ML20059C753
| ML20059C753 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/27/1990 |
| From: | Underwood K GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20059C648 | List: |
| References | |
| NUDOCS 9009050307 | |
| Download: ML20059C753 (29) | |
Text
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ENCLOSURE'
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POST: REPUBLTNG 0UTAGE
[
.STARTUPLTESY REPORT
- UNIT
1 CYCLE 13 -
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-EDWIN - I.- HATCH' NUCLEAR : POWER PLANT -
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Prepa re d by: /
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K.A. - U nderwood Reactor Engineer I
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_ PLANT E.
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HATCH t
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STARTUP TEST; REPORT.
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H 1.0': INTRODUCTION ij.
1.1 - PURPOSE l
ThisLreport consists ofLa' summary:of selected static, y.
and dynamic reactor core performance tests conducted j
!,i prior to.and during the beginning-of-cycle startup of Hatch Unit 1 Cycle 13.
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1.2-PLANT DESCRIPTION D
The Edwin.I-. Hatch Nuclear Power Plant Unit 1 Eis-a General Electric-design' single-cycle boiling' water a
reactor (BWR/4).
Hatch Unit 1 is rated at 2436:MW(th)y with a generator rating at'this power of 810 MW(e)'.
The'plantcis located on the south side of the Altamaha.
-i River, Southeast of the intersection of the river.with U.S.~ Highway #1 in.the Northwestern sector of Appling County,: Georgia..
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4 1.3L POST-REFUELING OUTAGE STARTUPITEST DESCRIPTION-The Edwin;I. Hatch-Nuclear Power Plant Unit'11 resumed commercial operation on 06/06/90 after completing a.
109; day' refueling / maintenance outage.; The following:
core performance tests were performed ~as part of the post-refueling-outage startup test program:
Core Verification Control Rod Drive Friction Testing Control Rod Drive Timing Full Core-. Shutdown Margin 4 Demonstration
-Cold critical Eigenvalue Comparison Whole1 Core LPRM.Celibration APRM Calibration Control ~ Rod Scram Time Testing Reactivity Anomaly Calculation n
The-purpose for, a brief-description of, and acceptance criterionLfor each of-the tests listed above.is enumerated in Section 3 of this report.
1.4 POST-REFUELING OUTAGE STARTUP TEST ACCEPTANCE CRITERIA Where applicable, a definition of the' relevant acceptance criteria'for the-test is given-and is.
designated either " Level 1" or " Level 2".
A Level 1 criterion normally relates to the value of a process variable which is used as the basis for the reload safety analysis with supplements prev 4.ously submitted to the. Nuclear Regulatory Commission and/or.which are affected by the limiting condition for operation in the Unit's Technical Specifications.
A Level 2 criterion is associated with expectations related to the design performance of systems or components.
If a Level 2 criterion is not satisfied, operating and testing plans would not necessarily be.
altered.
Investigations of the measurements and of the analytical technique used for'the prediction would be initiated l
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2.0; CYCLE DESION:
SUMMARY
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.2.1~
CORE DESION.
SUMMARY
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Cycle,13 was: designedLto operate approximately - 412
'j ef fective: full; power days (EFPDs) at rated conditions,-
with an' additional 8 EFPDs1available from increased
. core. flow..
One hundred and eighty fresh fuel-bundles i
were loaded in'a. conventional core. configuration.
With the exception of the four ANF 9x9" Lead Fuel
-U Assemblies,Lall fuel assemblies loaded ~in the-interior i
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- of the core in Cycle'13 have barrier 1 cladding which
- permits the elimination of PCIOMRs.
Control rod sequence exchanges are to'be performed;at-core =
exposure increments of 2000 mwd /ST.
a 2.2 REACTIVITY / THERMAL LIMIT MARGINS
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- The.two parameters'which describe the global behavior-E of the core' reactivity throughout'the cycle are-bot-
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excess reactivity (HER) and cold shutdown margin (CSDM).
The beginning-of-cycle (BOC)-hot excess pls reactivityJis 2.07%.
This is also theLpeak HER of.the t
y cycle.. The' minimum predicted cold shutdown margin of 1.97% occurs at BOC.
f LHOR/ design margins.were relaxed slightly in an attempt to maximize cycle energy generation.
