP can be mnually'or automtically controlled.
APPLICABLE SAFETY MALYSES The operation of the Reactor Coolant Recirculation System is an initial condition assumed in the desi;n basis loss-of-coolant accident (ICCA) (Ref. 1).
During a ICCA, the operating RIPS are all assumed to trip at time zero due to a coincident loss of offsite power. The subsequent core flow coastdown will be innediate and rapid because of the relatively low inertia of the pumps. Powever, the RIPS are assumed to have sufficient flow coastdown characteristics to mintain fuel therml mrgins during abnormal operating transients-(Ref. 2), which are analyzed in Chapter 15 of the FSAR..For conservatism, no credit is taken for the increased inertia supplied by the two M/G sets. hat feed six of the RIPS.-.
With at least nine of the ten. RIPS _in operation the-IOCA analysis includes all potential power and flow operating points from which an event might te initiated. With eight or less RIPS in operation the-LOCA analysis-assu".ptions do not include all potential operating states so that additional restrictions are-necessary regarding reactor _ power based on the number l
of pu.ps actually operating, i
Reactor-internal pumps operating. satisfy Criterion 2 i
of the NRC Policy Statement.
ABWR STS B 3.4.1 6/5/92 5:21 PM
m. _
m RIPS Operating B 3.4.1 DR7 AFT BASES-LOO At least nine RIPS are reqaired to be in operation, or with onit/ seven or eight RIPS operating, reactor power is restricted to 90% and 95% RTP, respectively. This ensures that all potential initial power and flow operating states have been accounted for in either the ILCA or transient analysis.
AoPLICABILITY In MDDES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.
In FDDES 3, 4, and 5, the consequences of an accident are reduced and the flow and coastdown characteristics of the RIPS are not irportant, ACTIONS A.1, A.2 With eight RIPS in operation and the reactor power level greater than 95% RTP the assumptions of the LOCA and transient analyses are not met. Either a-condition of nine RIPS in operation must be restored, or else reactor power must be reduced to less than or equal to 95% RTP. A Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is specified in either case, based on engineering judgement considering the time required to reasonably complete the Required Action. As noted, the-provisions of ICO 3.0.4 are not applicable for this Condition to allow entry into FDDE 1 or 2 with less than nine RIPS in operation. This is acceptable because a' Condition of eight RIPS in operation.
provides sufficient core flow for all but very high power conditions.
B.1, 9.2 With seven RIPS in operation and the reactor power level greater.than 90% RTP the assumptions of the IOCA and transient analyses are not met. Either a condition of nine RIPS-in operation must be restored, I
ABWR STS B 3.4.1-3 6/5/92 5:21 PM
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i RIPS Oparating B 3.4.1 DRAFT i
BASES or else reactor power must be reduced to less than or
.squal to 90% RTP. A Completion Tine of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is specified in either case, based on engineering
}
judgement considering the time required to reasonably complete the Required Action. As noted, the provisions of LCO 3.0.4 are not applicable for this Condition. This is acceptable because a Condition of eight RIPS in operation provides sufficient core flow-for all but very high power conditions.
c.1, c.2 D.1, D.2 With less than seven RIPS operating the steady. state power and flow characteristics of the core have not been fully analyzed. Therefore, even at reduced power i
levels, continued operation is allowed for only a i
short time while an attempt is made to -restore at least seven pumps to operating status.
For the case i
of 5 or 6 RIPS in. operation, reactor power must be reduced to less than 25% RPT because of potential long term stability concerns.
-A Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is specified, based on engineering judgement considering the time required to reasonably complete the Required Action. With less than 5 RIPS operating j
the unit must be brought to FDDE 2 due to the lack of detailed analysis of the actual flow distribution with i
less than half of the RIPS in operation providing i
forced flow at higher power -levels. A Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is specified,- based on engineering.
?-
judgement considering the time required to reasonably ~
complete the Required Action. Furthermore, in each-case a condition of at' least seven RIPS in operation
[
must be restored such that the unit is-returned to conditions that-have been fully analyzed for long term power operation.
-A Ccepletion Time of-12 hours'is specified from the time ~1t is first discovered that there are less than seven RIPS in operation. This is based on the low prrbability of a design basis occurring during tht time period and because the l
potential consequences of such have been substartially; reduced by the concurrent reduction in reactor power level.
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With the Required Action and associated Completion Time of Conditions A, B, C or D not met,.the unit'is I
required to be brought to a FIE in which the LCO does not apply. To achieve this status the plant must be brought to BODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
In this condition, AWR STS B 3.4.1-4 6/5/92 5:21 PM
~
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RIPS Operating-1 B 3.4.1 1
DRAFT 1
BASES j
the RIPS are not required to be operating because of
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the reduced severity of DBAs and minimal dependence on forced flow characteristics, The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach FDDE 3 from full power conditions j
in an orderly manner and without challenging plant systems.
i i
SURVEILIANCE PEQUIREF22 TIS j
This SR ensures that the number of RIPS in operation i
and the corresponding reactor power level is consistent with the assumptions of the applicable i
design bases analyses and with core power and flow characteristics that have been analyzed for long term.
operaticn. This Surveillance is required to be performed once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Operating experience j
with previous Bh3 designs-has demonstrated that a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency for this type of surveillance is adegaate.
l l
REFERENCES i
j.
1.
ABh3 SSAR, Section (6.3.3.7.1].
i 2.
ABh3 SSAR, Section ' [5.5.1.5].
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3.5.1 DRAFT 4
1 3.5 EbERGE!JCY CORE CO2LItJG SYSTEMS (ECCS) 3.5.1 ECCS - Cperating LCO 3.5.1 Each ECCS injection subsystem and the Automatic Depressurization System (ADS) function of [eight] S/RVs shall be OPERABLE.
I APPLICleILITY:
!CCE 1, FDDES 2 and 3, except ADS valves are not reqaired to be OPEFABLE with reactor steam dme pressure 5 50 psig and RCIC is not required to be OPERABLE with reactor steam dm e pressure $ 150 psig.
l ACTIOt?S COtDITION RECUIED ACTIOt1 CCt@LETION TIME A. One ECCS injection A.1 Restore ECCS injection 30 days subsystem inoperable in subsystem to CPERABLE 4
any division.
status.
B. Two ECCS injection B.1 Restore one ECCS 14 days subsystema! inoperable, injection. subsystem to
+
each in a different OPERABLE status division.
I C. Both ECCS injection C.1 Restore one ECCS 7 days subsystems inoperable in injection subsystem to any one division.
OPERABLE status D. Three iajection D.1 Restore one inoperable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> subsystems inoperable, subsystem to OPERABLE each in a different status.
division.
4 E. Required Action and E.1 Be in BODE 3 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, B, L4 C or D not ret.
E.2 Be in FODE 4.
36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />s-(continued) 1 ABWR STS 3.5.1-1 6/4/92 3:59 PM
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ECCS-Operating 3.5.1 DRAFT l
CONDITION REQUIPID ACTION CCMPETION TIME F. Three injection F.1 Enter If0 3.0.3 Innediately subsyste:ns inoperable, two of which are in the j
sa e division, 3
i Four or nore injection i
subsystems inoperable.
3 G.
---NOTE G.1
--NC/TE l
This Condition nuy exist Provisions of 140 3,0.4 concurrently with are not applicable.
Conditions A through D.
Restore ADS valve to Prior to entry i
One ADS valve OPERABE status.
into BODE 2 inoperable.
following next 4
FDDE 5 entry.
H.
NOTE-H.1 Verify at least two high Intnediately l
This Condition nay exist pressure ECCS injection concurrently with subsystems are OPEFABE.
Conditions A through D.
,RQ Two ADS valves H.2 Restore one ADS valve to 30 days inoperable.
OPERABE status.
r I. Three or more ADS valves I.1 Be in Mode 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.
RC i
2 l
I.2 Reduce reactor steam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Required Action and dome pressure to 5 50 j
associated Completion psig.
Tine of Condition H not-met.
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l ABYR STS 3.5.1-2 6/4/92 3:59 PM
v ECCS-Operating 2
3.5.1 l,
DRAFT i
SURVEILLANCE PECCIREFENTS A
SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify for each ECCS injection subsystem tne 31 days l
piping is filled with water from the pump a
discharge valve to the isolation valve.
!ME--- - - -
LPFL subsystems nay be considered OPERABLE during alignment and operation for decay heat re:rval when below 135 psig in BODE 3, if capable of being nanually realigned and not -
otherwise inoperable.
l Verify that each ECCS subsystem manual, power-31 days cperated and autccatic valve in the flow path, that is not locked, sealed or otherwise secured in position, is in its correct position.
SR 3.5.1.3 Verify ADS (air receiver] pressure 2 161 psig.
31 days SR 3.5.1.4 verify each ECCS pu:rp (except for RCIC) develops In accordance the spacified flow rate against a system head with the corresponding to the specified reactor pressure:
Inservice Testing Program or 92 days SYSTEM HEAD i
CORRESPONDITE
'IO PIACTOR
[
SYETEM FLOW RAE
_EEES9M OF -
LPFL 2 4200 gpm 2-
-40 psig HPCF 2 800 gpm 2 1177 psig-SR 3.5.1.5
- inrE Not required to be performed until 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s-dfter reactor staAm-dome pressure is 2 920 psig.
