View source for W3F1-2017-0064, Startup and Power Escalation Report for Cycle 22

{{Adams
| number = ML17268A135
| issue date = 08/29/2017
| title = Startup and Power Escalation Report for Cycle 22
| author name = Meiklejohn S
| author affiliation = Entergy Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000382
| license number = NPF-038
| contact person = 
| case reference number = W3F1-2017-0064
| document type = Letter type:W
| page count = 18
}}

=Text=
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~Entergy Entergy Operations, Inc.
17265 River Road Killona, LA 70057-3093 Tel 504 739 6685 Fax 504 739 6698 iiarrel@enten:iv.com John P. Jarrell, Ill Manager, Regulatory Assurance Waterford 3 W3F1-2017-0064 August29,2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

==Subject:==
Startup and Power Escalation Report for Cycle 22 Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

==Dear Sir or Madam:==

In accordance with Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical Specification 6.9.1.1, Entergy Operations, Inc. (Entergy) is submitting the attached summary report for plant startup and power escalation testing for Waterford 3 conqucted at the beginning of Cycle 22 operation. Waterford 3 resumed commercial power operation on June 2, 2017, following the completion of Refueling Outage 21. As part of the refueling outage scope, eight Lead Use Assemblies (LUAs) which implement some minor mechanical changes were installed in the reactor core. This report summarizes the results of the WSES-3 Cycle 22 startup physics test program, which includes the impact of the introduction of the LU As.
There are no new commitments contained in this submittal.
If you have any questions, or require additional information, please contact John Jarrell at 504-739-6685.
Attachment 1: Waterford Steam Electric Station Unit 3 Startup and Power Escalation Report for Cycle 22

W3F1-2017-0064 Page 2 cc:    Mr. Scott Morris Acting Regional Administrator RidsRgn4MailC~nter@nrc.gov NRC Senior Resident Inspector Frances.Ramirez@nrc.gov Chris.Speer@nrc.gov NRC Project Manager for Waterford 3 April.Pulvirenti@nrc.gov

