ML18088A920

Startup Test Report

ML18088A920
Person / Time
Site:	Saint Lucie
Issue date:	12/31/1976
From:	Florida Power & Light Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML18088A920 (268)

Text

FLORIDA POWER AND LIGHT COMPANY ST. LUCIE NUCLEAR POWER PLANT UNIT, I DOCKET NO. 50-335, LICENSE NO. DPR-67 STARTUP TEST REPORT DECEMBER, 1976

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TABLE OF CONTENTS Section ~Pa e

1.0 INTRODUCTION

AND

1.1 INTRODUCTION

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1.2 SlRKARY 1.2.1 INITIAL FUEL LOAD 2 1.2.2 POST CORE HOT FUNCTIONAL TESTS 2 1.2.3 INITIAL APPROACH TO CRITICALITY 3 1.2.4 LOW POWER PHYSICS TESTS 3 1.2.5 ESCALATION TO POWER TESTS 3 2.0 INITIAL FUEL LOAD 3.0 POST CORE HOT FUNCTIONAL TESTS 22 3.1 CEDM/CEA PERFORMANCE TESTS 24 3.1.1 PURPOSE 24 3.1.2 TEST RESULTS 24 3.

1.3 CONCLUSION

S 25 3.2 REACTOR COOLANT SYSTEM PUMP FLOW AND COASTDOWN TEST 27 3.2.1 PURPOSE 27 3.2.2 TEST RESULTS 27 3.

2.3 CONCLUSION

S 28 3.3 REACTOR COOLANT SYSTEM LEAK TESTS 33 3.3.1 PURPOSE 33 3.3.2 TEST RESULTS LEAK TEST 33 3.3.3 TEST RESULTS LEAK RATE TEST 33

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3.4 CONCLUSION

S 33 3.4 PRIMARY AND SECONDARY WATER CHEMISTRY 34 3.4.1 PURPOSE 34 3.4.2 TEST RESULTS 34 3.4.2.1 CHEMISTRY CONTROL 34 3.4.2.2 BASELINE CORROSION 34 3.4.2.3 CHEMICAL SHOCK TREATMENT 35 3.

4.3 CONCLUSION

S 35 3.5 INCORE INSTRUMENTATION FUNCTIONAL 'TESTS 36 3.5.1 PURPOSE 36 3.5.2 TEST RESULTS 36 3.5.2.1 FIXED DETECTORS 36 3.5.2.2 MOVABLE DETECTORS 36 3.

5.3 CONCLUSION

S 37 3.6 REACTOR COOLANT SYSTEM PIPING THERMAL EXPANSION AND RESTRAINT 38 3.6.1 PURPOSE 38 3.6.2 TEST RESULTS 38 3.

6.3 CONCLUSION

S 38 3.7 REACTOR COOLANT SYSTEM AND STEAM GENERATOR 'INSTRUMENTA-

"'"TION'CALIBRATION'CHECKS 39

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C Section ~Pa e 3.7.1 PURPOSE 39 3.7.2 TEST RESULTS 39 3.

7.3 CONCLUSION

S 39 3.8 PRESSURIZER CONTROLS FUNCTIONAL TEST 40 3.8.1 PURPOSE 40 3.8.2 TEST RESULTS 40 3.8.3 .CONCLUSIONS 40 3.9 MAIN GENERATOR AIR FLOW TEST 41 3.9.1 PURPOSE 41 3.9.2 TEST RESULTS 41 3.

9.3 CONCLUSION

S 41 3.10 REACTOR COOLANT SYSTEM HEAT LOSS 42 3.10.1 PURPOSE 42 3.10.2 TEST RESULTS 42 3.

10.3 CONCLUSION

S 43 4.0 INITIALAPPROACH TO CRITICALITY 44 5.0 LOW POWER PHYSICS TESTS 53 5.1 CRITICAL BORON CONCENTRATION MEASUREMENTS 57 5.1.1 PURPOSE 57

'5.1.2 TEST RESULTS 57 5.

1.3 CONCLUSION

S 57 5.2 CRITICAL BORON CONCEiiTRATION AND SOLUBLE BORON WORTH MEASUREMENTS 59 5.2.1 PURPOSE 59 5.2.2 TEST RESULTS 59 5.

2.3 CONCLUSION

S t 59 5.3 CHEMICAL AND RADIOCHEMICAL TESTS 61 5.3.1 PURPOSE 61 5.3.2 TEST RESULTS 61 5.3.2.1 BASE LINE CORROSION 61 5.3.2.2 DISSOLVED OXYGEN 61 5.3.2.3 FISSION AND ACTIVATION PRODUCT BUILDUP 62 5.3.2.4 LITHIUM BUILDUP 62 5.3.2.5 DEMINERALIZERS (D.F.) 62 5.

3.3 CONCLUSION

S 63 5.4 TEMPERATURE COEFFICIENT OF REACTIVITY MEASUREMENTS 64 5.4.1 PURPOSE 64

5. 4.2 . TEST RESULTS 64 5.

4.3 CONCLUSION

S 64 5.5 NON-OVERLAPPED REGULATING AND SHUTDOWN CEA GROUP WORTH MEASURVKNTS 66 5.5.1 PURPOSE 66 5.5.2 TEST RESULTS 66 5.

5.3 CONCLUSION

S'VERLAPPED 66 5.6 REGULATING CEA GROUP WORTH MEASUREMENTS 77 5.6.1 PURPOSE 77'7 5.6.2 TEST RESULTS 5.

6.3 CONCLUSION

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Section ~Pa e 5.7 PRESSURE COEFFICIENT OF REACTIVITY'MEASUREMENTS 79 5.7.1 PURPOSE 79 5.7.2 TEST RESULTS 79 5.

7.3 CONCLUSION

S 79 5.8 DROPPED CEA'WORTH MEASUREMENTS 80 5.8.1 PURPOSE 80 5.8.2 TEST RESULTS 80 5.

8.3 CONCLUSION

S 80 5.9 EJECTED CEA WORTH MEASUREMENTS 82 5.9.1 PURPOSE 82 5.9.2 TEST RESULTS (ZERO POWER 6 FULL POWER) 82 5.9.3 "'ONCLUSIONS 82 5.10 STUCK CEA WORTH MEASUREMENT 84 5.10.1 PURPOSE 84 5.10.2 TEST RESULTS 84 5.

10.3 CONCLUSION

S 84 5.11 PART. LENGTH CEA GROUP MEASUREMENTS 85 5.11.1 PURPOSE'EST 85 5.11.2 RESULTS 85 5.

11.3 CONCLUSION

S 85 6;0 POWER ASCENSION TESTS 86 6.1 T V A 88 6.1.1 PURPOSE 88 6.1.2 TEST RESULTS 88

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1.3 CONCLUSION

S 88 6.2 MAIN GENERATOR EXCITATION SYSTEM INITIAL'OPERATION 89 6.2.1 PURPOSE 89 6.2.2 TEST RESULTS 89 6.

2.3 CONCLUSION

S 89 6.3 20% POWER TRIP TEST AND AUXILIARY'TO STARTUP TRANSFORMER AUTO TRANSFER'TEST 90 6.3.1 PURPOSE 90 6.3.2 TEST RESULTS 90 6.

3.3 CONCLUSION

S 90 6.4 PLANT POWER CALIBRATION 91 6.4.1 PURPOSE 91 6.4.2 TEST RESULTS 91 6.

4.3 CONCLUSION

S 91 6.5 POWER RANGE SAFETY AND CONTROL SUBCHANNEL CALIBRATION 93 6.5.1 PURPOSE 93 6.5.2 TEST RESULTS 93 6.

5.3 CONCLUSION

S 93 6.6 SHIELDING EFFECTIVENESS AND PLANT RADIATION LEVEL MEASUREMENTS 94

6. 6.1 PURPOSE 94 6.6.2 TEST RESULTS 94 6.

6.3 CONCLUSION

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Section ~Pa e 6.7 CHEMISTRY AND RADIOCHEMISTRY 'TESTS AT POWER 96 6.7.1 PURPOSE 96 6.7.2 TEST RESULTS 96 6.7.2.1 PRIMARY 96 6.7.2.2 SECONDARY 96 6.

7.3 CONCLUSION

S 97 6.8 FIXED INCORE DETECTOR ALARM SETPOINTS 99 6.8.1 PURPOSE 99 6.8.2 TEST RESULTS 99 6.

8.3 CONCLUSION

S" 6.9 REACTIVITY COEFFICIENT MEASUREMENTS POWER'h'MODERATOR 99'00 T&iPERATURE 'OEFFICIENTS 6.9.1 PURPOSE 100 6.9.2 TEST RESULTS 100 6.

9.3 CONCLUSION

S 100 6.10 . TOTAL RADIAL PEAKING FACTOR 102 6.10.1 PURPOSE 102 6.10.2 ~ TEST RESULTS 4 102 6.

10.3 CONCLUSION

S 102 6.11 XENON FOLLOW MEASUREMENTS 104 6.11.1 PURPOSE 104 6.11.2 TEST RESULTS 104 6.

11.3 CONCLUSION

S 104 6.12 PSUEDO EJECTED CEA POWER DISTRIBUTION MEASUREMENT 108 6.12.1 PURPOSE 108'08 6.12.2 TEST RESULTS 6.

12.3 CONCLUSION

S 108 6.13 CORE POWER DISTRIBUTIONS 110 6.13.1 PURPOSE 110 6.13.2 TEST;RESULTS 110 6.13.2.1 FUEL ASSEMBLY POWER FRACTION 110 6.13.2.2 6.13.2.3 AXIAL POWER DISTRIBUTION ill 111 6.13.3 6.14 PEAK LHR CONCLUSIONS GENERATOR TRIP WITH SHUTDOWN OUTSIDE CONTROL ROOM ill 118 6.14'.1 PURPOSE 118 6.14.2 TEST RESULTS 118 6.

14.3 CONCLUSION

S 118 6.15 STEAM GENERATOR FEEDWATER HAMMER TEST 119 6.15.1 PURPOSE 119 6.15.2 TEST RESULTS 119 6.

15.3 CONCLUSION

S. 119 7.0 UNUSUAL EVENTS 120 7.1 HIGHER THAN PREDICTED'COOLING'WATER DISCHARGE'CANAL LEVELS 121 7.2 'CEDM 44 INOPERABILITY AT LOW TEMPERATURES 123 7.3 APPARENT LOW REACTOR'COOLING'PUMP FLOW 124 7.4 HIGHER'THAN'PREDICTED'CONTAINMENT'RADIATION'LEVELS 125 7.5 POWER DISTRIBUTION'ANOMALY 126

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l. 0 INTRODUCTION AND

SUMMARY

1.1 Introduction This report fulfills the requirement of Regulatory Guide 1.16 which states that a Startup Test Report will be submitted to the NRC within 9 months of initial criticality. Initial criticality was April 22, 1976. Due to a flux distribution problem, comp-letion of our test program has been delayed. Therefore, this report covers only testing through the 50% plateau (see below for description of test program) then discusses the flux distri-bution problem and resolution. A supplementary report will be submitted on the remainder of the test program as required by Regulatory Guide 1.16.

The Startup Test Program was organized and administered by Florida Power and Light Company (FP&L) personnel assisted by Combustion Engineering (CE) Startup Engineers on-site and home office personnel in Windsor, Connecticut (CE, Windsor). The Startup Test Program consisted of several phases. CE commented on the test results from each phase. Then, the Facility Review Group (FRG) reviewed the results of each phase. Composition of the FRG is defined in our Technical Specifications, Section 6.5.1. Any test results falling outside of acceptance criteria were resolved prior to beginning the next test phase. The test phases were as follows:

(1) Initial Fuel Load (2) Post Core Hot Functional Tests (3) Initial Approach to Criticality (4) Low Power Physics Tests (5) Escalation to Power Tests 20% Plateau (6) Escalation to Power Tests 50% Plateau

• (7) Escalation to Power Tests - 80% Plateau

• (8) Escalation to Power Tests -100% Plateau Maximum licensed reactor core power level (100%) is 2560 MWt.

The Startup Test Program began March 3, 1976 with the loading of the first fuel assembly into the reactor vessel and was ter-minated due to the flux distribution anomaly July 10, 1976.

• Not included in this report. Results of these tests and any repeated during our Startup and ascent to power after resolution of the flux distribution problem will be in-cluded in our Supplementary Startup Report (s).

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1.0 INTRODUCTION

AND

SUMMARY

(Cont.)

1.1 Introduction (Cont.)

Since this unit is identical in core characteristics to other CE 2560 MWt plants, .the testing performed, by Calvert Cliffs Unit No.l .

and Millstone Point Unit No. 2 formed a basis for elimination of some phyaics testing..at..St. Lucie,Unit No. l.

1.2 Summary 1.2.1 Initial Fuel Load Fuel loading commenced on March 3, 1976 and was, completed on March ll, 1976.. A sizable portion of this time was spent in non-productive activities. The largest non-productive time consumption (approximately 15% of total time) was associated with fuel handling equipment mal-function and maintenance. There was a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> delay while inspecting fuel elements A047 and B032, which may have come into physical contact as A047 was being lowered into the core. None of the inspected fuel assemblies showed visible signs of damage. Fuel loading was done 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day by three crews working 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts. Without any delays or equipment problems, an experienced crew could load 12 to 14 fuel assemblies per shift.

A 1.2.2 Post Core Load Hot Functional Tests Post Core Load Hot Functional Testing (PCHF) commenced with plant heatup on March 26, 1976. PCHF was completed on April 20, 1976. In addition to the tests required, by the FSAR and Regulatory Guides, other tests were per-formed due to maintenance and previous inability to comp-lete tests. These tests were Reactor Coolant System (RCS)

Expansion, RCS and Steam Generator instrument calibration checks, RCS Heat Loss, Pressurizer Controls Functional Test and Main Generator Windage test. All the tests met acceptance criteria with the exception of RCS Flow and CEDM 44 performance testing. RCS flow was measured by. AP to be slightly (5%) below the required value. The Operating License currently limits power level to 90% of full power and the problem is still under evaluation. CEDM 44 is a full length Control"Element.Assembly (CEA) in Shut-down Group A. It would not operate properly (withdraw) during cold testing, but did trip properly under all conditions. CEDM 44 has been replaced and was tested upon return to operation after resolution of the flux distribu-tion anomaly.

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Page 3 1.2.3 Initial Approach to Criticality The initial approach to criticality commenced at 0815 on April 20, 1976. The reactor was declared critical at 0830 on April 22, 1976. The only problems of consequence en-countered were with the CEDM control system and the digital data processing system (DDPS). The CEDM control system gave unwarranted "deviation" and "out of sequence" alarms.

The DDPS was reading CEA height in error. New values of CEA levels were entered into the DDPS and the situation was corrected. A slow RCS dilution followed CEA withdrawal.

Measured RCS soluble boron concentration at criticality was in close agreement with that which was predicted and well within the acceptance criteria.

1.2.4 Low Power Physics Tests The Low Power Physics Test (LPPT) phase commenced on April 22, 1976. The LPPT phase was completed on April 30, 1976.

There were no significant delays or occurrences. Most LPPT measurements were in close agreement with predictions and all were within acceptable limits.

1.2.5 Power Ascension Testing The Power Ascension Testing began on May 4, 1976. The testing progressed through the completion of the 50%

power plateau. Power was increased to 80% before it was decided to withdraw from power ascension testing due to the development of a reactivity anomaly in the form of an unacceptable azimuthal tilt and increased axial peaking.

The reactor was shut down on July 10, 1976. A second LPPT program (mini-LPPT) was run in mid-July 1976 to evaluate the flux distribution anomaly. (See Section 7).

II Page 4 2.0 INITIAL FUEL LOAD At 0410 hours0.00475 days <br />0.114 hours <br />6.779101e-4 weeks <br />1.56005e-4 months <br /> on March 3, 1976, fuel assembly No.l containing neutron source No.l was loaded into core location X-ll. Fuel loading was com-pleted at 1426 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.42593e-4 months <br /> on March 11, 1976 when fuel assembly No.217 was loaded into core location V-7.

Table 2.0-1 and Figure 2.0-1 show the fuel loading sequence. Figures 2.0-2 and 2.0-3 show fuel assembly location and CEA location by their respective serial numbers. Table 2.0-2 gives core design cha'racteristics.

Neutron count rate was monitored during loading on four separate detector channels, Hide Range Log Channels B and D plus two temporary detectors.

Temporary Detector A in location V-7 and Temporary Detector B in location V-15 were installed prior to core loading. In step 140-B .

detector B was moved to location D-15. Independent plots of inverse count rate versus the number of fuel assemblies loaded were maintained to ensure the reactor remained subcritical at all times during loading.

Fuel loading was conducted with the spent'uel pool dry, refueling cavity full to the top of the fuel transfer tube flange and the reactor vessel filled to above the vessel nozzles but below the internals support ledge. A refueling boron concentration of >1720ppm boron was maintained with shutdown cooling flow through the core in a'ccoidance with the Technical Specifications at all times, No major problems were encountered during fuel loading. Numerous minor problems were encountered with fuel handling equipment, resulting in a total loss of approximately 2 days. These problems were various in nature and all were corrected by plant maintenance per-sonnel under the direction of the vendor representative on site.

Three Reportable Occurances were generated relating to initial core loading and are described as follows and in Licensee Event Reports Nos. 335-76-1, 335-76-2, and 335-76-4.

335-76-1 During initial core loading, the water level in the Refueling Cavity was found to be approx. 2 inches below the top of the Fuel Transfer tube. This was in conflict with spec.-3.9.4.c, which requires no direct access from containment to outside atmosphere incapable of auto-matic isolation. The immediate corrective action was to suspend core alterations until the level was restored and to require more frequent surveillance of the water level in the Refueling Cavity.

