ML20101Q300

Cycle 5 Startup Test Rept Summary

ML20101Q300
Person / Time
Site:	LaSalle
Issue date:	07/08/1992
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20101Q298	List: ML20101Q296 ML20101Q300
References
NUDOCS 9207140228
Download: ML20101Q300 (18)

Text

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l c l ATlACBMENIA i

Summary of Unit E Cycle 5 Startup Test Program .

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l SUMMAI1Y LaSalle Unit 2 Cycle 5 began commercial operation on April 12,1992 followino a refueling and maintenance outage. The Unit 2 Cycle 5 core loading consisted of 224 fresh fuel bundles (192 GE9B-P8CWB302-9GZ-100M 150 T and 32 GE98 P8CWB300 9GZ-100M 150 T) and 540 reload bundles. The same bundle l

design being loaded for Cycle 5 was previously loaded for Unit 2 Cycle 4 operation.

Unit 2 Cycle 5 had 21 LPRM strings replaced with General Electric NA 300 LPRM strings. No control blades were replaced for Unit 2 Cycle 5, however,34 General Electric control blades were shuffled to optimize control blade lifetime. All applicable test results (neutron instrument calibration, computer monitoring results) indicate expected core perictmance with the new fuel design.

A comprehensive startup testing program was performed during startup and power ascension. The startup program included:

local and in sequence shutdown margin tests.

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- reactivity anomaly calculations at initial critical and fell power.

- nuclear lnstrument performance verifications (SRM, IR VI, APRM response and overlap checks).

- instrument calibrations (LPRM, APRM, TIPS, core flow).

- control rod drive friction and full core scram timing.

- LPRM responses to control rod movement.

- process computer verification, comparison to off line calculation.

recirculation system performanco data.

- baseline stabifity data acquisillon.

The startup test program was satisfactorily completed on June 5,1992. All test data was reviewed in accordance with the applicable test procedures, anrj exceptions to any results were evaluated to verify compliance with Technical Specification limits to ensure the acceptability of subsequent test results.

A ctartup test report must be submitted to the Nuclear Regulatory Commission (NRC) within 90 days following resumption of commercial power operation (in accordance with Technical Specification 6.6.A.1). The startup test report presented in this report (Attachment B) contains results (evaluations) from the following tests:

Core Verification

- Single Rod Suberitical Chock

- Control Rod Friction and Settle Testing Control Rod Drive Timing

- Shutdown Margin Subcritical Demonstration

- Shutdown Margin Test (In-sequence critical)

- Reactivity Anomaly Calculation (Critical and Full Power)

- Scram insertion Times

- Core Power Distribution Symmetry Analysis A full evaluation of the startup test program is included with the evaluation of LTP-1600 37 (On-Site Review 92-22), Unit Startup Test Program. Data from each startup is available at LaSaite Station.

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i AT.IAG1MEhlT_B LaSalle County Nuclear Power Station Unit 2 Cycle 5 Startup Test Report l

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ZNLD/1091/5:

j i LTP-1700-1, CORE VERIFICATK)N 1

PURPOSE I

The purpose of this test is to visually verify that the core is ,

loaded as intended for Unit 2 Cycle 5 operation.  !

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I CRITERIA The as-loaded core must conform to the cycle core design used by the Core Management Organization (Nuclear Fuel Services) in the reload licensing analysis. The core verification must be observed

! by a member of the Commonwealth Edison Company Nuclear Fuel .

Services staff. Any discrepancles discovered in the loading will i os promptly corret'ied and the affected areas reverified to ensure -

proper core loading prior to unit startup.

Conformance to the cycle core design will be documented by a

permanent core serial number map signed by the audit participants.