Standard thermal margin-design-goals can be readily demonstrated.at the' expense of some. cycle energy.
Cycle 13 is the first cycle of operation at Plant
~
Hatch-to incorporate channel bow ^ effects en thermal limits per NRC.Bulletin 90-02.
These' effects will be
-t sg implemented by adjusting the bundle R-factors such L
that the calculated bundle CPR will be increased;by i
L approximately:0.1 delta CPR.
l 2.3-FUEL
SUMMARY
/ CORE LOADING PATTERN DESCRIPTION Hatch'l Cycle 13 is a conventional core loading which L
was designed to-achieve 8900. mwd /ST.
The-loading pattern is quadrant mirror symmetric.
The Reload 12 batch of fresh fuel contains a total of~180 bundles loaded in the interior of the core.
These bundles are GE9B with 3.15 weight parcent U-235.
New features of the OE9B design (relative to GE7B)-are:
'2.3.1 A redesigned spacer for greater MCPR margin and reduced pressure drop, 2.3.2 A higher limit for linear heat generation rate, 2.3.3 A largo central water rod for more efficient fuel utilization, b
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"n' l-4 2.3.4ifAfgreaterffuelLrod propressurization and-
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1 enrichment for highertdischarge: burnup,:
22 315i Axializoning foff uranium-enrichment ~ and ' gadolinia:
J
. concentration for-power-shaping and: improved fuel-
- j efficiency, and,
- )
2;3.6 Redesigned upper and lower tie plates for improved j
bundle flow.-
j Design; features such as: axial uranium enrichment and-5
' gadolinia concentration are optional and-are not inc3uded-in the design-of the-Reload 12. bundles-utilized _inicycle 13.
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SUMMARY
- 0F POST-REFUELING OUTAGE STARTUP TEST RESULTS:
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3i1= ' CORE VERIFICATION 3.1.1 l Purpose-J l
To verify that all fuel assemblies have;been-properly loaded into,the reactor _ core as per.the licensed final loading pattern 11ncluding fuel-bundle. location, orientation, and seating.;
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fs p 3.1. 2 : ' Acceptance' Criteria-3.1.2.1 Level 1 criteria: ' Esch: fuel aheambly must' be-verified to be.in its proper locationtas:
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specified by the General Electric' final loading pattern-(Licensed Core)fand be H
correctly seated in its respective cell.
3.1.2,2 Level'2 criteria: EN/A fa L
3.1'3-Test Description j
m The Hatch. Unit 1 Cycle 13 core verification was'.
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performed by use of an underwater TV camera to l{,
visually inspect the location (by bundle serial L
number-identification), orientation,.and seating j
of each of the 560 fuel assemblies that comprise-R the as loaded core.
l 3.1.4 Test Results' I
The core verification?was' performed on 4/18/90 in:
L accordance with engineering, procedure p
42FH-ERP-014-OS, Fuel Movement'0peration.
l Videotapes of the core loading indicated all fuel assemblies were'in'tne proper location with proper L
L orientation.
Fuel bundle seating verification L
indicated that all bundles were properly seated.
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R3. 2 CONTROLiROD' DRIVE FRICTION TESTING.
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3.2.1 Purpose
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t Toidemonstrate that'the. control rod drive l system' l
operates properly following.the completion:of a 4
core alteration.
InLparticular, this. functional
, test demonstrates the absence of excessive friction in the contro11 rod drive'from internal:
drive obstructions following extensive control rod drive maintenance / replacement.
3.2.2 Acceptance criteria e
3.2.2.1' - Level :1 criteria:
The differential-pressure.
variation of all control' rod drives to be testedimust be-less than or eqN11 to 15 paid for continuous insertion.
If tnis criterion cannot-be satisfied, then a settling-testimust o
be performed in which case the differential
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settling pressure should not be less;thanJ30-psid over the full stroke.
Lower differential pressures in the-settling test are indicative of excessive friction.
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3.2.2.2 Level 2 criteria:
N/A 3.2.3 Test Description control rod drive friction testing is normally-performed on all control drives that have been replaced or have undergone extensive maintenance-repair'during the refueling outage.
