HL-1781, Plant Ei Hatch Unit 2,Post Refueling Outage Startup Test Rept,Unit 2,Cycle 10

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Plant Ei Hatch Unit 2,Post Refueling Outage Startup Test Rept,Unit 2,Cycle 10
ML20082H858
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 08/22/1991
From: Beckham J
GEORGIA POWER CO.
To:
References
HL-1781, NUDOCS 9108260336
Download: ML20082H858 (28)
v • d • e


Text

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. Georgia Ibwer Company i O g. 40 inverrma Cente Pa%e i e' Itst 0";ce th 'WS I Gam ngham, kabama 3S %

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n J. T. Beckham, Jr.

bCOIMlilbOWCf Vice heudent -NJtNr ' '

um Pwject HL-1781 002072 August 22, 1991 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 PLANT HATCH - UNIT 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 PLANT STARTUP TEST

SUMMARY

REPORT Gentlemen:

In accordance with the requirements of the Plant Hatch Unit 2 Technical Specifications, Section 5.9.1.1, Georgia Power Company hereby submits a Plant Startup Test Summary Report. The report presents a summary of selected static and dynamic reactor core performance tests conducted prior to and during startup from the Spring 1991 Unit 2 maintenance / refueling outage. The unit is currently operating in Cycle 10 This report is being submitted as required by the Unit 2 Technical Specifications to document the first reload quantity application of the GE 9 fuel design in the Unit 2 reactor. As expected, the testing identified no unexpected behavior of_the reactor core and instrumentation.

If you have any questions in this regard, please call this office, Sincerely,

/ ..

6;5. T. Beckham, Jr. k JKB/cr

Enclosure:

Post Refueling Outage Startup Test Report - Unit 2 Cycle 10 cc: (See next-page.)

T 9508260336 910822 PDR PI / -

P ADOCK 05000366 d PDR j ,g'f j d'/r I g

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Georgia Power d U.S. Nuclear Regulatory Commission August 22. 1991 Page Two

Enclosure:

cc: Ecoraia Power _fompany Mr. H. L. Sumner, General Manager - Nuclear Plant NORMS U.S. Nuclear Reaulatory Commission. Washinaton. D.C.

Mr. K. Jabbour, Licensing Project Manager - Hatch U.S. naclear Reaulatory Commission. Reaion 11 Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert, Senior Resident Inspector - Hatch 1

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POST -R E F U E L I N G OUTAGE '

STARTUP T m T REPORT UNIT  ? CYCLE 10 O 1

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%) PLANT E.-1. HATCH UNIT 2 CYCLE 10 STARTUP TEST REPORT

1.0 INTRODUCTION

i 1.1 PURPOSE' This report is submitted to the NRC as required by Unit Two Technical Specification 6.9.1.1 because-a reload. batch of new fuel with a design never used before:in this core (GE98) has been loaded into Unit 2 Cycle 10.

This report consists of a summary of selected static and dynamic reactor core performance tests conducted prior tojand during the beginning-of-cycle startup of Hatch Unit ~2-Cycle 10.

1.2 PLANT DESCRIPTION

ry The Edwin I. Hatch Nuclear Power Plant' Unit-2 is a

'UE General Electric design single-cycle boiling water reactor (BWR/4). Hatch Unit 2 is rated at 2436 MW(th) with.a generator rating at.this power of 822 MW(e).

The plant is located on the south side of the Altamaha River, Southeast of the intersection of the river with U. G _ !!ighway # 1 in the. Northwestern sector of Appling County, Georgia.

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kY L1.3 POST-REFUELING-0UTAGE STARTUP TEST DESCR7PTION The Edwin I. Hatch Nuclear Power P.\ ant Unit 2 resumed' l commercial operation-on 06/02/91 after completing a 75 D day refueling /saintenance outage. The following core

  • performance tests were performed as part of the post-refueling outage startup test program:

Cors= Verification control Rod Drive Friction Testing Control Rod Driva Timing '

Full Core Shutdown Margin Demonstration critical Eigenvalue Comparison- l Whole Core LPRM Calibration APRM1 Calibration .