RESULTS AND DISCUSSION The Unit 2 Cycle 5 core verification consisted of a core height -

check performed by the fuel handlers and two videotaped passes of the cure by the nuclear groua. The height check verifles the proper seating of the assembly in the fuel support piece while the q videotaped scans verify proper assembly orientation,-location, and:

seating. Bundle serial numbers and orientations were recorded during the videotaped scans, for comparison to the appropriate tag ,

, boards and Cycle Management documentation. On March 13 1992 the

,' core was verified as being properly loaded and consistent with e Commonwealth Edison Nuclear Fuel Services Cycle 5 Cycle Management l Report and the Final Station use Loading Plan.10n March 15,1992 a -

partial inventory was performed on four fuel bundles that were re-: +

channeled when friction testing (LTP 700-2)'showed excessive friction between control rod 30-03 and the four surrounding

, bundles. On Msrch 15,1992, the videotapes were reviewed by the-

! Lead Nuclear Engineer to reverify all bundle ID's, orientation, and seating. -

A serial number inventory was also performed on the Unit 1 and Unit' 2 fuel pools on March 16,1992 and concluded on March 23,1992 to verify that the fuel pool contained the proper bundles. The fuel ro^ contahad no bundles which should have been loaded into the W . ' reacfar. >

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LTP-1000-30, Single Hod Subcritical Cixx:k i PURPOSE The puraose of this test is to demonstrate that the Unit 2 Cycle 5 core wil remain subcritical u3on the withdrawal of the analytically determined stroiegest contro rod.

d CRITERIA The core must remain suberitical, with no significant increase in SRM readings, with the analytically determined strongest rod fully withdrawn.

4 RESULTS AND DISCUSSION The analytically determined strongest rod for the Beginning of Cycle 5

, of Unit 2 was determined by Nuclear Fuel Services to be rod 22-31. On 4

March 13,1992, with a Unit 2 moderator temperature of 75.87 degrees

Fahrenheit (as read from computer point B741, cleanup system inlet temperature), rod 22 31 was single notch withdrawn to the full out position (48) and the core remained subcritical with no significant increase in SRM readings. The satisf actory completion of LTP-1600-30, Single Rod Subcritical Check, allows single control rod withdrawals for control rod testing provided moderator tem aerature is greater than or equal to 7L.87 degrees Fahrenheit. This in'ormation is documented on LTP-1600-30, Attachment 8, Unit Instructions for Single Control Rod i Movement, of which a copy was given to the Unit 2 NSO and the Shift Engineer.

Subsequent to the performance of the Single Rod Subcritical Check all control rods were withdrawn individually to the full out position and the core remained subcritical with no significant increase in SRM readings at any time.

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ZNLD/1932/2

L1P-700 2, CONTROL ROD FRICTION AND SETTLE TESTING PURPOSE 1

The purpose of this test is to demonstrate that excessive friction

, does not oat between the control rod blade and the fuel assemblies during operation of the control rod drive (CRD) following core alteratiens.

CRITERIA With the final cellloading complete for the fuel assemblies in a control cell, the differential pressure across the CRD drive piston should not vary by more than 15 psid during a continuous insertion.

9 If the drive piston differential pressure during a continuous insert varies by more than 15 psid, an individual notch (insert) settling pressure test must be performed on the CRD. The differential settling pressure for an individual notch test should not be less 1 than 30 psid, nor should it vary by more than 10 psid over a full stroke.

. RESULTS AND DISCUSSION Control Rod Drive (CRD) Friction testing commenced after the completion of the core load verification and single rod suberitical check, and was completed on March 16, t 992. Continuous insert friction traces were obtained for all 185 CRDs. Control rod 30 03 exhibited high friction during the test. The surrounding four bundles were rechanneled and the Control rod was tested satisfactorily.

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LOS-RD SRS, CONTROL ROD DRIVE TIMING PURPOSE The purpose of this test is to check and set the insert and withdrawal times of the Control Rod Drives (CRDs), in addliion, this surveillance will provide verification that each control rod blado is coupled to it's respective CRD mechanism.