Verify, with RCIC steam stcply pressure 5-(
)
92 days psig and 2 (
) psig, the hCIC pu:rp can develop a flow rate 2 800 gpm against a system head corresponding to reactor pressere.
l
-(continued)
ABWR STS 3.5.1-3 6/4/92 3:59 PM er e
r s A env e
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ECCS-Cperating 3.5.1 DRRFT l
I l.
SURVEILIRJCE FREQUDJCY SR 3.5.1.6
- --NTE--
- = - -
I Not recyired to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter teactor steam dcce pressure is 2150 psig.
i Verify, with RCIC steam supply pressure 5 [
]
18 months i
psig, the RCIC pu.~p can develop a flow rate i
2 800 gpm against a system head corresponding to l
reactor pressure.
--t M E l
Vessel injection may be excluded.
i j
verify each ECCS subsystem actuates on an actual 18 months or simulated automatic initiation signal.
1 4
- tmE Valve actuation nay be excluded.
Demonstrate that the ADS actuates on an actual 18 months or simulated automatic initiation signal.
1 4
SR 3.S.1.9
=-
NOTE Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam dome pressure is 2- [ ] psig.
Demonstrate that each ADS valve opens when 18 months, on canual.ly actuated at reactor steam dome pressure a STAGGERED 2[
] psig.
TEST BASIS for each valve I
solenoid e
s ABWR STS 3.5.1-4 6/4/92 3:59 FM
~, _.....
ECCS System B 3.5.1 DRA.FT BASES B 3.5 EMERGENCY CORE CCOLING SYSTEMS B 3.5.1 Eccs - % ratina 2
f EASES 4
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BACKGRGRO i
The ECCS is designed, in conjunction with the primary and secondary containment, to limi* the release of radioactive naterials to the et.viroment following a loss-of-coolant accirient (ICCA) :
The ECCS directs water to both inside and outside the core shroud to cool the core during a IOCA. The ECCS network is conposed of the High Pressure Core Flooder (HPCF) system, the Reactor Core Isolation Cooling.(RCIC) system, and the Low Pressure Flooder (LPFL) mode of the Residual Heat Removal (RHR) system.
he ECCS also consists of the Automatic Depressurizath System (ADS). The suppression pool provides the ;equired source of water for the ECCS. Although no credit is 4
taken in the safety analyses for the condensate storage tank (CST), it is capable of providing a 4
source of water for the RCIC System and the HPCF subsystems.
On receipt of an initiation signal, ECCS pumps autoratically start; simultaneously the system aligns and the pumps inject water, taken either from the CST -
or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcme ~by the discharge pressure of the ECCS pumps. Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if tr.e system
{
is not needed. The discharge. pressure of each of the HPCF pu ps alme.9t irrrnediately exceeds th t of the RCS, a
l
- and the punps inject coolant into the sparger above i
the core. Cace the steam d G an RCIC turbine has
- accelerated, the RCIC pump discharge pressure also j
quickly exceeds that of the RCS, and the pump injects coolant into the reactor pressure vessel (RPV) via one
?
of the feedwater. lines.
If the break is small,'RCIC 1-or either of the HPCF pumps will maintain' coolant
- inventory while the RCS is still pressurized and, thus, vessel level.
If RCIC and HPCF *sil, they are backed up by ADS in combination with LPCF.
In this-,
event, the ADS timed sequence would be allowed to' time out and open the selected safety / relief valves l
ABWR B 3.5-1 C/5/92 6:15 PM il y
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(
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.g py
i ECCS System B 3.5.1 DRAFT RASES 4
(S/RVs), depressurizing the RCS and allowing the LPCF to overcome RCS pressure and inject coolant into the vessel.
If the break is large, RCS pressure initially-i drops rapidly, and the LPCF subsystems cool the core.
l~
Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool is circulated through the RHR heat exchangers ccoled by the Reactor Cooling Water (RCW)
System. Depending on the location and size of the break, portions of the ECCS may be ineffective; however, the overall design is effective in cooling j
the core regardless of the size or location of the piping break.
4 The RCIC System ic also designed to operate either 4
automatically or manually following RPV isolation accompanied by a loss of coolant fbw from e
i feedwater system to provide adequate core cooling and
- l l_
control of FPV water level. Under these conditions, the HPCF and RCIC systems perform similar functions.
The RCIC System design requirements ensure that the criteria of Reference 1 are satisfied.
The ECCS injection systems are arranged in three i
separate divisions each comprised of a high pressure i
and low pressure subsystem, ECCS Division 1 consists of the RCIC system and LPFL-A. ECCS Division 2 consists of LPFL-B and HPCF-B.
ECCS Division 3 consists of HPCF-C and LPFL-C.
All ECCS subsystems are designed to ensure that no-j -
single active component failure will prevent automatic j
initiation and successful operation of the minimum required ECCS subsystems.
i LPFL is an independent operating mode of the RHR-system. There are three LPFL subsystems. Each LPFL' subsystem (Ref. 2) consists of a_ motor-driven pump,-a heat exchanger, piping and valves to transfer water from the suppression pool to the reactor vessel. Each-LPFL subsystem has its own suction and discharge piping. The water is injected into the' reactor vestal outside tne core shroud, via feedwater line B for LPFL subsystem A, and via dedicated LPFL subsystem inlet j
no :les and spargers for LPFL subsystems B and C.
The
-LPFL subsystems are designed to provide core cooling at low reactor vessel pressure. Upon receipt of an
?
initiation signal, each LPFL pump is aut( matically i
a ABWR B 3.5 6/5/92 6:15 FM i
1 m
m-m
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-s,,.,,,,,,p.
.,-,,,,,,g,,,y.
,e
%y,,ypp,%~
ECCS System
+
B 3.5.1 DPM T BASES started (approximtely (
) seconds after AC power is available). When the RPV pressure drops sufficiently, LPFL flow to the RPV begins. PRR system valves in the LPFL flew path are auto m tically positioned to ensure the proper flow path for water from the suppression-pool, through the FRR heat e.xchanger, to inject into the RPV. A discharge test line is provided to route water from and to the suppression pool to allow-testing of each LPFL pump without injecting water into the RPV.
There =re two HPCF subsystems. Each HPCF System (Ref, a) consists of a motor-driven pump, a flooder sparger above the core, and piping and valves-to transfer water from the suction source to the sparger.
D suction piping is provided from the CST and the suppression pool.
Pump suction is norm lly aligned to the CST source to minimize injection-of suppression pool water into the RPV.
If the CST water supply is low, however, or the suppression pool level is high, an automtic transfer to the suppression pool water source ensures a water supply for continuous operation of the HPCF System. The HPCF System is designed to provide core cooling over a wide range of RPV pressures (0 to 1177 psid, vessel-to suction source).
Upon receipt of an initiation signal, the HPCF pump autemtically starts (from AC power) and valves in the flow path begin to open.
Since tha HPCF System is designed to operate over the full rar.ge of expected RPV pressures, HPCF flow begins as soon as the necessary valves are open. -A full flow test line is provided to route water from and to the suppression pool to allow testing of the HPCF System during norm l operation without injecting water into the RPV.
The RCIC System consists of a steam-driven turbine-pump unit, piping and valves to provide steam to the turbine, as well as piping and valves to transfer a
water from the suction source to the core via the feedwater system line.
Suction piping is provided from the CST and the suppression pool. Pump suction is normily aligned to the CST to minimize injection of suppression pool water into the RPV.
If the CST water supply is low or the suppression.pcol level;is.
1 high, however, an automatic transfer to the suppression pool assures a water supply for' continuous operation of the RCIC System. The steam supply to the turbihe'is piped from main steam line B, upstream of the inboard main. steam line isolation valve (Ref. 4).
ABWR B 3.5-3 6/5/92 6:15-PM A
ECCS System B 3.5.1 DPJWT PAFES Tha RCIC System is designed to prcvide core cooling for a wide range of reactor pressuras, (
) to
[
] psig. Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed.
hs the RCIC flew increases, the turbine cor. trol valve is autoratically adjusted to maintain design flow.
Exhaust steam from the RCIC turbine is oischarged to the suppression pool. A full flow test line is provided to route water from and to the suppression pool to allow testing of the RCIC System during nomal operation without injectir.g water into the RPV.
The ECCS pumps e e provided with minimum flow bypass lines, vhich discharge to the suppression pool. The valves in these lines autceanically open to prevent pu:p damage due to overheating when other discharge line valves are closed or RPV pressure is greater than the LPFL pump discharge prescures following system initiation, To ensure rapid delivery of water to the RPV and to minimize water ha:rmer effects, the ECCS discharge line " keep fill" systers are d3 signed to raincain all pu:p discharge lines filled with water.
The ADS (r'.ef. 5 consists of 8 of the 18 SRVs.
