ML092380151
| ML092380151 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/18/2009 |
| From: | Weber T Arizona Public Service Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 102-06048-TNW/CJS | |
| Download: ML092380151 (101) | |
Text
Technical Specification 5.5.14 Palo Verde Nuclear Generating Station Thomas N. Weber Department Leader Regulatory Affairs Tel. 623-393-5764 Fax 623-393-5442 Mail Station 7636 PO Box 52034 Phoenix, Arizona 85072-2034 102-06048-TNW/CJS August 18, 2009 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2, and 3 Docket Nos. STN 50-5281529/530 Technical Specifications Bases Revision 50 Update Pursuant to PVNGS Technical Specification (TS) 5.5.14, "Technical Specifications Bases Control Program," Arizona Public Service Company (APS) is submitting changes to the TS Bases incorporated into Revision 50, implemented on July 22, 2009.
The revision insertion instructions and replacement pages are provided in the Enclosure.
No commitments are being made to the NRC by this letter. Should you need further information regarding this submittal, please contact Russell A. Stroud, Licensing Section Leader, at (623) 393-5111.
Sincerely,
ýkldý f-
ý TNW/RAS/CJS/gat
Enclosure:
PVNGS Technical Specification Bases Revision 50 Insertion Instructions and Replacement Pages cc:
E. E. Collins Jr.
J. R. Hall R. I. Treadway NRC Region IV Regional Administrator (enclosure)
NRC NRR Project Manager (enclosure)
NRC Senior Resident Inspector for PVNGS (enclosure)
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance A UIA Callaway 0 Comanche Peak
- Diablo Canyon 0 Palo Verde 0 San Onofre 0 South Texas 0 Wolf Creek
ENCLOSURE PVNGS Technical Specification Bases Revision 50 Insertion Instructions and Replacement Pages
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)T:2009.07.17 16:27:27 -0700' Insertion Instructions for the Technical Specifications Bases Revision 50 REMOVE PAGES Cover page List of Effective Pages 1/2 through 7/8 B 3.0-15 / B 3.0-16 B 3.0-17 / B 3.0-18 B 3.1.5-9 / B 3.1.5-10 B 3.1.7-3 / B 3.1.7-4 B 3.2.1-1 / B 3.2.1-2 B 3.2.2-1 / B 3.2.2-2 B 3.2.3-1 /B 3.2.3-2 B 3.2.4-1 / B 3.2.4-2 B 3.2.5-1 / B 3.2.5-2 B 3.3.1-23 / B 3.3.1-24 B 3.3.2-1 / B 3.3.2-2 B 3.3.2-9 / B 3.3.2-10 B 3.3.10-11 / B 3.3.10-12 through B 3.3.10-21 / B 3.3.10-22 B 3.3.11-7 / Blank B 3.3.12-1 / B 3.3.12-2 B 3.4.4-1 /.B 3.4.4-2 B 3.4.10-1 / B 3.4.10-2 B 3.5.6-5 / Blank B 3.6.1-1 / B 3.6.1-2 INSERT PAGES Cover page List of Effective Pages 1/2 through 7/8 B 3.0-15 1 B 3.0-16 B 3.0-17 / B 3.0-18 B 3.1.5-9 /B 3.1.5-10 B 3.1.7-3 / B3.1.7-4 B 3.2.1-1 / B 3.2.1-2 B 3.2.2-1 I B 3.2.2-2 B 3.2.3-1 / B 3.2.3-2 B 3.2.4-1 / B 3.2.4-2 B 3.2.5-1 / B 3.2.5-2 B 3.3.1-23 / B 3.3.1-24 B 3.3.2-1 / B 3.3.2-2 B 3.3.2-9 / B 3.3.2-10 B 3.3.10-11 / B 3.3.10-12 through B 3.3.10-21 / Blank B 3.3.11-7 / Blank B 3.3.12-1 / B 3.3.12-2 B 3.4.4-1 / B 3.4.4-2 B 3.4.10-1 / B 3.4.10-2 B 3.5.6-5 / Blank B 3.6.1-1 / B 3.6.1-2 I
B 3.6.7-1 / B 3.6.7-2 through B 3.6.7-5 / Blank B 3.7.1-1 / B 3.7.1-2 B 3.7.4-1 / B 3.7.4-2 B 3.7.4-3 / B 3.7.4-4 B 3.7.11-1 / B 3.7.11-2 through B 3.7.11-5 / B 3.7.11-6 B 3.8.1-7 / B 3.8.1-8 B 3.8.1-11 /B 3.8.1-12 B 3.8.1-23 / B 3.8.1-24 through B 3.8.1-31 / B 3.8.1-32 B 3.8.1-35 / B 3.8.1-36 B 3.8.1-41 / B 3.8.1-42 B 3.8.1-45 / B 3.8.1-46 B 3.8.3-3 / B 3.8.3-4 None B 3.7.1-1 / B 3.7.1-2 B 3.7.4-1 / B 3.7.4-2 B 3.7.4-3 / B 3.7.4-4 B 3.7.4-5 / Blank B 3.7.11-1 / B 3.7.11-2 through B 3.7.11-9 / Blank B 3.8.1-7 / B 3.8.1-8 B 3.8.1-11 /B 3.8.1-12 B 3.8.1-23 I B 3.8.1-24 through B 3.8.1-31 / B 3.8.1-32 B 3.8.1-35 / B 3.8.1-36 B 3.8.1-41 / B 3.8.1-42 B 3.8.1-45 / B 3.8.1-46 B 3.8.3-3 / B 3.8.3-4 2
PVNGS Palo Verde Nuclear Generating Station Units 1, 2, and 3 Technical Specificatlon Bases Revision 50 July 22, 2009 S
enso ~Digitally sined by Stephenson, Carl J(Z05778)
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Reason: This is an accurate copy of the original document.
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Date: 2009.07.17 15:15:09 -07'00' I
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0 PALO VERDE UNITS 1, 2,
AND 3 8
Revision 50 July 22, 2009
LCO Applicability B 3.0 BASES LCO 3.0.8 the supported system-occurting while the snubber(s) are (continued) not capableof performing their associated support,..,....
function.
LCO 3.0.8 requires that risk be assessed and managed. K Industry and NRC guidance on the implementation of 10 CFR 50.65(a)(4)(the Maintenance Rule) does not address seismic risk.
However, use of LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled.;:
and emergent issues are properly addressed.
The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerabil'ity of systems and components when one or more snubbers are not able to perform their associated support function.
In order to utilize LCO 3.0.8, the restrictions li:stded below shall be met.
- 1.
When LCO 3.0.8 is used, confirm that at least one train (or subsystem) of systems supported by'the non-functional snubber(s) would,.remain capable of performing their required safety.*or support..
functions for postulated design:.loads othertijit seismic loads.
LCO 3.0.8 does not apply to.-.....
non-seismic snubbers.
- 2.
When LCO 3.0.8 is used, a record of the des-igo function of the nonfunctional snubber(s) (i.e.
seismic vs. non-seismic), implementation of the-.i*, :i applicable LCO 3.0.8 restrictions, and the.;
associated plant configuration shall be avail'able, on a recoverable basis for NRC inspection.
- 3.
When LCO 3.0.8.a is used, at least one AFW train (including a minimum set of supporting equipment requi.red for its successful operation) or some,-'
alternative means of core cooling, not assocfated with the non-functional snubber(s), must be available.
- 4.
When LCO 3.0.8.b is used, at least one AFW train.
(including a minimum set of supporting equipment.
required for its successful operation) not associ~ated with the non-functional snubber'(s), or some `
alternative means of core cooling (e.g., fire water system or "aggressive secondary cooldown" using 'the steam generators) must be available.
'MLO VERDE-UNITS 1,2:3
-,. - ý B-3. 0-15 REVISION-50
SR Applicability B 3.0 BASES B 3.0 SURVEILLANCE REQUIREMENT,(SR) APPLICABIY'ITY_ :.,
SRs-SR 3'..0.1 thnrough-SR 3.,.O4"est~als*.J.he~general requirements applicable to al.l Specifications and, apply at all times,
.unless otherwise, stated.
SR 3.0.1 SR 3.0.1 establishes the;requirement that SRs must be met during the MODES~or other specified conditions in the Applicability for which;:the requirements.-of the LCO apply, unless otherwise.specified,in the-individual SRs.
This Specification is to ensure that;Surveillances are performed to verify the OPERABILITY of systems and components, and that. variables are, within specifiedlimi mts.
Failure to meet a' Surveillance within,the specified Frequency,. in accordance
-., with SR 310,2,,constitutes: a failureto~meet an LCO.
Surveillances may be performed.by means-of any series of sequential, overlapping, or total steps: provided the entire
-,Surveillance is performed within.the specified Frequency.
Additionally, thedefini-tions related-to instrument testing (e.g'.:, CHANNEL CALIBRATION) specify, that these tests are "preformed by means, of-any ser.ies-of sequential, overl appi ng, or total., ste'ps.J J:'.*
- ' J " ""
stems and ccmponent s,!;arie :assLumed-.to,be' OPERABLE when the "associ-ated,-SRs aVieý beeni"me,. Nothing i n thi s Specification,.
J to,_be-construed as implying that systems or components are OPERABLE when:
- a.
The systems or components are known to be inoperable, stt ilhstl:
- mreeti, ng -the
- SRs; or,
- b.
.- The'-requirerrieris of:.the--Surveillance(s) are known to be not met.ýbetween:-required-*Surveillance performances.
Surveillances'do.nothave to be performed when the unit is
--in a; MODE or other*-specified.condition for which the requirements of the associated,.LCO are not applicable, unless otherwise specified.
The SRs associated with a Special Test Exceotion,,(STE)-'are! only applicable when the
-STE is~used as an'ailowab-le exception to the requirements of a Specification,.
(continued)
PALO VERDE UNITS 1,2,3 B 3.0-16 PO E UI 1 3B3 1REVISION"50
SR Applicability B 3.0 BASES SR 3.0.1 Unplanned events may satisfy the requirements (including (continued) appl i cabl e accep.tance cri teri a)ý.ýfor;.ai.'agifven.,R.;:
R.Inq this
,cas.e,, the unplanned event.may be credited as fulfilling
.the performance of.the SR.:Thi allowance includes those SRs'"whose'"performance is normaljly'precliuded in a given MODE or other specifiedcondition.
Survei llances, "including SurVeillaraces invoked by Required-Actions...,do not :have to -beperformed on inoperable
-equipment because:.the ACTIONS define the remedial measures
.that apply:
Surveillances have to-be met and performed in accordance withiSR::3-..0.2,..prior.to returning equipment to
..- 'IOPERABLE'.*.status.,.
LUpon' completion of maintenance, appropriate post maintenance testing is 'required to declare equipment OPERABLE.
This
-.includes ensuring applicable..Surveillances are not failed
.and their most recent performance is in~accordance with
'SR 3.0.2. :Post maintenance testing may not be possible in
.,'the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been establi-shed.
I.n these.situations, the equipment may' be considered OPERABLE. provided, testing has been satisfactorily completed to,the. extent' possible and the equipment is not otherwise believed to be incapable of performing! Jts function. - This-wi l1.zallow operation to proceed to a 'MODE.*:.ri other.speci..fied..condition where other necessaryý..post*.maintfenance,tests.canbe'completed.
Some. examples of this process are:
a Auxi li ar{-Feedwater. (AFW) pump turbine maintenance during refueling that requires testing at steam l l'pressures> 800-psi,. -,,-.However, iflother appropriate
"- testing.'is satisfactorily completed, the AFW System can be considered OPERABLE.
This allows startup and
'other necessarý/(testing to proceed until the plant reaches.the:,steam:..pressure.required to perform the
- 'test~ing.. :,::,2*.
,.,§
- b:
High Pre§sure:-Safety.-'Injection (HPSI) maintenance during shutdoWn', that-requires system functional tests at a specified pressure., Provided other appropriate testing is satisfactorily completed, startup can proceed with HPSI considered OPERABLE.
This allows operation to reach the specified pressure to complete the necessary post maintenance testing.
(continued)
-PALO VERDE UNITS 1,2,3 B 3.0-17
- . i.ý " ý REVISION, 50
SR Applicability B 3.0 BASES SR 3.0.2 SR 3.0.2 establishes the requirements for-meeting the specified Frequency for SurVeillances and any Required
.Actiontwi~th a Completion Time thatrequires the periodic performancebof the Requiired'Action, on a-- once per..'
SR 3.0.2 permits a 25%extension of thelinterval specified in the Frequency.
This extension facilitates Surveillance
-scheduling and-considers-plant operating-conditions that may not be suitable for conducting the Surveillance..(e.g.,
transient conditions or other ongoing Surveillance or..,
maintenance-activities).
The 25% extehsion'does noýtsignific.ntlyidegrade the reliabiility.'that're'sUlts~from erf6drmjhg-ithe Surveillance at its specified Frequency:, This is based on the recognition that the most probable resUlt6f 'any particular Surveillance being performed is the verification of conformance with the SRs.
The. exceptions to SR 3.0.2 are those Surveillances for which the'25% extension of the interval specified in the Frequency does not apply.
These exceptions are stated in the individual Specifications..
An example of where SR 3.0.2 does not apply is the Containment Leak Rate Testing Program.
I *..
.~,
s,,Astated in SR 3.0:2. 1thýe,25% extension also does not apply t
te' in i tial portin.- o aperiadic Completion Time that
,,requi res perf6rmance e -a o'nce
"'er.
K basis.
The 25%
extension app"les..
ao,.h p'er prormance after the initial S..erfbrmance.
The 'ri fi1 lperforrance of the Required Action, whether 8tii aparticul ar,/Survei I lance or some other remedial action, is consider'ed asingle action with a single Completion Time.
One reason..for not allowing the 25%
extension to this Completion T me 'is that such an action
, usually verifies thatro loss of function has occurred by checking *the stat*u*6ofredundant or diverse components or accompi i shes the' 4 ci orIc of" the`inoperable equi pment in an Sat&rnative; 'mahnre.r I
(continued)
PALO.VERDE UNITS 1,2.3
- . B-3.0-18 P R SBREVISION 49
- ,'.ii* :* i r CEA Alignment B 3.1.5 BASES ACTIONS Dl (continued).
cod*ld re'sult'in' ýa'situation out si~de the design basis and immediate action'would be.'required to prevent any potential fuel damage.
Immediately opening the.reactor trip breakers minimizes theseeffects.
SURVEILLANCE REQU I REMENTS SR 3.1.5.1 J
.1, -:
Verification that individualiCEA positions are within 6..6inches (indicated reed switch positions) of all other
.CEAs:.Zin.thegroup ata a12 hour Frequency allows the operator tob..d'etect a.CEA'thatý:is beginning to deViate from its
.. expected position.' The specified Frequency takes into
'account'other CEA position information that is continuously available to the operator in they control. room, so that during actual CEA motion, deviations can immediately be detected.
- SR 3.
1'.5.21 OPERABILI-TY of at leas,,t two CEA~position indicator channels i,-s required :to determine CEA positions ',and thereby ensure
.compli ance,.w.thh*.&EýE a I f
.grn.nt and i nsertion limits.
The CEA fulli n6 and ',f,0H ul:out ilimitýsphdovi'de an additional.
'n I i
ndepIe-de*
1me.an§ordeermiing the CEA positions when the iCEAs-areu-at,ieitBr thei, r fully ins'erted or fully withdrawn
.Veri fyi ng esch ii strength CEA is;. trippable would require
-.that each CEA be-tr~lp~p'ed..
In MODES *1 and 2 tripping each full strength C'EA' Wod'lid resultih :radial or axial power tilts, or oscillations7 Therefore individual full strength CEAs are exercised every 92 days to provide increased confidence that all full strength CEAs continue to be trippable, even if they are not regularly tripped.
A movement of 5 inches is adequate to demonstrate motion without exceeding the alignment limit when only one full strength CEA is being moved.
The 92 day Frequency takes into consideration other information available to the operator in the control room and other surveillances being performed more frequently, which add to the determination of OPERABILITY of the CEAs (Ref. 3).
Between required (continued)
-PAL'O VERDE UNITS 1,22,3
ýB ý.3. 1. 5 -9 REVISION 46
CEA Alignment B 3.1.5 BASES SURVEILLANCE:`-,
SR,' 3.1: 5.3 3.(continued)':'
REQUIREMENTS performances of SR 3.1.5.3, if afCEA(s:),i~s.discovered to be immovable but remains trippable and aligned, the CEA is "consi'dered to. be. OPERABLE.
At anytime,. if a CEA(s) is immovable, a..determination, of the trippability (OPERABILITY) of.that.CEA(s)must be made, and appropriate action taken.
SR
- 3.
1.5.4.
Performance of a CHANNEL FUNCTIONAL TEST of each reed switch position transmitter channel ensuresthe channel is OPERABLE and capable of-indicating CEA position: Since this test ----
must be performed when the reactor.is shut down, an 18 month,
.. Frequency.to-:be coincident with refuel'ingoutage was.
selected.
Operating:experience :has shown'that the~se/:.
components usually pass this Surveill'ance'when performed at a Frequency of once every 18 months..: Furthermore, the Frequency takes into account other.factors, which determine the-OPERABILITY of the CEA Reed Switch.Indication System.
These factors include:
a..-,
Other, more frequentlY performed surveillances that help to verify, OPERABILITY;
- b.
On-line diagnost'ics perform'ed automatically by the CPCs,.CEACs,.
arid the P! ant Computer which include CEA position compai: soný, and sensor validation; and
- '"**:*c'.
The CHANNEL CAL~IBR, *1ONs~for*:the CPCs.(SR 3.3.1.9) and CEACs (SR"3.3.3L4 **nput,channels that are performed at 18 month.jntervals and is an overlapping test.
-S
.5.5 Verification of full strength CEAdrop times determines that the maximum CEA drop.time.permitted.is consistent with the
..assumedydrop time used i't he s-afety analysis (Ref. 3).
'Measuring drop timesp,.6rto reactor criticality, after
-reactor vessel headremr6val,.-ensures the reactor internals and CEDM will not interfere with CEA motion or drop time,
'and that no degradati.dn i:n thesesystems:.has occurred that would adversely.affed!,C.OEA..motion or. drop time.
Individual CEAs whose drop times are,,greater than, safety analysis assumptions are not OPERABLE.
This SR is' performed prior to criticality due to. themplant conditions needed to perform the SR and the potential 'for an unplanned plant transient if the Surveillance wereperformed with thereactor at power.
The 4 second CEA drop time is the maximum time it takes for a fully withdrawn individual full strength CEA to reach its 90% insertion position when electrical power is interrupted to the CEA drive mechanism with RCS T o1d greater than or equal-to 550F and.all reactor-coolan?.pumps operating..-
(continued)
PALO VERDE UNITS 1,2,3 B 3.1.5-10 REVISION 50
.2 Regulating CEA Insertion Limits 8 3.1.7 BASES BACKGROUND event of a CEA ejection :accident,. and the, shutdown.and.'.
(continued) regulating bank insertion limits ensure the requi.red.SDuis,
-' Operation wit~in the subject; LCO lidmits will prevent fuel c'laddi~ng failuresthat would breach~the primary fission product barrier~and release fissi~on* products to the reactor coolant in the event of a LOCA, loss of flow, ejected CEA, or other accident requiring termination by a Reactor rotective System trip.functi on.
APPLI CABLE.:,....
SAFETY ANAL.YSES.,
TZhe;fuel cliaddi'ng must ;not sustain damage :as a result of
- normal operation (Condition I).
and anticipated operational
.occurrences.(Cdndition II)...
<The acceptance criteria for the regulating CEA'insertion; part length or part strength CEA
.insertionASI, and Tq LCOs preclude core power di~stributi'ons from occurring that would, violate the following fuel design criteria:
a; During.a large break LOCA, the peak cladding temperature must not >xceed a limit of 2200°F, I0 CR50.46.(Re~f.2)':
b..*
Duri~ng CEg rmi.'o.*&atlon--eventsW there must be at least S a 95%Zproba~i*Tity'at "a 95% dohfidence level (the 95/95
...DNB criteri o*) 5*hat, the hoqt fuel rod in the core does
., *not. exp*e~ce &&acDNB 66hd~i~ti on;
- c.
During an eject*d *CEA abidenti the fission energy input to the fuel must not, exceed 280 cal/gm (Ref. 3):
anad
,:*.m
.[,,
- t';
,'i dl,..
The CEAs hmus;.tY~e.bapable of shutti~ng down the reactor with :a miriii:imi.rm::quihed SDM,. wich:t~he highest worth
. *.:"L '"
.. :..C[A. stuck lfiii.i,5,/wthdrawn G0mlC.26 (Ref. 1).
Regulat:ing I]CEA positli~on;.*ASI, anid Tm are f~rocess variables I
...!that together Icha~acte~fhize ~and control the three dimensional
,pdower* distribution*:of the reactor core._
i'. '"
Fuel cladding..damage d~oes not occur when:.the core is
",operate'd outside these.L-COs during normal operation.
.;?
-::,..',.(conti nued)
- PALO VERDE UNITS 1,2-,3B3..3REION 0
B 3.!.7-3
- REVISION 50
Regulating,.CEA Insertion Limits B 3.1.7 BASES APPLICABLE SAFETY ANAL (continue YSES d) accident occur with simultaneous violationof one or more of these. LCOsI' 1Changes in" the'power distribution'"can'dause' increased power peaking and corresponding increased local, LHRs *..'
The SDM requ'ireme t'is ensuredby'limiitingthe regulating and shutdown.'CEA inserti'on limits, so that.,the allowable inserted worth of the CEAs" isuch thatsUfficient reactivity is avaldble inh.the CEAS to shut down the reactor to hot-zero"power with' a&reactivlity 'margi:n-that assumes the maximum'worth.CEA remainsj.fully'withdr'awn upon trip The"most limi'ting' SDM"requirementsfor:MODE 1 and 2 conditions at-BOC are. determined.,by-the.requirements of several transients, e.g., Loss of.FlowSeized Rotor, etc.
However, the., most limiting SDM requirements for. MODES 1.and 2 at EOC come from just one transient,.Steam Line Break (SLB).
The requirements of the SLB event at EOC for both the.;fuil.lpower and.no load conditions'are significantly l.arger"than those of any other event'at that time in cycle and, also, considerably larger than"the most limiting requirements at BOGC:',
I Although the most. lfmtivlng ýSDM.,requfrements at EOC are much
' :larger than6 :those'a'tBOOý,n.the avai:l able SDM obtained via the scramming, of the CEAs are also substantially larger due to the mbch low4er boron,'t *ro ntratJi'on. at EOC.
To verify that adeq'uat& SDM Tarbe aVa i ITb!Ti;{hroughout the cycle to satisfy the changing requirement.*!'7:c-lcu~lat.,ions are performed at both BOC and EOC.
