Text

Supplementary Startup Test Report
	ML18088A919
Person / Time
Site:	Saint Lucie
Issue date:	05/31/1977
From:	; Florida Power & Light Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML18088A919 (136)
	v • d • e

FLORIDA POWER AND LIGHT COMPANY ST. LUCIE NUCLEAR POWER PLANT UNIT 1 gackot+ 0 4- -~ ~F00 Qggtg) 0 11 l Oa(~5'f 90cumen

- h<GuvTG";(9 CKH FiH DOCKET NP, gP 335 LICENSE NO. DPR-67 SUPPL22iENTARY STARTUP TEST REPORT MAY, 1977

.0

TABLE OF CONTENTS SECTION PAGE

1.0 INTRODUCTION

AND

SUMMARY

1.1 1.2 1.2.1 1.2.2 1.2.3 1.2.4 1.

2.5 INTRODUCTION

SUi%fARY CORE RELOAD POST CORE HOT FUNCTIONAL TESTS APPROACH TO CRITICALITY LOW POWER PHYSICS TESTS POWER ASCENSION TESTING 1

2 2

3 3

2.0 3.0 CORE RELOAD POST CORE HOT FUNCTIONAL TESTS NONE INCLUDED IN THIS REPORT.

4.0 APPROACH TO CRITICALITY 17 5.0 LOW POWER'HYSICS TESTS 26 5.1 5.2 5.3 5.4 5.5 5.6 5.7 5.8 REACTIVITY COMPUTER RESPONSE CHECK PURPOSE TEST RESULTS CONCLUSIONS CEA SYMMETRY PURPOSE TEST RESULTS CONCLUSIONS CRITICAL BORON CONCENTRATION MEASUREMENTS PURPOSE TEST RESULTS CONCLUSIONS TEMPERATURE COEFFICIENT OF REACTIVITY MEASUREMENTS PURPOSE TEST RESULTS CONCLUSIONS NON-OVERLAPPED REGULATING CEA GROUP WORTH MEASUREMENTS PURPOSE TEST RESULTS CONCLUSIONS OVERLAPPED REGULATING CEA GROUP WORTH MEASU1UBKNTS PURPOSE TEST RESULTS CONCLUSIONS CEA 7-1 WORTH MEASUREMENT GROUP 7~ 100" PURPOSE TEST RESULTS CONCLUSIONS SHUTDOWN K&GIN VERIFICATION PURPOSE TEST RESULTS CONCLUSIONS 28 31 32 36 38 41 44

SECTION 6.0 6.1 6.2 6.3 6.4 6.5 6.6 6.7 6.8 6.9 6.10 6.11 6.12 POWER ASCENS ION & POST 100%

TRANSIENT TESTS FLUX DISTRIBUTION MONITORING PLANT POWER CALIBRATION PURPOSE TEST RESULTS CONCLUSIONS FIXED INCORE DETECTOR ALARM SETPOINTS PURPOSE TEST RESULTS CONCLUSIONS MODERATOR TEMPERATURE COEFFICIENT AND POWER COEFFICIENT PURPOSE TEST RESULTS CONCLUSIONS TURBINE VALVE TESTS PURPOSE TEST RESULTS CONCLUSIONS PARTIAL LOSS OF FLOW TEST PURPOSE TEST RESULTS CONCLUSIONS POWER DEFECT AND XENON WORTH AFTER SHUTDOWN PURPOSE TEST RESULTS CONCLUSIONS TOTAL LOSS OF FLOW/NATURAL CIRCULATION TEST PURPOSE TEST RESULTS CONCLUSIONS STEAM GENERATOR FEEDWATER HAMMER TEST PURPOSE TEST RESULTS CONCLUSIONS LOSS OF OFF-SITE POWER AND LOAD REJECTION PURPOSE TEST RESULTS CONCLUSIONS 10% LOAD REDUCTION-TURBINE RUNBACK PURPOSE TEST RESULTS CONCLUSIONS TURBINE TRIP TEST PURPOSE TEST RESULTS CONCLUSIONS PAGE 45 49 51 55 56 57 60 61 62 64

SECTION

6. 13

6. 14 6.15 6.16 6.17

6. 18 6.19 PAGE 66 74 GENERATOR TRIP TEST PURPOSE TEST RESULTS CONCLUSIONS LOAD CYCLE TEST 67 PURPOSE TEST RESULTS CONCLUSIONS STATIC CEA DROP 69 PURPOSE TEST RESULTS CONCLUSIONS DYNAMIC CEA INSERTION 72 PURPOSE TEST RESULTS CONCLUSIONS AUTOMATIC CONTROL SYSTEM CHECKOUT STEAM GENERATOR LEVEL CONTROL CEA REGULATING SYSTEM AUTOMATIC TURBINE CONTROL AND LOAD SWING TEST 73 PURPOSE TEST RESULTS CONCLUSIONS NSSS ACCEPTANCE RUN PURPOSE TEST RESULTS CONCLUSIONS SHIELDING EFFECTIVENESS AND PLANT RADIATION LEVEL MEASUREiMENTS 75

6. 20 6.21 6.22

6. 23

6. 24 PURPOSE TEST RESULTS CONCLUSIONS CHEMISTRY & RADIOCHEMISTRY TESTS AT POWER PURPOSE TEST RESULTS-RCS TEST RESULTS-SECONDARY SYSTEMS CONCLUSIONS TOTAL RADIAL PEAKING FACTOR PURPOSE TEST RESULTS CONCLUSIONS TURBINE OVERSPEED TRIP TEST PURPOSE TEST RESULTS CONCLUSIONS XENO'OLLOW PURPOSE TEST RESULTS CONCLUSIONS EFFLUENT MONITOR CORRELATION PURPOSE TEST RESULTS CONCLUSIONS 77 80 81 82 87

SECTION PAGE

6. 25 7.0 7.1 DDPS CALORIMETRIC AND DDPS SNAPSHOT PURPOSE TEST RESULTS CONCLUSIONS COUNTS ON ORIGINAL STARTUP REPORT RCS FLOW COASTDOWN 88 90 91

Page 1

1.0 INTRODUCTION

AND

SUMMARY

Introduction This report fulfills the Requirements of Regulatory Guide 1.16 which states that a Startup Test Report will be sub-mitted to the NRC within 9 months of initial criticality and every three months thereafter until the unit has been declared commercial and startup testing is completed.

Ini-tial Criticality was April 22, 1976.

The initial Startup

Report, submitted January 21, 1977, covered the Start Test

Program, thru the 50% power level of power ascension testing.

As described in that report, testing was stopped in July,

1976, due to a flux anomaly which required replacement of the burnable poison pins in the fuel assemblies.

This re-port covers the core reload, return to 50% power, the re-mainder of power ascension and completion of startup test-ing.

Also, the unit was declared commercial at 0001, De-cember 21, 1976.

Startup testing was essentially completed on March 24, 1977 with the completion of the Nuclear Steam Supply System warranty run.

The Startup Test Program was organized and administered by Florida Power

& 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 Techni-cal Specifications, Section 6.5.1.

Any test'esults fal-ling outside of acceptance criteria were resolved prior to beginning the next test phase.

The test phases were as follows:

(1)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

Core Reload.

Post Core Load Hot Functional Tests.*

Approach to Criticality.

Low Power Physics Tests.

Escalation to Power Tests 50% Plateau.

Escalation to Power Tests 80% Plateau.

Escalation to Power Tests 100% Plateau.

Post 100% Plateau (Transient)

Tests.

None included in this report.

Page 2

1.0 INTRODUCTION

AND

SUMMARY

(cont) 1.1 Introduction (cont)

Kmimum licensed reactor core power level (100%) is 2560 Mwt.

The Startup Test Program recommenced November 4, 1976, with the loading of the first fuel assembly into the reactor vessel and was completed March 24, 1977.

FPGL re-quested and received permission from Inspection and Enforce-

ment, Region II, to submit this report by May 30, 1977, rather than April 22, 1977.

This allowed time to include in this report all testing not previously reported and allows both FPL and the NRC to close out the startup phase of St. Lucie 81 without another supplementary report.

l. 2 Summary 1.2.1 Initial Fuel Load Fuel loading commenced on November 4, 1976, and was completed on November 15, 1976.

A sizable portion of this time was spent in non-fuel loading activities.

The largest period of non-fuel loading time (approxi-mately 41% of total time) was associated with in-specting fuel elements B-069 and A-015, and A-049 and B-001, which came into physical contact with each other (B-069 into A-015, etc.).

None of the inspected fuel assemblies showed visible signs of damage.

Fuel loading was done 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day by three crews working 8-hour shifts.

Without any delays or equip-ment problems, an experienced crew could load 12 to 14 assemblies per shift.

1.2.2 Post CORE LOAD Hot Functional Tests All post core load hot functional testing was com-pleted during initial startup and reported in the first Startup Report.

1.2.3 Initial Approach to Criticality.

The initial approach to priticality commenced on December 2, 1976.

The reactor was declared criti-cal on December 4, 1976.

The CEA's were withdrawn, followed by a slow RCS dilution to criticality.

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

The only problems of consequence encountered were with the CEDM System.

CEA 1-29 "dropped" four different times during dilution.

Dilution was secured and CEA 1-29 was retrieved each time. This problem has been resolved and has not reoccurred.

0

Page 3

1.0 INTRODUCTION

AND

SUMMARY

(cont)

1. 2 Su~ru '(cont) 1.2.4 Low Power Physics Test 1.2.5 The Low Power Physics Test (LPPT) phase commenced on December 4, 1976.

The LPPT phase was completed on December 9, 1976.

There were no significant delays or occurrences.

Most LPPT measurements were in close agreement with predictions and all were within acceptable limits.

Power Ascension Testing Power Ascension Testing began December

9. 1976.

The program consisted of a 2-week ascent to 50%

power, about 2 weeks at 50% power, ascent to 80%

power, ascent to 90% power, and when the license was amended, ascent to 100% power.

The slow as-cent to 50% and the hold at that power were to allow ample time for monitoring of in-core flux distribution to verify that the flux distribution anomaly no longer existed.

The results of this monitoring'were satisfactory.

After initial op-eration at 100% power, the two 100% trip 'tests were performed.

Testing was then performed at 100%

steady state followed by plant response (transient) testing.

