NL-15-0905, Startup Test Report for Cycle 24

Startup Test Report for Cycle 24

ML15163A291
Person / Time
Site:	Hatch
Issue date:	06/12/2015
From:	Pierce C Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-15-0905
Download: ML15163A291 (21)

Text

Charles R. Pierce Southern Nuclear Regulatory Affa1rs Director Operating Company, Inc.

40 Inverness Center Parkway Post Off1ce Box 1295 A

B1rmmgham, AL 35201 Tel 205 992.7872 Fax 205.992.7601 SOUTHERN COMPANY June 12, 2015 Docket Nos.: 50-366 NL-15-0905 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report Ladies and Gentlemen:

In accordance with FSAR requirements, Southern Nuclear Operating Company (SNC) hereby submits the Unit 2 Startup Test Report for Cycle 24. This report summarizes the startup testing performed on Unit 2 following the twenty-third refueling outage. The report is required due to the first use, other than as lead use assemblies, of GNF2 fuel assemblies loaded for Cycle 24.

The tests demonstrate the successful operation of the Plant E. I. Hatch Unit 2 reactor with the introduction of the GNF2 fuel.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

Rt.e~Jy f!;d, C. R. Pierce Regulatory Affairs Director CRP/RMJ

Enclosure:

Edwin I. Hatch Nuclear Plant - Unit 2 GNF2 New Fuel Introduction Startup Test Report for Cycle 24

U.S. Nuclear Regulatory Commission NL-15-0905 Page 2 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President- Hatch Mr. M.D. Meier, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Mr. G. L. Johnson, Regulatory Affairs Manager- Hatch RTYPE: CHA02.004 U.S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager - Hatch Mr. D. H. Hardage, Senior Resident Inspector- Hatch

Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report Enclosure GNF2 New Fuel Introduction Startup Test Report for Cycle 24

Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24

1.0 INTRODUCTION

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

Section 13.6.4. This report summarizes the startup testing performed on Unit 2 following the twenty-third refueling outage. This report is being submitted due to a reload batch of 224 GNF2 fuel assemblies that were loaded for Cycle 24. The GNF2 fuel design has not previously been utilized in Unit 2 except as Lead Test Assemblies (LTAs).

This report consists of a brief 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 Plant Hatch Unit 2 Cycle 24. These tests demonstrate the successful operation of the Unit 2 reactor with the introduction of the GNF2 fuel design into production use.

1.2 Plant Description The Edwin I. Hatch Nuclear Power Plant Unit 2 is a General Electric design boiling water reactor (BWR/4). Plant Hatch Unit 2 is rated at 2804 MW(th) with a generator rating at this power of 920 MW(e). 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 The Edwin I. Hatch Nuclear Power Plant Unit 2 resumed commercial operation on March 14, 2015, after completing a 33-day refueling/maintenance outage.

The following core performance tests were performed as part of the post-refueling outage startup test program:

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Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24

• 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 The purpose for, a brief description of, and the acceptance criteria for each of the tests listed above is 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. Level2 criteria do not constitute a test failure or acceptance; they serve as information only.

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Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 2.0 CORE DESIGN

SUMMARY

2.1 Cycle/Core Summary The Cycle 24 design achieves a full power energy of 15.616 GWd/ST or 625.3 effective full power days (EFPDs) at 2804 MWth. This energy includes cycle extension from increased core flow. Two hundred and twenty-four (224) fresh GNF2 bundles, divided into four streams having enrichment that varies from 3.98 w% to 4.11 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 200.0 MWd/ST hot excess reactivity is 1.494%, the early cycle minimum HER is 1.478% at 500.0 MWd/ST, and the mid-cycle peak HER is 1.647% at 9,200.0 MWd/ST. The minimum cold shutdown margin of 1.64% occurs at BOC for the as-to-be-loaded core loading based upon an EOC 23 shutdown at 17.056 GWd/ST cycle exposure. Calculated core parameters are delineated in Table 2.1.

Target rod patterns were developed at reasonable exposure increments and 2,300 MWd/ST sequence exchange intervals. Design margins to thermal limits were met for all exposures.

2.3 Fuel Summarv Table 2.2 provides a list of all fuel batches loaded in Cycle 24. Note that all fuel contains axially varying fuel lattice types.

All returning once-burned fuel and twice-burned assemblies are equipped with the Defender' Debris Filter LTP, leaving four (4) returning thrice-burned assemblies with the standard GE14 debris filter LTP.

Four thrice-burned GE14 bundles, equipped with standard debris filters, identified below were originally loaded in Hatch-2 Cycle 18 and discharged after Cycle 20.