Infessence, the functional test measures the, differential pressure across the drive piston during a normal
' insertion stroke.
If necessary, a settle test, which measures the differential' settling pressure h
of each notchs is performed on a control rod drive j
during a. withdrawal or insertion stroke.
3.2.4 Test Results control rod friction testing was performed on 4
4/18/90 for twenty-four control rod drive units.
Twenty of these drive units were replaced during the outage.
The testing was performed under engineering procedure 42IT-011-001-OS, Control Rod 1
Friction Testing.
The test results indicated that all of the control rod drives were satisfactory either by the normal insertion differential pressure test or the settle test.
A summary of the results of the control rod friction testing is given in Attachment 1.
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. 3.3-CONTROLLROD1 DRIVE TIMING
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- 3.3.1.Purppse To demonstrate that thefcontrol rodLdrive-system c?arates' properly-.following the.-completion of a corefalteration._ In particular,;this functional test verifies that the insertJand withdrawal ~
in capability of-the-control rod drive, system is within acceptable limits.
3.3.2 Acceptance Criteria 3.3.2.1 Level 1 Criteria:
The insert and withdraw-m
-drive time-for each control rod ~ drive must be between 38.4 and 57.6 seconds.
In: the event-that:a control rod < drive fails to meet'this criteria, then the' applicable drive must be adjusted and a-new criteria of 43.2 to 52.8-seconds is applied to the adjusted drive.
3.3.2.2>
Level 2 Criteria:
N/A 3.3.3 Test Description Control rod drive timing is performed once per operating cycle-on all control rod. drives.
Normal withdrawal and insertion times =are recorded for
-each of the drivestunder normal drive water pressure.
If acceptable withdrawal and/or insertion cannotJbe.obtained for normal drive water pressure, then the respective needle valve for.the applicable ~ withdrawal and/or insertion-stroke must be adjusted until an acceptable drive 4
time is. achieved in accordance with the'above criteria..
3.3.4 Test Results control red drive timing was performed on 4/28/90 for all 137 control rod drives in accordance with operations procedure 34SV-C11-004-15, CRD Timing.
Each' control rod drive was determined to have, or adjusted (where necessary) to have, a normal insertion and withdrawal speed as required.
A 3,
summary of the results of this functional test is given in Attachaer.t 2 e
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[3. 4 FULL CORE' SHUTDOWN _ MARGIN _DEMONSTRATIONE
- 3. 4.1; Purpose
+l To demonstrate that the reactor can be made J
subcritical for any-reactivity. condition during Cycle'13-operation-with.the: analytically-5
_ determined highest worthTcontrol rod,n capable of'
-withdrawal,-fully withdrawn;and all: other rods
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fully inserted.
3. 4.~ 2 LAcceptance Criteria 4 y E
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Level 1 Criteria:
The fully-loaded core must be suberitical by at least 0.'38%_ delta,kLwith the-
- analytically 1determinedLhighest-worth control rod, capable of withdrawal,Lfully withdrawn and all' n
other rods fully inserted'at the most reactive-
- condition during the cycle..
Level 2 Criteria:
N/A-4
- 3. 4. 3! Test Description The full core shutdown margin demonstration was
. performed. analytically during the Hatch. Unit 1 Cycle-13 BOC in-sequence critical with the reactor E
core in a-xenon-free state.
To account for reactivity effects-such as; moderator temperature,
, reactor period,-and one rod out criterion, correction-factors are used to-adjust the startup y
condition to cold conditions with the highest worth control rod fully withdrawn.
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3.4 4 Test Results-The full core shutdown margin demonstration was performed on 06/01/90 in accordance with core calculation procedure 42CC-ERP-010-OS, Shutdown Margin Demonstration.
Results of this calculation yielded a cold shutdown margin of 2'.06% delta k.
The minimum cold shutdown _ margin was also 2.06%
delta k because BOC is the most reactive point:in y
this operating cycle.
A summary of the shutdown
+
margin demonstration is given in Attachment 3 of this report.
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'3 51l COLD 1 CRITICAL EIGENVALUE COMPARISON-
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- 3. 5.1D : Purpose j
a To-compare the_ critical eigenvalue calculated using;the actual col'd, xenon-free critical control i
rod configuration '(corrected-- for' moderatora i
temperature and reactor period reactivity effects)
'to'theEcold critical eigenvalue assumed'in the l
cycle management analyses.