After a thorough investigation, two possible causes for the event were found. The first is that during electrical checkouts on the re-fueling canal sump pump, more water may have been pumped than had been realized. The corrective action was to cancel the blue tag (startup test) clearance on the pump and clearance tag the pump to the Nuclear Plant Supervisor. The second possible cause is an incorrect valve lineup, causing a gradual decrease in the water level.

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Page 5 During subsequent refilling operations, this lineup error may have been corrected. The corrective action was to re-verify the valve lineups affecting refueling cavity water level. The water that was lost was identified to have ended up in the Equipment and Chemical Drain Tanks-Due to the limited number of pathways to transmit this water, the two above situations are considered the only probable causes.

335-76-2 During initial core loading, the spent fuel crane overload interlock setpoint was found to have drifted above 2000 pounds. This is con-trary to the surveillance requirements of Technical Specification 4.9.7.

The immediate corrective action was to stop core loading.

The cause of the occurrence was malfunction of the crane interlock.

The interlock was malfunctioning such that it would occasionally prevent a fuel assembly from being lifted. Conversely, when tested, the interlock would occasionally permit loads in excess of 2000 pounds to be lifted. This indicated an intermittent condition. The immed-iate action was to propose that the requirement for the interlock be temporarily suspended to allow fuel loading to continue. The proposal included administratively limiting the permissible crane load to 2000 pounds by limiting the objects which could be lifted to a fuel ele-ment assembly or a control element assembly, neither of which weighs in excess of 2000 pounds. The temporary suspension, effective until midnight, March 19, was granted on March 5, 1976, by letter from the NRC Division of Project Management. The malfunction was later found tb be due to improper assembly of the load sensor which caused inconsistent operation. The interlock Fias been repaired and functions properly.

335-76-4 During initial core loading, a containment purge fan was started and containment pressure became subatmospheric causing a Containment Vac-uum Relief Valve to open. This was in conflict with the wording of Technical Specification 3.9.4 which requires that, during refueling operations, there be no direct access from containment to the outside atmosphere which is incapable of automatic isolation. In order to protect the containment structure from excessive vacuum during all

. modes of operation, the Containment Vacuum Relief Valves, by design, do not receive a Containment Isolation Signal (CIS). They do close again when containment pressure approaches atmospheric; however>>

they are not capable of being closed automatically by CIS. The immed-iate corrective action was to suspend core loading until the valves were closed.

Page 6 Specification 3.9.4 does not consider the unique function of the Con-tainment Vacuum Relief Valves. Even though containment integrity is not exactly as described in the specification when one of these valves opens, it should be noted that the valves and their associated check valves do close automatically when containment pressure approaches atmospheric.

Thus, actuation of the relief valves for the purpose of performing their design function (protection of the containment from excessive vacuum) does not violate the concept of containment integrity be-cause there is no outflow of air from containment to the outside atmosphere through the relief valves. A 'request for license ammendment

{Appendix A, Technical Specifications) to allow for correct operation of the Containment Vacuum Relief Valves has been submitted and is awaiting NRC action.

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Page 7 TABLE 2.0-1 FUEL ASSR'PLY LOADING SE UENCE HTEF NO. FUEL ASSKiBLY NO. CEA NO. CORE LOCATION B009 Neutron Source 1 X-11 C031 Y-10 C032 Y-12 C105 40 X-13 B067 W-13 A025 72 M-ll B057 W-9 C-104 31 X-9 C-007 Y-8 10 C-114 X-7 A002 W-7 A010 30 13 B017 V-11 14 A038 39 V-13 15 A004 47 W-15 16 C106 X-15 17 C023 Y-14 18 C016 X-16 19 B020 W-16 20 A015 51 V-16 21 B069 T-16 22 A048 46 T-15 23 B010 T-13 24 A049 T-ll

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Page 8 TABLE 2. 0-1 CONT 'D.

E STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 25 B001 T-9 26 A008 24 27 B012 T-6 28 A013 18 V-6 29 B080 C027 X-6 31 C008 X-5.

32 C212 14 W-5 33 B046 V-5 34 A018 T-5 35 B013 S-5 36 A054 17 S-6 37 B045 S-7 38 A055 S-9 39 B026 S-11 40 A061 S-13 41 B035 S-15 42 A053 50 S-16 43 B071 S-17 44 A019 H T-17 45 B037 V-17 C208 H-17 IP 47 C001 X-17 48 A031 54 R-17

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Page 9 TABLE 2.0-1 CONT'D.

STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 49 B007 R-16 50 A037 45 R-15 51 B004 R-13 A030 34 R-11 53 B038 R-9 54 A059 23 R-7 55 B040 56 A056 13 R-5 57 B064 N-5 r 58 A064 N-6 59 B056 60 A034 29 N-9 B051 N-ll 62 A057 38 N-13 63 B065 N-15 64 A060 N-16 65 B077 N-17 66 A043 L-17 67 B044 68 A068 44 L-15 69 B054 L-13 70 A032 33 L-ll 71 B025 L-9 72 A041 ,22 L-7

5 Page 10 TABLE 2.0-1 CONT'D STEP NO. ASSEMBLY NO. CEA NO. CORE'OCATION 74'UEL 73 B053 L-6 A044 L-5 75 B059 76 A058 J-6 77 B058 J-7 78 A050 28 J-9 79 B055 J<<ll 80 A045 37 J-13 81 B028 J-15 82 A063 J-16 83 B061 J-17 84 A042 G-17 85 B031 G-16 86 A051 43 G-15 87 B033 G-13 88 A052 32 G-ll 89 B070 G-9 90 A039 21 G-7 91 B005 G-6 I

92 A035 12 G-5 B052 F-5 A065 16 F-6 95 B072 F-7 96 A067 F-9

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Page 11 TABLE 2.0-1 CONT'D.

STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 97 B011 F-11 98 A066 F-13 I

99 B015 F-15 100 A069 F-16 101 B063 F-17 102 A062 E-17 103 B016 E-16 104 A036 42 E-15 105 B003 E-13 106 A012 E-ll 107 B032 E-9 108 A047- 20 E-7 109 B002 E-6 110 A027 E-5 B024 D-5 112 A029 15 D-6 113 B014 D-7 114 A046 27 D-9 115 B048 D-ll 116 A007 36 D-13 117 A017 48 D-16 118 B021 D-17 119 C206 52 C-17 120 B064 C-16

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Page 12 TABLE 2.0-1 CONT'D.

STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 121 A001 41 C-15 122 B075 C-13 123 A023 71 C-11 124 B022 C-9 125 A003 19 C-7 126 B027 C-6 127 C209 C-5 128 C003 B-5 129 C017 B-6 130 C108 B-7 131 C115 26 B-9 132 B050 NEUTRON SOURCE 2 B-ll 133 C101 35 B-13 134 C113 B-15 135 C037 136 C009 B-17 137 C006 A-14 138 C026 A-12 139 C018 A-10 140 C010 A-8 141 B068 V-15 142 C012 143 C202 67 V-4 144 B043 T-4

Page 13 TABLE 2 0-1 CONTrD STEP NO. FUEL ASSEHBLY NO. CEA NO. CORE LOCATION 145 A014 10 S-4 146 B041 A040 09 N-4 148 BO46 L-4 149 A011 08 J-4 150 B060 151 A016 07 F-4 152 B030 153 C203 66 D-4 154 C019 C-4 155 C029 D-3 156 C207 03 E-3 157 B066 F-3 158 A033 04 G-3 159 B047 J-3 160 A024 70 L-3 161 B029 162 A022 05 R-3 163 B074 S-3 164 C210 06 165 C015 V-3 166 C002 T-2 167 C035 S-2 168 C102 R-2

II Page 14 TABLE 2.0-1 CONT'D.

STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 169 C116 02 N-2 170 B079 L-2 171 C110 01 J-2 172 C103 G>>2 173 C040 F-2 174 C014 E-2 175 C005 H-1 176 C030 K-1 177 C011 M-1 178 C021 P-1 179 C034 W-18 180 C205 69 V-18 181 B049 T-18 182 A020 59 S-18 183 B078 184 A009 58 N-18 185 B006 L-18 186 A006 57 J-18 187 B018 G-18 188 A021 56 F-18 189 B019 E-18 190 C201 68 D-18 191 C020 C-18 192 C038 D-19

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Page 15 TABLE 2 0-1 CONTtD STEP NO. 'FUEL'ASSEMBLY'NO. CEA NO. CORE LOCATXON 193 C211 60 194 B034 F-19 195 A005 196 B076 J-19 197 A026 73 L-19 198 B008 N-19 199 A028 62 R-19 200 B039 201 C204 63 T-19 202 C036 V-19 203 C013 T-20 204 C033 S-20 205 C109 R-20 206 C111 65 N-20 207 B023 L-20

'08 C112 64 J-20 209 C107 G-20 210 C039 F-20 211 C024 E-20

'212 C022 H 21 213 C025 K-21 214 C028 H-21 215 C004 P-21 216 B073 D-15 217 BO42 V-7

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Page 16 TABLE 2 0-2 FIRST CYCLE CORE DESIGN CHARACTERISTICS Nuclear Characteristics Fuel Management 3-Batch, Mixed Central Zone Average First Cycle Burnup, MWd/MTU 15,400 U-235 Enrichment, w/o Batch A (69 assemblies) l. 93 Batch B (80 assemblies) 2.33 Batch C (68 assemblies) 2.82 Mechanical Characteristics Fuel Assemblies No. of Fuel Rods Poison Rods Poison Rods Batch Assemblies .I No./Ass No./Batch A 69 176 80 164 12 960 40 176 C-(Low Concen-tration B4C loading) 12 164 12 144 C+(High Con-centration B4C loading) 16 164 12 192 217 1296 Fuel Rod Array, Square 14 x14 Fuel Rod Pitch, Inches 0.580 Spacer Grid Type Leaf Spring Material Zircaloy-4 Number per Assembly 8 Retention Grid Type Leaf Spring Material Inconel Number per Assembly 1

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Page 17 Wei ht of Contained Uraniun k U Batch A 395 Batch B 368 Batch C ( oisoned) 368 Batch C un oisoned) 395 Outside Dimensions Fuel Rod to Fuel Rod, Inches 7.980 x 7.980 Fuel Rod Fuel Material (Sintered Pellets) U02 BATCH A BATCH B UO BATCH C U02 Pellet Diameter, inches .3 5 .3 5 Pellet Dish Depth, inches 0. 029 0.015 0.029 Pellet Dish Diameter, Inches 0. 2725 .2 15 Pellet Len th, inches 0 5 Pellet Density, g cc 10. 1 1 ~ 93 Pellet Theoretical Density, g cc 10.9 1 ~

e et Densit . t eoretica 5.

Stack Height Density, g cc 10.0 1 ~

Clad Material Zircaloy-Glad JD, inches Clad OD, nominal inches 0. 0 0.

a T ic ness, nom na inc es Diametral Cap, cold, nominal , inches Active Length, inc es 13 13 1 ~

Plenum Len t inches Burnable Poison Rod Active Len th, inches 122.7 Material B C-Al 0 Pellet Diameter. inches 0.376 Clad Material Zircalo -4 Clad ID inches 0.388 Clad OD inches 0.440 Clad Thickness, (nominal) inches 0.026 Diametral Ga cold nominal inches 0.012 Control Element Assembl CEA) Full Len th Part Len th um er Number of Absorber Elements per Assembly Type Cylindrical Rods Cylindrical Rods Clad Material . Inconel 25 Inconel 25 Clad Thickness, inches 0.040 0.0 0 Clad OD, inches 0. 948 0. 948 Poison Material Total Element Len th 161 161 Poison material is primarily B4C-A1203 Several CEA's finger's have a combination of A1203'and B4C-A1203 i I

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Page 18 Control Element Drive Mechanisms(CEDM Sin le CEA Dual CEA PLCEA Number of drive mechanism 49 '12 Stroke, inches 137 137 137 S eed inches er minute 40 20 30 Dro time seconds 90% .insertion 2.5 2.5 N.A.

Core Arran ement Number of Fuel Assemblies in Core Total 217 Number of CEA's 81 Number of Fuel Rods 36 896 Number of Poison Rods 1 296 CEA Pitch Min. inches 11.57 Spacing between Fuel Assemblies, Fuel Rod Surface to Surface inches 0.200 Spacing, Outer Fuel Rod Surface to Core Shroud inches 0.204 H draulic Diameter, Nominal Channel, feet 0.0444

'otal'Flow Area (excluding Guide Tubes)

'S ; Ft.) 53.5 Total Core Area, s . ft. 101.1 Core E uivalent Diameter inches 136 Core Circumscribed diameter inches 142.5 Core Vol'ume liters 32 600 Total Fuel Loadin k U 82 850 Total Fuel Wei ht ounds UO 207 200 Total Wei ht of Zircalo Clad ounds 44 700 Tot'al Heat Transfer Area, s . ft. 48,420 Fuel .Volume (including pellet dished ends), cu. ft. 330. 2

Page Florida Power 5 Light Company .

REACTOR FUEL LOCATION St. Lucie Plant Unit No. 1 FIGURE g. P-'I LOADING CE UENCE 'ORE Y X W.V T S R P NMLKJ HG F E D C B A g gi 215 214 213 212 203'04 205 206 207 208 209 210 211 202 201 200 199 198 197 196 195 194 193 192 179 180 181 182 183 184 185 186 187 188 189 190 191 47 46 45 44 43 48 65 ,83 101 102 118 119 136 15 18 19 20 2I 42 49 67 82 85 100 103 117 120 135 14 ~

16 15 141 22 41 50 63 68 81 86 99 104 216 121 134 13 17 137 12

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-z g 166 167 168 169 170 171 172 173 174 178 177 176 175

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Page 20 Florida Power 5 Light Company REACTOR FUEl LOCATION St. Lucie Plant Unit No. I FIGURE 2. 0-2.

FUEL ASSEMBLY SERIAL. NUMBERS AND CORE LOCATIONS Y X It. V T S R P NML K J HG F E 0 C B A I 61 C004 C028 C025 C02" 013 C033 C109 C111 B023 C11 C10 C039 c024 C036 C204 BO39 A028 BOO A026 B07 AOO B034 C211 C038 C034 C205 B04 A020 B078 A009 B006 A00 B01 A021 B019 C201 C020 gn 001 C208 8037 A019 B071 A031 B077 A043 B06 A04 B063 A062 B021 C206 C009 16 016 B020 A015 B069 A053 B007 A060 B044 A06 B03 069 B016 017 062 CO37 14C023 106 A004 B068 A048 B035 A037 B065 A068 B02 A05 B015 A036 B073 A001 C113 C006 g

j2 105 B067 A038 B010 A061 B004 A057 B054 A045 B03 066 B003 A007 B075 C101 I," C032 C031 B009 A025 B017 049 B026 A030 051 032 B055 A052 B011 A012 048 C026 023 B050 C018 9

I97 C007 C104 B057 A010 BOO 055 B038 A034 025 A050 B070 A067 B032 046 022 C115 C010 C114 A002 B042 AOO B045 A059 B056 041 B058 A039 B072 A047 014 003 C108 C027 B080 A013 B01 054 B040 A064 053 A058 B005 A065 B002 029 027 C017 6 C008 C212 B046 Aol 013 A056 B064 044 B059 A035 B052 A027 024 209 C003 C012 C202 B04 014 B041 A040 036 AOll B060 A016 B030 203 019 C015 C21 074 A022 B029 024 B047 A033 B066 C207 029 C002 C035 C102 C116 079 Clio C103 C040 Col4 0021 COll C030 C005

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Florida Power 5 Light Company Page 21 REACTOR FUEL LOCATION St. Lucie Plant Unit No. I Figure: 2.0-3 <

CONTROL ELE1ENT ASSEHBLY SERIAL NUHBERS AND CORE 'LOCATIONS YX It.VTSRPNMLKJHG'FE DC BA 65 64, 63 62 61 60 55'4 69 59 50 58 57 53 56 68 52 47 46 45 43 40 39 38 37 36 35 72 34 33 32 B 71 31 30 29 28 27 26 25 24 23 22 21, 20 .19 18 17 16 15 14 13 A 12 67 10 66 70 Normal CEA orientation:: Serial 8 on NW web Exception: CEA's 70, 73, F and G will have their serial 8 on SW web

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Page 3.0 POST CORE HOT FUNCTIONAL TESTS (PCHF Several of the tests required prior to initial criticality require installation of the fuel and all reactor internals as a prerequi-site. These tests (Post Core Hot Functional Tests) are conducted after initial fuel loading and complete the prerequisites for ini-tial criticality. A list of the required tests follows:

(1) Mechanical and instrument tests on Control Element Drive Mechanisms (CEDM) and Control Element Assembly (CEA) position indicators. This includes rod drop time measurements (cold and hot) under various Reactor Coolant System Flow conditions; (2) Reactor protective trip circuit and manual scram tests; (3) Final leak tests of RCS; (4) Primary and Secondary Water, Chemistry; (5) Neutron response check of source range monitors; (6) Mechanical and electrical tests of incore instrumentation, in-cluding incore detector resistance readings and testing of moveable incore instruments; (7) RCS flow determination tests; (8) Pressurizer effectiveness tests; and (9) Vibration monitoring per Regulatory Guide 1.20 The above tests except Items (2), (5), (8) and (9) are described in Sections 3.1 through 3.5. The instrumentation portions of Item (2)

Reactor trip circuit. were performed as a~rerequisite of the Initial Core Loading procedure and actual CEA trips and manual scram tests were included in the CEDM testing of Item (1). In addition, this testing was repeated as a prerequisite to the Initial Approach to Criticality (IAC). Item (5) Neutron Response Check of Source Range Monitors was performed as a prerequisite to Core Loading and function-ally. checked during IAC. Item (8) Pressurizer Effectiveness is discussed with the Pressurizer Controls Functional Test, Item 3, listed below. Item 9 Vibration Monitoring for this plant was comp-leted during pre-core load hot functional testing. We are-taking internal vibration measurements with installed equipment, but these are not the same type of measurements as discussed in Regulatory Guide 1.20.

