ML18023A158
| ML18023A158 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/18/2018 |
| From: | Lawrence D Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 18-003 | |
| Download: ML18023A158 (14) | |
Text
Dominion Nuclear Connecticut, Inc.
Rope Ferry Rd., Waterford, CT 06385 Mailing Address: P.O. Box 128 Waterford, CT 06385 dom.com U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 JAN 1 8 2018 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3 STARTUP TEST REPORT FOR CYCLE 19 Serial No.
MPS Lic/MLC Docket No.
License No.18-003 RO 50-423 NPF-49 Pursuant to Millstone Power Station Unit 3 Technical Specification 6.9.1.1, Dominion Nuclear Connecticut, Inc. submits the enclosed Startup Test Report for Cycle 19.
If you have any questions or require additional information, please contact Mr. Jeffry A.
Langan at (860) 444-5544.
D. C. Lawrence Director, Nuclear Safety and Licensing - Millstone
Enclosure:
(1)
Commitments made in this letter: None
cc:
U.S. Nuclear Regulatory Commission Region I Administrator 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 R. V. Guzman Senior Project Manager - Millstone Power Station U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Serial No.18-003 Docket No. 50-423 Page 2 of 2
ENCLOSURE STARTUP TEST REPORT FOR CYCLE 19 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3
~erial No.18-003 Docket No. 50-423
1.0
SUMMARY
Serial No.18-003 Docket No. 50-423 Enclosure, Page 1 of 11 Pursuant to Millstone Power Station Unit 3 (MPS3) Technical Specification (TS) 6.9.1.1, a summary of the MPS3 Cycle 19 startup testing, performed following completion of the fall 2017 refueling outage, is provided.
2.0 INTRODUCTION
The MPS3 Cycle 19 fuel reload was completed on November 2, 2017.
The attached core map (Figure 1) shows the final core configuration. Reference 6.3 documents that Cycle 19 uses a low leakage loading pattern (L3P) consisting of 84 new Region 21 fuel assemblies, 85 Region 20 once-burned fuel assemblies, and 24 Region 19 twice-burned fuel assemblies. All 193 fuel assemblies in the Cycle 19 core are the Westinghouse 17x17 robust fuel assembly (RFA-2) design.
The 84 new Region 21 fuel assemblies are comprised of 44 fuel assemblies enriched to 4.10 weight ~ercent Uranium-235 (w/o U235), 32 fuel assemblies enriched to 4.80 w/o U2 5, and eight AXIOM' lead test assemblies (L TAs) enriched to 4.1 O w/o U235. The L TAs are the same mechanical design as the rest of Region 21 with the exception of the clad material. The top and bottom regions of all fuel assemblies in the Cycle 19 core are comprised of a 6-inch annular blanket region enriched to 2.6 w/o U235. Placement of the new fuel assemblies in the designated fresh fuel assembly locations was made in a random fashion in order to prevent power tilts across the core due to systematic deviations in the fresh fuel composition.
The 109 re-inserted fuel assemblies were ultrasonically cleaned during the fall 2017 refueling outage. The purpose of the ultrasonic fuel cleaning was to remove adhered crud (primarily nickel and iron-based deposits) from the surface of fuel rods that have previous core exposure in order to reduce the probability of occurrence of crud-induced power shift (CIPS).
Every fuel assembly in Cycle 19 contains an insert. The inserts consist of 61 rod cluster control assemblies (RCCAs), 130 thimble plugs, and two secondary source assemblies.
Subsequent operational and testing milestones were completed as follows:
Initial Criticality
- Low Power Physics Testing completed Main Turbine Online 49% Power Testing completed 74% Power Testing completed 100% Power Testing completed November 13, 2017 November 13, 2017 November 14, 2017 November 15, 2017 November 16, 2017 November 20, 2017
3.0 FUEL DESIGN Serial No.18-003 Docket No. 50-423 Enclosure, Page 2 of 11 All of the 193 assemblies in the Cycle 19 core are of the RFA-2 design. This fuel design is the same as that used in Cycle 18.
