ML20117N433

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Rev 0 to NF-908.03, Brunswick Unit 2,Cycle 12 Startup Rept
ML20117N433
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 04/30/1996
From: Geyer E, Siphers J, Thomas R
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20117N432 List:
References
NF-908.03, NF-908.03-R, NF-908.03-R00, NUDOCS 9606200404
Download: ML20117N433 (10)
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Text

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i CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2Cl2 Startup Report Page 1 of 9, Revision 0 t

BRUNSWICK UNIT 2, CYCLE 12 STARTUP REPORT t

April 1996 i

8 f/fff, Prepared by: [ Roger L ThogJr Reviewed by:

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ErifJ. Geyer' Approved b. b M Z2- $Q John T. Siphlrs K:G'ROJECN12C l 2'STA RTRIrnSIL REP.WP 9606200404 960606 PDR ADOCK 05000324 p

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CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2Cl2 Stanup Report Page 2 of 9, Revision 0 1.0 Introduction This repon summarizes observed data from the initial Brunswick Unit 2 Cycle 12 (B2Cl2) startup tests. The Cycle 12 core employs the new gel 3 fuel type, which among other design differences represents a change from an 8x8 to 9x9 fuel rod array.

Pursuant to the requirements of Section 6.9.1.1 of the Unit 2 Technical Specifications, 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: ".. installation of fuel that has a different design or has been manufactured by a different fuel supplier."

This report shall include results of startup tests following core reloading as described in the UFSAR.

2.0 UFSAR Section 14.4.1. Item 1: Core Loading Verification A Core Loading Pattern Verification was performed per BNP Engineering Procedure ENP-24.13. " Core Verification." The core was verified to be loaded in accordance with the Recommended Full Core Loading Pattern.

3.0 UFS AR Section 14.4.1. Item 2: TIP Onerability and Core Power Svmmetry a.

TIP Uncertainty A TIP uncertainty determination was completed according to BNP Engineering Procedure PT 50.3 "Tip Reproducibility and Uncertainty Determination." The acceptance criteria for this test requires the TIP Total Noise Uncertainty to be s7.10%. The measured uncertainty was 1.72%, thus meeting the criteria.

b.

Core Power Symmetry Core power symraetry is indirectly verified via the standard traversing in-core probe (TIP) uncertainty measurement performed per PT 50.3, described in Section 3.0.a.

Direct power symmetry measurement utilizing computed bundle powers is no longer performed at Brunswick with the improved POWERPLEX core monitoring system.

POWERPLEX methodology does not require core symmetry. Therefore, the Core Power Syrnmetry Test was replaced by a more appropriate Predicted Versus Measured Bundle Power Test. The test results and acceptance criteria are provided in c. below.

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CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2Cl2 Startup Report Page 3 of 9, Revision 0 3.0 UFS AR Section 14.4.1. Item 2: TIP Ooerability and Core Power Svmmetrv (cont...)

c.

Predicted Versus Measured Bundle Powers BNP Engineering procedure PT 50.0," Reactor Engineering Refueling Outage Testing,"

was revised to replace the Core Power Symmetry Test ( 15% symmetric bundle power agreement acceptance criteria) with a Predicted Versus Measured Bundle Powers test.

This test compares the MICROBURN-B design code's calculation of predicted bundle powers to the plant process computer's measured bundle powers. The comparison must verify that the absolute difference between measured and predicted bundle powers meets the acceptance criteria of s8.64%. Bundles located in peripheral control cells or uncontrolled peripheral locations are excluded.

The acceptance criteria was met with the maximum absolute difference measured as 2.99%.

4.0 UFSAR Section 14.4.1. Item 3: Control Rod Mobility

~

l 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 BNP Engineering Procedure PT 90.2,

" Friction Testing of Control Rods." Control rods were verified to complete full travel without excessive binding or friction and the reactor was observed to remain suberitical during the withdrawal of each control rod.

b.

Scram Time Testing Scram Time Testing was performed for each control rod prior to exceeding 40% power per BNP Engineering Procedure PT 14.2.1," Single Rod Scam Insertion Times Test."

The acceptance criteria for this test are found in Technical Specifications 3.1.3.2,3.1.3.3, and 3.1.3.4. The maximum 90% insertion time was measured as 2.898 seconds meeting the 7.0 seconds acceptance criteria of Technical Specification 3.1.3.2. Acceptance criteria for the Core Average Scram Insertion and Maximum Average 2x2 Scram Insertion times were also met as illustrated in Attachment 1.

