ML19308D693

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Startup Rept
ML19308D693
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 07/27/1977
From:
FLORIDA POWER CORP.
To:
References
NUDOCS 8003120824
Download: ML19308D693 (260)


Text

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l FLORIDA POWER CORPORATION C,

CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET No. 50-302  ;

LICENSE NO.'DPR-72 STARTUP REPORT l

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JULY 27, 1977

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1 ABSTRACT Crystal River Unit 3 is a 2452 MWe, Babcock and Wilcox PWR with an e 855 MWe Westinghouse tandem compound turbine-generator. It is located on the Gulf Coast of Florida with a saltwater cooled condenser. This report covers fuel loading and startup testing. Important chronology was:

USNRC License DPR-72 Issued December 3, 1976 Fuel Loading December 4-7, 1976 Initial Criticality January 14, 1977 Initial Nuclear Elertric Power January 30, 1977 15% Full Power Reached February 1, 1977 * '

40% Full Power Reached February 27, 1977 Unit Declared Commercial March 13, 1977 75% Full Power Raached March 14, 1977 4

100% Full Power Reached April 1, 1977 f

All Testing Successfully Completed April 26, 1977 Based on evaluation of unit etartup and power escalation testing, the report comes to the conclusica that the unit is operating as designed and may be safely operated at full rated power.

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,, TABLE OF CONTENTS Section -

Page 1.0- INTRODUCTION AND

SUMMARY

1.1-1

1.1 INTRODUCTION

1.1-l' 1.2

SUMMARY

1.2-1 1.2.1 INITIAL FUEL LOADING 1.2-1 1.2.2- TESTING PRIOR TO POWER ESCALATION 1.2-1 1.2.3 POWER ESCALATION TESTS 1.2-2 2.0 INITIAL FUEL LOADING 2.0-1

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3.0 TESTING PRIOR TO POWER ESCALATION 3.0-1 3.1 REACTOR COOLANT PUMP 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 3.1-3 3.

1.4 CONCLUSION

S 3.1-3 3.2 CONTROL ROD DRIVE TRIP TEST 3.2-1 3.2.1 PURPOSE 3.2-1 s 3.2.2 TEST METHOD 3.2-1 3.2.3 ~

EVALUATION OF TEST RESULTS 3.2-2 3.

2.4 CONCLUSION

S 3.2-2 3.3 PRESSURIZEP. 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 3.3-1 3.

3.4 CONCLUSION

S 3.3-1 3.4 ZERO POWER PHYSICS TEST 3.4-1 3.4.1 PURPOSE 3.4-1 3.4.2 TEST METHOD 3.4-2 3.4.3 EVALUATION OF TEST RESULTS 3.4-3 3.4.3.1- INITIAL CRITICALITY 3.4-3 3.4.3.2 NUCLEAR INSTRUMENTATION OVERLAP 3.4-4 3.4.3.3 SENSIBLE HEAT DETERMINATION 3.4-5 3.4.3.4 REACTIVITY CALCULATIONS 3.4-5 3.4.3.5 "ALL RODS OUT" CRITICAL BORON CONCENTRATION 3.4-6 3.4.3.6 CONTROL ROD GROUP WORTHS 3.4-6 3.4.3.7 SOLUBLE POISON WORTHS 3.4-8 3.4.3.8 EJECTED CONTROL ROD WORTH 3.4-8 3.4.3.9 STUCK CONTROL ROD WORTH 3.4-9 3.4.3.10 SHUTDOWN MARGIN DETERMINATION 3.4-10 3.4.3.11 TEMPERATURE COEFFICIENTS OF REACTIVITY 3.4-10 3.4.3.12 TEMPERATURE NORMALIZATION CONSTANTS 3.4-12 3.4.4 C0JCLUSIONS 3.4-12

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TABLE OF CONTENTS (CONT'D)

Section Pm

, 4.0- POWER ESCALATION TESTS 4.0-1 t

. 4.1 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER TEST 4.1-1 I 4.1.1 PURPOSE 4.1-1

4.1.2 TEST METHOD 4.1-2 a 4.1.3 ~ EVALUATION OF TEST RESULTS 4.1-2 4.

1.4 CONCLUSION

S 4.1-3 4.2- BIOLOGICAL SHIELD SURVEY TEST 4.2-1 4.2.1 PURPOSE 4.2-1 4.2.2 TEST METHOD 4.2-2 4.2.3 EVALUATION OF TEST RESULTS 4.2-3 4.

2.4 CONCLUSION

S 4.2-4 4

4.3 PE. ACTIVITY COEFFICIENTS AT POWER TEST 4.3-1

! 4.3.1 PURPOSE 4.3-1 4.3.2 TEST METHOD. 4.3-1 4.3.2.1 TEMPERATURE AND MODERATOR COEFFICIENTS 4.3-1 4.3.2.2- POWER DOPPLER AND DOPPLER COEFFICIENTS 4.3-2 4.3.2.3 DIFFERENTIAL ROD WRTH AT POWER 4.3-3 4.3.3 EVALUATION OF TEST RESULTS 4.3 4.

3.4 CONCLUSION

S 4.3-5 4.4 CORE POWER DISTRIBUTION TEST 4.4-1 4.4.1 PURPOSE 4.4-1 4.4.2 ' TEST METHOD 4.4-2 4.4.3 EVALUATION OF TEST RESULTS 4.4-2

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i 4.4.3.1 NORMAL OPERATING CORE POWER DISTRIBUTIONS 4.4-2

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4.4.3.2 WORST CASE MINIMUM DNBR AND MAXIMUM LHR CALCULATIONS 4.4-3 4.4.3.3 QUADRANT POWER TILT AND AXIAL POWhh IMRALANCE 4.4-6 4.4.'4 CONCLUSIONS 4.4-7 4.5 UNIT LOSS OF ELECT 1(ICAL LOAD TEST 4.5-1 4.5.1 PURPOSE 4.5-1 4.5.2 TEST METHOD 4.5-1 i 4.5.3 EVALUATION OF TEST RESULTS 4.5-1

! 4.

5.4 CONCLUSION

S 4.5-2 l

4.6 TURBINE / REACTOR TRIP TEST 4.6-1 4.6.1 PURPOSE 4.6-1 4.6.2 TEST METHOD 4.6-1 4.6.3 EVALUATION OF TEFT RESULTS 4.6-2 4.6.3.1 REACTOR TRIP TEST AT 40% FULL POWER 4.6-2

( 4.6.3.2 TURBINE TRIP TEST AT 75% FULL POWER 4.6-2 l 4.6.3.3 TURBINE TRIP TEST AT 100% FULL POWER 4.6-3 4.

6.4 CONCLUSION

S 4.6-4 4.7 INCORE DETECTOR TEST 4.7-1 4^.7.1 PURPOSE 4.7-2 4.7.2 TEST METHOD 4.7-2 4.7.3 EVALUATION OF TEST RESULTS 4.7-3 4.7.3.1 INCORE DETECTOR CRECKOUT 4.7.3.2 4.7-3 INCORE DETECTOR SIGNAL CORRECTIONS 4.7-3 4.7.3.3 RADIAI! CORE TOWER DISTRIBUTION COMPARISONS 4.7-5

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TABLE OF CONTENTS (CONT'D)

Section Pa_ge 4.7.3.4 W0kJ CASE MAXIMUM LHR CALCULATION 4.7-5 4.7.3.5 INCORN DETECTOR RESPONSE 4.7-6 4.7.4 CONCLUSl ')NS 4.7-7 4.8 POWER IMBALANCE DETECTOR CORRELATION TEST 4.8-1 4.8.1 PURPOSE 4.8-1 4.8.2 TEST METHOD 4.8-1 6.8.3 EVALUATION OF TEST RESULTS 4.8-2 i

4.

8.4 CONCLUSION

S 4.8-3 4.9 NSSS HEAT BALANCE 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 4.9.3.1 PRIMARY AND SECONDARY HEAT BALANCE CALCULATION 4.9-2 -

4.9.3.2 REACTOR COOLANT FLOW DETERMINATION 4.9-2 4.

9.4 CONCLUSION

S 4.9-3 4.10 -UNIT LOAD STEADY STATE 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 4.

10.4 CONCLUSION

S 4.10-2 4.11 UNIT LOAD TRANSIENT TEST 4.11-1

< 4.11.1 PURPOSE 4.11-2 l 4.11.2 TEST METHOD 4.11-2 l

. ~4.11.3 EVALUATION OF TEST RESULTS 4.11-3 4.11.3.1 ICS TRANSIENT TEST AT THE 40% POWER LEVEL -

4.11-4 4.11.3.2 ICS TRAN3IENT TEST AT THE 75% POWER LEVEL 4.11-4 l 4.11.3.3 ICS TRANSIENT TEST AT THE 100% POWER LEVEL 4.11-4 4.

11.4 CONCLUSION

S 4.11-4 4.12 SHUTDOWN FROM OUTSIDE CONTROL ROOM TEST 4.12-1 l 4.12.1 PURPOSE 4.12-1 l 4.12.2 TEST METHOD 4.12-1 l 4.12.3 EVALUATION OF TEST RESULTS 4.12-1 i 4.

12.4 CONCLUSION

S 4.12-2 4.13 PSEUDO CONTROL ROD EJECTION TEST 4.13-1 4.13.1 PURPOSE 4.13-1 4.13.2 TEST METHOD 4.13-1 4.13.3 EVALUATION OF TEST RESULTS 4.13-2 4.

13.4 CONCLUSION

S 4.13-2 4.14 DROPPED CONTROL ROD TEST 4.14-1 4.14.1 PURPOSE 4.14-1 4.14.2 TEST METHOD 4.14-1 4.14.3 EVALUATION OF TEST RESULTS 4.14-2 4.

14.4 CONCLUSION

S 4.14-3 4.15. LOSS OF OFFSITE POWER 4.15-1 4.15.1 PURPOSE 4.15-1 4.15.2 TEST METHOD 4.15-1 4.15.3 ' EVALUATION OF TEST RESULTS 4.15-1 4.15.4- -CONCLUSIONS 4.15-2 111

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1.0 INTRODUCTION

AND SmefARY 1.1 INTRODUCIION Florida Power Corporation Crystal River Unit 3 was issued facility operating license DPR-72 by the Nuclear Regulatory Comnission on December 3, 1976 for operational modes 5 and 6 (i.e. cold shutdown and refueling). The first fuel assembly was inserted into the core on December 4,1976 and initial fuel load-ing was completed on December 7,1976. This license was later extended to a full power license on December 30, 1976 with an upper power limitation of 5%

full power initially imposed.

Zerc Power Physics Test commenced with initial criticality on January 14, 1977 and continued 5 days until a. severe electrical shortage due to cold weather caused it to be interrupted. Testing resumed on January 27, 1977 and was successfully completed on January 29, 1977. This program was conducted at one isothermal reactor coolant temperature of 532 F.

Following the completion of zero power physics testing, initial power level escalation was started on January 29, 1977, and further power level escala-tions 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 Achieved 15 February 1, 1977 j 40 February 27, 1977 75 March 14, 1977 100 April 1, 1977 The zero power physics and power escalation test programs were done to comply with the requirements of section 13.4 of Crystal River Unit 3 Final Safety Analysis Report (FSAR). These test programs were designed to provide adequate assurance that the unit could be operated in a safe and efficient manner.

Throughout these programs, detailed written procedures specifying the sequence of tests, the parameters to be measured, and the conditions under which each test was to be performed were followed. These procedures have provided the data necessary for the analyses made and conclusions drawn in this report.

This report is prepared and submitted in accordance with Technical Specification 6.9.1 and addresses unit startup and power escalation testing through 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> on April 26, 1977. At this time, all scheduled power escalation testing has been successfully completed.

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1.2

SUMMARY

Crystal River Unit 3 was declared to be in commercial operatian on March 13, 1977 at the completion of 40% full power testing. The nuclear steam supply system was designed by the Babcock and Wilcox Company and was the seventh in this series of systems to be placed Lato service. The tandem compound turbine generator was supplied by the Westinghouse Electric Company.

During the test program, the unit was operated at power levels up to and including 100% full power. 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 affect unit safety.

The deficiencies encountered have been of a nature that would be expected dur-ing the initial startup of a unit of this size and, based on an evaluation of unit startup and power escalation test, it has been demonstrated that the unit may be safely operated at full power.

The startup and power escalation testing, as addressed by the various major sections of this report, is summarized below.

1.2.1 INITIAL FUEL LOADING Initial fuel loading began on December 4,1976 and was completed on December 7, 1976. Loading of the core was accomplished under seni-dry conditions with the taactor vessel water level maintained above the fuel at the upper mid plane of the reactor vessel hot leg. A total elapsed time of 95.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> were required to complete fuel loading with less than 65.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> actually utilized to transfer the fuel assemblies. The delays experienced were primarily contributed to adjustment or repairs to fuel handling equipment. Overall, fuel loading was conducted in a safe and orderly manner with a significant increase in the fuel assembly loading rate experienced as crew familiarization and equipment reli-ability improved.

1.2.2 TESTING PRIOR TO POWER ESCALATION Following initial fuel loading of Crystal River Unit 3, certain testing was conducted prior to power escalation. This testing was conducted in January 1977 and included the Reactor Coolant Punp Flow and Flow Coastdown Test, the Control Rod Drive Drop Time Test, the Pressurizer Test, and the Zero Power Physics Test. A brief summary of each of these tests follows:

REACIOR COOLANT FLOW AND FLOW COASTDOWN TEST The measured reactor coolant flow rate during this test for four pump operation was approximately 109.9 percent of the design value. This flow provides adequate margin to both the mnimum and minimum allowable flow rates. A later flow detennination (110.0%) is given in Section 1.?.3, NSSS Heat Balance Test.

The measured reactor coolant flow rate versus time during the loss of four reactor coolant pumps was above the minimum allowable flow coastdown curve.

An acceptable snubber position for the reactor coolant flow signal was estab-lished at position (4). This position results in delay times well within the acceptance criteria limits of 1.00 seconds for multi-pump trips and 1.25 seconds for single pump trips.

1.2-1

CONTROL ROD DRIVE DROP TIME TEST N

0 The rod drop time at 532 F and full flow conditions averaged about 1.249 seconds '

which is well below the acceptance criteria value of 1.660 seconds as stated in section 3.1.3.4 of Technical Specifications. Also, as would be expected, all drop times under no-flow conditions were shorter than full-flow conditions.

PRESSURIZER TEST The pressurizer spray flow was set at 190.15 gpm which is within the acceptance criteria limit of 190 +19/-6 gpm. The pressurizer. spray bypass flow was set at 1.56 gpm which is within the acceptance criteria limit of 1.00 +2.00/-0.25

, gpm.

ZERO POWER PHYSICS TEST Zero power physics testing on Crystal River Unit 3 commenced with initial criticality on January 14, 1977. Part way through the testing, the FPC system

load was impacted by a severe cold spell. The license was at that time restricted to 5.0% full power. The decision was therefore made to interrupt testing to save the approximate 30 MWe required to run the reactor coolant pumps and other plant equipment. Testing was finally concluded after the afore-mentioned eight day delay on January 29, 1977. The progrts which was intended

, to verify the nuclear design parameters of the core prior to escalation into 0

the power range was performed at 532 F and 2155 Psig. In general, the Zero Power Physics Test was conducted with favorable agreement between measured and predicted results. A summary of each measurement performed is given in ~j Section 3.4, Table 3.4-13.

1.2.3 POWER ESCALATION TESTS Following the completion of zero power physics testing, initial power escala-tion commenced on January 29, 1977 with the first nuclear electrical power produced at 1800 on January 30, 1977. The power escalation test program was conducted at four major ' test plateaus of 15, 40, 75, and 100% full power with minor testing performed at intermediate power levels as required by the con-trolling procedure for power escalation.

Test results review was performed concurrently with testing. This allowed immediate review and approval by the Test Working Group (TWG) and Plant Review Committee (PRC) . Escalation to a new power level always followed promptly behind completion of testing. A brief summary of each of these tests follows:

NUCLEAR INSTRUMENTATION CALIBRATION AT POWER TEST l The power range channels were calibrated to indicate within 2.0% full power of the heat balance and to within + 3.5% offset of the full incore system several times during the startup program. (These calibrations were required due to variations in reactor coolant system boron concentration, cold leg temperature, control rod configuration and xenon buildin and burnout) . Early in the power

, escalation test program some difficulties were experienced until procedural s

errors were resolved and instrumentation personnel experience increased. Also near the completion of the 40% full power plateau, the procedural method was ]

modified from an on-line to off-line technique to assist in removing time dependent drift. In spite of these difficulties, the acceptance criteria were met in all calibrations required by the power escalation procedure.

1.2-2

1 l

l BIOLOGICAL SHIELD SURVEY TEST Upon completion of the Biological Shield Survey Test, the folicwing conclusions j were reached: l (1) Radiation and high radiation areas have been identified and posted.

(2) Radiation levels at various points around the nuclear unit have been estab-lished for . future reference and comparison.

(3) Areas in which radiation levels exceed their design values are undergoing further design evaluation. In the meantine, they have been appropriately ~

identified and marked.

REACTIVITY COEFFICIENTS AT POWER TEST The measured results at 40, 75, and 100% full power indicate that the moderator coefficient of reactivity will always be negative during power operation above 95% full power. Comparison between predicated and measured temperature coefd ficients of reactivity showed favorable agreement.

Analyzed data for the power Doppler coefficient versus power level indicates that the least negative coefficient is -1.02 x 10-4Ak/k/%FP which is sufficiently more negative than the acceptance criteria of -0.55 x 10 4Ak/k/%FP. The total power Doppler deficit from this measured data is projected at -1.10%Ak/k.

I CORE POWER DISTRIBUTION TEST Four normal operating' equilibrium xenon core power distributions taken at 15, 40, 75, and 100 percent full power were examined. The maximum radial peaking factors were not more than 5.0% greater than predicted by PDQ-7. The maximum total peaking factors were not more than 7.5% greater than predicted by PDQ-7.

Comparisons between measured and predicted radial and total core power distritu-tions as expressed in peaking factors on a 1/8 core power distribution showed favorable agreement at;all core locations.

The worst case minimum DNBR and maximum LHR measured as part of this test were subject to various types of analysis. On the bases of this study, the following were determined:

(1) A worst case minimum DNBR of 3.03 and maximum LHR of 12.7 kW/ft was measured at 100% full power. This is within their respective acceptance criteria limits af 1.30 and 18.00 kW/ft.

(2) ~ Acceptable core conditions at the trip setpoint of the next power level of escalation were verified prior to escalating reactor power.

(3) ' Margin analysis on the min 4== DNER and maximum LHR values at the LOCA and design overpower limits resulted in substantial margins.

' The results of the quadrant tilt and axial power imbalance calculations for a variety of different core power distributions taken during the power escalation test program yield the following conclusions:

1 1.2-3 I

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-(1) All maxi m m' quadrant power tilts determined during normal power operation were well within the Technical Specifications Limit of 4%. A typical 3 value observed was around +1.80%.

(2) The reactor protection system will provide sufficient protection against exceeding DNBR and LHR limits when the delta flux amplifier has a gain factor of 3.90.

UNIT LOSS OF ELECIRICAL LOAD TEST Upon completion of the Unit Loss Of Electrical Load Test at 100% full power, the

, following conclusions were reached:

(1) The unit successfully sustained a net load loss at 100% full power without any fuel or equipment damage.

(2) Unit response to the load loss indicates that the integrated control system successfully ran reactor power back to 15% full power at an average ramp i

rate 16.8% FP/ min with large margins to RPS trip 'setpoints.

(3) Turbine speed and frequency oscillations were experienced after the runback i to 15% full power, thus causing the acceptance criteria on frequency control to be exceeded. -Further investigation revealed that since the oscillation magnitude obcerved was small it did not represent a hazard to i the operating equipment on the unit.

TURBINE / REACTOR TRIP TEST Upon completion of the reactor and turbine trip portions of Turbine / Reactor Trip Test at 40, 75, and 100% full power, the following conclusions were reached:

(1) Analysis of the unit parameters during the reactor and turbine trip transients indicate that all acceptance criteria were met.

l (2) Unit . response during the turbine trip portions demonstrated that the integrated control system can successfully run reactor power back to 15%  ;

full power with large margins to RPS trip setpoints.  ;

(3) The reactor trip caused the turbine to trip which ensures that cooldown

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rates less than 1000 F/hr can be maintained following reactor trips at power. {

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-(4) Af ter adjustments, the feedwater block valves closure times were approxi-mately 28 seconds. This is rcquired to be less than 30 seconds by the procedural acceptance criteria. 1 INCORE' DETECTOR TEST i

Upon. completion of Incore Detector Testing at 40 and 75% full pcwer, the fnI-lowing was concluded:

(1) The incore monitoring system has performed as expected during the startup )

period. . Verification of . flux shapes on synunetric grou, Fq and ineme ../

detector response has been excellent.

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' (2) Computer corrections to the uncorrected detector signals were found to

, be consistent with hand calculations.

(3) The ability of the computer to determine the radial core power distribu-tion was demonstrated.

(4)~ Comparison of computer and hand-calculated worst case LHR indicates that

< the computer is slightly in error. No change to the computer is planned since the error is overshadowed by the 24 percent conservatism presently applied.

POWER IMBALANCE DETECTOR CORRELATION TEST

- Upon completion.of power imbalance detector correlation testing at 40 and 75%

full power, the following were concluded:

(1) The measured offset correlation function between the full incore system .

and each out-of-core detector was determined to be a linear relation-ship.; The average correlation slope measured was'1.08.

(2) A gain factor of 3.90 on each out-of-core power range detector difference amplifier will yield an acceptable slope relationship to the full incore system.

(3) The power / imbalance / flow envelope as set in the reactor protection system will protect the reactor from exceeding minimum DNBR and -4='m LHR thermal limits, when a gain factor of 3.90 is utilized.

(4) The backup recorder can provide an acceptable measurement of core imbalance.

A measured correlation slope of approximately 1.00 was observed.

NSSS HEAT BALANCE TEST All primary and secondary heat balance calculations met their respective acceptance criteria of being within + 2% full power of the hand calculated value.

The primary reactor coolant system flow rate on Crystal River Unit 3 was set at 110.0 percent of design flow, which is within the minimum and maximum allow-able flow values of 105.0 and 115.3 percent of design.

UNIT LOAD STEADY STATE TEST The average of the measured unit parameters during the test period fell within their respective =4n4='=/ maximum limits, except for OTSG outlet steam pressure which was indicating 20 Psig high and therefore controlling low. This was 3

corrected by recalibrating the pressure transmitter controlling steam header pressure.

Analysis of unit parameter stability indicates that all variables are relative-I ly stable. The maximum variation in unit parameters was experienced around 70%

full. power.

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UNIT LOAD TRANSIENT TEST All transients were performed without exceeding the limits of the Crystal River Unit 3 Technical Specifications and Sections 1 and 2 of Plant Limits And Pre-cautions. In addition, each was completed without causing the reactor protec-tion system to actuate.

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 each transient was excellent.

SHUTDOWN FROM OUTSIDE CONTROL ROOM TEST The test proved that the reactor can be brought to and maintained in a safe hot standby condition from locations outside the control, room by the normal shift complement.

PSEUDO CONTROL ROD EJECTION TEST The measured worth of the most reactive control rod was found to be 0.20%Ak/k which is less than the acceptance criterion of 0.65%Ak/k.

Analyzed core power distribution and thermal hydraulic data indicated a large perturbation to the steady state core power distribution, as was expected. No core limits were exceeded.

DROPPED CONTROL ROD TEST j Upon analysis of the Dropped Control Rod Test data, the following conclusions were deduced:

(1) Analyzed core power distribution and thermal hydraulic data .provided results indicating sufficient 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 quadrant power tilt of

+13.48%.

(2) The measured worth of the control rod which is calculated to produce the most adverse thermal effects in the core, if it is inadvertently dropped, was found to be -0.12%Ak/k.

l (3) The integrated control system accurately detected the asymmetric control l ' rod with appropriate alarms and initiated reactor runback. The power lere? was below 60.0% full power in 36.7 seconds.

i (4) The control rod withdrawal inhibit was shown to limit power to less than l 60% full power when an asymmetric control rod condition exists.

LOSS OF OFFSITE POWER TEST l

l The test proved the plant's ability to sustain a loss of offsite power. A i

problem in steam generator level control during the transient is still under -

l study. In the meantime, the emergency procedure for loss of offsite power has the operator monitor steam generator water levels and maintains balanced feed-flow between steam generators.

1.2-6

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2.0 INITIAL FUEL LOADING

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-Fuel loading was initiated with the insertion of fuel assembly 3C04 into the core on December 4,1976 and was completed with the loading of 3C53 on December 7, 1976. A total elapsed time of 95.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was required to complete fuel load-ing. However, less than 65.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> were actually utilized to transfer the fuel assemblies with the remaining time required primarily to adjust or repair fuel handling equipment. Figure 2.0-1 depicts the final configuration of the core at the conclusion of initial fuel loading. Table 2.0-1 provides the initial sequence for Crystal River Unit 3.

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 BF3 proportional counting systems furnished by Babcock and Wilcox. Prior to their use, each detector system was calibrated and source checked to ensure proper response. During fuel loading independent plots of

-inverse neutron count rate ratios versus the number of fuel assemblies loaded were maintained from_the output of these detectors as each fuel assembly was -

. loaded - see Figures 2.0-2 through 2.0-5. New base count rates were established for a detector whenever a. detector or neutron startup source was moved. When fuel assemblies are being loaded in the vicinity of a detector, geometric effects (i.e. neutron transmission vice multiplication) create large changes in count rate on the affected detector. Whenever this occurred, the procedure allowed the next two most conservative detectors to be used to predict the number of fuel assemblies to criticality. Using this technique, no partial fuel assembly insertions were required during Crystal River Unit 3 initial fuel loading.

Initial fuel loading at Crystal River Unit 3 was a semi-dry operation with the reactor vessel water level maintained above the fuel at the upper mid plane of i the reactor vessel hot leg. The semi-dry loading improved visibility of the fuel assemblies during manipulations and provided accessability to the vessel j flange area when repositioning the temporary detectors. Radiation levels were j not overly restrictive due to the lower water level in the fuel transfer and l spent. fuel pool canals. The maximum radiation level measured was 10 mrem /hr (n0) )

and 1 mrem /hr (8 + y) at the fuel handling bridge during the transfer of the I fuel assemblies containing the neutron startup sources. l Several minor problems were encountered during the initial fuel loading. A l discussion of these problems and their resolution is given in Table 2.0-2. The effects of these problems in some cases resulted_in delaying the fuel loading process. Figure 2.0-6 shows these delays as a function of time into fuel

+

loading-with each delay noted. From this Figure, it is seen that there were no major delays which stopped fuel loading for a prolonged period.

In spite of the above problems and delays, the fuel assembly loading rate increased significantly during the loading operation as crew familiarization and equipment reliability improved. Figure 2.0-7 is a plot of the frequency

~ distribution of the fuel assemblies loading time intervals. As can be seen, the mean loading time per assembly was 22.1 minutes.

In summary, initial fuel loading at Florida Power. Corporation Crystal River 7 Unit 3 was completed in 95.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and except for the minor problems and delays i

noted, fuel. loading proceeded in an orderly and smooth manner.

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l Fuel Assembly Loading Sequence i CRA = Control Rod Assembly ^D APSRA = Axial Power Shaping Rod Assembly i ORA = Orifice Rod Assembly BPRA = Burnable Poison Rod Ascembly ASSEMBLY Core l Step . Type ID# Feature ID# Position Action l 0-1 Support Detector A 14-H Insert

0-2 Support Detector B 10-P Insert 1 Fuel 3C04 Neutron Source 110 14-N Insert 2 Fuel 3C20 Neutron Source 111 2-D Insert 3 Fuel 3C55 ORA 168 13-0 Insert
4 Fuel 3C37 BPRA B47 13-N Insert 5 Fuel 3C35 BPRA B52 12-0 Insert 6 Fuel 3C48 ORA 170 14-M Insert .

7 Fuel 3A04 CRA C54 12-N Insert 8 Fuel 3A27 CRA C51 13-M Insert 9 Fuel 3B05 BPRA B66 12-M Insert 10 Fuel 3B39 BPRA B42 13-L Insert 11 Fuel 3B28 BPRA B68 11-N Insert 12 Fuel 3A42 APSRA A06 12-L Inser t 13 Fuel 3A21 CRA C50 11-M Insert 14 Fuel 3B50 BPRA B62 11-L Insert 15 Fuel 3B41 BPRA B40 12-K Insert 16 Fuel 3B15 BPRA B65 10-M Insert ,)

17 Fue1 3A44 CRA C44 10-L Insert i 18 Fuel 3A20 CRA C39 ll-K Insert 19 Fuel 3B25 BPRA B15 10-K Insert i 20 Fuel 3B33 BPRA B18 9-L Insert 21 Fuel 3B02 BPRA B37 11-H Insert 22 Fuel 3A48 APSRA A08 10-N Insert 23 Fuel 3A01 CRA C49 9-M Inser t 24 Fuel 3B23 BPRA B46 9-N Insert 25 Fuel 3B61 BPRA B43 8-M Insert 25-1 Support Detector B 10-P Remove 25-2 Support Detector B 7-M Insert 26 Fuel - 3A34 CRA C58 11-0 Insert 27 Fuel 3B21 BPRA B51. 10-0 Insert 28 Fuel 3C29 ORA 112 12-P Insert 29 Fuel 3C33 ORA 171 ll-P Insert r 30 Fuel 3A33 CRA C40 9-K Insert

! 31 Puel '3A47 CRA C43 8-L Insert L 32 Fuel 3A54 CRA C32 10-H Insert 33 Fuel 3B59 BPRA B14 8-K Insert

[ 34 Fuel 3B55 BPRA Bil 9-H Insert l .35 Fuel- 3B37 CRA C31 8-H Insert j

36 Fuel 3B32 BPRA B17 7-L Insert 37 Fuel 3B51 BPRA B08 10-G Insert

, 38 Fuel 3A45 CRA C37 7-K Insert

! 39 Fuel 3A39 CRA C25 9-G Insert

40 Fuel 3B58 BPRA B10 7-H Insert ~J i

l l Table 2.0-1

Fuel Assembly Loading Sequence ASSEMBLY Core Step g ID# Feature ID# Position Action 41 Fuel 3B57 BPRA B07 8-G Insert 42 Fuel 3A26 CRA C24 7-G Insert 43 Fuel 3B48 BPRA B13 6-K Insert 44 Fuel 3B20 BPRA B04 9-F Insert 45 Fuel 3A56 CRA C30 6-H Insert 46 Fuel 3A30 CRA C19 8-F Insert 46-1 Support Detector A 14-H Remove 46-2 Support Detector A 10-F Insert 47 Fuel 3B10 BPRA B06 6-G Insert 48 Fuel 3B18 BPRA B03 7-F Inst.rt 49 Fuel 3A38 CRA C18 6-F Insert 50 Fuel 3B01 BPRA B36 5-H Insert 51 Fuel 3Ld BPRA B30 8-E Insert 52 Fuel 3A16 CRA C23 5-G Insert 53 Fuel 3A12 CRA C13 7-E Insert -

54 Fuel 3B13 BPRA B59 5-F Insert 55 Fuel 3B49 BPRA B56 6-E Insert 56 Fuel 3A23 CRA C12 5-E Insert 57 Fuel 3B07 BPRA B33 4-G Insert 58 Fuel 3B53 BPRA B27 7-D Insert 59 Fuel 3A09 APSRA A03 4-F Insert 60 Fuel 3A06 APSRA A01 6-D Insert 61 Fuel 3B30 BPRA BS5 4-E Insert 62 Fuel 3B03 BPRA B53 5-D Insert 63 Fuel 3A53 CRA C08 4-D Insert 64 Fuel 3B11 BPRA B31 3-F Insert 65 Fuel 3B26 BPRA B22 6-C Insere 66 Fuel 3A11 CRA C11 3-E Insert l 67 Fuel 3C28 BPRA B26 3-D Insert 68 Fuel 3C14 ORA 173 2-E Insert )

69 Fuel 3A28 CRA C04 5-C Insert 70 Fuel 3C11 BPRA B21 4-C Insert 71 Fuel 3C52 ORA 175 3-C Insert 72 Fuel 3C40 ORA 109 4-B . Insert  ;

73 Fuel 3C46 ORA 176 5-B Insert 74 Fuel 3A50 CRA C42 6-L Insert I 75 Fuel 3A41 CRA C36 5-K Insert 76 Fuel 3B43 BPRA B61 5-L Insert

)i 77 Fuel 3A03 CRA C29 4-H Insert 78 Fuel 3B36 BPRA B39 4-K Insert 79 Fuel 3A07 APSRA A05 4-L Insert 80 Fuel 3A25 CRA C22 3-G Insert 81 Fuel 3B24 BPRA B35 3-H Insert 82 Fuel -3A08 CRA C35 3-K Insert 83 Fuel 3B42 BPRA B41 3-L Insert 84 Fuel 3C16 CRA C17 2-F Insert 85 Fuel 3B60 BPRA B05 2-G Insere 86 Fuel 3C43 CRA C28 2-H Insert 87- Fuel 3B54 BPRA B12 2- K Insert 88 Fuel 3C15 CRA C41 2-L Insert 89 Fuel 3C41 ORA 177 1-F Insert i

Table 2.0-1 (Continued)

. _ . - . - ._.._._m.~_-. ~ - . .. - - _ - _ . _ . . _

't Fuel Assembly Loading Sequence ASSEMBLY Core Step g IDd Feature ID# Position Action .

90 Fuel 3C08 ORA 183 1-G Insert 91 Fuel 3C19 ORA 188 1-H Insert 92 .

Fuel - 3C10 ORA 189 1-K Insert 92-1 Support Detector B 7-M Remove 92-2 Support Detector B l-L Insert 93 Fuel 3C05 CRA C61 10-P Insert 94 Fuel 3C54 ORA 190 10-R Insert 95 Fuel 3A51 CRA C57 9-0 Insere 96 Fuel 3B27 BPRA B20 9-P Insert 97 Fuel 3C21 ORA 192 9-R Insert 98 Fuel 3A40 CRA C53 8-N Insert 99 Fuel 3B34 BPRA B50 8-0 Insert 100 Fuel 3C18 CRA C60 8-P Insert 101 Fuel 3C07 ORA 193 8-R Insert  ;

102 Fuel 3A17 CRA C48 7-M Insert - '

103 Fuel 3B46 BPRA B45 7-N Insert 104 Fuel 3A02 CRA C56 7-0 Insert 105 Fuel 3B43 BPRA B19 7-P Insert 106 Fuel 3C27 ORA 196 7-R Insert 107 Fuel 3B08 BPRA B64 6-M Insert 108 Fuel 3A10 APSRA A07 6-N Insert 109 Fuel 3B47 BPRA B49 6-0 Insert 110 Fuel 3CO2 CRA C59 6-P Insert 111 Fuel 3C56 ORA 200 6-R Insert j

=.

112 Fuel 3A18 CRA C47 5-M Insert 113 Fuel 3B09 BPRA B67 5-N Insert 114 Fuel- 3A31 CRA C55 5-0 Insert 115 Fuel 3C26- ORA 201 5-P Insert 116 Fuel 3B29 BPRA B63 4-M Insert 117 Fuel 3A05 CRA CS2 4-N Insert 118 Fuel 3C22 BPRA B48 4-0 Insert 119 Fuel 3A55 CRA C46 3-M Incert 120 Fuel 3C09 BPRA B44 3-N Insert 121 Fuel 3C59 ORA 203 3-0 Insert

-122 Fuel -3C03 ORA 204 2-M Insert 123 Fuel 3C51 ORA 206 2-N Insert 124 Fuel 3A13 CRA -

C26 11-G Insert 125 Fuel 3B14 BPRA B60 11-F Insert 126 Fuel 3A36 CRA C33 12-H Insert 127 Fuel '3B19 BPRA B34 12-G Insert 128 Fuel 3A49 APSRA A04 12-F Insert

- 129 Fuel 3A46 CRA- C38 13-K Insert 130 Fuel 3B31- BPRA B38 13-H Insert 131 Fuel 3A35 CRA C27 13-G Insert 132 Fuel 3B44 BPRA B32 13-F Insert 133 Fue1 3C32 CRA C45- 14-L Insert 134 Fuel 3B56 BPRA B16 14-K Insert 135 Fuel 3C45 CRA C34 14-H Insert 136- Fuel 3B38 BPRA B09 14-G Insert 137 Fuel 3C31 CRA C21 14-F Insert 138 Fuel 3C57 ORA 207 15-L Insert Table 2.0-1 (Continued)

Fuel Assembly Loading Sequence ASSEMBLY Core Step Type IDd Feature - ID# Position Action 139 Fuel 3C13 ORA 209 15-K Insert 140 Fuel 3C06 CRA 213 15-H Insert 141 Fuel 3C12 ORA 214 15-G Insert 141-1 Support Detector A 10-F Remove 141-2 Support Detector A 15-F Insert 142 Fuel 3A29 CRA C20 10-F Insert

. 143 Fuel 3C24 CRA C01 6-B Insert 144 Fuel 3C42 ORA 217 6-A Insert 145 Fuel 3A15 CRA C05 7-C Insert 146 Fuel 3B04 BPRA B01 7-B Insert 147 Fuel 3C44 ORA 220 7-A Insert 148 Fuel 3A19 CRA C09 8-D Insert 149 Fuel 3B35 BPRA B23 8-C Insert 150 Fuel 3C17 CRA CO2 8-b Insert 151 Fuel 3C01 ORA 221 8-A Insert -

152 Fue l - 3A24 CRA C14 9-E Insert 153 Fuel 3B22 BPRA B28 9-D Insert 154 Fuel 3A52 CRA C06 9-C Insert 155 Fuel 3B40 BPRA B02 9-B Insert 156 Fuel 3C36 ORA 222 9-A Insert 157 Fuel 3B16 BPRA B57 10-E Insert 158 Fuel 3A37 APSRA A02 10-D Insert 159 Fuel 3B17 BPRA B24 10-C Insert 160 Fuel 3C25 CRA C03 10-B Insert 161 Fuel 3C38 ORA 224 10-A Insert 162 Fuel 3A14 CRA C15 11-E Insert 163 Fuel 3B12 BPRA B54 11-D Insert 164 Fuel 3A22 CRA C07 11-C Insert 165- Fuel 3C23 ORA 225 11-B Insert 166 Fuel 3B06 BPRA B58 12-E Insert 167 Fuel 3A32 CRA C10 12-D Insert 168 Fuel 3C49 BPRA B25 12-C Insert 169' Fuel 3A43 CRA C16 13-E Insert I 170 Fuel 3C50 BPRA B29 13-D Insert 171 Fuel 3C60 ORA 227 13-C Insert 172 Fuel 3C47 ORA 229 14-E Insert 173 Fuel 3C39 ORA 23 0 14-D Insert l 174 Fuel 3004' ORA 110 14-N Remove l 174-1 Fuel 3C04 ORA 110 4-P Insert l 175 Fuel 3C30 ORA 231 14-N Insert 176 Fuel 3C20 ORA Ill 2-D Remove 176-1 Fuel 3C20 ORA Ill 12-B Insert 177 Fuel 3C34 ORA 232 2-D Insert 173 Support Detector B 1-L Remove 178-1 Fuel 3C58 ORA 233 1-L Insert 179 Support Latector B 15-F Remove 179-1 Fuel 3C53 ORA 234 15-F Insert

i. .

l l

Table 2.0-1 (Continued) l l

l Summary Of Problems Experienced During Initial Fuel Loading.

4 Problem Number Problem Tvoe Description Of Problems Encountered And Corrective Actions i

1 NI-2 Scaler Timer During.the loading of the first four fuel assemblies, the count rate I observe,by the scaler timer on NI-2 did not produce the expected increase in count rate. It was taken out of service and replaced with a spare while the fifth fuel assembly was being loaded. The problem i was the result of an improper gain adjustment.

i 2 TYGON Tubing Leak The auxiliary neutron monitoring cables are pressurized to 20 psig to prevent moisture damage during fuel loading. Gas leaks were found at 3

the TYCON tubing connection to the auxiliary detector (A) (minor leak) and at the TYGON tubing connaction to the electronics of auxiliary

! e detector (A) (major leak) . The major leak was patched with tape and E- ' caulking compound.

.c y 3 Loading Assemblies During t,he loading of 3C32 into core location L-14 the assembly would o Into Core not seat. The assembly was removed from the core, inspected, and l, found to be okay. The decay heat pump was stopped and after cable manipulation the assembly seated.

! Fuel assembly 3A43 did not seat on the initial try. The second l .

attempt, however, seated the assembly without any problems.

I

( 4 Spent Fuel Bridge' (KEY) The keys which hold the drive wheels on the drive shaft fell '

out of the keyways. There is a coupling on each end of the bridge. It was felt that one key fell out prior to fuel loading and only ese wheel was driving the bridge. When the second key fell out. th. *

{ bridge came to a halt. The problem was resolved by replacing and

( staking the keys in the keyways.

l l

t L

l

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Summary Of Problems Experienced During Initial Fuel Loading Problem Number Froblem Type Description of Problems Encountered And Corrective Actions 4 Spent Fuel Bridge (SET SCREW) The set screw in the motor coupling lcosened twice during fuel loading. Each time the set screw was retightened.

5 Main Fuel Bridge (HOIST MOTOR) The Main Fuel Bridge hoist motor overheated and stopped operating. This was corrected by installing a fan for cooling.

(CABLE) The Main Fuel Bridge cable takeup caused problems and g required the operators to make sure the cable did not get run over by

g. the main bridge.

E g (HYDRAULIC TELESCOPE CYLINDER) The Main Fuel Bridge stopped operation when the hydraulic telescopic cylinder drifted causing the control rod

[ test light to come on and activate an interlock. The rod grapple reset switch was depressed to remove the interlock.

R a 6 Fuel Transfer (MOTOR COUPLING SET SCREW) The failure of a set screw on the Y upender

[- Mechanisms between the motor coupling and the shaft allowed the shaft to turn freely. The set screw was removed, two holes were drilled and tapped and two set screws installed to secure the coupling on the shaft.

(HYDRAULIC TUBE) The hydraulic tube for the Y transfer carriage became entangled on the takeup reel twice during fuel loading.

(HYDRAULIC AIR LEAK) A Itydraulic air leak on the Y transfer carriage took this system out of service before the last scheduled move.

Loading was completed using the X carriage.

7 Auxiliary Fuel Bridge There was some trouble latching on to the auxiliary neutron detector.

The decay heat pump was secured and movement of the hoist was guided by visually observing the operation with binoculars.

Summary Of Problems Experienced During Initial Fuel' Loading i

Problem Number Problem Type Description Of Problems Encountered And Corrective Actions i ,

8 Observable Countrate During the loading of the last fuel assembly, difficulty was

. experienced in obtaining a detectable countrate on auxiliary detector A. After trying several positions at the top of the

core, the detector was removed and held over the edge of the 1 4

reactor vessel, thereby providing an adequa'.e countrate. -

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Final Fuel Loading Distribucion For Crystal River Unit 3 A - 3C42 3C44 3C01 3C36 3C38 0217 0220 0221 C222 0224 8

3Cu 3m 3C24 3804 3C17 mo 3C25 X23 3C20 0109 0176 Cool 800 1 C002 8002 C002 0225 0111 C 3C32 3C11 3A23 3826 3A13 3B35 3A32 3817 3A22 3049 3C60 0175 8021 C004 8022 C005 8023 C006 8024 C007 B025 0227 D 3C34 3C23 3A53 3803 3A06 3853 3A19 3822 3A37 3B12 3A32 3C50 ;C39 0232 8026 C003 8053 A001 B027 C009 5023 A002 8054 C010 8029 0230 E 3C14 3A11 3830 3A23 3B49 3A12 3552 3A24 3816 3A14 3B06 3A43 2047 0173 C011 8055 C012 8056 C013 B030 C014 B057 C015 B058 C016 3229

, 3C41 3C16 3811 3A09 3813 3A38 3518 3A'O 3B20 3A29 3814 3A49 3844 3C31 3C33 0177 2017 8031 A003 8059 C018 3003 C019 B004 CO20 8060 A004 8032 2021 C234 G 3C08 3860 JA25 3B07 3A16 3510 3A26 3B57 3A39 3B51 3A13 3819 3A35 3333 2012 0183 9005 c022 8033 CO23 5006 CO24 8007 CO25 8008 C026 B034 CO27 009 0214 H- 3C19 3C43 3324 3A03 3801 3A56 3858 3837 3553 3AS4 3B02 3A36 3831 3C45 3026 0188 2029 3035 CO29 F)36 C030 8010 C031 8011 C032 B037 C033 B038 2034 0213 K 3C10 3B54 3AOS 3836 3. 1 3848 3A45 3859 3A33 3825 3A20 3B41 3A46 3556 3C13 0189 m 3 cm en3a en A ent3 cm 3ntc C040 ents m3a en C038 m c30o 3C58 3C15 3B42 3A07 3843 3A50 3832 JA47 3833 3A44 3850 3A42 3539 3C32 3037 0233 :041 8041 A005 B061 C042 3017 C043 8018 C044 8062 A006 8042 :245 02C7 l

g 3CO3 3A55 3B29 3A18 3808 3A17 3B61 3A01 3815 3A21 3B05 3A27 3C48 0204 C046 B063 C047 B064 C046 6043 c649 8065 C050 8066 C051 0170 .

N 3C51 3C09 3A05 3809 3A10 3B46 3A40 3B23 3A48 3828 3AC4 3C37 1C30 3206 3044 C052 B067 A007 B045 C053 B046 A008 8068 C054 5047 C231 0 3C59 3C22 3A11 3B47 3A02 3B34 3A51 3B21 3A34 3C35 3C55 0203 8048 C055 8049 C056 B050 C057 3051 C058 5052 0163 P 3CD4 3C26 3C02 3B45 3C18 3827 3C05 3C33 3C29 0110 0201 C059 2019 C060 3020 C061 0171 0112 g 3C56 3C27 3C07 3C21 3C54 0200 0196 0193 0192 0190 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 3A01 through" 3A56 = 1.93 vt 7. het Assemblies 3801 through 3861 - 2.54 we ?. Puel Assemblies 3C01 throu$ 3C60 - 2.83 vt % Fuel Assemblies B001 through B020 - 1.34 vt % Burnable Poison Assemblies B021 through B052 = 1.18 vt % surnable Poison Assemblies 8053 through B068 = 1.01 we ** Burnable Pof son Assemblies C001 through C061 = Control Rod Assemblies A001 through A008 = Axial Power Shaping Rod Assemblies

, 0109 through 02.34 = Orifice Rod Assemblies Figure 2.0-1

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a 3.0 gSTINGPRIORTOPOWERESCALATION Following initial fuel loading of Crystal River Unit 3, certain testing was conducted prior to power escalation. This testing was conducted in January 1977 and included the Reactor Coolant Pump Flow and Flow Coasedown Test, the Control Rod Drive Drop Time Test, the Pressurizer Test, and the Zero Power Physics Test. This section of the report presents the results of these tests.

In mil cases, all applicable Technical Specifications requirements were met.

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3.0-1 1

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3.1 REALTOR COOLANT FLOW AND FLOW COASTDOWN TEST Prior to Power Escalation, the reactor coolant pump flow and flow coastdown test was performed to ensure the functional capabilities of the reactor coolant system and the reactor coolant pumps during steady state flow and flow coast-down conditions.

This test was conducted at hot conditions of 532 F and 2155 Psig with the core installed. The intent was to measure the normal flow rate with four reactor coolant pumps operating and the minimum flow rate with two and three reactor coolant pumps operating. In addition it also measured the flow coastdown and time delay characteristics for the loss of four, two and one reactor coolant pumps from the normal operating conditions.

