BSEP 09-0066, Cycle 19 Startup Report

Cycle 19 Startup Report

ML091800019
Person / Time
Site:	Brunswick
Issue date:	06/25/2009
From:	Mentel P Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 09-0066
Download: ML091800019 (11)

Text

Progress Energy JUN 2 5 200 SERIAL: BSEP 09-0066 U. S. Nuclear Regulatory Commission ATTFN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit No. 2 Docket No. 50-324/License No. DPR-62 Cycle 19 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," Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., submits the enclosed Brunswick Unit 2 Cycle 19 Startup Report. The report is required as a result of installation of fuel that was manufactured by a different fuel supplier (i.e., the AREVA ATRIUM-10 fuel type) during the spring 2009 refueling outage.

No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Mr. Gene Atkinson, Supervisor - Licensing/Regulatory Programs, at (910) 457-2056.

Sincerely, Phyllis N. Mentel Manager - Support Services Brunswick Steam Electric Plant WRM/wrm

Enclosure:

Brunswick Unit 2, Cycle 19 Startup Report Progress Energy Carolinas, Inc.

Brunswick Nuclear Plant POBox 10429 Southport, NC 28461

Document Control Desk BSEP 09-0066 / Page 2 cc (with enclosure):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510

BSEP 09-0066 Enclosure Brunswick Unit 2 Cycle 19 Startup Report

BRUNSWICK UNIT 2, CYCLE 19 STARTUP REPORT June 2009 Hahn, Gregory Prepared By

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Prepared by: 2009.06.12 10:00:06 Gregory C. Hahn (BWR Fuel Engineering)

Thomas, Roger 2009.06.12 13:09:07 -04'00' Reviewed by:

Roger Thomas (BWR Fuel Engineering)

Westermark, Hans Reviewed By Reviewed by: 2009.06.15 07:46:19 -04'00' ".

Hans Westermark (BNP Reactor Engineering) 0* Oigned Dgitally byMur ray,WOiim0.[Bilr M urray, W illiam R. (Bi~l*\

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, nuU.r, Reviewed by: t Dl: 2W*9W5 15oll.32:$4-04W'00 William Murray (Regulatory Affairs)

Blom, Michael Management Approval..

Approved by: 2009.06.15 10:32:44 -04'00' Mike Blom (Superintendent, PWR Fuel Engineering)

Progress Energy Nuclear Fuels Management & Satbty Analysis Section B2C19 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 19 (B2C19) startup tests. The Cycle 19 core represents the first loading of thlc AREVA ATRIUM-I 0 fuel type in Unit 2. 238 ATRIUM-10 fuel assemblies have been loaded.

Pursuant to the requirements of Section 13.4.2.1 of the BSEP I & 2 Updated Final Safety Analysis Report (UFSAR), 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) 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" 2.4 OFH- 1I, "Refueling" 2.5 OPT-14.2.1, "Single Rod Scram Insertion Times Test" 2.6 OPT-14.3.1, "Insequence Critical Shutdown Margin Calculation" 2.7 OPT-14.5.2, "Reactivity Anomaly Check" 2.8 OPT-50.0, "Reactor Engineering Refueling Outage Testing" 2.9 OPT-50.3, "TIP Reproducibility And Uncertainty Determination" 2.10 OPT-90.2, "Friction Testing of Control Rods" 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." The core was verified to be loaded in accordance with the analyzed B2C19 core design.

Progress Energy Nuclear Fuels Management & Sat'ety Analysis Section B2C19 Startup Report Page 3 of 8, Revision 0 4.0 UFSAR Section 14.4.1, Item 4A: "liPOperability and Bundle Power-tvaluation

a. TI P 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." Total radial TIP measurement uncertainty at high core thermal power (CTP) (>80% CTP) was 1.007% and total nodal TIP measurement uncertainty was 1.768%. These radial and nodal uncertainties were also determined at medium core thermal power (40% to 80% CTP) and were 0.909% and 1.594%,

respectively. All 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." The radial deviation at high core thermal power (>80% CTP) was 1.778% and the nodal deviation was 3.945%. These radial and nodal deviations were also determined at medium core thermal power (40% to 80% CTP) and were 1.971% and 6.129%, respectively. All 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." The radial monitored power uncertainty at high core thermal power (>80% CTP) was 2.288% and the nodal monitored power uncertainty was 2.992%. These radial and nodal uncertainties were also determined at medium core thermal power (40% to 80% CTP) and were 2.644% and 3.380%, respectively. All 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." Bundles located in peripheral control cells or uncontrolled peripheral locations are excluded. The maximum radial difference was calculated to be 4.67% at medium power (40% to 80%

CTP). All results met the test acceptance criteria.

