ML20209E036
| ML20209E036 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/08/1999 |
| From: | Jamie Benjamin COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9907140126 | |
| Download: ML20209E036 (16) | |
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Commonu calth I'dnon Conip.ni) 1.aNille Georratmg Mation 2(dll Nortli Jist flo.id M r.nDes.11,61.4 i 19 5" 4
'Irl H15 WC61 July 8,1999 United States Nuclear Regulatory Commission Attention: Document Contro! Desk Washington, D.C. 20555 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374
Subject:
Startup Test Report Summary Enclosea for your information and use is the LaSalle County Station Unit 2 Cycle 8 Startup Test Report Summary. This report is submitted in accordance with Technical Specification NPF-18, Section 6.6.A.1.
LaSalle Unit 2 Cycle 8 began commercial operation on April 11,1999 following a refueling and maintenance outage. The Unit 2 Cycle 8 core loading consisted of 256 fresh Siemens Power Corporation ATRIUM-98 fuel bundles and 508 reload bundles manufactured by General Electric.
The startup test program was satisfactorily completed on May 1,1999. 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 and to ensure the acceptability of subsequent test results.
Attached are the evaluation results from the following tests:
- Core Verification j
- Single Rod Subcritical Check j
- Control Rod Friction and Settle Testing
- Control Rod Drive Timing
- Shutdown Margin Test (In-sequence critical)
- Reactivity Anomaly Calculation (Critical and Full Power)
- Scram insertion Times b
- Core Power Distribution Symmetry Analysis
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9907140126 990700 PDR-ADOCK 05000373 P
PDR A t:nionn compam
July 8,1999 U.S. Nucl: r R:gulltory Commission l-Ppge 2 -
i-Should you have any questions conceming this letter, please contact Frank A. Spangenberg, Ill, Regulatory Assurance Manager, at (815) 357-6761, extension 2383.
Respectfully, Jeffrey A. Benjamin Site Vice President LaSalle County Station Enclosure -
cc:
Regional Adm!nistrator-NRC Region ill l
NRC Senior Resident inspector - LaSalle County Station 1
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1 LaSalle Unit 2 Cycle 8 Startup Test Report Summary 1
LaSalle Unit 2 Cycle 8 began commercial operation on April 9,1999 following a refueling and maintenance outage. The Unit 2 Cycle 8 core loading consisted of fresh fuel bundles manufactured by Siemens Power Corporation (128 SPCA9-381B-13GZ7-80M and 128 SPCA9-384B-11GZ6-80M) and 508 reload bundles made by General Electric. In addition,6 LPRM strings were replaced with General Electric NA-300 LPRM strings. Twenty-seven control blades were replaced with General Electric Duralife 215 and General Electric Marathon designs, and 29 control blades were shuffled. All applicable test results (neutron instrument calibration, computer monitoring results, etc.) indicate expected core performance with the new fuel 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 1,1999. 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 tosts:
- Core Verification
- Single Rod Suberitical Check
- Control Rod Friction and Settle Testing
- Control Rod Drive Timing
- Shutdown Margin Test (In-sequence critical)
- Reactivity Anomaly Calculation (Critical and Full Power)
- Scram insertion Times
- Core Power Distribution Symmetry Analysis 1
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l' Findings and Recommendations
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l Based upon the preceding discussion and the review of the startup test report, the "LaSalle County Nuclear Power Station Unit 2 Cycle 8 Startup Test Report" i.e submitted to the NRC in accordance with Technical Specification 6.6.A.1.
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LTP-1700-1, Ccre Verification Purpose 1
The purpose of this test is to visually verify that the core is loaded as intended for Unit 2 Cycle 8 operation.
Criteria l
I The as-loaded core must conform to the cycle core design used by the Core
' Management Organization (Nuclear Fuel Management) in the reload licensing analysis.
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The core verification must be observed by a member of the Commonwealth Edison J
Company staff. Any discrepancies discovered in the loading will be promptly corrected and the affected areas re-verified to ensure proper core loading prior to unit startup.
Conformance to the cycle core design will be documented by a permanent core serial l
number map signed by the audit participants.
Results and Discussion s
Core verification was performed concurrently with core load. The Unit 2 Cycle 8 core verification consisted of a core height, assembly orientation, assembly location, and -
assembly seating eneck performed by Reactor Services and Reactor Engineering.