CRITERIA The insert and withdrawal times of a CRD should be 48 +/ 9.6 seconds (between 38.40 and 57.60 seconds). However, General Electric recommended that LaSalle change this criteria to 40 to 56 seconds for insert times and 46 to 58 seconds for withdrawal times in the cold shutdown conditions (depressurized). This change might avoid adjustments of the CRD velocity during rated reactor operation.

RESULTS AND DISCUSSION All CRDs were tested between 03 25 92 and 04 0192. All control rod drives demonstrated normal times during the performane e of this test.

A coupling check was also successfully performed on ecch drive during the timing process.

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ZNLD/1932/4

l l LTS-1100-14, SHUTDOWN MARGIN (SDM) SUBCRITICAL DEMOi4STRATION d

i i PURPOSE The purpose of this test is to demonstrate, using the adjacent rod suberitical method, that the core loading has been limited such that

] the reactor will oa subcritical thropghout the operating cycle with >

4 the strongost control rod in the full out position (position 48) and 4

all other rods fully inseriod.

CRITERIA j l If a SDM of 0.38% A K/K (0.38% A K/K + R) cannot be demonstrated with j the strongest control too Nily withdrawn, the core loading must be altered to meet this marg i. R is the reactivity difference between t the core's beginning of c ;le SDM and the minimum SDM for the cycle, j The R value for Cycle 5 is 0.0 % A K/K, with the minimum SDM

! occurring at 0.0 MWD /ST:nto the cycle.

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RESULTS AND DISCUS'910N On April 2,1992, the local SDM demonstration was successfully performed using control rods 22 31 and 26 35. Control rod 26 35 is of. -

diagonally cycle. Nuclear adjacent to 22-31, Fuel Services provided, theinstrongest rod at beginning Pacl<aDe, the Cycle Startup rod worth information (for control rods 22 31 and diagonally adjacent

! . rod 26 35) and moderator temperature reactivity corrections to support this test. Using the supplied information, it was determined i

that with control rod 22 31 at position 48 and rod 26 35 at position j 16, a modarator temperature of 160.0 degrees F, and the reactor

subcritical, a SDM of 0.617% A K/K was demonstrated. The SDM

. demonstrated exceeded the 0.38% A K/K required to satisfy Technical

!- Specification 3.1.1, and maintained sufficient margin to the l calculated SDM for the core at beginning-of-cycle (2.082% A K/K) to -

i avoid criticality during the test.

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LTS-1100-1, SillfiDOWN MARGIN TEST PURPOSE The purpose of this test is to demonstrate, from a normalin-sequence critical, that the ento loading has boon limited such that the reactor will be subcritical throughout the operating cycle with i the strongest control rod in the full-out position (position 48) and all other rods fully inserted.

j CRITERIA 1

If a shutdown mar 0i n (SDM) of .38% A K/K (0.38% A K/K + R) cannot bo demonstrated with the strongest control rod fully withdrawn, the core loading must be altered to meet this margin. R is the reactivity difference betwoon the coro's beginning-of cycle SDM and the minimum SDM for the cycle. The R value for Cycle 5 is 0.0% A K/K, so a SDM of 0.38% A K/K must bo demonstrated, i

4 i RESULTS AND DISCUSSION The beginning of cycle SDM was successfully determined from the i initial critical data. The initial Cycle 5 critical occurred on April 2.1992 on control rod 22 43 at position 08, using an A 2 4 sequence. The inoderator temperature was 165 degrees F and the reactor period was 72 seconds. Using rod worth information, moderator temperature reactivity corrections, and parlod reactivity corrections supplied by Nuclear Fuoi Services (In the Cycle Start Package), the beginning-of cycle SDM was deteimined to be 2.bs A K/K (see Tablo 1). The SDM demonstrated exceeded the 0.38%

A K/K required to satisfy Technical Specification 3.1.1.

The calculction was also performed for the April 8,1992 critical.

The reactor went critical on control rod 14 27 at 8, a moderator temperature of 165 degrees F and a reactor period of 184 seconds.

The Shutdown margin was calculated to be 2.682% A K/K.

