ML20134J328
| ML20134J328 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/29/1997 |
| From: | Geyer E, Siphers J, Thomas R CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20134J310 | List: |
| References | |
| NF-908.02, NF-908.02-R02, NF-908.02-R2, NUDOCS 9702120088 | |
| Download: ML20134J328 (10) | |
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CP&L h:ucar Fuels Management & Safety Analysis Secnon BICil Startup Report File: NF-908.02 Page I of 10. Revision 0 l
l BRUNSWICK UNIT 1, CYCLE 11 NEUTRONICS STARTUP REPORT l
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4 January 1997 1
Prepared by: M/I //MN)
[ Roger . TI)(nas, Jf. '
Reviewed by: bd u i/2.9/97 7 Eric J. dey'er ' l l
Approved b & //f 7b7 John T7Siphers I perintendent, BWR Fuel Analysis l i
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9702120088 970205
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CP&L Nuclear Puels Management & safety Analysis section File: NF-908.02 MCII startup Repon Page 2 of 10. Revision 0 1.0 Introduction This repon summarizes observed data from the initial Brunswick Unit 1, Cycle 11 (BICI1) stanup tests. The Cycle 11 core employs the new GE13 fuel type, which among
'other design differences represents a change from an 8x8 to 9x9 fuel rod array. Also for Cycle 11, core rated thermal power is being increased from 2436 to 2558 MWt.
Pursuant to the requirements of Section 6.9.1.1 of the Unit 1 Technical Specifications, a summary repon of plant startup and power escalation testing shall be submitted to the NRC should any one of four conditions occur. Conditions (2) - (4) apply:
(2): " amendment to the license involving a planned increase in power level",
and (3): " installation of fuel that has a different design or lias been manufactured by a different fuel supplier", and (4): " modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant."
This repon shall include results of neutronics related stanup tests following core reloading as described in the UFSAR. This repon is not intended as a stand-alone document, but will be included with the power up-rate startup report for Cycle 11.
2.0 UFSAR Section 14.4.1. Item 1: Core Loading Verification A Cost 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 Operability and Core Power Symmetry
- a. TIP Uncertainty, A TIP uncenainty determination was completed according to BNP Engineering Procedure Of"T-50.3,"Tip Reproducibility and Uncertainty Determination." The acceptance criterion for this test requires the TIP Total Noise Uncenainty to be 57.1%. The measured uncertainty was 2.47%, thus meeting the criteria.
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CP&L Nuclear Fuels Management & safety Analysis section File: NF-908.02 l BiCll Startup Report Page 3 of 10. Revision 0 i
i b. Core Power Symmetry Core power symmetry is indirectly verified via the standard traversing in-core probe (TIP) uncertainty measurement performed per OPT-50.3, described in Section 3.0.a. -
Direct power symmetry mearai anent utilizing computed bundle powers is no longer j performed at Brunswick with the improved POWERPLEX core monitoring system.
i POWERPLEX methodology does not require core symmetry. Therefore, the Core Power S
Symmetry Test was replaced by a more appropriate Predicted Versus Measured Bundle Power Test. The test results and acceptance criteria are provided in c. below. ;
! c. Predicted Versus Measured Bundle Powers BNP Engineering procedure OPT-50.0, " Reactor Engineering Refueling Outage Testing,"
i was revised to mplace the Core Power Symmetry Test ( 15% symmetric bundle power agreement acceptance criterion) 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 criterion of 58.64%. Bundles located in peripheral control cells or uncontrolled peripheral locations are excluded.
The acceptance criteria was met with the maximum absolute difference measured as ;
3.26%.
I 4.0 UFSAR Section 14.4.1. Item 3: 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 BNP Engineering 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 BNP Fuel Handling Procedure OFH-11," Refueling."
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c CP&L Nuclear Fuels Management & safe'y Analysis section File: NF-908 02 B1Cil startup Report Page 4 of 10. Revision v 4.0 UFS AR Section 14.4.1. Item 3: Control Rod Mobility (cont...)
