ML19322C130

Startup Rept,Unit 3,for 750314

ML19322C130
Person / Time
Site:	Oconee
Issue date:	03/14/1975
From:	DUKE POWER CO.
To:
References
NUDOCS 8001090548
Download: ML19322C130 (176)

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e t 6 F DUKE POWER C O M P.A N Y OCONEE NUCLEAR STATION DOCKET NO. 50-287 LICENSE NO. DPR-55 i l ,STARTUP REPORT UNIT 3 ( MARCH 14, 1975 .i i i i 4 6 l t m

t 8 p H> as Fm O m OO2 -4 fft 2 -4 M 4 D

e e TABLE OF CONTENTS ( Section Page

1.0 INTRODUCTION

AND

SUMMARY

1.1-1

1.1 INTRODUCTION

1.1-1 1.2

SUMMARY

1.2-1 1.2.1 GENERAL 1.2-1 1.2.2 INITIAL FUEL LOADING 1.2-1 1.2.3 TESTING PRIOR TO POWER' ESCALATION 1.2-1 1.2.4 POWER ESCALATION TESTS 1.2-2 2.0 INITIAL FUEL LOADING 2.0-1 3.0 TESTING FRIOR TO POWER ESCALATION 3.0-1 3.1 REACTOR COOLANT FLOW AND FLOW COASTDOWN TEST 3.1-1 3.1.1 PURPOSE 3.1-1 3.1.2 TEST METHOD 3.1-1 3.1.3 EVALUATION OF TEST RESULTS 3.1-1 3.

1.4 CONCLUSION

S 3.1-1 3.2 CONTROL ROD DRIVE DROP TIME TEST 3.2-1 3.2-1 3.2.1 PURPOSE 3.2.2 TEST METHOD

3. 2-1 3.2.3 F'

LUATION OF TEST RESULTS 3.2-1 3.2.4 sNCLUSIONS 3.2-2 3.3 ZERO POWER PHYSICS TEST 3.3-1 3.3.1 PURPOSE 3.3-1 3.3.2 TEST METHOD 3.3-1 3.3.3 EVALUATION OF TEST RESULTS .3.3, 3.3.3.1 Initial Criticality 3.3-1 1 3.3.3.2 Nuclear Instrumentation Overlap 3.3-2 3.3.3.3 "All Rods Out" Critical Boron Coiteentration 3.3-3 3.3.3.4 Control Rod Group Worths 3.3-3 d 3.3.3.5 Soluble Poison Worths 3.3-4 3.3.3.6 Ejected Control Rod Worth 3.3-4 3.3.3.7 Stuck Control Rod Worth 3.3-5 3.3.3.8 Temperature Coefficient of Reactivity 3.3-5 3.

3.4 CONCLUSION

S 3.3-6 4 4.0 POWEP ESCALiTION TESTS 4.021 i L 4.1 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER 4.1-1 '4.1.1 - PURPOSE 4.1-1 4.1.2 TEST METHOD -4.1 4.1.3 EVALUATION OF TEST RESULTS -4.1-2 4.

1.4 CONCLUSION

S 4.1-2 4.2 BIOLOGICAL SHIELD SURVEY 4.2-1 4.2-1 4.2.1 . PURPOSE 4.2.2 TEST METHOD _4.2-1 4.2.3 EVALUATION OF TEST RESULTS 4.2-1 1 e

o e Section Pajge 4.

2.4 CONCLUSION

S 4.2-1 4.3 REACTIVITY COEFFICIENTS AT POWER 4.3-1 4.3.1 PURPOSE 4.3-1 4.3.2 TEST METHOD 4.3-1 4.3.3 EVALUATION OF TEST RESULTS 4.3-1 4.

3.4 CONCLUSION

S 4.3-2 4.4 CORE POWER DISTRIBUTION 4.4-1 4.4.1 PURPOSE 4.4-1 4.4.2 TEST METHOD 4.4-1 4.4.3 EVALUATION OF TEST RESULTS 4.4-1 4.4.3.1 Steady-State Core Power Distributions at Equilibrium Xenon, Normal Operating Control Rod Configt tation Conditions 4.4-2 4.4.3.1.1 Radial and Total Peaking Factors 4.4-2 4.4.3.1.2 Minimum DNBR 4.4-3 4.4.3.1.3 Maximum Linear Heat Rate 4.4-3 4.4.3.1.4 Quadrant Power Tilt and Axial Imbalance 4.4-3 4.4.3.2 Core Power Distribution at 75% FP for Different Control Rod Configurations 4.4-3 4.

4.4 CONCLUSION

S 4.4-4 4.5 TURBINE / REACTOR TRIP TEST 4.5-1 4.5.1 PURPOSE 4.5-1 4.5.2 TEST METHOD 4.5-1 4.5.3 EVALUATION OF TEST RESULTS 4.5-1 4.5.4 CONCLdSIONS 4.5-2 4.6 ROD WORTH AT POWER 4.6-1 4.6.1 PURPOSE 4.6-1 4.6.2 TEST METHOD 4.6-1 4.6.3 EVALUATION OF TEST RESULTS 4.6-2 4.

6.4 CONCLUSION

S 4.6-2 4.7 POWER IMBALANCE DETECTOR CORRELATION TEST 4.7-1 4.7.1 PURPOSE 4.7-1 4.7.2 TEST METHOD 4.7-1 4.7.3 EVALUATION OF TEST RESULTS 4.7-2 4.

7.4 CONCLUSION

S 4.7-2 4.8 NSSS HEAT BALANCE 4.8-1 4.8.1 PURPOSE 4.8-1 4.8.2 TEST METHOD 4.8-1 4.8.3 EVALUATION OF TEST RESULTS 4.8-2 4.8.3.1 Primary and Secondary Heat Balance Calculation 4.8-2 4.8.3.2 Reactor Coolant Flow Determination 4.8-2 4.

8.4 CONCLUSION

S 4.8-3 4.9 UNIT LOAD STEADY-STATE TEST 4.9-1 4.9.1 PURPOSE 4.9-1 4.9.2 TEST METHOD 4.9-1 4.9.3 EVALUATION OF TEST RESULTS 4.9-2 4.

9.4 CONCLUSION

S 4.9-2 4.10 UNIT LOAD TRANSIENT TEST 4.10-1 4.10.1 PURPOSE 4.10-1 4.10.2 TEST METHOD 4.10-1 4.10.3 EVALUATION OF TEST RESULTS 4.10-2 l 11

4 o o- ~. f-Section Page 4.10.3.1 Integr.ated Control System Transient Test at 40% FP 4.10-f 4.10.3.2 Integrated Contral System Transient Test at 75% FP 4.10-2 4.10.4-CONCLUSIONS 4.10-2 4.11 PSEUDO CONTROL ROD EJECTION TEST 4.11-1 4.11.1 PURPOSE 4.11-1 4.11.2 TEST METHOD 4.11-1 4.11.3 EVALUATION OF TEST RESULTS 4.11-1 4.

11.4 CONCLUSION

S 4.11-2 4.12 DROPPED CONTROL ROD TEST 4.12-1 4.12.1 PURPOSE 4.12-1 4.12.2 TEST METHOD 4.12-1 4.12.3 EVALUATION OF TEST RESULTS 4.12-2 4.

12.4 CONCLUSION

S 4.12-2 ef 9 a 6 4 111 I

o e LIST OF TABLES Table Title 3.1-1 REACTOR COOLANT FLOW AT REFERENCE CONDITIONS OF 532 F AND 2155 PSIG FOR VARIOUS PUMP COMBINATIONS 3.3-1 COMPARISON OF CALCULATED ROD WORTH WITH THAT MEASURED AND EXPECTED BY THE R0D DROP METHOD 3.3-2 COMPARISON OF CALCULATED AND MEASURED CONTROL ROD GROUP WORTHS AT A MODERATOR TEMPERATURE OF 532 F AND APSR'S AT 20% WD 3.3-3 DIFFERENTIAL BORON REACTIVITY WORTH MEASUREMENTS DURING ZERO POWER PHYSICS TEST 4.0-1 LIST OF TESTS PERFORMED DURING POWER ESCALATION 4.1-1

SUMMARY

OF NUCLEAR INSTRUMENTATION CALIBRATIONS AT POWER PER-FORMED AS REQUIRED BY THE POWER ESCALATION PROGRAM AND SECTION 4.1 0F TECHNICAL SPECIFICATIONS 4.2-1

SUMMARY

OF MAXIMUM AND AVERAGE DOSE RATES 4.3-1

SUMMARY

OF MEASURED, CALCULATED, AND PREDICTED COEFFICIENTS OF REACTIVITY AT POWER 4.4-1

SUMMARY

OF MEASURED CORE POWER DISTRIBUTION RESULTS AT EQUILIBRIUM XENON CONDITIONS FOR VARIOUS CONTROL ROD PATTERNS AND CORE POWER LEVELS OF 15, 40, 75, AND 100 PERCENT FULL POWER 4.4-2 MINIMUM DNBR AND MMIMUM LHR WORST CASE UNCERTAINTY FACTORS 4.4-3 MEASURED CORE POWER DISTRIBUTION RESULTS AT 15% FP 4.4-4 MEASURED CORE POWER DISTRIBUTION RESULTS AT 40% FP 4.4-5 MEASURED CORE POWER DISTRIBUTION RESULTS AT 75% FP 4.4-6 MEASURED CORE POWER DISTRIBUTION RESULTS AT 100% FP ,~ 4.4-7 COMPARISON OF MEASURED AND PREDICTED MAXIMUM RADIAL AND TOTAL i PEAKING FACTORS 4.4-8 MEASURED CORE POWER DISTRIBUTION RESULTS AT 75% FP DURING CONTROL ROD MANEUVERING, WITH GROUPS 5-8 AT 100, 25, 00, AND l'0% WD, RESPECTIVELY 4.4-9 MEASURED CORE POWER DISTRIBUT' ION RESULTS AT 75% FP DURING CONTROL ' ROD MANEUVERING, WITH CROUPS 5-8 AT 100, 75, 03, AND 19% WD, RESPECTIVELY iv

e o ~ LIST OF TABLES (Cont'd) Table Title 4.4-10 MEASURED CORE POWER DISTRIBUTION RESULTS AT 75% FP DURING CONTROL ROD MANEUVERING, WITH CROUPS 5-8 AT 100,100, 24, AND 04% WD, RESPECTIVELY 4.4-11 MEASURED CORE POWER DISTRIBUTION RESULTS AT 75% FP DURING CONTROL ROD MANEUVERING, WITH GROUPS 5-8 AT 100, 100, 75, AND 24% WD, RESPECTIVELY 4.4-12 MINIMUM DNBR AND MAXDIUM LHR ANALYSIS FOR STANDARD CORE POWER DISTRIBUTION TEST = 4.5-1

SUMMARY

OF MINIMUM AND MAXIMUM DEVIATIONS IN UNIT PARAMETERS DURING THE PERFORMANCE OF TURBINE REACTOR TRIP TEST AT 40% FP ~ i 4.6-1 MEASURED DIFFERENTIAL ROD WORTHS DURING REACTIVITY COEFFICIENTS AT POWER MEASURDfENTE 4.7-1 MINIMUM DNBR AND MAXDIUM LikR ANALYSIS FOR POWER DiBALANCE DETECTOR CORRELATION TEST 4.8-1

SUMMARY

OF HEAT BALANCES PERFORMED DURING POWER ESCALATION TESTING 4.8-2 RESULTS.0F REACTOR COOLANT SYSTDI PRIMARY FLOW MEASUREMENTS 4.9.1 AVERACE UNIT PARAMETERS AT VARIOUS TEST PLATEAUS FOR OCONEE UNIT 3 DURING STEADY-STATE CONDITIONS 4.9-2 MAXI"UM DEVIATION IN AVERAGE UNIT PARAMETERS AT VARIOUS TEST PLAT AUS FOR OCONEE 3 DURING STEADY-STATE CONDITIONS 4.10-1 GENER L SU101ARY OF TRANSIENTS REQUIRED BY UNIT LOAD TRANSIENT TEST 4.10-2 TRANSIENT DATA OBTAINED DURING TiiE PERFORMANCE OF UNIT LOAD TRANSIENT TEST AT 40% FP j 4.10-3. TRANSIENT DATA OBTAINED DURING Tile ~ PERFORMANCE OF UNIT LOAD . TRANSIENT TEST AT 75% FP i 4.11-1 DETERMINATION OP PSEUDO EJECTED. CONTROL ROD WORTH 4 4.11-2

SUMMARY

OF CORE POWER DISTRIBUTIONS AND THERMAL-HYDRAULICS DATA TAKEN DURING THE PSEUDO RCD EJECTION TEST t 4.11-3 MEASURED CORE POWER DISTRIBUTION RESULTS WITH PSEUDO EJECTED CONTROL R0D 0% WITHDRAWN FROM '1HE CORE AT 40% FP 4.11-4: MEASURED' CORE POWER DISTRIBUTION RESULTS WITH PSEUD 0 EJECTED CONTROL' ROD 100% WITHDRAWN FROM THE CORE AT 40% FP ^' F V

o LIST OF TABLES (Cont'd) Table Title 4.12-1

SUMMARY

OF CORE POWER DISTRIBUTION AND THERMAL HYDRAULICS DATA TAKEN DURING THE DROPPED C01TIROL ROD TEST 4.12-2 MINIMUM DNBR AND L AXIMUM LHR ANALYSIS FOR THE DROPPED CONTROL ROD TEST 4.12-3 MEASURED CORE POWER DISTRIBUTION RESULTS WITH DROPPED CONTROL ROD AT 77% WD AT 40% FP 4.12-4 MEASURED CORE POWER DISTRIBUTION RESULTS WITH DROPPED C0h"IROL ROD ' AT 50% WD AT 40% FP 4.12-5 MEASURED CORE POWER DISTRIBUTION RESULTS WITH DROPPED CONTROL ROD AT 0% WD AT 40% FP e 9 + e 0 D vi 4

,c v 8 a LIST OF FIGURES Figure Title 2.0-1 FINAL FUEL LOADING DISTRIBUTION FOR OCONEE UNIT 3, CORE 1 3.1-1 ACCEPTANCE CRITERIA FOR REACTOR COOLANT FLOW DURING PUMP COASTDOWN' 3.1-2 MEASURED REACTOR COOLANT FLOW RATE FOLI4 WING THE TRIP OF PUMP 3B1 FROM FOUR PUMP INITIAL CONDITION AT 532 F D 3.1-3 MEASURED REACTOR COOLANT FLOW RATE FOLLOWING THE TRIP OF PUMPS 3Al AND 3B1 FROM FOUR PUMP INITIAL CONDITION AT 532 F 3.1-4 MEASURED REACTOR COOLANT FLOW RATE FOLLOWING THE TRIP OF ALL PUMPS FR(E FOUR PUMP INITIAL CONDITION AT 532 F 3.1-5 MEASURED REACTOR COOLANT FLOW RATE FOI, LOWING,THE TRIP OF PUMP 3Al FROM THREE PUMP INITIAL CONDITION AT 532 F 3.1-6 MEASURED REACTOR COOLANT FLOW RATE FOLLOWING THE TRIP OF ALL PUMPS FROM THREE PUMP INITIAL CONDITION AT 532 F 3.3-1 INVERSE MULTIPLICATION FOR DETECTORS NI-l AND NI-2 VERSUS TIME FOR APPROACH TO INITIAL CRITICALITY 3.3-2 INVERSE MULTIPLICATION FOR DETECTORS NI-l AND NI-2 VERSUS RCS BORON CONCENTRATION FOR APPROACH TO INITIAL CRITICALITY I 3.3-3 CORE POWER VERSUS DETECTOR RESPONSE 3.3-4 CONTROL ROD GROUP LOCATIONS FOR CORE BURNUP UP TO 250 EFPD 3.3-5 NORMALIZED DIFFERENTIAL WORTHS OF CONTROL ROD GROUPS 5-7 VERSUS WITHDRAWAL POSITIONS, FOR BEGINNING-OF-LIFE,.ZERO POWER CONDITIONS 3.3-6 NORMALIZED INIEGRAL WORTH OF CONTROL iis0D GROUPS 5-7 VERSUS WITH-DRAWA. POSITION, FOR BEGINNING-OF-LIFE, ZERO POWER CONDITIONS 3.3-7 NORMALIZED INTEGRAL WORTH OF CONTROL ROD GROUP 8 VERSUS WITH-DRAWAL POSITION, FOR BEGINNING-OF-LIFE, ZERO POWER CONDITIONS 3.3-8 COMPARISON OF MEASURED AND PREDICTED IliTEGRAL REACTIVITY WORTH OF-SOLUBLE BORON VERSUS BORON CONCENTRATION 3.3-9 INVERSE TIME VERSUS MEASURED REACTIVITY FOR THE DETERMINATION OF i THE EJECTED CONTROL R0D REACTIVITY WORTH 3.3-10' COMPARISON OF MEASURED TEMPERATURE CdEFFICIENTS AND CALCULATED MODERATORCOEFFICIENTSOFREACTIVITY}0THEIRRESPECTIV,EPRE-4 DICTED VALUES AT BOL, ZERO POWER, 532 F, CONDITIONS 4.0-1 AVERAGE DAILY POWER LEVEL VERSUS TIME DURING THE POWER ESCALATION TEST PROGRAM t t vii L.

t LIST OF FIGURES (Cont'd) Figure Title 4.1-1 NUCLEAR INSTRUMENTATION FLUX RANGES 4.3-1 TEMPERATURE COEFFICIENT OF REACTIVITY VERSUS BORON CONCENTRATION FOR CRITICAL BORON AND ROD CONDITIONS AT 579 F, 2155 PSIG AND 0 EFPD 4.3-2 MOLERATOR COEFFICIENT OF REACTIVITY VERSUS BORON CONCENTRATION FOR CRITICAL BORON AND ROD CONDITIONS AT 579 F, 2155 PSIG, AND 0 EFPD 4.3-3 POWER D0PPLER COEFFICIENT OF REACTIVITY VERSUS POWER LEVEL 4.4-1 POWER-IMBALANCE ENVELOPE 4.4-2 MEASURED RADIAL CORE POWER DISTRIBUTIONS ALONG THE X-Z PLANE WITH NORMAL OPERATING CONTROL ROD CONFIGURATION 4.4-3 COMPARISONOFPREDICTEDANDMkASUREDRADIALPEAKINGFACTORSAT STEADY-STATE, 2-D EQUILIBRIUM XENON, 15% FP CONDITIONS 4.4-4 COMPARIS0N OF PREDICTED AND MEASURED TOTAL PEAKING FACTORS AT STEADY-STATE, 2-D EQUILIBRIUM XENON,15% FP CONDITIONS 4.4-5 COMPARISON OF PREDICTED AND MEASURED RADIAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 40% FP CONDITIONS 4.4-6 COMPARISON OF PREDICTED AND MEASURED TOTAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 40% FP CONDITIONS 4.4-7 COMPARISON OF PREDICTED AND MEASURED RADIAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 75% FP CONDITIONS 4.4-8 COMPARISON OF PREDICTED AND MEASURED TOTAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 75% FP CONDITIONS 4.4-9 COMPARISON OF PREDICTED AND MEASURED RADIAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 100% FP CONDITIONS 4.4-10 COMPARISON OF PREDICTED AND MEASURED TOTAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON,100% FP CONDITIONS 1 4.4-11 MEASURED RADIAL CORE POWER DISTRIBUTIONS ALONG THE X-Z PLANE OBTAINED DURING CONTROL ROD MANEUVERING AT 75%.FP i .4.4-12 HOT CHANNEL MINIMUM DNBR VERSUS CORE POWER LEVEL viii

o LIST OF FIGURES (Cont'd) Figure Title 4.6-1 REACTIVITY WORTH OF CONTROL RODS VERSUS WITHDRAWAL POSITION, ZERO POWER, BOL, APSR'S AT 20% WITHDRAWN, 532 F AND 2155 PSIG-ROD GROUPS 5, 6, 7, 8 4.6-2 RF *TIVITY WORTH OF CONTROL RODS VERSUS WITHDRAWAL POSITION, z a POWER, BOL, APSR'S AT 35% WITHDRAWN, 532 F AND 2155 PSIG-ROD GROUPS 5, 6, 7, 8 4.6-3 REACTIVITY WORTH OF CONTROL RODS VERSUS WITHDRAWAL POSITION, ZERO POWER, BOL, APSR'S AT 100% WITHDRAWN, 532 F AND 2155 PSIG-ROD GROUPS 5, 6, 7 4.6-4 REACTIVITY WORTH OF CONTROL RODS VERSUS WITHDRAWAL POSITION, FULL POWER, BOL, APSR'S AT 20% WITHDRAWN, 579 F AND 2155 PSIG-ROD GROUPS 5, 6, 7, 8 ^ 4.6-5 REACTIVITY WORTH OF CONTROL RODS VERSUS WITHDRAWAL POSITION, FULL POWER, BOL, APSR'S AT 35% WITHDRAWN, 579 F AND 2155 PSIG-ROD GROUPS 5, 6, 7, 8 4.6-6 REACTIVITY WORTH OF CONTROL RODS VERSUS WITHDRAWAL POSITION, FULL POWER, BOL, APSR'S AT 100% WITHDRAWN, 579 F AND 2155 PSIG-ROD GROUPS 5, 6, 7 4.7-1 REACTOR PROTECTION SYSTEM POWER / IMBALANCE / FLOW TRIP ENVELOPE 4.7-2 ACCEPTANCE CRITERIA BETWEEN INCORE AND OUT-OF-CORE OFFSETS FOR POWER IMBALANCE DETECTOR CORRELATION TEST AT 75% FP 4.7-3 INCORE OFFSET VERSUS OUT-OF-CORE OFFSET DURING THE POWER UfBALANCE SCAN AT 40% FP WITH A GAIN FACTOR OF 3.35 4.7-4 INCORE OFFSET VERSUS OUT-OF-CORE OFFSET DURING THE POWER IMBALANCE SCAN AT 75% FP WITH A GAIN FACTOR OF 3.75 4.7-5 MINIMUM DNBR, MAXIMUM LHR, TOTAL PEAKING FACTOR, AhT RADIAL PEAKING FACTOR MEASURED DURING THE IMBALANCE SCAN AT 40% FP 4.9-1 REACTOR COOLANT SYSTEM TDIPERATURE VERSUS POWER LEVEL WITH FOUR REACTOR COOLANT PUMPS OPERATING 4.9-2 STEAM GENERATOR OUTLET PRE'SSURE VERSUS POWER LEVEL 4.9-3 STEAM GENERATOR STARTUP LEVEL VERSUS POWER LEVEL 4.9-4 STEAM GENERATOR OPERATING RANGE VERSUS POWER LEVEL, 4.9-5 STEAM GENERATOR STEAM TEMPERATURE VERSUS POWER LEVEL ix s

LIST OF FIGURES (Cont'd) Figure Title 4.9.6 FEEDWATER TEMPERATURE VERSUS POWER LEVEL 4.9-7 TOTAL F5EDWATER FLOW VERSUS POWER LEVEL 4.9-8 REACTOR COOLANT AVERAGE TEMPERATURE VERSUS POWER LEVEL 4.10-1 UNIT LOAD TRANSIENT TEST AT 40% FP 4.10-2 UNIT LOAD TRANSIENT TEST AT 40% FP 4.10-3 UNIT LOAD TRANSIENT TEST AT 40% FP 4.10-4 UNIT LOAD TRANSIENT TEST AT 40% FP 4.10-5 UNIT LOAD TRANSIENT TEST AT 75% FP 4.10-6 UNIT LOAD TRANSIENT TEST AT 75% FP 4.10-7 UNIT LOAD TRANSIENT TEST AT 75% FP 4.10-8 UNIT LOAD TRANSIENT TEST AT 75% FP 4.10-9 UNIT LOAD TRANSIENT TEST AT 75% FP 4.11-1 EJECTED ROD WORTH AND LOCATION AT 40% FP PLATEAU 4.11-2 MEASURED RADIAL CORE POWER DISTRIBUTIONS FOR CONTROL ROD POSITIONS OF 0% WD AND 100% WD 4.12-1 DROPPED R0D WORTH AND POSITION AT 40% FP PLATEAU 4.12-2 DIFFERENTIAL AND INTEGRAL ROD WORTH FOR CONIROL ROD ASSEMBLY AT CORE LOCATION H-12 VERSUS ROD POSITION DURING DROPPED CONTROL R0D TEST 4.12-3 QUADRANT POWER TILT, INCORE AXIAL OFFSET, MINIMUM DNBR, AND MAXIMUM LHR VERSUS CONTROL R0D POSITION DURING DROPPED CONTROL ROD TEST 4.12-4 COMPARISON OF MEASURED RADIAL CORE POWER DISTRIBUTIONS WITH DROPPED CONTROL ROD AT 77, 50, AND 0% WD FROM THE CORE AT 40% FP ~

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1.0 INTRODUCTION

AND

1.1 INTRODUCTION

On July 19, 1974, the Atomic Energy Comsfssion issued Facility Operating License DPR-55 to Duke Power Company for Cconee Nuclear Station Unit 3. The first fuel assembly was inserted into the core on July 20, 1974, and initial fuel loading was completed on July 28, 1974. ion September 5, 1974, 0conee Nuclear Station Unit 3 successfully achieved initial criticality. Zero Power Physics testing, which commenced on September 5, 1974, was successfully completed on September 9, 1974. This Se8C program was conducted primarily at the reactor coolant temperature of 532 F. Following the completion of Zero Power Physics testing, initial power level escalation was conducted on September 11, 1974, and further power level escalations occurred as required testing was satisfactorily completed. Major power levels', as defined by the power escalation testing sequence, were initially achieved as follows: Power Level (Percent of Full Power - %FP) Date 15 September 11, 1974

4..