4.0 LOW POWER PHYSICS TESTING The Low Power Physics Testing program for Cycle 19 was completed using the procedure in Reference 6.1 which is based on the Westinghouse dynamic rod worth measurement (DRWM) technique described in Reference 6.4. This program consisted of the following: control and shutdown bank worth measurements, critical boron endpoint measurements for all rods out (ARO),
and ARO moderator/isothermal temperature coefficient measurements. Low power physics testing was performed at a power level below the point of nuclear heat to avoid nuclear heating reactivity feedback effects.
4.1 Critical Boron Concentration The critical boron concentration was measured for the ARO configuration. The measured values include corrections to account for differences between the measured critical rod configuration and the ARO configuration. The review and acceptance criteria of +/-500 and +/-1000 percent milliRho (pcm), respectively, were met for the ARO configuration.
Summary of Boron Endpoint Results Measured Predicted M-P Acceptance (ppm)
(ppm)
(ppm)
Criteria (pcm)
All Rods 2007 1993
+14
_+/-1000 Out (ARO)
(-90.5 pcm) 4.2 Moderator Temperature Coefficient Isothermal temperature coefficient (ITC) data was measured with Control Bank D at 199.5 steps withdrawn. The review criteria of +/-2 pcm/degrees Fahrenheit (°F) to the predictions were met. The ARO moderator temperature coefficient (MTG) of
+0.06 pcm/°F was calculated by subtracting the design Doppler temperature coefficient (-1.74 pcm/°F) from the measured ITC of -2.25 pcm/°F, and adding the delta (~) ITC correction value of +0.57 pcm/°F (~ITC corrects the MTG at the measurement conditions to the minimum temperature for criticality value of 551 °F).
The TS limit of MTG < +5.0 pcm/°F at ARO hot zero power (HZP) was met.
Serial No.18-003 Docket No. 50-423 Enclosure, Page 3 of 11 Isothermal/Moderator Temperature Coefficient Results Measured Corrected M-P Acceptance (pcm/°F)
Predicted (pcm/°F)
Criteria (pcm/°F)
(pcm/°F)
ARO ITC
-2.25
-3.00 0.75 NA ARO MTG 0.06 NA NA MTG< +5.0 4.3 Control Rod Reactivity Worth Measurements The integral reactivity worths of all RCCA control and shutdown banks were measured using the DRWM technique. The review criteria of the measured worth is +/-15% or 100 pcm of the individual predicted worth, whichever is greater, and the sum of the measured worths is +/-8% of the predicted worths.
The DRWM rod worth acceptance criteria is defined as: the sum of the measured worths (M) of all banks shall be greater than or equal to 90% of the sum of their predicted worths (P).
Control Bank Integral Worth Results Measured Predicted M-P
% Difference (pcm)
(pcm)
(pcm)
(M-P) / P Control Bank A 877.3 873.8 3.5 0.4 Control Bank B 579.4 581.2
-1.8
-0.3 Control Bank C 729.5 757.7
-28.2
-3.7 Control Bank D 696.4 631.9 64.5 10.2 Shutdown Bank A 411.5 401.6 9.9 2.5 Shutdown Bank B 878.5 890.9
-12.4
-1.4 Shutdown Bank C 429.8 407.7 22.1 5.4 Shutdown Bank D 434.5 406.8 27.7 6.8 Shutdown Bank E 69.8 69.5 0.3 0.4 Total 5106.7 5021.1 85.6 1.7 The measured results of the individual bank worths and the total control bank worth showed excellent agreement with the predicted values. All individual and total worth review criteria were met. The acceptance criteria for the sum of the measured rod worths (greater than or equal to 90% of the sum of the predicted worths), was met.
5.0 POWER ASCENSION TESTING Serial No.18-003 Docket No. 50-423 Enclosure, Page.4 of 11 Testing was performed at specified power plateaus of approximately 49%, 74%
and 100% Reactor Thermal Power (RTP). Power changes were governed by operating procedures and fuel preconditioning guidelines.