- b. Scram Time Testing
' Scram Time Testing was performed for each control rod prior to exceeding 40% power per BNP Engineering Procedure OPT-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 2.892 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 0.815 seconds, I thus meeting the ODYN Option B time requirement of 0.861 seconds. ODYN Option B MCPR limits were therefore installed following the test.
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 Kea 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 OPT-14.3.1, "Insequence Critical Shutdown Margin Calculation." The initial BOC shutdown margin was measured as 1.362% Ak/k compared to a predicted value of 1.13% Ak/k, an absolute difference of 0.232% Ak/k. The acceptance criterion for minimum shutdown margin is defined in Technical Specification 3.1.1, which requires the shutdown margin be 2: 0.38% 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,-0.03% Ak/k (R), was applied to the BOC measured shutdown margin. This resulted in an inferred minimum shutdown margin for Cycle 11 l of 1.332% Ak/k. Therefore, the acceptance criterion is met.
- b. Cold Critical Eigenvalue (Keg)
The measured cold critical Kea was inferred as 1.00475 by nodal simulator code calculations with actual critical conditions as input. The predicted cold critical R;, was 1.00237 giving a measured vs. predicted difference of-0.238% Ak/k. Theref.;re, the acceptance criterion requiring agreement within 11% Ak/k is met. I K:\PRoJECnB I Cl l\STARTRimsU_ REP. doc
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i CP&L Nuclear Fuels Management & safety Analysis section File: NF-908.02 i i BICll startup Report Page 5 of 10. Revision 0
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5.0 UFSAR Section 14.4.1. Item 4: Reactivity Testina (cont...)
- c. Reactivity Anomaly A reactivity anomaly test was performed at near rated (2435.4 MWt or 95.2%) conditions per BNP Engineering Procedure OPT-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 1% Ak/k.
The measured and predicted values for CRD were 0.%8 and 0.056, respectively, an absolute difference of 0.012. Since for Cycle 11,1% Ak/k is equivalent to 0.035 CRD, the acceptance criterion is met.
6.0 Additional Testine Results As a matter of course, key testirs c.nd checks beyond those specified in the UFSAR are performed during initial startup and power ascension. These " standard" tests are described in items a. and c. below. Because of Cycle 1l's increase in rated power, three hold-points were added between 2436 and 2558 MWt to perform thermal limits checks.
Results of these additional tests are provided in item b. below,
- a. Core Monitoring Software Comparisonc to Design Code 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 Attxhment 2. The acceptance criteria specified in 017f-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. Hold-Point Thermal Limits Checks i
For Cycle 11 three hold-points were added to perform additional testing: Hold Point 1. - l 95.2% (2436 MWt), Hold Point 2. - 97.0% (2481 MWt), and Hold Point 3. - 100% i (2558 MWt). I Hold Point l'. acceptance criteria required that all measured thermal limits be < 0.98.
Hold Point 2. acceptance criteria required that all measured thermal limits be < 0.98 and !
the diff:rence between measured and predicted thermal limits be < 0.10. i l
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- CP&L Nuclear Fuels Management & safety Analysis section File
- NF-908.02
' f B tCll startup Report Page 6 of 10. Revision 0
- 6.0 Additional Testing Results (cont...) ,
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- Hold Point 3. acceptance criteria required that all measured thermal limits be < 0.98 ead !
j the difference between measured and predicted thermal limits be < 0.10. l
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4 At the time of this repon, Cycle 11 was limited to 95% power due to unresolved licensing !
i analyses connected to the uprated power of 2558 MWt. Therefore, only Hold Point 1. has :
{ been successfully completed with all measured thermal limits being <0.98. When Hold i Points 2. and 3. are reached and successfully completed, the additional information will !