Control Rod Scram Time Testing Reactivity Anomaly calculation The purpose for, a brief description of, and the acceptance criteria for each of the tests listed above is enumerated-in Section 3 of this report.

l' . 0 POST-REFUELING-00TAGE'STARTUP TEST-ACCEPTANCE CRITERIA

-s ,Where1 applicable,ca definition ~ofmthe relevant:

acceptance' criteria for the test is;given and is designated either "Lovel 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 previously-submitted to'the NuclearLRegulatory Commission and/or which are affected;by the limiting conditioncfor operation in the-Unit's Technical Specifications.-

.A Level 2 criterion is associated'with expectations related1to'the design performance of systems or components. :If~a Level 2 criterion is not satisfied, operating and testing plans:would not-necessarily be j). altered. Investigations of the measurements and of' the1 analytical technique used for.the prediction would be initiated.

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2.0 CYCLE DESIGN

SUMMARY

2.1 CORE DESION

SUMMARY

Cycle 10 was designed to operate approximately 420 effective full power daya (EFPDs) at rated conditions, with an additional 9 EFPDs available from increased core flow. One hundred and seventy six fresh fuel bundles were loaded in a conventional core configuration. With the exception of the four ANF 9x9 Lead Fuel Asstablies, all fuel assemblien loaded in the interior of the core in cycle 10 have barrier cladding which permits the elimination of PCIOHRs.

Control rod sequence exchanges are to be performed at core exposure incrementa of 200) HWd/ST.

2.2 REACTIVITY /THERHAL LIMIT HARGINS The two parameters which describe the global behavior of the core reactivity throughout the cycle are hot excess reactivity (HER) and cold shutdown margin (CSDH). The beginning-of-cycle (BOC) hot excess reactivity is 1.87%. The peak HER of 2.08% occurs at (3

ss/ approximately 5.0 mwd /ST. The minimum predicted cold shutdown margin of 1.40% occurs at BOC.

Thermal margin design goals of 10%, 10%, and 7% for MFLPD, HAPRAT, and MFLCPR, respectively, were generally met throughout the cycle. To maximize cycle energy, LHOR design margins were relaxed approximately 1%. Standard thermal margin design goals can be readily demonstrated at the eFpense of Some cycle energy.

2.3 FUEL

SUMMARY

/ CORE LOADING PATTERN DESCRIPTION Hatch 2 Cycle 10 is a conventior.a1 core loading which was designed to achieve 9050 mwd /ST. The Roload 9 batch of fresh fuel contains a total of 176 bundles of approximately 3.14% enrichment, loaded in the interior of the core. The loading pattern is quadrant mirror symmetric. Four of the reload bundles are OE11 LUAs with the remainder being GE9B. New features of the GE9B design (relative to OE7B) 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 large central water rod for more efficient fuel utilization,

t 2.3.4 A greater fuel rod propressurization and enrichment for higher discharge burnup, ,

2.3.5 Axial zoning of uranium enrichment and I gadolinia concentration for power shaping and improved fuel efficiency, and, 2.3.6 Redesigned upper and lower tie plates for improved bundle flow.

Design features such as axial uranium enrichment and gadolinia concentration are optional and are not included in the design of the Reload 9 bundles utilized-in cycle 10.

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O 3.0

SUMMARY

OF POST-REFUELING OUTAGE STARTUP TEST P.ESULTS 3.1 CORE VERIFICATION 3.1.1 Purpose To verify that all fuel assemblies havn been properly loaded into the reactor core as per the licensed final loading pattern including fuel bundle location, orientation, and seating.

3.1.2 Acceptance Criteria 3.1.2.1 Level 1 criteria: Each fuel assembly must be verified to be in its proper location as specified by the General Electric final loading pattern (Licensed Core) and be correctly seated in its respective cell.

3.1.2.2 Level 2 criteria: N/A 3.1.3 Test Description The Hatch Unit 2 Cycle 10 core verification was performed by use of an underwater TV camera to visually inspect the location (by bundle serial number identification), orientation, and seating of each of the 560 fuel assemblies that comprise the as loaded core.

3.1.4 Test Results Core verification was performed on 5/7/91 in accordance with engineering procedure 42FH-ERP-014-05, Fuel Movement. The initial videotaping of the core identified four misplaced bundles (ANFLT1, ANFLT2, ANFLT3, and LYH569) and one bundle in need of resenting (LYH569). On 5/8/91, the misplaced bundles were moved to their correct locations and LYH569 was resented.

