NL-03-1386, Startup Test Report

Startup Test Report

ML031780018
Person / Time
Site:	Hatch
Issue date:	06/24/2003
From:	Sumner H Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-03-1386
Download: ML031780018 (19)

Text

H.L Sumner. Jr. Southern Nuclear Vice President Operating Company, Inc.

Hatch Project Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 SOUTHERIL June 24, 2003 COMPANY Energy to Serve Your World' Docket No.: 50-366 NL-03-1386 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report Ladies and Gentlemen:

In accordance with the requirements of Unit 2 Final Safety Analysis Report Section 13.6.4, Southern Nuclear Operating Company (SNC), Plant Hatch, hereby submits the Unit 2 Startup Test Report for Operating Cycle 18. This report summarizes the startup testing performed on Unit 2 following the seventeenth refueling outage. The report is required due to the first use, other than as lead use assemblies, of GE14 fuel assemblies.

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

If you have any questions, please advise.

Sincerely, H. L. Sumner, Jr.

HLS/OCV

Enclosure:

Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 cc: Southern Nuclear Operating Company Mr. J. D. Woodard, Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch Document Services RTYPE: CHAO2.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. S. D. Bloom, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18

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 seventeeth refueling outage. This report is being submitted due to a reload batch of 232 GE14 fuel assemblies that were loaded for Cycle 18. The GE14 fuel design has not previously been utilized in bulk on Unit 2, although four GE14 Lead Use Assemblies (LUAs) successfully completed their third cycle at the end of Cycle 17.

This report consists of a summary 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 18. These tests demonstrate the successful operation of the Unit 2 reactor with the introduction of the GE14 fuel design into production use.

1.2 Plant Description The Edwin I. Hatch Nuclear Power Plant Unit 2 is a General Electric design single-cycle boiling water reactor (BWR/4). Plant Hatch Unit 2 is rated at 2763 MW(th) with a generator rating at this power of 900 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 30, 2003, after completing a 30 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 Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18

• Core Verification

• Control Rod Drive Timing

• Full Core Shutdown Margin Demonstration

• 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. Level 2 criteria do not constitute a test failure or acceptance; they serve as information only.

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Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 2.0 CYCLE DESIGN

SUMMARY

2.1 Core Design Summarv The Unit 2 Cycle 18 core is designed to operate approximately 634 effective full power days (EFPDs) at rated thermal power conditions, which includes extension from increased core flow and an expected mid-cycle power uprate of 1.5% rated thermal power. The fuel is arranged in a conventional core loading designed to achieve 16435 MWD/sT incremental energy exposure. All 232 fresh GE14 assemblies have an enrichment of 3.98 wI%. The loading pattern in this cycle is octant symmetric.

2.2 Reactivity/Thermal Limit Margins The two parameters that describe the global behavior of the core throughout the cycle are hot excess reactivity (HER) and cold shutdown margin (CSDM).

The beginning of cycle (BOC) + 200.0 MWD/sT HER is 1.68%, and the peak HER is 2.75% at 9500.0 MWD/sT.

The expected minimum CSDM of 1.366% AK occurs at 13.0 MWD/sT for the as-burned, as-loaded core. In-sequence critical calculations do not identify any high notch worths around the expected critical rod pattern at BOC. The Hatch-2 Cycle 18 calculated core parameters are delineated in Table 2.1.

Target rod pattern recommendations are calculated in 0.5 GWD/sT exposure increments.

Thermal margin design goals of 10%, 10%, and 7% for MFLPD, MAPRAT, and MFLCPR, respectively, are met throughout the cycle for these rod patterns. However, some MAPRAT and MFLCPR problems are expected late in the cycle when withdrawing deep control rods beyond notch 10. This issue can be addressed through additional load reductions for pattern adjustments.