October 15, 1974 75 November 16, 1974 I 100 Dec ember 16, 1974 These test programs were designed to provide adequate assurance that the unit could be operated in a safe and efficient manner and were carried out in accordance with the requirements of the Oconee Nuclear Station Final Safety Analysis Report. Detailed written procedur2s, which specified the sequenze of tests, the parameters to be measured, and the conditions under which each test was to be performed were followed throughout the test program. This report addresses unit startup and power escalation testing through 2400 hours on January 15, 1975. At that time, all or part of several power escalation tests remained to be completed as follows: (a) Unit Loss of Electrical Load Test (b) Turbine / Reactor Trip Test - 100% FP portion (c) Unit Load Transient Test - 100% FP. portion (d) Loss of Control Room Test Following completion of the above items, appropriate information, supplementary to this report, will be compiled and submitted concerning the results of the testing. This report is prepared and submitted in accordance with the Technical Specifi-cation 6.6.1.1 and summarizes unit startup and power escalation testing and the results thereof for Oconee Nuclear Station Unit 3. 1.1-1

t 4 t 1.2

SUMMARY

1.2.1 GENERAL Oconee Unit 3 is the third of a series of nuclear steam supply systems, designed by the Babcock and Wilcox Company and rated at 886 Mwe (net), to be placed into service by Duke Power Company. As of 2400 hours on. January 15, 1975, the unit has been operated at power levels up to and including 100% FP. In general, the performance of the unit has been acceptable. Testing and operation of the Nuclear Steam Supply System has revealed few items which were other than predicted and none which adversely affected unit safety. The deficiencies encountered have been of a nature that would be expected during the initial startup of a unit of this magnitude; and based on an evaluation of unit startup and power escalation testing, it is concluded that the unit may be safely operated at full rated power. t Oconce Unit 3 secrtup and pgwer escalation testing, as addressed,by the various major sections of this report, is summarized below. 1.2.2 INITIAL FUEL LOADING The initial fuel loading was initiated on July 20, 1974, and was compbeted on July 28, 1974. Minor equipment problems delayed the fuel loading operation by over four days. In spite of this delay, the initisl fuel loading was completed in eight days in a safe and orderly manner. 1.2.3 TESTING PRIOR TO POWER ESCALATION Following initial fuel loading of Oconee Unit 3 acd prior to power escalation, the Reactor Coolant Pump Flow and Flow Coastdown Test, the Control Rod Drive Drop Time Test, and the Zero Power Physics Test were conducted. In all cases, applicable Technical Specification requirements were met. A brief sammary of each of these tests follows: (a) Reactor Coolant Pump Flow and Flow Coastdown Test The measured steady-state reactor coolant flow rates for all pump com-binations tested were within the specified minimum and maximum values. The measured reactor coolant flow during pump coastdowns satisfied the applicable acceptance criteria for all test conditions. (b) Control Rod Drive Drop Time Test Th'e rod drop times were below the Technical Specification limits of 1.66 seconds at full flow and 1.40 seconds at no flow conditions. 1.2-1

s (c) Zero Power Physics Test Zero Power Physics Testing commenced with initial criticality on September 5, 1974,and was completed on September 9, 1974. The nuclear design parameters measured during the testing were in good agreement with their predicted values. 1.2.4 POWER ESCALATION TESTS Following the completion of Zero Power Physics Testing, initial power escalation commenced on September ll, 1974, with the first electrical power produced at 0733 on September 18, 1974. Subsequent power level escalations occurred when required testing was satisfactorily completed. The power escalation test program was conducted at four major test plateaus of 15, 40, 75, and 100% FP with minor testing performed at intermediate power levels as required by the controlling procedure for power escalation. A brief summary of each of these tests follows: (a) Nuclear Instrumentation Calibration at Power Calibration of the power range channels was required several times during the startup program due to changes in xenon worth, reactor coolant system boron concentration, and rod configuration. In all instances, it was possible to calibrate the power range channels to within i 2.0% FP to the reactor thermal power and to within 15.0% to the in-core axial of fset. The trip setpoint error for the high flux level trip bistable was verified to be less than 1 0.5% FP. Thus, all acceptance criteria for nuclear instrumen-tation calibration at power were met. (b) Biological Shield Survey The maximum combined dose rate measured during the power escala**'n testing was 4.9 mrem /hr., which was well within the acceptance criterion of 100 mrem /hr., in all accessible areas. (c) Reactivity Coefficients at Power The test results indicated that the toderator coefficient is nega-tive during operation at or above 95 %FP. Analyzed data for the power Doppler coefficient versus power level indi-cated that the least negative coef ficient is -0.65 x 10-4 Ak/k/%FP. Using the measured values for the power Doppler coefficient, the power Doppler deficit from 0 to 100% FP is determined to be -0.92% ak/k. (d) Core Power Distribution Analysis of core power distributions taken at 15, 40, 75, and 100% FP for the normal operating control rod positions showed that the maximum radial and the maximum total peaking factors were not more than 5.0 and 7.5 percent of the respective predicted peaking factors. Comparison of the measured and predicted radial power distributions indicated that Oconee Unit 3 has a flatter radial profile and lower maximum peaking factors than had initially been expected. 1.2-2 l

o Detailed analysis of the measured DNBR's indicated substantial thermal margins, seven under worst case conditions. Analysis of the measured maximum linear heat rates showed that the maximum linear heat rates met the LOCA and the central fuel melt criteria. All quadrant tilts and power imbalances measured during the test were within their specified limits. Analysis of the core power distributions taken at,75% FP for various ' conErol rod configurations indicated acceptable core thermal conditions, even when the worst, case uncertainty factors were applied. (e) Turbine / Reactor Trip Test Upon completion of the reactor trip portion of the Turbine / Reactor Trip Test at 40% full power, the following conclusions were made: (1) The reactor trip caused the turbine to trip, ensuring that cooldown rates less than 100 F/hr can be emintained during reactor trips at power as required by Technical Specification 3.1.2.3. (2) All acceptence criteria except for the turbine bypass setpoint trans-ferring to 1025 psia were met. An evaluation indicated that the turbine bypass setpoint transferred to approximately 995 psia, which was deter-minted to be acceptable since it is within 30 psia of the acceptance criterion. (3) Unit response to the reactor trip indicates that the Integrated Control System adequately controlled unit parameters during the trip. (f) Rod Worth at Power The measured diffciential rod worths at 40% FP were within the specified maximum and minimum values. The hot power integral rod worth curves developed from predicted rod worth data and Zero Power Physics Test results were found to be adequate when used in reactivity balances. Comparison of these integral curves against measured results using the fast insertion / withdrawal technique gave favorable results. (g) Power Imbalance Detector Correlation Test Upon completion of power imbalance detector correlation testing at 40 and 75% FP, the following conclusions were drawn: (1) The slope of the equation relating the incore and out-of-core imbalances vs. independent of the technique used to produce the imb.iance. (2) The observed in-core to out-of-core imbalance correlation on Oconee Unit 3 was a linear rela,tionship with a normalized slope of 0.275. 6 Using this correlation it was determined that a gain factor of 3.450 was the most desirable bain factor to be set into the delta-flux circuit. I 1.2-3 0

e I (3) The imbalance trip envelope as set in the Reactor Protection System will protect the reactor core from exceeding the minimum DNBR and maximum LHR limits. (h) Nuclear Steam Supply System Heat Balance All primary and secondary heat balance calculations met their respecti acceptance criteria, except for the unit computer calculated primary power at 65% FP. The discrepancy in this primary heat balance calculation was due to erroneous inputs to the unit computer's heat balance program. After correcting the erroneous inputs, the test was reperformed acceptably at 65% FF. The.hrimary reactor coolant system flow rate on Oconee Unit 3 was measured to be 112.44 + 0.79 percent of. design flow. (i) Unit Load Steady-State Test The average of the measured unit parameters during the test period fell within their respective minimum and maximum limits, although steam generator "A" steam temperature was lower than predicted., Analysis of unit parameter stability indicates that all variables are relatively stable, even though feedwater flow did not meet the stated acceptance criteria. However, neither steam generator temperature or feedwater flow stability has any adverse effect on the safe operation of the unit. (j) Unit Load Transient Test Fro:i analysis of all test data taken, the following conclusions may be made from the 40 and 75% FP sections of the Unit Load Transient Test: (1) All transients were performed without.sxeceding the limits of the f' Unit 3 Technical Specifications. (2) All transients were campleted without causing the Reactor Protective System to actuate. (3) The ability of the Integrated Control System to control unit parameters (i.e., power, unit average temperatures loop temperature mismatch, and turbine header pressure) during the transient was ex'cellent. h 1.2-4 1

(k) Pseudo Control Rod Ejection Test Tha measured worth of the most reactive control rod was found to be 0.40% Ak/k, which is less,than' the maximum value of 0.50% Ak/k specified in Section 3.5.2 of the Technical Specifications. As expected, the pseudo rod ejection produced a large perturbation to the steady-state core power distribution. The resulting maximum linear heat rate (kw/f t), minimum DNBR, and maximum power peak measured were 8.32, 5.54, and 3.18, respectively. (1) Dropped Control Rod Test Upon analysis of the Dropped Control Rod Test data, the following con-clusions were made: (1) Analyzed core power distribution and thermal hydraulic data indicated sufficient margin to minimum DNBR and maximum linear heat rate limiting criteria. The perturbation to the stesdy-state power distribution was as expected with a maximum tilt of 13.97 percent. (2) The measured worth of the control rod which produces the most adverse thermal effects in the core, if it is inadvertently dropped, was found to be 0.09% Ak/k. (3) The Integrated Control System accurately detected the asymmetric control rod, activated appropriate alarms, and initiated designated control actions. \\ 1.2-5

w e ND r" "2 O- > -4 0-> 2 r-OmCmr t I 0 I J l b e

O O 2.0 INITIAL FUEL LOADING Fuel, loading was initiated with the insertion of fuel assembly 3C28 into the core on July 20, 1974, and was completed with the loading of 3C06 on July 28, 1974. Figure 2.0-1 depicts the final configuration of the core at the con-clusion of initial fuel loading. f Neutron count rate was monitored throughout the fuel loading sequence utilizing the unit's two nuclear instrumentation source range channels, NI-l and NI-2, and two temporary incore proportional counters. Independent plots of the inverse neutron count rate ratio were maintained corresponding to the output from each of these detectors. In summary, the initial fuel loading of Oconee Unit 3 was completed in eight days, with only, minor equipment problems, and was carried out in a safe and orderly manner. 4 6 7 k 1 2.0-1 I

's FINAL FUEL LOADING DISTRIBUTION FOR OCONEE UNIT 3, CORE 1 I Vest A 3C03 3C23 3C51 3C40 3C08 B 3C33 3C42 3Cl3 3356 3C54 3B31 3C55 3C37 3C09 014 C085 B078 C062 3071 C063 011 C 3C32 3C29 1A06 3B29 3A12 3B08 3A03 3330 3A32 3C30 3C19 B109 C084 B110 C097 B091 C086 B092 C064 B093 D 3C58 3C49 3A30 3B10 3A28 3807 3A25 3503 3A46 3B25 3A33 3C25 3C59 B108 C083 8130 A016 5118 C098 Bill A009 B123 C065 B094 E 3C44 3A13 3828 3A10 3B24 3A40 3336 3A27 3322 3A16 3B13 3A04 3C12 C082 B129 C096 B138 C113 B119 C099 B131 C087 B124 C066 F ___ 3C04 3C17 3B50 3A34 3B16 3A39 3B38 3A21 3B44 3A47 3B23 3A45 3B49 3C50 3C06 C081 8107 A015 B137 C112 B086 C114 B079 C100 B132 A010 B095 C067 C,,, 3C34 3B60 3A02 3801 3A20 3B43 3A53 3B52 3A32 3833 3A34 3319 3A01 3B51 3C24 B077 C095 B117 C111 B085 C121 5087 C115 B080 C101 B112 C088 3072 H __. 3C18 3C22 3B15 3A35 3335 3A22 3B39 3B41 3B45' 3A36 3B42 3A38 3305 3C48 3C15 Y North W C080 B106 C110 B122 C120 5090 C122 B088 C116 B120 C102 3096 C068 ' South 3B18 3A41 3B37 3C10 K 3C21 3859 3A42 3821 3A31 3B34 1A51 3B47 3A08 3B32 3A50 B076 C094 8116 C109 B084 C119 B089 C117 B081 C103 3113 C089 B073 L 3C01 3C56 3B57 3A17 3B09 3A48 3B46 3A26 3553 3A49 3B20 3A29 3B58 3C41 3C02 C079 B105 A014 B136 C108 B083 C118 B082 C104 B133 A011 B097 C069 M 3C45 3A56 3B27 3A14 3B17 3A24 3B40 1A37 3814 3A44 3B26 1A05 3C2,7 C078 B128 C093 B135 C107 B121 C105 B134 C090 8125 CG70 N 3C60 3C20 3A19 3811 3A18 3B04 3A23 3202 1A15 3B12 3A07 3C46 3C57 B104 C077 3127 A013 Bil5 C106 B114 A012 B126 C071 B098 G 3C05 3C43 3A55 3B4$ 3A43 3B06 3All 3B61 3A09 3C38 3C31 B103 C076 B102 C092 B101 C091 B100 C072 B099 P 3r78 3C26 3C47 3354 3C11 3555 3C14 3C39 3C36 013 C075 B075 C074 B074 C073 012 3C07 3C52 3C16 3C53 3C35 l l Zl East l l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 i l 3A01 through 3A36 = 2.01 wt : Fuel Assemblies i 3B01 through 3823 = 2.60 wt % Fuel Assemblies 3B29 through 3361 = 2.67 we % Fuel Asse=blies 3C01 through 3C60 = 3.00 we

• Fuel Assemblies 5071 through 8090 = 1.43 wt
• Burnable Poison Assemblies 8091 through B122 = 1.26 'wt : Burnable Poison Assemblies i

5123 through 8138 = 1.09 we : Burnable Poison Asseanlies I C062 through C122 = Control Rod Assemblies A009 through A016 = Axial Power Shaping Rod Assemblies 0011 through 0014 = Orifice Red Assemblies

s Figure 2.0-1 O

3.0 TESTING PRIOR TO i POWER ESCALATION i l 4 e 1 1 O I

= 4 3.0 TESTING PRIOR TO PONER ESCALATION Following initial fuel loading of Oconee Unit 3 and prior to power escalation, the Reactor Coolant Pump Flow and Flow Coastdown Test, the Control Rod Drive Drop Time Test, and the Zero Power Physics-Test were conducted. A brief description of these tests and their results are presented in this section of the report. In all cases, applicable Technical Specification and FSAR requirements were met. t r 4 4 3 e 3.0-1 j r * -- v'

a 3.1 REACT 0R COOLANT FLOW AND FLOW COASTDOWN TEST 3.1.1 PURPOSE Prior to power escalation, the Reactor Coolant Pump Flow and Flow Coastdown Test was performed with the reactor core installed. The purpose of this test was to verify the functional capabilities of the Reactor Coolant System and the reaccor coolant pumps by determining the steady-state reactor coolant flow for various reactor coolant pump operating combinations and the reactor coolant flow during vari ~ r specified reactor coolant pump coastdowns. ~ 3.1.2 TES1 -;THOD Reactor coolant steady-state flow war determined by means of both the unit computer and the loop flowmeter AP cells. For each operating pump combination, five sets of steady-state data were read from the computer, and the in?icatea flow, temperatures, and pressures were averaged and the average values used to calculate properly compensated flow rates. From the loop flow meter AP 4 cell indications, the reactor coolant flow was. calculated as follows: _. b Flow = C AP }Bi f Vs Where: C = Flow Meter Coefficient = 397,100 f AP = Indicated AP (psi) V = Specific volume at reference conditions (68 F,14.7 psia) V, = Specific volume at system conditions Various combinations of reactor coolant pumps were operated,and steady-state flow data acquired. Subsequently, one oc more of the operating pumps were tripped,and data were recorded during the ensuing reactor coolant flow transient. Reactor coolant flow, at various times during the coastdown - i transients, was determined from loop flowmeter data. 3.1.3 EVALUATION OF TEST RESULTS Table 3.1-1 gives the minimum and maximum allowable flow rates for three dif-ferent pump combinations, along with the corresponding measured flowrates. It can be seen that the measured flowrates are well within the acceptance 4 criteria. Figure 3.1-1 shows the minimum acceptable reactor coolant flow versus time for both single-pump and multi-pump coastdowns. Measured reactor coolant flow versus time is presented in Figures 3.1-2 through 3.1-6 for typical coastdowns ' from three and fcur pump. icitial operating conditions. 3.

1.4 CONCLUSION

S The measured steady-state-reactor coolant flow rate was well within the specified minimum and maximum values for all pump combinations tested. \\ 3.1-1 -l

s c The measured reactor coolant flow during pump coastdowns satisfied the applicable acceptance criteria for all test conditions. 4 i \\ i i i 1 l 3.1-2

2 O REACTOR COOLANT FLOW AT REFERENCE CONDITIONS OF 532 F AND 2155 PSIG FOR VARIOUS PUMP COMBINATIONS Minimum Maximum Acceptable Acceptable Measured Pump FgowRate Flow Rate FgowRate 6 Combination (10 lbm/hr) (10 lbm/hr) (10 lbm/hr) Four Pumps 138.4 153.4 145.6 Three Pumps 103.2 153.4 108.4 One Pump Each Loop 67.8 153.4 [2.5 Table 3.1-1

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e s 3.2 CONTROL ROD DRIVE DROP TIME TEST 3.2.1 PURPOSE The purpose of the Control Rod Drive Drop Time Test was to verify the inte-grated, functional trip capability of the Control Rod Drive System and to determine for each control rod assembly the total elapsed drop time from the initiation of a trip signal until the egntrol rod assembly was three-fourths inserted. 3.2.2 TEST METHOD This test was conducted at various combinations of reactor coolant flow, pressure, and temperature as follows: Test Condition Flow Pressure Temperature 1 No Flow 2; 350 psig i 400 F 1 1700 psig 4 1 00 F 2 One Pump Each Loop 3; 350 psig 1 1700 psig 3 Four Pumps 2155 1 30 psig 532 + 10 F At each condition, Control Rod Groups 1 through 7 were driven sequentially to the fully withdrawn position. A manual trip of all control rod drives was then initiated and, coincidentally, a time signal was provided to the data logging equipment. As each control rod assembly reached the three-fourths insertion position, a second time signal was provided to the data logging devices from a switch located on each control rod drive's position indicator tub e. The total elapsed time from the initiation of a trip signal until s three-fourths insertion was then determined for each control rod drive from the data acquired. The test was repeated a second time for all control rods at each test condition. The drop time for each rod was recorded along with its core location and pertinent Reactor Coolant System conditions. 3.2.3 EVALUATION OF TEST RESULTS l For no flow condition, control rod E-ll recorded the. shortest insertion time ~ of 1.080 seconds, and rod K-03 recorded the longest insertion time of 1.123 seconds. For a second trip under the same system conditions, control rod ,l -D-08 recorded the shortest insertion time of 1.092 seconds,"and rod H-06 recorded the longest insertion time of 1.129 seconds. For test condition 2, rods D-08 and E-ll were the fastest.with 1.145 seconds ; and H-06, the slowest with 1.188 seconds. For the second trip, rod D-08 was the fastest with 1.139 seconds; and H-06 and F-06, the slowest with 1.181 i seconds. 4 3.2-1

o 1 e I I Under full flow condition, control rod E-11 had the smallest insertion time l of 1.231 seconds, and rod H-06 had the greatest insertion time of 1.264 i seconds. For a second trip, rods D-08 and E-13 had the smallest insertion time of 1.233 seconds, and rod E-07 had the greatest insertion time of 1 266 s eco nds. Rods E-13 (the fastest) and H-06 (the slowest) were dropped an additional ten times at full flow condition. The average drop time for E-13 was found to be l 1.216 seconds with a maximum deviation of 0.025 seconds, and the average drop time for H-06 was 1.245 seconds with a maximum deviation of 0.042 seconds. i 3.

2.4 CONCLUSION

S All measured rod drop times were well below the Technical Specification limits of 1.66 seconds at full flow and 1.40 seconds at no flow conditions. i i i

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s 3.3 ZERO POWER PHYSICS TEST 3.3.1 PURPOSE The purpose of the Zero Power Physics Test was to verify the nuclear design parameters used in the safety analysis, the Technical Specification limits, and the operational parameters. This test was conducted after initial fuel loading and before escalation into the power range. Testing was performed at hot conditions (532 F and 2155 psig). The test included the following measurements: (a) Initial criticality (b) Nuclear instrumentation overlap between the source and intermediate range (c) "All Rods Out" critical baron concentration (d) Differential and integral rod worth (e) Differential and integral boron worth (f) Ejected control rod worth (g) Stuck rod worth (h) Temperature coefficients as a function of boron concentration 3.3.2 TEST METHOD Initial criticality was achieved by control rod withdrawal and boron dilution of the Reactor Coolant System after the system had been heated to hot con-ditions using the reactor coolant pumps. During the approach to initial criticality, plots of inverse multiplication versus boron concentration and time were maintained by using channels NI-l and NI-2 of the unit's nuclear ins trumentation. Af ter achieving criticality, the nuclear power was in-creased and the source and intermediate range nuclear instrumentation overlap was verified to be in excess of one decade. During the same increase in power the nuclear heatup point was determined,and the upper power limit for Zero Power Physics testing was established. The Zero Power Physics testing then proceeded with the measurement of the core physics parameters. The unit computer and a reactimeter were utilized for reactivity determination during these measurements. 3.3.3 EVALUATION OF TEST RESULTS 3.3.3.1 Initial Criticality Initial criticality was achieved on September 5, 1974, 2111 hours, at reactor coolant conditions of 532 F and 2155 psig. The procedure for the approach to initial criticality consisted of the sequential withdrawal of the control rod groups and subsequent deboration as follows: (a) Control rod group withdrawal: Groups 1-4 100 percent withdrawn Group 8 100 percent withdrawn Group 5 100 percent wi:hdrawn Group 6 100 percent witha. awn Group 7 75 percent withdrawn. 3.3-1

s (b) Deboration from 1954 ppr5 to 1674150 ppab using a feed and bleed rate of approximately 70 g-as per minute. (c) Deboratio.. From 1674 i 50 rpmb to the critical boron concentration using a feed and bleed rate of approximately 40 gallons per minute. Control Rod Group 7 inser* ' to maintain criticality if required. Throughout the approach to criticality, two independent plots of inverse multiplication versus reactivity addition (control rod withdrawal and boron dilution) were maintained corresponding to the two startup range neutron detectors. The inverse multiplication was obtained from the ratio of the initial average count rate to the average count rate at the end of each reactivity addition. As indicated above, deboration was carried out in essentially two steps. First, deboration from an initial boron concentration greater than 1800 ppmb to 1674 1 50 ppmb was accomplished using a 70 gpm letdown and makeup rata. This step brings the reactor coolant system boron concentration to within 100 ppmb of the predicted critical boron concentration (1574 ppmb). Boron concentration changes during this peri >d were determined by taking reactor coolant samples at least every 30 minutes. The second deboration step was accomplished at a lower makeup and letdown rate of 40 gpm. Again, Reactor Coolant System boron samples were taken at least every 30 minutes, and plots of inverse multiplication versus boron concentration and time were maintained. Deboration was terminated upon initial criticality, and Control Rod Group 7 was inserted as necessary to maintain the reactor critical. Plots of inverse multiplication versus time and boron concen'tration are presented in Figures 3.3-1 and 3.3-2. In sommary, initial criticality occurred at a Reactor Coolant System boron concentration of 1548 ppmb and was achieved in a safe and orderly manner. Analyzed results indicate good agreement between predicted and measured criticality endpoints. 3.3.3.2 Nuclear Instrumentation Overlap Technical Specifications state that prior to operation in the intermediate nuclear instrumentation range, at least a one decade overlap between the source range and intermediate range must be gbserved. This means that before the source range count rate equals 10 cps, the intermediate range must be on scale. If the one decade overlap is not obtained, the approach to 10-9 anps on the intermediate range cannot be continued until the situation has been corrected. In addition, the following number of channels must be in operation fo r the test program to continue beyond initial criticality: -Channels Available Minimum Operating Source Range NI 2 2 Intermediate Range NI 2 2 To satisfy these overlap requirements after initial criticality was reached, ~ core power was slowly in? aased until the intermediate range channels came s 3.3-2

s s e on scale. Detector signal response was then recorded for both the inter-mediate and source range channel. This was repeated for two more decades until the source range signals approached 106 cps. The results of the nuclear instrumentation overlap measurement at 532 F and 2155 psig are shown in Figure 3.3-3. The core power indicated in this figure was estimated from the system heatup rate. A heatup rate of 36 F/hr' has been shown to be equivaler.c to a 10 MWt heat addition to the Reactor Coolant System. Examination of Figure 3.3-3 reveals that the overlap between the source and intermediate range detector readings is approximately 2.27 decades and that the linearity, overlap,and absolute output of bqth sets of detectors are within specifications and are performing satisfactorily. 3.3.3.3 "All Rods Out" Critical Boron Concentration The "All Rods Out" critical boron concentration was measured at the isothermal temperature test plateau of 532 F. The measurements were made with the Control Rod Group 7 partially inserted. The measured boron concentration was adjusted to the "All Rods Out" condition using the results of rod worth measurements to determine the reactivity worth, in terms of boron concentration, of the inserted control rods. The measured "All Rods Out" critical boron concentration at a moderator temperature of 532 F is 1554 ppm. This value compares quite favorably with the predicted value of 1582 ppm. Therefore, the dif ference between the measured and predicted "All Rods Out" critical boron concentration is well within the acceptance criterion of i 100 ppm. 3.3.3.4 Control Rod Group Worths The Oconee Unit 3 initial control rod group configuration is shown in Figure 3.3-4. The beginning-of-life control rod group reactivity worths were measured at zero power conditions for the normal rod withdrawal sequence with Group 8 at 20 percent withdrawn. The measured group worths were then compared with their calculated values, which were based on Group 8 positions either at 35 percent withdrawn or at 100 percent withdrawn. The boron swap meth'od was employed to determine the differential and integral worths for Groups 8, 7, 6, and part of 5. This method consisted of setting up a suitable boration or deboration rate and compensating the. reactivity change by small step changes in control rod group positions to maintain the reactor critical. The rod drop method was used to determine the worth of control rod groups not measured by boron swap. For this measurement the reactor was made critical with all the control rod groups.co be measured out of the core and at a power level near the Zero Power Physics Test upper power limit. The control rod groups being measured were. then tripped, and the resulting reactivity insertion was measured. 3.3-3

J o Based on the experience with Oconee Units 1 and 2, it was expected that the rod drop measurements on Oconee Unit 3 would yield values approximately 74 percent of the calculated value when considerably more than 1% Ak/k resetivity was being inserted. The Oconee Unit 3 results were only marginally consistent with this expectation. Table 3.3-1 compares the actual group worth with that expected and measured by the rod drop method. The expected value was computed by applying a normalization factor of 0.74 to the calculated value. The results show that the measured value was within 9.2 percent of the expected value. The measured control rod group worths for a moderator temperature of 532 F ate tabuiated in Table 3.3-2 along with the corresponding calculated values. Based on the results of the rod drop measurement considered above, the calculated worths were used as the best estimate for Groups 1 through 4. In the case of Group 5, the total worth was determined by extrapolating the measured reactivity worth (by boron swap) when this group was moved from 100 percent withdrawn to 43 percent withdrawn. The results indicate good 4 agreement between the measured and calculated group worths. The measured normalized differential and integral reactivity worths for Control Rod Groups 5 through 7 are plotted in Figures 3.3-5 and 3.3-6, respectively, for 532 F and Group 8 at 20 percent withdrawn conditions. The integral worths were calculated by integrating the differential worth curves given in Figure 3.3-5. The normalized integral reactivity worth of Centrol Rod Group 8 was also measured and is presented in Figure 3.3-7. l 3.3.3.5 Soluble Poison Worths i Measurement of the soluble poison differential worth was made at 532 F, zero power conditions. The measured value was determined by summing the j incremental reactivity values measured during the rod worth measurements over a known boron concentration range. I t The result of the soluble poison differential worth measurement is given in i Table 3.*3-3. The integral reactivity worth curve shown in Figure 3.3-8 was then established by integration of the measured differential worth and by The reactivity balance calculations performed during the test program. integral worth curve at an average coolant temperature of 579 F, obtained from reactivity balance during power operation, is also shown in Figure 3.3-8. In summary, the measured differential boron worth at zero power conditions i was within 5.7 percent of c.he predicted worth. Also, the measured integral boron worths were in close agreement with the predicted values. 3.3.3.6 Ejected Control Rod Worth Pseudo ejected control rod reactivity worth was measured at zero power conditions of 532 F, and 2155 psig for the control rod at core location F-02. The purpose of this measurement was to verify the safety analysis calculations relating to the assumed accidental ejection of the most reactive control rod during power operation. The acceptance criterion for i the ejected control rod test is that the reactivity worth of the most reactive control rod does not exceed 1.0% ak/k at hot zero power conditions.' 3.3-4 r n-- m ,e., ,e- ,-.,m, - - ~ - ~.- ,m r

s The ejected rod worth of rod F-02 was measured by the rod drop method. The reactor was made critical with the control rod F-02 at 100 percent withdrawn. The rod was then dropped,and reactivity was recorded every 0.2 seconds for approximately 4.0 minutes following the rod drop. The worth of the ejected rod was then determined by plotting the measured reactivity versus inverse time and extrapolating it to zero inverse time, as shown in Figure 3.3-9. The ejected rod worth was determined to be 0.77% Ak/k, and the acceptance criterion was adequately met. 3.3.3.7 Stuck Control Rod Worth The purpose of the stuck rod worth measurement is to verify that the calcu-lated stuck rod worths are conservative compared to the measured worths. i The safety rod,at core location h-02 was selected for the stuck rod worth measurement. In order to determine the stuck rod worth, two rod drop measurements were made. During one rod drop measurement, all the rods of the safety groups (Groups 1 through 4) were dropped, and the resulting reactivity was measured. During the other lod drop measurement, the designated stuck rod remained fully withdrawn, and the remaining rods of the safety groups were dropped. The reactivities measured during the two rod drop measurements were then corrected for uncertainties in the rod drop measurement, and the difference between the two corrected reactivities was taken as the stuck rod worth. J The measured stuck rod worth is 2.64% Ak/k, which is much less than the calculated value of 4.27% Ak/k. Therefore, the calculated stuck rod worth is conservative. I 3.3.3.8 Temperature Caef ficient of Reactivity ) The temperature coefficient of reactivity for the Ocone'e Unit 3 core was measured for various reactor coolant boron concentrations at 532 F isothermal conditions. For a particular reactor coolant boron concentration, the temperature co-efficient of reactivity was determined by measuring the core reactivity changes when the reactor coolant temperature was increased and then decreased by approximately 10 F. Figure 3.3-10 contains the measured and predicted values for the temperature coefficient of reactivity at zero power conditions.,The moderator tem coefficients shown in this figure were calculated by adding 0.16 x 10 gerature Ak/k/0F to the measured temperature coefficients of reactivity. The deviations between the measured and predicted temperature coefficients of reactivity were well within the acceptable limits of + 0.4 x 10-4 Ak/k/ F. In addition, the calculated moderator coefficients were within the require-ments of Technical Specification 3.1.7. 3.3-5

f f 9 3.