3.1.1 PURPOSE The purpose of the reactor coolant flow and flow coastdown test are listed below: ,

(a) To measure the reactor coolant flow and flow coastdown characteristics for selected pump combinations at 532 0 F and 2155 Psig with the core in-stalled.

(b) To compare reactor coolant flow with design calculations and to verify adequate core flow.

(c) To verify that the reactor coolant system flow instrumentation response time is within acceptable limits.

Four acceptance driteria are specified for the reactor coolant flow and flow coastdown test as listed below:

(1) Steady state total reactor coolant system flow with the core and orifice rods installed when corrected to 5320F and 2155 Psig shall be within the limits specified in, Table 3.1-2.

(2) The delay time between the snubbered and unsnubbered flow response after a flow reduction equivalent to the flux to flow ratio of 1.044 must be less than 1.25 seconds for the loss of one reactor coolant pump and 1.00 seconds for the loss of two and four reactor coolant pumps from four pump initial conditions at 532 F and 2155 Psig.

(3) The reactor coolant system flow when corrected to 532 F and 2155 Psig, must be greater than or equal to the flow versus time relationship in Figure 3.1-1.

(4) During the operation of four reactor coolant pumps, steady state loop flows shall be within 2 percent of each other.

3.1.2 TEST METHOD Reactor coolant steady state flows were determined by means of the unit com-puter which calculated the coolant flow from the loop flow meter AP cells as follows:

3.1-1

=

Fm Cf (APm/Ym) is EQ. (3.1-1)

Where: Fa = Flow Rate At Measured Conditions, MPPH )

Cg = Flow Mater Coefficient, 407030

  • APm = Measured Delta Pressure, Psig Va " SP ecific Volume At Measured Conditions, Ft /lb. 3 For each pump operating combination given in Table 3.1-1 ten sets of steady state data was taken on the unit computer and the values averaged to determine the coolant flow rates. These results were then corrected to reference condi-tions of 532 F and 2155 Psig by using the equation below:

=

Fr Fm (Vm/Vr) EQ. (3.1-2)

Where: Fr = Flow Rate At Reference Conditions, MPPH .

Fm = Flow Rate At Measured Conditions, MPPH

, Vr =

Specific Volume At Reference Conditions (532 0 F), Ft 3 /lb.

i l

Vm =

Specific Volume At Measured Conditions, Ft 3 /lb.

The measured flow rates corrected to the reference conditions were then comp- .

ared to their respective acceptance criteria.

i Reactor coolant flow coastdown versus time was determined from temperature

! compensated square root extractor which solved equation 3.1-1. For each pump trip combination given in Table 3.1-1, steady state data was obtained. Sub-sequently, all or a portion of the operating pumps were tripped and data was recorded during the ensuing reactor coolant flow transient. Steady state data.

was again taken following the flow transient. The measured reactor coolant flows at various times during the coastdown transients were then normalized to their initial value, plotted versus time, and compared to the acceptance criteria.

Delay time analysis between unsnubbered and snubbered flow signal response was r

done for each reactor coolant flow coastdown. For each pump trip combination given in Table 3.1-1, data was collected for snubbered posit ~ons 3 and 4, and the time delay relative to the unsnubbered flow signal was < Iculated versus time. The time delay. at a flow reduction of the flux to flow ratio (1.044) was then~ determined and compared to the acceptance criteria.

3.1.3 EVALUATION OF TEST RESULTS Table 3.1-2 gives the minimum and ==rimum allowable flow rates for three dif-forent pump combinations, along with the measured flow rates for each listed condition. It can be seen that the measured flow rates are well within the acceptance criteria, and that the flow imbalance with four reactor coolant pumps was also well within the acceptance criteria of 2 percent.

  • This coefficient was changed after the test program to reflect -he final J
flow determination in Section 4.9.

3.1-2 1

7__ _ , . . . - _

.- : ,~

..-,.--._z _

Figure 3.1-1 shows the minimum acceptable reactor ecolant flow rate versus time for a four pump coastdown. Also included ar a tl.e measured reactor cool-anc flows versus time for the coastdowns of four, two and one reactor coolant pumps from an initial four pump operating condition. It can be seen that during the four pump coastdown the flow remained above the acceptance curve which was normalized to the steady state initial flow rate of 109.9% design flow.

Delay time analysis during the four, two and one reactor coolant pump trips from an initial four pump operating condition is shown in Figures 3.1-2, 3.1-3 and 3.e4 with the results tabulated in Table 3.1-3 for snubber positions 3 and 4. The data taken at snubbered position 4 indicate that the resulting time delay was well below the acceptance criteria.

3.

1.4 CONCLUSION

S The measured reactor coolant flow rate is approximately 109.9% of the design value. This flow provides adequate margin to both the maximt t and minf=2m .

allowable flow rates. This flow was preilminnry. A final f*.ow measurement was performed as part of the NSSS Heat Balance Test.

The measured reactor coolant flow rate versus time for a loss of four reactor coolant pumps was above the minimum allowable flow coastdown curve.

Snubber position 4 combined adequate filtering with 'a delay time response well within the acceptance criteria.

I l

i 3.1-3 l

l

I i

Summary Of Testing Done During The Performance Of Reactor Coolant l, Pump Flow And Flow Coastdown Test j

Test Pumps Running Pumps Tripped Data Number Al A2 B1 B2 Al A2 B1 B2 Reported General Comment j A. Reactor Coolant Pump Steady State Flow

{

l 1 X X X X Tab. 3.1-2 Normal Four Pump Flow '

2 X X X Tab. 3.1-2 Minimum Three Pump Flow.

T 3 X X Tab. 3.1-2 Minimum Two Pump Flow l

r B. Reactor Coolant Pump Flow Coastdowns.

H '

$ 4 X X X X X X X X Fig. 3.1-1 Loss of Four Pumps w

, e u 5 X X 'l X X X Fig. 3.1-1 Loss of Highest Flow Pump in Each g Loop During Four Pump Operation e

i "

6 X X X X X Fig. 3.1-1 Loss of Highest Flow Pump During Four Pump Operation l

C. Delay Time Analysis During Flow Coastdowns.

7 X X X X X X X X Tab. 3.1-3 Response of Snubber Position 3 Fig. 3.1-2 and 4 to the Loss of Four Pumps 8 X X X X X X Tab. 3.1-3 Response of Snubber Position 3 i

Fig. 3.1-3 and 4 to the Loss of Two Pumps 9 X X X X X Tab. 3.1-3 Response of Snubber Position 3 Fig. 3.1-4 and 4 to the Loss of One Pump I

/

! C -

C , _.)

^

Comparison Of Measured Reactor Coolant Flow Rates For Various Pump Combinations To The Acceptance Criteria At Reference Conditions of 532 F 0And 2155 Paig Minimum Maximum Acceptable Acceptable Measured Absolute Pump Flow Rate Flow Rate Flow Rate' Flow Imbalance Combination (106 1ba/hr) (1061bm/hr) (1061bm/hr) (%)

Four Pumps 142.1 153.3 148.7 'O.47

.-i

g. .

7 Three Pumps 105.8 153.3 111.7 NA F

T w

One Pump Each Loop 69.6 153.3 73.4 1.72 6

Note: Design core flow rate is equal to 135.3 X 10 1bm/hr at 532 F and 2155 Psig.

Note: The measured flow rates stated above are preliminary. A fintel flow measurement was performed as part of the NSSS Ileat Balance Test see Section 4.9.

t. -

i >

f

! Comparison Of Measured Time Delays Between Snubbered And Unsnubbered Reactor Coolant Flow .

Rates For Various Pump Combinations From Four Pump Initial Conditions of 532 F.

0 And 2155 Psig 4 t

Time Reactor Trip Time Delay Steady State Pump Trip Snubber Should Occur At Trip Noise Level ,

Combination Position (Seconds) (Seconds) (%)

3 Four Pumps 3 1.03 0.91. 0.67 '{

4 0.63 'l.01 H One Pump Each Loop 3 1.30 1.08 0.56 '

h 4 0.66 0.80 ,

i j

T' One Pump 3 2.59 1.98 0.71 '

4 0.99 1.01 t

i i

Note: The " Time Reactor Trip Should Occur " is defined as that time after the pump trip when the

nsnubbered reactor coolant flow rate has been reduced to a flux to flow ratio of 1.044.

i Note: The " Steady State Noise Level " is defined as the absolute percentage variation in the reactor 3 coolant flow rate at two standard deviations of the mean value.

! Note: The acceptance criteria on time delay for a two and four pump trip, and for a one pump trip is less than 1.00 and 1.25 seconds respectively. ,

i 4

L L ,

J

Measured Reactor Coolant Flow Rate Following The Trip i Of One, Two, And Four Reactor Coolant Pumps From Four l i Pump Initial Conditions of 532 F0And 2155 Psig I

,.- .. i -

~

Pump (s) rnn 2 *,

Tripped 7= -- - - __ -

=

=- -= ~ ~ n~: .. \

w t_ -

" ' - mJW,,

."_ ir-3 :12 w -w B1 M +_t  %  % "L_  ;

f.r = = _ :1 . 'hc D '. Ti' 2 -

t

-- E - + _=  !

w 1 Al, B1

=

& =--

g Four Pump Trip Flow -

'{ +_ __- _

g - ,, .

Coastdown Acceptance  :

's. zy . _,

l Crit'eria At 109.9% of 1.g  : Al, A2 l 3g Design Flow 71, .B1, B2 E - '\

g _: _

1._

$ . I 1 )

g t_w

&  ? '. .

l EE .2 _. -

= - -

, 3 _

.~. u. . 2 d.. .. . }

Time After Pump Trip, Seconds '

i 1

l Note: The Measure Flow Coastdown Data Is Taken From Snubber Position (4).

i l

Figure 3.1-1 1

. . . .. . . - . . . - . . - - . . - . - - . . ~ - - - . . - . . - ~

Measured Time Delay Between Snubbered And Unsnubbered Reactor Coolant Flow Rates Following The Loss Of Four Reactor Coolant Pumps From Four Pump Initial Conditions Of 532 0F And 2155 Psig RCP Trippedi 'A1]'A2,B1,B2' W .-. .

Snubber Position (3) - O Snubber Position (4) - A _

=-

827- 3; 8 o. !

y :=,. ~: ~

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j .i: v' c.-

- -' _' _r1

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F '~.=

,# _S_. =

~

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-n .=-

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Time After Pump Trip, Seconds Figure 3.1-2

Measured Time Delay Between Snubbered And Unsnubbered Reacter Coolant Flow Rates Following The Loss Of Two Reactor Coolant Pumps From Four Pump Initial Conditions Of 5320F and 2155 Psig

= = .

T*T RCP Trippedi X1TB1'!.___,

~2 E

'55uS6er Position (3) - O

-t 6

- Snubber Position (4) -- A .

,g ; _ ,.

c g .

O g-m- N.

a.

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8 u - a_ ,

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'= = 2C 1

9 E  ?

Time After Pump Trip, Seconds l

l l

l Figure 3.1-3  !

- . _ . _ _ _ . . __ . . . ~ . _. . . _ . . . . . . . . - _ _ . . ..

Measured Time Delay Between Snubbered And Unsnubbered Reactor Coolant Flow Rates Following The Loss Of One Reactor Coolant Pump From Four Pump Initial Conditions of 532 F And 2155 Psig

^-

ET RCP Tripped: .B1 . ____

--y --

~~

' Snubber Position (3) - O WF_

Snubber Position (4) -- A --

=- - --

03 MA - .

"J -

E y N2  :

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- . . 5 5 Time After Pump Trip, Seconds Figure 3.1-4

3.2 CONTROL ROD DRIVE DROP TIME TEST

' l 3.2.1 PURPOSE '

i l

The purpose of the Control Rod Drive Drop Time Test was to verify the function- l al trip capability of the Control Rod Drive System. This was donc by fulfilling the following test object 1yes:

(a) . To verify, for each control rod assembly, that the total elapsed drop time l from the initiation of the trip signal until the control red assembly was three-fourths inserted, was less than a predetermined limit at different unit conditions.

(b) To verify the repeatability of the drop time measurement by dropping the fastest and slowest control red assemblies an additional ten times each.

(c) To verify that the partial length rods (APSR's) cannot be tripped.

Two acceptance criteria are specified for the Control Rod Drive Drop Time Test as listed below:

(1) The individual safety and regulating control rod assembly drop times at zero and full flow, and 1 5250F is less than 1.400 and 1.660 seconds, respectively.

(2) The partial length control rods (APSR's) do not trip on a trip command.

3.2.2 TEST METHOD The Control Rod Drive Drop Time Test was performed using strip chart recorders to time the rod drops. Each control rod group was pulled to 100% withdrawn and then dropped into the core using the manual trip pushbutton. A zero time signal was furnished to the test recorders for each control rod assembly from a contact on the manual trip switch. A second signal to indicate three-fourths insertion was furnished to the recordets by a reed switch located on the posi-tion indicator tube of each control rod drive. The test was conducted at various combinations of temperature and flow at the following defined test conditions:

Test Condition Flow Temperature Mode 1 No Flow 0 1 280 F Cold Shutdown 2 One Pump Each Loop 1 2800F Cold Sh'utdown 3 No Flov 0 1 525 F Hot Standby 4 Four Pumps 1 525 F Hot Standby After the drop time measurements on all the groups were completed, the rods with the fastest and slowest trip insertion times at test condition 4 were tripped ten additional times to demonstrate repeatability of the measurement.

Measurements.were also performed on.the Group 8 control rods to verify that they did not drop into the core when power to the control rod drive trip breaker undervoltage coils was interrupted.

3.2-1

3.2.3 EVALUATION OF TEST RESULTS The results of the rod drop times for the four test conditions stated above are presented in Tables 3.2-1 and 3.2-2. As can be seen, the fastest dropped control rods were H-12, F-14, B-06, M-03 and E-13, and the slowest dropped control rods were C-11, M-09 and H-10 during the performance.

Analysis of the drop times showed that the slowest dropped control rods under zero and full flow conditions yielded drop times of 1.092 and 1.275 seconds respectively, which were well within the acceptan,ce criteria of 1.400 seconds for zero flow and 1.660 seconds for full flow.

Rod H-10 (the slowest) and E-13 (the fastest) at test condition 4 were dropped an additional ten times and produced rod drop times of 1.243 + 0.007/-0.009 and 1.262 + 0.008/-0.013 seconds, respectively.

3.

2.4 CONCLUSION

S The rod drop times were well within the acceptance criteria stated in section 3.1.3.4 of the Technical Specifications, of 1.660 seconds at full flow and greater than 5250F conditions. Also, as would be expected, all drop times under no flow conditions were shorter than under flow conditions.

s

. j

\.

3.2-2

. . . _ . - ~ - - . ._.

Control Rod Drive Drop Times For The Fastest And Slowest Control Rod At Each Test Condition r

Test Pumps Mean Drop Fastest Rod Slowest Rod Condition Running Time (sec) Location Time (sec) Location Time (sec) 1 0 1.099 H-12 1.080 C-11 1.125 2 2 1.167 11-12, F-14, B-06 1.145 C-11 1.195

-3 0 1.075 H-12 1.051 M-9 1.092 U

O 4 4 1.249 M-3, E-13 1.230 11 - 1 0 1.275 Y -

w Note: The fastest control rod (E-13) and the slowest control rod (11-10) at test condition (4) were dropped an additional ten times. The results are given below:

Fastest control rod time (sec) = 1.243 + 0.007 / -0.009 Slowest control rod time (sec) = 1.262 + 0.008 / -0.013 Note: The mean drop time is the average of all 61 control rod drive drop tJuca. As would be expected, the mean drop time increases with increasing reactor coolant flow.

Sumunary Of Control Rod Drive Drop Time Measurements Obtained During The S rformance Of Control Rod Drive Drop Time Test b

Core Rod Rod Rod Drive Drop Times In Secor de At Test Condition - _f, Position Group Number (1) (2) (3) (4) jl?

E9 1 1 1.100 1.180 1.085 1.263 -l .

Gil 1 2 1.110 1.170 1.078 1.266 -

K11 1 3 1.098 1.170 1.075 1.249 M9 1 4 1.110 1.180 1.092 1.272 M7 1 5 1.110 1.180 1.080 1.264 KS 1 6 1.120 1.170 1.083 1.256 G5 l~ 7 1.120 1.150 1.070 1.253 f'

E7 1 8 1.100 ,1,160 1.078 1.254 i I F8 2 1 1.095 1.170 1.080 1.265- g,

rio 2 2 1.090 1.175 1.080' 1.265 t

' a H10 2 3 1.100 1.185 1.080 1.275 S' L10 2 4 1.110 1.185 1.077 1.257 f $ L8 2 5 1.100 1.180 1.080 1.263 w L6 6 1.100 1.170 }

}

2 1.081 1.257 i y 11 6 2 7 1.090 1.180 1.080 1.270 i w F6 2 8 1.100 1.180 1.085 1.265 3 l i

C9 -

3 1 1.095 1.170 1.082 1.246 C13 3 2 1.100 1.170 1.085 1.245 K13 3 3 1.100 1.170 1.088 1.252 09 3 4 1.085 1.160 1.078 1.232 I, 07 3 5 1.100 1.170 1.085 1.246 I' K3 3 6 1.095 1.180 1.085 1.245 3

l G3 7 .1.090 1.160 1.077 1.245 C7 3 8 1.095 1.170 1.087 1.246 l 1

i B8 4 1 1.085 1.170 1.070 1.259 {

D12 4 2 1.095 1.175 1.072 1.259 i H14 4 3 1.100 1.180 1.072 1.262 i N12 4 4 1.085 1.165 1.070 1.250 i P8 4 5 1.095 1.180 1.082 1.269 f N4 4 6 1.090 1.170 1.070 1.250 [

11 2 4 7 1.085 '

1.170 1.070 1.267 D4 4 8 1.095 1.180 1.071 1.268 f .

[

1

c_

Summary Of Control Rod Drive Drop Time Measurements Obtained During The Performance Of Control Rod Drive Drop Time Test Core Rod Rod Rod Drive Drop Times In Seconds At Test Condicion Position Group Number (1) (2) (3) (4)

C11 5 1 1.125 1.195 1.087 1.264 C9 5 2 1.100 1.165 1.063 1.250 E13 5 3 1.105 1.160 1.067 1.230 M13 5 4 1.100 1.165 1.075 1.250 K9 - 5 5 1.090 1.165 1.068 1.252 011 5 6 1.100 1.165 1.070 1.243

-05 5 7 1.100 1.160 1.070 1.248 :

K7 5 8 1.100 1.160 1.057 1.241 I M3 5 9 1.085 1.155 1.055 1.230 g K3 5 10 1.110 1.170 1.081 1.265

$ G7 5 11 1.100 1.170 1.073 1.260

'7 d C5 5 12 1.100 1.170 1.070 1.258 D8 6 1 1.100 1.150 1.070 1.256 h Ell 6 2 1.110 1.170 1.078 1.266

,s H12 6 3 1.0 80 1.145 1.051 1.232 O

M11 6 4 1.f>20 1.160 1.062 1.252 k -

N8 6 5 1.100 1.155 1.070 1.248 a

M5 6 6 1.090 1.155 1.658 1.254 H4 6 7 1.100 1.170 1.079 1.263 ES 6 8 1.095 1.155 1.069 1.259 H8 7 1 1.100 1.160 1.075 1.255 B10 7 2 1.095 1.150 1.069 1.236 F14 7 3 1.095 1.145 1.069 1.243 L14 7 4 1.100 1.160 1.079 1.244 P10 7 5 1.110 1.155 1.073 1.245 P6 7 6 1.115 1.160 1.085 1.256 L2 7 7 1.110 1.160 1.075 1.252 F2 7 8 1.115 1.155 1.088 1.249 B6 7 9 1.100 1.145 1.062 1.242 i

)

3.3 PRESSURIZER TEST 3.3.1 PURPOSE Presaurizer operational testing was conducted after initial fuel loading and prior to initial criticality. The purpose was to set pressurizer spray and bypass spray flows at preacribed setpoints.

Two acceptance criteria are specified for the pressurizer test as listed below:

(1) The pressurizer spray flow must be set at 190 + 19 / -6 gpm.

(2) The pressurizer bypass spray flow must be set at 1.00 + 2.00 / -0.25 gpm.

3.3.3 TEST METHOD -

The technique used to set the pressurizer spray and bypass flows was based up-on balancing the heat input to the heat losses from the pressurizer. Initial steady state pressure and temperature conditions were established, in the pres-surizer without spray or bypass flow. The power input from the pressurizer heaters necessary to maintain steady scate conditions was recorded. The ad-ditional heat input required to balance spray and bypass flow was then calcu-lated using Equation 3.3-1.

F =

KVfs (Q - Qo)

Hfp-Es f EQ. (3.3-1)

Where: F =

Spray Or 3ypass Spray Flow, gpm K = Conversion Constant, 425.36 gpm BTU /Ft3 kW Vfs " Specific Volume Of Spray Water, Ft 3/lbm Q = Heater Input With Spray Flow, kW Qo

=

Heater Input Without Spray Flow, kW Hgp = Enthalpy Of Pressurizer Water, BTU /lbm Hfs = Enthalpy Of Spray Water, BTU /lbm Heat input to the pressurizer from the haaters was then increased by the amount calculated. The bypass and spray valve flows were increased to balance the additional heat input and maintain the pressurizar temperature and pressure at their initial values. An initial nominal pressure of 1300 psig was chosen for the spray flow measurement to ensure adequate heater capacity to overcome the 190 gpm spray flow and maintain steady state conditions.

3.3.3 EVALUATION OF TEST RESULTS The measured results from setting the pressurizer spray and spray valve bypass flows are listed in Table 3.3-1. The bypass and spray flows were set at 1.56 gpm and 190.15 gpm, respectively. The measured pressarizer heat loss was in excess of 169 kW at system conditions of 5320F and 2155 psig.

3.

3.4 CONCLUSION

S The pressurizer spray flow was set within the acceptance criteria limit of 190 + 19 / -6 gpm. The pressurizer spray bypass flow was set within the acceptance criteria limit of 1.00 + 2.00 / -0.25 gpm.

3.3-1

, ~n -, - - +,-

! I i' l' Measured Results For Determination Of Pressurizer Spray And Bypass Spray Flows ,'

i i

i l

. RCS COOLANT RCS COOLANT PRESSURIZER HEATER i CONVECTIVE SPRAY ACCEPTANCE VALVE TEMPERATURE PRESSURE TDiPERATURE POWER LOSSES FlhW CRITERIA POSITION (DEG F) (PSIG) (DEG F) (KW) (KW) (CPM) (CPM) (TURNS) j i I j A. Pressurizer Bypass Flos 5.easurement, RCV-149 I

! 534.0 2145.0 647.0 198.6 169.6 1.56 1.00 1/8 l

j +2.00/-0.25  ;

I i, f

f t

I r

i

.) .

.s i l y o

5 B. Pressurizer Spray Flow lleasurement (With Bypass Flow), RCV-14F

i w

! O 533.0 1303.0 580.0 1503.1 195.6 190.15 190 1j4 ,

+19/-6

,! I i i l

i i

l

  • s 4

1 4

~

I C o .

3.4 ZERO POWER PHYSICS TEST Zero power physics testing on Crystal River Unit 3 commenced with initial criticality on January 14, 1977. Part way through the testing, the FPC grid electrical system was impacted by a severe cold spell. The license was at that time restricted to 5.0: full power. The decision was therefore made to interrupt testing to save the approximately 30 MWe required to run the reactor coolant pumps and other plant equipment. Testing was finally concluded after the aforementioned eight day delay on January 29, 1977. The program which was intended to verify the nuclear design parameters of the core prior to escala-tion into the power range was performed at 532 0F and 2155 Psig.

Each section of the report addresses a specific test which was conducted during the zero power physics testing program. The analyzed results are presented and comparisons made to predicted values and Technical Specification limits.

The specific sections include the following:

Initial criticality -

Nuclear instrumentation overlap Sensible heat determination Reactivity calculations "ALL RODS OtTI" critical boron concentration Control rod group worths Soluble poison worths Ejected control rod worth Stuck control rod worth Shutdown margin determination Temperature coefficient of reactivity Temperature normalization constant 3.4.1 PURPOSE The Zero Power Physics Test was performed to verify the nuclear design para-meters used in the safety analysis, the Technical Specification limits, and the operational parameters. The purpose of the test was to direct testing at the 532 F and 2155 Psig plateau and included the following:

(a) To direct the initial approach to criticality (b) To measure selected reactor physics parameters at zero power following initial fuel loading and before escalation into the power range.

(c) To verify that source and intermediate range nuclear instrumentation have sufficient overlap.

(d) To determine the reactor coolant system temperature normalization constant.

Four acceptance criteria are specified for the Zero Power Physics Test as listed below:

(1) Measured values when compared to the predicted values shall agree within the tolerance bands given in Table 3.4-1.

3.4-1

(2) Shutdown margin with the measured vorst case stuck rod fully withdrawn shall be more negative than -1.00% ak/k.

(3) Measured worst case ejected rod shall not be in excess of +1.00% ak/k.

(4) Moderator ccafficient shall be less positive than +0.90 X 10"Ak/k/0F.

3.4.2 TEST METHOD Initial criticality was achieved by control rod group withdrawal and boron dilution of the reactor coolant system after the system had been heated up to hot conditions using the reactor coolant pumps. During the initial approach to criticality, plots of inverse multiplication versus boron concentratioa, rod group position, domineralized water added, and time were maintained by using the source range nuclear instrumentation channels NI-l and NI-2. After each rod group withdrawal and at every 30 minutes during deboration, the plots were updated and criticality was predicted. After reaching criticality, the nuclear power was increased and the source and intermediate range nuclear in- .

strumentation overlap was verified to be in excess of one decade. During the same increase in power, the point of sensible heat was determined and the cur-rent range under which Zero Power Physics Test would be conducted was estab-lished. Upon completion of this test, steady state conditions were obtained and the reactor coolant md hot leg temperatures were normalized to indicate a core AT of zero.

Measurement of the following physics parameters at 532 F sud 2155 Psig was theo begun:

3

)

(a) "All rods out" boron concentration -

(b) Temperature coefficient of reacitivity at three boron concentrations.

(c) Differential and integral rod worth of Groups 8, 7, 6, 5, 4 and part of Group 3 by boron swap.

(d) Total rod worth of part of Group 3 and Groups 2 and 1 by rod drop techniques.

! (e) Differential boron worth over large boron changes.

(f) Stuck rod worth by static rod swap techniques.

(g) Determination of the minimum shutdown margin.

i l (h) Ejected rod worth by static rod swap techniques.

Measurements of reactivity during this test were accomplished by the Babcock &

i l

Wilcox Reactineter which utilizes periodic neutron flux sampling from an inter-mediate range nuclear instrumentation channel to calculate reactivity. In the case of the Reactimeter, the neutron flux is sampled every 0.2 seconds and the reactivity result is updated after each sample. The Reactimeter can also sample as of ten as once every 0.2 seconds, the analog or digital signals representing ,

reactor and unit parameters other than neutron flux. This data logging cap- d ability was used extensively throughout the test program.

3.4-2 7_ .- _ - . . - - . . _ , . . . _ . . .

3.4.3 EVALUATION OF TEST RESULTS 3.4.3.1 INITIAL CRITICALITY Initial criticality was achieved on January 14, 1977 at 0845 hours0.00978 days <br />0.235 hours <br />0.0014 weeks <br />3.215225e-4 months <br /> at reactor coolant conditions of 5320F, 2155 Psig, and 1394 ppmB. Control rod groups 1 through 4 had previously been withdrawn as part of the heatup procedure with reactor coolant boron concentration at 1777 ppmB. The approach to critical began by withdrawing control rod groups 8, 5, and 6 to 100% and group 7 to 75%

withdrawn. The reactvr coolant system was then deborated until the measured boron concentration fell 50 ppmR below the original predicted value of 1521 ppmB. After an evaluation was made, deboration continued until initial criti-cality was achieved.

Throughout the approach to criticelity, inverse multiplication curves versus time, reactor coolant system boron concentration, and quantity of demineralized water added were maintained for the source range detectors by two independent persons. At the end of each control rod group withdrawal and every 30 minutes -

during the deboration cycle, the count rate was taken from each source range detector by way of the scaler-counters. The ratio of the initial average count rate to the count rate at the end of each reactivity addition was plotted. ,

, The approach used to obtain initial criticality is outlined below in three basic steps.

(1) Control rod group withdrawal:

4 Groups 1 - 4 already at 100% withdrawn Group 8 100% withdrawn Group 5~ 100% withdrawn Group 6 100% withdrawn Group 7 75% withdrawn (2) Deboration from 1777 ppmB at a letdown and makeup flow of 70 gpm until

, the plot of boron concentration versus time indicated a reactor coolant l system boron concentration of 1471 ppmB. At that time demineralized water addition was suspended; the reactor coolant system and makeup tank l were allowed to come to equilibrium (i.e. approximately 1440 ppmB); and i and evaluation was made:

(a) A lab standard was used to insure that the measured boron concen- l trations were valid. t (b) Rod position indication was checked to verify that all control rods were fully withdrawn except Group 7.

(c) The 1/M plots were checked for consistency and reasonability.

Criticality projections indicated that an additional 30-50 ppm boron reduction was required to reach criticality.

(d) Possible error in the predicted critical boron concentration was

, speculated. This value was later found to be in error and official-17 revised to 1434 ppmB.

3.4-3

- . - - .- ,. . ~ . _

_,_ _ . _ _ _ _ . _ _ m~.._,

O1 (e)~ Agreement between Babcock and Wilcox and Florida Power Corporation to continua de% ration was obtained after each reviewed the above results.

]

(3) Deboration 'from the above conditior. to, criticality at a me.keup and let-down rate of 40 gym. Control Rod Group 7 inserted to maintain critical-ity as required. Throughout both deboration steps a plot of reactor coolant system boron concentration versus time was maintained as shown e

in Figure 3.4-1.

The aoove procedure was followed, with initial criticality occurring at a reactor ecolant system boron concentration of 1394 ppmB. The plots of inverse multiplication versus time, reactor coolant boron concentration, reactivity, removed, dominera11 zed water added are shown in Figures 3.4-2, 3.4-3 and 3.4-4 a & b, respectively, and utilize the data from each of the source range channels.;

From these plots, the excellent response of the source range detector to sub-critical multiplication can be seen.

In summary, initial criticality was obtained in an orderly manner. Analysis of the measured critical-boron concentration of 1394 ppmB and the revised predicted value of 1434 ppmB indicated favorable agreement with the difference ascribed to a slightly larger measured differential boron worth at 5320F and 2155 Psig than predicted.

3.4.3.2 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 j range and intermediate range must be observed. This means that before the source range count rate equals 105 cps the intermediate range must be on scale.

If the one decade is not observed, the approach to the intermediate range can-not be continued until the situation has been corrected.

To satisfy the above overlap requirements after initial criticality was reached, core power was slowly increased until the intermediate range channels came on scale. Detector signal response was thus recorded for both the intermediate and source range channel. This was repeated for another decade to assist in determining the number of decades of overlap.

The results of the Nuclear Instrumentation overlap at 532 F and 2155 Psig have i

been tabulated in Table 3.4-2 and plotted in Figure 3.4-5. Framination of Table 3.4-2 shows that the overlap between the source and intermediate range is constant at an average value of 2.21 decades which is well above the =4n4- =

of one decade overlap stated in Technical Specifications. Examination of Figure 3.4-5 shows that the linearity, overlap, and absolute output of both the inter-mediate and source range detectors are within specifications and performing satisfactorily.

3.4.3.3 SENSIBLE HEAT DETERMINATION

-Determination of the intermediate range current level at which the production of sensible nuclear heat occurs is important to the Zero Power Physics Test program in that it establishes the upper power level limit. Thus by restrict-ing reactor power operation to a level reduced by a factor of three below the I

sensible heat level, the effects of temperature feedback are eliminated in the measurement of physics parameters.

3.4-4'

~ -~ - - . . . . _ _ . _ _ _ _ _-.

1he test method was to increase the 12termediate range current level in one third decade increments until the detection of sensible heat. Since turbine bypass valves and pressurizer levels were maintained constant, heat production in the core was observed by change in the core outlet temperature and makeup tank level. To ensure that the changes observed were caused by the core power production, the heatup rate was' plotted versus intermediate range current level.

The point at which there was a definite heatup rate was 6.75 X 10-8 amps.

Therefore, during the test 6.75 X 10-8 amps on channel NI-3 was defined as the

" sensible heat" point. From this the upper zero power physics test current limit was established at 2.25 X 10-8 amps.

. After setting the upper zero power physics test current limit, temperature feed-back measurements were performed to verify that the above limit was adequate.

This was done by adding a +20 pcm over a current range of 1.5 X 10-8 ca 3.0 X 10-8 anps and verifying that it remained constant. The results indicated that no temperature feedback effects were present over the current range indicated.

An additional part of the measurement was to ensure that all power range channels were indicating between 1.0 to 2.0 percent full power at the point of sensible

, heat. This was done to ensure that the overpower trip protection was adequate.

At the point of sensible heat, the indication on NI-5, NI-6, NI-7 and NI-8 were 1.3, 1.2, 1.2 and 0.8 percent full power. Thus, all power range channels except NI-8 were within their recommended tolerances. Since all channels were set near i their *=m sensitivity no change in the calibration of this channel was per-formed.

During evaluation of test results, additional evaluation was performed. Figure 3.4-6 shows these results. There was a very small heatup rate at relatively icw powers. Using this curve it is possible to establish a relationship between current and power. Thus, 5.4 X 10-8 amps is about 1 We and the .

2.25 X 10-8 amps used as the upper lielt is 0.4 Wt.

3.4.3.4 REACTIVITY CALCULATICNS s

Reactimeter is the name given to the Babcock and Wilcox reactivity meter which solves the one-dimensional, inverse kinetics equation with six delayed neutron groups for core net reactivity based upon periodic samples of neutron flux.

In addition to reactivity and neutron flux, the Reactimeter can also record 23 other analog and digital signals from the plant.

After initial criticality and prior to the first physics measurement, an on-line functional check of the reactimeter was performed to verify its readiness for use in the test program. After steady state conditions with a constant neutron flux were established, a small amount of negative reactivity was in-serted in the core by inserting control rod group 7. Stop watches were used to measure the doubling time of the neutron flux and the reactivity inserted was determined from period-reactivity curves. The measurement was repeated for several values of reactivity inserted by rod group 7, from 10.025 to 1 0.075% ak/k. The reactivities determined from doubling time measurements were

, then compared with the reactivities calculated by the reactimeter.

[ The results of the reactimeter verification measurements are summarized in Table 3.4-4. In each case except for the -0.075%Ak/k reactivity insertion ,

3.4-5

l l

, the reactivity calculated by the reactimeter was within the acceptance criteria limit of i 5% of the reactivity determined from doubling times. Since the reactimeter checkout did not meet the acceptance criteria at -0.075% Ak/k, reactivity additions were limited to 10.050% Ak/k for all measurements except rod drops during the remainder of the test program.

3.4.3.5 "ALL RODS OUT" CRITICAL BORON CONCENTPATION

, The "all rods e.t" critical boron concentration was measured at one isothermal

temperature t 3t plateau of 5320F. The measurements actually were made with rod group 7 partially inserted, but the measured boron concentration was adjust-ed to the all rods out condition using the results of rod worth measurement to )
determine the reactivity worth, in terms of ppm boron, of the inserted control I

rods.

The results are tabulated in Table 3.4-3. These results show that the measured boron concentration was within the revised acceptance criteria of 1445 +/-

, 100 ppmB. -

3.4.3.6 CONTROL ROD GROUP WORTHS

} The layout of the core according to the standard alpha-numeric mes,h showing the initial location of the control rod groups and the location of the 52 incore detector strings is given in Figure 3.4-7.- The number of rods in each

group and the reactivity control function of each group is listed below.

Rod Group No. No. of Rods Control Function 1 8 Safety 2 8 Safety 3 8 Safety 4 8 Safety 5 12 Power Doppler i 6 8 Power Doppler

' 7 9 Transient 8 8 Axial Power Shaping

, 69 Predicted beginning of life control rod group reactivity worths for the normal withdrawal sequence were supplied at two unit conditions of 532 F and 2155 Psig and 5790F and 2155 Psig, and at two APSR positions of 35 and 100 percent

withdrawn. These predictions were made using the PDQ-7 code with either a i two or three dimensional model of the core. Measurements of the beginning of life control rod group reactivity worth for the normal withdrawal sequence were determined during zero power physics testing at unit condition of 532 F and 2155 Psig, with APSR's at 35 percent withdrawn, using the boron swap and rod

! drop method. The boron swap method was used to determine the integral and differential worth for groups 8, 7, 6, 5, 4 and part of group 3. This method consisted of setting up a deboration rate and compensating for the change in reactivity by small step changes in rod group positions. The calculation of reactivity on a continuous basis is made by the Reactimeter. The output of reactivity in terms of percent milli k -(PCM) is recorded on a strip chart and also on a magnetic tape by the Reactimeter.

3.4-6 e ~ _. _ _.

The rod-drop method was used to determine the worth of the control rod groups not measured by boron swap. For this measurement, the reactor was adjusted to a critical state with all of the control rod groups to be measured out of the core and at a power level near the Zero Power Physics Test upper current limit.

. The control rod groups were then simultaneously dropped into the core and the neutron flux and resulting reactivity recorded by the Reactimeter every 0.2 seconds. An uncorrected value for the reactivity introduced by the rod drop

. was obtained by averaging reading between 20 and 52 seconds after the drop.

The corrected reactivity worth was then obtained by multiplying by a correction factor (1.28) to account for the change in the spatial flux shape during the rod drop. The correction factor was determined 8y plotting measured rod drop reactivity worth versus inserted group worth as measured by the boron swap method. The results of this analysis are shown in Figure 3.4-14.

t The results of both the predicted rod group worths and the measured group worths are tabulated in Table 3.4-5 for a moderator temperature and ' pressure of 532 F 2155 Psig respectively. The total worth of groups 1, 2 and 43% of group 3 was

, measured by the' rod drop method. The sum of groups 1 and 2 worths was then -

determined by subtracting the group 3 partial worth based on extrapolation of I

the boron-swap data. Finally, the individual worths of these groups were

, established by assuming that the percentage deviation between the measured and predicted value was identical on both groups (i.e. groups 1 and 2) . The results are presented in Table 3.4-6. Comparison between measured and predicted rod worth shows groups 5-8 agreed within -3.1 percent and groups 1-4 agreed within

-28.3 percent. The general acceptance for rod worths calls for analysis and resolution of measurements more than 20% different from predictions. Groups 1-4 did not meet this acceptance criteria. However, when the low worth of those groups is combined with the low stuck rod worth (3.4.3.9) an acceptable shut-down margin exists. Also presented in Table 3.4-6 are the expected control rod group worths at 5790F- and 2155 Psig with the APSR's at 35% wd. These values were obtained by applying the percent deviation between the measured and pre-dicted worths at 5320F to the predicted worths at 5790F.

The results of the measured and normalized differential rod worth shapes for i

control rod groups (3-4) and (5-7) at 5320F and 2155 Psig are plotted for group

(8) at 35% wd in Figures 3.4-8 and 2.4-9, respectively. As can be seen the transient, power doppler and safety groups all had similar shapes. The normal-ized differential rod worth shape for control rod group (8) was also measured and is shown in Figure 3.4-10. Normalized integral rod worth shapes were then i developed by integrating the above differential shapes and are presented in Figures 3.4-11 through 3.4-13.

3.4.3.7 SOLUBLE POISON WORTHS Soluble poison in the form of dissolved boric acid is added to the moderator to pt; vide additional reactivity control beyond that available from the control l rods. The primary function of the soluble poison control system is to control the excess reactivity of the fuel throughout each core life cycle.

The reactivity worth of the boric acid in terms of ppm boron was predicted by

! using both the one-dimensional LIFET Code and the multi-dimensional PDQ-7 Code as developed or modified by Babcock & Wilcox. The predictions were at three moderator temperature and pressure conditions of 300 0F and 800 Psig, 5320F and 2155 Psig, and 5790F and 2155 Psig.

3.4-7

- - - _-- , ----m , -

, , , , .,v -s v - . - . - - m- n. - - .., . --= ., =me--

1 e l l

Measurement of the s"oluble poison differential worth has been completed at 532 F and 2155 Psig. The measured value was determined by s==ning the incremental reactivity values measured during the rod worth measurements over a known boron '

concentration range from 1403 to 916 ppmB.

The result of the differential soluble poison worth measurement is tabulated in Table 3.4-7 and plotted in Figure 3.4-15 against the predicted value. This I measured value along with various reactivity balance calculations performed during the test program were then used to establish the integral boron reactivity worth curve shown in Figure 3.4-16. .

In sununary, the measured differential boron worth at zero power was within 5.3 percent of the predicted worth which is within the acceptance criteria of i 10 percent.

3.4.3.8 EJECTED CONTROL ROD WORTH Pseudo ejected control rod reactivity worth was measured at hot, zero power -

conditions of 5320 F, and 2155 Psig for control rod 7-8. The purpose of this measurement was to verify the safety analysis calculations relating to the assumed accidental ejection of the most reactive control rod which is fully inserted in the. core during normal power operation. The acceptance criterion for the ejected control rod test is that the reactivity worth of the most re-active control rod does not exceed 1.0%Ak/k at hot zero power conditions. Since the ejected control rod worth increases as more rods are inserted into the core, the measurement was performed at the minimum allowable rod po'sition of 45% wd on group 5 as specified by Technical Specifications.

}

The ejected worth of control rod 7-8 was measured by the rod swap method. The .

test method was to initially position Group 5 to between 45 to 55 percent with-drawn with the ejected control rod 100% withdrawn. Then the ejected rod and control rod groups 5 and 6 were swapped until the ejected rod was 0% withdrawn.

The change in Groups 5 and 6 positions was then converted to reactivity by using a previously generated rod worth to curve to determine the measured ejected rod worth.

- Since the measured rod position did not precisely equal group 5 at 45% with-drawn, the measured ejected rod worth was adjusted to the Technical Specifica-tions position by adding 0.02%Ak/k as determined using the equation below:

Where: p(ER) = F X p (R) + p (M) EQ. (3.4-1) i p(ER) = The measured ejected rod worth at the maximum allowable j inserted group worth.

F = Empirically derived constant of 0.45 which relates ejected rod worth and total inserted group worth.

p (R) ' = The initial worth of Group 5 minus the worth of Group 5 at 45% wd.

p (M) = The measured worth of the ejected rod which corresponds to the worth changes in Groups 5 and 6. *

)

3.4-8

. , - -7 , . .+ . -., #wg- .w,_,-e- . , , , , .

l

, To ensure that the above ejected rod worth was conservative, it was multiplied l by an uncertainty factor of 1.05. The results of the above analysis is given  :

in Table 3.4-8. l The predicted and measured ejected control rod reactivity worth at the Technical Specifications rod position of 45% wd on group 5 (i.e. inserted group worth of .

-2.95%Ak/k) are tabulated in Figure 3.4-17 which also gives the dore location of the ejected control rod relative to the out-of-core intermediate range detectors. The measured and worst case ejected control rod worth were deter- ,

mined to be 0.44 and 0.46%Ak/k, respectively. l l

3.4.3.9 STUCK CONTROL ROD WORTH l The control rod calculated to have the highest withdrawn worth with all ot'aer

groups fully. inserted is rod 4-3 in core location H-14, or those symmetric to

, it. On Crystal River Unit 3, a stuck rod worth of +4.00%Ak/k was predicced with the APSR's at 35% withdrawn.

The stuck control rod worth was measured during the Zero Power Physics Test program using the rod swap method. In this method c:.e stuck rod was swapped out of the core with control rods and boron until e critical state was achieved

. with the stuck rod fully withdrawn, the APSR's at 35% wd and the remainder of the control rod groups fully inserted. The stuck rod was then driven into the core and the reactor retaken critical using the normal withdrawal sequence hold-ing reactor coolant system horon ecustant. The amount of reactivity removea by the control rods to regain criticality is equal to the stuck rod worth. The worth of the control rod groups withdrawn from the core was then determined by the rod drop mettod.

The result of the stuck rod worth measurement is presented in Table 3.4-9.

The measured worth of the stuck rod was determined to be 2.34%Ak/k, which establishes a worst case stuck rod worth of 2.57%Ak/k (i.e. , measured value increased 10.0 percent as required by procedure for conservatism). Comparison of the predicted and measured stuck control rod reactivity worths are tabulated la Figure 3.4-18 which also gives the core location of the stuck control rod relative to the out-of-core intermediate range detectors. The large difference between the predicted and measured worth is attributed to the difficulty in PDQ-7 to predict individual rod worths in heavily rodded cores.

The measured stuck rod worth is below the acceptance criteria of i 30 percent of the predicted value. However, this is in the conservative direction.

3.4.3.10 SHUTDOWN MARGIN DETERMINATION Shutdown margin is defined as the instantaneous amount of reactivity by which the reactor is or would be suberitical from its present condition assuming:

(1). The worst case stuck rod is fully withdrawn (2) No change in APSR position (3) All control rod groups (safety and regulating) are fully inserted 3.4-9

. . - . - - . - - ~. - -- - --.__.- .. - - _ . . . . . - -~

l i

s Technical Specification section 3.1.1.1 states that the shutdown margin shall i not be'less than -1.00%Ak/k during all modes of Unit Operation after fuel is 3 loaded in the core. The predicted =4n4== shutdown margin during pit critic-  ;

ality at beginning of life conditions is projected to occur at 532 F and 2155 l j Psig with a control rod position of group 5 at 45% withdrawn. j l

I Upon determination of the necessary core physics parameters, as measured at l

5320 F and 2155 Psig during the Zero Power Physics Test, a calculation to deter- 1 mine the minimum shutdown margin was performed at the condition specified above
as shown in Table 3.4-10. The results using the method ' called out in the test i procedure indicate that a conservative minimum shutdown margin of -2.93%Ak/k was present which adequately satisfies the acceptance criteria of -1.00%Ak/k. J A review of the test results by Babcock and Wilcox revealed that the procedure applied an uncertainty to the stuck rod worth in two different steps of the procedure. Elimination of this double uncertainty yields a minimum shutdown ,

margin of -3.39%Ak/k. '

j 3.4.3.11- TEMPERATURE COEFFICIENTS OF REACTIVITY f

The temperature coefficient of reactivity is defined as the fractional change in the excess reactivity of the core per unit change in core temperature. The l temperature coefficient is normally divided into two components as shown in Equation (3.4-2).

" T =M M + " D

{ EQ. (3.4-2) r Where: " T =

Temperatua Coefficient of Reactivity

" M

=

Moderator Coefficient of Reactivity D = Doppler Coefficient of Reactivity The technique used to measure the 532 F and 2155 Psig isothermal temperature j

coefficient at zero power was to first establish steady state conditions by i

maintaininr reactor flux,~ reactor coolant pressure, turbine header pressure and core average temperature constant, with the reactor critical at approximately 1 X 10-8 amps if a negative feedback effect was expected from a temperature i decrease or at approximately 1 X 10 ? amps if a positive feedback effect was expected. Equilibrium boron concentration was established in the reactor coolant system, make-up tank and pressurizer to eliminate reactivity effects due t to boron changes during the subsequent temperature swings. The reactimeter and I the brush recorders were connected to monitor selected core parameters with the l reactivity value calculated by the react 1 meter and the core average temperature

) displayed on a two chanr.d recorder.

l Once steady state conditions are established, a negative heatup rate was  ;

started by opening the turbine bypass valves. As the reactivity changed about l

+ 40 PCM, control rods were moved in step changes to keep the reactivity values on scale and also to keep core power in the range necessary to produce an ad-equate intermediate range signal. After the core average temperature decreased by about SoF coolant temperature and reactivity were stabilized. This process  ;

l was then reversed except that the core average temperature was increased by }

n/ l l about 100F. After stabilizing coolant temperature and reactivity, the core '

average temperature was then returned to its original value. The measurement l

!- 3.4-10 l

^ Ah* en-+u.m=.i.-- .e _ , _

of the temperature coefficient from the data obtained was then performed by dividing the change in reactivity by the corresponding changes in core temperature i over a specific time period. .

