RA06-041, Cycle 12 Startup Test Report Summary

Cycle 12 Startup Test Report Summary

ML061780187
Person / Time
Site:	LaSalle
Issue date:	06/16/2006
From:	Landahl S Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA06-041
Download: ML061780187 (11)

Text

Exelon Generation Company, LLC www.exeloncorp.com E:xel~r,,

LaSalle County Station Nuclear 2601 North 21" Road Marseilles. IL61341-9757 RA06-041 June 16, 2006 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 LaSalle County Station, Unit 1 Facility Operating License No. NPF-1 1 NRC Docket No. 50-373

Subject:

LaSalle 1 Cycle 12 Startup Test Report Summary Enclosed for your information is the LaSalle County Station (LCS) Unit 1 Cycle 12 Startup Test Report. This report is submitted in accordance with Technical Requirements Manual Section 5.0.b.

LaSalle County Station Unit 1 Cycle 12 began operation on March 18, 2006, following a refueling and maintenance outage. The Unit 1 Cycle 12 core loading consisted of 280 fresh Framatome-ANP (FANP) Atrium-10 fuel bundles, 290 Global Nuclear Fuel (GNF) once-burned GE14 fuel bundles and 194 twice-burned FANP Atrium-10 fuel bundles. Also installed in the Unit 1 Cycle 12 reactor were eight new GE/Reuter-Stokes NA-300 Local Power Range Monitors (LPRMs), and fifteen new General Electric Marathon Control Rod blades.

Attached are the evaluation results from the following tests:

- Reactor Core Verification

- Single Rod Subcritical Check

- Control Rod Friction and Settle Testing

- Control Rod Drive Timing

- Shutdown Margin Test (In-sequence critical)

- Reactivity Anomaly Determination

- Scram Insertion Times

- TIP Measurement Uncertainty

- Reactor Recirculation System Performance

June 16, 2006 U.S. Nuclear Regulatory Commission Page 2 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.

Should you have any questions concerning this letter, please contact Mr. T. W.

Simpkin, Regulatory Assurance Manager, at (815) 415-2800.

Respectfully, S. R. Landahl Site Vice President LaSalle County Station Attachment cc: Regional Administrator - NRC Region III NRC Senior Resident Inspector - LaSalle County Station

ATTACHMENT Page 1 of 9 Reactor Core Verification Purpose The purpose of this test is to visually verify that the core is loaded as intended for Unit 1 Cycle 12 operation.

Criteria The as-loaded core must conform to the cycle core design used by the Core Management Organization (FANP & Nuclear Fuels) in the reload licensing analysis. 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 number map signed by the audit participants.

Results and Discussion Core verification was performed concurrently with core load and shuffle per core verification guideline NF-AA-330-1001. The Unit 1 Cycle 12 core verification consisted of a core height, assembly orientation, assembly location, and assembly seating check 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 7 through March 8, 2006, the core was verified as being properly loaded and consistent with the LaSalle 1 Cycle 12 Design Basis Loading Plan per Transmittal of Design Information (TODI) # NF0500160, Revision 2. This was documented in WO# 00704587.

ATTACHMENT Page 2 of 9 Single Rod Subcritical Check Purpose The purpose of this test is to demonstrate that the Unit 1 Cycle 12 core will remain subcritical upon the withdrawal of the analytically determined highest worth control rod.

Criteria In accordance with LTP-1600-30, the core must remain subcritical, with no significant increase in SRM readings, with the analytically determined highest worth rod fully withdrawn.

Results and Discussion The analytically determined highest worth rod for the Beginning of Cycle 12 for Unit 1 was determined by Nuclear Fuels to be control rod 10-19 per TODI# NF0600037, Revision 1. On March 13, 2006, with a Unit 1 moderator temperature of 85 OF, control rod 10-19 was withdrawn to the full out position (48) and the core remained subcritical with no significant increase in SRM readings. This information is documented on LTP-1600-30, Attachment A.

ATTACHMENT Page 3 of 9 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 (CRD) following core alterations.

Criteria Appropriate acceptance criteria are provided in LOS-RD-SR7 and include limits on rod settle time (less than 10 seconds), scram times from position 45 to position 05 (less than 1.25 times the full core baseline), and full stroke insertion time criteria (dependent on seal leakage and normal insertion time).

Friction testing shall be performed on the respective control rod drives(s) when any condition listed below is applicable:

• After relocation or replacement of the CRDM.

• After relocation or replacement of control rod blades.

• After maintenance or modification of an installed CRDM that could affect the performance of the drive.

• Meeting Channel Distortion criteria established in TODI NF04001 11, Revision 8.

• The Unit Nuclear Engineer or CRD System Engineer determines that friction testing is appropriate.

Results and Discussion 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 identified by the above criteria as the rod met the appropriate acceptance criteria. The testing was completed on March 15, 2006 and is documented in LOS-RD-SR7.

ATTACHMENT Page 4 of 9 Control Rod Drive Timing Purpose The purpose of this test is to check and set the insert and withdrawal speeds of the control rod drives (CRDs).

Criteria Per LOS-RD-SR5, "Control Rod Drive Timing," the acceptance criteria for the withdraw times (full-in to full-out) is between 45 and 60 seconds and insert times (full-out to full-in) is between 40 and 55 seconds. Per UFSAR Section 4.6.1.1.2.5.2, the maximum allowed withdrawal speed is 5.14 inches /second (minimum withdrawal time of 28 seconds).

