ML20115D878

Cycle 8 Startup Test Rept

ML20115D878
Person / Time
Site:	LaSalle
Issue date:	07/08/1996
From:	Ray D COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9607160005
Download: ML20115D878 (14)

Text

Commonwealth 1:diwn Corapany 12thalle Generating Station 2S)I North list Road Marseilles 11.61.411-9757

• Tel Hi sd57W"61 July 8,1996 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 l 1

Subject:

LaSalle County Nuclear Power Station Unit 1 Unit 1 Cycle 8 Startup Test Report NRC Docket No. 50-373  :

The Attachment to this letter presents the LaSalle Unit 1 Cycle 8 Startup Test Report. This report is being submitted in accordance with Technical Specification 6.6.A.1. Additional startup test results are available at i LaSalle Station.

If there are any questions or comments concerning this letter, please refer them to me at (815) 357-6761, extension 2212.

Respectfully,

%N D.J. Ray Station Manager LaSalle County Station Enclosure cc: H. J. Miller, NRC Region Ill Administrator i

M. P. Huber, NRC Senior Resident inspector - LaSalle  ;

D. M. Skay, Project Manager - NRR - LaSalle l F. Niziolek, Office of Nuclear Facility Safety - IDNS Central File  !

9607160005 960708 /

PDR ADOCK 05000373 P PDR A linicom company

LaSalle Unit 1 Cycle 8 Startup Test Report )

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SUMMARY

LaSalle Unit 1 Cycle 8 began commercial operation on April 12,1996 following l a refueling and maintenance outage. The Unit 1 Cycle 8 core loading consisted of 248 l fresh fuel bundles (104 bundles of fuel type GE98-P8CWB343-12GZ-80M-150-CECO and I 144 bundles of fuel type GE98-P8CWB34210GZ-80M 150-CECO), and 516 reload bundles. The Cycle 8 reload bundle is the same bundle design that was previously loaded in Unit 1 Cycle 7. However, the reload bendles utilize 80 mil channels, in lieu of the 100 mil channels used previously. This channe! design _ change was approved per 10CFR50.59 prior to startup, in addition,6 LPRM strings were replaced with General Electric NA-300 LPRM strings. Twenty control blades were replaced for Unit 1 Cycle 8, and 15 additional control blades were shutiled. All applicable test results (neutron instrument calibration, computer monitoring results) indicate expected core performance with the new core design.

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

-in-sequence shutdown margin test.

- reactivity anomaly calculations at initial critical and full power.

- nuclear instrument performance verifications (SRM, IRM, 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.

- baseline stability data acquisition.

The startup test program was satisfactorily completed on May 28,1996. All test data was reviewed in accordance with the applicable test procedures, and exceptions to any results were evaluated to verify compliance with Technical Specification limits to ensure the acceptability of subsequent test results.

A startup 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 review contains results (evaluations) from the following tests:

- Core Verification

- Single Rod Subcritical Check

- Control Rod Friction and Settle Testing

- Control Rod Drive Timing

- Shutdown Margin Test (In-sequence critical) l l

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- Reactivity Anomaly Calculation (Critical and Full Power)

- Scram insertion Times

- Core Power Distribution Symmetry Analysis FINDINGS AND RECOMMENDATIONS Based upon the preceding discussion and the review of the startup test report, the "LaSalle County Nuclear Power Station Unit 1 Cycle 8 Startup Test Report" is submitted to the NRC in accordance with Technical Specification 6.6.A.1.

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LTP-1700-1, CORE VERIFICATION PURPOSE The purpose of this test is to visually verify that the core is loaded as intended for Unit 1 Cycle 8 operation.

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 discrepancies discovered in the loading will be promptly corrected 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 1 Cycle 8 core verification consisted of a core height check performed by the fuel handlers and a videotaped pass of the core by the nuclear group. The height check verifies the proper seating of the assembly in the fuel support piece while the videotaped scan verifies proper assembly orientation, location, and seating. Bundle serial numbers and orientations were recorded during the videotaped scan, for comparison to the appropriate core loading map and Cycle Management documentation. On March 23,1996, the core was verified as being properly loaded and consistent with Commonwealth Edison Nuclear Fuel Services LaSalle 1 Cycle 8 Design Basis Loading Plan.

A serial number inventory was also performed prior to core load on the Unit 1 Discharge Queue on February 23,1996 to verify that the discharge queue contained the proper bundles. The discharge queue contained no bundles which should have been loaded into the Unit 1 reactor.