3.5.2 Acceptance Criteria ~
p 3. 5 '. 2.1 Level I criteria:
N/A if ti 3.5.2.2: Level 2 criteria:-
N/A
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3.~ 5 3 Test Description-The cold critical eigenvalue is th'e assumedcvalue-of~the1 PANACEA 3-D simulator model Keff at.which p
criticality isDachieved with the reactor in a j.
xenon-free state and the-~. coolant at:68 degrees F.
j This value is: determined based on historical data C
t and used for cycle management' analyses by the BWR
. Core Analysis Group of southern: Company Services L
in Birmingham, Alabama.
Once the actual critical state is. achieved during the beginning of cycle-startup, the1 applicable data is sent.to the BWR Core Analysis Group'and,the actual (corrected for-moderator temperature And reactor period g
reactivity effects) cold critical eigenvalue is calculated.
This va'lue is thenicompared tocthe~
assumed critical eigenvalue as a method of j
validating rod worth and shutdown' margin L
. calculations throughout the cycle.
The actual critical eigenvalue is also entered into a database for predicting future cold critical p
eigenvalues.
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- 3'.5241' Test Results.
for Hatch Unit =if The beginning-of-cycle startup/90.
Cycle.13:was; performed on?6/1 LThe:following:
reactor core conditions were observed when a-
"criticalistate was achieved:
~
Sequence-A2' RSCS 0roup111 l Fully 1 withdrawn 7
RSCSfGroupL2:
Fully withdrawn--
. RSCS Group:3 5'contro11 rods fully' withdrawn andsthe 6th>
- controlt rod -- ( 3 4-07 )
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withdrawn to. notch 6.
Moderator Temperature 175 degrees F
-Reactor 1 Period?
155.8 sec Control' Rod Density 0.738
-Calculated MCSDM
-2.06% delta'k:
ALcold. critical eigenvalue of-1.00324 was calculated 1from the actual-critical data given above.- This compares well to an assume'd-value of 1.0025.
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.3.6
- WHOLE-CORE ~ LOCAL POWER RANGE-MONITORi(LPRM') CALIBRATION.
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- 3. 6.1E ' Purpose
.i Tol determine (1) The LPRM calibratinn constants.
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such that when multiplied by the actual LPRM' S
readings will produce calibrated LPRM readings proportional to: the traversing in-core probe (TIP) signal = readings at the LPRM locations'and (2)-The BASE and;BASELP arrays which'contain.the machine normalized. full powerLadjusted TIP signals at'every?
node and LPRM detector location,,respectively.
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3.6.2 Acceptance criteria.
3'.6.2.1 Level 1 criteria:
N/A Ji
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3.6.2.2 Level 2 criteria:
N/A i
3.6.3' Testing Description l
The<whole core LPRM calibration and-BASE E
' distribution calculation determines the LPRM:
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- calibration constants and the. BASE and BASLP distributions'used in the axial power-distribution i
calculations.
In. essence,-the TIP systemiis used:
p in conjunction with the process computer to generate the axial 1 distribution of
. machine-independent / core average power-independent i
TIP.. signals.
The-axial-distribution of-~ machine a
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normalized full power: adjusted TIP-data is used to generate LPRM calibration constants required for i
TIP normalized LPRM~ readings.
l L ig In. addition, machine normalized full power adjusted LM TIP readings are generated at every axial node B
(BASE. distribution) and-at every LPRM detector
[
location (B ASLP ' distribut ion ).
These arrays are
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l used as input data in the core p
.e calculation / monitoring. programs to accurately calculate the power distribution at every node in li the core.
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13.'6.4 Test Results; s
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Whole= core'LPRM Calibration and-EASE: distribution a
-was performed'in.accordance with' Engineering.
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o-procedure'42CC-ERP-015-05,iOD1 andt0D2 NUMAC TIPL t
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operationist approximately; 2 4. 5%,13 7. 2%,. 50.~ 3%, - and.
