NL-19-0719, GNF3 New Fuel Introduction Startup Test Report

GNF3 New Fuel Introduction Startup Test Report

ML19168A133
Person / Time
Site:	Hatch
Issue date:	06/17/2019
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-19-0719
Download: ML19168A133 (16)

Text

~ Southern Nuclear Cheryl A. Gayheart Regulalory Affa1rs D1rector 3535 Colonnade Parkway Birmingham. AL 35243 205 992 5316 lei 205 992 7601 fax JUN 1 7 2019 cagayhea@ soulhernco.com Docket Nos.: 50-366 NL-19-0719 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 GNF3 New Fuel Introduction Startup Test Report Ladies and Gentlemen:

In accordance with FSAR requirements, Southern Nuclear Operating Company (SNC) hereby submits the Startup Test Report for Cycle 26 for Hatch Nuclear Plant Unit 2 (HNP2). This report summarizes the startup testing performed on HNP2 following the twenty-fifth refueling outage.

The report is required due to the first use, other than as lead use assemblies, of GNF3 fuel assemblies loaded for Cycle 26.

The tests demonstrate the successful operation of the HNP2 reactor with the introduction of the GNF3 fuel.

This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.

Respectfully submitted, Director, Regulatory Affairs Southern Nuclear Operating Company CAG/tle/scm

Enclosure:

GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 cc: Regional Administrator, Region II NRR Project Manager- Hatch Nuclear Plant Senior Resident Inspector- Hatch Nuclear Plant RType: CHA02.004

Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report Enclosure GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26

Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26

1.0 INTRODUCTION

1.1 Purpose and Summary The Edwin I. Hatch Nuclear Plant Unit 2 (HNP2) Startup Test Report is submitted to the Nuclear Regulatory Commission (NRC) in accordance with regulatory commitments contained in the HNP2 Final Safety Analysis Report (FSAR)

Section 13.6.4. This report summarizes the startup testing performed on HNP2 following the twenty-fifth refueling outage. This report is being submitted due to a reload batch of 208 GNF3 fuel assemblies that were loaded for Cycle 26. The GNF3 fuel design has not previously been utilized in HNP2 except as Lead Test Assemblies (LTAs).

This report consists of a summary of the core design followed by summaries of selected static and dynamic reactor core performance tests conducted prior to and during the beginning-of-cycle startup of HNP2 Cycle 26. These tests demonstrate the successful operation of the HNP2 reactor with the introduction of the GNF3 fuel design in production use.

1.2 Plant Description HNP2 is a General Electric design boiling water reactor (BWR/4) and is rated at 2804 MW(th) with an approximate output of 920 MW(e), gross. The plant is located on the south side of the Altamaha River, southeast of the intersection of the river with U.S. Highway #1 in the Northwestern sector of Appling County, Georgia.

1.3 Post-Refueling Outage Startup Test Description HNP2 resumed commercial operation on March 21, 2019, after completing a 45-day refueling/maintenance outage. The following core performance tests were performed as part of the post-refueling outage startup test program:

• Core Verification

• Control Rod Drive Timing

• In-Sequence Critical Shutdown Margin Demonstration

• Cold Critical Eigenvalue Comparison

• LPRM Calibration

• APRM Calibration

• Control Rod Scram Time Testing

• Core Performance

• Reactivity Anomaly Calculation E-1

Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 The purpose for, a brief description of, and the acceptance criteria for each of the tests listed above are enumerated in Section 3 of this report.

1.4 Post-Refueling Outage Startup Test Acceptance Criteria Where applicable, a definition of the relevant acceptance criteria for the test is given and is designated either "Level 1" or "Level 2."

Acceptance Criteria:

Level 1 criteria: Data trend, singular value, or information which relates to Technical Specifications margin and/or plant design in such a manner that requires strict observance.

Level 2 criteria: Data trend, singular value, or information relative to system or equipment performance which does not fall under the definition of Level 1 criteria.

Failure to meet Level 1 criteria constitutes failure of the specific test. The Test Lead is required to resolve the problem, and if necessary, the test is repeated. Level 2 criteria do not constitute a test failure or acceptance; they serve as information only.

2.0 CORE DESIGN

SUMMARY

2.1 Cycle/Core Summary The Cycle 26 design achieves a full cycle exposure of 16.558 GWd/ST or 683 effective full power days (EFPDs) at 2804 MW(th). This energy includes cycle extension from increased core flow and final feedwater temperature reduction.