The last test was the 200 hour0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months NSSS accept-ance run.

Page 4

2.0 CORE RELOAD At 0010 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months on November 4, 1976 fuel assembly number B-009 contain-ing neutron source number 1, step number 1 (of Appendix A of Operating Procedure

1600022, Revision 3), was loaded into core location X-11.

Fuel loading was completed at 2215 hours0.0256 days 0.615 hours 0.00366 weeks 8.428075e-4 months on November 15, 1976 when fuel element number C-010, step number 217, was loaded into core loca-tion A-S.

Table 2 0-1 and Figure 2 0-> show the fuel loading sequence.

Figures 2 0-2 and 2'-3 show fuel assembly location and CEA location by their respective serial numbers.

It should be noted that in the original initial core load (cycle 1) fuel assembly C-112 containing CEA 64 was in core location J-20, and fuel assembly C-107 was in core location G-20.

In this initial core load (cycle 1A) fuel assembly C-107 contain-ing CEA 64 is in core location J-20, and fuel assembly C-112 is in core location G-20.

This change was done per Combustion Engineering Com-pany's request because fuel assembly number C-112 contained a slightly depressed guide tube.

The core desiygn characteristics are the same as for cycle 1, and are available in Table 2.0-2 of the original Startup Test Report (dated

December, 1976).

Neutron count rate was monitored during fuel loading on two separate detectors, Wide Range Log Channel C and Wide Range Log Channel D.

In-dependent plots of inverse count rate versus the number of fuel assemblies loaded were maintained to ensure the reactor remained subcritical at all times during fuel loading.

Fuel loading was conducted with the spent fuel pool full of borated water.

A refueling boron concentration of > 1720 ppm boron was maintained with shutdown cooling flow through the core in accordance with the Techni-cal Specifications.

Fuel loading took approximately 290 hours0.00336 days 0.0806 hours 4.794974e-4 weeks 1.10345e-4 months .

A few problems occurred, two of which will be addressed at this time.

The fuel handling equipment malfunctioned numerous times.

Florida Power and Light Company's main-tenance department repaired the upender, refueling machine and new fuel machine as required.

Equipment down time resulted in a loss of approxi-mately 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

During movement of the fuel bundles within the core area fuel assembly B-069 came into physical contact with fuel assembly A-015, and fuel assembly A-049 came into physical contact with fuel assembly B-001.

The subsequent removal, inspection and replacement of these four assemblies resulted in an approximate 120-hour delay.

Miscel-laneous other delays accounted for another ll hours.

,Page 5 TABLE

2. 0-1 FUEL ASSEMBLY LOADING SE UENCE STEP NO.

FUEL ASSEMBLY NO.

B009 C031 C032 C105 B067 A025 CEA NO.

Source 1

40 72 CORE LOCATION X-11 Y-10 Y-12 X-13 W-13 W-ll 10 B057 C104 C007 C114 A002 25 W-9 X-9 Y-8 X-7 12 14 15 16 17 18 19 20 22 B042 A010 B017 A038 B068 A004 C106 C023 C027 B080 A013 C016 B020 30 39 47 18 V-7 V-9 V-ll V-13 V-15 W-15 X-15 Y-14 X-6 W-6 V-6 X-16 W-16

Page 6

.TABLE 2.0-1 (Cont.)

FUEL ASSEMBLY LOADING SE UENCE STEP NO.

25 26 FUEL ASSEMBLY NO.

A015 C008 CEA NO.

51 CORE LOCATION V-16 X-5 27 28 C212 B046 W-5 V-5 29 30 31 32 C001 C208 B037 C012 X-17 W-17 V-17 W-4 33 34 35 36 C015 C202 C034 C036 C205 67 69 V-3 V-4 W-18 V-19 V-18, 38 39 40 41 42 43 45 C013 C204 B049 A019 B069 A048 B010 A049 63 46 T-20 T-19 T-18 T-17 T-16 T-15 T-13 T-11 46 B001 T-9 47

~

~s A008 B012 24 T-7 T-6

Page 7

TABLE 2.0-1 (Cont.)

FUEL ASSEMBLY LOADING SE UENCE STEP NO.

49 50 52 FUEL ASSEMBLY NO.

A018 B043 C002 C210 CEA NO.

CORE LOCATION T-5 T-4 T-2 T-3 53 C035 S-2 54 B074 S-3 A014 10 56 B013 A054 17 S-5 S-6 58 B045 S-7 A055 S-9 60

~

si 62 B026 A061 B035 S-11 S-13 S-15 63 A053 50 S-16 64 B071 S-17 65 A020 59 S-18 66 B039 S-19 67 C033 S-20 68 C109 R-20 69 70 A028 B078 62 R-19 R-18 71

~

~z A031 B007 54 R-17 R-16

0

Page 8

TABLE. 2.0-1 (Cont.)

FUEL ASSEMBLY LOADING SE UENCE STEP NO.

73 74 75 76 FUEL ASSEMBLY NO.

A037 B004 A030 B038 CEA NO.

45 34 CORE LOCATION R-15 R-13 R-ll R-9 77 A059 23 R-7 78 B040 R-6 79 A056 13 R-5 80 B041 R-4 81 A022 R-3 82 C102 R-2 83 C021 P-1

~

ss 86 C116 B029 A040 N-2 N-3 N-4 87 B064 N-5 88 A064 N-6 89 B056 N-7 90 A034 29 N-9 91 B051 N-ll 92 A057 38 N-13 93 94 B065 A060 N-15 N-16 95 B077 N-17 A009 58 N-18

Rage 9

TABLE 2.0-1 (Cont.)

FUEL ASSEMBLY LOADING SE UENCE STEP NO ~

FUEL ASSEMBLY NO.

CEA NO.

CORE LOCATION 97 98 B008 C111 65 N-19 N-20 99 100 C004 C028 P<<21 M-21 101 B023 L-20 102 103 104 105 106 107 108

~

109 110 112 113 114 A026 B006 A043 B044 A068 B054 A032 B025 A041 B053 A044 B036 A024 73 44 22 70 L-19 L-18 L-17 L-16 L-15 L-13 L-ll L-9 L-7 L-6 L-5 L-4 L-3 115 116 117 B079 C011 C030 L-2 M-1 118 119

~

120 C005 C110 B047 H-1 J-2 J-3

.TABLE 2.0-1 (Cont.)

FUEL ASSEMBLY LOADING SE UENCE Page 10 STEP NO.

FUEL ASSEMBLY NO.

CEA NO.

.CORE LOCATION 121 A011 J-4 122 123 124 125 126 B059 A058 B058 A050 B055 28 J-5 J-6 J-7 J-9 127 128 129 130 131 132 134 135 136 137 138 139 140 141 142 143

~

l44 A045 B028 A063 B061 A006 B076 C107 C025 C022 C112 A005 B018 A042 B031 A051 B033 A052 B070 37 57 32 J-13 J-15 J-16 J-17 J-18 J-19 J-20 K-21 H-21 G-20 G-19 G-18 G-17 G-16 G-15 G-13 G-ll G-9

0 0

TABLE 2.0-1 (Cont.)

,Page FUEL ASSEMBLY'LOADING'SZ UENCE STEP NO.

FUEL ASSEMBLY NO.

CEA NO.

CORE LOCATION 145 146 147 148 A039 B005 A035 B060 12 G-6 G-5 G-4 149 150 151 152 153 154 155 156

~

~sr 158 159 160 161 162 163 164 165 166 167 168 A033 C103 C040 B066 A016 B052 A065 B072 A067 B011 A066 B015 A069 B063 A021 B034 C039 C024 C211 B019 16 49 56 60 G-2 F-2 F-3 F-5 F-6 F-7 F-9 F-ll F-13 F-15 F-16 F-17 F-18 F-19 F-20 E-20 E-19 E>>18

TABLE 2.0-1 (Cont.)

Page 12 FUEL ASSEMBLY LOADING SE UENCE STEP NO.

FUEL ASSEMBLY NO.

CEA NO.

CORE LOCATION 169 170 A062 B016 E-17 E-16 171 A036 42 E-15 172 B003 E-13 173 A012 E-ll 174 B032 E-9 175 A047 20 E-7 176 B002 E-6 177 A027 E-5 178 B030 179 C014 E-2 180

~

~si 182 183 C207 C029 C203 B024 66 E-3 D-3 D-4 D-5 184 A029 15 D-6 185 B014 D-7 186 A046 27 D-9 187 188 189 190 191

~

ice B048 A007 B073 A017 B021 C038 36 48 D-ll D-13 D-15 D-16 D-17 D-19

TABLE 2.0-1 (Cont.)

page 13 FUEL ASSEtGLY LOADING SE UENCE STEP NO.

FUEL ASSEMBLY NO.

CEA NO.

CORE'LOCATION 193 194 195 196 C201 C020 C206 B062 68 D-18 C-18 C-17 C-16 197 198 199 200 201 202 203 204

~

205 206 A001 B075 A023 B022 A003 B027 C019 C209 C003 C017 41 71 19 C-15 C-13 C-ll C-9 C-7 C-6 C-4 C-5 B-5 207 208 209 210 211 212 C108 C115 B050 C009 C037 C113 26 Source 2

B-7 B-9 B-11 B-17 B-16 B-15 213 C101 35 B-13 214 C006 A-14 215

~

225 217 C026 C018 C010 A-12 A-10 A-8

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ST.

LUCIE PLANT, UNIT NO.

1 CONTROL ELEMENT ASSEMBLY SERIAL NUMBERS AND CORE LOCATIONS Page 16 Figure 2.0-3

.NORTH 21 63 62 65 73 64 60 20 19 69 59 58 57 68 55 51 50 53 49 48 16 40 S-l 31 47 72 25 39 30 46 24 34'3 38 29 44 33 22 37 28 43 32 21 42 20 36 27 71 19 S-2 26

15 14 13 12ll 10 9

8 14 18 17 13 A

/

12 16 67 10 70 66 I

I Y

X W

V T

S R

PN ML KJ HG F

E D

C B

A Normal CEA Orientation:

Serial

// on NW Web Exception:

CEA's 70, 73, F and G have Serial

// on SW Web

Page 17 APPROACH TO CRITICALITY FOLLOWING CORE RELOAD Criticality was achieved on December 4,

1976 at Reactor Coolant System (RCS) conditions of 533oF and 2263 psia.