In addition, those items or systems which required maintenance or had testing deferred from pre-core hot functional tests were tested during PCHF. This included:

(1) Taking measurements of RCS expansion;

Page 23 3.0 POST CORE HOT FUNCTIONAL TESTS (PCHF) (Cont.)

(2) RCS and Steam Generator instrument calibration checks; (3) Pressurizer Controls functional test; (4) Checking Main Generator air flow and (5) RCS heat loss test.

The above tests are described in Section 3.6 through 3.10.

All test results met their acceptance criteria with the exception of RCS flow rate and the performance of CEDM 44. After considera-tion of instrument and measurement errors and uncertainties, RCS flow rate was slightly (5%) below that required by Technical Speci-fications. An Interim Operating License change allowed opera-tion up to 60% rated thermal power. Further analysis resulted in an Operating License change allowing operation up to 90% rated thermal power. Plow was later evaluated using a plant power calori-metric approach and this indicated RCS flow is, in fact, above the required value. As of December, 1976, this subject has not been fully resolved.

CEDM 44 is a shutdown rod in Group A. Initial performance testing revealed that this CEA could not be withdrawn at cold conditions.

Further testing demonstrated that this CEA functioned satisfactor-ily at hot (operating) conditions. A Technical Specification change deleting permission to go criticalbelow 515 F RCS temperature for testing was approve'd, allowing criticality and further testing. Con-siderable extra testing was done and evaluated. Since all testing indicated proper performance at elevated temperatures (> 400 F) and the CEDM never failed to trip, FPGL did not feel the CEDM required replacement. However, in compliance with License Amendment 4, this CEDM was replaced during the shutdown for resolution of the power distribution anomaly (See Section 7.0). It has beenretested satisfactorily as applicable during startup and return to the Power Ascension Test Proj"am. A requcot or"license amendment will be subm'tted requesting reapproval of the mception allowing criticality below 515o>, for tooting.

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Page 24 3.1 CEDM/CEA Performance Tests 3.1.1 Purpose The Control Element Drive Mechanism/Control Element Assembly (CEDM/CEA) performance tests were performed to accomplish the following objectives:

(1) To demonstrate proper functioning of the CEA's and CEDM's under various Reactor Coolant System (RCS) tempera-ture, pressure and.flow conditions.

(2) To provide measured CEA withdrawl, insertion and drop time data which will serve as comparison standards for future performance tests.

(3) To perform a check of the position indication system and to establish proper functioning of the CEA op-erating and interlock lights.

(4) To verify proper operation of the upper gripper coil power supply paralleling switches.

(5) Verification that PLCEA's.(part..length CEA's) do not drop.

3.1.2 Test Results The CEDM/CEA performance tests were con-ducted at RCS temperatures and pressures of 260 F/475 psia and 532 F/2250 psia. These tests consisted of:

(1) Verification that all single, dual, and non-tripping coil current traces indicate proper operation.

(2) A check of CEDM/CEA withdrawl speed.

(3) Timing of rod drops from full out t'o 90% insertion of all regulating and shutdown CEA's under various RCS flow conditions.

(4) Additonal drops of the fastest and slowest CEA's.

(5) A check of all CEDM/CEA position indication, operating and interlock lights.

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Page 25 3.1 CEDM/CEA Performance Tests (Cont'd) 3.1.2 Test Results (Cont'd)

(6) A check of the upper gripper coil par-alleling switch circuit for regulating and shutdown CEA's.

CEDM witfidrawal and insertion traces were analyzed and adjustments were made to Coil Power Programmers during the test to ensure acceptable operation of the CEDM's. Discrepancies in rod position indica-tion were corrected by adjusting computer setpoints and progr'amming and replacing defective reed switch stacks.

The results of the CEDM drop time test were found to be acceptable and are given in Table 3.1-1.

3.1.3 Conclusions The results of the CEA/CEDM test were acceptable with the exception of CEDM 44 during cold testing.

CEDM 44 had an apparent defective lower gripper coil or latch and would not work consistently during cold testing. CEDM 44 was replaced and testing completed with all results evaluated to be acceptable. (See Section 7). This testing demonstrated proper operation of the CEDM's and verified they would perform their control and shutdown"functions.

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TABLE 3.1-1 Page 26 ROD DROP TIME RESULTS 260 F/475 psia Drop Time to 90% Full Insertion 3 um flow 2 um flow 0 um flow Fastest 81 46 2.08 sec. 1.90 sec. 2.00 sec.

a2 8 2.15 sec. 1.96 sec. 1.90 sec.

Slowest 81 51 2.36 sec. 2.24 sec. 2.17 sec.

82 69 2.44 sec. 2.35 sec. 2.24 sec. 532 F/2250 psia Drop Time to 90% Full Insertion (Full Flow) -Specification <3.3 Seconds Fastest CEA 81 46 2.02 sec.

a2 9 2.08 sec.

Slowest CEA 81 6 2.36 sec.

a2 69 2.36 sec.

Drop Time to 90% Full Insertion (Zero Flow)

Fastest CEA 81 46 1.71 sec.

9 1.74 sec.

Slowest CEA '8l 61 2.07 sec.

82 69 2.07 sec.

Page 27 3.2 Reactor Coolant'S stem Pum Flow and Coastdown Test 3.2.1 ~Pur ose This test was conducted to determine the following Reactor Coolant System (RCS) characteristics:

(1) Reactor Coolant System flow rates and pressure drops around the reactor coolant system.

(2) Reactor Coolant Pump Coastdown flow characteristics.

(3) Reactor Coolant System flow input to Reactor Protec-tion System (RPS) Low Flow trip unit noise character-istics.

(4) Pressurizer Spray characteristics for various 2, 3, and 4 pump combinations.

3.2.2 Test Results RCS flow measurements were taken at a RCS temperature and pressure of 532F and 2250 psia respectively using installed and temporary instrumentation. All instruments were cali-bration checked or calibrated at least twice, once before the test and once after. Measurements were taken for RCP combinations as listed in Table 3.2-1. All data collected was corrected for pressure and temperature and zero flow conditions. The flow coastdown portion of the test con-sisted of monitoring pump D/P, Vessel D/P 'and Steam generator D/P. Coastdowns were coordinated by turning off the common timing pulse for a brief interval and restarting it shortly before the pump trip. Recorder traces of RCS flow versus time after pump trips were made for each RCP combination listed in Table 3.2-1 as indicated under transient data.

Figure 3.2-1 shows a typical flow coastdown curve computed from RPS flow innut.

Spray Data was taken as pei Table 3.2-2. Manual control was, taken of pressurizer spray valves so as to demand a full open signal. Flow through one or the other spray line was verified by decreased pressurizer pressure and increased soray.line temperatures. Spray lines are equipped with check valves which prevent reverse flow from the pressurizer to the RCS loops. Loop temperature increases which would indicate reverse flow were not observed. Table 3.2-2 and Figure 3.2-2 are the results of the pressurizer spray test.

3.2.3 Conclusion Evaluation of test data indicated that all test data were satisfactorv although the resultant flow was about 5% less than that necessary to guarantee the 370,000 gpm required

Page. 28 3.2 Reactor Coolant S stem, Pum Plow and'Coastdown'TestContinued 3.2.3 Conclusion Continued by the Technical Specifications. An interim License amend-ment was issued limiting operation to 90% power based on this apparent low flow. 'he safety analysis was modified to cover a minimum acceptable,RCS flow rate of 354,000 gpm nendina final verification of flow through calorimetric determination. A calorimetric RCS flow determination was performed at 80% power. The flow determination produced a measured RCS flow of 399,000 gpm corresponding to 123.1+

4.6% of design flow. This flow rate is above the minimum technical specification value of 113.9% design flow even with allowance for measurement uncertainties. This data, with supporting information, has been submitted to the NRC with a request for license amendment allowing operation at 100% power.

It should be noted 'that, although RCS low flow trip set-points are derived from this lQ? instrumentation, they are based on a specified relative change (decrease) in the output signal of the instruments and thus are independent of RCS flow measurements. See Section 7.3 for further dis-cussion.

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Page 29 TABLE 3.2-1 REACTOR COOLANT PRP FLOW AND COASTDOWN'TEST"SE UENCE RCPs STATIC NOISE DATA' ' 'ATA'RANSIENT SPRAY DATA RUNNING DATA lA1 1A2 1B1 1B2 X NONE X

'X'

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lAl lA2 1Bl X' lAl lA2 1B1 1B2 1A2 1B1 1B2 1Bl 1B2 X 131'A1 131 'X

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Page 32 TABLE 3.2-2 SPRAY TEST RESULTS F SPRAY LINE TEMPERATURES MAXIMUM RATES OVERALL RATE INITIAL FINAL OPERATING PUMPS PSI/MINUTE PSI/MINUTE 1Bl 1B2 1Bl 1B2 4

Pumps [1A1 1A2 1B1 1B2 225 156 480 520 510 530 1A1 1A2 1Bl 160 87 487 500 , 520 520 3 1A2 1B1 1B2 200 120 475 485 510 510 Pumps 1Bl 1B2 340 174 425 405 470 470 lA1 1A2 1B2 150 62 445 460 440 510 lA1 lA2 none none No change 2 1Bl none none No change Pumps 1B2 47 23 506 495 525 495, 1A2 1B1 none none No change 1A2 1B2 none none No change 1B1 1B2 300 81 475 480 500 500

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Page 33 3.3 RCS Leak Tests 3.3.1 Purpose A leak test of the Reactor Coolant System (RCS) was performed to check for indications of abnormal leakage from the primary system and a leak rate test was performed to demonstrate sen-sitivity of the water inventory balance procedure.

3.3.2 Test Results Leak Test The leak test of the RCS was performed at 2330 (+ 20) PSIA and consisted of a visual inspection of the entire RCS, with par-ticular emphasis on the following areas:

(1) Re'actor Coolant Pump (RCP) Seal Area (2) Reactor Vessel Head Seal (3) Steam Generator Manways (4) Control Element Drive Mechanisms (CEDM's)

(5) In-Core Instrument Penetrations (6) Valve stem packing and body to bonnet joints (7) Pressurizer heater penetrations The inspection of these areas showed no abnormal leakage and the test was considered satisfactory.

3.3.3 Test Results Leak Rate Test A water inventory balance (over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) was performed to deter-mine RCS leak rate. Immediately afterward, a,l gpm known leak was initiated through the sample system and another 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. inventory balance test run. Close agreement (within 1/10 gpm) between the two (disregarding the lgpm known leak) indicated the accuracy of the method and that this method will detect leaks in the range of cgpm or less.

3.3.4 Conclusions The RCS leak test showed that the primary system was tight after reactor vessel reassembly following fuel load and no abnormal leakage should be expected. The RCS leak rate test demonstrated the water inventory balance method works satisfactorily and will detect leaks in the range of 4gpm or less.

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Page 34 3.4 Prima and Seconda Water Chemistr 3.4.1 Purpose To establish, monitor, and control primary and secondary water chemistry during plant heatup and conduct of Post Core Hot Functional (PCHF) Tests using normal chemistry operating pro-cedures. Baseline data to support the Low Power Physics and Escalation to Power Test phases was obtained. Also, Baseline Corrosion data was obtained and a chemical shock treatment was performed to aid in evaluating apparently low RCS flow.

3.4.2 Test Results 3.4.2.1. Chemistry Control All primary and secondary water chemistry results during PCHF were either acceptable or when acceptance criteria were not met, corrective action was insti-tuted to achieve acceptable conditions.

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3. 4. 2. 2 Baseline Corrosion.

(1) Suspended Solids (S.S.) in the RCS were generally

<.01 ppm. The highest S.S. value observed was 0.25 ppm after a heatup from 260 to 530 F.

(2) A five gallon filter sample of reactor coolant was analyzed by Z-ray diffraction. The results are as follows: (Relative %)

Top Filter Bottom Filter Element Oxide Element Oxide Silica 44.2 50.'1 15.7 17.7 Aluminum 21.2 21.1 67.7 67.3 Phosphorus 2.4 2.9 ND Sulfur 3.6 4.7 4.7 6.2 Chlorine 2. 9 1.5 ND Potassium 1.0 0. 65 ND Calcium 0.2, 0.13 ND Chromium ND Titanium 1.7 1.5 ND Iron 20.6 15.6 ll.8 8.9 Nickel 0.5 0.34 ND Molybdenum 1.7 1.4 ND No foreign materials (organics) were indicated.

(3) During PCHE the Steam generators were intermittently fed with deaerated water from the Condensate Storage Tank (CST) and chemicals were injected manually to control pH and a hydrazine residual. The main feed and condensate systems were not operated. Consequent-ly there were no S.S. problems. S. S. values were generally in the 0.1-0.2 ppm range. Highest value of S. S. observed'was 0.5 ppm. Blowdown was run at maximum permissible rate dependent upon makeup water supply.

II Page 35 3.4.2.3 Chemical Shock Treatment This evolution was performed to aid in evaluation of possible causes for the lower than expected RCS flow.

1) Prior to shock test: S.S. were generally

<.Ol ppm. On 4-15-76 8 1700 the S.S. value was 0.25 ppm resulting from heatup from 260 F to 530 F.

2) Chemical Addition: On 4-16-76 8 1340 six (6) gallons of Hydrazine were added yielding 40 ppm N H4, and 1627 grams of LiOH was added in-creasing Lithium from 0.8 to 2.0 ppm. An in-crease in pH from 6.05 to 6.82 was observed.

3) Resulting S.S.: Ten minutes after chemical addition S.S. were 0.24 ppm. This result is believed to be erroneous. It appeared that part of the bottom filter had stuck to the drying tray thereby increasing the relative weight of the top filter. Visual inspection of the top filter indicated no discoloration and under magnification there appeared to be no significant particles, About three hours after chemical addition a four (4) liter filter showed 0.03 ppm.

About twenty-two (22) hours after chemical addition a 5 gallon filter showed 0.038 ppm.

4) It was concluded that the shock test pro-duced no significant crud burst.

3.4.3 Conclusions Overall the data collected indicated that good chemistry control was maintained even with the variable plant conditions required by the test program. During baseline testing for Reactor Coolant System (RCS) particulate level and soluble'orrosion product's, no unusual or unexpected results were obtained. The lack of a crud burst from the chemical shock test implied that RCS pre-conditioning had been effective and demonstrated that there was no excessive buildup of crud (corrosion products) within the RCS.

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Page 36 3.5 Incore Instrumentation Functional Tests 3.5.1 Purpose These tests were performed to:

(1) Obtain fixed incore thermocouple data (2) Measure fixed incore detector leakage resistance (3) Verify proper installation of the moveable incore detector transfer and drive machines and detector thimbles (4) Verify proper operation of the moveable incore detector drive system (5) Obtain path length measurements for use during future operation and (6) Verify proper operation, of incore detector drive system when controlled by the Digital Data Processing System (DDpS).

These tests will verify the system is ready for actual use for operations during LPPT and power ascension.

3.5.2 Test Results 3.5.2.1 Fixed Detectors The thermocouple data was compared to other plant instrumentation and was satisfactory.

The minimum leakage resistance of the fixed incore detectors was a factor of 2 higher than required. One detector string cable connector of 45 was faulty. In addition, one detector string was apparently damaged during installation and never functioned during initial startup and ascent to power. It has since been replaced.

3.5.2.2 Moveable Detectors Proper operation of the system was verified by performing all evolutions the system wa's cap-able of, including measuring full path lengths and the satisfactory completion of this not only verifies proper operation but also proper in-stallation.

NOTE: The thermocouples, by design, are encapsuled and isolated from RCS flow. Therefore, at power they are subject to preferential gamma

'heating and are not an accurate representation of RCS coolant temperatures.

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Page 37 3.5 Incore Instrumentation Functional Tests (cont.)

3.5.2.2 (cont.)

DDPS control of the moveable incore drive system was demonstrated by traversing all paths and ordering the DDPS to attempt to traverse impossible paths. The system functioned properly, including the printout of evolution in progress and "error" messages when improper commands were given.

3.5.3 Conclusions These tests demonstrated that the thermocouples were satisfactory for indication of actual incore temperature at zero powei and that the leakage resisteance of the fixed in-core detectors was well above specification and should present no operational problems due to shorting out.

Also verified was the ability of the drive system to properly move the detectors to the various required locations in either manual control or when automatically controlled by the DDPS. Satisfactory completion of all these evaluations, including verifying all fixed detectors to have leakage resistance greater than 1 x 107 ohms, demonstrated the incore instrumentation to be ready for use during Low Power Physics Testing and Power Ascension.

I Page 38 3.6 Reactor Coolant S stem Pi in Thermal E ansion and Restraint 3.6.1 Purpose This test was repeated during PCHF as many restraints and snubbers had been added or modified since original performance of the test. The purposes were to verify that:

(1) piping systems were free to expand ther-mally as designed; (2) spring hangers were not bottomed out or unloaded; (3) hydraulic snubbers and pipe rupture re-straints did not unduely restrict thermal movements; (4) mechanical snubbers did not lock-up; and (5) vent and drain lines did not vibrate ex-cessively.

3.6.2 Test Results The components described in Items 1 through 4 were ob-served and/or measurements taken by the Contractor't ambient, 260 F, 360 F, 470 F and 532 F and after cool-down to ambient. Also, mechanical snubbers were checked during the heatup to those temperatures. Vent and drain lines were checked during system operation to verify the (new) restraints prevented excessive vibration. Early in the heatup FP&L observers noted that 5 mechanical assem-blies were locked up. (Licensee Event Report 335-76-9, 4/12/76). This prompted addition of Item 4 to the offi-cial test program. The cause was confirmed to be corro-sion and all mechanical snubbers were replaced with snub-bers of different design from a different vendor which are not susceptible to this problem. (Follow-up LER 335-76-9, 6/10/76). The locked snubbers were the only problem of any significance noted during this test.