77%' power.'~ LPRM calibrationfoonstants, BASE.and
'BASLP arrays, were. computed:by.the_ process--computer:
1 and. subsequently used.successfully by the process' a
. computer to calculate the nodal power distribution-1 s
and the core thermal limits.
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- 3. 7 - APRM CALIBRATION h
3.7.1 Purpose h
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- To: calibrate the APRM system lto actual core thermal
. power,-as determined by a heat balance.:
3.7.2 Acceptance criteria i
3.7.'2.1 Level 1 criteria:
The-APRM-readings'must be
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m-within a. tolerance of 2% of< core: thermal power >
t as determined from:a heat balance.
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3.7.2.2 Leveli2 criteria:
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3.7.3 Test Description The APRM gains are adjusted after major power level changes,_if required, to read the actual core y
thermalipower as determined by'a1 heat balance i_n accordance with procedure 34SV-SUV-021-OS,=APRM Adjustment to Core Thermal Power. -The heat: balance <
' required for the calibration process may be
-obtained from the process computer program P1 o
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'(Periodic Core Evaluation) and OD3 -(Core Thermal Power and APRM Calibration) or from a manual heat
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balance in accordance with core calculation s
procedure 42CC-ERP-001-1S,-Core Heat Balance-Power f
Range.
3.7.4 Test Results
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APRM calibration was. performed-in accordance.with plant procedure 34SV-SUV-021-OS, APRM-Adjustment to Core ~ Thermal Power at approximately 18%, 24%,-28%,
37%, 50.2% and 85.9% power..Each APRM was i
calibrated within a 2% tolerance to read core' L'
thermal power as calculated by the process computer.
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3'. 8 l CONTROL RODLSCRAM. TIME TESTING.
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' 3. 8. li Purpose To demonstrate that the control rod drive' system-j functions as designed with respect to scram insertion times following the completion of core
~ alterations.
1 3.8.2:l Acceptance Criteria i
3.8.2.1: -Level 1 criteria i
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. a ), 'The average = scram insertion time for all
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operable control rods from theJfully withdrawn;
- po-3. tion,-based-on de-energization of the. scram
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pilot, solenoids, with reactor steam dome a
pressure-above 950 psig-shall not exceed.the following:
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l Notch Position Average.
s from Fully Insertion Withdrawn Time (secs) 46 0.358-36 1.096 26 1.860 06 3.419
-(b)
The average. scram insertion time, from the
- )
fully withdrawn position, for the 3-fastest control rods in.t h group of four. control rods arranged in a 2x ray, based on the m
de-energization. g,fthe scram pilot solenoids',
i shall not exceed (- e following:
Notch Position
-Average >
'from Fully Insertion Withdrawn Time (secs) 46 0.379 36 1.162 26 1.972
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06 3.624
.(c)
The maximum scram insertion time of each control rod, from the fully withdrawn position to position 06, based on the de-eni.gization of the' scram pilot solenoid, shall ne-exceed 7.0 seconds.
Level 2 criteria:
N/A
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',1 Test 1 Description H
.3.8.3 3
The' control-rodidrive scram time testingiwas s
performed'in accordance with engineering procedurel
~
- 42SV-C11-001-15, Control Rod Scram: Testing,.L.with.
-the steam. dome pressure above'950 psig.- The. test 1
consists of scramming each1 control rod, collecting.
the-resultingfscram time data, and analyzing thof ~
l data in accordance'with.the procedure-to ensureL 9
compliance with.the acceptance; criteria noted' above.
3.8.4 Test-Results
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All control rod drives were~ tested.in'accordance ch,
-with engineering procedure 42SV-C11-001-1S,. Control E-Rod Scram. Testing,.with the steam dome pressure'
-l greater lthan'950 psig.
A summary of the results is
=i giveniin Attachment-4 of this-report, j
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' 3. 9.
REACTIVITYLANOMALY CALCULATION I
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Purpose:
+
e To; check-for possible reactivity _ anomalies as the p
core excess reactivity changesLwith exposure.
3.9.1L Acceptance Criteria 3.9;1.1, Level 1 criteria:
The corrected control rod density lshall not differ from a-control rod density equivalentsby moreLthan plus or minus-1%' delta k.
i 3. 9.1. 2 ' Level 2. criteria:
N/A p
3.9.2 Test Description L
During_the BOC-startup following a refueling outage
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and every month thereafter, a-reactivity anomaly calculation is performed to monitor the core reactivity < ring the cycle.