The initial RCS boron concentration was 1760 ppm.

The Approach to Criticality began by withdrawing the CEA's in specified increments 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 890 ppm.

The Initial Criticality Following Refueling procedure limits RCS boron concentration dilution below 800 PPH.

The procedure cautions that "criticality shall be anti-cipated whenever CEA's are being withdrawn or boron dilution opera-tions are in progress".

Throughout the approach to criticality, two (2) independent sets of

'nverse 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 hours is shown in Figures 4.0-3 and 4.0-4.

After achieving criticality, Control Element Assembly (CEA) Group 7 was used to control neutron flux.

Conditions were stabilized at 10-4~ power and the critical data shown in Table 4.0-4 was re-corded and compared with predicted values.

In summary, initial criticality following core reload was achieved in a safe and orderly fashion.

There was acceptable agreement be-tween the measured and predicted critical boron concentrations.

The predicted boron concentration at criticality was 837 ppm.

The actual value of 890 ppm was well within the PSL acceptance tolerance of + 100 ppm.

TABLE 4.0-1 Page 18 CEA WITHDRAWAL SEQUENCE

STEP CEA GROUP INCHES WITHDRAWN B

1/M D

1/M 8.8.la 8.8. lb 8.8.2a 8.8.2b 8.8.3 A

68 136 68 136

>132 1.040 0.980 1.050 1.010 0.970 0.992 1.020

1. 050 0.980 1.030 8.9.1 8.9.2 8.9.3 8.9.4 8.9.5 3

4 5

68 UEL

<54 122

<40

.,UEL

<107

<26 UEL

<93

<11

1. 010 0.980 1.020 0.990 0.790 1.010 1.010 1.000

0. 980 0.820 8.9.6 8.9.7 UEL

<79 UEL

<54 0.780 0.770 0.780

0. 790 8.9.8 122

<40 0.770 0.770 8.9.9 UEL 68 0.780 0.760

8. 9. 10 8.9.12 UEL 68 0.740 0.772 0.790 0.801

Steps IAC with O.P.

0030221 NOTE:

UEL = Upper Electrical Unit

FIGURE 4.0-1 Page 19 ST.

LUCIE UNIT 1 INITIALAPPROACH TO CRITICALITY CYCLE 1A, 532 F, 2250 PSIA WIDE RANGE LOG CHANNEL B RCS BORON > 1720 PPM 1.4 1.3 1.2 1.0 0.9 0.8 0.7 5o 0.6 0

0.5 R

0.4 0.3 0.2 0.1 0.0 a

C4 H

F)

A cV CO CO CO CO CO Ch Ch CO CO CO CO CO CO CO Fl Ch CO CO Ch Ch Ch M

Ch CO CO CO CO CO CO CO CO Ch CO CEA WITHDRAWAL STEPS IAC WITH O.P.

0030221

0

FIGURE 4.0-2 Page 20 ST. LUCIE UNIT 1 INITIALAPPROACH TO CRITICALITY CYCLE 1A, 532 F, 2250 PSIA HIDE RANGE LOG CHANNEL D RCS BORON

> 1720 PPM 1.4 1.3 1.2 1.0 0.9 0.8 0.7 o

0.6 0.5 0,4 0.3 0.2 0.1 0.0 cd W

C CV C4 CO CO CO CO CO Ch Ch

~

CO CO CO CO CO CO CO Ch Ch CO CO lA C)

Ch Ch CO CO CO Ch CO CO O

cv Ch Ch Ch CO CO CO CEA WITHDRAWAL STEPS IAC >11TH O.P.

0030221

0

Page 21 TABLE 4.0-2 DILUTION TIME (MINUTES) 20 50 R.C.S.

BORON CONCENTRATION 1770'776 CHANNEL B 1/M VALUE

'.000 0.

,885 QiANNEL D 1/M VALUE 1 000

0. 917 14 3

12 0

0.792

0. 759

0. 675

0. 563 0.7 0

0.6

0. 685 0.612

0. 576 1157 0.

2 494 519 539 559 579 599 619 629 639 659 980 967 950 937 934 903 894 890

0. 286

0. 270

0. 259

0. 237

0. 194 0.163

0. 100

0. 052

0. 283

0. 266

0. 248

0. 193 0.1 5

0.108 0.055

FIGURE 4.0-3 Page 22 ST. LUCIE UNIT 1 INITIALAPPROACH TO CRITICALITY CYCLE 1A, 532 F, 2250 PSIA WIDE RANGE LOG CHANNEL B CEA GROUP 7 AT 68 INCHES 1.4 1.3 1.2 1.0 0.9 0.8 0.7 P) 0.6 0

0.5 R

0.4 0.3 0.2 0'. 1 0.0 0

1 2

3 4

5 6

7 8

9 10 ll 12 DILUTION TIME (HOURS)

0

FIGURE 4.0-4 Page 23 ST.

LUCIE UNIT 1 INITIALAPPROACH TO CRITICALITY CYCLE 1A, 532 F, 2250 PSIA WIDE RANGE LOG CHANNEL D CEA GROUP 7 AT 68 INCHES 1.4 1.3 1.2 1.0 0.9 0.8 0.7 o

0.6 0.5 M

0.4 0.3 0.2 0.1 0.0 0

1 2

3 4

5 6

7 8

9 10 11 12 DILUTION TIME (HOURS)

Page 24 TABLE 4.0-3 ST. LUCIE UNIT 1 CYCLE 1A RCS BORON CONCENTRATION VS DILUTION TIME 1800 1700 1600 1500'400 1300 1200 1100 1000 900 800 0

1 2

3 4

5 6

7 8

9 10 11 12 DILUTION TBK (HOURS)

0

Page 25 TABLE 4.0-4 PARAMETER INITIAL CONDITION MEASURED VALUE RCS TEMPERATURE ( F)

RCS PRESSURE (PSIA)

RCP'S OPERATING 532 2250 533 2262.5 CEA GROUPS WITHDRAWN INCHES A

136 136 136 136 P-1 136 136 P-2 136 136 136 136 136 136 136 136 136 136 136 136 136 136 68 68 PREDICTED MEASURED RCS BORON CONCENTRATION (PPM) 837 890

Page 26 5.0 LOW POWER PHYSICS TESTS LPPT)

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 conjunction 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 reactivity 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.

All raw test data was collected,

reduced, and analyzed on site.

In all cases, measured data met applicable acceptance criteria.

CE, Windsor provided backup support for data analyses.

LPPT was completed satisfactorily during the original plant startup.

Due to the flux distribution anomaly and fuel assembly poison pin replacement described in the original report, it was deemed prudent to repeat selected portions of LPPT.

This testing and the flux monitoring (Section 6.1) performed during the slow ascent to power and throughout the test program confirmed that the flux distribution anomaly had been resolved satisfactorily.

Page 27 TABLE 5.0-1 LOW POWER PHYSICS TEST SCHEDULE Reactivit Com uter Res onse Check CEA Symmetry Unrodded Critical Boron Concentration Isothermal Temperature Coefficient (CEA Group 7 approximately 115 inches)

CEA Group 7, 6 and 5 worth (non overlap)

Isothermal Temperature Coefficient (CEA Group 5 approximately 10 inches)

CEA Group 4, 3 and 2 worth (non overlap)

Isothermal Temperature Coefficient (CEA Group 1 approximately 126 inches)

Sequential Worth of CEA Groups 2, 3, 4, 5,

6 and 7 (overlap mode)

Center CEA Worth Measurement, GP7=100"

Reference:

Zero Power Physics Tests After Reload Preoperational Procedure No. 0110097

Page 28

5. 1 REACTIViTY COMPUTER RESPONSE CHECK Purpose The reactivity computer is an analog computer that calculates reactivity by solving the kinetics equations of reactivity.

This exercise was not necessarily a test, but rather a verification of appropriate constants, and of equipment functioning properly. It was also necessary to know what flux (power) levels corresponded to the operability range of the scale on the reactivity computer, and this correlation check was performed at this time, in the Low Power Physics Test Schedule.

Test Results

-2.

The reactor was critical at a power level

< 10 Reactivity was being controlled by CEA group 7.

A small amount of reactivity was introduced into the core, by group 7 withdrawal.

The reactor period was calculated using the equation T

At

-.'ln (P/Po).

This was repeated a couple of times, and the measured reactivity (from the trace generated on the chart of the reactivity computer) was compared to the design reactivity (using calculated T and design curve hp).

The Lower Power level was determined by decreasing power until the trace on the pen on the chart recorder on the reactivity computer was "Noisier" than 0.3c, peak to peak.

The upper limit was below the point of adding nuclear heat.

Conclusions The reactivity computer did respond properly.

The acceptance criteria (correction factor 1.00 + 0.10) was satisfied.

The lower and upper power levels were defined.

Page 29 TABLE 5, l-l REACTIVITY COMPUTER REACTIVITY RESPONSE CHECK REACTOR PERIOD**

216 Seconds 222 Seconds 345 Seconds STRIP CHART hp 4.550 4.560 3.000 DESIGN CURVE hp 4.800 4.650 3.200 CORRECTION FACTOR*

1.055 1.020 1.067

CF = DESIGN hp /

Recorder hp

~ T ~ ht /

1n (P/Po)

Page 30 TABLE 5. 1-2 REACTIVITY COMPUTER LOWER, MIDRANGE AND UPPER POWER LEVEL WIDE RANGE RPS AVERAGE POWER

(%)

0.500 X 10 1.500 X 10 2.250 X 10 KIETHLY INPUT

'I-9 (AMPS) 0'20 X 10 0.058 X 10 0.100 X 10 KIETHLY INPUT NI-10 (AMPS) 0.016 X 10 0.042 X 10 0.070 X 10 REACTIVITY COMPUTER STRIP CHART 30 65 100 CEA GP 7 POSITION (INCHES'ITHDRAWN) 93 93 91

Page 31 5.2 CEA SYMMETRY Purpose After fuel loading and reactor vessel reassembly, it is necessary to insure CEA's are coupled to their respective CEDM shafts.

During this

test, the reactivity worths of the symmetric FLCEA's (full length CEA's) were compared to each other.

That is, one CEA's worth was compared to the worth of another CEA that is symmetric to it.

This was done to insure uniform reactivity control within a CEA group.