3.6.3 Conclusions This test verified that the piping restraint systems met all acceptance criteria. That is, it demonstrated that: piping was restrained only by the installed re-straints, snubbers or hangers; that insulation was not deformed excessively by interferences; that piping dis-placements were as predicted by design (within tolerances) and that mechanical snubbers -remained free and moved properly during heatup. These results verified that piping is properly restrained for system heatup and op-eration but is not over-restrained which could lead to overstress and eventual'ailure.

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Page 39 3.7 Reactor Coolant System and Steam Generator Instrumentation.

Calibration Check 3.7.1 Purpose The purpose of this test was to check the calibration of the named instrumentation by correlating indications for a given Parameter (i.e. Tcold) between the Digital Data Processing System (DDPS) Reactor-Turbine Gageboards (RTGB), Reactor Protection System (RPS) and the Engineered Safeguards System (ESF).

3.7.2 Test Results Readings were taken from the locations listed above for certain parameters at 7 different plant conditions, from ambient temperature with the RCS vented to normal operating temperature and pressure. The parameters taken were:

(1) RCS Tcold hot (3) RCS pressure (4) RCS flow 1 (5) Charging flow (6) Letdown flow (7) Letdown temperature (8) Containment Pressure (9) Steam Generator Pressure (10) Steam Generator Level Although all these parameters are not monitored at all the locations in 3.7.1, all are monitored at two or more of those locations. Data was evaluated throughcut .the test to ensure acceptable results.

3.7.3 Conclusions The acceptance criteria for this test were met. All the safety channel correlations agreed within instrument accuracy tolerances. This gives assurance that the instrumentation provides accurate signals to control and protection systems and gives proper indications to guide the, operators.

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Page 40 3.8 Pressurizer Controls Functional Test 3;8 1 . Purpose The objectives of this test were as follows:

(1) Verify the operation of the power operated relief valves at 2400 +0 -25PSIA (simulated pressure signal) and demon-strate that the quench tank would condense and cool the discharge of these valves.

(2) Establish proper settings for the continuous spray valves.

(3) Verify the operation of the spray valves at design pres-sures with the RCS at 532 +2oF and 4 RCP's operating.

(4) Verify that system pressure can be reduced at the design rate using spray flow. Rate of increase previously had been verified and is part of 5 below also, (5) Verify design alarm and control settings and that the level and pressure controls maintain the pressurizer within desired limits.

(6) Demonstrate cooldown of the pressurizer using auxiliary spray.

Test Results All of the above were satisfactorily completed. The rate of pressure decrease using normal spray was 80 psi/minute which exceeded the minimum required value of 65 psi/minute. All valves and controls (spray and power operated reliefs) operated properly.

And, the quench tank accepted the power operated relief valve discharge for 30 seconds without o~ver ressurization or rupture disc problem. This discharge was the first opportunity for ob-servation of thermal expansion for the involved lines and tank and this was completed satisfactorily.

3.8.3 Conclusions Item 2~ setting of continuous spray valves, was implicitly included in Item 6, verification of control within design

. 1Mits, as well as documented complete within the body of the procedure. Items 1 and 3 through 6 were acceptance criteria for the test and were completed satisfactorily. Also, note that additional spray valve testing was performed in conjunction with RCS flow testing (Section 3.2)

II Page 41

3. 9 Hain Generator Air Flow Test 3.9.1 Purpose The purpose of this procedure was to collect data for the generator vendor to use to verify that design changes in the generator ventilation (cooling) system would prevent an unstable aerodynamic condition (blower surging) during operation.

3.9.2 Test Results Air pressure, air temperature and barometric data were taken at various locations in the generator at two different values of generator revolutions per minute.

(25% and 38% of full speed). Steam generated by the Reactor Coolant Pump heat was used 'to roll the turbine generator. The data was extrapolated to full speed.

Comparison with the expected values'(acceptance criteria) verified that the ventilation (cooling) system would function properly.

3.9.3 Conclusions This test verified the design of the ventilation system and demonstrated that the system would provide proper cooling gas flow without aerodynamic instabilities such as blower surging or other flow related problems.

II Page 42

3. 10 Reactor Coolant System Heat Loss Measurements 3.10.1 Purpose The purpose of this test was [to determine the RCS (including pressurizer) heat loss to the environment (containment). This number is an input'o the calorimetric calculation which determines reactor thermal power. Also, it provides a check of the efficiency of the thermal insulation installed on the system.

3.10.2 Test Results The pressurizer heat loss was determined by establishing very stable RCS conditions, isolating pressurizer spray, and measuring the power input to the pressurizer heaters and the time intervals during which they were energized over a given period of time. Initial and final data were taken to verify pressurizer operating conditions remained stable. The calculated energy input was then equal to the heat loss. This value was then used as part of the data to determine RCS heat loss.. This value was some-what higher than expected but did not pose a safety or operational concern.

RCS heat loss was determined by establishing stable RCS conditions with a known steam generator level (mass) and isolating feedwater flow. RCS temperature was maintained steady by dumping steam through the Atmospheric Steam Dumps until steam generator level (mass) had been reduced to a predetermined level. Knowing the mass and enthalpy of the steam and having collected data on all other sources of energy to and from the RCS, including pressurizer heat loss as determined above, it was possible to calculate the total RCS heat loss. This steam down test was performed twice.

In addition, two feed up tests were performed. The only differences were that no steam was removed and feedwater was added to control RCS temperature and increase steam generator level (mass) a given amount. It proved impossible to regulate feedwater precisely enough to maintain the very tight RCS temperature band (normal Tco]d,+ 1 F) required for the heat loss calculation. However, the calculated results were of the same order of magnitude and provided assurance that the steamflow heat loss determination method was correct.

Page 43 3.10 Reactor Coolant System Heat Loss Measurements (cont.)

3.10.3 Conclusions The results of the 2 steam down tests agreed very well with each other, with the vendor's predicted results and with results from similar plants. This assured that the RCS heat loss value was proper for inclusion in the thermal power calorimetric calculation. It also confirmed that the design and installation of the RCS thermal insulation were satisfactory.

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Page 44 4 ' INITIAL APPROACH TO CRITICALITY Initial criticality was achieved on 4-22-76 at Reactor Coolant System (RCS) conditions of 532 F and 2250 psia. The initial RCS boron concentration was 1729 ppm. The Initial Approach to Criti-cality (IAC) began by withdrawing the CEA's in specified incre-ments with count rate data taken after each increment. During this withdrawal CEA Group control and group interlocks were verified to be functioning properly. Criticality was subsequently achieved by deborating the RCS to a boron concentration of 935 ppm.

Throughout the approach to criticality, two (2) independent sets of inverse multiplication plots were maintained. Two plots of inverse count rate versus RCS boron concentration were maintained during the dilution phase. Periodically, count rates were obtained from each Wide Range Log Channel (WRLC). The ratio of initial average count rate to the count rate at the end of each time increment was the value plotted.

The CEA withdrawal sequence and intervals are shown in Table 4.0-1.

The inverse count rate versus CEA position points for two WRLC are shown in Figures 4.0-1 and 4.0-2. The inverse count rate versus RCS dilution time in'ours is shown in Figures 4.0-3 and 4.0-4.

After achieving initial criticality, Control Element Assembly (CEA)

Group was used to control neutron flux. Gonddtdons were stebdldzed at 10- ( ~ power and the critical data shown'n Table 4.0-3 was recorded and compared with predicted values.

In summary, initial criticality was achieved in a safe and orderly fashion. There was good agreement between the measured and predicted critical boron concentrations.

Page 45 TABLE 4.0-1 CEA WITHDRAWAL SEQUENCE STEP

• CEA GROUP INCHES WITHDRAWN B 1/M D 1/M 12.8.l.a 68 1.019 l. 00

12. 8.l.b A 136 l. 022 0.971 12.8.2.a 68 1.055 1.044 12.8.2.b 136 1.118 0.952 12.8.3 > 132 1.085 0.981 12.9.1 68 1.00 0.975 12.9.2 1.036 0.931 2 54 12.9.3 122 1.081 0.968 40 12.9.4 UEL 0.821 0. 750 107 26 12.9.5 UEL 0.845 0.819 93 ll 12.9.6 UEL 0.881 0. 824 79 12.9.7 UEL 0.855 0.777 54 12.9.8 6 122 0.813 0. 726

'7 40 12.9.9 6 UEL 0.852 0.786 7 68 12.9.10 UEL 0.833 0.750 12.9.12 68 0.790 0.763

• Steps from TAC Procedure, 83200086 NOTE: UEL = Upper Electrical (position) Limit

D Page 46 TABLE 4.02 DILUTION TIME DILUTION TIME RCS BORON (Minutes) (Houurs) CLOCK TIME CHEM. ANAL. B 1/M D 1/M 30 60 90 2130 1528 0.816 0.887 2200 1459 0.795 15 2230 1389 0.692 0.726 180 210 2 0 240 4-22-760000 1220 0. 593 0.504 250 0010 0. 50 2 260 0020 0. 16 270 0030 280 0040 0.46 .4 8 290 0050 0.471 0 496'4 300 0100 1134 0.408 310 0110 0.466 0.444 320 0120

'30 0130 1112 0. 470 0.452 340 350 0150 360 02 390 420 0 0 450 480 04 510 0 30 540

'70 0530 0600 1039 1017 0.339 0.315 0.331 0.308 600 10 630 0630 1015 0.312 0.294 660 0700 998 0. 268 0.282 670 0710 0. 260 0.252 680 0720 24 0.229 0730 0.235 0.221 700 0740 974 0.201 0.197 710 0750 0.168 0.174 12 0800 960 0.146 << 0.136 0810 0.079 0. 079 0820 944 0.018 0.019 0830 5

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Page 47 TABLE 4.0-3 PAEUQKTER INITIAL MEASURED CONDITION VALUE RCS TEMPERATURE 532'F 532oF RCS PRESSURE '2250 PSIA 2250 PSIA RCP'S OPERATING CEA GROUPS WITHDRAWN, IN INCHES I

A UEL UEL UEL UEL.

P-I UEL UEL P-2 UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL 5 UEL UEL UEL 68 68 ~

PREDICTED MEASURED RCS BORON CONCENTRATION (PPM) 936 935

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Page 48 ST. LUCIE UNIT 1 INITIAL APPROACH TO CRITICALITY BOL, 1st CYCLE, 532oF, 2250 PSIA WIDE RANGE LOG CHANNEL B 1.4 1.3 1.2 1.0 o 0.9 0

0.8 O

M 0.7 0.6 o 0.5 0.4 M

0.3 0.2 0.1 0.0 Q C4 CO Cl C4 5) ~ CV CO Ch CO CO CO CO CO Ch Ch Ch Ch Ch Ch Ch II ~

CV hl c4 c4 H cV cV Steps from IAC Procedure, f/3200086 Figure 4.0-1

Page-49 ST. LUCIE UNIT 1 INITIAL APPROACH TO CRITICALITY BOL, 1st CYCLE, 532oF, 2250 PSIA WIDE RANGE LOG CHANNEL D 1.4 1.3 1.2 1.0 CD 0.9 o

H 0.8 0.7 0.6 O

CD ca 0.5 R

0.4 0.3 0.2 0.1 0.0 CV 0O 0O 00 OO 00 Ol CJl Ch Ch Ch CV C4 N CV N CV Steps from TAC procedure, /I3200086 Figure 4.0-2

I Page-X9 ST. LUCIE UNIT 1 INITIAL APPROACH TO CRITICALITY BOL, 1st CYCLE, 532oF, 2250 PSIA WIDE RANGE LOG CHANNEL D 1.4 1.3 1.2 1.0 CD 0.9 oM 0.8 g Oo7 0.6 8

cn 0.5 R

0.4 0.3 0.2 0.1 0.0 CO co 00 " 00 00 00 Ch Ch Ch Ch Ch M Ch Ch Ch CV CV CV N CV & & C4 CV CV CV Steps from IAC procedure, 83200086 Figure 4.0-2

II Page 50 ST.. LUCIE UNIT 1 INITIAL APPROACH TO CRITICALITY BOL y 15 't CYCLE y 532 F y 2250 PS IA WIDE RANGE LOG CHANNEL B ZERO POWER; CEA GROUP 7 AT 68 INCHES 1.4 1'. 3 1.2 1.0 0.9 CD 0

CD 0.8 0.7 0.6 O

0.5 CC 0.4 H

0.3 0.2 0.1 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 DILUTION TIME (HOURS)

Figure 4 .0-3

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page 51 ST. LUCIE UNIT 1 INITIAL APPROACH TO CRITICALITY BOL, 1st CYCLE, 532oF, 2250 PSIA WIDE RANGE LOG CHANNEL D ZERO POWER, CEA GROUP 7 AT 68 INCHES 1.4 1.3 1.2 1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 DILUTION TItfE (HOURS)

FIGURE 4.0-4

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Page 52 ST. LUCIE UNIT 1 BOL 1st CYCLE REACTOR COOLANT SYSTEM BORON CONCENTRATION VS DILUTION TIME 1800 1700 1600 1500 R

O H

1400 CD CD 1300 O

CC O

CD 1200 1100 1000 ESTIMATED CRITICAL 936 BORON CONCENTRATION 900 0 1 2 3 4 5 6 7 8 9 10 11 12 DULUTION TIME (HOURS)

Figure 4.0-5

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Page 53 5.0 LOW POWER PHYSICS TESTS (LPPT)

The St. Lucie Unit 1 initial core consists of two hundred seven-teen (217) fuel assemblies each containing one hundred seventy-six (176) fuel rods/burnable poison rods and five (5) Control Element Assembly (CEA) guide tubes. Fuel assemblies are divided into three (3) distinct groups by enrichment, Type A, B and C.

Twelve (12) fuel rods in all Type B and twenty-eight (28) Type C fuel assemblies are replaced with burnable poison rods.

Table 2,0-:2 tabulates this and other important core design characteristics.

In addition to soluble boron in the Reactor Coolant System (RCS),

reactivity control is provided by eight-one (81) CEA's. CEA's are inserted into and withdrawn from the core by means of sixty-nine (69) Control Element Drive Assemblies (CEDM's). Twelve (12) CEDM's are attached to dual CEA's. Figure 5.0-1 shows the core location of the CEA's. The CEDM's are arranged into eleven (11) CEA Groups. Those Groups are further defined by function. CEA Groups A and B are Shutdown Groups. CEA Groups 1 through 7 are Regulating Groups. CEA Groups P-1 and P-2 are Power Shaping Groups. Figure 5.0-2 displays the relative core location of the CEA Groups.

CEA Group movement is restricted -as a function of power level in order to insure that CEA configurations not analyzed for in the safety analysis do not occur. The mechanism for this re-striction is a so-called Power Dependent Insertion Limit (PDIL) curve found in the Technical Specifications. Automatic con-trol features as well as operating instructions prevent inser-tion of CEA Groups into the core below this PDIL curve. The lower the reactor power, the greater the CEA insertion allowed.

LPPT consists primarily of the measurement of reactivity worths of phenomena which can vary the critical condition of the core.

To speed the collection of this data, as well as to enhance its accuracy, an analog computer which solves the kinetics equation for reactivity was used. Several techniques were used in con-

)unction with this reactivity computer to measure CEA worths.

The soluble boron swap technique consisted of a continuous or slug dilution or boration of the RCS simultaneous with small compensating reactivity changes in CEA position. The reactor was kept near critical during this evolution, and the reac-tivity computer provided a signal which could be trended and correlated with CEA position as a function of time. A CEA trip technique was also used in conjunction with the reactivity computer. The rapid change in reactivity caused by a CEA or CEA Group trip was correlated with reactivity change detected by the reactivity computer.

II Page 54 5.0 LOW POWER PHYSICS TESTS (LPPT) (Cont'd)

A11 raw test data was collected, reduced, and analyzed on site. In all cases, measured data met'pplicable acceptance criteria. CE, Windsor provided backup support for all measured data analyses.

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Page 55 FIGURE 5.0-1 ST. LUCIE UNIT 1 CORE LOCATION OF THE CEDE's BOL, 1st CYCLE (CEA LOCATIONS ARE GIVEN IN FIGURE 2.0-3) ss/

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Page 56 FIGURE 5.0-2 ST. LUCIE UNIT 1 RELATIVE CORE LOCATIONS OF THE CEA GROUPS BOL 1st CYCLE

'7 A P2 Pl 2 P2, B

B Pl Pl 7 B

A P2 Pl P2 NORTH

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Page 57 5.1 Critical Boron Concentration Measurements 5.1.1 Purpose Critical boron concentration measurements were performed at various CEA positions at RCS temperature and pressure. The pur-relatively'onstant pose of these measurements was to obtain an as-measured value for the excess reactivity loaded in the core and to provide basis for verification of predicted CEA Group reactivity worths.

5.1.2 Test Results Boron concentration values were averages of mul-tiple chemical analysis measurements made during periods of stable reactor coolant system (RCS) boron concentration. Boron end point technique was used when required. This method borates (dilutes) CEA's out near UEL+ (in-near LEL**).

After RCS conditions stabilize, and the RCS boron concentration has been chemically analyzed, the CEA's are quickly moved to UEL (to LEL),

reactivity stabilized; and CEA quickly moved back in (out) to their bite position. The re-activity change (reactivity being plotted on a recorder of a reactivity computer) is measured.

The amount of reactivity added (subtracted) is converted, via boron worth, to an equivalent PPM. .This L PPM is added to (subtracted from) the measured boron concentration. This techni-que gives a safe, fast and accurate method of determining critical boron concentrations at hard to achieve CEA positions (relatively low reactivity worths at UEL's and LEL's).