be prepared. I
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i l c. Hot Full Power Eigenvalue l J
l After establishing a sustained period of near full power (2436 MWt) equilibrium ,
operation, the design and core follow Hot Full Power Eigenvalues (Ken) are compared. l At 203 MWD /MT the core follow Ken was calculated as 1.00392 and the design Kerr is 1.00384. The difference between the core follow and design values is -0.008% Ak/k which is well within the 11% Ak/k reactivity anomaly requirements.
i 7.0 Summary i Evaluation of the Brunswick Unit 1, Cycle 1I stanup 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 at the reduced power of 95% of the new rated power l of 2558 MWt. The startup and initial operating conditions and parameters compare well l 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 BICI1. All l prescribed and additional tests, which were completed, met their acceptance criteria.
Additional testing at Hold Points 2. and 3, as described in Section 6.0, Item b. of this report, will be completed when further power escalation is permissible. At the conclusion of this additional testing, the results will be reported in accordance with Section 6.9 of the Technical Specifications.
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CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.02 BlC11 Startup Report Page 7 of 10. Revision 0 Attachment I to the B1C11 Startup Report Results of Control Rod Scram Time Testine ;
Core Average Scram Insertion Time i Technical Specification 3.1.3.3 Insertion Position / Notch Tech Spec Limit Average Measured ,
(sec) Insertion Time l p(sec) l 5% 46 0.358 0.305 20 % 36 1.096 0.815 50 % 26 1.860 1.340 90 % 6 3.419 2.455 Maximum Average 2x2 Scram Insertion Time Technical Specification 3.1.3.4 Insertion Position / Notch Tech Spec Limit Average Measured (sec) Insertion Time ,
(sec) l 5% 46 0.379 0.312 20 % 36 1.162 0.843 50 % 26 1.971 1.399 90 % 6 3.624 2.561 I
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CP&L Nuclear Fuels Management & Safety Analysis Section File: NF-908.02 BlCil Startup Report Page 8 of 10. Revision 0 Attachment 2 to the B1C11 Startup Report Core Monitorine Software Comparisons to Desien Code Medium Power Testing Plateau 75.9% CMWT, %NOV11,97:13:09,48 MWD /MTU Thermal Limit POWERPLEX MICROBURN-B Difference Acceptance On-Line Design Code Criteria Monitoring MFLCPR 0.857 0.864 0.007 i0.15 MAPRAT 0.8% 0.939 0.133 10.15 MFLPD 0.649 0.760 0.111 10.15 High Power Testing Plateau 94.9% CMWT, %NOV12,00:53:41,64 MWD /MTU Thermal Limit POWERPLEX MICROBURN-B C;fference Acceptance On-Line Design Code Criteria Monitoring MFLCPR 0.807 0.817 0.010 10.10 MAPRAT 0.869 0.877 0.008 i0.10 MFLPD 0.821 0.835 0.014 i0.10
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CP&L Nuclea- Fuels Management & Safety Analysis Section BICll Startup Report File: NF.908.02 Page 9 of 10. Revision 0 Attachment 2 to the B1C11 Startup Report (cont...)
Medium Power Testine Plateau Statenoint Report BRUNSWICK-1 WK-9646 96NOV11-07.13.09 48 MWD /MTU TRIGR*USEP REV=MAY96 CORE PERFORMANCE LOG - SHORT EDIT B1C11 BOC to EOC-2205 MWD /MT ODYNB POW DEP MCPR CALCOLATION TYPE : NORMAL CONVET4ENCE : TIGat? SYMMETHY : FULL I CTP CALCULATION : HEAT BALANCE CYCLE : 11 STATE CONDITIONS !
