RA-20-0217, Cycle 23 Startup Report

Cycle 23 Startup Report

ML20177A282
Person / Time
Site:	Brunswick
Issue date:	06/25/2020
From:	Salazar S Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-20-0217
Download: ML20177A282 (10)

Text

Brunswick Nuclear Plant 8470 River Rd SE Southport, NC 28461 June 25 22, 2020 Serial: RA-20-0217 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit No. 1 Renewed Facility Operating License No. DPR-71 Docket No. 50-325 Unit 1 Cycle 23 Startup Report Ladies and Gentlemen:

In accordance with the Brunswick Steam Electric Plant (BSEP), Updated Final Safety Analysis Report (UFSAR), Section 13.4.2.1, Startup Report, Duke Energy Progress, LLC (Duke Energy), is providing the enclosed Brunswick Unit 1, Cycle 23 Startup Report, dated June 2020.

The report is required as a result of the first loading of Framatome ATRIUM 11 fuel in Unit 1 during the Spring 2020 refueling outage.

This letter and the enclosed Startup Report contain no regulatory commitments.

Please refer any questions regarding this submittal to Mr. Stephen Yodersmith, Brunswick Regulatory Affairs, at (910) 832-2568.

Sincerely, For Sabrina Salazar Manager - Nuclear Support Services Brunswick Steam Electric Plant

U.S. Nuclear Regulatory Commission Page 2 of 2

Enclosure:

Brunswick Unit 1, Cycle 23 Startup Report cc (with enclosure):

Ms. Laura Dudes, NRC Regional Administrator, Region II Mr. Andrew Hon, NRC Project Manager Mr. Gale Smith, NRC Senior Resident Inspector Chair - North Carolina Utilities Commission

RA-20-0217 Enclosure Brunswick Unit 1, Cycle 23 Startup Report

BRUNSWICK UNIT 1, CYCLE 23 STARTUP REPORT June 2020 Prepared by:

Peter Noel (Brunswick Nuclear Design )

Reviewed by:

Jason Krupp (Reactor Systems - BNP)

Approved by:

Robert St. Clair (Manager Brunswick Nuclear Design)

Duke Energy, Nuclear Fuels Engineering, Brunswick Nuclear Design B1C23 Startup Report Page 2 of 7, Revision 0 1.0 Introduction This report summarizes observed data from the Brunswick Steam Electric Plant (BSEP) Unit 1, Cycle 23 (B1C23) startup tests. The Cycle 23 core represents the first refuel batch loading (226 assemblies) of the AREVA ATRIUM 11 fuel type in Unit 1 (Reference 2.11).

Pursuant to Section 13.4.2.1 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 Updated Final Safety Analysis Report, Revision 26 2.2 BSEP Technical Specifications 2.3 0ENP-24.13, Core Verification (CMPR-BNP-0ENP-24.13---028040148) 2.4 0FH-11, Refueling (CMPR-BNP-0FH-11---027868739) 2.5 0PT-14.2.1, Single Rod Scram Insertion Times Test (CMPR-BNP-0PT-14.2.1-083--

028040150) 2.6 0PT-14.3.1, Insequence Critical Shutdown Margin Calculation (CMPR-BNP-0PT-14.3.1-021--028040156) 2.7 0PT-14.5.2, Reactivity Anomaly Check (CMPR-BNP-0PT-14.5.2-039-028040157) 2.8 0PT-50.0, Reactor Engineering Refueling Outage Testing (CMPR-BNP-0PT-50.0-047-

-028040158) 2.9 0PT-50.3, TIP Uncertainty Determination (CMPR-BNP-0PT-50.3-025--028040159) 2.10 0PT-90.2, Friction Testing of Control Rods (CMPR-BNP-0PT-90.2-032--028040220) 2.11 BNEI-0400-0035, Brunswick Unit 1 Cycle 23 Cycle Management Report, Revision 0.

2.12 0PT-14.1A, Control Rod Coupling Check and CRD Testing (CMPR-BNP-0PT-14.1A--

-027897887) 3.0 UFSAR Section 14.4.1, Item 1: Core Loading Verification A Core Loading Pattern Verification was performed per BSEP Engineering Procedure 0ENP-24.13, Core Verification (Reference 2.3). The core was verified to be loaded in accordance with the analyzed B1C23 core design.

4.0 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 Procedures 0PT-90.2,

Duke Energy, Nuclear Fuels Engineering, Brunswick Nuclear Design B1C23 Startup Report Page 3 of 7, Revision 0 Friction Testing of Control Rods (Reference 2.10) and 0PT-14.1A, Control Rod Coupling Check and CRD Testing (Reference 2.12). Control rods were verified to complete full travel without excessive binding or friction. In a prerequisite to 0PT-90.2, the reactor was observed to remain subcritical during the withdrawal of the most reactive rod per the BSEP Fuel Handling Procedure 0FH-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 0PT-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 core average 20% insertion time measured was 0.789 seconds which is faster than the analyzed nominal insertion limit of 0.829 seconds (Reference 2.1).