It has been determined that calculations at-these "two tiimes `-in cy' cle a're;ý.suffi~cient since the differencesbetweeh Thea"'
"lab'leSDM-and:the:limiting SDM requirements are stal at these. times in the cycle.
The measurement, df CEA*' bank'worth performed as part of the Startup. Testing Prog,;ariV'demonstrates that the core has expected shUtdown caoSabil ity- ;Consequently, adherence to LCOs 3416 and'3.1-,.'
i*roides.assurande,.that the available SDM at any time in cyc~le wi'll: exceed'-the Ii miting SDM requirements 'at th'att time'-in thelcycle, (conti nued)
PALO VERDE UNITS 1,2,3 B 3.A.7-4
.REVISION.: 48
LHR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Linear Heat Rate (LHR)
BASES*
'.2 BACKGROUND
... The purpose of this. LCO-is to limit the core power distribution to, the initial values assumed in the accident analyses.
- Operation,.wi~thin the limits imposed by this LCO
..limits'or prevents potential.fuel cladding, failures that
-, couldý breach, the,primary fiss§ion product' barrier and release fission roducts to the.,-reactor coolant in the event of a Loss Of Coolant Accident (LOCA),
ejected-Cntrol Element Assembly (CEA) accident, or other postulated accident
.'.requiri~ng-terminationby a Reactor Protective System (RPS)
. trip..function.
This LCO limits the damage to the fuel
..cladding during an accident by ensuring that the plant is operating within acceptable bounding conditions at the onset of a transient.
Methods of controlling the power distribution include:
a,.
- Using full,strength,-part length, or part strength CEAs to alter the axial power di.stribution;
- b.
Decreasing :,CEA: -insertion by: boration, thereby improving the.r'.adil power, d~is ilnution: and
- c.
Correct-ing:.o'ff-.,op.timum conditfon
,,.g.,
a CEA drop or
-mlsoperatli( Qvqff The. unit),that cause, margin dAeg radatl-on s:ý~.
The core power -di-s~tribution i s.controlled so that, in conjunction wi~tb, th.er core operatihng parameters (e.g., CEA insertion and-a.]JgnFenýit limits).Lthe.power distribution does not result-in vio-latiJn of this-LCO..
The limiting safety system settings.andh.is-LCOWare based on the accident analyses: (Refs.,- 1 ;alnd;.2),., so. that specified acceptable fuel
-design limits are.n.,t n
exceeded as a result of Anticipated Operational. Occurrences.-(AOOs),
and the limits of acceptable consequences are.:not. exceeded for other postulated accidents.
Limiting power distribution skewing over time also minimizes xenon distribution skewing, which is a significant factor in controlling the axial power distribution.
(continued)
.PALO VERDE UNITS 1,2,3 B 1,2". 1-1
ýREVISION 50
LHR B 3.2.1 BASES BACKGROUND (continueed)
Power distribution is a product of multiple parameters, various combinations of which may produce acceptable power distributions.
Operation within the design limits of power distribution:is -accomplished: by generating operating limits on the LHR and Departure from Nucleate Boiling (DNB).
Proximity'to'the DNB':condition is. expressed by the Departure from Nucleate Boiling Ra'to (DNBR),
defined as the ratio of
-the cladding-surface heat fl.ux.required to.,cause DNB to the
-actual. cladding surface:heat,-flux..
Thenmihimum DNBR value during both normal Qperati~on and-AOOs. is the DNBR Safety
-!Limit as calcu,))atedý,by the.CEý-.iCorrelation (Ref.
- 3) and corrected, for..such,.factorsias rpd.bow,,and'grid spacers.
It is accepted.as~an, appropriate margin, to, DNB for all
- operating conditions.-
There are two systemsthat monitor core power distribution online:
the Core Operating Limit Supervisory System (COLSS) and the Core Protection Calculators (CPCs).
The COLSS and CPCs that monitor the core power distribution are capable of verifying that the LHR and the DNBR do not exceed their limits:.TheCOLSS,.performs,this..function by continuously
.montoring.,the core power.distribution and calculating core power operating limits corresponding to the allowable peak LHRý amndtDNBR..-:The CPsi,,perform.-thi S function by
.,acorilously-ca~lcu.!at~ing,-*anactual
-.value Of DNBR and Local Power Density ([PD) for comparison with the respective trip
, ', setpoi-nts:..
The COLSS indicates contji,,.uously:to:pthe operator how far the core is from the operating limits and provides an audible alarm i,f an operatingi-mit. is, exceeded..;.: Such a condition signifies,a.,reduc:tion,-.,,,n.,the: capabjlj.ty.of the plant to withstand an.:anticipated-transient,,but..does not necessarily imply an immediate vi,61ati'oh 6ffuel'design limits.
If the margin to fuel;..-des-ign limitscontinues.to-decrease, the RPS ensures that-the~spec.ified acceptable fuel design limits are not exceededtby ini.tjia'ti.ng 4, reactor...trip.
The COLSS,-,continually :enerates an-assessment of the calculated margin for specified LHR and DNBR limits.
The data required for these assessments include measured incore neutron dflux, CEA positi~ons, and. Reactor, Coolant System
..(RCS) inlet temperatur.e,,pressure, and flow.
(continued)
PALO VERDE UNITS 1,2,3 B 8.2.1-2 REVISIONI:10
F B 3.2.ý B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Planar Radial Peaking Factors (F.)
BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to: the initi-al:-val'ues assumed in the accident
-analyses.
Op'ation withirn-the, limits imposed by this LCO
`-:either limits or pirevents~potential fuel cladding failures thatd#ould breadh' the primary fission product barrier and release fission products to the reactor coolant in the event ofl a uoss,.Of Coolant Acci dent (LOCA),.loss of flow accident, ejected C&ntrol"ý Elemrnnt, Assembly (CEA) accident, or other
'postul ated acciden'Ltrequiiring termination by a Reactor Protective System (RPS) trip function..-' This LCO limits damage to the fuel cladding during an accident by ensuring thatlthe pl~ant'is operati:ng within acceptable conditions at the'onset of a transient.
Methods of controlling the power distribution include:
I
-I
- a.
Using full, strength,',Ipart length,;or part strength CEAs'to alterthe axi'al',power distribution:
b
-Deceassing CEA' insertion by boratin., thereby I 'i mprov-ing theýrd'al d
power, distribution; and
- c.
Correcting off optimum conditions.(e:g., a CEA drop or misoperation of the unit) that cause margin The; core-powerg!*ribution
,is controlled' so that, in c6nj Onctilon" i tho:.tothetr. -Core'operating parameters (CEA insertionrand-ai'trgen* '
,limits)', the power distribution does
'n6t,'resbIlt irn v'io'Iathiin of-this LCOD Limiting safety system
.'settings ahd this,"LCO'are, based' on' the'accident analyses (Refs-Y'aid 2:), :st-"bath;'specified.acceptable fuel design l.i mi ts" are:&'not. exce did':as a' result of Anti ci pated Operational Occurrences, (AOOs),
and the limits of acceptable consequences'are-nhOt exceeded for other postulated accidents-.
Limiting power di:.str-ibution skewing over, time also minimizes
'xenon distribution, skewing, which is a significant factor in controlling axial power distribution.
Power distribution is a product of multiple parameters, various combinations of (continued)
PALO'.VERDE' UNITS 1,2,3 B
2-. 2-1 REVISION 50
F B3.2 BASES BACKGROUND which'may produce acceptable power'distributions.
Operation (continued) within the design limits of power distribution is accomplished by generating operating limits on Linear Heat
-Rate:(LHR).'and.Departure--frbm-Nucleate Boiling :(DNB).
I
'-I' Proximityto the-DNB condition is expressed by the Departure from Nucleate Boiling Ratio (DNBR),
defined as the ratio of
-the cladding:.sUr'facb heat-flux required-.to cause DNB to the actual cladding~surface;heat flux.
The:minimum DNBR value
'during both normal operationand AOOs i.s,.the DNBR Safety
.Limit as calculated: b0 the CE'-1Correlation (Ref.
- 3) and corrected for such":factors*as ýrbd bow.and.grid spacers 7 and it is accepted as an appropr iatemargin-to DNB for all operating conditions..,
l.
The're are two system* that monitor cbre.,power distribution online:-' the Core Operating: Limi.t Supervisory System (COLSS) and the Core Protection Calculators,;(CPCs).
The COLSS and CPCs that monitor, the_ core power distribution are capable of ver'ifying that-the LHR and the DNBR do not exceed their limits.
The COLSS performs this function by continuously
- monitoring thecore.'power distribution and calculating core power'operating~limits corresponding to the allowable peak LHR andDNBR values.
The CPCs perform this function by con't-inuoulIy tý.cutu latia:ig,-ac.tuai,values of DNBR and Local Po".er'Density-'(LPD)*.fr.mparison'*with the respective trip setpoints.
SDNBR~pna.Ity..factor~s i*,3 hcincluded in both the COLSS and CPC DNBR calculations to accr m*dte the effects of rod bow.
The amount of rod bow in each assembly is dependent upon the average burnup,exper-4;'ed bythat'assembly.
Fuel "assemblies'that-average burnup experience greater-rod"bow..
C goh,ýýrse.ly', fuel assemblies that receive lower 'than.average,ýburniuTptexperience less rod bow.
In design calculationSi-forla reload core,.each batch of fuel is
'assigneda penalty5-aopplied to.',.the*max'imum integrated planar radial-.power peak,-" of. the, batch... This penalty is correlated
- with!theamount,-bf. rodow. determined-from the maximum average assembly.bur nup of.the.batch.- A single net penalty for the COLSS and CPCs is then determined from the penalties associated with each batch that comprises a core reload, accounting for the. offsetting margins' due to the lower radial power peaks&'i-n the higher burnup.batches.
The COLSS indicates continuously to the operator how far the core is to the operating limits and provides an audible (continued)
-PALO VERDE UNITS 1,2,3 B 1.3:-2.2-2 REVISION,--10
B 3.2 3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 BASES AZIMUTHAL POWER TILT (q)
BACKGROL ND The-purpose ofIthis-LCO.AstO.limi t the core power distribution tothe initialvalues assumed in the accident analyses.
Operation within, the, limits imposed by this LCO
._either limits or iprevents poteitial fuel cladding failures that could breach,-the pri.mary fission product barrier and re-easefission products,.to, the reactor.coolant in the event ofa Loss Of Coolant.Accident -(LOCA), loss of flow accident.
ejected: Contrpol-Element,Assembly (CEA) accident, or other postulated accident'requiring termination by a Reactor Protective System (RPS) trip function.
This LCO limits the amount of damage to. the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditi.ons at.the, onset. of a transient.
Methods of controlling the power distribution include:
a.ý Using full strength;,.'part length,,.or part strength CEAs to.alter the;..axial power. distribution:
, b.
Decreas.ing.: CEA insertion by. boration, thereby
,.mprovngthe-ada.1 power.distribution; and
- c.
Correcting off optimum conditions, (e.g.,
a CEA drop or misoperatA~op ofthe'unit), tha~t.jcause margin O'.
egradait.j QoRs.-*.*
r,".
..r, Thee core power di, tibution.is control led so that, in conjunction iwith,-Pther,..core, operating parameters (e.g.,
'insertion andval.,gnment limits');- the. power distribution does not result'-in, yioilatlon. of-,th-is,,LCO.
The limiting safety system settings and.'thjs'.LCO'are based on the accident analyses (Refs...1* anrd:420,':.so..,that specified acceptable fuel
- .design limits are-.not',eceeded:.as a, result of Anticipated OperationalOccur.epces-,(AOOs),and the limits of acceptable consequencesare.:no.tiexceeded for other postulated accidents..
Limiting power distribution", skewing' ove& time also minimizes xenon distribution: skewing,.which is a significant factor in controlling axial power distribution.
(continued)
PALO VERDE UNITS 1,2,3
-.B.3.2.3-1 REVISION 50
T B 3.2.3 BASES BACKGROUND (continued)
Power distribution is a product of multiple parameters, various combinations of which may produce acceptable power:
"di~stributions.. Operation within-.the.design.limits of power distribution is accomplished by generating operating limits onb the Linear Heat Rate (LHR). and the-Departure from
- NLucleate` B il':ing (DNB),.
Proximity to "the. DN condi-tion is expressed by the Departure fromNucleate'.BOiiingRa'tio (DNBR),
defined as the ratio of
,,the cladding,surfacei-heat.flux required, to" cause DNB to the actual cladding 'sUrface hea.t.flux,.
The minimum DNBR value during both' hofmal.:.operation, and;:AOOs; is'the DNBR Safety Limit as'calculated'by'the CE-1. Correlation (Ref. 3) and corrected for suth-fat*offs asrodbowiandgrid spacers, and it is accepted as an appropriate margi:n to DNB for all operating conditions I
There are two systems that monitor core power distribution online:
the Core Operating Limit Supervisory System (COLSS) and 'the Core' Protection Calcul ators (CPCs') -... The COLSS and CPCs that monitor the core power distribution are capable of vehifyinrg,,that the LHR and the' DNBR:do not exceed their l imits'., The COLSS performs this function by continuously monitoring the core power distribution and calculating core p*"* Wr oper'ating 'linitsM' rrespondi.ng to the allowable peak
- 'L'HR 'a'nd DNBR`:,The)CPCs -pe6rform,,nthi's, function by continuously calculating actual values of DNBR and Local Power Densit-y (LFD)-or.. 'oriiparis-on with the respective trip A DNBR penalty factor is included in the COLSS and CPC DNBR caculation 'to aceonibdi-b6, the.*effects of: rod bow.
The
.amount'. of rod bow'd;`n h aIssembly' is *dependent upon the
'" average~burnup exp&T*icea b, the, assembly.
Fuel assemblies that incur higher-tharihtverage burnup experience greater magnitude-:f rod boqý ;Conversely,: fuel. assemblies that receive lowerl-than:avierage burnup experience less rod bow.
-'In design caldul~atiosfor,a reload;core,, each batch of fuel is assigned a,.peraltý.'
plied tosthe maximum integrated
-planar radial power peN2 of the batch.
This penalty is correlated with the amount of rod bow:that is determined from the maximum average assembly burnup of the batch.
A single het-penalty.:for the COLSS and CPCs is then determined from the penalties associated with each'batch that comprises a core reload, accounting;for.the offsetting margins caused by the lower radial power peaks in the higher burnup batches.
(conti nued)
PALO VERDE UNITS 1,2,3 I 3.1 2.3-2 REVI S lON- 0
DNBR B 32.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 Departure from Nucleate Boiling Ratio (DNBR)
BASES BACKGROUND The-purpose of,this LCO is to l.imit, the core power distribution to the initial, value assumed in the accident analyses.
Specifically, operation within the limits imposed by-this LCO either l-imits or prevents potential fuel cladding failure*s-that, could breach the primary fission product barri~er and release,fiss,ion products to the reactor coolant inthe event, of a. Los.s Of Coolant'Accident (LOCA),
laoss of flowaccident,ejected Control Element Assembly
.CEA) acci'dent'. or other postulated accident requiring
.,termination,-by.a Reactor P~rotective System (RPS) trip function.
This.LCOlimits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at the onset of a
° Methods of contro~llingthe power distribution include:
- a.
Using full strength,,:part.length,- or. part strength CEAs to alter, the ;.axi,.al power d4istributiodn;
-b.
Decreasing,..E-A.iserti onby boraiton', thereby improving.-'*e 30.a;ial. power.dis-tlriybutlion; and C
o or ect,ir**n,.p'p t mum condirtioný.'](e-.g.,
a CEA drop or misoperation of. the unit) that,cause margin degradations.
The core..powerdyis-tr_,i~but-i-on-is controlJ.d so that, in
-,:.;"cojuctbniwi -l',o**:ie, coe.p rt-Jng par..ameters (e, g.,
, n t anda nt,,i*mits):, the.power distribution does not result, in.
- v.
ioil:aj,-qpn of-.this.LCO.
-ThO limiting safety system settings;,arjid theirFs: LCO are based'on the accident analysJis,.-(Re1fs-..1,-,ard,.2).--.so that specified acceptable fuel design imits:ar,,;n texceeded-as a result of Anticipated Operational OccurrencesJ,(AOOs).-and the" limits of acceptable
... consequencesare not,-exceeded.for,:other postulated
.accidents..,
. I Limiting power distriibutJio-skewing :over time also minimizes the xenon distribution skewing, which~is a significant factor in'control.lirng,axial power distribution.
(continued)
PALO..VERDE UNITS 1,2,3
,B..3.2,4-1 REVISION 50
DNBR B 3.2.4 BASES BACKGROUND Power distribution is a product of multiple parameters, (continued) various combinations of which may produce acceptable power distributions.
Operation within.the design limits of power, "distribution is accomplished by'generating-operating limits on the Linear Heat Rate (LHR). and the Departure from nucleate boi:lin g (DNB).
Pfoxi'mi'ty to "the:' DNBi condi-tion is expiressed by the DNBR, defined as."the Yatio of, the,:claddihg surface heat flux required :to cause DNB to the actual cladding surface heat flux.- The min~imum DNBR'va:lue 'during-.both normal operation and AOOs i s;.the' DNBR.Safety. Limi-t -as-calcul ated by the CE-I Correlation (Ref: 3) 'and.corrected for)-*such factors as rod bows and'.grid spacers-7"an'd1it is. actepted'-.as an appropriate margin to. DNB for all operating Cobidi't-ioris;.
There are two systems' that' monitor corepower distribution online:
the Core Operating Limits-Supervisory System (COLSS),and the Core Protection Calculators. (CPCs).
The COLSS and CPCs that monitor-the core power distribution are
.capable, of verifying that the LHR and DNBR do not exceed
'..their':limi ts.
The COLSS performs this function by cont-i uousiy ilonl'itori ngr-the:'core power di stri buti on' and calculating core Dower onerating limits corresponding to the a!olowable peak'-LHR:ana d DBRi TheCPCs perform this function y' dontinuously ca" 1 tu
'an
-ctual value of DNBR and LPD for comDajriscn with tine..resoecti ve trip setpoints.
S.DNBR-penaltyfatc:;"
`'
I.-ed'in:'both the COLSS and CPC DNBR calculation to acco a6date*'thb'eeffects.of rod bow.
The amount of rod bow in each assembly is dependent upon the average burnup exper.lence -b':-ýthat'-assembjly.
Fuel
." assembliest~~"e' tanu er.thdn a*,erage.-burnup experience
,ssmbp s,,tinuý
"' 'grea'ter maghitude *ofo'dbow.,
Conversely, fuel assemblies hat receivee-lower, 'xt11i'ra'ge buirnup eperience less rod bow. l nýclesign c Li-1 7-11L'-,bn's~for a-'reload'core, each batch
'of fuel ;is assi:gnrdfriieralLyt'th'at is-applied to the maximum integrated, planar.r*dvil po'ern peak:of the batch.
This penalty.'is correiatca-"';ii`th the 'amount of~rod bow that is determined from the maxirnum 'average assembly burnup of the batch.
A single net penalty for the COLSS and CPCs is then determined from the penaltiesassoci'ated with each batch that comprises. a. core reioad, accounting for the offsetting margins due to' t~ie lower radial power peaks in the higher burnup batches.
(continued)
PALO VERDE UNITS 1,2,3 B-3-. ý. 4 -2 REVISIONi'10
B 3.2 POWER DISTRIBUTION LIMITS Bm 3.2.5 "AXIAL SHAP'E-"INDEX (ASI).
BASES AS I B 3.
2.5 BACKGROUND
The Ourpose of 'th-is LCO. i-s t~o limi.t the" core power distribution to the initial values assumed in the accident a.nal~ysis.
Operation within the limits imposed by this LCO either limit.s or'p prevents..potential fuel cladding failures
.that,'could. breach the primary fission product barrier and release fission product',toth reactor coolant in the event of1a Loss Of Coolant.Accident,(LOCA), loss of flow accident, ejected Control ElementAssembly (CEA) accident, or other
,postulated. accidenti:.requirjing,'termination. by a Reactor Protectiv, System (RPS),.trio function'.
This LCO limits the amount of damage,to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at *the onset of a transient.:
-Methods of controling, the.axial power diistribution include:
- a.
Using full strength, part lergth, or part strength CEAs to alter the a~ial, power d0stribution:
7 b'i.".Decreasingh,,CtA inserti on by borationi, thereby impropvig pt eaxialp6wer~distributio0n; and
- c.
Cbr'rec h'ing-ff optimum cnditli6n6 (e'.g.,
a CEA drop or
,misoper,,atiop,if the.,uni.t) that ccause margin p
r..
1,
.The-: core, power,distr,.ibution is controlled so that, in conjunction witW~pher: cdre, oper.ting epa rameters (CEA
.*"insertion and...alý.igniment1.-.lim*,ts), 'the powe*r distribution does S-not result in vio.06 ain-o.f,..hisLCO.. The limiting safety system s:ettings]a.ar.e.-.':$ed,_on the accident, analyses (Refs. 1 and 2), so that spe:'i.fi:.d'acceptable.fuel.design limits are not exceeded as,a-r's[t-lt of.Ahti ipated Op~erational
-.,Occurrences,(AOQs)..,idl'the'*linits of, acceptable consequences
.,,are: not exceeded,'fq.r',.ther.postulated,.accidents.
Limiting power diistribu]ti on skewing over.,,time also minimizes
' xenon distributio-n skewing, which. is a significant factor in controlling' axial,power distribution.
(continued)
,PALO VERDE-UNITS 1,2,3 B.3.,. 2'. 5 -1
.REVISION.50
ASI B 3.2.5 I
BASES BACKGROUND Power distribution is,.a product of m'ltfplle parameters-,
(continued) various combinations of which may produce acceptable power-distributions.
Operation within the design:limits.of power.
distribution is -accomplished by'generating bperating limits
-onthe inearHeat.Rate (LHR)*and the Departure from Nuqcqleate-Boil ing.,(DNB).
Proximity to the ONB condition is e pressed by the Departure from Nucleate"Boiling Rato.(DNBR), defined as the ratio of the cl'adding 8Urface heat flux required to cause DNB to the actual cladding surface heat flux:
The minimum DNBR value during both~normaloperation and' AOO is the DNBR Safety Limit as calculated. b.,the CE-1,Correlation (Ref. 3), and corrected'for;such'factors as rod bow and grid spacers, and it is.accepted-as:.an appropriate',.margin to DNB for all operating conditions.-
There are two systems that monitor core power distribution online:
the Core'Operating Limit Supervisory System (COLSS) or the Core Protection Calculators (CPCs).
The COLSS and CPCs moniftorthe. corempower distribution and are capable of
,ver.ifying that the LHR and DNBR do not exceed their limits.
'The.COLSS' performs.this functi'on by continuously monitoring the core power di;s§tri.Lht*on and, calculating core power hoperating limits,orrespnd
-g" to the allowable peak LHR and
. -NBR.,Q"The :CPCs per"tormthis.. unction by continuously calculating actual values.sofDNBR and local power density (LPD) for comparj. son,.wjthnthe respective trip setpoints.
