BSEP 11-0071, Cycle 20 Startup Report
| ML11189A035 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/30/2011 |
| From: | Noel P Progress Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| BSEP 11-0071 | |
| Download: ML11189A035 (8) | |
Text
BRUNSWICK UNIT 2, CYCLE 20 STARTUP REPORT June 2011 Prepared by:
Reviewed by:
Reviewed by:
Reviewed by:
Approved by:
Noel, Peter 2011.06.20 09:25:53:-04'00' Peter Noel (BWR Fuel Engineering)
Earp Jr, Dennis.
2011.06.20 09:56:13 -04'00' Dennis Earp Jr. (BWR Fuel Engineering)
Westermark, Hans 2011.06.20 13:59:35 -04'00' Hans Westermark (BNP Reactor Engineering)
Murray, William R. (Bill) 2011.06.20 12:30:10 -04'00' William Murray (Licensing/Regulatory Programs)
Stroupe, Charles For Roger Thomas 2011.06.21 06:19m],9 -04'600 Roger Thomas (Supervisor - NFM&SA)
Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C20 Startup Report Page 2 of 8, Revision 0 1.0 Introduction This report summarizes observed data from the Brunswick Steam Electric Plant (BSEP)
Unit 2, Cycle 20 (B2C20) startup tests. The Cycle 20 core represents the first loading of the AREVA ATRIUM I0XM fuel type in Unit 2. A firesh fuel batch size of 226 ATRIUM 1OXM fuel assemblies has been loaded (Reference 2.11).
Pursuant to Section 1.3.4.2.1 of the BSEP 1 & 2 Updated Final Safety Analysis Report (UFSAR) (Reference 2.1), a summary report of plant startup and power escalation testing shall be submitted to the NRC should any one of four conditions occur. Condition (3) of the referenced requirements applies:
(3):
"installation of fuel that has a different design or has been manufactured by a different fuel supplier."
This report shall include results of neutronics related startup tests following core reloading as described in the UFSAR.
2.0 References 2.1 BSEP UFSAR 2.2 BSEP Technical Specifications 2.3 OENP-24.13, "Core Verification" (PGN RMS 4659172) 2.4 0FH-1 1, "Refueling" (PGN.RMS 4652182) 2.5 0PT-14.2.1, "Single Rod Scram Insertion Times Test" (PGN RMS 4659168) 2.6 OPT-14.3.1, "Insequence Critical Shutdown Margin Calculation" (PGN RMS 4659108) 2.7 OPT-14.5.2, "Reactivity Anomaly Check" (PGN RMS 4659117) 2.8 0PT-50.0, "Reactor Engineering Refueling Outage Testing" (PGN RMS 4683253) 2.9 OPT-50.3, "TIP Reproducibility And Uncertainty Detennination"(PGN RMS 4659156, RMS 4659160) 2.10 0PT-90.2, "Friction Testing of Control Rods" (PGN RMS 4659142) 2.11 BNP Calculation 2B21-0646, "B2C20 Cycle Management Report", Revision 0.
3.0 UFSAR Section 14.4.1, Item 1: Core Loading Verification A Core Loading Pattern Verification was performed per BSEP Engineering Procedure OENP-24.13, "Core Verification" (Reference 2.3). The core was verified to be loaded in accordance with the analyzed B2C20 core design.
Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C20 Startup Report Page 3 of 8, Revision 0 4.0 UFSAR Section 14.4.1, Item 4A: TIP Operability and Bundle Power Evaluation
- a.
TIP Measurement Uncertainty Radial (bundle or 2D) and nodal (3D) gamma TIP measurement uncertainties were determined in accordance with BSEP Periodic Test Procedure OPT-50.3, "TIP Uncertainty Determination"(Reference 2.9). Total radial TIP measurement uncertainty at high core thermal power (CTP) (>80% CTP) was 0.748% and total nodal TIP measurement uncertainty was 1.153%. These radial and nodal uncertainties were also determined at medium core thermal power (40% to 80% CTP) and were 1.095% and 1.620%, respectively. The results met the test acceptance criteria.