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i O 3.2 CONTROL ROD DRIVE FRICTION TESTING ,

3.2.1 Purpose To demonstrate that the control rod drive system operates properly following the completion of a core alteration. In particular, this functional test demonstrates the absence of excessive friction in the control rod drive from internal drive obstructions following extensive enr. trol rod i drive maintenance / replacement. l 3.2.2 Acceptance criteria 3.2.2.1 Level 1 criteria: The differential pressure variation of all control rod drives to be tested munt be less than or equal to 15 psid for continuous insertion. If this criterion cannot be satisfied, then a settling test must be performed in which case the differential settling pressure should not be less than 30 paid over the full stroke. Lower differential pressures in the settling test are indicative of excessive friction.

O(j 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. In essence, the functional test measures the differential pressure across the drive piston during a normal insertion stroke. If necesarry, a settle test, which measures the differential settling pressure of each notch, is performed on a control rod drive during a withdrawal or insertion stroke.

3.2.4 Test Results control rod friction testing was completed on 5/9/91 for nineteen control rod drive units.-The testing was performed under engineering procedure 421T-C11-001-0S, control Rod 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 At.tachment 1.

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3.3 CONTROL ROD DRIVE TIMING 3.3.1 Purpose To demonstrate that the control rod drive system operates properly following the completion of a -

core alteration. In particular, this functional test verifies that the insert and withdrawal f capability of the control rod drive system is within acceptable limits.

3.3.2 Acceptance criteria l

t 3.3.2.1 Level 1 Criteria: The insert and withdraw i drive time for each control rod drive must be between 38.4 and 57.6 seconds. In the event that a control rod drive falls to meet this )

criteria, then the applicable drive must be  !'

adjusted and a new criteria of 43.2 to $2.8 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 r withdrawal and insertion times are recorded for each of the drives under normal drive water pressure. If acceptable withdrawal and/or l insertion cannot be obtained for normal drive  ;

water pressure, then the respective needle valve for the applicable withdrawal and/or insertion stroke must be adjusted until an acceptabic drive time is achieved in accordance with the above criteria. j 3.3.4 Test Results Control rod drive timing was completed on 5/10/91 for all 137 control rod drives in accordance with 3 operations procedure 34SV-C11-004-1S, CRD Timing. J Each control rod drive was determined to have, or was adjusted (where necessary) to have, a normal [

insertion and withdrawal speed as required. A summary of the results of this functional test is given.in Attachment 2.

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3.4 FULL cohE SHUTDOWN MAR 0!N DEMON 3TRATION 3.4.1 Purpose To demonstrate that the reactor can be made suberitical for any reactivity condition during cycle 10 operation with the analytically determined highest worth control rod, capable of withdrawal, fully withdrawn and all other rods fully inserted.

3.4.2 Acceptance criteria Level 1 criteria: The fully loaded core must be suberitical by at least 0.38% delta k with the analytically determined highest worth control rod, capable of withdrawal, fully withdrawn and all other rods fully inserted at the most reactive condition during the cycle.

Level 2 criterlat N/A 3.4.3 Test Description The full core shutdown margin demonstration was O performed analytically during the Hatch Unit 2 cycle 10 B00 in-sequence critical with the reactor 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 condition to cold conditions with the highest worth control rod fully withdrawn.

3.4.4 Test Results The full core shutdown margin demonstration was performed on 05/31/91 in accordance with core calculation procedure 42cC-ERP-010-OS, shutdown Margin Demonstration. Results of this calculation yielded a cold shutdown margin of 1.50% delta k.

The minimum cold shutdown margin was also 1.50%

delta k because cold _ shutdown margin this operating cycle is a minimum at Boc. A summary of the shutdown margin demonstration is given in Attachment 3 of this report.

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3.5 COLD CRITICAL E!OENVALUE COMPARISON 3.5.1 Purpose Tu compare the critical eigenvalue calculated using the actual cold, xenon-free critical control ,

rod configuration (corrected for moderator temperature and reactor period reactivity effects) to the cold critical eigenvalue assumed in the cycle management analyses.