It is designed to provide depressurization of the prirary system during a small b' ak IJ?CA if HPCF and RCIC rail or are unable to maintain requited wati level in the PTV. ADS operation reduces the RPV pressure to within the cperating pressure range of the low pressure ECCS subsystems (LPFL), so these subsystems can provide core eccling. F.wh ADS valve is supplied with i
pneu:ratic power from a nitrogen accumulator located in the drywell. The atmospheric control system (AC3) supplies the nitrogen necessary to both directly actuate tne ADS valves unoer nonral conditions (when pneunatic power from the accumulators is not needed),
and to ensure the accumulators remain charged for use in emergency actuation.
5 APPLICABLE S UETY ANALYSES The ECCS perfomance is evaluated for the entire spectrum of break sizes for a postulated IOCA. The accidents for which ECCS operation is required are presented in References 6, 'i and 8.
The required analysea and assu~ptions are defined in 10 CFR 50 (Ref. 9), and the results of these analyses are described in Reference 10.
m ECCS System
-B 3.5.1 DRPET BASES This LCO helps.to ensure that the following accept.ance criteria for the ECCS, established by 10 CFR Sv.46 (Ref. 11), will be ret following a WCA assuming the worst case single active ccqx:nent. failure -in the ECCS:
Maximum fuel element cladding temperature is a.
5 2200 F; b.
Maximum cladding oxidation is 5 0.17 times the total cladding thickness before oxidation; c.
Kuimum hydrogen generation from :irconium-water reaction is 5 0,01 tir.es the nypothetical amount that <>uld be generated if all of the metal ~in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volune, were to react; d.
The core is maintained in a coolable geometry; and Adeq: ate long-term cooling capability is e.
maintained.
The limiting single failures are discussed in Reference 12. For any WCA,.. failure of ECCS subsystems in Division 2-(HPCF-B and LPFL-B) or I
Division 3 (HPCF-C and LPFL-C) due to. failure of the associated diesel generator is the most severe failure. One '.DS valve failure is aaalyzed as the single failure for events requiring ADS operation, however, the above single failure of a diesel-generator and the associated motor driven ECCS injection subsystems in that division, is a mre limiting failure. The remaining OPEPABLE ECCS subsystems provide the capability to. adequately cool
- the core and prevent excessive fuel damage. An additional function of the RCIC system is to respond to transient events by providing makeup coolant-to the reactor vessel.
Further "best estimate" ECCS analyses were performed to support the success criteria utilired in the ABhR PRA (Ref. 13). These analyses were performed to
- determine the minimum amount' of ECCS equipment tnat must operate for the plant to still treet the 10.CFR 50.46 acceptance criteria. listed above,.but using more-realistic analysis assurrptions. The analyses were performed using the sane calculational methods as were..
used for the design basis analyses, with the'following exceptions:
ABwR B 3.5-5 6/5/92
-6:15 PM-i i
ECCS System B 3.5.1 DRAFT g ES a.
Reactor scram and ECCS actuation occur on high drywell pressure; b.
Events are initiated from nontal reactor water level; c.
A nominal decay heat curve is utilized; d.
A realistic critical flow nodel (Maody Homogeneous) is utilized for the assumed break; anQ
)
e.
Re-wet occurs at a AT of 450 F.
Thesc "best estimate" analyses demonstrated that
" success" (i.e., no violation of the above stated 50.46 lindts) under postulated accident scenarios was achieved provided any one motor driven ECCS injection subsystem was available for injection into the RPV.
For scenarios requiring depressurization, " success" was achieved with the availability of three SR/Vs (whether actuated in the ADS mede or otherwise).
Thus, under these more realistic analyses conditions, the ECCS is -able to perfor.a its intended safety function, even under various situations with some eqaipment initially out of service or unavailable due to multiple postulated failures.
The ECCS satisfies Criterion 3 of the NRC M licy Statenent.
Ifos Each ECCS injection suosystem and (eight] ADS valves are required to be OPERABLE. The ECCS subsysters are defined as the three LPFL subsystems, the two HPCF subsystems, and the RCIC System. The high passure ECCS subsysters are defined as the RCIC System and the two HPCF subsystems.
With fewer than the required number of ECCS subsysters-OPERABLE during a limiting design basis IOCA concurrent with the worst case single failure,,the margins to the limits specified in 10 CFR 50.46 (Ref. 11) would be reduced and in sone cases such limits could potentially be exceeded. All ECCS subsysters must therefore be OPERABLE to be consistent with the design basis analyses (Ref.10) that have been performed to satisfy the single failure criterion ABWR B 3.5-6 6/5/92 6:15 P!4
ECCS System B 3.5.1 DRAFT BASES required by 10 CTR 50.46 tref. 11). The ECCS is supported by other syste.Ts that provide automatic ECCS initiation signals (Ir0 3.3.5.1/2, " Emergency Core Cooling System (ECCS) Instrumentation / Actuation Logic"), cooling ano service water to cool rooms containing ECCS eqaipment (LCO 3.7.1, " Ultimate Heat Sink," and LCO 3.7,2, " Reactor Cooling Water (RCW)
System / Reactor Service Water (kSW) System"), and electrical power (LCO 3.8.1, "AC Sources--operating,"
and LCO 3.8.4, "DC Sources--operating").
The OPERABILITY of the RCIC System further provides adequate core cooling such that actuation of any of the remaining ECCS subsystems is not reqaired in the event of RPV isolation accompanied by a loss of feedwater flow. The RCIC has sufficient capacity to maintain RPV ire Mry during an isolation-event.
A LPFL subsysteA my be considered OPERABLE during alignment and operation for decay heat renoval when below the actual RHR cut-in permissive pressure in FODE 3, if capable of being manually realigned (renete or local) to the LPFL mode and not otherwise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsyste.Ts can provide the required core cooling, thereby allowing cperation of an RHR shutdown cooling loop when necessary,
)
APPLICABILITY All ECCS subsystems are reqaired to be OPERABLE during BODES 1, 2, and 3 when there is considerable energy in the reactor core and core cooling would be required to prevent fuel danage in the event of a break in the primary r.ystem piping. Further,-the RCIC System is required to be OPERABLE since it is the primary water source for core cooling when-the reactor is-isolated-and pressurized.
In FDDES'2 and 3 with reactor steam dome pressure < 150 psig, RCIC is not required to be OPERABLE since the other ECCS subsystems can provide sufficient flow to the vessel. In FDDES 2:and 3, the
--ADS function is not required when pressure is-s,150 psig, because the low pressure ECCS subsystems (LPFL) are capable of providing flow into the RPV below this-pressure. ECCS requirements for BODES 4 and 5 are-specified in ICO 3.5.2, "ECCS-Shutdown" t
-ABWR B 3.5-7 6/5/92 6:15 PM
ECCS System B 3.5.1
+
DRAFT BASES ACTIONS Ad.
If any one ECCS injection subsystem is inoperable,.the inoperable subsystem must be restored to OPERABLE status within 30 days.
In this Condition, the reraining OPERABLE subsystems provide adequate core cooling during a lOCA.
However, overall ECCS reliability is reduced, and a single failure in one of the remining CPEPABLE subsystems concurrent with a IOCA r.ay result in the ECCS not being able to perform its safety function consistent with traditional design basis analyses. Nonetheless, even given a single failure in one of the remining OPEPABLE subsystems concurrent with a LOCA, there will always be at least one ECCS subsystem available to inject water into the PPV. Analyses using "best estimate" assuntions.
(Ref. xx) show that this situation is acceptable '. rom an overall risk perspective. The 30 day Completion Time is thus based on the overall redundancy previded -
by the ECCS and its continued ability to perforan its intended safety function, while assuring a return to full ECCS capability in a-reasonable time so as to not significantly inpact overall ECCS reliability.
Ed.
With two ECCS injection subsystems inoperable, each belonging to a different divi,~on, at least one ECCS injection subsyst e ust be restored to OPERABLE stav's within 14 t'
In this Condition, the reraim.ng OPERABL. absysters provide adequate core cooling during a IOCA. However, overall ECCS reliability is reduced, and a single failure in one of the re.Taining C/ERABLE subsystems concurrent with a IOCA ray result in the ECCS not being able to perfom its safety function consistent with traditional design basis analyses. Nonetheless, even given a-single failure in one of the remaining OPEPSBLE subsystems concurrent with a IOCA, there will always be at'least one ECCS subsystem available to inject water into the-RPV. Analyses using "best estimate" assumptions.
(Ref. 13) show that this situation is acceptable from an overall risk perspective. However, since the ECCS availability is reduced relative to Condition A, a more restrictive Conpletion Time is imoosed. The 14 day Crepletion Time is based on the overall redundancy prc" Med by the ECCS and its continued ability to ABWR B 3.5-8 6/5/92 6:15 PM
ECCS System B 3.5.1 DRATT BASES perform the intended safety function, while assuring a return towards full ECCS capability in a reasonable
- ime so as to not significantly 1:rpact overall ECCS reliability.
Cd With two ECCS injection subsyste:ts inoperable, both belonging to the same division, at least one ECCS injection subsystem must be restored to CPERABLE status within 7 days.
In this Condition, the remaining OPEFABLE 2bsystems_ provide adequate core cooling during a 10CA. However, overall ECCS reliability is reduced, and a single failure in one of the re: raining OPERABLE subsystems concurrent with a LOCA may result in the ECCS not being able to perform its safety function consistent with traditional design basis analyses. Nonetheless, during a IOCA there will always be at least-one ECCS subsystem available to inject water into the RPV, and in most postulated scenarios this will be the case even given a single failure in one of the remaining OPERABLE subsystems.