Two hundred eight (208) fresh GNF3 bundles, divided into four streams having enrichment varying from 4.04 w% to 4.28 w% U-235 enrichment, were loaded in a conventional core configuration for a 24-month fuel cycle.

2.2 Calculated Reactivity/Thermal Limit Margins The two parameters which describe the global behavior of the core throughout the cycle are hot excess reactivity (HER) and cold shutdown margin (CSDM). The 0.0 MWd/ST hot excess reactivity is 2.06%, the early cycle minimum HER is 1.88% at 500.0 MWd/ST, and the mid-cycle peak HER is 2.00% at 2,700.0 MWd/ST. The minimum CSDM of 1.24% occurs at BOC for the as-to-be-loaded core loading based upon an EOC 25 shutdown at 16.831 GWd/ST cycle exposure. Calculated core parameters are delineated in Table 2.1.

Target rod patterns were developed at reasonable exposure increments and 2,200 MWd/ST sequence exchange intervals. Design margins to thermal limits were met for all exposures.

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Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 2.3 Fuel Summary Table 2.2 provides a list of all fuel batches loaded in Cycle 26. Note that all fuel contains axially varying fuel lattice types.

Four once-burned GNF3 LT As and 348 other GNF2 bundles are carried over from Cycle 25. All returning once-burned fuel and twice-burned assemblies are equipped with the Defender' Debris Filter LTP.

Table 2.1, Cycle Calculated Parameters Parameter Value SOC Core Average Exposure 16,325 MWd/ST Cycle Core Weight 115.7393 ST Daily Full Power Exposure Capability 24.23 MWd/ST Cycle Energy EUP Rated (DOR) 15,764 MWd/ST 651 EFPD Projected Rated (DOR) 1 15,404 MWd/ST 636 EFPD Projected Total (EOC) 2 16,558 MWd/ST 683 EFPD Uncertainty in Energy (+/- 0.3 % ~k) +/- 342 MWd/ST Cold Shutdown Margin soc 1.24  % l1k R 0.00 %Ak S 0.30 %Ak Hot Excess Reactivity soc 0 MWd/ST 2.06 %Ak Early Cycle Min 500 MWd/ST 1.88 %Ak Mid-Cycle Peak 2,700 MWd/ST 2.00 %Ak E-3

Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 Table 2.2, Fuel Bundle Batch Descriptions Loaded in Cycle 26 Batch IAT QTY 10 Range Bundle Type Label Fresh Fuel H2R25 25 24 YLU825 - YLU848 GNF3-P1 ODG3B416-13GZ-83AV-150-T6-4598 (NSF)

H2R25 26 32 YLU745- YLU776 GNF3-P1 ODG3B409-14GZ-83AV-150-T6-4596 (NSF)

H2R25 27 48 YLU777 - YLU824 GNF3-P1 ODG3B428-12GZ-83AV-150-T6-4597 (NSF)

H2R25 28 104 YLU641 - YLU744 GNF3-P1 ODG3B404-15GZ-83AV-150-T6-4595 (NSF)

Once Burned Fuel H2R24 6 56 YLM101 - YLM156 GNF2-P10DG2B395-15GZ-100T2-150-T6-4447 (NSF)

H2R24 7 64 YLM 157 - YLM220 GNF2-P1 ODG2B397-14GZ-1 OOT2-150-T6-4448 (NSF)

H2R24 8 64 YLM221 - YLM284 GNF2-P1 ODG2B414-14GZ-1 OOT2-150-T6-4449 (NSF)

H2R24 9 40 YLM285 - YLM324 GNF2-P1 ODG2B414-12GZ-1 OOT2-150-T6-4450 (NSF)

H2R24 10 4 YLM325 - YLM328 GNF3-P10DG38404-12G7.0-83AV-150-T6-4451-LUA(NSF)

Twice Burned Fuel H2R23 20 12 YLE246- YLE253 GNF2-P10DG2B398-15GZ-100T2-150-T6-4314 YLE258- YLE261 H2R23 21 20 YLE286 - YLE305 GNF2-P1 ODG2B402-14GZ-1 OOT2-150-T6-4315 H2R23 22 48 YLE318 - YLE333 GNF2-P1 ODG2B400-13GZ-1 OOT2-150-TS-4316 YLE342 - YLE357 YLE366 - YLE381 H2R23 23 36 YLE382- YLE417 GNF2-P10DG2B411-14GZ-1 OOT2-150-T6-4317 H2R23 24 8 YLE462 - YLE469 GNF2-P1 ODG2B411-14GZ-1 OOT2-150-T6-4317 (NSF) 3.0

SUMMARY

OF POST -REFUELING OUTAGE STARTUP TEST RESULTS 3.1 Core Verification 3.1.1 Purpose To verify all fuel assemblies have been properly loaded into the reactor core as per the licensed final loading pattern, including fuel bundle location, orientation, and seating.