The worthiest (according to design calculations)

CEA was measured at this time, also.

The PLCEA's (partial length CEA's) were tested for operability only.

Test Results The initial conditions before starting this test were reactor critical, with CEA group 7 approximately 95 inches withdrawn.

The first designated CEA was diluted to LEL (lower electrical limit) and CEA group 7 adjusted to zero reactivity.

By rod swapping, that is inserting another CEA and withdrawing the CEA that was in the core, keeping reactivity and reactor power within limits, without moving CEA group 7, the relative worth of the newly inserted CEA can be compared to the relative worth of the newly withdrawn CEA.

As different symmetrical CEA groups are compared, the first CEA is inserted to LEL, and reactivity zeroed using CEA group 7.

The remaining CEA's in that group are then swapped and compared to each other.

The worthiest, by design calculations, CEA was B-7.

The worth was measured by, diluting B-7 to LEL, measuring reactivity, as it was being stepped into the core.

The worth was also measured by borating B-7 out.

The two values were averaged, and the resultant worth was used as the test result.

Each individual PLCEA was inserted to 120 inches and withdrawn.

This was done to verify that negative reactivity was being inserted into and withdrawn from the core.

Conclusions The acceptance criteria for the test were met with one exception.

The CEA symmetry check revealed 2 CEA's with relative worth more than speci-fied when compared to its symmetric "sisters."

CE has evaluated these results and verifies that their relatively lower worth is due to their location in the area of greatest fuel depletion.

(They are in the high power area of the previous core tilt which led to fuel reconstitution.)

The FRG concurs with this analysis and accepts all the results of this test.

FP<L Reactor Engineering also agrees with CE's analysis.

The highest worth CEA, B-7, was worth 0.158

%Ap, (acceptance criteria calls for equal to or less than 0.231 %Ap).

8

Page 32 5.3 Critical Boron Concentration Measurements Purpose Critical boron concentration measurements were performed at various CEA positions at relatively constant RCS temperature and pressure.

The purpose of these measurements was to ob-tain an as-measured value for the excess reactivity loaded in the core and to provide basis for verification of pre-dicted CEA Group reactivity worths.

Test Results Boron concentration values were averages of multiple 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 (di-lutes)

CEA's out near UEL* (in near LEL**). After RCS con-ditions 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 reactivity change (reactivity being plotted on a recorder of a reactivity computer) is measured.

The amount of reactivity added (sub-tracted) is converted, via boron worth, to an equivalent l PPM.

This I PPM is added to (subtracted from) the measured boron concentration.

This technique gives a safe, fast and accurate method of determining critical boron concen-trations at hard to achieve CEA positions (relatively low reactivity worths at UEL's and LEL's).

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

Page 33 TABLE 5.3-1 CRITICAL BORON CONCENTRATION AT VARIOUS CEA POSITIONS CEA CONFIGURATION 5 LEL 2 LEL 7

100" MEASURED CBC (PPM) 912 788 540 907 PREDICTED CBC (PPM) 866 742 504 RCP' OPERATING RCS T~ERATURE ('F) 532 532 533 533 RCS PRESSURE (PSIA) 2259

" 2265

2260 2260 CEA GROUP Pl

POSITION, UEL UEL P2 UEL UEL UEL UEL UEL UEL UEL UEL UEL LEL UEL UEL UEL LEL UEL UEL UEL LEL LEL 6

LEL LEL LEL LEL 100" 2104 1701 1247 1113 DATE 12/6/76 12/8/76 12/8/76 12/9/76

Page 34 5.4 Tem erature Coefficient of Reactivit Measurements 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 coefficient cannot be measured directly but it can be derived from a measurement of the isothermal temperature coefficient.

Test Results Isothermal temperature coefficient measurements were conducted at several different CEA withdrawal configurations and boron concentrations.

Measured values for each condition are the re-sult 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 feedback.

Re-actor 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 measurements and com-parisons with predicted valves.

Agreement between measured and predicted values indicates acceptance criteria has been met.

Technical Specification 3.1.1.4a specifies that the moderator temperature coefficient shall be less positive than +0.5 x 10 4 L k/k/ F whenever power

< 70%.

The FSAR accident analysis assumes various values within the range of +0.5 to -1.4 x 10"4 Ak/k/ oF for beginning of core life.

Conclusions For all cases, the measured values of isothermal temperature coefficient are within the FSAR and Technical Specifications acceptance criteria, and are therefore acceptable.

TABLE 5.4-1 Page 35 ISOTHERMAL TEMPERATURE COEFFICIENTS AT VARIOUS CEA POSITIONS RCS Boron, PPM 909 780 542 CEA Configuration (Desired) 78 115" 58 10" 18 126" Measured MTC*

+0. 26

-0.25

-0.90 Technical Specification MTC Limit* " <0.50 9<0. 50

<(0~~5 RCS Temperature, F

532 532 532 RCS Pressure, PSIA 2262 2260 2258 CEA Group Position +*

z3 Pl P2 151 l TIME u.s" 0226 2300 1516 DATE 7 Dec 76 07 Dec 76 SDec 76 X 10 4 AK/K

/oF L ~ Lower Electrical Limit, U ~ Upper Electrical Limit

Page 36

5. 5 NON-OVERLAPPED REGULATING CEA GROUP WORTH MEASUREMENTS Purpose During reactor operations, nearly all excess reactivity is held down by soluble boron concentration in the Reactor Coolant System and burnable poison shim rods in the fuel assemblies.

Additional hold down and reactivity control is provided by moveable Control Element Assemblies (CEA).

The CEA group worths were measured over the range of group 7 UEL to group 2 LEL, at the various cor-responding Reactor Coolant System boron concentrations.

Test Results All CEA Group reactivity worths were measured using a soluble boron swap method, either dilution or boration, to maintain criticality while inserting CEA Groups in increments.

The reactivity trace generated by this evolution was reduced to obtain the relationship between CEA Group positions from full in to full out and integral reactivity worth at these positions.

The results of this test were compared to the results of the identical test performed on cycle 1

(Ref. Start Up Report dated December 1976, Section 5.5) to insure similar CEA worths.

Conclusions Amendment 1O to the Facility License requires that "individual CEA group worths will be measured for CEA groups 7, 6, 5, 4, 3, and 2.

The measurement of the reactivity worth for regulating CEA group 1

and shutdown CEA group B will be made only if the previously measured worths of CEA groups 7, 6, 5, 4, 3, and 2 vary from the design cal-culations by more than the acceptance criteria.

The acceptance criteria are that the measured individual group worth varies from the calculated worth by less than either 10% or O.l %6K/K (whichever limit is larger) and that the measured cumulative worth of CEA groups 7 through 2 is within 10% of the calculated cumulative worths."

The acceptance criteria was met.

S'ee chart on next page.

TABLE 5.5-1 NON-OVERLAPPING CEA GROUP WORTHS CYCLE 1,

%5K/K CYCLE lA,

%6K/K CEA GROUP, DESIGN WORTH MEASURED ACCEPTABLE LIMIT LIMITS MET MEASURED ACCEPTABLE LIMIT LIMITS MET l.451

1. 297 1.088 1.8141 Yes 1.470

1. 306 1; 596 2 Yes 0.575 0.517 0,431 0.719 1 Yes 0.500 3

0.475 0.675 Yes 4.

1.375 1.262 1.031 1.718 1 Yes 1.300 2

1.238 1.513 Yes 0.329 0.311

0. 247 0. 411 1 Yes 0.395 0.229 0.429 3 Yes 0.505 0.470 0.379 0.631 1 Yes 0.548 3

0.405 0.605 Yes 0.739 0.707

0. 554 0. 923 1 Yes

0. 709 0.639 0.839 Yes ACCUMULATIVE 4.974 NOT APPLICABLE 4.922 4.477 5.471 'es 1 = Design + 25%

2 = Design + 10%

3 = Design + 0.1 %6K/K

Page 38 5.6 OVERLAPPED REGULATXHG CEA GROUP NORTH MEASUREMENTS Purpose Reactor Power level may be controlled by sequential insertion or with-drawal of Regulating Control Element Assemblies (CEA).

Percent of overlap is selected so as to insure a relatively constant insertion rate of positive or negative reactivity over the full range of CEA Group movement.

Technical Specifications allow CEA Group insertion as a function of reactor power level.

The integral reactivity worth curve for Regulating CEA Groups 2, 3, 4, 5, 6 and 7 in an overlapped mode was measured.

This measurement was made at a Reactor Coolant System (RCS) temperature of 532 F.

Principal purpose of this test was to compare the reactivity worth of the overlap mode for this fuel loading (Cycle 1A) with the reactivity worth of the overlap mode for the original fuel loading (Cycle 1).

Test Results The overlapped integral reactivity worth of CEA Groups 2, 3, 4,

5, 6

and 7 was measured using a soluble boron swap method to maintain criticality while sequentially inserting CEA Groups in increments.

The reactivity trace developed by this CEA Group movement was reduced to obtain the relationship between CEA Group positions and integral reactivity worth at those positions.

Figure 5.6-1 displays the over-lapped integral reactivity worth curve.

Conclusions The overlap CEA Group worth measurements for Cycle 1A compaxed satis-

~actorily. with overlap CFA Grou~. measurements for Cycle l. On this basis, and the conclusion reached in Section 5.5, it was judged not to generate new. individual.CEA.groug worth curves but to use those curves generated during Low Power Physics Tests in Cycle 1.

5. 00 ST.

LUCIE UNIT 1 CYCLE lA, 532

, 2250 PSIA SEQUENTIAL CEA WORTN Figure 5.6-1

4. 00

3. 00

2. 00

1. 00 oo O

0 30 60 Group 3

0 30 60 90 120 136 0

Group 5

30 60 90 120 136 0

30 60 90 120 136 I

I Il~

Group 7

0 30 60 90 120 136 Group 2

Group 4

INCllES OF WITNDRAWAL Croup 6

3-11-77

5.00 ST.

LUCIE UNIT 1 CYCLE 1, 5320, 2250 PSIA SEqVENTIAL CEA WORTll Figure 5.6-2

4. 00

3. 00 0

0

2. 00

1. 00 CO 0

30 60 90 120 136 l~+

Group 1

Groilp 3 0

30 60 90 120 136 0

30 60 90 120 136 Gro Ip I 0

30 60 90 120 136 l-Group 2

0 30 60 90 120 136 Croup 4

INCllES OF HITllDRA'ilAL 0

30 60 90 120 136 G

6 Group 6

3-11-77

Page 41 i

5.7 CEA 7-1 WORTH MEASUR1ÃENT GROUP 7 ~ 100" Purpose This measurement was done to get the differential worth of CEA 7-1, which will be used later in the Moderator Temperature Coefficient/Power Coefficient tests.