3.4 CONCLUSION

S Zero Power Physics Testing commenced on September 5,1974, with initial criticality occurring at 2111 hours. The Zero Power Physics Test was completed on September 9, 1974, with good agreement between measured and predicted parameters. The results of each of the measurements performed during the Zero Power Physics Test are summarized as follows: (a) Initial Criticality The initial criticality was achieved in a safe and orderly manner, and the measured boron endpoint was in agreement with the predicted value. (b) Nuclear Instrumentation Overlap Nuclear instrumentation overlap was verified to be in excess of two decades between the source and intermediate range. The minimum acceptable overlap is one decade. (c) "All Rods Out" Baron Concentration The measured "All Rods Out" boron concentration was 1554 ppm as compared to the predicted value of 1582 ppm. (d) The measured reactivity worths for Control Rod Groups 5, 6, 7, and 8 were in good agreement with the calculated values. A maximum deviation of 7.3 percent of the calculated value was observed for the Group 6 worth. The rod drop method, used for the safety groups, produced only marginal agreement with expected worths. (e) Soluble Poison Worths The measured dif ferential boron reactivity worth at an average boron concentration of 1395 ppm was 1.00% ak/k as compared to the predicted-value of 1.06% Ak/k/100 ppm. (f) Ejected Control Rod Worth The ejected rod worth was measured to be 0.77% Ak/k,which satisfies the Technical Specification 3.5.2.3 requirement. (g) The measured stuck rod worth was 2.64% Ak/k as compared to the calcu-lated value of 4.27% ak/k, thus verifying the available shutdown margin to be more than adequate. (h) Temperature Coefficient of Reactivity The deviations between the measured and predicted temperature coefficeints of reactivity were much less than the maximum allowable deviation of +~ 0.4 x 10~4 ok/k/ F. The calculated moderator coefficients were well within the requirements of Technical Specification 3.1.7. 3.3-6

COMPARISON OF CALCULATED ROD WORTH WITH THAT MEASURED AND EXPECTED FROM ROD DROP METHOD Position-Calculated Expected Measured Deviation From Deviation From Control Rod Interval Worth Worth Worth Calculated Worth Expected Worth Group (% wd) (% Ak/k) (% Ak/k) (% Ak/k) (%) (%) 1 100 to 0 0.89' 2 100 to 0 3.01 3 100 to 0 0.74 -7.16 5.30 4.81 -12.8 -9.1 4 100 to'0 1.86 gi 5 43.to 0

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s' s AT A MODERATOR TDiPERATURE OF 532 F AND APSR'p0L ROD GROUP WORTHS COMPARISON OF CALCULATED AND MEASURED CON S AT 20 PERCENT WITHDRAWN I Calculated (1) Measured I) Percent ( ) Group No. Rods Worth, % Ak/k Worth, % Ak/k Deviation 1 8 -1.07 -1.07 NA 2 8 -2.90 -2.90 NA 3 8 -0.79 -0.79 NA 4 8 -1.71 -1.71 NA 5 12 -1.08 -1.08 0.0 6 8 -1.23 -1.14 -7.3 7 9 -1.21 -1.23 +1.6 8 8 -0.38 -0.39 '+2.6 Total 69 -10.37 -10.31 -0.6 d Note (1): Calculated worths.of control rod groups (1-8) are based on APSR's at 35 percent withdrawn. Note (2): Measured worths of control rod groups (5-8) are by boron swap method. Note (3): NA indicates that deviations are not applicable to rod groups (1-4), since predicted worths are used. I . Table 3.3-2

.o CORD POWER VERSUS DETECTOR RESPONSE Intermediate Range Detector Response, amps 10-11 10-10 10-9 10-8 10-7 10-6,. 10-3 10-4 8 10 j / 7 10 7 J l' 6 / 10 / / db j / Average Intermediate + 7 105 / Range Channel at 532F and 1520 ppm. r / / W W me 104 ma i / / 3 j ] O /" f y 10 r g / Average Source Range __._- / Channel at 532F and 3 / 1520 ppm. 2 / 10 / / ' 101 f / / 0 10 s v 10 -1 ~J-M l -2 l 10 100 1 10 102 103 104 105 106 107 Source Range Detector Response, counts / seconds i Figure 3.3-3

J 0 CONTROL R0D GROUP LOCATIONS FOR CORE BURNUP UP TO 250 EFPD A I (7) (4) (7) a f(3) C (5) (5) (3) D (4) (8) (6) (8) (4) l (5) (6) (1) (1) (6) (5) r-(7) (8) (2) (2) (2) (8) (7) '~ i (3) (1) (5) (5) (1) (3) (4) (6) (2) (7) (2) (6) (4) ~ (3) (1) (5) (5) (1) (3) ~ (7) (8) (2) (2) (2) (8) (7) M (5) (6) (1) (1) (6) (5) N (4) (8) (6) (8) (4) i o (5) (3) (3) (5) (7) (4) (7) R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 4 l (X)y Control Rod Cro sp Number FIGURE 3.3-4 1

.s .o NORMALIZED DIFFERENTIAL WORTHS OF CONTROL ROD GROUPS 5-7 VERSUS WITHDRAWAL POSITION, FOR BEGINNING OF LIFE, ZERO POWER CONDITIONS

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_ _ _ _ ~ =/ if 4 4.0 POWER FSCALATION TESTS 4 Following the completion of Zero Power Physics Testing, initial power escala-tion commenced on September 11,1974, with the first electrical power produced at 0733 on September 18, 1974 The power escalation test program was con-7 ducted at four major test plateaus of 15, 40, 75, and 100 percent full power with minor testing performed at intermediate power-levels as required by the controlling procedure for power escalation. Power escalations to the various power levels occurred when required testing was satisfactorily completed. The major power levels were achieved as follows: i Power Level (Percent of Full Power - %FP) Date 15 September 11, 1974 40 October 15, 1974 75 . November 16, 1974 100 December 16, 1974 j The tests reported in this section cover all power escalation testing as of { 2400 hours on January 15, 1975. A listing of tests performed during the i initial power escalation period, along with the appropriate section number of this report and the power level at which they were performed,is given in Table 4.0-1.' Figure 4.0-1 shows the unit's average daily power levels during 1 the power escalation testing period. In the following subsections, all the tests performed during the initial 1 power escalation period, along with their pertinent data and the conclusions drawn from the test results, are presented. I -l l } i '4.0-1 t f 1 ,,, O .~

LIST OF TESTS PERFORMED DURING POWER ESCALATION Report Test Power Level, % FP 3 100 Section Tests 3,5 15 30 40 65 75 90 4.1 l.NuclearInstrumentationCalibrationatPowerTest x x x x x '4.2 Biological Shield Survey Test x x x x 4.3 Reactivity Coefficients at Power Test x x x 4.4 Core Power Distribution Test x x x x 4.5 Turbine / Reactor Trip Test x gi -4.6 Rod Worth at Power Test x x x e EI 4.7 Power Imbalance Detector Correlation Test x x 'o 4.8 Nuclear Steam Supply System lleat Balance Test x x x x x x x x 4.9 Unit Load Steady-State Test x x x x x x x x 4.10 Unit Load Transient Test x x ! Pseudo Rod Ejection Test 4.11 x 4.12 Dropped Control Rod Test x x S

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l l 4.1 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER I 4.1.1 PURPOSE j The purpose of nuclear instrumentation calibration at power was to calibrate j the power range nuclear instrumentation power indication to the reactor thermal power, and the axial offset indication to the incore axial offset. l Two other purposes were to adjust the high power level trip setpoint when required by the Power Escalation Test Procedure and to verify that at least one decade overlap existed between the intermediate and power range nuclear instrumentation. i The acceptance criteria for nuclear instrumentation calibration at power were l that the power range nuclear instrumentation indicate the power level to be j within i 2.0% FP of the reactor thermal power and the axial offset to be ] within 5.0% of the incore offset and that the high power level trip bistable 1 be set to trip at the desired value within 1 0.5% FP. j 4.1.2 TEST METHOD i j The nuclear instrumentation calibration at power was performed several times during the testing period. i In order to calibrate the power range nuclear instrumentation, the top and-4 bottom linear amplifier gains were adjusted so that the power level indicated by the power range nuclear instrumentation channels was equal to the reactor thermal power within i 2.0% FP. The most conservative value for the reactor thermal power was obtained from the core AT power indication of the unit computer, and the power range channels were calibrated to this core AT power. i When the gains were adjusted, their ratios were kept the same if the axial offset was within i 5.0% of the axial offset determined by the incore monitoring system; otherwise the ratio of the gains was also varied to correct { the offset mismatch. During calibration of the power range channels, data were also taken to verify 4 the overlap between the intermediate and power range channels. The. required overlap was a minimum of one decade between these two nuclear instrumentation channels. 1 When directed by the Power Escalation Test Procedure and/or the unit startup. procedure, the high flux trip bistable setpoint was adjusted. The major settings during power escalation are given below: Test Plateau, %FP Bistable Setpoint, %FP 15 35 40 50 75 '80 100 105.5 6 Thef trip setpoint was then verified by placing each power range channel in the test mode and slowly increasing an input power. signal until each, bistable -4.1-1

2 / i tripped. The difference between the actual power at which the bistable tripped and the tri.setpoint power, the trip setpoint error, was recorded during each adjustment. 4.1.3 EVALUATION OF TEST RESULTS An analysis of test results indicated that changes in Reactor Coolant' System baron concentration, control rod configuration, and xenon buildup or burnout affected the power indicated by the nuclear instrumentation. This was expected since the power range nuclear instrumentation metsures reactor neutron leakage, which is directly related to the above changes in system conditions. Changes in these system conditions resulted in an increase or decrease of approximately 3 to 5% FP of the power level indicated by the power range channels. Therefore, it was necessary to calibrate the power range nuclear instrumentation; and with each calibration, the acceptance criteria were met without any difficulty. Table 4.1-1 is a summary of the calibration data taken at different power levels during power escalation testing. In all cases the nuclear instrumentation was adjusted to within i 2.0% FP of the reactor thermal power and to within i 5.0% of the incore axial offset. The high flux level trip bistable was adjusted to 35, 50, 80, and 105.5% FP prior to escalating the power to 15, 40, 75, and 100% FP, respectively. Acceptance criteria of adjusting the setpoint to the above values within 1 0.5% FP were mes each time without difficulty. The maximum trip setpoint error observed was 0.06% FP when the high flux trip was set at 50% FP. The overlap measured during the startup program included the total span of the power range, thus exceeding the minimum requirement of a one-decade over-lap. Figure 4.1-1 shows the overlap of all three nuclear instrumentation channels. 4.

1.4 CONCLUSION

S The power range channels were calibrated to within i 2% FP to the reactor thermal power and to within i 5.0% to the incore axial offset several times during the startup program. These calibrations were required because of changes in xenon worth, reactor coolant boron concentration, and rod configuration during the power escalation test program'. The overlap between the intermediate range and power range nuclear instrumentation was verified to be in excess of the minimum requirement. The trip setpoint error for the high flux level trip was. also verified to be less than 0.5% FP. The calibration procedure has therefore been thoroughly tested and has proven extremely satisfactory. Acceptance criteria for nuclear instrumentation cali-bration at power were met in all instances. 4.1-2

SLImlARY OF NUCLEAR INSTRUMENTATION CALIBRATIONS AT POWER PERFORMED AS REQUIRED BY Tile POWER ESCALATION PROGRAM AND SECTION 4.1 0F TECllNICAL SPECIFICATIONS Core AT incore Max imum ' Power Before And Af ter Calib. %FP offset Before And Af ter Calib. % Power Offset Quad Tilt (%FP) (7) (%)_ NT-5 NT-6 NI-7 NI-8 NT-5 NT-6 NT-7 NT-8 11.07 14.78 14.59 14.25 14.28 -13.73 -10.69 -11.65 -9.87 11.82 (1) 12.72(1 1 12.72Cl ) 12.63(1) 12.9 1(1 )+ 1.26(1j + 3.22CL: + 5.46(1 1 +4.11(1) i 14.78

13. _N 13.53 13.59 13.97

+27.08 +25.42 +30.10 +28.85

15. 32 14.78 14.84 14.62 14.87

+19.v1 +18.73 +18.40 +20.17 39.69 - -33.87 40.24 36.72 36.78 36.9 1 37.91 -34.12 -32.79 -34.46 -42.79 39.2 7 -38.76 +0.24 38.78 38.78 38.59 38.62 -38.45 -37.31,,,-34,75. -38.27 40.00 - 5.59 +0.20 44.9 1 44.81 44.53 44.59 - 7.50 -10.24 - 8.13 -10.16 0 40.28 - 8.14 +0.20 40.00 40.00 40.03 40.50 -11.48 -13.83 -12.72 - 9.79 41.31 + 2.38 -0. 34 41.16 41.31 41.i2 41.74 +27.33 +25.9 5 +27.35 +26.81 f-41.31 + 3.15 -0.33 41.28 41.19 41.22 41.44 + 1.89 + 1.68 + 1.36 + 1.81 y 66.52 -24.87 0.24 62.19 62.09 62.91 62.66 -17.59 -16.96 -17.58 -17.11 55.41 57.28 57.34 57.31 57.34 -12.12 -11.72 -11.39 -11.44 71.44 ~-25.36 +0.18 75.44 75.22 75.22 75.31 -18.97 -18.11 - 19. 82 - 19.1 7 74.39 - 9.69 +0.17 74.28 74.28 73.84 73.87 - 9.84 -10.4 3 -10.08 - 9.69 75.23 -13.82 40.21 .74.84 74.87 74.44 74.78 -26.72 -22.45 -23.82 -22.23 76.01 -15.61 +0.21 74.97 75.03 75.34 76.19 -14.46 -13.82 -14.39 -13.90

77. 19

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SUMMARY

OF N!! CLEAR INSTRUMENTATION CALIBRATIONS AT POWER PERFORMED AS REQUIRED liY Tile POWER ESCA1.ATION PROGRAM AND SECTION 4.1 OF TECllNICAL SPECIFICATIONS 4 Co re A T Incore Maximma Power Before And Af ter Calib. %FP Offset Before And After Calib. % Powe r O f fse t Quad Tilt NI-5 NI-6 NI-7 NI-8 NI-5 NI-6 NT-7 NT-8 (typ)' ry) ty) 97.07 -0.90 40.16 95.53 95.28 97.69 97.50 -0.29 -0.43 -2.40 -3.40 97.51 -0.18 +0.17 97.03 97.06 96.91 97.12 -0.97 -0.77 -0.68 -0.83 Note (1): Values after calibration g E-r N 8n, 6

Detector Neutron Flux, nv O, O, O O O O 5 5 5 5 5 5 a w c m m N m e 3 w 1 i I i t 1 3 l 3 l l 3 3 I I I I I I I 1 I g 5, 5, 5 5 5 5 5 5 5 5 5 i e 9 w ,e ao m m s w w Reactor Power, 7 MnV Counts per Second E m Source w

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i 4.2 BIOLOGICAL SHIELD SURVEY 4.2.1 PURPOSE The purpose of the Biological Shield Survey was to measure radiation levels in all accessible locations of the unit adjacent to the biological shield and to obtain baseline radiation levels for cocparison with future ceasure-ments of radiation levels during operation. Two acceptance criteria were specified for the Biological Shield Survey Test: (a) Dose rates (combined gamma and neutron) in normally accessible areas in the Auxiliary Building and Reactor Building during full power operation must be less than or equal to 100 mrem /hr. (b) Areas found not satisfying the above criterion will be given special consideration by Health Physics with regard to future personnel entry. 4.2.2 TEST METHOD Background surveys were conducted at all locations of interest prior to initial criticality. Surveys were conducted at the following power levels: 0, 15, 40, 75, and 100 percent full power. Inside the Reactor Building, all accessible floors, stairways, and landings were surveyed; and the highest reading was recorded. All areas inside the Auxiliary Building and the restricted area yard adjacent to the Reactor Building were also surveyed. 4.2.3 EVALUATION OF TEST RESULTS A summary of the maximum and average dose rates observed is given in Table 4.2-1. The maximum combined dose rate, 4.9 mrem /hr, was less than 5 percent of the maximum allowable. 4.

2.4 CONCLUSION

S Since the maximum combined dose rate measured during the Power Escalation Test Program, 4.9 mrem /hr, is well within the acceptance criterion of 100 mrem /hr, the Biological Shield meets all design criteria. 4.2-1

-. ~ _. 9 i SU?B1ARY OF FtAXIMUtl AND' AVERAGE DOSE RATES Inside Reactor Building Adjacent to Reactor Building - Ph ase Date }1a ximum, mrem /hr Average, mrem /hr Maximum, mrem /hr Average, mrem /hr Gamma Ne ut ron Camma Neutron Gamma Neutron Camn Neutron

Background

06/07/74 <.02 < 1.0 <.02 < 1. t> <.02 < 1.0 .02 < 1.0 0% FP 09/06/74 0.03 < 1.0 0.02 < 1.0 0.05 < 1.0 0.02 < 1.0 15% FP 09/17/74 0.02 < 1.0 0.01 < 1.0 0.16 < 1.0 0.03 1.0 40% FP 10/19 /74

0. 34

< 1.0 0.2 < 1.0 0.15 < 1.0 0.02 < 1.0 75% FP 11/17/74 1.4 < 1.0 0.7 < 1.0 0.2 < 1.0 0.04 1.0 0 100% FP-12/17/74 2.4 2.5 0.8 < 1.0 1.8 < 1.0 0.04 1.0 m u e 4 e

i. j i 4.3 REACTIVITY COEFFICIENTS AT POWER d. 4.3.1 PURPOSE The purpose of this test was to measure reactivity coefficients during power operation at 40, 75, and 100 percent full power. The following coefficients were either measured or calculated from the de.ta obtainedi I (a) Temperature coefficient of reactivity, defined as the fractional change , in the reactivity of the core per unit change in fuel and moderator tempe rature, (b) Moderator coef ficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in moderator temperature. (c) Power Doppler coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in power. (d) Doppler coefficient of reactivity, defined as the fractional change in the 4 reactivity of the core per unit change in fuel temperature. 4.3.2 TEST METHOD Measurements of the temperature coefficient and power Doppler coefficient i were made at each of the major test plateaus during the power escalation test program. The major test plateaus were at core. power levels of 40, 75, and 100 percent full power. The power Doppler coefficient was measured at the 90% FP test plateau also. The moderator coefficient was calculated by subtracting the Doppler co-efficient from the measured temperature coefficient, and the Doppler coef ficient was calculated from the measured power Doppler coef ficient and the predicted fuel temperature variation with power level. 4 Rod worth measurements were executed prior to reactivlty coefficient measure-ments in order to generate rod worth data. For temperature coefficients, y average reactor coolant temperature was increased or decreased by about 5 F. and data recorded. For power Doppler coef ficients, power was increased or j decreased by about 10 percent full power and data recorded. From the measured temperature and power Doppler coefficients,'the moderator and-Doppler coef ficients were calculated. 4.3.3 EVALUATION OF TEST RESULTS f The values of the measured temperature coef ficients and calculated moderator coef ficients at power are plotted in~ Figures 4.3-1 and 4.3-2 respectively. 1 For comparison the corresponding predicted values are also shown in these figures. A tabulation 'of the measured temperature coefficients and the calculated moderator coefficients is given in Table 4.3-1 along with the predicted values. I 1 Exaritation of the calculated moderator coefficients plotted in Figure 4.3-2 revs..s.that the limit of a non-positive value at or above 95 percent full i power will not be exceeded unless_the soluble poison concentration exceeds l 1290 ppm. Measured-results of the soluble poison concentration versus Cycle 1 lifetime for equilibrium xenon all rods.out, beginning-of-life i 4.3-1

conditions indicate a maximum boron concentration of 1140 ppm. These results confirm that during power operation at or above 95 percent full power the moderator coefficient will be negative. The results of the measured and predicted power Doppler coefficient of reac-tivity are plotted in Figure 4.3-3. A tabulation of the measured, calculated, and predicted Doppler coefficients and power Doppler' coefficients are also presented in Table 4.3-1. TheacceptancecriterionforthemeasuredpowerDopplegcoefficient-isthat the coef ficient must be more negative than -0.55 x 10-Ak/k/% FP. Figure '4.3-3 shows that all measured power Doppler coefficients are below this value and that the acceptance criterion is adequately met. In addition, all calculated Doppler coef ficients were also below their respective acceptance criterion. The measured power Doppler coefficient is slightly lower than the predicted value, and this results in a lower power Doppler reactivity deficit than originally predicted. The total measured reactivity deficit between 0 and 100 percent full power is estimated at -0.92% Ak/k as compared to the predicted value of -1.32% ak/k. This difference is a net gain of +0.40% ak/k in core excess reactivity available for Cycle 1 lifetime. 4.

3.4 CONCLUSION

S The test results indicate that the moderator coefficient will be negative during power operation at or above 95% FP. Analyzed data for the power Doppler coefficient versus power level indicate that the least aegative coefficient is -0.65 x 10-4 Ak/k/% FP. The total power Doppler deficit from 0 to 100% FP from this measured data is estimated to be -0.92% ak/k. 4.3-2

.~.

SUMMARY

OF MEASURED, CALCULATED, AND PREDICTED COEFFICIENTS OF REACTIVITY AT POWER ^ Average Rod Position, % wd Boron CorA Coefficient of Reactivity, 10-4 Ak/k Conc. Burn up Power Level 6 7 8 (pom) (EFPD) ?Ieasured Predicted Calculate =d Pre di cted (%FP) 1-5 A. Temperature and !!oderator Coef ficient Tempe rature Moderator 40 100 76 01 18 1111 6.1 -0.33 -0.14 -0.17 40.06 74 100 74 00 09 1024 14.2 -0.49 -0.30 -0.37 -0.14 97 100 91 14 03 1055 34.1 -0.46 -0.38 -0.35 -0.13 B. Power Doppler and Doppler Coef ficient Power Doppler Doppler U 36 '100 74 00 18 1111 6.1 -1.00 -1.38 -0.16 -0.22 70 100 71 00 09 1024 14.2 -0.72 -1.29 -0.12 -0.21 69 100 84 09 12 1030 15.9 -0.75 -1.30 -0.12 -0.21 92 100 90 14 03 1055 34.1 -0.66 -1.20 -0.11 -0.19 93 100 81 06 05 1032 35.1 -0.66 -1.20 -0.11 -0.19. 87 100 68 00 13 993 39.9 -0.68 -1.23 -0.11 -0.20 2

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,o .e. e POWER DOPPLER COEFFICIENT OF' REACTIVITY VERSUS POWER LEVEL -T:

6..

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p:

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.:. = t =. n + = :.= : . rt==q= rut = - rc-! =2.qn,

= c.

r t-r:t= n==

r :=

=F= =j=1==i=-l-ni=-== F =-~ ni==t===r----t.===turcat_.._ 3._-- j o E=Fi-A [. ;._ ["i. . _ M.. ~.1:1.... _f.M._ Measured Power Doppler _...Z ZT .u.g. r. 3 ,7 ~...:,.-- .r. _..._ . -,,Z ir.Z:=:=......+J nf= Coe f ficient Results M. ma. =. _::=. :p= =. t.=.=.:. _....z i

n. ;.

t=;.- : 1 40 %FP O y =..,r=.,, =. =_=. ; :=

. u= p2_:==g.=;}=t

t -- 11. 75 %FP e=

e

a

e-

u... p

i- -

m jaTdi;,_gp2;g;iEE[E]E i y 90 %FP

sig gfF

=ipE=EjE={:f:ig? EEi= =EjE J g=- a --- 100 %FP

24: =

a= = = =. = - m:--g= =l=--~2=g=u_==,==a=,.=. =. 2=;n._ ; 2.l-*^- =:=p--- =---.. _.. _..2. t -.-. 5 ~- - * - ' ~ ' : =4. 4

r. a.

t-~~ l

==t= c HEi-f=4= i EE=EEi2 HEiE=EE!E}F=-IVEEEiE:li=aF =t=E =in - El"E

=.

l===c=n.d =_r0 D =:==

==: =" ====f -- =y2-

=

=. = =. . 21_. -..a__ ). a.. h_....::. b)(h i.-. = r ~- M =t rf= 60._. m~ t = _. 3._. _....._._7__16 = t=n . u :: r ~ u;8 c_._..,.._._g_._..v__" ... t- _ - .y_,,.__,.. g... ._=__x,.t----- .y. n_. Power Level, %FP d t Figure 4.3-3

4.4 CORE POWER DISTP.IBUTION 4.4.1 PURPOSE Steady-state core power distribution measurements were intended to verify that the core axial power imbalance, quadrant power tilt, minimum DNBR, and maximum linear heat rate do not exceed their specified limits. Acceptance criteria specified for the Core Power Distribution Test included: (a) The combination of reactor power and reactor power imbalance does not exceed the safety limits as defined by the locus of points established in Figure 2.3-2C of the Technical Specifications (Figure 4.4-1 of this report). (b) DNBR is greater than 1.55. (c) The quadrant power tilt does not exceed four percent, except during certain physics testing. (d) The measured maximum linear heat rate multiplied by the worst case uncertainty factor is less than 17.5 kw/ft. (e) The measured maximum linear heat rate, multiplied by the worst case uncertainty factor and then extrapolated to the Technical Specification power-imbalance envelope, is less than the 19.8 kw/ft central fuel melt limit. (f) The highest measured radial peaking factor is not more than 5.0% greater than predicted; and the highest total peaking factor, not mere than 7.5% greater. 4.4.2 TEST METHOD ~ When required by the test program, core power distributions were determined af ter establishing steady-state conditions at the required power level and at rod configurations specified by the controlling procedure for power escalation. In order to compare the measured power distribution to pre-dicted results, some cases required either two-dimensional or three-dimensional xenon equilibrium. To assure three-dimensional xenon equilibrium, the in-core axial imbalance must be determined to be within i 3 percent of the equilibrium imbalance value for an eight-hour period while the part-length control rods are maintained at a constant position. In other cases, data were taken according to required conditions for the individual tests. 4.4.3 EVALUATION OF THE TEST RESULTS Steady-state core power distributions were measured for normal operating control rod configurations at each of the major power escalation test plateaus. Additional core power distribution measurements were made at 75 4.4-1

? percent f ull power for dif f e' rent control rod group configurations. In the following sections analysis of these two core power distribution measurements are presented. Analyses of other core power distribution measurements required in conjunction with other power escalation tests are presented elsewhere in this report. Table 4.4-1 summarizes core power distributions taken during the Core Power Distribution Test. 4.4.3.1 Steady-State Core Power Distributions at Equilibrium Xenon, Normal. Operating Control Rod Configuration Conditions Steady-state, equilibrium xenon core power distributions were measured at each of the major power escalation test plateaus for control rod configurations for normal operation of the unit. The measurements covered the following control rod patterns and core power levels: Case Power Level Control Rod Position (% Wd) Equilibrium Number (% FP) 1-5 6 7 8 Xenon 1 15 100 74 00 34 Yes/2-D 2 40 100 75 02 19 Yes/3-D 3 75 100 79 04 30 Yes/3-D 4 100 100 90 15 00 Yes/3-D Measured core power distributions for these four cases are shown in Tables 4.4-3 through 4.4-5. These tables give a complete 1/8 core power distribution map using the corrected signal outputs frem 203 incore detectors located in 29 d if f erent fuel assemblies which describe the entire core assuming eighth core symmetry. A summary of the results of each of these core power distri-bution measurements is given in Table 4.4-1, which tabulates the core power level, the control rod positions, core burnup, boron concentration, axial imbalance, maximum quadrant tilt, maximum LHR, minimum DNBR, and power peaking data for each measurement. 4.4.3.1.1 Radial and Total Peaking Factors The results of the four core power distribution measurements indicate a maximum radial peaking factor between 1.35 and 1.47 and a maximum total peaking factor between 1.79 and 2.07. The maximum radial peaking factor appears to decrease with increasing power for a given control rod con-figuration, due to thermal feedback. Figure 4.4-2 shows the degree of power flattening on the radial power profile, as observed along the X-Z plane for various power levels. As can be seen, a relatively flat radial power distribution is observed at 100% FP. In all cases, the maximum total peaking factor is well below the design value of 2.67. The measured radial and total peaking factors are compared with their calculated values in Figures 4.4-3 through 4.4-10 to demonstrate the degree of agreement between the measured and predicted power distributions. As can be seen from these figures, f avorable agreement was observed, except on the location of the highest radial and total peaks. However, in all instances, the measured maximum radial and maximum total peaking factors were not more than 5.0% and 7.5% of the respective predicted values. A comparison of the measured and the predicted maximum radial and maximum total peaking factors - are shown in Table 4.4-7. 4.4-2

4.4.3.1.2 Minimum DNBR Minimum DNBR values were calculated by the unit computer as part of the standard core power distribution calculations. The minimum DNBR values obtained during the core power distribution measurements are plotted in Figure 4.4-12. For each measured minimum DNBR a worst case minimum DNBR was calculated by subtracting the worst case uncertainty (0.68) from the measured minimum DNBR. The measured and worst case minimum DNBR's were then extrapolated to the boundary of the Technical Specification power-imbalance envelope limits. The measured, worst case, and extrapolated minimum DNBR's are tabulated in Table 4.4 All measured and worst case minimum DNBR's prior to and after extrapolation were above the design limit of 1.55. A minimum margin of 106 percent from the design limit of 1.55 was noted for the extrapolated minimum DNBR, and a minimum margin of 65 percent was noted 3 for the extrapolated worst case minimum DNBR. 4.4.3.1.3 Maximum Linear Heat Rate Maximum linear -heat rate (LHR) values were calculated by the unit computur as part of the standard core power distribution calculations. For each measured r'aximum LHR the worst case maximum LHR value was calculated by multiplying the measured maximum LHR by the worst case uncertainty factor (given in Table 4.4-2) and compared with the LOCA limit of 17.5 kw/ft. The measured and the worst case maximum L4R's were then extrapolated to the boundary of the Technical Specification power-imbalance envelope limit and compared with r the central fuel melt limit of 19.8 kw/f t. The results of these analyses are presented in Table 4.4-12. All worst case maximum LHR's and a.1 ' extrapolated worst case maximum LHR's met the respective LOCA and central fuel melt. acceptance criteria. 4 4.4.3.1.4 Quadrant Power Tilt and Axial Power Imbalance j Table 4.4-1 shows the maxi =um quadrant power tilts measured by the in-core detector system during the Core Power Distribution Test. All the measured-i quadrant power tilts are well within the 4 percent Technical Specification limit. The axial power imbalances measured during.the core power distribution test i are also shown in Table 4.4-1. These power imbalances in combination-with the corresponding values of the reactor power are well within the Technical-Specification power-imbalance safety limits. 4.4.3.2 Core Power Distribution at 75% FP for Different Control Rod Configurations Steady-state core power distributions at -75% FP and equilibrium xenon ) conditions were measured for different control rod group positions. These measurements were made to study the effects of control rod positions on' core l_ power distribution. The measurements covered the following control rod: 1- -patterns: i .4.4-3 a l

j e' .a' Case Power Level Control Rod Position (%Wd) Equilibrium Number (% FP) 1-5 ,6, 7, 8, Xenon 1 75 100 25 00 10 Yes/2-D 2 75 100 75 03 19 Yes/2-D 3 75 100 100 24 04 Yes/2-D 4 75 100 100 75 24 Yes/2-D The measured minimum DNBR, maximum LHR, and the highest radial and total peaking factors are given in Table 4.4-1 along with other pertinent data. The Tables 4.4-8 through 4.4-11 contain the measured core power distributions based on 1/8 core symmetry. The results indicate a maximum radial peaking factor between 1.34 and 1.56 and a maximum total peaking factor between 1.77 and 2.24 for the four cases studied. The radial peaking factor appears to be strongly dependent on the control rod group position at constant power. Figure 4.4-11 shows the effect of the control rod group positions on the radial power distribution as observed along the X-Z plane. As can be seen, the highest radial peaking was observed at the center of the core (core location H-08) when Group 7 was at 75 percent withdrawn. In'all cases, the observed maximum total peaking factors were much less than 2.67, the design 1 value. Worst case analyses were performed for the measured maximum LHR's and minimum DNBR's. The results of thesa analyses showed that the worst case maximum LHR's and the ' extrapolated worst case LHR's were less than their respective acceptance limits. In a similar manner, the measured, worst case, and extrapolated worst case DN3R's were well above the design limit of 1.55. 1 4.