5.1.3 Conclusions Results indicate that measured boron concentration were in adequate agreement within predictions and well within the acceptance criteria of + 100 PPM.

• UEL Upper Electrical Limit

• + LEL Lower Electrical Limit

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Page 58 TABLE 5.1-1 CEA CONFIGURATION ARO 7 LEL 5 LEL 4 LEL 2 LEL B LEL MEASURED CBC (ppm) 962 906 845 746 605 523 PREDICTED CBC (ppm) 963 906 841 733 579 RCS TEMPERATURE, F 533.0 532.8 533.7 535.3 534.7 533.7 RCS PRESSURE, psia 2255 2255 2255 2256 2250 2250 CEA GROUP POSITION Pl UEL UEL UEL UEL UEL P2 UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL LEL UEL UEL UEL UEL UEL LEL UEL UEL UEL LEL LEL UEL UEL UEL UEL LEL LEL UEL UEL UEL LEL LEL LEL UEL LEL LEL LEL LEL UEL UEL LEL LEL LEL LEL 7 UEL LEL LEL LEL LEL LEL 1352 1602 2027 0818 2155 0600 DATE 4/24/76 4/24/76 4/24/76 4/25/76 4/25/76 4/27/76

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Page 59 5' Critical Boron Concentration and Soluble Boron cnorth Heasurements 5.2.1 Purpose Soluble boron in the form of dissolved boric acid in the Reactor Coolant System provides variable reactivity control over the life of a core. It can supplement the reactivity control provided by CEA Groups. However, its principal function is to com-pensate for burnup of excess reactivity as core de-pletion proceeds. The critical boron concentration for various CEA configurations were measured in order to develop a relationship for determination of the soluble boron reactivity worth. CEA Group hold down values were also measured and are 'presented in Section 5.5 5.2.2 Test Results CEA Group integral reactivity worths were measured using a soluble boron swap technique. In addition, the soluble boron concentration at the end point of several of those CEA configurations was also measured. Soluble boron samples were independently analyzed by FP&L and by CE-Mindsor. A comparison of measured boron worths for these several CEA con-figurations/ reactor coolant system (RCS) boron con-centrations is listed in Table 5.2-1.

5.2.3 Conclusions The agreement between measured and predicted critical boron concentrations and between measured and predicted soluble boron worths are adequate and well within the acceptance criterion.

II Page 60 TABLE ~5. 2~1-COLUMN 1 COL%BI 2 COLUMN 3 COLUMN 4 COLUMN 5 MEASURED CRITICAL CEA CON- CHANGE NON-OVERLAP MEASURED SOLUBLE BORON CONCENTRATION FIGURATION IN CBC CEA WORTH BORON WORTH TABLE 5.3-1 TABLE 5.5-1 COLUMN 3  : COLIRQl 4 PPM BORON APPM  % A1c/k PPM /  % Ak/1c 962 ARO 906 7 LEL 56 0. 707 79.21 845 5 LEL 61 0.781 78.10 746 4 LEL 99 78.45 605 2 LEL 141 1.814 77.73 Vg AVERAGE 78's37

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Page 61 5.3 Chemical and Radiochemical Tests 5.3.1 Purpose Chemical and radiochemical analyses of the Reactor Coolant System (RCS) and Steam Generators were performed to determine baseline corrosion data, fission and activation product levels and buildup, to detect failed fuel and various 'impurities which could enhance corrosion.

5.3.2 Test Results 5.3.2.1 Baseline Corrosion Data collection, commenced during PCHF, continued during LPPT.

(1) Suspended Solids (S.S. in the RCS were generally

<.01 ppm. The highest S. S. value observed was 0.25 ppm after a heatup from 260o to 530oF.

(2) During Low Power Physics Testing the Steam gener-ators were intermittently fed with deaerated water from the Condensate Storage Tank (CST) and chemicals were injected manually to control pH and a hydrazine residual. The main feed and condensate systems were not operated. Consequently there were no S.S.

problems. S. S. values were generally in the .

0.1-0.2 ppm range. Highest value of S. S. observed was 0.5 ppm. Blowdown was run a maximum permissible rate dependent upon makeup water supply.

5.3.2.2 Dissolved Oxygen (D.O.)

(l) Prior to heatup and LPPT, D.O. was scavenged by hydrazine and nitrogen purges of the Volume Control Tank (VCT).

(2) Prior to and during LPPT, D.O. was generally

<.005 ppm and at no time did it exceed the limit of O.l ppm when >250 F.

(3) On two occasions prior to LPPT D.O. approached the limit reaching a maximum of .07 ppm. These increases were due to large volume dilutions and borations. In both cases hydrazine was immediately added.

(4) During LPPT, 0.01=ppm was the highest D.O. detected.

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Page .62 5~3 Chemical and Radiochemical Tests (cont) 5.3.2.3 Fission and Activation Product Buildup (1) Fission Products (uci/ml) max. min.

I 133 4.4 x 10 7 2.7 x 10 7 Xe 135 9.3 x 10 2.6 x 10 (2) VCT gases detected (uci/cc) max Ar 41 1.8 x 10 1.49 x 10 Kr 88 2.42 x 10-6 7

Xe 133 9.1 x 10 Xe 135 3.6 x 10 (3) Activation Products (uci/ml)

I max. min.

Ar 41 3.5 x 10 6 7.9 x 10 F 18 3.6 x 10 5 3.0 x 10 6 Na 24 7.8 x 10 7.3 x 10 Gross. Act. 8.2 x 10 5 9.6 x 10 5.3.2.4 Lithium: Approximately one week prior to LPPT lithium was added to 0.75 ppm, then to 1.5 ppm.

Several subsequent additions were made to main-tain >1.0 ppm.

Due to numerous dilutions and borations, lithium buildup was not evident until well into the power ascension program.

5,3.2.5 Demineralizers: RCS purity and lithium were maintained by use of a 1:1 LiOH borated mixed bed resin.

A second demineralizer was loaded with HOH borated resin. This was used on one occasion during Power Ascension for lithium removal. Due to the purity of reactor coolant, D.F.'s were low and no meaning-ful data was accumulated.

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Page 63 5.3 Chemical and Radiochemical Tests (cont,)

5.3.3 Conclusions Evaluation of the test results showed no abnormal or unexpected conditions in the RCS or steam gen-erators and provided good baseline data for future use. Also, it was verified that, for the hot standby conditions of the LPPT phase, the chemistry control methods worked quite satisfactorily.

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Page 64 5.4 Tem erature Coefficient of Reactivit Measurements, 5.4.1 Purpose The moderator temperature coefficient of reactivity can be either negative or positive, depending upon the magnitude of the Reactor Coolant System boron concentration. The moderator temperature coeffi-cient cannot be measured directly but it can be derived from a measurement of the isothermal temp-erature coefficient.

5.4. 2 Test Results Isothermal temperature coefficient measurements were conducted at several different CEA with-drawal configurations and boron concentr'ations.

Measured values for each condition are the result of averaging data from several segments of the heatup and cooldown phases of the measurement.

Throughout the measurements, reactor power was maintained below the point of adding nuclear heat to minimize the confusing effect of doppler feed-back. Reactor Coolant System ramp temperature changes were affected by proper positioning of turbine bypass or atmospheric dump valves.

Table 5.4-1 summarizes the results of the measure-ments and comparisons with predicted valves. A-greement between measured and predicted values indicates acceptance criteria has been met. Tech-nical Specification 3.1.1.4a specifies that the mod-erator temperature coefficient shall less positive than +0.5X10-4 Ak/k/oF whenever power <70%.

5.4.3 Conclusions For all cases, the measured values of isothermal temperature coefficient are within the acceptance criterion of'0.5Z10-4 Ak/k/ F of the predicted value, and is therefore acceptable.

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Page 65 TABLE 5.4-1 5 LEL 2 ~ 10" 7 ~ 120" CEA CONFIGURATION MEASURED MTC, xl0 hK/K / F -0. 25 -0.79 +0. 25 ..

TECHNICAL SPECIFICATION LIMIT MTC <0.50 <0. 50 <0.50 RCS NOMINAL TEMPERATURE, F 532 532 532

'CS PRESSURE, psia 2250 2250 2250 CEA GROUP POSITION Pl UEL UEL UEL P2 UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL UEL 1 PI I UEL UEL LEL UEL UEL LEL UEL LEL LEL UEL LEL LEL UEL LEL LEL ~1 20ll 0100 0200 0900 DATE 4-25-76 4-26-76 4-30-76

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Page 66 5.5 Non-Overla ed Re ulatin and Shutdown CEA Grou Worth Measurements 5.5.1 Purpose During reactor operations, nearly all excess reactivity is held down by soluble boron concentration in the Reac-.

tor Coolant System and burnable poison shim r'ods in the fuel assemblies. Additional hold down and reactivity control is provided by moveable Control Element Assemblies (CEA). These CEA's are arrayed in symmetrical groups a-bout the core (see Figure 5.0-1). The number of CEA's in each Regulating and Shutdown CEA Group and the func-tion of that group is described in Table 5.5-1. The CEA Group worths were measured in a non-overlapped mode over the full range of their movement at various Reactor Cool-ant System boron concentrations.

5.5.2 Test Results All CEA Group reactivity worths were measured using a soluble boron swap method, either dilution or boration, to maintain criticality while inserting or withdrawing CEA Groups in increments. The reactivity trace generated by this evolution was reduced to obtain the rel'ationship between CEA Group positions from full in to full out and integral'eactivity worth at these positions.

For Shutdown CEA Group A, integral worth was measured using the soluble boron swap method in combination with a group trip method. The combination of methods allows extra- total integral worth of Group A to be determined from polation of measured data without decreasing shutdown margin below the Technical Specification limit.

The integral worths of all Shutdown and Regulating CEA Groups were measured at 532 F. These results are com-pared with predicted values in Table 5.5-1. In addition,

.the integral reactivity worth curves developed at 532oF for all Shutdown and Regulating CEA Groups are displayed in Figures 5.5-1 through 5.5-9.

5.5.3 Conclusions The measured CEA Group integral reactivity worths are in good agreement with predicted values and are well within acceptance limits.

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Page 67 TABLE 5.5-1 REACTIVITY CNORTH % hK/K ACCEPTANCE CEA GROUP NUMBER OF CEA'S FUNCTION PREDICTED MEASURED LIMITS Pl SHAPING P2 SHAPING A

• 16 SAFETY 4.120 4.520 3.090-5.150

• • 8 SAFETY 0.406 0.425 0. 305-0. 508 REGULATING 0.687 0:648 0.515-0.859 REGULATING 1.451 1.297 1.088-1.814 REGULATING 0.575 0.517 0.431-0.719 REGULATING 1.374 1.262 1.031-1.718 REGULATING 0.329 0.311 0.247-0.411 REGULATING 0.505 0.470 0.379-0.631 9 REGULATING 0.738 0.707 0.554-0.923

(

• EIGHT SETS OF DUAL CEA'S)

(** FOUR SETS OF DUAL CEA'S )

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page 68 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532 F, 2250 PSIA CEA GROUP 1 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-1

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Page 69 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532 F, 2250 PSIA CEA GROUP 2 1.50 1.40 1.30 1.20 1.10 1.00 A

0. 90 O

0.80

~ 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0 0 30 40 5 6 0 80 9 1 0 110 120 1 0 1 0 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-2

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Page 70 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532~F, 2250 PSIA CEA GROUP 3 0.75 0.7 0.6 0.6 0.5 0.5 0.4 0.4 0.3 0.3 0.2 0.2 0.1 0.1 0.0 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-3

I Page 71 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532 F, 2250 PSIA CEA GROUP 4 1.50 1.40 1.30 1.20 1.10 A

1.00 No4 0.90 C) 0.80 0.70 0.60 0.50 0.40

0. 30 0.20 0.10 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-4

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Page 72 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532 F, 2250 PSIA CEA GROUP 5

0. 75 0.70

0. 65 0.60 0.55 A

A 0.50

<<3 ai4 0.45 p

tD 0.40 M 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-5

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Page 73 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532 F, 2250 PSIA CEA GROUP 6

0. 75 0.7 0.6 0.6 0.5 0.5 A

0.4 a%~

0.4 g

O 0.3 0.3 H

0.25 0.2 0.1 0.1 0.0 0 10 20 30 40 50 6 0 80 90 100 110 120 130 140 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-6

page 74 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532OF, 2250 PSIA CEA GROUP 7 0.75 0.70 0.65

- '"0.60 0.55 0.50 0.45 0.40 0.35 0.3

0. 25 0.2 0.1 0.1 0.0

,0 140 20 30 40 50 60 70 80 . 90 100 110 120 130 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-7

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Page 75 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532 F, 2250 PSIA CEA GROUP B 0.75 0.70

0. 65

0. 60

0. 55

0. 50

0. 45 A

b4 0.*40

0. 35

0. 30

0. 25 0.20

0. 15 0.10

0. 05 1 2 30 40 50 6 70 80 90 100 110 1 0 1 0 1 0 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-8

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Page 76 ST. LUCIE UNIT 1 INTEGRAL CEA GROUP WORTH BOL, 1ST CYCLE, 532oF, 2250 PSIA CEA GROUP A 7.5 7.0 6.5 6.0 5.5 5 '

4 5 o 4.0 3.5 3.0 2.5 2.0 1.5 1.0 0.5 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 CEA WITHDRAWAL (INCHES)

FIGURE 5.5-9

I Page 77 5.6 Overla ed Re ulatin CEA Grou Worth Measurements Purpose Reactor Power level may be controlled by sequential insertion or withdrawal of Regulating Control Ele-ment Assemblies (CEA). Percent of overlap is select-ed so as to insure a relatively constant insertion rate of positive or negative reactivity over the full range of CEA Group movement. Technical Speci-fications allow CEA Group insertion as a function of rea'ctor power level. The integral reactivity worth curve for Regulating CEA Groups 1,2,3,4,5,6, and 7 in an overlapped mode was measured. Maximum allowed insertion at zero power being at approxi-mately 68 inches withdrawal on CEA Group 4.

This measurement was made at a Reactor Coolant Sys-tem (RCS) temperature of 532oF. Principal purpose of the measurement being to develop an integral worth curve for use in making estimated critical condition calculations prior to a reactor startup.

Reactor startup's are performed at a nominal RCS temperature of 532 F.

5.6.2 Test Results The overlapped integral reactivity worth of CEA Groups 1,2,3,4,5,6 and 7 was measured using a soluble boron swap method to maintain criticality while sequentially withdrawing CEA Groups in incre-ments. The reactivity trace developed by this CEA Group movement was reduced to obtain the relation-ship between CEA Group positions and integral re-activity worth at those positions. Figure 5.6-1 displays the overlapped integral reactivity worth curve.

5.6.3 Conclusions The Overlapped Integral CEA Worth curve derived from this measurement has been adequate for use as an operational tool, and has been placed in the Plant Curve Book.

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ms ST. LUCIE UNIT 1 BOL; 1ST CYCLE, 532 F, 2250 PSIA SEQUENTIAL CEA WORTH 5.00 4.00 3.00 2.00 RADIO 1.00 0 50 60 90 120 157 0 50 80 90 120 157 0 50 60 90 120 157 0 50 60 90 120 157 GROUP 1 GROUP 3 GROUP 5 ROU 7 0 50 60 90 120 157 O 5O 6O 9O 12O 157 0 50 60 90 120 157 GROU 2 GROUP 4 GRO 6 INCHES OF WITHDRAWAL FIGURE 5.6-1

I Page 79 5.7 Pr'essure Coefficient of Reactivit Measurements Purpose The pressure coefficient of reactivity can be either negative or positive depending on the magnitude of the Reactor Coolant System boron concentration.

5.7.2 Test Results The pressure coefficient of reactivity was not measured at PSL-1. However, the measurement was done at sister plants and the nominal value was

-5 x 10 7 ~k/k/psi.

5.7.3 Conclusions Results indicate that the pressure coefficient of reactivity is several orders of magnitude smaller than the temperature coefficient of reactiv'ity and will be considered relatively insignificant.

Page 80 5.8 Dro ed CEA North Measurements 5.8.1 Purpose A dropped CEA under power operation conditions will reduce reactor power level and distort the core power distribution. The reactivity worth of the most reactive CEA is measured from a full power CEA configuration in order to verify safety analysis.

5.8. 2 Test Results The dropped .CEA integral reactivity worth measure-ment was performed simultaneously with a check of core symmetry. The integral reactivity worth of each CEA was measured using a CEA swap technique.

This measurement was compared with that of all symmetric CEA's in its CEA Group in order to de-tect any unexpected core asymmetry. No signifi-cant asymmetry was noted and thereby gave addi-tional assurance of proper assembly and core load-ings.

The most worthy dropped CEA from a full power CEA configuration was CEA 8-9. Xts measured integral reactivity worth is compared with predicted worth in Table 5.8-1.

5.8.3 Conclusions The integral reactivity worth of the most reactive dropped CEA from a full power CEA configuration was determined to be slightly more than predicted and was well within the acceptance criteria.

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TABLE 5.8-1 SELECTED CEA WORTHS 4

CEA SWAP TECHNIQUE CEA REACTIVITY WORTH, q 7-1 10.80 7-38 6. 71 7-68 5.92 6-14 11. 85'

- 3 11.74

'4- 50 9.55 3-69 5.62 2-20 9.38 1-33 10. 32 B-9 17. 65 A-49 10. 90 Pl 11 6. 24 P2- 35 6.41 REACTIVITY WORTH CEA BORATE DILUTE AVERAGE AVERAGE PREDICTED LIMIT B-9 21.00 19.70 20.350 0.142 %hp 0.138 %hp 0. 231%hp

Page 82 5.9 E ected CEA Worth Measurements 5.9.1 Purpose F.S.A.R. safety analysis states that the maximum re-activity worth of any one ejected CEA in the core sha3,1 ~ot exceed 0.92% Ak/k at hot zero power at the beginning of core life and not exceed 0.23% 8c/k at full power at beginning of life.