Since anticipated-
=j operation or ananticipated~ events may place the j
L reactor in a. condition other than that.for which
- j the baseline = anomaly. curve was developed, the_
l actual control rod density is corrected i or 1
f off-rated conditions. 'The corrected control rod
' density is then compared to the reactivity anomaly 1
curve'provided in the Cycle Management Report to ensure.that-the corrected-control rod density:is 1
within a plus or minus 1% delta k acceptance band 1
about the curve, j
- 2. 9. 3-The reactivity anomaly calculation was performed in accordance with 42CC-ERP-007-OS, Reactivity Anomaly:
Calculation, on 6/29/90.
The corrected control' rod density was well within the acceptance criteria range as specified above.
The results of-this calculation are given in Attachment 5'of this H
report.
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CONTROL ROD FRICTION TEST RESULTS!
DRIVE RX.-
CRD DIFF INSERTION DIFF. PRESS SETTLE TEST LOCATION PRESS.
PRESS.
REQUIRED.
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270 66.6 53.3 13.3 NO.
l 06-15 0
270 66.7 52.7 14.0 NO ~
l 06-23 0
270 84.0-60.0 24.0 YES 06-19 0
270 70.0 58.0 12.0 NO 14-15 0
270 70'0 56.7 13.3' NO 42-11 0
270
-73.3 58.3 15.0 NO 42-19 0'
270
-71.7 58.3 13.4 NO-46-23
'O 270 71.7 56.7 15.0 NO 50-31 0
270 70.0
- 56.7 18.3 YES~
50-27 0
270 73.0 58.0 15.0
- NO 34-43 0
270 70.0-61.5 8.5 NO 38-39 0
270 67.3
. 58.7 8.6 N0' i
34-35 0
270 75.0 60.0 15.0
- NO 30-35 0
270 66.7 56.7 10.0 NO 34-31' O
270 63.0, 54.0-9.0L
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ATTACHMENT 1 CONTROL ROD FRICTION TEST RESULTS DRIVE RX.
CRD'DIFF INSERTION'DIFF. PRESS'
. SETTLE--TEST' LOCRTION PRESS.
PRESS.
. REQUIRED' 34 0 270 68.0' 56.0.
12.0 NO 26-51 0
270 71.0 56.0-15.0 NO 26-35 0
270 69.0 57.0 12.0 NO 22-35 0
270 72.0 58.0,
.14.0 NO 22-51 0
270
- 73.0 60.0 13.0 NO 22-47 0
.270 70.0 57.0
-13.0 NO 4
22-39 0-270 68.0 56.0 12.0 N0' l
14-39 0
270 71.0 56.0 15.0
~NO 06-35 O
270 71'.0-56.0 15.0 NO l
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' ATTACHMENT 2-
. ~
..m
- + Y
CRD. TINING RESULTS DRIVE WITHDRRH INSERT-
. DRIVE
'HITHDRRH
~
INSERT LOCATION TINE TIME-LOCRTION TIME TIMEi
~
18-51 41.1 48.9 10-43 48.7 50.3-22-51 42.6 49.0 14-43 40.4 143.9--
26-51 51.8*
47.9k 18-43 44.3
.47.8 30-51 44.0 49.5 E2-43 39.2 42.1 34-51 51.4-51.4 26-43 41.8
'49.2 10-47 49.3 47.6 38-43
-40.9 55.0-I 14-47 50.3 44.1 34 -43 44.19' 43.3e 18-47 47.9 50.4 38-43 45.4 50.3 22-47 48.3*
44.3*
42-43 43.5 50.8
~
5k.O 46.7 46 42.9 45.1 26-47 39.8 52.9 30-47 41.8 54.3 06-39 i
34-47 47.25 49.5*
10 46.8 47.4 38-47 47.5 50.2 14-39 43.5e-38.7-42-47 45.6 54.6 18-39 49.3 44.7 l
l 06-43 46.3
.51.3 22 49.3e
-45.2e
~
I
- RDJUSTED DRIVE l
=
=
~, [ ---
~
- n. s w
~
~
ATTACHf1ENT it:
1
-CRD TIMING RESULTS DRIVE NITHDRRW INSERT DRIVE-MITHORAN-INSERT LOCATION TINE TINE LOCATION iTIME TINE-
~
26-39 41.4 52.7 38.50.8e
'-46.9 r 30-39 50.4 54'.8 42-35
-46.1 47.41 34-39 42.0 56.3 46-35 42.3 47.6 38-39 49.3e 45.6*
50-35 46.4*
44.9 w -
~ !
