BSEP 15-0048, Cycle 22 Startup Report

Cycle 22 Startup Report

ML15201A153
Person / Time
Site:	Brunswick
Issue date:	06/29/2015
From:	Pope A Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 15-0048
Download: ML15201A153 (11)

Text

Brunswick Nuclear Plant

@DUKE ENERGY. P.O. Box 10429 Southport, NC 28461 JUN 2 9 2015 Serial: BSEP 15-0048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit No. 2 Renewed Facility Operating License No. DPR-62 Docket No. 50-324 Cycle 22 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, Inc. is submitting the enclosed Brunswick Unit 2, Cycle 22 Startup Report, dated June 2015. The report is required as a result of the first loading of AREVA ATRIUM 11 Lead TestAssemblies during the Spring 2015 refueling outage.

No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

Sincerely, Anr ette H. Pope Director - Organizational Effectiveness Brunswick Steam Electric Plant

Enclosure:

Brunswick Unit 2, Cycle 22 Startup Report, June 2015 L~Y' gA~~ti~4L

U.S. Nuclear Regulatory Commission Page 2 of 2 cc (with enclosure):

U.S. Nuclear Regulatory Commission, Region II ATTN: Mr. Victor M. McCree, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission ATTN: Ms. Michelle P. Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510

BSEP 15-0048 Enclosure Brunswick Unit 2, Cycle 22 Startup Report, June 2015

BRUNSWICK UNIT 2, CYCLE 22 STARTUP REPORT June 2015 Prepared by:

Ryan Wells (BWR Fuel Engineering)

Reviewed by:

6yan Wester (Reactor Systems - BNP)

Approved by:

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 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 22 (B2C22) startup tests. The Cycle 22 core represents the first loading of the AREVA ATRIUM 11 fuel type in Unit 2. Eight (8) ATRIUM 11 lead test assemblies have been loaded (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 24 2.2 BSEP Technical Specifications 2.3 OENP-24.13, "Core Verification" (DEP RMS 5638826) 2.4 0FH-11, "Refueling" (DEP RMS 5668959) 2.5 OPT-14.2.1, "Single Rod Scram Insertion Times Test" (DEP RMS 5694473) 2.6 OPT-14.3.1, "Insequence Critical Shutdown Margin Calculation" (DEP RMS 5694466) 2.7 OPT-14.5.2, "Reactivity Anomaly Check" (DEP RMS 5694449) 2.8 0PT-50.0, "Reactor Engineering Refueling Outage Testing" (DEP RMS 5694441) 2.9 OPT-50.3, "TIP Uncertainty Determination"(DEP RMS 5694431) 2.10 OPT-90.2, "Friction Testing of Control Rods" (DEP RMS 5694469) 2.11 CMR U2 CYCLE 22, "Brunswick Unit 2, Cycle 22, Cycle Management Report,"

Revision 0.

2.12 OPT-14.1A, "Control Rod Coupling Check and CRD Testing" (DEP RMS 5673880) 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 B2C22 core design.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Startup Report Page 3 of 8, Revision 0 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 OPT-90.2, "Friction Testing of Control Rods" (Reference 2.10) and OPT-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 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-1 1, "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 core average 20% insertion time measured was 0.795 seconds which is faster than the analyzed nominal insertion limit of_< 0.862 seconds.

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 OPT-14.3.1, "Insequence Critical Shutdown Margin Calculation" (Reference 2.6). The cycle minimum SDM was determined to be 2.097% Ak/k compared to a predicted cycle minimum SDM value of 1.80% Ak/k (Reference 2.11), resulting in an absolute difference of 0.297% 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 the SDM at beginning-of-cycle (BOC). The acceptance criterion for minimum SDM is

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Startup Report Page 4 of 8, Revision 0 defined in Technical Specification 3.1.1, which requires the SDM be Ž 0.38% Ak/k during the entire cycle. Since the cycle minimum SDM was determined to be 2.097%

Ak/k for B2C22, the acceptance criterion is met.

b. Reactivity Anomaly A reactivity anomaly test was performed at near rated conditions (2919.8 MWt or 99.9%

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 keff were 0.9999 and 1.0000 (Reference 2.11), respectively, an absolute difference of 0.01% Ak/k.