Attachment 1 to W3F1*2017*0064 Waterford Steam Electrlc Station Unit 3 Startup and Power Escalation Report for Cycle 22 to W3F1-2017-0064 Page 1 of 15 TABLE OF CONTENTS Page 1.0      Introduction ..........................................................................................2 2.0      Reactor Core Description ........................................................................2 3.0      Low Power Physics Testing ..................................................................... 3 3.1    Initial Criticality .............................................................................3 3.2    Critical Boron Concentration Measurement.. ...................................... 3 3.3    Isothermal Temperature Coefficient Measurement.. ............................. 3 4.0      Power Ascension Testing ........................................................................ .4 4.1    Fuel Symmetry Verification ............................................................. 4 4.2    Core Power Distribution Measurement. ............................................ .4 5.0      Operational Testing ................................... .- ............................................ 5 5.1    Isothermal Temperature Coefficient Measurement. .............................. 5 6.0      Conclusion ...........................................................................................6 7.0      References ........................................................................................... 7 8.0      Tables 8-1    Waterford-3 Cycle 22 Design Core Loading Description ........................ 8 8-2    CECOR Results for Fuel Symmetry Verification ...................................... 10 8-3    GETARP Results at 68% Power with Westinghouse Predictions ............. 12 8-4    GE TARP Results at 100% Power with Westinghouse Predictions ........... 14
_J to W3F1-2017-0064 Page 2of15 1.0      Introduction This report summarizes the results of the Waterford 3 Cycle 22 startup physics test program, as it pertains to the incorporation of eight (8) Lead Use Assemblies (LUAs) into the core design. This program included pre-critical tests as well as those conducted during low power physics testing (LPPT), power ascension, and at full power. While all these tests performed as part of this program were completed satisfactorily, not all test results are included in this summary. Only those tests deemed necessary to demonstrate acceptance of the measured core physics parameters are included.
The objective of these tests was to demonstrate that, during reactor operation, the measured core physics parameters would be within the assumptions of the Final Safety Analysis Report (FSAR) accident analysis and within* the limitations of the plant Technical Specifications, as well as to verify the nuclear design calculations. It was also the intent of these tests to demonstrate adeqt,1ate conservatism in the Cycle 22 core performance with respect to the WSES-3 FSAR, Technical Specifications, Cycle 22 Core Operating Limits Report (COLR),
and Cycle 22 Reload Analysis Report.
2.0      Reactor Core Description The Waterford-3 Cycle 22 core design incorporates eight (8) LUAs. These assemblies, designated as sub-batch GU, use Combustion Engineering (CE) 16x16 Next Generation Fuel (NGF) fuel design with changes to the grids to increase the robustness of the design. The following is a general description of these changes, as Westinghouse considers the specifics to be proprietary:
* A modified outer strap to add grid-to-rod-fretting margin on peripheral row fuel rods;
* A grid strap manufactured using an alternative stamping method to align with Westinghouse fuel production and operating experience; and
* A slight change to the outer strap tab to aid In disrupting crud deposition on two of the peripheral row fuel rods.
For Waterford-3 Cycle 22, the feed region GG will consist of:
* 16 type GA assemblies, each with 48 Integral burnable absorber rods
* 4 type GB assemblies, *each with 88 Integral burnable absorber rods
* 12 type GC assemblies, each with 124 Integral burnable absorber rods
* 60 type GD assemblies, each with 136 Integral burnable absorber rods
* 8 type GU assemblies, each with 124 Integral burnable absorber rods The Cycle 22 core makes use of a low-leakage fuel management scheme in .
which four (4), sixteen (16), eight (8), twelve (12), and eight (8) previously burned to W3F1-2017-0064 Page 3of15 Sub-Region DA, DG, EA, ED, and EE assemblies are each placed on the core periphery, respectively. One (1) previously burned Sub-Region AD assembly is placed in the center location. The one hundred (100) fresh Region GG (Sub-Regions GA, GB, GC, GD, and GU) assemblies are located throughout the interior of the core, where they are arranged with other previously burned Region DD and EE fuel assemblies in a pattern that minimizes power peaking, and reduces both core leakage and the total neutron fluence to the reactor vessel.
See Table 8-1 for additional enrichment information.
All one hundred (100) assemblies of the Region GG fuel incorporates 2.0X ZrB2-coated Integrated Fuel Burnable Absorber (IFBA) rods utilizing a nominal B-10 loading of 3.14 mg/in in the central region of the rod and uncoated, fully-enriched annular pellets in the bottom and top six (6) inches of the fue,1 stack. Some of the previously burned fuel (Region AA) loaded in Cycle 22 used seven (7) inch CL!tbacks.
3.0      Low Power Physics Testing 3.1     Initial Criticality Following each refuel, initial criticality is achieved by boron dilution. An Estimated Critical Boron Concentration (CBC) is calculated for Regulating Control Element Assembly (CEA) Group Pat 100 inches withdrawn and all other regulating and shutdown CEA groups withdrawn to their upper electrical limits. Dilution is then commenced. For Cycle 22, the estimated CBC was calculated to be 1236 ppm. Criticality was achieved at a CBC of 1242 ppm and Group P at 100 inches withdrawn.
3.2     Critical Boron Concentration Measurement The purpose of this test is to verity the.critical boron concentration for the All Rods Out (ARO) CEA configuration of the startup test predictions.