3.5.2 Acceptance criteria j l

3.5.2.1 Level 1 criteria N/A t

3.5.2.2 Level 2 criteria N/A 3.5.3 Test Description The cold critical eigenvalue is the assumed value l of the PANACEA 3-D simulator model Keff at which criticality is achieved with the reactor in a xenon-free state and the coolant at 68 degrees F.  ;

O- This value is determined based on historical data and used for-cycle management analyses by the BWR l Core Analysis Group of Southern Nuclear Operating ,

company in Birmingham, Alabama. Once the actual j critical state is achieved during the beginning of ,

cycle startup, the opplicable data is sent to the -

i BWR Core Analysis aroup and the actual-(corrected for moderator temperature and reactor period i

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reactivity effects) cold critical eigenvalue is calculated. This value:is then compared to the 4 assumed critical eigenvalue as a method of validating rod worth and shutdown margin -

calculations throughout the cycle. The actual'  ;

. critical eigenvalue is also entered into a- i database for_ predicting future cold' critical 'i eigenvalues.

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3.5.4 Test Results for Hatch Unit 2 The cyclebeginning-of-cycle startup/91.

10 was performed on 5/31 The following reactor core conditions were observed when a critical state was achieved:

Sequence A2 RSCS Oroup 1 rully withdrawn RScs Group 2 13 control rods fully withdrawn and the 14th control rod (22-19) withdrawn to notch 18 Hoderator Temperature 175 degrees r Reactor Period 159.0 sec i control Rod Density 0.7783 l calculated McSDH 1.50% delta k i A cold critical eigenvalue of 1.0033 was calculated from the actual critical data given above. This compares well to an assumed value of 1.0030.

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3.6 WHOLE CORE LOCAL POWER RANGE MONITOR (LPRM) CALIBRATION 3.6.1 Purpose To determine (1) The LPRM calibration constants  !

such that when multiplied by the actual LPRM readings will produce calibrated LPRM reedings 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 power adjusted TIP signals at every  ;

node and LPRM detector location, respectively.

3.6.2 Acceptance Criteria 3.6.2.1 Level 1 criteria. N/A 316.2.2 Level 2 criteria: N/A 3.6.3 Testing Description The whole core LPRM calibration and BASE distribution calculation determines the LPRM calibration constants and the BASE and BASLP

( )- distributions used in axial power distribution calculations. The axial distribution of machine normalized full power adjusted TIP data is used to generate LPRM calibration constants required for TIP normalized LPRM readings.

In addition, machine normalized full power adjusted

-TIP readings are generated at every axial node (BASE distribution) ano at every LPRM detector location (BASLP distribution). These arrays are used as input data in the core calculation /

monitoring-programs to accurately calculate the power distribution at every node in the core.

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4 O 3.6.4 Test Results Whole core LPRH Calibration and BASE distribution was performed in accordance with Engineering procedure 42CC-ERP-015-08, OD1 and OD2 NUMAC TIP Operation at approximately 25%, 50%, 794 and 100%

power. LPRM calibration constants, BASE and BASLP arrays, were computed by the process computer and subsequently used successfully by the process computer to calculate the nodal power distribution and the core thermal limits.

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3.7 APRM CALIBRATION 3.7.1 Purpose To calibrate the APRM system to actual core thermal power, as determined by a heat balance.

3.7.2 Acceptance Criteria 3.7.2.1 Level 1 criteria: The APRM readings must be within a tolerance of 2t of core thermal powsr as determined from a heat balance.

3.7.2.2 Level 2 criteria: N/A 3.7.3 Test Description The APRM gains are adjusted after major power level changes, if required, to read the actual core thermal power as determined by a heat balance in accordance with procedure 34SV-SUV-021-05, APRM Adjustment to Core Thermal Power. The heat balance required for the calibration process may be obtained from the process computer program P1 O (Periodic Core Evaluation) or 003 (Core Thermal Power and APRM Calibration), or from a manual heat balance in accordance with operations procedure 34SV-SUV-025-OS, Core Heat Balance-Power Range.

3.7.4 Test Results APRM calibration was performed in accordance with plant precedure 34SV-SUV-021-OS, APRM Adjustment to Core Thermal Power at approximately 14%, 2 5 *. , an.

100% power. Each APRM was calibrated within a 2 ;.

tolerance to read core thermal power as calculated by the process computer.

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3.8 CONTROL ROD S03AH T!HE TESTIN0 i 3.8.1 Purpose "

I To demonstrate that the control rod drive system functions as designed with respect to scram insertion times following the completion of core alterations.