Bundle serial numbers and orientations were recorded during the videotaped scans for comparison to the appropriate core loading map and Cycle Management documentation. On March 30,1999, the core was verified as being properly loaded and consistent with the Commonwealth Edison Nuclear Fuel Management LaSalle 2 Cycle 8 Design Basis Loading Plan.
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LTP-1600-30, Single Rod Subcritical Check Purpose The purpose of this test is to demonstrate that the Unit 2 Cycle 8 core will remain subcritical upon the withdrawal of the analytically determined strongest control rod.
Criteria 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 Cycle 8 for Unit 2 was determined by Nuclear Fuel Management to be rod 18-31. On March 30,1999, with a Unit 2 moderator temperature of 101 degrees Fahrenheit, rod 18-31 was 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, Single Rod Subcritical Check, allows single control rod withdrawals for control rod testing. This information is documented on LTP-1600-30, Attachment A.
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LTP-700-2, Control Rod Friction and Settle Testing 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
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(CRD) following core alterations.
Criteria With the final cell loading complete for the fuel assemblies in a control cell, the drift alarm shall not be received when moving the control rod from position 00 to 02, and then to 04. Only those assemblies which were greater than 30,000 mwd /ST exposure and control rod drives that underwent maintenance were tested.
Results and Discussion i
Control Rod Drive (CRD) Friction testing commenced after the completion of the core load verification and single rod subcritical check. There was no indication of excessive friction on the control rods tested as described above since none of the rods tested produced a drift alarm. The testing was completed on March 31,1999.
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LOS-RD SR5, Control Rod Drive Timing ourpose The purpose of this test is to check and set the insert and withdrawal speeds of the Control Rod Drives (CRDs). In addition, this surveillance will provide verification that each control rod blade is coupled to its respective CRD mechanism.
Criteria The insert and withdrawal times of a CRD should be 48 i 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 seonds for withdrawal tirnes in the cold shutdown conditions (deprescurized) to give iqaeation of seal wear.
LOS-RD-SR5, Control Rod Drive Timing, currently requires withdraw times to be between 50 and 58 seconds and insert times to be between 40 and 48 seconds. This change might avoid adjustments of the CRD velocity during rated reactor operation.
GE analysis for LaSalle Station (DRF A12-00038-4) allows a maximum withdrawal speed of 5.14 in/sec (28 seconds full stroke of rod).
Results and Discussion All CRD testing was completed after control rod drive replacements with blade guides inserted and the core defueled on November 6,1996. Control rods 06-31,10-19,18-59, 26-55, 30-39, 34-03, 42-39, 54-15, and 58-23 had withdrawal times faster than 50 seconds (but greater than 38.4 seconds) due to degraded drive seals. The following surveillance evaluation was performed allowing the use of the control rods:
During LOS-RD-SRS, Control Rod Drive Timing, the above listed control rods were unable to be adjusted within the timing range estabSshed by the surveillance. LOS-RD-SRS requires adjustment of the 1(2)C11-D001-120 or 1(2)C11-D001-123 Directional Control Valves (DCV) to make timing adjustments. These control rods had their respective 120 DCV fully adjusted and the times were still unable to be set within the surveillance limits. The surveillance times were established to account for seal wear during the course of the cycle, as well as to enhance rod movements. Although the withdrawal times for the control rods listed above are out of tolerance as specified in LOS-RD-SR5, the as-left times are acceptable, even with the.120 DCV fully adjusted.
The time range established in the surveillance (50-58 second range) is given to avoid future timing adjustments during power operation and to provide margin during rod movement to alleviate the potential for control rod double notching.
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n-The design control rod drive speed as defined in the control rod design specifications is 48.0 i 20%. This speed is designed to ensure no rod movement anomalies. Because the times obtained during this surveillance are within the design specifications, there are no operability concerns with these control rod drives. Action Requests have been generated, except for rod 58-23 (49.9 sec), to replace these drives during L2R08.
L These control rods will be trended by the CRD system engineer during cycle eight for any abnormal behavior.
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LTS-1100-1, Shutdown Margin Test 1
Parpose 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 remain suberitical throughout the operating cycle with the strongest worth control rod in the full-out position (position
- 48) and all other rods fully inserted.