- b.
Measured and Calculated TIP Comparison Radial and nodal deviations between measured and calculated TIP data were determined in accordance with BSEP Periodic Test Procedure OPT-50.3, "TIP Uncertainty Determination" (Reference 2.9). The radial deviation at high core thermal power
(>80% CTP) was 1.721% and the nodal deviation was 3.163%. These radial and nodal deviations were also determined at medium core thermal power (40% to 80% CTP) and were 2.03 1% and 3.826%, respectively. The results met the test acceptance criteria.
- c.
Monitored Power Uncertainty Radial and nodal monitored power uncertainties were determined in accordance with BSEP Periodic Test Procedure OPT-50.3, "TIP Uncertainty Determination" (Reference 2.9). The radial monitored power uncertainty at high core thermal power
(>80% CTP) was 2.369% and the nodal monitored power uncertainty was 2.958%.
These radial and nodal uncertainties were also determined at medium core thermal power (40% to 80% CTP) and were 2.613% and 3.288%, respectively. The results met the test acceptance criteria.
- d.
Bundle Powers This analysis compares the MICROBURN-B2 predictions of bundle powers to the plant process computer's measured bundle powers in accordance with BSEP Periodic Test procedure OPT-50.0, "Reactor Engineering Refueling Outage Testing" (Reference 2.8).
Bundles located in peripheral control cells or uncontrolled peripheral locations are excluded. The maximum radial difference was calculated to be 2.4 1% at medium power (40% to 80% CTP). The results met the test acceptance criteria.
Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C20 Startup Report Page 4 of 8, Revision 0 5.0 UFSAR Section 14.4.1, Item 2: Control Rod Mobility Control rod mobility is verified by two tests: friction testing and scram timing. The results of these tests and their acceptance criteria are described below.
- a.
Friction Testing Friction Testing was performed prior to startup per BSEP Periodic Test Procedure OPT-90.2, "Friction Testing of Control Rods" (Reference 2.10). Control rods were verified to complete full travel without excessive binding or friction. In a prerequisite to OPT-90.2, the reactor was observed to remain subcritical during the withdrawal of the most reactive rod per the BSEP Fuel Handling Procedure OFH-11, "Refueling" (Reference 2.4).
- b.
Scram Time Testing Scram Time Testing was performed for each control rod prior to exceeding 40% power per BSEP Periodic Test Procedure OPT-14.2. 1, "Single Rod Scram Insertion Times Test" (Reference 2.5). The acceptance criteria for these tests are found in Technical Specification 3.1.4 (Reference 2.2). The control rods had a scram time of* 7.0 seconds and thus were considered operable in accordance with Technical Specification 3.1.3. The maximum measured 5%, 20%, 50%, and 90% insertion times are given in Attachment 1 of this report.
The average 20% insertion time measured was 0.794 seconds which is faster than the analyzed nominal speed limit of< 0.862 seconds.
6.0 UFSAR Section 14.4.1, Item 3: Reactivity Testing Reactivity Testing consists of a shutdown margin (SDM) measurement, reactivity anomaly check, and measured critical kln. comparison to predicted values. The results of these tests are provided below with the acceptance criteria.
- a.
Shutdown Margin SDM measurements were performed per BSEP Periodic Test Procedure OPT-14.3. 1, "Insequence Critical Shutdown Margin Calculation" (Reference 2.6). The cycle minimum SDM was determined to be 1.662% Ak/k compared to a predicted cycle minimum SDM value of 1.62% Ak/k (Reference 2.11), resulting in an absolute difference of 0.042% Ak/k. The cycle minimum SDM is determined by subtracting the maximum decrease in SDM which occurs at 0.0 GWD/MTU cycle exposure (R = 0.0% Ak/k) from
Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C20 Startup Report Page 5 of 8, Revision 0 the SDM at beginning-of-cycle (BOC). The acceptance criterion for minimum SDM is defined in Technical Specification 3.1.1, which requires the SDM be > 0.38% Ak/k during the entire cycle. Since the cycle minimum SDM was determined to be 1.662%
Ak/k for B2C20, the acceptance criterion is met.