42-39 42.8 44.6 02-31 42.8 47.5
-I 46-39 39.2 52.1' 06 42.4 51.2 02-35 47.2e 51.2e 10 ~39.7
- 45.1
~
06-35 53.5 42.9 14-31 49.3 44.7 i) 10-35 41.8 45.8 18-31' 42.8:
-54.4 e
14-35 47.7-47.7 22 47.6 47.7 18-35 39.3 48.9 26-31 49.2 45.7
.I 22-35 38.5 42.8 30-31--
51.2e
- 50.2e ^
26-35 44.1e 51.9
- 34-31 43.9 47.0e 30-35 49.2e-54.3 138-31 52.5 48.1 34-35 43.8 51.3 42-31
- 44.4-
.46.901 l
i-
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- -RDJUSTEDLDRIVE-i
-m-
_fA,...._...
e'
,._. ~ _.i J.
(.
e.
' i.
a-V ATTACHF1ENT 2 CRD tit 1ING RESULTS DRIVE WITHDRRN INSERT DRIVE WITHDRAM INSERT.
I LOCRTION 1
TIME I
TIME LOCRTION TIME TIME-
._ _ N;_ __ __._
46.9 54.3 46-31 47.7 57.8
___ ' 02-23
_..___ e
. ___ _. I 50-31 51.05
'43.8 06-23 49.0' I
49.1 I
_L 4
_u_.
t l
^10-23 42.8 4 8. --
02-27 39.1 47.5......L' I
14-23 50.2 51.4
. l l7_
._ a n.. _. _=. :...
06-27 47.75 44.1*
10-27 42.3 1
51.5 l
18-23 40.1 49.9 22-23 41.3
-46.9 l
14-27 53.7-50.6 I
26-23 5 l. 3 47.1 I{*.
=
18-27 38.8 47.4-1 30-23 43.5*
49.O 22-27 54.4 46.5
_. 1
._ l _
l
- 56.5-34-23 41.9 l
51.3.
26-27 42.8 1
46'.6 47.6 30-27 46.6 48.8 38-23 j
i 42-23 48.1 1
43.1..
34-27 50.9e l
57.3 46-23 48.9*
38.4 l
1 j.f g
j 38-27 48.8 l
48.3 46.0 49.6 l
.If
'50-23 42-27 50.2 51.2 51.8 46-27 44.1 46.6 1
02-19
. 40.5 -
50-27 45.9e 47.8*-
06 50.4 I
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i L
- v e ADJUSTED' DRIVE.
Y'
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+
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+
[
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~,
T-ATTACHMENTL2 CRD TIMINGLRESULTS
~
DRIVE WITHDRAW INSERT.
-DRIVE.
WITHDRAM.
INSERT;-
n LOCRTION TIME TIME:
1.0 CATION TIME TIME-~
10-19 42.'4 46.8*
22-15 49.4 48.9; 14-19 48.8*
48.7*
26-15
- 42.4 -
52.2' 18-19 39.4 55.4
.30-15 46.6.
146.0-22-19 44.7 54.4 34-15 49.5
-l 54.1:
i 26-19 44.5 50.0 38-15 53.3 42.7--
30-19 49.7 56.2 48-15 41.5 41.6:
j._
i 34-19 44.9p 46.0*
1 46-15 41.4 42.s I
.g 38-19 40.1 45.5*
06 41.7 48.3e 42-19 49.7*
47.2*
10-11 47.4 57.0 46.1-L 56.2 46-19 i
47.9*
50.0 j
14-11 53-19 50.45 53.0 19-11 47.6 1
53.5 s
-_ _._. x 06-15 46.5e 46.2*
I 22-11~
l 50.O*
48.O*
j
-._..___._____.___4______
._____-_.._._.__.._.._._.g..__.