3.8.2 Acceptance Criteria  !

3.8.2.1 Level 1 criteria:

(a) The average scram insertion time for all  ;

operable control rods from the fully withdrawn position, based on de-energization of the scram  ;

pilot solenoids, with reactor steam done  ;

pressure above 950 psig shall not exceed the  ;

following:  !

Notch Position Average j from Fully Insertion Withdrawn Time (secs)  !

46 0.358  !

, -( ) 36 1.096 l

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26 1.860 $

l 06 3.419 f (b) The average scram insertion time, from ths  !

fully withdrawn position, for the 3 fastest ,

control rods in each group of four control rods arranged in a 2x2 array, based on the l de-energi:ation of the scram pilot solenoids,  !

shall not exceed the following: l

-Notch Position Average l from Fully Insertion  ;

Withdrawn Time (secs)  ;

46 0.379 36 1.162 l 26 1.972  !

06 3.624 .

(c) The maximum scram insertion time of each  :

control rod, from the fully withdrawn position {

l to position 06, based on the de-energi:ation of ,

l the scram pilot solenoid, shall not exceed 7.0 3 seconds. l

() Level 2 criteria: N/A j

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() 3.8.3 Test Description The control rod drive scram time testing was performed in accordance with engineering procedure 42SV-c11-001-25, control Rod Scram Testing, with the steam dome pressure above 950 psig. The test consists of scramming each control rod, collecting the resulting scram time data, and analyzing the data in accordance with the procedure to ensure compliance with the acceptance criteria noted above, 3.8.4 Test Results All control rod drives were tested in accordance with engineering procedure 42SV-c11-001-25, control Rod Scram Testing, with the steam dome pressure greater than 950 psig. A summary of the results is given in Attachment 4 of this report.

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I O 3.9 REACTIVITY ANOMALY CALCULATION 3.9.1 Purpose To check for possible reactivity anomalies as the core excess reactivity changes with exposare.

3.9.1 Acceptance criteria 3.9.1.1 Level 1 criterlat The corrected control rod density shall not differ from the predicted control rod density equivalent by more than plus or minus it delta k.

3.9.1.2 Level 2 criteria: N/A 3.9.2 Test Description During the BOC startup following a refueling outage and every month thereafter, a reactivity anomaly calculation is performed to monitor the core reactivity during the cycle. Since anticipated operation or unanticipated events may place the reactor in a condition other than that for which

_( ) the bascline anomaly curve was developed, the actual control rod density is corrected for off-rated conditions. The corrected control rod density is then compared to the reactivity anomaly curve provided in the Cycle Management Report to ensure that the corrected cotstrol rod density is within a plus or minus 1*. delta k acceptance band about the curve.

3.9.3 The reactivity anomaly calculation was performed in accordance with 42CC-ERP-007-OS, Reactivity Anomaly calculation, on 6/11/91. 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 report.

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e ATTACHMENT 1 CONTROL ROD FRICTION TESTING REACTOR PRESSURE: 0.0 PSIG CRD DlFFERENTIAL PRESSURE: 260.0 PSIG DRIVE LOCATION INSERTION'OlFFERENTIALPRESSURE SETTLE TEST MAX. I MlN. I DlFF. REQU1 BED 26 15 82 68 14 NO 18 19 83 68 15 NO 22 19 85 70 15 NO 22 31 81 68 13 NO 18 35 79 67 12 NO i 26 39 86 71 15 NO 18 39 94 70 24 YES 22 43 80 73 13 NO 10 15 78 65 13 NO 10 27 82 68 14 NO 14 27 82 69 13 NO 22 27 87 72 15 NO l 34 15 80 71 15 NO  !

30-19 87 72 15 NO  !