Isothermal temperature coefficient measurements were conducted at three different reactor coolant boron concentrations during the Zero Power Physics test program.

The results of the measurements are summarized in Table 3.4-11 along with the predicted values which are included for comparison. In all cases the measured results compared favorably with the predicted values. All measured temperature coefficients of reactivity were within the acceptance criteria of i 0.40 X 10 4 Ak/k/9F of the predicted value.

The moderator coefficient cannot be directly measured in an operating reactor because a change in moderator temperature causes a similar change in the fuel temperature. However, since the moderator coefficient has safety implications, it is an important reactivity coefficient. As specified in Technical Specific-tions the moderator temperature coefficient shall not be positive at power levels above 95 percent full power and shall,be less than +0.9 X 10-4ak/k/ F at all other power levels. To obtain the moderator coefficient from the measured temp-erature coefficient, a Doppler correction of 0.6 X 10-4ak/k/ F must be added.

Determination of the moderator coefficient hae been completed and shows that this coefficient is well within the limits stated above.

In conclusion, all temperature coefficients of reactivity that were measured at the 532 F and 2155 Psig plateau were within the acceptance criteria of i 0.40 X 10-4Ak/k/0F of the predicted value. In addition, calculation of the modera-tor coefficient indicates that it is well within the requirements of the Tech-nical Specification 3.1.1.3.

3.4.3.12 TEMPERATURE NORMALIZATION CONSTANTS Isothermal temperature normalization constants on the reactor coolant system cold and hot leg were determined as part of.the Zero Power Physics test program.

The intent of this measurement is to normalize the AT across the core to zero

.and to verify that these temperature indications are reasonable. The impor-j tance for normalization lies in the fact that the AT across the core is directly related to core thermal power. Thus, if normalization is performed, this quantity can be used during power operation for power level determination.

Measurements of the normalization constants for the cold and hot leg RTD tempera-ture indication on the primary system were performed at 5320F and 2155 Psig, and are reported in Table 3.4-12.

In all cases only minor corrections were necessary with the maximum deviation from the average of all RID readings of -2.20F.

3.

4.4 CONCLUSION

Zero Power Physics Test commenced with initial criticality on January 14, 1977 and continued 5 days until a severe electrical shortage due to cold weather caused it to be interrupted. Testing resumed on January 27, 1977 and was success-fully completed on January 29, 1977. The program which was intended to verify ,

the nuclear design parameters of the core prior to escalation into the power i~

range'was performed at 532"F and 2155 Psig. The Zero Power Physics Test was conducted with good agreement between measured and predicted'results on all physics parameters except the reactivity worths of groups (1-4) and the stuck rod. 'A' summary of each of the measurements performed during the Zero Power

! Physics Test is given in Tabla 3.4-13.

I I-l:

, 3.4-11

)

Acceptance Criteria Deviation Limits Between Measured And Predicted Values Allowable Deviation Between Core Physics Parameters Predicted And Measured Values A. Control Rod Worths Group Worths (1 - 8) i_20%

e

.h O

Total Worth (1 - 8) i 15%

7 Ejected Rod Worth i 30%

Stuck Rod Worth i 30%

B. Temperature Coefficients 4 i

1 0.4 X 10 Ak/k/0F C. Differential Boron Worth i 10%

! D. All Rods Out Boron Concentration i 100'ppmB

- - - - _ _ ~ _ - - - _ - - -

)

t i

Summary Of Nuclear Ins.trumentation Overlap Measurements During Zero Power Physics Test ~ l i

i Source Range Indication Intermediate Range Indication Average SR Average IR Data Indication Indication Overlap -

Set NI-1 (CPS) NI-2 (CPS) NI-3 (AMPS) NI-4 (AMPS) (CPS) (AMPS) (Decades)

.j 01 8.0 X 103 8.0 X 103 1.1 X 10-11 1.3 X 10-11 8.0 X 103 1.2 X 10-11 2.18 02 8.0 X 104 8.0 X 104 1.2 X 10-10 1.5 X 10-10 8.0 X 10 4 1.4 X 10-10 2.23 ,

Average.= 2.21 5?

tr i y F

T p

{

l Note (1): Overlap is obtained between the Source and Intermediate ,

lJ Range by using the average indications in the equation j below.

})

t Overlap = ( 6 - Log SR ) + ( Log IR + 11 ) {l l

I s .

All Rods Out Critical Boron Concentration (PPM Boron)

Moderator Predicted Measured Temperature Results Results o

532 F 1445 1403 (1) .

H E-

.y*

T w

Note (1): This number is based on the as measured boron concentration with a 2 ppmB correction to account for the fact the rods were not quite at 100% withdrawn.

b

,( -j j Comparison Of Reactimeter And Doubling Time (DT) Reactivity Measurements i ll - ,

ii ' i

-! I J DT Converted Reactimeter - Absolute Case DT Reactivity Reactivity Error

! No. (Sec) (% A k/k) (% O k/k) (%)

1 i

1 228.0 -0.029 -0.028 3.6 I.

u 2 219.4 +0.024 +0.025 4.0 '

u 3 147.7 -0.047 -0.046 2.2 4 96.3 +0.049 +0.047 4.3

?' E?

1

-$a .5 102.7 -0.078 -0.073 6.9 6 50.3 +0.082 +0.079 3.8 t k L. <

l 1

NOTE: The Absolute Error Between DT Converted Reactivity And Reactimeter Reactivity Is +

Defined By The Equation Below: {

I E (%) =

100 (f R -

f DT) / (.9 R)

>. NOTE: Since The React 1 meter Checkout Failed To Meet The Acceptance Criteria At -0.078 % A k/k, l Reactivity Additions And Insertion Were Limited To + 0.050 % 4 k/k for all measurements except rod drops during the remainder of the test program. l 4

i j

i f' b ( , _.)

Comparison Of Predicted And tkasured Control Rod Group Reactivity Worth A.. Moderator Temperature at 532 F. APSR's at 35% Withdrawn Predicted Measured Percent (1)

Rod Number Worth Worth. Deviation

- Croup Of Rods __ (ZAk/k) _( %Ak/k) (%)

1 8 -1.05 -0.80 -31.8 2 8 -3.20 -2.43 -31.8 3 8 -0.75 -0.59 -27.1 4 8 -1.75 -1.43 -22.4 5 12 -1.16 -1.14 -1.8 6 8 -1.27 -1.21 -5.0 7 9 -1.17 -1.11 -5.4 g 8 _J! -0.36 -0.36 -0.0

g. Total 69 -10.71 -9.07 -18.3 5

u,' B. Moderator Temperature at 579 F, APSR's at 35% Withdrawn

  • t f, Predicted Estimated (2) Percent (1)

Rod Numbe r Worth Worth Deviation Group of Rods (%Ak/k) '(%Ak/k) (%)

1 8 -1.38 -1.05 -31.5 2 8 -3.69 -2.80 -31.8 3 8 -0.77 -0.61 -27.1 4 8 -1.55 -1.27 -22.4 5 12 -1.47 -1.44 -1.8 6 8 -1.43 -1.36 -5.0 7 9 -1.13 -1.07 -5.4 8 _Jl, -0.42 -0.42 0.0 Total 69 -11.84 -10.02 -18.3 NOTE (1): Percent deviation is determined assuming that the measured values are correct'.

NOTE (2): Estimated worth is determined by applying the percent deviations between predicated to measured results at 532 F to the predicted results at 579 F.

9 0

1

1:

j--

.  ?

.,5 i 1

[ Determination Of The Remaining Worth Of The Safety Groupa Not Measured By Boron Swap Using The Rod Drop Hethod

^

., f  !

t Measured Corrected To Position Predicted Rod Drop Boron Swap Absolt'.:e Rod Group Interval Worth Worth Worth Error Number (%wd) (%Ak/k) (%Ak/k) (%Ak/k) (%)

a j- 1 100 to 0 -1.05 -0.80 F l

i l ~

2 100 to 0 -3.20 -2.43 -

i .

9 3 43 to 0 -0.56 -0.42 O

tr

. TOTAL -4.81 -2.85 -3.65 -31.8 T

i l

Note: The Boron Swap Worth Is Equal To 1.28 Times The Rod Drop Worth. I This constant (1.28) Was Established From Figure 3.4-14 And Relates The Rod Drop Worth To Boron Swap Worth, i .

1 -

A

Differential Baron Reactivity Worth Heasurements During Zero Power Physics Test Rod Position, % wd Measured Average Delta Boron Differential Worth %Ak/k/ppmB Boron Conc. Boron Conc. Boron Conc. Worth 1 2 3. 4 5 6 7 8 (nom) (nom) (pom) '(%Ak/k) ' Measured Predicted 100 100 100 100 100 100 100 100 1403 1160 487 5.48 0.0113 0.0108 100 100 35 0 0 0 0 35 916

~

i U

E 4

v' 4

Note
The linear least squares best fitted differential boron worth using all steady
state data points yielded a values of 0.01142%Ak/k/ppmB.

[

5 i

l i

Determination Of The Ejected Rod Worth At Zero Power, BCu, And 532 F Conditions 4

Inserted RCS Boron i Rod Position, %wd Ejected Rod Change In Ejected Rod Worth,%Ak/k i Reactivity Joncentratior Position Reactivity 1 2 3 4 5 6 7 8 (%Ak/ k) (PpmB) h'asured Worst Case

(%Ak/k) (%Ak/k) 100 100 100 100 51 0 0 35 2.89 1181 100 0.42 0.44 0.46 i

i 100 100 100 100 100 16 0 35 2.47 0 l

l*

.g

.e 5

F.

7 oz

(

)

I Note: The measured ejected rod worth has been normalized to the Technical Specifications minimum allowable rod index of 45% wd by the addition of 0.02%Ak/k.

i i

Note: The worst case ejected rod worth is obtained by application of an uncertainty factor of 1.05.

1 i

E i t i V C ,

J I

Determination Of The Stuck Rod Worth At Zero Power, BOL, And 532 F Conditions 1-Rod Position, ~4wd Withdrawn RCS Boron Stuck Rod Change In- Stuck Rod Worth,(%Ak/k)

Reactivity ;oncentratiot Position Reactivity 1 2 3 4 5 6 7 8 (%Ak/k) (PPmB) (%wd) (%Ak/k) Heasured Worst Case 15 0 0 0 0 0 0 0 0.10 819 100 2.34 2.34 2.57 0 34 0 0 0 0 'O 0 2.44 o Mi E

o la

+

Note: The measured stuck rod worth was obtained by dropping group 1 and part of group 2 from the above conditions and by adjusting this value for the amount of reactivity withdrawn at the start of the measurement.

(SR) =

(1.28) X (1.91) - 0.10 =

2.34%Ak/k Note: The worst case stuck rod worth is obtained by application of an uncertainty factor of 1.10 4

i 4

4

l.

4 l-I Determination Of Minimum Shutdown Margin 1.

The Reactor must have a minimum reactivity shutdown margin of -l% Ak/k at all times with the k highest reactivity worth control rod stuck in itc most reactive position of 100% withdrawn. '

i l

1.0 Measured Rod Worth available, if trip occurred at the maximum allowable inserted group worth (i.e. Group (5) at 45% withdrawn). t Group 1 (at 100% wd) =

-0.80%Ak/k i

Group 2 (at 100% wd) = -2.43%Ak/k a

h Group 3 (at 100% wd) = -0.59%Ak/k o

e

{ Group 4 (at 100% wd) =

-1.43%Ak/k

~

by Group 5 (at 4_.% wd) =

-0.86%Ak/k TOTAL =

-6.ll%Ak/k ,

The above value should be reduced by ten percent (10%) for measurement uncertainties -5.50%Ak/k 2.0 The worst case stuck rod worth The above value is obtained by increasing the measured stuck rod worth by ten 1 percent (10%) for measurement uncertainties +2.57%Ak/k {

3.,0 The minimum shutdown margin (SUM of above terms) -2.93%Ak/k

+

NOTE: The shutdown margin should be more negative than -1.0%Ak/k f f

i, r

i -V ( ,

J  !

2 - cry Of Measured, Calculated, And Predicted Coefficients Of Reactivity At Zero Power i Rod Position, %wd Average RCS Boron Reactivity Coefficients. 1.0E-04 Ak/k/"F Temperature Concentration Temperature Coe f ficient Moderator Coefficient 1 2 3 4 5 6 7 8 (der F) (ppmB) Measured Predicted Calculated Predicted 100 100 100 100 100 100 86 100 532 1400 +0.36 +0.21 +0.52 +0.37-100 100 100 100 07 0 0 35 532 1070 -0.61 -0.69 -0.45 -0.53 100 100 35 0 0 0 0 35 532 916 -0.64 -0.48 -0.48 -0.32 Y

a ta l **

Note: The Predicted Temperature And Moderator Coefficient Were Adjusted For The Difference Between Predicted And Actual All-Rods-Out Critical Boron Concentration (-0.17 X 10-4 Ak/k/ F).

Note: The calculated moderator coefficient has been determined from the measured temperature coefficient of -0.16 X 10-4 Ak/k/ F.

4 5

  • 1

Reactor Coolant System Cold And Hot Leg RTD Temperature Normalization Constants

-i Temperature Computer Normalization Percent Unit Parameter Transmitter Point ID Constant CN Error i (Den. F) (Y)

RCS LP (A) Hot Leg Temperature RC4A-TT4 R212 +1.6 +0.30 i

l RCS LP (f) Hot Leg Temperature RC4A-TTI R212 +0.8 +0.15 i i

j RCS LP (B) Est Leg Temperature RC4B-TT4 R213 -2.1 -0.40 RCS LP (B) Hot Leg Temperature RC4B-TT1 R213 -2.2 -0.41 7 RCS LP (A1) Cold Leg Temperature RC5A-TT1 R214 +0.6 +0.11 4 RCS LP (A2) Col'd Leg Temperature RCSA-TT3 R215 +0.5 +0.09

?r-y RCS LP (B1) Cold Leg Temperature RCSB-TTl R216 +0.5 +0.09 RCS LP (B2) cold Leg Temperature RC5B-TT3 .R217 +0.5 +0.09 ,

i f

Comparison Of Measured And Predicted Results Obtained During Zero Power Physics Test

\

Physic Parameter Unito Measured Predicted Acceptance Reeutremente Comparison

1. All Rode Out Critical Boros ppas 1403 1445 Heasured value must be within 100 OK (42 ppm?)

ppab of predicted value

2. Sensible Heat amps 2.25 x lo-a yo ,
3. NI Overlap decades 2.21 Creater than 1.0 decade OK ( > 1.0)
4. Control Rod Group )forth I A k/k Percent deviation of group worth Croup 8 -0.36 -0.36 should be less than 20I. OK (0.02)

Croup 7 -1.11 -1.17 OK (-5.4%)

Croup 6 -1.21 -1.27 OK (-5.02)

Croup 5 -1.14 -1.16 OK (-1.81) g Group 4 -1.43 -1.75 NOT OK (-22.41) b Group 3 -0.59 -0.75 NOT OK (-27.11).

O Croup 2 -2.43 -3.20 NOT OK (-31.82)

Croup 1 -0.80 -1.05 NOT OK (-31.82) w

, T0TAL -9.07 -10.71 Percent deviation of total worth a should be less than 151 NOT OK (-18.1%)

w 5 Shutdown Margin I Ak/k -2.93 Measured value must be less than

-12 A k/k OK ( -1.0)

6. Stuck Rod Worth (4-3) I A k/k -2.34 -4.00 Percent deviation of stuck rod worth should be within 30% of the predicted value. NOT OK (-70.92)
7. Ejected Rod Worth (7-8) I Ak/k +0.4/ +0.47 Percent deviation of ejected rod worth must be' within 301 of the predicted OK (-6.8%)

value. Also value must be more OK (<+1.0)

"*8* * "" * *

8. Temperett.re Coefficient 10-4 Ak/k'F At 1400 ppaa +0.36 + 0. 21 Hessured value must be within 0.4 OK (.15)

At 1070 ppas -0.61 -0.69 X 10-4 A k/k/8F of the predicted OK (.08)

At 916 ppen -0.64 -0.48 value. OK (.16) e

I

\i +

', 'j

{ Comparison Of Measured And Predicted Results Obtained During Zero Power Physics Test  !

t; j

!i 1

[.  !

i f Physic Parameter Units Measured Predicted Acceptance Requirements Casparison l [

l' I

'! 9. Moderstor coefficient 104 A k/k/*F Maximum positive moderator coefficient

') must be less than 0.9 1 10-4 A k/k/*F ,  ;

. At 1400 ppmR +0.52 +0.37 og ( <.90) )

+ At 1070 ppmB -0.45 -0.53 OK ( c.90)

d. At 916 ppmR -0.48 -0.32 , OK ( <.90)

II

10. Differential Boron llorth
' I A k/k/ ppat At 1160 ppmB .

.j -0.0113 -0.0107 Percent deviation must be less thaa OK (-5.6) }

H 10%  ;

h i

.l 0 I id i

  • t &  !

l t

, H t,a i

! m M Es Percent Deyfation is defined by the following formula I

! n  !,

g ,

i rt, 1 Deylation = Measured Value - Predicted Value X 100

.? g, Measured Value t

i I

i I

1 3  ;

I C ( ,

m ) I

RCS Boron Concentration Versus Time For Approach To Initial Criticality

} l{: I

.)'

Tl '

L

-l Measured RCS Boron I

^

IConcentration - 0

! i

41. .

y  ; -)' O t'l ,,

,i sk ,

p.- 1,,g.

l

@ .- 4 w [-

L,,

g

. d i

-i

@ U .

4 T 8 et u 1.i

@ ll l

u. ' '

. Stopped Demineralized Water 7, hri lll Addition At 0615 And Started H

g.

b

.Taking Pressurizer Samples m

0 4

.i.

gg _ i!

l.' <

, 4 ll .

!Deboration Restarted At 0750 "

f

~

..!.l.l..L f I

I" 4l i

..i , ,

11:!O:. UJI:':1 a:4 :i:l 5

l:1 , Mi[I :l[I :l!llnl (j! Illi) d :l .:l l O I . l. . .l l J l TIME (EST),llours i

c l

__ ___________ ___ - - - - - - - - - - - - - ^ ~ - ' ^ ^ ^ ^ ^ ^^

4 Inverse Multiplication For Detectors NI-l And NI-2 Versus Time For Approach To Initial Criticality i

! I

" o a ,.:> ,

i  !

NI-l Data Point - a  !

u NI-2 Data Point - = l l

y ll l

R "

, , l

' " e c8 h l U

w il ,

y y 4 . l, >

'E n -

4 a i U

d }

u # ii  :

, y4 ,

j b' e il j ,!

y ,, ' Criticality At 0845  !

E f M

t: , i ql -

.In l- .

' ' 'I o o ,,'

-; ii u i h li i, '

4] , 4.' .

h -

4:.3 :l:l i; '

l[I :i .[ l :i i i pi  ;, ; , ;,

Time (EST),llours fI

!I L t J

Inverse Multiplication For Detectors NI-l And NI-2 Versus RCS Boron Concentration For Approach To Initial Criticality i l l' l l

ii i 4.m il ,

1 NI-l Data Point -- e I ir NI-2 Data Point - +

a u -

e ,.,..

,7 S a

0 h -

-l g M-a' t .

r-n Projected Critical Boron ' n '~

" Concentration At 1394 ppm

-[_

w a l l h

, ya <i-.- ,

f, e &

O. ,

y e A x !

l 'li p i a 1 H . . -

1-

, ti D: .i . . .: ' :1 D S 41 .

1 L 1 i RCS Boron Concentration, ppm

?:

t i

1 Inverse Multiplication For Detectors NI-1 And NI-2 Versus Reactivity Removed For Approach To Initial Criticality -

1 l

1 fb I  !

u [

T1

, u i [ i NI-l Data Point -- m j ll NI-2 Data Point -- * ,

!- i

.h s

6. .n. .

f o ,

g n

g. II ,

m Il li

<l, j

n o-

$N H r1 : ll i

.I' i

l A p' w M~ g i i y j--II ,

, t ir

$ *' ll -

I B

e .l I

G i i , ll .

. L ll Note: Groups (1 - 4) Were Withdrawn Prior To lleat-up.' ji

... i

'Ir I

I ...

.:1. II -

O

l i.

!'i !I i i !I ll< !I.  !!< fi ).ll ,

l .[L ,l l! ,,5 (( l , .[

U' .N S ..

Reactivity Removed, % A k/k '

48 5 6 7

-e . .  :  ;

r i

~

Group Withdrawal -

Deboration  ;  ! l l 't .

b  !

Inverse Multiplication For Detectors NI-1 And HI-2 Versus Domineralized Water Added For Approach To Initial Criticality l

l

-,l ,- ;i.. ,

' isn .

, Data Point -- x :

.N1-1 '

[ *I

'HI-2 Data Point d .

.,\,.

'{ .

I IF

,84 ,

'f

}

p g 4 i "- .

  • 2 a a

et 4 o p u u Il S. -

i 4 x - -

Il o' .

$. ll g

11 ll n1 ,\

'll a, ..I .

bI I

--i l

!d '
0 < < ! 'O 1;! : ,:i

, i  ; !

n -; , ..:q 3 , ;,5 g g ,

Demineralized Water Added, Callons

- .. - . . . - ~ . . . . - . . -. . . - - . . . ...

Core Power Versus Detector Response

  • m Intermediate Range Detector Response, amps 10-11 10-10 10 -9 10-8. 10-7 10-6 10-5 10-4 8

10 , . , ,

, . iii > .

Core Power Det'er b d By Using Sensible Heat '

[

10 7

Experimental Heat Up Rate. ,

(1 X 10+6 W = 3.6 F/hr) ,-

, / l 10 6 jd s ___

p s

f verage Intermediate 105 ,/ Range Channel At 532 F, I e 2155 Psig And 1377 PpmB

/

/ J l e l'04 ,

! s l

y / >

a / /

6 6 ,

U 3 10 3

l .

/ 1

)

o / 2 r i k / a

/ u i P. M Y l 0 10 2 / ' Average' Source' Range' ___

i .e Channel At 532 F, 2155 :::

i

/ Psig And . 377 ppmB ---

e l l 101 ,

s' 10 0 , ,

e . . ,

/ $ I 4 l

/ I i I I 1o-1 ,- , l

/ 1

{ t I  ;

r 10

-2 l- l 100 10 1 102' 103 104 165 106 107 Source Range Detector Response, counts / seconds .

Figure 3.4-5

Reactor Coolant System Heatup Rate Versus Intermediate Range Current Level For The Determination Of The Point Of Sensible Heat.

_ II - _-

l a

.. L- f p_+e - -

g . . . _ f g y " RCS Heatup Rate - A ^.'

9' ..-^,

E mr- l

,l e

.ZT .-

l U

f-:. ^

j'.

a. .

3 .7

%==*t-, l 2

i 0*.e---

os .E f.

?

~

.._=.

/

f

~

2 = =

_ -. '?

Average Intermediate Range Current Level, 1 X 10-8 g ,

1 0

Figure 3.4-6

,.%...4 ew.,5 c+-6+e w=- - 9M b + " 'O ^ ' ' " * # * ' ' '

Cycle 1. Core Locations Of Control Rods, Incore Monitor Asse$blies, And Out-Of-Core Detectors.

3 x

1 2 3 4 5 6 s to 7 9 11 12 15 14 is

' i i i i i i i S NI-t g st-3 g

$ NI-5 CRJ-9 Ch' C27-2 9 N1-7 3 C15-1; b

CR3-E 3%

FM CRS-:

lR4-8 CR8-d 22 h -C l

i

_ CR6-1 C'JLa] :14-2 -

D 32 22 . Sh j ;RS-lc

_ CR6-8 C -8 2- C?JL." CR5-3 j 34j 7 ' R 26; R7-8 CR6- ,

CR2-8 _ C'A' CR2-2  : .h2. CR7-:

63 43 ~y 2Q

! CRJ-J c' -

'_ , , _ CRS-1 .

C'X C'- C"J.a2 i 32 8J l'J ' ~C M2J .

, y _ :14-/ CRb-i _

R2-1 C' - _ CR2-.
Rb-3 CR4 s 4

37; 102 2J E Y l1 CR.3-6 :3 -

CAS-d CAS-t C '

_ CRJ-J l l' .

L9 20j -- E

R M CR8-6 CR M CR2-5 22-4 _ .Rd-3 CR7-4 33 IL L8j .

""" E CRM CR6-4 CF ' 9 C' -

CR6-4 :15-4 -

>0 2 L6 1] 69; M Cuct R8-5 C b::,1 , :R8-4 CR4-4

> 1; 1]g L5J N CB,2 i 33-5 CR.3-4 CR5-6 -

f 4EA

R Z
.1. CR4-5 f

CR7-5 hPJ y

G NI-8 2

$ st.6 R

' 42 4 52

$,NI-4 ,

  • $ N1-2 Z

2 -

Total Core Monitor '

2 -

Total Core And Symmetry Monitor Syumstry Monitor z -

Incore Monitor Number Control Rod Number Y of Control Rod Group I CRX-Y -

Figure 3.4-7

_ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __. n._ __

Normalized Differential Shapes Of Control Rod Groups 3-4 Versus Withdrawal Position, For Beginning Of Life, Zero Power Conditions

'=

~~

Group (3)

_, ___ Data Point - o r- A Group (4) Data Point - y

[-

1-t i _

i- 5' w

m _

6, i-

- r d .

3 T-  ;

i@

w v.

2a _

E ___ .bf-3 .

a2 .

e

= '

o N

'1 H- +g_ i e

c.

T-e "

.c -

m' ,

i__

gg- i-4 -

8 u .. . =-

c - - -- .

OU m-e ~ _~

44 0

='t~K i-i_

.___ _.___ _ _ _ ___ m

- :5 ':E- SO 2p IG)

Group Position, % Withdrawn Figure 3.4-8

- . . . . ~ - . . . - . . . . - . . - . . . . . . . . . . .

Normalized Differential Shapes Of Control Rod Groups 5-7 Versus Withdrawal Position, For Beginning Of Life, Zero Power Conditions ,

" " ~ ' -

?=. .

i lk Group (5) Data Point - A

((3  : Group (6) Data Point -- O

.p-g Group (7) Data Point - 0. _

5 l.s i

-E-1 i=

5.,4  : 'E L -

w L 27

^

u =

u o  :

3

% = $

g _- -

s _m_

M p i=

s

a. -

[ ^_2 , . .

en :h h ~

i h o

8-m.

H v.e = 'J

&  ! E'. -

U e

-t- !

' = . =:

N  : V%

4 w .

=

= v.

= - - _

i

?. "'".  :. ". 1, . " ' . W.". ~ 6f; l .a- ,.

Group Position, % Withdrawn Figure 3.4-9

Normalized Differential Shape Of Control Rod Group 8 Versus Withdrawal Position, For Beginning Of Life, Zero Power Conditions a

LA .C. -

-- Group (8) Data Point -- o H +56 -t_

w -

%_ 'z u

e '-

3 - v.:

H 2 r: w '.

Ne r.

8 .te 2 .

v4 u ++-~ 4 2 0 =

w .

O .--

& -n-= n-x , :. "

m .

gm_ T-p __

O- --

we .-

Oi . . -'

n e4. .,_2

,e4 . .

. b.

u --

0 ..

g6 a h e5  !

' _ 9_ .-

1

. p n. l

, . 1 l_;- 0

:,0 +:- O 5:- 1 =__-

Group Position, % Withdrawn l

l l

l l

Figure 3.4-10 1

1 L

Normalized Integral Shapes Of Control Rod Groups 3-4 Versus Withdrawal Position, For Beginning Of Life, Zero Power Conditions -

~ '

Group (3) Data Point - o _; M Group (4). Daca Point - y g"

. -3 5;- -

- E' u G-O- ='

n j=

a 5

u =

$ ~_r n -

l h

ti + l

? J

&! 1 0 I e: _

,A .:

O P y 5

[

a =

.c _-

l-

_5 1: .

ww-L "

Group Position, % Withdrrwn Figure 3.4-11

Normalized Integral Shapes Of Control Rod Groups 5-7 Versus Withdrawal Position, For Beginning Of Life, Zero Power Conditions

  • = ,- ~ Group (5) Data Point - A Group (6) Data Point -- O =-

Group (7) Data Point - O _

-=' ,

_ . =_

E g0 ._- __

y _=_

s #

$u =d a

0-2 -=',=

N 7'

E cn +: _

7 g 5 -

$ l 1 ,.-

=

E

~

U E m ___

=

^'

EO E-e

. e -

e:G Group Position, % Withdrawn Figure 3.4-12

Normalized Integral Shape Of Control Rod Group 8 Versus Withdrawal Position, For Beginning Of Life, Zero Power Conditions

)

EM -. -

Group (8) Data Point -- o

[

a Pi gO.~'" --T

_-1 e_ m

'~

a W:~ -

H .

'.=

H __f.0_ .5.

_ __ f

, ' ~  ? ~l I 6 ' -

s ti

  1. -)

o

--- b T

y- -

1u bv =

U ~=T u=

'\ .  :

l

,1GG 4 .d

.__1

~

1 b 0 r '

ing Group Position, % Withdrawn l

I l

l Figure 3.4-13

l l

l Measured Reactivity Worth Versus Inserted Group Worth For The determination Of The Rod Drop To Boron Swap Correction Factor 1

1 I

1 vsa.. --9e

. 2-s All Rods Drop Data Point - O 4 -

g -

w

x 3- E
o. t
2 .

N 3u Inverse' Slope = 1.28

=-

E .

'Y g-- x. .

m. .

a o

  • -._^

E =.,'

h. -

v

!= ^

G -

~

Boron Swap Inserted Group Worth, %Ak/k I

1 Figure 3.4-14

Comparison Of Heasured And Predicted Dif ferential Reactivity Worth of Soluble Boron Versus Boron Concentration For Beginning Of Life Conditions

!l

]

-.,1  ; 4 A.

a' Si o' -~3., .I

}

E!

N . 300 F.

-l . i!

U .d i y t ic

- Measured at Sa2F -

n o i~ ... ... , ,

illl W 3 ' '

I

. p y , ,, ,,,

c ---l

.ta g o -

L+1. - l

-f I ,

e o l '

w m ,

La '- 532 F

}u t l --I

, h-l I 579 F-d l M

A i

, !I - -

J d

l i i

)..s,f l 7 , n .. hii i hf l h h

. RCS Boron Concentration, ppm  !-

t t

4 I

Comparison of Measured And Predicted Integral Reactivity Worth Of Soluble Boron Versus Boron Concentration For Beginning Of Life Conditions l 11 1111111111111 1 '300 F-l L ,, P

-532 F

' Critical' Boron Worth' During:

r ,

p" I J, IZero Power Physics -- O

' ~~

, ',,e i

,-p ,

'l ,,<- ,

- " , 579 FL

' lgy ,,(

a  :

Ji .

1 '. ., ' ,

, e: .

s  : i ',, - ,

, .- i x l '

i d I f

^

l N [

, , ,3. ,,

[

' ' 'T

? fu .;! .

. C ',

(,, , c '

I....."

n, ,

{ -

g  ;. , ,

7 e ,t.

e ,,

[

,C

' e

,ta @ ,

'{,l

~ ' '

7 s m

o. -3 .

, e-'

i ,,

J"I d -

l cn

,U.. ' -

f g i , , ,

n l ,

u .

, e ,

no , ,,

J e < k,, ,

u I e

a l 1,. a,, ,,,, - -

g .q. .. .

l -

,,j '

-r .t li l

' I T%  ; -  !.

l l . . . ) i .

RCS Boron Concentration, ppm

. . . - . . . . . ~

Ejected Control Rod Reactivity Worth At Zero Power, Beginning Of Life, And 532 Degree F Conditions.

g NI-3

/

/

/

/

/

/

A /

B /

2 C -

l D -- /

E ,'

F -

_ - f .

G - " -

n -

K -

,/

L - /

M f #.

N ,

o -- e

.)

P l ,#

l R ,s'

,' I I I I I

/

r

/

/

le 2 3 4 5 6 7 8 9 10 11 12 13 14 15 M NI-4 CONTROL CORE CONTROL ROD W0ra ', % Ak/k

_ ROD POSITION PREDICTED MEASURED 7-8 F-02' O.47 0.44 l

Figure 3.4-17

Stuck Control Rod Reactivity Worth At Zero Power, Beginning Of Life, And 532 Degree F Conditions NI-3

/

/

/

/

A /

l B /

C -

/ l D ,

l l E / l l r -

/ II .!

o -

. II I u - ..- - .

K -

j l l t -

, II x .' lI N s l l

.0 -- /

l P  ! _/

l R /-

/ l l 1 l I

/

s

/

1/ 2 3 4 5'6 7 8 9 10 11 12 13 14 15 k NI-4 CONTROL CORE CONTROL ROD WORTH, % ak/k ROD POSITION PREDICTED MEASURED 4-3 H-14 4.00 2.34 Figure 3.4-18

4.0 POWER ESCALATION TESTS Following the completion of Zero Power Physics Testing, initial power escalation commenced on January 29, 1977 with the first nuclear electrical power produced at 1800 on January 30,1977 (a small amount of power was produced for six minutes on July 17, 1976, during turbine testing as part of Hot Functional Testing). The power escalation test program was conducted at four major test plateaus of 15, 40, 75, and 100% full power with minor testing performed at intermediate power levels as required by the controlling procedure for power escalation. All testing was completed by 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> on April 26, 1977. The average daily power level history during the power escalation test program is shown in Figure 4.0-1. The major power levels were achieved as follows: -

Power Level (Percent of Full Power - % FP) Date Achieved

, 15 February 1, 1977 40 February 27, 1977 -

75 March 14, 1977 100 April 1, 1977 Test results review was performed concurrently with testing. This allowed inunadiate review and approval by the Test Working Group (TWG), Plant Review

, Connaittee (PRC), and Testing / Operations management. Escalation to a new power level always followed promptly behind completion of the approval process.

( The test program was written to comply with the Technical Specifications and structured ia accordance with NRC Regulatory Guide 1.68, November, 1973. The tests reported in this section cover all required power escalation testing. A -

brief summary of tests reported, 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. These tests and the reports written on each, represent all unit and physics parameters necessary for experimental verification to assure that continued operatic.n of Crystal River Unit 3 is within the limits as imposed by Technical Specificctions.

A summary of performe ce indicators during the test period are tabulated in Table 4.0-2. As can be seen, a net time for the test program was 87.6 days.

This time includes 23 days for the one major delay encountered.

l 4.0-1

-._s_.

_ .4 __- ; -- , ,  ; , , _ u , _ o. . . . . . . . . _ .: - ~ _ J

1 Average Daily Power Level Versus Time During The Power Escalation Test Program f

i I '

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's Summary Of Power Escalation Tests Reported In Section 4.0 Report Test Power Levels Section Title of Section 0-5 15 25 40 65 75 92 100 4.1 Nuclear Instrumentation Calibration at Power Ter.t X X X X X X s 4.2 Biological Shield Survey Test X X X 4.3 Reactivity Coefficients At Power Test X X X 4.4 Core Power Distribution Test X X X X g 4.5 Unit Loss Of Electrical load Test X cr E 4.6 Turbine / Reactor Trip Test Turbine Plus Reactor X s-Turbine Only X X o 4.7 Incore Detector Test X X b ,

4.8 Power Imbalance Detector Correlation Test X X 4.9 Nuclear Steam Supply System Heat Balance Test X X X X X 4.10 Unit Load Steady State Test X X X X X X X X 4.11 Unit Load Transient Test X X X

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4.12 Shutdown From Outside The Control Room X*

4.13 Pseudo Rod Ejection Test X 4.14 Dropped Control Rod Test X X 4.15 inss Of Offsite Power Test X*

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  • Note: Done from 15% full power as part of the 100% full power plateau.

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7 Summary Of Power Escalation' Performance Indicators 1

Test Testing Major Net j Plateau Duration Delays Time Unit Capacity Factor *

.; (%FP) (Days) (Days) (Days) Heasured Estimated 15 3.3 0 3.3 0.66 0.65 40 14.3 23.0 37.3 0.75 0'.85 75 19.3 0 19.3 0.74 0.78 i

T?- 100 27.7 0 27.7 0.62 0.75 i 5

c j' Total 64.6 23.0 87.6 0.69 0.76 sa e

  • Based on total MWh that could have been produced at a given test plateau.

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4.1 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER TEST The power range instrumentation has four linear channels, (NI-5, NI-6, NI-7, NI-8) originating in four detector assemblies, each of which contains two un-compensated ion chambers. These ion chambers are positioned to represent the top and bottom halves of the core. The individual currents from the chambers are fed into individual lin' e ar amplifiers. The sum of the top and bottom neutron signals is the total reactor power. The difference between the top and bottom neutron signal multiplied by a gain factor is the power imbalance of the core. The outputs are directly proportional to reactor power and cover the range of 0 to 125% full power for the total power and -62.5 to +62.5% full power for the power imbalance. Each channel supplies power level information to the reactor protection system (RPS) and to the integrated control system (ICS) .

The gain of each linear amplifier is adjustable, providing a means for calibra-ting the outputs against a thermal heat balance and an incore axial imbalance measurement. The location of the out-of-core detectors relative to the reactor core is shown in Figure 4.1-1.

4.1.1 PURPOSE The purpose of the Nuclear Instrumentation Calibration At Power Test was to calibrate the power range nuclear instrumentation to within 2.0% full power of the thermal power as determined by a heat balance and to within 3.5% incore axial offset as determined by the incore monitoring system. Additional purposes during the power escalation program were as follows:

(a) To adjust the high power level trip setpoint when required by the power escalation test procr.fure.

(b) To confirm that the nuclear instrumentation calibration procedure can adequately adjust the power range channels to a heat balance and an in-core offset.

(c) To verify that at least one decade overlap exists between the intermediate and power range nuclear instrumentation.

Four acceptance criteria are specified for the Nuclear Instrumentation Calibra-tion At Power Test as listed below:

(1) The power range nuclear instrumentation indicates the power level within 2.0% full power of the heat balance and within 3.5% incore axial offset.

The incore axial offset criteria does not apply below 30% full power.

(2) The high power level trip bistable is a; f.:, trip at the desired value within a tolerance of +0.00 to -0.31% full power.

(3) The power range nuclear instrumentation can be calibrated adequately using the nuclear instrumentation calibration procedure.

(4) The overlap between the intermediate and power range is in excess of one decade overlap.

4.1-1

4.1.2 TEST METHOD 3

As required during power escalation, the top and bottom linear amplifier gains were adjusted to make the power range nuclear instrumentation channels indicate the measured heat balance power. The most conservative indication of reactor thermal power was obtained using weighted average heat balance on the unit computer. The gains were adjusted keeping their ratios the same if the offset was within tolerance of plus or minus 3.5 percent as determined by the incore monitoring system. If the offset needed to be adjusted, the gains were adjust-ed to correct the offset and heat balance mismatch at the same time.

During each adjustment, data was also taken to verify overlap between the in-termediate and power range channels. The. required overlap was a minimum of one decade between these two nuclear instrumentation ranges.

After all required testing was completed and approved at a major test plateau, the power escalation procedure then directed the high flux trip bistable set-point to be re-adjusted. The major settings during power escalation are given -

below:

Test Plateau (%FP) Bistable Setpoint (%FP) 15 35 40 50 75 85 2

100 104.7 The verification of the trip setpoint was then performed by placing each channel j in the test mode and slowly inreasing a test power signal until each bistable tripped. The difference betseen the actual power at which the bistable tripped and the setpoint trip power was the trip error recorded during each adjustment.

1 A 1. 3 EVALUATION OF TEST RESULTS An analysis of test results indicates that changes in reactor coolant system boron, cold leg temperature, control rod configuration, and xenon buildup or burnout affected the power indication as observed by the nuclear instrumentation.

This was as expected since the power range nuclear instrumentation measured reactor neutron leakage which is directly related to the above changes in system condicions. As system conditions were changed, this resulted in a nuclear power range increase or decrease which of ten fell outside the calibration tolerances, thus requiring them to be calibrated. Early in the power escalation test pro-gram, some difficulties were experienced until procedural errors were resolved and instrumentation personnel experience increased. Also near the completion of the 40% full power plateau, the procedural method was modified from an on-line to off-line technique to assist in removing time dependent drift. However, each time that it was necessary to calibrate the power range nuclear instrumen-tation, the acceptance criteria of calibration to the heat balance power of i 2.0%FP was met without any difficulty. Also each time it was necessary to l calibrate the power range nuclear instrumentation, the i 3.5% axial offset I criteria as determined by the incore monitoring system was also met. Table

! 4.1-1 is a summary of the data taken during calibration at different power levels during power escalation testing.

4.1-2 i

The performance of the power imbalance detector correlation test has indicated that an acceptable relationship between incore and out-of-core imbalance on off-set is obtained when a gain factor of 3.90 is employed. A detailed report on this gain factor adjustment is included in section 4.8 of this report.

The high flux level trip bistable was adjusted to 35, 50, 85, and 104.7% FP prior to escalation of power to 15, 40, 75, and 100% FP, respectively. Accep-tance criteria of adjusting the setpoint to the above values within + 0.00 to

-0.31% FP was met each time without difficulty. The maximum trip error record-ed was -0.08% FP when setting the high flux trip at 100% FP.

The actual overlap measured during the startup program exceeded the one decade overlap requirement. The overlap between the Latermediate and power range channels covers the total span of the power range. Figure 4.1-2 shows the over-lap of all three nuclear instrumentation channels.

4.1.4 ' CONCLUSIONS The power range channels were calibrated to within 2.0% full power and to within 3.5% axial offset several times during the startup program. Total power was determined by a thermal heat balance, and axial incore offset was determined using the incore monitoring system as input to the calibration. The calibra-tions were required due to changes in xenon worth, reactor coolant system boron concentration and rod configuration during the power escalation test program.

The calibration method has therefore, been thoroughly tested and has proven satisfactory.

The high flux level trip bistable was adjusted prior to each power level escalation to its respective trip setpoint. Verification of trip setpoint in all cases indicated that the maximum error in the setting is well within the acceptance criteria.

4.1-3

. 1 Summary of Nuclear Instrumentation Calibrations Performed At Power During ,

The Power Escalatior Test Program Pow r Offse ad T 1 wer e re An Af er ai . Offset Before And Af ter Calib, %

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(%FP) '(%) (%) NI-5 NI-6 NI-7 NI-8 NI-5 NI-6 NI-7 NI-8 i r

15.38 + 5.07 NA 15.95 15.28 14.91 15.19 + 6.37 + 7.02 + 8.84 + 6.18 15.08

- 0.86 NA 14.91 14.78 14.44 14.69 - 2.21 - 2.03 + 0.13 - 1.84 26.72 +10.97 -1.03 24.07 22.35 22.22 23.10 +31.10 +15.20 +17.70 +10.60 25.65 - 0.97 -1.03 26.69 26.66 26.88 27.04 - 3.56 - 3.34 - 2.27' - 2.74 40.43 - 1.41 -0.93 40.60 40.40 40.90 40.90 - 6.33 - 5.45 - 5.31 - 4.99 s 38.94 - 4.52 -0.93 39.60 39.60 39.73 39.7 - 6.06 - 6.14 - 4.21 - 5.16

Ek E 76.60 + 0.67 -2.22 74.33 74.02 73.77 75.33 + 0.07 + 0.11 + 0.37 - 1.92 ,

s- 76.50 - 1.30 -2.07 76.67 76.52 76.42 76.71 - 2.14 - 0.43 + 0.14 - 0.03 , ,

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H 91.02 - 4.46 -2.18 89.18 88.90 88.93 89.08 - 8.12 - 6.21 - 5.58 - 5.95 90.78 - 9.48 -2.37 92.02 91.71 91.24 92.55 - 7.09 -10.54 - 6.52 - 7.44 99.62 - 4.70 -2.28 97.62 98.56 96.81 98.40 - 1.84 - 1.84 - 1.37 - 2.28 100.47 - 7.73 -2.28 101.90 101.52 101.52 101.71 -10.54 -10.54 -10.40 -10.39 f

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4.2 BIOLOGICAL SHIELD SURVEY TEST The biological shield on Crystal River Unit 3 is designed to attenuate radia-tion levels throughout the unit, from direct and scattered radiation, to the dose rate levels which comply with the radiation exposure limits specified in the Final Safety Analysis Report (FSAR). These limits are listed below for each type of zone classification:

Zone Classification Dose Rate (mrem /hr) Tvpical Location Unlimited Occupancy 1 0.5 Office, Control Room and Turbine Building Normal Continuous Occupancy 1 2.5 Auxiliary Building:

Accessible Areas Controlled, Limited Access $15.0 Valve Galleries iontrolled, Limited Access 3 .25 Reactor Building: j Accessible Areas Restricted Access >100 Restricted Access:

Inside Secondary Shield To ascertain that the above design criteria in regard to radiation fielas have 7 been met, testing of the installed biological shield was required during the power escalation test program.

This report presents the results of the biological shield survey test performed on Crystal River Unit 3 at 0, 40, and 100 percent full power. The test program was based on ANSI N18.9-1972 Program for Testing Biological Shielding in Nuclear Reactor Plants.

4.2.1 PURPOSE The general purpose of the biological shield survey test was to measure dose rates in all accessible locations of the unit adjacent to the biological shield.

The specific purposes were:

(a) To measure and record radiation levels of the normally accessible areas of the nuclear unit.

(b) To identify those locations where radiation levels do not reet design criteria.

(c) To obtain base-line radiation levels at various power levels for compari-son with future measurements of activity buildup with operation.

l (d) To identify and label radiation and high radiation areas.

Three acceptance criteria are specified for the Biological Shield Survey Test and are listed below:

4.2-1 l

l

,. __ _ ~-,

- . . - . . _ . - - . - ~ . .- - - -

(1) Radiation levels of the normally accessible areas of the nuclear unit have been measured with these locations not meeting the design criteria noted. 7 (2) Base-line radiation levels at various pcwer levels have been established for future reference and comparison.

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(3) Radiation and high radiation areas have been identified and posted.

4.2.2 TEST METHOD The biological shield is defined as the effective shielding between the measure-

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ment location and the reactor vessel. Its principle shielding components are the primary shield wall, the secondary shield wall, the reactor building and the auxiliary building.

During the initial escalation in power from 0 to 100% FP, biological shield measurements at power levels of 0, 40, and 100% FP were performed to verify its ability to attenuate gamma and neutron radiation. ,

The test was performed by dividing all accessible areas into three different groups, each requiring a survey. The particular group surveys were:

(1) Reactor Building Interior Survey (2) Reactor Building Periphery Survey (3) Selected Unit Locations Survey .

J The reactor building interior survey was performed by a walk through survey of the normally accessible areas of the reactor building interior. At least one dose rate measurement was recorded for each stairway flight and each intermediate landing. If during the walk through, locally high radiation levels were en-

  • countered a complete scan of this area was performed. These areas were surveyed by traversing the survey instrument approximately waist-high and recording the highest reading observed.

The reactor building periphery sung was intended to scan normally accessible areas adjacent to the reactor building. Prior to performing this survey, accessible portions of the reactor building periphery were divided into a grid network. Each grid was then assigned a unique designation number of the form LLL-T-SSS, where:

LLL - Denoted the major floor, ground, or roof elevation in feet T- Denoted the grid type:"A" indicated a grid from floor elevation to 4 feet above and "B" indicated a grid from 4 feet above floor elevation to 8 feet above SSS - Denoted the section sequence number During the measurement the entire extent of each grid section was scanned by holding the detector approximately 2 inches from the surface, with the highest reading encountered recorded. s

. The selected unit location survey consisted of 60 locations throughout the unit.