Initially, CEA's are ARO except for Regulating CEA Group P at greater than 130 Inches withdrawn. Group P Is withdrawn to the upper group stop and the residual worth Is measured using a reactivity meter. The measured ARO CBC for Cycle 22 was 1267 ppm. The predicted ARO CBC for Cycle 22 was 1257 ppm.
3.3     Isothermal Temperature Coefficient Measurement When applying the Startup Test Activity Reduction (STAR) test program, the Moderator Temperature Coefficient (MTC) of reactivity Is calculated at Hot Zero Power (HZP) by adjusting the predicted MTC to account for the difference between actual boron concentration and the boron concentration associated with the test prediction. The resultant MTC Is extrapolated to 70% and 100% power to ensure compliance with the COLR. All values were within the limits of the COLR which meets the MTC Alternate Surveillance acceptance criteria. Results are shown In Table 3.3-1.
to W3F1-2017-0064 Page 4of15 Table 3.3-1 WSES-3 Cycle 22 MTC Alternate Surveillance Test Extrapolated           Acceptance Value*                Criteria*
MTCm(0%)**                -0.242           -3.9 < MTC < 0.5 MTC (70%)                -0.828           -3.9 < MTC < 0.0 MTC (100%)                -1.132          -3.9 < MTC < -0.2
                                                   -4     0
                             *All values are x10 Lip I F.
                             **This value was not extrapolated but corrected to account for the difference in t;>oron concentration.
4.0      Power Ascension Testing 4.1  Fuel Symmetry Verification Prior to exceeding 30% full power, fuel symmetry verification must be performed to ensure that no detectable fuel misleadings are present.
Assembly power data is obtained by executing CECOR, a computer code used to construct three dimensional assembly and peak pin power distributions from incore detector signals. Each instrumented assembly power is compared with the average of its symmetric group and a percent difference is calculated. The acceptance criterion states this difference must be less than or equal to 10%. The largest percent difference from average observed was approximately 5.67%. See Table 8-2 for CECOR output.
4.2  Core Power Distribution Measurement The purpose of this test is to verity that selected measured core power distribution parameters agree with the predicted core power distribution parameters at both the 68% and 100% power levels. These parameters Include the measured radial power distribution, axial power distribution, planar radial peaking factor (Fxy), Integrated radial peaking factor (Fr), core averaged axial peaking factor (Fz), and three-dimensional (3-0) power peaking factor (Fq). A snapshot Is taken and CECOR executed to obtain assembly power data. The comparisons were made using the GET ARP program and the results are shown In Tables 8-3 and 8-4, and summarized In Tables 4.2-1 and 4.2-2.
The acceptance criteria states that for the measured radial power distribution, the total RMS (Root Mean Square) error between measured and predicted relative power densities (RPO) for all assemblies must be less than 5.0%. Also, for each assembly with a predicted relative power density less than 0.9, the percent difference between measured and predicted must within +/-0.10 RPO or 15%, whichever Is greater. For those to .
W3F1-2017-0064 Page 5of15 assemblies with predicted relative power densities greater than or equal to 0.9, the percent difference between measured and predicted must be less than or equal to 10%. For the axial power distribution, the RMS error between measured and predicted relative power densities must be less than 5%. Additionally, for all four peaking factors, measured and predicted values must agree to within 10%. All acceptance criteria were met at both the 68% and 100% power levels and are summarized in Tables 4.2-1 and 4.2-2.
Table 4.2-1 WSES-3 Cvcle 22 68% Core Power Distribution Results Westinghouse                                   Acceptance Measured* % Difference Predicted                                         Criteria Radial RMS                N/A        1.2332             N/A              S5.0%
Axial RMS                N/A        2.6063             N/A              S5.0%
Fxv             1.4290        1.4457          1.1657          +/- 10.0 %
Fr             1.4150       1.4047          -0.7275          +/- 10.0 %
Fz             1.0870        1.0869         -0.0119          +/- 10.0 %
Fa              1.5290       1.5515           1.4727          +/- 10.0 %
                                *RMS values in %.
Table 4.2-2 WSES-3 Cycle 22 100% Core Power Distribution Results Westinghouse                                   Acceptance Measured* % Difference Predicted                                         Criteria Radial RMS                 N/A       0.7526             N/A              S5.0%
Axial RMS                 N/A        1.1680            N/A              S5.0%
F11y             1.4100       1.4204          0.7341          +/-   10.0 %
_____.Et______             1.3940       1.3819         -0.8647          +/- 10.0 %
Fi:             1.0890       1.0968          0.7187          +/- 10.0 %
F
                          . --**--**--() *-*- ---
1.5110       1.5413----- -------*----------
2.0025               --%
                                                                                                 -:t:-10.0  --
                                *RMS values in%.
5.0      Operational Testing 5.1   Isothermal Temperature Coefficient llTCl Measurement Prior to reaching 40 Effective Full Power Days (EFPD) core burnup, an Isothermal Temperature CoefficienVModerator Temperature Coefficient (ITC/MTC) test must be conducted to verify compllance with Technical Specification and COLR requirements. Initially, power Is reduced to to W3F1-2017-0064 Page 6of15 approximately 99.5% to allow temperature fluctuations necessary for the test. The RCS average temperature is increased and decreased by approximately 4 &deg;F and the power change is measured. This process is repeated to obtain sufficient data to determine an average rate of change of power with temperature. This value is multiplied by a predicted Power Coefficient to arrive at an average ITC.
The MTC is then calculated by subtracting the predicted Fuel Temperature Coefficient (FTC) from the measured average ITC. Additional calculations include MTC linear extrapolations to 70% and 100% power at the current burn up and an extrapolation to 100% power at peak boron and the end of cycle (EOC).