FLOW RATES CORE PARAMETERS NUCLEAR LIMITS LOCATION GMWE 679.67 WT 42.2 CMEQ 0.3412 P-PCS -0.47 39 24-04 CMWT 1940.6 (75.9%) WTSUB 44.58 CAEQ 0.1936 FCBB 1.894 PR 1000.5 PSIA WTFLAG 7 CAQA 0.1162 CMPF 2.368 37-24-04 1 DHS 31.09 WFW 8.07 CAVF 0.4951 CMFLCPR 0.857 39-22 WT 42.19 (54.8%) WD 18.29 CAPD 39.1615 P=1.454 F=1.376 i
I CRD 0.092 RWL 186.5414 CMAPRAT 0.806 37-22-04 CYCEXP 48 MWD /MTU ERATIO 0.93 CDLP 5.7262 P=0.874 F=0.828 MEASURED / CALCULATED LPRM READINGS DPCC 10.0814 CMFLPD 0.649 37-22-04 AVG: 9.53% MAK: 28.044 KEFF 1.0003 CMFLEX 0.786 01-30-10 LOCATION 1 2 3 4 5 6 7 8 9 10 11 12 AKIAL REL POWER 0.55 1.18 1.20 1.15 1.17 1.18 1.15 1.12
- is 0.96 0.81 0.45 REGION REL POWER 0. 94 1. 03 0. 95 1. 02 1. 06 1. 03 0. 9 4 1. 01 , . M RING REL POWER 0.88 1.15 1.05 1.15 1.13 1.10 0.72 APRM GAFS 1.00 0.98 1.03 0.99 0.99 1.00
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- NUCLEAR LIMITS BY REGION " * * *'*
- 7 l 8 l 9 0.831 13 -38 l 0.844 31-40 l 0.849 39-38 0.626 09 36-11 l 0.619 21-38-04 l 0.635 39-38-04 0.790 09-36-11 l 0.780 19-44-11 l 0.788 43-36-11 4 l 5 l s ..........
0.830 13-22 [ 0.783 31-22 l 0.857 39-22
- 0.627 09-20-11 l 0.584 27 24-04 l 0.649 37-22 04
- 0.793 09-20-11 l 0.720 27-24-04 l 0.806 37-22-04
- KAPRAT
- 1 l 2 l 3 l 0.827 15-12 l 0.846 31-14 1 0.837 37-12 j 0.635 17-10-11 l 0.643 31-16-04 l 0.645 37-14-04 0.800 17-10-11 1 0.795 19-10-11 l 0.796 31-14-04 l l
. . . . . . . . . . . *
- COffrROL ROD DATA ' * " * * * * " *
- j 02 06 10 14 18 22 26 30 34 38 42 46 50 51 -- -- -- -- -- 51 47 -- -- -- -- -- -- -. -- .. 47 i 43 -- -- -- 30 -- 18 -- 30 -- -- -- 43 39 . . . .. .. .. .. .. .. .. .. .. 39 35 -- -. 30 -- 10 -- 08 -- 10 -- 30 -- -- 35 31 -- -- -- -- -- -- -- -- -- -- -- -- -- 31 27 -- -- 18 -- 08 -- 22 -- 08 -- 18 -- -. 27 23 .. -. .. .. .. .. .. .. .. .. .. .. . 23 19 -- -- 30 -- 10 -- 08 -- 10 -- 30 -- -- 19 15 .. .. .. .. .. .. .. .. .. .. .. 15 11 -- -- -- 30 -- 18 -- 30 -- -- -- 11 07 -- -- -- -- -- .- -. -- +- 07 03 -- -- -- -- -- 03 02 06 to 14 18 22 26 30 34 38 42 46 50 K:\ PROJECT \BICll\STARTRi%SU REP. DOC
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CP&L Nuclear Fuels Management & Safety Andysis Section File: NF 908.02 B1Cil Startup Report Page 10 of 10. Revision 0 Attachment 2 to the BIC11 Startup Report (conte..)