'These areas were confined to the following locations:

4.2-2 L! _ _ _ . . _ _ _ _ _ _ -

(1) Auxiliary Building (2) Access Corridor To Reactor Building (3) Intermediate Building (4) Control Complex (5) Turbine Building (6) Reactor Building At each location only gamma radiation measurements were required provided the reactor building periphery survey indicated no derectable neutron radiation.

During the survey ~ instrumentation was held approximately waist-high when taking readings.

Prior to performance of the above shield surveys at each power level, calibra- l tion of the survey instruments were performed. This calibration was also re-peated at the completion of the surveys at each power levels. The surveys were obtained utilizing portable ionization and Geiger-Muller Detectors for gamma and BF3 detectors for neutron.

4.2.3 EVALUATION OF TEST RESULTS The reactor building interior survey was confined to floor elevations of 95 and 119 feet, since admittance within the secondary shield was prohibited when reactor power was greater than 1.0E-07 amps on the intermediate range detectors

(%1% FP). On these two reactor building levels, points which represented whole body exposure were established alphabetically from A to Z as shown in Figures I

4.2-1 and 4.2-2. Additional areas were also monitored whenever significant localized radiation was observed. During the survey, all locations except E, N, O and P met the design requirement of less than or equal to 25 mrem /hr as shown in Table 4.2-1. Of these only "E" was of a sufficient level to be considered a high radiation zone (i.e. >100 mrem /hr). The, problem observed at "E" was that the shield penetration for the cavity cooling fans was providing a path for neutron and gamma strasmiag. Additional sources of high radiation levels were

'also caused by numerous localized penetrations which were providing paths for neutron and gasmia streaming. In some cases, these localized areas did exceed 100 mRam/hr. -

The reactor building periphery survey was conducted at elevations of 95,119, 143, 162, and 195 feet and incorporated the survey of 582 grids. Each grid location at which radiation levels were observed during the shield test are listed in Table 4.2-2, with the levels measured at each plateau listed. As can be seen, only gamma type radiation was observed. The radiation levels in general have l been lower than expected, with only a few areas having levels higher that the design value of less than or equal to 2.5 mrem /hr. The high values are attribu-table to piping within the auxiliary building.

The selected unit location survey was designed to monitor the radiation levels at specific points throughout the unit. The intent was to verify that normally accessible areas were within their respective design values. Table 4.2-3 presents the radiation levels monitored at these locations for the three test l plateaus. In all cases, the dose rates measured were within their designed values.

l 4.2-3

. -~ - - - . . - - - - - . . - - . . . - . . . ..

4.

2.4 CONCLUSION

S

- 3 Upon completion of the Biological Shield Survay Test, the following conclusions were reached:

(1) Radiation and high radiation areas have been identified and posted.

(2) Radiation levels at var 1ous points around the nuclear unit have been established for future reference and comparison.

(3) Areas in which radiation levels exceed their design values are undergoing

further design evaluation. In the meantime they have been appropriately identified and marked.

9 I

's.

4.2-4

s l, Heasured Dose Rates On Crystal River Unit 3 During The Reactor Building Interior Survey l

POINT DESIGN 0% FP PLATEAU 40% FP PLATEAU 100% FP PIATEAll INDEX DCSE RATE DOSE RATE, mrem /hr DOSE RATE, mrem /hr DOSE RATE, mrem /hr (X). (mrem /hr) NEUTRON GAMMA TOTAL NEllTitON CMIMA TOTAT. PFUTitON CAMMA TOTA 1.

A 1 25.0 ND ND ND ND 0.5 0.5 < 0.5 < 0.5 < 1.0 B 1 25.0 ND ND ND ND 1.5 1.5 0.5 < 0.5 < 1.0 C 1 25.0 ND ND ND ND 1.5 1.5 3.0 2.5 5.5 D 1 25.0 ND ND ND 5.0 30.0 35.0 5.0 1.5 6.5 E < 25.0 ND ND ND 25.0 5.0 30.C 125.0 20.0' 145.0 F < 25.0 ND ND ND ND 1.5 1.5 .2.0 3.0 5.0 i

G 525.0 ND ND ND ND 2.0 2.0 1.0 4.0 5.0 H < 25.0 ND ND ND ND 3.5 3.5 0.5 6.0 6.5 I 525.0 ND ND ND ND 0.5 0.5 < 0.5 < 0.5 < 1.0 J < 25.0 ND ND ND ND 0.5 0.5 < 0.5 < 0.5 < 1.0 K 7 25.0 ND ND ND 0.5 0.5 1.0 < 0.5 0.5 < 1.0 s L 7 25.0 ND ND ND ND 0.'S 0.5 < 0.5 < 0.5 < 1.0

$- M 7 25.0 ND ND ND 1.0- 0.5 1.5 1.5 0.5 2.0 o N 525.0 ND ND ND ND 15.0 15.0 < 0.5 55.5 <56.0 4- O < 25.0 ND ND ND ND 30.0 30.0 < 0.5 55.5 <56.0 w P < 25.0 ND ND ND ND 45.0 45.0 < 0.5 80.0 <80.5 A Q 525.0 ND ND ND ND 0.5 0.5 < 0.5 0.5 < 1.0 R 1 25.0 ND ND ND ND 0.5 9.5 < 0.5 10.0 <10.5 S 1 25.0 ND ND ND ND 5.0 5.0 < 0.5 15.0 <15.5 T 1 25.0 ND ND ND ND 2.0 2.0 < 0.5 0.5 < 1.0 U 1 25.0 ND ND ' ND 1.5 0.5 2.0 2.5 2.0 4.5 V 1 25.0 ND ND ND ND 0.5 0.5 < 0.5 3.0 < 3.5 W 1 25.0 ND ND ND ND 20.0 20.0 < 0.5 18.0 <18.5 X 1 25.0 ND ND ND ND 0.5 0.5 < 0.5 0.5 < 1.0 Y 1 25.0 ND ND ND ND 0.5 0.5 < 0.5 3. 5 < 1.0 Z 1 25.0 ND ND ND ND 0.5 0.5 < 0.5 0.5 .< 1.0 No u ; The definition of (ND) is none detected 1

e

l' Measured Dose Ratcs On Crystal River Unit 3 During The Reactor Building Periphery Survey i

l POINT DESIGN 0% FP PLATEAU 40% FP PLATEAU 100% FP PIATFAU INDEX DOSE RATE DOSE RATE, mrem /hr DOSL RATE, mrem /hr DOSE RATE, mrem /hr (LLL-T-SSS) (mrem /hr) NEUTRON CAMMA TOTAL NEUTRON CAMMA TOTAL NEUTRON CAMMA TOTAL 095-A-001 12.5 ND ND ND ND ND ND ND 0.20 0.20 095-B-001 12.5 ND ND ND ND 0.50 0.50 ND 0.20 0.20 095-A-002 12.5 ND ND ND ND ND ND ND 0.20 0.20 095-B-002 12.5 ND ND ND ND ND ND ND 0.60 0.60 095-A-003 12.5 ND ND ND ND ND ND ND 0.60 0.60 095-B-003 12.5 ND ND ND ND ND ND

  • ND 0.60 0.60 095-A-004 12.5 ND ND ND ND ND ND ND 0.80 0.80 '

095-B-004 12.5 ND ND ND ND ND ND ND 0.80 0.80 i 095-A-005 12.5 ND ND ND ND 0.50 0.50 ND- 2.00 2.00 i 095-B-005 12.5 ND ND ND ND 0.50 0.50 ND 2.00 2.00  :

095-A-006 12.5 ND 1 ND ND ND ND ND ND 2.50 2.50 s 095-B-006 12.5 ND ND ND ND 0.50 0.50 ND 3.50 3.50 i C- 095-A-007 12.5 ND ND '

ND ND ND ND ND 2.50 2.50 E 095-B-007 12.5 ND ND ND ND 2.00 2.00 ND 3.50 3.50

- 095-A-008 12.5 ND ND ND ND ND ND ND 2.50 2.50 L 095-B-008 12.5 ND ND ND ND 1.50 1.50 ND 4.00 4.00 h 095-A-009 12.S ND ND ND ND ND ND ND ' 2.50 2.50 095-B-009 12.5 ND ND ND ND 0.50 0.50 ND 5.00 5.00 095-A-010 12.5 ND ND ND ND ND ND ND 2.50 2.50 095-B-010 12.5 UD ND ND ND ND ND ND 3.50 3.50 095-A-011 12.5 ND ND ND ND ND ND ND 2.00 2.00 095-B-011 12.5 ND ND ND ND ND ND ND 2.50 2.50 095-A-012 12.5 ND ND ND ND ND ND ND 2.00 2.00 095-B-012 12.5 ND ND ND ND ND ND ND 2.50 2.50 095-A-013 12.5 ND ND ND ND ND ND ND 1.50 1.50 oP 1 J.3 12.5 ND ND ND ND ND ND ND 1.50 1.50

  • I bya-A-014 12.5 ND ND ND ND ND ND ND 1.50 1.50 095-B-014 12.5 ND ND ND ND ND ND ND 1.50 1.50

'095-A-015 12.5 ND ND ND ND ND ND ND 1.00 1.00 s 095-B-015 12.5 ND ND ND ND ND ND ND 1.00 1.00 095-A-016 12.5 ND ND ND ND ND ND ND 1.00 1.00 Note: Only grid sections with measurable radiation reported ,

( Note: The definition of (ND) is none detec j

m Heasured Dose Rates On Crystal River Unit 3 During The Reactor Building' Periphery Survey POINT DESIGN 0% FP PLATEAU 40% FP PLATEAU 100% FP PLATEAU INDEK DOSE RATE DOSE RATE, mrem /hr DOSE RATE, anem/hr DOSE RATE, mrem /hr (LLL-T-SSS) (mrem /hr) NEUTRON GAMMA TOTAL NEUTRON CAMMA TOTAL NEUTRON GAMMA TOTAL 095-B-016 12.5 ND ND ND ND ND ND ND 1.00 1.00 095-A"017 12.5 ND ND ND ND ND ND ND 1.00 1.00 095-B-017 12.5 ND ND ND ND ND ND ND 1.00 1.00 095-A-018 12.5 ND ND ND ND ND ND ND 1.00 1.00 095-B-018 12.5 ND ND ND ND ND ND ND 1.00 1.00 095-A-019 12.5 ND ND ND ND ND ND ND 0.18- 0.18 095-B-019 12.5 ND ND ND ND ND ND ND 0.18 0.18 095-A-020 12.5 ND ND ND ND ND ND ND 0.18 0.18 095-B-020 12.5 ND' ND ND ND ND ND ND 0.18 0.18 095-A-021, 12.5 ND ND ND ND ND ND ND 0.16 0.16 g 095-B-021 12.5 ND ND ND ND ND ND ND 0.18 0.18 g 095-A-022 12.5 ND ND ND ND ND ND ND 0.16 0.16 a 095-B-022 12.5 ND ND ND ND ND ND ND 0,16 0.16 f- 095-A-023 12.5 ND ND ND ND ND ND ND 0.14 0.14 y 095-B-023 12.5 ND ND ND ND ND ND ND 0.14 0.14 w 095-A-024 12.5 ND ND ND ND ND ND ND 0.08 0.08

^ 095-B-024 12.5 ND ND ND ND ND ND ND 0.08 0.08 h 095-A-025 12.5 ND ND ND ND ND ND ND 0.06 0.06 rt . 095-B-025 12.5 ND ND ND ND ND ND ND 0.06 0.06

$ 095-A-026 12.5 ND ND ND ND ND ND ND 0.04 0.04 095-B-026 12.5 ND ND ND ND ND ND ND 0.04 0.04 095-A-027 12.5 ND ND ND ND ND ND ND 0.04 0.04 095-B-027 12.5 ND ND ND ND ND ND ND 0.04 0.04 095-A-028 12.5 ND ND ND ND ND ND ND 0.02 0.02 095-B-028 12.5 ND ND ND ND ND ND ND 0.02 0.02 095-A-093 12.5 ND ND ND ND ND ND ND 0.10 0.10 095-B-093 12.5 ND ND ND ND ND ND' NP 0.04 0.04 095-A-094 12.5 ND ND ND NL ND ND n 1.00 1.00

, 095-B-094 12.5 ND ND ND ND ND ND ND 0.50 0.50

! 095-A-095 12.5 ND ND ND ND ND ND ND 0.02 0.02 4 s 095-B-095 12.5 ND ND ND ND ND ND ND 0.14 0.14

! Note: Only grid sections with measurab'le radiation reported j

4 Note: The definition of (ND) is none detected

l I

Heasured Dose Rates On Crystal River Unit 3 During. [

i The Reactor Building Periphery Gurvey POINT DESIGN 0% FP PLATEAU 40% FP PLATEAU 100% FP Pl.ATEAU

. INDEK DOSE RATE DOSE RATE, mrem /hr DOSE RATE, mrem /hr DOSE RATE, (mrem /hr)

(LLL-T-SSS) (mrem /hr) NEUTRON GAMMA TOTAL NEUTRON GAMMA TOTAL NEUTRON GAMMA TOTAL 095-A-096 < 2.5 ND ND ND ND ND ND ND 0.02 0.02 095-B-096 7 2.5 ND ND ND ND ND ND ND 0.02 0.02 119-A-004 }[ 2.5 ND ND ND N9 1.50 1.50 ND 4.50 4.50 119-B-004 di 2.5 ND ND ND ND 2.00 2.00 ND 4.00 4.00 119-A-005 5, 2. f ND ND ND ND 1.00 1.00 ND 2.00 2.00 119-B-005 5, 2. 5 ND ND ND ND 1.00 1.00 ND 3.00 1.00 119-A-006 5, 2. 5 ND ND ND ND 1.00 1.00 ND 1.50 1.50 119-B-006 5, 2. 5 ND ND ND ND 1.00 1.00 ND 2.00 2.00 119-A-007 5, 2. 5 ND ND ND ND 1.50 1.50 ND 2.00 2.00 119-B-007 < 2.5 ND ND ND ND 2.00 2.00 ND 2.00 2.00 H 119-A-008' II 2.5 ND ND ND ND 1.50 1.50 ND 1.50 1.50 b 119-B-008 7 2.5 ND ND ND ND 1.00 1.00 ND 1.50 1.50 t $~ 119-A-009 ][ .25 ND ND ND ND 1.50 1.50 ND 0.50 0.50 f- 119-B-009 5, 2. 5 ND ND ND ND 1.00 1.00 ND 1.00 1.00 N 119-A-010 < 2.5 ND ND ND ND 1.00 1.00 ND 1.00 1.00 eb 119-B-010 ][ 2.5 ND ND ND ND 1.00 1.00 ND 1.00 1.00 s 119-A-011 < 2.5 ND ND ND . ND 1.50 1.50 ND 1.00 1.00 9 199.-B-011 72.5 ND ND ND ND 1.50 1.50 '

ND 1.00 1.00

' $, 119-A 012 3[ 2.5 ND ND ND ND 1.50 1.50 ND 1.00 1.00 f3 119-B-012 5, 2. 5 ND ND ND ND 1.50 1.50 ND 1.00 1.00 119-A-013 5, 2. 5 ND ND ND ND- 1.00 1.00 ND 1.00 1.00 119-B-013 5, 2. 5 ND ND .ND ND 1.50 1.50 ND 1.50 1.50 119-A-014 5, 2. 5 ND ND ND ND 1.50 1.50 ND 1.G0 1.00 119-B-014 jL 2.5 ND ND ND ND 1.50 1.50 ND 1.00 1.00 119 .1-015 5, 2. 5 ND ND ND ND 1.50 1.50 ND 1.00 1.00 119-b-015 ji 2.5 ND ND ND ND 1.00 1.00 ND 1.00 1.00 .

119-A-016 5, 2. 5 ND ND ND ND 1.00 1.00 ND 1.00 1.00

, 119-B-016 5, 2. 5 ND ND ND ND 1.00 1.00 ND 1.00 1.00

,119-A-017 5, 2. 5 ND ND ND ND 0.50 0.50 ND 1.00 1.00 119-B-017 5, 2. 5 ND ND ND ND 0.50 0.50 ND 1.00 1.00 i 119-A-018 jL 2.5 ND ND ND ND 1.00 1.00 ND 1.00 1.00 Note: Only grid sections with measurable radiation reported i Note: uf Thedefinitionof(ND)isnonedet( j

7 Measured Dose Rates On Crystal River Unit 3 During The Reactor Building Periphery Survey POINT DESIGN 0% FP PLATEAU 40% FP PLATEAU 100% FP PLATEAU INDEX DOSE RATE DOSE RATE, mHem/hr DOSE RATE, mrem /hr DOSE RATE, mrem /hr (LLL-T-SSS) (mrem /hr) NEUTRON GAMMA TOTAL NEU GON GAMMA TOTAL NEUTRON GAMMA TOTAT.

119-B-018 1 2.5 ND ND ND ND 1.00 l'.00 ND 0.02 0.02 162-A-004 1 2.5 ND ND ND ND ND ND ND 0.02 0.02 162-B-004 < 2.5 ND ND ND ND ND ND ND 0.04 0.04 162-A-005 52.5 ND ND ND ND ND ND ND 0.04 0.04 162-B-005 1 2.5 ND ND ND ND 0.06 ND ND ND 0.06 162-A-006 < 2.5 ND ND ND ND ND ND ND 0.06 0.06 162-B-006 52.5 ND ND ND ND ND ND ND 0.06 0.06 162-A-007 1 2.5 ND ND ND ND ND 0.06 162-B-007 ND ND 0.06

-< 2.5 ND ND ND ND ND ND ND 0.06 0.06 162-A-008. 1 2.5 ND ND ND ND ND ND ND 0.06 0.06 ff 162-B-008 1 2.5 ND ND ND ND ND ND ND 0.06 0.06

$ 162-A-009 1 2.5 ND ND ND ND ND ND ND 0.04 0.04

  1. . 162-B-009 1 2.5 ND ND ND ND ND
  • ND ND 0.04 0.04 162-A-010 1 2.5 ND ND ND ND ND ND ND 0.02 0.02 162-B-010 1 2.5 ND ND ND ND ND ND ND 0.02 0.02 8

8 e

g .

Note: Only grid sections with measurabic radiation reported Note: The definition.of (ND) is none detected

t Measured Dose Rates On Crystal River Unit 3 During ,

The Selected Unit Locations Survey j

. 2 i  !

POINT DESIGN 0% FP PLATEAU 40% FP PLATEAU 100% FP PLATEAU INDEX DOSE RATE DOSE RATE, mrem /hr DOSE RATE, mrem /hr DOSE RATE, mrem /hr (XX) (mrem /hr) NEUTRON CAMMA TOTAL NFUTRON GAMMA TOTAL NEUTRON GAMMA TOTAL A. Auxiliary Building l 01 < 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 1 02 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 03 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5

  • 04 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 05 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 06 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 07 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 H 08 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 EE 09 72.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 Y 10 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 .

11 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 w 12 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 b 13 7 2.5 ND ND ND ND < 0.5 < 0.5 NR < 0.5 < 0.5 14 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 15 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 16 7 2.5 ND ND ND ND 1.5 1.5 NR < 0.5 < 0.5 ' '

17 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 18 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 19 72.5 ND ND ND ND ND ND NR < 0.5 < 0.5 20 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.3 21 7 2.5 ND ND ND ND 1.5 1.5 NR I.5 1.5 22 72.5 ND ND ND ND ND ND NR < 0.5 < 0.5 l 23 52.5 ND ND ND ND ND ND NR < 0.5 < 0.5

< 0.5 < 0.5 24 < 2.5 ND ND ND ND ND ND NR 25 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5

26 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 j B. Access Corridor To Reactor Building 27 < 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5

}

i Note: The definition of (ND) is none detected I

(.- . Note: The definition of (NR) is not requ- I )

. C .

/

i Heasured Dose Rates On Crystal River Unit 3 During The Selected Unit Locations Survey POINT DESIGN 0% FP PLATEAU 40% FP PIATEAU 100% FP Pl.ATEAU '

INDEX DOSE RATE DOSE RATE, mrem /hr DOSE RATE mrem /hr DOSE RATE, mrem /hr (XX) (mrem /hr) '1EUTRON GAMMA TOTAL HFilTRON CAMMA TOTAL NEUTRON CAMMA TOTAL 28 < 2.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 1 29 < 2.5 ND ND ND ND ND ND NR < 0. 5 < 0. 5 30 52.5 ND ND ND ND ND ND NR < 0. 5 < 0. 5 31 1 2.5 ND ND ND ND ND ND NR < 0. 5 < 0. 5 32 1 2.5 ND ND ND ND ND ND NR < 0. 5 . < 0.5 C. Intermediate Building -

33 1 2.5 ND ND ND ND ND ND NR < 0. 5 < 0. 5 l 34 1 2.5 ND ND ND ND ND ND NR < 0. 5 < 0. 5 35 < 2.5 ND ND ND ND ND ND NR < 0. 5 < 0. 5 s 36 7 2.5 ND ND ND ND ND HD NR < 0. 5 < 0.5

, EI- 37 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0. 5 E 38 7 2.5 ND ND ND ND ND ND NR < 0. 5 < 0. 5

,c- 39 [2.5 ND ND ND ND ND ND NR < 0.5 < 0. 5

.l { D. Control Complex 1 m 40 < 0.5 ND ND ND ND ND ND NR < 0. 5 < 0. 5

, 9 41 7 0.5 ND ND ND ND ND ND NR < 0. 5 < 0.5 S- 42 70.5 ND ND ND ND ND ND NR < 0.5 < 0.5 d'

43 7 0.5 ND ND ND ND ND ND NR < 0.5 < 0.5 44 7 0.5 ND ND ND ND ND ND NR < 0.5 < 0.5

! 45 7 0.5 ND ND ND ND ND ND NR < 0.5 < 0.5 46 7 0.5 ND ND ND ND ND ND NR < 0.5 < 0.5 47 < 0.5 ND ND ND ND ND ND NR < 0.5 < 0.5 48 7 0.5 ND ND ND ND ND ND NR < 0.5 < 0.5 49 _7 0.5 ND ND ND ND ND WD NR < 0.5 < 0.5 E. Turbine Building -

50 < 0.5 ND ND ND ND ND ND NR < 0.5 < 0.5 j 51 50.5 ND ND ND ND ND ND NR < 0.5 < 0.5 i

. Note
The definition of (ND) is none detected l Note: The definition of (NR) is not required 1 '

Heasured Dose Rates On Crystal River Unit 3 During The Selected Unit Locations Survey 3 4

l t

l POINT DESIGN 0% FP PLATEAU 40% FP PLATEAU 100% FP PLATEAU .'

INDEX DOSE RATE DOSE RATE, mrem /hr DOSE HATE, mrem /hr DOSE RATE, mRe'm/hr

, (XX) (mrem /hr) NEUTRON GAMMA TOTAL NEUTRON CAMMA TOTAL NEUTRON GAMMA TOTAI,

'i

'f 52 < 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 53 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5

,: 54 55

{2.5

< 2.5 ND ND ND ND ND ND NR < 0.5 < 0. 5 1 ND ND ND ND ND ND NR < 0.5 < 0. 5

!! 56 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0. 5

.i7 7 2.5 ND ND ND ND ND ND , NR < 0.5 < 0.5 i' 58 7 2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 59 {2.5 ND ND ND ND ND ND NR < 0.5 < 0.5 l

F. Reactor Building.

a E 60 1 25.0 ND ND ND ND ND ND 3R < 0.5 < 0. 5 E

Y w

8

> 0. ,

n. ,

8 Il 4

i .

i Note: The definition of (ND) is none detected r Note: The definition of (NR) is not requif C

} ( .

~ Location f Whole Body Exposure Points Inside The Reactor Building At Elevation of 95 Feet

.L N

s J

\ \

Y 0 Zh

/ I K \ I

-- 00 t W

I I EiY E L i -

4 I h2 f I r!

O o l REACTOR w.3sgL l 0

, il 5

=

u!

~

-s -

W e g

SG=S8 a

N IMT.

+m o n .m s

Figure 4.2-1 d

-es = * *--

  • -e--

3 opee We e w eg .

l l j

[ Location Of Whole Body Exposure Points Inside The )

! Reactor Building At Elevation of 119 Feet m '

( l l

I I

l i

b N i @ .

SOJ3A l W CTOR g (

N

,) .

RREssumst En b i a  :.:

l

.)

Ib Q l w I medd a

l

= .

1 y

% REACTOR CAvtTY COCL.PAug g

Wtf . il U T ME f C REACTOR SLOG.

PAN AS$36Y E1A c, n, n .

mm ,

t l

I I

I Figure 4.2-2

+

4.3 REACTIVITY COEFFICIENTS AT POWER TEST 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, and to verify that they were conservative with respect to the FSAR. The following coefficients were either measured or calculated from the data obtained.

(a) Temperature coefficient of reactivity, defined as the fractional change in the reactivity of. the core per unit change in fuel and moderator temp-erature.

(b) Moderator coefficient 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 reactivity of the core per unit change in fuel temperature.

Three acceptance criteria are specified for the Reactivity Coefficients at Power Test and are listed below:

(1) The moderator coefficient of reactivity measured at 40 and 75% full power -

shall be less than +0.90 X 10-%k/k/ F.

(2). The moderator coefficient of reactivity measured at 100% full power shall l be less than 0.00 X 10-%k/k/0F. 1 (3) The power Doppler coefficient of reactivity shall be more negative than

-0.55 X 10-%k/k/%FP.

4.3.2 TEST METHOD Reactivity coefficient measurements were made during the power escalation test program at core power levels of 40, 75, and 100% full power. No measurements were made at 15% full power because of the change in control mode from changing TAVE to constant TAVE which occurs around 15% full power. This change makes reactivity coefficient measurements impractical at 15% full power.

Differential rod worth measurements were executed at each steady state plateau during the reactivity coefficient measurement in order to generate rod worth data for the specific test conditions. For temperature coefficients, average reactor coolant temperature was increased or decreased about 50F and data re-corded. For power Doppler coefficients, power was decreased about 5 percent full power and data recorded. From the measured temperature and power Doppler coefficients, the moderator and Doppler coefficients were calculated.

4.3.2.1 TEMPERATURE AND MODERATOR COEFFICIENTS

~

The tamperature coefficient of reactivity is defined as the fractional chang'e in the reactivity of the core per unit change in fuel and moderator temperature.

The temperature coefficient is normally divided into two components as shown in equation 4.3-1.

4.3-1

[- .

_ - ~ = .. .- ..m. _ .,

, ,,,m

aT = ag + aD EQ. (4.3-1)

= Temperature Coefficient of Reactivity Where: aT og = Moderator Coefficient of Reactivity a = Doppler Coefficient of Reactivity

, D The moderator coefficient cannot be directly measured in an operating reactor because a change in the moderator temperature also causes a similar change in the fuel temperature. Therefore, the moderator coefficient must be calculated

using equation 4.3-1 after the temperature and Doppler coefficients have been determined.- Technical Specifications, section 3.1.1.3, requires that the mod-erator coefficient be less than +0.90 X 10- M k/k/0F at power levels below 95%

full power and non-positive at or above 95% full power.

t Temperature and moderator coefficients were theoretically predicted at various power levels as shown in Figures 4.0-1 and 4.3-2 using the distributed modera- -

tor and fuel temperatures instead of the isothermal values which were used for zero power physics predictions. For these predictions, the normal mode of operation with critical boron and rod conditions was assumed which set the average core moderator temperature equal to 5790F.

The measurement method used at power is to change the reactor coolant tempera-ture setpoint at the reactor control station with the integrated control system in automatic effecting an approximate 50F change in the reactor coolant temp-erature. The reactivity change caused by the temperature change of the core

)

was measured by recording the change in the position of the controlling control rod group and converting this change to reactivity using differential rod worth values measured during the test. Prior to running the test, steady state equilibrium xenon conditions including a stable boron concentration and no significant control rod motion during the last 30 minutes prior to taking data were required as prerequisite systgm conditions.

. 4.3.2.2 POWER DOPPLER AND DOPPLER COEFFICIENTS The Doppler coefficient relates the change in core reactivity to a correspond-ing change in fuel temperature. The Doppler effect is a negative reactivity contribution arising from the Doppler broadening of the U-238 neutron capture

cross sections in the resonance or high energy region. An increase in fuel i temperature increases the effective absorption of neutrons within the' fuel which J

decreases the core excess reactivity. The Doppler coefficient of reactivity is not measured directly in an oper ting reactor because the fuel temperature cannot be-readily measured. However, the Doppler coefficient must be determined in order to calculate the moderator temperature coefficient. Thus an indirect approach is taken: The power Doppler coefficient of reactivity is measured and the-Doppler coefficient is calculated using the calculated fuel temperature change associated with a power level change according to Figure 4.3-3. Mathe-matica11y, the Doppler coefficient is related to the power Doppler coefficient as follows:

aD = (AP/ATE ) aPD EQ. (4.3-2)

Where: aD =

Doppler Coefficient of Reactivity l aPD

=

Power Doppler Coefficient of Reactivity 4.3-2 z . . ._. ._.

\

AP/aT g = Change in Power per Unit Change in Fuel Temperature The power Doppler coefficient relates the change in core reactivity to a cor-responding change in power. Theoretical predictions of the power Doppler coef- ,

ficient were made using a three-dimensional PDQ code with thermal feedback. l The predicted power. Doppler coefficients using this code are presented in l Figure 4.3-4.

The measurement method used was to change the reactor power level 5 percent full power. This change in power level was initiated by manually decreasing the reactor power at the reactor master control station. After obtaining ap-proximately ten minutes of steady state data at the reduced power level, reactor power was returned to the initial power.

The calculation of the power Doppler coefficient uses the measured change in j the controlling rod group position converted to an equivalent reactivity value i and the measured change in reactor power determined by using the normalized core AT which is the primary side heat balance. .

4.3.2.3 DIFFERENTIAL ROD WORTH AT POWER l The method by which the differential rod worth was determined at power is the fast insertion / withdrawal method. In this measurement, the controlling rod group is inserted for approximately six seconds, followed immediately by a withdrawal for approximately six seconds. Since the total elapsed time is on the order of the primary loop recirculation time, the moderator temperature effects are eliminated and the reactivity versus time is essentially a combina-tion of the effects due to the control rod motion and the fuel power variation.

A typical plot of the response of the reactivity, rod position, and fuel power is shown in Figure 4.3-6.

Determination of the differential rod worth was then found by using the measured reactivity and rod positions, compensating the data by a predicted fuel power correction factor. Mathematically, this is expressed as:

do/dh = (CF) (202 - pl - p3) / (2H2 - H1 - H3) EQ. (4.3-3)

Where: H1 = CRA (s) position (%) prior to motion i

H2 = CRA (s) position (%) after the insertion but prior to withdrawal H3 = CRA (s) position (%) after the withdrawal is terminated p1 = Reactivity prior to CRA motion (ak/k) p2 = Reactivity af ter the insertion but prior to the withdrawal (ak/k)

, p3 = Reactivity af ter the withdrawal is terminated (ak/k) l CF = Fuel Power Correction Factor (none) j The fuel power correction factor accounts for the time delay involved in fuel j temperature change during the measurement and is given by the equation below:

l CF =

1 + (CC) (P1)

Where: CF =

~ Fuel Power Correction Factor (none)

CC =

Correlation constant (BOL = 0.0053%FP-1)

P1 = Fuel power prior to motion (%FP) 4.3-3 i ,

_. .. . m .. m. _

-, - . . - - , . - . - - - . - . .- . ~ . ~ a- _

By application of.the above two equations, the differential rod worth during

.'the reactivity coefficient at power test.was determined.

7 '

4.3.3 ' EVALUATION OF TEST RESULTS The results of the measured temperature and calculated moderator coefficients at power are tabulated in Table 4.3-1 and plotted in Figures 4.3-1 and 4.3-2 which also shows the predicted. temperature and moderator coefficient results.

A tabulation of the measured data used to determine the temperature coefficient at the 40, 75, and 100% full power plateaus is also furnished for each measure-ment in Tables 4.3-2 through 4.3-4.

Examination of the calculated moderator coefficients as plotted in Figure 4.3-2 indicates that the limit of a non-positive value will not be exceeded at 95 percent full power unless the soluble poison concentration is in the range of 1120 ppmB. Measured results show that the soluble poison concentration for equilibrium-zenon all-rods-out condition is about 1020 ppmB. This confirmed that during power operation at or above 95 percent full power the moderator .

coefficient will be negative even if all control rods are withdrawn from the Core.

The results of the measured and predicted power Doppler coefficient of reactivity are tabulated in Table 4.3-1 and plotted in Figure 4.3-4. A tabulation of the measured data used to determine the power Doppler coefficient at the 40, 75, and 100% full power plateau is also furnished for each measurement in Tables 4.3-5 through 4.3-7.

TheacceptancecriterionforthemeasuredpowerDogplercoefficientisthatthe j coefficient must be more negative ~than -0.55.X 10' Ak/k/%FP. Figure 4.3-4 shows that all measured coefficients are below this value and that the accep-tance criterion is adequately met. In addition, all calculated Doppler coef-ficients were also within their respective acceptance criteria.

Examination of the measured power Doppler coefficient indi' cates a slightly more negative value than predicted which results in a reactivity deficit versus power more negative than originally predicted. A comparison of the measured and predicted power Doppler reactivity deficit versus power level is shown in Figure 4.3-5. The total measured reactivity deficit between 0 and 100 percent full power is estimated at -1.10 percent ak/k in cors excess reactivity available for Cycle 1 lifetime.

The results of the differential rod worth measurement performed as part of the Reactivity Coefficient At Power Test are shown in Tables 4.3-8 through 4.3-10.

It should be noted that the maximum value measured was 0.0202%Ak/k/%wd which is well within the maximum allowable design value of less than 0.0303%Ak/k/%wd at hot full power.

To ensure that these differential rod worths were correct, verification of the correlation constant used in the fuel power correction factor was also performed at 40, 75, and 100% full power. In all cases, favorable agreement was obtained.

~

4.

3.4 CONCLUSION

S The measured re di:s at 40, 75, and 100% full power indicates that the moderator coefficient of reactivity will be negative during power operation above 95% full .

4.3-4

power. Comparison between predicted and measured temperature coefficients of

. reactivity showed favorable agreement.

Analyzed data for the power Doppler coefficient versus power level indicates that the least negative coefficient is -1.02 X 10-4ak/k/%FP which is sufficiently

.below the acceptance criterion of -0.55 X 10-4ak/k/%FP. The total power Doppler deficit from this measured d.ata is projected at -1.10%Ak/k.

I e

4 9

3 r

-3 4.3-5

.s%-,%,.,% .p. - - - - - s - --

- Summary Of Heasured, Calculated, And Predicted Coef ficients Of Reactivity At Power 4

Average Incore Rod Position, 7, wd Boron Core Coefficient of Reactivity, 10-0 Ak/k Power Offset Conc. Burnup (IFP) (1) 1-5 6 7 8 (ppm) (EFPD) Heasured Predicted Calculated Predicted A. Temperature And Moderator Coefficient Temperature Moderator

.I 40.0 -2.0 100 96 21 18 1034 1.9 -0.34 -0.24 -0.14 -0.06 70.8 +0.6 100 88 10 17 956 7.7 -0.52 -0.40 _o,34 -0.25 93.6 -2.1 100 94 15 8 916 18.3 -0.46 -0.49 -0.28 -0.36 Y

a-8

~ .

t j ,B. Power Doppler And Doppler Coefficient Power Doppler Doppler 36.9 -2.0 100 94 19 18 1034 1.9 -1.12 -0.99 -0.20 -0.18 71.8 +0.6 100 88 13 17 956. 7.7 -1.05 -0.97 -0.18 -0.17 91.1 -2,1 100 93 12 8 916 18.3 -1.02 -0.96 -0.18 -0.17 k

8

1 N*

4 1

i l-Temperature Coefficient Measurement Calculation At 407. Full Power c

l-TDIPERATultE COErr!Cigitr ANALYSIS .

REACTIVITY C00rrROI.t.ING PARAMETERS (Up) ____ _ (Down) __ g Tg 578.58 583.57 ar  !

q. (inittst ava reactor coolant temp.)
f 'T2 ( IC8I 8V8. E88CE F CO I8at E8"P.) SA1.57 576 Al r

+6.76 or '

AT=(Tg - T2 ) -4.89

[. u, (initial contratting caa Croup rosition) 96.48/20.90 96,49/20.90 tud

{

x2 (final contratting cRA croup rosition) 96.49/20.90 96.50/20.90 two j, 4'

Au (change in rod position during measurement) 0.01/0.00 0.01/0.00 twa l

ak/k/aui (atfrerential rod worth values) 0.0168 0.0168 lzak/k/aa eg (inittet power tevet) 40.76 39.16 tre j, N r2 (fin 81 power level) 39.16 41.01 tre I ct 7 Ar (change in power due to temperature change) -1.60 +1.85 _

IFF

. PrD (Power doppler coefficient) -0.0112 -0.0112 zak/k/ar ,

W  !

b Baron Concentration 1034 1034 ppna i O n. g (initial menon reactivity worth) -2.0214 -?_0245 Zak/k ose 2 (final menon reactivity worth) -2.0245 -2.0256 zak/k j

  • (ak/k/au) (au) 0.0000 0.onno .zak/k e ("rD) (AP) +0.0179 -0.0208 Zak/k e op,, (avg. reactivity change due to Xe) -0.0006 -0.0011 :Iak/k e Apeacess reactivity inserted due to temperature change +n_nnni +0.0020 Zak/k  ;.

t

.; ano - total of

  • itema abovo +0.0174 -0.0217 Zak/k a 7 (temperature coefficient) - ano/ AT -0.0036 -0.0032 Iak/k/aT 1

3 g

(

Temperature Coefficient Heasurement Calculation At 757. Full Power TEMPERAft*E COEFFICisNT ANALYSIS BEACTIVITY CONTROL.I.1NC PARAMETERS _U)

(P (Down) g Tg .(initial avs. reactor coolant temp.) 578.79 583.55 *r 72 (fin.1 avs. reactor coolant temp.) 583.55 577.96 *r AT=(Tg - T2) -4.76 +5.59 *r 88.49/10.51 N, ,(initial controllins CaA Croup Fosition) 87.82/9.73 zud H2 (final e ntrolling CRA Croup Foottion) 87.82/9.73 87.48/9.40 Zwd AH (change in god position during measurement) -0.67/-0.78 -0.34/-0.33 twd ak/k/aN (differential rod Wrth values) -

0.0112 0.0il2 Zak/k/au Pg (initial power level) 72.84 ,, 69.44 %FP h F2 (final Power level) 69.44 71.6} IFP AP (change in power due tn temperature change) -3.40 +2.17 IFP k

, PFD (P wer doppler coef ficient) ,

-O_010s -0.0105 zak/k/ae Baron Concentration 956 956 r pi.a P me g (initial menon reactivity worth) -2.5642 -2.5652 zak/k

  1. xe 2 (final menon reactivity worth) -2.5652 -2.5671 Zak/k e (ak/k/aH) (au) -0.0082 -0.0038 x3kfg
  • ("tD) (AP) +0.0357 -0.0227 gf,kfg
  • Ap ,, (avg. reactivity change due to xe)' -0.0010 -0.0019 Zak/k
  • Apencess reactivity inserted due to temperature change -0.0008 _+0.n0n4 Zak/k ano - total of
  • seems above +0.0251 -0.0280 Zak/k at (temperature c efficient) - ano/AT -0.0054 -0.0050 zak/k/AT 8

-u__mm.mm- +-

Temperature Coefficient Heasurement Calculation At 1007. Full Power s

TEMPERATLRE COEFFICIENT ANALYSIS REACTIVITY CONTROIl,YNC PARAMETERS (Up) (Down) g  ;

578.69 584.12 or  !

T1 .(initiat avs. reactor coolant temp.)  ;

T 2 III"*I *"8* '**** ' ***I'"" **"P'I 5NE 1R 79.02 r ,

I ATa(T -T) 2

+5.49 -5.10 or i Hg ,(initial controlling CEA Croup Position) 93.24/14.30 94.77/15.53 2,a j i

H2 (II"*I * ""'*III"8 C'* C'*"P rosition) 95.30/16.29 93.26/13.66 tw4 AH (change in rod position during measurement) +2.06/+1.99 1.51/1.87 twa g ak/k/ali (dif ferential rod worth values) 0.0146 0.0141 zak/k/aH N. el (init181 Power level)  %.M M.M M f

E P2 (final p wer level) 94.72 92.86 tre

< p j

Ar (change in power due to temperature cha nge) +0.53 +0.16 tre i e

& PPD (Power doppler coef ficient) -0.0102 -0.0102-zak/k/ar 1 Boron Concentration 916 916 ppaa I

P x. 3 (initial menon reactivity worth) -2.7201 -2.7198 1ak/k 8 x. 2 (flaal =*aon reactivity worth) -2.7201 -2.7198 zak/k j

e (ak/k/an) (AH) +0.0296 -0.0238 zagfg I

f

  • ' PPD) (AP) -0.0054 -0.0016 Zak/k Ap,, (avg. reactivity change due to Xe) 0.0000 0.0000 zak/k a hpexcess reactivity inserted due to temperature change _+0.0001 +0. 001 'l ' Zak/k i

l BHO = total of

  • items above +0.0243 -0.0241 Zak/k -

f l a, (temperature coef ficient) = RH0/ AT -0.0044 -0.0047 zak/k/ar l 'Y

~

f Power Doppler Coefficient Heasurement Calculation At 407. Full Power POWER DOPPIER cOEFFictElfr ANALYSIS

., REACTIVITY coefraol.LINc PARAMETERS . (Ik>wn) (Up) t! NITS rg (initial power level) 39.98 33.77 tre r, (final power level) 33.77 40.21 tre AP4(Pg -r) 2 6.21 -6.44 tre ug (inteint contrattins car croup rosition) 95.69/20.23 92.43/16.73 tua

. n2 (fi"*1 e nerotting caA croup rosition) 92.43/16.73 96.88/21.58 tud AH (change in rod position during measurement) -3.26/-3.50 4.45/4.85 twa Ak/k/aB (dtfierentist rod worth values) 0.0181 0.0166 Zak/k/au h 73 (initial avg. reactor coolant temperature) 578.30 578.14 *e T2 (final avg. reactor coolant temperature) 578.14 578.35 "F v

AT (change in temperature due to power change) -0.16 +0.21 "F b "T (temperature coefficient) -0.0034 -0.0034 Zak/k/AT noron concentratton 1034 1034 pp.a P xe 1. (initial menon reactivity worth) -2.0355 -2.0438 zakfk

  1. xe 2 (finat xenon reactivity worth) -2.0438 -2.0475 zak/k
  • (ak/k/au) (ati) -0.0613 +0.0772 zak/k
  • (o T) ( AT) +0.0005 -0.0006 Zak/k
  • A p xe (ava. reactivity change due to ze) -0.0083 -0.0037 Zak/k
  • A p excess reactivity inserted due to power change -0.0003 -0.0009 2ak/k nuo - total of
  • itema above -0.0694 +0.0720 rak/k
"to (Power doppler coefficient = RH0/AF) -0.0112 -0.0112 zak/k/ap i

{' .

i.

Power Doppler Coefficient Heasurement, Calculation At 757. Full Power i f

i-

,!WOL DOPP!ER COgFFICIEffT ANALYSIS +

REACTIVITY COWTROLA.ING PARAMETERS (Dmin) (Up) UNITS Pg (initial power level) 74.13 69.ss tre j P, (final power level) 69.78 74.09 tre

! Ar=(eg - P2) 4.35 -4.54 tre i

f Hg (initial controlling CRA Croup Position) 88.19/14.04 86.66/12.32 two t-H2 (final controlling CRA Group Position) 85.70/11.25 89.09/14.96 twd AH (change in rod position during measurement) -2.49/-2.79 2.43/2.64 tud i

Ak/k/au (differential rod worth values) 0.0187 0.0185 zak/k/4H t'

!H i n Tg (initial ave. reactor coolant temperature) 578.01 578.66 *r b

j.

l 72 (final ava, reactor c olant temp *e rat ure) 578.40 578.73 ar F

4

,e AT (change in temperature due to power change) +0.39 +n.n7 "r "T (temperature coef ficient) -0.0052

'f

-0.0052 rak/k/a7 f Boron Concentration 956 956 pena Pm. 3 (initial menon reactivity worth) -2.5667 -2 . '. ,' . c 23gfg ,

P me 2 (final menon reactivity worth) -2.5667 -2.5749 23gjg f

  • (ak/k/all) (aH) -0.0494 +0.0469 23gfg i e (* T) ( AT) -0.0020 -0.0003 rak/g e A p ua (avg. reactivity change due to Xa') 0.0000 -0.0004 23gjg e O P excess reactivity inserted due to power change +0.0008 0.0000 23gfg l ano - tor.1 of
  • stems above -0.0466 +0.0461 23,jg

! 'to (Power doppler coefficient - Raio/Ap) -0.0107 -0.0102 Zak/k/aP I

i

(x 'r Power Dop;,ler Coef ficient Heasurement Calculation At 1007. Full Power POWER DOrelER COErrICIENT ANALYSIS BEACTIVITY ColtrROI.I.INC PARAMETERS (Down) (Up) UNITS Pg (initial power level) 93.66 88.34 tre r, (final power level) 88.34 94.11 trr Ar(e -r) g 2 5.32 5.77 tre un (initist controtting Ca4 Croup roattion) 93.93/14.10 90.56/10.37 twa

-l u2 (final c ncr 111ng CaA Group rosition) 90.56/10.37 95.36/15.10 twa AH (change in rod position during measurement) -3.37/-3.73 +4.80/+4.73 twa Ak/k/all (differential rod worth values) 0.0136 0.0134 zak/k/au Tg (initial avg. reactot coolant temperature) 578.77 578.88 r

.I

n .

$ T2 (final avg. reactor coolant tengerature) 578.88 578.69 a r

'i

  • p AT (change in temperature due to power change) +0.10 -0.19 "r "T (temperature coefficient) -0.0046 -0.0046 zak/k/ar

,.i Boron concentration 916 916 pens Pa. g (initial menon reactivity worth) -2.7341 -2.7386 Zak/k

'! P me 2 (final menon reactivity worth) -2.7386 -2.7434 2ak/k a (Ak/k/All) (All) -0.0483 +0.0639 xak/k.

a (a T) ( AT) -0.0005 +0.0009 rak/k a A p ue (avg. reactivity change due to Ke) -0.0045 -0.0048 Zak/k a

O p excess reactivity inserted due to power change -0.0005 -0.0002 23kfg Rito = total of a items above *

"to (tower Joppler coefficien: = RIIO/AP) -0.0101 -0.104 2ak/k/ar i

8

i i3 li I!

i

, Sununary 7f Measured Differential Rod Worths At 40% Full Power i b

I a l  !

Initful Correction \

Power Factor Control Rod Croup' Position Average Bifferentin Reactivity Position llor e s.