3 38 31 82 70 12 NO l (d 34 39 88 73 15 NO 42 47 100 77 29 YES 38 35 83 71 12 NO 50 35 84 70 14 NO

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ATTACHMENT 1

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U CONTROL ROD FRICTION TESTING NOTCH OPERATION SETTLE DIFFERENTIAL PRESS.. PSID NOTCHING NOTCHING DRIVE DRIVE IN _ DUI 1E-32 dE42 48-46 00-02 43 45 46 44 02 04 43 45 44 42 04-06 45 43 42 40 06 08 45 43 40-38 08 10 45 42 38-36 10-12 45 43 36-34 12 14 45 46 34 32 14 16 45 46 32 30 16 18 45 45 30-28 18 20 45 44 28 26 20-22 46 44 26 24 22 24 46 44 24 22 24 26 45 44 22 20 26-28 45 43 O '

20-18 28 30 45 45 44 43 18 16 30-32 16-14 32 34 45 44 14 12 34 36 45 41 12 10 36 38 45 42 10 08 38-40 45 41 08 06 40 42 46 40 06 04 42 44 46 40 04-02 44 46 40 44 02-00 46-48 46 44

ATTACHMENT 2 CONTROL ROD DRIVE TIMING

(

  • INDICATES AN ADJUSTED DRIVE)

DELVE LOCATION WITHDRAYLI1ME lHSERT TIME (SEC) (SEC) 22 19 45.77' 43.2 34 03 49.60' 45.0 30 03 48.48' 47.12*

22 31 42.7 43.9 26 03 41.0 51.8 22 03 39.5 55.3 18 35 42.5 41.5 18 03 43.2 39.0 42 07 53.1 39.4 38 07 42.5 50.51' 34 07 39.0 42.6 30-07 44.2 38.4 26 07 46.0 43.0 26 39 48.0 46.38' 22-07 53.2 48.4 18 19 45.62* 43.3 10-39 55.3 49.40' O 18-07 14-07 44.3 52.8 43 45-46.2 10 07 51 2 39.6 22 43 44.1 43.5 46 11 46.1 42.0 42-11 41.9 48.3 38 11 52.4 51.5 10 15 43.2 45.09' 34 11 48.4 41.5 30 11 49.9 43.3 26 11 49.9 41.9 10 27 48.7 50.00*

22 11 51.9 45.31*

10 11 42 4 50.4 14 27 u.9 38.7 14 11 47.1 39.5 10-11 45.2 51.4 06 11 40.6 49.0 22 27 53.9 45.33' 46 15 48.5 52.0 42 15 48.1 41.9 38 15 48,1 40.7 30 15 44 34' 56.7 26 15 39.0 41.5 22 15 42.8 51.6 O 1815 14 15 45.9 43.5 42.3 40.3

ATTACHMENT 2 j O Con 1not nod Dnive riui,m

(

  • INDICATES AN ADJUSTED DRIVE)

QBLVE LOCATION WlIliDBAYLTAiE luSERT TIME (SEC) (SEC) 06-15 55.5 41.0 50 19 49.9 52.1 46-19 45.04' 47.2 42-19 39.2 43.3 38 19 46.1 45.4 34 19 45.8 44.5 26 19 48 8 41.1 14 19 40.0 42.3 10-19 42.1 39.5 06 19 43.7 40.4 34 15 47.1 46.94*

02 19 3P.2 51.0 50-23 49.2 30.6 30-19 47.8 48.51' 46-23 d5.22' 54.6 38 31 45.9 49.35' 42 23 38.8 53.3 0 38 23 34-39 39 s 43.2 53 5 48.1 34 23 45.1 44.0 30-23 46.3 40.4 26 23 44.13* 43.1 42-47 51.9 45.77' 22 23 49.5 43.4 18 23 40.0 46.8 14 23 48.4 41.5 10 23 39.2 46.3 06 23 49.3 49.6 02 23 45.25' 44.89' 38-35 44.4 39.0 50-27 46.1 48.42' 46 27 50.4 39.3 50-35 48.9 45.53' 42 27 47.2 44.3 38 27 39.2 45.2 34 27 49.4 42.9 30-27 41.1 51.1 26-27 46.0 49.5 18-27 49.6 41.6 06-27 44.3 45.89*

02-27 41.9 53.2 50-31 42.46 44.02 0 46 31 42 31 38.88 47.36 40.95 45.49

ATTACHMENT 2 O

V CONTROL ROD DAIVE TIMING

(

  • INDICATES AN ADJUSTED DRIVE)

DRIVE LOCAllOR y/1111 DRAW TIME lt1 SERT TIME (SEC) (SEC) 34 31 54.52 47.87 30 31 48.5 47.28 26 31 46.86 39.98 18 31 47.75 40.89 14 31 40.79 47.85' 10 31 44.82 45.07' 06 31 42.30 47.31 02 31 48.35 46.26 46-35 41.36 56.14 42 35 46.27 46.66*