The average 20% insertion time measured from the low power testing was 0.839 seconds, thus meeting the ODYN Option B time requirement of 0.861 seconds. ODYN Option B MCPR limits were therefore installed at BOC.

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CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2Cl2 Startup Report Page 4 of 9, Revision 0 5.0 UFSAR Section 14.4.1. Item 4: Reactivity Testing Reactivity Testing consists of a shutdown margin measurement, reactivity anomaly check, and measured critical K,y comparison to predicted values. The results of these tests are provided below with the acceptance criteria.

a.

Shutdown Margin Shutdown margin measurements were performed per BNP Engineering Procedure PT 14.3.1,"Insequence Critical Shutdown Margin Calculation." The initial BOC shutdown margin was measured as 1.958% Ak/k compared to a predicted value of 1.768% Ak/k, an absolute difference of 0.1907c Ak/k. The acceptance criteria for i

minimum shutdown margin is defined in Technical Specification 3.1.1, which requires the shutdown margin be 2 0.387c Ak/k for the entire cycle. To calculate the minimum shutdown margin for the cycle, the maximum predicted decrease in shutdown margin over the cycle relative to BOC,-0.6557c Ak/k, was applied to the BOC measured shutdown margin. This resulted in an inferred minimum shutdown margin for Cycle 12 of 1.3037c Ak/k. Therefore, the acceptance criteria is met.

b.

Cold Critical Eigenvalue (K,y)

The cold critical K,y was calculated by applying the 0.190% Ak/k reactivity difference between measured and predicted shutdown margin to the predicted K,y of 1.00237. The cold critical K,y is therefore 1.00427. Since the acceptance criteria requires predicted versus measured agreement within 17c Ak/k, and the deviation was measured at 0.1907c Ak/k, the acceptance criteria is met.

i c.

Reactivity. Anomaly A reactivity anomaly test was performed at rated conditions per BNP Engineering Procedure PT 14.5.2," Reactivity Anomaly Check." The acceptance criteria is defined by Technical Specification 3.1.2 which requires the difference between actual and predicted control rod density (CRD) not exceed l'7c Ak/k. The measured and predicted values for CRD were 0.038 and 0.029, respectively, an absolute difference of 0.009.

Since for Cycle 1217c Ak/k is equivalent to 0.027 CRD, the acceptance criteria is met.

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CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2Cl2 Startup Report Page 5 of 9, Revision 0 I

6.0 Additional Testing Results j

Additional key testing and checks beyond those specified in the UFSAR are performed during initial startup and power ascension. The results are provided below.

I Core Monitoring Software Comparisons to Design Code a.

i Thermal limits calculated by the online POWERPLEX Core Monitoring Software System were compared to those calculated by the MICROBURN-B design code at medium and high power levels. The results of these comparisons and the POWERPLEX statepoints are provided as Attachment 2. The acceptance criteria specified in PT 50.0 require the two codes' thermal limits agree within 0.15 for medium power testing and 0.10 for high power testing. The acceptance criteria were met.

b.

Hot Full Power Eigenvalue After establishing a sustained period of full power equilibrium operation the design and core follow Hot Full Power Eigenvalues (K,y) are compared. At 267 MWD /MT the core follow K,y was calculated as 1.00393 and the design K,y is 1.00470. The difference between the core follow and design values is -0.077% Ak/k of reactivity which is well within 1% Ak/k reactivity anomaly requirements.

7.0 Summarv l

Evaluation of the Brunswick Unit 2, Cycle 12 startup data concludes the core has been loaded properly, the behavior of the new GE13 fuel design can be accurately predicted, and the core 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 shutdown margin demonstration indicates adequate shutdown margin will exist throughout B2Cl2.

All prescribed and additional tests met their acceptance criteria.