81 CF all lil 11 2 D it pl p2 p) pp HA f (1 FP) (NGNE) (IWD) (1 WD) (IWD) (IWD) (PCH) cr lepp gi si i - (PCH) (PCH) (PCH) (IWD) (l's it/ mp) s[

A. Tempe reture And Hoderator Coel ficient - t 1

~

h 40.76 1.220 20.2 -18.3 20 1 -3.7 +10.0

e. 0.0 -25.0 -60.0 19.2 19.8

, 40.76 1.220 20.1 18.2 19.8 -3.5 0.0 1

-24.0 + 9.0 -57.0 19.4 19.9 i

,l

! i W

tt H '

I L.

k B. Power Doppler And' Doppler Coefficient 39.98 1.212 20.8 19.0 20.7 -3.5 -3.5 -22.0 + 8.4 -51.1 19.9 17.7 39.98 1.212 20.1 18.3 20.0 -3.5 -1.4 -44.1

-19.2 + 7.1 19.2 15.3 [

33.77 1.186 16.5 14.7 16.1 -3.2 -1.0 -23.4 + 6.1 -52.1 15.5 19.3 l i

33.77 1.186 17.0 15.3 16.7 -3.1 +1.0 -22.5 + 6.8 -52.8 16.1 20.2 l 40.21 1.213 21.1 19.5 20.9 -3.0 -2.0

. -20.8 + 9.0 -48.6 20.3 19.6 40.21 1.211 20.1 18.2 20.0 -3,7 -1.8 '

-23.0 + 8.1 -52.3 19.1 17.1 t

l 3

Ilotes CF = (1 + 0.0053 P1)

Dit = (2112 - ut - H3) i DP = (2pe - pl - p3) uA = (2ua + ul + m3) / 4 l

G

< )

{

Summary Of Measured Differential Rod Worths At 75% Full Power Initla! Correction Power Factor Control Pod Group Posittora Average liif f erent in React ivit y Position PL CF Unr t h lit 11 2 H) 13 11 Pt p2 p) pp

(! FP) DIONL) na cr 3,pf ri n

(%WD) (1 WD) (%WD) (%Wil) (PCH) (PCH) (PCil)

(PCH) (IWis) (rssst ato A. Temperature And Hoderator Coel ficient 72.84 1.386 10.6 8.9 10.6 -3.4 ~ 0. 0 - 9.1 + 6.3 -24.5 9.8 10.0 72.84 1.386 10.2 8.4 10.1 -3.5 0.0 -10.5 + 4.9 -25.9 9.3 10.3 69.44 1.368 9.2 7.2 9.1 -3.9 0.0 -14.1 + 6.4 -34.6 8.1 12.2 69.44 1.368 9.1 7.2 9.0 -3.7 0.0

_ -12.7 + 7.6 -33.0 8.1 12.2 2

0 5

,e B. Fower Doppler And Doppler Coef ficient

[ 74.13 1.393 13.6 11.8 13.6 -3.6 0.0 -18.0 +11.3 -47.3 12.7 18.3 74.13 1.393 13.5 11.8 13.5 -3.4 0.0 -17.0 +10.0 -44.0 12.7 18.0 69.55 1.369 11.7 9.9 11.8 -3.7 0.0 -19.3 +11.0 -49.6 10.8 18.3 69.55 1.369 11.7 10.0 11.7 -3.4 0.0 -19.3 +11.2 -49.8 10.9 20.0 74.09 1.393 9.1 7.4 9.2 -3.5 0.0 -16.3 +10.9 -43.5 8.3 l' 3 74.09 1.393 9.2 7.4 9.2 -3.6 0.0 -18.7 + 9.7 -47.1 8.3 18.2 Hotes CF = (1 + 0.0053 art)

DH = (2112 - HL - H3)

D9 = (2p pt - p3) ma = (2ua + ut + m3) / 4

i i

{

{i Sununary Of Measured Differential Rod Worths At 100% Full Po3er, I

1 .

3 i!

l l

i i

Initial Correction 1

, Power Facter control Rod Croup Positten Average Isif f erent ia Reactivity Position unri ti P1 0F ,

11 1 11 2 ill Dl4 PI p2 p) pp isa I (X FP) Ok)NE) (%Wu) (1 WD) (%WD) (ZWD) (PCH) cr gepp p si (PCH) (PCH) (PCH) (IW:)) t e4 ts/ swis)

A. Temperatura And tioderator Coef ficient a ~ __

f t

94.19 1.499 14.20 12.50 14.20 0.00 -12.50

-3.20' +8.30 -33.30 13.35 15.60 94.19 1.499 14.30 12.50 14.30 0.20 -13.00

-3.60 +9.00 f

-35.20 13.40 14.66 94.72 1.502 16.00 14.10 15.90 0.00

-3.70 -12.75 +9.00 -34.50 15.03 14.01 94.72 1.502 15.60 13.60 15.30 -3.70 0.00 -13.00 +9.00 -35.00 14.53 14.21 92.86 1.492 13.50 11.60 13.25 -3.55 0.20 -13.00 +8.20 -34.40 12.49 14.46  ;

92.86 1.492 13.30 11.50 13.30 -3.60 0.20 -12.00 +9.00 -33.20 12.40 13.76 E' I D* - i g'

, e ,

B. Power Doppler And Doppler Coef ficient tf 93.66 1.496 14.10 12.30 14.10 -3.60 0.30 s -12.00 9.00 -33.30 13.20 13.84 o 93.66 1.496 14.10 12.20 14.10 -3.80 0.00 -12.30 8.90 -33.50 13.15 13.19 88.34 1.468 10.80 9.00 10.70 -3.50 0.50 -12.00 8.00 -32.50 9.88 13.63 -

88.34 1.468 10.70 9.00 10.70 -3.40 0.20

-11.50 8.60 -31.80 9.85 13.73 [

94.11 1.499 14.50 12.50 14.30 -3.80 0.00

. -12.00 /.90 -31.90 13.45 12.58 '

94.11 1.499 14.30 12.50 14.30 -3.60 0.30 t

-11.80 9.20 -33.10 13.40 13.78 il i

i meta Cr - (1 + 0.0053 arl)  !

Du - (21:2 - ut - n))

DP - (2A - pl - p3) uA - (2ua + ut + m3) / 4- ,

t' U .s

.)

q L .

  • 4 h

Temperature Coefficient Of Reactivity Versus Boron Concentration For Critical Boron And Rod Conditions At 579"F, 2155 PSIC, And 0 EFPD i I l

n

l l l i ,<:.

I

n. .n:.;- "  !

3 E . -li- I

<l I. [ .

1, '.j;;, , i =

~ ,

.. -g -QT.:

3 Predicted Tempdrature , '. -

I.. -

, +: ,  :: Coefficient Results

? y- 15.%FP " "

\ '.-7

..t T 3..0 y liO %FP

"{j

, 3-

@ T 100 %FP '

', I w ,g '

1 s

..I'..

y, s s h

I .e

..ps " '

y f' ,,  :, , ,

? <

g,...rf..f .

~-

Y

~

O. l..) y---~T[.3", ,

, -~1 v .. , , i-

.'.T-u , , - .-

B - --

Measured Temperature

$ ., q, '

Coefficient Results -

O -- 10 1 %FP O

H -

o

, IF - '

E1 -- 75 %FP A -- 100 %FP

- I , D, l} ,,, 'd.'

l

'lll ;lij

~

f.. I, '

l l l l l l 4 it j'li !I l jl l

[ I l . l l l RCS Boron Concentration, ppmB L

i .

Hoderator Coefficient Of Reactivity Versus Boron Concentration For Critical  !

Boron And Rod onditions At 579 F, 2155 PSIG, And O EFPD i

-l l i

l i i  !

, -R ,i:  ; j

.i.

all ,

4

' si. }

O s '

a ,,

e ..T 4

son x

I l q- , ,

a l

, - e

' l ,

H Predicted Moderator eg

' ' r f

Coefficient Results , .

}

i

% ~--n i -

15;%FP y ,

{ J- h0 %FP Q ,

l

}

g h 100 %FP ,,." " ""

n ,O 3 .

h

' ,e di '

h*..-,,,, c "Q',

w d..n ' ' ,

i ..i

} E. , 8_ ,

l .-f 1,

,, ie d '

i o "

l" " ..

1 ,..n ,ij "

Q i

l 3- " ';

' Calculated Moderator j  ?. ..n ,

Coefficient' Resulta -

E' l o -- h0 %FP j S..g  ;

e- 75 %FP ,

j a -- 100 %FP

? l l .n .  !- -

1 i l

1 a  ! i

'l

_.i. l i i <

  • ; I Q:-  : L :l :i-:t  ! .:lD j RCS Boron Concentration, ppmB  :
i

?  :

i 6 2

i

)

Avt. rage Fuel Temperature Versus Power Level At 0 EFPD 1

l i

i l

H a n--

/-- ,

i,,_ \

RC ( AP/ATg ) = 0.lf6 7,FP/dF ,

\

-"/

A  :/

o _-

,m m . /

u=~1= /__

o  :

u  :

/4

  • /

S  :'

N 1"O_- _

/ ~

v.

2: -

@' .[

c c;;- ..

f Eie o-  :

/'

<. 7 I'

7."... j

=

i

-een-

- ^

0 20 '0 9: =0 'm Power Level, 7. Full Power Figure 4.3-3

.-.. - - ... ..-.n .

..~ . .-....

1 I

I l

1 Power Doppler Coef ficient of Reactivity Versus Power Level

-m, rr -

Acceptance Criterion Of 0.55X10-4Ak/k/%FP _

n :.1 s

.w __

4

.e

'o e -

g~ _

a J -

g Predicted _._,-j__.,

%~

. r_ _ .

['

_ 1 = == =

_/"; j 8 _ ;" &_ --

='

_-M -

Or-O sh- __

u -

E Measured Power Dogpier g

Coefficient Result s O- 40 %FP O - 75 %FP

_. .= n A 100 %FP _.

-?-- --_ _ - .

-. c_ =_ _ = -.- - - .

ruu.___

Power Level, % Full Power

~)

^]

Figure 4.3-4

- . . . - . . . _ . - . . - - - . - . ..... . ~ . -

Power Doppler Reactivity Deficit Versus Power Level 7_ . .I _

_= .n.-

s. .

a ~

.d f'

) - --.

+

.f-4 -,,. wrw,

4 G _Ef-f---+--

yuC +

Measured _s_,-- , + - - ,

,A W am-+ arr ,

-- .'I. - - ~

p- -

h> j' Pred'i cted --

94 . - t---/ -

u o_ n- .r

,. +..,. ,"-- _

+-

$. ;c -f -

  • ,7/ -.- -

E ~- .=

-0. 2. ,g' E#

_=n.

/

m- - . - - . -- ,-.._..

___ 1

-w, _.

. , - - . . _ _ _ l

-0 20 60 sc 100 - - l Power Level, % Full Power l

1 l

../ ,

1 Figure 4.3-5 l l

t lj i ,l.  ; - 4lg .j iI8l  !-t; . P C OaprOo.

f nW4WMN= OMOCo P a a mCSH mOCe" 0CNTm >

h oWg D*%m@MMOnWpP @ o. gON h xaWMCnaBeU" >n E ceN

a P

F -

ll

% l r lfr -

e $/

w s fV o l P

ljlI b\

q l t ,i e

u -

F d l w

% I n ,

o f,,

.e, q'l i il\

t  !

i' , '

i

,l i r s

, ngl 9's o 'a P  !

a',l ,

h

,2 p 1 u

o r ,j .

G l l

M liq lWfpl  ;

C l P l l Il .il

,q,*i'n,.,,Iq

,j

'I.'

y t  !'

'p i 8, i

v '

-,', .,'hi t ,

c a

e llO ,.

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i IL.

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i ,

l

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i i i

.r.[,[.

  • rW$m m , 8am mWpM9 ULi*

. t ilt,* .i' O

4.4 CORE POWER DISTRIBUTION TEST Core power distributions were taken as required during the power escalation test program as a part of the following individual tests.

Core Power Distribution Test Power Imbalance Detector Correlation Test Incore Detector Test Pseudo Ejected Control Rod Test Dropped Control Rod Test Continuous monitoring of the core power density at 364 core locations is accom-p11shed by the incore monitoring system. This system is comprised of 52 detec-tor strings each having 7 individual neutron detectors equally spaced at seven axial elevations in the center of 52 fuel assemblies. This system is capable of producing detailed core power distributions for either eighth core or quarter core symetry conditions. A detailed description of the incore monitoring system and the calibration test results are presented in Section 4.7. The output of the incore detecto,rs is connected to the unit computer and is corrected for back-ground, fuel depletion and the as-built dimensions to provide accurate outputs of relative neutron flux. The computer output of the corrected signals is used to develop core power distributions which provide power peaking information necessary to determine core performance in terms of DNBR and LHR.

8 Because the incore monitoring system does not immediately respond to prompt changes in core conditions, the above limits are maintained on the bases of core power and core power imbalance as measured by the out-of-core nuclear instrumen-tation. Implementation of the core power and core power imbalance safety limits, in terms of the reactor protection setpoints, is shown in Figure 4.4-1.

, The results of the core power distributions taken during the test program as part of the power escalation sequence are discussed by dividing this section into

, three subsections as follows:

(1) Normal Operating Core Power Distributions

, (2) Worst Case Minimum DNBR And Maximum LHR Calculations (3) Quadrant Power Tilt And Axial Power Imbalance Numerous core power distributions were taken'during power escalation testing.

Generally, the detailed analyses presented herein are those required by the Core Power Distribution Test, TP 7 1 800 11. Additional analyses of the core power distributions taken as a part of other tests are presented in their appropriate sections.

4.4.1 PURPOSE The purposes of tha steady state core power distribution measurements are as follows:

(a) To measure core power distribution and thermal-hydraulic data at the major test plateaus as required by the power escalation test program.

4.4-1

.. ... .~ . -

- .- ._ - -. ~ _ _ .

i j-(b) To compare the measured and predicted core power distributions at 40, 75, and 100% full power.

(c) To verify acceptable core thermal-hydraulic parameters at 40, 75, and 100% full power.

Four acceptance criteria are specified for the Core Power Distribution Test and are listed below.

(1) The core power distribution and thermal-hydraulic parameters have been measured, evaluated, and deemed reasonable.

(2) The highest measured radial and total peaking factors are not more than 5.0 and 7.5 percent greater than predicted, respectively.

j (3) The measured worst case minimum DNBR is greater than 1.30 and the measur-ed worst case marinnan LHR is less than 18.00 kW/f t.

(4) The extrapolated worst case minimum DNBR is greater than 1.30 and the extrapolated worst case marimum LHR is less than 19.70 kW/ft, or the extra-polated imbalance falls outside the power imbalance trip envelope as shown

in Figure 4.4-1.

i 4.4.2 TEST HETHOD i

When required by the test program, computer printouts of the core power distri-bution and thermal hydraulics conditions were obtained after establishing steady state conditions at the required power level and rod configurations as deter- / ,

., mined by the Controlling Procedure for Power Escalation, TP 7 1 800 00. In '

order to compara measured results to predicted results, some cases required two-dimensional or three-dimensional equilibrium xenon. When three-dimensional equilibrium xenon was required, the APSR's were maintained at a constant posi-

. tion and the axial incore imbalance was maintained within +/-2% FP of zero for approximately eight hours prior to taking data.

4.4.3 EVALUATION OF THE TEST RESULTS 4.4.3.1 NORMAL OPERATING CORE POWER DISTRIBUTIONS Normal operating, equilibrium xenon core power distributions were measured and predicted for various operating control rod patterns at each of the major power escalation test plateaus. Measured results of four core power distributions j covering various control rod patterns and core power levels are tabulated in detail in Tables 4.4-2 through 4.4-5. These tables give complete 1/8 core i

power distribution maps using.the corrected signal outputs from 203 incore detec-tors located in 29 different fuel assemblies which describe the entire core, assuming eighth core synunatry. A summary of each measured core power distribu-tion presented in the above tables is given in Table 4.4-1 which tabulates the core power level, control rod positions, core burnup, boron concentration, axial imbalance, maximum quadrant tilt, maximum LHR, minimum DNBR and power peaking data ~for each measurement.. The measurements covered the following control rod

_ patterns and core power levels

3 J

0 4.4-2

Case Power Level Control Rod Position, %wd Xenon Equil.  ;

Number  % FP Cp (1-5) Gp (6) Cp (7) GP (8) Status 1 15 100 94 19 20 Yes/2-D 2 40 100 95 18 18 Yes/3-D 3 75 100 97 18 08 Yes/3-D 4 100 100 86 10 08 Yes/3-D The results of these measured core power distributions taken at operating control rod configurations indicate a maximum radial peaking factor between 1.36 and 1.42, and a maximum total peaking factor between 1.62 and 1.92 for the four cases studied. Examination of the maximum radial peaking factor in-dicates that this value is a monotonically decreasing function of power level at a given control rod configuration. This is basically ascribed to thermal feedback effects between 15 and 100 percent full power. Figure 4.4-2 shows the degree of power flattening on the radial power profile as observed along the X-Z plane at row 08 for various power levels. As can be seen, a relatively flat radial power distribution at 100 percent full power was observed. Exam- .

ination of the maximum total peaking factor indicates that it also is a function of power level and strongly dependent on the incore offset and core lifetime, at a given control rod configuration. As quoted above, the range of total peaking factors observed during normal operation was well below the design maximum total peaking factor of 2.67.

Core power distribution predictions at steady state conditions were predicted using the three-dimensional PDQ-7 code with thermal feedback. The results of the steady state PDQ predictions are given in th Physics Test Manual for different unit conditions in terms of the maximum radial and total peaking factors. The four cases reported in this section have been compared (measured versus predicted) in Table 4.4-6 to demonstrate the degree of agreement. In all instances, measured maximum radial and total peaking factors were within the acceptance criteria of not being more than 5.0 and 7.5 percent greater than those predicted by PDQ-7. A comparison of the measured and predicted core power distribution using (1/8) core analysis, was also performed on an assembly type basis (lump burnable poison, fuel enrichment, rod and unrodded, or a combination of the aforementioned). These results are presented in Figures 4.4-3 and 4.4-4. As can be seen, PDQ-7 is adequately predicting the radial and total peaking factors independent of the location and assembly type.

In summary, the comparison between the measured and predicted radial and total power distributions as expressed in peaking factors indicated favorable agree-ment, well within the acceptance criteria. Examination of the measured core power distribution relative' to that predicted by Puy-/ indicates that Crystal River Unit 3 has a better power distribution (i.e. flatter radial power profile and lower maximum peaking factors) than had initially been predicted.

4.4.3.2 WORST CASE MINIMUM DNBR AND MAXIMUM LHR CALCULATIONS To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal peaking conditions. The two primary core thermal limits which are indicative of fuel thermal performance are fuel melting and departure from nucleate boil-s ing. These limits are independent; each must be evaluated to ensure core safety for a given power peaking situation. Fuel melting is basically a function of 4.4-3

- - - - _ . - . - . _ . . - - - - - . - ~ . - . - - _- -

the local power generated in the fuel which is a combination of radial and axial peaking. The parameter used to define this limit is the maximum linear heat rate given in terms of kW/ft. .)

}

The upper boundary of the nucleate boiling region is termed " departure from i nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperature and the 1 possibility of cladding failure. The local DNB ratio, (DNBR), defined as the l ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB.

Although DNB is not an observable parameter during reactor operation, the ob- -

servable parameters of neutron power, reactor coolant flow, temperature and pressure can be related to DNB through the use of the W-3 correlation. The W-3 correlation has been developed to predict DNB and the location of DNB for uniform and non-uniform axial heat flux distributions. The minimum value of the DNBR,~during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. A DNBR of 1.30 corresponds to 94.5%

probability at a 99% confidence level that.DNB will not occur; this is consider-ed a conservative margin to DNB for all operational conditions.

Technical Specifications Limits related to DNBR and LHR limits of 1.30 and 19.70 kW/ft respectively, are given in terms of core axial imbalance, power level, reactor coolant flow, temperature and pressure. Figure 4.4-1 gives the protec-tive limits in terms of core power level and axial imbalance for four reactor coolant pump operation.

This section presents the results of the worst case minimum DNBR and maximum LHR ,

calculations based on measured core power distributions taken during the power

{ escalation test program through the 100 percent full power test plateau. The

! analyses presented are based on the output of the unit computer where measured i data is printed out in a reduced form ,and/or analyzed by the various computer  ;

j packages. Both, worst case minimum DNBR and maximum LHR calculations were per- t l formed by the unit computer and printed as part of the standard core power distribution analysis.

WORST CASE MINIMUM DNBR DETERMINATION i

l' The worst case minimum DNBR values were calculated by the unit computer for each core power distribution taken as part of the power escalation test program.

The results of various worst case minimum DNBR values calculated at each test plateau under normal rod configurations are plotted in Figure 4.4-5. These results indicate that all measured values were above the design worst case minimum DNBR versus power level and well above the minimum acceptable value of 1.30.

Four normal operating equilibrium zenon core power distributions as required oy The Core Power Distribution Test were obtained during the power escalation sequence. Each distribution was subjected to the following analysis:

(a) From each core power distribution, the worst case measured minimum DNBR was selected.

T (b) .)

The measured worst case minisman DNBR's were then extrapolated to the over-l power trip setpoint and corrected for axial peak location and magnitude.

4.4-4 g- m._ . -_ _ . ,

___,_m

(c) Verification of acceptable core conditions at the present and next power level of escalation were then performed relative to 1.30.

(d) The measured worst case minimum DNBR's were then extrapolated to the LOCA and design overpower power level and corrected for axial peak location and magnitude.

(e) Margin analysis was then performed on the extrapolated worst case minimum DNBR value relative to the +4== value of 1.30.

The results of these analyses are presented in Table 4.4-7. All measured and worst case minimum DNBR prior to and after extrapolation were above 1.30. In addition, all cases studied resulted in substantial DNBR margins. A minimum worst case minimum DNBR margin of 96.2 and 53.8 percent were observed from the LOCA and design overpower limits, respectively.

WORST CASE MAXIMUM GR DETERMINATION Worst case maximum GR values were calculated by the ut.it computer for each standard core power distribution taken as part of the power escalation test program. These calculated results were then multiplied by an additional factor of (1.02), to bring the position of the worst case calculation to nine feet above the bottom of the corc. Comparison of each case relative to the LOCA limit of 18.00 kW/ft was then performed. In all cases, the above limit was met during the power escalation test program.

Four normal operating equilibrium xenon core power distributions as required by the core power distribution test were obtained during the power escalation sequence. Each distribution was subjected to the following analysis:

(a) From each core power distribution, the measured worst case maximum LHR was selected.

(b) The measured worst case ==4== LHR values were multiplied by an addition-al factor of (1.02) and then extrapolated to the overpower trip setpoint.

(c) Verification of acceptable core conditions at t.he present and next power level of escalation was then performed relative to the acceptance criteria values of 18.00 and 19.70 kW/ft, respectively.

(d) The measured worst case mnimum GR values were multiplied by an addition-al factor o.f (1.02) and then extrapolated to the LOCA and design over-power levels.

(e) Margin analysis was then performed on the extrapolated worst case maximum GR values relative to the LOCA and des'.gn overpower limits of 18.00 and 19.70 kW/ft, respectively.

The results of these analyses are presented in Table 4.4-7. All worst case i

' values prior to and af ter the extrapolation to the overpower trip setpoint were within their respective acceptance criteria of 18.00 and 19.70 kW/ft, respectively.

In addition, all cases studied resulted in adequate LHR margins. A minimum worst case maximum GR margin of 22.0 and 21.7 percent were observed from the LOCA and design overpower limits, respectively.

4.4-5

. . . . - . - ~ . --_ -

4.4.3.3 QUADRAFf POWER TILT AND AXIAL POWER IMBALANCE s

QUADRANT POWER TILT Quadrant power tilt limits have been established in the Technical Specifications.

i These limits, when used in conjunction with the control rod position limits, assure that the design peak. heat rate criterion is not exceeded during normal power operation.

l Quadrant power tilt is defined by the following equation and is expressed in j percent.

t

- l' Power in Any Core Quadrant x 100 Quadrant Power Tilt =

, Average Power of All Quadrants ,

EQ. (4.4-1) i During the startup testing program, marimum quadrant power tilt was determined i using the corrected signals from the 16 synsnetric incore monitoring assemblies.

! Table 4.4-1 shows the maximum quadrant power tilt for each standard core power distribution taken as required by the Core Power Distribution Tes,t. These marimum quadrant power tilts are representative of the values observed during j the startup test program.

i During the performance of the Power Imbalance Detector Correlation Test, Se quadrant power tilt as indicated by the corrected signals from the 16 symatrie incore monitoring assemblies was shown to be relatively independent of the degree of incore axial offset. This trend was as expected since quadrant power

! tilt theoretically is not dependent on the magnitude of the incore axial offset. ,

The results of this study for the 40 and 75% full power plateaus are shown in .y i

Figure 4.4-6.

AXIAL POWER IMBALANCE i

Results from the Standard Core Power Distributions taken at 40 and 75% FP during the performance of the Power Imbalance Detector Correlation Test, show that the imbalance trip envelope (Figure 4.4-1) of the reactor protective l Systen is sufficient to protect the unit from exceeding the DNBR and the LHR limits under all core imbalance conditions when a gain factor of 3.90 is set into the delta flux amplifier. In addition, analyses indicate that the largest l thermal margins (measured by DNBR and LHR) exist when a negative 5.0 to 10.0 incore axial offset is present.

During the startup test program, the calculation of axial core imbalance was performed by the unit computer using signals from the 364 incore detectors. A minimum and maximum core imbalance value over the test period of -29.80 and

+5.91% FP was .obs uved during the performance of Power Imbalance Detector Correlation Test. It should be noted the core preferred to have a negative imbalance and achieving positive imbalances of any magnitude was difficult.

The core imbalances measured in conjunction with the core povoa distributions  !

of this section are shown in Table 4.4-1. These results confirm that under normal operating rod configurations no difficulty in maintaining an approximate zero percent full power imbalance was observed. .

l o

4.4-6

i 4.

4.4 CONCLUSION

S Four normal operating equilibrium xenon core power distributions taken at 15, 40, 75, and 100 percent full power under various core rod positions and incore axial imbalance were examined. The maximum radial and total peaking factors were not more than 5.0 and 7.5 percent greater than those predicted by PDQ-7.

Comparison between measured and predicted radial and total core power distribu-tions as expressed in peaking factors on a 1/8 core power distribution showed favorable agreement on all core locations.

The worst case minimum DNBR and mari== LHR measured as part of this test were subject to various types of analyses. On the bases of this study, the following were determined:

(a) All measured worst case minimum DNBR and maximum LHR were within their respective acceptance criteria of 1.30 and 18.00 kW/ft.

(b) Acceptable core conditions at the trip setpoint of the next power level of escalation were verified prior to escalating reactor power.

(c) Margin analysis on the minimum DNBR and maximum LHR values at the LOC".

and design overpower limits resulted in substantial margins.

The results of the quadrant power tilt and axial power imbalance calculations for a variety of different core power distributions taken during the power escalation test program yield the following conclusions:

\

(a) All maximum quadrant power tilts determined during normal power operation were well within the Technical Specifications Limii: of 4%.

(b) The reactor protective system will provide sufficient protection against exceeding DNBR and LHR limits when the delta flux amplifier has a gain factor of 3.90.

i

. -4.4-7 1

-- . ~.

~ ,,

Surunary 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, 7.wd Core Boron Axial Maximum Thermal Maximum Ievel Burnup Conc. Imb. Tilt Date Time (7.FP) 1-5 6 7 8 (E FPD) (ppmB) (7.FP) (7.) DNBR LilR P (R) P(RXAl 02/01/77 0620 16.20 100 94 19 20 0.28 1155 +1.47 +0.53 21.64 2.23 1.42 1.92 03/01/77 0206 40.21 100 95 17 18 1.56 1031 -0.82 +0.94 10.13 4.44 1.40 1.62 03/15/77 2230 73.94 100 97 18 08 7.62 960 +0.61 +0.74 4.66 8.93 1.36 1.79 04/03/77 1240 99.65 100 86 10 08 18.27 908 -0.91 +1.60 3.03 12.70 1.37 1.83 5

0

~

_. .___ . _ _ _ . _ _ . _ . _ . ~ _ . _ _ _ . _ _ - . . _ _ . _ . . .._

Measured Core Power Distribution Results At 15 % Full Power s

, +

Control Rod Group Positions Gps 1-5 100 % wd GP 7 19 % wd Gp 6 - 94 % vd GP 8 20 % wd Core Power Level 16.2 % FP Boron Concentration 1155 ppmB Core Burnup E EFPD Axial Imbalance +1.47 % FP i Xenon Conditions

, Equilibrium Conc. _Yy,3_ Yes or No Reactivity Worth -g % ak/k ,

Max Quadrant Tilt q%

1/ 8 Core Incore Weigh ting Fuel Assy. De tector Pmax/ P/P Fuel ':

rnenH nn Number Factor go3en Assembly  !

H-08 1 1 1.13 0.92 G-08 2 4 1.60 1.27 F-08 4 4 1.71 1.27 E-08 10 . 4 1.92 1.1A D-0 8 14 4 1.70 1.23

C-08 21 4 1.76 1.31 B-0 8 30 4 1.85 1.42 A-08 37 4 1.19 0.92 '

G-09 3 4 1.64 1.2a i F-10 12 4 1.76 1.97 l E-11 26 4 1.54 1.07 l D-12 41 4 1.32 0.93 -

C-13 52 4 0.73 0.52 F-09 6 8 1.87 1.37 E-09 5 8 1.70 1.23 l D-09 15 8 1.81 1.29 _

C-09 29 8 1.48 1.09 B-09 31

{

8 1.36 1.05  !

A-09 45 8 1.01 0.78 E-10 17 8 1.85 1.30 D-10 27 8 1.52 1.30 C-10 28 8 1.34 0.98 B-10 44 8 0.82 0.66 A-10 46 8 0.55 0.46 D-11 33 8 1.59 1.11 C-11 42 8 3.93 n_g7 B-ll 49 8 0.82 0.64 '

C-12 48 8 1.11 n.A9 s-12 31 5 0.61 0.46

)

Table ' 4.4-2

-4 .

Measured Core Power Distribution Results At 40 % Full Power Control Rod Group Positions Gps 1-5 100 % wd GP 7 17 % vd Gp 6 95 % vd GP 8 18 % wd Core Power Level 40.2 % FP Boron Concentration 1031.ppmB Core Burnup 1.56 EFPD Axial Imbalance -0. 82 % FP Xenon Conditions Equilibrium Conc. YES Yes or No Reactivity Worth -M7 % ak/k Max Quadrant Tilt +0. 94 1/ 8 Core Incoce Weigh ting Pmax/ P/P Fuel Fuel Assy. Dete ctor Factor Assembly i nc s nn Number JcoE Ec H-08 1 1 1.36 0.96 G-08 2 4 1.46 1.26 F-08 4 4 1.50 1.27 E-0 8 10 4 1.61 1.33 D-08 14 4 1.53 1.23 ,,,

C-08 21 4 1.56 1.31 l l B-08 30 4 1.62 1.40 l A-08 37 4 1.10 0.94 G-09 3 4 1.46 1.19 F-10 12 4 1.56 1.26 E-ll 26 4 1.37 1.07 D-12 41 4 1.22 0.96 C-13 52 4 0.69 0.59 F-09 6 8 1 e 1. u E-09 5 8 1.50 1.22 D-09 15 8 1.59 1.26 C-09 29 8 1.34 1.10 B-09 31 8 1.21 1.06 A-09 45 8 0.92 0.79 E-10 17 8 1.62 1.26 D-10 27 8 1.36 0.98 C-10 28 8 1.22 0.98 B-10 44 8 1.01 0.67 A-10 46 8 0.60 0.49 D-11 33 8 1.39 1.10 C-11 42 8 1 19 n Ro B-ll 49 8 0.78 0.66 C-12 48 l 8 1.04 0.84 l 7 B-lZ 31 5 0.59 0.49 l l

Table 4.4-3  !

. . . , _ . - - -~ ~ j

Measured Core Power Distribution Res}u'lts At 75 % Full Power Control Rod Group Positions

! Gps 1-5 100 % wd GP 7 18 % vd Gp 6 97 % vd GP 8 08 % vd Core Power Level 73.9 % FP

, Boron Concentration 960 ppmB Core Burnup 7. 62 EFPD Axial Imbalance +0. 61 % FP Xenon Conditions i

Equilibrium Conc. YES Yes or No Reactivity Worth -2.57 % Ak/k Max Quadrant Tilt +0.74%

1/8 Core Incore We2.gh ting Pmax/ P/) Fuel Fuel Assy. De tector Factor 7 core Assembly T.n e n ei nn Number E cal H-08 1 1 1.21 0.93 G-08 2 4 1.58 1.25 F-08 4 4 1.67 1.26 E-08 10 4 1.77 1.31 D-0 8 14 4 1.79 1.26 C-08 21 4 1.73 1.33 B-08 30 4 1.73 1.36 ',

. A-08 37 4 1.20 0.92 G-09 3 4 1.54 1.22 F-10 12 4 1.66 1.28 E-11 26 4 1.59 1.15 D-12 41 4 1.35 1.01 C-13 52 4 0.76 0.58 F-09 6 8 1.73 1.11 E-09 5 8 1.65 1.25 D-09 15 8 1.70 1.25 l C-09 29 8 1.42 1.11 B-09 31 8 1.29 1.03 A-09 45 8 0.94 u.75 E-10 17 e 1.76 1.31 D-10 27 8 1.45 0.98 C-10 28 8 1.36 1.03 ,

B-10 44 8 0.82 0.62 l A-10 46 8 0.60 0.47 D-11 33 8 1.50 1.09 l

C-11 42 8 1.22 0.89 l B-11 49 8 0.85 0.65 C-12 48 8 1.11 0.82 5-1Z 31 5 0.63 0.48

~

Table 4.4-4f L L ________ r_TTT~~'~~ ~~ T Y TT -~ ~ ~

Measured Core Power ' Distribution Results At 100 % Full Power Control Rod Group Positions Gps 1-5 100 % wd GP 7 10 % wd Gp 6 g6 % vd GP 8 8 % vd '

Core Power Level 99.80% FP Boron Concentration 908 PPmB Core Burnup 18. 27 EFPD Axial Imbalance -0. 91 % FP ,

Xenon Conditions  !

Equilibrium Conc. YES Yes or No Reactivity Worth -2.75 % ok/k

, Max Quadrant Tilt +1; 60 %

1/8 Core Incore Weigh ting Pmax/ P/P Fuel Fuel As sy. De tector ~i tnr ca n , Number Factor goryen Assembly H-08 1 1 0.92 1.27 G-0 8 2 4 l'.28 1.64 F-08 4 4 1.26 1.69 E-0 8 10 4 1.34 1.83 D-0 8 14 4 1.29 1.82 C-08 21 4 1.34 1.76 r

B-0 8 30 4 1.33 1.69 I A-08 37 4 0.88 1.17 G-09 3 4 1.20 1.52 F-10 12 4 1.30 1.69  !

E-11 26 4 1.17 1.65 D-12 41 4 1.01 1.37 C-13 52 4 0.58 0.76  !

F-09 6 8 1.31 '1.74 I E-09 5 8 1.27 1.68 D-09 15 8 1.37 1.81 C-09 29 8 1.11 1.48 B-09 31 8 1.02 1.25 A-09 45 8 0.71 0.92 E-10 17 8 1.33 1.79 D-10 27 8 0.98 1.50 C-10 28 8 0.96 1.28 B-10 44 8 0.63 0.73 A-10 46 8 0.46 0.60 D-11 33 8 1.07 1.49 .

C-11 42 8 0.90 1.17 B-11 49 8 0.64 0.86 C-12 48 8 0.81 1.10 B-12 31 6

! 0.47 0.63 Table 4.4-5 l

l l

5

i.  !

l Comparison Of Measured And Predicted Maximum Radial And Total Peaking Factors f'

Maximum Predicted Maximum Heasured Case Power Peaking Factors Peaking Factors Percentage Error Number Imvel Radial Total Radial Total Radial Total (dim) (%FP) (dim) __(dim) (dim) (dim) (%) (%)

[

1 15 1.49 2.03 1.42 1.92 -4.93 -5.73 {

< \

l 2 40 1.41 1.74 1.40 1.62 -0.71 -7.41 '

i l 3 75 1.41 1.74 1.36 1.79 -3.68 +2.79 -

, 4 100 1.46 1.85 1.37 1.83 -6.50 -1.09 ,

~

,y cr l E-t i NOTE: Percentage Error = ' Measured - Predicted" x 100 Measured ,

NOTE
The acceptance criteria requires that the radial and total peaking factors are not more than 5.0 and 7.5 percent greatcr than predicted, respectively.

l l

i . ,

Minimum DNBR And Maximum LilR Analysis For Core Power Distribution Test Data Analysis Power Incore Axial Peak Axial Peak Extrapolated W rat Case Worst Case Level Offset Magnitud* tocation Imbalance Maximum LHR Ninimum Dalaa N'"I . (IFP) (1) (None) (Segment) (2 pp) (kW/ ft) (None)

Measured Point 3.6.20 + 9.07 1.35 '

5 + 1.47 2.23 21.64 Overpower Trip Setpoint *0.00 + 4.54 6.68 5.98 1DCA Limit 102.00 + 9.25 14.04 2.55 Design Overpower Limit 112.00 +10.16 15.42 2.00 Measured Point 40.21 - 2.04 1.17 5 - 0.82 4.44 10.13 Overpower Trip setpoint 85.00 - 1.73 9.39 4.45 IACA Limit 102.00 - 2.09 11.26 3.75 Design overpower Limit 112.00 .2.28 12.37 3.35 (Y

< o' e .

O Nessured Point 73.94 + 0.82 1.42 5 + 0.61 8.39 4.66 ,

p Overpower Trip Setpoint 104.70 + 0.86 11.88 2.70

  • IACA Limit 102.00 + 0.84 11.57 2.89 i

Desi8n overpower Limit 112.00 + 0.92 12.71 2.39 i

I Measured Point 99.65 - 0.91 1.37 5 - 0.91 12.70 3.03 Overpower Trip Setpoint 104.70 - 0.95 13.34 2.72 LOCA Lielt 102.00 - 0.93 13.00 2.82 Design Overpower Limit 112.00 - 1.02 14.27 2.42

-q s 9

i

Reactor Protective System Maximum Allowable Envelope For Four Pump Operation

___=

5

-100-  :~: .'.

/ _

_/ I' .

-m- -l m.

H 3m.- ? "

s 3 -- 53.

u o _g -

m

_': 0 .

__*^-

_ = -

= ..

=

10 - ~- E -30;-20 -10  :- +1s --2 0 - 20 -d e --se. _

~

Power Imbalance, %FP Figure 4.4-1

Comparison Of Measured Radial Core Power Distributions Obtai id During The I Performance of Standard Core Power Distributions Test As Taken Along The X-Z Plane At ' Row 08 [

2.1 2.1  !

21.8 i d 1.8 s 1.5 1.5  ;

i

    • M M 1.2 k 1.2 4 " " ' 2M dk g 0.f \ \( h g 0.d \! i m

s.

g 0.6 0.6

-4 Power Level

=

16.20 2 FP IPower Level 0.3 = 40.21 I FP

} 0.3 Core Burnup = . 0.28 EFPD - Core Burnup = 1.56 EFFD "

0.0 *

!  ! 0.0  !  !  !  !  !  !  !

A B C D E F C K L M N O 11 P R A B C D E F G II K L M N O P R ,

y Radial Position. Fuel Assemblies Radial Position. Fuel Assemblice  !,

u i

u ,

2.1 2.1 I d 1.8 im so.d1.8

% 4 1.5 N4 1.5

".2 1 *d ' ^ -- **

1.2 N '" "b dW' #"

h h g 0.d i g 0.926

. 46  ;

A. . A.

a 0.6 s 0.6 M [ Power Level = 73.96 I FP - d Power Level = 99.65 2 FP

}0.3 Core Burnity

= 7.62 EFPD 0.3 Core Burnup = 18.27 EFFD

~

'i O.0  !  !  !  !  !  ! 0.0 I I I I I  ! I A B C D E F G H K L M N O A i

P R B C D E F C II K L M N O P R '

Radial Position. Fuel Assemblica Radial Position. Fuel Aoucablies

. L L, .

._ )

Comparison Of Predicted And Measured Radial Core Power Distribution Results At Steady State, 3-D Equilibrium Xenon, 40 7.FP Conditions Predicted Conditions I

Control Rod Group Positions Core Power Level 40.0 %FP i Gas 1-5 100.0 %. wd Boron Concentration NA ppmB GP 6 95.8 % wd Core Burnup 4.0 EFPD I Gp 7 20.8 % vd Axial Imba . mee +0.24 %FP l Cp 8 29.2 % wd Max Quadran: Tilt 0.00 % 1 Measured Conditions Control Rod Group Positions Core ?ower Lnel 40.2 %FP Gps 1-5 700  % wd Boron Concentration 1031 ppmB GP 6 os  % wd Core Burnup 1.56 EFPD Gp 7 17  % wd Axial Imbalance -0.82 %R Gp 8 is  % wd Max Quadrant Tilt +0.94  %

H G F E D C B A 8

1.01 1.25 1.22 1.41 116[,L.2.8 1.33 0.01 g 0.S6 1.26 1.27 1.33 1.23 1.31 1.40 0.94

.1$ 1.41 1.20 1.28 1.01 0.97 0.76 9 1. 1.35 1.22 1.26 1.10 1.06 8

0.79 10 f1 1.33 0.97' O.97 0.71 0.49

1. 1.26 0.98 0.98 0.67 0.49 e s s n l'

'.9 1.14 0.83 0.72 1.0 1.10 0.89 0.66 12

0. 0 0.89 0.55 l 0.9 0.84 0.49 0.'63 1

0.5h N .

14 15 b

I Control Rod Group (5-8)

X.XX Predicted Results X.XX Massured Results Figure 4.4-3

Comparison Of Predicted And Measured Total Peaking ' Core Power Distribucion Results At Steady State, 3_D Equilibrium Xenon, 40 7.FP Conditio,ns _

7 4

Predicted Conditions Centrol Rod Group Positions Core Power I4 vel 40.0 %FP Gps 1-5 100.0 % wd Boron Concentration NA ppm 3

Gp 6 95.8 % vd Core Burnup 4.0 EF7D I GP 7 20.8 % wd Axial Imbalance +0.24 %77 Gp 8 29.2 % wd Max Quadrant Tilt 0.00 %

l Measured Conditions Control Rod Group Positions Core Power Level 40.2 %FP Cps 1-5 100 % wd Boron Concentration 1031 ppmB Cp 6 95  % wd Core Burnup 1.56 EFFD Gy 7 17  % vd Axial Imbalance -0.82 %FP Gp 8 18  % wd Max Quadrant Tilt +0.94 %

I 1

H G F E D C B A 8 1.35 1.43 1.47 1.74 1.45' M4 1.55 10fl _g 1.8 1.46 1.50 1.61 1.53 1.56 1.62 1.10 9

l'. 8 ' 1.71 1.50 f.62 1.22 1.14 0.91 8

1.4 1.62 1.50 1.59 1.34 1.21 0.92

} 51 1.71 1.34 8 1.18 1.04 7 0.61

].

10 1. 1.62 1.36 1.22 ' 01

, 0.60 7~s l'. 1 , 1.47 1.01 LO.66 11 1. 1.39 1.12 0.78 i 1. 3 1.08 0.66 12 1.2 1.04 0.59 N

0.'77 i

0.6k '

l  !

3 l 14 l

l 15 3 Note: The results represent the maximum b -

total peaking in each assembly ,

1 l

f I control Rod Group (5-8) l l X.XX Predicted Results '

X.XX Measured Resuits j Figure 4.4-4

Hot Channel Minimum DNBR Versus Core Power Level

_=g __w._, ~ =b ~

___m _

w ~ ~ ~

_ Measured Val'ues Calculated--_--

~~By The Unit Computer 0 --- ~ ~~

~_

r

=._= .

=w ' '

=.

(= -

1:-

_w g_ __g_

z A__..,_ i _

e 2

ve. = s- _-

- = -

E '-

i y - n. , ,_ -

x -

g _ .-

c .

e _.w _

.c_y,- -

g._.---__ _

a -~ _ ._

@ ==rg- - - -

'_ i-

-4 ___

. = - -

= G_.: ' ' .'__.

. .=>-e-g_ .

-w w  %- -

T-

_w- --

.- m , ___ _ _ . . . . _ . . . _ . .

P_ ' I NK .9 S&en---7M- moo ==un mm- J-6 .a + s Core Power Level, 7. Full Power Figure 4.4-5

Maximum Quadrant Power Tilt Versus Full Incore Offset During The Performance Of Power Imbalance Detector Correlation Test m, g -

Technical Specification Limit +4%

ea Wr _.

u

T w a 0%

4- r.-_-.

g it- 5 _r_  ; .__

m y O _

~

E ./

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4.5 UNIT LOSS OF ELECTRICAL LOAD TEST

/

4.5.1 PURPOSE The purpose of this test was to ensure that the primary / sect.ndary systems can safely undergo a net load loss at 100% full power. This was accomplished by verifying the automatic response of the reactor and auxiliary systems during a generator separation from the transmission system.

Six acceptance criteria are specified for the Unit Loss Of Electrical Load Test, and are listed below:

(1) Unic 3 auxiliary transformer supplies power to the 6900 volt reactor auxiliary busses 3A and 3B, the 4160 volt unit busses 3A and 3B, and associated motor control centers.

(2) Automatte power reduction to 15% full power at a nominal rate of 20% full power per minute and less than a maximum rate of 50 %FP/ min. "

(3) The unit frequency will peak at less than the overspeed trip point and decay back to a set frequency in 40 to 50 seconds.

(4) The loss of electrical load does not result in fuel damage (leakage).

No increase in RC coolant letdown activity as measured by RM-Ll.

(5) Reactor coolanc pressure / temperature relationship must remain within the protective system envelope of Figure 4.5-1 (6) Minimum total feedwater flow does not decrease belaw 5.0 percent of the ,

normal value (10.2 MPPH). '

i 4.5.2 TEST METHOD I l

This test in m1ved separating unit 3 generator from the line and measuring the system responoe during the ensuing transient. The turbine-generator output '

should drop to house load and continue to supply unit power. However, the generator separation might also cause the reactor / turbine to trip. In any case, the primary intent was to verify that the unit can withstand a loss of electrical 1 load.

, This test was conducted from 100 percent full power as scheduled by the power escalation test procedure.

4.5.3 EVALUlTION OF TEST RESULTS On April 22, 1977 at approximately 2000, Unit Loss Of Electrical Load Test was performed by tripping the generator. output breakers. Upon the separation, the integrated control system smoothly ran reactor power back from 100 to 15%

full power within 5.0 minutes at an average ramp rate of 16.8% FP/ Min. As shown in Figures 4.5-2 and 4.5-3, excellent control of primary and secondary parameters during the transient was observed. In all cases, largr margins to RPS trip setpoints were experienced. After reaching 15% full power, the unit

' continued to supply power in the fully integrated mode of control until hand  !

control of the turbine was taken at 7.8 minutes into the transient.

4.5-1

During the performance of this test, some difficulty was experienced which should be noted. These areas are listed below: 3 (1) During the reactor runback, the low load feedwater block valve stuck at approximately 50% shut. This caused a moderate loop temperature mismatch to occur around five minutes into the transient.

'(2) The turbine experienced speed oscill.itions prior to and after reaching 15% full power. The magnitude of the turbine speed variations was approximately + 40 RPM.

Even though the above did not affect the successful runback of the unit, the second problem did cause the acceptance criteria on unit frequency control not to be met, since .it did not decay back to a set frequency in 40 to 50 seconds.

Further investigation into this difficulty revealed that the frequency oscil-lation magnitude was small and did not constitute a hazard to the operating equipment on the unit.

The results of the test when compared to the acceptance criteria =4n4== and ,

maximum parameter. tolerances are shown in Table 4.5-1 which also gives the mini-mum and maximum deviation in primary and secondary unit parameters both prior to and after the trip. As can be seen, all measured values fell within their allowable limits. An additional acceptance criteria was that the trend in plant fission product activity from RM-L1 did not increase.' Indication on

' RM-L1 (i.e. RCS letdown) revealed a before and after count rate of 3.4 x 103 cpm. These results confirm that no change in the activity of the coolant system (i.e. no fuel failure) was observed. J s i 4.