This is within the +/-1% Ak/k acceptance 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 0.99793 by applying the period correction of -0.00053 to the nodal simulator code calculated keff value of 0.99846 using actual critical conditions as input. The predicted BOC cold critical keff was 0.9950 (Reference 2.11) resulting in a measured to predicted absolute difference of 0.29% Ak/k. Therefore, per Technical Specification 3.1.2, the acceptance criterion requiring agreement within +/-1% Ak/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 OPT-50.3, "TIP Uncertainty Determination" (Reference 2.9). Total radial TIP measurement uncertainty at the medium-power testing plateau (40% to 80% CTP) was 1.910% and total nodal TIP measurement uncertainty was 2.468%. These results met the test acceptance criteria of

• 2.9% and < 4.7%, respectively, and in accordance with OPT-50.0, "Reactor Engineering Refueling Outage Testing" (Reference 2.8), OPT-50.3 was not performed at the high-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 OPT-50.3, "TIP Uncertainty Determination" (Reference 2.9). The radial deviation at the medium-power testing

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Startup Report Page 5 of 8, Revision 0 plateau (40% to 80% CTP) was 2.316% and the nodal deviation was 3.500%. These results met the test acceptance criteria of

• 4.0% and _<11.6%, respectively, and in accordance with OPT-50.0, "Reactor Engineering Refueling Outage Testing" (Reference 2.8), OPT-50.3 was not performed at the high-power testing plateau.

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 medium core thermal power (40% to 80% CTP) was 2.245% and the nodal monitored power uncertainty was 2.809%.

These results met the test acceptance criteria of

• 5.3% and

• 5.4%, respectively, and in accordance with OPT-50.0, "Reactor Engineering Refueling Outage Testing" (Reference 2.8), OPT-50.3 was not performed at the high-power testing plateau.

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 B SEP 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 3.70% 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.

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 188.6 MWD/MTU on April 13, 2015, the predicted and core follow Hot Full Power Eigenvalues (keff) were compared (Reference 2.8). The core follow keff was calculated as 0.9999 and the predicted kff was 0.9997. The absolute difference between the predicted

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Startup Report Page 6 of 8, Revision 0 and core follow values is 0.02% Ak/k which is within the +/-1% Ak/k reactivity anomaly requirements.

8.0 Summary Evaluation of the BSEP Unit 2, Cycle 22 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 B2C22. The UFSAR prescribed and additional tests met their acceptance criteria.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Startup Report Page 7 of 8, Revision 0 Attachment 1 to the B2C22 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 0.331 20% 36 1.080 0.943 50% 26 1.830 1.567 90% 06 3.350 2.882

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Startup Report Page 8 of 8, Revision 0 Attachment 2 to the B2C22 Startup Report Core Monitoring Software Comparisons to Predictions Medium Power 58.6% CMWT, April 7, 2015 Thermal Limit POWERPLEX MICROBURN-B2 Absolute Acceptance On-Line Predicted Difference Criteria Monitoring CMFLCPR 0.669 0.670 0.001 < 0.061 CMAPRAT 0.461 0.462 0.001 < 0.164 CMFDLRX 0.564 0.564 0.000 < 0.164 High Power 99.8% CMWT, April 13, 2015 Thermal Limit POWERPLEX MICROBURN-B2 Absolute Acceptance On-Line Predicted Difference Criteria Monitoring CMFLCPR 0.843 0.850 0.007 < 0.041 CMAPRAT 0.740 0.741 0.001 < 0.109 CMFDLRX 0.825 0.828 0.003 < 0.109