Criteria If a shutdown margin (SDM) of 0.38% delta K/K + R cannot be demonstrated with the strongest worth control rod fully withdrawn, the core loading must be 8;tered 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 0.007% detta K/K, so a SDM of 0.387% 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 6 critical occurred on April 9,1999 on control rod 26-39 at position 14, using an A-2 sequence. The moderator temperature was 120.3 degrees F and the reactor period was 182 seconds. Using online core monitoring software (Powerplex) and reactivity data that was supplied by Nuclear Fuel Management, the beginning-of-cyck SDM was determined to be 1.683% delta K/K (see Table 1). The SDM exceeded t. ' O.387% delta K/K that was required to satisfy Technical Specification 3.1.1.
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i TABLE 1 Shutdown Margin Calculation i
Critical Rod: 26-39 @ 14-Moderator temperature: 120.3 degrees F Reactor period: ' 182 seconds Computed cold critical eigenvalue (1) 1.0091 Period Reactivity C?rrection (2)
.000359876 Ker with strongest rod out (3)
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Calculated shutdown margin SDM = [(1)-(2)-(3)] x 100 = 1.683 % AK/K i
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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.
Criteria in accordance with Technical Specification 3.1.2, the reactivity equivalence of the difference between the actual critical control rod configuration and the predicted critical control rod configuration shall not exceed 1% delta K/K. If the difference does exceed 1% delta K/K, the Core Management Engineers (Nuclear Fuel Management) 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.
Results and Discussion Two reactivity anomaly calcu ations were successfully performed during the Unit 2 Cycle 8 Startup Test Program - one from the in-sequence critical and one from steady-state, equilibrium conditio:.s at approximately 100 percent of full power.
The initial critical occurred on April 9,1999, on control rod 26-39 at position 14, using an A-2 sequence. The moderator temperature was 120.3 degrees F and the reactor period was 182 seconds. Using rod worth information, moderator temperature reactivity corrections, and period reactivity corrections supplied by Nuclear Fuel Management, the actual critical was determined to be within -0.37% detta K/K of the predicted critical (see Table 2). The anomaly determined is within the 1% delta K/K allowed by Technical Specification 3.1.2.
l The reactivity anomaly calculation for power operation was performed on April 20, 1999. The data used was from 99.1% power at a cycle exposure of 142.7 MWD /MT at equilibrium conditions. The expected K n supplied by Nuclear Fuel Management was 1.0029. The actual K n was 1.0072. The resulting anomaly was -0.43% AK/K. This value is within the 1% AK/K criteria of Technical Specification 3.1.2.
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TABLE 2 Initial Criticality Comparison Calculations BOC cold critical K.n (from Cycle Management Report)
(1) 1.0050 Period Reactivity Correction (2) 0.000359876 Core monitoring eigenvalue at actual temperature l
(3) 1.0091 Reactivity Anomaly RA
= [ (1) - (3) + (2) ) x 100 = -0.37% AK/K I
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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 posit _ ion (48), 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 0.43 39 0.86 25 1.93 05 3.49 l
l 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 I
Results and Discussion 1
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Scram time testing was successfully completed during the reactor pressure vessel l
hydro on April 4,1999. All control rods were scram timed from full out. All control rod 12
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scram timing acceptance criteria were met during this test. The results of the testing l
are given below:
Average Maximum Average j
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of all rods'(sec)
Two-by-Two Array (sec) 45 0.330 0.371 39 0.628-0.661 25 1,339 1.368 05.
'2.398 2.461 The data for Thermal Limits Determination (SPC Methods) are the core average times.
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LTP-1600-17, Cor7 Power Distribution Symmetry Analysis Purpose The purpose of this test is to verify the core power symmetry.
Criteria The ' value of the total measured TIP uncertainty must be less than the critical value x
at the 1% confidence level. This results in an acceptance criteria of 36.19.
The gross check of the TIP signal symmetry should yield a maximum deviation between symmetrically located pairs of less than 25%.
Results and Discussion Core power symmetry calculations were performed based upon data obtained from a full core TIP set (OD-1) performed on April 16,1999 at approximately 75% power. The X value was 3.27, which satisfies the test criteria of 36.19. The maximum deviation between symmetncal TIP pairs was 6.52%, which is within the 25% acceptance criteria.
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