Test Results The reactor was critical with CEA group 7 approximately, 100 inches withdrawn.

CEA 7-1 was borated to UEL (upper electrical limit),

diluted to LEL (lower electrical limit), then borated to UEL.

As these steps were performed, the reactivity was being monitored on the reactivity trace on the reactivity computer.

The final boration left CEA 7-1 back at CEA group 7 height.

Conclusions Enough raw data was collected to generate an integral rod worth curve for CEA 7-1.

The measured value of CEA 7-1 was 0.1066% p

CEA 7-1 DIFFERENTIAL ROD WORTll llZP, 532 F, 22GO PSIA Figure 5.7-1 20 15 a

10 IC) 0 0

20 40 10 12 140 INCllES OI'ITlIDRAWAL

CEA 7-1 IHTECRAL ROD WORTll

11ZP, 532 F, 2260 PSIA Eigure 5.7-2 0.10 0.106 0.09 0.08 0.07 0.06 a%4 0.05 0. 04

0. 03 0 ~ 02

0. 01 6

8 INCllES Oli WITIIDRAWAL 10 120 140

Page 44 SHUTDOWN KaRGLl VERIFICATION Purpose The purpose of this test is to verify the shutdown margin at the zero power dependent insertion limits (ZPDIL) assuming the highest worth CEA is stuck out.

Test Results The reactivity worths of shutdown group A,, shutdown group B, regulat-ing groups withdrawn to the ZPDIL, and the highest worth individual CEA were compiled from test measurements taken during fuel cycle l.

Conclusions Section 3.1.1.1 of the Technical Specifications specifies that the shutdown margin should be greater than or equal, to 2.45%ay.

The value that was measured was 5.18%kg.

Table 5.8-1 Worth of Reeulatin CEA's Withdrawn at t Worth of Shutdown CEA's, Group A 3.

/

4.520

%ho Worth of Shutdown CEA's, Group B

0.425

%do Total Available Rod Worth at the ZPDIL Highest Worth Stuck CEA 8.280

%6p 3.100

%ho Actual Shutdown Margin 5.180

%6p

Page 45 6.0 POWER ASCENSION TESTS The power ascension tests were conducted to determine the as-built 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 performed at the 20, 50, 80 and 100% power plateaus.

Several other tests were performed at the 14, 25, 30, 40, 60, 70, 75 and 85% power levels.

Power ascension tests through 50% power on fuel cycle 1 was reported in the Startup Test Report, dated

December, 1976.

The power ascension testing for fuel cycle 1A repeated some of the same tests below 50% power and completed the 50 to 100%

spectrum, also.

The Preoperational Test Procedure Number 0010180 titlted "POWER ASCENSION SEQUENCING DOCUMENT" was used as the guideline for this phase of testing.

Two revisions were made to this procedure during testing to accommodate the re-ascent to power after fuel reconstitution, correct minor editorial errors, and revise the sequence of testing to best fit plant availability and system load demand.

Page 46 TABLE 6.0-1 POWER ASCENSION TESTING, CYCLE 1*

PROCEDURE TITLE & NUMBER (NOTE:

P Preoperational, 0=Operating) 14

% POWER 20 0

30 40 50 SIMULATED CEA EJECTION STEAM BYPASS P

0110087 P

0810080 START UP TRANSFORMER GENERATOR EXCITATION P

0910081 P

0910085 NUCLEAR & AT POWER CALIB GEN. TRIP OUTS.

CONT.

ROOM 0

1200051 P

1400093 PRIMARY CALORIMETRIC 0

3200020 FIXED INCORE ALARM SETPOINT 0

3200050 MOD TEMP COEF

& PWR COEF TCT.

RAD. PEAK FACT.

(FT)

PWR RNG CONT SUBCH CALIB 0

3200051 0

3200054 P

3200080 FORCED XENON RADIATION SHIELDING EVAL CHEM & RADCHEM S/G FEEDWATER HAMMER P

3200087 P

3300081 1

P 3400081 P

0700080A 1 = Performed once at this power level 2 = Performed twice at this power level, usually once at the beginning and again gust prior to leaving

Reference Page 87 of Startup Test Report for St. Lucie Unit 1, Cycle 1, dated

December, 1976

Page 47 LE 6.0-2 POWER ASCENSION TESTING, CYCLE 1A PROCEDURE TITLE AND AMBER NOTE:

P=Preo erational 0=0 eratin E

uent Monitoring a or@metric aps ot oa yc e tat c roy yn c CEA Insertion

. Load Reduction ow et.

y a

roc.

Partial Loss of Flow Total Loss of Flow/Nat Circ S

G Feedwater Hammer APD

& SA Baseline Nuclear

& hT Power Calib Linear Power Ran e Calib.

Auto Control S stem Seconda Samole NSSS.Acce tance Run Generator Tri Turbine Tri Loss of Offsite Power Fixed Incore Alarm Set oint Mod Tem Coef

& &n Coef Power Defect Radiation Shieldin Eval.

Chem

& Radch m

Chemistr 77 DDPS DDPS 0010195 0110088 0110089 0110090 0120051 012008 0120084 1130021 12000 2200 2

4000 30080 2

082 210008 2 000 320 0

3300081

TABLE 6.0-2 (Cont.)

POWER ASCENSION TESTING, CYCLE 1A POWER LEVEL Chem.

77 Cal.

DDPS Sn DP 2

2 2

P 011008 P

0110089 P

0110090 0

1 1

P 0120081 P

0120084 P

0700080B 0

1130021 0

1200051 0

1220052 P

1400084 P

1730080 P

2100082 P

2100089 P

2100090 P

2100091 0

3200050 0

3200051 P

3200084 P

3300081 2

2 2

1 1

2 1

1=Performed once at this power level 2=Performed twice at this power level

S~To be completed sometime during Power Ascension Testing

Page 49 6.1 Flux Distribution Honitorin Due to the flux distribution anomaly and poison pin replacement a program of slow ascent to power with a long observation period at 50% was used for the return to power operations.

Over a week was used in ascent to 50% power and the plant remained at 50%

power for over three weeks.

The monitoring program to verify absence of the anomaly is discussed below.

During Cycle 1, an azimuthal tilt developed due to the failure of burnable poison pellets.

During the repair/down-time the fuel element assemblies were reconstituted.

The reconstituted fuel was reloaded into the core,

and, except for two fuel assemblies exchanging core locations, the core was identical to the initial clean fuel.

To distinguish this re'constituted fuel and subsequent testing and operation from the original clean fuel, the designa-tion "Cycle 1A" was used.

As Cycle lA was returned to the 50% power plateau to resume testing in the Power Ascension Testing program, the azmuthal power tilt was closely monitored.

Table 6 1-1 represents a few of the many snapshots that were taken.

It should be noted that there was extensive time taken to return to the Power Ascension Testing

program, to insure adequate time to observe that the fuel was "burning evenly."

Although Table 6.1-1 ends on January 11,

1977, Tq was still closely monitored during the remainder of the Power Ascension Tests and no indication of any anomaly has been observed.

Page 50 SELECTED Tq VALUES FROM CYCLE lA TABLE 6. 1-~1 DATE TIME

% POWER Tq F

GROUP 7

09DEC76 DEC76 11DEC76 12DEC76 13DEC76 15DEC76 16DEC76 1 DEC76 18DEC76 19DEC76 21DEC76 22DEC76 23DEC76 24DEC76 2000 1127 06 0 0849 1106 1820 0936 1631 1

0 1230 2020 0628 0810 1548 21 30 31 32 42 50 50'1 51 51 51 51 0.016

0. 01 0.013 0.012

.014 0.011 0.011 0.009 0.009 0.009 0.008 0.008 0.008 1.358 1.287 1.288 1.296 1.297 1.295 1.302 1.302 1.307 1.302 0

1.300

1. 304 1.314 ARO*

ARO ARO ARO ARO ARO ARO ARO ARO AR ARO ARO ARO 7 at 100 25DEC76 26DEC76 27DEC76 28DEC76 DEC 6 0802 0707 0600 1530 0728 51 51 51 51 51 0.008

0. 008

0. 007 0.006 1.306 1.302 7 at 72 1.349 7 at 100 7 at 100 1.350 7 at 100 30DEC76 1DEC76 01JAN77 02 JAN77 04JAN77 05JAN77 06JAN77 08JAN77 JAN 10JAN77 2142 0015 1116 1515 1116 1930 0300 1004 1230 1530 1048 6 0 51 51 51 51 51 51 51 51 51 60 61 0.006 0.007 0.006

0. 006 0.006

0. 006

0. 007

0. 007 0.006

0. 007 0.005

+All Rods Out 1.299 1.286 1.302 1.306 1.300 1.286 1.305 1.300 1.309

1. 307 1.307 7 at 72 ARO ARO ARO ARO ARO ARO ARO

Page 51 6, 2 PLANT POWER CALIBRATION Purpose The purpose of the test was to:

(1)

Determine core thermal power by means of a,Reactor Coolant System heat balance.

(2)

Adjust the Power Range Safety Channels and A T Power Reference Calculators to agree with the thermal energy balance calculations.

(3)

Perform, when necessary, a calibration of the Safety and/or Control Power Range Chan-nels.

Test Results Feed flow Calorimetrics were conducted through power ascension.

The Calorimetrics were used to calibrate nuclear instrumentation.

The Power Range Safety Chan-nels and AT Power Reference Calculators were adjusted to agree within 0.5% of the Calorimetric calculations.

These adjustments were performed periodically through-out power ascension.

Conclusions I

Calibration of the Power Range Safety and Control Sub-channels was accomplished acceptably at each major test plateau.

The intent of the Ex-Core Nuclear Instrument Calibration was to adjust nuclear power, AT power and the Calorimetric to within 0.5% of each other, and this was done.

It should be noted that the Calorimetric program had been slightly revised during the fuel reconstitution shutdown.