IProgrcs,' Energ5y Nuclear Fuels Management & Safety Analysis Section B2C 9 Starnup Report Page 4 of 8, Revision 0 5.0 UFSAR Section 14.4.1, ltem 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." Control rods were verified to complete full travel without excessive binding or friction. In a pre-requisite to OPT-90.2, the reactor was observed to remain subcritical during the withdrawal of the most reactive rod in BSEP Fuel Handling Procedure OFH-11, "Refueling."

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 Scramn Insertion Times Test."

The acceptance criteria for this test are found in Technical Specification 3.11.4. All 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 I of this report.

The average 20% insertion time measured from the low power testing was 0.816 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 kerr 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." The cycle minimum SDM was determined to be 1.0515% Ak/k compared to a predicted cycle 'Minimum SDM value of 1.09% Ak/k, an absolute difference of 0.0385% 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 defined in Technical Specification 3.1.1.1, which requires the SDM be Ž> 0.38% Ak/k during the entire cycle.

Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C 19 Startup Report Page 5 of 8, Revision 0 Since for B2CI9 the cycle minimum SDM was determined to be 1.0515% Ak/k, the acceptance criterion is met.

b. teactivity Anomaly A reactivity anomaly test was performed at near rated conditions (2901.8 MWt or 99.3%)

per BSEP Periodic Test Procedure OPT-14.5.2, "Reactivity Anomaly Check." The acceptance criterion is defined by Technical Specification 3.1.2, which requires that the reactivity difference between monitored and predicted core ker- be within +/-1% 9Ak. The measured and predicted values for keff were 0.9980 and 1.0025, respectively, a difference of 0.45% Ak/k. This is within the +1% Ak requirement.

c. Cold Critical Eigenvalue (kcrr)

The measured BOC cold critical krt-f per BSEP Periodic Test Procedure OPT-] 4.3.1, "Insequence Critical Shutdown Margin Calculation" was inferred as 0.99604 by nodal simulator code calculations with actual critical conditions as input (including period correction). The predicted BOC cold critical kff was 0.99600 resulting in a measured to predicted difference of 0.004% 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. The results of these comparisons and the POWERPLEX statepoints are provided as Attachment 2. All results met the test acceptance criteria.

b. Hot Full Power Eigenvalue After establishing a sustained period of full power equilibrium operation at 163.0 MWD/MTU on May 5, 2009, the predicted and core follow Hot Full Power Eigenvalues (ken-f) were compared. The core follow klt, was calculated as 0.9981 and the predicted ken- was 1.00235. The difference between the predicted and core follow values is 0.425% Ak/k which is within the +/-1% Ak/k reactivity anomaly requirements.

Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C 19 Startup Report Page 6 of 8, Revision 0 8.o Summnary Evaluation of the BSEP Unit 2, Cycle 19 startup data concludes-the core has been loaded properly and is operating as expected. The startup and ifitia1 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 B2C 19. All. UFSAR prescribed and additional tests met their acceptance criteria.

. . a Ilrogress Energy Nuclear Fuels Management & Safety Analysis Section 132C 19 Startup Report Page 7 of 8, Revision 0 Attachment 1 to the B2C19 Startup Report Results ot Control Rod Scram Time TestinE 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 0.328 20% 36 1.080 1.003 50% 26 1.830 1.685 90% 06 3.350 3.083

Progress Energy Nuclear Fuels Management & Safety Analysis Section B2C 19 Startup Report Page 8 of 8, Revision 0 Attachment 2 to the B2C19 Startup Report Core Monitoring Software Comparisons to Predictions Medium Power Testing Plateau 66.9% CMWT, May 01, 2009 Thermal Limit POWERPLEX *MICROBURN-B2 Absolute On-Line Predicted Difference Monitoring CMFLCPR 0.741 0.753 0.012 CMAPRAT 0.746 0.767 0.021 CMFDLRX 0.697 0.724 0.027 CMFLPD 0.516 0.531 0.015 High Power Testing Plateau 99.3% CMWT, May 05, 2009 Thermal Limit POWERPLEX MICROBURN-B2 Absolute On-Line Predicted Difference Monitoring CMFLCPR 0.842 0.853 0.011 CMAPRAT 0.810 0.814 0.004 CMFDLRX 0.892 0.896 0.004 CMFLPD 0.698 0.703 0.005