- b.
Reactivity Anomaly A reactivity anomaly test was performed at near rated conditions (2915.5 MWt or 99.7%
of rated power) per BSEP Periodic Test Procedure OPT-14.5.2, "Reactivity Anomaly Check" (Reference 2.7). The acceptance criterion is defined by Technical Specification 3.1.2, which requires that the reactivity difference between monitored and predicted core keff be within +/-1% Ak/k. The measured and predicted values for krff were 1.0003 and 0.9975 (Reference 2.11), respectively, an absolute difference of 0.28% Ak/k.
This is within the +/-1% Ak/k requirement.
- c.
Cold Critical Eigenvalue (keff)
The measured BOC cold critical keff per BSEP Periodic Test Procedure OPT-14.3.1, "Insequence Critical Shutdown Margin Calculation" (Reference 2.6), was inferred as 0.99487 by nodal simulator code calculations with actual critical conditions as input (including period correction). The predicted BOC cold critical kef" was 0.9940 resulting in a measured to predicted difference of 0.087% Ak/k. Therefore, per Technical Specification 3.1.2, the acceptance criterion requiring agreement within +/-1% Ak/k is met.
7.0 Additional Testing Results As a matter of course, key testing and checks beyond those specified in the UFSAR are performed during initial startup and power ascension. These "standard" tests are described in items (a) and (b) below.
- a.
Core Monitoring Software Comparisons to Predictions Thermal limits calculated by the online POWERPLEX Core Monitoring Software System were compared to those calculated by MICROBURN-B2 predictions at medium and high power levels (Reference 2.8). The results of these comparisons and the POWERPLEX statepoints are provided as Attachment 2. The results met the test acceptance criteria.
- b.
Hot Full Power Eigenvalue After establishing a sustained period of full power equilibrium operation at 186.4 MWD/MTU on April 27, 2011, the predicted and core follow Hot Full Power Eigenvalues (kfrr) were compared. (Reference 2.8). The core follow kerr was calculated as
Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C20 Startup Report Page 6 of 8, Revision 0 1.0002 and the predicted keff was 1.0001. The difference between the predicted and core follow values is 0.01% Ak/k which is within the +/-1% Ak/k reactivity anomaly requirements.
8.0 Summary Evaluation of the BSEP Unit 2, Cycle 20 startup data concludes the core has been loaded properly and is operating as expected. The startup and initial operating conditions and parameters compare well to predictions. Core thermal peaking design predictions and measured peaking comparisons met the startup acceptance criteria. The BOC SDM demonstration indicates adequate SDM will exist throughout B2C20. The UFSAR prescribed and additional tests met their acceptance criteria.
Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C20 Startup Report Page 7 of 8, Revision 0 to the B2C20 Startup Report Results of Control Rod Scram Time Testing Maximum Measured Scram Insertion Time Technical Specification 3.1.4 Insertion Position/Notch Tech Spec Maximum Measured "Slow" Limit Insertion Time (seconds)
(seconds) 5%
46 0.440 0.325 20%
36 1.080 0.855 50%
26 1.830 1.417 90%
06 3.350 2.589
Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C20 Startup Report Page 8.of 8, Revision 0 to the B2C20 Startup Report Core Monitoring Software Comparisons to Predictions Medium Power 65.3% CMWT, April 21, 2011 Thermal Limit POWERPLEX MICROBURN-B2 Absolute On-Line Predicted Difference Monitoring CMFLCPR 0.772 0.780 0.008 CMAPRAT 0.565 0.562 0.003 CMFDLRX 0.773 0.778 0.005 CMFLPD 0.377 0.375 0.002 High Power 99.8% CMWT, April 27, 2011 Thermal Limit POWERPLEX MICROBURN-B2 Absolute On-Line Predicted Difference Monitoring CMFLCPR 0.875 0.884 0.009 CMAPRAT 0.789 0.777 0.012 CMFDLRX 0.899 0.883 0.016 CMFLPD 0.498 0.489 0.009