A DNBR penalty factori T'ludad i both the COLSS and CPC The amount of rod bow ir6each*assembly is dependent upon the ave'rage burnup'exp'ri e4c b td-hat assembly.
Fuel assemblies,-thatirnur higher,'than average burnup experience
,-greater rod bow.-. Conversely, fuel assemblies that receive lower than' average bu`rnup' experience less rod bow.
In design'Calculatirs!,foýra rel'o'ad core, each batch of fuel is assigned a penalty that., is,applied to the maximum integrated planar radial power"peak of the batch.
This penalty is correlated with the'amount of'rod bow that is determined from the 'maximum ave'rage 'assembly burnup of the batch.
A single-net penalty'for.the.COLSS and CPC is then determined frnom the penalties associated with each batch that comprises a acord' rel6ad; accoun'ting for the offsetting margins due to the. lower radia.power-peaks:in-the higher burnup batches.
(continued)
`PALO'VERDE UNITS 1,2.3 B 3.2.5-2 A.VR.. U 2 B:
REVISION.O
RPS Instrumentation - Operating B 3.3.1 BASES; B
SAFETY ANALYSES
.Desi~gn Bas,is-Definition (continued),
- 5.
Departure` from-Nucleate Boiling Rati'o (DNBR)
Low
- 1 The CPCs perform'.the: caiculationsrequired to derive the DNBR and LPD parameters,:and their associated RPS trips.
The DNBR -. Low and LPD - High trips provide plant
.."'protectonr during the:foll.owing'AOOs and assist the ESF systems 'in the mittigation of the following
- accidentsý.,".
'Z i.' " '
The DNBR -,Low t-riip.provides'protection against core
,damage due, to the occurrence of locally saturated conditions in the,.limiting (hot) channel during the following events and is the-primary reactor trip (trips the reactor first),,for, these events:
Decrease in Feedwater Temperature:
Increase in Feedwater Flow; Increased Main Steam Flow (not due to steam line rupture) Without Turbine Trip.;:'
Increa'sed..Main:Steam 'Flow.(no~t due to steam line r
r upture:.) iWth a Cohcurrent Sing e Failure of an
'Act've.,cdn~ponent; Steam LeBreak With,Concurrent Loss of Offsite
. ",AC!
Powermj.,
.. '-Lo~ssf-*PNm* AC PbWer.
9, Partia a
- Loss-*S opf. Forced, Reactor.Cool ant Fl ow:
Tota'bal*;, o'f Forced ReactorCoolant Flow; S
S~ingle,1Rac'br, Coolant, PumY.(RCP) Shaft Seizure:
Uncontro-1o'ed CEA Withdrawal,,From Low Power; Unconitro611ed.CEA Withdrawal 'at.Power-S
'CEA Misopernat-n:.,CEA Dr, op; CEA'Mi.op6ýrthin;ý Part Length or Part Strength CEA S~bgroup Drop:;
'Primary*.Sampiple'or Instrument'Line Break; and
- .',Steam.'Generator'Tube Rupture.
In the above.'.l,,1'st, ony. t'he steam'iline break, the
.steam generator.tuberupture, theRCP shaft seizure, and the s'ample Or instrument line break are accidents.
The rest are AOOs.
I (continued)
- PALOD VERDE UNITS 1,2,3 1'ý B 3.3.1-23 REVISION,50
RPS Instrumentation - Operating B 3.3.1 BASES APPLICABLE
- 15.
Departure from Nucleate Boiling Ratio (DNBR)-Low SAFETY ANALYSES (continued)
... In.the safety ana-lyses for-transients involving reactivity andpower di-stribution anomalies, credit may-be taken for the CPCVOPT, auxiliary trip algorithm in l.ieu of,the RPS VOPT. trip function.
The exact trip credited '(CPC or RPS).,isdocumented'in chapter 15 of the UFSAR. under.the individual event sections.
The CPC VOPT auxiliary itr,,ip,acts through the CPC DNBR-Low and.LPD-High trip contacts to, provJde over power protecti on.
W.h6n credit fs'.takeh for the CPC VOPT algor~ithm, *the CPC:VOPT
'setpoints,"installed in the plant are based on the 'afetyf.a'nalyses and may differ from, the RPS VOPT allowablevalues and nominal setpoints.
The setpoints :associated with the CPC VOPT are controlled via Addressable 'Constants (TS Section S5.4.1) and Reload Data Block Constants (Ref. 8 and 13).
The CPC VOPT,auxiliary trip algorithm may provide protection.against core damage during the following events.:.
- Uncontrolled
,EA,Withdrawal From. Low Power (AGO);
UncontrO].l eQ..CEA.Withdrawal at Power (AOO):
Single".CEA W0lidhdiraala" Within:Deadband (AOO);.
Steam BypWsaC6Cihti'ý6bbSystem Misoperation (AOO);
CEAEjecti on.,4Accde nd M
. Main..Steam.L~pýL,Break (Accident).
(continued)
PALO VERDE UNITS 1,2,3
ý, B 3.3.1-24 REVISION. -.35
RPS Instrumentation - Shutdown B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Reactor Protective System (RPS)
Instrumentation -- Shutdown BASES BACKGROUND The RPS initiates a reactor. tri'p: to. protect against violating:the core fuel design1-imits and reactor coolant pressure boundary.j(RCPB) integri;ty.during anticipated operational~occurrences.(AOOs).,.:By. tripping the reactor, the RPS also assists the.rEngineered Safety Features systems i n..mi ti gating ýacci dents.
The protection and rbni~toring,syste!ms have been designed to ensure-saf#e operationr of the. reactor.
This is achieved by speciýying lJimiting safe.ty system settings (LSSS) in terms
.of parameters directly monitored by the RPS, as well as LCOs on~oth0r reactor system parameters and equipment performance_.
'Except for trip Functions 2 and 3, the LSSS defined in this Specification as.*the Allowable Value, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding.acceptable limits during Design Basis Accidents. (DBAs).
For Trip Functions 2 and 3, the UFSAR Trip Setpoi nt is th6 LSSS'.
During AOOs,'.whici:)a e i thbse6 events'expected to occur one or
- - more times dur.,ingtne phlantli-fe, the acceptable limits are:
. Thde.depaprtu.enfrom nucleate boiling ratio shall be Srainetinedo abve theSafety Limit (SL) value to vprev 'den*t`d'.tur fr'm-nucleate boiling:
-""Fuelcente -`-,-- meltirngsHal lnot occur; and The Reactor Coolant System pressure SL of 2750 psia shall not be exceeded.
Maintaining the parameters within the above values ensures that the offsite dose will be within the 10 CFR 50 (Ref. 1) and 10 CFR 100 (Ref. 2) criteria during AOOs.
Accidents are events that are analyzed even though they are not expected to occur during the plant life.
The acceptable limit during accidents is that the offsite dose shall be maintained within an acceptable fraction of 10 CFR 100 (Ref. 2) limits, Different accident categories allow a different fraction of these limits based on probability of occurrence.
Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.
(continued)
PALO VERDE UNITS 1.2,3
.- ".
- B 3.ý3.2-1 REVISION 50
RPS Instrumentation - Shutdown B 3.3.2 BASES BACKGROUND (continued)
The RPS is segmented into four 1 These-,modules are:
-Measurement-channels; Bistable trip',1Units; RPS~toqic; and nterconnected modules.
Reactor trip circuft-breakers (RTCBs).
This LCO applies to the'Logarithmic Power Level - High trip
-infMODES 3,.,1, and 5 wiith -the RTCBs closed and the CEAs capable of withdrawail.
In MODES 1.-nid. 2,, this trip function is addressed In LCO 3..3.1,.. Reactor Protective System (RPS)
Instrumentation, -
Operating.".LCO 3..3.",12. "Boron Dilution Alarm System (BDAS),"
applies.whe~thel
. RTCBs are open.
1j, This LCO applies to the Steam Generator #1 and the Steam Generator #2 Pressure-Low trip in MODE 3, with the RTCBS closedand the CEAs capable of withdrawal.
In MODES 1 and 2, this trip function is addressed in LCO 3*3.1, "Reactor Protective System (RPS)
Instrumentation-Operating."
- ')
i*,
r4 IJ (continued)
PALO VERDE UNITS 1,2,3
- -B 13.2-2 I ;:".
ý'.REVISJON 0
RPS Instrumentation -. Shutdown B 3.3.2 BASES APPLICABILITY (continued)
.The Steam Generfator #Pressure-Lowj.and the Stbam Generator #2 Pressure-Low trfps, RPS Logic, RTCBs, and Manual Trip are required in MODE,3 with the RTCBs closed, to provide protection for large.:MSLB events in MODE 3. The Steam Generator Press'ure-Low'trip in this lower MODE is addressed in this LCO:
The RPS'Logic in MODES 1,2,3.4.
and 5 is addressed.in.LCO 3.3.4,,Reactor Protective System (RPS)
Logic "ahd'Trip Initiation.
ýTh e,appl i cabilifty :fbrothe,-Logarithmic Power Level-High function i's moddified by;. a-Note: that allows the trip to be 0bypssed:wwhtei.'logarithmic power is > 1E-4% NRTP, and the bypass is' automatically removed.when logarithmic power is
' "1E-4% NRTP.
'1 I,
ACTIONS The most common causes of channel inoperability are outright failure or drift:of the bistable, or process module sufficient to ekceed the tolerance: allowed by the plant specific setpoint analysis.,.-Typically, the drift-is found to be small and results in a delay of actuation rather than a total loss of function.
This determination is generally made during the performance of a CHANNEL FUNCTIONAL TEST when the process instrument is set up for adjustment to bring it to within specification.
If the trip setpoint is less conservative than the Allowable Value stated in the LCO, the channel is declared inoperable immediately, and the appropriate Condition(s) must be entered immediately.
In the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value, or the excore logarithmic power channel or RPS bistable trip unit is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the unit must enter the Condition for the particular protection Function affected.
(continued)
?PALO"VERDE UNITS 1,2,3 B,2ý-3.2-9 REVISION 50
RPS Instrumentation -. Shutdown B 3.3.2 BASES ACTIONS (contir With a channel process measurement circuit that affects nued) multiple functional units iroperablelor-in test,-bypassdr trip a11 associated functional units as listed below:
PR6CESS MEASUREMENT CIRCUIT.
FUNCTIONAL UNIT (Bypassed or Tripped)
Steam Generator Pressure-Low
.Steam"Generator Pressure - Low (RPS)
... Steam'Generator #2 Level - Low (ESF)
Whe
.o ---
Steam Generator #2 Level - Low (ESF)
SWhen thenumber of inoperab]e-channels in a trip Function exceeds that specified in any related Condition associated
.-with the same, :trip.Function..-.then the plant is outside the safety analysis, Therefore,:'LCO.3.!0.3 is immediately entered, if applicable in the current MODE of operation.
A.1,and A.2 Condition A applies to the-failure of a single trip channel or associated instrument channel inoperable in any RPS function.
The RPS coincidence logic is two-out-of-four.
If one channel
.i s.inoperab]e, operationin MODES.:3, 4, and 5 is allowed to con-icontinue'. OroV'iding: the inoperable'.channel is placed in Ibypass or tfip in1 h6.ur (Required Action A.1).
"I h The `-l hour a-Iloted to-bypass-,rtrip the channel is
- s,,f ici entý,to`.al t
tor"take all appropriate actioný.>for"6the&f6-Ie'I-n `
e, fanh'6eI while ensuring that the risk involved, in operating wi-th the failed channel is acceptable.
-The failed. channei4must*be',restored to OPERABLE status prior
- to entering MODE 2;':fdT:io4ing.the next MODE 5 entry.
With a
' -hannel 'byp'assed:..:theýxtco.'incidencee -logic is now in a to-out-of-three o.gatibn. The Completion Time is based on adequate channe.l: touichannel, independence, -which allows a two-out-of-three channel operation since no single failure will cause or prevent a re.actor trip.
The intent of this requi.r.ement -is.that should a failure occur that cannot be repalired during power operation, then continued operation is allowed without requiring a plant shutdown.
However, the,:failure needs to be repaired during the next MODE 5 outage. ;, Allowing the unit to exit MODE 5 is acceptable, as the appropriate retest may not be possible until normal operating pressures and temperatures are achieved.
If the failure occurs while in MODE 5, then the problem needs to be resolved during that shutdown, and OPERABILITY restored prior to the subsequent MODE 2 entry, (continued)
PALO VERDE UNITS.,2,3 B 1 *3. 2' 10 REVISION-38
PAM Instrumentation B 3.3.10 BASES LOC (Continued).
10Pressur'izer Level C:
Pressurizer Level is used to determine whether to
..,terminate Safety Injection (SI). if still in progress, or to reinitiate SI if it has been stopped.
Knowledge of pressuirizer water-level"is also used to verify the plant.conditions necessary to establish natural circulation in the RCS and to verify that the plant is maintained1in0a safe shutdown condition.
.'At PVNGS, Pressurizer Level instrumentation consists
- 1'"
Iof the following,-
RCA-LT-11OX RCB-LT-110Y
- 11.
Steam Generator Water Level Steamrt Generator Water Level is provided to monitor operation of decay heat removal via the steam geheratO.rsý.
T' e Categbry.I indication of steam generator levelis the wide rahge level 1 nstruumentatj %. ý,The wide range level covers a span
..of 1,43 i s
oiy! the ower tubesýheet to 55.5 inches above the'iS.*am",se'paratpr deck}:< 4. 1 WidRange Steam Generator Level is a Type A variable b'
, because. the:opert.ator *must manually' control steam
.generatLor.1le eldUring*,a Steam Generator Tube Rupture (SITGR) ievent, tQoensure, steam generator tube coverage.
- , At`PVNGS wide:; fiange, Steam Generator Level Instrumentatilort consi'sts of:~
SGA-LT -JI13A-2 SGB-LT-1113B SGG-LT-1113Cý SGD-LT-'I113D-,
SGA-LT'1123A,
" SGC LT'I1-23C SGD-LT-1123D (continued)
PALO VERDE UNITS 1.2,3 B 3.3.10-11 REVISION 50 PALO VERDE UNITS 1,2,3 4B 3.3.10-11 i
,REVISION 50
PAM Instrumentation B 3.3.10 BASES LCO (continued)
- 12.
Condensate StorageTank (CST) Level CST l
- evelisprovided to enure:.water supply for AFW.
The CST provides the ensured' isafety grade water supply for the AFW System.
Inventory is monitored by a 3 ft. to 50 ft. level-indication.
CST Level is displayed on a contrOdl room indicator.
At PVNGS CST Level Instrumentation consists of:
CTA-LT-35 K
CTB-LT-36
-i 13, 14, '15, 16.
Core Exit Temperature.
Core-Exit Temperature.is provjded for verification and long term surveillance Of core cooling.
"An.nevaluation was made of-tthe minimumnumber of valid
,,core ex.it thermocouples necessary for inadequate core
-:cool tngcdetection.. The -evaluation determined the reduced,.complement of Qore exit thermocouples L,.,necessary.,todetecti,ýinitial-core recovery and trend
.,::the :*ensuil ng'icor.e-j::, iatup., --,heevaluations account for core nonuniformities including incore effects of the radial decay power distribution and excore effects of condensate runback in the h ot.legs and nonuniform inlet temperatures.
neseý tp~t'ons..
adqut
'Based ori.,thes eevLuations,.adequate or inadequate core cool i ng' de6'ect6ion.is ensu0r6ed with two vali d core
_-.,exit thermocourlESper. quadrant.
. The design':
'ofthe
'incohe: Instrumentation System
.inludes 61 Typ 51K(
ch'romel alumel) thermocouple within e; a.
- 1. :,: t."". ". ":
- eac
.h'qf
'the 61-,:i-Hc6ýe-ins,trument detector assemblies.
The juncti.or of:eaehli~thermocouple is located a few inches above the fuel assembly, inside a structure that supports :drd shields the incore instrument
'detector assembly strirg from flow forces in the
-outlet plenum'-regiorn.These core exit thermocouples monitorthe temperature of the reactor coolant as it exits: the !fuel.asgemblies.
The.core exit therm6couples have a usable temperature range from 32F-.to 23000F, although accuracy is reduced at temperatures above 18000 F.
(continued)
PALO VERDE UNITS 1,2,3 B ý13:10-142 REVISION 51,0
PAM Instrumentati on B 3.3.10 BASES LCO (continued)
- 17.
Steam Generator Pressure Steam-Generator pressure Indi~ation is provided for Steam Generator. pressure veri-fication.
At PVNGS Steam
,, Generator. Pressure Instrumnentation consists of:
SGA-PT-71013A SGB-PT-1013B.
SGC_-PT-1013C SGD.-PT7EO13D SGA-PT-1023A SGB-PT-1023B SGC-PT-1023C SGD-PT-1023D!
- 18.
týReactor Goolant, System-Subcool.ing Margin Monitoring The RCS Subcooling Margin Monitor is a portion of the Inadequate-Core Cooling (ICC)
Instrumentation required eby Item II.F.2 in NUREG-0737; the post-TMI Action Plan.
The ICCiinstrumentation enhances the ability of the Operator to. antic~ipate 'the approach to, and recovery fror,.ICC, At PVNGS.RCS subcooling Margin
-Moni tori ng.l08strumentati on..consists of:
QSPDS'A"
.QSPDS B :
,r Each channelgof QSPDS processing equipment will
,caculatý tr1**, wi.qfing satur.ation margin parameters:
a')
RS Satl urt i6 h:Mrgin n'
'temperature margin based
.,. on the.difference between..saturation temperature
'and the,m0x'i.mj'um RTD, tempe'rature taken from the hot andc'6,1d-4ý.Lpg
' This._lgorithm uses the hottest
-RCS tempe'ratu*re (Thot or Tcold) and pressurizer
..pressur~e, (PT,1102)j:to-complete the calculation.
b)
CET,,Saturatior, Margin temperature margin based on. the difference between the saturation temperature.-rand the representative core exit temperature,calculated from the CET's.
A representative CET value is first calculated (and displayed on the B02 trend recorder) for the input
-temperatOre., -This is coMpared to pressurizer pressure (PT-102) to complete the saturation S(continued)
PALO VERDE UNITS 1,2,3 3.3.10-13 REVISIOW,50
PAM Instrumentation B 3.3.10 BASES
(.1 LCO
- 18. Reactor Coolant System-Subcooling Margin Monitoring (continued) -,-
2' marg* n ca'alculationx.i. Minimum requirements for CET
-...operabil:ity.must.be'met,before the CET Saturation
., Monitor can be consideredloperable.
c*.
UpperVýHead'Saturation Margin - temperature margin based on'the difference' between the saturation S*"
temperature and theý unheated'junction thermrnocouples.(U.HJTC')
temperature.-., Thi.s algorithm
-. uses,the hofttestoft three upper unheated thermocouples, from. RVLMS.along with pressurizer pressure (PT-102) to Compl'ete the margin calculation.
One OPERABLE Subcoo&:ing'Margin Monitor Channel consists of one RCS Saturation *Margin indicator and one CET Saturation margin indicator.
These indicators
'shall be from'-the-same channel.
Additionally, for any CET Saturation monitor indicator to be considered OPERABLE, the CET:*s.'for-that channel must also be operable'.
1i9)' Reactor Coolant'stm Acti vityv I
.The -RCS.ActlvItY rWvdes an ind:ication of fuel c'addi'ng fai!,ure, Tis ind c ate" "dradation of the first of three o.Sdc'ri hers'-tof'ihsif..n' product release to
'the environme'nt'.
ffih three bar'riers'-to fission product release are (1) fuel 'cla"dding, (2) primary coolant pressureboundary,, and (3) containment, At PVNGS-the 'RCSý.Actfjv%ý,i tInstrumentation consists of:
I
...20;
,SQA-RU-1 50,.
.ý,S B
15 HPSI Syst
-7F Io I.
"HPSI System fioW.indication is provided for HPSI flow
.'veri ficat'ion. "-",..
(continued)
PALO VERDE UNITS 1,2.3 B 31-3.:10-14 PL VD UT 1 3B : 1 1REVISION, 50
PAM Instrumentation B 3.3.10 BASES LCO
- 20. 21 HPSI: System Fl6w (continued).
HPSI System.f.low.,.i,s. a Type.A variable because the operator must manuallybalance the HPSI flow between the hot.and cold!Iegs'when switching from cold leg injection, to a combined cold/hot leg injection in SLIpport. of`LOCA.Long. Term Cooling to prevent boron precipitation in stagnate core areas.
Monitoring of I
thesejinstruments.i.s not. required for initial operation Of.
HPSI fldw:, At PVNGS:" HPSI System Cold Leg Flow S1 id-ication,..consi sts' of J-SIB-FT-0311 J-SIB-FT-0321 J-SIA-FT-0331 J--SIA FTý-0341 AtPVNGS. HPSI System Hot of:
J-SIA-FT-0390 Leg Flow indication consists J-SIB-FT-0391 Two channels are required to be OPERABLE for all but one Function.
Two OPERABLE channels ensure that no single failure wi thin the, PAM in pupent.ation.Q.r its. auxiliary supporting features.'or.- plowe*-§.s
- 'e*,_
concurrent with' failures that are a
,.conditi.on of or.. re.su.1tjfrom a specific accident, prevents the
",ope-raors f rom "bs0 te foprmation necessary for them to, det6rmifne.t4J.*safety. s tatus of the plant and to bring th'e. olanh to and' nal ttýin it in a safe condition following
,IniTable,3.3,.;lO"Jl":the..Q.eption to.,the two channel requirement is COntainrr meht IsoTat'on `Vatlve"Position.
Two OPERABLE channels of core'exit thermocouples are required for each channel in each quadrant to-provide indication of radial distribution of.the,..coolant temperature rise across representative regions of,-the core..
Power distribution symmetry
,was considered in..determining the specific number and locations provided for diagnosifsof local core problems.
Plant specific evaluations in response to Item iI.'F.2 of NUREG-0737 (Ref. 3) have determined that any two thermocouple pairings per quadrant, satisfy these requirements.
Two sets of two thermocouples in each quadrant ensure a single failure will not disable the ability to determine the radial temperature gradient.
(continued)
I-.:.
PALO VERDE UNITS 1,2,3 B,'.1.3;:10-15
-REVISION 50
PAM Instrumentation B 3.3.10 BASES......
LCO (conti nued)
For loop and steam generator related variables, the required
-information is individual loop temperature and individual"
-steam generator level.:- In these cases two-channels are required to'be OPERABLE.for'"each.loop of steam generator to redundantly provide the. necess'ary information.