5.4 CONCLUSION

S l Upon completion of The Unit Loss Of Electrical Load Test at 100% full power, the following conclusions were reached:

(1) The unit successfully sustained a net load loss at 100% full power with-out any fuel or equipment damage.

(2) Unit response to the load loss indicates that the integrated control I system successfully ran reactor power tack to 15% full power with large I margin to RPS trip setpoints. l 1

(3) Frequency oscillations were experienced after the runback to 15 % full power, thus causing the acceptance criteria on frequency control not to be met. Further investigation resulted in the conclusion that since the oscillation magnitude observed was small it did not represent a hazard 4

to the operating equipment.

l I

l 4.5-2

._ _ , 4 m.

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g MS Cenerator A Imvel Inches 142.3 145.8 30.2 149.9 w

08 HS Cenerator B Imvel Inches 147.1 149.1 24.5 147.6 c.

u HS Cenerator A Temperatura 'F 593.1 594.4 581.2 601.5 I

H MS Cenerator 5 Temperature 'F 593.4 593.6 584.6 600.5 MS Cenerator A Pressure PSIC 911.7 913.0 837.9 1049.2 MS Cenerator B Fressure FSic 914,9 916.4 835.5 1051.8 IW Isop A Temperature 'F 447.8 448.5 410.1 448.2 FDW Imop B Temperature *F 446.0 446.4 389.5 446.1 FDW Total Flow MPPli 10.1 13.2 0.7 10,2 0.5 Turbine Hender Presourc FSIG 896.4 901.2 835.2 1058.6 Turbine Speed RPM 1796.0 1798.0 1732.0 1928.0 Note: T'he definition of (NR) is not recorded

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, 4.6 TURBINE / REACTOR TRIP TEST i

The Turbine / Reactor Trip Test consisted of two distinct parts: a reactor trip with subsequent turbine trip at 40 percent full power and turbine trips in which the reactor was not tripped at 75 and 100 percent full power. Its purpose was to measure the plant response during and after a deliberate reactor or turbine trip from power. In order to minimize the number of reactor and/or turt ine trips, unit performance data adequate for the analysis of this test was monitored continuously during power operation. An unscheduled trip might have been substituted for a scheduled trip. (This did not happen, however).

The Turbine / Reactor trip test provided data that was used to optimize the per-formance of the integrated control system and provide baseline data for compari-son with future performance data. In addition, it also provided data for use in evaluating the effectiveness of the main steam line snubbers during closure of the main turbine stop valves.

~ i 4.6.1 PURPOSE The general purpose of this test was to measure and evaluate the transient response of the unit during and following a reactor / turbine trip from 40% rated power and a turbine trip from 75 and 100% rated power. Specific purposes were as follows:

(a) Test pressurizer level control, reactor coolant pressure control and reactor coolant temperature control during a reactor or turbine trip.

(b) Test feedwater control and OTSG level control during a reactor or turbine trip.

(c) Test mMn steam pressure and feedwater temperature control during a reactor or turbine trip.

(d) Test main turbine response during a reactor or turbine trip.

The' acceptance criteria for the Turbine / Reactor Trip Test are divided into the individual acceptance criteria for the reactor trip and turbine trip as given in Table 4.6-1.

4.6.2 TEST METHOD The reactor trip portion of the test was performed at 40% full power by manually tripping the reactor from the control room console. Various unit parameters 1

were monitored using the Reactimeter and six Brt"h recorders throughout the i ensuing transient. Additional performance data was also obtained using the unit computer post trip review program which can printout data thirty minutes prior to and after such an occurrence. Specific data was tabulated as shown in Table l 4.6-2 and comparison to the acceptance criteria was performed. )

The turbine trip portion of the test was performed at 75 and 100% full power by i manually tripping the turbine from the control room console. Unit parameters were monitored during the transient in the same manner as for the reactor trip.

4.6-1 l l

-_r-_ _ ._. . . . _ . , , .: .

l l

l Specific data was then tabulated as shown in Tables 4.6-3 and 4.6-4 and compari-son to the acceptance criteria was performed. 3 During the performance of both parts of the test, the response of primary and secondary parameters were monitored and plotted to verify that the transient performance of the integrated control system to a major pertubation was accept- l able. This included evalua' ting the impact of the spray flow, BTU limits, safety valve settings and various other controlled quantities on the transient response. j l

4.6.3 EVALUATION OF TEST RESULTS The evaluation of the test results of the Turbine / Reactor Trip Test has been subdivided into three sections: one for the reactor trip test at 40% full power, one for the turbine trip test at 75% full power, and one for the turbine trip test at 100% full power. In each instance, all the acceptance criteria for this test were met as stated in Table 4.6-1.

4.6.3.1 REACTOR TRIP TEST AT 40% FULL POWER

~

On March 9, 1977 at approximately 1458, the Turbine / Reactor Trip Test was per-formed by tripping the reactor. This immediately caused an automatic turbine trip, which transferred house load to the startup transformer. However, during i this transfer, the feedwater pump tachometer experienced a momentary loss of l power causing a zero speed indication to the feedwater pump controller which tripped the pump. Since feedvater flow was lost, the control room operator put the unit in manual and took control to maintain pressures, temperatures, levels and flows. Thus the ability of the integrated control system to maintain ,

, the above quantities was not ascertained. To correct the above problem on '

future reactor trips, the power supply to each feedwater pump tachometer has been connected to a vital bus.

The response of the primary and secondary unit parameters both during and after the transient is shown in Figures 4.6-1 and 4.6-2. As can be seen, adequate -

control of the unit was maintained with no unstable conditions developing. Even though sufficient control of the unit was achieved, periodic lifting of safety valve MSV-33 at approximately 1000 Psig (i.e. 50 Psig below its setpoint) was contradicting the ability of the integrated control system turbine bypass valves to control main steam pressure at 1010 Psig. The status of all main safety valves during the test is given in Table 4.6-5. It should be noted that MSV-33 blev for 348 seconds, which had a strong effect on the primary system cooldown rate.

The results of the test were compared to the acceptance criteria minimum and maximum tolerances as shown in Table 4.6-2. Also presented are the minimum and maximum primary and secondary unit parameters both prior to and after the trip.

As can be seen, all measured values fell' within their allevable limits.

4.6.3.2 TURBINE TRIP TEST AT 75% FULL POWER on March 30, 1977 at approximately 0814, the Turbine / Reactor' Trip Test at 75%

full power was performed by manually tripping the turbina. The unit successfully ran reactor power back to 15% full power within 2.8 minutes after the trip with an average ramp rate of approximately 21%FP/ Min. As shown in Figures 4.6-3 and 4.6-4, excellent control of primary and secondary parameters during the transient

)

was experienced. After the reactor power had leveled out around 15% full power, 4.6-2

periodic oscillation in the primary side unit parameters at a two minute period were present. This, in part, has been attributed to the fact that reactor power is less than that recommended for the fully integrated control mode.

During the turbine trip, some difficulty was experienced when FWV-29 and MV-30 would not close. The control room operator took manual control and shut the block valves in 216 and 142 seconds, respectively. Since this performance did net meet the acceptance criteria, the torque and limit switches on these valves were adjusted. To ensure that this adjustment was adequate, the turbine was brought back on-line at 15% full power and tripped again. This time WV-29 and FWV-30 took 28.0 and 27.5 seconds, respectively, to close (i.e. time from a red light to green light indication using a stop watch).

The results of the test were compared to the acceptance criteria minimum and mari== tolerances as shown in Table 4.6-3. Also presented are the minimum and maximum primary and secondary unit parameters both prior to and after the trip.

As can be seen, all measured values fell within their allowable limits.

The above test results were also reviewed by the Babcock and Wilcox Control Analysis Section, so that improvements in the ICS transient performance could be incorporated prior to transient testing at 100% full power. As a result of this review, the ICS modifications listed in Table 4.6-6 were incorporated prior to the 100% turbine trip test.

4.6.3.3 TURBINE TRIP TEST AT 100% FULL POWER 1

On April 18, 1977 at approximately 1100, the Turbine / Reactor Trip Test at 100% .

full power was performed by manually tripping the turbine. The unit success-

, fully ran reactor power back to 15% full power within 4.0 minutes at an average ramp rate of approximately 21% FP/ Min. As shown in Figures 4.6-5 and 4.6-6, excellent control of primary and secondary parameters during the transient was observed. In all parameters, large margins from RPS. trip setpoints were experienced. After the reactor power had leveled out around 15% full powet periodic oscillations in the primary side unit parameters on a two minute pem _

were present. These oscillations were similar to those experienced during the j turbine trip test at 75% full power.

, The results of the test were compared to the acceptance criteria min 4-= and maximum tolerances as shown in Table 4.6-4. Also presented are the minimum sud maximum primary and secondary unit parameters both prior to and after the trip.

As can be seen, all reasured values fell within their allowable limits.

During the test, the closure times on the feedwater block valves FDW-29 and FDW-30 were also recorded using a stop watch. The results indicate that the closure times were less than 30.0 seconds.

4.

6.4 CONCLUSION

S Upon-completion of the reactor and turbine trip portions of Turbine / Reactor Trip Test at 40, 75, and 100% full power, the following conclusions were reached:

(o 4.6-3

(a) Acalysis of unit parameters during the transient indicates that all acceptance criteria were met. ,

(b) Unit response to a manual turbine trip indicated that the integrated control system can successfully run reactor power back to 15% full power with large margins to RPS trip setpoints.

(c) The manual reactor trip caused the turbine to trip which ensures that cooldown rates less than 1000F/hr can be maintained following reactor trips at power.

(d) After adjustments the feedwater block valves closure times were verified.

to be less than 30 seconds.

g J

4.6-4

~

'% r Susunary Of The Turbine / Reactor Trip . Test Acceptance Criteria i

f Acceptance Criteria Het

  • Critsria Reactor / Turbine Turbine only Turbano Only Number Procedural Acceptance criteria 40 I FP 75 % FP 100 2 FP 1 The reactor trip causes the turbine / generator to trip Yes NA NA 2 The reactor coolant pressure remains above 1800 psig and below 2355 peig Yes Yes Yes 3 The main steam pressure at the OTSC outlet must not exceed 1155 peig Yes Yes Yes 4 The final feedwater temperature remains above 1850F and below 470*F Yes Yes Yes H

@ 5 The pressuriser level does not 30 below 0 inches or above 320 inches Yes NA HA s

s- 6 The pressurizer level does not go below 40 inches or above 320 inches HA Yes Yes a

M 7 The turbine speed does not reach or exceed its overspeed trip point of 1998 RPM. Yes Yes Yes 8 The OTSG 1evel remains abovs 8 inches and below 384 inches Yes Yes Yes l

9 The reactor coolant pumps remain in operation Yes Yes Yes 10 Automatic high pressure injection is not initiated by the engineered safeguard system. Yes Yes yes 11 Main Feedwater block valve FDW-29 and FDW-30 close l within 30 seconds of receiving the closure signal NA Yes Yes Note: The definition of (NA) is not applicable i

e a

. p. _ _

1 Summary Of Minimum And Maximum Unit Parameters During The Performance of Turbine / Reactor Trip Test At 40% Full Power i

i i

l Test Data Frior To Trip Test Data After Trip Acceptance Criteria 1.imits

! Unit Parameters Units Minimum Hautmum Minimum Haulmum Minimum Maximum i

RCS Average Temperature *F 578.7 579.0 548.5 577.5 RCS Coolant Pressure P!!!C 2134.9 2153.6 1921.8 2153.6 1800.0 2355.0 f' .

I RCS Pressurizer level Inches 198.8 201.0 63.1 201.0 0.0 320.0  !

RCS Inop Temperature Hismatch "F -0.2 +0.2 -1.1 +6.4 I

4 RCS H.ekeup Tank level Inches 64.0 64.1 29.9 64.0 1 MS Cenerator A Inval -Inches 55.0 57.7 20.2 56.5 S.O 384.0 HS Cenerator 8 level Inches 53.7 56.0 12.2 55.1 8.0 384.0 g MS Cenerator A Temperature 'F 587.9 588.2 581.4 599.0 su HS Cenerator B Temperature *F 588.5 588.7 588.9 602.4 y MS Cenerator A Fressure PSIG 879.9 885.3 884.6 1040.8 1155.0 i

N HS Cenerator B Pressure PSIC 881.3 886.6 885.6 1021.2 1155.0 FDW toop A Temperature *F 367.8 368.0 226.4 367.9 185.0 4i0.0 f WW Isop B Temperature *F 365.9 366.5 358.1 366.4 185.0 470.0

. Turbine IIcador Pressure PSIG 884.8 890.3 889.1 1047.9 i

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Summary Of Minimum And Maximum Unit Parameters During The Performance Of Turbine / Reactor Trip Test At 75% Full Power Test Data Prior To Trip Test Data After Trip . Acceptance Criteria Limita Unit Parameters Unita Minimum Haximum Minimum Maximum Minimum Maximum ICS Average Temperature *F 579.5 579.5 576.8 591.0 i RCS Coolant Pressure PdIC . 2153.8 2158.9 1923.7 2320.3 1600.0 2355.0 RCS Pressurizar 1mvel . Inches 200.3 200.5 164.1 266.1 40.0 320.0 RCS teop Temperature Hismatch "F 0.0 40.3 -6.8 +3.3 RCS Hakeup Tank level Inches 62.3 62.3 48.5 64.8 HS Cenerator A Imval Inches 99.8 101.0 25.E 112.2 8.0 384.0 HS Cenerator B I4 vel Inches 101.0 101.9 75.8 90.0 8.0 384.0 MS Cenerator A Temper u ure *F 592.1 592.3 581.5 606.7 Y HS Cenerator 8 Tempe ature *F 594.2 594.3 389.8 609.7 -

cr o MS Cenerator A Pressure PSIC 891.9 893.8 859.0 1035.0 1155.0

  • HS Cenerator B Pressure PSIC 894.7 896.7 909.0 1055.0 1155.0 FDW Ioop A Temperature *F 423.7 423.7 305.3 423.7 185.0 470.0 o

FDW Ioop B Temperature F 422.6 422.6 339.5 422.5 185.0 470.0 Turbine Header Pressure PSIC 892.0 894.2 860.9 1076.1 Turbine Speed RPM 1794 1796 1085 1795 1998.0' I

Summary Of Minimum And Maximum Unit Parameters During The Performance Of Turbine / Reactor Trip Test At 100% Full Power i

~} .

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'l Test Data Frior To Trip Test Data After Trip Acceptance Criteria Limitas l

Unit Parameters Units Minimum Haximum Minimum Maximum Minimum

.{ Maximum t

'll RCS Average Temperature 'F 578.6 572.7 576.0 586.1 . l i  !

RCS Coolant Pressure FailG 2126.9 2133.9 1975.3 2280.0 1800.0 2355.0 lli i j RCS Fressuriser Imvel Inches 199.1 199.8 173.2 '240.5 40.0 320.0 .

i RCS toop Temperature Hismatch 'F -0.6 10.0 -3.7- +3.3 l o

RCS Hakeup Tank Inval Inches NR HR NR NR  ;

MS Cenerator A 1mvel Inche s 145.7 149.4 26.6 159.3 8.0 384.0 j HS Cenerator B Imvel Inctes 147.9 150.6 24.8 142.7 C.0 384.0 MS Cenerator A Temperature 'F 592.5 592.6 583.9 605.8 f N HS Cenerator B Temperat.are 'F 592.8 592.9 583.0 603.5 tr '

o MS Cenerator A Pressure PSIG 912.0 913.2 888.4 1049.1 1155.0 s~ ,

  • MS Cenorator B Fressure PSIC 916.4 917.5 899.2 1035.8 1155.0 m

S- FDW !aop A Temperature F 488.6 449.3 288.8 449.0 185.0 470.0 l FDW toop B Temperature *F 446.9 447.3 294.7 422.5 185.0 470.0 Turbine Header Pressure PSIG 897.4 901.1 883.8 1061.9 Turbine speed RPM 1794.0 1795.0 862.0 1794.0 1998.0  ;

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Main Steam Valves Status During The Performance Of Turbine / Reactor Trfp Test At 40, 75 And 100 % Full Power '{

Main Steam Safety Valve Orifice 40 % FP Plateau 75'I FP Plateau 100 2 FP Plateau i Valve Number Lift Setting Size Delay Time Lift Time Delay Time Lift Time Delay Time Litt Time (Pnnoi (Psie) (Inchen) (Secondn) (Seconds) (Secondo) (Secord.) (Sec onda) (Seconds)

Hain Steam Line Al MSV-34 1050 4.515 NA NA 10 245 6 180 MSV-38 1070 4.515 NA NA 11 185 6 180 MSV-43 1090 4.515 NA NA 11 130 6 180 HSV-40 1100 3.750 NA NA. NA NA NA NA Main Steam Line A2-MSV-33 1050

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240 a MSV-46 1100 4.515 NA NA NA NA NA NA o'

H 5

h8 Main Stesu Line B1 t,n HSV-35 1050 4.515 5 60 10 245 6 240 MSV-39 1070 '4.515 NA NA 11 185 6 240 MSV-44 1090 4.515 NA NA 11 130 6 240 MSV-47 1100 4.515 NA NA NA NA NA NA Main Steam Line B2 MSV-36 1050 4.515 5 60 10 245 6 210 MSV-41 1070 4.515 NA NA 11 185 6 210 HSV-45 1090 4.515 NA NA 11 130 6 j

210 MSV-48 1100 3.750 NA NA NA NA NA NA l

) Note: The definition of (NA) is not applicable Note (*): Actual setting was approximately 1000 psig

,f .

1 1

Summary Of ICS Modifications Reconumended As A Result Of Turbine Reactor Trip l Test At 75% FP. Modifications Incorporated Prior To 100% Trip Test.

.it

.3

'l Change Number

  • ICS Modifications Reason For Modification

!' 1. Feedwater pump speed kicker circuit was added to. To prevent a signif* cant reduction of feedwater flow I

] loop A and B FDW pumps. This increased feedwater to the steam generators immediately after tripping the I i pump speed about 20% in the first four esconds turbine or rejecting electrical load.

.! after tripping the turbine or a loss of electrical d load.

l I

j 2. Added the requirement that both atmospheric dump To reduce peak steam pressure after a trip. This '?

I valves open whenever steam pressure exceedes 1025 peig is important since peak steam and reactor coolant

, pressure are strongly related to each other. 2 H I u- 3. Added three back-to-back diode "deadband" filter To elistinate system " hunting" due to noisy error [

y

.; circuite signals.

, 4. Changed the characterization of the grid frequency To limit the large frequency error correction to the j m error signal to the unit load demand algnal so that the maximum step change to ill.D will only be + 10%

unit load demand signal.

.l

[  ;

not f 100% ,

5. Changed pressurizer spray valve logic so that the To assure quick response of the Pressurizer Spray valve receives an immediate "open" signal for five System to high reactor coolant pressure. ,

seconds after either a turbine trip or a loss of '

electrical load.  !

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4.7 INCORE DETECTOR TEST The incore monitoring system consists of 52 detector assemblies each containing 7 neutron detectors equally spaced axially. Of these 52 detector assemblies, all contain rhodium self-powered neutron detectors (SPND) with inconel lead wires. Each also contains one inconel lead wire detector without rhodium for use in background corrections. The power distribution within the core is measured at 364 locations by use of these neutron detectors (7 axial positions in the 52 detector assemblies). The outputs of the SPND's are connected to the unit computer which provides the unit operators with power distribution infor-mation. In addition, 48 detector signals are available in the control room on two 24 point recorders for backup in case the unit computer is not operational.

The incore monitoring system does not perform direct protective actions or direct control actions. -However, the incore Lastrumentation is used to cali-brate the top and botto 1 out-of-core power range detectors in terms of imbclance (power in top half of core minus power in bottom half of core). Furthermore,

, the incore instrumentation provides detailed power distribution data that could be used to alert the reactor operators to any adverse power distribution change or trend during operation, and to provide power distribution and fuel burnup data.

The incore monitoring assemblies are placed at preselected radial positions in the core as shown in Figure 4.7-1. Each assembly contains seven equally spaced flux detectors corresponding to seven axial elevations in the core to provide measurement of axial flux shape. As shown in Figure 4.7-1, 17 detector assem-blies are positioned to act as symmetry monit.m .; the remaining 35 assemblies with 5 of the 17 symmetry monitors, monitor other fuel assembly positions assuming core symmetry.

The unit computer utilizes a variety of correction factors which are used to correct the signal of the self-powered neutron detector. The computer provides readouts of the incore detector signals on typewriters and individual digital display.

In order to ascertain that the incore monitoring system and the Bailey 855 unit computer when coupled together were giving an accurate description of the distribution of core power, these systems were qualified during the power escalation test program by comparing certain computer calculated values with similar values calculated by hand. It was not expected that this comparison would give identical values since a number of simplifying assumptions were used in the hand calculation.

The incore monitor system and the Bailey 855 unit computer were qualified for both the 40 and 75% full power test plateau, with a preliminary qualification done at 15% full power. Once the unit c omputer was qualified, it was then used

.in lieu of hand calculations for testing at that test plateau. Similarily once .

it was qualified at 75% full power, it was then used for all testing thereaf ter.  !

The results of the incore detector testing done during the test program as part of the power escalation sequence are discussed by dividing this section into five subsections .as follows: l 1

(1) Incore Detector Checkout l l

4.7-1 1

(2) Incore Detector Signal Corrections m

(3)' Radial Core Power Distribution Comparison (4) Worst Case Maximum Linear Heat Rate (LHR) Calculations (5) Incore D= ,,ctor Response 4.7.1 PURPOSE-The general purpose of this test is to verify that the incore monitoring system and the Bailey 855 unit computer together provide an accurate description of the core power distribution. This is accompli",hed by fulfilling the following individual purposes: -

(1) To verify that the incore detector outputs are reasonable and to identify

any which have failed or have questionable outputs.

(2) To verify that the Dailey 855 unit computer is properly performing the necessary corrections to the incore detector signals.

(3) To verify that the Bailey 855 unit computer is calculating the radial core power distribution and worst case maximum LHR correctly.

Five acceptance criteria are specified for Incore Detector Test and are listed below:

(1). The hand and computer calculated values for the sensitivity, depletion, ./

and background corrected signals shall agree within i 2 percent.

(2) The hand and computer calculated values for the segment powers shall agree within i 3 percent.

(3) The hand and computer calculated values for the radial peaking factors shall agree within i 5 percent.

(4) The hand calculated worst case LHR is less than or equal to the computer worst case IllR.

i (5) Any incore detectors which have failed or are questionable have been j identified, located, and flagged for further evaluatioc. l 4.7.2 TEST METHOD l

The Incore Detector Test was performed at the 40 and 75% full power test plateaus concurrently with the core power distribution test. After three-dimensiona.' .

equilibrium xenon was established, data was obtained using the performance < ata I output program on the unit computer. This included a listing of the following:

- (1) Uncorrected-detector signals (2) Sensitivity, depletion, and background corrected detector signals ,

. (3) - Enrichment, rodded /unrodded, and LBP corrected segment power

.]

4.7-2

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(4) Radial core power distribution (5) Thermal hydraulic data This data was then used in a hand calculation to verify that the computer values for corrected SPND, segment power, radial power distribution and worst case MLHR was being performed satisfactory. To reduce analysis time, hand calcula-tions were performed on a representative sample of four incore monitor assemblies (i.e.1, 7, 35 and 52) when checking the ability of the unit computer to calcu-late the various corrections.

4.7.3 EVALUATION OF TEST RESULTS 4.7.3.1 INCORE DETECTOR CHErKOUT Verification of proper incore detector output was accomplished by comparing the uncorrected detector output and the background detector output of detectors located in similar locations. Since the shape of the curve was the variable -

being compared, all detector outputs were normalized to the average detector output per assembly. Tables 4.7-1 and 4.7-? show the results of this compari-son at the 40 and 75% full power test plateau, respectively. In all cases, similar flux shapes and normalized background detector outputs for the indica-ted grouping were obtained, except for IC (17) and IC (32) BG which both had lower values than expected at the 75% full power test plateau. An investiga-tion on these two detectors were performed with the following conclusions reached:

(1) IC (17) level 7 -- This detector was found to be acceptable and thus was not locked out of the computer. The lower than expected inciation was the result of comparing uncorrected level 7 SPND's which inherently have large background corrections.

(2) IC (32) BG -- This background detector was locked out of the computer.

The lower than expected indication is believed to be cable related. How-ever, no definite conclusion was reached during the test program.

A comparison was also performed between 52 background detector outputs for consistency. The results of this review indicated that the background detectors were relatively constant at an average ratio of background signal to average assembly rhodium signal of 0.17 and a standard error of 0.03.

4.7.3.2 INCORE DETECTOR SIGNAL CORRECIIONS  ;

The unit computer utilizes a variety of correction factors which are applied to correct the signal from the self-power neutron detectors. During the power escalation test program, hand calculations were performed to verify that the computer was performing these corrections as required. This section covers the results of this investigation. 1 SENSITIVITY, DEPLETION. AND BACKGROUND CORRECIIONS I To verify proper operation of the unit computer in performing sensitivity, depletion, and background corrections to the rhodium incore detectors, the un-corrected detector outputs and the background detector outputs along with the 4.7-3 ,

~w a individual detector sense and charge factors were used to calculate the cor-rected signals for incore strings 1, 7, 35 and 52. These results were then w compared to the corrected SPND values as calculated by the unit computer and a percentage deviation determined. Tab'ss 4.7-3 and 4.7-4 show the results of these comparisons at 40 and 75% full power, respectively. As can be seen, good agreement between the computer and hand calculated values were obtained with all percentage deviat1~on within the acceptance criteria limit of i 2.00 percent. The maximum difference observed was +1.92 percent at the 40% full power test plateau.

ENRICHMENT RODDED /UNRODDED, AND LBP CORRECTIONS During the performance of the Incore Detector Test, hand calculations were per-formed using the computer sensitivity, depletion, and background corrected detector signals and the fuel enrichment, rodded /unrodded, and LBP correction factors given in Table 4.7-5 to generate the segment power for 28 detectors located in incore strings 1, 7, 35, and 52. These particular detectors were chosen because they constituted a representative sample for the corrections being applied. The equation used to perform the above correction is given below:

P i

=

St XF AXF XF t 2 XF7 XF8 EQ. (4.7-1)

Where: P i =

Segment power at level (1), MWt Si =

Corrected detector output at level (1), Amps

= Fit factor, assumed to be 1.000 at BOL FA F1 = - Megawatt thermal to current conversion factor F2

=

Fuel enrichment correction factor /

F7

=

Non rodded / rodded correction factor Fg =. Lump burnable poison correction factor The hand and computer calculated values at 40 and 75% full power were then

. compared as shown in Tables 4.7-6 and 4.7-7, respectively. As can be seen, good agreement between the computer and hand calculated values were obtained with most values meeting the acceptance criteria limit of 13.00 percent. The maximum difference observed,was -3.02 percent at the 75% full power test plateau.

It is important to note the results show a consistent error indicative of the approximate factor used in the hand calculation plus a random error well within the allowable band. - Since the segment powers are later normalized to the heat balance, this consistent error is eliminated.

INSULATION LEAKAGE CORRECTIONS The f- tlation leakage correction factor is Lacorporated to account for the det .or current which leaks to the sheath of the detector assembly instead of being_ transmitted to the computer. To ensure that insulation leakages were small and negligible, direct measurements were performed to determine these factors on various detectors in incore strings 9, 11, 17, 19 and 23. These had low resistances during the pre-operational incore detector checkout. The equation used to calculate the insulation leakage factor for each detector is given below:

3

)

-4.7-4

9 F

L

1 + R X g X 10 EQ. (4. 7-2) av Where: Fg = Insulation leakage factor R = Resistance, 20 X 10 3 Ohms AI

Delta Current (I,- I_), Amps AV = Delta Voltage, 20 Volts l

The results of f:he above investigation revealed a maximum measured value of i 1.002 on the selected group of detectors measured. However, these results do  !

imply that the computer insulation leakage factors assumed to be 1.00 at BOL  !

have so far adequately described this correction. i 4.7.3.3 RLIAL CORE POWER DISTRIBUTION C0" PARIS 0N To confirm that the unit computer was calculating the radial core power distri- -\

t bution correctly, hand calculation was done using segment powers (i.e. SPNDAA I values) from the unit computer. The calculational method was to assum2 that l 1/8 core symmetry existed which allowed the analysis to be performed using 29 I characteristic fuel assemblies. From the hand and computer calculated data, the i percentage deviation between the two methods was then determined for comparison 1 to the acceptance criteria. Figure 4.7-2 shows the percentage deviation observ-ed during performance of this test at 40 and 75% full power. As can be seen, the maximum deviations at 40 and 75% full power were +0.70 and +0.53%, respective-ly, which is well within the acceptance criteria limit of + 5.00%.

4.7.3.4 WORST CASE MAXIMUM LHR CALCULATIONS As part of the unit computer checkout, hand eniculations for determining the worst case maximum LHR were also performed in conjunction with the above 1/8 core analysis. After selection of the fuel assembly which produced the maximum total peaking factor, the worst case maximum LHR was then determined and compared to the value generated by the unit computer. The equation utilized to perform the hand calculation was based on using the maximum LHR times the worst case uncertainty factor as shown below:

WC MLHR = WCF X MLHR EQ. (4.7-3) where the definition of the above terms is expressed by the following two equations:

P MLHR =

R x A x Q Rate x FNT x FOP EQ. (4.7-4)

NA x NP x AL where: MLHR =

Mav4== Linear Heat Rate (kW/ft)

P RxA

= Total Peaking Factor

=

Q Rate Rated Core Thermal Power (2452 x 103 kW)

FNT =

Fraction of power generated in,the fuel (0.973)

IUP =

Fraction of power level (power in percent /100)

NA =

Number of fuel assemblies in the core (177)

NP =

Number of fuel pins in each fuel assembly (208)

AL =

Active length of each fuel pin (12 ft.)_

4.7-5

Us . -i.,.. ., ' . e _y -ww m ,u -

> WCF =

FDxPgxPg x FQ" x PS x Up xU N EQ. (4.7 'i)

N Where: WCF = Worst Case Factor FD = Fuel Densification Factor (1.021)

Pg = Radial Local Peaking Factor (1.050, 0.00% B 4C)

(1.082,1.01% B C) 4

('. 098,1.18% B4c)

(1.094, 1.34% B 4C)

Pg = Axial Local Peaking Factor (1.026)

FQ" =

Hot . Channel Factor (1.014)

PS = Power Spike Factor -(1.075)

Up = Power Uncertainty Factor (1.020)

UN

= Nuclear Uncertainty Factor (1.075)

Table 4.7-8 presents the results of these hand calculated worst case maximum LHR for each case along with the unit computer worst case value. As can be

  • seen, good agreement was obtained between the hand and computer calculated

, values. However, the hand calculated values exceeded the computer calculated values which does not meet the acceptance criterion.

A review of the han'd and computer calculational methods has revealed that the

unit computer is slightly in error in its method of . determining the maximum

, total peaking factor. This error had been isolated to the two basic difficulties i listed below:

(1) The unit computer selects the fuel assembly with the highest radial ,

peaking factor. This fuel assembly however, may not contain the highest >

total peaking factor required in the calculation.

(2) The unit computer uses a segment averaged totel peaking factor instead of the maximum total peaking factor observed. Thus an additional factor i of 1.04 (i.e. average segment power to peak) should be incorporated in the computer value.

,No change to the unit computer is planned since the small error introduced by

. using the present output is overshadowed by - the 24 percent conservatism presently applied through various factors. However, as a result of this study,
Babcock and Wilcox has issued a revised method of obtaining the worst case LHR from the unit computer which corrects the, problems noted'above.

4.7.3.5 INCORE DETECTOR RESPONSE i

The incore' detectors have performed accordin's to design during the startup test-

! -ing. The' signal response from the detectors has been measured continuously since power operation began on January 29, 1977. The signal from the detectors is almost exactly proportional to power. Figure 4.7-3 is a plot of measured average,incore detector signal in nanoamps versus core power level.

'During the startup period a few detectors as measured by the unit computer had low or questionable readings either as a result of a bad detector or cable problems. The problems were worked on and by the end of the test period, there 3 remained'only two detectors with low and/or questionable outputs. These detectors /

are listed below:

I' 4.7-6

. #  :~~.- ,.o 3,,- w ar.e ,-...a, . , .~,,,,- n,n. p rav m. , e- w.n , .c-

(1) IC (32), BG (2) IC (49), BG In each case the Bailey 855 is substituting a background signal from another detector to be used in the calculation.

4.

7.4 CONCLUSION

S Upon completion of Incore Detector Testing at 40 and 75% full power, the following was concluded:

(1) The incore monitoring system has performed as expected during the startup period. Verification of flux shapes on symmetric grouping and incore detector response has been excellent.

(2) Computer corrections to the uncorrected detector signals were found to be consistent with hand calculations.

(3) The ability of the computer to determine the radial core power distribution was demonstrated.

(4) Comparison of computer and hand calculated worst case LHR indicates that the computer is slightly in error. No change to the computer is plannned since the error introduced by using the present output is overshadowed by the 24 percent conservatism presently applied.

+

e

4.7-7

Comparison Of Incore Honitoring Assemblies Flux Shapes At 40% FP

, Using The Indicated Grouping Civen In Incore Detector Test i

i j Incore Fuel Group Detector Assembly Detector Current To Average Detector Current Per Assembly l Numbe r Number Tocation Level 1 Level e Level 3 level 4 Ievel 5 Level 6 Imvel 7 (BC) 3 01 05 E-09 .815 1.041 1.036 1.185 1.234 1.101 .588 .143 ,

07 E-07 .835 1.075 1.040 1.159 1.236 1.071 .582 .147

' 09 C-05 .834 1.063 1.038 1.177 1.228 1.089 .571 .142 11 K-05 .815 1.076 1.030 1.166 1.256 1.082 .576 .156 13 H-07 .814 1.044 1.049 1.178 1.232 l'.087 .596 .164

! 16 H-09 .820 1.051 1.033 1.176 1.250 1.083 .586 .149 19 K-11 .827 1.048 1.012 1.171 1.267 1.088 .587 .145

25 G-11 .835 1.070 1.037 1.184 1.247 1.073 .554 .142 02 23 F-13 1.011 1.080 .911 1.114 1.225 1.080 .579 .156 35 F-03 .992 1.087 .904 1.111 1.228 1.078 .600 .165 g 39 L-03 1.015 1.089 .895 1.115 1.244 1.062 .579 .170

'l 50 L-13 1.031 1.077 .909 1.136 1.207 1.075 .565 .159 Q.

) y 28 C-10 1.023 1.050 .919 1.160 1.249 1.046 .553 177

'j , 32 C-06 1.008 1.104 .907 1.096 1.238 1.064 .585 .159 i

y 43 0-06 1.002 1.070 .913 1.136 1.242 1.077 .561 .169

,! 4 47 0-10 1.012 1.092 .901 1.129 1.225 1.042 .599 .173 1

!f 03 06 F-07 .852 1.081 1.087 1.159 1.207 1.049 .566 .166

'i

,1 08 G-06 .866 1.071 1.049 1.171 1.218 1.050 .574 .166 04 15 N-09 .883 1.038 .971 1.172 1.260 1.085 .587 .163 2

20 K- L2 .859 1.029 .979 1.175 1.253 1.097 .50P .168 lI i 05 17 H-10 .889 1.037 .963 1.180 1.284 1.101 .547 .186 3 18 L-11 .848 1.026 .955 1.173 1.283 1.119 .597 .171 l 06 22 G-13 .913 1.089 1.019 1.150 1.244 1.017 .567 .143

" .904 1.091 1.000 1.144 29 C-09 1.220 1.073 .568 .141

-i l

l e  ;

i t .

s Comparison Of Incore Honitoring Assemblies Flux Shapes At 40% FP Using The Indicated Grouping Given In Incore Detector Test incore Fuel Detector Current To Average Detector Current Per Assembly Group Detector Assembly Number Number Location Level 1 Level 2 Level 3 Level 4 level 5 I2 vel 6 revel 7 (BC) 07 24 F-12 .959 .959 .604 1.201 1.415 1.210 .650 .148 27 D-10 .987 .810 .629 1.262 1.415 1.242 .655 .147 08 31 B-07 .998 1.177 1.023 1.121 1.142 .990 .541 .161 36 G-02 .996 1.171 1.023 1.109 1.139 .986 .576 .166 09 33 D-05 .900 1.038 .919 1.141 1.261 1.120 .621 .172 34 E-04 .888 1.057 .919 1.146 1.275 1.109 .605 .160 10 38 L-02 1.547 1.216 .878 .978 1.041 .872 .468 .198

  1. 44 P-06 1.564 1.172 .863 .997 1.043 .895 .467 .201 ,

E

  • 40 .911 1.141 11 H-03 1.101 .973 1.228 1.075 .570 .140

.#' 42 0-05 .902 1.081 .963 1.160 1.262 1.065 .568 .157 7

" 12 01 11- 0 8 1.473 1.158 .944 1.008 1.038 .870 .508 .220 E

8 13 02 11 - 0 9 .978 1.130 1.056 1.131 1.162 .975 .569 .177 lb 14 03 G-09 .883 1.101 1.108 1.162 1.180 .1.051 .514 .152 15 04 F-08 .824 1.087 1.094 1.186 1.189 1.042 .577 .143 16 10 11- 0 5 .835 1.080 1.060 1.184 1.218 1.047 .576 .175 17 12 L-06 .823 1.056 1.047 1.159 1.243 1.087 .586 .149 18 14 N-08 .833 1.061 1.031 1.209 1.244 1.076 .547 .156 19 21 11- 1 3 .877 1.106 1.068 1.154 1.194 1.050 .551 .169 e

e

}

?

Comparison Of 'Incore Monitoring Assemblies Flux Shapes At 40% FP Using The Indicated Grouping Civen In Incore, Detector Test

-j i!

Incore Fuel Detector Current To Average Detector Current Per Assemblu Croup Detector Assembly Number' Number incarfon Level 1 Level 2 Level 3 tevel 4 Level 5 Level 6 Ievel 7 (nc) 20 .26 E-Il .879 1.026 .969 1.197 1.278 1.111 .541 .154 s

, 21 30 B-08 .908 1.158 1.090 1.145 1.162 1.066 .521 .186 i 22 37 H-01 .895 1.171 1.093 1.170 1.140 .991 .541 .179 i 23 41 N-04 .808 1.050 1.003 1.152 1.264 1.111 .612 .155 f

.942 1.172 1.089 1.136 1.163 .998 .500

!! 24 45 R-07 .197

?

.N' .

l' $

0 25 46 R-10 1.104 1.226 1.043 1.094 1.087 .957 .489 .167 -

1 f' 26 48 0-12 .849 1.063 1.016 1.167 1.243 1.051 .610 .167 Y

" 27 49 H-14 1.065 1.177 .991 1.136 1.170 1.007 .455 .087 E

O 28 51 D-14 .860 1.092 1.074 1.174 1.201 1.063 .538 .164 lb 29 52 C-13 .804 1.077 1.059 3 172 1.238 1.070 .581 .182 Average = 0.163-Standard Error = 0.020 4

7 l , .

s-Y Comparison Of Incore Monitoring Assemblies Flux Shapes At 75% FP Using The Indicated Grouping Civen In Incore Detector Test Incore Fuel Detector Current To Average Detector Current Per Assembly Group Detector Assembly Numbe r Number incation Level 1 Level 2 Level 3 Imvel 4 Level 5 Level 6 Level 7 (BC) 01 05 E-09 0.664 0.899 1.139 1.325 1.270 1.131 0.572 0.159 07 E-07 0.687 0.926 1.153 1.312 1.287 1.066 0.568 0.164 09 G-05 0.682 0.922 1.164 1.333 1.283 1.081 0.534 0.154 11 K-05 0.659 0.951 1.140 1.330 1.326 1.051 0.543 0.166 13 M-07 0.654 0.911 1.174 1.333 1.272 1.074 0.582 0.192 16 H-09 0.674 0.932 1.142 1.325 1.317 1.054 0.556 0.152

' 19 K-11 0.679 0.913 1.101 1.290 1.328 1.113 0.576 0.149 25 C-11 0.686 0.936 1.133 1.359 1.307 1.078 0.501 0.160 02 23 F-13 0.824 0.878 1.023 1.320 1.290 '1.104 0.561 0.222 35 P-03 0.811 0.883 1.046 1.327 1.311 1.075 0.547 0.185 39 L-03 0.846 0.933 '1.045 1.355 1.329 1.020 0.472 0.188 g 50 L-13 0.786 0.893 1.037 1.324 1.291 1.076 0.592 0.181 o' 28 C-10 0.833 0.920 1.035 1.348 1.307 1.031 0.52.c 0.198 o' 32 C-06 0.787 0.883 1.066 1.325 1.316 1.090 0.532 0.026

.'* 43 0-06 0.817 0.904 1.052 1.361 1.296 1.030 0.541- 0.191 47 0-10 0.855 0.909 1.034 1.330 1.259 1.071 0.542 0.201 03 06 F-07 0.735 0.982 1.177 1.293 1.273 1.048 0.491 0.183 .

08 G-06 0.736 0.979 1.155 1.316 1.256 1.036 0.521 0.193 04 15 N-09 0.709 0.877 1.113 1.363 1.326 1.061 0.550 0.178 20 K-12 0.699 0.870 1.086 1.360 1.304 1.055 0.596 0.173 05 17 M-10 0.746 0.894 1.129 1.437 1.408- 1.106 0.280 0.224 18 L-11 0.685 0.879 1.076 1.338 1.328 1.106 0.589 0.188

06 22 C-13 0.761 0.943 1.112 1.307 1.328 1.000 0.549 0.151 29 C-09 0.750 0.967 1.099 1.278 1.278 1.068 0.560 0.143 i,

a

Comparison Of Incore Monitoring Assemblies Flux Shapes it 75% FP

'Using The Indicated Grouping Civen In Incore Detector Test l l

Incore . Fuel Croup Detector Assembly Detector Current To Average Detector Current Per i sembly Number Number Location Level 1 Level 2 Level 3 Ietel 4 Level 5 Level 6 tevel 1 (BC) 1.

07 24 F-12 0.764 0.540 0.736 1.533 1.527 1.261 0.639 0.169 27 D-10 0.770 0.576 0.802 1.493 1.473 1.257 0.630 0.170 08 31 B-07 0.856 1.071 1.099 1.256 1.207 0.995 0.517 0.176 36 C-02 0.841 1.073 1.117 1.224 1.171 0.981 0.594 '0.190

' 09 33 D-05 0.705 0.820 1.058 1.378 1.344 1.100 0.596 0.189 34 E-04 0.695 0.852 1.070 1.362 1.330 1.114 0.577 0.166 10 38 L-02 1.346 1.150 0.971 1.122 1.105 0.864 0.442 0.219

{

44 P-06 1.354 1.135 0.970 1.130 1.108 0.893 0.410 0.236 ,

  • 40 11 H-03 0.720 0.931 1.114 1.356 1.286 1.041 0.551 0.149 .

. 42 0-05 0.710 0.923 1.095 1.373 1.339 1.032 0.527 0.168

[

12 01 H-08 1.338 1.184 0.975 1.106 1.090 0.859 0.448 0.265 E

{ 13 02 H-09 0.890 1.068 1.096 1.226 1.263 0.949 0.507 0.207 f 14 03 C-09 0.800 1.031 1.207 1.262 1.221 1.085 0.894 0.161 15 04 F-08 0.703 0.994 1.179 1.326 1.241 1.016 0.540 0.158 16 10 H-05 0.707 0.975 1.152 1.351 1.272 1.023 0.521 0.179 17 12 L-06 0.671 0.926 1.149 1.793 1.285 1.121 0.556 0.156 18 14 N-08 0.680 0.941 0.948 1.416 1.331 1.099 0.585 0.160 19 21 H-13 0.738 0.998 1.139 1.302 1.249 1.052 0.523 0.186 i

i C ( _)

Comparison Of Incore Monitoring Assemblies Flux Shapes At 75% FP Using The Indicated Grouping Given In Incore Detector Test Incore Fuel Detector Current To Average Detector Current Per Assemblir Croup Detector Assembly Number Number incarfon level 1 Invel 2 level 3 Ievel 4 Level 5 Level 6 gl 7 (BC) 20 26 E-Il 0.668 0.837 1.084 1.379 1.346 1.025 0.561 0.156 21 30 B-08 0.796 1.059 1.175 1.270 1.206 1.017 0.477 0.182 22 37 11 - 0 1 0.789 1.091 1.142 1.313 1.168 0.967 0.529. 0.182 23 41 N-04 0.611 0.881 1.130 1.338 1.328 1.105 0.607 0.169 24 45 R-07 0.834 1.099 1.151 1.265 1.224 1.013 0.414 0.205

.s

% 25 46 R-10 0.938 1.165 1.142 1.265 1.054 0.922 0.509 0.189 o

  • 26 48 0-12 0.695 0.942 1.112 1.347 1.283 1.030 0.586 0.277 Y

N 27 49 M-14 0.915 1.075 1.099 1.320 1.230 0.998 0.359 0.079 8

g 28 51 D-14 0.740 0.972 1.160 1.317 1.257 1.052 0.503 0.193 n

[ 29 52 C-13 0.654 0.945 1.115 1.290 1.312 1.093 0.590 0.226 Average = 0.179 Standard Error = 0.038 l

l 1l 1

ll Deviation Between Computer And lland Calculated Segment Powers At 40% FP _l f

f Incore Detector Detector Detector SPND-UC BKGC BECC SPND DDt.FC SPND-MC $p;DhbC Purcamtage i Assembif 1.evel Sense Charge Computer lland lland lland Itand computer 11eviation (None) (Eagment) (None) (Coulombs) (nA) (nA) (nA) (None) (nA) (nA) (2) .