Thermal-hydraulic parameters, nuclear parameters, and related instrumentation were monitored throughout the power ascension. Data was compared to previous cycle power ascension data and engineering predictions, as required, at each test plateau to identify calibration or system problems. The major areas analyzed were:
Core Performance Evaluation: Flux mapping was performed at approximately 49%, 74% and 100% RTP using the moveable incore detector system. The resultant peaking factors and power distribution were compared to TS limits to verify that the core was operating within its design limits. All analysis limits were met and the results are summarized in Section 5.1.
Nuclear Instrumentation Indication: Overlap data was obtained between the intermediate and power range nuclear instrumentation channels. Secondary plant heat balance calculations were performed to verify the nuclear instrumentation indications.
lncore/Excore Calibration: Scaling factors were calculated from flux map data using the single point calibration methodology. The nuclear instrumentation power range channels were re-scaled at approximately 74% and 100% RTP.
Reactor Coolant System (RCS) Flow: The RCS flow rate was measured at approximately 100% RTP using a secondary calorimetric heat balance for each loop with the steam generators as the control volumes. The calculated RCS flow rate met the TS requirements and is reported in Section 5.3.
5.1 Power Distribution, Power Peaking and Tilt Measurements The core power distribution was measured through the performance of a series of flux maps during the power ascension, as specified in Reference 6.2. The results from the flux maps were used to verify compliance with the power distribution TSs.
A low power flux map at approximately 49% RTP was performed to determine if any gross neutron flux abnormalities existed. At the 74% RTP plateau flux map, data necessary to perform an excore to incore calibration via the single point methodology was obtained. Per TS surveillance 4.3.1.1, Table 4.3-1, Functional Unit 2, Note 6, a flux map at approximately 100% RTP was performed for an excore to in core calibration.
The 100% RTP map also verified core power distributions were within the design limits.,
Serial No.18-003 Docket No. 50-423 Enclosure, Page 5 of 11 A summary of the measured axial flux difference (AFD) and incore tilt for the flux maps, performed during the power ascension, is provided below. Additional tables provide comparisons of the most limiting measured heat flux hot channel factor (Fa) and nuclear enthalpy rise hot channel factor (Ft.h), including uncertainties, to their respective limits from each of the flux maps performed during the power ascension. The most limiting Fa reported is based on minimum margin to the steady state limit that varies as a function of core height.
As shown below, all TS limits were met and no abnormalities in core power distribution were observed during power ascension.
Summary of Measured Axial Flux Difference and lncore Tilt Power Burnup Rod lncore Position. AFD (%)
(%RTP)
(MWD/MTU)
(steps)
Tilt 48.5 13.0 216 5.420 1.0085 73.8 27.5 216 3.297 1.0097 99.9 203.6 216 0.642 1.0087 Comparison of Measured FQ to FQ RTP Limit Power
- Burnup Measured FQ F0 RTP steady Margin to.
(%RTP)
(MWD/MTU) state limit Transient Limit 48.5 13.0 N/A N/A N/A 73.8 27.5 1.8884 3.3983 13.6 %
99.9 203.6 1.8588 2.6026 17.1 %
Comparison of Measured F ah to F ah Limit Power Burnup FC:h Fah Limit
(%RTP) (MWO/MTU) 48.5 13.0 1.486 1.831 73.8 27.5 1.446 1.711 99.9 203.6 1.421 1.586 Figures 2, 3 and 4 are the measured power distribution maps showing percent difference from the predicted power for approximately 49%, 7 4% and 100% RTP plateaus. These figures show there is good agreement between the measured and predicted assembly powers.
5.2 Boron Measurements Serial No.18-003 Docket No. 50-423 Enclosure, Page 6 of 11 Hot full power ARO boron concentration measurements were performed after reaching equilibrium conditions. The measured ARO, hot full power, equilibrium xenon, boron concentration was 1344 ppm with a predicted value of 1372 ppm.
The predicted to measured difference was -163 pcm which met the acceptance criteria of +/-1000 pcm.
5.3 Reactor Coolant System Flow Measurement The RCS flow rate was determined using a secondary calorimetric heat balance for each loop with the steam generators as the control volumes.