4.4 CONCLUSION

S Comparison of the four steady-state core power distributions, taken at-15 40, 75, and 100 %FP with the normal control rod configurations, with the predicted distributions showed that the measured values for the maximum radial peaking factor and the maximum total peaking factor were not more than 5 pe :ent and 7.5 percent of the predicted values, recpectively. The mes ared maximum total peaking factors were much less than the design va.de of 2.67. Because of the flatter radial power profile and lower maximum peaking factors, the observed core power distribution was more favorable than had initially been predicted. All measured anl worst case minimum DNBR's were well above the minimum design limit of 1.55 even af ter extrapolation to the boundary of the Technical Specification power-ir' alance envelope limits. A minimum margin of 106 percent and 65 percent respectively was noted for the measured and worst case extrapolated minimum DNBR. All measured and worst case LHR's resulted in adequate margins for LOCA limited and central fuel melt limited LHR's. 1 The oilerved quadrant power tilts and axial power imbalances were well within acceptable limits. 4.4-4 e - w w

0 Analysis of the core power distributions taken at 75% FP for various control rod positions verified acceptable thermal conditions, even when the worst case uncertainty factors were applied to the measured minimum DNBR's and maximum linear heat rates. .dP s 4.4-5

9

SUMMARY

OF MEASURED CORE POWER DISTRIBUTION RESULTS AT EQUILIBRIUM XENON CONDITIONS FOR VARIOUS CONTROL ROD PATTERNS AND CORE POWER LEVELS OF 15, 40, 75, AND 100 PERCENT FULL POWER. Power Rod Position, %wd Core Boron Axial Maximum Thermal Max 1 mum,9y gp, Level Burnup Conc. Imb. Tilt Min. Max. g,, n ar,- Time (7FP) 1-5 6 7 8 (EFPD) (ppm) (7FP) (7) DNRM THR Radial Total A. Stel dy-St< tc Core l'ower 1.istrit utions with Normal 01. crating Control R( d Configi.ratior 09-20-74 1104 15.14 100 74 00 34 0.8 1242 -3.46 -1.63 28.00 1.88 1.47 2.07 10-23-74 1210 41.44 100 74 00 19 4.4 1110 -3.68 +2.03 10.16 4.46 1.41 1.79 11-19-74 2153 75.79 100 79 04 30 12.1 1037 -14.00 +1.76 5.75 8.53 1.37 1.82 12-18-74 1401 95.58 100 90 15 00 32.8 1020 -3.75 +1.57 3.83 11.28 1.35

1. ?0 I!

u-E B. Cori Power Distributions at 75 %FP ft r Different Control Rod Configur:.tions 12-02-74 2057 74.60 100 75 03 19 21.0 1030 -14.99 +1. 64 4.59 8.41 1.38 1.84 12-04-74 1933 76.57 100 25 00 10 22.5 960 - 2.56 +1.41 5.03 8.48 1.42 1.77 12-05-74 0502 76.18 100 100 24 04 22.8 1090 + 0.08 +1. 80 4.88 9.21 1.35 1.94 12-06-74 1432 75.34 100 10 0 75 24 23.7 1146 - 1. 80 +1.57 4.46 9.68 1.56 2.24 s I P

f - MINIMUM DNBR AND MAXD!UM LHR WORST CASE UNCERTAINTY FACTORS A. Minimum DNBR Worst Case Uncertainty Re duce the measured value of the minimum DNBR by a worst ' case of 0.68 to account for the following: Radial Uncertainty & Heat Balance Error 0.25 Inlet Temperature Error 0.03 System Pressure Error 0.16 Fuel Densification 0.09 Axial Model Correction & Axial Uncertainty 0.03 4% Quadrant Power Tilt 0.12 Total 0.68 These ite=s represent uncertainties and conservativisms which, when sub-tracted from the measured DNBR, result in a worst case DNBR number which will be conservative even if all e'rrors and uncertainties were actually present at the same time, at full magnitude, and in the worst directicn. 4 B. Maximum LHR Worst Case Uncertainty Factors The worst case uncertainty factor which is applied depends on the percent weight of lumped burnable poison in the fuel assembly as 4 follows: Ascembly Lumped Burnable Wo rs t Cas e Type Poison Weight Uncertainty Factor 1 0.00% 1.357 2 1.09% 1.402 3 1.26% 1.411 4 1.43% 1.419 The above factors. assume the worst case' conservatism factors ms. listed below: Maximum Power Spike Factor - = 1.100 (9 feet above bottom of core) Fuel Densified Active Height, ft = 11.638 Maximum Quadrant Tilt Factor = 1.074 (4% Quadrant Tilt) Radial Local Peaking Factors = 1.050 (Depends on LBP enrichment)' l.085 -1.092 1.098 These items represent uncertainties which when multiplied.by the measured ' maximum LHR, result in the worst case LHR which'will be conservative even if all uncertainties' were actually present at-the same time, at full mag-nitude, and in the worst direction. ' Table 4.4-21

MEASURED CORE POWER DISTRIBUTION RESULTS AT 15% FP Centrol Rod Group Positions Gps 1-5 100.0 % wd GP 7 00.0 % wd Gp 6 74.0 % vd GP 8 34.0 % ud Core Power Level 15.1 % FP Boron Concentration 1242 PPM Core Burnup 0.8 EFPD Axial Imbalance -3.5 % FP Xenon Conditions Equilibrium Conc, Yes Yes or No Reactivity Worth -1.24 % ak/k Max Quadrant Tilt -1.63 % 1/8 Core Incore Weigh ting Paax/ P/P Fuel Fuel Assy. De te cto r Factor {o cag_ Assembly cor i nc,H nn Numbe r m H-0 8 1 1 1.42 1.07 G-08 2 4 1.76 1.29 F-08 4 4

1. 74 1.24 E-0 8 10 4

1.86 1.29 D-0 8 14 4 1.69 1.16 C-08 21 4 1.82 1.28 B-0 8 30 4 2.07 1.47 A-08 37 4 1.38 0.98 G-09 3 4 1.72 1.22 F-10 12 4 1.76 1.25 E-11 26 4 1.58 1.08 D-12 41 4 1.35 0.93 C-13 52 4 0.81 0.58 F-09 6 0 1.83 1.32 E-09 5 8

1. 71 1.21 D-09 15 8

1.78 1.20 C-09 29 8 -1.53 1.08 B -09 31 8 1.46 1.06 A-09 45 8 1.15 0.83 E-10 17 8 1.82 1.28 D-10 27 8 1.53 0.97 C-10 28 8 1.42 0.98 B-10 44 8 1.01 0.73 A-10 46 8 0.66

0. 49 D-11 33 8

1.59 1.07 C-11 42 8 1.23 0.84 B-11 49 8 o,go o,71 C-12 48 8 1,19 0,g5 B-12 31 6 0.69 0.50 Table 4.4-3

e MEASURED CORE POWER DISTRI3UTION RESULTS AT 40% FP Control Rod Group Positions Gps 1-5100.0 % ud GP 7

1. 8 % ud Gp 6 75.0 % ud GP 8 19.4 % wd Core Power Level 41.1 % FP Boron Concentration 1110 PPM Core Burnup' 4.4 EFPD Axial I= balance

-3.7 % FP Xenon Conditions i Equilibrium Conc. Yes Yes or No Reactivity Worth -2.06 % 4k/k Max Quadrant Tilt +2.03~ 4 1/ 8 Co re Incore Wei gh ting P=ax/ P/P Fuel Fuel Assy. De te c to r Factor {corg Assembly t nc,s en Nurle r ca H-0 8 1 1 1.39 1.11 G-0 8 2 4 1.62 1.27 F-OS 4 4 1.63 1.24 E-0 8 10 4 1.58 1.24 D-0 8 14 4 1.50 1.11 C-08 21 a 1.65 1.27 B-0 8 30 4 1.79 1.41 A-08 37 4 1.20 0.94 G-09 3 4 1.59 1.25 F-10 12 4 1.67 1.33 E-11 12 6 4 1.34 1.01 D-12 41 4 1.15 0.93 C-13 52 4 0.77 0.60 F-09 6 0 1.64 1.29 E-09 5 8 1.59 1.26 D-09 15 3 1.44 1.17-C-09 29 8 1.43 1.14 B -0 9 - 31 8 1.27 -1.03 A-09 45 9 1.01

0. 80 i

E-10 17 8

1. 72 1.34 D-10 2/

8 1.27 0.99 C-10 29 8 1.26 -0.98 4 l B-10 44 8 0.91 0.72 A-10 46 8 0.60 0.49 D-ll 33 l 8 1.38 1.06 C-ll 42 .8 1.06 0.84 B-ll-49-S o,90 0,71 C-12 48 8 1,07 0,85 B-12 51 5 0.65 0.51 Table 4.4-4 u--

MEASURED CORE POWER DISTRIBUTION AT 5% FP Control Rod Group Positions Cps 1-5 100.0 % vd GP 7 03.5 % wd Gp 6 78.9 % wd GP 8 30. 3 % wd Core Power Level 75.2 % FP Boron Concentration 1037 PPM Core Burnup 12.1 EFPD Axial Irbalance -14.0 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth -2.60 % Ak/k Max Quadrant Tilt +1. 76 % 1/ 8 Co re Incore Weigh ting Pmax/ P/P Fuel Fuel Assy. De tec to r Factor Peore Assembly i n c,.-4 n, Nu-be r Incal H-0 8 1 1 1.42 1.14 G-08 2 4 1.64 1.25 4 1.72 1.26 F-08 4 E-0 8 10 4 1.70 1.24 D-0 8 14 4-1.58 1.18 C-08 21 4 1.65 1.27 B-0 8 30 4 1.82 1.37 A-08 37 4 1.21 0.92 G-09 3 4 1.66 1.28 F-10 12 4 1.76 1.36 E-ll 26 4 1.45 1.01 D-12 41 4 1.27 0.92 C-13 52 4 0.75 0.59 F-09 6 8 1.77 1.30 E-09 5 8 1.71 1.29 D-09 15 8 1.69 1.18 C-09 29 8 1.47 1.13 B-09 31 8 1.29 0.99 A-09 45 8

1. 0.1 0.78 E-10 17 8

1.80 1.34 D-10 27 8 1.50 1.03 c-10 28 8 1.34 0.99 B-10 44 8 0.94 0.70 A-10 46 8 0.61 0.47 D-11 33 8 1.48 1.06 C-11 42 8 1.15 0.83 B-ll 49 8 0.91 0.70 .C-12 48 8 1.10 0.84 B-12 l 51 6 0.65 0.50 Table 4.4-5

MEASURED CORE POWER DISTRIBUTION RESULTS AT 100% FP Control Rod Group Positions Gps 1-5 10'0.0 % wd GP 7 15.0 % wd Gp 6 90.0 % wd - GP 8 00.0 % wd Core Power Level 95.6% FP Boron Concentration 1020 PPM Core Burnup

32. 8EFPD Axial Imbalance

-3.8 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth -2.76 % ak/k Max Quadrant Tilt +1.57% i 1/8 Core Incore Weigh ting Pmax/ P/P Fuel gory Assembly Fuel Assy. De te c to r Factor i n e.n H nn Nu.be r ca ' H-0 8 1 1 1.48 1.21 G-08 2 4 1.61 1.23 F-08 4 4 1.69 1.25 E-0 8 10 4 1.75 1.27 D-0 8 14 4 1.64 1.25 C-08 21 4 1.73 1.28 B-0 8 30 4 1.69 1.31 A-08 37 4 1.14 0.89 G-09 3 4 1.57 1.25 F-10 12 4

1. 80 1.34 E-11 26 4

1.54 1.14 D-12 41 4 1.23 0.90 C-13 52 4 0.78 0.57 F-09 6 8 1.73 1.29 E-09 5 8 1.74 1.27 D-09 15 8 1.66 1.22 2-09 29 8 1.48 1.10 "'d-09 31 8 1.25 0.96 A-09 45 8 0.97 0.77 E-10 17 8 1.92 1.35 D-10 27 8 1.57 1.02 C-10 28 8 1.38 1.01 B-10 44 8 0.88 0.71 A-10 46 8 0.57 0.47 D-11 33 8 1.53 1.08 C-ll 42 8 1.18 0.83 B-11 49 8 0.94 0.71 C-12 43 8 1.12 0.85 B-12 31 S 0.66 0.50 Table 4.4-6

• w:

9 COMi'ARIS0ti OF MEASURED AND PitEDICTED MAXIMUM RADI AL AND TOTAL PEAKING FACTORS Powe r Maximum Predicted Maximum Measured Percentage Error Level Peaking Factors Peaking Factors Radial Total ~(%FP) Radial Total Radtal

Total,

(%) (%) 15 1.48 2.00 1.47 2.07 - 0. 7 + 3.4 40 1.47 1.88 1.41 1.79 - 4.3 - 5.0 75 - 1.51 2.03 1.37 1.82 -10.2 -11.5 100 1.41 1.81 1.35 1.92 - 4.3 + 5.7 .;n s. Note: Percentage Error = ' Measured - Predicted-x 100 1-Measured Note: During the 100 %FP core power distribution measurement, the part-length rods were at 00 % withdrawn instead of the predicted position of 20.8 % withdrawn. 3. Note: The'seco'nd and third. highest measured radial and total. peaking factors also j il met the acceptance criteria. M e 1 a. -

MEASURED CORE POWER DISTRIBUTION RESULTS AT 75% FP DURING CONTROL ROD MANEUVERING, WITil CROUPS 5-8 AT 100, 25, 00 AND 10 % WITliDRAWN, RESPECTIVELY Control Rod Group Positions Gps 1-5 100 % wd GP 7 0.0 % wd Gp 6 25.4 % wd GP 8 9.8 % wd Core Power Level 76.6 % FP Boron Concentration 960 PPM Core Burnup 22.5 EFPD Axial Imbalance -2.6 % FP Xenon Conditions Equilibrium Conc. Yas Yes or No Reactivity Worth -2.61 % ak/k Max Quadrant Tilt +1.41 % 1/ 8 Co re Incore Weigh ting Paax/ P/P Fuel Fuel Assy. De te cto r Factor Pcom Assembly inc2etnn Numbe r Lo c al H-0 8 1 1 1.57 1.15 G-08 2 4 1.73 1.32 F-08 4 4 1.71 1.35 E-0 8 10 4 1.45 1.21 D-0 8 14 4 1.21 0.92 C-08 21 4 1.48 1.22 B-0 8 30 4 1.77 1.42 A-0 8 37 4 1.25 1.00 G-09 3 4 1.73 1.41 F-10 12 4 1.72 1.37 E-11 26 4 1.01 0.76 D-12 41 4 1.10 0.90 C-13 52 4 0.81 0.61 F-09 6 8 1.71 1.37 ~ E-09 5 8 1.57 1.27 D-09 15 8 1.38 1.12 C-09 29 8 1.40 1.10 B-09 31 8 1.29 1.03 A-09 45 8 1.08

0. 86 E-10 17 8

1.52 1.27 D-10 27 8 1.38 0.98 C-10 28 8 1.35 1.01 B-10 44 8 0.93 0.73 A-10 46 8 0.65 0.51 D-11 33 8 1.29 0.99 C-ll 42 8 1.14 0.85 B-11 49 8 0.99 0.75 C-12 48 8

1. 15

0. 89 B-lZ 51 6

0.71 0.54 Table 4.4-8

MEASURID CORE POWER DISTRIPUTION RESULTS AT 75% FF DURING CONTROL RQD MAlfEUVERING, WI.' CROUPS 5-8 AT 100, 75, 03 Alm 19 % WITHERAWN, RESPECTIVELY Control Rod Group Positions Gps 1-5 '100.0 % wd GP 7 2.7 % vd Cp 6 75.2 % wd GP 8 18. B_% ud Core Power Level 74.6 % FP Boron Concentration 1030 PPM l Core Burnup 21.0 EFPD Axial Imbalance -15.0 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth -2.57 % ak/k Max Quadrant Tilt +1. 64 % 1/8 Core Incore Weigh ting Pmax/ P/P Fuel Fuel Assy. De te c to r Factor {cery Assembly 3 i nc,H en Numbe r car H-0 8 1 1 1.53 1.01 G-08 2 4 1.71 1.22 F-08 4 4 1.75 1.28 E-08 10 4 1.74 1.26 D-0 8 14 4 1.58

1. 19 C-08 21 4

1.66 1.28 B-0 8 30 4 1.84 1.35 A-08 37 4 1.22 0.92 G-09 3 4 1.73

1. 30 F-10 12 4

1.80 1.38 E-11 26 4 1.41 1.01 D-12 41 4 1.26 0.91 C-13 52 4 0.74 0.58 F-09 6 0 1.83 1.31 i E-09 5 8 1.73 1.29 D-09 15 8 1.69 1.21 C-09 29 8 1.47 1.11 B-09 31 8

1. 30 0.97 A-09 45 8

1.05 0.79 E-10 17 8 1.76

1. 35 D-10 27 8

1.39 1.02 C-10 28 8 1. 30 0.99 B-10 44 8 0.95 0.69 A-10 46 8 0.63 0.47 D-11 33 8 1.42 1.06 C-11 42 8 1.15 0.85 B-11 49 8 0,91 0,70 C-12 48 S 1.11 0.83 3-12 l 31 S 0.65 0.50 Table 4.4-9

k MEASURED CORE POWER DISTRIBUTION RESULTS AT 75% FP DURING CONTROL P.0D MANEUVERING, WITH GROUPS 5, 6, 7, AND 8 AT 100% wd,100% wd, 24% wd, ~AND 4% wd, RESPECTIVELY Control Rod Group Positions Gps 1-5 100.0 % vd GP 7 24.3 % vd Gp 6 100.0 % wd GP 8 3.5 % wd Core Power Level 76.2 % FP Boron Concentration 1090 PPM Core Burnup 22.8 EFPD Axial Imbalance +0.1 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth -H % ak/k 7 Max Quadrant Tilt +1.81 % 9 1/8 Core Incore Weigh ting Pmax/ P/P Fuel 3 Fuel Assy. De te cto r Factor {cor$ Assembly f nnmH nn Numbe r ca H-08 1 1 1.53 1.07 G-08 2 4 1.54 1.15 4 F-08 4 4 1.67 1.20 E-0 8 10 4 1.75 1.24 D-0 8 14 4 1.65 1.17 C-08 21 4

1. 75 1.29 B -0 8 30 4

1.73 1.35 A-08 37 4 1.16 0,92 G-09 3 4 1.56 1.21 I F-10 12 4 1.83 1.31 E-11 26 4 1.55 1.09 D-12 41 4 1.25 0.90 C-13 52 4 0.78 0.56 F-09 6 0 1.72 1.24 E-09 5 8 1.76 1.25 D-09 15 8 1.67 1.21-C-09 29 8 1.51 1.13 B-09 31 8 1.26 1.01 A-09 45 8 1.00 0.81 E-10 17 8 1.94 1.34 D-10 27 8 1.57 1.01 C-10 28 8 1.38 1.03 B-10 44-8 1.13-0.80 A-10 46 8 0.69 0.51 D-11 33 8 1.56-1.09 C-11 42 8 1.16 0.78 ) i B-11 49- ,8 0.93 0.74 C-12 48 8 1.13-0.85 B-12 { 31 6 0.66 0.50 } Table 4.4-10 i ^

e MEASURED CORE POWER DISTRIBUTION RESULTS AT 75% FP DURING CONTROL ROD MANEWERING, WITH GROUPS 5, 6, 7, AND 8 AT 100% vd,100% vd, 75% wd, AND 24% wd, RESPECTIVELY Control Rod Group Positions l Cps 1-5 100.0 % wd GP 7 78. 4 % ud Gp 6 100.0 % wd GP 8 16.1 % ud Core Power Level 75. 3 % FP Boron Concentration 114GPPM Core Burnup 2 3.7 EFPD Axial Imbalance -1.8 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No ) Reactivity Worth -2.58 % Ak/k Max Quadrant Tilt +1. 5 % 1/8 Core Incore Weigh ting Pmax/ P/P Fuel j Fuel Assy. De te cto r Factor ?co$ Assembly t nn,H m Nu.be r Eca j H-0 8 1 1 2.24 1.56 j G-0 8 2 4 1.77 1.23 i F-08 4 4 1.55 1.12 E-0 8 10 4 1.51 1.11 i D-08 14 4 1.46 1.07 C-08 21 4

1. 75 1.26 B-08 30 4

1.93 1.40 A-08 37 4 1.36 0.98 G-09 3 4 1.62 1.20 F-10 12 4 1.60 1.17 E-11 26 4 1.32 0.92 D-12 41 4 1.11 0.83 C-13 52 4 0.75 0.54 F-09 6 8 1.56 1.14 E-09 5 8 1.52 1.12 D-09 15 8 1.52 1.11 I C-09 29 8 1.56 1.13 B-09 31 8 1.58 1.11 A-09 45 8 1.26 0.90 E-10 17 8 1.68 1.19 D-10 27 8 1.41 0.96 C-10 28 8 1.53 1.09 B-10 44 1.61 1.06 o A-10 46 8 0.90 0.64 D-11 33 8 1.42 0.99 C-11 42 8 1.24-0.91 I -B-11 49 8 1.22 0.85 j i C-12 48-8 1.11 0.81 i B-12 51 5 0.70 0.51 Table 4.4-11 1 n 1 --c.

~ MINIMUM UNBR AND MAXIMUM LIIR ANALYSIS FOR-CORE POWER DISTRIBtTTION TEST Power Inco re Power at Mivimum THR. Vu/ft Minimiin DNBR. dfm . Assembly Envelope Vorst Worst Level Offset (y yp) (7) Type Limits (%FPl Measured Case Measured Case A. Steady-S. ate Core l'ower Distr.butions Wi th Normal Operating Cantrol Rod Configurat ion Measured 15.14 -22.85 1 1.88 2.55 28.00 27.32 Extrapolated 92.20 11.45 15.54 3.20 3.00 Measured 41.44 - 8. 89 1 4.46 6.05 10.16 9.48 Extrapolated 101.94 10.97 14.89 4.30 3.80 tkasure d ' 75.59 -18.67 1 8.53 11.57 5.75 5.08 Ext rapolated 94.92 10.71 14.34 4.70 4.10 a Measured ' 95.58 - 3.93 2 11.28 15.82 3.82 3.14 5 Extrapolated 105.92 12.50 17.53 3.83 2.72 P .7 B. Core Pow <-r Distribii tions at 7S% FP for IJ ifferent Control Rod 2cnfigurat ons iC Measured 74.60 -20.09 1 8.41 11.41 4.59 3.91 Extrapolated 93.98 10.59 14.38 3.60 2.95 Heasured 76.57 - 3.34-1 8.48 11.51 5.03 4.35 Extrapolated 106.41 11.78 15.99 3.75 3.10 Measured 76.18 + 0.11 2 9.21 12.92 4.88 4.20 Ext rapolated 107.00 12.94 18.14 3.60 2.95 Meas ured 75.34- - 2. 39 1 9.68 13.14 4.46 3.78 Extrapolated 107.00 13.75 18.66 3.20 2.55 Note: 'Ibe worst case uncerta uty factor s are obta:ned from T able 4.4-2, \\

. 1. o. POWER-IMBALANCE ENVELOPE I 4 .- r rt: rt.

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= Power Imbalance, %FP Figure 4.4-1 L

MEASURED RADIAL CORE POWER DISTRIBUTIONS All1NG THE X-Z PI ANE WITil NORMAL. OPERATING CONTR01, ROD CONFIGURATION 2.1 2.1 d1.8 . d 1.8 .m N 15 N 15

1. N A

M X 1.2 k MW w y 'Nr' Nr ,. 1.2 qg wr 0.9' {

0. 9' S

a. ~ g 0.6 Power Level - 15.1 ZFP ~ Q Power Level = 4'1.4 ZFP' 0.6 g 0.3 Core Bumup = 0.8 MD _ y 0.3 Core Burnup = 4.4 EFPD _ 0.0 0.0 A B C D E F G II K L M N O P R A B C D E F G 11 K L M N O P_ R Radial Position Fuel Assenh11es Radial Position Fuel Assemblies nj 5 r2 t~ tJ 2.1 2.1 21.8 d1.8 1A th N 1.5 N 1.5 4 \\ / I 4 i ,. 1 2 k,,' "wr'O k, 1.2 p. ) g r { 0.'f { 0.% O o. k a 0.6 Power Level = 75.8 ZFP - g 0.6 Power Level = 95.6 %FP ' h .3 Core Burnup = 12.1 EFPD_ y 0.3 4

• " " ~

0 0 l 0.0 0.0 A B C D E F G K L M N O P R' A B C D E F G K L M N O P R Radial Position, Fuel Assenblies Radial Position, Fuel Assemblies l 1 i

e COMPARISON OF PREDICTED AND MEASURED RADIAL PEAKING FACTORS AT STEADY-STATE, 2-D EQUILIBRIUM XENON,15% FP CONDITIONS Predicted Conditions Control Rod Group Positions Core Power Level 15.0 gyp Gps 1-5 100.0 % wd Boron Concentration NA ppm Cp 6 75.0 % wd Core Burnup h.0 EFFD Gp 7 00.0 % wd Axial Imbalance -2.o %EP Cp 8 37.'s % wd Max Quadrant Tilt 0.00 Measured Conditions Control Rod Group Positions Core Power Level 15.1 %FP Cps 1-5 100.0 % wd Boron Concentration 1242 Fpm Gp 6 74.0

7. vd Core Burnup 0.6 EFPD Gp 7

00.0 % vd Axial Imbalance -3.5 %FP Gp 8 34.0 % wd Max Quadrant Tilt -1.63 % i H G F E D C B A f 1 '. 1 31 1.3 ..l.h5

1. 6' 1.25 1.28 q.o2 g

a m 1.07 1.29 1.2h 1.20 1.16 1.28

1. h'7 0.08 5

1. 5 1.h8 1.2h 1.26 1.01 0.95 0.76 9

l.2. 1 32 1.21 1.20 1.08 1.06 0.83 s a i 1.'T 1.32 0 98 0 9h 0.6h 0.h7 10 1.2 1.28 0 97 0.98 0.73 0.L9 i N e 3

1. 7 1.08 0.80 0.68 11 1.0 1.07 0.8h 0 71 b,89 0.86 0.55 12 0.