3.1.2 Acceptance Criteria Levell criteria: Each fuel assembly must be verified to be in its proper location and orientation as specified by the final loading pattern (Licensed Core) and be correctly seated in its respective cell.

Level 2 criteria: N/A 3.1.3 Test Description The HNP2 Cycle 26 core verification was performed by use of underwater TV cameras to visually inspect the location (by bundle serial number identification),

orientation, and seating of each of the 560 fuel assemblies that comprise the as-loaded core.

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Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 3.1.4 Test Results Core verification was performed on February 22, 2019, in accordance with engineering procedures for fuel movement. The visual inspection confirmed all bundles were in their correct location and orientation, and no bundles required reseating.

3.2 Control Rod Drive (CRD) Timing 3.2.1 Purpose To demonstrate the CRD system operates properly following the completion of a core alteration. In particular, this functional test verifies that the insert and withdrawal capability of the CRD system is within acceptable limits.

3.2.2 Acceptance Criteria Level 1 Criteria: The insert and withdrawal drive time for each CRD must be between 43.2 and 52.8 seconds. In the event that a CRD fails to meet these criteria, the applicable drive must be adjusted and new criteria of 45.4 to 50.2 seconds are applied to the adjusted drive.

Level 2 Criteria: N/A 3.2.3 Test Description Control rod drive timing is performed once per operating cycle on all CRDs.

Normal withdrawal and insertion times are recorded for each of the drives under normal drive water pressure. If acceptable withdrawal and/or insertion cannot be obtained with normal drive water pressure, then the respective needle valve for the applicable withdrawal and/or insertion stroke must be adjusted until an acceptable drive time is achieved in accordance with the above criteria.

3.2.4 Test Results Control rod drive timing was completed on March 12, 2019 for all137 CRDs in accordance with plant operating procedures for CRD timing. Each CRD was determined to have, or was adjusted (where necessary) to have, a normal insertion and withdrawal speed as required.

3.3 In-Sequence Critical Shutdown Margin Demonstration 3.3 .1 Purpose To demonstrate the reactor can be made subcritical for any reactivity condition during Cycle 26 operation with the analytically determined highest worth control rod capable of withdrawal, fully withdrawn and all other rods fully inserted.

3.3.2 Acceptance Criteria Levell Criteria: The loaded core must be subcritical by at least 0.38% ilK with the analytically determined highest worth control rod capable of E-5

Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 being withdrawn, fully withdrawn, and all other rods fully inserted at the most reactive condition during the cycle.

Level 2 Criteria: N/A 3.3.3 Test Description The in-sequence critical shutdown margin demonstration was performed immediately following the HNP2 Cycle 26 BOC initial criticality with the reactor core in a xenon free state. To account for reactivity effects such as moderator temperature, reactor period, and the one rod out criterion, correction factors were used to adjust the startup condition to cold conditions with the highest worth control rod fully withdrawn.

3.3.4 Test Results The in-sequence critical shutdown margin demonstration was performed on March 18, 2019 in accordance with core calculation procedures for shutdown margin demonstration. Results of this calculation yielded a CSDM of 1.64% .1 K.

The minimum CSDM was also 1.64% 1:!. K because CSDM this operating cycle is a minimum at BOC. A summary of the shutdown margin demonstration is given in Attachment 1 of this report.

3.4 Cold Critical Eigenvalue Comparison 3.4.1 Purpose To compare the critical eigenvalue calculated using the actual cold, xenon free critical control rod configuration (corrected for moderator temperature and reactor period reactivity effects) to the cold critical eigenvalue assumed in the cycle management analysis.

3.4.2 Acceptance Criteria Level 1 Criteria: The cold critical eigenvalue calculated using actual critical data shall not differ from the design cold critical eigenvalue by more than +/- 1% I:!.K.

Level 2 Criteria: N/A 3.4.3 Test Description The cold critical eigenvalue is the assumed value of the PANACEA 3-D core simulator model Kerr at which criticality is achieved with the reactor in a xenon free state and the coolant is 68 degrees F. This value is determined based on historical data and used for cycle management analysis by core analysis personnel. Once the actual critical state is achieved during the beginning of cycle startup, the applicable data are provided to core analysis personnel, and the actual (corrected for moderator temperature and reactor period reactivity effects) cold critical eigenvalue is calculated. This value is then compared to the assumed critical eigenvalue as a method of validating rod worths and shutdown margin E-6

Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 calculations throughout the cycle. The actual critical eigenvalue is also entered into a database for predicting future cold critical eigenvalues.