Analyses using "best estimate" assumptions (Ref. 13) show that this situation is acceptable from an overall risk perspective. -However, since the ECCS availability is reduced relative to Condition A and B, and because ccmplete single failure cap 2 ility is not retained for all cases, a more restrictive Completion Time is iriosed.
The-.7 day Completion Time is based.
on the overall (but reduced) redundancy provided by t
the ECCS and its continued ability to perform the intended safety function, while assuring a return towards full ECCS capability in a reasonable time so as to not significantly impact overall ECCS reliability.
D.d With three ECCS injection subsystems inoperable, each belonging to a different division, at least one ECCS injection subsystem must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
In this Condition,-the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However,- overall ECCS eliability is reduced, and a single failure in one of
+Se remaining OPERABLE subsystems concurrent with a MCA may result in the ECCS not being able to perform its safety function. Nonetheless, during a IOCA there will always be at least one ECCS subsystem available ABWR B 3.5-9 6/5/92 6:15 PM i
i i-ECCS System-
-B 3.5.1 DRAFT i
BASES
[
to inject water into the RPV, and.in rest postulated-scenarios this will be the case even given a single failure in one of the remining OPERABLE subsystems.
Analyses using "best estimate" assumptions (Ref. 13) show that this situation is acceptable from an overall i
risk perspective. However, since the ECCS availability is reduced relative to Conditions A, B and C, and because complete single failure capability-is not retained for all cases, a mare restrictive-4 i
Completion Time is imposed. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion
~
t Time is based on the overall (but rec'uced) redundancy-provided_ by the ECCS and its continued ability to perform the intended safety function, while assuring a return towards_ full ECCS-capability in a reasonable time so as to not significantly i:rpact overall ECCS reliability, i
l E.1 E.2 I
If any Required Actions and associated Completion Times of Condition A, B, or C are not met, the plant i_
must be brought to a FODE in which _the LCO _ does not apply. To achieve this status,~the plant must be; brought to at least FODE 3 'within 12 ' hours and in FODE 4.within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Corpletion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without i
challenging plant systems.
I 1
J When multiple ECCS subsystems are inoperable, as j
stated for Condition F, the plant is in a Condition outside of_the design basis accident analyses and one that cannot be supported by "best estimate" analyses or overall reliability arguments. Therefore, ICO:
l 3.0.3 must be entered imediately.
GL The.LCO requires [eight] ADS valves to be OPEBABLE to i.-.
provide the ADS function. Traditional design basis analysis (Ref.10). includes an evaluation of the effect of one ADS valve out of service. Per this' analysis, using conservative assumptions, operation'of only (seven] ADS valves will provide the required depressurization. Analyses using "best estimate" 7
d asst"tpt!ons (Ref.13), however, demonstrates that only-ABWR B 3.5-10 6/5/92 6:15 PM i
-.,y
_-,,m,
-m.,m-,--,-,.,_...,,,,m.,,_.
. - - -., _ - ~,... -
.,.,..-.-~,,,...,-.,~.-,.....d.~,,L,,.~m
4 ECCS System i
B 3.5.1 DRA' T l
BASES-three ADS valves a 3 required for the ADS to i
st.;cessfully perform its function. Therefore, with one ADS valve inoperable, the overall reliability of j
the ADS is relatively unaffected. Thus, the stated Conpletion Tine allows for continued operation until-the nex-FDDE 5 entry (i.e., refueling outage), where l
the necessary plant conditions exist to affect repair.
'This situation is acceptable, even for an extended-period of time, from an overall risk perspective. As noted, the provisions of LCO 3.0.4 are not applicable i
for this Condition to allow recovery from shutdowns of short duration prior to the next refueling outage.
{
Furthermore, as ncted, this Condition is allowed to j
exist concurrently with Conditions A, B, C or'D.
l This Required Action assumes that the valve inoperability is due to a_ failure located in an inaccessable area of the plant. Since the intent of this LCO is to have all ADS valves operable at all i
times, all ADS valve failures which can be repaired in areas of the plant accessable during norrral operation should be repaired within a reasonable period of time.
1 M.,
uo j
With two ADS valves inoperable, at least one ADS valve j
must be restored to OPERABLE status within 30 days.
i Additionally, it must be verified innediately that at least two high pressure ECCS injection subsystems i
(HPCF or RCIC) are OPEhABLE.
In'this Condition, although overall ADS reliability is reduced, the remaining OPERABLE valves provide adequate depressurization capability. Furthermore,= sufficient nigh pressure ECCS capability is assured such that during postulated LOCAs the ADS function would not be needed, even for those breaks that do not result in j
rapid depressurization~of the RPV. The 30 day Completion Time is thus based on.the overall i
redundancy provided by the ECCS and its continued ability to perform its intended safety function, while assuring a return towards full ADS capability in a reasonabla e W so as_to not significantly i@ act onrall AD.
Hability. As noted, this Conditio,n is i
allowed to --c concurrently with Conditions A,-
B, C c,r D.
4 i
.~,_ _., _ _. _, _. _., _.. _ _
a.
- ~
-.a..
ECCS System B 3.5.1 DRAFT PASES I.1.
I.2 If any Required Actions and associated CTpletion Times of Condition G or H are not met, the plant must be brought to a FODE or condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least FDDE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reactor steam dame pressure reduced to 5150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Cccpletion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly m nner and without challenging plant systems.
SURVEILIJJJCE PEQUIPI! CTS FR 3.5.1.1 5
The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the p=p discharge lines of the RCIC system, and the LPFL and HPCF subsystems full of wt.ter ensures that the systems will perfo m properly, injecting their full capacity into the 3 actor Coolant System (RCS) upon demnd. This will also prevent a water harimer following an initiation signal. One acceptable methcd of ensuring the lines are full is to vent at the high points. The 31-day Frequency is based on operating experience, on the procedural ec-*rols governing system operation, and on the grawal nature of void buildup in the ECCS piping.
SR 3.5.1.2 Verifying the correct alignment for mnua;. power-operated, and autcratic valves in the ECCS flow paths provides assurance tk
- the proper flow paths will exist for ECCS operation. 'his Sk does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verif'ed to be in the correct position prior to locking, sealing, or Jocuring. A valve that receives an initiation signal u allowed to ce in a non-accident position provided the valve will automtically reposition in the proper stroke time. This SR does not require any testing or valve.ranipulation; rather, it involves verification that those valves capable of potentially being ABWR B 3.5-12 6/5/92 6:15 PM l
ECCS System B 3.5.1 DRAFT BASES mispositioned are in the correct position.
This SR d:es net apply to valves that cannot to inadvertently misaligned, such as check valves.
For the RCIC System, this SR also includes the steam flow path for the turbine and the flow controller position.
The 31-day Frequency of this SR was derived from the Inservice Testing Prcgram requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated unoer procedural control and tecause irproper valve alignment would only affect a single subsystem. This Frequency has heen shown to te acceptable through operating experience.
This SR is modified by a Note that allows an LPFL subsystem to be considered CPEPR3LE during alignrent and operation for decay heat renoval when below the MR cut-in permissive pressure in FDDE 3, if capable of being tranually realigned (remote or local) to the LPFL rrode and not otherwise inoperable. This allows operation in the MR shutdown cooling trode during FDDE 3 if necessary.
ER 3.5.1.3 Verification every 31 days that ADS (air receiver]
pressure is > (161) psig assures pneu:ratic pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The designed pneurratic suppy pressure requirements for the accumulator are Mch that, following a failure of the pneumarie supply to the accumulator, at least one valve actuation can occur with the drywell at design pressure, or five valve actuations can occur wfth the drywell at atrtospheric pressure (Ref. 5). The ECCS safety analysis assurres only one actuation to achieve the depressuriration required for operation of the low-pressure ECCS. This minimum required pressure of (161) psig is provided by the Attrospheric Control System (ACS). The 31-day Frequency takes into consideration administrative control over operation of the ACS and alarms for low pneumatic pressure.
SR 3.5.1.4 SR 3.5.1.5. SR 3.5.1.6 The perfortrance requirements of the ECCS purips are determined through application of the 10 CFR 50, ABWR B 3.5-13 6/5/92 6:15 PM l
.--_=
)
ECCS System B 3.5.1 DRAFT i
l BASES Appendix K criteria (Ref. 9). Periodic surveillance is performd (in accordance with the American Society of Machanical Engineers (ASME) Code,Section XI i
require.~ents for the ECCS pumps) to verify that the ECCS pugs will develop the flow rates required by the respective analyses. The ECCS pum flow rates ensure that adequate core cooling is provided to satisfy the 1
acceptance criteria of 10 CFR 50.46 (Ref.11). The i
RCIC pu."p flow rates also ensure that the system can naintain reactor coolant inventory during pressurized conditions with the RPV isolated.
