RA17-047, Submittal of Cycle 17 Startup Test Report Summary
| ML17132A056 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/12/2017 |
| From: | Trafton W Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA17-047 | |
| Download: ML17132A056 (8) | |
Text
Exelon Generation RA17-047 May 12, 2017 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555
Subject:
LaSalle County Station, Unit 2 Renewed Facility Operating License No. NPF-18 NRC Docket No. 50-374 LaSalle 2 Cycle 17 Startup Test Report Summary LaSalle County Station 2601 North 21 51 Road Marseilles, IL 61341 815-415-2000 Telephone www.exeloncorp.com Enclosed for your information is the LaSalle County Station (LSCS) Unit 2 Cycle 17 Startup Test Report. This report is submitted in accordance with Technical Requirements Manual Section 5.0.b.
LaSalle County Station Unit 2 Cycle 17 began operation on March 9, 2017, following a refueling and maintenance outage. The Unit 2 Cycle 17 core loading consisted of 268 fresh Global Nuclear Fuel GNF-2 fuel bundles, 263 once-burned Global Nuclear Fuel GNF-2 fuel bundles, 4 once-burned Global Nuclear Fuel GNF-3 fuel bundles, and 229 twice-burned Global Nuclear Fuel GNF-2 fuel bundles. Also installed in the Unit 2 Cycle 17 reactor were 8 new GE/Reuter-Stokes NA-300 Local Power Range Monitors (LPRMs) and 3 new General Electric Ultra HD Control Rod blades.
Attached are the evaluation results from the following tests:
- Reactor Core Verification
- Control Rod Friction and Settle Testing
- Shutdown Margin Test (In-sequence critical)
- Reactivity Anomaly Calculation (Critical and Full Power)
- Scram Insertion Times
- Reactor Recirculation System Performance
May 12, 2017 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.
There are no regulatory commitments in this letter. Should you have any questions concerning this report, please contact Mr. Guy V. Ford, Jr., Regulatory Assurance Manager, at (815) 415-2800.
Respectfully,
!/ILp:ft-wmiam J. Trafton Site Vice President LaSalle County Station Attachment cc:
Regional Administrator-NRC Region Ill NRC Senior Resident Inspector-LaSalle County Station
Purpose ATTACHMENT Page 1 of 6 Reactor Core Verification The purpose of this test is to visually verify that the core is loaded as intended for Unit 2 Cycle 17 operation.
Criteria The as-loaded core must conform to the cycle core design used by the Core Management Organization (GNF & 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 per NF-AA-330-1001 "Core Verification Guideline". The Unit 2 Cycle 17 core verification consisted of a core height, assembly orientation, assembly location, and assembly seating check.
Bundle serial numbers and orientations were recorded during the video recorded scans for comparison to the appropriate core loading map and Cycle Management documentation. The core was verified as being properly loaded and consistent with the LaSalle 2 Cycle 17 Core Loading Plan, Revision 5. This was documented in WO# 1815073-01. During core verification, a metal sliver was identified on the fuel channel installed on bundle JYY 402. The sliver was removed and the bundle was dispositioned by Nuclear Fuels for operation in the L2C17 core. This is documented inlR3981119.
Purpose ATTACHMENT Page 2 of 6 Control Rod Friction and Settle Testing 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 (CAD) following core alterations.
Criteria Appropriate acceptance criterion for beginning of cycle rod settle times is provided in LOS-RD-SR?, "Channel Interference Monitoring". The control rod settle test acceptance criterion is less than or equal to 3 seconds. The testing population consists of all control rods.
Results and Discussion CAD Friction Testing commenced after the completion of the core load verification. All 185 control rods settled in less than 3 seconds, which is documented in WO# 1815091-05 and WO# 1815071-01.
Purpose ATTACHMENT Page 3 of 6 Shutdown Margin Test 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 strongest worth control rod in the full-out position and all other rods fully inserted.
Criteria In accordance with L TS-1100-1 and Technical Specification 3.1.1 "Shutdown Margin (SDM)", if a shutdown margin (SDM) of 0.38% ~k/k + R cannot be demonstrated with the strongest 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 17 is 0.00% ~k/k per the LaSalle Unit 2 Cycle 17 Cycle Management Report, so a SDM of 0.38% ~k/k must be demonstrated.
Results and Discussion The beginning-of-cycle SDM was successfully determined from the initial critical data.
The initial Cycle 17 critical occurred on March 8, 2016, on control rod 22-35 at position 16, using an A2 sequence. The moderator temperature was 168 °F and the reactor period was 250 seconds. Using L TS-1100-1 and the LaSalle Unit 2 Cycle 17 Cycle Management Report, the SDM was determined to be 1.582% ~k/k. This was documented in L TS-1100-1, Attachment A and WO# 1815072-01. The SDM was greater than the minimum 0.38% ~k/k that is required to satisfy the Technical Specifications.
Purpose ATTACHMENT Page 4 of 6 Reactivity Anomaly Calculation 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-LA-715, "Critical Predictions with 30 Monicore", NF-AB-760, "Reactivity Anomaly Determination", and Technical Specification 3.1.2, "Reactivity Anomalies", the reactivity equivalence of the difference between the actual critical control rod configuration and the predicted critical control rod configuration and the difference between the actual and predicted reactivity of the control rod configuration at full power steady state conditions shall not exceed 1 % ~k/k. If the difference exceeds 1 % ~k/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 2 Cycle 17 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% full power.
The initial Cycle 17 critical occurred on March 8, 2016, on control rod 22-35 at position 16, using an A2 sequence. The moderator temperature was 168 °F and the reactor period was 250 seconds. The expected kett supplied by Nuclear Fuels was 1.0020. The actual kett was 1.00152. The resulting anomaly was -0.048% ~k/k. The anomaly determined is within the 1 % ~k/k required for beginning-of-cycle conditions as stated in NF-LA-715. This was documented in NF-LA-715, Attachment 3 and WO# 1815075-01.
The reactivity anomaly calculation for full power steady state operation was performed.
The data used was from 99.8% power at a cycle exposure of 97.2 MWD/sT at equilibrium conditions. The expected kett supplied by Nuclear Fuels was 1.0075. The actual kett was 1.0071. The resulting anomaly was 0.04% ~k/k. This value is within the 1% ~k/k criteria of Technical Specifications. This was documented in NF-AB-760,, and WO# 1818801-01.
Purpose ATTACHMENT Page 5 of 6 Scram Insertion Times 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 LOS-RD-SR12, "Scram Insertion Times" and Technical Specification 3.1.3, "Control Rod OPERABILITY" and 3.1.4, "Control Rod Scram Times", 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. Also, 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 45 39 25 05 Results and Discussion Scram Time to Notch Indicated (seconds) 0.52 0.80 1.77 3.20 Scram testing was performed per WO# 1818799-06 and WO# 1814984-01. Results of testing are given below.
Notch Position 45 39 25 05 Core Average Scram Times of all CRDs (sec) 0.297 0.585 1.271 2.303 These results also meet the "Option B" Scram Speeds referenced in the Unit 2 Cycle 17 Core Operating Limits Report {TRM Appendix J).
ATTACHMENT Page 6 of 6 Reactor Recirculation System Performance Purpose The purpose of this test 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-1600-13, "Recirculation System Performance", and Technical Specification 3.4.3, "Jet Pumps", the performance curves used in conjunction with reactor recirculation (RR) 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 RR data was collected during the L2C17 startup. Data was obtained from computer points for all the points of interest to evaluate the RR System performance. The RR performance curves were updated for L2C17; no significant changes from L2C16 were noted in the curves. This is documented in WO# 1826535-01.