3.4.4 Test Results The beginning of cycle startup for HNP2 Cycle 26 was performed on March 18, 2019. The observed reactor core conditions when a critical state was achieved are listed in Attachment 1.

The results of the PANACEA case show the temperature and period-corrected cold eigenvalue to be 1.0056. This is 0.41% ~K above the design value of 1.0015 and is well within the +/- 1.0% ~K acceptance criteria.

3.5 Local Power Range Monitor (LPRM) Calibration 3.5.1 Purpose To calibrate the local power range monitors (LPRMs) by fine-tuning gain adjustment factors (GAFs) such that LPRM readings are equivalent to Traversing Incore Probe (TIP) detector readings. The TIP measurements, in turn, are proportional to the axial flux distribution at selected intervals over the regions of the core where the LPRMs are located. TIP readings are of high precision to allow reliable calibration of LPRM gains.

3.5.2 Acceptance Criteria Level 1 Criteria: All detector GAFs:::; 40.

Level 2 Criteria: N/A 3.5.3 Test Description The LPRM channels were calibrated to make the LPRM readings proportional to the neutron flux in the narrow-narrow water gap at the chamber elevation. This calibration was performed in accordance with engineering procedures for LPRM calibration.

3.5.4 Test Results Using site procedures, LPRMs were successfully calibrated at 100% power.

LPRM Gain Adjustment Factor Values for all operable LPRM channels were within specified limits.

3.6 APRM Calibration 3.6.1 Purpose To calibrate the APRM system to actual core thermal power as determined by a heat balance.

3.6.2 Acceptance Criteria Level 1 criteria: The APRM readings must be within a tolerance of 2% of core thermal power as determined from a heat balance.

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Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 Level 2 criteria: N/A 3.6.3 Test Description The APRM gains are adjusted after major power level changes, if required, to read the actual core thermal power as determined by a heat balance performed in accordance with plant operating procedures for APRM adjustment to core thermal power. The heat balance required for the calibration process was obtained from the process computer program OD3 (Core Thermal Power and APRM Calibration) or from the Official Monitor case in accordance with plant operating procedures.

3.6.4 Test Results APRM calibration was performed in accordance with plant operating procedures at approximately 7%, 14%, 18%,24%,40%,59%, and 66% of Rated Thermal Power. Each APRM was calibrated within a 2% tolerance to read core thermal power as calculated by the heat balance.

3.7 Control Rod Scram Time Testing 3.7.1 Purpose To demonstrate that the CRD system functions as designed with respect to scram insertion times following the completion of core alterations.

3.7.2 Acceptance Criteria Level 1 criteria:

(a) The individual scram insertion time for all operable control rods from the fully withdrawn position, based on de-energization of the scram pilot solenoids, with reactor steam dome pressure above 800 psig shall not exceed the following:

From Fully Individual Rod Withdrawn Maximum Insertion To Notch Position Time (sec) 46 0.44 36 1.08 26 1.83 06 3.35 (b) The individual control rods with scram times in excess of those listed in (a) above are to be declared as SLOW with the following restrictions:

1. No more than 10 operable control rods are declared SLOW.

2. No more than 2 operable control rods that are declared SLOW occupy adjacent locations.

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Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 (c) The maximum scram insertion time of each control rod, from the fully withdrawn position to position 06, based on the de-energization of the scram pilot solenoid, shall not exceed 7.0 seconds.

Level 2 criteria: N/A 3.7.3 Test Description The CRD scram time testing was performed in accordance with engineering procedures for control rod scram testing, with the steam dome pressure above 800 psig. The test consists of scramming each control rod, collecting the resulting scram time data, and analyzing the data in accordance with the acceptance criteria noted above.

3.7.4 Test Results All CRDs were tested in accordance with engineering procedures for control rod scram testing, with the steam dome pressure greater than 800 psig. A summary of the results is given in Attachment 2 of this report.

3.8 Core Performance 3.8.1 Purpose To evaluate core performance parameters to assure plant thermal limits are maintained during power ascension to rated conditions.