10-15 44.2 l
44.6*
26-11 56.2 l-46.O
-._._q 55.2^
k-51.5 14-15 50.6e I
48'.2*
i 30-11 J l, _._-
. -. ~ - - _. 1_,
a.:_:--
4 34-11:
40.4-
^
145.2*-
18-15 52.2 46.2*
. __. v
- RDJUSTED-DRIVE-
~
_ i;
-~.:..,...
~'
r.
,.fr.%-
- ~
e..
i ATTACHMENT 2 CRD TIMING RESULTS-m
.-c.
DRIVE WITHDRAW INSERT' l DRIVE-WITHORAM.
INSERT' LOCATION TIME TIME LOCATION--
TIME TIME 38-11 42.7 50.3' 30-03
- 46.6-53.9
? 0.5 56.1 4
42-11 51.4 42.8 ^
34-03 46-11 39.3 51.2*
10-07 44.7 53.3 14-07 44.5 49.7e 18-07 47.5 47~.7e 22-07 43.5 49.7.
26-07 47.6 45.3 30-07 45.1 53.7 34-07 37.9 46.0*
38-07 43.9 55.8 42-07 46.9 54.2 18-03 51.8e 45.Be 22-03 44.8 45.3:
26-03 41.5 47.55 a
~
- LADJUSTED DRIVE L
.i
r: ;
L ATTACHMLNT 3
m FULL CORE SHUTDOWN MARGIN DEMONSTRATION 1
K 0.9803 SRO 7
-2 K
1.0030 CRIT 3
CONTROL ROD DENSITY 0.738 4
RX.
COOLANT TEMPERATURE 175*F
+
t.
5 REACTIVITY CORRECTION FOR TEMPERATURE
.0017A K 6
REACTOR PERIOD 155.8 SEC
~
7 REACTIVITY CORRECTION FOR E. I O D 0.0004A K 8-COLD SHUTDOWN MARGIN (2-1+5-7) c.06
%oK 9
VALUE OF R
0.0
%aK 10 MINIMUM COLD SHUTDOWN MARGIN (8-9) 2,06
%oK j
11 TECH SPEC REQUIRED l
SHUTDOWN MARG 1N 0.38
%AK J
' I ATTACHMENT-4 SCRAN TIME TESTING RESULTS - BO C13'
~
i NOTCH SLOWEST SCRAM AVERAGE SCRAM AVERAGE SCRAM POSITION INSERTION TIME INSERTION TIME INSERTION TIME CRITERIA SINGLE ROD 2X2 ARRAY SINGLE ROD 2x2 ARRAY SINGLE 2x2 ARRAY 3 FASTEST 3 FASTEST ROD 3 FASTEST 46 399
.332 305 298 358
.379 36 1.001
.850 825 815 1.096 1.162 26 1.859 1.406 1.349 1.323 1.860 1.972 06 3.439 2.554 2.484 2.442 3.419
.3.624
-p
- - -'e-'
yc a
m
.yy ym-w-g er D
.e-pow,-
w+%.------gy-- -
4 tw,-
- w e<w e-t g-m,-
p
'O' t
ATTACHMENT 5 REACTIVITY ANOMALY CALCULATION RESULTS Date Performed:
June 29, 1990 Unit 1 Cycle 13 Sequence A2 CMWT:
2429.9 WT:
80.50 ' Mlb/hr Dome Pressure:
998.7 pelo DHS:
21.6 BTU /lb,,
Actual Control Rod Deneity:
0.0912 Corrected CRD = Actual CRD + Correction Correction:
- 3.0187 E-1 *(1 -(CMWT/2438))
+ 1.9 8'/ 5 E-1 *(1 -(WT/ RATED CORE FLOW))
+ 2. 6 5 63 E-3
- (DESIGN INLET SUSCOOLING - DHS).
+ 6.584 3E-5
- (1020 - PR)
--O.0037 Corrected CRD = 0.0875 Predicted CRD = 0.0951
+ 1 % Value - O.1415
-1 % Value = 0.0487
. r' 1