5.0 UFSAR Section 14.4.1, Item 3: Reactivity Testing Reactivity Testing consists of a shutdown margin (SDM) measurement, reactivity anomaly check, and measured critical keff 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 0PT-14.3.1, Insequence Critical Shutdown Margin Calculation (Reference 2.6). The cycle minimum SDM was determined to be 0.944% k/k compared to a predicted cycle minimum SDM value of 1.19%

k/k (Reference 2.11), resulting in an absolute difference of 0.246% k/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% k/k) from 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% k/k during the entire cycle. Since the cycle minimum SDM was determined to be 0.944% k/k for B1C23, the acceptance criterion is met.

b. Reactivity Anomaly A reactivity anomaly test was performed at near rated conditions (2921.6 MWt or 99.95% of rated power) per BSEP Periodic Test Procedure 0PT-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% k/k. The measured and predicted values for keff were 1.0025 and 1.0010 (Reference 2.11), respectively, an absolute difference of 0.15% k/k. This is within the 1% k/k acceptance requirement.

c. Cold Critical Eigenvalue (keff)

The measured BOC cold critical keff per BSEP Periodic Test Procedure 0PT-14.3.1, Insequence Critical Shutdown Margin Calculation (Reference 2.6), was inferred as 0.99299 by applying the period correction of -0.00052 to the nodal simulator code calculated k eff value of 0.99351 using

Duke Energy, Nuclear Fuels Engineering, Brunswick Nuclear Design B1C23 Startup Report Page 4 of 7, Revision 0 actual critical conditions as input. The predicted BOC cold critical k eff was 0.99250 (Reference 2.11) resulting in a measured to predicted absolute difference of 0.049% k/k. Therefore, per Technical Specification 3.1.2, the acceptance criterion requiring agreement within 1% k/k is met.

6.0 UFSAR Section 14.4.1, Item 4: 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 0PT-50.3, TIP Uncertainty Determination (Reference 2.9). Total radial TIP measurement uncertainty at the high-power testing plateau

(>90% CTP) was 0.887% and total nodal TIP measurement uncertainty was 1.713%. These results met the test acceptance criteria of 2.9% and 4.7%, respectively, and in accordance with 0PT-50.0, Reactor Engineering Refueling Outage Testing (Reference 2.8), 0PT-50.3 was not performed at the medium-power testing plateau.

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 0PT-50.3, TIP Uncertainty Determination (Reference 2.9). The radial deviation at the high-power testing plateau (>90% CTP) was 1.727%

and the nodal deviation was 2.898%. These results met the test acceptance criteria of 4.0% and 11.6%, respectively, and in accordance with 0PT-50.0, Reactor Engineering Refueling Outage Testing (Reference 2.8), 0PT-50.3 was not performed at the medium-power testing plateau.

c. Monitored Power Uncertainty Radial and nodal monitored power uncertainties were determined in accordance with BSEP Periodic Test Procedure 0PT-50.3, TIP Uncertainty Determination (Reference 2.9). The radial monitored power uncertainty at the high-power testing plateau (>90% CTP) was 2.295% and the nodal monitored power uncertainty was 2.575%. These results met the test acceptance criteria of 5.3% and 5.4%, respectively, and in accordance with 0PT-50.0, Reactor Engineering Refueling Outage Testing (Reference 2.8), 0PT-50.3 was not performed at the medium-power testing plateau.

d. Bundle Powers This analysis compares the MICROBURN-B2 predictions of bundle powers to the plant process computers measured bundle powers in accordance with BSEP Periodic Test procedure 0PT-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 2.00% at medium power (40% to 80% CTP). This result meets the test acceptance criteria of 8.9%.

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.

Duke Energy, Nuclear Fuels Engineering, Brunswick Nuclear Design B1C23 Startup Report Page 5 of 7, Revision 0

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 630.8 MWD/MTU on April 22, 2020, the predicted and core follow Hot Full Power Eigenvalues (k eff) were compared (Reference 2.8). The core follow keff was calculated as 1.0016 and the predicted keff was 1.00076. The absolute difference between the predicted and core follow values is 0.08% k/k which is within the 1% k/k reactivity anomaly requirements.

8.0 Summary Evaluation of the BSEP Unit 1, Cycle 23 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 B1C23. The UFSAR prescribed and additional tests met their acceptance criteria.

Duke Energy, Nuclear Fuels Engineering, Brunswick Nuclear Design B1C23 Startup Report Page 6 of 7, Revision 0 Attachment 1 to the B1C23 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 0.358 20% 36 1.080 0.860 50% 26 1.830 1.417 90% 06 3.350 2.626

Duke Energy, Nuclear Fuels Engineering, Brunswick Nuclear Design B1C23 Startup Report Page 7 of 7, Revision 0 Attachment 2 to the B1C23 Startup Report Core Monitoring Software Comparisons to Predictions Medium Power 40.6% CMWT, March 29, 2020 Thermal Limit POWERPLEX MICROBURN-B2 Absolute Acceptance On-Line Predicted Difference Criteria Monitoring CMFLCPR 0.762 0.774 -0.012 0.061 CMAPRAT 0.436 0.434 0.002 0.164 CMFDLRX 0.621 0.619 0.002 0.164 High Power 94.8% CMWT, March 31, 2020 Thermal Limit POWERPLEX MICROBURN-B2 Absolute Acceptance On-Line Predicted Difference Criteria Monitoring CMFLCPR 0.883 0.891 -0.008 0.041 CMAPRAT 0.681 0.676 0.005 0.109 CMFDLRX 0.763 0.762 0.001 0.109