The following parameters were measured:
RCS pressure Hot leg temperatures Cold leg temperatures Feedwater temperatures Feedwater flow rates Feedwater pressure Steam generator pressure Steam generator blowdown was not isolated during the data acquisition period.
Per TS surveillance 4.2.3.1.3, the RCS flow was measured within 7 days after exceeding 90% RTP. The measured flow at 100% RTP was 404,664 gallons per minute (gpm) with a minimum required flow of 379,200 gpm. All TS limits were met.
6.0 REFERENCES
6.1 SP 31008, Rev. 010-00, "Low Power Physics Testing (ICCE)"
6.2 EN 31015, Rev. 006-00, "Power Ascension Testing of Millstone Unit 3" 6.3 ETE-NAF-2017-0125, Rev. 000, "Millstone Unit 3 Cycle 19 Nuclear Design Report" 6.4 WCAP-13360-P-A, Revision 1, "Westinghouse Dynamic Rod Worth Measurement Technique"
7.0 FIGURES Serial No.18-003 Docket No. 50-423 Enclosure, Page 7 of 11 1
Core Loading Pattern, Millstone Unit 3 - Cycle 19..................................... 8 2
INCORE Power Distribution - 48.5%, Millstone Unit 3 - Cycle 19.............. 9 3
INCORE Power Distribution - 73.8%, Millstone Unit 3 - Cycle 19............. 10 4
INCORE Power Distribution - 99.9%, Millstone Unit 3 - Cycle 19............. 11
R p
N M
19B 19A W69 W47 19A 21B 21B W28 Y61 Y70 19A 21B 20B W49 Y73 X75 19A 21B 21.A 20B W2fi Y49 Y02 X82 20A 21B 20A 20B X14 Y67 X04 X57 20A 21.A 21A 20A X32 Y30 Y17 X43 20A 21B 20B 21A X40 Y54 X64 Y11 20A 21A 21A 20A X29 Y40 Y20 X41 20A 21.B 20A 20B X21 Y68 X07 X56' 19A 21B 21A 20B W33 Y51 Y24 X71 19A 21B 20B W44 Y75 X66' 19B 2IB 21B W70 Y64 Y76' 19A 19A W38 W51 FIGURE 1 CORE LOADING PATTERN MILLSTONE UNIT 3 -
CYCLE 19 L
K J
H G
F E
I 19A 20A 20A 20A 20A 20A 19A W30 Xl.5 X13 X53 X17 X20 W31 21B 21B 21A 2IB 21A 21B 21B Y46 Y58 Y31 Y53 Y34 Y60 Y48 21A 20A 21A 20B 21A 20A 21A Y03 Xl.2 YiO X65 Y12 x03 Y21 20B 20B 20A 21.A 20A 20B 20B X72 X60 X47 Y22 X37 X54 X74 21.A 20A 21C 20A 21C 20A 21A Y27 X09 Y78 X52 Y80 X30 Y05 20A 21.A 20B 21A 20B 21A 20A X18 Y35 X83 Y42 X84 Y43 X08 21C 20B 21A 20A 21A 20B 21C Y81 X77 Y13 x24 Y14 X76 Y82 20A 21A 20A 20A 20A 21A 20A X51 Y32 x36 X49 X22 Y39 X44 21C 20B 21A 20A 21A 20B 21C Y83 X79 Y08 X19 Y07 X78 Y77 20A 21A 20B 21A 20B 21A 20A X10 Y38 X73 Y33 X68 Y35 X16 21A 20A 21.C 20A 21C 20A 21A Y19 X26 Y84 X42 Y79 xos Y28 20ll 20B 20A 21A 20A 20B 20B xas X58 X45 Y15 X38 X59 X69 21A 20A 21A 20B 21.A 20A 21A Y23 Xl.1 Y06 X62 Y26 XOi Y04 21B 2Ill 21A 2Il3 21A 21B 21.B Y52 Y62 Y29 Y56 Y41 Y63 Y47 19A 20A 20A 20A 20A 20A 19A W17 X28 X35 x46 X27 X25 W34 oo D
19A W43 21B Y72 20B X67 20B X81 20B X55 20A X39 21A Y01 20A X50 20B X61 20B xeo 20B X70 21B Y71 19A W42 LEGEND REGION ASSEMBLIES I ~DI Region Identifier 19A 20 FUel Assembly Identifier 19B 4
20A 53 20B 32 21A 44 21B 32 21C 8
Serial No.18-003 Docket No. 50-423 Enclosure, Page 8 of 11 C
B A
1 19A 2
W18 21B 19B 3
Y66 W68 21B 19A 4
Y74 W46 21A 21B 19A 5
Y16 Y50 W35 20A 21B 20A -6 X02 Y57 X23 21A 21A 20A -7 Y09 Y44 X31 20B 21B 20A 8
X63 Y55 X49 21A 21A 20A 9
