RA14-025, Cycle 16 Startup Test Report Summary
| ML14115A231 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/25/2014 |
| From: | Karaba P Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA14-025 | |
| Download: ML14115A231 (11) | |
Text
RA14-025 April 25, 2014 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Unit 1 Facility Operating License No. NPF-1 1 NRC Docket No. 50-373
Subject:
LaSalle 1 Cycle 16 Startup Test Report Summary Enclosed for your information is the LaSalle County Station (LSCS) Unit 1 Cycle 16 Startup Test Report. This report is submitted in accordance with Technical Requirements Manual Section 5.0.b.
LaSalle County Station Unit 1 Cycle 16 began operation on March 1, 2014, following a refueling and maintenance outage. The Unit 1 Cycle 16 core loading consisted of 288 fresh Global Nuclear Fuel GNF-2 fuel bundles, 296 once-burned Global Nuclear Fuel GNF-2 fuel bundles and 180 twice-burned AREVA Atrium-10 fuel bundles. Also installed in the Unit 1 Cycle 16 reactor were 12 new GE/Reuter-Stokes NA-300 Local Power Range Monitors (LPRMs), and 4 new General Electric Ultra HD 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 Calculation (Critical and Full Power)
- Scram Insertion Times
- Core Power Distribution Symmetry Analysis
- Reactor Recirculation System Performance 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. Guy V. Ford, Jr., Regulatory Assurance Manager, at (815) 415-2800.
Respectfully, P. Karaba 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 16 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 and shuffle per core verification guideline NF-AA-330-1001. The Unit 1 Cycle 16 core verification consisted of a core height, assembly orientation, assembly location, and assembly seating check. Bundle serial numbers and orientations were recorded during the videotaped scans for comparison to the appropriate core loading map and Cycle Management documentation. On February 20, 2014, the core was verified as being properly loaded and consistent with the LaSalle 1 Cycle 16 Core Loading Plan, Revision 3. This was documented in WO# 01523072-01.
ATTACHMENT Page 2 of 9 Single Rod Subcritical Check Purpose The purpose of this test is to demonstrate that the Unit 1 Cycle 16 core will remain subcritical upon the withdrawal of the analytically determined strongest Control Rod.
Criteria In accordance with LTP-1 600-30, 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 16 for Unit 1 was determined by Nuclear Fuels to be Control Rod 50-19 per TODI# NF140090, Revision 0. On February 21, 2014, with a Unit 1 moderator temperature of 87°F, Control Rod 50-19 was withdrawn to the full out position (48) and the core remained subcritical with no significant increase in SRM readings. This information is documented in WO# 01523066-01.
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 LOS-RD-SR7 requires beginning of cycle rod settle time in less than or equal to 3 seconds. The testing population consists of all control rods:
Results and Discussion CRD Friction Testing commenced after the completion of the core load verification and single rod subcritical check. All 185 control rods met the appropriate acceptance criteria of less than or equal to 3.0 seconds. The testing was performed on February 21-26, 2014, and is documented in WO# 01523022-01.
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 LOS-RD-SR5, Control Rod Drive Timing, preferred beginning of cycle 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.
Results and Discussion Control rod timing per LOS-RD-SR5 was performed satisfactorily for all 185 CRDMs on February 25-26, 2014, and is documented in WO# 01629862-01. None of the rod withdrawal speeds were faster than the LOS-RD-SR5 preferred 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 strongest worth Control Rod in the full-out position and all other rods fully inserted.
Criteria In accordance with LTS-1 100-1 and Technical Specifications, if a shutdown margin (SDM) of 0.38%
Ak/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 16 is 0.00% Ak/k per the LaSalle Unit 1 Cycle 16 Cycle Management Report, Revision 2, 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 16 critical occurred on February 27, 2014, on Control Rod 14-19 at position 12, using an A sequence.
The moderator temperature was 178 OF and the reactor period was 257 seconds. Using LTS-1 100-1 and the L1 C16 Cycle Management Report, Revision 2, the SDM was determined to be 1.716% Ok/k.
This was documented in LTS-1 100-1, Attachment A and WO# 01523071-01. The SDM was greater than the minimum 0.38% Ak/k that is 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-LA-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 and the difference between the actual and predicted reactivity of the Control Rod configuration at full power steady state conditions shall not exceed 1 % Ak/k. If the difference exceeds 1 % Lk/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 16 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 16 critical occurred on February 27, 2014, on Control Rod 14-19 at position 12, using an A sequence. The moderator temperature was 178 OF and the reactor period was 257 seconds. The expected keff supplied by Nuclear Fuels was 1.0020. The actual keff was 1.00146. The resulting anomaly was 0.054% Ak/k. The anomaly determined is within the 1 % ©k/k required for BOC conditions as stated in NF-LA-715. This was documented in NF-LA-715, Attachment 3 and WO# 01523106-01.
The reactivity anomaly calculation for full power steady state operation was performed on March 5, 2014. The data used was from 99.9% power at a cycle exposure of 64.3 MWD/ST at equilibrium conditions. The expected keff supplied by Nuclear Fuels was 1.0080. The actual keff was 1.0070. The resulting anomaly was 0.10% Ak/k. This value is within the 1 % Ak/k criteria of Technical Specifications. This was documented in NF-AB-760, Attachment 1, and WO# 01523064-01.
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 LOS-RD-SR12 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. 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 Scram Time to Notch Indicated (seconds) 45 0.52 39 0.80 25 1.77 05 3.20 Results and Discussion Scram testing was completed on March 2, 2014 per WO# 01526650. Results of testing are given below.
Notch Position Core Average Scram Times of all CRDs (sec) 45 0.298 39 0.585 25 1.270 05 2.310 These results also meet the "Option B" Scram Speeds referenced in the Unit 1 Cycle 16 Core Operating Limits Report (TRM Appendix I).
ATTACHMENT Page 8 of 9 Core Power Distribution Symmetry Analysis Purpose The purpose of this test is to verify the core power symmetry.
Criteria In accordance with NF-AB-707, the TIP uncertainty value must be less than 6%.
Results and Discussion Core power symmetry calculations were obtained on March 7, 2014 based upon data obtained from a full core TIP set (OD-1) at approximately 100% power. The TIP uncertainty value was 3.00%. This was documented in WO# 01706646-01.
ATTACHMENT Page 9 of 9 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-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 L1 C16 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 Li C16; no significant changes from L1 C15 were noted in the curves. This was completed on March 12, 2014, and is documented in WO# 01528032-01.