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4 LTP-1600-30, Single Rod Subcritical Check PURPOSE The purpose of this test is to demonstrate that the Unit 1 Cycle 8 core will remain subcritical upon the withdrawal of the analytically determined strongest control rod.

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

RESULTS AND DISCUSSION The analytically determined strongest rod for the Beginning of Unit 1 cycle 8 was

determined by Nuclear Fuel Services to be rod 38-35. On March 24,1996, with a Unit 1 moderator temperature of 78.5 degrees Fahrenheit, rod 38-35 was single notch withdrawn to the full out position (48) and the core remained subcritical with no

significant increase in SRM readings. The satisfactory completion of LTP-1600-30, j Single Rod Subcritical Check, allows single ~ control rod withdrawals for control rod testing provided moderator temperature is greater than or equal to 78.5 degrees Fahrenheit. This information is documented on LTP-1600-30, Attachment A.

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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LTP-700-2, CONTROL ROD FRICTION AND SETTLE TESTING 1

PURPOSE  !

The purpose of this test is to demonstrate that excessive friction does not exist between the control rod blade and the fuel assemblies during operation of the control rod drive I (CRD) following core alterations, j I

CRITERIA With the final cell loading 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.

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 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 q load verification and single rod subcritical check, and was completed on April 15,1996. l Continuous insert friction traces were obtained for all 185 CRDs. All rods tested Satisfactorily.

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l LOS-RD-SRS, CONTROL ROD DRIVE TIMING  !

PURPOSE

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The purpose of this test is to check and set the insert and withdrawal times of the  !

Control Rod Drives (CRDs). In addition, this surveillance will provide verification that each control rod blade is coupled to it's respective CRD mechanism.

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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) to give indication of seal wear. This change might avoid adjustments of the CRD velocity during rated- i reactor operation.

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RESULTS AND DISCUSSION All CRDs testing was complete on March 26,1996. Control rods 38-19,46-27,46-55 and 02-19 had withdrawal times faster than 46 seconds (but greater than 38.4 seconds) ,

due to degraded drive seals. The above listed control rods directional control valves )

were tested and found to be operating properly. These control rods are being evaluated for replaced during the next refueling outage.

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LTS-1100-1, SHUTDOWN MARGIN TEST PURPOSE The purpose of this test is to demonstrate, from a normal in-sequence critical, that the core loading has been limited such that the reactor will be subcritical throughout the operating cycle with the strongest control rod in the full-out position (position 48) and ,

all other rods fully inserted.

CRITERIA If a shutdown margin (SDM) of .38% delta K/K (0.38% delta K/K + R) cannot be demonstrated with the strongest control rod fully withdrawn, the core loading must be altered to meet this margin. R is the reactivity difference between the core's beginning-of-cycle SDM and the minimum SDM for the cycle. The R value for Cycle 8 is 0.0% delta K/K, so a SDM of 0.38% delta K/K must be demonstrated.

RESULTS AND DISCUSSION ,

The beginning-of-cycle SDM was successfully determined from the initial critical data.

The initial Cycle 8 critical occurred on April 12,1996 on control rod 42-31 at position ,

18, using an A-2 sequence. The moderator temperature was 164 degrees F and the reactor period was 242 seconds. Using rod worth information, moderator temperature j reactivity corrections, and period reactivity corrections supplied by Nuclear Fuel j Services (in the Cycle Startup Package), the beginning-of-cycle SDM was determined i to be 2.311% delta K/K (see Table 1). The SDM demonstrated exceeded the 0.38%

delta K/K required to satisfy Technical Specification 3.1.1.

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TABLE 1 SHUTDOWN MARGIN CALCULATION Critical Rod = 42-31 @ 18 Predicted Keff at Critical Rod Pattern (1) 1.0096 K.,

Period Reactivity Correction From CMR (2) 0.00025 delta K/K Moderator Temperature Reactivity Correction (3) -0.0018 delta K/K Keff with strongest rod out from CMR

-(4) 0.98444 delta K/K Shutdown Margin Keff:

SDM Keff = [ (1) - (2) + (3) - (4) ] x 100

= 2.311 % delta K/K

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i l LTS-1100-2, CHECKING FOR REACTIVITY ANOMALIES  ;

! PURPOSE The purpose of this test is to compare the actual and predicted critical rod configurations to detect any unexpected reactivity trends.

l CRITERIA 1

! I In accordance with Technical Specification 3.1.2, the reactivity equivalence of the difference between the actual control rod density and the predicted control rod density shall not exceed 1% delta K/K. If the difference does exceed 1% delta K/K, the Core '

Management Engineers (Nuclear Fuel Services) will be promptly notified to investigate the anomaly. The cause of the anomaly must be determined, explained, and corrected for continued operation of the unit.