H-08 1 0.990 121.152 320.80 4.94 325.74 1.0110 329.32 333.66 +1.32 2 1.002 122.820 242.70 15.89 258.59 , 0.9987 258.25 262.48 +1.64 3 3 0.990 123.870 183.50 24.16 207.66 1.0106 209.86 213.88 +1.92 i

4 0.999 122.309 194.50 31.52 226.02 1.0015 226.36 225.45 +0.92 5 0.986 123.328 192.00 39.05 231.05 1.0147 234.45 235.27 +0.35 6 0.994 122.181 150.00 45.73 195.73 1.0065 197.00 197.23 +0.12 7 0.996 122.053 64.20 49.90 114.10 1.0042 114.58 115.17 +0.51 E-07 1 1.005 123.328 310.10 2.91 313.01 0.9959 311.73 313.57 +0.59 2 1.001 122.892 390.70 11.30 402.00 1.0001 402.04 403.6' + 0. 4 0 3 1.006 121.539 370.20 20.39 390.59 0.9951 388.68 390.1 'O.43 y 4 0.991 120.958 400.20 421.20 29.65 39.56 429.85 460.76 1.0102 1.0022 434.23 461.77 434.97 464.08

').17

.0.50

)

c' 5 0 ,999 122.309 351.00 400.63: 402.14 +0.38 7 5 7

0.999 1.000 122.047 122.565 163.00 48.83 55.00 399.83-218.00 1.0020 1.0004 218.09 218.51 60.19 y F-03 1 0.984 120.245 226.30 2.50 228.80 1.0169 232.67 234.61 +0.5J 2 0.985 120.373 241.20 9.04 250.24 1.0159 254.22 256.89 +1.05 f 3 0.991 121.088 194.50 15.13 209.63 1.0096 211.64 213.7a +0.98 4 0.9tl6 121.281 237.10 21.15 258.25 1.0149 262.10 262.67 +0.22 5 0.987 120.245 257.20 28.06 285.26 1.0139 289.23 290.36 +0.39 l

6 1.001 120.023 215.80 34.62 250.42 0.9996 250.32 25%.78 +1.78 7 0.986 120.505 100.40 39.00 139.40 1.0145 141.42 141.75 +0.23 C-13 1 1.008 123.709 99.80 1.19 100.99 0.9923 100.21 101.35 +1.13 l 2 1.007 121.668 130.70 4.67 135.37 0.9934 134.48 135.74 + 0. 94 l 3 1.010 122.309 125.50 8.54 214.04 0.9904 132.75 133.54 +0.60 4 1.006 122.437 135.20 12.48 147.68 0.9944 146.85 147.70 +0.58 5 1.002 122.820 139.40 16.65 156.05 0.9984 155.80 156.03 +0.15 6 1.001 122.692 114.00 20.50 134.50 0.9993 134.41 134.85 +4 33 i 7 0.999 122.437 50.00 23.00 73.00 1.0011 73.08 73.24 +0.22  !

l Corrected $FMD's $lgaal Calculation (Detector (1) On Assembly (j))

semo-eC = +

( sema-pC, acs,,j ) x ( DDarC,,3 )

Weres Sonettivity Asad Deplettee Correction le Given By:

DbtrC, = ( Charge )/( Charge - SPND-UC E .8642-04) I ( Sense, )

bteres Background Correctica la Cives Sys BCE, , =

( ScK. ) I (( f.J . l/(CI Als)) ' #-

1

/

Daviation Between Computer And Hand Calculated Segment Powers At 75% FP Incore Detector Detector Detector SPND-UC BKGC BKGC SPND DDEPC SPND-DC SPND-DC Percentage Assembly Level Sense Charge Computer it.and liand 11and lland Compster beviation (Hone) (Segment) (None), (Coulombs) (nA) (nA) (nA) (None) (r.A) .(nA) (%)

11-08 1 0.990 121.152 462.50 8.77 421.27 1.0134 477.59 485.19 +1.59 2 1.002 122.820 393.00 29.27 422.27 1.0007 422.57 429.23 +1.58 3 0.990 123.870 298.80 45.81 344.61 1.0122 348.81 353.35 41.30 4 0.999 122.309 335.80 61.01 396.81 1.0034 398.16 401.07 40.73 5 0.986 123.320 310.50 76.54 387.04 1.0164 393.39 395.28 +0.48 6 0.994 122.181 218.90 89.29 308.19 1.0076 310.53 311.46 +6.30 7 0.996 122.053 65.50 96.10 161.60 1.0045 162.33 162.57 #0.15 E-07 1 1.005 123.328 421.60 4.46 426.06 0.9979 425.17 428.76 10.84 2 1.001 122.892 555.80 17.64 573.44 1.0029 575.10 578.59 10.6a 3 1.006 121.539 684.50 34.38 718.88 0.9988 718.02 720.19 +0.30 4 0.991 120.958 752.30 53.85 806.15 1.0145 817.84 819.14 +0.16 H 5 0.999 122.309 724.30 73.94 804.08 +0.12 798.24 1.0061 803.11 h 6 0.999 122.047 570.50 91.48 661.98 1.0050 665.29 665.82 10.08 y 7 1.000 122.565 251.50 102.60 354.10 1.0018 354.74 354.94 10.06 c- F-03 1 0.984 120.245 294.70 3.65 298.35 1.0184 303.84 307.03 +1.05 .

( 2 0.985 120.373 326.60 13.44 340.04 349.04 +0.86 1.0176 346.02 s 3 0.991 121.088 374.00 24.36 398.36 405.06 40.50

1.0118 403.06 4 0.986 121.281 469.50 37.68 507.18 1.0176 517.78 10.23 516.11 5 0.987 120.245 443.50 52.04 495.54 1.0164 504.37 +0.14 503.67

. 6 1.001 120.023 350.00 64.44 414.44 1.0015 420.4/ +1.30 415.06 7 0.936 120.505 155.30 72.30 227.60 1.0153 231.08 231.22 10.06 C-13' 1 1.008 121.709 137.80 139.87 139.87 0.9930 138.89 140.32 +1.03 2 1.007 121.668 194.10 202.50 202.50 0.9944 201.37 202.70 +0.66 3 1.010 122.309 223.80 240.16 240.16 0.9917 238.17 239.11 +0.39 4 1.006 122.437 251.70 277.13 277.13 0.9958 275.97 276.57 10.22 5 1.002 122.820 246.10 281.06 281.06 0.9997 280.98 281.35 40.13 6 1.001 122.692 190.90 234.24 234.24 1.0003 134.31 234.47 40,07 7 0.999 122.437 77.80 126.30 126.30 1.0015 126.49 126.55 40.05 Corrected SPND's Signal Calculation (betector (1) on Ameeably (j))

smo-aC - + acs,

( srun-uC, ) x ( cor.rC,,3 )

Where Senettivity And Depletion Correction is Cavan By:

DOErC = ( Charge )/( Charge g - SFND-UC E .8ME44) I ( Sense )

Where Background Correction la Given bys acs, , = ( ncu, ) x (( q' d,da )/(( 7 ga.)) .

d i

t Enrichment, Rodded /Unrodded, And LBP Correction Factors .

F Factor Fuel Enrichment LBP Enrichment Factor  :

FA All All '1.000 F1 All All 4.847 X 10-3 F2 1.93 All 0.768 2.54 0.943 2.83 1.021 1

F7N All All 1.000 1.93 0.00 0.966 2.54 0.00 0.961 g 2.83 0.00 0.965 e

EI F7R All All 1.000 c-1.93 0.00 1.026 u 2.54 0.00 1.009 di 2.83 0.00 1.007 F8 All 0.00 1.000

. 2.54 1.01 1.030 1.18 1.031 1.34 1.031 l

2.83 1.18 1.021 Note: The definition of the above factors are given below:

FA = Fit Factor, assumed to be 1.000 at BOL F1 = Megawatt Thermal to NANOAMP Conversion Factor F2 = Fuel Enrichment Correction Factor F7 = Non Rodded / Rodded Correction Factor F8 = Lump' Burnable Poison Correction Factor

( ,

_.)

h Deviation Between Computer And Hand Calculated Sensitivity, Depletion, And Background Corrected SPND Signals At 40% FP In.: ore Detector SPND-8C Enrichment. Rodded /UnroJJed, And 1.EP correction Factors p 3pgi) g 33 39,g ,gg, ,

4ssembly 1.evel computer (Ser, ment)

FA F1 F2 F7 F3 Iland Computer Deviatton (None) (nA)

(Nunc) (itWt/nA) (None) (None) (None) (Hilt) OlVL) (I) 1 1 333.66 1.000 4.847 X 103 0.943 0.961 N 1.000 1.466 1.45a -1.02 2

  • 262.48 1 009 R 1.181 1.196 +1.27 3 213.88 0.986 0.975 -1.12 4 228.45 1.054 1.041 -1.23 ' --

C'nup 7 rod 235.25 5 1.085 1.072 -1.20 2.5% % U235 6 197.23 0.910 0.899' -1.21 0.00 % B4C 7 115.17 0.531 0.524 -1.32 7 1 313.57 1.000 4.847 X 163 0.768 0.966 1.000 1.128 1.107 -1.86 2 403.65 1.451 1.424 -1.86 3 390.35 1.404 1.377 -1.92 4 434.97 1.533 N croup 1 rud 1.564 _1,98 tr 1.93 % U23., 5 464.08 1.669 1.634 -2.10 0 0.00 % H c 6 402.14 1.446 1.416 2.07 4 7 218.58 0.786 0.769 -2.16 L 35 1 234.61 1.000 4.847 X 163 0.943 1.000 1.031 1.106 1.081 -2.26

& 2 256.89 . 1.211 1.183 -2.31 3 213.71 1.007 0.984 -2.28 1,nr Typi 2 4 262.67 1.237 1.209 -2.26 2.54 % U235 5 290.36 1.368 1.335 -2.41 1.18 % a C 6 254.78 - 1.201 1.a71 -2.50 4 7 141.75 0.668 0.651 -2.54 52 1 101.35 1.000 4.847 X 163 1.021 0.965 1.000 0.484 0.477 -1.45 2 135.74 0.6'.8 0.640 -1.23 3 133.54 0.638 0.629 -1.41 4 147.70 0.705 0.a96 -1.28 2.H1 % U235 5 156.03 0.745 0.735 * -1.34 6 134.85 0.644 0.635 -1.40 n.00 % B 4 c 7 73.24 0.350 0.344 -1.71 Note (*); This detector when group 7 is at 19% vd is 47.3 percent rodded thus the appropriate F7 value is 0.984 Notes Deviation (Z) ? Computer - Hand' it 100 lland _

'l

! Deviation Between Computer And lland Calculated Sensitivity, Depletion, And

j. Background Corrected SPND Signals At 75% FP l

i lueore Detector SI'!du-SC Enrictuac nt. Rodded /UnroJJed, And Ltr Correction V. actors Satu-M SPND-M Percentage j .W embly Level Computer II.uW Guput u FA F1 F2 Y7 F8 N vi.ition I.None) (Segment) (nA)

(kne) (MWc/nA) (Nonc) (kne' (tlone) (W8I IWC) II) i 1 1 485.19 1.000 4.847 x 103 0.943 0.961 14 1.000 2.131 2.104 -1.2/

i 2

  • 429.23 1.009 R 1.931- 1.949 40.93 I

3 . 353.35 1.630 1.604 -1.60 4 401.07 1.850 1.819 -1.68 Croup 7 rod 5 395.28 1.823 1.792 -1.70 2.5'. I U235 6 311.46 1.436 1.411' -1.74 O.00 % B4C y 162.57 0.749 0.736 -1.73

, 7 1 428.76 1.000 4.847 1 163 0.768 0.966 1.000 1.542 1.509 -2.14 2 578.59 2.081 2.036 -2.16 3 720.19 2.590 2.532 -2.34 Y Croup 1 rod 4 819.14 2.946 2.878 -2.31 cr 1.93 I U235 5 804.03 2.891 2.822 -2.34 6 665.82 2.394 E 0.00 I n C 4 7 354.94 1.276 2.335 -2.46 1.243 -2.59

. LI 35 1 307.03 1.000 4.847 E 103 0.s43 1.000 1.031 1.447 1.406 -2.70

. 2 349.00 1.645 1.599 -2.80 3 405.06 1.909 1.855 -2.8) 8.Er Type 2 4 517.28 2.438 2.367 -2.91

+ 2.54 1 U235 5 504.37 2.377 2.306 -2.99 i

1.18 1 5 C 6 420.40 1.981 1.921 . -3.03 4 7 231.22 1.090 1.057 -3.03 i' 52 1 140.32 1.000 4.847 x 107 1.021 0.965 1.000 0.670 0.660 -1.49 2 202.71 0.969 0.953 -1.65

j 3 239.11 1.142 1.123 -1.66 4 276.57 1.321
  • 1.299 -1.67

' 1.344 -1.71 i

    • 3 g gy33 5 281.35 1.321 6 234.47 1.120 1.100 -1.79 0.DO Z 8 4C 7 0.604 0.594 -1. 66' 126.55 Note (*): This detector when group 7 is at 19% vd is 47.3 percent rodded thus the appropriate F7 value is 0.984 trote Deviation (1) = Computer - Iland x 100 Hand ,

\ /

Comparison Of Computer And Hand Calculated Worst Case Maximum LilR's Power Incore Maximum Maximum P (RxA), dim Maximum LilR, kW/ft

  • Date Time Level Offset Tilt

(% FP) (%) (%) Computer

  • Hand Computer lland 03-01-77 0206 40.21 - 2.09 + 0.94 1.57 1.62 4.44 4.63 03-15-77 2230 73.94 + 1.11 + 0.74 1.70 1.79 8.93 9.39 d'

E o

Y o>

l Note (*): The unit computer values have been reduced by their respective radial local peaking factor which was 1.05 for these two cases.

Incore Detector Core Location w

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Comparison Between Computer And Rand Calculated Radial l Core Power Distribution At 40 And 75% Full Power i

i i

A I

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-0.38 +0.13 +0.19 +0.25 + 0.29 +0.25 +0.1: +0.19

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+0.18 +0.25 +0.27 +0.2 5+0.12 +0.12 +0.5 3 g, +0.34 +0.20 +0.22 -0.41 -0.35 +0.70

+0.28 +0.20g +0.12 +0.13 -0.33 -0.11 g +0.39 +0.1E + 0.06 -0.3.

+0.25 +0.14 + 0. 22 +0.35 5 +0.4: -0.24 +0.1c

+0.25 +0.28 4 . 21 0 +0.39

+0.09 F

R I 1 2 3 ,4- 5 6 7 8 9 10 11 12 13 14 15 I.II 40% Tull Fover I.II 75% Full Power Note: Deviation (2) = 'Ceeeuter-Han[ z 100 M .

Figure 4.7-2

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Average Incore Detector Signal In Nanoamps Versus Core Power Level In Percent Full Power 3

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Figure 4.7-3 4

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l 4.8 POWER IMBALANCE DETECTOR CORRELATION TEST Labalance of the neutron flux in a reactor results from temperature distribu-tions, fuel depletion, xenon oscillations, or control rods positioned in the core. The amount of imbalance

is determined using input signals from the out-of-core detectors, it is essen-tial that they are calibrated to read the true imbalance as determined from the incore detectors.

This report presents the results of incore imbalance to out-of-core imbalance induced on Crystal River Unit 3, Cycle 1 at 40 and 75 percent full power during the performance of Power Imbalance Detector Correlation Test.

4.8.1 PURPOSE The Power Imbalance Detector Correlation Test had the following four objectives:

(a) To measure the relationship between core offset

  • as indicated by the out-of-core power range nuclear Lastrumentation detectors and as indicated by the full incore monitoring systam.

(b) To provide sufficient information to adjust the NI/RPS differential amplifiers should the measured ccrrelation indicate a need.

(c) To verify that an acceptable relationship between the backup recorder system and the full incere monitoring system indication of core offset was d) served.

(d) To verify acceptable core thermal-hydraulic parameters at each imbalance condition induced.

Three acceptance criteria are specified for the Power Imbalance Detector Correlation Test and are listed below:

(1) The measured correlation between each out-of-core detector offset to full incore monitoring system offset lies within the acceptable region shown in Figure 4.8-2. The correlation slope shall be greater than or equal to 1.00 and the correlation intercept should be less than 13.5% offset.

(2) The measured relationship between the full incore monitoring system offset and the backup recorder offset lies within the acceptable region shown in Figure 4.8-3.

I (3) The measured worst case minimum DNBR is greater than 1.30 and the measured worst case maximum LHR is less than 18.00 kW/ft.

4.8.2 TEST METHOD During the Power Escalation Sequence at Crystal River 3, as part of the test program, imbalance measurements were made to detennine the acceptability of the out-of-core detectors to detect imbalance and to establish a basis for verifying

  • Imbalance = Power in top of core - power in bottom of core
  • Offset = Imbalance Total Core Power 4.8-1 l

~ ,- ,

that DNBR and LHR limits would not be exceeded while operating within the flux / '

delta flux / flow envelope set in the reactor protection system. These imbalance measurements were made at different power levels with a gain factor of 3.90 applied to the multiplier on the delta flux amplifier output.

In performing the test, AP'SR's were positioned to obtain the desired full incore imbalance with reactivity compensations made by control rad groups 6 and/or 7.

At both test plateaus, offset as indicated by the full incore system, out-of-core system, and backup recorder system was recorded as given in Tables 4.8-1 and 4.8-2. From this data, plots of average out-of-core offset and backup recorder offset versus full incore offset were maintained as shown in Figures 4.8-4 through 4.8-7. Worst case minimum DNBR and maximum LHR data was monitored during the test. The rssults are plotted versus full incore offset in Figures 4.8-8 and 4.8-9.

Based upon previous startup experience, the relationship between incore offset and out-of-core offset was determined to be a linear equation of the form below:

OCO = M x ICO + B EQ. (4.8-1)

Where: OCO =

Out-of-Core Offset (Percent)

ICO =

Incore Offset (Percent)

M = Slope of Relationship B = Intercept at Zero ICO The experimental slope and intercept could then be obtained using a linear '

least squares fit from the data obtained. If the measured slope of the relation-ship of ICO to OCO was determined to be unsatisfactory; i.e. <1.00, the gain of the out-of-core power range detector should be adjusted. The relationship of measured slope to gata factor is as follows:

GF =

(M2/M1) x GFo EQ. (4.8-2)

Where: GF = Desired Gain Factor M2 = Desired Slope (i.e. >1.00)

M1 = Measured Slope GFo =

Present Gain Factor (i.e. 3.90)

Verification of the adequacy of the power imbalance system trip setpoint was performed in conjunction with worst case analysis on each minimum DNBR and m aimum LHR measured. The technique used was to extrapolate each measured point to the power / imbalance / flow enverlope boundary limits given in Figure 4.8-1. In this way the adequacy of the imbalance system trip setpoints to protect the unit from exceeding thermal-hydraulic limits could be verified.

4.8.3 EVALUATION OF TEST RESULTS The measurement of the offset correlation function between the full incore system and each out-of-core detector was determined during imbalance scans by APSR's and control rod group 6 and/or 7 at 40 and 75% full power to be a linear 3 relationship on all power range detectors. Figures 4.8-2 and 4.8-6 show the /

average response in offset between the out-of-core power range detectors and the full.incore system. Test data indicated that all measured offsets from the out-of-core power range detectors fell within the acceptable areas of the 4.8-2

curve during the performance of the test. For each power range detector, a linear least squares fit was applied to the measured data points to obtain a value for the slope and intercept of the observed relationship. The results of these calculations are tabulated in Table 4.8-3. In all cases, the measured slopes were greater than the M h n allowable value of 1.00 which verified the utilization of a 3.90 gain factor for the difference amplifiers.

The ability of the backup recorder to follow full incore offset was also verified as part of this test by collecting backup recorder data and performing the necessary calculations. The results of this analysis are plotted in Figures 4.8-5 and 4.8-7 for the 40 and 75% full power cases, respectively. In all cases, the calculated offsets from the backup recorders fell within the acceptable area which satisfied the test acceptance criteria. The measurement of the slope between the backup recorders and the full incore yielded approxi-mately 1.00.

During all phases of testing, worst case minimum DNBR and maximum LHR were -

recorded against incore offset and a plot maintained as shown in Figures 4.8-8

. and 4.8-9. The most limiting value observed on the worst case minimum DNBR and maximum LHR was 3.81 and 12.35 kW/f t, respectively which is well within the procedural acceptance criteria. As can be seen from the plots, both worst case minimum DNBR and maximum LHR are strong functions of incore offset. It can also be concluded that the greatest thermal margins are present when an incore offset between -5 to -10 percent is present.

The measured worst case minimum DNBR and maximum LHR were then extrapolated to the power / imbalance / flow envelope limits for verification of the imbalance system trip setpoints. A summary of this analysis for the 40 and 75% full power plateaus is presented in Tables 4.8-4 and 4.8-5. The extrapoalted worst case values were then compared to the DNBR and LHR limits of 1.30 and 19.70 kW/ft, respectively. As can be seen from these tables, all worst case values were within the acceptance criteria with adequate margins.

4.

8.4 CONCLUSION

S Upon completion of power imbalance detector correlation testing at 40 and 75 percent tull power, the following were concluded:

l (a) The measured offset correlation function between the full incore system 2

and'each out-of-core detector was determined to be a linear relationship.

(b) A gain factor of 3.90 on each out-of-core power range detector difference amplifier will yield an acceptable slope relationship to the full incore system.

(c) The power / imbalance / flow envelope as set in the reactor protection system will protect the reactor from exceeding minimum DNBR and maximum LHR thermal limits, when a gain factor of 3.90 is utilized.

(d) The backup recorder can provide an acceptable measurement of core imbalance.

4.8-3

Summaary Of Test Data Obtained During The Performance Of Power Imbalance Detector Correlation Test At 40% Full Power Power Date Rod Position, %wd Incore Offset, % Out-Of-Core Offset, % W rst Case Worst Case latvel Maximum LHR Minimum DNBR Time (%FP) 1-5 6 7 8 Full Backup NI-5 NI-6 NI-7 NI-8 (kW/ft)

(Dim) 03-05 ~/7 40.20 100 90 13 1 +14.7 +17.2 +21.5 +19.9 +22.4 +19.6 5.67 8.26 i 1311 -

03-05-77 .40.25 100 99 23 17 + 7.1 + 9.4 + 7.7 + 7.0 + 9.2 + 7.3 4.83 9.18 1708 1

03-05-77 41.06 100 100 25 19 - 2.0 0.0 - 3.6 - 4.0 - 1.9 - 3.1 4.50 9.97 >

1832 l

t 03-05-77 40.43 100 100 25 24 -13.0 -10.8 -14.8 -14.9 -12.9 -13.7 5.00 10.03  !

1902 -

.-t {

E 03-05-77 40.35 100 100 26 20 i

-22.9 -20.4 -24.9 -24.9 -23.1 -23.5 5.91 8.41 i

., o 2156 03-05-77 40.75 100 100 21 32- -37.7 -36.8 -39.5 -39.2 -37.2 -37.3 6.66 7.41 E 2238 03-05-77 40.83 100 93 16 45 -48.7 -46.1 -47.8 -48.0 -47,2 -47.1 6.64 7.37  ;

2338 4

I 4

_ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - , - _ - - v -

ll Summary Of Test Data Obtained During The Performance Of Power Imbalance Detector Correlation Test At 75% Full Power Power Rod Position, %wd Incore Offset, % W rst Case Worst Case Date Imvel Out-Of-Core Offset, %

Maximum LHR flinimum DNBR Time (%FP) 1-5 6' 7 8 Backup NI-6 NI-7 Full NI-5 NI-8 (kW/ft) (Dim) 03-28-77 77.03 100 81 5 0 + 0.2 + 0.5 + 3.9 + 4.9 + 6.4 + 4.2 11.22 3.81 0650

. 03-28-77 76.94 100 89 15 18 - 1.3 - 0.2 - 1.5 - 0.0 9.55

+ 0.8 + 0.1 4.24 0814 03-28-77 77.04 100 91 15 21 - 8.2 - 7.0 - 9.9 - 8.0 - 7.4 - 7.4 8.74 4.58 0950 03-28-77 76.69 100 91 15 21 -12.9 -11.4 -15.0 -13.0 -12.4 -12.4 9.40 1050 4.84 g

E-y 03-28-77 76.43 100 91 15 23 -19.9 -18.6 -22.7 -20.5 -20.1 -19.3 10.21 4.72

p. 1138 co 4 03-28-77 76.34 100 91 15 26. -26.8 -28.7 -29.6 -27.5 -27.0 -26.1 10.97 4.39~'

1226 03-28-77 76.55 100 90 13 28 -34.5 -32.9 -37.9 -35.7 -35.2 -33.9 11.79 1320 4.03 03-28-77 76.67 100 90 13 29 -38.9 -37.9 -42.2 12.35 1405

-39.9 -39.4 -38.1 3.82 4 .

O

.. i

'l 5

Summary Of Best Fitted Correlation Slopes And Intercepts Measured On Each Power Range Channel During Tha Performance Of Power Imbalance Detector Correlation Test Power Gain Test Correlation Slope, Dim Correlation Intercept, %

Range Factor Data Be st Standard Acceptance Be st Standard Calibration ,

Channel Se t tir.g Sets Fitted Error Criteria Fitted Error Tol er ance

  • A. Power Imbalance Test At 40 Percent Full Power ,

5 3.90 7 1.07 0.05 > 1.00 +1.12 1.40 +3.50 6 3.90 7 1.05 0.05 > 1.00 +0.41 1.50 +3.50 7 3.90 7 1.07 0.05 > 1.00 +2.62 1.18 +3.50 8 3.90 7 1.03 0.04 > 1.00 +0.99 1.03 19 50 Y

tr I g'

] ,,

B. Power Imbalance Test At 75 Percent Full Power

.h Y l

5 3.90 8 1.13 0.03 > 1.00 -0.72 0.61 +3.50 6 3.90 8 1.10 0.03 > 1.00 +2.17 0.77 +3.50 g

! 7 3.90 8 1.12 0.04 > 1.00 +3.14 0.92 +3.50

,4

] 8 3.90 8 1.05 0.02 > 1.00 +2.10 0.57 f).50 Note (*): A correlation intercept within +/- 3.5% offset is indicative of proper NI calibration during the test. A value outside this tolerance does not however invalidate the test i- . .

1 (-- ( '

J l l

Worst Case Minimum DNBR And Haximum LHR Analysis For Power Imbalance Detector Correlation Test At 40% Full Power Data j Power Incore Axial

  • Envelope Limits Worst Case Worst Case

-Analysis level Offset Peak P AP Haximum LHR Hinimum DNBR (None) (%FP) (%) (Dim) (1 FP) (1 FP) (kW/ft) (Dim)

Heasured 40.20 +14.71 1.48 (5) 5.67 8.26 Extrapolated 100.92 +14.85 14.23 2.71 Heasured 40.25 + 7.13 1.27 (4) -

4.83 9.18 Extrapolated 104.30 + 7.44 12.52 2.89 Hessured 41.06 - 1.99 1.19 (4) 4.50 9.97 Extrapolated 104.30 - 2,08 11.43 3.08 Heasured 40.43 -13.00 1.30 (4) 5.00 10.03 Extrapolated 104.30 -13.56 12.90 3.33 a

g. Heasured 40.35 -22.87 1.54 (3) 5.91 8.41 g Extrapolated 98.30 -22.48 14.40 3.17 h

Heasured 40.75 -37.67 1.74 (3) 6.66 7.41 1 Extrapolated 84.91 -31.99 13.88 3.44 Heasured 40.83 -48.70 1.79 (2) 6.64 7.37 Extrapolated -

72.90 -35.50 11.86 4.37 Note (*): The number in parenthesis represents the segment location of the axial peak

3

~

l Worst Case Minimum DNBR An.1 Maxim w. U!R Analysis For Power Imbalance Detector Correlation Test At 75% rull Power Data . Power Incore Axial

  • Envelope Limfts Worst Case Worst Case Analysis Ievel Offset . Peak ,

P AP Maximum UIR Minimum DNBR

, (None) (%FP) (%) (Dim) (% FP) (7. FP) (kWlft) (Dim)

Heasured 77.03 + 0.15 1.59 (4) 11.44 3.81 Extrapolated 104.30 + 0.17 15.49

~

Heasured 76.94 - 1.33 1.34 (4) 9.74 4.24 Extrapolated 104.30 - 1.39 13.20 2.73 Hessured 77.04 - 8.16 1.23 (4) 8.91 4.58

} Extrapolated 104.30 - 8.51 12.06 2.78 Heasured 76.69 -12.93 '1.32 (4) 9.59 4.84 j Extrapolated 104.30 -13.49 13.04 3.16

$ ~

Heasured 76.43 -19.90 1.44 (4) -

10.41 4.72 Extrapolated 101.54 -20.21 13.83 3.47 Hessured 76.34 -28.6! 1.55 (3) 11.19 4.39

,j .

E trapolsted 92.62 -26.54 13.58 3.68

'l '

Heasured 76.55 -34.54 1.66 (3) 12.03 4.03 Extrapolated 87.46 -30.21 13.74 3.73

,?..

I Heasured* 76.67 -38.88 1.75 (3) 12.60 3.82

] Extrapolated 84.00 -32.66 13.80 3.96

,j u

l!

Note (*); The number in parenthesis represents the segment 3 location of the axial peak 1 I f

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l.

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Acceptance Criteria Between Full Incore And Out-Of-Core Offset For Power Imbalance Detector Correlation Test N 1

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Acceptance Criteria Between Full Incore And Backup Recorder

- ~

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4.9 NSSS HEAT B,ALANCE TEST The unit computer determine the core thermal power by performing ooth a primary and secondary heat balance. To do these calculations, the computer employs a core thermal power analysis program (CTPA). In addition, the IBM 5100 has also been programmed to calculate the core thermal power using the temperature rise across the core called the core AT power. The core AT power can be con-tinuously displayed to the control operator. The primary heat balance and the core AT power both employ primary flow, which is initially normalized to the measured flow determined during the performance of Reactor Coolant Flow And Flow Coastdown Test (Section 3.1). The primary and secondary heat balance should give similar results for the reactor thermal power. If they do not, the problem is assumed to. lie in this initially measured primary flow. The primary and secondary heat balance are set equal to each other and a better calculation of the primary flow is obta.ined. However, the primary flow deter-mination by this heat balance method will be accurate only when there is sufficient enthalpy rise across the core, thus requiring high power to make the -

measurement. The test, therefore, checked out heat balance routines until after steady state 75% full power was reached. At that time, primary flow determinations were begun. Data from 75, 92, and 100% full power were used to determine primary flow. The actual adjustment to this value was made at the end of 100% full power testing.

4.9.1 PURPOSE The purposes of this test are briefly discussed below:

(a) To verify computer calculated primary and secondary side heat balance ccmputations.

(b) Provide baseline data for comparison with subsequent heat balance checks.

(c) To determine reactor coolant loop , flow from primary and secondary heat balance calculations.

Four acceptance criteria are specified for nuclear steam supply heat balances as listed below:

(1) The primary heat balance performed by hand calculation and by the unit computer must agree to within 2% full power of each other.

(2) The secondary heat balance performed by hand calculation and by the unit computer must agree to within 2% full power of ,each other.

(3) The weighted average heat balance performed by hand calculation and by the unit computer must agree to within 2% full power of each other.

(4) Reactor coolant flow with four pumps operating as determined from a primary and secondary heat balance at 100% full power is between 105.0 to 115.3 percent of design flow.

4.9.2 TEST MEIHOD Primary and secondary heat balances as required in the NSSS Heat Balance Test were performed at 15, 40, 75, and 100% full power. At each of the above test 4.9-1 4

~

.; --.. T

~ ~

L . . - - . .-

plateaus, steady state conditions were established as indicated below:

(a) Turbine Header Pressure 1 9.0 Psig 'D (b) Pressurizer Level i 5.0 Inches (c) RCS Pressure . 1 20.0 Psig (d) RCS T. (AVE) 1 2.0 0F

{

(a) Power Level i 2.0% FP (f) Feedwater Flow i 2.0%

After obtaining steady state conditions, data collection was started. This comprised of starting the unit computer to printout the unit parameters listed in Table 4.9-1 and at the same time the core thermal power analysis program (CIPA) was printed out. The unit parameters obtained were then input into the ,

, primary, secondary and weighted average heat balance equations 4.9-1, 4.9-2 i and 4.9-3 in Table 4.9'-2.

1

~

The reactor coolant flow measurement using the primary / secondary heat balance method was performed at various power levels above 65% full power. Once steady

, state conditions were established, the necessary data was collected every 10 l minutes for 30 minutes. This data was then averaged and substitued into Equations 4.9-4 in Table 4.9-2 to obtain the primary flow in each loop and the

= -

total primary flow.

j 4.9.3 EVALUATION OF TEST RESULTS The NSSS Heat Balance Test Procedure is the procedure used for determining the thermal output of the reactor core and the primary reactor coolant flow. There are'two methods used for core thermal output: primary side heat balance and

secondary side heat balance. This section covers the results of each and i presents the determination of reactor coolant flow.

l 4.9.3.1 PRIMARY AND SECONDARY HEAT BA ANCE CALCULATION The results of the primary, secondary and weightad average heat balance measure-mants performed from 15 to 100% full power are presented in Table 4.9-3. Good agreement was found between the hand and computer calculated heat balances.

Comparison of the hand calculated values with respective computer calculated i

values resulted in an average and maximum deviation of 0.97 and 1.83% full power respectively. These values are well within the procedural acceptance criteria l of 2.0% full power.

l Examination of the primary and secondary heat balance above 40% full power j indicate the primary heat balances were consistently lower than the secondary heat balances. This difference is due mostly to the indicated reactor coolant flow being slightly less than the actual flow.

4.9.3.2 REACTOR COOLANT FLOW DETERMINATION Crystal River Unit 3 was designed for a minimum primary coolant flow rate of l '105.0-percent. design flow. Even though a greater flow rate-than the minimum

-4.9-2 .

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will provide excess DNB protection, the primary flow rates should be less than the core lift limits. The core lift limit to the primary flow rate as specified by the Babcock and Wilcox Company is 115.3 percent of design flow.

The basis of the flow measurement is the application of a heat balance across the primary and secondary sides of the two steam generators. The test con-sisted of monitoring the thermal-hydraulics data for 30 minutes averaging the various quantities, and substituting them into Equation 4.9-4 in Table 4.9-2 to determine the primary flow.

Table 4.9-4 shows the values of the reactor coolant primary flow determined using this method for four pump operation. As can be seen, the average values of the rea: tor coolant primary flow measured was 109.3 percent of the design flow. ' Comparison of this value to the acceptance criteria in Section 4.9.1 indicates that an acceptable flow rate is present on Crystal River Unit 3.

The accuracy of the flow measurement depends on the precision of the various -

instruments used to measure the thermal-hydraulics data. An error analysis was perforned and indicates that the error in the primary flow measurement due to the instrument tolerances and uncertainty in the flow coefficients and the ambient heat loss measurements is i 2.0 perc'ent of the design flow.

The priniary flow measured on Crystal River Unit 3 was therefore bounded at 109.3 1 2.0 percent of the design value. A more detailed analysis of flow was also performed by the Babcock and Wilcox Company. As a result of this inves-tigation, the primary flow was set at 110.0 percent of the design value.

4.

9.4 CONCLUSION

S All primary and secondary heat balance calculations met their respective acceptance criteria.

The primary reactor coolant system flow rate on Crystal River Unit 3 was set at 110.0 percent of design flow, which is within the minimum and maximum allow-able value as specified by the acceptance criteria.

4.9-3

. . - - .- - .-. .nc- -s . ~

List Of Terms Used During The Performance Of NSSS Heat Balance Test At Power Qp = Power using primary heat balance (BTU /hr)

Qs = Power using secondary heat balance (BTU /hr)

Qw = Power using weighted primary / secondary heat balance (BTU /hr)

' Qpumps = Total pump power (BTU /hr)

A Wp = Prir,ary side Loop (A) flow (Ibm /hr)

B Wp = Frimary side Loop (B) flow (lbm/hr)

A W fdw = Feedwater Loop (A) flow (1bm/hr)

B W = Feedwater Loop (B) flow (Ibm /hr) fdw W = Letdown flow from reactor coolant system (lbm/hr)

. A )

H = Outlet enthalpy for Loop (A) (BTU /lbm) .)

out B

H = Outlet enthalpy for Loop (B) (BTU /lbm) out A

H = Inlet enthalpy for (A) cold leg (BTU /lbm) in I B H

in

= Inlet enthalpy for (B) cold leg (BTU /lbm)

A H = Outlet steam enthalpy for OTSG (A) (BTU /lbm) st l

l B L H = Outlet steam enthalpy for OTSG (B) (BTU /lbm) st l

A

_H = Inlet feedwater enthalpy for OTSG (A) (BTU /lbm) fdw L H = Inlet feedwater enthalpy for OTSG (B) (BTU /lbm) fdw '

H = s_

Enthalpy of the J.etdown flow from RCS reactor coolant (BTU /lbm)

Table 4.9-1 o

. - - . . . . - - . ~ . - - .

List Of Terms Used During The Performance Of NSSS Heat Salance Test At Power

-P = Nuclear instrumentation power level (7.FP)

NI P = Core AT power level (7.FP)

AT AT e = Coolant temperature rise across the core (Deg. F) d 4

t s

Table 4.9-1 (Cont'd)

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  • i 6

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y EQ (4.9-3) Weighted primary / secondary heat balance calculation Y

~

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Wp =WA+WB=Wfa,(Hft p - Hfdw)/(hout + Hfn)

  • W dw (Hft- Hfdw)/(Hfut-Hfn)

EQ (4. 9-5) Core AT power calculation PAT = 2.292 ( ATc - 6.,912 ) + 15 (Equation Used During Test Program)

)

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Summary Of Heat Balance Performed During Power Escalation Testing Test Core AT Computer And Hand Calculated Heat Balances, IFP Plateau Date Time Power Prinary Secondary Weighted

(%FP) (% FP) Computer Hand domputer fland Computer Hand 15 03-31-77 2123 16.47 16.01 16.75 14.58 14.10 16.01 16.71 40 02-27-77 2338 43.62 40.88 42.35 38.76 40.27 40.23 41.72 75 03-14-77 2100 75.65 73.98 73.84 76.71 74.88 75.98 74.47 100 04-01-77 1440 99.09 96.49 95.07 99.00 99.19 98.94 99.15 '

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Determination Of Reactor Coolant System Primary Floh l

Power Calculated Primary Flow, 1 x 106 lb/hr Percent of Flow Date Time Level 3esign Flow Imbalance I (% FP) Loop (A) Loop (B) Total De sign (%) (g)

A. Summary Of Flow Calculations Performed At And Above 75% Full Power l , t 03-27-77 1536 75 69.56 72.40 141.96 129.75 , 109.4 4.00 03-27-77 1723 75 69.58 72.41 141.98 129.77 109.4 3.99 03-29-77 1808 92 70.33 72.78 143.11 130.33 109.8 3.45 04-01-77 1113 100 71.20 72.37 142.56 130.65 109.2 1.64 ,

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4.10 UNIT LOAD STEADY STATE TEST 4.10.1 PURPOSE The purpose of the Unit Load Steady State Test was to measure reactor coolant system and steam generator steady state parameters as a function of power to compare wi::h design predict' ions, equipment, and system limits. Specific purposes are as fo11cvs:

(a) To measure RCS and MSG parameters with four (4) reactor coolant pumps operating from 0 to 100% full power.

(b) To measure primary and secondary system operating parameters for compari-son with future performance.

(c) To verify the ability of the integrated control system to meet the design-ed performance under steady state conditions. .

(d) To verify that the adjustable steam generator downcomer restrictors are not causing flooding of the feedwater nozzles during normal operation.

Two acceptance criteria are specified for the Unit Load Steady State Test as listed below:

(1) All recorded steady state parameters, as a function of power level, are within their respective minimum and maximum limits as given in Figures 4.10-1 through 4.10-9.

(2) No indications of downcomer flooding have been observed during power escalation from zero to 100% full power.

4.10.2 TEST METHOD This test measures the steady state behavior of the RCS and the MSG's r.t various power levels. The power levels used during the performance of Power Escalation Testing were 0, 15, 25, 40, 65, 75, 92 an' 100 percent full power of 2452 megawatts thermal. At each of the above po: .c levels, steady state conditions were established and data required in Table 4.10-1 was taken for 30 minutes.

From this data the average unit parameters were calculated and plotted on Figures 4.10-1 through 4.10-9 and comparisons between the measured data and the design curves were made in order to verify the acceptance criteria. Unit para-meter stability during steady state operation was measured in terms of the deviation from the average value of the parameter during the test. These cogarisons were done to verify the integrated control system ability to main-tain steady state conditions and to try to eliminate any oscillations.

Price to escalation in unit power, the steady state parameters were extrapolated to the next plateau in order to estimate the acceptability of each at the next plateau. If the extrapolation indicated that a limit might be exceeded or a substantial deviation from the predicted existed, the condition was evaluated prior to escalating unit power.

i i

I 4.10-1

.. . _ _ .- - ____ _ _ . _ _ . -.- . - - ~ ~

0 i

4 i

! At power levels of approximately 40, 75, and 100% full power, the integrated 2

system was also operated in the turbine following and reactor /0TSG following 3

1. modes of control for a period of 30 minutes and steady state data recorded. -

This data was then averaged sad the results evaluated. Adjustments were then made to the integrated control system as neceasary to improve unit steady state response.

1 .

In addition to the above data, sceam generator downcomer temperature was also

monitored during the escalation of core power from 0 to 100% full power, to verify that steam generator downcomer restrictors were not causing flooding.

i This was proven by confirming that no sudden decrease in this temperature in-l dication was observed.

! 4.10.3 EVALUATION.0F TEST RESULTS i

Table 4.10-1 shows the unit average parameters of the primary and secondary system measured over the test period for the various power levels. Comparison between the measured data and the expected design data was performed by plot- -

ting the averaged unit parameters against the expected design curves. As can j be seen in Figures 4.10-1 through 4.10-9, all measured unit average parameters j

fell within their respective minimum / maximum boundaries except for steam genera-i tor outlet pressure. The deficiency was found to be in the calibration i of the transmitter controlling steam header pressure (i.e. nominally 885 Psig) .

A calibration check revealed it to be controlling at an actual pressure of

, approximately 20 Psig below the design setpoint.

1 During this test, unit stability was measured by determining the deviation of .

unit parameters from their respective average value. Table 4.10-2 is a listing _)

of the atandard deviation of these variables over the various test power levels.

j From this analysis it was concluded that flows, temperatures, and pressures were stable within the following average '.imits while operating in the fully integrated mode of control.

RCS Temperatures 1.30F of an average value RCS Pressure 15.5 Psig of an average value Main Steam Pressure 3.5 Psig of an average value RCS Flow 0.6% of an average value Feedwater Flow 2.7% of an average value The above results confirm that the integra*ed control system adequately control-led the unit during the startup test program. However, it should be noted that some difficulty was experienced at 75% full power due to secondary side imposed os cillations. These problems were eliminated by increasing the deadband on the ,

turbine header pressure error from i 5 to i 7.5 Psig, and by slowing down'the response of the feedwater pump speed controller by changing the gain and AP error deadband.

i The results of the checkout of steam generator downcomer flooding versus power level from the O to 100% full power data indicates that no flooding is present.

4.

10.4 CONCLUSION

S The average of the measured unit parameters during the test period fell within their respective minimum and maximum limits, except for OTSG outlet steam pressure

,) '

which was indicating 20 Psig high and therefore controlling low. This was cor-

, rected by recalibrating the pressure transmitter controlling steam header pressure.

j 4.10-2

. . = . = = _ _

4

-- Analysis of unit parameter stability indicates that all variables are relatively stable. The maximum variation in unit parameters was experienced around 70%

full power.

O e

S 9

1 1

\L/

l l-.

l 4.10-3 l

,- Average Unit Parameters At Various Test Plateaus For Crystal River Unit 3 During Steady State Conditions.

. i i

Average Unit Parameter Unit Parameter Units 0%FP 15%FP 25%FP 40%FP 65%FP 75%FP 92 %FP IGO%FP f

RC Cold Img Al NR Temperature F 532.09 573.57 573.10 569.52 564.26 562.95 559.80 558.35 f-RC Cold Img A2 NR Temperature F 531.81 573.63 572.95 569.34 563.77 562.55 559.52 557.94 '

RC Cold Img B1 NR Temperature F 531.36 574.08 573.04 569.36 564.07 562.74 559.61 558.18 RC Cold Ieg B2 NR Temperature 531.19 573.87 572.74 569.47 563.84 562.55 559.41 f:

F 557.75 RC Hot Leg A NR Temperature F 532.90 581.37 584.17 587.27 595.53 597.53 600.55 602.63 l RC llot Leg B NR Temperature F 532.90 580.88 584.51 587.77 594.63 596.56 599.72 601.50 (

RC Average Temperature F 532.29 577.43 578.58 578.32 579.42 579.87 579.87. 580.06 RC Loop Temperature Hismatch F +0.67 -0.38 +0.14 -0.01 -0.06 -0.10 -0.15 -0.20 {

RC Loop A Coolant Flow HPPH 70.01 67.78 68.30 68.68 69.10 68.98 69.52 69.76 RC Loop B Coolant Flow HPPH 70.77 69.30 69.40 69.84 70.47 70.01 70.42 70.68 g RC Pressurizer Imvel Inches 111.80 127.53 134.88 134.64 134.15 134.94 134.65 134.81 Q. RC Ioop A NR Prc::sure Psig 2150.17 2158.32 2139.93 2143.27 2147.44 2147.87 2139.10 2150.25 y N1 Power Imvel 7. FF 0.14 14.15 25.50 40.49 67.07 74.15 90.72 97.20 FDW Imop A Flow 0.57 0.63 1.01 1.89

, HPPH 3.32 3.81 4.62 5.04 l g FDW loop B Flow HPPH 0.02 0.52 1.15 1.96 3.40 3.85 4.69 5.09 f FDW Loop A Temperature P 219.40 289.97 326.23 364.05 411.50 423.67 438.44 445.53

" FDW Ieop B Temperature 219.18 288.70 325.69 363.38 410.20 422.27 437.36 F 444.86 HS Cenerator A S/U Ievel Inches 29.58 26.36 37.71 54.98 94.28 103.70 131.11 143.76 HS Cencrator B S/0 Irvel Inches 29.61 25.82 37.13 54.42 104.29 114.52 143.09 l

156.38 HS Cenerator A OP level  % 4.87 6.75 9.86 14.91 31.14 35.64 47.92 54.05 [

6 HS Cencrator B OP Imvel  % 6.20 8.70 11.90 16.29 28.80 34.28 46.56 52.81 I HS Cenerator A Pressure Paig 831.00 868.20 878.95 883.89 895.22 888.00 897.20 888.75 - '

HS Cenerator B Pressure Psig 861.10 873.30 865.25 870.64 897.89 890.62 900.80 894.00 MS Cenerator A Temperature F 100.25 580.13 584.06 No Data 592.72 593.90 594.35 593.83 [

MS Cencrator B Temperature F 145.30 580.04 584.66 No Data 593.01 593.94 594.38 594.15 '

HS Cenerator A D/C Temperature F 526.40 532.19 531.80 532.62 535.14 534.85 536.36 535.58 i HS Cencrator B D/C Temperature F 530.30 530.98 530.58 531.50 534.31 534.07 535.77 535.06 i Turbine Header Pressure Psig 869.30 884.44 880.33 882.56 887.10 891.60 887.49 883.38  !'

h Note: The Above Results Are Based D- The Data Acquisition Center Output C y ,

)

s Haximum Deviation In Average Unit Parameters At Various Test Plateaus For Crystal River Unit 3 During Steady State Conditions.

, Standard Deviation of Averane Unit Paramerer Unit Parameter Units 0%FP 157.FP 257.FP 40%FP 651FP 751FP 92 IFP 100% FP RC Cold leg Al NR Temperature F 0.07 0.24 0.51 0.26 0.27 0.18 0.17 0.09 RC Cold Img A2 NR Temperature "F 0.08 0.41 0.48 0.25 0.20 0.14 0.09 0.18 RC Cold leg B1 NR Temperature F 0.06 0.41 0.43 0.25 0.34 0.16 0.11 0.07 RC Cold leg B2 NR, Temperature F 0.06 0.41 0.46 0.27 0.30 0.14 0.16 0.16 RC Hot Leg A NR Temperature F 0.08 0.40 0.44 0.28 0.13 0.18 0.11- 0.09

  • RC Hot Leg B NR Temperature F 0.06 0.38 0.42 0.27 0.19 0.16 0.11 0.09 RC Average Temperature F 0.07 0.60 0.45 0.26 0.24 0.16 0.13 0.14 l RC loop Temperature Hismatch F 0.07 0.37 0.47 0.26 0.28 0.16 0.13 0.13 i RC Imop A Coolant Flow HPPil 0.32 0.17 0.19 0.13 0.20 0.21 0.19 0.11  !

RC loop B Coolant Flow HPPil 0.39 0.08 0.26 0.23 0.28 0.18 0.15 0.22 a RC Pressurizer Imvel Inches 0.63 0.85 0.86 0.67 0.36 0.62 0.38 0.24 S RC NR Pressure Psig 1.93 11.77 15.42 8.71 5.90 2.95 1.45 1.91 E HI Power level  % FP 0.11 0.09 0.68 0.36 0.25 0.41 0.26 0.79 4-FDW Inop A Flow HPPli 0.01 0.00 0.03 0.02 0.07 0.03 0.02 0.03 s FDW Inop B Flow HPPil 0.00 0.04 0.03 0.04 0.09 0.02 0.01 0.03

?