0.85 0.50 x 0163 13 o, s 14 15 I E X.XX Predicted Results X.XX Measured Results Figure 4.4-3

s , COMPARISON OF PREDICTCD AND MEASURED TOTAL PEAKING FACTORS AT STEADY-STATE, 2-D EQUII.1BRIUM XENON, 15% FP CONDITIONS Predicted Conditions Control Rod Group Positions Core Power Level 15.0 %FP Cps 1-5 100.0 % vd Boron Concentration HA ppm Gp 6 75.0 % wd Core Burnup h.0 EFPD Gp 7 00.0 % wd Axial Imbalance -3.9 %FP Gp 8 37.5 % wd Max Quadrant Tilt 0.00 % M_easured Conditions Control Rod Group Positions Core Power Level 15.1 gyp Gps 1-5 100.0 % vd Boron Cencentration 12h2_ ppm Gp 6 Th.0 % vd Core Burnup 0.8 _ EFPD Gp 7 00.0 % wd Axial Imbalance -3.5 %FP Cp 8 'h.0 % wd Max Quadrant Tilt -1.63 % i H G F E D C B A 1.98 1.21' 2.00,1.86 1.69 1.82 2.07 1.38 g g 1.Ch 1 76 1 74 5 9 1.%7 1 98 1.62 1 73 1.83 1 71 1.78 1.53 1.46 1.15 s 's 8 7 1.71 1.h1 10 1.i 1.82 1.53 1.h2 1.01 0.66 1 's 6 5 11

1. '

1.5 1.59 1.23 0.c9 N, i 1 3)s 1.19 0.69 x 13 0'(8 0.81N s 14 15 E X.XX Predicted Results X.XX Measured Results Figuru 4.4-4

U y COMPARISON OF PREDICTED AND MEASURED RADIAL PEAKING FACTOR $ AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 40% FP CONDITION,S Predicted Conditions Control Rod Group Positions Core Power Level h0.0 gyp Gps 1-5 100.0 % ud Boron Concentration IIA ppm Cp 6 75.0 % wd Core Burnup 4.0 EFPD Cp 7 00.0 % vd Axial Imbalance h.6 %F? Cp 8 30.8 % vd Max Quadrant Tilt 0.00 % Measured Conditions Control Rod Group Positions Core Power Level 41.4 gyp Cps 1-5 100.0 % wd Boron Concentration 1110 ppm Cp 6 75.0 % wd Core Burnup h.4 EFPD Cp 7 1.d % wd Axial Imbalance -3.7 %FP Cp 8 19.h % wd Max Quadrant Tilt +2.03 % H G F E D C B A 7 6 0.98 1.20 1.27 1.h3 1 12,_M8_

1. M Q1._. g 8

1. E., 1.27 1.24 1.2h 1.11 1.27 1.h1 0 9h N

s 1 1.h7 1.21 1.27 1.00 0 97 0.78 9 1.2 1.29 1.26 1.17 1.1h 1.03 0.80 s 4 7

1. L 1.32 0.95 0.9h 0.6h 0.h8 10 13 1.3h 0.99 0 98 0.72 0.h9 i

's 6 5

1. 2 1.11 0.82 0.71 17 1.0 1.06 0.8h 0.71 D0 0.91 0.58 09 0.85 0.51 x

0367 0.k i s 14 15 l E X.XX Predicted Results X.XX Measured Results Figure 4.4-5

U ~ COMPARISON OF PREDICTED AND MEASURED TOTAL PEAKING FACTORS AT STEADY-$ TATE, 'l-D EQUILIBRIUM XENON, 40% FP CONDITIONS Predicted Conditions Control Rod Group Positions Core Power 14 vel LO.0 %FP Gps 1-5 100.0 % vd Boron Concentration ;iA ppm Gp 6 75.0 % wd Core Burnup 4.0 EFPD Cp 7 00.0 % wd Axial Imbalance h. t> %FP Cp 8 30.6 % vd Max Quadrant Tilt O.CO % Measured Conditions Control Rod Grcup Positions Core Pcwer Level 41.4 gn Gps 1-5 100.0 % wd Boron Ccncentration 1110 ppm Gp 6 74.h % wd Core Burnup 4.4 EFPD Gp 7 1.o % wd Axial Imbalance -3.7 %W Gp 8 19.h % wd Max Quac' rant Tilt +2.03 % H G F E D C B A 1.21' l.60 1.60 1.88 15,h[d8 '.70 '. o' g S

1. f) 1.62 1.63 1.58 1.50 1.65 1.79 1.20 f

l' l.86 1.59 1.74 1.32 1.23 0.99 9 8 15 1.6h 1.59 1.hh 1.h3 1.27 1.01 8 1

1. 1 1.81 1.kl 1.28 0.82 0.60 1.

1.72 1.27 1.26 0 91 0.60 'f. ' 8 1.55 1.08 0 90 6 3 I 3 1.3 1.38 1.06 0.90 N l.' l.16 0.73 12 1.1 1.07 0.65 x 0.'85 13 0.Th s 14 i' 15 l E X.XX Predicted Results X..u Measured Results l Figure 4.4-6 l

'd g COMPARISON OF PREDICTED AND MEASURED RADIAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 75% FP CONDITIONS Predicted Conditions Centrol Rod Group Positions Core Power Level 75.0 %FP Cps 1-5 100.0 % wd Boron concentration NA ppm Gp 6 75.0 % vd Core Burnup 15.2 EFPD Gp 7 00.0 % vd Axial Imbalance -13.0 %FP Gp 8 30.8 % ud Max Quadrant Tilt 0,00 % Measured Conditions Control Rod Group Positions Core Power Level 75.8 %FP Gps 1-5 100.0 % wd Boron Concentration 1037 ppm Cp 6 78.9 % vd Core Burnup 12.1 EFPD Gp 7 03.5 % vd Axial Imbalance -14.0 _ %FP A 1.76 % Cp 8 30.3 % ud Max Quadrant Tilt H G F E D C B A 1.00 1.32 1.30 1.46 1.13' 1.25 1.28 ! 0.90 1.1T 1.25 1.26 1.24 1.18 1.27 1.37 0.92 '.29 1.51 1.24 1.27 0.99 0.92 0.74 9 8 1. 1.30 1.29 1.18 1.13 0.99 0.78 .27 1.34 0.98 0.94 0.6f 0.46 1.. 1.34 1.03 0.99 0.70 0.47 '.0f 1.12 0.81 0.69 11 1. 1.06 0.83 0.70 M.91 0.91 0.58 0.k 0.84 0.50 'Q.68 13 O.s4 s 14 15 I E X XX Predicted Results X.XX Measured Results 1 1 Figure 4.4,

O COMPARISON OF PREDICTED AND MEASURED TOTAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 75% FP CONDITIONS Predicted Conditions Control Rod Group Positions Core Power Level 75.0 %FP Gps 1-5 100.0 % vd Boron Concentration NA ppm Gp 6 75.0 % wd Core Burnup 15.2 EFPD Cp 7 00.0 % wd Axial Imbalance -13.0 %R Gp 8

30. 8

% ud Max Quadrant Tilt 0.00 % Measured Conditions Control Rod Group Positions Core Power Level 75.8 %FP Gps 1-5 100.0 % vd Boron Concentration 1037 ppm Gp 6 78.9 % vd Core Burnup 12.1 EFPD Gp 7 03.5 % wd Axial Imbalance -14.0 %R Cp 8

30. 3 % wd Max Quadrant Tilt

+1. 76_ % H G F E D C B A 1.29' 1.72 1.74 2.04 1.67' 1.76 1.72 1 19 f 8 a --- 1.42, 1.64 1.72 1.70 1.58 1.65 1.82 1.21 ' i.645 2.03 1.75 1.88 1.39 1.24 t).9 8 9 1 1.77 1.71 1.69 1.47 1.29 1.01 i 7 77 1.99 1.53' 1.36 0.84 0.60 10 1. 1.80 1.50 1.34 0 94 0.61 'i.625 1.68 1.163 0.94 11 1 1.48 1.15 0.91 129 1.24 0.78 e 1.Q 1.10 0.65 '06.9 1 0.% i s 14 15 i E X.XX Predicted Results X.XX Measured Results Figure 4.4-3

o 1 COMPARISON OF PREDICTED AND MEASURED RADIAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUILIBRIUM XENON, 100% FP CONDITIONS Predicted Conditions Control Rod Group Positions Core Power Level 100.0 %FP Gps 1-5 100.0 % vd Boron Concentration tons ppm Gp 6 87.5 % wd Core Burnup 33.0 EFPD Cp 7 12.5 % vd Axial Imbalance +1.9 %FP Gp 8 20.8 % wd Max Quadrant Tilt 0.00 % Measured Conditions Control Rod Group Positions Core Power Level 95.6 %FP Cps 1-5 100.0 % wd Boron Concentration 1020 ppm GP 6 90.0 % wd Core Burnup 32.8 EFPD Cp 7 15.0 % wd Axial Imbalance -3.8 %FP Gp 8 0.0 % vd Max Quadrant Tilt +1. 57_ % 2 H G F E D C B A t.03 ', 1.,29 1.24,,1.40 1.14 1.23, 1.28 q.89 g g 1.25 1.27 1.25~ 1.28 1.31 0.89 1.21. 1.23 1 20* 1.41 1.20 1.24 1.01 0.96 0.76 i 1. 1.29 1.27 1.22 1.10 0.96 0.77 i s

• {

1.29 0.98 0.99 0.73 0.50 10 .? A 1.35 1.02 1.01 0.71 0.47 't 7' 1.11 0.85 0.74 8 11 1.1 1.08 v.33 0.71 1 0.90 0.57 2 0.9 0.85 0.50 h65 O.5h s 14 15 I X.XX Predicted Results X.XX Measured Results Figure 4.4-9 i

e g COMPARISON OF PREDICTED AND MEASURED TOTAL PEAKING FACTORS AT STEADY-STATE, 3-D EQUTLIBRIUM XE? ION,100% FP CONDITIONS Predicted Conditions. Control Rod Group Positions Core Power Level 100.0 %FP Gps 1-5 100.0 % wd Boron Concentration 1004 ppm Gp 6 87.5 % vd Coru Burnup 33.0 EDD Gp 7 12.5 % wd Axial Imbalance +1.9 %FF Cp 8 20.8_ % wd Max Quadrant Tilt 0.00% Measured Conditions Control Rod Group Positions Core Power Level 95.6 %R Cps 1-5 100.0 % wd Boron Concentration 1090 ppm 1?.R EFPD Gp 6 90.0 % wd Core Burnup ~ -1_9 %FP Gp 7 15.0 % wd Axial Imbalance Gp 8 0.0 % wd Max Quad-Tilt +1. 5 7 % H G F E D C B A 1.26' 1.,58 1.57,,1.81 1.52' 1.58, 1.59 1,.10 g 8 1.48 1.61 1.69 1.75 1.64 1.73 1.69 1.14 f O' 1.79 1.58 1.67 1.32 1.19 0.93 9 1.5, 1.73 1.74 1.66 1.48 1.25 0.97 9 1.76 1.44 1.31 0.80 0.62 10 1.8 1.92 1.57 1.38 0.58 0.57 'I 49 1.53 1.12 0.94 6 3 8 g 9i 1.5 1.53 'i.19 A 1 22 1.17 0.73 i 1.2 1.12 0.66 x 0184 O.7k j s 14 15 I E X. XX Predicted Results jX.XX Measured Resuits 1 Figure 4.4-10

~ a MEASur D RADI AL CORE POWER DISTRIIiUTIONS ALONG Tile X-Z PIANE 011TAINED DURING CONTROL ROD MANEl'JJERING AT 75 %FP + 2.1 r 2.1 ,,d1.8 no.d1.8 s 1.5 13 1.2 i ri 12 k-UMM MML ..J \\ / A* 8 0.9 O./ .e g 0.6 Rod Group Position - .- 0.6 Rod Croup Position - 0.3 3 0-o,3 l l l l l 0.0 0.0 A B C D E F G 11 K L M N O P R A B C D E F G 11 K L M N O P R Radial Position, Fuel Assephlies Radial Position, Fuel Assemblies Em T C 2.1 2.1 d 1.8 d 1.8 n o. ea. l'. 5 -R x N 1.5 ~ I "* 1 2 s' 12 k,rd - vc' ,r y g g w s a,,- g g ( g 0.Y g 0.9 .-e 0.6 R d Group Position - ) 0.6 Rod Group Position - p. 4 1-5 6 7 8 4 1-5 6 7 8 100 100 24 04 - '" 0.3 100 100 75 24 - "#"0.3 l I I I I I I I I I I 0.0 0.0 A B C D E F G I! K L M N O P R A B C D E F G 11 K L M N O P R Radial Position, Fuel Assemblies Radial Position, Fuel Assess 11es

o, HOT CHAaNEL MINIMUM DNBR VERSUS CORE POWER LEVEL (=,.t= =_.tu_n=..t.:_-E.==.t=.. =_:. u..n....=..ta enta 2Ep:2 :n.t= i=:=d' - t --t=m.e!-= _ ;__ rn; :e t, . g . t L_

---+nx '-- :--

t- - '--ru... u. .. -.. - =. - - - ..... j=n r=$n.

nn: rain.::=;~.:t= Ir *Cr

rf:

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1 4.5 TURBINE / REACTOR TRIP TEST 4.5.1 PURPOSE The purpose of the Turbine / Reactor Trip Test at 40% FP was to measure unit resp 3nse during and af ter a deliberate reactor trip and specifically as follows: (a) Test pressurizer level control, react 1r coolant pressure control,and reactor coolant temperature control during a reactor or turbine trip. (D) Test feedwater control and steam generator level control during a reactor or turbine trip. (c) Test main steam pressure control during a reactor or turbine trip. (d) Test proper transfer of reactor coolant pump power supply from trans-former 3T to CT3 during a turbine trip. 4.5.2 TEST METHOD The Turbine / Reactor Trip Test at 40% FP consisted of a reactor trip with subsequent turbine trip. During unit startup, a number of reactor or turbine trips occur due to primary and/or secondary problems. To document these events adequately, the unit computer is equipped with a post trip review program which monitors data 30 minutes prior to and af ter such occurrences. To obtain additional Jaca during a trip, data logging equipment was run continuously during periods of non-testing. Additional data monitored continuously during power operation as a part of normal unit operation were also used to verify that the response during the trip was within specifications. The method for performing this test (i.e., monitoring continuously sufficle'nt data for analysis of a trip in the event that a trip which will fulfill the test requirements does occur) precluded the necessity of a deliberate trip of the unit. However, if, at 40 percent full power, no adequate reactor trip had occurred, the reactor would have been tripped and the test performed as scheduled. 4.5.3 EVALUATION OF TEST RESULTS On. November 9, 1974, a reactor trip occurred on high flux following a mal-function in the Integrated Control System while the unit was operating at 40% FP. Minimum and maximum values of primary and secondary unit parameters recorded following the trip are tabulated in Table 4.5-1 and compared to applicable acceptance criteria. Analysis of this data indicated that all acceptance criteria were met except for the transfer of the turbine bypass system setpoint to 1025 psia. Subsequent evaluation indicated that the setpoint transferred to approximately 995 psia and controlled there until manual control was taken. 4.5-1

4.

5.4 CONCLUSION

S Upon completion of the reactor trip portion of Turbine / Reactor Trip Test at 40 percent full power, the following conclusions were madp: (a) The reactor trip caused the turbine to trip, ensuring that cooldown rates less than 100 F/hr can be maintained during reactor trips at power as required by Technical Specification 3.1.2.3. (b) All a*ceptance criteria except for the turbine bypass setpoint trans-ferring to 1025 psia were met. An evaluation indicated that the turbine bypass s.2tpoint transferred to approximately 995 psia, which was determined to be acceptable since it is within 30 psia of the acceptance criterion. (c) Unit response to the reactor trip indicated that the Integrated Control System adequately controlled unit parameters during the trip. 4.5-2

SUMMARY

OF MINIMUM AND MAXIMUM DEVIATIO'4S IN UNIT PARAMETERS DURING Tile PERFORf!ANCE OF TURBINE / REACTOR TRIP TEST AT 40% FP l Test Data Test Data Ar.ceptance Criteria Prior to Trip After Trip Minimum Maximum Minimum Maximum Minimum Maximum ,RCS Average Temperature ("F) 578.6 579.5 538.3 573.8 Reactor Coolant Pressure (PSIG) 2128.5 2158.7 1781.1 2210.6 1650 NA Pressurizer Level (Inches) 210.0 218.0 61.0 232.0 40 300 151.8 3%(1) 3%(1) RCS Total Flow (MPPil) 142.5 l 143.2 145.2

LD Storage Tank Level (Inches) 66.8 68.7 11.1 66.4 l

Steam Generator A Temp. ("F) 569.2 570.6 524 ', 573.8 g Steam Generator B Temp. ( F) 570.4 572.3 524.7 570.2 i cr I 895.2 964.5 1025.0 l NA NA y Turbine lleader Pressure (PSIG) 880.9 RCS 1.oop Temperature Misma tch ("F)( 40.4 -0.7 <-10.0 +2.1 [ Steam Generator A Level (Inches) 38.0 42.5 2.0 104.0 l NI Power Level (%FP) 39.7 50.0 0.0 0.0 Note (1): Since reactor coolant flow is a function of :he reactor coolant cold leg temperature, ~ the change in flow must be analyzed using ti e percent of design flow which is independent of cold leg temperature variations. Using this analysis, the change in flow between the data taken prior to the trip and after the trip was determined to be less than one percent. Note (2): Loop temperature mismatch = A Loop - B Loop \\ ~

o 4.6 ROD WORTH AT POWER 4.6.1 PURPOSE The purpose of this test was to measure the differential rod worths at power and to establish integral rod worths at hot power conditions. The acceptance criteria specified for rod worth measurements at power were that the maximum differential reactivity worth of any control rod group be less than 3.03 x 10-4 ak/k/%wd at hot power conditions and that the minimum dif ferential worth of any control rod group be greater than 8.35 x 10-6 ak/k/ %wd except for Group 5 positions below 25 percent withdrawn and Group 7 positions above 75 percent withdrawn. 4.6.2 TEST METHOD The differential rod worths at hot power conditions were measured by the fast insertion / withdrawal method. In this technique, the controlling group was inserted for six seconds and then withdrawn for six seconds. The dif-ferential rod worths were obtained from the measured core reactivity changes and changes in control rod group positions and by using the equation: f = (2 0 2 - pl - p3)/(2H2 - Hi - H3), where: H1 = CRA(s) position (%) prior to insertion, H2 = CRA(s) position (%) after the insertion but prior to withdrawal, H3 = CRA(s) position (%) after withdrawal was terminated, al = reactivity prior to CRA insertion (ak/k), p2 = reactivity after incertion but prior to withdrawal (ak/k), and p3 = reactivity after the withdrawal was terminated (ak/k). This technique was e= ployed for determining the dif ferential rod worths required by the reactivity coefficient measurements at power and fcr determining the rod worths required by the pseudo rod ejection and dropped control rod measurements. The integral rod worths at hot power conditions with the part-length rods at 35%wd were calculated from the predicted rod worths at hot power con-ditions by assuming that tne same percentage deviations between the measured and predicted rod worths at zero power conditions would exist at hot power conditions. Since the Oconee Units 2 and 3 have similar type cores, the integral rod worths for part-length rods at 35 percent and 100 percent withdrawn were assumed to be the same as those obtained f rom Oconee Unit 2 measurements. The integral rod worth curves thus gencceted were verified by performing reactivity balances at power and also ey integrating the measured differential rod worths where applicable. 4.6-1

4.6.3 EVALUATION OF TEST RESULTS The measured differential rod worths at power are given in Ta' ale 4.6-1. The measured maximum and minimum differential rod worths are 0.0179% Ak/k/ Lad and 0.0066% ak/k/%wd, respectively; and therefore, the acceptance criteria are satisfied. The integral rod worth curves for zero power conditions developed from data measured during Oconee Units 2 and 3 zero power physics testing are shown in Figures 4.6-1, 4.6-2, and 4.6-3. The hot power integral worth curves generated from the predicted hot power rod worths by applying the observed percentage deviation between measured and predicted zero power rod worths are shown in Figures 4.6 4, 4.6-5, and 4.6-6. The shapes of these hot power integral rod worth curves have been established based on the results of the reactivity balances performed at various control rod positions and based on the shape of the predicted normalized integral rod worth curves at power. During power escalation testing at 40% FP, differential rod worth measurements were performed in conjunction with the Pseudo Rod Ejection and Dropped Control Rod Tests. The ejected rod worth and the dropped rod worth were determined by integrating the measured differential rod worth curves. These measured rod worths were compared to the values obtained from the integral rod worth curves as follows: Rod Worth Obtained APSR Position APSR Position From Integral for Measured Measured Rod for Calculated Rod Worth Rod Worth (%wd) Worth (% ak/k) Rod Worth (%wd) Curve (% Ak/k) 32.4 0.24 35.0 0.28 21.0 0.09 20.0 0.10 Thus, the measured rod worths at power agree reasonably well with those ob-tained from the integral rod worth curves. It should be noted that the integral rod worth curves have been well established since the reactivity balances performed by using these curves resulted in negligible reactivity anomalies. 4.6.4 CG.:CLUSIONS The measured dif ferent'ai rod worths at power were within the specified minimum and maximum values. The hot power integral rod worth curves developed from predicted rod worth data and Zero Power Physics Test results predicted rod worths adequately. 4.6-2

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4.7 POWER IMBAL\\NCE DETECTOR CORRELATION TEST 4.7.1 PURPOSE The Power Imbalance Detector Correlation Test had two objectives: to deter-mine the relationship between the out-of-core and the in-core imbalances and to verify the adequacy of the Reactor Protection System imbalance trip set-points. The acceptance criteria specified for the Powet Imbalance Detector Correlation Test are listed below: (a) The measured and worst case maximum linear heat rates are less than the LOCA limit of 17.5 kw/ft. (b) The measured and worst case maximum linear heat rates when extrapolated to the boundary of the Technical Specification power-imbalance envelope (shown in Figure 4.7-1) are less than the central fuel melt limit of 19.8 kw/ft. (c) The measured and extrapolated minimum DNBR's are greater than 1.55. (d) the power imbalances measured during the 75% FP test should fall within the acceptance region shown in Figure 4.7-2. 4.7.2 TEST METHOD Imbalance measurements were made to determine the acceptability of the out-of-core detectors' measuring power imbalances and to verify that DNBR and LHR linits would not be exceeded while operating within the flux / delta flux / flow envelope set in the Reactor protection System. These imbalance ...easurements were made at 40% FP and at 75% FP with different gain factors applied to the multiplier on the delta flux amplifier output. In performing the test at a particular power level, the part-length rod position was varied to obtain the desired in-core imbalance, and reactivity compensations were made by control rod Groups 6 and/or 7. For each required in-core imbalance, the out-of-core imbalance, the core power distribution, and other pertinent data were taken. Using these data, plots of in-core offset versas out-of-core offset were obtained. In addition, plots of minimum DNBR and maximum LHR versus in-core offset were maintained to ensure safe core thermal conditions during the test. From Oconee Units 1 and 2 experience, the relationship between in-core offset and out-of-core offset was determined to be a linear equation of the form: OCO = M x ICO + B, where: OCO = Out-of-Core Offset (percent), 1CO = Incore Offset (percent), M = Slope of Relationship. B = Intercept at Zero ICO. 4.7-1

From the plots of in-core of fset versus out-of-core offset shown in Figures 4.7-3 and 4.7-4 for the imbalance scans at 40% FP and 75% FP, the experimental slope could be determined. Since the slope of the correlation is proportional to the gain factor of the delta-flux amplifier, the gain factor necessary to meet the acceptance criteria may be determined from the measured slope by using the equation: GF = M /M, 3 y where: GF = Gain Factor, Mt = Normalized Slope (i.e., slope of the correlation with GF = 1.0), M2 = Desired Slope. In order to verify that the power-imbalance trip setpoint on the Reactor Protection System provides adequate margins for limiting LHR and DNBR, the measured and worst case maximum LHR and minimum DNER were extrapolated to the boundary of the Technical Specification power-imbalance envelope. (The worst case uncertainty factors for minimum DNBR and maximum LHR are dis-c ussed in Sections 4.4. 3.1.2 and 4.4. 3.1.3, respectively. ) 4.7.3 EVALUATION OF TEST RESULTS Figures 4.7-3 and 4.7-3 show the measured out-of-core offset plotted against the in-core of f set for the imbalance measurements at 40% FP and 75% FP, respectively. It can be seen that both sets of data fit straight line with a normaliaed slope of 0.275. Using this measured slope of 0.275. 4 ac-ceptance criteria slope of 0.920 (see Figure 4.7-2), the minimum gain factor is found to be 3.345. Since this gain factor would give only marginal as-surance that the acceptance criteria would be met, it was decided to set the gain factor at 3.45. This gain factor would insure out-of-core of fset readings indicative of in-core readings but would not place unnecessary restrictions on the allowable out-of-core of fset. The minimum DNBR's and maximum LHR's measured during the 40% FP imbalance measurement are shown in Figure 4.7-5 along with the maximum radial peaking factors and the maximum total peaking factors. It can be concluded from this figure that the most favorable core thermal conditions are obtained when the in-core axial offset is -15 percent. The measured minimum DNBR's and maximum LHR's were analyzed for the worst conditions and were evaluated at the boundary of the Technical Specification power-imbalance envelope. The results of this analysis, shown in Table 4.7-1, indicate that the acceptance criteria have been met with adequate margins. 4.

7.4 CONCLUSION

S Upon completion of power imbalance detector correlation testing at 40 and 75% FP, the following conclusions were drawn: (a) The slope of the equation relating the in-core and out-of-core imbalances was independent of the technique used to produce the imbalance. 4.7-2

(b) The observed in-core to out-of-core imbalance correlation on Oconee Unit 3 was a linear relationship with a normalized slope of 0.275. Using this correlation, it was determined that a gain factor of 3.45 was the most desirable gain factor to be set into the delta-flux circuit. (c) The imbalanc.e trip envelope as set in the Reactor Protection System will protect the reactor core from exceeding the minimum DNBR and maximum LHR limits. 4.7-3

MINIMUM DNBR AND MAXIMUM LilR ANALYSIS FOR POWER IMBALANCE DETECTOR CORRELATION TEST Power Incore Assembly Power At Maximum I.IIR. ku/ f t Minimum D BR Level Offset Type Envelope M.3asured Heasured (% FP) (%) Limits Case Case Measured 42.36 -9.93 2 5.28 7.04 8.85 8.17 Extrapolated 101.15 12.60 17.68 3.67 3.20 Fk as ured 41.44 -26.21 2 4.72 6.62 -10.40 9.72 Extrapolated 90.13 10.27 14.39 4.89 4.46 Measured 41.52 + 1.89 2 5.19 7.28 9.04 8.36 Extrapolated 107.00 13.38 18.76 3.46 2.92 Meas u re d - 41.30 - 8.73 2 4.41 6.18 10.51 9.83 . E? Ext rapola ted 102.07 10.90 15.28 4.70 4.10 Meas ured

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==:== = ;= b=r=: _=:.._..==:==Eh:=r i. : h= h=~. =_=

._n=h :-+ =._

. :=...;=._W = ;=l"==;===N t =-

.=~.= JA=I=:

= =i . a==1 l== ===c===u=- Absolute Incore Axial Offset, % Figure 4.7-2

0 INCORE OFFSET VERSUS OUT-OF-CORE OFFSET DURING POWER IMBALANCE SCAN AT 40% FP WITH A GAIN FACTOR OF 3.35 ~ El v :..nta njn un *n".. a-til:tti

=I.,n

.=..w a.. r_ nn =

+::4= ::::

n un I:n

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n. :

nn . =.....

n.. -.

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=: =pr-n; pm - :!=..n =p=

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u..h.... n..

-tr =... - *. ~- ~.. -

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n;}=.:=t:/: - ::n,u:l; n

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m:

an

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E:

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g

r d ii 42 iib =

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2- = iir i[NE!!!jMd 32 !?E

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w s is iEi si iE 2 stas

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4

t-

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: nn =

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t >: ::,n..

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n.. } =. =.. r.. _.. r...

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INCORE OFFSET VER5US OUT-OF-CORE OFFSET DURING 'THE POWER IMBALANCE SCAN AT 75% FP WITH A GAIN FACTOR OF 3.75 .= tn

.t =

n1. .-dr-n ut

tit = nut = :nt=

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t-.r. =;=

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a

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=

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=g-w w - lcu;r===

.= c. ;_ =.un. u.; =. _,a;;g..,=- - =;

,_g. r r 1= w =t =

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g. =.c=:=;. l _. :b=y. - -;n-p;{=- nf_--i-{=_4 2;: t r i.-- ;: ".;..ct: 4_.j =i=j r- ;

.t n ' Incore Offset, % Figure 4.7-5

4.S NSSS HEAT BAIANCE 4.8.1 PURPOSE' The purpose of this test included: (a) To determine the core thermal power by calorimetric methods. (b) To determine the primary Reactor Coolant. System flow by the Heat Balance Me th od.