In the.c.ase.of.Containment Isolation..,Valve Position, the importantinformatiorv.is the-status of the containment penetrations. "The LCO requiresonea position indicator for each active containment' i'solation valve..:, This is sufficient to redundantly veri~fy:the: isolation status.of each isolable
.penetration ei.ther.via,:indi:cated status-,of the active valve and prior, knowledge of the, passiv.e.v.alvelor via system boundary status.
If a normally active containment isolation valve is known to be closed and deactivated, position indication is not needed to, determine status.
Therefore, the position indication for valves in this state is not required to be OPERABLE.
APPLiCABILITY
.The.PAM instrumentation LCO is applicable in MODES 1, 2, and 3.:
These variables,are related to' the diagnosis and
"';pý-eplanned~actions requir~ed to'miti~gate DBAs.
The
,-,..applicable. DBAsareassed-to' 'occur in-MODES 1, 2, and 3.
-. -,,-In MODES.4,. 5, and 6',,ji,n t-conditions are such that the l,
i keli hood.b! an ev ý d
rurrng, that would require PAM instrumehta6tiorn"is l1ow"t.t refor'e,. 'PAM instrumentation is not required to beWFEXABiE;inthese MODES.
ACTIONS A Note has been added in the ACTIONS to clarify the
'appli cation:of ýComplet-ho Time rules.. -.The Conditions of this Specification 'may,;be. entered. independently for each Function--listed in, T&ae:
3: -;,1*
The.Completion Time(s) ocf.ethe inoperable c te.](s:) of:a: Function will be tracked separately: fort each Fr-,-Action, starting. from the time the Condi tion. was entered! or that Function.
(
.o i.n
].
'_.(continued)
PALO VERDE UNITS 1,2,3 B 3 1 A 16 REVISIOW501
PAM Instrumentation B 3.3.10 BASES ACTIONS A.1 (coti nued)....
(,When'one or more Functions have one required chanrnel that is ino'p.erable., : the, required inopera.ble *channel must be restored to.,OPERABLE,status within.30 days.
The 30 day
.Cobmpletion Time'is' based on operating'experience and takes
- 'into account; the remaining OPERABLE channel (or in the case of a Function that.has-only-one required channel, other non-RegulatoryGuide,1 97*.instrument channels to monitor Sthe Function),, the passive nature of -the instrument (no critical]automatic action, is assumed *t o occur from these J instruments),
and thelow probability of an event requiring PAM instrumentation during this interval.
- B.1I -.
This Required Action specifies -initiation of actions in accordance with Specification 5.6.6, which requires a written report-to be submitted to the Nuclear Regulatory...
Commission.
This report discusses the results of the.root..
cause evaluation of the inoperability and identifies proposed restorative Requi red'Actions'. -This Required Actidn is appropriateifn lieu of a shutddwn requirement,
- h.
given the iikelii1d.df plant conditi'ons 'that would require information pr0o,'d-Jbýy _thi~s instrumentation.
- Also, f..fternatidnalR ve6e ct- l.orýis Iar, ei:jde~ntified before a loss
.. ifunctlonfaca a.f
,ty cohdi ti on -occurs.
, Whn orneor more uh.,ctions'-have-.two required channels inoper'able'(-i 'e.:,.- we:channe'ls, inoperable in the same Function)', 'one*.: cianh, a
- ftJ-in the Functi on. s~houl d be restored to OPERABLE.std'ti si-tiiin. 7. days.;, The: Completion Time of 7 Tdays is based-on.-: t' relati-vely; low probability of an event" requi r.i ng' PA'RM iTnstrumentation -operati on and the availability of alternate means to obtain the required information.
Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation, (continued)
-PALO VERDE UNITS 1,2,3 REVISION-50
PAM Instrumentation B 3.3.10 BASES ACTIONS C.1 (continued)
Therefore,'Yrequiring restoration of one' inoperable channel of the Function :l imi"ts the,:risk that :the.PAM Function will be";in..a degraded,.conditionr::should an"acc6,entodcur.
This RequiredAction" diects. entry into the appropriate Condition referenced.in'Table'3:3..1iO0-i:
The applicable Condition, referencedi.in the"Table isFunction dependent.
Each time Required ActinC.,1:.'is not met,.:and the associated Completion Time has expired, Condition D is entered for that channel and provides f-or transfer to. the appropriate.....
subsequent. Condition.
E.1 and E.2 If the Required Action and associated Completion Time of Condition C are not met and Tabl'e 3.3.10-1 directs entry into Condition E, the plant must be 'brought to a MODE in which the LCO does.n66t., apply:.To,..achieve, this status, the plan t 'must be: broudht.lco. ;at least:,.MODE 3. within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and t6o MODE,-4 :withir, :12 'houirS.
r,!The-allowed Completion Times
.;.-are reasobnable,..basedi, operating,exper ience, to reach the equlred plantcondit.*s:fcm fulI power conditions in an orderly y.mnner and te.t't. chall engi.ngplant systems.
(continued)
PALO VERDE UNITS 1,2.3 1 3`3:3.10-18 REVISI.ON.50
PAM Instrumentati on B 3.3.10 BASES ACTIONS (conti nued)
F.1 S..o.Alternatemeansof-monitoring. Reactor-Vessel Water Level, RCS.Activity, and Containment Area Radiati]on have been
- developed andtested.- Thesealiternate:means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time.
If these alternate means are used, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6.
The report provided to the NRC should discusswhether the alternate means are equivalent
... :ito the installed-PAM channels:, justify.the areas in which
..,:they-are not-equivalent; and: provide a schedule for
.restoring the normal PAM channels.
SURVEILLANCE REQUIREMENTS A Note at the beginning of the SR table specifies that the following SRs apply to each PAM instrumentation Function found in Table 3:3.10-1.
SR 3.3.10ý.1,
-.. Performance of thea CHANNEL CHECK.once -every 31 days ensures
- -I
.that ai gross m'faiil.teofo i nstrumentation :has not occurred.
A CHANNEL CHECK ý&sL rvzrrally a compari son of the parameter indicated, onvone. chanineel to-_a:simrilair panrameter on other channel*s<:* It-4s~aed.on-the-asumptionthat instrument channels,.ioni tonili-the; same parameteF'shouldd read approximately the same value.
Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious.
A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability.
If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
If the channels are within the criteria, it is an indication that the channels are OPERABLE.
(continued)
- .PALO VERDE UNITS 1,2.3
- ;. B ý-3. 3.':-10 -19
,-,REVISION 50
PAM Instrumentation B 3.3.10 BASES SURVEILLANCE SR 3.3.10.1 (continued)
REQUIREMENTS If the channels are normally.off scale during times when" surveillance is required, the-CHANNEL CHECK will only verify that they are off scale inthe same-direction.
Current loop channels are verified to be reading at the bottom of.the, range, and not.fal*led.downscale.
The Frequency of 31 days is based upon plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event.
The CHANNEL,CHECK supplements less formal, but more frequent, checks of channel during normal operational use of the displays associated with this LCO's required channels.
SR 3.3.10.2 A CHANNEL CALIBRATION is performed every 18 months or approximately every refueling.
CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor.
The Surveillance verifies the channel responds to the measured parameter within the necessary range and accuracy.
A Note excludes the neutron detectors from the CHANNEL CALIBRATION.
For the Containnent Area Radiation instrumentation, a CHANNEL CALIBRATION as described in UFSAR Sections 18.II.F.1.3 and 11.5.2.1.6.2 will be performed.
,The calibration of the Containment.Isolation Valve (CIV) position indication channels will consist of verification that the position indication changes from not-closed to closed when the valve is actuated to its isolation position by SR 3.6.3.7.
The position switch is the sensor for the CIV position indication channels.
The Frequency is based upon operating experience and consistency with the typical industry refueling cycle and.
is justified by the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift.
PALO VERDE UNITS 1,2,3
-B 3.31.10-20 P V D I 1 33 -REVISION?, 50
PAM Instrumentation B 3.3.10 BASES REFERENCES
- 1.
UFSAR Section 1.8, Table 18-1.
- 2.
Regulat'ory.mGuide,1.97, Revisi6ol 2.;
3' NUREGO7,37.
Supplelnent,,.
7, 7ý PALO'VERDE: UNITS 1,2,3
-B 3.3,10-21
- , REVISION 50
t","
-.5-
-5 Thils. page, intentional ly.jleft blank 1~'
~...
"I S
.-. ~,2i I.
-t
Remote Shutdown System B 3.3.11 BASES SURVEILLANCE SR 3.3.11.2 REQUIREMENTS (continued)
SR 3.3.11.2 verifies that each required Remote Shutdown System transfer switch and control circuit performs its intended function.
The intended functions are:
- 1) To isolate the circuit from the control room.
- 2) To provide the capability to operate the equipment from the remote shutdown location.
This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of the equipment from the remote shutdown panel is not necessary.
The Surveillance can be satisfied by performance of a continuity check.
This will ensure that if the control room becomes inaccessible, the plant can be brought to and maintained in MODE 3 from the remote shutdown panel and the local control stations.
The 18 month Frequency is based on the need to perform this Surveillance under the 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 demonstrates that Remote Shutdown System contro'l.channels seldom fail to pass the Surveillance when performed at a Frequency of once every 18 months.
SR 3.3.11.3 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor.
The Surveillance verifies that the channel responds to the measured parameter within the necessary range and accuracy.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
REFERENCES
- 2.
PALO VERDE UNITS 1,2,3 B 3.3.11-7 REVISION 50
4,. 1.
C
- I Thspaeintent~ional ly, blank :
4'Q.
~I;-.
I
~
-I
Boron Dilution Alarm System (BDAS)
B 3.3.12 B 3.3 INSTRUMENTATION B 3.3.12 Boron Dilution Alarm System (BDAS)
BASES BACKGROUND The Boron Dilution Alarm System (BDAS) alerts the operator of a boron dilution event in MODES 3, 4, 5 and 6.
The boron dilution alarm is received at least 15 minutes prior to criticality in Modes 3-5 and at least 30 minutes prior to criticality in Mode 6 to allow the operator to terminate the boron dilution.
In MODES I and 2 protection for a boron dilution event is addressed by LCO 3.3.1, "Reactor Protective System (RPS)
Instrumentation-Operating."
In MODES 3 and 4 with the CEAs withdrawn, LCO 3.3.2, "Reactor Protective System (RPS)
Instrumentation-Shutdown," provides protection.
The BDAS utilizes two channels that monitor the startup channel neutron flux indications.
If the neutron flux signals increase to the calculated alarm setpoint a control room annunciation is received.
The setpoint is automatically, 1owered to;a fixed amount above the current flux level signal.
The alarm setpoint will only follow decreasing or constant flux levels, not increasing levels.
Two channels of BDAS must be OPERABLE to provide single failure protection and to facilitate detection of channel failure by providing CHANNEL CHECK capability.
APPLICABLE SAFETY ANALYSES The BDAS channels are necessary to monitor core reactivity changes.
They are the primary means for detecting and triggering operator actions to respond to boron dilution events initiated from conditions in which the RPS is not required to be OPERABLE.
The OPERABILITY of BDAS channels is necessary to meet the assumptions of the safety analyses to mitigate the consequences of an inadvertent boron dilution event as described in the UFSAR, Chapter 15 (Ref.
1).
The BDAS channels satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii).
(conti nued)
PALO VERDE UNITS 1,2,3 B 3.3.12-1 REVISION 15
Boron Dilution Alarm System (BDAS)
B 3.3.12 BASES (continued)
LCO The LCO on the BDAS channels ensures-that adequate, information is available to mitigate.the consequences of a boron dilution event.
Alarm capability in the "at the-controls::area-" of-the Control.: Room :is-required for-ýa BDAS channel to.be considered 'operable.. Prompt RESET of the
!alarm is required to maintain operability.
A A'
minimum. of -two BDAS-channels,:ýre required to be OPERABLE.
Because the BDAS utilizes;the:excore.:startup channel instrumentationas its detection system the OPERABILITY of
,the excore startup channeT is~also~part-of the OPERABILITY of the BDAS.
I',
- ?
. *., i APPLICABILITY The BDAS must be OPERABLE in. MODES;3, 4.,..5 and 6 because the safety analysis assumes this~alarmlwill be available in these MODES to alert the operator to take action to terminate the boron'dilution... In-:MODES:1 and 2, and in MODES 3, 4,. and,5, with the RTCBs shut, and the CEAs capable offwithdrawal, the-logarithmic power monitoring channels are addressed as part of.the RPS.in-LCO 3.3.1, "Reactor Protective System (RPS)
Instrumentation - Operating" and LCO 3.3.2,. "Reactor,:P.rP-tecti~ve System (RPS)
Instrumentation-
'Shtldown~i-
-. he'
.d.re~quirementts.
"o*" *,rce range, neutron flux monitoring in "MODE -6';are *addfessed-'id*O3.9.2
-"Nuclear IT.
nstrumentation..",The excore :startup zchannels provide nc-utronr fl ux,coverag etending an;-additional one to two decades.; belh'w'fthe :"logarith.mi:cý channels-f,.or use during
.shutdoWrn 'and. refueli~ng,-,"when: neutron flux may be extremely The Applicability.,.ts) modified by.a Note that the BDAS is requi~red i,n MODE-31wJthin, 41 hour4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br />, after,.the neutron flux is w thin the startup rage fol owi ng a,reactor shutdown.
This a
alowss the neut'ron' f.*uixu l:evel to decay, to a level within the range-of the excore startup channels and for the operator to initialize the-.BDAS... Neutronf]ux.is defined to be within the startup range following a reactor shutdown when reactor power 'is 2E-6% NRTP or: less.. : -,..
-(contln~ued)
'PALO-VERDE UNITS 1,2,3
-B. 3.3.12-2 REVISION,50
RCS Loops -
MODES 1 and 2 B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.4 RCS Loops -
MODES 1 and 2
- , :j",....,. ":":
BASES BACKGROUND The primary function,of the RCS is removal of the heat generated in the fuel due to the fission process and transfer of this heat, via the steam generators (SGs),
the secondary plant.,
- The,secondary functions of.the RCS include:
to
- a.
Moderating the neutron energy level to the thermal state,_toJin~creas~ethe probability of fission; b., ',Improving the,-neutron economy by acting as.a
.',,reflector::;
- c.
- Carrying the'soluble neutron-poison, boric acid:
- d.
Providing a second barrier against fission product release:to the environment;_ and e'.
Removing't-he.',heat generated 'in the fuel due to fission product decay following a unit, shutdown.
1*
The..:RCS configur,*tjonfor. heat. t.ranspprt uses two RCS loops.
.Each'RCS loop*tnrtai,ns.-,aýSG.:and two Reactor Coolant Pumps
.. (RCPs)
., An RCP:-:.* 's,),;!*caLed.,in: each.ofi-,the two SG cold legs.
Thepump flow rat,.e,' !s been sized, to provide core heat
-removal wJith-appocnr, atecmargjin to.Departure from Nucleate
.Boi ling (DNB) during' power. operation. and for anticipated transients originating from power operation.
This Specification requires two RCS loops with both RCPs in operation" in eachý-floop.,%,-The intent of:,the Specification is to re'quirecoreihe~aT,,, rJemoval,with.. for.ced flow during power operati on.: ' Specify.,rng.twoa RCS !.oQps provides the minimum necessary: paths' (,W6* Srs:), f or -he at. removl j'
s APPLICABLE SAFETY ANALYSES Safety analyses, contaihl.various assumpt1ions for the Design Bases Accident (DBA) initial conditions including RCS pressure, RCS temperature, reactor power level, core parameters, and safety system setpoints.
The important
-(conti nued)
- LbPALO A
VERDE UNITS 1,2,3
- B 3.4.4-1
..,:.REVISION 0
RCS Loops -MODES 1 and 2 B 3.4.4 BASES APPLICABLE SAFETY ANALYSES (continued) aspect for this LCO is the reactor coolant forced flow rate, which is represented by the number of RCS loops in service.
.The reactor coolant pumps'provide sufficient:forced circulation flow through the.reactor coolant system to assure-adequate heat'removal from the-reactor core during power operation'.' The.plant is designed to operate with both reactor coolantjloops and associatedý'reactor coolant pumps in.operation, and,.maintain,a'departure'from nucleate boiling ratio (DNBR) above the'D-BR'.Safety Limit during all normal
,operations and..antiCipated transiefts..The safety analyses
,that are-of-Tmostlimportance'to RCP operation are the total los's of reactor coolant flow,.imngepUmpo:locked
- rotor, single pump (broken shaft or coastdown),* and rod withdrawal events (Ref. 1).
RCS Loops -
MODES' 1andi'2 satisfy Criteria 2 and 50..36 (C)(2)(ii) 3 of 10 CFR LCO I I I::,.
,r.
The.,purpbseof. this'LC, is 6t requihe adequate forced flow for tore heat removal,"','Flow is represented by having both "RCSlo6ps with both RCPsinWeah Idop in operation for removal of heat by the two. SGs.
To meet safety analysis
-,acceptance criteria-frOQ'NB,-1four pumps are required at
ýate -power' " --
£...
Each OPERABLE Vwop' Cor s.
1o.fRCPs.providing forced flow for heat transoorLu to an SG that"is OPERABLE.
SG, and
,..hence RCS,!O OPERABLITwith regard.to SG water level is
.ensure'd~by~the Reactor.P.oct "ei Systemn'(RPS) in MODES 1 and 2.'
(continued)
PALO AERDE'UNITS 1,2,3
-B3..4. 4-2 REVISION'50
Pressurizer Safety Valves-MODES 1, 2, and 3 B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 Pressurizer Safety Valves BASES BACKGROUND
'The purpose.,of the four-spring loaded-pressurizer safety Valves is to provide RCS overpressure protection.
Operating n..i conjunction.with the Reactor:Protective System, four valves are used toensure that the Safety Limit (SL) of 2750 psia is ndot-xceeded for analyzed'transients during i.Operation in MODES!12 and'3.
One safety *valve used for MODE 4.. For MODE,5,-.and,.MODE 6 with the head on.
.overpressure protection.iis provided by'operating procedures and the.CO 3A4.113,- Low,.Temperature Overpressure Protection (LTOP) System.
The-self actuated pressurizer safety valves are designed in accordance with the requirements set forth in the ASME, Boiler and Pressure Vessel Code,Section III (Ref. 1).
The required lift pressure is 2475 psia +3%,
-1%.
The safety valves discharge steam from the pressurizer to a-quench-tank located in the containment.
.The discharge flow is indicated by an increase'in.,temperature'.downstream of the safety valves and by ad increasein the quench"tank temperature and
.'level
.. The lift ettingsfor-theaiambientc-c*o Ldtions associated with MODES 1, 2, and 3. This requires eitherthat the valves be set
" hot or that a corQelption between hot and cold settings be
-,estab*:is
- d.
i
'd
-The. pre'ssu'ri zer>.*;fe'tyva]ves' 'are. part o~f-the primary success path and~iitigatethe effects of postulated accidents.
OPERABILITY of the safety-v'alves ensures that the RCS pressure will be limited to 110% of design pressure.
The consequences of exceeding the ASME pressure limit (Ref. 1) could include damage to RCS components, increased leakage, ora requirement to perform additional stress analyses prior to resumption of reactor operation.
(continued)
PALO--VERDE..UNITS 1,2,3
. ;,B:..3..4,,;10-1 RE-VISION. 50
Pressurizer Safety Valves-MODES.1,. 2, and 3 B 3.4.10 BASES APPLICABLE SAFETY ANAL' All-accident analysesin the UFSAR that.require safety valve (SES actuation assume operation of four pressurizer safety val:ves..
to limit:i.ncreasing reactor-coolan.t~pressure.
The overpressure.,protecti~on-analysis is,'also'based on operation of, four safety valves, and assumes that-the.valves open at the hig.h:-range of the.,settihg'(2475 psia.+'3%).
These valves must accommodate pressurizer pressure and volume insurges.that could occur dur*ngtransients due to decrease in heat removal by the secondary, systems, reactivity and power distribution anomalieý, -andincreases in RCS i.nvento.ry,;..Single failue of a sfety" valve' is neither assumed in the accident ahalysis...h6rnr'equired to be addressed by the.ASME Code.
Compliance'With this speci-fication is. reqUired-to.ensure6that -the accident analysis and designbasis calculatifons r~main valid.
The pressurizer safety va'vesi-,satisfy Criterion 3 of 10 CFR 50.36, (c)(2)(ii).
LCO W
-The fou'r pressurizer.safetyvalves are set to open at 25 psia. less than RCS designpr.essure (2475 psia) and within the ASME specified toleancO to avoid exceeding the maximum RCS design pressure SL, to maintain accident analysis
-to cor'p-ly with ASME Code requirements.
. The-1imiit-protected b,,,i s,
"s*
ci ficatioh.is the Reactor Coolant; Pressure' Bbuhý Rd "PsL'of.110%
of design
.pressure...,,Inoperab lJy ',f one" or; more,.valves could result in exceeding the" -Sl.,it r'rainient W'ere'tb occur.
The consequences of exc'eedihg the'ASHE pressure limit could include damage... to.one__or.jnore.RCS components, increased leakage, or additional stress analysis being required prior to resumption..of.-reaq.c: i-op:eration.
APPLICABILITY In-MODES 1, 2,
a'nd-3,-uPERABILITY of four valvesis.-required because the combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents.
MODE 3 is conservatively included, although the listed accidents may not require four safety valves for protection.
(continued)
PALO'VERDE UNITS 1,2,3
-B 3.4.AI-2
.-:REVISION 7
TSP B 3.5.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3:5.6.2 "(continued)
The sample wei'fht and volume correspond to the design minimum concentration"of TSP expected-post LOCA in the contai nment:sbmOj.s.
The limjiting'concentration occurs when the LCO minimum TSP'volume Of 524 cubic feet, weighing about 25, 325 pounds at the intalIed:bulk density, is dissolved into the maximumrecirculation. fluid mass of approximately 7,690,750 pothdsK. Which i's'about 920,000 gallons at room temperature.
Th'6 bboron-cohcent'ration 'of the test water is
- -,the highest pos'sible with the maximum expected recirculation sump volume*.
Ag itacti O' 6of: the test 'solution is prohibited since an adequ-te','stand'ard for the' agitation intensity cannot be specified.
The test time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is necessary to allow time for the dissolved TSP to naturally difffuse through the sample solution.
In the post LOCA containment sump, rapid mixing would occur, significantly decreasing the actual amount of time~before the required pH is achieved.
This ensures compliance with UFSAR Section 6,1.1.2 which requires containment sump pH..to be greater th'an or equal to 7.0 andi less than or e*qual.to8.5-within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s-after a
-Recirculation Actbti~on S'ignal. (RAS).
The tempera1tiu6' 6fM.35 q'9 0F'was- 'hosen,:for the borated
- sol.fc T
6 hat-is'the*minimum temperature
" ~ ~
~
~
~
+.