5.9. 2 Test. Results A pseudo ejected CEA reactivity measurement was made at the rounded zero power dependent insertion limit.

CEA groups 7,6,5 and 4 were fully inserted; that is, at their lower electrical limits (LEL). The remaining CEA groups were fully withdrawn; that is, at their up-per electrical limits (UEL). Certain CEA's from the three inserted CEA groups were selected and their re-spective reactivity worths measured. The measurement technique was of borating the first selected CEA to UEL, and then CEA "swapping" between the succeeding CEA's, to measure their integral reactivity worths.

The most reactive CEA was 5-3. The results of this test are tabulated in Table 5.9-1.

A pseudo ejected CEA reactivity measurement was made at the full power dependent insertion limit. CEA group 7 was at approximately 74 inches. The other CEA groups were at their UEL's. CEA group 7 contains the center CEA and eight symmetrically located CEA's.

The center CEA, and one other group 7 CEA, was measured for reactivity worth using the borating/swapping tech-nique described in the preceding paragraph. The most reactive CEA was 7-1. The results of this test are tabulated in Table 5.9-2.

5.9.3 Conclusions The measured values of both pseudo egected CEA condi-tions were less than the predicted reactivity worths and were within the acceptable limits. Measured values were also less than the limit assumed (or set) by acci-dent analysis.

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Page 83 TABLE 5.9-1 TABLE 5.9-2 RCS TEMPERATURE, F 532 532 RCS PRESSURE, PSIA 2250 2250 NUCLEAR POWERS (0.1 (0.1 POWER DEPENDENT INSERTION LIMIT 'OUNDED ZERO FULL CEA GROUP POSITION UEL

• UEL UEL UEL UEL-LEL ** UEL

~

LEL 'EL LEL UEL LEL 174 INCHES WITHDRAWN CEA/WORTH (% Ak/k) 5-3 / 0.176 7-1 / 0.0322 CEA/WORTH (% Ak/k) 6-15 / 0.302 7/39 / 0.023 CEA/WORTH (% Ak/k) 7-1 '/ 0.083 PREDI CTED WORTH (% Ak/k) 0.329 0.050

+ 25% PREDICTED WORTH 0.247 0.411 0.038 0.063 W/IN PREDICTED RANGE YES F.S.A.R.'AFETY ANALYSIS LIMIT (% Ak/k) 0.920 0.230 LESS THAN FSAR LIMIT YES YES

• Upper Electrical Limit

• • Lower Electrical Limit

• • • OUTSIDE OF RANGE, BUT STILL ACCEPTABLE PER C.E. LETTER F-SF-842 DATED 5/4/76

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Page 84 5.10 Stuck CEA Worth Measurement 5.10.1 Purpose Technical Specifications state that available shutdown margin shall not be less than 2.45%, with the highest worth CEA stuck out, whenever the reactor is critical.

The reactivity worth of the most reactive stuck CEA was measured in order to verify the validity of predicted stuck CEA- reactivity'orths.

5.10.2 Test Results CEA A-48 was predicted to be the most worthy stuck CEA. The measurement technique consisted of CEA Group A combined trips with and without the "stuck" CEA stuck in the full out position. The reactivity data from these measurements was then used in combi-nation with data from the CEA Group A, reactivity worth measur'ement described in Section 5.5.2 to ex-trapolate an integral reactivity worth for CEA A-48.

This preliminary on-site evaluation resulted in a measured stuck CEA reactivity worth of 3.109% Ak/k.

A subsequent"off-site analysis of the measured data by CE-Windsor resulted in a stuck CEA reactivity worth of 3.146% Ak/k.

5.10.3 Conclusions The preliminary on-site analysis of measured data indicated that CEA A-48 reactivity ~orth was within

+ 25% of the design Hot Zero Power stuck CEA worth of of 2.7% Ak/k. Subsequent off-site analysis of data by CE-Windsor has verified this on-site conclusion.

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Page 85 5.11 Part Len th CEA Grou Measurements 5.11.1 Purpose The two groups (Pl and P2)of part length CEA's are designed for axial power shaping purposes, for the control of quadrant tilt and for the, control of offset. Their function is, as they are being with-drawn from LEL to UEL, to add negative reactivity, zero reactivity and finally positive reactivity.

5.11.2 Test Results PSL-1 is administratively prohibited from moving the part length CEA's from UEL. The reactivity worths are not needed, and therefore, were not measured.

5.11.3 Conclusions Measurements have been taken at other similar-design reactors and the results are available, if this information is ever needed.

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Page 86 6.0 Power Ascension Tests The power ascension tests are conducted to determine the as-built plant characteristics during steady state and transient operations from 0% to 100% power and to demonstrate that the plant is capable of withstanding the accidents and transients.

analyzed in the FSAR. Tests requiring steady state power were to be performed at major plateau's of 20, 50, 80 and 100% power.

Several other tests were to be performed at 14, 25, 30, 40, 60, 70, 75 and 85% power plateau's.

During power ascension testing an imbalance in the power pat-tern was discovered. At the beginning of the 80% plateau, it was decided to withdraw 'from power ascension testing.

The tests performed up through and including the 50% plateau are described in this start-up report, ref. Table 6.0-1. All other testing, including tests repeated due to the flux anomaly problem, will be described in one or more supplementing reports.

See Section 7 for further discussion.

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Page 87 TABLE 6.0-1 PROCEDURE TITLE & NUMBER  % POWER

'4$

(NOTE: P=Preoperationa1, 0 Opqratjng ) 5 14 20 0 30 40 50 SIMULATED CEA EJECTION 0110087 STEAM BYPASS 0810080 START UP TRANSFORMER 0910081 GENERATOR EXCITATION 0910085 NUCLEAR & AT POWER CALEB 1200051 2 2 GEN. TRIP OUTS. CONT. ROOM 1400093 PRIMARY CALORIMETRIC 3200020 2 2 FIXED INCORE ALARM SETPOINT 3200050 MOD TEMP COEF & PWR COEF 3200051 TOT. RAD. PEAK FACT. (Fr,~) 3200054 PWR RNG CONT SUBCH CALIB 3200080 FORCED XENON 3200087 RADIATION SHIELDING EVAL 3300081 1 CHEM & RADCHEM 3400081 S/G FEEDWATER HAMMER 0700080A

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Page 88 6.1 Steam B ass Valve Test and Turbine Generator Startu 6.1.1 Purpose The purpose of the test was to verify the proper operation of the steam bypass valves and control system. This includes valve stroke'time in the "quick-open" mode and automatic control of steam generator pressure. At the completion of this test the first turbine generator startup was then performed.

6.1.2 Test Results All valves met'or exceeded the specified 3 seconds maximum opening time. The valves operated satisfactorily in manual then easily met the acceptance criteria of maintaining 'nd pressure within 25 psi of setpoint operating in the automatic mode. This portion of the test included using the steam

'bypass valves to control reactor power up to about 14% steam demand. The turbine generator startup was then performed satisfactorily. During this startup the steam bypass valves

'ere used to control steam generator pressure.

6.1.3 Conclusions The sy'tem met or exceeded all acceptance criteria. In addition, during its first actual operation it performed satisfactorily to control steam 'generator pressure during the initial Turbine Generator Startup. The turbine and generator performed properly although the generator was not loaded at this time due to other necessary testing (See Main Generator Excitation System Initial Operation, Section .6.2).

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Page 89 6.2 Main Generator Excitation S stem Initial 0 eration 6.2.1 Purpose

'he purpose of this test was to perform initial energization of the'Main Generator Excitation System and the Generator and verify that the system would:

(1) flash the main generator field (2) increase and decrease voltage with the regulator out of service (base adjust) and in service (voltage adjust)

I (3) transfer the regulator in and out of service and remove the generator field and excitation circuits from service.

6.2.2 Test Results The test was performed after initial turbine generator startup and no problems of any significance were experienced. The Excitation System met all acceptance criteria (as listed above) and performed 'well in both manual and automatic modes of

,operation.

6.2.3 Conclusions The Excitation System will function properly to maintain generator voltage (nominal 22,000 volts) under all expected normal and .

transient operational conditions.

Page 90 6.3 20% Power Tri Test and Auxiliar to Startu Transformer Auto Transfer Test 6.3.1 Purpose The purpose of this test was to measure plant response to a reactor trip from 20% power and to verify proper transfer of the 4160 and 6900 volt A.C. buses from the auxiliary (in plant) to the startup (off-site) transformer.

6.3.2 Test Results The reactor trip initiated a turbine trip resulting in the interceptor steam valves and turbine valves closing and the generator field and output breakers opening. The automatic transfer functioned properly for the safety related buses but the non-safety related bus 1A1 breaker did not'lose to energize the bus from the startup transformer. A wiring error was discovered and corrected. On a later plant trip (2 days later) all buses transferred properly including the lAl bus.

Required transfer time was 0.3 seconds or less and actual transfer times were 0.05 seconds or less. Decay heat was satisfactorily removed by the Steam Bypass Control System and the Feedwater Regulating System reduced feedwater flow to about 5% flow as desired.- Also, Reactor Coolant System pressure was automatically controlled to less than 2500 psia (design pressure) throughout the transient.

6.3.3 Conclusions The acceptance criteria (given above) were all properly met.

The conditions at the completion of the transient were all as intended; RCS pressure less than 2400 psia, pressurizer cool-down less than 200 F per hour, Steam Generator pressure less than 985 psia, RCS cooldown less than 100 F/hour and electrical power being supplied by the startup transformers. This verifies that all involved systems performed properly to control a plant trip from 20% power.

il Page 91 6.4 Plant Power Calibration 6.4.1Puxpose The purpose of the test was to:

I (1) Determine core thermal power by means of a primary plant heat balance.

(2) Adjust the Power Range Safety Channels and AT Power Reference Calculators to agree with the ther-mal energy balance calculations.

(3) Perform, when necessary, a calibration of the Safety and/or Control Power Range Channels.

6.4.2Test Results Feed flow Calorimetrics using hand calculations were con-ducted at the 20 and 50% power plateaus". The Calori-metrics were used to calibrate nuclear instrumentation and to verify the plant computer core thermal power cal-culations. The Power Range Safety Channels and Reference Calculators were adjusted to agree with-CL'ower in 0.5% of the Calorimetric calculations. These adjustments were performed at the 20, 30, 40 and 50%

power plateaus.

Initial calibration of the Power Range Safety Channels was conducted during the 20% test plateau using the Keithley pico-ammeters as a standard for subchannel calibration.

Adjustment of the Power Range Contxol Channels was also completed at the 50% power plateau.

6.4.3Conclusions V

Hand calculations of core thermal power pointed out some minor discrepancies in the computer calculations. After correction of the deficiencies, computer calculations proved to be reliable and accurate. Calibration of the Power Range Safety and Control Subchannels was accomplished acceptably at each major test plateau. The, intent of the Ex-Core Nuclear Instrument Calibration was to adjust nuc-lear power, AT power and the Calorimetric to within 0.5%

of each other. Plant operating procedures contain in-structions for hand calculations in case of computer fail-ure and/or to verify computer calculations.

Il Page 92 FIGURE 6.4-1

'-:.'j DDPS CALORIMETRIC FOR NUCLEAR & AT. POWER CALIB. O.P. 1200051 DATE NOMINAL REACTOR POWER, CALORIMETRIC POWER 5-8-76 20 20.76 5-9-76 20 22.55 5-12-76 20 19.23 5-13-76 30 30. 06 5-14-76 30 30. 12 5-15-76 40 39.76 5-20-76 50 50.63 5-20-76 50 48.73 5-21-76 50 49.00 5-23-76 50 47. 69 5-24-76 50 49.80 5-25-76 50 49.80 5-26-76 50 50.07 5-27-76 50 50. 19 PRIMARY CALORIMETRIC, OPERATING PROCEDURE 3200020 IJ NOMINAL REACTOR POWER, 20 50 Qc, BTU / HR 1.971 e 9 4.110 e 9 ACTUAL REACTOR POWER, 23.7 49. 6 DATE 5-9-76 5-20-76

I Page 93 6.5 Power Ran e Safet and Control Subchannel Calibration 6.5.1 Purpose The purpose of this procedure was to calibrate the upper and lower power range excore detectors and provide (input) Shape Annealing Factors (SAF's) to be used by the Subchannels to calculate the Axial Shape Indices (ASI's).

The excore power range detectors must be accurately calibrated in order to monitor the in-core flux profiles as determined by the incore flux detectors. At 20% power,conservatively assumed value's of SAF were input to allow operation until 50%

power. At 50%, actual observed SAF's were input. In both cases the ASI's calculated from the Safety Subchannels signals to the Reactor Protection System were required to be within 5% of the computer calculated values based on the incore detector signals.

6.5.2 Test Results All Subchannels calibrated properly at both 20% and 50%

power although it was noted that detector signals were somewhat lower than expected requiring amplifier gain to be increased for proper operation and calibration. Using the assumed values of SAF, the RPS calculated values of ASI were well with the 5% limit when compared to the actual com-puter calculated ASI's. And, using the observed SAF values, the RPS calculated values of ASI agreed quite well with the actual ASI's (well within the 5% tolerance).

6.5. 3 Conclusions This procedure is used periodically to calibrate the sub-channels and verify that the outputs are correct and the results confirmed that the procedure and equipment will properly calibrate the Power Range Subchannels. Also this procedure assured safe operation between 20% and 50% power and at 50% power verified that the ASI's calculated by the Power Ran'ge Excore Subchannels are in fact an accurate representation of the actual in-core flux profiles.

Page 94 6.6 Shieldin Effectiveness and Plant Radiation Level Measurements 6.6.1 Purpose The test was conducted to accomplish the following objectives:

(1) Determine background radiation levels prior to plant startup.

(2) Evaluate the adequacy of plant radiation shielding.

(3) Determine radiation levels throughout the plant "at various power levels.

6.6.2 Test Results A comprehensive series of gamma and neutron dose rate level surveys was performed during initial startup, low-power-physics testing and power'escalation. Dose rates at select-ed points, both inside and outside of the Radiation Controll-ed Area, were determined. Dose rates at each point were de-termined at power levels of 0% (Background), 1xl0  %, 5%, 20%

and 50%. The final survey planned for the 100% power level has not been completed, since the facility has not operated at full power as of December, 1976.

Radiation dose rate levels at each point-were compared for different power levels to verify that a linear relationship existed. This was done to ensure an extrapolation of dose rates to 100% power could be considered valid allowing for identification of potential problem areas. It was also felt that it would be in keeping with AIdLRA"to limit dose for per-sonnel performing surveys in areas where access during full-power operation would be highly restricted due to high and possibly variable dose rates.

In addition to personnel performing surveys, two special neutron monitoring systems were available allowing dose rate determinations to 20 Rem/hour (neutron only).

Design, dose rates utilized for comparison are specified in the St. Lucie FSAR, Chapter 12, Section 12.1.1.

General area gamma dose rates for all areas around the con-tainment were less than 0.1 mrem/hour and neutron levels were less than 0.5 mrem/hour with the exception of the per-sonnel hatch and the two entrance doors to the containment annulus. These measurements do not include electrical and mechanical penetration areas in the reactor auxiliary

• ALARA As Low As Reasonably Achievable

l Page 95 6.6 Shieldin Effectiveness and Plant Radiation Level Heasurements (Cont.)

6.6.2 Test Results (Cont.)

building or the fuel handling building or spent fuel pool area. These enclosed equipment spaces may experience variable dose rates. Therefor'e, in addition to the ini-tial measurements which verified that these areas were not Radiation Areas, they are surveyed periodically to ensure proper control on a continuous basis.

Dose rates extrapolated to 100% power indicated approx-imately 3 mrem/hour and 10 mrem/hour contact with the southwest and northeast annulus doors respectively (com-bined~a/ dose rates). At the containment entrance hatch combined neutron and gamma dose rates extrapolated to 100%

power indicated 1 mrem/hour outside the airlock, 22 mrem/

hour in the airlock, and 80 mrem/hour in the containment at the airlock exit. Neutron/gamma dose rate ratios in containment were variable from approximately 1 to as high as 20 at some points along the cavity edge.

6.6.3 Conclusions Maximum general area gamma and n'eutron dose rates as deter-mined by shielding effectiveness surveys at the St. Lucie Plant were generally consistent with criteria presented in Section 12.1.1 of the FSAR for areas outside the exterior containment wall and outside the Radiation Controlled, Area boundary. The presence of possible projected neutron streaming problems in the containment, RAB and personnel emergency escape hatch areas was reinforced by the results of the surveys performed to date.

As noted in FSAR Section 12.1, the analytical and empirical results to date indicate the need for shielding and so the design effort for St. Lucie 1 has preceded the measurement program. Following completion of the St. Lucie 1 neutron streaming measurement program, the adequacy of the proposed shield design will be evaluated against the actual measured data. The proposed reactor cavity neutron shield consists of nylon-neophrene covered bags holding ordinary light water (non-borated). The water bags will be fabricated by the B. F. Goodrich Company. The bags are vertically installed by prefolding them in a manner such that no gaps will exist between bags as the bags are filled.

The 100% power level shielding effectiveness survey will be performed as soon as practicable after the unit reaches full power operation and will be repeated following installation of neutron shielding currently being considered to reduce neutron streaming from the gap around the reactor vessel.

I Page 96

6. 7 Chemistr and Radiochemistr at Power 6.7.1 Purpose

1) To ensure that the primary and secondary systems water chemistry meet the criteria set forth in the St. Lucie I Chemistry Procedures Manual for system protection.