• The acceptance criteria specified for this test included:

(a) All calculated values of the core thermal power derived from the primary heat balance should agree within 5 percent of each other. (b) All calculated values of the core thermal power derived from the secondary heat balance should agree within 5 percent of each other. (c) The reactor coolant flow for four-pump operation is between 102.3 and 113.4 percent of the design flow. 4.8.2 TEST METHOD Primary and secondary heat balances were performed at all power levels required by the power escalation testing sequence. At each of these test plateaus, steady-state conditions were established as follows: (a) Average RCS Temperature, constant f 1.0 F (b) Feedwater Flow, constant 1 1.5 percent of average value (c) Reactor Power, consti t i 1.0% FP The plant computer was set up to monitor the primary and secondary side thermal-hydraulics data; and af ter steady-state conditions were established, the data were collected. Calculation of the core thermal power was then performed by the plant computer and by a special program on a time-share computer. Hand calculation of the core thermal power was also performed by using the average thermal-hydraulics data. The reactor coolant flow measurements were performed at 75% FP using the precision heat balance method. In addition to the plant computer, dead-weight gauges, calibrated-flow nozzles, and precision' resistance bridges were used for precision measurement of the primary and secondary side thermal-hydraulics parameters. When steady-state conditions were established, the necessary data were collected. The averaged data were-then substituted into the heat balance equation for each Reactor Coolant System loop, and appropriate loop flow rates were obtained. The sum of these loop' flow rates gave the total primary flow rate. 4.8-1

~ 4.8.3 EVALUATION OF TEST RESULTS 4.8.3.1 Primary and Secondary Heat Balance Calculation Table 4.8-1 shows the results of the primary and secondary heat balances from 15 to 100% FP. At 15% FP only primary heat balance values were used to determine the core thermal power. The secondary heat balance is inaccurate at low power levels because of insufficient enthalpy changes across the steam generators. However, from 30 to 100% FP, the secondary heat balance was utili=ed as another means of determining the core thermal power. As shown in Table 4.8-1, all primary and secondary heat balances met their respective acceptance criteria, except for the unit computer calculated pri-mary power at the 65% FP test plateau. During this particular heat balance, the values of the cold leg temperatures input into the heat balance programs on the unit computer were in error, causing the primary power as calculated by these programs to be incorrect. However, the remaining primary heat balances and all the secondary heat balances at this test plateau were correct and within acceptance. After correcting the erroneous inputs, the test was reperformed acceptably at 65% FP. In most cases studied, the values of the core thermal power derived from the secondary heat balances were slightly higher than those derived from the primary heat balances. These differences were attributed to the slightly lower than normal primary flow measured by the unit computer. Therefore, prior to escalation to 100% FP the primary flow constant on the plant computer was adjusted such that the indicated Reactor Coolant System flow was con-sistent with the flow measured by precision heat balance. 4.8.3.2 Reactor Coolant Flow Determination a I The Reactor Coolant System primary flow measured during the Reactor Coolant ] Flow and Flow Coastdown Test (Section 3.1) was approximately 108 percent of the design flow. Following the initial power escalation to 15% FP, there j was a slight increase in the primary flow indicatea by the unit computer. I Since the measurement of primary flow at low power levels by the heat balance cethod would be inaccurate due to insufficient enthalpy rise across the core, precision measurements of the primary flow were carried out at 75% FP. 1 Table 4.8-2 shows.the values of the Reactor Coolant System primary flow for four-pump operation determined by the precision heat balance method. The average value of the two measurements is 112.44 percent of the design flow and is within the specified maximum and minimum limits'. The maximum error in the measured primary flow, taking into account instrument tole,rances and other uncertainties, is estimated to be fp.79 percent of the design flow. a Therefore, the Reactor Coolant System primary flow rate is established as 112.44 i 0.79 percent of design flow. 4 1 4.8-2

t .4.

8.4 CONCLUSION

S i All primary and secondary heat balance calculations met their respective acceptance criteria, except for the unit computer calculated primary power-at 65,% FP. The discrepancy in this primary heat balance calculation was due to erroneous inputs to the unit computer's heat balance program. After correcting the erronerous inputs, the test was reperformed acceptably at 65% FP. The primary Reactor Coolant System flow was measured to be 112.4 + 0.79 percent of the design flow. 1 a 4 i 4 4 1.. i ] + 0 d j t 4.8-3 1 i l

b

SUMMARY

OF llEAT BALANCE PERF0PJ1ED DURING POWER ESCALATION TESTING Test Co re A T Primary lleat Balance, % FP Secondary Heat Balance,% FP Accept. Plateau Date Time Power Unit Time-share Unit. Time-Share Met lland Cal. PHB/SIIB (% FP) (% FP) Computer Computer lland Cal. Computer en mn r i n.r 15 10-10-74 0905 14.89 16.71 16.88 16.70 (1) (1) (1) Yes/NA 1 30 10-14-74 0735 28.93 29.60 29.86 29.89 30.72 30.73 30.50 Yes/Yes 40. 10-22-74 2245 40.74 40.83 40.02 39.80 42.54 42.53 42.41 Yes/Yes 65 11-15-74 0650 63.91(2; 65.16(2; 60.68 60.48 65.67 63.35 63.18 No/Yes 75 11-17-74 1115 75.10 73.19 72.22 71.87 75.54

74. 39 74.54 Yes/Yes Y

90 12-15-74 2345 89.71 88.67 87.42 87.22 90.58 89.68 89.73 Yes/Yes 100 12-18-74 1000 98.61 97.62 95.80 95.64 96.22 95.98 96.02 Yes/Yes 7 Note (1): At power levels less than 30% full power, the secondary heat balance canr.ot be considered valid due to insufficient feedwater floa through the steam generator o Note (2): The Computer (CTPA) and the core AT power programs use a combination'of the wide range and narrow range RTD readings for the cold leg temperature. Dur!ng this particular. heat balance, the values of the wide range RTD temperature readings were in error. l l l l I I i 'l l 1 J 4

Es RESULTS OF REACTOR C00lANT SYSTDI PRDLARY FLOW MEASURDIENTS 6 Power Calculated Primary Flow,1 x 10 lb/hr Percent of Flcw Desi n Flow Imbalance Date Time Level (y pp) loop ( A) Loop (B) Total Design (yf (7) 11-18-74 1545-1645 70.10 72.38 73.51 145.89 129.73 112.46 0.77 11-18-74 1715-1815 70.61 72.53 73.42 145.95 129.73 112.42 0.61 H X r T - ra Note: An analysis of the error in the flow measurement was performed based on the tolerances of the leistruments used to measure the thermal hydraulic data and the uncertainties in the flow coefficients and the ambient heat loss measurements. The errors and the un-certainties used are given below: Flow Constant In The Railey Flow Nozzles +/- 0.5Z Ambient lleat losses +/- 50.Z Temperature +/- 0.2 Degree F Feedwater Pressure +/- 1.0 Psi Main' Steam Pressure' +/- 2.0 Psi Reactor Coolant Pressure +/- 25. Psi Flow Nozzle Pressure Drop +/- 0.1 inch of itG Thl's analysis has shown that the total error and uncertainty in the above r.easurement is +/- 0.79 percent of the design flow. m

4.9 UNIT LOAD STEADY-STATE TEST 4.9.1 PURPOSE The purpose ef the Unit Load Steady-State Test was to measure Reactor Coolant System and Steam Generator steady-state parameters as a function of power and to compare them with design predicticz.s and equipment and system limits. Specific purposes are as follows: (a) Determination as to 5 hether an adjustment is required to steam generator level to conform with average temperature versus power requirements during power escalation to 15% FP. (b) Measurements of Reactor Coolant System and steam generator parameters with four reactor coolant pumps operating during escalation f rom 15 to 100% FP. (c) Measurements of primary and secondary system operating parameters for comparison with future performance data. Three acceptance criteria specified for Unit Load Steady-State Test are listed below: (a) At all times, the steam generators, Feedwater System, and Reactor Coolant System exhibit stable operation with no significant oscillatory or unusual behavior detected. (b) All recorded steady-state parameters, as a function of power level, are within their respective minimum and maximum limits as given in Figures 4.9-1 through 4.9-7. (c) Steady conditions are maintained as follows: RC Average Temperature Steady within + 2 F of a mean value RC Inlet AT(c) Steady within i 1 F Turbine Header Pressure Steady within i 9 psi Neutron Error Within i 1% FP of a mean value Feedwater Flow Within i 1% of a mean value Pressurizer Level Steady within i 6" H O of a mean value 2 RC Pressure Steady at 2155 50 psig 4.9.2 TEST METHOD The power levels used during the performance of Power Escalation Testing were 15, 30, 40, 65, 75, 90, and 100 "FP. At each power level, steady-state conditions were established, and data were recorded. ~ From the data the average unit parameters were calculated and comparison between the measured data and the design curves were made in order to verify the acceptance criteria (see Figures 4.9-1 through 4.9-7). Unit parameter stability during steady-state operation was measured in terms of the deviation from the average value of the parameter during the test. 4.9-1

In addition to these data, during the initial power escalation from 0 to 15% FP, the steam generator level limit was set to maintain reactor coolant average temperature versus power level, as shown in Figure 4.9-8. 4.9.3 EVALUATION OF TEST RESULTS Table 4.9-1 shows the unit average parameters of the primary and secondary systems measured over the test period for the various test power levels. . Comparison between the measured data and the expected design data was then performed by plotting the averaged unit parameters against the expected design curves. As can be seen in Figures 4.9-1 through 4.9-7, all unit average parameters measured fell within their respective minimum / maximum boundaries. However, steam generator A outlet temperature was observed to deviate from the expected design value at power levels above 75 percent full power. Upon investigation, it was found that this steam generator and some of its as-sociated piping had not been properly flushed. Preliminary data taken after the feedwater header ring was cleaned indicated better agree =ent. During the test period, unit stability was measured by determining the deviation of unit parameters from their respective average value. Table 4.9-2 is a listing of the maximum deviation of these variables during the test at various power levels. As can be seen, all values measured were well within acceptance, except for feedwater flow. From this analysis it was concluded that flows, temperatures, and pressures were stable within the following average maximun limits: Temperatures 1 1 F of an average value Coolant Pressure i 15 psig of an average value Steam Pressure 15 psig of an average value RCS Flow 1% of an average value Feedwater Flow i 3.6% of an average value The results of the 0 to 15 %FP test as shown in Figure 4.9-8 indicate that an acceptable steam generator level setpoint has been obtained. 4.

9.4 CONCLUSION

S The average of the measured unit parameters during the test period fell within their respective minimum and maximum limits, although steam generator A steam temperature was lower than predicted. Analysis of unit parameter stability indicates that all variables are relatively stable, even though feedwater flow did not meet the stated acceptance criteria. However, neither steam generator temperature nor feedwater flow stability has any adverse effect on the safe operation of the unit. i 4.9-2

AVERACE UNIT PARMIETERS AT VARIOUS TEST PLATFAUS FOR OCONEE UNIT 3 DURING STEADY-STATE CONDITIONS Average Unit Parameter Unit Parameter Units 15% FP 30% FP 40% FP 65% FP 75% FP 90% FP 100% FP RC Cold Leg Al NR Temperature F 574.82 571.97 $69.06 563.85 562.03 559.47 357.43 RC Cold Leg A2 NR Tenperature F 573.93 571.21 569.07 564.92 562.33 559.18 557.30 RC Cold Leg B1 NR Temperature F 574.63 571.20 569.10 564.78 562.19 559.00 557.09 RC Cold Leg B2 NR Temperature F 574.50 571.01 568.77 (1) 562.40 559.13 557.13 RC llot lag A NR Temperature F 582.63 585.58 587.68 592.96 595.21 599.03 600.85 RC llo t Leg B NR Temperature F 582.49 585.21 587.78 (1) 595.65 599.28 601.87 OF 578.52 578.38 578.37 578.83 578.84 579.17 579.10 RC Average Tempe rature RC Loop Temperature Mismatdi F 0.19 0.39 0.13

0. 39 0.12 0.26 0.25 RC Loop A Coolant Flow FTPli 70.36 70.48 70.64 70.90 70.9 8 71.15 71.21 RC Loop B Coolant Flow FTPII 71.58 71.78 71.90 72.41 72.55 72.83 72.90 p

RC Pressurizer Level (UC) In ches 151.48 154.00 153.04 152.47 152.46 151.44 151.60 RC Loop A NR Pressure Psig 2164.08 2151.18 2162.57 2145.58 2146.52 2147.41 2165.64 RC Loop B NR Pressure Psig 2142.55 2130.84 2142.41 2154.74 2155.56 2153.07 2173.09 Neutron Error % FP 0.08 0.29 0.68 0.42 0.48 0.30 0.88 I FDW Total Flow PTPil 1.33 2.90 4.07 6.62 7.80 9.65 10.41 s i' FDW Final Temperature F 235.59 311.04 347.67 413.28 427.89 445.37 452.34 O MS Cencrator A S/fl Imvel Inches 23.93 37.06 51.03 78.38 97.63 121.95 136.54 PG Generator B S/U Level Inches 26.00 39.06 52.60 79.34 99.15 125.25 141.21 FG Gene rator A OP Level 6.82 10.53 14.22 25.95 33.36 44.74 51.91 MS Cenerator B OP Level 7.29 10.93 14.90 25.73 31.73 45.84 53.28 MS Cenerator A Pressure Psig 891.13 889.06 894.42 895.08 901.51 904.73 901.14 t$ Cenerator B Pressure Psig 887.00 883.69 887.36 886.45 892.98 894.02 893.95 MS Cenerator A Temperature F 578.30 586.91 587.24 589.65 586.93 578.28 578.31 16 Cenerator B Temperature 7 5 77.72 583.91 587.20 592.16 59 3.84 595.31 590.11 Turbine IIcader Pressure Psig 888.69 883.74 883.45 879.14 885.68 885.56 881 15 Hote (1): During the performance of thir test, these computer points were unavailable for recording.

.~. s o ~ MAXIMUM DEVI ATION IN AVERACE UNIT PARAMETERS AT VARIOUS TEST PLATEAUS FOR OCONEE UNIT 3 DURING STEADY-STATE CONDITIONS Maximum Deviation of Average Unit Parameter thii t Parame t e r Units 15% FP 30% FP 40% FP 65% FP 75% FP 90% FP 100% FP RC Cold lag Al NR Temperature F 0.71 0.88 1.13 0.49 0.44 0.52 0.65 RC Cold Leg A2 NR Temperature F 0.59

0. 80 1.10 0.44 0.44 0.52 0.62 HC Cold Iag B1 NR Temperature "F

0.66 0.68 0.97 0.47 0.38 0.49 0.52 RC Cold Leg B2 NR Temperature "F O.68 0.67 0.78 (1) 0.65 0.71 0.56 RC Hot Leg A NR Tempe rature F 0.69 0.66 0.64 0.65 0.24 0.48 0.46 RC Hot leg B NR Tempe rature F 0.68 0.62 0.85 (1) 0.24 0.43 0.42 RC Average Temperature F 0.66 0.61 0.71 0.22 0.30 0.51 0.49 RC Loop Temperature. Mismatdt F 0.21 0.60 0.29 0.40 0.16 0.05 0.06 RC Loop A Coolant Fl ow 1.20 0.88 0.86 0.78 0.60 1.01 0.90 RC Loop B Coolant Flow 0.88 0.88 0.59 1.05 10.86 0.72 0.78 i RC Pressurizer Level (UC) In ches 1.56 1.68 1.04 0.16 0.48 0.72 0.56 RC Loop A NR Pressure Psig 22.52 16.63 25,62 3.61 8.52 11.99 10.64 E' RC Loop B NH Pressure Psig 22.46

16. 34 25.36 3.67 8.56 11.93 10. 49 Neutron Error

% FP 0.21 0.46 0.69 0.42 .0.48 0.19 0.22 y FDW Total Flow 6.23 4.31 6.81 2.17 3.01 2.34 2.44 9 4 FDW Final Temperature F 8.09 1.79 3.18 0.09 0.13 0.16 0.44 MS Cenerator A S/U Level Inches 0.45 0.27 0.88 3.12 2.83 0.50 0.96 MS Cenerator B S/U Level In ches 0.15

0. 34 0.20 3.06 2.65 0.50 1.04 MS Cencrator A OP Icvel 0.20 0.37 0.60 0.25 0.33 0.22 0.79 MS Cenerator B OP Icvel 0.19 0.27 0.24 0.54 0.21 0.50 0.44 F6 Generator A Pressure Psig 5.37 2.44 3.42 8.58 5.24 1.77 2.89 FG Generator B Pressure Psig 5.25 2.37 3.04 9.45 5.77 1.99 3.21 F6 Generator A Tenpe rature F

0.18 0.45

0. 39 (2) 0.36 1.36 1.14 MS Cenerator B Temperature F

0.38 0.52 0.53 0.15 0.17 0.40 0.66 Turbine llender Pressure Psig 0.34 0.51 1.60 3.71 4.58 0. 89 2.55 Note (1):.During the performance of this test, these computer points were unavailable for recording. Note (2): The unit computer did not print out the minimum and maximum deviation of this unit parameter. I I I I I I I j

REACTOR COOL \\NT SYSTEM TEMPERATURE VERSUS POWER LEVEL WITH FOUR REACTOR COOLANT PUMPS OPERATING ( ~

y

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p.Ei!

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1. E=NMv =Es=i =fiE= =EtEE sa='

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1. =...i= :ETw

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= ~s r 4 - gju; -i===h==t=i Ej== ia [-: -p=- gjug = g-i: gjp . :..,i. =. L.=..,:_;..m.. g.: g.n_._ ug...=.g._ _ ggg..,g 2.

~;-y~{ _ a.g.

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p j[--

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h. j r.

E ~~iFi i:p. uj T(COLD) --- A. 5; ; . ng.: qip

s

,. 4 2 _. =+4 _q = 2;paz pp. Fiijs;jc i 2. gn- = g. -- q - i:- - !+ pj. sg= uigi p}pj--p--g q_ +.-

4:-

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lc

=c !Fi-t- F=i uiiilqil :H{J f n_ j=f: i. i:[. ip p gp ]. 3.. ;.;e..+.. ...c 2 w- = ay. ._,.u,._ m: q 7.._. .,...n..t. .,..q g m + . 3. '.:r.: 2.:. r ~ 1;:n 2n -m[..:2!.- _:I 1.ja. r -t:;; 2.p: :::c.d.1r.... l. ~Ic.r. u - r.. _2 ) :..:.2 t =4.a:Jz : - i, - 3 .:r Power Level, % Full Power 1 Figure 4.9-1

STEAM CENERATOR OUTLET PRESSURE VERSUS POWER LEVEL .ii - f[:i

t:

' {;- "l -l. pl: .- [ - l_ - i

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.j.. l... l l p v' p: -!u ' Measured OTSG Outlet Pressures 3 i: "iE:

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-i -

0TSG (A)---- O I p~I.

+ ! :

t-

Fi i~ii

- OTSG (B) --- El I" ii.: i.M ..q:u .ij:'.. iip; ni!ME: =T rp al. 4: tl : X - ~ p' .:i ! .;- t.i .:t ' m- =in _ (1:; t

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Power Level, % Full Power Figure.4.9 o e STEAM CENERATOR STARTUP LEVEL VERSUS POWER LEVEL 4 1

n. =f",:.. =}=.;;. = tn. _

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a. t.=-- -.:t--

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n..in.-- : =... :. =-.1 a!.- ,.: = -:: .r.. :.n..,- ~4.-

n. j. =..

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• ==iE Ji!!i 9-f lii!!

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4:ijm x_i -.i.iB : alii:

ri-!E ni = o... E..::" T'r,~l:'...

., t ::...

T_ Expected

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ir:-

nim ij/MHl:t li5 a 1G0.i--

l._ l=_=___i: 'Eis : =i -? =ii=

_.. _/..9 E- + !..!= =!_ _ E i = "=li=Fi: r:Fu-

==in- ... = l;j EFE g gn:E MER ;+;ii== 'n;u = p =- citi-

ip+ E= -4. : i i' -i:f4i::p:

4-FijH 3 =.ir: l-- rhj

fA.

Ein ir:i - ~pll: u F"- "ti:- -dir-art;:. ..;.:. Minimum 1 p:_ . ;u.. tir.. ..i.. . a... a;; fa - - -a.- 4 e rsu .t- .r.n-u w . r --

r z.

g,:... a.. tj PE-- "i:E: 9" - ' gi- -g ._:p: Hji:-lMpf tih t= - P: L'j=- pj== .-ph 5 "i-i"_: :-E=p;phiJ "@ ej-uri=%: " :H - i+i +

h:--

u j g ': ci-i" _nh D i@ji

j.:

'..:j.: .;u a: .:iji..:jpe' j.. .={ c.

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qn

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b.

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n

il

{:

uJn lli ' s- _;a ,.u....

ic

=:=

.n .: p.: EM 'E +P Zr '" ' 1 Measured OTSG Startup Ranges :.1 " 7

=;:

. ::=. ,,. p :.; 40TSG (A) --- @ 'FhE '..-f"i: =.:.". :' 9. ti.4c.-i!F =. :. "Hi.ci OTSG (B) --- O ~ i. i. }i. "i t t M

w

.t.<.

x;n.

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=1. _

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x K. z y. =n r. I ' c r". l. :- ' ' .T'*- 21: 2L' ' m

1:['?

• - -E nn, di^fE-Mitin ~iiM._

7c:; u. n;jur ci i ~j

hi lfi T k tESH = $ 'i.;.

"T .. JiF 7E: i 552' i i .!i!- lihl ' 3_ c _; _ !!i~: .:: t.= ';;. PVn:;.: unt:Wr r --M ..ju W) ..i:. W.: t. : .. ::.: ply V :. t n.r .;i ( . t.: g.,

t.

/_l= l=$.c.an tL;.:1 =lu

. :.J.;: :: -

.:;. );::;a Power Level, % Full Power i 1 Figure 4.9-3

STEAM GENERATOR OPERATING RANGE VERSUS POWER LEVEL n tar W=i:= =-- t= =i:n: : nutnr =;= =n=

r=t=;=txt

= anter =:==

+

t

5 . r a n:- nr: =;- =tunhx.=trtr" mtr* marr n-- =diEj ij$ =ilin liiipi - Idiir Q:;== x.-hin EM = i-i:!iit . _ - "5l#E =q= ; !sL=.f A9E ihic. - yjg - hjg:- -Eniii digMEME E"9 - M".jEi-. ni:L E!Mj@j= - iEE =... "..... =........f.=............-t.=. .. ~... =.... ~ i. a.... =. =.-_ =.....::==...r, =... +r.. : ...y... =.. :_=.

. t u..

=. + =.. n =.. =....:.=. .... =. -... a r E.i'. 9..! =.. E. _. l._iiE..i.ii.=i.j E_. E_.n;=.=.. _n.=is...i.="..ii.s_.:_i!"l=E_ Eii._E 2.T._ E..' l -.;g- _ t Q _.... - --. g 7

t.

_g, r x =.====.= {gn ;g , n. ; _, ;, gn ;

g _- -

.g- _ =1c. =u- _ qm== ;n;-,- t - -nim-

-t =

E.= cur =t ..I -rtu ec - 5 =w:7.=

Measured OTSG Operating Ranges p=ra..:_..

u;=.aq:- = j =-- i lent rt= Maximum = t=. =;_.: OTSG (A) --- O =37:- --gh=- t-- += t=_ L-=t-- ec =t._. = ....=..OTSG (B) ---. c

=-..=... =..

.t.u.r.. *t.=. =..T=...=.. 3t r... s

u. u:n,..........=..

m#=.3=.... + = =...:_= t...- ..f.-_..._._..!._.L.... U ..v... _ +.,.. . _...=: =. -.. }.... !=..=.:=u=_=t.=.=. =.n=s~.:::..._.n.._ :=.=. =..=.... . =... r m.=. w= :.:_:_= =__:.=_=.= m.,- t m

u..w g... ;

.~t_-- ..t.,_,,, c. r_.. .g._._, ._j.... _ _... _-_- g- - ....g

c..;._g.._

.t.._ ._t... EEEE EiEEEi"E =-liif M E= i5iM E[#f E55 EINiE!"l5i= ii#i"i;Mi"EE H _ _ _. _ _ _.r+=..=. m; =_. =_n._..._._:=. .=t=._..=..=.t.=. -..... _.. ..... _.=..t=.,<._=.r._--.. _.. ,=__=;... ~,... _ .=_} =.. p._::=.. r. n. =_ =_._m=._ r g,._..._ =...*.;.=... =.$.=.. =..;.=. _.i._. =. n=.. y....... f x.. ;,..s..._ _g..._.........l.....[=......_h__..._ g .g m.. 2_n_ j.=... ............ _.z.n.;nr' ....l.=_. t - _.. t _ _- E!E EiElEfM'liEEEli =Ei=d=i!E 'E lEl'=. I"..g~l"M i5 u:T =isC;r EIM E o :-.e= - = ..g n.L...... q:.....}1_=..- _...i=... =

=m f=...

-4.. .... j r, =.......=t=n; a =. r nr z.._. -J- . T.:...r:: Expected '. ::: :-"r.. T rj =.:.:::.. - :d. ::... #1_f Z..!=.......1. l.=i.::.;.*EZ. . ~ : 1. E tr =. ;=.. =.. p==. t. _._, : - :g =_p. n =_.+t=.. -..+t=.... t _. n.t... g-.-...... -.. -. _ - p Nh[I hiiN ENN'i U. b r - El5=!b!b Eb!b Ch"khEiiEif5: I!! b NfUII!b D... t..+

.E. E..=~ *t_==..f. E..

==.. i.ne.. M_ =i."=i.== =. F=. =.=_.i=..=+. 141=

==i=_A =. 1.__+_ V. =.. i.=..-3. p...m E i L. .t u { "siMI -Eil5i "Mi54Il5 $ME "EEii "5ti"l; "EE Elf if :8:i Wi" fi~55 AE

1. Mc Ei:=

=iE# t EMis,El== iBTs 512 -Ms-Eii-EiEi!E=i E=i= = c =fu =1:n = nt= =p=- nai:.-

:

anpu: =:r

: 2.. =:

un+ =: --2 r._. n.re!nc - --in

• lu r e. :

--~ n - ".=-- =.r m: t --- = r. u-r-" n: -"-t--- c =g=:. =rn .. =t u.. =p

n...

2_..:..

n-

=3-- t=-*m "' p=

• nin=z

n.a

ar_. _n .r J= ."O= J

t

t

NJr =ir .4=I= g =:=. .. F. :. E. E. !.=..=... ; =...=. t. = =. =.. p. F

6.. 4._ ["=ix:- Eii.iEjE. jE._E!=JE. i=_ E- =. j =-

cy i - b b '41nimum

=. m =.=rr

=..L==.t=..-- =...:. n -=..t.=..._=. t....b'...L._._t... _ t. _r = =.. "t.._.. _ -. . h _r =y n

g

.=. I r._ =;= 22 =t 2:=

_.n =

1.p=. t.: t= 4 =:.u.:n.r:.=;. : _ .n .= t a -"1. pra rtrx ..t=- nr-

. u--

r ::= t. : ---

n2 =

Efijf E!lE Elis-Ep=1=i =Ej==1EEiE}=:fr. 2"=

i:E : =tE Ei=

~~igii MiiEElr$.E F5i5 EM@i5Es isi@iME#l55kud5(iii""M5 Elds:n155

==t==}==.u= r===:e=:===E=t=;:=l=u:=:w=: L 2=i===w.r: -m w=2. = a r==+ 2= u=h.p=r==l.c.

x nn;

mt=

;

:===

= " = an Power Level, % Full Power Figure 4.9-4

STEAM GENERATOR STEAM TEMPERATURE VERSUS POWER LEVEL 1:- r.n. l_n.-.:i.'.' H...I~ilN~.. n.. =..t. t= ... p.

.*=._ ' m : r..... n t :.

u p._. .-.:.. ; : {1. : .j p.. 3._. .-3 n .;g n n..

r. 1!+ -6:ip -

n RF

+jE-. 2 p; p[-E: =gh; =;llh-..j Ei =i

== 2 :("I Rid 5. ti.' Mi. d si!di: -if5 Elli' =dI5= lil!Ei! iCIEE =E[EE 5 m;t q.=3+7g~ Maximum .m. m... :: : n.. t r..=. :. r..- n t. :.:

r..a t=...=.3._.

3,. . ]. 7- .i.. u q .+ _,p .: g g .2..:....:. =...c. :

r. b.. r._r! =.

r. L,=. =...:=..

=... t.:n..

n.

• r... !...

.. ! s.. -,. 2 =. r. =_:.r- ..t .n.t. --.n.t =..- 1 ygg'j: E2ij 7 ='jr i@: "!=" E:EE rijjfr i:!f =iy.ggjii i=iiiiii iEfgf . c..

!. =_. -- x. t...

c.r= =t., .a:v.= p.e.t...3....... =. _ :=..:.:. =... p ...g

y..

iiisix. =jji~a r=t . 2:9. i:!, _ i ;Q.jj-l;iu ublj,6 ';j:' !;lij hhigi;Eiir n=jE! I ' Oi=h:5 Expected f5!!5E "$Ei.