J The pump flow rates are verified against a system head that is equivale-r to the RPV pressure expected during a I4CA. The fic. tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both-at the higher and lower operating-renges of the system. The total system pu::p outlet pressure is adequate to overcone the elevation head pressure t
between the pump suction and the ve sel discharge, the piping friction losse., and RPV pressure present during LCCAs. These values may be established duri.ng preoperational testing.
Since the required reactor steam dame pressure nust be available to perform SR 3.5.1.5 and SR 3.5.1.6, sufficient time is allowed after adequate pressure is achieved to perform these SRs. Reactor startup is i
allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time to satisfactorily perform the Surveillance'is short. The reactor pressure is allowed to be j
increased to nomal operating pressure since it is assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes which state the surveillances are only required to be perfomed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the specified reactor steam done pressure is reached.
A 92-day Frequency for SRs 3.5.1.4 and 3.5.1.5 is consistent with the Inservice Testing Program requirements. The 18-month Frequency for SR 3.5.1.6 is based on the need to perform this Surveillance under low reactor pressure conditions that apply i
during plant startup following a plant outage.
Operating experience has shown that these cccponents usually pass the SR when performed on the 18-month i
ABWR B 3.5-14 6/5/92 6:15 PM i
ECCS System B 3.5.1 DRPET BAers Fregaency, which is based on the refueling cycle.
Therefore, the Frequency was concluded to be acceptable from a reliability standp:> int.
FR 3. 5.1 "'
The ECCS subsystems are required to actuate automtically to perform their design functions.
This surveillance test verifles that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCF, RCIC and LPFL will cause the systems or subsystems to operate as desioned, including actuation of the system throughout-its emergency operating sequence, automtic punp startup, and actuation of all automtic valves to their required positions. This test also ensures that the RCIC System and HPCF subsystems will automtically restart on an RPV low water level (Level-2 and lowl 1.5, respectively) signal received subsequent to a i@V high water level 17evel 8) trip and that the suctions are automticady transferred from the CST to the suppression pool.
The 18-nonth Frequency is based on the need to perform
~
this Surveillance under conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Cperating experience has shown that these components usually pass the SR when perforned at the 18-month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is ncdified by a Note that excludes vessel injection during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the F2V is not required during the Surveillance.
S3 ?.5.1.9 The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test (logic only) is performed to verify _that the ADS logic operates as.
designed when initiated either by an actual or ABWR B 3.5-15' 6/5/92 6:15 PM
ECCS System B 3.5.1 DRAFT PAeES j
simulated initiation signal, causing proper actuation of all the required corTonents.
The 18.~cnth Frequency is based on the need to perform this Surveillance under conditions that apply during a plant outage and the potential for an unplanned transient if t.he Surveillance were perforned with the reactor at power.
Cperating experience has shown that these corponents usually pass the SR when perforned at the 18-tronth Frequency, which is based on the refueling cycle. Therefore, the Fregaency was concluded to be acceptable frca a reliability j
standpoint.
n This SR is modified by a !Jote that excludes valve i
actuation. This prevents an RPV pressure blowdown.
SR 3.5.1.9 A tranual actuation of each ADS valve is perforned to 4
verify that the valve and solenoids are functioning properly and that no blockage exists in the S/RV discharge lines. This is demonstrated by the response of the turbine control or bypass valve, or by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adegaate reactor steam dome pressure must be available to perform this test to avoid damging the valve.
Sufficient time is therefore allowed, after the required pressure is echieved to perform this test. Adequate pressure at which this test is to be perforned is (950] psig (the
[
pressure reconmended by the valve manufacturer).
l Reactor startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per AS!E requirements, prior to valve installation. Therefore, this SR is modified by a Note which states the surveillance is required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam dome pressure is 2 (950).
The Frequency of 18-months on a STAGGERED TEST BASIS ensures that t th solenoids for each ADS valve are alternately tested.
The Frequency is based on the need to perform this Surveillance under conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience bas shown that these conponents usually pass the SR when perforned at the 18-month Frequency, i
i ABWR B 3.5-16 6/5/92 6:15 PM
ECCS System B 3.5.1 DRAFT which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFEFINCES 1.
10 CFR 50, Ippendix A, GDC 33.
2.
ABhR SSAR, Section 6.3.2.2.4.
3.
ABhR SSAR, Section 6.3.2.2.1.
4.
ABhR SSAR, Section 6.3.2.2.3.
5.
ABhR SSAR, bection 6.3.2.2.2.
6.
IBAR SSAR, Section 15.2.8 7
ABhR SSla, Section 15.6.4 8.
ABhR SSAR, Section 15.6.5 9.
- 10. ABhR SSAR, Section 6.3.3.
11, 10 CFR 50.46.
- 12. ABhR SSAR, Section 6.3.3.3.
- 13. ABhR SSAR, Chapter 19.
I ABhR B 3.5-17 6/5/92 6:15 PM
FJE Suppression Ecol Cooling System 3.6.2.3 bbddNb.,
3.6 CO!EAI! MEIC SYSTFIS 3.6.2.3 Fesidual Heat Remval (FEFJ Suppression Pcol Ccoling LCD 3.6.2.3 Three FRR suppressica poc' cooling subsystems shall te CFEFABLE.
AF FLICt.dILITi
!ECES 1, 2, and 3.
ACTICGS CCUDITICU FIQUIFED ACTIO!J CCtTLETIO!! TI!E A.
C'ne PRR suppression A.1 Restore FRR 30 days pcol cooling subsystem suppression pool incpe ra.ble.
cooling subsystem to GPEFABLE status.
B.
Two PHR suppression B.1 Restore one FHR 7 days pcol cooling subsystems suppression pool inoporable.
cooling st system to OPEFABLE status.
C.
Req; ired Action and C.1 Be in IE E 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Comoletion Tir:e of Ccndition A or B not cet.
C.2 Se in !CCE 4 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 3
Three PRR Suppression Pool Cooling subsysters inoperable.
tf/
y 9'
ABWR STS 3.6.2.3-1 6/4/92 2:58 FM
4 PJE Suppression Pcol Cooling System 3.6.2.3 DRAFT 4
l SLMIILIMCE PICUIPDEJTS 4
St%IILIMCE FPICUErJCY f
SR 3.6.2.3.1 verify each PJB suppression pool cooling 31 days subsystem ranual, power operated, or i
autcratic valve that is not locked, sealed or
{
otherwise secured in position, is in the correct position or can be aligned to the correct position.
SR 3.6.2.3.2 Verify each PJR pump develops a flow rate In accordance 2 4200 gpm through the associated heat with Inservice
,1 exchanger while operating in the suppression Testing.
pool cooling node.
Program, or 92 days i
i b
1 i
i P
I 1
l i
l l
t I
i-ABWR STS 3.6.2.3-2 6/4/92
.-2:58 PM-
.-....m.
___._____..m.
PRR Suppression Pool Cooling System B 3.6.2.3 DRAFT 4
i PASES i
f B 3.6 CotCAItrin SYSTEMS i
B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling 3
System i
j BASES i
i BACKGROUtJD Following a Design Basis Accident (DBA), the PRR
}
Suppression Pool Cooling System remves heat from the suppression pool. The suppression pool is designed to absorb the sudden input of heat from the primary system.
In the long term, the pool' continues to absorb residual i
heat generated by fuel in the reactor core. Scene mans must be provided to remve heat from the suppression pool so that the te@erature inside the primary containment I
remains within design limits. This function is provided by three redundant RHR suppression pool cooling.
subsystems. The purpose of this LCO is to ensure that.
j all subsystems are OPERABLE in applicable FDDES, i
Each RHR subsystem contains a pug and a heat exchanger, i
which are mnually initiated and independently controlled. Each RHR subsystem performs the_ suppression l
pool cooling function by circulating water from the l
suppression pool through the respective RHR. heat-exchanger and returning it to the suppression pool.
i Reactor Cooling Water (RCW), circulating through the shell side of the heat exchangers, exchanges heat with 4
the suppression pool water and discharges this heat to j
the Reactor Service Water (PSW) System,.which then in j
turn rejects the heat to the external heat sink.
t-The combined heat-removal capability of two of the' three l
RHR subsystems is sufficient to meet the overall DBA pool cooling requirement for icss-of-coolant accidents (IDCAs) and transient events such as a turbine trip'or stuck-open safety / relief valve- (S/RV). The heat-removal capability F
of a single RHR subsystem is sufficient to meet the overall. pool cooling requirements for S/RV leakage and reactor core isolation cooling (RCIC) testing as these events increase suppression pool temperature mre slowly.
l The RHR Suppression Pool Cooling System is also used to lower the suppression pool water bulk temperature followina'such events, 4.
l' l
ABWR STS B 3.6.2.3-1 6/5/92L -5:22'PM a_.,+.~
- ~ ~
._,,-___,_,.,...;p y,,-.-,.,_.,v..m,_...
,..._~m
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_. _ _ _ _. ~.. _ _ _. _...
1
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PER Suppression Pool Ccoling System B 3.6.2.3 1
DRAFT i
4 l,
BASES APPLICABLE SAFETY ANALYSES i
Peference 1 contains the results of analyses used to l
predict primry containment pressure and temperature following large-and smll-break If)CAs. The intent of
{
the analyses is to demnstrate that the heat-remval j
capacity of the PRR Suppression Pool Cooling System is 1
adequate to maintain the primary containment conditions i
within design limits. The tine history for suppression pool temperature is calculated to demonstrate that the
}
mximum temperature remins below the design limit.