They are being incorporated into Hatch-2 Cycle 24 for a fourth cycle of operation on the core periphery.

One once-burned GE14 bundle identified below was originally loaded in Hatch-1 Cycle 25. It was discharged for post outage inspection after one cycle of operation and reloaded into Hatch-2 Cycle 23. It will complete a third cycle of operation during Cycle 24.

Four GNF2 Lead Test Assemblies (LTAs) were originally loaded in Hatch-2 Cycle

22. The GNF2 LTAs are designed to mimic a sibling GE14 bundle design. The GNF2 LTAs contain ZIRON clad fuel rods and standard NSF channels. They will complete their third cycle of operation during Cycle 24.

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Enclosure to N L-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 Table 2.1 Cycle Calculated Parameters 1

BOC Core Average Exposure 15,016.56 MWd/ST Cycle Core Weight 112.2764 ST Daily Full Power Exposure Capability 24.9741 MWd/ST Cycle Energy ~ ElU:lQ:i!.!r~

23 EUP (rated) ' 629.9 15,730 MWd/ST 2

Achieved (rated) 625.3 15,616 MWd/ST 3

Total Energy (with coastdown) 663.0 16,558 MWd/ST Uncertainty in Energy +/-428 MWd/ST Cold Shutdown Margin BOC 1.94  % t!k R 0.00  % t!k B 0 .30  % t!k Hot Excess Reactivity BOC 200.00 MWd/ST 1.49  % t!k Early Cycle Min 500.00 MWd/ST 1.48  % t!k Mid-C cle Peak 9,200.00 MWd/ST 1.65 %t!k 1- BOC CAVEX based on projection to an EOC 23 cycle exposure of 17,056 MWd/ST.

2- Rated power at 105% core flow.

3- Energy based on July 16, 2014 Nuclear Fuel Plan of Record E-4

Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 Table 2.2 Fuel Batches Loaded in Cycle 24 QTY Bundle Type Label Fresh Fuel 32 GNF2-P10DG2B398-15GZ-100T2-15D-T6-4314 32 GNF2-P10DG2B402-14GZ-100T2-15D-TG-4315 72 GNF2-P10DG2B400-13GZ-100T2-15D-T6-4316 80 GNF2-P10DG2B411-14GZ-100T2-15D-TG-4317 8 GNF2-P10DG2B411-14GZ-100T2-15D-T6-4317 Once Burned Fuel 64 GE14-P10DNAB393-14GZ-100T-150-T6-4182 64 GE14-P10DNAB406-18GZ-100T-150-T6-4183 24 GE14-P10DNAB418-16GZ-100T-150-T6-4184 72 GE14-P10DNAB423-15GZ-100T-150-T6-4185 Twice Burned Fuel 48 GE14-P10DNAB419-16GZ-100T-15D-T6-3392 28 GE14-P10DNAB395-14GZ-100T-150-T6-3391 23 GE14-P10DNAB402-15GZ-100T-150-T6-3389 4 GE14-P10DNAB423-15GZ-100T-150-T6-2876 4 GNF2-P10DG2B401-14GZ-100T2-15D-T6-3394 1 GE14-P10DNAB423-15GZ-100T-150-T6-2876 Thrice Burned Fuel 4 GE14-P10DNAB398-4G7.0/11G6.0/1G2.D-100T-15D-T6-2620 E-5

Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 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 Level 1 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 Hatch Unit 2 Cycle 24 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.

3.1.4 Test Results Core verification was performed on February 28, 2015, 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 CAD system operates properly following the completion of a core alteration. In particular, this functional test verifies that the insert and withdrawal capability of the CAD system is within acceptable limits.

3.2.2 Acceptance Criteria Level1 Criteria: The insert and withdrawal drive time for each CAD must be between 43.2 and 52.8 seconds. In the event that a CAD 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.

Level2 Criteria: N/A E-6

Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 3.2.3 Test Description Control rod drive timing is performed once per operating cycle on all CADs.

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 8, 2015 for all 137 CADs in accordance with plant operating procedures for CAD timing. Each CAD 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 24 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 Level 1 Criteria: The loaded core must be subcritical by at least 0.38% 11 K with the analytically determined highest worth control rod capable of 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 Plant Hatch Unit 2 Cycle 24 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 10, 2015 in accordance with core calculation procedures for shutdown margin demonstration. Results of this calculation yielded a cold shutdown margin of 1.861% 11 K. The minimum cold shutdown margin was also 1.861% 11 K because cold shutdown margin this operating cycle is a minimum at BOC. A summary of the shutdown margin demonstration is given in Attachment 1 of this report.