Iri '

EFi ="! - ! iil!= iMi -!E' : Ext Ei=i - px E-@u. afr 2.FE- "~i..Q';f.pfljj, ai}4Elijn =};;rF!ijiii; ;=.:ijEi; g e,....._--.E. i.dEEiiPr ... t s' T ' ' '.. ;. ". ".. j "..= @ =_. = iE.i.did..i.:.;.:. -ii. i. j.=. i. i.EE.jE.l_i 7 o Hi=E :"li = frig ='}EF y.' .gi;i.,. Eppy.t at:; -

:.pt.;.....r..jEr iEEE cI --5ii~i _EFr fi= fI '

a O*MZi~ ~li5iC

Fa?%IliiMEfi5i!E= =lii...i_..

. _EE t -F.... w__ .c... ki..,l-FE. hC-!5~5hZ:C 20dT--.CiEC ?!'E=0 .-_,-...f_=.... =t.. [ hC[ 3 1FF fF ff '3?E --t--- 2 ShI: =liE-EE dpl-.==HE = B= ~EIMn E'i!i i- =liEki -@MEdfi EigliliE i i E:9)= 5Ehu. =;: f f..y .dj=2j}gchi=p=iiE par-dE: fili =ijE@ 4 hijF -Eji= a , =., =.. - =...:=",. .p. 3.- n n...t =. =...;.n - ;- u.n- .. r._.t :- -

n. t.n. x..l =. _.: =. 33 =.... t'rx

=..t'=_... e : r. _rr_. n g e iE ~= e SL - ---iE' iEiEi'i /G..E: =ii!;ii E"id2 =455 ihEf: =hFi" ri"-jiE"!!!EfMiEd5EE . tr.... . = +n_. =.r.=..=_ =...r. s-:t.=... =...=. .... r.... ....,2..._.:n.. t _=. ,e _=_ =_ =_. j _=_. :.

r..r. p p g =..

_ =. =.. u.r u =:.:.:. =..... _u. I m T j- . =1;J' . ='.! _.8 3.' ~- C ;.-

:;t ;.:;.:m t n; _ -- ; ;.

Ja t;=.rJ ).{. g. uur er :-

-1=

nr

2., -.

n = yu ca n. = Minimum

== ,r =r oNN !bfi ~EN h55NIN!1 =5'i5bfiii5fbhib!!U5fi55 EEIS ih 5 hNE EfiU { = =FiF =ji llhg..i :"#jii = die 27 jig : n;.- !;ip .ppi- -ih E-[liij E =as - kia +iIlV::-lii+jhr Tui: iiH* =!9 - i:Hin:=r Eii " in* n=thi5 a city s =g/ p c gce =ig; 4 g9;;je;iE g qi:; m

gig

....._ =......... ..l. .... j, j[ = g...

c._

_=.c -_ q;.. : h-... -..... g _.... en ..rn. m : 1 ,:t - - c.g M 2::= or-j n r o==I=-

h. :!-~-- U.jf q!,t E... i =..=.H. i.i.=.=il.iE.. t =.:d Ec'E.i.Ei.js..-E=- :==. i..i:---a. i

.E E. ' E:t==.. "g u =i== '=FF 5l[fc i-!ib 2 I~iFl !=iE:' ic! : tiii !Ti= - i ci ilri liini:in =.,: n- =_.p,r-a. ff s';u-=:. 2_2 .ap:.

n.. i.a.u.. -'t un

.r r... .c. r i/[x lSi li!:r =S != !:Fi: : Et s!E

da EEn usii!:

i '. EEi= .=N = i k:-- =. _. ; =fi: - =:}E E:jE;.i !Ex :=iEr EEih='i:E,LE(E". Fa!i: }jjg =! git

u.n _ u.. n

.. =.=. =. - {=.r--..

3. :2

..,.......t.=... =1=..,.. .=::: =..;.=_ =...t=...~.=..;...=_ ::=.. _ xtu. ra u =:nz. -mz.--.-.._ ,a=. W. : q=:l" = = .r-Measured OTSG Steam Temperatures =t=" r r/.y...i E J. 2.;

n:;n- - - - OTSG (A) --- O

" d. u -- 2.; ; =:= ; 2 n = ii@i: 2#iiEf i:= vi"~ "EiE-OTSG (B)---- O c5j!5l Mj}irj =i3! ;g u-iih-ii=@@igjEll@? $ NM7 kjr Qj[TI _ $i' ]l[fj g;;.3 d --. i = ri=n iyiu

};j: q:p 7-

?pu Qi ry x-. f:px-nint- -- s.-p H. rt... 3..f 2 i ra p. J..V q:.. x ~ r - r-- -;= r

L.

..t..,,,;1;i.; ..,n -- 4.m :in": w :*-

t.

,,,,.., a :

t ,u =.juntax-,2nl= l -

.t = lMn,:.2

.".jun.;. u } = ;=[._: r.. ..dn utn sn ? tc ni=:t;W ut a :.W:a.. riw n .:d =.. ) =s = ; ir- . 24n =. ; nn Power Level, % Full Power Figure 4.9-5

e s i FEEDWATER TEMPERATURE VERSUS POWER LEVEL i l i L....n

.1

~ l:"'" n " M n'.t:. n ! --

t

-in* f-!:

• =f-*: 'n.-:1x;* : ::-!nn' =:'i 02 :={: ;'~"

n;*in :.3-. t:

- t ::

• t"::1.u:J::-

- tn" - nt. 2 3 :* n -- %n :- id. i.

ij.i.

Ji.!

C:1,;i.

. i.i.i.:- : n. d.E. - 3:!nx ti. l.,ii. i: u. n.. " -H.E..ijE..n. H.iii.i.i...u.E..ji.E... '3: b!f _fi I:f iN bf ~ ki f~ ~52i'ii- _!ilil i!Mf;hi ?!N n!?i- ~ d-E P!2Mi Ei3f hlN i'i!!!'iflil5 l !EN5 i!!i! N!!m.iENf-d i ) 15Ei Iilii i:L" iIi Ulf "!!i "d h! !!Ii ~ "Mi I i . l !!4 =5fii - i' 5:=i -i@ilil!Ei5 ~iii!E 55fis "iiEi IMit -5ilin Hb!' i:Ilin un!:hi i l i44= =E!!!E Maximum Y si= 4!# 4!!k =!!i= !ils =usi 5.ti 4iii=i==is - = Jt:jn-- j@:E Li2}ij tihi- 'p;;; ..: gn -j@i" f!!!F :ni c; EijjpEpiEjjt GjiWiEExpected j !~i!!?! i!l5 SIN . idii ild -i!!ii'lia!IiL =iiEif $f#N..=. T.Eik 5!i5 j gi-Np rjiFp - ;]~ q{j :=@ihiQkjMg.n;nh%:iEjQ@pfEj Num ~ 3 [: .. ;j];{E"h fjiiii UE {_5 , ;: q ' t h. !.7!25 - I; l i '--, li 12[e "hdji 1 3 gi=".:...g== C j= p? !=: ryp@""nju $ igg---i n. pp, p ti =F - -jM;j-jp_ _.' " ;IC'. z. g..... - -i4E" CNUih 22323- _.. _! ii'~l ~~~~'t"2 4.a gr;.

1.. ! ~

g }.:.

.. t - : - b -+. id9 -:i:? - ":ijn:,[t i-- - i.ii.". - :227...ui n i. " d;?;;; m:{.=.; mij..i -i. :. { E..E.. c t. .t. y

tc " tie MF 'ii-fii= "b"i +eiE :-iiiii- :idM4EtE=

I :r

ii,

$ -pjE.

'.}l!

..((. illi ; j.,j".;[j;j.!!'i: ~.g}i!" 7E;.En- -_gg~.j gij[j I -Ejig;' -ijj;j 7 nu;;;st..J:;.n. :nC.:n:4;. ~ r~ r f.

.... : /' '-

- * ~ - . : ;.t un l.a :s_. 1:.  :...r:n. F " '

1'-

.: t :: .:t M '=.l=1 g ..t. / . *, pa' j .t [~ ~ I' _4 N '!' ~ f 'I? Z: j.J .:j-. j.q

hp..r,jihi.Cphi

!}d. .dij i-fU: qih[,..~iififiU p[i:

lj'.

i.hj iff

• 'jif!]

I

j:;

'[.

N ki.

{-*

nh! n*:{pr .2-

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.{i-4 2 --

-' n g:n..n. -t- -U: n,-

r

,::P ' - -- n. - *p-. u '

n. a c..- '

- n t. u. -t. r.:.

n'}" :

-' pn ~.- ....:f; -.-... -.::t.:: y u 1,9 g N i- !!Ii li!M 'lih!

$l5 axi - ni'fhsiirI'5i '5Ii:---

42:i iiiij n RSEE 'riN'55:' i' d@i i--jij -dii=.idpii : iiFi -injui --p". :.q(( jq =ifj{j

I i !--

irjii

."i!i..... Measured Feedwater'E!Eu %p- '?;r.- ..{..

tr 2p...

4. -

t

{.

E"

...jii - Temperature --- @ %!E2 g@g...7 .[.

ijgii

@jg

- !;i

j-L.....?jjij.

K

((i zj7' j:q, r ;;g.... ..,:=,an ;; :.g pr = t n. r

q :ip:q

- + -

p up g

_y :. 4 q.; i.j - .y.: it; c..[ qi. 4.; h.1-: +1 mI :.. - 9 . 3m + , + 6.n.,. iT.. i.u a" ..g:.%w d.i.+ ,e.. i, .:i. i. :. j i. . f. . t :: } w..t... .t

. w,.. ].

.... : : 3. wvp... g.. c. :.2-0

". j., p., ..=:g:::' n j ~::. g:::ap ...r i. .i ..:. { C. - - ". :}:.. . l. :.:: gn..r.:g.::, q.: Power Level, % Full Power i i i Figure 4.9-6

TOTAL FEEDWATER FLOW VERSUS POWER LEVEL r. .i - . a.. } _. 3.. x}. .c.. 4 .g.. g

i :

..n. i.x n n..n u. n.m.. tn tr . at:.- q . :p.- 1=[i i j!- E F-t T: ii =ij: - iifii iiliE iij?"; ~ 5$x- ~ !n "ir.

f' alii

k

• ~

il-i" ' id4 iir'i

. - -j yy "-t-f

- j-i 3:

(ll ..p;

jf p-j-((j idi:

- F- = F. _.. :

t. :.

.t.: r = : !. =. Hin

.f r.

p. i-

n. jr..- in.,. -ii

j..f.r.j

.-=2:u ...;il:. t yin giua

i+

.;i.; HM: J Mi. .];. pl.;

$n!!

7 =::_.

n.. t u.. u..

n.. ln.

u. t,d. mln. Maximum e nr:-

n.. -. n.g:...

s g ' ~4. -4 nin it! upii-diw ur.. di!E= "=t:ing;uj;jn y.:i !? Expected + y

Hi

'i-i 7;; '. i;jg ,t i ijij;.; -- ;p= : :; iia ngj=~- jj-id,a:ig. Hp $ a:i 7

in

- ;r i:i:: Minimum c q'

dE 4: :

Eli: fij-- gn.

19f? f' i.f.W:=[i

C e F lii: I!;; "iL Ef! Mili fyill?,.$ ~d f 3 p::j.in..p;=.pi-inn p_-.: _!id_.. ;;jis.. iEling:; :.(n.;r:g.jif.{yj _ imp 3 O :. ;.. -

. p. :

- E!. : --. p:

.J r..

i..j. !.

.I. iiin..i.:.5:. i, n.

b..,M. ; E. i[ i..,i /

.j t i. d.. f.il.iiiiin:. 3 g :7g. --- ii' il 2Fi: niij jlg 2:pnr j:. d ~ ii.:*iiji'ij:jii[:[th .-h}llf: jdfEIi 1 -E-hithh!!~!:t" ME~ ^ di: _ U h n-~:!!ir '-!!![d I : {i~- i!iis i=lli d d i O rite -N;l E: tNfik:i4" Yi / :,I N E5 i I- ~ !4E i"St -O=f i! li5 7!il5:i d ' t: I =:= uran.l u: n,==.1=:,.. nci.. +f n!...,. 2 = / - 50 i =}.. ...f=.. :=i.n n .. t. 2-:.2 .. n .c .=:n .x.. nut g u= n. C#1 ? 93E.;f 214:.-}jmd"-- Ja-ig:j niijr'.Q7:K-} / ~(gl-ijijhi 24jn. 4, _ ur 2.;c:=. ;t: : t. ......, / Hi7. -.; / i e. . I.....

=n.

nn.. - n =:- ann.. m. : =:- "2

t :

-r

ni;

" l=.i t -- fE - iSi:L n;?- 1-iff '/i,1 i j:- up-

4;

-lM u

f li: =! i;' lif j.c T r!#-..4/[je $ - "j:

jii

t

!@i .g ; gf iiFi

~i.l. t

ll:Mpp

+-ihj[g};. .. t:: 7;} ; jf;

l)ll l'@=

i? ; --fif"fij=;in - f'[e[;-l%=-j --

-}t: p j-" i- -- j;n

f:

=" , 5-!3 N5}fr[5 @!!N5!E-Y/i~ 5:~iid I"h--d2E :@ir' i:Id -i 5 Ei== F#iT=i@ E 1E= F = M V =isitE-ii "la=&:1 uiu 1J O %=J 9 E-M =i = H #i= T M i:# L r - --,-i :!:!;i ?E! t ?in = .r. r._.. r.a.._n. . _.- 3. j,e .... t ; _.rn c t . : ~;.-... ...t..- .p:

t..:

r. E... j ~~ -U:9 B. r ~~ l. i. ii. "- t-M r

=tn-2 22-M t;..,. E:(..

t _t

,. w.,...

' "" ' i~ ...r--l :: At.h=t-r:

;.;f EhEl=HE Measured Feedwater i=i-8 9.Bgj pii=/:;

=i" M""f = : ")h-i --i)EY5:'!EM Flow --- O E!ME O -f. %. t. ; l...:. i.d..--.. ii.i...N.!. i_!.!.bi-Fidiid!';d.. M :! - - _.. D....

• P b_- L 4.. i. i. ~ '. ~- ~ t. d. H..r..._'.

Ci... 7. t. i. b. - _ + n.. f l L .t. Fi _F =(.. niiiif Ej ii !#E lfi. :--Mi'i%2ijiiT.n yniif.1.': ' 'iht - 1:iii

l;'i

.u. =t.nr ' m..:1.... _=.t...,.,. :I..Lnn... n.1". : -. :i:: =_.:.r._ n..': a.. ::- n_n. n. -..a_t.:m- -.c:* r-....: =!. :.. 3,,r1..=. ...m ....n

d..-

.' : :.d'.r_.

n rP

a:-

ur" " I~. - nnt-- - }n* na:.: .t. ..n.:jgi.p. ..j u y:: :t - - - d"u n .q:.}. :' 1 ab u. ..t i n. t. -c Power Level, % Full Power Figure 4.9-7 y,e. r-

e e

=*= nut =1 = :::: nnt = nn+=fu:*:-

}.n= ~=rf=r-:=:d

n}.nu + +

=...*. t +-+- t:ti-t

........=T=.nt =t+."..: tr u t=.. t = rn nr rnt n-

t

= .a._. ...g*..

• ..g..

T.=. g-~..... - n!. _*;. ~~.. { ~.~ n.=...1.:.*.*. n. t.!*_*. :.*_*;. t. n_::* _+- 1.**n. .g._ !!=...:n..: $. n.*- ..n. g n.._.. n*t. = -.. "_n_..~sg~*** . nt= . t r-r.. ;; = =;= r.=. ..njn.. nnt= nrin:

f= =*nn

..... = ::

n i==

n:-f =t-n =trc .:; 64 = t n--==n-

1a u..

f

== t== t.= nnt.

=t=

.: W @ nntn - +nn ctu ' ngn. =l =.~~ -~: ! =- =* ~f r-- :--d: n. + + - - -~

n'i n -

t :

• • t : -

=t:r n = r- ! ~: rt n .: o

---

t n n M =.=

~t r t n

4=2.r:

.: M I ..::4nn..t tuu ..%. :... n. ;.= nnn. =j==.... .u. 3 =;;. ::.:.j= ....t. d :n 1 O I n..j un nnn= ..._n T =.. =..i.g nn..i=. m..t. =. nt n- !Tn -nt=

t:

.y.~- -:.+.-.-- nn. = .M ....t =l.. ..s =.... _., = r. t=. :.g 0.=..*. =tn.n. 0 .....:t. =_ .. _ {=. y =.. _ t. : : n.n. _- nnrr =t: o m. r t=. =.. _. =. _r

n. :;= x.;n.\\: ;p.c.: g utnn =p;n j= t_. n:

lll g Q:;- n t: O.O -~ - . - ;.1: -t " r. = t].....= t = Winn::-.:c;nr-nnn= .: 0 u n. 4 .: n.= U3 cc =*:=:n= h dNC. 5 5555N O $ 5 i $5N5M55Mi;i G @ts IS W n;..... g....... . g... j =, nn ut _. ..m \\.. __[,n g un tn-- ~nt: 0 O C '*IC 22 C -~ 2 7 I> g nnt=

tn

?gn n'N:~:.. ning =. ::.=. _pung:=n.=g u H n.u.=j.n. nn+~ = =:=;< 0 nut=

=p -- ..n r.. c3 =

. = nnfr-++-t =.: b =r

- 4. n g ..= r = +n ;; T U H :n:{ tun 2= W n.;1=

nt..g uni =n:.1;\\qI"A ry t=.=:::

nie r'-. y :=::t= o t > t =fE ; O

-t

'U ~3 3

• n:t= =t=

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s i 4.10 UNIT LOAD TRANSIENT TEST [ 4.10.1 PURPOSE t The purposes of the Unit Load Transient Test are listed below: (a) To observe and record certain RCS and secondary system parameters during i transient conditions imposed on the unit. 1 (b) To demonstrate the ability of the Integrated Control System to maintain control of the unit during transients at different transient rates and in different modes of control. (c) To aid in the tuning of the Integrated Control System. 4.10.2 TEST METHOD During the power escalation testing on Oconee Unit 3, the Unit Load Transient Test was performed at three different power plateaus: 40, 75, and 100% FP. Table 4.10-1 gives a general summary of all transients required by the Unit i Load Transient Test. The transients were performed with the integrated Control System in three separate modes of. control: (a) The fully integrated control mode. (b) The turbine-following control mode. (c) The reactor / steam generator - following control mode. Transients in each mode of control were performed at ramp rates between 1 and i 5% FP per minute to assist in verification of the acceptance criteria. At the 75 percent full power plateau, one additional transient was performed to check reactor runback due to a manual trip of a feedwater pump while the Integrated Control System was in the fully integrated mode of control. i R The desired transient response of the unit required that reactor power could be varied at the design ramp rate of 5% FP per minute at 40 and 75% FP and of 10% FP per minute at 100% FP and return to steady-state operation without exceeding the unit parameters deviation given below: TRANSIENT LIMITS Turbine Header Pressure +/-50 psig. Reactor Coolant Average Temperature +/-3 deg F l Loop Temperature Mismatch +/-5 deg F-STEADY-STATE LIMITS Turbine Header Pressure +/-9 psig Reactor: Coolant Average Temperature +/-l deg F Loop Temperature Mismatch +/-2 deg F j Reactor Power Error. -+/-l'%FP 1 Throughout this test, Integrated Control System tuning-was performed when the. above unit parameters indicated that tuning was necessary.in order to optimize ~ the transient response of the Integrated Control System. 1The technique of inducing each transient was'to decrease reactor power from the test plateau to a predetermined lower ' power level; then af ter_ establishing i 4.10 -.

steady-state conditions, reactor power was increased to the test plateau. During all transients, the variation of pertinent primary and secondary system parameters during negative and positive power ramps were monitored and recorded. 4.10.3 EVALUATION OF TEST RESULTS 4.10.3.1 Integrated control System Transient Test at 40% FP The Unit Load Transient Test was conducted at 40% FP to evaluate the ability of the Integrated Control System to accomplish a smooth negative and positive change in power level at a ramp rate of 5% FP per minute while operating in the fully integrated, turbine following, and reactor / steam generator following l modes of control. The data taken during each transient (i.e., power, unit average temperature, loop temperature mismatch, and turbine header pressure) were then analyzed. The behavior of the unit during the negative and positive ramps in power level is presented in Figures 4.10-1 through 4.10-4 and summarized in Table 4.10-2. All acceptance criteria were met during this transient. 4.10.3.2 Integrated Control System Transient Test at 75% FP The Unit Load Transient Test was conducted again at 75% FP to evaluate the ability of the Integrated Control System to accomplish a smooth negative and positive change in power level at a ramp rate of 5% FP per minute while j operating in the fully integrated, turbine-following, and reactor / steam generator following modes of control. l The data taken during each transient (i.e., power, unit average temperature, loop temperature mismatch, ar.d turbine header pressure) were then analyzed. The behavior of the unit during the negative and positive ramps in power level is presented in Figures 4.10-5 through 4.10-9 and summarized in Table 4.10-3. All acceptance criteria were met during this transient. 4.

10.4 CONCLUSION

0 i From analysis of all test data taken, the following conclusions may be made from the 40 and 75% FP sections of the Unit Load Transient Test. (a) All transients were performed without exceedir.g the limits of the Unit 3 Technical Specifications. (b) All transients were completed without causing the Reactor Protective Systen to actuate. (c) The ability of the Integrated Control System to control unit parameters (i.e., power, unit average temperature, loop temperature mismatch, and turbine header pressure) during the transient was excellent. 1 -4.10-2

a GENERAL SUteMRY OF TRANSIF,!TrS FEQUIRED BY UNIT LOAD TAX;SIENT TEST Power Power Test Accept Transient ICS Mode of Decrease Increase Ramp Rate Rate Group Operation (%FP) (%FP) (%/ min) (%/ min) A. Unit Load

• Transient Test at h0% Full Pcwer 1

Fully Integrated 40 to 30 30 to 40 1 to 5 5 4 2 Turbine-Following 40 to 30 30 to 40 1 to 5 5 3 Reactor / Steam Generator-40 to 30 30 to 40 1 to 5 5 Following 4 Fully Integrated 40 to 20 20 to 40 1 to 5 5 B. Unit Load Transient Test at 75% Full Pcwer 5 Fully Integrated 75 to 65 65 to 75 1 to 5 5 6 Turbine-Following 75 to 65 65 to 75 1 to 5 5 7 Reactor / Steam Generator-75 to 65 65 to 75 1 to 5 5 Following 8 Fully Integrated 75 to 55 55 to 75 1t' 5 5 9 Fully Integrated 75 to 55 Table 4.10-1

TRANSIENT DATA.nETATMED DI! RING Tile PERFORMANCE OF . UNIT 1.0AD TRANSIENT TEST AT 40 PERCENT FUI.I. POWER Transient ICS Mode of Power Rate,%/ min AT(AVE), F AT(C), F ATilP, PSI Accept Met Number Ope ration (%FP) Ave Max Min Max Min Max Min Max (Yes/No) 1.1 Fully Integrated 40 to 32 -4.0 -6.8 -1.2 40.1 -0.4 +0.1 00 +27 Yes 32 to 41 +3.6 +4.8 -0.8 +1.0 -0.5 40.4 -17 +05 Yes 2.1 Turbine Follwing 41 to 31 -2.1 -3.1 -1.3 ^O.4 -0.4 40.3 -10 +05 Yes 31 to 41 +2.6 +4.5 -0.8 +0.6 -0.3 40.2 -07 406 Yes 2.2 Turbine Following 40 to 30 -4.3 -4.9 -1.2 40.2 -0.5 40.2 -06 406 Yes 31 to 41 +5.2 + 7. 0 -0.8 40.6 -0.8 40.2 -03 +11 Yes a 3.1 Re ac tor /S t e a m 40 to 28 -4.1 -7.0 -1.2 40.4 -0.4 40.1 -13 44 8 Yes O Cenerator 30 to 40 +4.1 +5.5 -0.8 +1.2 -0.4 40.6 -43 +14 Yes 7 Following 3.2 Reactor / Steam 41 to 31 -3.9 -5.2 - 1. 2 0.0 -0.5 40.1 -04 +23 Yes Generator 31 to 41 +5.2 +5.8 -0.8 +1.0 -0.4 +0.1 -28 +07 Yes O Following 4.1 Fully Integrated 40 to 20 -6.2 -7.9 -1.4 -0.1 -0.8 0.0 -04 +18

Yes, 22 to 42

+5.6 +7.4 -0.7 40.8 -0.5 +0.5 -23 -04 Yes a \\ . -. =

TRANSIENT DATA OBTAINED DURING Tile PERFOlufANCE OF UNIT LOAD TRANSIENT TEST AT 75 PERCENT FULL POWER Rate. %/nin A T(Ave),UF AT(C). F AT11P, PSI Accept Met Transient ICS Mode of Powe r Nunber Ope ration (%FP) Ave Max Min Max Min Max Min Mn (Yes/No) 5.1 Fully Integrated 74 to 63 -4.2 -5.7 -0. 8 40.2 -1.0 0.0 -11 +21' Yes 65 to 75 +4.1 +5.0 -0.5 +0. 5 -0. 4 0.0 -15 +17 Yes 6.1 Turbine Followint 74 to 65 -3.9 -5.4 -1.0 +0.2 -0. 5 -0.1 -13 +16 Yes 65 to 73 +4.3 +6.6 -0.9 +0.8 -0.6 +0.1 -10 +10 Yes 7.1 Reac to r/ Steam 75 to 62 -4.7 -6.3 -1.1 +0.2 -0. 6 0.0 -23 +40 Yes Generator 64 to 73 F3.1 +4.4 -0.7 +0.1 -0.6 to.1 -34 +11 Yes D Following 8.1 Fully Integrated 73 to 55 -5.1 -5.7 -1.2 -0.2 -0.6 0.0 -12 429 Yes 58 to 76 F4.5 +5.6 -0.6 +0.8 -0.8 FO. 2 -18 +17 Yes [~ 9.1 Fully Integrated 75 to 61 -20.3 -27.0 -1.5 to.4 -1.4 +0. 6 -17 +26 Yes 6 (FUP Trip)

UNIT LOAD TRANSIENT TEST AT 40% FP Maximum Power Increase Rate:+4.8 %/ Minute Transient Numbe r: 1.1 Maximum Power Decrease Rate:-6.8 %/ Minute ICS Mode: Fully Integrated

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t.= f== [.=_,_ w.

=.;.=.

• : f..ntu :=[=j=:ra m.=

;={==:=

un;n r = =r} rn. Ti:::e, Seconds Figure 4.10-9

'f i 4.11 PSEUDO CONTROL ROD EJECTION TEST 4 4.11.1 PURPOSE t The purpose of this test was to verify the safety analysis relating to the accidental ejection of a control rod which is normally inserted in the core during full power operation. In the Pseudo Control Rod Ejection Test this was accomplished by determining the worth of.the most reactive control rod which is normally inserted in the core during full power operation and by i analyzing the core power distribution when the most reactive control rod is i fully withdrawn from the core. ] 4.11.2 TEST METHOD I The control rod at core location H-08 (Figure 4.11-1) was selected for the i Pseudo Control Rod Ejection Test, and the test was conducted at 40% FP. The reactivity worth of this control rod was determined by measuring the core reactivity change when the single control rod was gradually withdrawn from f 0% wd position to 100% ud position. During this control rod withdrawal, Control Rod Group 6 was inserted as necessary to maintain he reactor power level. The core reactivity change due to Control Rod Group 6 movement was i determined by measuring the differential worth of Control Rod Group 6 by' the withdrawal / insertion method for every 10 percent change in the control rod I position and by integrating the differential worth between the initial and final Control Rod Group 6 positions. Other contributions to the core reactivity change, due to changes in xenon worth, boron concentration, temperature, and power level, were also taken into account to determine the ejected rod worth. j In order to study the effect of the withdrawn control rod on the core thermal ] parameters, core power distribution and thermal-hydraulics data were taken corresponding to the control rod positions of 0% wd and 100% wd. Since the control rod is located at the center of the core, the core power distributions j were analyzed using 1/8 core symmetry. 4.11.3 EVALUATION OF TEST RESULTS The analysis of the core reactivity changes during the control rod withdrawal is shown in Table 4.11-1. The results of this analysis show that the pseudo ejected control rod worth is 0.4% Ak/k, which is less than the Technical i Specification limit of 0.5% ak/k. Table 4.11-3 and 4.11-4 show the core power distribution for the contro5 1 rod positions of 0% wd and 100% wd, respectively; and a comparison of the corresponding radial power distribution is presented in Figure 4.11-2. The core thermal parameters measured during the pseudo control rod ejection are. tabulated in Table 4.11-2 along with the power peaking parameters. A maximum linear heat rate of 8.32 kw/ft, a minimum DNBR of 5.45 and a maximum power peak of 3.18 were measured with the control rod at 100% wd. These values were measured at the core location H-08,Lthe fuel assembly containing the ejected rod. The effect on quadrant tilt was small since the control rod was located at the center of the core. The maximum quadrant tilt increased from 1.88% to 2.27% following the pseudo rod ejection. 'The. increase i 4.11-1'

in the axial offset (mor*e negative) was the result of Control Rod Group 6 movement from 75.0% wd to 51.8% ud. 4.