H!nh Power Testina Plateau Statenoint Report BRUNSWICK-1 WK-9646 96NOV12-Oc.53.41 64 MWD /MTU TRIGRelHR REV*MAY96 CORE PERFORMANCE LOG -- SHORT EDIT B1C11 BOC to EOC-2205 MWD /MT ODYNB POW DEP MCPR CALCULATION TYPE : NORMAL CONVERGENCE : TIGHT SYMMETRY FULL CTP CALCULATION : HEAT BALANCE CYCLE 11 STATE CONDITIONS FLOW RATES CORE PARAMETERS NUCLEAR LIMITC LOCATION GMWE 851.54 WT 73.4 CMEQ 0.2509 P-PCS 0.79 29-22-05 CMWT 2428.5 ($4.9%) WTSUB 74.75 CAEQ 0.1438 FCB8 2.075 PR 1024.9 PGIA WTFLAG 2 CAQA 0.1454 CMPP 2.378 37-24-04 DHS 20.13 WFW 10.39 CAVF 0.4396 CMFLCPR 0.807 39-22 WT 73.36 (95,3%) WD 32.12 CAPD 49.0070 P=1.396 F=1.200 CRD 0.090 RWL 186.8425 CMAPRAT 0.869 17-10-11 CYCEXP 64 MWD /MTU ERATIO 0.93 CDLP 15.9690 P=0.974 F=1.000 MEASURED / CALCULATED LPRM READINGS DPCC 21.1050 CMFLPD 0.821 37-22-04 AVG: 2.994 MAX: 16.56% KEFF 1.0050 CMFLEX 0.787 01-30 10 LOCATION 1 2 3 4 5 6 7 8 9 10 11 12 AXIAL REL POWER 0.50 1.12 1.21 1.18 1.20 1.22 1.19 1.15 1.08 0.94 0.78 0.43 REGION REL POWER 0.95 1.03 0.95 1.02 1.09 1.02 0.94 1.00 0.94 RING REL POWER 0.96 1.19 1.06 1.16 1.12 1.10 0.70 APRM GAFS 0.99 0.99 0.99 0.98 0 99 0.98
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- NUCLEAR LIMITS BY REGION "'" * *
- 7 l 8 l 9 P 789 13-38 l 0.801 23-38 l 0.795 39-38 0.803 09-36-11 l 0.784 19-44-11 1 0.794 39-38-04 0.862 09-36-11 l 0.843 19-44-11 l 0.847 43-36-11 4 l 5 l 6 * * * * * " * "
0.794 13-22 l 0.754 31-22 l 0.807 39-22
- 0.802 09-20-11 l 0.735 27-24-04 l 0.821 31-22-04
- 0.863 09-20-11 l 0.771 27-24-04 ,l 0.866 37-22-04
- MAPRAT *
.........--......+...........................................e.,
1 l 2 l 3 0.791 15-12 l 0.800 21-14 l 0.788 37-12 0.810 17-10-11 l 0.820 31-16-04 l 0.812 37-14-04 0.869 17-10-11 l 0.862 19-10-11 l 0.852 37-14-04
" * " * * * * "" CON'!7tOL ROD DATA * " * * * ' * " . .
02 06 10 14 18 22 26 30 34 38 42 46 50 51 -- -- -- -- --
51 47 -- -- -- -- -- -- -- -- -- 47 43 -- -- -- 30 -- 18 -- 30 -- -- -- 43 39 -- -- -- -- -- -- -- -- -- -- -- 39 35 -- -- 30 -- 10 -- 08 -- 10 -- 30 -- -- 35 31 .. -. .. .. .. -- .. .. .. -. .. .. -. 31 27 -- -- 18 -- 08 -- 30 -- 08 -- 18 -- -- 27 23 -- -- -- -- -- -- -- -- -- -- -- -- -- 23 19 -- -- 30 -- 10 -- 08 -- 10 -- 30 -- -- 19 15 -- -- -- -- -- -- -- -- -- -- -- 15 11 -- -- -- 30 -- 18 -- 30 -- -- -- 11 07 -- -- -- -- -- -- -- -- -- 07 03 -- -- -- -- --
03 )
02 06 10 14 18 22 26 30 34 38 42 46 50 i
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