An error was introduced which caused Calor-imetric power to be 2.5% lower than actual,.

(at 80%

power).

This was found and corrected before the plant exceeded 80% power.

The rest of the program was checked to verify no further errors existed.

(LER 335-77-4 dated February 18, 1977).

Page 52

6. 3 FIXED INCORE DETECTOR ALARM SETPOINTS 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 monitored by one segment level of the incore detector.

Results The procedure instructs that a set of data, taining the score 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 overpower ratios are cal-culated.

Through various factors and constants, an incore detector alarm setpoint is generated for each of the 180 incore detectors.

This was repeated per-iodically as power level was increased to ensure up-to-date setpoints.

It was also done when the Peak Linear Heat Generation Rate interim limit due to STRIKIN II coding errors was lifted.

Conclusions Each of the alarm setpoints was entered into the Digital Data Processor.

The operability of the Digi-tal Data Processor was checked, and each of the fixed incore detectors was determined to have a valid alarm setpoint.

Page 53 MODERATOR TEMPERATURE COEFFICIENT AND POWER COEFFICIENT Purpose The purpose of these tests were to determine the at-power moderator temperature coefficient (MTC) and power coefficient at the 50, 80 and 100% nominal power levels.

'Zhe Technical Specifications have set limits for temperature coefficient.

Combustion Engineering design values

(+10%)

were used for'ower coefficient limits ~

Test Results At the beginning of these (three) plateaus, the plant was steady state at the nominal power level with CEA group 7

= 100 inches withdrawn, with CEA 7-1 in the manual individual (MI) control mode.

The test at each plateau was done in two parts.

The moderator temperature coefficient (MTC) was calculated by holding power constant and lowering RCS Tave.

The resultant reactivity addition was compensated by insertion of CEA 7-1.

The RCS Tave was returned to nominal value with the resultant negative reactivity addition being compensated by CEA 7-1 withdrawal.

This technique was repeated several times and the average temperature coefficient values were used.

The power coefficient was measured by holding the RCS Tave constant and lowering power.

The resultant reactivity change was compensated by move-ment of CEA 7-1.

The power was returned to nominal value and CEA 7-1 again was used to compensate for reactivity changes.

This technique was repeated several times and the average power coefficient values were used.

Conclusions The reactivity worth of CEA 7-1 is known.

By measuring CEA 7-1 height change (hh) we have, in effect, measured reactivity change (hp).

By comparing the two variables in each case (hp/A F) and (hp/h%)

we have the respective coefficients.

It should be noted that the reactivity change associated with the variable Tave process is actually the isothermal temperature coefficient (ITC).

ITC is made up of two components, moderator temperature coefficient (MTC) and fuel temperature coefficient (FTC).

ITC MTC + FTC.

FTC is also known as doppler feedback.

ITC is what was measured.

The FTC value that was used was taken from design values.

MTC is calculated by sub-tracting FTC from ITC.

MTC = ITC FTC.

Moderator temperature coefficient values and power coefficient values at different power levels are presented along with their respective limits, on the table on the next page.

The measured values of MTC and Power Coefficient met all acceptance criteria.

Page 54 TABLE 6.4-1 AT POWER DETERMINATION OF MODERATOR TEMPERATURE COEFFICIENT AND POWER COEFFICIENT Nominal Reactor Power 50%

80%

100%

Moderator temp. coef.

A k/k /

F Ak/k /

oF Ak/k /

F Measured x 10

-Tech-Spec Limit, x 10 4

-0.35

>-2.25

<+0.50

-.022

>-2. 25

<+0.20

-0.307

>-2.25

<+0.20 FSAR Limit, x 10

>-1. 40 BOL

> -1. 40 BOL

<+0. 50

<+0.50

<+0. 50 BOL ~ Beginning of Life Power Coefficient

Measured, x 10 C.E. design, x 10 Ak/k /

-1.06

-1. 00 A k/k

-1.10

-1.025

/%

Ak/k

/

-0.965,

-0. 98 Date 04 Jan.

77 25 Jan." 77 07 Mar.

77

Page 55 6.5 TURBINE VALVE TESTS Purpose The purpose of performing this evolution was to verify capa-bility of plant systems to perform this evolution at power.

Test Results The test recommended by the turbine vendor is essentially a

stroke test.

That is, the valves are cycled closed and open to verify freedom of movement.

There are two valves (Reheat and Interceptor) in each of 4 lines from the high pressure to the low pressure turbine via the Moisture Separator Re-heaters.

These were simply closed and slowly reopened, one set at a time.

The throttle valves (4) which are primarily for startup and quick isolation of the turbine were tested in the same manner.

The governor valves which directly con-trol turbine loading are tested in single valve control mode with impulse (1st stage) pressure control in automatic.

In single valve control mode the 4 governors act together in response to a single signal.

When one valve was closed and

reopened, the other three acted automatically to maintain turbine load essentially constant.

Conclusions The evolution went smoothly.

It demonstrated capability to test the valve at power and that our procedural requirement to reduce power to 90% is more than adequate to prevent ex-ceeding power limits during the minor transients imposed by this testing.

This evolution was performed during the in-itial ascent to 80% power.

Page 56 6.6 PARTIAL LOSS OF FLOW TEST Purpose The purpose of this test is to observe plant response to a partial loss of reactor coolant flow while at power and verify that the Reactor Protective System low flow trip units initiate a reactor trip.

Test Results With the plant at 80% power, one Reactor Coolant.

Pump was turned off, reducing flow to about 80% of normal full flow.

The RPS initiated a low flow trip and all 4 of the low flow trip channels responded to the low flow condition.

Overall plant response was as expected with no significant problems.

Conclusions The plant responsed properly to the trip from 80% power.

The RPS low flow trip units are capable of sensing reduced flow within the required time limits.

Page 57 6.7 POWER DEFECT 'AND XENON WORTH AFTER SHUTDOWN Purpose Various plant and reactor parameters affect core reactivity changes.

Temperature, RCS boron concentration, CEA position and other variables have a pronounced affect on reactivity.

Power level also contributes to reactivity changes.

The higher the power level, the more negative reactivity is in-serted into the core.

This phenomenon is called "power de-fect".

One of the purposes of this test was to measure power defect.

When a reactor is critical, one of the by-products, both directly due to fission and due to decay of other fission produced

elements, is xenon.

Xenon has a very high neu-tron absorption cross section and its presence in a reac-tor has a direct affect on reactivity.

Because xenon ab-sorbs neutrons, xenon inventory in a core is a function of both power history and present power level.

The sev-erest effect of xenon follows a rapid reactor shutdown from power (i.e.,

a reactor trip).

The second purpose of this test was to measure the xenon worth after a shut-down as a function of time.

Test Results'he reactor was at 80% steady state, equilibrium xenon and ARO when it was tripped.

Within an hour, the reactor was brought critical and at zero power.

The position of CEA group 7 was noted.

The difference in CEA position from 80% power ARO partially represents the reactivity due to being at 80% power (i.e.,

power defect).

As xenon production was being compensated by CEA withdrawal, the reac-tivity steps were measured on the reactivity computer.

The amount of reactivity addition to return to ARO was calcu-lated and used in the power defect measurement section of this test.

CEA group 7 was then diluted in to approximately 20 inches withdrawn.

Negative reactivity insertion due to xenon production was compensated by group 7 withdrawal.

A record of reactivity insertion and time after reactor trip was kept.

A chart of these two items plotted against each other was made.

This chart fulfills the xenon worth measurement section of this test.

Page 58 6.7 PONER DEFECT AND XENON NORTH AFTER SHUTDOWN (cont)'onclusions The power defect was actually in two parts.

The first part was the negative reactivity that was inserted due to xenon production during the hour between the reactor trip and criticality.

This value can be infered from the size, shape, magnitude,and time on the xenon worth curve (see next paragraph).

The second part was the negative reactivity that was compensated for by withdrawal of CEA group 7 to ARO.

The power defect from 80% power was 0.75

%Ap.

The peak value of xenon worth was 1.78

%ho eight hours following the reactor trip.

The plot of the measured xenon worth versus time compared to the predicted xenon worth versus time, is on Figure 6.7-1.

From Figure 6.7-1, it can be seen that there is very good agreement between the predicted and actual value.

For instance, the dif-ference between predicted and actual peak xenon was less

. than 2%.. The.acceptance criteria. were met..

0

Page 59 FIGURE 6.7-1 ZION WORTH 1.8

1. 7

l. 6 1.4 1.3 PREDICTED SURED 1.2 1.0 clP 9

a@4 0.8 C

0.7 0.6 0.5 0.4 0.3 0.2 0.1 2

3 4

5 6

ELAPSED TZ'fE FROif TRiP (HOURS)

Page 60 6.8 TOTAL LOSS OF FLOW/NATURAL CIRCULATION TEST Purpose The purpose of this test was to verify that the RPS would initiate a reactor trip on low flow after a total loss of RCS flow and to verify that natural circulation occurs.

Test Results With the plant operating at 40% power, all 4 RCP's were turned off simultaneously.

A reactor trip from low flow then occurred followed by tur-bine and generat'or trip.

The plant was then taken to hot standby conditions except that the RCP's were not restarted until natural circulation flow was verified.

Starting before the trip, AT power data was taken until one hour, 15 minutes after the trip.

AT power initially dropped to nearly zero, returned to the pre-trip value and then started decreasing.

For the last 45 minutes befor'e pumps were restarted, AT power slowly but steadily decreased (as did Thot),

verifying that natural circulation'was occurring.

This allows controlled decay heat removal/cooldown without RCP flow.

Conclusions The RPS low flow trip units initiated the trip and all 4 channels detected the trip condition within the required time.

Verification of natural circulation flow and initiation of a low flow trip by the RPS satisfied the acceptance criteria given in the Purpose.

0 0

Page 61 6.9 STEAN GENERATOR FEEDMATER ~KR TEST Purpose The purpose of this test was to verify the absence of water hammer in the steam generator feedwater piping when the steam generator was drained below the feed ring and then refilled.

This was repeated, as the method of pre-venting rapid draining of the feed ring was modified since performance of the test during the first ascent to power.

Test Results Following a trip from 40% reactor power (after several days of operation at 80%), level in one steam generator was reduced below the feed ring and held at that level or lower for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Then the level was raised to nor-mal at the maximum flow of two motor-driven Auxiliary Feed-water Pumps (785 gpm).