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TABLE 1 SilOTDOWN MARGIN CALCUt/sTION The Foilowing data is from the April 2,1992 critical.

Critical Rod - 22 43 @ 08 Worth of Strongest Rod - 0.02004 A K/K (1)

Worth of Control Rods Withdrawn to Obtain Criticality:

24 Group i rods at 48 = 0.03550 A K/K '

24 Group 2 rods at 48 - 0.01747 A K/K 6 Group 3 rods at 04 - 0.0005 A K/K 18 Group 3 rods at 08 - 0.00175 A K/K Temperature Correction -0.00165 A K/K (5) for Tm = 165'F Period Correction - 0.00075 A K/K (0) for Period = 72 seconds Shutdown Marain Koff:

SDM Keff - 1.0000 + 1) - (2) - (3) - (4) - (5) + (6)

= 0.97316 A I K SDM - (1.000 - (SDM Keff))

• 100 - 2.084% A K/K k

ZNLD/1932/7

LTS1100-2,' CHECKING FOR REACTIVITY ANOMALIES L PURPOSE 0 The purpose of this test is to compare the actual and predicted critical rod configurations to detect any unexpected reactivity - 3 trends, s 5

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CRITERIA a-

! In accordance with Technical Specification 3.1.2, the reactivity equivalence of the difference between the actusi control rod i density and the predicted control rod density sha!! not exceed 1%

l -- 4 K/K. If the difference does ox:eed 1% A K/K, the Cars '

! Management Engineers (Nuclear Fuel Services) will be promptly a notified in hvestigate the anomaly. The cause of the anomaly must i be deterrnhsd, eglained, and corrected for continued operation of

the unit. 1 RESULTS AND DISCUSSION 3 Three reactivity anomaly calculations were successfully performed during the Unit 2 Cycle 5 Startup Test Program, two from in-sequence criticals and a third from' steady-state, equilibrium conditions et approximately 100 percent of full power.

e The initial critical occurred on April 2,1992, on control rod 22-- t 4 43 at position 08, using an A-2 sequence. The moderator _

temperature was 165 degrees F and the reactor period was 72 seconds. Using rod worth information, moderator temperature -

reactivity corrections, and period reactivity corrections supp!ied by Nuclear Fuel Services (in the Cycle Startup Package), the actual .

critical was determined to be within 0.602% A K/K of the -

predicted critical (see Table 2). The anomaly determined is within -

the 1% A K/K allowed by. Technical Specification 3.1.2; The calculation was also performed for the April 8,1992 critical.-

The reactor went critical on cor*ot rod 14 27 at 8, a moderator .

temperature of 165 degrees I nd a ^ eactor period of 184 seconds.

The calculated Reactivity'Anouv,y :as 0.600% A K/K.

The third reactivity anonialy calculation, for power operation, was =

performed using' data from May 6,1992 st 100% power at a a :le: _

exposure of 308.1 MWD /ST, at equilibrium conditions. The predicted notch inventory supplied by Nuclear Fuel Services was 149 notches.

-The actus! corrected notch inventory was 105.4 notches. Using the l notch worth provided by Nuclear Fuel Services, the resulting .  ;

anomaly was 0.148% A K/K. This value is dthin the' 1% i

'A K/K criteria of Technical SpecUlcation 3, i y

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I INITIAL CRITICA~LITY COMPARISON CALCULATIONS ~-

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__ ITEM ' -- -. - - . .A K/K -

i. Keff with all rods in at 68 degrees F = 10,95314 *-

[ - Reactivity inserted b 24 group 1 rods at position 48 =

0,03556 * ,

Reactivity inserted b 24 group 2 rods at position 48 - - :0.01747

• i Reactivily inserted b . 6 group 3 rod at position 041 - 0.0005 .