Results and Discussion LOS-RD-SR5 timing was performed for all CRDs requiring post maintenance testing. Initial timing was completed on March 17, 2006 and is documented in WO# 00704860. Speed control valves for CRDs initially outside LOS-RD-SR5 acceptance criteria were adjusted. As left CRD speeds are within the UFSAR criteria.

ATTACHMENT Page 5 of 9 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 remain subcritical throughout the operating cycle with the highest worth control rod in the full-out position and all other rods fully inserted.

Criteria In accordance with LTS-1100-1 and Technical Specifications, if a shutdown margin (SDM) of 0.38%

Ak/k + R cannot be demonstrated with the highest worth 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 12 is 0.0% Ak/k per ANP-2528(P), "Startup and Operations Report LaSalle Unit 1 Cycle 12," transmitted by NF TODI#

NF0600080, Revision 0, so a SDM of 0.38% Ak/k must be demonstrated.

Results and Discussion The beginning-of-cycle SDM was successfully determined from the initial critical data. The initial Cycle 12 critical occurred on March 17, 2006, on control rod 34-55 at position 22, using an A-2 sequence. The moderator temperature was 169 0F and the reactor period was 389 seconds. Using rod worth information, moderator temperature reactivity corrections, and period reactivity corrections supplied by Nuclear Fuels in TODI# NF 0600080, the SDM was determined to be 1.238% Ak/k. The SDM exceeded the 0.38% Ak/k that was required to satisfy the Technical Specifications.

ATTACHMENT Page 6 of 9 Reactivity Anomaly Determination 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 NF-AB-715, NF-AB-760, and Technical Specifications, 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 at full power steady state conditions. If the difference does exceed 1% Ak/k, the cause of the anomaly must be determined, explained, and corrected for continued operation of the unit.

Results and Discussion Two reactivity anomaly calculations were successfully performed during the Unit 1 Cycle 12 Startup Test Program. One reactivity anomaly calculation is from the in-sequence critical and the other is from steady state, equilibrium conditions at approximately 100 percent of full power.

The initial critical occurred on March 17, 2006, on control rod 34-55 at position 22, using an A-2 sequence. The moderator temperature was 169 OF and the reactor period was 389 seconds. Using rod worth information, moderator temperature, reactivity corrections, and period reactivity corrections supplied by Nuclear Fuels, the actual critical was determined to be within 0.0033% Ak/k of the predicted critical. The anomaly determined is within the 1% Ak/k required for BOC conditions as stated in NF-AB-715. This was documented in NF-AB-715, Attachment 3.

The reactivity anomaly calculation for full power steady state operation was performed on March 23, 2006. The data used was from 99.9% power at a cycle exposure of 98.2 MWD/MT at equilibrium conditions. The expected k0, supplied by Nuclear Fuels was 1.0030. The actual ke, was 1.0001. The resulting anomaly was 0.289% Ak/k. This value is within the 1% Ak/k criteria of Technical Specifications.

ATTACHMENT Page 7 of 9 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.

Criteria In accordance with LTS-1 100-4 and Technical Specifications, 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.

No more than 12 operable control rods shall be "slow" in accordance with the below table. In addition, no more than 2 operable control rods that are "slow" shall occupy adjacent locations.

When the scram insertion time of an operable control rod from the fully withdrawn position (48), based on de-energization of the scram pilot valve solenoids as time zero, exceeds any of the following, that control rod is considered "slow":

Notch Position Scram Time to Notch Indicated (seconds) 45 0.52 39 0.80 25 1.77 05 3.20 Results and Discussion Scram testing was successfully completed on March 20, 2006 per WOs 00704489 and 00875960. All 185 rods were scram timed during the reactor pressure vessel leakage testing (Hydro) prior to startup or at power prior to exceeding 40% rated thermal power. One control rod (30-07) was determined to be "slow." The results of the testing are given below.

Notch Position Core Average Scram Times of all CRDs (sec) 45 0.358 39 0.654 25 1.363 05 2.431 These results also meet the "Nominal" Scram Speeds referenced in the Unit 1 Cycle 12 Core Operating Limits Report (TRM Appendix I).

ATTACHMENT Page 8 of 9 TIP Measurement Uncertainty Purpose The purpose of this test is to verify the core power symmetry.

Criteria In accordance with NF-AB-716, the Z2 value of the total measured TIP uncertainty must be less than the critical value at the 1% confidence level, i.e., 36.19 for 19 TIP pairs.

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 March 24, 2006, at approximately 100% power. The TIP set was performed with all 5 TIP machines operable. All traces were obtained. The X2 value was 6.32, which satisfies the test criteria of 36.19 for 19 pairs. The maximum deviation between symmetrical TIP pairs was 4.83%, which is within the 25% acceptance criteria.

ATTACHMENT Page 9 of 9 Recirculation System Performance Purpose The purpose of this procedure is to collect sufficient baseline data at the beginning of cycle to establish the following relationships:

- core thermal power vs. total core flow

- recirculation total drive flow vs. total core flow

- core plate flow vs. total core flow

- recirculation flow control valve position vs. loop drive flow

- jet pump readings vs. loop drive flow Criteria In accordance with LTP-1 600-11, LTP-1 600-13, and Technical Specifications, the performance curves used in conjunction with reactor recirculation system flow and differential pressure data will establish baseline data to determine if possible jet pump or recirculation pump degradation exists.

The established baseline performance curves will also be used to verify jet pump operability to determine if jet pump anomalies exist.

Results and Discussion Reactor recirculation data was collected during the LIC12 startup. Data was obtained from computer points for all the points of interest to evaluate the RR System performance. No significant changes from LIC1 1 were noted in the LI C12 RR performance curves. This was completed on April 25, 2006 and is documented in WO# 00730895.