Reference xx contains discussion of additional analyses that was performed to support PRA success criteria for the long term heat renoval function. The intent of these analyses was to predict primary containment pressure and temperature following low probability events beyond the
[
DBA and to determine the minimum heat-removal capacity required to maintain the primry containment conditions
}
within its. ultimte capacity. The results are used to i'
establid the minimum amunt PER (Suppression Pool l
Cooling) system equipment required to prevent ultimate containment failure under such beyond DBA events.
J FRR suppression pool cooling satisfies Criterion 3 of the NRC Policy Statement.
1 i
I LCO i
During a DBA, a minimum of two PRR suppression pool cooling subsystems are required to mintain the primry I
containment peak pressure and temperature below design i
limits (Ref.1). To ensure that these requirements are j
met, three PRR suppression pool cooling subsystems must i
be. OPERABLE with power from three safety-related independent power supplies. Therefore, in the event of.
g i
an accident, at least two subsystems are OPEPABLE assuming the worst case single. active failure. An BRR 4
suppression pool cooling subsystem is OPERABLE When one the_ pump, heat exchanger, and associated piping, valves, instrunentation, and controls are OPERABLE.
L ABWR STS B.3.6.2.3-2 6/5/92 - 15:22 PM i
-.-w e
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r pw.,,.,e,
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+
PRR Suppression Pool Cooling System B 3.6.2.3 DRAFT BASES APPLICABILITY In FDDES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containwnt and cause a heatup and pressurization of primry containment.
In FDDES 4 and 5, the probability and consequences of these events are reduced due to the pressure and tegerature limitations in these FDDr.S.
Thererore, PRR suppression pool cooling is not regnred to be OPEPABLE in FDDES 4 or 5.
ACTIONS A1 With one PRR suppression pcol cooling subsystem j
inoperable, the inoperable subsystam must be restored to OPEPABLE status within 10 days.
T.
this Condition, the remining PRR suppression pool cool.'ng subsystems are adequate to perfor:n the primry containmnt cooling function. However, the overall reliability is reduced because a single failure in one of the two OPEPABLE i
subsystem would result in reduced primry containment cooling capability, which under some conditions (e.g.,
high service water tenperature) my not be sufficient to meet DBA containment cooling requiremnts. The 30-day l
Cmpletion Time was chosen in light of the redundant PER suppression pool cooling capabilities afforded by the. two l
OPEPABLE trains and the low probability of a DBA occurring during this period. --Additionally, analyses of beyond DBA events demonstrates that one PRR suppression t
pool cooling subsystem'is adequate to mintain containment conditions well.bclow its ultimate capacity i
IL1 i
j With two PER suppression pool cooling subsystems inoperable, at least one inoperable subsystem must be
-restored to OPERABLE status within-7 days.
In this Condition, the remining Ph suppression pool cooling -
subsystem affords significant primary containe nt cooling capability and would be sufficient to maintain containment conditions well below its ultimate capacity.
However, the overall reliability is reduced because a single failure in the one OPERABLE subsystem could result in a substantial loss of primary containmnt cooling--
caoability. The 7-day Completion Time was chosen in l
ABWR STS B 3.6.2.3-3 6/5/92
-5:22 PM l
-=
U PRR Suppression Pool Cooling System B 3.6.2.3 DRAFT i
BASES s
1 light of the redundant PER suppression pool cooling
)
capability afforded by the OPERABLE train and the low j
probability of a DBA occurring during this period.
c.'
ed c.2 1
l The plant must be placed in a ! ODE in which the IEO does not apply if the inoperable RHR suppression pool cooling subsystems cannot be restored to OPERABIE status in the i
associated Completion Tines, or if all three PRR i
suppression pool cooling subsystems are inoperable. This is done by placing the plant in at least ! ODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in toDE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed l
Completion tines are reasonable, based on operating experience, to reach the required plant condition from i
full power conditions in an orderly manner and without challenging plant systems.
SURVEILIRJOE PEQUIPE!MS i
i Verifying the correct alignrent for manual, power-operated, and autonatic valves in the PRR suppression l
pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR i
does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were 1
verified to be in the correct position prior to locking, l
. sealing or securing. A valve is also allowed to be in the non-U cident position provided it can be aligned to its accident condition. This SR does not require any testing or valve nanipulation; rather, it involves 4
i verification of those valves capable of potentially being mispositioned, are in the correct position. This SR does 4
not apply to valves that cannot be inadvertently i,
misaligned, such as check valves.
The 31-day Frequency of this SR was developed based on-the Inservice Testing Program requirenents to perform vals testing at least once per 92 days. The Frequency of 3 -days is further -)ustified because the valves are i
operated under procedural control,.imptuper valve position would affect only a single subsystem, and.the probability of an event-requiring initiation of the system is-low.. This Frequency has been shown to be
-acceptable through operating experience.
i
?
i ABWR STS B 3.6.2.3-4 6/5/92 5:22..PM i
.., _ -. ~,.,.,. - - -
PRR Suppression Pool Cooling System B 3.6.2.3 DRAFT BASES SR 3.6.2.3.2 Verifying that each PRR pu::p develops a flow rate 2 (4200] gpm while operating in the suppression pool ecoling mode with flow through the associated heat exchanger ensures that pup performnce has not degraded during the cycle. Flow is a norml test of centrifugal pu~p performnce required by Section XI of the ASME Code (Ref. 2).
This test confim.s one point on the pu~.p design curve, and is indicative of overall perfomance.
Such inservice inspections confirm cceponent OPERABILITY, trend perform nce, and detect incipient failures by indicating abnonnal performnce. The Frequency of this SR is in accordance with the Inservice Testing Program, or 92 days.
REFEPDJCES 1.
ABWR SSAR, Section (6.2.1.1.3.3].
2.
ASME Boiler and Pressure Vessel Code,Section XI,
" Rules for Inservice Inspection of Nuclear Power Plant Ca:ponents," Amrican Society of Mechanical Engineers.
ABWR STS B 3.6.2.3-5 6/5/92 5:22 PM
- E * #
a h %,;... _
e 9
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k
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7 L
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9 h
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e 4
4
)
i 8
i..
s 4
P 4
i L
l d
I e
l] -
I d
t
't J
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,,;--w,,
__%w%%,,_
.,,yg,_
g,,yg g y m m e, m qw.,p.je9y
.I i
I RCW/RSW System 4
3.7.2 DRAFT 1
3.7 PLAtTT SYSTDS 3.7.2 Remnr B"ildin' C "lin? Wut t
r (904) and E*3"Or 9"ildin' E* "iO" Kater (PSW) Systers i
4 140 3.7.2 Division 1, 2, and 3 RCW and RSW subsystems shall be OPERABII.
l f
APPLICABILITY:
!GES 1, 2, and 3.
ACTICNS CONDITICN REQUIPID ACTION CCMPIITION TIME A. One RCW punp, one PSW A.1
NCfrE---------
l punp and/or one RCW/RSW
- l. Provisions of 140 heat exchanger 3.0.4 are not inoperable in the sane applicable if this subsystem, in one or Condition exists for nore subsystems.
only one RCW/RSW subSyEtem, i
- 2. Enter applicable Conditions and Required Actions of 140 3.4.9, " Residual Heat Removel-! ODE 3" for RHR shutdown cooling nade inoperable by RCW/PSd.
- 3. Enter applicable Conditions and Required Actions of LCO 3.6.2.3, "RHR Suppression Pool Cooling" for containnent cooling
. trade inoperable by RCW/RSW.
Restore the inoperable-30 days RCW/RSW component (s) to OPEPABLE status.
(continued)-
'5:35 PM
_...m 1
l RCW/RSW System 3.7..
DRAFT
)
CCRJDITIOt1 FIC01 PED ACTICt1 CCt@LETICt1 TIFE l
B. One RCW/RSW subsystem B.1
!DIT-----------
inoperable ior reasons
- 1. Enter applicable other than Ccndition A.
Conditions and Required Actions of LCO 3.4.9, " Residual Heat Renoval-FDDE 3" for PRR shutdown ecoling made inoperable by RCW/P54.
- 2. 3ter applicable 3
Conditions and Required Actions of j
LCO 3.5.1, "ECCS-
}
Operating" for ECCS injection subsystem (s) made inoperable by 1
RCW/P54.
1
- 3. Enter applicable Conditions and I
Required Actions of ico 3.6.2.3, "PRR Suppression Pcel
.i Cooling" for containment cooling-made inoperable by i
RCW/P54.
1
- 4. Enter applicable Conditions u d.
Required Actions of
. ICO 3. 8.1, "AC i
{
Sources-Operating" for
~
diesel generator nade inoperable by RCW/PJJW.
4
~
Restore the inoperable 7 days RCW/PSW subsystem to CPERABLE status.