The acceptance criteria re9uires that for any core burnup the MTC be less positive than 0.0 x 1o* l::..p I &deg;F at 70% power, more negative than -0.2 x 10*4 l::..p I &deg;F at 100% power, and less negative than -3.9 x 10"4 l::..p I &deg;F at any power. Additionally, the measured average ITC must agree with predictions to within+/- 0.5 x 10-4 t::..p I &deg;F. All acceptance criteria were met and are summarized in Table 5.1-1.
Table 5.1-1 WSES-3 Cycle 22 Variable TAvG Test Results Entergy*                         Acceptance*
Measured*
Predicted                            Criteria ITC             -1.0562           -1.20629           +/- 0.5 MTC (70%)**                  N/A          -0.672816   -3.9 < MTC < 0.0 MTC ( 100% )**                N/A           -1.12682   -3.9 < MTC < -0.2 Peak Boron MTC N/A           -0.84836   -3.9 < MTC < -0.2 (100%)**
EOC MTC N/A           -2.45198   -3.9 < MTC < -0.2 (100%)**
                                *All values are x10*4 t::..p I &deg;F.
                              ** MTC values at 70%, 100%, Peak Boron, and EOC 100% are extrapolated.
6.0      Conclusions Based upon the successful completion of all startup tests required, specifically those described above, and the proximity of core physics parameters to predicted values, It Is concluded that the measured core parameters verify Cycle 22 nuclear design calculations and demonstrate adequate conservatism with respect to the limits and requirements of the WSES-3 FSAR, Technical Specifications, Cycle 22 Core Operating Limits Report (COLR), and Cycle 22 Reload Analysis Report.
to W3F1-2017-0064 Page 7of15 7.0      References 7.1   WSES-3 Technical Specifications 7.2   WSES-3 Cycle 22 Core Operating Limits Report (COLR) 7.3   WSES-3 Final Safety Analysis Report (FSAR) 7.4   NF-WTFD-17-6, "Waterford 3 Cycle 22 Final Reload Analysis Report, Revision 1" 7.5   WSES-3 Procedure NE-002-002, Variable Tavg Test 7.6   WSES-3 Procedure NE-002-003, Post-Refueling Startup Testing Controlling Document 7.7   WSES-3 P_rocedure NE-002-030, Initial Criticality 7.8  WSES-3 Procedure NE-002-050, Critical Boron Concentration Measurement 7.9  WSES-3 Procedure NE-002-110, Fuel Symmetry Verification 7.10 WSES-3 Procedure NE-002-140, Core Power Distribution Measurement 7.11  NF-WTFD-17-30, "Startup Test Predictions for Waterford-3 Cycle 22" 7.12  CE020017-00043, "WSES-3 Cycle 22 Variable Tavg Test Predictions" 8.0     Tables
[See Attached]
to W3F1-2017-0064 Page 8of15                                             Table 8-1 Waterford-3 Cycle 22 Core Loading Description Number of U0 2 Rods    Nominal    Zr8 2 Rods     Shim              Number Sub-     Number of   Pattern ID .                                                  Fuel Rods per      Enrichment      per      Loading             ofZrB 2 Batch ID   Assemblies                                                                (Including Assembly      (wt.%)     Assembly     (Zr8 2)             Rods ZrB, Rods)
PATl6321FB          176         4.86           8        2.0X      2944       128 GA           16 (48 IFBA)          12         4.46         40         2.0X       832      640 PATl6481FB          124         4.86         60         2.0X        736     240 GB           4 (88 IFBA)         24          4.46         28         2.0X       208       112 GC           12 PATl6361FB          112         4.86         72         2.0  x    2208      864 (124 IFBA)          0          4.46         52         2.0 x       624     624 PATl6501FB          92          4.56         92         2.0 x     11040    5520 GD          60 (136 IFBA)          8           4.16        44         2.0 x      3120    2640 PAT16361FB          112          4.86        72         2.0 x      1472     576 GU            8 (124 IFBA)          0           4.46        52         2.0 x      416      416 Total        100                                                                      23600    11760 PAT16331FB          164          4.38         20        2.llX      36811    4011 EA          20 (611 lFBA)        12          3.98        40         2.llX      1040     1100
     *Jm             II l'ATl6491FB         116          4.311        611       2.11 x     1472      544 (112 IFBA)           II         3.98        44         2.0  x      416     352 EC            4 PATl6.l21FB         176          3.1)11        II       2.0  x      736       32 (411 IFHAI        12          .Ull         411       2.0X        2011     160 El>          411 l'ATl6491FU         116          3.911        (Ill      2.0  x     8832     32(14 (112 IFBA)           II         3.Sll        44        2.0 x      24%. 2112 l'ATI 6.l<1IFB      112          .l.98        72        2.0 x      2944     II 52 l\E          16 (124 IFBA)          0          3.58         52        2.0 x       1132     832 Totnl        *}(1                                                                     22<1~6    %411 to W3F1-2017-0064 Page 9of15                                            Table 8-1 Waterford-3 Cycle 22 Core Loading Description Number of U0 2 Rods     Nominal   ZrB 2 Rods     Shim              Number Sub-      Number of   Pattern ID                                                  Fuel Rods per      Enrichment      per      Loading             ofZrB 2 Batch ID   Assemblies                                                              (Including Assembly      (wt.%)     Assembly     (ZrB 2)             Rods ZrB, Rods)
PAT16321FB        176         4.53          8         2.0X       736       32 DA           4 (48 IFBA)        12         4.23         40         2.0X       208       160 PATl6501FB        92           3.83        92         2.0X      2944      1472 DO           16 (136 IFBA)        8          3.53         44         2.0X       832      704 Total         20                                                                     4720     2368 AD PAT16351fB        136          3.90        48         2.0 x      184       48 (100 IFBA)        ()          3.50        52         2.0X        52       52 Total                                                                                236       100 Grund 217                                                                     51212    23876 Total}}