Y25 Y37 X33 20A 21B 20A 10 X06 Y59 X34 21A 21B 19A 11 Yl.8 Y45 W40 21B 19A 12 Y69 W45 21B 19A 13 Y65 W23 19B 14 W65 15 ENRICHMENT 4.10 4.95 4.10 4.95 4.10 4.80 4.10
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INCORE Power Distribution - 49.5~
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CYCLE 19 p
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FIGUP.Ii! 3 Serial No.18-003 Docket No. 50-423 Enclosure, Page 1 O of 11 IUCORB Power Distribution - 73.1!%
MILLSTONJ2 U?i!T 3 -
CYCLE 19 p
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0.6 0.8 0.8 r, 0.4/
1238 L218 1.142 *"12a2 1.187 1.170 0.44'~ -10 0.8
- 0. j LO,1.1./
2.5 O.Q O.l:i 1219 i.153 '1.347 1.340 1.274 1.0:"'4 llo.312 - 11 2.6 2.2 :, f.9.1 2.0 2.3 D.B \\. O.Ej l'.j,;3
- {t.j 1.m 1.338 *"1,.,-~
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2.3 1.7 '- 1.9..;
0.9 02 1238 1.152 1.260 1.177 0.881 /o.zn
-13 t.O 1.4 l.2 1.9 1.1....-o.3,;
1.020 lt's_ 157 1.093,,0_50g 0.282
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-0.2 0.8
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-0.7
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0.465
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,-0.8.,1 l?IGURB 4 Serial No.18-003 Docket No. 50-423 Enclosure, Page 11 of 11 INCO:RB Power Distribution - 99.9%
MILLSTONE UliIT 3 -
CYCLE 19 p
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L K
I 0.2,04 0.444
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0.2 0.9 f\\.. 1.1,,.
0.8 0.3 1.1S7 v;.217 1.245 1,_,2s i.21:2,
-1.1 r.....-0.7./
0.6 '1.7.,1 o.,;i
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0.5 1.6 1.:,
1.12.s 1.107 1.266 Vu31 i.231
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D Measured Power
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H lloAac 0-477
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1.187 1.103
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1.224 /"1.197
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rj _23) 1.086
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0.7 1.252 1.096 L4 1.3 v;.,.:)
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1.5 1.21 i /'1.132 I
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1.12c 1.257 0.9 1.6 1.22.::: '1.23i
-0.3 \\... 0.4.,1 1'"1.01 ~ 1.199
,-0.2/ -0.1 0.46.S /0.485
-0.3 \\. -0. i,,1 G
F E
D C
8 A
I 0.460 t;°:i_44g 0.313
-1.9 '--0.5/ -0.5 1.013 Li62 LO?Q 0.497 0271
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-0.3
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2,
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U13 i.266 i.30i 1.313 1.136 O.W1
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2_1 1.3
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1.5 1.6 o_g,0.9/
'1.134 i.2-54 i.314 '1.32f 1.150 O..E(H
-12
'1.7/ 2.0 12, 1.2/
0.3 0.1 1230 i.i2,8 l.239 1.160 0.876 'o.275
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-14
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0.4 0-468 0.444 0.314 15
-0.3
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