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Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 Table 2.1 Cycle 18 Calculated Parameters BOC Core Average Exposure 12,057.0 MWD/sT Cycle 18 Core Weight 108.232 sT Cycle Energy (Rated Power) 16,435.0 MWD/sT (634.0 EFPDs)

Uncertainty in Energy +314.0 MWD/sT

-320.0 MWD/sT Cold Shutdown Margin BOC 1.683 %

R 0.047%

B 0.300%

Hot Excess Reactivity 200 MWDlsT (min) 1.68%

9500 MWD/sT (peak) 2.75%

2.3 Fuel Summary All fuel assemblies loaded in Cycle 18 have barrier cladding. A set of "soft-startup" preconditioning guidelines have been established and are applied to selected fuel assemblies during the first sequence of cycle operation. These fuel assemblies are chosen for preconditioning because they have been moved from lower power regions of the core in the previous cycle to higher power regions this cycle. In addition to "soft-startup" guidelines, additional Interim Preconditioning Guidelines have been established as a result of fuel failures at other BWR plants sharing similarities in fuel design with Hatch.

Table 2.2 provides a list of all batches loaded in Plant Hatch Unit 2 Cycle 18.

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Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 Table 2.2 Fuel Batches Loaded in Hatch-2 Cycle 18 IAT Bundl ID Batch Range Bundle Nomenclature H2R1 5 4 100 YJU405-YJU408 GE13-P9DTB378-6G5.0/6G4.0/lG2.0-OOT-YJU421-YJU436 146-T YJU445-YJU448 YJU453-YJU460 YJU469-YJU472 YJU481-YJU484 YJU493-YJU500 YJU505-YJU512 YJU521-YJU524 YJU533-YJU548 YJU553-YJU568 YJU573-YJUS76 YJU585-YJU588 H2R1 5 5 44 YJU397-YJU404 GE13-P9DTB378-6G5.0/6G4.0/lG2.0-lOOT-YJU409-YJU412 146-T YJU417-YJU420 YJU465-YJU468 YJU473-YJU476 YJU485-YJU492 YJU529-YJU532 YJU569-YJU572 YJ.U577-YJU580 H2RI6 7 104 YJZ745-YJZ748 GE13-P9DTB378-6G5.0/6G4.0-lOOT-146-T-YJZ753-YJZ756 2398 YJZ761-YJZ792 YJZ797-YJZ808 YJZ817-YJZ864 YJZ869-YJZ872 H2R16 8 56 YJZ873-YJZ928 GE13-P9DTB378-6G5.0/6G4.0/lG2.0-lOOT-146-T-2402 H2R16 9 24 YJZ749-YJZ752 GE13-P9DTB378-6G5.0/6G4.0-lOOT-146-T-YJZ757-YJZ760 2398 YJZ793-YJZ796 YJZ809-YJZ816 YJZ865-YJZ868 NL-03-1386 E-5

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 TAT Bundl ID l Batch e Range Bundle Nomenclature H2R17 13 160 JLH-00 1-LH160 GE14-P1ODNAB3984G7.0/lOG6.0-1OOT-150-

_____ _____ ~~T-26 15 H2R17 14 48 JLH161-JLH208 GE14-PlODNAB3984G7.0/1 IG6.0-1 OOT- 150-

_ ____ ____ ____ T -26 16 H2R17 15 24 JLH213-JLH236 GE14-PIODNAB3984G7.0/1 G6.0/lG2.0-

_____I l1OOT-150-T-2617 NL-03-1386 E-6

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 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 as specified by the General Electric 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 18 core verification was performed by use of an underwater TV camera 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 A full core verification was completed on March 19, 2003, in accordance with engineering procedures for fuel movement. The verification showed that all bundles were in their correct locations, in the correct orientation and properly seated.

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Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 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: N/A Level 2 Criteria: The insert and withdrawal drive time for each CRD must be between 38.4 and 57.6 seconds. In the event that a CRD fails to meet this criteria, the applicable drive must be adjusted and a new criteria of 43.2 to 52.8 seconds is applied to the adjusted drive.

3.2.3 Test Description Control rod drive timing is generally 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 28, 2003, for all 137 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, with the following exceptions:

(1) Seventeen control rods could not be moved from the full-in position using normal drive water pressure. By procedure, drive water pressure was increased in discrete steps until control rod movement was successful. These control rods could only be moved at the elevated drive water pressure and sixteen were successfully timed after movement was established (see item 2).