.epced.atei*if]f*teshutor"cooling heat exchanger's diriY' hel iriti'al';phaseof.this accident when the TSP is di's#lv'61 into 'solution REFERENCES
- 1.
PVNGS operatirnsl'-*cense.'amendmentnumbers 110, 102 and 82 for Units 1, 2 and.3, respectively, and associated
..... NRC..Safety-.E:aluation dated.December 1., 19.96...
.P4LO`VERDE UNITS 1,2,3
- B 3.5.&16-REVISION 50
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This page intentionally blank I~I,.
'I 7..
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Containment B 3.6.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment BASES BACKGROUND The containment consists of the concrete Containment Building (CB), its steel liner, and the penetrations through this structure.
The structure is designed to contain radioactive material that may be released from the reactor core following a design basis Loss of Coolant Accident.
Additionally, this structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.
The containment is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow dome roof.
The cylinder wall is prestressed with a post tensioning system in thevertical and horizontal directions, and the dome roof is prestressed utilizing a two way pattern of tendons, which are an extension of the continuous vertical tendons.
The inside surface of the containment is lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions, The concrete CB is required for structural integrity of the containment under Design Basis Accident (DBA) conditions.
The steel liner and its penetrations establish the leakage limiting boundary of the containment.
Maintaining the containment OPERABLE limits the leakage of fission product radioactivity from the containment to the environment.
SR 3.6.1.1 leakage rate requirements comply with 10 CFR 50, Appendix J, Option B (Ref.
1), as modified by approved exemptions.
The isolation devices for the penetrations in the containment boundary are a part of the containment leak tight barrier.
To maintain this leak tight barrier:
- a.
All penetrations required to be closed during accident conditions are either:
- 1.
capable of being closed by an OPERABLE automatic containment isolation system, or (continued)
PALO VERDE UNITS 1,2,3 B 3.6.1-1 REVISION 0
Containment
. B 3.6.1 BASES (continued)
BACKGROUND (continued)
- 2.
closed by manual val.ves, blincfdflanges.or de-activated automatic valves secured in their closed positions,.except as provided in LCO 3.6..3, "Contai nment Isolationi ValvWs";
- b. '
Each, aj-r lock!;is*OPERABLE-,'except as, provided.in
-LCO33..6.2, "Containment.Air.Locks"2:, and C. `-All equipment hatches, are closied.f I
APPLICABLE Thd safety design basis for the containment is that the SAFETY ANALYSES-containment. must' w~ithstand. t.he pressures and temperatures of the limiting DBA without exceeding t.hedesign leakage rate.
The,limiting.DBAs that -result in
/release of radioactive material within containment,,.are a Loss.;.Of Coolant Accident (LOCA)'
a Main SteamLine Break (MSLB)"
a feedwater line break, and a control.element assembly ejection accident
.. (Ref.- 2.).
.In-the-analysisof each of these accidents, it is
.assumed that containment,is:OPERABLE such that release of
--fission products to,the-envi.ronment is controlled by the rate.of containment leakage.
The containment was designed with"an al lowable ieakage'-rate of 0.
1% of-' containment aai r iZb
'iass per day(Ref.i3)Thjisleakage rate. is definedin 10
- "-CFP 50,,Appendix:',
Optii.
B-(Ref., 1), as La; the maximum alFowable contai.nmen-le.kage: rate,at, the calculated maximum vpeakrconta.inmen..t pre*stune *(Pa) of.-52.,0 psig for units operating at3876, MWt.TPR R;-58 0,.ps-ig, fo'r;units operating at 3990 MWt RIP, wh~ich r*r,.l~ts, from :the iimiting design basis LOCA.
.-Satisfacty eakage-rate,-a test results,,are a requirement for the establifshment.of, containment OPERABILITY.
The containment-sati sfites -Cri:terion 3-of 10 CFR 50.36' LCO Containment OPERABILITY is'.maintained, by limiting leakage to
, 1.0 L,, except. prior,to.the.first startup after performing a required Containment.Leakage Rate Testing Program leakage test. At.this time, the applicable leakage limits must be met..
(continued)
PALO VERDE UNITS 1,2,3 B 3.6.1-2 REVISION 49 CORRECTED PAGE
MSSVs B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Safety Valves (MSSVs)
BASES BACKGROUND The primary purpose of the MSSVs is to provide overpressure protection fohrthe secondary system.
The MSSVs also provide protection against overpressurizing the Reactor Coolant Pressure Boundary (RCPB) by providing a heat sink for the removal of eliergy.ýfrom the-Reactor:Coolant System (RCS) if the preferred heat sink, provided by the Condenser and Circulating Water System;--is--not available-..
F ive MSSVs "are located on;each of the four main steam lines, ou-*
tsi'de' contaVinment,,, upstream of.. themain steam. i solati onp..
valves', as.de~scribed in the UFSAR, Section 5.2 (Ref.
1).
The MSSV rated capacity passes the full steam flow at
-102% RTP (.106%t÷+`:2%for instrument error) with the valves full open."' This meets the requirements of the ASME Code,Section III (Ref. 2).- The MSSV design includes staggered setpoints, according to Table 3.7.1-2.. in the accompanying LCO, so that
.only the number Of valves needed will actuate.
Staggered setpoints reduce the potential:for.:valve chattering if there Jis insufficient steam pressure. to ful.ly; open all valves.
APPLICABLE:-"
The design ;basjs7 for;. the MSSVs. comes from Reference 2; its SAFETY 'ANALYSES purpose is tol.Il.iisecd(ndany.. system pressure to _< 110% of desigh pr'eýssr.e.ssd nt'passingý100% of design steam flow.
This
'desigiIhbasisý:"i.s ýsuffiicitent, to, -co-pe..:wi,.t/hany Anticipated hOperatioal
.Occdrn't-&' (AO.).or -accident, considered in the Desi4n -Basis* Aci ednt (DBA).and transient analysis.
The events that challenge the MSSV relieving capacity, and
- .,thus 'RCS press~ureare.,those.characterized as decreased heat
,"removal events,:.-,and are pres.ented.ini the FSAR, Section 15.2 (Ref. 3).
Of these, the full power Loss Of Condenser Vacuum (LOCV) event 'i;s t-hOl;bl ji tng.AOO.;.,An LOCV isolates the turbine and condenser, and terminatesinormal feedwater flow to the steam generators.
Peak Main Steam System and Reactor Coolant System (RCS)--pressure occur before delivery '.of S,..
auxiliary feedwater to the steam generators.
The peak pressures become high!enough to-actuate'both the Main Steam Safety Valves (MSSVf) 7and:Pressurizer Safety Valves, but remain less than 110% 1f the. design (1397 and 2750 psia for main steam system and RCS, respectively).
The LOCV Secondary Peak Pressure event is the limiting decrease in heat removal transient for determining the maximum allowed thermal power with inoperable MSSVs.
(continued)
PALO'VERDE UNITS 1,2,3 B 3.7.1-1 REVISION'28
MSSVs B 3.7.1 BASES K;XJH APPLICABLE SAFETY ANALYSES The limiting accident for peak RCS pressure is the full (continued) power feedwater' line break (FWLB),
inside containment, with the faiIure'of-thebackfl1ow.:check-.valve-i-n...the:-feedwater line from the affected steam.generator., Water from the affectked steam 'generat6ris 'assumed, to be! lost through 'the' break:with'minimal--addition'al heat transfer from the RCS.
,With'heat removal limited-tb the'una'ffected steam generator,
-the reduced'heat' transfer dauses an -increase in RCS temperature, and the.resui1ting RCS fluid expansion causes an
'increase in* pressure'..'. The.Ancrease in:Main Steam and Reactor Coolant System pres'sure Is mitigated by the relief capacity of the Main Steam' Safety :Valves (MSSVs) and pressurizer safety valves.
The peak`p ressures do not exceed 120% of the design pressure (1524 psia and 3000 psia for main'steam and RCS, riespe.c-tiVely).,fThese results were found acceptable by, the.'NRC ba'sed" 'n the lTbw probability of the event.
In MODE 3. 'one MSSV per 'steam generator (two total) have sufficient relieving capacity to dissipate core decay heat and reactor coolant pump heat to limit secondary system pressure to less than".or equal to 110% of design pressure,
.. as required by ASME Code,Section III (Ref. 2).
A minimum
-of twd'oMSSVs: per stea6ýebe'rat&r are're 'ired to be operable
. i r"*i n
de*3.*irn case'f"-siog] e failure of..one of the valves ir n-ether N stem enea/a
,Jhe. ASSV.satisf.~C~r~6rno 3 of',-10CFR 50.36 (c)(2)(ii).
d LCO
"'This LCO' requi res a"1 "MVs;tobe!OPERABLE in compliance Reference 2, even.,hough this-is',not a requirement of Sthe-DBA--an aIY s iis,.-
Tr11 i's" s*because Ope'ration with less than th&ý--fuil-number orf ýMS3ýSV-requires limitations on allowable th&~ ~
~to DB-nl'i
ý ' 2 beasTprtonwt esta THERMAL -POWER (to'meCRee'"erence 2Zrequirements),
and
'djusttment "to.te, Rtea'"dr Prbtective&System trip setpoints in Modes 1-and 2:',Thesý;Ll, imitations are according to those shOwn ýiri':Table' 3.7 1'**'andARequired Action A.2 in the accompanying LCO.
Since* the VOPT is not required to be aopeanbleyin MODE iace the VOPTo nSS 3.3r1 and 3.3.2, a note
'has been-addedto-Table 3T.7.1-1 stating that the VOPT setpoint is.notrequired to be reset in MODE 3.
An MSSV is considered inoperable if it fails to open upon demand.
The OPERABILITY of:the MSSVs is defined as the ability to open within the setpoint'ýtoierances', relieve steam generator (continued)
'PALO VERDE UNITS 1,2,3
.: - 7 " -7.1-2 B -3.
ý REVISION'50
ADVs B 3.7.4 B 3.7 PLANT SYSTEMS B 3.7.4 Atmospheric Dump Valves (ADVs)
BASES BACKGROUND
!'The ADVs prbvide a.safety.grade method fo,r cooling the unit to. Shutdown Cbool.i~ng :(SDC)..System entry conditions, should the preferred h'eat sink. via the Steam Bypass Control System to the condenser not be. available, as. discussed in the FSAR, Section'10..3 (Re*f*.1). This'is done in conjunction with the
, Auxiliary Feedwate System providing. cooling water from the
-:Condensate Storage Tank-,(.CST)..,The ADV.s may also be Sirequired to meet *the design cooldown rate during a normal cooldown-...
~1)
Four-ADV l hes. are provided.-. Each ADV.line consists of one
-ADV and an associated biock, valve.
One ADV line per steam generator is required to meet the assumptions in the safety analyses.
The ADV block valves are not required to be closed in the event.of a stuck open.ADV.
- 4 The AD/s are &equipped with 'pneumatic.,'controllers to permit control of the/cO0]:own rate.
The ADVs are pr'o'viaed,.wi.th a pressur.ifzed.gas supply of
.bottlednitroenýthat, o6n a ldss of pressure in the normal instrument air s y:ýi autombaticall. ysuppl ies nitrogen to operate the ADVs.'
The nitr6gen supply i's sized to provide
- I I. suffic:ient.,press.R:uizedgas to2 operate the ADVs for the time required for RCS" coold6wrl to the'SDC System entry conditions,.as-des*tribed-in UFSAR Appendi x-- 5C.;-," Natural.
ICi rculation Coo,.*wnn Analysis..
The Appendix 5C analysi,s..
sý
- ibased:
onth'e,',a~ssuuipt'ions' a'nd-condi tf~ons i n the NRC's..
-Br anch Technical -PosJition. (BTP).,RSB 5-1,:"Design IRequi rementsof* te si-dualP Hea-t Removal System." RSB 5-1 is an attachment-<.to 3ý,S1-andard:.eviewPlan (SRP) 5.4.7,
.-'Residua-l Heat.RemPqva i.. (RHR)
System-.", and identifies RHR System requi rement$s:-that ensure conf!drmance with General IDesign_%Criteria..(iGOC),- 34, "Residual Heat. Removal."
The PVNGS RSB 5-1 -cooldown scenario described in UFSAR Appendix 5.C is.
basedL on -a natural. circulation cool down with
.both steam generators,-(-SGs).'.available,- using safety-grade equipment, assuming a loss of offsite power, a limiting single failure (assumed to be a diesel generator failure),
'-and with minimal operiator-actions outside the control room,
.-as approved.by.the NRC.- The RSB 5-1. c.ooldown duration was (continued)
PALO'VERDE UNITS 1,2,3
-B, 3:.7. 4-1I REVISION 50
ADVs
-B 3.7.4 BASES BACKGROUND (continbued)
'stablished during actual testing performed in January.
-1986;,
and was conifirmed.through,,subsequent. analyses to address>, steam 'generator! rep'l acement and? power uprates.
.A.',desc'ription of 'theJADVs:-i's 'found in-.Reference I.. -The ADVs
.*.require'both DC.sources and:class ACiý:instrument power to be considered OPERABLE..
In addition, handwheels are provided for local manual, peration.:
APPLICABLE.
SAFETY, ANAL The design basis of -the `ADVs i-sest'ablished by the
.YSES capability to'cool the un'it.to,SDC' System-entry conditions.
A cooid'own.rate of '75F ý'per hour.'isobtai'nableby one or
'both steam generators. :Thits design, isadequate to cool the unit to.SDC System entry..,condi.tionswith only one.ADV and
-one steam generator, Iutilizing the cooling water supply available in the CST.
Cooldown scenarios using a single
,ADV may require a combination of, the available nitrogen ubpp)Y*a*'dical-lahual rperatio"or'other actions:.
a',?
Alternatives for cooldow*n and for 'ADV.operation beyond the RSB 51,scenario have'. biee evaluated using probabilistic rfsk 'analysils (PRA).,* s"I '.rtct'.the.
resolution of Unresolved Safety'Issue (USI) 'A-45'" "ShitdoWn. Decay Heat Removal Requirements."
USI A-45 Was'ssubs'umed into the Individual
,'. ant;Examination (!P..)*'Th~ch. used PRA techniques and was submitted to the NRC* in'response to Generic Letter 88-20.
The IPE considered various operator actions and the use of hon-safety 'related equTpment,-:and'concluded' that-there are.
no si~gn.i'ficant, heati pemoval: vulpnerabilities at PVNGS.
Operator actionsi~tba ldc*-lj operate,theADVs are not credited in the :UFSA1ojApte" 15"accidehnt analyses but are described in the EO-o n..-safety related equipment such as the supplemental :njitrogen,,-supply could also be used during xtendedcool down, ss*uati.,,
l...
In the accident analysis presented in' the UFSAR, the ADVs are assumed to be used by the operator to cool down the unit to SDC System entry conditions for accidents accompanied by a loss of offsite power.
Prior to the operator action, the Main Steam Safety Valves (MSSVs) are used to maintain steam generator pressure and temperature at the MSSV setpoint.
This is typically 30 minutes following the initiation of an event.
(This is less for Steam Generator Tube Rupture (SGTR) events as detailed below).
The limiting events are those that render one steam generator unavailable for RCS heat removal, with a (continued)
PALO VERDE UNITS 1,2,3
- B::3,.: 7.4-2
- REVISION.50
ADVs B 3.7.4 BASES I
APPLICABLE SAFETY ANALYSES coincident los,o.f offsite-,power:..:-this results from~na.Atuhbine (conti nued)
-trip-.
Typical--i:ni ti at! ng, events 'fal l i ng i nto this dategdry are
...a ma-in steam' line. break upstream of: the.mai n steam i sol ati on valves, and a feedwater line break.
For the SGTR and SGTRLOP
.events, ADV's arejassumed to-be opened two minutes post trip to prevent cycling of Main Steam SafetyValves (MSSVs) and they
.,remain open.unti2 the affected SG is,isolated.
From then on, the ADVs on the unaffected SG.:.is-us.ed till shutdown cooling entry conditions are reached, The *imiting design basis event for nitrogen supply capacity
.s' t-is hei-'RSB"5-1"natura'l circulati6ri'-cooldown scenario-described A.above
-Thi s scenariO inicludes an initial period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at
- Thot standby conditi6ns follow`d by natura circulation I
. cool.down'. for' 9.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. until..SDC entry conditions are achieved., -Eac6h.:ADV-is required to have a nitrogen supply that
- .spports" ADV operation for a total of. 13.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
The Steam Generator witha Loss 6f Offsite Power and a Single Failure (SGTRLOPSF)-eventý assumes an'-ADV on the affected SG sticks open 2 minutes post trip for the duration.
The
.credited operatogr'ation of~directingauxiliary feedwater to the affected'SGIQ'ep*-the-tUbes covered-,
Thus the majority of
. the heat retova]iduring this'evenf'is c6nducted through the affected.SG AQV.,
...The ADVs-satlsf&*Cr'telion!30 f:10 CFR 50.36 (c)(2)(ii).
i LCO One ADV(li is..r 9red-"tot`be OPERABLE On each steam generator Udondjct ajiun-i1t cool down following an event in
"* which-one steam':-h 'rat6r 'beomes unavailable.
Failure to meet 'the
'C0-can, ri*-
-t' in thfe inab'il.i-ty-to cool the unit to SDC System erftryý.dbdiitionsýfollowing an* event in which the condenser is ýuriaýAJMabl e for use with--the Steam Bypass Control System..
~
~
4 (continued)
- PALO VERDE UNITS 1,2,3 B.,1.7.4-3 REVISION 50
ADVs B 3.7.4 BASES LOC An ADV is considered OPERABLE when i.t. is capable of (continued) providing a controlled relief of the main steam flow, and is
.capabl e1offul ly' open*n n and.-closing on demand.
APPLICABILITY,
.',rIn.MODES"1, 2, and 3 and in MODE 47 when steam generator is being relied up o n' for heat'..removal, the ADVs are required to be OPERABLE.-
In MODES-5 and 6, ah SGTRMi'gnot.a:credible event.
ACTIONS
-,With one required ADV line inoperable, action mus{tbe-7taken to restore the OPERABLE:status within-72 hours.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time takes into accouht the availability of a
a nonsafety grade backup in the Steam'Bypass Control System and kSSVs.,
- .Bi..-,
.. :2 Wi]th two requi red "ADi' n~esinoperable (one in each steam
-generator), action mfit be' taken to-restore one of the ADV lines to OPERABLE status.
As the block valve can be closed to--isola1te an' ADV;-some-m'epairs may--be possible with the-unit at power.
The. 2.4 ur Completion Time is reasonable to repair inoperable ADV ine,- based :,on the availabIl-ity of
. the Steam Bypass.Contro-l.System and.MSSVs, and the low probability of an event occurring dd'ingtlii S petiO4d that requires the ADV lines.
(continued)
PALO VERDE UNITS 1,2,3 I a Bý 3.7.4-4 REVISION 50
ADVs B 3.7.4 BASES ACTIONS C. 1 and C.2 2
' If the,ADV lines: cahnot be restore'd to. OPERABLE status within the associated Completion Time*' the unit must be placed in a MODE in which the LCO does not apply.
To achieve this-status, the unit must be placed in at-least MODE 3.within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,. and in MODE 4, without reliance on I the:steam,IgetneIrat rbr for hea~t removal",- within 24.hours.I The allowed Completion Times 'are reasonable, based on operating experience, to reach the required unit conditions from full
...power conditiors..,in an orderly manner and without
'Challenging unit'Systefms.
SURVEILLANCE REQUIREMENTS SR 3'7.4.1
,1To perform a"contr'olledcooldown of the RCS, the ADVs must be able to.be'opened and throttled through their full range.
This SR ensures the ADVs are tested. through a full control cycle at least once per fuel cycle., Performance of inservice testing or use of an ADV during a unit cooldown may satisfy this requirement.
Operating experience has shown that these components usually pass the SR when performed at. the-,18,month. Frequen cy,.
Therefore, the Frequency...i]s. a~d
~l e.
from a' rel' iability standpoint.
d, REFERENCES-
- 1.
- UFSAR S" 1
".'3 PALO'WVERDE UNITS 1,2,3 7
B 1.7.4-5
,REVISION 50
V1
- t.
~1~
II This.pagel.,intentionallyIleft blank
- 1*.~
.;-'*.c'
- I*
C I,,
.°
CREFS B 3.7.11 B 3.7 PLANT SYSTEMS B 3.7.11 Control Room Essential Filtration System (CREFS)
BASES BACKGROUND The CREFS provides a protected environment from which occupants can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke.
The CREFS consists of two independent, redundant trains that recirculate and filter the air in the control room envelope (CRE) and a CRE boundary that limits the inleakage of unfiltered air.
Each CREFS train consists of a prefilter, a High Efficiency Particulate Air (HEPA) filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodine), and a fan.
Ductwork, valves or dampers, doors, barriers, and instrumentation also form part of the system.
A second bank of HEPA filters follows the adsorber section to collect carbon fines, and provides back-up in case of failure of the main HEPA filter bank.
The CRE is the.area,.within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions.
This area encompasses the control room, and may encompass other non-critical areas to.which frequent personnel access or continuous occupancy is not necessary in the event of an accident.
The CRE is protected during normal operation, natural events, and accident conditions.
The CRE boundary is the combination of walls, floor, roof, ducting, doors, penetrations, and equipment that physically form the CRE.
The OPERABILITY of the CRE boundary must be maintained to ensure that the inleakage of the unfiltered air into the CRE will not exceed the inleakage assumed in the licensing basis analysis of design basis accident (DBA) consequences to CRE occupants.
The CRE and its boundary are defined in the Control Room Envelope Habitability Program.
The CREFS is an emergency system.
Upon receipt of the actuating signal(s), normal HVAC to the CRE is isolated, and the stream of ventilation air is mixed with outside air and recirculated through the filter trains of the system.
The prefilters remove any large particles in the air, to prevent excessive loading of the HEPA filters and charcoal adsorbers.
(continued)
PALO VERDE UNITS 1,2,3 B 3.7.11-1 REVISION 50
CREFS B 3.7.11 BASES 2i BACKGROUND""
(continued)
,Actuation of CREFS aligns-the system for recirculation of.
the air.withilnthe:CRE through the redundant trains of HEPA" aand-,charcoal.-f'lters.,
Actuation ofthe:CREFS also initiates pressurization and filtered ve~ftilation.6f the air supply to the CRE.
,Outside, air. is combined:., and,. fItered with the air being
- recirculated from,the..CREE Press'r ization of CRE minimizes
. infiltration ofunfilteredair from"a]-L,.the surrounding areas adjacent to: the CRE bo0udary.
.- The airý-entering the CREis,ccntinudusly monitored by
,.rad.iation detectors..,, One detector output above the setpoint will cause actuatioh. of.theCREFS trains.