11.4 CONCLUSION

S The measured worth of the most reactive control rod was found to be 0.40% ak/k, which is less than the maximum value of 0.50% ak/k specified in Section 3.5.2 of the Technical Specifications. As expected, the pseudo rod ejection produced a large perturbation to the steady-state core power distribution. The resulting maximum linear heat f ate (kw/f t), minimum DNBR, and maximum power peak measured were 8.32, 5.54, and 3.18, respectively, i 4.11-2

DETERMINATION OF PSEUDO EJECTED CONTROL ROD WORTil Rod Cont rol Rod Position,%wil Power Boron Coolan t Xenon Pull Average T(AVE) Worth Nu:nbe r Position dp/dh Position Level Conc. (% Wd) 1-5 6 7 8 x yp PPM Deg F % AK/K % ud % AK/K/%wd 00 100.0 75.0 00.0 32.4 40.52 1090 578.4 -2.0844 10 100.0 74.2 00.0 32.4 1 75.2 .0131 20 100.0 72.6 00.0 32.4 2 73.6 .0112 30 100.0 68.8 00.0 32.4 3 69.8 .0093 40 100.0 .6 00.0 32.4 4 66.7 .0089 50 100.0 6,.8 00.0 32.4 5 64.7 .0091 E 60 100.0 59.8 00.0 32.4 6 61.0 .0088 [ 70 100.0 58.6 00.0 32.4 7 59.6 .0090 7 c-- 80 100.0 55.8 00.0 32.4 8 56.7 .0091 90 100.0 55.3 00.0 32.4 9 56.3 .0111 100 100.0 51.8 00.0 32.4 43.71 1104 577.8 -2.0794 10 52.6 .0128 Differential Reactivity Differential Re. activity Heactivity Balance Unit Parameters Balance Co.*fficients contribeition of (me) = -0.005 Zak/k -0.0050 Zak/k A H = +23.2 % vd df/dH = +0.0102 Z Ak/k/Zwd +0.2366 Z Ak/k A A P = -3.19 % FP a PD = -0.0094 % Ak/k/ZFP +0.0300 Zak/k A T = +0.60 F af - -0.0033 % Ak/k /F -0.0020 % Ak/k "a = -0.0102 i Ak/k/PPMB +0.142M Za k/k A s -14.0 PPMs total = +0.4024 % Ak/k n a a I I I f

SUlflARY OF CORE POWER DISTRIBUTIONS AND TIIERE\\L-IlYDRAULICS DATA TAKEN DURING PSEUDO ROD EJECTION TEST Rod Incore Quadrant Tilt (%) Pcue r Maximum Minimum AssemblyAssembly Maximum Peak (Dim) Position Of' set LilR DNBR LocattorExit Temi-(I wd) (%) WX XY WZ ZY hkN) (kw/ft) (dee F) Radial Total 00 -3.28 -1.36 -1.25 +1.88 +0.37 40.52 4.51

10. 11 B-08 609.18 1.40 1.78 100

-28.05 -1.08 -1.27 +2.27 +0.09 43.71 8.32 5.54 11 - 0 8 632.43 2.36 3.18 Y O .L e i m-

MEASURED CORE POWER DIS *IRIBUTION RESULTS WITil PSEUDO EJECTED CONTROL R0D 0 7. WITifDRAWN FROM Tile CORE AT 40% FP Control Rod Group Positions Gps 1-5 100.0 % wd GP 7 00.0 % vd Gp 6 75.0 % wd GP 8 32.4 % ud Core Power Level 40.5 % FP Boron Concentration 1090 PPM Core Burnup 7.6 EFPD Axial Imbalance -1. 3 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth -2.05 % !.k/k Max Quadrant Tilt +1.88 % 1/ 8 Co re Incore Weigh ting Psax/ P/P Fuel Pco$ Assembly Fuel Assy. De te cto r Factor -tnc n H en Nu-her Lo c a H-0 8 1 1 1.46 1.14 G-08 2 4 1.65 1.28 F-08 4 4 1.65 1.26 E-0 8 10 4 1.59 1.24 D-0 8 14 4 1.52 1.07 C-08 21 4 1.65 1.27 B-0 8 30 4 1.78 1.40 ~ A-08 37 4 1.23 0.95 G-09 3 4 1.63 1.28 F-10 12 4 1.69 1.35 E-11 26 4 1.25 0.97 D-12 41 4 1.13 0.92 C-13 g 52 4 0.73 0.59 F-09 )~ 1.66 1.31 6 0 E-09 5 8 1.59 1.28 D-09 15 8 1.48 1.17 C-09 29 8 1.42 1.13 B-09 31 8 1.26 1.02 A-09 45 8 1.02 0.81 E-10 17 8 1.65 1.34 D-10 27 8

1. 39 1.04 C-10 28 8

1.20 0.98 B-10 44 8

0. 89 0.72 A-10 46 8

0.61 0.48 D-11 33 8 1.29 1.04 C-ll 42 8 1.02 0.83 B-11 49 8

0. 89 0.71 C-12 48 8

1,04 o,g4 B-12 31 6 0.63 0.50 Table 4.11-3

i ~ j MEASURED CORE PCWER DISTRIBUTION RESULTS WITH PSEUDO EJECTED CONTROL ROD 100% WITHDRAWN FROM THE CORE AT 40% FP Control Rod Group Positions i Gps 1-5 _100.0 % wd GP 7 00.0 % vd 4 Gp 6 51.8_% ud GP 8 32.4_% ud Core Power Level _43.7 % FP Boron Concentration 1104 PPM Core Burnup _7.6 EFPD Axial Imbalance -12. 3 % FP Xenon Conditions Yes Yes or No Equilibrium Conc. Reactivity Worth -2.05 % Ak/k i Max Quadrant Tilt +2.27 % 1/ 8 Core Incore Weigh ting Pmax/ P/P Fuel Fuel Assy. De te cto r Factor 7corg Assembly i ncnH en Nurbe r oca-H-0 8 1 1 3.18 2.36 G-08 2 4 2.63 1.90 F-08 4 4 2.18 1.51 I E-0 8 10 4 1.97 1.30 D-O 8 14 4 1.71 1.03 C-08 21 4 1.73 1.16 B-0 8 30 4 1.84 1.27 A-08 37 4 1.21 0.86 G-09 3 4 2.35 1.71 F-10 12 4 2.04 1.46 E-11 26 4 1.53 0.98 l D-12 41 4 1.32 0.83 C-13 52 4

0. 75 0.53 F-09 6

0 2.18 1.52 E-09 5 8 1.92 1.33 D-09 15 8 1.83 1.12 C-09 29 8 1.52 1.04 B-09 31 8

1. 31 0.92 A-09 45 8

1.02 0.73 E-10 17 8 2.01 1.33 D-10 27 8 1.69 1.00 C-10 28 8 1.39 0.91 B-10 44 8 0.96 0.66 A-10 46 8 n_o 0.44 D-11 33 8 1.55 0.95 C-11 42 8 1.19 0.76 B-11 49 8 0.92 0.64 i C-12 48 8 1.12 0.76 B-12 31 6 0.65 0.46 i. Table 4.11-4 l

EJECTED ROD WORTH AND LOCATION AT 40% FP PLATEAU X A B C D E F G H W-y K L M N O P l R I I I I I Z l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Core Control Rod Worth, %ak/k Position Predicted Measured H-08 0.49 0.40 Figure 4.11-1

MEASURED RADIAL CORE POWER DISTRIBUTIONS FOR CONTROL ROD POSITIONS OF 0% WD AND 100% WD Measured Conditions With 7-1 At 00 % Withdrawn _ Control Rod Group Positions Core Power Level 40.5 %IT Gps 1-5 100.0 % wd Boron Concentration 1090 ppm Gp 6 75.0 % wd Core Burnup 7.6 _ EFPD Cp 7 00.0__% wd Axial Imbalance _-1.3 %FP Cp 8 32.4_ % wd Max Quadrant Tilt +1. 88 % Measured Conditions With 7-1 At 100 % Withdrawn Core Power Level 43.7 %FP Control Rod Group Positions Cps 1-5 100.0 % wd Boron Concentration 1104 ppm Gp 6 51.8 % wd Core Burnup 7.6 EFPD Cp 7 00.0_% wd Axial Imbalance -12. 3 %FP Gp 8 32.4 % vd Max Quadrant Tilt +2.27 % H G F E D C B A 1.14 1.28 1.26 1.24 1.0/ 1.27 1.40 0.95 g 4 g 2.36, 1.90 1.51 1.30 1.03 1.16 1.27 0.86 28 1.31 1.28 1.17 1:13 1.02 0.81 9 1. 1.52 1.33 1.12 1.04 0.92 0.73 35 1.34 1.08 0.98 0.72 0.48 10 1. 1.33 1.00 0.91 0.66 0.44 9/ 1.04 0.85 0.71 11 0. 0.95 0.76 0.64 M.92 0.84 0.50 i 12 0. 0.76 0.46 g M.59 13 0.K x 14 15 b X.XX Rod 7-1 At 0 % Withdrawn X.XX Rod 7-1 At 100 % Withdrawn Figure 4.11-2

4.12 DROPPED CONTROL ROD TEST 4.12.1 PURPOSE The purpose of this test was to verify the safety analysis of an accidentally dropped control rod, normally withdrawn from the core during full power operation. The primary test objective was then to determine the reactivity worth of the dropped control rod, and other objectives included the following: (a) To measure the core power distribution with an asymmetric control rod at 50% and 0% withdrawn. (b) To demonstrate that the position indicator alarm and asymmetric rod alarm indicated an asymmetric control rod. (c) To demonstrate that control rod withdrawal is inhibited beyond a pre-determined power. (d) To demonstrate that reactor power is automatically reduced to a pre-determined power when a control rod is asymmetric (nine inches from group average position). 4.12.2 TEST METHOD The dropped control rod worth measurement was performed at 40% FP by moving the designated control rod (see Figure 4.12-1) from 77% wd to 0% vd and by compensating the resulting reactivity change with Control Rod Group 6 to maintain the reactor power level. During this exchange, dif ferential rod worth measurements were performed by the fast withdrawal / insertion method for every 10 percent insertion of the control rod, and the data were plotted as shown in Figure 4.12-2. The integral worth of the control rod was then obtained by integration of this dif ferential rod worth curve. The worth of the dropped control rod was also obtained by converting the resulting change in the controlling control rod group position. In order to study the effect of the dropped control rod on the core power distribution and core thermal conditions, core power distribution data were obtained for the control rod positions of 77% wd, 50% wd, and 0% ud. The radial and axial core power distributions were analyzed using full core analysis since 1/8 core symmetry did not exist when the off-center control red was inserted into the core. Objectives (b), (c), and (d) were acccmplished at 75% FP by moving the designated control rod out of the group average position by seven inches and verifying that the position indicator alarm and asymmetric rod alarm indi-cated an asymmetric control rod. The control rod was then moved out of the group average position by nine inches, and the automatic reactor runback to a power level less than 60% FP was verified. Actuation of the control rod withdrawal inhibit mode was then verified during this asymmetric rod condition. 4.12-1

4.12.3 EVALUATION OF TEST RESULTS The results of the dropped control rod worth measurement are shown in Figure 4.12-2. Integration of the differential rod worth data results in a dropped control rod worth of 0.09% ak/k, which compares favorably with the value of 0.10% Ak/k obtained from the. change in the controlling control rod group position. The measured axial offset, quadrant power tilt, minimum DNBR, and maximum linear heat rate are tabulated in Table 4.12-1 and are displayed in Figure 4.12-3. As was expected, the quadrant power tilt increased and the axial offset decreased as the control rod was gradually inserted into the core. A maximum positive power tilt of 13.97% in the ZW quadrant and a maximum negative power tilt of 13.03% in the XY quadrant were measured when the control rod was fully inserted into the core. Analysis of the core thermal conditions was performed taking into account the worst case uncertainty factors for minimum DNBR and maximum linear heat rate. These results are tabulated in Table 4.12-2. Extrapolation of the worst case minimum DNBR and maximum LHR to 102% FP yielded a worst case minimum DNER margin of 20% and a worst case maximum LHR margin of 25% when the control rod was at 0% wd from the core. The results of the core power distribution analysis for the control rod positions of 77% wd, 50% wd, and 0% wd are tabulated in Tables 4.12-3 through 4.12-5. Examination of the radial peaking factors, obtained from this analysis and shown in Figure 4.12-4, indicate that a severe flux perturbation has occurred as the control rod is fully inserted into the core. The asymmetric rod runback portion of this test was performed at 75% FP. When the control rod was inserted'seven inches farther from the group average position, the position indicator alarm and the asymmetric rod alarm both actuated, thus indicating an asy= metric control rod. When the control rod was further insetted into the core until it became asymmetric with its average group position (nine inches from the group average position), the reactor automatically raa back f rom an initial power level of 74.7% FP to a power icvel less than 60% FP in 28 seconds. A test to verify control rod withdrawal inhibition was then successfully conducted. 4.

12.4 CONCLUSION

S Upon analysis of the Dropped Control Rod Test data, the following conclusions were made: (a) Analyzed core power distribution and thermal-hydraulic data indicated suf ficient margin to minimum DNBR and maximum linear heat rate limiting criteria. The perturbation to the steady-state power distribution was as expected with a maximum tilt of 13.97 percent. (b) The measured worth of the control rod which produces the most adverse ' thermal effects in the core, if it is inadvertently dropped, was found to be 0.09% ak/k. 4.12-2 I

(c) The Integrated Control System accurately detected the asyr: metric control rod, activated appropriate alams, and initiated designated control actions. t 1 1 I l 1 ) I 4.12-3 e m c-r - +a-

SUMMARY

OF CORE POWER DISTRIBUTION AND TiiERMAL liYDRAULICS DATA TAKEN DURIllG Tile DROPPED CONTROL ROD TEST Quadrant Tilt (%) Power Maximum Minimum Assembly Assembly Maximum Peak Rod g MR DNBR Location Exit Tem ' Position (t) WK XY Wz zy (%FP) (Kw/ft) (deg F) Radial To tal 77 -3.73 - 0.90 - 1.20 + 1.88 + 0.22 41.57 4.77 E-06 608.27 1.37 1. 89 9.79 B-08 609.18 1.38 1.87 50 -1.31 & 4.17 - 6.20 & 6.94 - 4.91 47.92 5.70 8.19 K-05 612.30 1.45 1.96 00(l) -1.20 +11.01 -13.03 &l3.97 -11.95 42.00 5.86 (2) 11 - 0 2 1.75 2.28 U t 5 e l-Data Source: Quadrant Tilts lland Calculations, 16 Symmetric Assembifes Maximum Peaks = Unit Computer Thermal Data Unit Computer Incore Of fset Unit Computer Note (1): Due to computer problems during this phase of the test, the minimum DNBR, assembly location, and maximum radial and total peaking factors were calculated by B&W incorporating an improved fitting routine. The maximum LilR was hand calculated from these results. Note (2): See Note (2), Table 4.12-2.

_~_~___ _ MINIMUM DNBR AND M4XIMUM LIIR ANALYSIS FOR Tile DROPPED CONTROL ROD TEST Powe r Incore Assenh ly Maximum LliR, Kw/Ft Minimum DNBR W rst Rod Level Iob alance Type W rat Measured Measured Position Case Case 77 Meas ured 41.48 -1.55 2 4.77 6.69 9.79 9.11 Normalized 40.00 -1.49 2 4.59 6.44 10.20 9.49 Extrapolated 102.00 -3.80 2 11.73 16.44 4.01 3.47 50 Measured 47.92 -0.63 1 5.70 7.73 8.19 7.51 Normalized 40.00 -0.52 1 4.78 6.49 10.04 9.23 Extrapolated 102.00 -1.34 1 12.13 16.45 3.82 3.18 00 Measured 42.00 -0.50 1 5.86 7.95 Normalized 40.00 -0.48 1 5.58 7.57 - d Extrapolated 102.00 -1.22 1 14.23

19. 31 (2)

= 8' C .L 1 Note (1): The worst case uncertainty f actor which is applied is obtained from Table 4.4-2. Note (2): The Babcock and Wilcox Company performed the minimum DNBR worst case analysis using larger errors and uncertainty. The results indicate an extrapolated minimum DNBR margin of 20 percent as calculated relative to the Technical Spec-ification limit. No as measured value was furnished. I

m_ . _ _ _. ~. _ 4 MEASURED CORE POWER DISTRIBUTION RESULTS WITil DROPPED CONTROL ROD AT 77% WD AT 40% FP y Control Rod Group. Positions Cor_ Pov>r Level 41.57 %FP Axial Imbalance -1.55 %FP Gps 1-5 _100 % vd Cp 7 01 % wd Baron concentration 11.07 ppm Max Quadrant Tilt + 1.88% Cp 6 77 % wd GP 8 21 % wd Core Burnup 7.4 EFPD Xenon Equilibrium Yes Fuel Incore Fuel Incore Assembly Detector Pmax/Pcore P/P Fuel Assembly Detector Pmax/Pcore P/P Fuel I.oca t ion Number Local Assembly Location Number Local Asser hiv 11 - 0 8 1 1.45 1.12 D-10 27 1 _ 19 1.01 11 - 0 9 2 1.64 1.24 C-10 28 1.35 0.99 G-09 3 1.63 1.25 C-09 29 1.53 1.14 F-08 4 1.67 1.24 H-08 30 1.87 1.38 E-09 5 1.70 1.28 B-07 31 1.31 1.01 F-07 6 1.71

1. 30 C-06 32 1.33 0.98 E-07 7

1.74 1.30 D-05 33 1.47 1.06 G-06 8 1.69 1.29 E-04 34 1.40 1.03 G-05 9 1.73 1.30 F-03 35 1.31 0. 89 erg ig-05 10 1.63 1.24 G-02 36 1.34 1.01 K-05 11 1.79 1.35 11- 0 1 37 1.25 0.95 L-06 12 1.79 1.35 L-02 38 0.99 0.76 h M-07 13 1.76 1.32 L-03 39

1. 39 1.02 N-08 14 1.61 1.05 M-03 40 1.19 0.87 l

N-09 15 1.54 1.18 N-04 41 1.21 0.93 M-09 16 1.74

1. 30 0-05 42 1.13 0.84 M-10 17 1.84 1.34 0-06 43 0.96 1.02 1.

18 1.72 1.34 P-06 44 1.37 0.72

1. 7f 1.35 R-07 45 1.05 0.80 K-11 19 K-12 20 1.55 1.18 R-10 46 0.63 0.48 11 - 1 3 21 1.70 1.28 O-10 47 1.36 1.00 G-13 22 1.50 1.14 0-12 48 1.14 0.85 g

F-13 23 1.32 0.98 M-14 49 0.96 0.71 F 24 1.46 0.97 L-13 50 1.31 0.97 G-Il 25 1.77 1.37 D-14 51 0.68 0.50 E-11 26 1.39 0.99 C-13 52 0.79 0.59

MEASURED CORE POWER DISTRIBUTION RESULTS WITil DROPPED CONTROL ROD AT 50% VD FR0tt T11E CORE AT 40% FP Control Rod Group Positions Core Power Level 47.92 %FP Axial Imbalance -0.63 1FP Cps 1-5 100 % ud Cp 7 05

7. ud noron Concentration 11.07 ppm Max Quadrant Tilt + 6.94 %

Cp 6 84 % wd GP 8 21__% ud Core Burnup 7.4 _EFPD Xenon Equilibrium Yes Fuel incore Fuel Incore Assembly Detector Pmax/Peore P/P Fuel Assembly Deter 1r Pmax/Pcore P/P Fuel I.oca t ion Number 1.o ca l Assembly 1.ocation Number Dir al Asserbiv 11-08 1 1.43 1.12 0-10 27 1.41 1.02 11 - 0 9 2 1.56 1.21 C-10 28 1.36 1.01 c-09 3 1,56 1.23 C-09 29 1.57 1.17 F-08 4 1.66 1.26 B-08 30 1.91 1.45 E-09 5 1.70 1.29 B-07 31 1.37 1.06 l'-07 6 1.77 1.35 C-06 32 1.42 1.03 E- 0 7 7 1.81

1. 36 D-05 33 1.60 1.14 s

G-06 8 1.75 1.35 E-04 34 1.54 1.12 h G-05 9 1.85 1.38 F-03 35 1.40 0.94 H-O'i 10 1.74 1. 32 C-02 16 1.44 1.06 K-05 11 1.96 1.45 11 - 0 1 37 1.33 1.01 15 1.-06 12

1. 89 1.41 L-02 38 1.07 0.81 e

M-07 13 1.84 1.37 L-03 39 1,51 1,10 N-08 14 1.64 1.07 M-03 40 1.30 0.93 N-09 15 1.56 1.01 N-04 41 1.32 0.99 M-09 16 1.74 1.31 0-05 42 1.20 0.86 M-10 17

1. 80

1. 34 0-06 43 1.46 1.08 1.-11 18 1.54 1.34 P-06 44 1.01 0.76 K-11 19 1.49 1.22 R-07 45 1.09 0.84 K-12 20 1.22 1.01 R-10 46 0.64 0.50 11-1 3 21 1.38 1.06 0-10 47 1.37 1.02 G-13 22 1.20 0.97 0-12 48 1.09 0.83 F-13 23 1.08 0.88 M-14 49 0.83 0.65 F-12 24 1.25 0.89 L-13 50 1.07 0.87 G-11 25 1.46 1.21 D-14 51 0.61 0.47 E-Il 26 1.38 0.99 C-13 52 0.75 0.57

MEASURED CORE POWER DISTRIBUTION RESULTS WITil DROPPED CONTROL ROD AT 0% WD FROM Tile CORE AT 40% FP Control Rod Group Positions Core Power Level 42.00 %FP Axial Imbalance -0.50 %FP Gps 1-5._100 % wd GP 7 07 % wd 15oron Concentration 1107 ppm Max Quadrant Tilt +13.97 % ~ Gp 6 87 % wd GP 8 21 _% wd Core Burnup _7. 4 EFPD Xenon Equilibrium 7 es, Fuel Incore Fuel Incore l Assembly Assembly Detector Pmax/Peore P/P Fuel Detector Pmax/Pcore P/P Fuel 1.ocation Number Local Assembly Location Number local Assenhly 11 - 0 8 1 1.47 1.17 D-10 27

1. /.?

1.09 11 - 0 9 2 1.49 1.17 C-10 28 1.38 1.02 C-09 3 1,53 1.20 C-09 29 1.61 1.20 F-08 4 1.61 1.28 B-08 30 1.97 1.51 E-09 5 1.73 1.29 11 - 0 7 31 1.42 1.12 F-07 6 1.83 1.40 C-06 32 1.49 1.10 E-07 7 1.88 1.41 D-05 33 1.70 1.22 G-06 8 1.84 1.42 E-04 34 1.63 1.20 [.-05 9 1.96 1.47 F-03 35 1.47 1.02 g Pr _y.;.15 10 1.84 1.41 C-02 36 1.53 1.18 -i' 11 2.03 1.53 11 - 0 1 37 1.40 1.10 E ~ UI 12 1.99 1.49 L-02 38 1.14 0.91 c-h [ H-o/ 13 1.91 1.43 L-03 39 1.61 1.19 N_-jl 14 1.66 1.09 M-03 40 1.37 1.01 N-09 15 1.58 0.98 N-04 41 1.40 1.06 M-09 16 1,76 1.31 0-05 42 1.27 0.93 M-10 17 1.80

1. 30 0-06 43 1.52 1.14 L-11 18 1.49 1.23 P-06 44 1.06 0.83 K-11 19 1.37 1.04 R-07 45 1.12 0.88 K-12 20 1.03 0.78 R-10 46 0.66 0.31 11 - 1 3 21 1.02 0.78 0-10 47 1.39 1.03 c-13 22 0.99 0.76 0-12 48 1.07 0.80 F-13 23 1.00 0.75 M-14 49 0.77 0.58 F-12 24 1.19 0.77 L-13 50 0.99 0.74 G-ll 25 1.35 1.04 D-14 51 0.58 0.43 E-Il 26 1.35 0.95 C-13 52 0.72 0.54

DROPPED ROD WORTH AND LOCATION AT 40% FP PLATEAU X A B C D E F G H W-y K L M N 0 P l R I I I I I I I Z l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 i Core Measured Position Control Rod Worth, % ak/k H-12 0.09(1) 0.10(2) j Note (1): Integrated differential rod worth results. Note (2): Control Rod Group 6 swap results. l Figure 4.12-1

e DIFFERENTIAL AND INTEGRAL ROD WORTH FOR CONTROL ROD ASSDiBLY AT CORE LOCATION H-12 VERSUS ROD POSITION DURING DROPPED CONTROL ROD TEST ~ " " = m:t.e ..t= =tn= =tn=t:mim =t=

}

9'ii: = tam ent-t nett-ert,r =t

m:t:r un.:n -din net = rrn=- nr.. =nct m =_. n nt =_. .: :r=.tunt=n = t =_. 1 .. _ +..., =t=_ ....g. .. g. .....i_ . v. U. 2U..Nm=.. t,.._: =....t=.=. n.n..%........p. _.4 -a.-~ -... ~..u":d- 0 . _.."-"_-~. "._# b... __.1.. . n.. d..= =....=_. l.2.:.. P

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f

q= a_,.

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208 n

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== =.

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aup= =n =:=== :n =t=

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• 5, 1.51 i:n 1:H 2:81 8:H e

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9 0 e 9 ) O e f, t I l l l 1 i l l M r-,,-.,,-m., ,-m+ r

t' DUKE .P0WER C 0 M P A N.Y OCONEE NUCLEAR STATION UNIT 3 DOCKET NO. 50-287 LICENSE No. DPR-55 i STARTUP REPORT i SUPPLEMENT 1 ~l

• l 4

1 JUNE 12,-1975 i l I y .e

Sl*1 INTRODUCTION On March 14, 1975, the Startup Report for Oconee Nuclear Station Unit 3 was submit ted. The report addressed unit startup and power escalation testing through 2400 hours, January 15, 1975. At that time all or part of several power escalation tests remained to be completed. This supplement provides a summary of the tests completed after 2400 hours, January 15, 1975, and prior to 2400 hours, May 31, 1975. S Sl-1

S1.2 UNIT LOAD TRANSIENT TEST AT 100% FP The Unit Load Transient Test was conducted at the 100% FP plateau to evaluate the ability of the Integrated Control System to accomplish a smooth negative and positive change in power level at a ramp rate of 10% FP per minute while operating in the fully-integrated, turbine-following, and reactor / steam generator-following modes of control. This portion of the test was performed on January 19, 1975. The acceptance criteria of demonstrating the above maneuvering capability without an actuation of the Reactor Protective System or violation of any operating limits were satisfied. Analysis of data showed that no additional ICS tuning was required. The Unit Load Transient Test is now complete through the 100% FP plateau. 4 Sl-2

Sl.3 UNIT LOSS OF ELECTRICAL LOAD TEST The purpose of t Unit Loss of Electrical Load Test was to measure the unit response during and af ter a deliberate trip of the generator breakers and to verify the turbine overspeed control. The test was performed on April 30, 1975, while the unit was at 98% FP; and all acceptance criteria were met. l l l Sl-3

)

F '1 Il 5 1 O a

DUKE POWER COMPANY OCOEE NUCLEAR STATION 4 WIT 3 f DOCKETNo.50-287 1 LIGNSE No. DPR-55 ? 4 STARTLP EPORT 4 SUPPLEttvr 2 l 5 AucusT25,1975 f 3 4 l r

4 o e S

2.1 INTRODUCTION

On June 12, 1975, Supplement 1 to the Startup Report for Oconee Nuclear I Station Unit 3 was submitted. That supplement addressed the unit startup and power escalation testing through 2400 hours, May 31, 1975. At that time, all or part of several power escalation tests remained to be com-pleted. This supplement provides the summary of the tests completed during the period from 2400 hours, May 31, 1975, to 2400 hours, July 31, 1975. During this period, che-remaining power escalation tests were completed. l S2-1 1 4 -w

o S2.2 TURBINE / REACTOR TRIP TEST The purpose of the Turbine / Reactor Trip Test was to measure the unit response during and after a deliberate turbine or reactor trip from power, and to verify the proper transfer of the reactor coolant pumps' power supply from the unit auxiliary transformer to the unit startup transformer upon a turbine trip. The Turbine / Reactor Trip Test was completed on July 11, 1975, by performing the turbine trip portfon of the test at full power, and all acceptance criteria were met. S2-2 j

O t / S2.3 LOSS OF CONTROL ROOM TEST The purpose of the Loss of Control Room Test was to demonstrate that the unit could be brought to a hot shutdown condition from outside the control room. The test was initially performed on January 10. 1975, with the unit at 12.8 % FP. A discrepancy that resulted from the test was satisfactorily resolved on July 3, 1975. S2-3 L}}