Level was again reduced and held below, the feed ring for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Then level was raised to normal 'at the design -flow rate for one -motor-driven Auxiliary Feedwater Pump (300 gpm).

The HRC had requested decay heat equivalent to that 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after shutdown from at least 2 days of operation at 30% power or greater.

This corresponds to slightly less than 5

megawatts (Mw).

Actual decay heat was calculated to be 7.2 Mw for the first test run and 6.9 Mw for the second.

The behavior of the feedwater piping was monitored by:

observers inside the containment, installed RCS noise monitoring equipment, a test pressure transducer and

recorder, and by measurements of line and restraint positions before and after each test.

Ho evidence of feedwater hammer was observed.

Conclusions The absence of any evidence of steam generator feedwater hammer indicated that St. Lucie Unit 1 is not susceotible to the problem.

Page 62 6;10 LOSS OF OFF-SITE POWER AND LOAD REJECTION Purpose The purpose of this test was to evaluate unit reliability during a load rejection and a loss of off-site AC power.

Test Results With the unit at 20% power and off-site AC power supply breakers

open, the unit was separated from the system distribution grid.

This was done to evaluate response to this 120 MW load rejection and determine the unit capability regarding carrying its own auxiliary loads (about 40 MW) while divorced from the grid.

This in-formation was of interest to FP6L and did not adversely affect the main objective of the test which was response to loss of off-site power.

After the unit carried its own auxiliaries briefly, the generator was tripped.

This demonstrated unit response to a trip and loss of off-site power.

The general ac-ceptance criteria was that systems required during loss of off-site power function as designed.

The specific acceptance criteria were:

2 ~

3.

4.

5.

6.

7.

Decay heat is satisfactorily removed and Tavg is reduced to the no-load value (532 F) or less.

Steam generator levels can be manually re-stored and maintained within limits by feed-ing with the auxiliary feedwater pump.

At least one emergency diesel generator starts and supplies power.

At least one Intake Cooling Water pump starts and operates.

At least one Component Cooling Water pump starts and operates.

RCS pressure shall not exceed 2500 psia during the test.

The RCS shall achieve final conditions of pressure

<2400 psia with pressurizer cooldown not exceeding 200 F in one hour and steam generator pressure

<985 psig with RCS cooldown not exceeding 100 F in one hour.

Page 63 6.10 LOSS OF OFF-SITE POWER AND LOAD REJECTION Cont Conclusions The unit was cap'able of handling the load rejection with the steam dumps controlling pressures and carry-ing its own auxiliaries thus providing data for eval-uation and demonstrating ability to handle separation from the grid at low power levels.

All the acceptance criteria (1-7 above) were met and the plant was returned to and maintained in hot standby without off-site AC power.

The systems required during loss of off-site AC power did function properly as designed.

It should be noted that both diesels did start and supplied power and two ICW and CCW pumps started, as

designed, thus exceeding the minimum acceptance criteria.

Page 64 6 ll 10% LOAD REDUCTION TURBINE RUNBACK Purpose The purpose of this test was to verify turbine runback would occur on a (simulated) rod drop and to determine plant response resulting from a load reduction.

Test Results An actual'od drop occurred with the plant operating at 90%.

Turbine runback was properly initiated and plant performance was satisfactory.

This met the intent of the proce'dure and documentation was gathered to support the proper performance of the systems involved.

Conclusions Turbine runback is properly initiated by a dropped rod and plant response is satisfactory.

8

Page 65 6.12 TURBINE TRIP TEST Purpose The purpose of this test was to verify control systems perform as designed to bring unit to hot standby conditions following a trip from 100% power.

Test Results The test was to demonstrate the following items would occur following the turbine trip.

1.

The generator and reactor tripped.

2.

RCS pressure did not exceed 2500 psia.

3.

Steam dump and bypass systems returned the plant to nominal hot standby conditions RCS temperature about 532 F.

4.

Steam generator pressure did not exceed 1025 psig.

5.

Steam generator levels were manually returned to normal hot standby level.

These items did all occur properly after the turbine trip, verifying the controls maintained the unit within design parameters.

Conclusions Satisfactory completion of the acceptance criteria 1-5 above demonstrated that the control systems performe as designed, and bring the unit to hot standby conditions following a trip from 100% power with minimum operator action required.

6.13GENERATOR TRIP TEST Page 66 Purpose This test demonstrated that the unit could accept design load rejectionthat is, a loss of generator load from 100% power.

Test Results The acceptance criteria for this test were:

2.

3 ~

4.

RCS achieves stable conditions,

<2400

psia, S/G pressure

<985 psig.

The turbine does not exceed its design overspeed of 111%.

RCS pressure does not exceed 2500 psia during the test.

Steam generator pressure does not exceed 1025 psig during the test.

In addition, the turbine and reactor protective systems must terminate the transient before any limiting set-points are exceeded.

Conclusions The reactor tripped and initiated turbine trip on high pressurizer pressure at the proper setpoint.

Steam generator levels were restored to normal and maintained by normal operator action.

The 4 acceptance criteria were satisfactorily met also.

This demonstrates that the unit can accept a design (100%) load rejection and be brought to hot standby conditions with minimum operator action.

0

Page 67

.14 LOAD CYCLE TEST Purpose This test was conducted to evaluate plant and nuclear para-meters in a typical load cycle operation.

FSAR Table 4.4-2 provides limiting values for the three dimensional point~cak-ing factor F, and the integrated radial peaking factor Fr (also identified as the total enthalpy rise factor, AH hot channel/hH core, F<H).

Test Results The plant was initially at approximately 85% power, steady state.

The power was decreased, at the nominal rate of 30%/

hour, to approximately 40%.

The power was increased, at the nominal rate of 30%/hour, to approximately 85%.

Snapshot paper tapes containing incore detector readings, CEA positions and calorimetric powers, were taken at the 85%,

40%,

and 85%

plateaus.

Various plant pressures, temperatures and flows were recorded on strip chart recorders.

Conclusions From the data that was monitored on the strip chart recorders (see Table below) and the nuclear measured via snapshot/GINCA (see Table on next page), it was determined that all levels of acceptance criteria had been met.

Table 6.14-1 Parameters monitored by Stri Chart Recorders Pressurizer pressure Pressurizer level S/G 1A1 Pressure S/G 1A1 level S/G lAl Feedwater flow S/G 1B1 Pressure S/B 1Bl level S/G 1B1 Feedwater flow S/G lA1 Steam Flow S/G 1B1 Steam Flow Loop 1Al T (NR)

Loo 1Bl T NR Loop 1Al or 1A2 Tc (NR) 1st Stage Pressure Loo 1Bl or 1B2 T NR Generator

~i Nuclear Power NI-9

Page 68 TABLE 6. 14-2 LOAD CYCLE TEST (conducted March 19, 1977)

POWER (X)

FAH 0100 0200

77. 35 83.57 0300 0400 0500 0530 0600 0700 0730 0800 0900 0930 1000 1030 1100 83.57

80. 51

82. 91 82.62 72.48 45.19 37.28 47.11 76.79 80.12 83.30

82. 27 81.81 1.57 1.61 1.55 1.55 1.56 1.28 1.30 1.28 1.28 1.28 FSAR LIMIT

<2'1

<2. 02

Page 69

6. 15 STATIC CEA DROP Purpose The purpose of this test was to observe the effects of a con-trol element assembly (CEA) insertion and withdrawal on core power distribution.

The subsequent azmuthal power tilt (Tq) oscillation and dampening, and the effects on linear heat rate was also observed.

Test Results The test was conducted at 50% power.

At this power level, the trends could be seen, but the fuel would not be subjected to as large power transients as would be encountered at high-er power levels and margins from peaking would be greater.

Initial conditions were all rods out, 50% nominal power and equilibrium xenon.

The static measurement consisted of di-luting rod B-7, a dual shutdown CEA, to its lower electrical limit (LEL), while holding reactor power constant.

The di-lution/insertion took approximately one hour..

The azmuthal power tilt grew until it reached a maximum. This required approximately four hours after LEL.

The CEA was then borated out, again holding reactor power constant.

-Boration/

withdrawal took approximately two hours.

Conclusions During the time the CEA was in the core, reactor protection sys-tem (RPS)

Channels A and D showed an increase in nuclear power, while Channels B and C showed a decrease in nuclear power.

Tilt (Tq) indication at the beginning of the test (before the CEA was inserted) was 0.1%.

The maximum value of Tq occurred

~ 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the CEA reached LEL and was

~20.2%.

As the CEA was withdrawn, Tq decreased in value.

The tilt oscillations continued, although dampened with time.

The curve of reactor

power, CEA B-7 position, and Tq plotted against time can be found on the next page.

The effects of the dropped CEA were measured and the ratio of afterdrop to before drop radial fuel peak is less than or equal to 1.154.

Also, an acceptable value of linear heat rate was not exceeded.

Fr (after drop)

/

Fr (before drop)

=

1.153

< 1.154 (FSAR Limit)

T T

Peak Linear Heat Rate

~ 8.2 KW/ft

Page 70 TABLE 6.15-1 STATIC CEA DROP TIME POWER B-7 RPS NUCLEAR POWER T

A B

C D

1800 1825 1830

51. 6

51. 9 U

109

0. 001

'51.4

52. 6

51. 5

52. 1

51. 5

52. 0 53.0 1900 1930 1935

51. 7

51. 6 50 10

0. 075

55. 3 58.3

50. 1 49.7

48. 9 47.7 54.9

57. 2 2000 2100 2200 51.3 51.0

51. 1

0. 136

59. 0

0. 161

60. 0

0. 181

60. 6

49. 2 48.5

47. 7 47.0 57.8 46.1 59.3 45.5 58.9 2300 51.1

0. 193

61. 2

47. 4

44. 6

59. 3 0000 51.0

0. 201

61. 3

46. 9

44. 1

59. 2 01 0 0115 0.