34 35 40.09 40.81 30-35 47.22 52.20 26 35 40.09 42.37 22 35 38.77 41.54 14 35 47.39 47.35 10 35 46.52 49.13 06 35 47.73 45.82 O o2 35 46-39 47 a 45.54 49.54-48.04 42 39 44.26 44.68 38 39 50.00 45.69 30-39 50.17 50.40 22 39 45.34 42.57 14 39 52.92 44.32 10 39 50.51 43.14 06 39 48.59 44.16 46-43 56.99 47.15 42-43 52.69 43.92 38-43 44.68 47.74 34 43 50.08 41.05 30-43 42.08 52.78 26-43 55.82 41.27 16-43 45.37 42.91 14-43 43.77 40.25 10-43 44.11 52.65 0643 42.78 43.83 38 47 41.04 39.92 34 47 44.47 39.05 30-47 42.84 45.16 26-47 47.66 40.42 22-47 51.28 49.73*

p 18 47 43.22 44.15 V 14 47 44.02 47.48 43.68 10-47 41.64

ATTACHMENT 2 O CONTROL ROD DRIVE TIMING

(' INDICATES AN ADJUSTED DRIVE)

DEVE LOCATIDlj y/ITHDRAYLIlfdE LtGERI_IJ1dE (SEC) (SEC) 34 51 39.90 43.02 30 51 43.12 48.34 26 51 48.64 47,03 22 51 44.32 44.55 18 51 43.01 46,74 O

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9 ATTACHMENT 3 O

FULL CORE SHUTDOWN MARGIN DEMONSTRATION 1 K- 0.9851 2 Kr , 1.0027 3 CONTROL ROD DENSi1Y 0.7783 4 REACTOR COOLANT TEMPERATURE 175' F 5 REACTIVITY CORRECTION FOR TEMPERATURE 0.0022/1 k 0 REACTOR PERIOD 159 sec 7 REACTIVITY CORRECTION FOR PERIOD 0.0004 o k 8 COLD SHUTDOWN MARGIN 1.5% o k 9 VALUE OF R 0.0% o k 10 MINIMUM COLD SHUTDOWN MARGIN 1.5% 4 k 11 TECH SPEC REQUIRED SHUTDOWN MARGIN 0.38% 6 k O

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l ATTACHMENT 4 (G

s' SCRAM TIME TESTING I NOTCH SLOWEST SCRAM AVERAGE S RAM AVERAGE SCRAM POSITION INSERTION TIME INSERTION '"'JE INSERTION TIME (SEC) (SEC) CRITERIA (SEC) ,

SINGLE 2X2 ARRAY l SINGLE 2X2 ARRAY ' SINGLE l 2X2 ARRAY ,

ROD 3 EASTEST ! ROD 3 FASTEST ! ROD __ ! 3 FASTEST l

46 0.343 0.304 0.282 0.277 0.358 0.379 36 0.880 0.840 0.791 0.785 1.096 1.162 26 1.512 1.410 1.315 1.304 1.860 1.972 00 2.832 2.611 2.426 2.403 3.419 3.624 i

,I

.v i

O l

ATTACitMENT 5 REACTIVITY ANOMALY CALCULATION Q

UNIT 2 CYCLE 10 SEQUENCE: A2 DATE PERFORMED: C/11/91 THERMAL POWER CMWT 2434.0 CORE FLOW (Mlb/hr) WT 75.15 RATED CORE FLOW (Mib'hr) 77.0 DOME PRESSURE (psia) PR 1005.0 SUBCOOLING (BTU /hr) DHS 19.3G DESIGN INLET SUBCOOUNG(BTU /hr) 18.7 CONTROL ROD DENSITY CRD 0.084 CORRECTED CRD. CRD + CORRECTION CORRECTION: 3.2070E 1 x (1 CMWT/2430)

+2.1326E 1 x (1 WT/ RATED CORE FLOW) 42.9931E 3 x (DESIGN INLET SUBCOOUNG DHS)

+7.2254E 5 x (1020-PR) = 0.004 CORRECTED CRD- 0.084 + 0.004 - 0.088 PREDICTED CRD= 0.085

+ 1% VALUE= 0.135 1% VALUE= 0.035 O

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