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Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 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% ~K Level2 Criteria: N/A 3.4.3 Test Description The cold critical eigenvalue is the assumed value of the PANACEA 3-D core simulator model Kett at which criticality is achieved with the reactor in a xenon free state and the coolant at 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 calculations throughout the cycle. The actual critical eigenvalue is also entered into a database for predicting future cold critical eigenvalues.

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Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 3.4.4 Test Results The beginning of cycle startup for Plant Hatch Unit 2 Cycle 24 was performed on March 10, 2015. 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 1.0022. This is 0.22% ~K above the design value of 1.0000 and is well within the +/- 1.0% ~K acceptance criteria. This also compares favorably to the value of 1.0000 which was actually used in the core design as an extra margin of conservatism due to the introduction of the GNF-2 fuel type.

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 lncore 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.

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Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 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.

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 003 (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 23%, 36%, 58%, 71%, 85%, 95%, and 100% of Rated Thermal Power. Each APRM was calibrated within a 2% tolerance to read core thermal power as calculated by the heat balance.

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Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 3.7 Control Rod Scram Time Testing

3. 7.1 Purpose To demonstrate that the CAD 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.

(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.

Level2 criteria: N/A E-11

Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 3.7.3 Test Description The CAD 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 CADs 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.

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Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 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.

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% 36% 58% 70% 85% 96% 100%

MFLCPR 0.760 0.730 0.820 0.913 0.861 0.864 0.886 MFLPD 0.642 0.547 0.798 0.790 0.856 0.847 0.832 MAPRAT 0.358 0.359 0.615 0.615 0.630 0.666 0.699 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 ke 11 shall not differ from the predicted core ke11 by more than +/- 1% .1 K.

Level2 criteria: N/A E-13

Enclosure to NL-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24 3.9.3 Test Description After 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 ke<< 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 ke<< for the reactor conditions obtained from plant instrumentation. A comparison of the monitored core ke<< to the predicted core ke<< at the same cycle exposure is used to ensure the difference is within a +/- 1% .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 March 19, 2015. The monitored core ke<< 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-15-0905 GNF2 New Fuel Introduction Startup Test Report for Cycle 24

4.0 CONCLUSION

S As indicated by the acceptable results of all the startup testing, operation of the Plant E. I. Hatch Unit 2 reactor is successful with the introduction of the GNF-2 fuel.

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ATTACHMENT 1 IN-SEQUENCE CRITICAL COLD SHUTDOWN MARGIN DEMONSTRATION Sequence A2 RWM Group 1 Fully Withdrawn RWM Group2 15 control rods fully withdrawn KsRo 0.98063 KcRIT 1.00455 Control Rod Density 0.7664 Reactor Coolant Temperature 125.0° F Reactivity Correction for Temperature -0.002 ~K Reactor Period 215 sec.

Reactivity Correction for Period 0.00031 ~K Corrected KcRIT 0.99924 ~K Cold Shutdown Margin 1.861% ~K Value of R O.Oo/o~K Value of B (conservative bias) 0.0030 ~K Minimum Cold Shutdown Margin 1.861% ~K Tech Spec Required Shutdown Margin 0.38%~K

ATTACHMENT 2 SCRAM TIME TESTING LOCATIONS TIME IN SECONDS TO NOTCH POSITION 46 36 26 06 Slowest Rods 30-43 0.358 0.906 1.447 2.513 30-43 0.358 0.906 1.447 2.513 30-43 0.358 0.906 1.447 2.513 26-31 0.280 0.835 1.396 2.591 Fastest Rods 06-19 0.228 0.700 1.198 2.257 42-27 0.249 0.698 1.159 2.083 42-27 0.249 0.698 1.159 2.083 42-27 0.249 0.698 1.159 2.083 Average (All 0.256 0.760 1.276 2.328 Rods)

ATTACHMENT 3 REACTIVITY ANOMALY CALCULATION UNIT 2 CYCLE 24 SEQUENCE: A2 DATE PERFORMED 03/19/2015 THERMAL POWER (MW!h) CMWT 2768.4 RATED THERMAL POWER (MW1h) 2804.0 CORE FLOW (Mib/hr) WT 78.93 RATED CORE FLOW (Mib/hr) 77.00 CORE XENON CONCENTRATION -2.20 XE/RATED 0.987 CYCLE EXPOSURE (MWD/sT) 82.1 EIGENVALUE 1.0056 PREDICTED keff = 1.0060

+1% VALUE= 1.0160 -1% VALUE= 0.9960