[ Reactivity inserted b 18 group 3 rods at position 08 = 0,00175' 4 Predicted Keff at actual critical rod pattern (68'F) - -- '1.00842-- ,

Reactivity associated with_the measured reactor' a

[- -- 0.00075 *-

period (period ctrrection for ~73 second period) i Reactivity associated with moderator temperature ? - --

4 (165 F actual,68'F predicted) - -0.00165 * -

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Reactivity Anomaly = [(predicted neff - 1) - (period . .

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!- correction) + (temperature correction)]

• 100%

40.602% A K/K ,

- LaSalle Unit 2 Cycle 5 Startup Package", supplied by Nuclear ; '

! Fuel Services.

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- LT61100% SCRAM' INSERTION TIMES p

PURPOSE i- The purpose of this test is to demonstrate that the control rod

scram insertion times are within the operating limits set forth by -

l- the Technical Specifications (3.1.3.2,3.1.3.3, 3.1.3.4).

p CRITERIA I The maximum scram insert!on time of each control rod from the fully -

withdrawn position (48) to notch poshion 05, based on de-energization of the scram pilot valve solenolds as time zero, shall- .

not exceed 7.0 seconds.

i The average scram 'nsertion time of all operable control rods from-

the fully wnhdrawn position (48), based on de-energization of the!

scram pilot valve solenoids as time zero, shall not exceed any of l the following:

,- Position Inserted From Average Scram Insortion j

[ FullyMillidamn._._. _T_imelSeco.nds) '

45 0.43

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0.86 =

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25  : 1.93 :

05 3.49' _

The average scram insertion time, from the fully withdrawn position -

. (48), for the three fastest control rods in each group of four J

two array, based on de-control rodsofarranged onergization the scraminpilot a two va bylve solenoids as time zero, shall .

2-not exceed any of the following:

i Position Inserted From Average Scram insertion- .

_EullyEllbdrawn Time _(Secords) 45- 10.45:

i 39 0.92:

l- 25 2.05,

. 05' 3.70-

_RESULTS AND DISCUSSION L- -

Scram testing was successfully perfcrmed April 14-15,1992. All: -

i control rods were scram timed from full out.-- All control rod scram:

i timing acceptance criteria were met during this test. Control rod 06-27. had its pilot valves replaced and tested satisfactorily.

Control rod 38-11 had a work request written for preentive -

maintenance and tested satisfactorily on April 21,1992. -

The results of the testing are given below.

Maximum Average -

Average Scram Times . Scram Times in a fosition oLalLORDL(sent) Iwo-by IwolnaylsesA) 0.324- 0.341-39 0.618- 0.6371 25 > 1.327 - 1.400-

05' 2.404 2.5?.0 -

Tau Ave (position 39) for Minimum Critical Power Ratio Lirnit~

. determination: 0.618 seconds.

.ZNLD/1932/10 '

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LTP-1600-17, CORE POWER DISTRIBUTION SYMMETRY ANALYSIS PURPOSE

! The purpose of this test is to verify the core power symmetry and the reproducibility of the TIP readings.

CRITERIA The total TIP uncertainty obtained by averaging the uncertainties for all data sets must be less than 8.7%

The gross check of the TIP signal symmetry should yield a maximum _

. deviation between symmetrically _ located pairs of less than 25%/

i RESULTS AND DISCUSSION:

i. Core power symmetry calculations were performed based upon data -

obtained from two full core TIP sets (OD-1)c The first TIP set was - '

! performed on May 4,1992 at approximately 100% power and the second on May 5,1992 at approximately 100% power.- The TIP uncertainty -

from the first data set at approxirnately 100% power was 3.774% with an average standard deviation of 5.338%. The TIP uncertainty from the second data set was 3.105% with an average standard deviation-4 of 4.391%. Both data sets exhibitedTIP uncertainties within the -

8.7% acceptance criteria.-

Table 3' lists the symm +trical TIP pairs, their core locations, and their respective average deviations. The maximum deviation between symmetrical TIP pairs was 14.32% forTIP pair 05 34, satisfying the criteria of the test (less than 25%).