(2) One control rod, 22-35, was found to have an excessively fast withdraw speed of 32.6 seconds when timed from notch position 00 to position 48. This condition, although deficient by procedural requirements, was deternined to be within acceptable limits for the Rod Withdrawal Error analysis.

Note: These CRD mechanisms have been documented via the Corrective Action Program and are currently being trended and evaluated for repair and/or replacement.

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Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 3.3 Full Core Shutdown Margin Demonstration 3.3.1 Purpose To demonstrate the reactor can be made subcritical for any reactivity condition during Cycle 18 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 0.3 8% A 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 full core shutdown margin demonstration was performed analytically during the Plant Hatch Unit 2 Cycle 18 BOC in-sequence critical 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 are used to adjust the startup condition to cold conditions with the highest worth control rod fully withdrawn.

3.3.4 Test Results The full core shutdown margin demonstration was performed on March 28, 2003, in accordance with core calculation procedures for shutdown margin demonstration.

Results of this calculation yielded a BOC cold shutdown margin of 1.439% AK. The minimum cold shutdown margin was calculated at 1.392% AK, since the MCSDM does not occur at BOC, but at 13.0 GWD/sT of exposure. A summary of the shutdown margin demonstration is given in Attachment 1 of this report.

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Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 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 I Criteria: N/A Level 2 Criteria: N/A 3.4.3 Test Description The cold critical eigenvalue is the assumed value of the PANACEA 3-D simulator model Keff at which criticality is achieved with the reactor in a xenon-free state and the coolant is 68 degrees Fahrenheit. 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 is 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 shutdown margin 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 intial beginning of cycle startup for Plant Hatch Unit 2 Cycle 18 was performed on March 28, 2003. The observed reactor core conditions when a critical state was achieved are listed in Attachment 1.

A cold critical eigenvalue of 1.002018 was calculated from the actual critical data given above. This compares quite well with the initial estimate for the cold critical eigenvalue of 1.002000.

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Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 3.5 Local Power Range Monitor (LPRM) Calibration 3.5.1 Purpose To calibrate the local power range monitors (LPRMs).

3.5.2 Acceptance Criteria Level 1 Criteria: Per plant procedures.

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 on April 7, 2003. Average 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.

Level 2 criteria: N/A NL-03-1386 -1

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 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 will be obtained from the process computer programs OD3 (Core Thermal Power/Flow Log) or CTP (Core Thermal Power program),

or from the Official Monitor case, or from a manual heat balance in accordance with plant operating procedures.

3.6.4 Test Results APRM calibration was performed in accordance with plant operating procedures at approximately 7%, 15%, 21%, 48%, 56%, 60%, 73%, 92% and 99% 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 I 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 greater than or equal to 800 psig shall not exceed the following:

Notch Position Average from Fully Insertion Withdrawn Time (sec) 46 0.44 36 1.08 26 1.83 06 3.35 NL-03-1386 E-12

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 (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.

3. No more than 20% of the control rods tested are determined to be SLOW.

(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. Scram times for all control rods were acceptable. A summary of the results is given in Attachment 2 of this report.

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Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 3.8 Core Performance 3.8.1 Purpose To evaluate the core performance parameters to assure plant thermal limits are maintained during the ascension to rated conditions.

3.8.2 Acceptance Criteria Level I criteria: The following thermal limits are S 1.000:

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 was increased, core thermal limits were evaluated at various levels up to 100%

rated thermal power. 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 regularly monitored during power ascension. The surveillance procedure was performed satisfactorily at various levels as indicated below:

Power l Thermal Limit Level MFLCPR MFLPD MAPRAT 20.2% 0.566 0.277 0.456 39.5% 0.806 0.462 0.678 47.7% 0.775 0.513 0.705 62.6% 0.829 0.582 0.761 71.9% 0.877 0.623 0.761 85.7% 0.855 0.740 0.823 97.0% 0.946 0.861 0.867 99.9% 0.946 0.898 0.885 NL-03-1386 E-14

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 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 corrected control rod density shall not differ from the predicted control rod density equivalent by more than + 1%A K.