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RESULTS AND DISCUSSION l

Two reactivity anomaly calculations were successfully performed during the Unit 1 l Cycle 8 Startup Test Program, one from the in-sequence critical and one from  ;

steady-state, equilibrium conditions at approximately 100 percent of full power.

The initial critical occurred on April 12,1996, on control rod 42-31 at position 18, using an A-2 sequence. The moderator temperature was 164 degrees F and the reactor period was 242 seconds. Using rod worth information, moderator temperature reactivity corrections, and period reactivity corrections supplied by Nuclear Fuel 1 Services (in the Cycle Startup Package), the actual critical was determined to be within i l 0.755% delta K/K of the predicted critical (see Table 2) in the conservative direction. I The anomaly determined is within the 1% delta K/K allowed by Technical Specification 3.1.2. However, this anomaly is larger than typically observed at LaSalle. The cause of the reactivity anomaly has been attributed to extended coast down of the previous cycle, a large top peak an the end of that cycle and the use of a course cross section instead of a finer cross section library that is also available. i The reactivity anomaly calculation for power operation, was performed using data f from May 25,1996 at 98.2% pov er at a cycle exposure of 213.4 MWD /ST, at '

equilibrium conditions. The predicted notch inventory supplied by Nuclear Fuel Services was 544.7 notches.

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The actual corrected notch inventory was 436.3 notches. Using the notch worth )

provided by Nuclear Fuel Services, the resulting anomaly was 0.19% delta K/K. This l value is within the 1% delta K/K criteria of Technical Specification 3.1.2.

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TABLE 2 l

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INITIAL CRITICALITY COMPARISON CALCULATIONS i Predicted Keff at actual Critical pattern  ;

i (1)- 1,0096 K, i Reactivity Period Correction ,

l (2) 0.00025 delta K/K Moderator Temperature Correction l

(3) -0.0018 delta K/K Reactivity Anomaly RA = [ ((1) - 1.0) - (2) + (3) ) x 100

= 0.755 % delta K/K l

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I LTS-1100-4, SCRAM INSERTION TIMES PURPOSE The purpose of this test is to demonstrate that the control rod scram insertion times are within the operating limits set forth by the Technical Specifications (3.1.3.2,3.1.3.3,  ;

3.1.3.4).

CRITERIA The maximum scram insertion time of each control rod from the fully withdrawn position (48) to notch position 05, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

The average scram insertion time of all operable control rods from the fully withdrawn position (48), based on de-energization of the scram pilot valve solenoids as time zero, j shaji not exceed any of the following: l Position Inserted From Average Scram Insertion Fully Withdrawn Time (Seconds) 45 0.43 39 0.86 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 control rods arranged in a two-by-two array, based on de- energization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:

Position Inserted From Average Scram insertion Fully Withdrawn Time (Seconds) 45 0.45 39 0.92 25 2.05 05 3.70

o RESULTS AND DISCUSSION Scram testing was successfully completed on April 4,1996. All control rods were scram timed from full out. . All control rod scram timing acceptance criteria were met during this test. The results of the testing are given below.

Maximum Average Average Scram Times Scram Times in a Position of all CRDs (secs.) Two-by-Two Array (secs.)

45 0.329 0.360 39 0.626 0.669 25 1.339 1.445 05 2.429 2.595 Tau Ave (position 39) for Minimum Critical Power Ratio Limit determination: 0.626 seconds.

,t 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 i 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%.

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RESULTS AND DISCUSSION l 1

Core power symmetry calculations were performed based upon data obtained from a  !

full core TIP set (OD-1) performed on May 28,1996 at approximately 100% power. l The TIP uncertainty was 2.75% with an average standard deviation of 3.714% which is within the 8.7% acceptance criteria.

The maximum deviation between symmetrical TIP pairs was 10.32% for TIP pairs 33-43 satisfying the criteria of the test (less than 25%). All of the other strings were under 6%

deviation.

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