I Table 4 lists the data calculated to determine the Random Noise Uncertainty and Geometric Noise. The Random Noise Uncertainty was determinad to be 0.867% and the Geometric Noise was determined to e be 3.328%.

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Additionally,~two full core TIP sets were performed at .- . .

-approximately 100% power on June 12,1992. Although they were not.

officially par

• of the Unit 2 Cycle 5 Startu a Test Program, core power symmetry calculations were rever fled from the results of~

these TIP sets. ihe calculations yielded a TIP uncertainty of; 3.207% with an average standard deviation of_4.535% for the first '

TIP set and a TIP uncertainty of 3.153% with an average standard -

deviation of 4.459% for the second TIP set.

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3 A discussion of the calculational methodology is provided below.

The method used to obtain the uncertainties consisted of calculating the average of the nodal BASE ratio of TIP pairs by:

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N,i hS ps l where Rij = the BASE ratio for the ith node of TIP pair j, n a number of TIP pairs = 19.

! Next, the standard deviation (expressed as a percentage) of these ratios is calculated by the following equation:

1x ra _%

c, on = % . , >, . , ( Rg - R)'

4 too j _ (IS 4 - O ,

!' The total TIP uncertainty (%) is calculated by dividingcA (%) by f2 because the uncertainty in one TIP reading is the desired

parameter, but the measured uncertainty is the ratio of two TIP :

readings.

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r TABLE 3 TIP SIGNAL SYMMETRY RESULTS All numbers shown are averages from two OD-1 data sats from 5-4-92 and 5-5-92 at 99.7% and 99.5% power, respectivel .

Symmetrical TIP Pair Absolute Percent thambME.(GQIe.LocatioD) Difference TIP Pair

__a_ b aLBASE# DAylation' 1 16-09) 6 (08-17) 3.02 4.09 2 24-09) 13 08-25 0.67 0.63 .

3 32-09 20 08 33 0.82 0.29 4 4009 27 08-41 2.36 3.47 5 48-09 34 08-49 5.08 14.32 8 24-17 14 16-25 0.10 0.27 9 32-1 21 16-33 1.73 1.61 10 40-1 28 16-41 1,63 1.26 11 48-17 35 16 49 4.38 4.18 12 56-17 40 16-57 2.50 1.95 16 32-25 22 24-33 2.08 1.82 17 40-25 29 24-41 5.81 5.93 18 48-25 36 24-49 1.36 1.29 19 56-25 41 24-57 0.42 4.09 24 40-33 30 32-41 0.41 0.97 25 48-33 37 32-49 2.22 1.84 26 53-33 42 32-57 4.05 7.29 32 48-41 38 40-49 0.15 0.10 33 56 41 43 40-57 2.62 5.10

1. where: Absolute Difference of BASE - BASEa- BASEb and BASE = fg[BASEq(K)

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BASEa - BASEb l* 100

- where: % Deviation -

0.5(BASEa + BASEb).

l ZNLD/1932/13

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! RANDOM NOISE UNCERTAINTY AND GEOMETRIC NOISE DATA

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[- Por LTP-160017, Attachment D, at approximately 99.9% thermal power, i the Random Noise Uncer1ainty and Geometric Noise Data- Analysis was - >

aerformed. The results of the calculations are presented below. The Random ^

Noise was determined to be 0.867%.

The Geometric Noise was determined to '

l be 3.328%.

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! 5- 124.71-i 6 121.26 l 7 116.36 i 8 -116.38:-

! 9 114.36 l 10 112.655 i 11 .111.93 l 12 112.87. '

13 111.77-i 14 107.40'

, 15 109.20-- ,

' 16 111.29 17 106.65:  !

i 18 -105.561 l 19 103.01-20 -93.80' l4 i- 21 82.31 4 i'i 22 72,16-l' s

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l 4 i AHACBMEliTS s

4 List of Rt.'erences s

1. NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel", Latest Approved Revision.

2. LaSalle 2 Cycle 5 Cycle Management Report, DRF Number LS2-0006 Volume 2, 4

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