Level 2 criteria: N/A 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. Since anticipated operation or unanticipated events may place the reactor in a condition other than that for which the baseline anomaly curve was developed, the actual control rod density is corrected for off-rated conditions. The corrected control rod density is then compared to the reactivity anomaly curve provided in the Cycle Management Report to ensure that the corrected control rod density is within a + 1% AK acceptance band about the curve.

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 April 5, 2003. The corrected control rod density was well within the acceptance criteria range as specified above. The results of this calculation are given in Attachment 3 of this report.

4.0 CONCLUSION

S As the results of the startup testing indicate, operation of the Plant E. I. Hatch Unit 2 reactor is successful with the introduction of the GE14 fuel.

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.Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 ATTACHMENT 1 FULL CORE SHUTDOWN MARGIN DEMONSTRATION Sequence A2 RWM Group 1 Fully Withdrawn RWM Group 2 11 control rods withdrawn full out (notch 48), 5 control rods withdrawn to notch 24, and the 12th control rod in the group (38-19) withdrawn to notch 28 KSRO 0.98317 KCRf 1.00441 Control Rod Density 0.7731 Reactor Coolant Temperature 1550 F Reactivity Correction for Temperature -0.0035 AK Reactor Period 187 sec.

Reactivity Correction for Period 0.00035 AK Cold Shutdown Margin 1.439% AK Value of R 0.047% AK Value of B (conservative bias) 0.003 AK Minimum Cold Shutdown Margin 1.392% AK Tech Spec Required Shutdown Margin 0.38% AK NL-03-1386 E-16

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 ATTACHMENT 2 SCRAM TIME TESTING LOCATIONS TIME IN SECONDS TO NOTCH POSITION 46 36 26 06 Tech Spec Limts 0.440 1.080 1.830 3.350 Slowest Rods with slowest notch identified in bold 30-23 0.293 0.818 1.353 2.508 42-35 0.288 0.856 1.410 2.562 46-27 0.284 0.852 1.449 2.666 Fastest Rods with fastest notch identified in bold 18-03 0.227 0.724 1.245 2.318 18-51 0.227 0.720 1.232 2.292 02-35 0.242 0.710 1.210 2.230 No control rods met the criteria to be decleared "SLOW" Average (All Rods) 0.256 0.775 1.312 2.412 NL-03-1386 E-17

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Startup Test Report for Cycle 18 ATTACHMENT 3 REACTIVITY ANOMALY CALCULATION UNIT 2 CYCLE 18 SEQUENCE: A2 DATE PERFORMED 04/5/03 THERMAL POWER (MWt) CMWT 2758.0 RATED THERMAL POWER (MWtb) 2804.0 CORE FLOW (Mlb/hr) WT 75.05 RATED CORE FLOW (Mlb/hr) 77.0 DOME PRESSURE (psia) PR 1049.4 RATED PRESSURE (psia) 1050.0 SUBCOOL1NG (BTU/lb) DHS 21.58 DESIGN INLET SUBCOOLING (BTU/lb) 21.70 CYCLE EXPOSURE (MWD/sT) 127.1 CONTROL ROD DENSITY CRD 0.0718 CORRECTED CRD = CRD + CORRECTION CORRECTION = -2.1857E-1 x (1.0-(CMWTIRATED CORE THERMAL POWER))

+1.3385E-1 x (1.0-(WT/RATED CORE FLOW))

+1.61764E-3 x (DESIGN INLET SUBCOOLING-DHS)

+4.08085E-5 x (RATED PRESSURE - PR)

CORRECTION = 0.0000 CORRECTED CRD = 0.0718 + 0.0000 = 0.07 18 PREDICTED CRD = 0.0728

+1% VALUE = 0.1160 -1% VALUE = 0.0296 NL-03-1386 E-18