(continued)
ABWR STS 3.7.2-2 6/5/92 5:35 PM
RCW/RSW Syst m 3.7.2 DRAFT CCUDITICN REQUIPED ACTION CCt@LETICN TIFE C. Two RCW/RSW Divisions C.1
17 E -----------
inoperable for reasons
- 1. Enter applicable other than Ccndition A.
Conditions and Regaired Actions of LCO 3.8.1, "AC Sources-Operating" for diesel generator made inoperable by RCW/RSW.
- 2. Enter applicable Conditions and Required Actions of Ir0 3.5.1, "ECCS-Operating" for ECCS injection subsystem (s) made inoperable by RCW/PSW.
- 3. Enter applicable Conditions and Pegaired Actions of ICO 3. 6.2.3, "RHR Suppression Pool Cooling" for containment ecoling made inoperable by RCW/RSW.
- 4. Enter applicable Conditions and Regaired Actions of LCO 3.4,9, " Residual Heat Removal-FODE 3" for RHR shutdcNn cooling made inoperable by RCW/P5d.
Restore one inoperable.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCW/PS4 subsystem to OPERABLE status.
.(continued)
ABWR STS 3.7.2-3 6/5/92 5:35 PM
e 4
RCW/RSW System 3.7.2 DRAFT COtOITICt1 FIQUIPID ACTICt1 CCt@LETICt1 TIbE D. Three RCW/P54 Divisions D.1 Be in FDDE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable for reasons g
other than Concition A.
j D.2 m
=ICII-----------
Only required if RCW/P5d i
Regaired Action and and UHS have sufficient associated Conpletion cooling capability to l
Tine of Condition A, B reach and naintain or C not net.
FDDE 4.
Be in FODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> frca discovery of
)
RCW/RSW and UHS capability
(
adegaate to reach and naintain j
FDDE 4.
4 4
1 i
i 3
I 4
A ABWR STS 3.7.2-4 6/5/92 5:35 PM 4
e--
g
-,-r
,-,,,w.
,,-,.e.
,.vw
.e c
n
RCW/RSW System 3.7.2 DRAFT SURVEILIRJCE RECOIFD'E!7TS SURVEILIR4CE FREQUE!?CY SR 3.7.2.1 Verify the water level in each FSN pump well of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the intake stracture is 2 [
] feet.
13CfrE-- ----------------
Isolation of flow to individual corponents does not render RC"d/PSW subsystem inoperable.
Verify each RCW/PSW subsystem nanual, power 31 days operated, and automatic valve in the flow path-servicing safety related systems or components that is not locked, sealed or otherwise secured in position is in the correct position.
SR 3.7.2.3 Verify each RCW/RSW subsystem actuates and/or 18 months reconfigures to the safety related mode of operation on an actual or eMalated initiation signal.
ABWR STS 3.7.2-5 6/5/92 5:35 PM l
__ __ _9
4 l
RCW/PSW System B 3.7.2 DRAFT BASES B 3.7 P1RTT SYSTE2G i
9 1.7.2 Epp."*cr CMiina Water ( P'"*d ) and Pey+0r Service Water (PSW)
DC.?.b i
BASES s
I BACFGRCWD 3
Tne rcd and PSW systems together are designed to provide i
l ccoling water for the r:emoval of heat frcin plant auxiliaries, such as Emrgency Core Cooling System (EOCS)
-I t'otors and pump seal coolers, Residual Heat Removal (PER)
System heat exchangers, standby diesel generators, and rcem coolers for ECCS equipnent required for a safe reactor shutdown following a Design Basis Accident (DBA)
)
or transient.
The combined RCW/RSW system also provides j
cooling to unit compe'... qts, as required, during nornal i
power operation, shutoc.3 and reactor isolation rodes.
During a DBA, nest of the equipment required for norml operation only is isolated from the rcd /RSW system, and cooling is directed only to safety-related equipmnt and i
to selected non-essential equipment such as control rod drive (CRD) pump oil coolers, instrument and service' air ccmpressor coolers, reactor water cleanup (RWCU) pump coolers and reactor internal pump -(RIP) M3 set coolers, all and any of which can then be msually isolated, if required. Additionally, the standby equipment (one RCW purp, one PSW pump and 'a one RCW/BSW heat exchanger) in each subsystem is either automatically started or valved i
into service upon receipt of a LOCA signal, t
The ccrnbined RCW/RSW system includes three seperate subsystems (A, B and C). Each subsystem consists of the
[ ultimate heat sink (UHS)), an independent cooling water header, and the associated pumps, heat exchangers, piping, valves and instrumentation. Each subsystem includes two RCW pumps, two RSW pumps and three rcd to RSW heat exchangers. Each subsystem is sized to provide sufficient cooling capacity to support the. required safety-related systems in its respective division during-safe shutdown'of the unit following a loss-of-coolant accident (IOCA).
Subsystems A, B and C are redundant and-service eaui,rvrant in Divisions 1, 2 and 3; respectively.
Cooling water is pumped from the [ UHS) by the RSW pug >(s) in each subsystem to the supply header serving the ABWR STS B 3.7.2-1 6/5/92 5:57 PM 2-.. _,, _ _ _
RCW/P54 System B 3.7.2 DRAFT PMES respective rcd /RSW heat exchangers. Af ter remving heat frtxn the respective RCW subsystem the water is pumped back to the (UFS).
In a seperate closed loop, cooling vuter is circulated by the pump (s) in each rcd subsyste:n through tre essential components to be cooled and back through the rcd'RSW heat exchangers. Thus, the heat removed from the components by the RCW is transferred to the RSW, and then ultim tely rejected to the UHS.
Subsystems A, B and C supply cooling water to redundant egaipment reqaired for a safe reactor snutdown.
Additional informtion on the design and operation of the RCW and RSW systems along with the specific equipment for which the combined RCW/F54 system supplies cooling water it pcov'ded in SSAR, Sections 9.2.11 and 9.2.15, and Table 9.2-4 (Ref. 1, 2 and 3), respectively. The RCW/P54 system is designed to withstand a single active or passive failure coincident with a loss-of offsite power without losing the capability to supply adequate cooling water to equipment reqaired for safe reactor shutdown.
Following a DBA or transient, the RCW/F5d system will operate automatically without operator action. Panual initiation of supported' systems is, however, performed for some cooling operations (e.g., shutdown ecoling).
APPLICABLE SAFETY ANALYSES The volume of each water source incorporated in a UHS complex is sized so that sufficient water inventory is available for all RCW/RSW system post-10CA cooling require 2 rents for a 30-cay period with no additional m keup water source available. The ability of the RCW/P54 system to support long-term cooling of the reactor or containment is assumed in evaluations of the equipment required for safe reactor shutdown presented in the SSAR, Sections 9.2.11, 9.2.15, 6.2.1.1.3.3 and Chapter 15 (Refs.1, 2, 4 and 5, respectively). These analyses include the evaluation of the long-term primm
~
containment response after a design-basis'10CA.
Th'e' RCW/PSA system provides cooling water fcr the PRR suppression pool cooling mode to limit suppression pool te:rperature and primary containment pressure following a IOCA. This ensures that the primry containment can perform its intended function of limiting the release of radioactive materials to the environm nt following a ABWR STS.
B 3.7.2-2 6/5/92 5:57 PM
RCW/PFd Systen B 3.7.2
@ g pg
==
BASES ILCA. The rcd /RSW system also provides cooling to other cocponents assumed to function during a LOCA (e.g., RHR and HPCF pu", ps). Also the ability to provide onsite emergency AC power is dependent on the ability of the rcd /P54 system to cool the DGs.
The safety analyses for.leng-term containment cooling were performed, as discussed in the SSAR, Sections 6.2.1.1.3.3 and 6.2.2.3 (Refs 4 and 6, respectively),
for a ILCA, concurrent with a loss-of-offsite power, and minimum available DG power. The worst-case single failure affecting the perfonunce of the RCW/RSW system is the failure.of one of the three standby DGs, which-would in turn affect one rod /P5d subsystem. The PSW flow assumed da the analyses is 2.63 M1b/hr to each RHPheat exchanger (SSAR, Table 6.2-2a, Ref. 7).
References 1 and.2 discuss RCW/RSW system performance during these conditions.
G
'l The combined RCW/RSW system, together with the UHS, satisfy Criterion 3 of the NRC Policy Statement.
KO The OPCPABILITY of-subsystem A (Division 1), subsystem B (Division 2) and subsystem C_ (Division 3) of the combined RCW/PSA system is required to et.s;*e the effective operation of the RHR system in v pv ug heat from the reactor, and the effective operation of other safety-related equipment during a DBA'or transient. Requiring all three subsystems to be OPERABLE ensures that at two of the three subsystems will be available in the event of a single failure. Each subsystem individually.is
.. designed _to provide adequate capability to meet cooling requirements of the minimum equipment required for safe shutdown.
A suosystem is considered OPEPABLE when:
c-a.
Both associated RCW pumps are OPERABLE, b - _both associated RSW pumpa are OPERABLE, c.
all three associated RCW/P5d heat exchangers are.