A single CREFS-train operat.rg at. a ff6w rate of :1000 cfm isdesigned to pressurize the CRE to.:0.125 inches water gauge relative to external areas adjacent to the CRE boundary., TheCREFS.-operation in"majntaining the CRE habitable is-discussedin the UFSAR, Section 6.4 (Ref. 1).
Redundant recirculatidon.trains provide the required filtration.
Normally 6.jen isolation dampers in the normal Cohtrol-Room HVAC System-are arranged in series pairs-so.
- -that the faiure-ofone-,damper,,to shut will not result in a
., reac;ofisolfatioi-',:,
REFS i's designed in accordance
. thSelismicC&ateg61',!j i,
en; oYi Ajqui remient.
The.CREFS 's. desigded
-t4naintaina habitable environment in the, CRE for 30 da`s-,Qf-cQhtin6o6sloccupancy after a Design Basis. Accidnt.,,(DBAYwithout exlceeding a 5 rem whole body., dose or -its -q*uii.*le'nt to any part.of the body to the CRE occupants i n the evenit of a large radioactive release.
APPLICABLE SAFETY ANALYSES The CREFS components. :re.,arranged in redundant, safety related ventilation trains.
The location of components and ducting within the CRE;-ensures.an>adequate supply of filtered air to all areas requiring access.
(continued)
PALO VERDE UNITS 1,2,3
ý 3.7.11-2
.-REVISION 50
CREFS B 3.7.11 BASES APPLICABLE SAFETY'ANALYSES (continued)
The CREFS provides airborne radiological protection.for.CRE.
'occupants, as.demonstrated by the"CREoccupant dose.,
'analysesforýthe'nmost limiting desighnýbasis accident
.'.fission produc.t;,r'el.ease' presented i n, the'UFSAR, Chapter 15 (Ref. 2).
The CREFS provides protection from smoke and hazardous chemicals to the CRE occupants: however, *hazardous chemi cals* a"re not..tored or*bsed onsite. in quantities sufficien to' 'necessitate CRE protectionh, as required by Regulatory' Guide.1.78-Ih addition, nearby industrial, military, and transportation facilities present no hazard "o th6 operation of PVNGS,";and-,there are' no site-related design bass: Sevents due to accidents at these facilities
('Ref. 1 and.'Ref. 3).
The evaluation of a smoke challenge demonstrates that it will not result in the inability of
'the CRE occupants' to control the reactor-either from the control room'or froni.the remote shutdown panel (Ref. 4).
The Worst case.singl'e active failure of a component of the CREFS, assuming a loss of offsite power, does not impair the ability of the system to perform its design function.
The CREFS satisfi'es.Criterion 3 of..10;.CFR 50.36 (c)(2)(ii).
LCO
- Two:. ndepend enri'
-_ý edundant trains of"'the CREFS are required t6 be'.O-bý!ABLEto'-ensur-t'rat"',t least one is availIab1'ief aslhg e' ative failUre dis'ables the other train.
Total system failure, such as from a loss of both
,'vent'ilationrtrlai'.roi"-",rom-an:ln inoperabl,6CRE boundary, could
ýr, 'es'l'tii n'] Jnxce.deing,"- ýose of 5 rem"'Whol b body or its
'equ.ilval ent to ahY-- bart, of the' body to the CRE occupants in the 'eventof'-a,a'e!r'adi'akctive release.
Each CREFS train is considered OPERABLE when the individual components necessary--to.- Timi t CRE.- occupant.-exposure are.
OPERABLE in both trains.
A CREFS train is considered
'OPERABLE Whenb'th' aqssc.ciated:
aý.:
Fan is OPERABL L
- b.
HEPA filters and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration functions; and I
- c.
Ductwork, valves. and dampers are OPERABLE, and air circulation can be maintained.
(continued)
PALOIVERDE UNITS 1,2,3 B 3.7.11-3
,_ :.' RE.V1,SION 50
CREFS B 3.7.11 BASES
?.;
LCO (continued Ih!"order'for the CREFS trains to-be6considered OPERABLEý,
i).
the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed,the caldculated dose:.,iin.the licensing basis consequence analyses.for DBAs,,and that-the CRE occupants are protected fromnhazardouss-chemicals and smoke.
- : The LCO -i s modi~fi edr byý a 'Note6 al I ow'ing: the CRE boundary to be opened intermitt-ently U'nder administrative controls.
This Note only, applies to: openings. in.the. CRE boundary that can be rapidly restored.*.toý:the design:condition such as
""doors,.hatches., floor: plugs..and accessipanels.
For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area.
For other openings--, these' controls should be pfocedbralized and-.consist of stationing a-dedicated
- i ndivi'dual at the6peni ng-7who 'irs i n co'ntinuous communication with,:the-operators in the'CRE.
This individual wil:l-have a-meithod to,rapidly'close the opening and to restore the CRE boundary integrity to the design condition when a need for CRE isolation is indicated.
APPLICABILITY In MODES 1, 2, 3, 4, and during movement of irradiated fuel
'.i-6
"-"- *ss~inbiies, -.the'REFS.',:;ulýt be'.-COPERABLE td ensure that the C
R, E REwi 1-1',b reaai ri-habi tb !,:doring'and.following a DBA.
In NODES 5 :and' 6, the RFSP is requiredto cope with the releasfroma.trptur*V awaste-gas: tank.
-.Movement of spent fue*;*;caSks.scontaining..irradiated fuel assemblies is.-not with1t. the scope of the Applicability of this,technical specif iat-ion: The movement of dry casks
"' contairning irradi atod fUel assemblies:wil1 be done with a
`ihgefai lure-orofb hrahdlling system 'and with transport equipmedt that wo*ýd...prcvnt any,credible accident, that
-could result:.in 4*aý'lease of radioactivity.
During movement -of i-rradiated fuel assemblies, the CREFS
'.must be OPERABLE tocope withthe release from a fuel handling accident.,
PALO VERDE UNITS 1,2.3
'B 3.7.11-4 (continued)
REVISION`50
CREFS B 3.7.11 BASES ACTIONS A;I
-.With' one CREFS.train inoperable. for.:reasons other than an
.,inoperable CREL boundary,,.. action must be.taken to restore
-OPERABLE statusý.within 7Tdays.
In.this.Condition, the remaining OPERABLECREFS train is. adequate to perform the CRE occupant protection function.
However, the overall reliability.iJszreduced because, a failure in the OPERABLE CREFS train couldresult in.loss of CREFS function.
The 7 day Completionj:Time is based on theTowprobability of a DBA occurring duringthis time period, and the ability of the remaining train ito provide the, required capability.
-B1. B.2, and B3.3
" If 'th6'eunfiltfer.d.',ailr leak'age of potentially contaminated air past:'tbe CREjboundary and into the CRE can result in CRE
.:occu ant radiological dose greater than the calculated dose of.the'licensingbasis analyses.of DBA consequences (allowed to be up to 5 rem wholebody-or its equivalent to any part of the body) or inadequate protection'of CRE occupants from hazardous chemicals or smoke, the CRE boundary is inoperable.
Actions must be taken to-restore an OPERABE-CRE-boundary within 90 days.
Duri;ng the;,eri,,hat.the,. CRE bounda'ry.; is considered
!J noperabl e,.: acti oapFnust:be;beiAniti ated to-i mpl ement mitigating actions to lessen the effect 6nCRE oc6"pants from the
..,potential hazardsf~adiological 'or chemical event or a
.challenge" from-s"oke--... Actions-.mus.t' be taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to verify that in the event of a DBA, the mitigating actions
..wi:ll ensurpe ethat.;CR-z.:.gccupant radiological exposures will not exceed the, ca.l.culraeedose, of the.icensing basis analyses of DBA, consequences.*..*nd-.that CRE ;ccupants, are protected from hazardous:.chemiQals 0nd smoke.
These mitigating actions (i.e...iactionst.haVt~are token1to offset the consequences of
. the.inoperabl.e.,CREbo.bQdary-) shouldbeb preplanned for impl ementat-ion.: u.por.;etny -into.the condition, regardless of whether entry is intentional ortunintentional.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time, is, reasonable ýbased upon the low probability of. a DBA occurring during this time period, and the use of mitigating actions.
The,-90 day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect (continued)
PALOýVERDE UNITS 1,2,3
ý B 3.7,.11-5 REVISION 50
CREFS B 3.7.11 BASES -
ACTIONS B.1, B.2': and B.3 (c*ntinued:) '
their abil ity'to control t6ereacto~rand' maintain it in a
.safe:"'.shutdown., cbnditionn hth e'event of a'DBA.
In addition, the90.day Cpleti.on T~it isa reasonable time to diagnose, plan and. 'possibly 'repair.and-te'st most, problems with the CRE b o u n d a ry 7 "
C.1 and C.2
,In MODE 1.'2'3.
3or 4!, if
'the'inoperable, CREFS or the CRE boundary cannot.be restored tc OPERABLE status within the required Compl-etibnTime, the'unit:m st-be placed in a MODE that mnPnimizes'th6eaccident'risk..i'.To achieve this status, the uni.t 'ust be placed i:n"..,at Ieast MQDE' 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
. and in MODE 5 within 36'ho, u si's' The allowed Completion Times.are reasonabl e,o base pera'ng perience, to reach 'the required unit condi~tions from full power conditions in an orderly manner *and without challenging-_
unit systems.
n-E-
" *D I ;,
'o 6...
.. i 10 -
Ac i n A"I c InMODE 5 or 6.,'if ReLured Action A'.1 cannot be completed w.,"ithin the'" r~qirne ;diipltioh Time! 'the OPERABLE CREFS train must be, :immediate'ly., p**cýýdinr{ tnie emierge'cy mode of operation n(i.e.,
fA runningvalves/dampers a6ligned to the post-CREFAS mode, etc.).
This action ensures. that the remaining train is OPERABLE, that no failures preventing automatic actuation.
will occur, and that" any-acttive fai lure' will be readily detected.
'.. E.. 1 arid E".
t
-'-.w
.', During mov'ement -of'irradiated fuel. assemblies, if required Action A."1 tann'ot-'btompleted wi'thin the required Completion Time, the OPERABLE CREFS train"must be immediately placed in the emergency mode of operation (i.e.. fan running, valves/dampers aligned'to'thepost-CREFAS mode, etc.) or movement of,-irradiated'fuel assemblies':must be suspended immediately.
The first' action "ensures that the remaining train is OPERABLE,' 'that no undetected'failures preventing system operation will, occur, and that any active failure will be readily detected.
(conti nued)
VERDE UNITS 1.2,3 B 3.7.11-6
. REVISION 50 I
PALO
C R E F S B 3.7.11 BASES ACTIONS E.1 and E.2 (continued)
..An alternative, to Requi.red 'Action El. 1..is, to immediately suspend
- -,actiVities that,could'`resuit 'in a rele'ase of radioactivity that:
,might require isolation oftheCRE.
This places the unit in a condition that minrimizesthe.accident risk.
This does not preclude themovement of fuel to a*safe position F.1 and F.2 If two CREFS trains become inoperable for reasons other than an inoperable'CRE boundary or one or more CREFS trains become
' in°operable due to an inoperable CRE boundary, during Mode 5 or
.6. or. duori ng.the, movement ý6f. i rradi.ated fuel assemblies,
.imrediate actiion' must be taken t suspend.activities that could release radioactiviity that, might enter.the CRE.
The Required Actions..lage the uLnit inra condition that minimizes accident risk.
These.actions do not preclude movement of fuel
- assemblies to.safe positions.
G.1 If both CREFS trains are inoperable in MODE 1, 2, 3, or 4 for
..reasons othepr.
pn inoperable.CRE.boundary (i.e.,
Condition-B*,.th6 *,EFS may not be capable of performing the
" intended fuhntid&'o d fh6e.nit.,ist inn a-' condition outside the accident ana yse~,' Therefore LCO3,.0 3must be entered
- ..'. mmed~iately."
SURVEILLANCE SR 3.7.11.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.
Since-the env'ironment and normal
. operati ng condi.tjQný,, onthi s, system.are.not severe, testing
-eachtrainonce e&ve~ym'nth brovides,,an",adequate check on this system.,
Monthly operateins tot _ 15 minutes todemonstrate the function of.,the.,system 45s required.
The, 31 day Frequency is basedon the known reliabiflity of the,.equipment, and the two train redundancy, available.
(continued)
-PALO VERDE UNITS 1,2,3 B 3.7.11-7 REVISION'50
CREFS B 3.7.11 BASES SURVEILLANCE REQUI'REMENTS (continued)
SR 3.7.11.2 I !: -ý 1 I
This ISR; Verifi es' that the-:reqoi red. CREFS :testing "i s' per~formed in accordance With. the Ventilation Filter Testing Program (VFTP)
' The CREFS~filter te~stsare in accordance
'with Regulatory Guide.1.52! (,Ref.. 5).
.:The VFTP includes testing HEPA filter perfornmancei charcoal adsorber efficiency: minimum system -f-low rateý, and the physical prolpertiesý-of the acti~vated charcoal (general use and foll1owing speci fi c: operatios). - Specific test Frequencies and "addi:ti onfal i nformati'on,.are&rdi scussed i n detai I i n the
ýVFTP.
I SR 3.7 11.3 I
li Thi s SR verifies t6aa e'ach*- REFS't. ai n starts and oe'rtesi on an actual or simulated actuation'isignal..' This includes verification that the system is a'utomatically placed into a filtration mode of operationwith'flow through the HEPA fi Iters and charcda'li-1ds6rber batiks. "The. Frequency of 18 months is based on industry operating experience and is consistent with.the typical,refueling cycle:
,D T 7 11 A
This SR verfiees the 0 jer,1 ity' of the CRE boundary by and:-iinto,-the CRE.,
Thj*:
tjfls of the testing are specified inthe Control *RobmoEn, eope Habitability Program.
The CRE is-considered.habitable when the radiological dose of CRE occupants calculated""in thelicensing basis analyses of DBA consequences:.is..n.o-.more than 5 rem.whole body or its equivalent to any part"ofthe b at ehdtheCREoccupantszare protected from hazardous chemicals and smoke.
This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences.
When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered.
Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident.
Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref 6) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 7).
(continued)
PALO VERDE UNITS 1,2,3 3 3.7.11-8 REVI S I.OIN : 50
CREFS B 3.7.11 BASES SU RE nt',rIrTt I hld/"12 MVLrLLM*kI L rR 3J.I.11.4. ý,LUIIL IIllU U)
QUIREMENTS
.:.,These.compensatory-measures may also: be:.used as mi.ti gati ng
-,actions as required by:Action B.2.
Temporary analytical
-methods may also be used as compensatory measures to
- restore operabil.ity. (Ref..,8).: Options for restoring the CRE boundary to OPE.RABLE; status..incl ude. changing the licensing, basis DBA consequence analysis, repairing the CRE boundary...or a comb-in~ation:-of:.these actions.
Depending on the nature cf. the,.problem and the corrective action, a full scope inleakage.tes~t may not [be necessary to establish that the CRE boundary has been restored to OPERABLE status.
REFEREHNCES:.,/
.. 1 UFSAR.,-Section 6.4..
2
,'UFSAR,' Ch'ap5ter-15.
-3.
3
'UFSAR, Section. 2.2.3..
- 4.
UFSAR, Section 9.4.
- 5.
Regulatory. Guide.1.52*(Rev. 2).
- 6.
- 7.
NEI 99-03, "Control Room: Envelope-Habitability Assessmeo_,".4une.20.01, 8*1<
Letteri'frdoaiEkYc-,-,uJ.
Leeeds:4'N RC.).;to.James W. Davis
-200l
- 4.i"NElDraft White Paper, e
Useof GenefWn&-;.Le'tter 91'-18' "Process and Alternative Source Terms in the Context of Control Room
'Habitabi~l~i"-ty ý"i-;(":ADAMS Accessidn No. ML040300694).
- I C. "
7PALO.VERDE UNITS 1,2,3 B '.'ý '17'. 1`1 -9 REVISION 50
ii'-:
~
C. -.
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- '(
t
-~
.12
?2..
a*- '-
t 2- -
-I" i
AC Sources - Operating B3.8.1 BASES APPLICABILITY:.
The AC power requirements for MODES 5 and 6, and during (continued) movement of irradiated fuel assemblies are covered in LCO 3.8.2, "AC Sources - Shutdown."
ACTIONS Condition A applies only when the offsite circuit is unavailable to commence automatic load sequencing in the event of a design basis accident (DBA).
In cases where the offsite circuit is available for sequencing, but a DBA could cause actuation of the Degraded Voltage Relays, Condition G applies.
A note prohibits the application of LCO 3.4.0.b to an inoperable DG.
There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable DG and the provisions of LCO 3.0.4.b which allows entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
A.1 To ensure a highly reliable power source remains with the one offsite circuit inoperable, it is necessary to verify the OPERABILITY of the remaining required offsite circuit on a more frequent basis.
Since the Required Action only specifies "perform," a failure.of SR 3.8.1.1 acceptance criteria does not result in a Required Action not met.
However, if a second required circuit fails SR 3.8.1.1, the second offsite circuit is inoperable, and Condition C, for two offsite circuits inoperable, is entered.
A.2 Required Action A.2, which only applies if the train (i.e.,
ESF bus) cannot be powered from an offsite source, is intended to provide assurance that an event coincident with a single failure of the associated DG will not result in a complete loss of safety function of critical redundant required features.
These features require Class 1E power from PBA-S03 or PBB-S04 ESF buses to be OPERABLE, and include:
charging pumps; radiation monitors Train A RU-29 and Train B RU-30 (TS 3.3.9), Train A RU-31 and Train B RU-145; pressurizer heaters (TS 3.4.9); ECCS (TS 3.5.3 and TS 3.5.4): containment spray (TS 3.6.6): containment isolation valves NCA-UV-402, NCB-UV-403, WCA-UV-62, and WCB-UV-61 (TS (continued)
PALO VERDE UNITS 1,2,3 B 3.8.1-7 REVISION 42
AC Sources - Operating B 3.8.1 BASES ACTIONS A.2 (continued)
I S3.6. 3).; auxiliary feedwater{ system.(TS;'3. 7..5); essential cool i ng wate'r system (TS*'3'.-7),; essentitalf spray pond system F.(TS.`3..7 8) essential chilled water system (JS 317.10):
.control.room essenti~al 'f.iltfration sys.tem (TS 3.7.11) control room emergency a~ir temperature control system (JS 3.7.12):
ESFpump room,. air exhaust cleabnup,.sysyem'.(TS 3"7.13);
shutdown. cooling""subsystems.(TS 3.4,6,, 3'4.7, 3.4.8, and 3.4.15)., and fuel' buildifng,'vientilation;
-Mode applicability
-isa s specified in each-apprbphiate.TS section,.
The Completion Time for Requi-re'd ýAction' A.2 is intended to allo the operato itime to evaluateiatid."repair any d'isc6vered 'inoperabil'iti'es.-.
s T'bisrCoipletfon Time also allo'ws f6r an-exce'pti on to thenormal" "time zero for
'beginning the;,allowed outage::time-clock.)"
In this Required
'Action',: the Completion..Time 'only begins 'on' discovery that both:'
a":, " The train has no offsite~power supplying its loads:
and "b.
A.;required featurs on ;the other train is inoperable.
U'I foat-any time during t-h..5'exis*tence -of Condition A (one offsite circuit inoperable) a redundant required feature subsequently becomes inoperable, this. Completion Time begins to be tracked.
Di scoveri ng' no coffsi te 06:wer -to.. one. train of the onsite Class '1E, Elec~trical Pow&ei'iDfstrlibut tonrSystem coincident' with one' or mor~e,i nop, e lble. orequ Ired;su'pport or supported featu'r6s, or both,,'Nhat.: re' associated wi.th the other train
.that has 'offsite,'powe) tes*,* tsLs in starting the Completion jTimes for the ReqLui,','ed':&,.tidn,
'Twenty-.four hours from the
,di sc6very'bf -these eveWs:',exl.sti ng concurrently is
~acceptable.becai §e, i"t n'irimizes risk whiTe allowing time for restoration. before.,sUbjet-ting,'the 'unit' -'to transients associ ated with s
- itd1wn.
I The remaining OPERABLE offsite circuit and DGs are adequate
.to supplyrelectricalpower to Train A'and Train B of the onsite Class. IE Distribution System.,
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time takes into account the component OPERABILITY of the redundant. counterpart-to the inoperable required feature.
-..(continued)
"PALO VERDEUNITS E-21.3B 8 3.8.1-8 REVISION 50
AC Sources - Operating B 3.8.1 BASES ACTIONS B.2 (continued) allowing time for. restorati~on before subjecting the unit to transients associiated with shutdown.
SIn ihi.s Conditidlnr, the -remaining :OPERABLE DG and offsite circuits are adequate to supply electri~cal power to the bnsite Class lE Distribution System.
Thus, on a component basis, single failur6 protecti~on for the-required feature's function may aVebeen *lost;.h6wever, function has not been
'lost.
The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time takes into account the
`OPERABILITY'of'the redundant couinterpart~to the inoperable required feature., Additionally, the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time takes into accourit the capaci'ty and~capability of the rema'fnng AC surces, a reasonable time for repairs, and the
,low probability of a DBA'occurring..during'this period.
if a DG has beenýdeclared inoperable and Condition B has been entered; and~during that inoperability a new problem with the inoperable DG is discovered, a separate entry into Condition B is not required for the new DG problem,
'Therefore, the Required Actions of Condition B would not apply to the new DG problem.
The new DG problem must be entered into the corrective action program and corrective actions specified in accordancewith the corrective action program.
Transportability must be addressed in a timely manner in dccordande wifth'the corrective action program.
B.3.1 and B.3.2 Required Action B.3.J1provides an allowance to avoid unnecessary te#stih *;f OPERABLE:IDGs:5.7 If'-it can be determined Ithat the -cause of"t66 inoperable DG do6s not exist on the
.,,OPERABLE.G.
SRD'3
`§'!?.2 does,' not have :totbe performed.
If t...he cause of ino~erbfity exists dnthe other DG, the other DG would-be deel:are "ilnhperable -upon discovery and Condition Eof LCO3.8.l wq,9"Ilb'bentered." Once-the failure is repaired, the"Qcomm aus:failu no longer exists and RequiredAction '.`.
ii#saisfi'ed' !If the cause of the initial i nope rable"-DGOcanhot bG-confirmed not to exist on the remaining DG, perfor.'ance of SR 3.8i 1.2ýsuffices to provide assurance of continued OPERABILITY of that DG.
In'the event the Inoperable DG is restored to OPERABLE status* prior tocompleting either..B.3.1,or B.3.2, the plant corrective action:program will continue to evaluate the common cause possibility.
This continued evaluation, however, is no longer under the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> constraint imposed while in Condition B.
(continued)
PALO VERDE UNITS 1,2,3 B 3.8.1-11 REVISION 43 CORRECTED PAGE. ON REV 50
"AC Sources - Operating B 3.8.1 BASES ACTIONS-".,
. B.31"and B.3.2. (ontinued);-'.