2) To correlate corrosion data and fission product buildup data to power levels.

6.7.2 Test Results:

Chemical and Radiochemical tests were performed as specified in the Chemistry Procedures Manual and Preoperation Test Procedure No. 3400081 (Chemical and Radiochemical Analyses 20, 50, 80, 100% power).

The plant reached 80% power for a short period of time and was shutdown due to a flux tilt problem.

Meaningful data was obtained for steady state powers of 20 and 50%. (Table 6.16-1).

6.7.2.1 Primary

1) All parameters were within limits except Lithium which was intentionally high for better filming during initial power operations.

2) A small fuel failure was suspected shortly before plant shutdown. On 6-22-76 the iodine ratio began increasing. Several days passed before a significant increase in Dose Equivalent Iodine 131 was seen. On 6-30-76 at 80% power the iodine ratio reached 0:4 and Dose Equivalent I131 reached 1.04 x 10 uc/ml. Also an increase in VCT fission gases was noted. The reactor was shutdown soon after so conclusive data is not yet available.

6.7.2.2 Secondary

1) Moisture carryover test was not performed due to premature plant shutdown.

2) Maximum steam generator blowdown was maintained when possible, foi the ma)or part, of power testing.

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Page 97

6. 7 (cont. )

6.7.2.2 (cont.)

3) Condensate and Feed Systems were flushed until solids were as low as possible prior to feeding the steam generators.

'4) It was found that pH additives were not necessary to control steam generator or feed pH while maintaining a feed system hydrazine residual of 30-40 ppb.

5) Suspended solids in the Feed System and steam generators were lower than expected until the mechanical shock of the High Volume Steam Dump Valves drove Feed Solids to 0.8 ppm and Steam Generators 1A to 44 ppm and 1B to 51 ppm.

Maximum blowdown was maintained resulting in a decrease in solids to specification.

6) At one point silica increased significantly and steam generator sample water became frothy.

This is believed to have resulted from initial condensate cleaning chemicals that could not be flushed during flush evolutions at that time.

Maximum blowdown was used to remedy this problem.

6.7.3 Conclusions Chemical Testing, both primary and secondary, showed that the plant was in a generally good condition. It also showed that chemistry could be controlled within specified limits and that out of specification parameters could be remedied in a timely manner.

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Page 98.

TABLE 6.7 1 RCS CHEHISTRY RESULTS AT STEADY STATE POWER 5-10-76 5-26-76 Analysis Limits 20% 50%

pH 4.5-10.2 6.28 6.7 Cond'uctivity Varies 8.3 16 Cl <.15 ppm <:05 <. 05

<.1 ppm <.05 <.05 D.O. .005 <.005 S.S. <.5 ppm <.01 <.01 Boron Varies 822 714 Lithium .2-2.0 ppiR .73 1.49 Diss. H2 10-50 cc/kg 31.8 47.2 II Gas Act. 7.32e-4 6.99e 3 Gross Act. 1.8le 2 4 3e Crud Act. 5 2e 6 . 2.1e-4 Tritium 5.6e I ratio .07 .08 DEq I-131 <1 uci/gm 4.4e 4 2.5e-4 Spectrum performed performed

l Page 99 6.8 Fixed Incore Detector Alarm Set pints 6.8.1 Purpose This was not a test, but more of an instrumentation setpoint procedure. The purpose was to calculate and adjust the fixed incore detector alarm setpoints.

The core is considered to be divided into four axial regions, each approximately one fourth the core height, and each encompassing the axial region moni-tored by one segment level of the incore detector.

6.8.2 Results The procedure instructs that a set of readings of the measurement of the incore detector readings, various temperatures, and CEA heights be taken.

The plant computer has a program called "SNAPSHOT" that will do this. Via hand calculations or a computer program called "GINCA", the Nodal over-power ratios are calculated. Through various fac-tors and constants, an incore detector alarm set-point is generated for each of the 180 incore de-tectors. This was repeated periodically as power level was increased to ensure up to date setpoints.

At the beginning of the 80% plateau, the setpoints were readjusted due to discovery of STRIKIN II computer code errors (See Section 6.13). An incorrect adjustment factor was used in the setpoint calculations resulting in nonconservative setpoints. This was noted about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> later and corrected. Review of the data verified no actual setpoints were violated. (LER 335-76-34, August 6, 1976) 6.8.3 Each of the alarm setpoints was entered into the Digital Data Processor. The operability of the Digital Data Processor was checked, and each of the fixed incore detectors was determined to have a valid alarm set point.

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Page 100

6. 9 Reactivity Coef ficient Measurements 6.9.1 Purpose A test commonly referred to as a Variable T avg Test was conducted to determine the Power Coefficient and the Isothermal Temperature Coefficient (ITC).

6.9.2 Test Results Variable T avg Tests were conducted at the 50% power plateau with the Control Element Assemblies (CEA) inserted to approximately 100 inches on CEA Group 7.

The test was conducted with the Power Coefficient and the Isothermal Temperature Coefficient (ITC) as sep-arate tests. During the ITC test, AT power was held constant and Reactor Coolant System (RCS) T avg was varied. T avg was decreased 10 F below original temperature, conditions stabiliz'ed, data recorded and temperature increased to original temperature, conditions stabilized, and data recorded. This cycle was repeated twice.

The Power Coefficient Test was conducted by holding T avg constant and decreasing gross electrical power by approximately 5%, conditions stabilized, data re-corded. Then gross electrical power was increased to the original power level, conditions stabilized and data recorded. This cycle was repeated twice.

The final power coefficient and ITC values were the average value of the runs conducted. The measured value and limits for the temperature and power co-efficient for the 50% power plateau are shown in Table 6;9-1.

6.9.3 Conclusions The measured value for Isothermal Temperature Co-efficient was well below its limit. The Power Coefficient was within the tolerance of its limit.

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Page 101 TABLE 6.9-1 NOMINAL REACTOR POVER 50%

MODERATOR TEMPERATURE COEFFICIENT Ak/k / F

-0.10 10 LIMIT <+ 0.50 x 10 ~j.5.g POWER DEFECT COEFFICIENT hk/k /  %

MEASURED -1.07 x 10 LIMIT ~

(-1 0+O.l)xx10

II Page 102 6.10 Total Radial Peaking Factor 6.10.1 Purpose This was done at 50% power to measure the total T T radial peaking factor, Fr. Fr is defined as the

'product of the unrodded planar peaking factor and the quantity one (1) plus the azmuthal tilt.

F,=Frx (1+Tq).

T P 6.10. 2 Test Results This was more of a 'measuremen't than it was a test.

As stated in Section 6.0, a power imbalance became apparent during the power ascension testing sequence.

This imbalance grew very apparent during testing at power levels greater than 50%. Selected data is pre-sented in Table 6.10-1.

6.10.3'onclusion The total radial peaking factor measurement'one at the 50% power level proved that the plant was opera-,

ting below the Technical'pecification limit of 1.36.

However, as previously stated, the azmuthal tilt/

total radial peaking factor at higher power levels was serious enough to warrent withdrawal from the power ascension testing program.

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Page 103 TABLE 6.10-1 i~~TO'fAL~RROTAO':.FRA1CCNG'ACTOR DATE 5/21/76 POWER, 50 0.006

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Page 104 6.11 Xenon Follow Measurements 6.11-.1 Purpose The purpose of this test was to 'obtain transient test data at the 50% power plateau for the purpose of eval-uating the Shape Annealing Factor (SAF) for each Power Range Safety Channel and to evaluate an induced free xenon oscillation.

6.11. 2 Test Results During the 50% test plateau, axial oscillations were induced in the core. These oscillations were monitored by the In Core Analysis GINCA computer program and the Axial Shape Index (ASI) calculated by the Reactor Protective System (RPS). All axial oscillations were convergent.

The SAF for each Power Range Safety Channel was also

'easured. The SAF corrects the detector signal to a'ccount for the distance from the excore detector to the reactor core and corrects for the signal received by the upper detector from neutrons generated in the bottom of the core and the signal received by the lower detector from neutrons generated in the upper part of the core. The SAF is determined by plotting ASI GINCA versus the ASI (EXT) as read from the Reactor Protec-tive System (RPS) during a xenon oscillation with all CEA's full out. The slope of the line resulting from this plot is the Shape Annealing Factor. ASI .is',Axial Shape Index, a ratio of the difference in power gen-erated in the lower and upper halves of the core to total core power. Table 6.11-1 summarizes the compari-sons between F.P.&L. and C.E. results. As measured SAF's were incorporated into RPS setpoints.

6.11.3 Conclusions The induced xenon oscillation test proved that induced xenon oscillations were self dampening to a stable pow-er distribution and that resultant DNBR and LHR were both well within acceptance limits.

The ASI's c'alculated by the RPS as well as GINCA were continuously observed and evaluated. Observation verified the adequacy of SAF's previously determined from measured data.

gi Page 105 TABLE 6.11-1 CHANNEL SHAPE ANNEALING FACTOR (SAF) INTERCEPT F.. P. W&C~E F,. ':P~i;i &'"-L.

A 3.45 3.46 -0.0004 -0.0012 3.76 3.77 +0'. 0019 +0.0009 3.89 3.92 +0.0090 -0. 0102 3.06 3.07 -0.0010 -0.0015 3.21 3.21 +0.0010 0 10 3. 67 3.68 +0.0070 +0.0060

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Page 106 TABLE 6.11-2 DATE TIME A. S. I. CEA GP 7 DATE TIME A. S. I. CEA GP 7 29HAY,76,. 2130 +0.011 ARO 30MAY76~, 2314 -0.149 ARO

/+~HA 76': 0624 +0.011 133 ~Y76~ 0020 -0.146 ARO 0721 +0.168 87 0116 -0.136 ARO 0813 +0.281 68 0200 -0. 123 ARO 0914 +0.312 68 0300 -0.107 ARO

'1016 +0. 334 68 0400 -0.087 ARO 1113 +0.345 68 '500 -0.067 ARO 1220 +0.337 73 0600 -0.044 ARO 1317 +0.265 88 0700 -0.024 ARO 1413 +0.133 116 0800 -0.004 ARO 1516 +0.056 ARO 0900 +0.017 ARO 1645 +0.005 ARO 1000 +0.034 ARO 1715 -0.033 ARO 1100 +0.048 ARO 1817 -0.075 ARO 1200 +0.061 ARO 1915 -0.102 ARO 1600 %0. Q7.9'" ARO 2020 -0.125 ARO TUN76 f 123 -0.010* ARO

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2110 -0.135 ARO 02 JUH76.,0812 +0.017** ARO 2214 "

-0.145 ARO + MIN. VALUE ** MAX..VALUE

5 ST. LUCIE UNIT 1 B.O.L., 1ST CYCLE, 532~F, 2250 PSIA A.S.I. vs TIME

+0.3 C4 O

Q 5 +0

<o~ 2 u ~ +0.1 H

00 H

-O.l

-0.2 22 2 2 4 6 8 10 12 14 16 18 20 22 2 2 4 6 8 10 12 14 16 18 20 22 2 5/29/76 5/30/76 5/21/76 FIGURE 6.11-1

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Page 108 6.12 Pseudo Ejected CEA Power Distribution Mea'surement 6.12.1 Purpose The purpose of this test is to measure the core power distribution resulting from a CEA being "ejected" from a full power dependent insertion limit (FPDIL) CEA con-figuration. The measurement results are compared with that predicted by Safety Analysis to verify its con-servatism.

6.12.2 Test Results When CEA's are inserted to the full power PDIL, only CEA Group 7 is in the core. With respect to relative reactivity worth, CEA Group 7 consists of three (3) types of CEA's. The center CEA and one peripheral CEA was ejected from the full power configuration and appropriate power distribution information col-lected using the following technique. Equilibrium xenon was established at a nominal 50% power level, and then, core power distribution information gathered from the in-core neutron detector's'. Two of CEA Group 7 CEA's were "ejected" from the core; the first by a soluble boron swap technique; and the other by a rod swap with the preceding CEA. Power distribution in-formation from the in-core detectors was gathered at the full-out cond'ition for each CEA. Power level was maintained at 50% throughout the test.

A three-dimensional power peaking factor was developed from measured data by extrapolating in-core detector signals to the "ejected" CEA locations. As had been expected, this resulted in a power peaking factor con-siderably below that predicted in safety analysis.

Safety analysis assumed an ejected CEA w'orth which was an order of magnitude greater than that measured during Low Power Physics Testing. Results of the on-site analysis for each CEA are presented in Table'.12-1..

6.12.3 Conclusions A preliminary on-site analysis of data indicates that the 3-D power peaks resulti'ng from pseudo ejected CEA power distribution measurements are considerably smal-ler than those calculated in the Safety Analysis. The results of this on-site evaluation when coupled with the results of ejected CEA worth measurements made during LPPT confirm that the Safety Analysis of the CEA Ejection Incident is very conservative.

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Page 109 TABLE 6.12-1 DATE JUNE ll, 1976 NOMINAL RCS TEMPERATURE 548. 7 F NOMINAL RCS PRESSURE 2253.4 PSIA NOMINAL NUCLEAR POWER 49.8  %

3-D FUEL ROD PEAK,= Fq CEA 7-1 l. 99 CEA 7-59 1.97 LIMIT < 5.10

Page 110 6.13 Core Power Distributions 6.13.1 Purpose Detailed core power distribution measurements were per-formed under steady state conditions during PAT to veri-fy that fuel assembly power fractions, axial power dis-tributions and peak linear heat rates were within accep-table limits.

The specific acceptance criteria applied to the measured core power distributions are listed below.

Fuel assembly power fraction: Fuel assembly power fraction is defined as the ratio of the average Linear Heat Rate (LHR) in a fuel assembly to the average LHR over the en-tire core.

(2) Axial power distribution: The measured core average axial power distribution shall be compared with the predicted distribution for general agreement.

(3) Peak linear heat rate: It shall be less than 12.7 KW/ft.

6.13.2 Test Results A summary of core power distribution results is presented in Table 6.13-1.

6.13.2.1 Fuel assembly power fraction: Steady state equilibrium xenon core power distribution measurements were performed at 20% and 50%

power plateaus. The definition of equili-brium xenon used in this section is:

The change in critical boron con-centration as determined from Re-actor Coolant System chemical ana-lysis, shall be less than 1% devia-tion from the average, of 3 con-tinuous samples, taken 15 minutes apart.

The analysis of the incore detector readings is performed by two computer programs. The first program automatically converts the voltage signal from the detector'o the correct neutron flux level. The Incore An-

.alysis(GINCA) program converts the neutron flux levels and various other reactor para-meters and, on demand, calculates several in-core data. The GINCA program assumes eighth

page 111 core symmetry. The two cases re-ported in this section are shown in Tables 6.13.1 through'6.13.3,.

6.13.2.2 Axial power distributions: At steady state equilibrium xenon, the core av-erage axial power distribution was determined at each of the major test plateaus using the GINCA program. Fig-ures 6.13-2 and 6.13-3 display the measured values of core average axial power distribution.

6.13.2.3 Peak LHR: The peak LHR is determined by the GINCA program As can be seen in Table 6 13-1 the peak LHR never exceeded the ac-ceptance criteria of 12.7 KW/ft. If the worst case'f 5.33 KV/ft is multiplied by 1.2991 to account for uncertainties, the resultant 6.9242 KW/ft is still acceptable.

These uncertainties include measurement-calc'ulational uncertainty, an engineering factor, effects of fuel densification and thermal expansion, and power measurement uncertainty.

6.13.3 Conclusions At steady state equ'ilibrium xenon, the. axial and radial core power distributions are within acceptable limits.

The peak linear heat rate does not exceed that allowed by Technical Specifications. It should be noted that the limiting peak LHR of 12.7 KW/ft is an interim limit imposed due to the discovery of coding errors in the vendor's STRIKIN II computer code. Results of the first

- corrected analysis (worst case LOCA) indicate tha't, when full reanalysis is complete, the permanent limit LHR will be higher thus making these results even for'eak more conservative.

Page 1l2 TABLE 6.13-1 MEASURED POWER, 19.38 49.64 CORE BURNUP, EFPH 13 200 BORON CONCENTRATION, PPM 827 730 LINEAR HEAT RATE, KW/FT 2.18 5.33 RADIAL PEAK, F 1.3257 1.3414 r

AXIAL PEAKt Fz 1.296 1.336 NOMINAL RCS TEMP, F 537 538 NOMINAL RCS PRESSURE, PSIA 2250 2245 CEA CONFIGURATION ARO ARO 2100 2130 DATE 5-.11-76 - 5-29-76

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PO!PER 19.38% TABLE 6.13-2 TIME 2100 Fuel Type 8-C-O Relative Power Knsit 0.662 DATE 5-11-76 Linear Heat Rate 1.619 f-C-0 8 C-2 0.662 1. 109 1.619 1.969 1-C-0 1-C-2 P.-B-O 1-A-0 0.563 1.060 1.054 1.006 1.459 1.900 1.800 1.669 X'-C-0 1-B-0 1-A-0 I-B-0 PA-0 0.767 0.993 0.992 1.122 1.059 1.616 1.726 1.648 1.856 1.742 1-C-1 1-A-0 i-B-0 1-A-0 1-B-0 8 A-0

1. 021 0. 980 1.104 1.050 1.166 1.084 1-C-0 2. 007 1.628 1.827 1.729 1.906 1.779 0.588 1.522 0-C-1 T-B-0 f-A-0 1-B-0 1-A-0 P-B-O f-A-0 1.149 1.084 1.033 1.153 1.081 1. 187 1.097 1-C-0 1.925 1.796 =1.701 1.885 1.755 l. 919 1.800

0. 766 1.606 f-B-0 9-A-0 l~B-0 1%A-0 1-B-0 fgA-0 8 B-0 5A-0 1.009 1.001 1.125 1.063 1.174 1.097 1.193 1.088 1.714 1.646 1.843 1.744 1.970 1.800 1.935 1.785

POWER 49.64%

TABLE 6.13-+3 TIME 2130 Fuel Type f-C-0 Relative Power Density 0.691 DATE ~ 5-29-76 Linear Heat Rate 4.240 1-C-0 1'.-C-2 0.691 1.067 4.240 4.712 1'-C-0 t 1-C-2 I'-B-0 1'-A-0

0. 596 1.024 1.004 1.049

3. 891 4.654 4.370 4.376 1'-C-0 1'-B-0 1-A-0 1-B-0 I-A-0 0.829 0.952 1.033 1.081 1. 092 4.460 4.199 4.318 4.490 4.524 8-C-I I-A-0 f-B-0 f-A-0 1-B-0 I-A-0

1. 027 1.038 1.065 1.086 1.114 1.112 5.108 4.357 4.465 4.501 4.622 4.615 f-C-0

'.623 4.125 1-C-1 1-B-0 I-A-0 1-B-0 1-A-0 1-B-0 1'-A-0

l. 132 1. 044 1. 071 1.097 1.086 1.122 1.124 1-C-0 4.913 4.426 4.459 4.519 4.458 4.637 4.647

0. 814 4.318 f-'B-0 1-A-0 I-B-0 9-A-0 8-B-0 1-A-0 1-B-0 f-A-0 0.995 1.049 1.080 1.091 1.114 1.118 l. 138 1.131 4.665 4.455 4.517 4.526 4.620 4.633 . 4.660 4.663

P Page 115 NORTH FIGURE 6. 13-1 LOCATION OF INCORE NEUTRON DETECTORS, 45 44 42 41 40 38 3?