_B_3.7.2 6/5/92 5:57: PM
1 rcd /PSW System B 3.7.2 j
DRAFT BASES 1
e.
the associated piping, valves, instru untation and 3
controls reqJired to perform the safety-related i
functicn are OPEFABLE.
i The isolaticn of the RCW/P54 system to components or i
system may render those components or systems inoperable, but does not affect the CPEPABILITY of the RCW/RSW system.
i APPLICABILITY 1
i In FDCES 1, 2, and 3, the RCW and RSW systems are j
reqJired to be OPEPABLE to support OPERABILITY of the equipnent serviced by the combined RCW/RSW system, and 4
i required to be OPEPABLE in tnese FCCES.
In FDDES 4 and 5, the OPERABILITY reqJirements of the RCW/P54 system are deterndned by the -systems it supports.
i l
ACTIONS f
Al If one RCW pump and/or one P54 pump and/or one RCW/RSW heat exchanger in the same subsystem is inoperable in one or-more subsystems, action must be taken to restore the inoperable component (s) to OPERABLE status within j
30 days.
In this condition sufficient redundant equipment is still available to provide cooling water to the required safety related components and sufficient heat re: oval capacity is still available to adequately cool most safety related loads. - In the degraded mode of this Condition, a subsystem may not be capable of renoving heat at the-required rate from the respective RER heat exchanger. However, with a mininum complement of one RCW pump, one RSW pump and two RCW/PSd heat exchangers,-a-subsystem is capable of performing its intended safety related cooling function except for long-term containment cooling.-
The 30-day Completion Tine is reasonable, based on the i
low probability of an accident occurring during the 30 days that a component is inoperable in one or more subsystems, the number of available redundant subsystems, the substantial-cooling capability still remaining in'a ABWR STS B 3.7.2-4
-6/5/92 5:57 PM
~
_. - ~
_ - ~ = -
RCW/RSW System B 3.7.2 DRAFT PASES 4
subsystem (s) in this Condition, and the expected high subsystem availability afforded by a system where rest of
}
the equipnent is nornally operating.
the Required Action is nodified by a Note indicating that the provisions of 140 3.0.4 are not applicable if this Condition exists for only one RCW/RSW subsystem. This is acceptable given the substantial degree of redundancy
)
provided by the RCW/RSW and supported systems and the i
i significant operational capability that still exists, even in this degraded condition.
4 The Regaired Act.on is further modified by two additional Notes indicating that the applicable Conditions of LCO 3.4. 9, " Residual Heat Removal (RHR)-FDDE 3," and LCO 3.6.2.3, "RHR Suppression Pool Cooling," be entered and Regaired Actions taken if the inoperable RCW/RSW subsystem results in an inoperable PFR-Suppression Pool Cooling or RHR-Shutdown Cooling, respectively.
This is in accordance with LCO 3.0.6 and ensures the proper actions are taken for these conponents.
B.l 4
If one RCW/PSW subsystem is inoperable for reasons other than Condition A, the subsystem must be restored to OPERABLE status within 7 days.
In this Condition, the renaining OPERABLE RCW/RSW subsysters are more than adequate to perform the required heat-removal function.
However,-the overall reliability-is reduced and a single failure in one of the OPERABLE RCW/BSW subsysters could result in a substantial reduction in the overall heat renoval capability. The 7-day Completion Time was developed taking into account the redundant capabilities afforded by the OPERABLE subsystems and the low probability of a DBA occurring during this period.
The Required Action is modified by four Notes indicating that the applicable Conditions of LCO 3.4.9, " Residual Heat Renoval (RHR)-BODE 3," ICO 3.5.1, "ECCS-Operating",
LCO 3,6.2.3, "RHR Suppression Pool Cooling," and ICO 3.8.1, "AC Sources-Operating," be entered and Required Actions taken if the inoperable RCW/RSW subsystem results in an inoperable RHR-Suppression Pool Cooling, ECCS injection subsystem (s),_RHR-Shutdown Cooling, or DG, respectively. This is in accordance with ICO 3.0,6 and ensures the proper actions are taken for these components.
5:57 PM
~
J 1
rcd /PSd System B 3.7.2 DRAFT j
BASES 4
If two RCW/RSW subsystems are inoperable for reasons other than Condition A, one RCW/RSW subsystem must be i
restored to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
In this Condition, the remaining OPERABLE RCW/PSd subsystem is I
adequate to perform the required heat-re!cval function.
However, the overall reliability is substantially reduced 4'
because a single failure in the remaining OPERABLE RCW/PSd subsystem could result in loss of the rcd /P54 2
function.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time was developed taking into account the low probability of a DBA occurring during this period and to allow for minor repairs that could restore OPERABILITY and avoid a forced shutdown that could result in potentially challenging unit systems.
i The Required Action is modified by four Notes indicating i
that the applicable Conditions of ICO 3.4.9, " Residual i
Heat Removal (RHR)-FDDE 3," Iro 3.5.1, "ECCS-Cperating",
5 I40 3.6.2.3, "RHR Suppression Pool Cooling," and ICO 3,8.1, "AC Sources-operating," be entered and Required Actions taken if the inoperable RCW/PSd subsystem results in an inoperable RHR-Suppression Pool Cooling, ECCS injection subsystem (s), RHR-Shutdown Cooling, or DG, 4
respectively. This is in accordance with Iro 3.0.6 and ensures the proper actions are taken for these components.
1 0.1 and D.2 If all three RCW/PSA subsystems are inoperable for reasons other than Condition A, or RCW/RSW subsystnms are inoperable in accordance with Condition A, B or C and cannot be rest.ored to OPERABLE status within the associated Completion Time, the unit must be placed in a i
tDDE in which the ILO does not apply. To achieve this status the unit must be placed in at least BODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and in FODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Required Action D.2 is modified by a Note precluding the~ requirement to place the unit in FODE 4 unless the PCW/RSW system and (UHS] have sufficient cooling capability to. reach and maintain FODE 4.
In this case, the unit should be maintained in FDDE 3 until this capability is restored.
Similarly, the Completion Time for reaching FDDE 4 has been modified to permit-36 hours from discovery of RCW/PSd cnd (UHS] capability adequate to reach and-maintain FDDE 4.
The allowed Completion Times are ABWR STS B 3.7.2 6/5/92 5:57 PM-
5 f
RCW/PSW System B 3.7.2 DRAFT j
BASES 1
reasonable, based on operating experience to reach the reqJired unit conditiens from full power "itions in an orderly mnner and without challenging uniu systems.
l SURVEILJCE REQUIPDOG'S FR 3.7.1.1 This SR verifies the water level (in each RSW pump well d
of the intake structure] to be sufficient for the proper
- operation of the PSW pumps (net positive sut; tion head and '
t
.put:p vortexing are considered in determining this limit).
The 24-hour Frequency is based on operating experience I
related to trending of the paramter variations during the applicable FDDES.
SR 3.7.1.2 Verifying the correct alignment for each manual, power-operated, and automtic valve in each RCW/RSW subsystem flow path provides assurance that the proper flow paths will exist for RCW/RSW operation. This SR does not apply to valves that are locked,-sealed, or otherwise-secured in po.sition, since these valves were verified to be in' the correct position prior to~ locking, sealing, or securing. A valve is also allowed to be in the nonaccident position and yet considered in the correct position, provided it can be automtically realigned to its accident position. This SR does not require any testing or valve manipulation; rather, it involves e
verification that those valves capable of potentially being mispositioned are in the correct position. This SR does net apply to valves that cannot be inadvertently misaligned, such as check valves.
This SR 1s modifled by.'a Note indicating that isolation of the RCW/PSW system to conponents or systems my render
~
those components or systems dmoperable, but does not affect the OPERABILITY of-the-RCW/PSW system.. As such,-
l when all-RCW/RSW pumps,- heat exchangers, valves and piping are all OPERABLE but a branch connection off of the main header is isolated,.the'RCW/RSW system is still.
1 ABWR STS B 3.7.2 C 5/92 5:5'hPM
..u.&-......----.a.....-
rcd /PSd System B 3.7.2 DRAFT BASES The 31-day Fregaency is based on engineering judgemnt, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.
SR 3. 7. L,1 This SR verifies the autocatic isolation valves of the RCW/RSW system will autte tically switch to the safety or energency position to provide cooling water exclusively to the safety-related equipment, ano limited non-safety related eqaipment, during an accident event. This is derenstrated by an actual or simulated initiation signal.
This SR also verifies the automatic start capability of the RCW and RSW pumps that are in standby and autcm tic valving in of the standby RCW/PSW heat exchangers (and auterratic start capability of required UHS active components) in each subsystem.
Operating experience !.as shown that these components usually pass the SR when performed on the 18-month Frequency. Additionally, to ensure _approximately egaal duty for all components within a subsystem, the normally operating and standby components are sheeduled to be alternated on a monthly basis. Therefore, this Frequency is concluded to e acceptable from a reliability a
standpoint.
FIFEPDJCES 1.
FSAR, Section (9.2.11).-
-2.
FSAR, Section (9.2.15).
3.
FSAR, Table (9.2-4).
4.
FSAR, Section (6.2.1.1.3).
5.
FSAR, Chapter 15.-
6.
FSAR, Sec.. ion (6.2.2.3).
7 FSAR, Table (6.2-2al. -
ABWR STS B 3.7.2-8 6/5/92 5:57 FM I
a