)
According to Generic Letfterý84-i5(Ref,' 7)" 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is reasonable'-to confirm'that'-the'OPERABLE DG.(s) is not affected-by the same"problemn asthe inoperable DG.
A' A
"in Condition B, th-e"remairi'in'g'OPERABLE* DG and offsite circuits are adequate'-to silsply electrical power to the onsite Class, 1E Distribdtiohi'.Systend" The>:10 day Completion Time takes into accoun'lihi"-capacity and"capability of the remaining AC sources,..'esonable'time for repairs, and the 16w probability o'f a' DBA occurring. durirg :this period.
When utilizing.an~extended..DG Completion.lime (a Completion Time greater, than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.,-andie,ssithan-or equal to 10 days), the compensatory',measures listed -below shall be implemented.;.Forplanned.maintenance..uti l.izing an extended Completion Time, the compensatory measures shall be implemented prior to entering Condition B. For an unplanned entry into.an extended Completion Time,. the compensatory measures shall be, implemented without.delay.
1... The redundant DG (-aliengwith all of its required
_,.systems&, subsystems;-,3.trr-a~ins,, components, and devices) will beverifPed..,QQPA,;
C -as-required by TS) and no
-di. scretionarymai'nten.r.nceactivities;w-ill be scheduled
,onr, the-,.redundant ('QPEABLE),
G "D.
2.:, -No disc etlonaryintpeoance, actlv;itles will be
, d'cedul'ed'on the' sl,aTr.' 1 l6 ac'kout generators (SBOGs).
- 3.
No discretionary maintenance activiti~es will be scheduled on the ste~rtup transformers.
- 4.
No discretionary maintenance-activitifes will be scheduled.inthe. APS switchyard or the unit's 13.8 kV
-" powen'rsupp-'ylini~rl transfbrmrs which could cause a
.l~ihe"'outage o, -haV'enge offsite pdwer availability to
'the unit uti biinz r 'I.-he' extended--DG CoImpletion Time.
S.. All -activity,,.inc.luding access,.in the Salt River Project-.(SRP) switchyard shall be c.losely monitored and controll'ed.,Di.scrpetionary maintenance within the
/ ',switchyard,.that could challenge offsite power supply availabi lity will be evaluated in accordance with 10 CFR 50,65(a)(4.) :and managed on a graded, approach according to risk significance.
- 6. The SBOGs will not be used for non-safety functions (i.e., power peaking to the grid).
(continued)
.PALO' VERDE UNITS 1,2,3
-.B 3". 8.IL-12 REVISION `48
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued The required steady.state frequency,range for the DG is, 60 +0.7/-0.3 Hz 'to be consistent*.with the--safety analysis to p.rovide.adequate. safety. injection f-low.., In accordance with the guidance provi.ded in RegulatbryGuide-1.9 (Ref. 3),
..,.where steady,stpate conditions'do not.exist (i.e.
transients),' the frequency range shoUld be restored to within +/- 2% of the 60 Hz nominal frequency (58.8 Hz to 61.2 Hz) and the voltage range should be restored to within +/- 10%
of the,41.60;,volts nominal:;voltage. (3740 volts to 4580,Volts).
T he: timed start is-s'atisfied when the DG
- a. h i
achieves, at, lea.,..3740 volts, and.58.8 Hz.
At these values, the DG output breaker, permissives are satisfied, and on detection, of buý'4indervoltage 6r loss ofpower, the DG breakers would-close., reenergizing its respective ESF bus.
i.
Steady state andtransient:,.voltage and frequency limits have riot been.,.adjusted for instrument accuracy Error values for specilficiinstruments are established by plant staff to derive -the,indicated values for the steady state and transient,voltage and frequency limits..--_
Specific MODE restraints.have been footnoted where applicable to,.each,:18 month SR. i The reason for "This Surveillance shall not be performed in MODE 1 or 2" is that during operation wivth the ýreactor: crtticaI, performance of this SR could.cauise p.erturbations to-.the EDS that could
,challenge ccnti.ýt.
adyrstate&operation and, as a result, unit. safetyt'systci~sor that performing the SR would remove a required D -I'frofrý,,rrvice,.,,The--*reason for "This Surveil~lance,.shaT].:.not be.performed. in MODE 1, 2, 3, or 4" i.ithat.,perf.o ip".ii.s iSR.wouId. remove a'required offsite circuitfrom'servl'e', pertu'rb the EDS, and challenge safety syV~tems'.
SR 3.8.1.1 This SR assures proper,_ crcui t contihuity for the offsite AC electrical power supp.Jy t ttheonsitedistribution network and i'ndicated.. avajilability~o,f offseý.rAC electrical power.
The breaker alignm entverifies that each breaker is in its co rrect.position~tb.,;eihsure that distribution buses and loads are connected totheIr preferred power source, and that appropriate: independence of offs-ite circuits is maintained.
The 7 day Frequency'i"s adequate since breaker position is not likely to change without the operator being aware of it
'and because its status is displayed in the control room.
(continued)
PALO VERDE UNITS 1,2,3 B 3.8ý, 1723
. REVJýSIONIý,,50
-l AC Sources
-. Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (conti nued)l SR 3.8.1.2 and SR 3.8.1..i '
These'SRs help to ensure~the ýavailabilitybf the standby' el ectOical power supply to-7mitigate DBAs' and transients and to mai ntaihn:* the unit.Jin. a.,s'afe'shutdown 'condition.
To minimize the wear on moving parts thatjdo not get iubricated when the.engine* i~s,;not runn-ing, these SRs are modified by a Note to ihdi*cate that all DGý starts for these Surveillances may be, preceded by an engine prelube period and followed -by a warmupriod.priorto' loading.
For the.ipurposesý'of' SR.3
- 8.'r.12 and SR !3.8'.1.7 testing, the bDGt are started fr~om standby co'ndition*.K,'.Standby conditions for-a DG mean that othe engine 1l ube, oi l'and cool ant temperatures are maintained-consistent with manufacturer redommend'ations!- Additli'on'ally1,.duri ng stdndby conditions the 'dies~l,enginbliIbe 1oiI iS c'i rclated-con'tinuously and the eiigine coolant iS'>ci~rculatedd-on and'off via thermostatic control.
v In order to reduce stress and-we'ar on diesel engines, the DG manufacturer recommends a-modified start in which the starting.peed of DGs fs-1`i,mited:, warmup is limited to this
.lbwer speed, and. the OG'.'regradually accelerated to
..*ynchrous soeedh pr h:"to loading.."This is the intent of Note 3, which is only appliýable when guch modified start
.procedures are recommended by the manufacturer.
SR 3.8. 12 I2Note 4 randS,
,8.1.7 Note 2 state that the steady*.state~'vol tage* a'Trequency-limits are analyzed v a.
] lues and' have-nbt beý!iY-'adjusted' for: instrument accuracy.
'The aalyzed' values.fdrthe'steady-statediesel generator voltage
.ami4re nd
- 4377.2 volts and the analyzed' values"for- ýthesteadyf'state diesel generator frequency limits are a 59.7 and.,- 60.7 hertz.
The indicated steady state diesel generator voltage and frequency limits, us'i ng-Ithe panel mnourft!tddifesel gnerator i nstrumentati on and adjust'ed for irtrumer f~error' are > 4080. and -< 4300 volts (Ref 12), and> 5*-;
and 6G 5 hertz (Ref.
13),
respectively'.
If"ldigital]- Maintenance-and: Testing Equipment (M&TE) is-'used instead of the panel, mounted diesel generator instrumentation, the :instrument error may be reduced, increasing the range :for the indicated"steady state voltage and frequency limits.
(continued)
PALO VERDE UNITS 1,2,3 REVISIONI-_50
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 and SR 3.8.1.7 (,continued)
REQUIREMENTS SR 3.8.1.7 requi.res that, at a 184 day Frequency, the..DG starts from s'tah'dby conditions with..the"engine atnormal keep-warm condit ions and achieves..required voltage and frequency withi'n" 10 seconds, and subsequently achieves steady state required voltage and frequency ranges.
The 10 second-startl requirement supports the assumptions of the design basis LOC a
nalysi's inthe FSAR,. Chapter 15 (Ref. 5).
A minimum. voltageiand f requency is 'specified rather than an upper and a lower limit'because a diesel engine acceleration at fullfuel %,.(such.'as during a fast.start) is likely to
,,,-. "overshoot" the upper limit i'niti lly and then go through
.;:,,;,..several oscillations prior vto l
votage.and frequency within the stated upper.and lower -bounds.-The time to reach
,.steady state," -'could ;exceed -10 *seconds and be cause to fail
,,,the SR.,, Howeverr on :*an.actual emergency start, the EDG would; reach minimum vol'tage and frequency in _<10 seconds at which time it would be loaded.
Application of the load will dampen the oscillations.
Thereforej'only specifying the minimum voltage and.frequency. (at which the EDG can accept load) demonstrates the necessarycapability of the EDG to satisfy.safety:ireqi-'i-rements without including a potential for failing,the-.Supgeillance.
Error, values for specific instruments, are established'.to derive indicated values in test proceduresc, ;,.D,
'I.;
While readhing minimum voltage'"and freq'Uehcy (at which the DG can accept loadý,,.-.'n. < 10,seconds is&an.immediate test of OPERABILITY, '.-the ajity of the-governor, and voltage regulator to-dach ieve 'qteady-state'. Oper'ation, and the time to do. so are impor.ant -indi.cators of conti hued OPERABILITY.
- Therefore, the.. t ime achieve steady state voltage and
,frequency will,-ý bem*prn*m,,itored.`asa. afuncti on of continued OPERABILITY.
-The 10 second start'equi:rement i:s-no&aop)licable to SR..3.8.1.;2 (see Note*,I hw hený a modif'ed.:start procedure as described aboves iuse.
'.-.,a modified start is not 'used, 10-:secondistart requirement '6f SR 3.8,..I*.7 applies.
The existing design, for a.CSAS actuation signal does not provide an emergency, mde stlart.to the DG.
A CSAS actuation signal cannot occur unt-il.a-fter..d SIAS actuation signal has already been generated.
(continued)
PALO VERDE UNITS 1,2,3
.B 3.8.1725 REVISION 50
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2,and SR 3.8.1.7,(continbed)Y.
REQU I ýM.LNIS -
1 1,
- I I Since SR 3.8.1.7 requires a.0,second'start, it is more r I strictive thanSR.3.8'.1"2, and it may'be-performed in lieu of SR3.8.1..2. This'i's thei Ttent of Note I of SR 3.8.1.2.
'.The normal 31 day Frequency for'SR 3.'8'.1.2'is consistent
`with Regulatory Guide 1.9 (-Ref. 3,)."The'184 day Frequency for.SR 3.8.1.7 i.s a reduction 'n cold testing consistent with.Generic Letter 84-15%'(Ref'.: 7).
These Frequencies provide adequate assuranc o#f 0DG OPERABILITY, while minimizing degradation-rýiztfng from testing.
..,,SR 3.8.1.3 This Surveillince.ver~ifiesthat the DGs, are capable of synchronizing-withithe offslite electrical.system and accepting loads of 90 to.,10.,percent (4950 - 5500 kW) of the continuous rating of the DG.
Consistent with the guidance provided in the Regulatory Guide 1.9. (Ref.
- 3) load-run test
..description, the 4950 - 5500 kW band will demonstrate 90 to
',100 percent of the continuous rating of the DG. The load band (4950 - 5500 kW).is',meant as guidance to avoid routine
-.' overloading of the ;engin.,.,: Loads' in excess of this band for special. testing may, be',,ip!formed withj.n the guidance of the generator capabil.ity.cue,,
Y
. " minimum run time-of.*_rn n iutes riSrequired to stabilize
- engine,)temperatures, wnl*'i,(2minimi.,ing the' time that the DG J s connected.,to thef s
otesburce.,.
The.normal 31.dayFxreuncy fo6 this Surveillance is consistent with Regu at.bry Gu ide 1.9 ý(Rf. 3).
.hi s SR ts modi.fied-b.,,four'Notes.
Note I indicates that diesel engineruns" for thisSur'veillance. may include gradual
- Sloading, as. recommende6d by thee manufacturer, so that mechanical s/tress, aiid4 e.ar n,the. diese]:engine are "minimi-zed.- Not*e,2.:'staý'Tes' that m6mentar*,
transients because
, cf,,changing bus ioa3ds'do*.not invalidate this test.
Note 3 indicates that this Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
Note 4 stipulates a prerequisite requirement for performance of this SR.
A successful DG start must precede this test to credit satisfactory performance.
(continued)
PALO VERDE UNITS 1,2,3
,B 3,.8.1-26 REVISIM50
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.-1. 4 REQUIREMENTS (continued)
This SR verifiesthat thereis enough usable fuel-oil-in fhe DTG Day Tank 'to r~un, the dieS*b1 generator':at-full load for a minimum of, 1 ho.urpl'us-10%.
The, surveillance is on fuel level since there is no direct indicator of volume.
Level is read in feet on the Main. Control Board indicators or in equivalent. units, oný local DG instrumentt~ion.
The source for..the rbk'-time requirement is the UFSAR Sec. 1.8 and Questfon 9A.9x.commitment to ANSI N195'1976.
That standard refers to the le~vel at whichfuel'is automatically added to the tank, For'the DG Day Tanks the "pump start" level is above the SR and so is additionally conservative, The 31 day Frequency is adequate to assure that a sufficient s
supply of fuel 'oil.is,,available, since low level alarms are provided arid...Unitý oper'ators would be aware of any large uses of fueloil~during this period.
Microbiologicael'-fouling is a major cause-of fuel oil degradation. j-There are numerous.bacteria. that can grow in
...fuel oil *and 6a'usefouling, but all must have a water "environmentin` o`ýer`t6 s urvive.
Removal. of water from the fuel oil day tarnk',bnce -every 92 days:,el.iminates the nece~ssary environipený for bacterial survival.
This is the most, effecti v:'*e','is of - drtrol~l ing microbiological fouling.
In-additidh
-et pelin*'na nes.tthez.potential'.,for water entrainment* i'nýýihe ftel, oil duning: DG,.operation.
Water may come from any of several sources, including condensation, ground water: rai'M',.water,contaminated fuel oil, and from breakdown of the`**1eu, oil bylbacteria.: Frequent checking for and removal of accumulated water minimizes fouling and provides dat:a r gard ig thewatertight integrity of the fuel oil system. The Sdieiil,ance Frequencies are established by Regulatory G'uide'- 1':137"(Ref., 9)., This SR is for preventive maintenance."' The tresence-of water:,d6es not necessarily represent fai'lure &ifhis'ýR prdvided the accumulated water
'is removed duringthe porformance of this Surveillance.
(continued)
-PALO:VERDE UNITS 1,2,3 B, 3.,. 8.. 1:- 2 7 REVISIONý50
.5 AC Sources.- Operating B 3.8.1 BASES
(..
SURVEILLANCE REQUIREMENTS SR 3.8.1.6 This Surveillance demonstrates that each required fuel oil tiransfer pump-operates and :transfers fuelV'oil from its associated'stbrage tank to';,its associated:day tank.
This is required.-*to support continuoussoperation.of standby power so urces. -This Survei.llan~eprovides*.aSsurance that the fuel oil transfe'r pump is'.OPERABLE:, the. fuel oil, piping system is intact., the fuel del-iverypi'ping is not obstructed, and the controls adcontrol,"systemstfor;automatic fuel transfer s
ystems are OPERABLE-.
,. Since;,the design of the fue],transfer,system is such that pumps: will, operate,,automatically in order.'to maintain an adequate. volume-of fuel oil in the day tank during or following DG testing,, a 31-.day:Frequencyujis appropriate, SR 3.8.1.7
- 1 T.
See SR.3.8.1.2.,
-SR 3.8.1.:8
-Transfer:of each 4.16,l,.P.7S&Fbus power. supply from the normal cffsite circuit totheaenrate offsite-circuit demonstrates
-he.LOPERABILITY o.of the,alternatecircuit.distribution network to power thei.:auto-con.,£e*.;*.d..-emer, gency loads.
The 18 month I 'Freqiency' of the Surveilance, is based on engineering
'judgment*L-,taking into',3nsideration the6unit conditions required to perform theoSurvei llance,; -and, is intended to be
.consi~stent with expectei-.,
cIfuelcyc!e lengths.
Operating experience has shown thlt,.thes*:comoonents usually pass the SR when performed at the 18 month Frequency.
Therefore, the Frequency was conclude*d' to be acceptable from a reliability' standpoint'.
This SR is modified by a Note. The reason for the Note is
. that duri*ng oper[ati on.ow,,ith the :reactor critical, performance of this SR.could :ca.s,.: perturbations to the electrical distribution systems that could challenge continued steady state operationýand.,-as a result, unit safety systems.
This restr.iction from.normal*y performing the surveillance in MODE 1 or.2 is.,,furtherzamplified, to allow the surveillance to be performedfor the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated (continued)
PALO VERDE UNITS 1,2,3 B 3.8.1-28 REVISION 41
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS SR 3.8.1.8 (continued)
-OPERABILITY concerns).provided an assessment determines
.. plant.safety i s-maintained.'or 6nhanced.. This assessment shall, as a mini~mum, c6nsiderfthe poten'tial outcomes and ltransients associated with.a faliled~surv~illance, a successful,:surveillance,, and a perturbation of the offsite or onsite system-when they are tied together or operated independently,'for. the, surveillance; as well as the operator procedures available to'cope With'.these. outcomes.
These shall be measured against the avoided risk of a plant shutdown and~sta.6rtup-.,tO determine that.plant safety is
-maintained or :enhanced.when the. surveillance is performed in MODE 1 or&2. *Risk insights.or deterministic methods may be used. for. this'.assessment.
C I
SR 3.8.1.9 Each DG is provided with an engine o'erspeed trip to prevent damage to the engine.
Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might uesUlt in a trip of the engine.....T *hsSurveillance demonstrates the DG load response charactd'istics-and capabil.ity,to reject the largest~single'l*a#1,,:..or equivalent o~d.-without exceeding predetermined~'vbage' and frequency-and while maintaining a S. :specified margin.o: 6"the.,overspeed.,trJp:..- Train A Normal
(-Water Chiller (.a,24kW) and: Train:-B 1AFW: pump (at 936 kW) are :the.bouhding.; Ao'ads ;for the DG A-and,.DG B to reject, respectively. Thbs'e v-..alues were established in reference 14.
T heis. Survei.lllace may be.ý accompl i shed by:
- a.
Trippoing ;.the' DG. output. breaker with the DG carrying greater than or equal to its lassoci:ated single largest post-accident load while solely supplying the bus; or
' Tripping its lassociated, single largest post-accident
'load with 'the; DG.solely. supplying the bus.
As required-by IEEE-*8308 (Ref. 11),.the load rejection test S *-is acceptable. if.the, increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed'trip setpoint,, or..15% above synchronous speed, whichever is lower:.
(continued)
PALO VERDE UNITS 1.2.3 B 3.8.1-29 REVISION 41 (CORRECTED PAGE)
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE n*n ITDCMEMTC SR 3.8.1.9 (continued)
I!,EL*U i.RLI'ILINI I.)
The time, voltage and frelquency tolerances specified in this SR are:derived from Regullatory'Guide`'1.9 (Ref. 3) recommendations for response during load sequence intervals.
The 3 sekonds specified is.1equal to!60%-of:,a typical 51 second: load sequence interva'l associated with sequencing of the -1rgest load.
Thev.o1]tage' and frequency specified are,.consistent with the design range of, the equipment
'powered by the DG.'
SR:,3*,871.9.:a corresponds to the maximum frequenty excursion, wh~f-le'ý.SR 3.8.1,-9.b and SR 3.8.1.9.c are the voltage and frequencyý':,*alues :the 'system must meet.
within 'three seconds,' folVoWing 1load 'rejection.
Error values for specific instruments are established to derive indicated values in test procedures.
The 18 month Frequency i sconsistent' with ýthe 'recommendati on Of Regulatory Guide 1.9 (Ref. 3).
This SR is-modified by 'a :Note'
ý-Th'e;reason for the Note is that performing this-SR would remove a'.required offsite circuit fro-mn service, perturb' 'the EDS',!and chal 1enge safety systems.
This SR is performed in emergency mode (not paralleled to the grid) ensuring that the DG is tested under load conditions that are as close,-to design basis conditions as possible.
This restriction from normally performing the
- survet~l I ance in Mode.i,.;.2*,3:- or 4 i s further ampl i fied to S:al 1 ow the surVei 11 ance*,J.o'6'eiiperformed, for the purpose of reeostablishing OPERAB.*.....,-*...(e.g.,. post. work testing foll~owing, corrective m*.Antenance, corrective modification, deficientor, i ncomp-letei*lwrvei 1ance:.testing, and other unanticipated OPERABI1ITconcerns) provided an assessment.
determines that plant,safety 1s maintadined or enhanced.
.-This assessment shall T.,,S Ias-a minimum, consider the potential outcomes and transient*-..assotiated with :a failed surveillance, a. succesfUl, surve~illance,'and a perturbation c-.
f-:the offs.ite or on.6
.,§eIstem When they are tied together
.. or operated indeperideiyi.,fd:r:.the surveil'lance: as well as the: operator..procedures.vailable to-ope with these
.-outcomes.*' These.sh&!.!,j be. measL-red a'gaihst the avoided risk of a plant shutdown. ar,'dýtar~tup to.:determine that plant safety is maintaine*d r enhanced when'the surveillance is performed in MODE 1, 2, 3, or 4.
Risk insights or deterministic methods may be used for this assessment.
(continued)
PALO:VERDE UNITS 1,2,3
. B 3.8..'1-30 REVISION.:50
AC Sources - Operating B 3.8.1 B
ASE S SURVEILLANCE REQUIREMENTS SR 3.8.1.9 (continued)
The followi ng.compensatory-measures, shall..be impleienrited prior to the. performanceof.this SR in MODE 1 or 2:
a'. Weather co6nd-itions will. be assessed, and the SR will not
.. be scheduled when severe weather conditions and/or unstable..g,rl d, conditions' ar~e predicted "or present.
b:. No discretionary, maintenance activities will be scheduled in the APS swiitchyard or*the unit's 13'8 kV power supply lines and, transformers Which could cause a line outage or chal~lenge-offsite power avail~abi;lity.to the unit performing this SR.
- c. All,actitvi'ty.- -including access, in the 'Salt River Project (SRP) switchyard' shall be closely monitored and controlled.
Discretionary maintenance within the
- swi~tchyard that-could challenge offsite power supply
.. availability will.be evaluated in accordance with
.10 C FR 50.65(a)(4) and managed on a graded approach according to ris(:significance.