FDW Ieop A Temperature F 0.13 0.34 1.23 0.22 0.52 0.09 0.07 0.09 FDW Imop 11 Temperature F 0.11 0.44 1.25 0.21 0.36 0.15 0.15 0.21 HS Cenerator A S/IIInvel Inches 0.17 0.30 0.88 1.04 1.91 1.58 0.46 0.18 HS Cenerator il S/U Icvel Inches 0.05 0.25 0.62 1.04 1.84 1.49 0.13 0.36 HS Cencrator A OP Invel 7. 0.02 0.10 0.40 0.46 0.39 0.17 0.23 0.22 HS Cencrator in OP Invel 7. 0.00 0.08 0.18 0.92 0.42 0.25 0.25 0.35 HS Cencrator A Pressure Psig 0.00 1.01 2.38 1.67 2.33 1.51 1.03 1.16

. HS Cenerator in Pressure Pstg 1.18 0.94 2.50 1.65 1.76 1.51 0.92 0.93 HS Cunerator A Temperature F 0.00 0.39 0.41 No Data 0.20 0.17 0.20 0.05 HS Cencrator il Temperature F 0.20 0.32 0.19 No Data 0.18 0.15 0.12 0.16 HS Generator A D/C Temperature F 0.00 0.15 0.23 ' O.57 0.19 0.09 0.12 0.14 HS Cenerator B D/C Temperature F 0.00 0.17 0.24 0.42 0.11 0.07 0.08 0.11 Turbina lleader Pressure Psig 1.49 1.15 2.52 1.79 2.60 3.41 2.39 1.37 i

Note: The Above Results Are Based On The Data Acquisition Center Output ,

.i

1 i

l l

l Reactor Coolant System Temperature Versus Power Level With Four. Reactor Coolant Pumps Operating

~ 1 5g. _._ Applicable From 15 To 100% FP

_ ._ Maximum

-'E

,r_,,--, ;T(HOT)

- Minimum

'-Y '_ ,

^~^

,;.C..

-_-_ pw----7_.=

g 1

~

-e w.

ea -

- _-- ',f--

o- --.

A _- p- -

, _.- s .- -

'^

8 i___f i.-

3 a2_ **- '

,i_. El---

. .% c, _ .c;---'J'-

Maximum g lt #m # e . t:

T(AVE)

Q ____. ,'__((-* _: _ _

Minimum

$ -MA j' _

g __ _

,__ t-t c.

um- -x

' N

~

un-- -

-- ~ _ s. %

. +-- . ;. -

h +-.- x . -,

w -

s x ._

p _

~' . YANw _

e, j -

_ _-m

$o $dT- i_ l'-' ~_ *QW Maximum u: i -w  ; N--A

-f--

T(COLD) u u-+

o .

Minimum U .-!1 g gg t:

v, ,

" ~

i: ~ Measured RCS Temperatures d! T(AVE) O

'j: T(HOT) u T(COLD) - A

.'{'

7..

T_ , ._

..____r_

=

.= =, , nc

_ .y Power Level, % Full Power Note: The minimum and maximum curves represent '

design limits only.

Figure 4.10-1

_, Steam Generator Outlet Pressure Versus Power Level wu

-Applicab'le,From O'To 100% FP MeaEured OTSG Outlet Pressures

~

g,31,y,

_ _ OTSG (A) O OTSG (B) G  :, -

mm- /-

g a w- _

w ,,-

m - j -Expected n - -

. . , ~

o

/

/m y=w  :

/t-E

""~ .__ #-  :

/ j- Minimus

y. -- -- u p-.- =

-un n/

ms u w ,_. -- -

,.,- y _.

o

-e w A

.'_-'77 r

_~

L 5 / d. T u _ f .' ' ~

{ Q 's~~ -.

$ m. .

~. s . - . . - - .._

g 7; Ey - t- -

o p .-

s E!

u @

"E820: __

Reading low because _ __

O the controllar was.

out of calibration.

S't 0 i- _

r. _-. _

-__._ _ ~__m . . . _ . _ . . _ .

~

u 26 ^G $D =b TW Power Level, % Full Power Note: At 0% full power OTSG (A) steam pressure was off scale at 831 Psig.

  • Note: The minimum and maximum curves represent design limits only.

Figure 4.10-2 1

i 1

1

Steam Generator Startup Level Versus Power Level

~ ~

=wr Applicable From 0 To 100'% FP

- . m. _ .

Measured OTSG Startup Ranges

OTSG (A) @ Maximum'

~

OTSG (B) G ,[~' _-

a =w ^

3o - ._

,5 _ _ __

Q g .1t?_. f 7,_

y _ . -'

3____ .

_-' ~B-- G, Expected oc -_ - - :y g.

u-W ",. - -

$ -" r , ., -'

u ,- -

M lMjk u

  1. M -,a Minimum C

, --.'4-..

W_ ,.~ , e '?

g  ?

o _.-'_.' ^

g #- -

e u --* &

,-~ . _.- :

m-- 4s '

_-,_.___m_-'

Ef =

_=2

.._ ,_ ,__ . _ =

- ,1 -

e:

~

2: Mi :  : 4:

Power Level, t Full Power Note: The minimum and maximum curves represent design limits only.

.)

Figure 4.10-3

Steam Generator Operating Range Versus Power Level

p,y. Applicable From 15 To 100% FP -

.~ .

~

_ Measured CTSG Operating Ranges e --

OTSG (A) O E OTSG (B) 3

$ m__a c -

0--

a e- > Maximum

y. s-
a. _.- . _ . _ _ -- d-- _--

y -

o _- _-" ~ _ . _ _ _ . _ -

<a -

y--.gg ie _

a---

o y

u f'

j .,

-+ Expected

,f ,- .

g - .

.y ;_

-n

_ c.r , -

h. f~

~

u ._ , ,- ___

S~~

m. a-Aai_  :

. m-_ .

$! i-_6'J- L7 --

c- rit--

0 ,-

a -

g--

q.._

- ,s mo -~~ _ .__. _ c; u _-

p _.

.N~ M~ _ - -

Minimum

- n .

==, 2 EET p a- - - .

h-- p---- .y---%-- &,._ . .. p .1 YL,- ^

,_ _ - _ _ _ _ _ _ _.___,____p__d E# . =: F.:: F: __ i ne :

' Power' Level, % Full Power Note: The minimum and maximum curves represent design limits only.

Figure 4.10,4

r

~

Steau Generator Steam Temperature Versus Power Level x

_ n_n- Applicable From_15 To 100% FP Maximun

~_m,. j' s

.+ r:1- _-E' Ng ~~.,- -__.'N.j.r_ -_- Expected

.a - ^ --_ 2.-

ca  ; /

. _f __

U f _

s a - .- - m _- _. - _
e. w gan .

e s

o. --- 'I '

a -

Minimum o

H le '~' ?? 5^

u s cn  :.

u o:

u. .

$my . 11!

c.

ei +-i

.u.

o g-sas ' i!

e  :

$T :f< Measured OTSG Steam Temperatures

~H.' ;0TSG (A) D

'OTSG (B)

O

=5E' 5

i i,

) _ -- . .%

w v,

=  : =  : 4:

Power Level, % Full Power Note: The =4n 4=_= and =av4=_= curves represent design ifmits only.

Figure 4.10-5

7.

Feedwater Temperature Versus Power Level

g. Applicable From 0 To 100% FP i---

Measured Feedwater -

Temperature @ __ _ _ _

m, L L .

_"_" ___ - Maximum

~-

,A d- -

- tExpected p- #SN _vW =5 V  ; ~~ ~ r--- Minimum

_,nn ____.

m. . _ _ _ . _ . . _ _ - - _ . .._.._ # _ _ _. _ _ . -

- ~

gn e

._ p ,,- 7 ,--

r--, __. 7. % - - =

o _ ,,

p -- -

,. __-_s_---

.g ,m  :----

u . x. -'-

3 f'-- e' y -eng f .

y Ls? ' /"

f

, -p ..--

o s ,i.-_,.-

Qcpf==- g- #_-__

24AA __ _ _ . _ _ - p.__C__+____4_J .

,._--. .- -- g _ . . , - . - _ _ - - - .._.

m-- -. - -

's -- w --

o__ - . __

o o -

44 A, N,,V t

==

~ "

__. m ._

_ g

_i __ . _ . _ _ . _ _ . __ . . _ _ _ _ . . , . _ . . _ _ _ _ _ _ _ _ .

__-_ c --

g,,=._____ _ _ _ _ _ _ -___

__._.._j_.,___.____ 4 . . _

6 _

= == 6e u sno==

Power Level, % Full Power Note: The minimum and maximum (arves represent design limits only.

Figure 4.10-6 u

. .. - . . . . . - . .. - _ - -. - . . - - - - - - - - ~ . ~

Total Feedwater Flow Versus Power Level 3 t ,

_4_- -_: +._. -

m e

_m f . _ __-

Applicable From 15 To 1 0% FP-.

-_i_ =t=- _ _ . - ._ . _ - . _ . - . . - - . -

mm _~__ _ __ .

~~---

-~^ -"

=_

= =*

Measured Feedwater C o n.-

Flow Q --

, , ' Maximum O

?

M --

- r_ ' ~ r - - -

u Q f_H Expected

[_ --

-i2/tb.--.

w --

/.5fM/ Minimum h- ._- __

_.__ _ . . . . . EMC I_T Q_

.--.-1-1_,--

.L -.a_-- r- - 2/= -- --.-

~ /_..

c +. .- - - - - _

o- .g y .:. =-c

_-- _- pW-.. = ,C_. s_.-- -

kh r -

/I! [- M

,N,_. ,_

o _m _. s - _

,/= . .

Nr- * ((

ui _ g_T/-T' ._

~--*'

= . lf=f' 3'

a- . . -

.r-E= ,/=,.^

e - + . , . -+

e --

=.- /

u. -

i_- EA: 75 -

' er .

^

u =u= . / :"

,W

w. _- .. - - . . _

l sf- '

- J E_

?-

-- c ,- ,a_. ,

_y- yy- . _ . - _

w _..._I-_.-_-__.__,._,_____.

i -m e. c trwy_

uc. ____,-.=_=_

=_ _ _.

Pos.er Level, % Full Po$er'~~

Note: The minimum and maximum curves represent design limits only. )

Figure 4.10-7

{ . . . ..

1 Turbine Header Pressure Versus Power Level

_syv--

Applicable From 15 To 100% FP t--_-_-

Measured Turbine Header A ' ~~

Pressure

-n.n-

=

w m

C4

  • g _ _ , -

O _

II __-

--- ~Z Maximum w

b .. .

h. -

A -

cf

=-

~

Expected

.a. . __

@ .m mA _t _ _ _ . .

.o r___

w y

__1-

- - _ _ _ = __ Minimum

n. --

ff^-

t l __

.-s._. , - - . _ a-

,e . _ _ ._

r.

7-n f.h -

.n

=, _

m_ _ n =__2__

Power Level, 7. Full Power Note: The minimum and maximum curves represent design limits only.

Figure 4.10-8

Reactor Coolant System Loop Temperature Mismatch Versus Power 14 vel With Four Reactor Coolant Pumps Operating

'l i

_m Applicabl'e From 0 To 100% FP

_.6..

Measured RCS Loop Temperature

-Mismatch A w -. n y_ ---

._ C Maximum 2u _~TC -

h

c a

{ ZLI -

z a

k-_

e f AW ee i -

.1

__._;= .- m

._i.____.

m .A. . . ,,, _ n . Expected }

o -~- j 3

E o

u__.,.

-._n us a

e e

m o:

S T_ A w

O t Minimum e -

M _+ c. -

GK

--n

.y _- .--___ __.__._ . _ - -

_ = . = = 60--

Power Level, 7. Full Power Note: The minimum and maximum curves represent .

design limits only.

Figure 4.10-9

4.11 ' UNIT LOAD TRANSIENT TEST The integrated control system at Crystal River Unit 3 is designed to produce the best megawatt response to megawatt demand consistent with the capabilities and limitations of the entire unit. The control philosophy is developed recognizing

- that each of the control variables has independent requirements that must be con-sidered separately and that'all variables must be coordinated into one system to produce automatic control. Figure 4.11-1 displays the conceptual organization of the integrated control system designed to operate the unit.with minimum opera-

, tor involvement. This complex control system can be thought of as having five

_ levels of unit or system control, ranging from local closed loop control of components to complete system control whose purpose is to generate the megawatts being requested.

The ICS incorporates the following different normal modes of operation, (Figure 4.11-2):

(a) Reactor /0TSG-following control mode defined as the turbine controlling -

electrical megawatts with the reactor / steam generator maintaining the 4

required steam conditions and following an index of the unit load to main-tain the required steam conditions.

t l . (b) Turbine-following control mode defined as the reactor / steam generator

! controlling thermal megawatt generation and the turbine following this 1- steam production to maintain the required steam pressure.

(c) Fully integrated mode or the integrated reactor / steam generator / turbine generator control mode for once-through steam generators designed to i

combine both of these basic control schemes in such a way that the advan-tages of both are accented and their disadvantages are minimized.

The ICS takes advantage of the rapid, accurate steam pressure control available with the turbine-following approach. At the same time, the ICS achieves the j rapid response to load changes available from the reactor-following scheme by '

prograammed borrowing and depositing of energy. in the steam generator during transient conditions.- The ICS thus provides improved unit stability by allowing a limited change in the energy stored in both. steam generators during a transient and ensuring that normal steam pressure and temperature are restored at the desired megawatt generation after the transient has occurred. The basic require-ment of the reactor / steam generator / turbine unit control is matching megawatt generation to unit load demand. The integrated reactor / steam generator / turbine control does this by coordinating steam flow to the turbine with the rate of steam generation.

The ICS also has the capability to automatically run back and/or limit the unit output (and thereby reactor power) when certain unit conditions would prevent operation at the desired power level.

When the ICS is operating in the fully integrated mode, it utilizes feed forward control signals developed from a megawatt demand signal to control both feed-water flow.and reactor power simultaneously with changes to the turbine control valves to produce the megawatts desired. If an imbalance exists between the flow of steam to the turbine and the steam production rate, the signals to the feed-water and reactor power systems are modified to shorten the interval of time 4.11-1

required to re-establish equal steam flow / steam production conditions. The ICS is therefore characterized by fast response and stable operation of the unit as T a whole.

4.11.1 PURPOSE  !

The purposes of the Unit Load Transient Test are listed below:

(a) To demonstrate the ability of the ICS to maintain control of the unit during power changes up to the design ramp rates af ter tuning and op-timization.

(b) To demonstrate prop,er ICS response to tripping one reactor coolant pump at 40% full power and restarting it at 20% full power.

(c). To demonstrate the proper ICS response during a runback initiated by tripping one main feedwater pump at 75% full power.

Three acceptance criteria are specified for Unit Load Transient Test as listed below:

(1) Reactor power can be varied at the design ramp rates without exceeding the limits stated in Technical Specifications and Sections 1 and 2 of Plant Limits And Precautions.

(2) Each transient must be completed without causing the reactor protective ,

system to actuate.

~

(3) The tripped reactor coolant pump will not restart until reactor power is-less than 30% full power.

4.11.2 TEST METHOD During the Power Escalation Sequence at Crystal River Unit 3, the Unit Load Transient Test was performed at three different power plateaus: 40, 75, and 100% of rated full power. Table 4.11-1 gives a general summary of all transients required by the Unit Load Transient Test. The transients were performed with the'ICS in four separate modes of control:

(a) The fully integrated control mode (b) The turbine-following control mode .

(c) The reactor /0TSG - following control mode (d) The turbine /0ISG - following control mode The turbine /0TSG-following control mode is not normally employed for changing power during unit - operation. However, application of this mode does help in tuning the ICS since it tends to isolate the response of various modules.

Transient testing at the 40% and the 75% full power plateaus was performed in g accordance with the test procedure and as specified in Table 4.11-1. Transients .)

in each mode of ICS control were performed at ramp. rates of approxLmately 5% FP 4.11-2

per minute to assist in verification of the acceptance criteria. Additior al transients to check unit response to a reactor coolant pump trip at 40% full power and a feedwater pump trip at 75% full power were performed.

Transient testing at the 100% full power plateau was also performed as specified in Table 4.11-1. Transients in each mode of ICS control were performed at ramp rates of approximately 5% FP per minute between 90 to 100% full power and at ramp rates of approximately 10% FP per minute between 45 to 85% full power.

These transients were then used to assist in verification of the acceptance criteria.

The desired transient response of the unit was that reactor power could be varied at the design ramp rates of 5% FP per minute at 40 and 75 percent full power and of 10% FP per minute ht 100 percent full power and return to steady state operation without exceeding the unit parameters deviations given below:

TRANSIENT STATE Turbine Header Pressure +/- 50 Psig .

Reactor Coolant Average Temperature +/- 5 deg F Loop Temperature Mismatch +/- 5 deg F STEADY STATE Turbine Header Pressure +/ -

9 Psig Reactor Coolant Average Temperature t/- 1 deg F Loop Temperature Mismatch +/- 1 deg F Reactor Power Error +/- 1% FP Throughout this test, ICS tuning was performed when the above unit parameters indicated that tuning was necessary in order to optimize the transient response of the ICS.

The technique of inducing each transient was to decrease reactor power from the test plateau to a predetermined lower power level, then after establishing ateady state conditions, reactor power was re-escalated to the test plateau. During all transients, the variation of pertinent primary and secondary system para-meters during negative and positive power ramps was monitored and recorded on Brush recorders and the Reactimeter.

4.11.3 EVALUATION OF TEST RESULTS

, The Unit Load Transient Test was conducted at 40%, 75% and 100% full power to evaluate the ability of the Integrated Control System to accomplish smooth negative and positive changes in power level while operating in the fully in-tegrated, turbine-following, reactor /0TSG-following, and turbine /OTSG-following modes of control. It was also done _ to observe the transient response of the unit'to the loss of a reactor coolant pump and a main feedwater pump.

The data taken during each transient (i.e. power, unit average temperature, loop temperature ptsmatch, and turbine header pressure) were then analyzed. This analyses included verifying that the transient response fell within the desired limits stated in 4.11.2.

The evaluation of the results of the Unit Load Transient Test has been subdivided into three sections, one for each test plateau.

4.11-3

- - - - w e w yr - --+-,.g y y _, - y., , - -- w- 7-m- , ..%9

t 4.11.3.1 ICS TRANSIENT TEST AT THE 40% POWER LEVEL 3

The behavior of the unit during the negative and positive ramps in power level is presented in Figures 4.11-3 through 4.11-7 and summarized in Table 4.11-2.

In every case, the desired transient response was observed, thus no ICS tuning was required.

The reactor coolant pump trip portion was performed after the completion of the above ICS transient. It was initiated from a steady state condition of approxi-mately 45% full power. Figure 4.11-8 shows the behavior of various unit para-meters versus time. From this Figure, it can be concluded that the AT(c),

controller smoothly accomplished the four pump to three pump transient. Verifi-cation of the reactor coolant pumps inhibit was then demonstrated by trying to start the pump at 32% full power. Reactor power was then reduced to 20% full power and the pump re-started.

In conclusion, all required transients were performed at 40% full power with the acceptance criteria on each met. ~

4.11.3.2 ICS TRANSIENT TEST AT THE 75% POWER LEVEL The behavior of the unit during the negative and positive ramps in power level is presented in Figures 4.11-9 through 4.11-13 and summarized in Table 4.11-3.

In every case, the desired transient response was observed, thus no ICS tuning was required.

The feedwater pump portion was performed after the completion of the above ICS '

transient. The test method was to trip one of the two main feedwater pumps '

and record the transient response of the unit. Figure 4.11-14 shows the behavior of various unit parameters versus time. Even though this resulted in a rapid runback in reactor power to 56% full power, the ICS adequately correct-ed the various error signals that developed and returned the unit to a steady state condition in approximately 3 minutes.

In conclusion, all required transients were performed at 75% full power with the acceptance criteria on each met.

4.11.3.3 ICS TRANSIENT TEST AT THE 100% POVER LEVEL The behavior of the unit during the negative and positive ranps in power levels is preser ced in Figures 4.11-15 through 4.11-21 and summarized in Table 4.11-4.

In every :ase, the desired transient response was observed, thus no ICS tuning was requ; red.

In conclusion, all required transients were performed at 100% full power with the acceptance criteria on each met. 1 r l 4.

11.4 CONCLUSION

S I l

From analyses of all the test data, the following conclusions may be made for the 40, 75, and 100% full power plateau sections of the Unit Load Transient Test.

(a) All transients were performed without exceeding the limits of the Crystal River Unit 3 Technical Specifications and Sections 1 and 2 of Plant Limits And Precautions.

4.11-4

b

^

(b) All transients were completed without causing the reactor protective system 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 and within the limits stated in 4.11.2.

s e

w 4

e e

% m' 4.11-5

General Summary of Transients Required By Unit Load Transient Test

.m Transient ICS Mode of General Power Power Ie s t Group Operation Comment Decrease Increase Ramp Rate (Dim) (Dim) (Dim) (% FP) (7.FP) (%/ min)

A ., Unit Load Transient Tc:st at 40% Full Power 1.1 Fully Integrated 40 to 30 30 to 40 5 1.2 Turbine Following 40 to 30 30 to 40 5 1.3 Reactor /0TSG Following 40 to 30 30 to 40 5 1.4 Turbine /0TSG Followins (3 cases: 40 to 30 30 to 40 5 1.5 Fully integrated ,

40 to 15 15 to 40 5 1.6 Fully Integrated (RCP Trip: (Reduce Power / Start RCP) 1.7 Fully Integrated 15 to 40 5 B. Unit Ioad Transient Test' at 75% Full Power 2.1 Fully Integrated 75 to 65 65 to 75 5

2. 2 Turbine Following 75 to 65 65 to 75 5 -

2.3 Reactor /OTSG Followins 75 to 65 65 to 75 5 2.4 Turbine /0T3G Following (3 Cases) 75 to 65 65 to 75 5 2.5 Fully Integrated 75 to 30 30 to 75 5 2.6 Fully Integrated (FWP Tripi 75 to 55 2.7 Fully Integrated 55 to 75 10 C. Unit Load Transient Test at 100% Full Power 3.1 Fully Integrated -

100 to 90 90 to 10C 5

3. 2 Turbine Following 100 to 90 90 .to 100 5 3.3 Reactor /0TSG Following 100 to 90 90 to 100 5 3.4 Turbine /OTSG Followins (3 Cases) 100 to 9090 to 100 5 Fully Integrated 100 to 85 NA 3.5 Fully Integrated 85 to 45 45 to 85 10 3.6 Turbine Following 85 to 45 45 to 85 10 3.7 Reactor /0TSG Following 85 to 45 45 to 85 10

(

b I

l

. Table 4.11-1 l

ICS Transient Data Obtained During The Performance Of Unit load Transient Test At 407. Full Power Unit Transient Response Unit Steady State Return

.Transien t ICS Mode of Powe r Rate.Z/ min A T( AVE) . F A T (C) , F A1HP, PSI AP. 1 FF AT(Ave), F A T(Cl . Fl ATit? PS1 Nuwbe r Operation (II?) Ave Max Hin Max Min Pax Mfn Haw Min Max flin Max Hfn Max 5 Hin Max 1.1 Fully Integrated 40 to 29 -2.6 -4.7 -1.8 +0.0 -0.1 to.4 +0.0 +32.6 -0.2 +0.1 -0.2 40.3 -0.1 +0.1 -1.3 +2.0 29 to 39 +2.8 +4.8 +0.0 +0.9 -0.4 10.0 -35.4 +0.0 -0.2 +0.1 -0.1 +0.1 -0.1 +0.1 -0.7 60.8 l'. 2 Turbine Following 40 to 30 -2.5 -3.1 -0.7 10.5 10.0 to.4 -1.8 +4.8 -0.1 +0.1 40.0 +0.1 -0.1 40.1 -0.8 +1.6 30 to 40 +1.6 +3.0 -0.2 +0.5 -0.2 to.O -1.5 +4.8 -0.1 +0.0 -0.2 0.0 -0.1 10.1 -1.5 t0.0 1.3 Reactor /0TSC Following 40 to 30 -2.3 -3.5 -0.4 to.1 +0.0 to.3 +0.n F45.3 -0.6 40.6 -0.4 +0.2 -0.1 +0.1 -2.7 t2.7 30 to 39 +2.0 +3.5 ,+0. 0 +0.2 -0.6 t0.0 -44.1 +0.0 -0.6 40.4 -0.2 +0.1 -0.1 +0.1 -5.6 t5.2 1.4 (1) Turbine /0TSO Following 41 to D -2.2 -2.5 -2.3 +0.0 -0.2 60.1 -4.8 +0.0 -0.1 +0.1 -0.6 +0.4 -0.1 +0.1 -4.0 14.0 g (Rode In Hand) 29 to a +2.2 +2.5 +0.0 +2.1 -0.2 to.1 +0.0 48.7 -0.6 +0.3 -0.3 +0.6 -0.2 +0.2 -3.1 t3.5 h

E 1.4 (2) Turbine /0TSC Following 40 to 29 -3.4 -3.6 -5.4 +0.0 -0.2 10.5 -12.9 +3.3 -0.3 +0.1 -0.1 40.2 -0.1 +0.1 -3.8 +3.4 (Reactor Demand In Hand: 29 to 39 +3.2 +4.1 +0.0 +4.7 -0.4 to.O -2.1 +7.8 -0.1 +0.0 -0.2 + 0. 4 -0.2 +0.2 -0.9 ti.2 W

1.4 (3) Turbine /0TSG Following -3.7 10.2 +0.0 -8.1 +0.0 +0.4 -0.2 + 0.1 +0.1 +2.0 4 (Rods & Reactor In Hand 40 to 29 -2.3 -2.4 to.8 -0.4 -0.1 -2.2 1.5 Fully Integrated 42 to 15 -4.5 -7.1 -G.8 -0.1 0.0 to.6 t0.0 t32.0 -0.9 +0.6 -0.3 +0.4 -0.2 +0.6 -2.5 t2.5

+4.3i +6.2 +0.0 +1.5 -0.9 +0.3 -27.8 +0.0 -0.6 +0.3 -0.1 +0.0 -0.7 +0.5 -8.0 t4.3 1.6 Fully Integrated 45 to 32 -5.1 -5.5 -1.4 +0. ' -0.6 to.0 -4.3 t18.5 -1.8 +2.3 -0.8 +0.5 -0.1 60.1 -5.2 &l2D (ROP Trip) 1.7 Fully Integrated 32 to 41 +6.3 +8.4 -0.9 +1.1 -5.6 -1.4 -38.4 -10.6 -0.1 +0.1 -0.2 40.3 -0.5 +0.5 -5.7 t1.7 4

-ICS Transient Data Obtained During The Performance Of Unit  ;

Inad Transient Test At 757. Full Power i

Uutt T  : lent Response Unit steady State Return  ;

Transien t ICS Mode of Power Rate.2/ min A T( AVE) , F A T(c) , F A T11P PST AP, 1 FP A T( Ave) , F A T(c) . F ATilP ' PSI [

Numbe r Ope ration (IFP) Ave Max HIn Max Hin Max Mtn Max Hin Max Hin Max Hin Max Hin Ma I

'l ,

l

] 2.1 Fully Integrated 77 to 67 -3.4 -4.9 -0.2 +0.3 -0.1 +0.2 0.0 +16.5 -0.1 +0.2 -0.1 +0.2 -0.1 +0.2 -6.9 +5.7 67 to 76 -2.4 -4.9 -0.3 +0.5 -0.3 +0.4 0.0 -18.4 -0.1 +0.1 -0.1 +0.2 -0.2 +0.2 -3.9 +4.2 i i s

2.2 Turbine Following 77 to 66 -2.4 -3.5 -0.9 +0.2 -0.1 +0.4 +9.2 +1.3 -0.3 +0.4 -0.1 +0.0 -0.1 '+0. 0 -1.6 +1.2 l 66 to 78 +2.9 +4.5 -0.4 +0.7 -0.1 +0.1 0.0 +15.4 -0.3 +0.2 -0.2 +0.2 -0.2 + 0. 0 -1.3 +2.4 j

l t

i 2.3 Reactor /0TSC Following 78 to 67 -2.7 -4.8 -0.4 +0.7 -0.4 +0.3 +0.0 +24.8 -0.3 +0.3 -0.2 +0.2 -0.2 +0.2 -5.5 +2.8 67 to 77 +2.2 +3.5 +0.0 +1.6 -0.1 +0.5 -1.9 +9.3 -0.4 +0.4 -0.1 +0.0 0.0 +0.2 -6.5 +7.8 H

> 2.4 (1) Turbine /0TSG Following 78 to 66 -2.2 -2.7 -2.4 +0.0 -0.3 +0.6 -8.1 +8.8 -0.3 +0.3 -0.1 +0.0 -0.2 +0.1 +3.2

-2.4 i e

(Rods In 11and) 66 to 77 +2.2 +3.5 +0.0 +1.6 -0.1 +0.5 -1.9 +9.3 -0.4 +0.4 -0.1 +0.0 0.0 +0.2 -6.5 +7.8  !

v g 2.4 (2) Turbine /0TSC Following 79 to 67 -2.3 -3.6 -3.2 +0.0 +0.0 +0.3 -15.4 0.0 -0.2 +0.3 -0.5 +0.5 -0.1 +0.2 -2.9 +3.9 e

e (Reactor Demand In Hand 67 to 77 +2.4 +3.2 t0.0 +2.7 +0.0 +0.4 -0.9 +5.0 -0.2 +0.2 -0.2 +0.2 -0.1 +0.1 -1.8 +2.1 ta 2.4 (3) Turbine /0TSG Following 77 to 66 -2.0 -2.8 -2.1 +0.0 -0.3 +0.1 -9.6 +c.0 -0.2 +0.2 -0.2 +0.7 -0.2 +0.2 -0.7 +1.1 (Rods & Reactor In Hand) 66 to 77 +2.5 +3.2 +0.0 +2.5 +0.0 +0.2 +0.0 t10.8 -0.4 +0.6 -0.2 +0.2 -0.1 +0.1 -4.6 +5.7 2.5 Fully Integrated 78 to 31 -4.2 -7.7 -0.4 +2.6 -1.6 +1.3 -22.9 F22.4 -0.7 +0.4 -0.5 +0.3 -0.8 +0.6 -9.8 t!O.6 31 to 75 +5.6 +11.7 -3.5 +0.5 -1.2 +1.1. -17.2 +0.0 -0.9 +0.9 -0.2 +0.1 -0.7 +0.7 -5.9 +5.3 2.6 Fully Integrated 77 to 56 -16.6 -20.4 -3.9 +3.5 -0.5 +4.1 -3.3 +12.8 -0.2 +0.1 -0.3 +0.2 -0.2 +0 1 -2.7 +3.5 (FDW Pump Trip) 2.7 Pully Integrated 50 to 76 48.6 *10.8 -2.5 +0.0 -0.3 +0.7 -8.1 +0.0 -0.4 +0.5 -0.1 +0.2 -0.3 +0.3 -7.0 +4.9 I

1 l'

m .

(

)

ICS Transient Data Obtained During The Performance Of Unit lead Transient Test At 1007. Full Power Unit Transient Responsa Unit Steady Statu Return Transient .ICS Mode of Fower Rate,%/ min AT( AVE), F AT(C). F ATHP PST A P . 7. FP A T( Avo) . F A T(Cl . F A TH P PSI Numbe r Ope ration (%FP) Ave Max H!n Max Hin Max Mfn Max Hin Max Mfn Max Hin Max Hin Hax 3.1 Fully Integt ated 97 to 85 -2.5 -5.3 -0.1 +0.7 -0.2 +0.4 -3.5 +16.4 -0.4 +0.3 -0.4 +0.2 -0.1 +0.3 -2.3 +4.2 85 to 97 +2.4 +4.6 +0.0 +0.6 -0.4 +0.3 -18. 1 +0. ') -0.5 +0.2 -0.1 +0.2 -0.4 + 0. 2 -2.3 +2.3 3.2 Turbine Following 98 to 85 -2.5 -3.4 -0.6 +0.5 -0.1 +0.4 -10.8 + 0. 0 -0.4 +0.2 -0.1 +0.1 -0.1 . + 0. 3 -1.0 +0.8 85 to 97 +2.0 +3.7 -0.7 +0.5 -0.2 +0.5 -5.5 +0.5 -0.4 +0.2 -0.1 +0.1 -0.2 +0.1 -2.0 +3.3 3.3 Reactor /0TSC Following 97 to 85 -2.4 -6.7 -0.1 +0.6 -0.2 +0.4 0.0 +36.6 -0.1 +0.1 -0.3 +0.2 -0.1 +0.1 -3.7 +2.9 85 to 96 +2.8 +4.6 -0.7 +0.0 -0.4 +0.0 -12.5 +4.8 -0.4 +0.3 -0.2 +0.1 -0.2 + 0. 0 -1.8 +0.8 3.4 (1) Turbine /0TSC Following 96 to 87 -1.7 -2.9 -3.2 +0.0 -0.3 tl.3 -7.2 +0.0 -0.5 +0.4 -0.5 +0.6 -0.2 +0.4 -4.7 +2.0 h (Rods In Hand) 87 to 95 +2.0 +2.9 -0.3 +2.0 -0.4 to.1 +0.0 +11.5 -0.2 +0.3 -0.1 +0.3 -0.2 +0.3 -3.7 +2.0 E

o y 3.4 (2) Turbine /0TSG Following 95 to 87 -1.7 -2.5 -3.5 +0.0 +0.0 +0.9 -7.6 + 0. 0 -0.4 +0.1 -0.5 +0.3 -0.1 +0.0 -5.7 + 3. 8 e-*

(Reactor Demand In lland 87 to 96 +1.8 +2.6 +0.8 +3.3 -0.1 +0.3 +4.0 +9. 5 -0.1 +0.2 -0.2 .+0.2 -0.2 +0.2 -2.3 +1.6 Y

3.4 (3) Turbine /0TSG Following 96 to 88 -1.4 -2,1 -3.4 +0.0 -0.2 +1.2 -8.8 +0.0 -0.2 +0.3 -0.5 +0.5 -0.4 +0.3 -2.2 +1.5 (Rods & Reactor In' Hand. 88 to 95 +1.5 +2.9 +0.0 +3.4 40.0 +0.5 0.0 +15.8 -0.1 + 0. 2 -0.1 + 0. 2 -0.2 +0.1 -1.5 t2.1 3.5 Fully Integrated 86 to 44 -7.1 -14.9 -0.9 4' i -1.4 +0.4 -3.1 +19.0 -0.2 +0.1 -0.1 +0.1 -0.2 +0.1 -1.8 +2.8

, 44 to 86 +8.0 t10.8 -1.3 -0.2 -0.4 60.8 -29.7 0.0 -0.7 +0.4 -0.2 +0.3 -0.4 +0.2 -2.2 +1.2 3.6 Turbine Following 87 to 41 -3.4 -10.0 -0.4 +2.3 -2.6 +0.5 -12.2 +7.2 -0.1 +0.1 + 0. 0 +0.0 40.0 +0.1 -0.1 +0.1 1

41 to 85 +3.7 +8.3 -1.9 +0.4 -0.3 +1.0 -7.2 +3.0 -0.4 +0.2 -0.1 +0.1 -0.4 +0.4 -4.4 +3.2 3.7 Reactor /0TSC Following 85 to 44 -2.2 -5.6 +0.0 +1.6 -0.7 +0.6 +0.0 +34.2 -0.5 +0.1 -0.2 +0.2 -0.1 + 0.1 -3.4 +7.3 42 to 82 +3.2 +6.6 -1.6 +0.5 -0.3 +1.7 -35.1 +5.9 -1.1 +1.7 -0.8 +0.6 -0.1 10.1 -5.6 +5.7

-.. .. .-- . . - . ._, . . . - - - - - . . ~ . . - - - - . . - - -

Integrated Control System Organization '

Load Dispatch System

  • Unit Load Demand (UL)

Level 4 Integrated Master (IC) r

./

Pressure Reactor 30 Level 3 Control Control Feedwater i Control I (IC) (RC) (FW) 9 1 9 i ,

Level 2 lH/Al lH/Al lH/Al [ggjki/AllH/Al 4 i i .

6 ,

Turbine Turbine Reactor FW FW By FW Gov'r 3yste SG#1 Feed SG#2 I"C*#f* Drives Interface Interf Valve Pumps Valve (IC) (IC) (RC) Interf Interf Interf l l Turb Level Turbine Bypass Rod SGpl Feed SG#2 1 Gov r yaiye Drives Valves Pumps Valves l

l l

l l

NOTE (*): Not used at Crystal River Unit 3 Figure 4.11-1 ,

- , . . . . # .,.% _ . - , , . . . . .y..

Integrated Control System Megawatt Megawatt Throt tle ' Reactor Demand Generation Pressure Power i

v v -

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  • Rate j Control -

Cen t rol Contro1 TURBINE . REACTOR / OTSG FOLLOWING MODE . FOLLOWING MODE Figure 4.11-2

UNIT LOAD TRANSIENT TEST AT 40 PERCENT FUI.L FO'a'ER Maximum Power Increase Rate: +4. 8 % / Minute Transient Number: 1.1 S

!!aximum Power Decrease Rate: -4. 7 % / Minute ICS Mode: Fully Integrated

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UNIT LOAD TRANSIENT TEST AT 40 PERCENT FULL POWER Maximum Power Increase Rate:+3.0 %/ Minute Transient Number: 1.2 Maxi =us Power Decrease Rate:-3.1 '%/ Minute ICS Mode: Turbine Followine

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Time, Seconds Figure 4.11-4

UNIT LOAD TRANSIENT TEST AI A PERCENT FULL POWER Maximum Power Increase Rate: +3. 5% / Minute Transient Number: M Maximum Power Decrease Rate: -3.5%/ Minute ICS Mode: Reactor /OTSC ^

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t Figure 4.11-19

UNIT LOAD TRANSIENT TEST AT 100 PERCENT FUI.L PO'aT.R 1

Maxi =u= Power Increase Rate: +8.3 %/ Minute Transient Number: 3.6 )

, Maximum Power Decrease Rate:-10.0 %/ Minute ICS Mede: Turbine Followine  !

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UNIT LOAD TRANSIENT TEST AT 100 PERCENT FULL POWER Maximum Power Increase Rate:+6.6 %/ Minute Transient Number: 3.7 Maximum Power Decrease Rate:-5.6 %/ Minute ICS Mode: Reactor /0TSG '

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l l

I r- 4.12 MMm'DOUN FROM OUTSIDE CONTROL ROOM TEST The shutdown from outside the control room test simulates an emergency situa-tion requiring evacuation of the control room. All plant controls are luft in automatic unless remote indication requires taking them into manual modes of operation.

4.12.1 PURPOSE The purpose of this test is to demonstrate that the unit can safely be brought to hot standby conditions from outside the control room.

4.12.2 TEST METHOD i l

The test was started from 15% power. Letdown flow was stopped and the control room evacuated by the normal shift complement of operators. The operators 1 manned their stations as shown in Table 4.12-1. A complete second set of op- 'l '

erators was left in the control room to assume plant control if the test fail-ed. The reactor was tripped remotely and the plant allowed to come to hot standby automatically. The operators outside the control room were to take control of various equipment if it was not performing adequately in automatic.

4.12.3 EVALUATION OF TEST RESULTS On April 16, 1977, the first run at shutdown from outside the control rocm was attempted and aborted after approximately 18 minutes due to feedpump AP oscil-lations. Main feedwater pump speed control had been shifted to hand by the operators _ in the control room early in the test. Af ter the reactor trip high leakage through the startup valves resulted in. overfeeding both steam genera-tors. The test was terminated due to loss of steam generater level control and greater than desired cooldown of the primary plaat. As a result of this ex-perience, the plant emergency procedure for shutdown from outside the control room was changed to require tripping the main feedpump remotely. This would allow the steam driven emergency feedwater pump to start and take over steam generator feed requirements.

On April 22, 1977, the test was repeated with the above modifications. This run was stopped after approximately 9 minutes due to low levels in both steam generators. Subsequent investigation revealed that initial conditions for steam pressure to the steam driven emergency feedwater pump were not met. This resulted in both steam generators being dry

  • until flow was established by the electrical driven emergency feedpump. The plant emergency procedure was modified to require the operator to check the steam driven pump and,,1f it is not operating properly, start the electrical driven pump. ,

On April 23, 1977, the test was run successfully. The control room was evacua-ted with the plant at 15% power. The running main feedpump was tripped remote-ly (which trips the main turbine). The reactor, however, was not tripped until at least one minute later, resulting in low steam generator levels. The operators

  • The steam generators were designed for 20 allowable thermal cycles equivalent s to being boiled dry.

4.12-1 I

~ _ _

started the electric driven feedwater pump, took manual control of both feed- ,

water startup valves and restored level in both steam generators. Twenty minutes into the test, an operator remotely added water to the makeup tank, otherwise the plant remained in a fully automatic mode of operation and came to a hot standby condition. The test was allowed to run for thirty minutes to verify that the operators outside the control room had complete control of the plant. At this time, plant parameters were at or near their final steady state values (see Figure 4.12-1) and the test was ended.

Although level and feedflow indication do not show zero, post test analysis Ladicates that the steam generators were dry about seven minutes. (See Figure 4.12-1). This occurred because of a combination of problems with reference legs, flows, and/or calibration errors. This can be verified by noting that during the dry period, main steam pressure was below the saturation pressure and recovered as soon as feedflow was re-established.

4.

12.4 CONCLUSION

S -

The test proved that the reactor can be brought to and maintained in a safe hot standby condition from locations outside the control room by the normal shift complement of operators.

)

4.12-2 TL 1

Indi .ioual Personnel Responsibilities During The Performance Of Shutdown From Outside The Control Room 7.est Assignment Number Personnel Title Individual Personnel Responsibilities 1 Nuclear Operator 1. Coordinate the shutdown of the unit at the remote shut-down panel by directing operators to perform specific functions at various stations around the unit. Start electrically driven emergency FW pump, if required.

2 Chief Nuclear Operator 1. Control RCS pressure through pressurizer spray valve RCV-14.

2. Trip main feed pump (s) and main turbine locally.

H

g. 3 Assistant Nuclear Operator 1. Check that the steam driven emergency feedwater pump is y (Plant) on line with adequate discharge pressure.
  • 2. Cycle MUV-30 or MUV-73 as necessary to maintain makeup g .

tank level.

Y r

4. Assistant Nuclear Operator 1. Maintain steam generator level with FWV-39 and FWV-40.
2. Control RCS outlec temperature with MSV-25 and MSV-26.

5 Nuclear Auxiliary Operator 3. Verify that power to the 6900 V and 4160V unit buses has transferred to the S/U transformers.

2. ' Assist any member of the operating crew as requested.

6 Shift Supervisor 1. To ensure that the unit shutdown is proceeding properly as observed from the remote control shutdown area.

2. Trip the reactor by tripping the CRD breakers.

,_ _ _ _ _ _ _ _ . - - - ^ ^ ~

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4.13 PSEUDO CONTROL ROD EJECTION TEST

/' During normal full power operation, a control rod assembly could only be eject-ed from the core if a physical failure of a pressure barrier component in the control rod drive assembly occurred. Such a failure would cause a pressure differential to act on a control rod assembly and rapidly eject the assembly from the core region. The resulting power excursion due to the rapid incraase in reactivity is ultimately limited by the Doppler Effect and terminated by reactor protection system trip. The severity of the rod ejection accident is dependent upon the worth of the ejected control rod assembly, the reactor power level, and the core power distribution. For these reasons, the maximum allow-able control rod assembly worth at power is set at 0.65%Ak/k. The only control rod group fully inserted during full power operation is group 7, which is used for load control. Other rod groups are maintained out of the core for rapid shutdown. Thus, the maximum worth of an ejected rod at power is limited to the most reactive rod of control rod group 7 which is rod 7-1 at core location H-08.

This selection was based on calculations presented in the FSAR, Section 3.2.2.

Figure 4.13-1 shows the location of this control rod and its location relative to each core quadrant.

4.13.1 PURPOSE 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 Ejected Control Rod Test, this was accomplished by determinirg the worth of the most reactive control rod which is inserted in the core during full power operation.

('

The acceptance criterion specified for the Pseudo Control Rod Ejection Test is that the most reactive control rod worth does not exceed 0.65%Ak/k.

4.13.2 TEST NETHOD The pseudo control rod ejection worth measurement was performed by static techniques, in which the reactor is maintained in an approximate critical state throug. control rod 7-1 and group 6 exchange. During this exchange, differential rod wcrth measurements were done on 8roup 6 for each 20% withdrawal of control rod 7-1 by the fast insertion / withdrawal method. The above data obtained was then used to determine the average differential worth of rod 7-1 as a function of rod position as shown in Figure 4.13-2. The integral worth of rod 7-1 was then obtained by integration of the differential rod worth curve using the trapezoidal nethod.

The effect of _the power distortion produced by the withdrawal of control rod  !

7-1 was measured using the unit computer performance data output program which printed out core power distribution and thermal hydraulics data at 22% and 100% withdrawn. Since control rod 7-1 is the center contrcl rod as shown in Figure 4.13-1, 1/8 core symmetry was present throughout the test. A comparison of the radial and total core power distributions measured before and after the pseudo ejection of control rod 7-1 is presented in Figure 4.13-3 and 4.13-4.

4.13.3 EVALUATION OF TEST RESULTS The test results at 40 percent full power in terms of reactivity were calculated by integrating the differeatial worth curve measured on rod 7-1 as it was 4.13-1

withdrawn as shown in Figure 4.13-2. An integration of the measured differen-tial worth curve gives an ejected control rod worth of 0.20%4k/k as compared to ,

the predicted value of 0.49%Ak/k. This value is well below the acceptance criterion of 0.65%Ak/k.

The effect of withdrawing control rod 7-1 on the core power distribution was determined by recording Core Power Distribution and Thermal Hydraulic data at 22% and 100% withdrawn as shown in Table 4.13-1. The graphical representation of radial peaking factors as a function of radial position along the X-Z plane at Level 08 is presented in Figure 4.13-5. This exemplifies the marked devia-tion in the radial peaking factor in the proximity of H-08 due to the pseudo ejection of control rod 7-1. The worst case maximum linear heat rate LHR (kW/

ft), the worst case minimum DNBR (dim) and the maximum total peaking factor (dim) measured with rod 7-1 at 100% withdrawn were 9.18, 4.87, and 2.89, rerpectively. These values were measured in core location H-08, the fuel asuembly containing the ejected rod. The effects on quadrant power tilt was swall, since control rod 7-1 was located in the center of the core. Maximum positive quadrant power tilts of +0.94 and +1.13 percent were measured in quadrant WX before and after ejection of control rod 7-1 respectively. The shift of the power to the bottom of the core as indicated by the incore offset was the result of the insertion of control rod group 6 from 94.2 to 64.8%

withdrawn and by the withdrawal of control rod 7-1 from 22 to 100% withdrawn.

Thus, the large change of offset did not totally result from the ejection of control rod 7-1.

4.

13.4 CONCLUSION

S The measured worth of t) sost ras etive control rod was found to be 0.20%4k/k which is less than the , star..e criterion of 0.65%Ak/k.

Analyzed core power distribution and therma' hydraulic data indicated a large perturbation to the steady state core power distribution, as was expected.

4.13-2

~ ~

I Summary Of Lare Power Distributions And Thermal Hydraulics Data Taken During The Performance Of Pseudo Rod Ejection Test Rod 7-1 Power Incore Worst Case Worst Case Position Level Offset Guadrant Tilt. I Maximum Peakina. Dim Maximum LHR Minimum DNBR

(% wd) (% FP) (%) WX XY WZ ZY Radial Total (kW/ft) (Dim) 22 40.21 - 0.02 +0.94 -0.93 +0.54 -0.60 1.40 1.62 4.44 10.13 100 46.61 -31.14 +1.13 -0.85 4 0.64 -0.92 2.07 2.89 9.18 4.87 N

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l 5

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