36 36 34 33 32 31 30 29 28 27 26 25 23 22 20 19 17 15 13 12 10 7'

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ST. LUCIE UNIT 1 19.38% POWER, %13 EFPH CORE AVERAGE AXIAL POWER DISTRIBUTION 1.500 1.400 1.300 1.296 1.200 1.100 1.000 0.900 0.800 5 0.700

~ 0.600 0.500 0.400 0.300 0.200 0.100 13.6 27.2 40.8 54.4 68.0 81.6 95.2 108.8 122.4 136.0 CORE HEIGHT INCHES FROM BOTTOM OF CORE FIGURE 6.13-2

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ST. LUCIE UNIT 1

• 49,64% POWER, +200 EFPH CORE AVERAGE AXIAL POWER DISTRIBUTION 1.500 1.400 1.336 1.300

'.200 1.100 1.000 0.900 o 0.800 0.700 0.600 0.500 0.400 0.300 0.200 0.100

13. 6 27. 2 40. 8 54.4 68.0 81. 6 95.2 108.8 122.4 136 CORE HEIGHT INCHES FROM BOTTOM OF CORE FIGURE 6.13-3

I Page .118 6.14 GENERATOR TRIP WITH SHUTDOWN OUTSIDE CONTROL ROOM 6.14.1 Purpose The purpose of this test was to demonstrate that the plant responds properly to a generator trip from 50%

power and that it can be safely shutdown to hot shut-down conditions from outside the control room without exceeding any safety'imits; 6.14.2 Test Results The acceptance criteria were:

(1) Successful shutdown of the plant from outside the control room by the normal complement of plant operators without exceeding any safety limits, including a check that all automatic trip related functions did, in fact, occur.

(2) Emergency communications successfully establish-ed between the Hot Shutdown Control Panel and various local operating stations .

(3) Successful boration (at least 10 ppm) from the remote operating station .

(4) Satisfactory removal of decay heat using the steam dump and maintaining reactor T avg at or below the hot standby temperature.

(5) Satisfactory control of steam generator level by manual control of the auxilliary feedwater sys-tem.

All the acceptance criteria were met with no significant problems.

6.14.3 Conclusions Satisfactory completion of this test demonstrates that the plant can be safely controlled and shutdown should the control room become inaccessible.

III Page,119 6.15 Steam Generator Feedwater Hammer"Test

6. 15. 1 Purpose The purpose of this test was to verify the absence of any water hammer in the steam generator feedwater piping when the steam generator was drained below the feed ring and then refilled.

6. 15. 2 Test Results Following a trip from 33% reactor power, level in one steam generator was reduced below the feed ring and held at that level or lower for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Then the level was raised to normal at the maximum flow rate (300 gpm) of one Auxiliary Feedwater pump. The behavior of the feedwater piping was monitored by:

observers inside containment, installed RCS noise monitoring equipment and by measurements of line and restraint positions before and after the test.

No evidence of feedwater hammer was observed.

6, 15.3 Conclusions The absence of any evidence of steam generator feedwater hammer indicated that St. Lucie Unit 1 was not susceptible to the problem.

Since performance of this test, the method of pre-venting the feedring from draining has been modified in accordance with 10CFR50.59. Therefore this test will be repeated upon return to power operations.

Page 120

7. 0 UNUSUAL EVENTS During the time interval covered by Startup Testing, several ma)or problems occurred which significantly affected testing.

The first 4 problems failed to delay testing only because the fifth, and most important one, created a delay of approximately six months. The problems were:

(1) Higher than predicted cooling water discharge canal levels (2) Inability to operate CEDM 44 when cooled down (3) Apparent lower than predicted Reactor Cooling Pump Flow (4) Higher than predicted Containment radiation levels (5) A power distribution anomaly These problems and their resolution are discussed in greater detail in the following Sections 7.1 through 7.5 respectively.

Ill Page 121 Higher Than Predicted Cooling Water Discharge Canal Level When all cooling water systems began to operate simultaneously, discharge canal levels were found to be higher than expected .

When this was observed to threaten spill-over from the canal banks (especially at high tide), evaluation was begun; Flow determinations for the main circulating water pumps were performed by FP&L's Power Resources Test Group to see if the pumps produced greater than design flow. This was, in fact, true so the pump discharge valves were throttled to gust above design flow rate. However, this was not the only problem as canal level still threatened to spill over the banks during high tide.

More extensive investigation/evaluation was undertaken. This included having divers photograph the interior of the canal outfall pipe . This 12 foot concrete pipe extends from the .

canal's end, under the dune line and beach for 1200 feet out to sea. This pipe had been completed for nearly a. year with virtually no flow through it. The photographs and TV video tape proved that there was extensive marine growth on the interior surfaces. In places this fouling was up to an inch deep. Under normal operating conditions, flow through the pipes and the small residual chlorine level allowed would help reduce this fouling although it cannot be entirely eliminated.

Although it was felt that this fouling was not the only problem, it was obvious that the growth must be removed to allow proper evaluation of the system. Therefore, Florida Power & Light Personnel designed and had fabricated a large pipe cleaning machine. During the early part of the shutdown for resolution of the power distribution anaomaly, the pipe was cleaned.

Once this was completed, the main circulating water pumps (turbine condenser cooling water) were run as much as possible to help prevent recurrence of the fouling. There were no cases of threatened spill-over observed for 2 months.

The LPPT and initial ascent to power= (December 1976) coincided with a period of higher than average tides compounded by a weather front also causing higher tides. Under these conditions, it again proved necessary to reduce cooling water flow to avoid spill over.

Again divers photographed the interior of the pipe . A thin layer of fouling was evident on most of the surface. It is believed that the fouling reduces flow not by reducing pipe cross sectional area but, by its non-uniformity., inducing turbulent flow and thus increasing the friction factor.

Page 122 7.1 Higher Than Predicted Cooling Water Discharge Canal Level (cont.)

This problem is still not resolved. In the future during certain high tides, it will be necessary to reduce turbine load and cooling water flow. Various alternatives are being evaluated but no final decision has been made.

NOTE: This problem raised concern that a "reverse" problem might arise for the intake canal. That is, at full flow and low tides the drawdown in the intake canal might be excessive. It has been verified that, although drawdown is greater than expected, it is not excessive and presents no safety or operational concerns.

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Page 123

7. 2 C~EDM During cold rod testing, this CEDM was found, to be inoperative (would not withdraw) at low temperatures. This CEDM was tested after heatup (as were all CEDM's) and found to operate very satisfactorily. It was, in fact, the second fastest rod in terms of the 90% insertion drop time. After a cooldown,with this CEDM out, it dropped satisfactorily and would withdraw some, indicating slight improvement. This rod never failed to drop properly and never failed to operate hot satisfactorily.

Amendment No. 4 to St. Lucie License No. DPR-67 was issued April 16, 1976 . This deleted a special test exemption allowing low temperature criticality for physics tests. It also required repair or replacement of CEDM 44, at the first shutdown expected to last 2 weeks or longer. This allowed hot criticality and continuation of the test program.

This CEDM was tested each shutdown and cooldown during testing.

It continued,to show operability improvement after each heatup/

cooldown cycle and again, never failed to trip, which is its only safety related function.

During the early parts of the shutdown for resolution of the power distribution anomaly, this CEDM was inspected mechanically.

No conclusive evidence was noted. As soon as it became apparent this shutdown would be lengthy it was decided to replace CEDM 44 even though it presented no safety concerns.

This CEDM has been replaced and was retested satisfactorily (cold and hot) upon return to operation. The NRC Division of Inspection and Enforcement has reviewed the work and retest documentation and considered it satisfactory. Assuming issuance of an Amendment to License deleting reference to CEDM 44 this-matter is fully resolved and will not be discussed in .,the supplementary Startup Report(s).

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Page 124 7.3 Apparent Low Reactor Cooling Pump Flow During Post Core, Load Hot Functional Testing, Reactor Coolant System Flow was determined, using Reactor Cooling Pump AP.

The original nominal design 100% flow was 324,800 gpm total.

The FSAR, safety analysis and Technical Specifications assume a total flow of 370,000gpm (113.92%). Actual measured flow was about 5% less than that needed to guarantee the Technical Specification required flow.

Considering this 5% reduction in flow, Florida Power & Light Company requested a license limit of 90% power (10% reduction) to allow continuation of testing during resolution of the problem. Amendment 5 to St. Lucie License DPR-67 allowed operation to 64 power as an interim measure. An Order for Modification of License, dated June 17, 1976, addressing vendor Strikin II computer code errors as well as the flow uncertainty, amended this limit to 90% power.

During Power Ascension Testing, a flow determination by calori-metric means was performed at 80% power. Actual flow was determined to be 123.1% (399,800 gpm) of nominal design 100%.

flow. Since measurement uncertainties are about the same using this method or using pump AP it is evident that actual flow is in fact proper to allow operation at 100% power. It should be noted that the AP method of flow determination is obviously dependent upon pump geometric configuration and can be signifi-cantly affected by minor variations in inner casing surfaces, test tap orientation and internal clearances. It is felt that this is the root. cause of the initial low measured flow. A calorimetric determination is, of course, independentof pump geometry.

Based on the above information, a request for license amendment has been submitted. to the NRC. The request also includes a detailed error analysis of the flow determination by calorimetric means, and other supporting information such as discussion of results from similar, plants.

Florida Power & Light Company feels this matter is now resolved.

Assuming favorable response from the NRC on the request for license amendment, this item will not be discussed in the supplementary Startup Report(s).

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Page 125 7.4 Hi her Than Predicted Contairimerit Radiation Levels.

Dose rates at selected points, both inside and outside the Rad-iation Controlled Area, were determined during Power Ascension Testing. These results indicated that the 50% levels could be extrapolated to 100% to give preliminary verification of an expected dose rate problem. 100% power dose rates have not yet been determined as the unit has not yet operated at 100% (see section 7.5) A dose rate problem was expected due to problems at similar plants.

Extrapolated results indicated 3 potential problem areas. These were the doors to the shield building/containment annulus and the containment personnel airlock. Dose rates (combined neutron and gamma) were 3 mrem/hr and 10 mrem/hr at the southwest and northeast annulus doors respectively. Dose rates were 1 mrem/hr outside the airlock, '22 mrem/hr inside the airlock and 80 mrem/hr inside containment. at the airlock. In addition, within containment, certain areas of the operating deck area (which should be accessible if necessary during operation) had estimated dose rates of up to 60 rem/hr (extrapolated).

This neutron streaming is primarily due to the large annular gap between the reactor vessel and the primary shield wall. This gap is to provide a vent path for a postulated LOCA but also provides a means for neutrons to scatter from the core midplane area.

These then scatter from the containment ceiling causing the excessive dose rates.

As this pxoblem was anticipated, a preliminary sheilding design has been prepared (see section 6.7). However, until 100% power dose rates are known, final evaluation of this design and a proposal to the Nuclear Regulatory Commission cannot be performed. Final resolution of this item will be discussed in our followup Startup Report(s).

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Page 126 7.5 Power Distribution Anomal On June 30, 1976, with the reactor at 80% power and all control rods out, a routine power distribution map gave the first in'di-cation of a small azimuthal power tilt. This was attributed at that time to detector errors or failure. It should be noted that at this time Technical Specifications for tilt and total radial testing in accord-peaking factor (FT) r were suspended for physics ance with the special test exceptions of the Technical Specifica-tions.

Within the next week a few incore alarms were received. During evaluation of these, it was found that the calculated alarm set-points were in error (LER 335-76-34, August 6, 1976)-and it was also determined that the previously indicated tilt was still pre-sent. The alarms were corrected. On July 6, 1976, plant power was reduced to 50% for routine cleaning of a condensate pump strainer. While at 50% power, it was determined conclusively tilt of approxi-.

from the power distribution that an azimuthal mately 4% was present along with an axial peaking value of 1.5, as compared to an expected value of C 1.35. This tilt was Technical veri-Speci-fied using the moveable incore detector system.

fications for tilt and total radial peaking factor (FT) r were reinstated. It should be noted heie that at no time was the plant in violation of any Technical Specification regarding azimuthal tilt or peaking. (LER 335-76-35, July 23, 1976)

On July 13, reactor power was reduced to about 10 % and a low power physics test program commenced. This program was a repeat of selected protions of the LPPT performed after initial startup.

At this time two theories were offered as possible explanations:

1) a selective deposition of crud on the fuel leading to local flow maldistributions; and 2) early burnout of the burnable poison pins in the fuel assemblies. The results of these tests (avail-able July 18) verified'hat the tilt was present, and that the core was more reactive (about .45%) than predicted. This second finding tended to support the early poison pin burnup theory. It was decided to open the reactor'vessel for inspections and a shut-down/cooldown was commenced. Over the next week, many discussions were held, data was reduced and evaluated and theories postulated.

None of this information could conclusively explain the existing phenomenon; therefore, on July 27, actual disassembly of the reactor began.

Representative fuel assemblies were removed from various areas of the vessel and inspected. The crud buildup theory was quickly dispensed with, as blisters and perforations were found on the poison pin cladding. More fuel assemblies were removed and in-".

spected. Sufficient flaws were found to statistically demonstrate that there was a core wide problem with the cladding of the burnable poison pins. It should be noted here that no evidence was noted of any fuel pin anomalies. Due to the core-wide poison pin problem, the plant was defueled.

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Page 127 After the discovery of these poison pin cladding failures, a new theory was postulated. This was that the failure allowed the boron within the rods to wash out and be lost or to migrate and redistribute within the poison pins. This boron loss/redistri-bution theory correlated much better than any other theories previously considered.

It was then necessary to resolve two major concerns: 1) what caused the cladding failure and 2) what must be done to return the plant to power operation. To aid in resolving the first concern, poison pins were removed from selected fuel assemblies.

These were submitted to on-site visual and eddy current testing.

Then they were sent to research laboratories to determine the cause of the cladding failures, the mechanisms of boron loss and redistribution, and verification that loss of boron had occured in some pins and that it could cause the observed results.

As a result of these laboratory/test reactor inspections, the cause of the failure was confirmed to be hydriding of the zircalloy cladding of the pins. This was caused by excessive moisture

!content within the pins. Under incore conditions of high temp-erature and neutron flux the moisture produced free hydrogen which attacked the cladding. It was proven that the perforations did result in loss/redistribution of boron from the affected poison pins under incore conditions. And, it was confirmed that this loss/redistribution of boron could create the conditions observed at the St. Lucie Plant.

Then, regarding resolution of the second concern, it was determined that on site replacement of the poison pins with new ones of much lower moisture content was the appropriate solution. At the time this decision was made, some of the pin removal equipment hadthere already been proven in removal of the pins for testing, So, was reasonable assur. ance the job could be done even though it had to be done under water in the spent fuel pool. The vendor s specifications and controls on moisture content were significantly tightened to avoid repetition of the original problem. Replacement of the poison pins resulted in fuel assemblies virtually iden-tical to the original ones except for minor fuel depletion (burnup).

Actual reconstitution (removal of old pins and installation of new ones) commenced on October 5, 1976. The basic procedure consisted of: drilling or cutting the flow plate webs above the poison old pins; cleaning and deburring the newly machined surfaces; removing pins using a template to ensure fuel pins were not removed; in-stalling new pins; installing a retention assembly over the flow plate; and final QC and FPSL acceptance inspection. This process is described in greater detail in the FP&L submittals leading up1976, to Ammendment 10 to St. Lucie License DPR-67, dated 3 December which allows resumption of power operations.

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Page 128 By November 3, 1976 this process was close to complete and core reload was commenced. By November 7, all but 2 assemblies were completed and on November 10, the last of the 108 assemblies had been reconstituted and core reload was continuing (supplementary LER 335-76'-35, December 17, 1976).

We have now resumed power ascention testing and thus far have seen no evidence of any anomalies except those directly related to the uneven fuel depletion (burnup) which resulted from 'the power tilt/

peaking. These have been minor in magnitude and should be self-correcting as plant operation (and fuel depletion) continue. The activit'ies after fuel reconsititution (fuel reload, initial criticality etc). will be described in our supplementary Startup Report(s).

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