3.8.2 Acceptance Criteria Level 1 criteria: The following thermal limits are ::;; 1.000 when ~ 24% RTP:

1. MFLCPR (Maximum Fraction of Limiting Critical Power Ratio)

2. MFLPD (Maximum Fraction of Limiting Power Density)

3. MAPRAT (Maximum Average Planar Linear Heat Generation Ratio).

Level 2 criteria: N/A 3.8.3 Test Description As power is increased, core thermal limits were evaluated at various levels up to 100%. In accordance with plant operating procedures for core parameter surveillance, demonstration of fuel thermal margin was performed. Fuel thermal margin was confirmed at each level before increasing reactor power further.

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Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 3.8.4 Test Results Thermal limits were continuously monitored during power ascension. The surveillance procedure was performed satisfactorily at various levels as indicated below:

Thermal Limit 24% 40% 55% 66% 79% 90% 100%

MFLCPR 0.717 0.692 0.805 0.917 0.807 0.770 0.894 MFLPD 0.775 0.563 0.638 0.810 0.815 0.847 0.867 MAPRAT 0.261 0.309 0.369 0.544 0.612 0.694 0.753 3.9 Reactivity Anomaly Calculation 3.9.1 Purpose To check for possible reactivity anomalies as the core excess reactivity changes with exposure.

3.9.2 Acceptance Criteria Level 1 Criteria: The monitored core Kerr shall not differ from the predicted core Kerr by more than +/-1% ~ K.

Level 2 criteria: N/A 3.9.3 Test Description Mter obtaining steady-state conditions following a BOC startup from a refueling outage and every month thereafter, a reactivity anomaly calculation is performed to monitor the core reactivity during the cycle. Verifying the reactivity difference between the monitored and predicted core Kerr is within limits provides assurance that plant operation is maintained within the assumptions of the DBA and transient analyses. The core monitoring system calculates the core Kerr for the reactor conditions obtained from plant instrumentation. A comparison of the monitored core Kerr to the predicted core Kerr at the same cycle exposure is used to ensure the difference is within a +/- 1% ~ K acceptance band.

3.9.4 Test Results The initial reactivity anomaly calculation for the cycle was performed in accordance with the engineering procedures for reactivity anomaly calculations on April3, 2019. The monitored core Kerr was well within the acceptance criteria range as specified above. The results of this calculation are given in Attachment 3 of this report.

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Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26

4.0 CONCLUSION

S As indicated by the acceptable results of all the startup testing, operation of the HNP2 reactor is successful with the introduction of the GNF3 fuel.

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Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 ATTACHMENT 1 IN-SEQUENCE CRITICAL COLD SHUTDOWN MARGIN DEMONSTRATION Sequence A2 RWMGroup1 Fully Withdrawn RWMGroup2 16 control rods fully withdrawn KsRo 0.9862 Reactor Coolant Temperature 139.5° F Reactor Period 455 sec.

Corrected KcruT 1.00559 Cold Shutdown Margin 1.64% AK Value ofR O.OAK Value ofB (conservative bias) 0.0030AK Minimum Cold Shutdown Margin 1.64% AK Tech Spec Required Shutdown Margin 0.38% AK E-12

Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 ATTACHMENT 2 SCRAM TIME TESTING LOCATIONS TIME IN SECONDS TO NOTCH POSITION 46 36 26 06 Fastest Rods 46-11 0.211 0.725 1.260 2.329 42-27 0.231 0.682 1.151 2.089 42-27 0.231 0.682 1.151 2.089 42-27 0.231 0.682 1.151 2.089 Slowest Rods 34-51 0.273 0.803 1.332 2.391 14-47 0.260 0.833 1.389 2.459 38-47 0.252 0.809 1.391 2.569 38-47 0.252 0.809 1.391 2.569 Average (All Rods) 0.243 0.752 1.271 2.306 E-13

Enclosure to NL-19-0719 GNF3 New Fuel Introduction Startup Test Report for HNP2 Cycle 26 ATTACHl\tiENT 3 REACTIVITY ANOMALY CALCULATION UNIT 2 CYCLE 26 SEQUENCE: A2 DATE PERFORMED 04/03/2019 THERMAL POWER (MWth) CMWT 2801.2 RATED THERMAL POWER (MWth) 2804.0 CORE FLOW (Mlblhr) WT 76.291 RATED CORE FLOW (Mlb/hr) 77.00 CORE XENON CONCENTRATION -2.22 XE/RATED 0.987 CYCLE EXPOSURE (MWD/sT) 161.5 EIGENVALUE Kerr 1.0081 PREDICTED Keff = 1.0084

+1% VALUE= 1.0184 -1% VALUE= 0.9984 E-14