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CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2Cl2 Startup Report Page 6 of 9, Revision 0 to the B2C12 Startup Report Results of Control Rod Scram Time Testine Core Average Scram insertion Time Technical Specification 3.1.3.3 Insertion Position / Notch Tech Spec Limit Average Aleasured (sec)

Insertion Time (sec) 57c 46 0.358 0.299 207c 36 1.096 0.839 507c 26 1.860 1.387 907c 6

3.419 2.582 hlaximum Average 2x2 Scram insertion Time Technical Specification 3.1.3.4 Insertion Position / Notch Tech Spec Limit Average Aleasured (sec)

Insertion Time (sec) 57c 46 0.379 0.311 207c 36 1.I62 0.879 507c 26 1.971 1.479 907c 6

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CP&1. Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2Cl2 Startup Report Page 7 of 9 Revision 0 to the B2C12 Startup Report Core Monitoring Software Comparisons to Design Code Medium Power Testing Plateau 70.3% CMWT,96 MAR 2316:04:35,49 MWI)/MTU Thermal Limit POWERPLEX MICROBURN-B Difference Acceptance On-Lir e Design Code Criteria Monitoring MFLCPR 0.853 0.840 0.013 0.15 MAPRAT 0.789 0.742 0.047 0.15 MFLPD 0.562 0.572

-0.010 0.15 Iligh Power Testing Plateau 92.7% CMWT,96M

  • it ?5,02:02:40,74 MWDiMTU Thermal Limit POWERPLEX MILKOBURN-B Difference Acceptance On-Li;.c Design Code Criteria Monitoring MFLCPR 0.796 0.807

-0.011 0.10 MAPRAT 0.862 0.857 0.005 0.10 j

MFLPD 0.772 0.729 0.043 0.10 K :\\PRoJEUN12Cl 2\\STA RTRIrnSU_RD'.WP

CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2Cl2 Startup Report Page 8 of 9. Revision 0 to the B2C12 Startup Report (cont...)

Medium Power Testine Plateau Statenoint Renort H M: WICK 2 W 9612 9(M AR23 - 16.04.35 49 MWD /MTU TElGP-CDlFIN LEV-JAN96 coeE FEGrokMANCE LOG 5aCRT EDli TLOO EOC to ECC 2000 MWD /MT 00Yh9 POW DEP MCPR CALCULATION TYPE NORMAL CONVERGENCE TIGHT SYMMETRY F L:LL CTP CALCULATICN HE AT P AL ANCE CYCLE 12 STATE LONDIT IONS FLOW RATE CORE PARAMETER 5 NUCLE AR LIMIT S LOCATION GMWE 568.44 WT 44.2 CME 0 0.3023 P-FC5 0.lB 27 20-15 CMWT 1711.6 (70.3%) WISLB 43.63 CAE0 0 lf.05 FCEB

!.729 F-C 970.5 F5!A WTFLAG 2 CACA 0.1024 CMPF 2.337 31-26-15 DHS 2t.61 WFW 6.99 CAVF 0.4494 CMFLCFR 0.e53 29 24 WT 44.23 (b7.4%) WD 18.67 CAPO 34.5448 P-1.512 f-1.360 CRD 0. 0'i l RWL 187.1539 CMAFRAT 0.7E9 33-26 15 CYCEXP 49 MWD /MTU ERAT 10 0.93 CDLP 7.5057 F-0.845 F-0.846 MEASUEED/ CALCULATED LFRM READlhGS DFCC 12.4093 CMFLFD 0.562 09 20-10 avg C.lR1 MAX:

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10 11 12 AXIAL REL POWER 0.49 1.05 1.14 1.14 1.23 1.27 1.27 1.22 1.11 0.93 0.75 0.42 REGICh EEL POWER 0.99 1.02 0.87 1.03 1.28 1.03 0.87 1.02 0.E8 klNa EEL POWER 1.34 1.35 1.24 1.17 1.09 1.04 0.68 APEM GAFS 0.99 0.99 0.98 1.00 0.96 0.96

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t CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.03 B2C12 Startup Report Page 9 of 9, Revision 0 to the B2Cl2 Startup Report l

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ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 NRC DOCKET NO. 50-324 OPERATING LICENSE NO. DPR-62 BRUNSWICK UNIT 2, CYCLE 12 STARTUP REPORT LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Carolina Power & Light Company in this document. Any other actions discussed in the submittal represent intended or planned actions by Carolina Power & Light Company. They are descr. bed to the NRC for the NRC's information and are not regulatory commitments. Please riotify the Manager-Regulatory Affairs at the Brunswick Nuclear Plant of any questions regarding this document or any associated regulatory commitments.

Committed Commitment date or outage

  • 1.

Incorporate the changes resulting from these Unit 2 Cycle 12 The next start-up tests into Sections 14.4.1 and 14.4.2.2 of the scheduled Updated FSAR.

revision to the Updated

FSAR, (Currently 5/97) i
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