SR "3.8. 1.'10 This Surveillanqe-demonstrates'the DG capability to reject a full 16"ad W'ithb'&.Vt.er'6peed tripping 'or exceeding the predete:rmin'ed voýt g6 lim-its:. :The, -G full load rejection
..may'occur beca'asu'(f!ia, sYstiem faultcor, inadvertent breaker tripping'., Tihiýi'ýSiej ei-l.ance e'-n'suires'proper engine generator
-od~resporis'e"inde, thesimUla-ted'test conditions.
This testIsimulatesýAt'e-loss` of the total: connected' load that the DG 'experfiences 'fol'lowing,'a ful,l.load rejection and verifies that' the "DG WiI-. nottrip upon loss of. the load.
These acceptance cri teriapho.ide DG damage protection.
While the DG is not expecfte"6'ý'6experience this. transient during an eventý and.'crintin"est*o*
b&'*av'ail-abl6>'e-this response ensures
.,that the DG' i'*
qnot
,qe'gr'adle'd" 4
.fo'r future-'appl i cati on, Si ncludi ng reconnec
' tohe busif the trip initiator can be -" orrect d OrIs
- ated.*""
.d",
(continued)
'PALO',VERDE UNITS 1,2,3
.B 3. 8_11 -31
- I REVISION,,50
,-, j AC Sources Operating
'B '3.8.1 BASES SURVEILLANCE REQUIREMENTS SR" 3.811. 10 (Continued)
I I brde'rtb ensure that.the ;DG i s tested' under load conditionrs'that are:as' close, t6'design. basis conditions as possible.,.testing Jis 'perfo'rmed `Using'design basis kW loading and maximum'kVAR l1ading'"permitted during testing.
These loads 'represent the i'hducti-ve loading that. the DG would experjernce to the extent. praeti'cable and is consist'ent with the guidance Of Regulatory:'Guide i:;9, (Ref.`, 3).
Consistent
,;wi th the& dui dance'; provi ided'" ih fhe'Regufla'toby' Gui de 'l :9 'fullI -
'loadrejection test. description',: the; 4950.,- 5500ý kW band' will demonstrate :the'DG'.*'Tcapability, t6 reject a load equal to 90 to' 100 percent of it'* continuous rating.
'Error values for specific *instruments are established to derive indicated values',in test procedures.
-Administrative limits have been,
-, placed..upon the Class.1E.4160 V,.buses due..to high' voltage
'concerns.
Asa result p
-wer'factor sdev1iat'ing much' from
.unity"are currently not ssib.le when' theiDG runs, parallel to the grid 'while 'the plant i's shutdown..To the extent practicable; VARs wi'llbe:provided'by theDG during this SR.
The 18 month:Frequency,isconsistent with the recommendation of Regulatory Guide 1.8,(Ref. 3).and is intended to be
,conisistent"with" expected 'tuel"cycle lengths.
.ThiSSR is, modi f ied.by.N,a.te, This Note ensures that the DG.is"'tested under loadeoriditti~ons that:.ware as close to d d e*,igrbasis conditions, aSpossible.. When synchronized with
.offsite-power, testing:o,-ould.be performed at a lagging
.. power'factor.-of <*,0.8f9*.Thispower factor -is representative of the-:aetual :inductiel-lloading a.'DG,.would see under design
,bas'is accidentcondittb'Is '<This power.,factor should be able
'-to be achieved when perform~ingl this SR at.-power and synchronized with'offs-'ittepower-byp-trarisferring house loads
'¢
' fr'om the 'auxiliary.t'r'-iformer to the-startup transformer in order:to lower,the C~ia8'-;1E'lbus voltage-..:Under certain
.':;hoditions. however': No'te..2, allows the surveillance to be conducted at a power factor other than < 0,89.
These conditions occur when grid voltage is.high, and the additional field excitation needed' to'.get'.the power factor to _< 0.89 results in. voltages on the. emergency busses that e too high. "This.~-uld occur "when performing this SR
..,while shutdown'and the i'bads on the startup transformer are
,'too 1light to loWer"'the. voltage suffi'ciently to achieve a
0..89 power factor,. ',Under these conditions, the power factor (continued)
PALO VERDE UNITS 1,2,3 B31.18.-"32 REVISION :50
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE n
I ITDCMfLTC SR 3.8.1.11 (continued)
LqU I rI
..,surveillance in MODE 1, 2.. 3., and 4 is.further amplified to al.low portions of the surveil-lance, tp, beperformed for the p...
urpose ' of ree s:ablshing OPERABILITY (e.g., post work testing following corrective maintenance,. corrective
,..,modification., deficient or incomplete surveillance testing, and other.unantlcipated OPERABILITY,.concerns) provided an assessment determines plant safety is maintained or enhanced.- This. assessmentshall..as a minimum, consider the potential outcomes.:and transients: '
associated with the failed
- ;N* partial, surveiinea,
- a.
successful partifal, surveillance, and a perturbation of.the oQffsi~te, or onsite system when they are
-tied together or-operated.independentlyfor the partial
- surveil.lance: as well asthe operator procedures available
- o. cope.,wi-th these outcomes;. These shall be measured against-the.--avoided ri'sskof a plant shutdown and startu to determi.ne that:plant safety is maintained or enhanced when portions of the surveillance are performed in MODE 1. 2, 3.
or.4.,Risk insights or deterministic methods may be used for this assessment.
Note 3 states that momentary voltage and frequency transients induced by load changes do not invalidate this-,test. Note 4 states that the steady state voltage and frequency limits-are analyzed.- values and have not been adjusted for instrument accuracy.
The analyzed
"*.values for thes.teady.state~diesel generator voltage limits are a 4000..and:ii-437V. :2 volts and.the-,analyzed val ues for the steady'state.liesel.:generatorj.requency limits are 59.,7-and
,::60:73:ihptz.- -The indicated steady state diesel generator voltage andfrequency. ]l.imits,,.,,using the panel mounted dieselgpfleator instrumentationand adjusted for instrument er-ror.:: a
- , 40801and, 4300. volts (Ref. 12), and
?:-. 59.9 and s60-.5Jhertz. (Ref.. 13)..respectively.
If digital
-Maintenance and T.;-t;ng Equipment (M&TE)Xis used instead of
-thelpanel mounted!.d;iesel,.generator instrumentation, the instrument e~rror~may;:,be reduced,,increasing the range for the indicated-steadyi.state voltage-and. frequency limits.
SR _3. 1..12 This S~urvei'llance demonstrates that the.'DG automatically starts and achieves ýhe required voltage., and frequency within.,the specified time (10 seconds) from the design basis accident (LOCA). silgnal, and subsequently achieves steady state required voltage and frequency ranges, and operates for a 5 minutes.
The 5 minute period provides sufficient time to demonstrate stability.
Error values for specific instruments for non-steady state (transients) are established to derive indicated values in test procedures.
(continued)
PALO VERDE UNITS 1,2,3
.B 3...,3 REVJSION 50
AC Sources - Operating SB 3.8.1 BASES SURVEILLANCE SR 3.8.1.12 (continued)
REQUIREMENTS RETU i
'Ikhe exis'ting deslgni for"'CSAS' actuation -signal does not
- "p
,rovilde an.emerged modeý*sta~rt:"to thbe-DG,.,
A CSAS actuation signal 'cannot occur' Ubnti'l" K4fter a SIAS adtuation signal has already'been "generat'ed. r'SR 3..8.1.12.d and SR 3.8.1.12.e ensure that'permanently connected loads and auto-connected emergency loads (auto-connecte'd through 'the automatic load sequencer) are energi~ze dfTP-m the offsite electrical power system on:an ESF signa~l Without-loss of offsite power.
The req'ui rement' to verify' t'he connection of permanent and auto-connedted emergency ioads is intended to satisfactorily show the relationship-of these-iloads to the offsite circuit loading logic.
In certain circumstances, many of these
"] :"**. "
loads cannot, actuallj be c6nnected or loaded without undue hardship or potential-: f6rC"nhdes'ired.-ope'ration.
For instance', ECCS.injedt'ton val'ves' arenot desired to be stroked open, high'pressure'qinjectiOn systems are not capable of-being operated at full, *flowl or SDC systems performing 'a, decay heat removal' function are not desired to be'>real'igned to -the: ECCSn mode-.of'operation.
In lieu of actual'..demonstration of connection and loading of loads, testirig' that adequately s',ows: the capabil-ity of the offsite circUit systemto.'perfo;-rV-tli[2se functions is acceptable.
rhis~tetlng may 2rnclud&'nyseriesl*of.sequential overlapping, or to t-,ae"PS tot o that;.the entire connection and-, dading sequencei-s v&rified 'to'. the. -extent possible
.ensi*
rinpc6wer is',:val61eto, the component.
The 2:.Frequericy of 18,.-6o'iths takes-1nto consideration unit cPndi ti onsrequi red t6 Pje;formý:the"Survei'l lance and is intended to be consistgft With' the expected fuel cycle 6ngths Operating--e4,rence~has shown 'that these components usual ly 'p'.sý-the SR when "performed at the
- 18. nonth Frequency-
,heore,, the Frequency was conclubded to be.acceiprabdi from: a reliability standpoint.
.(continued)
PALO,-VERDE UNITS 1,2,3 B.3 '.'8..l -3 6 A.VD UI 12"3. 6REVISION 50
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.15
,-!Thi.s.Surveillance demonstrates that the.diesel engin'e can restart from a hot condition,.ýsuch astsubsequent to shutdown
.. om normal Surveillances,,.and achieve the required voltage
.. and frequencywi1thin 10,setonds,-.and subsequently achieves
-s teady state required Voltage and frequency ranges.
Error values for specific instruments for non-steady state
,(transientsi) are-established, to derive indicated values in test procedures,.,'.The,1O second time is:derived from the requirements of the accident analysis to respond to a design basis l.arge.break.,LOCA.
.The 18. month Frequency is consistent~with. the recommendations of Regulatory Guide 1.9 (Ref. 3), paragraph 2.2.1.*,0.
'.- L*
-This SR is modified-by.three Notes.
Note; 1 ensures that the tLes~tis performedwwt.
the diesel sufficiently hot.
The Iload band-is pr~ovi'dedjto avo-id routihe overloading of the DG:.
Routine overloads may;.resultin more frequent teardown inspections inj accordance with vendor recommendations in order-to maintain DG:OPERABILI-TY.
Per, the guidance in Regulatory Guide l.9,ýthis SR would demonstrate the hot restart functional capability at fijl.l-load temperature conditions, after. the DG has operatedI.for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (or until S..
- .....i, operating. temperares* have..stabhilized)7-at full load.
Momentary transients due to changing bus loads do. not invalidate the;tesI*,* -Note,2, allows, all DG starts to be precededIby ;an ne~prellbC period tonminimize wear and tear on the. dies4dupidg testing,,6Not6.3' states that the ste~ady state volt*ge andr.feruencdl]"i mi ts are analyzed values and have not been adjusted'fOb ihstrument accuracy.
The; analyzed value*-jor the steady-state diesel generator voltage.imits;-are z.4000:and
.i4377.
2 "volts and the ana,lyzed values fq-:tbe.steady-state diesel generator frequency:1li.mits,are 759 7 and < 60.7 hertz. The indicated steady state'dieselý,.6herator voltage and frequency limits, using the panel,mounted..diesel.generator instrumentation and adjusted for-instrupmený,er~ror, are.,
4080and < 4300 volts (Ref.
12),
and ý 59.9 and 5 60.5'hertz (Ref.
13),
respectively.
If digital Maintenance and Testing Equipment (M&TE) is used instead of the panel mounted diesel generator instrumentation, the instrument error may be reduced, increasing the range for the indicated steady state voltage and frequency limits.
(continued)
- PALO VERDE UNITS 1,2,3
- .B -3."ý81J ýý41 REVISI*ON-5O
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REOUIREMENTS (con ti nued)
SR 3.8.1.16
-As reqfiiredtbyý'RegulatoryGUide 1.9,.(Ref., 3),
- paragraph!2.2'11, thisi-Sur-veillance ensures that the manual synchronization and loadtransfer from the DG'to the offsite source-cani be made and that'the DG can be'fetUrned to ready-to-load status whenoffsite~power*1is're'stored.
It also ensures that the auto-startlogic.isreset to allow the tG to reload if:a;subsequent*I'oss-of'offsite poweroccurs.
The DG isconsidered'to be in ready-to-load status when the DG is at rated speed.and voltage, in standby operation (running unloaded), the output breaker isopen.and~can receive an autoclose signal on'bus undervoltage, and 'the load sequence
-timers are-reset.;
The. Frequency. bf'18.months i~s..consistent: with the recommendations of.Reýuiatory'.iGuide 1919'(Ref. 3)- and t.akes into consideration unit conditions required 'to perform the Surveil ance.
l This SR is mod-ified-byla Note:-.
The.t.reason for the Note is that' performing 'the Surveitl,lance wobld remove a required offs-ite' circui-t from service,:tperturb the electrical distr'ibution system', and chailenge'safety systems.
This restriction from normaly-.i7'.performi ng the.surveillance in MODE 1, 2, 3,"and 4 i.sK U',5ther: amplified to allow the
. 'ur'i T1 anceto be:%per.,-'rm'hed for the purpose of reestabl i shing OPERABIU9.!_,'.(e.:g., postý work testing
'followi ng correctirve"ný,*Atenance, corrective modification, defi cient-or J4ncemple *lSFUrvei l.lance* testing, and other unanticipated OPERABILITY concerns) provi'ded an assessment determines plant safety is maintained or enhanced.
This assessment shal l-, as,'-
.iiiimum, consider the potential outcbmes'and transi erts3ssociated-with a faied surveillance, asucces.fui surveillance, and a perturbation of ithe& offsi te or
'ni tre.system.When. they are tied together
'oroperated indepee;dtl.y,for the-surveillance: as well as
.,,theOperator proce1ur
'ailable2-to cope with these outcomes., :,,1'heseP-shdl.'be measured against the avoided risk of (pl ant' shutdowriLiind stdartup,to determine that plant safety is maintained or.enhanced whenthe surveillance is
'performed in MODE iLor;2., Risk insights or deterministic methods may be used f6r:..this assessment.
(continued)
PALO VERDE UNITS 1,2,3
.B 3..8..1-42 REVISION-45
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.18 (continued)
REQUIREMENTS (continued)
.,or-operated independently.for the surveillance:.as well'.as 0the operator procedures avaIlable-to cope with these
....outcomes.
These..shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is mai.ntained or enhanced when the surveillance is performed in.MODE 1 or,2<Risk insights or deterministic methods may be;used-:for.this~assessment.,
SR 3.8.1.19
~
In the event of a DBA coinc-ident with a loss of offsite power, the DGs are requiredto supply the necessary power to I" -
-ESF-systems.,so-that*theefuel, RCS, and containment design are,,not xceeded.
This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.11, during a loss of offsite power actuation test.signal in conjunction with an ESF actuation si.gnal.
In lieu of actual demonstration of connection.and:ýloading~of loads, testing that adequately shows the capabiji.ty.of the DG.system to. perform these functions i s -accepttabl e.
- Thi s -testi ng may i ncl ude any series of,- sequehniial,
- overl apping, or -total steps so that the enti re conn&ecil:,wand. loading sequence i's verified. The conditions:.requitedlto:-perform the._urveillance and is intended;.to :be.onsistent,,with,. ar-expected fuel cycle length
- of :18, monthsý:.;!~ W Thi S.SR is modified by-three.Notes.
The.,reason for Note I is tormi.nimize.,,.wearŽ:and ;tea.r on the,,.DGs during testing.
For thepurpose -of thJ.r-testing. th.e,.DGs must be started from standby conditi ons-- that, is, -with the.engine coolant and oil
. continuously circu}Itd..and temperature.maintained
.j.cons~istentwi~th,,arfacnturer recommendatjons for DGs.
The
.," *:Teason f orý.Note 2:. i s -hatjper~f6rmi.n.
the SreilI ance would
.remove a irequi-red ofsitelcircult.from service, perturb the el.ectrical.,di,stnibution.system, and.challenge safety systems.
This restriction.from normally, performing the surveillance in"MODE 1. :2, 3, and 4J.s.further amplified to allow portions of the surveillance to be.performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or
.(continued)
PALO.VERDE UNITS 1,2,3 SB 3-.8....1-1,45 REVISION, 45
AC Sources -
Operating B 3.8.1 BASES SURVEILLANCE.
SR.3.8.1.19 (continued)
REQUIREMENTS enhanced.
This ass'essment shall, 'as 'minimum, consider the potential outcomes and transients associated with a failed partial surveillance, a successful partial surveillance and Sa :perturbation of the of.fsite or.onsite.system when they are.
tied::together or oper ated -independently-for the partial surveillance;-.as,,well -as the operator: procedures available
..to cope with these.outcomes;.,. These shal'l, be measured against~the avoided-rJ:sk of::a plant shutdown and startup to determine that plant safety, is maintained, or enhanced when
, portions of~the, surveill-an&e are performed in MODE 1, 2, 3,
'or 4.
- Risk insights, or deterministli methods may be used for this assessment.
Note 3 states that,the steady state voltage and frequency limits are analyzedvalues and have not been adjusted for instrument accuracy.
The analyze values for, the.steady-state diesel generator voltage limits are a 4000 and -5 4377.2 volts and the analyzed values for the-steady-state,.diesel generator frequency limits are -
59.7 and !5 60.7 hertz... The indicated steady state diesel generator voltage andsfrequency limits, using the panel mounted diesel generator~instrumentation and adjusted for p!",.instrument error, are-,>-.4080 and:' 4300 volts (Ref.12),
and 59-.9 and 5 60.5.hert,.-.Ref,.13), respectively.
If digital Maintenance and Testin.g,E~quipment (M&TE) is used instead of
-the panel mounted diese,q.generator. instrumentation, the instrument error, mayi b6 ouced..increasing the range for
.'theindicated -steadys.t.-,te.vpl tage and frequency limits.
ýS 1:8 7Ji -;
- .20 This.SurveilIlance'dennhsira tes that the DG starting "indeoendence, has not-,,been ;compromised::;..-Also, this
.Surveil l ance demonstfr, that each engine can achieve propep speed wi;th-in eh',-speoif ied time when the DGs are started sfmultaneou'8'y
ýError values for, specific instruments for non-steady state (transients) are established to derive indicated values in test procedures.
The 10 year Frequency is consistent with 'the recommendations
...,of:Regulatory.Guide-l.:91(Ref.
3), paragraph 2.3.2.4 and Regulatory'Guide,1.137Z,(Ref.,9).
- This SR. is modified'b'..two Notes, The'reason for Note 1 is
,to-minimize wear onthe DGduring testing.
Note 2 states that the steady state voltage and frequency limits are analyzed values and have not been adjusted for instrument accuracy.
The analyzed values for the steady-state diesel generator voltage limits are - 4000 and (continued)
PALO VERDE UNITS 1.2,3 B 3.8.1-46 REVISION 50
Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES APPLICABILITY air are required to be'within limits'.whenithe associated DG (continued) is required tobe OPERABLE...
ACTIC)NS
ý"The ACTIONS Table is modified by a Note. indicating that separate Condition entry :is allowed for each DG.
This is
- ac'ceptable,-since the Required Actions for each Condition provide apphopri:ate.compensatory actions'for each inoperable DG subsystem. Complyingwith the Required Actions for one inoperable DG: subsystem may allow for'continued operation,
,*,and ubsequent"nperable DG subsystem are governed by separate Conditioh entry and application of associated Required Actions.
A.1I IIInthisCorrdition (i.e',"< 80% indicated fuel level), the 7 day-fuel oil supplyfor a DG is not available.
- However, the Condition it festrictedto fuel oil level reductions that maintainat least a 6 day supply., These circumstances
.,may be caused.by'events such as full 'oadoperation required aft'ran inadve'tI rnt.
start while at minimum required level; or feed and!'b~le:e'd*9qperations,,.whichmay'be necessitated by increasing p'afrti-u."te -levels or anypnumber of other oil
,quality degrada-
- This restriction" allows sufficient
- "time for obtai-ni:the.requisite replacement volume and performing the analyses required prior to addition of fuel oil to the tank.
A period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Tis considered sufficient to complete restorati-on of the required level prior tO,ýdeclaringth..e DG. -noperable'.
This period is acceptablebase.d.Qnte remaining capac~ity (> 6 days or 7
11% indicated fj._je-'le-el)-. thejfact that procedures will be i:nitiated 'to obia.n;-neplenishment, and the low probability of, an e',ervtduringthis. rbief period.
B.1 With'lube oil inven'tdry < 2.5 inches visible in the sightglass,' suffiCibht iubricating.oil:-to support 7 days of continuous DG operation at full load conditions may not be available.
'However,"-the Condition is'restricted to lube oil volume'reductions that maintain at least a 6 day supply.
(continued)
PALO VERDE UNITS 1,2,3 B 3.8.3-3 REVISION 50
Diesel Fuel Oil, Lube Oil and Starting Air B 3.8.3 BASES ACTIONS B.1 (continued)
This restriction allows sufficient time to obtain the requisite replacement volume.
A period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is considered sufficient to complete restoration of the required volume prior to declaring the DG inoperable.
This period is acceptable based on the remaining capacity ( > 6 days), the low rate of usage, the fact that procedures will beinitiated to obtain replenishment, and the low probability of an event during this.brief period.
The normal level of lube oil is maintained'at mid-scale visible on the sightglass which ensures sufficient lube oil todsupport at least* 13.5 days of engine operation duri.ng periods when the DG is supplying maximum post-LOCA load demand as discussed.in.,.the.FSAR-(Ref. 1).. This is based on a conservative lube oil '6onsumption rateof 115 gallIn' per hour and 486 gallons of available lube oil between'the top of the lube oil suction pipe in the engine crankcase (minimum available level) and the mid-scale positi*on on the sightglass'. 252 gallons or 7 days of available lub6 oil is actually indicated at 1 inch visible in the sightglass.
With Ž 2.5 inches visible in the sightglass, a conservative supply.of lube oil is ensured for 7.days of full load operation.
C.1 This Condition is entered :as a result-of.a failure to meet the acceptance criterion of SR 3.8.3.3.
Normally, trending o.f particulate levels a1.1ows sufficient time to correct high particulateIlevels'prioý'}to reaching the limit of acceptability.
Poor.-sample procedures (bottom sampling),
contaminated sampling equipment, and errors in laboratory analysis can produce failures that do not follow a trend.
Since the presence of particulates does not mean failure of thefuel oil to burn properly in the diesel engine, and particulate concentration is unlikely to change significantly between Surveillance Frequency intervals, and proper engine performance has been recently demonstrated (within 31 days), it is prudent to allow a brief period prior to declaring the associated DG inoperable.
The 7 day Completion time allows for further evaluation, resampling, and re-analysis of the DGfuel oil.
(continued)
PALO VERDE UNITS 1,2,3 B 8.8.3-4 REVISION 0