0. 20 0.9

46. 5

43. 7

58. 8 0130

51. 5 30

60. 9 47.6

45. 0

58. 9 0200 0300 0314 0330 0400 0500 0600

51. 3

51. 9

52. 5

52. 5 58 117 0.137

0. 060

0. 018

0. 025

0. 059

58. 6

54. 3

53. 0

52. 5

51. 5

52. 1

48. 5

49. 5

50. 2

51. 2 53.1

52. 7

46. 5

'50. 3

49. 7

51. 1

53. 4

52. 7

57. 1

54. 7

52. 8

52. 7

52. 1

52. 4 0700 0730

52. 2

52. 4

0. 088

50. 4

0. 101

49. 9

53. 1

53. 6

53. 4

51. 1

54. 1

50. 7 0830 1030 1330 430 1530 16 0 1730 1830 1930

52. 4 51.3 1.2

52. 3

52. 5 51.5 51.4

51. 4 U

0. 118

0. 141

0. 143

0. 141

0. 130

0. 114 0.102

0. 081

0. 065 41

0. 022

49. 2

48. 3

47. 8 47.5 47.8 48.9

49. 3 49.4

49. 7 0.3

51. 8

54. 4

.6 54.5 54.6

55. 1

54. 5 53.5 52 ~ 4 51.6

50. 0

55. 1 55.4 55.5 55.4 55.

53.9

52. 5 51.5

49. 6

50. 4

49. 5

.2

49. 0 49-2 50.4

50. 4 50.4 50.7

51. 6 U ~ UPPER ELECTRICAL LAIT L = LOWER ELECTRICAL LIMIT

Azmuthal Power Tilt from Excore Detectors

AZHUTIIAL POWER TILT (Tq)

CEA II-7 POSITION (INCIIES WITIIDRAHN)

CALORIHETRIC POWEP

(/)

1800 0 0 0 0 0 0

0

~

0 0 0 O

I P

I h)

Ln M 0 h)

Ln O

Ln 0 Vt 0 Ln 0 0

~

M0 M

0 Ln Ch M 0 M 4) 0 0

0 0 0 0 0 Q 2000

~

2200-0000 0

0200-O 0400 R

0600 0800 P4 0

1000 1200 1400 3 600 1800 2000

Page 72 6.16 DYNAMIC CEA INSERTION Purpose'o demonstrate the ability of the reactor protective protection system (RPS) to detect a dropped control element assembly (CEA) under typical operating conditions.

To demonstrate that the control element drive motor (CEDM) limit,switches can independently detect a dropped rod and indicate this on the Core Mimic Display.

To demonstrate that the associated annunciator circuits respond to a dropped CEA.

To demonstrate that the digital data processor system (DDPS) will indicate a rod has been dropped on the CEA position log printout.

Test Results This test was performed at a nominal power level of 50%, with all rods out (ARO) and the turbine under valve position limit.

The control element drive system (CEDS) was placed in the manual individual (MI) control mode with CEA 7-1 selected.

The CEA was dropped by opening the 240 volt breaker in the coil power programmer.

CEA 7-1 was the CEA selected to meet the requirement regarding dropping the furtherest (from the excore detectors) detectable CEA.

Reference FSAR section 15.2.3.1.

The following plant and nuclear parameters were monitored on recorders throughout the test:

Loop lA Th Loop 1A Tc Loop 1A Steam Flow SG lA pressure SG 1A level Loop lA feed flow Pressurizer Pressure NI 9 upper NI 10 upper Loop'1B Th Loop 1B Tc Turbine First Stage Pressure SG 1B pressure SG 1B level Generator Output (MW)

Pressurizer Level NI 9 lower NI 10 lower Conclusions The RPS detected the rod tested, determined it to be dropped and indicated this on the RPS panel.

The associated annunciator circuits K9, K14, K18, K24, K27, K30 and K33 responded to the dropped rod.

The CEDM limit switches detected CEA 7-1 and verified on the Core Mimic Display that it was dropped.

The DDPS indicated on the rod position log that CEA 7-1 had been dropped, thus all acceptance criteria were met.

Page 73 6.17 AUTOMATIC CONTROL SYSTEM CHECKOUT, STEAM GENERATOR LEVEL CONTROL, CEA REGULATING SYSTEM AUTOMATIC TURBINE CONTROL A Purpose The purposes of this test were:

To demonstrate the capability of the CEA Regulating System to cause CEA insertion when required under steady state and normal transient operation.

To verify that the feedwater regulating system for 1A and 1B steam generators will give an adequate and stable response during steady state and expected transient conditions.

To verify smooth response of the turbine control system during load increases or decreases.

Test Results Testing was conducted at reactor power levels of 30%,

50%, and 90%.

Each series of tests at a given plateau consisted of a 2% turbine load reduction at a rate of 1/2%/min, and a 10% turbine load reduction at a rate of 1%/minute.

Both CEA Regulating System channels were tested in this manner.

In addition, at each test plateau, one channel was tested with a 10% turbine load reduction at a rate of 5%/minute, and the other channel tested by a 5% "step" reduction in turbine load at a rate of 200 MK/minute followed by a second such step.

Each test at a given plateau demonstrated the following:

The CEA Regulating System has the capability to decrease reactor power to minimize reactor - turbine mismatch during load reductions.

Steam generator level can be controlled in the automatic mode without significant oscillations during steady state operation and following transients.

The turbine control system performed the load increases and decreases required by the test in a controlled manner with no undamped or divergent oscillations as indicated by generator megawatts or turbine first stage pressure.

Conclusion Testing showed that the CEA Regulating System, steam generator level control system and turbine control system respond when in automatic mode to control plant conditions during normal steady state operation, minimize reactor-turbine mismatch during load reductions, and stabilize plant conditions subsequent to transients

~

Thus, all acceptance criteria as listed in the purpose, were satisfactorily met.

0

Page 74 6.18 NUCLEAR STEAM SUPPLY SYSTEM NSSS ACCEPTANCE RUN Purpose The purposes of the test were to:

(1)

Verify reliable steady-state full power capability of the Nuclear Steam Supply System portion of the plant at or above the 200-hour initial warranted power level of 2450 MWt.

(2)

Verify 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months continuous operation at the maximum warranted power level of 2560 MWt.

(3)

Verify steam moisture at warranted power did not exceed 0.2%.

Test Results The first 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of the warranty run was satisfactorily conducted at a power level at or above 2450 MWt.

The second 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of the warranty run was satisfactorily conducted at a power level of 2560 MWt.

The actual test was conducted over a period of 111 hours0.00128 days 0.0308 hours 1.835317e-4 weeks 4.22355e-5 months due to a power reduction caused by a main condenser tube leak.

This did not invalidate the 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months run as'he cause was not related to the NSSS.

Due to plant and test equipment availability we have not yet measured steam moisture.

Conclusions The NSSS was demonstrated capable of sustained operation during the warranted period of 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months .

Steam moisture will be measured as soon as practical but we will not issue a report unless significant problems are noted.

Page 75 6.19 Shieldin Effectiveness and Plant Radiation Level Measurements 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 shield-ing.

(3)

Determine radiation levels throughout the plant at various power levels.

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 selected points, both inside and outside of the Radiation Controlled Area, were determined.

Dose rates at each poin) were deter-mined at power levels of 0% (Background),

1x10 5%,

20%,

50% and 100%.

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

Due to excessive exposure rates found at 50% power, surveys were not taken inside containment at 80% and 100% power.

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 personnel hatch and the two entrance doors to the containment annulus.

These measurements do not include electrical and mechanical penetration areas in the reactor auxiliary building or the fuel handling building or spent fuel pool area.

These en-closed equipment spaces may experience variable dose rates.

Therefore, in addition to the initial measurements which verified that these areas were not Radiation Areas, they are surveyed periodically to ensure proper control on a continu-ous basis.

Page 76 6.19 (Con't.)

Conclusion Maximum general area gamma and neutron 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.

Surveys determined that areas above FSAR limits are those in'nd adjacent to the containment.

Other than the containment itself, the area which presents the greatest exposure prob-lem is the spent fuel handling 62'levation which has ex-posure levels of 100 mRem/hr.

(neutron plus gamma) where the FSAR specifies 2.5 mRem/hr.

maximum combined neutron plus gamma.

Of primary concern are the high levels found at the 62'leva-tion of the containment.

Exposures of 12 Rem/hr were found at 50% power.

However, no single areas were found other than the refueling machine, which were not in the same range of exposure rate.

This uniformity of dose rate on the 62'leva-tion is postulated to be from backscatter from the contain-ment vessel walls and ceiling.

The high dose rates in the containment severely limit and at most power levels prohibit personnel access to areas which require periodic or special inspection.

As noted in FSAR Section 12.1, the analytical results in-dicated the need for shielding and so the design effort for St. Lucie 1 preceded the measurement program.

We have now submitted to the NRC a report entitled "Neutron Streaming Report" letter L77-126 dated 4-25-77.

This report fully describes the neutron problem and our proposed solution(s) to correct it and gives more detailed neutron survey re-sults.

Page 77 6.20 CHEMISTRY AND RADIOCHEMISTRY TESTS AT POWER Purpose Chemical and Radiochemical analyses of the Reactor Coolant System (RCS) and the steam generators were performed to determine corro-sion data, fission and activation product levels and buildup and to detect failed fuel and various impurities which could enhance corrosion.

Also to ensure that primary and secondary water chem-istry meet the criteria set forth in the St. Lucie Unit 1 Chem-istry Manual for system protection.

Test Results Chemical and Radiochemical tests were performed in accordance with the St. Lucie Unit 1 Chemistry Manual and Pre-Operational Test Pro-cedure f/3400081.

This included samples at various power levels up

'o and including nominal 100%.

Primary (RCS) 1.

During and following re-startup activities all parameters were within specified limits with few exceptions as described below.

Iodine activities showed no new evidence of fuel failure.

2.

Lithium - Lithium was added in the form of LiOH during heatup and prior to criticality to 1.2 ppm.

Subsequent additions maintained Li concentrations at the high end of specifications to aid in building a protective corrosion film.

3.

Dissolved Oxygen (D.O.) Hydrazine (N2H4) was added to the primary system during the filland vent evolution to a maxi-mum of 30 ppm N2H4 along with purges of the VCT scavenged oxygen in the system.

Subsequent additions of N2H4 and VCT purges were performed until D.O. was generally <.005 ppm.

D.O. was maintained at this level through power ascension and at no time did D.O. exceed

.1 while >250 F.

4.

Suspended Solids (S.S.)

S.S.

were maintained generally

.01 ppm during startup and operation.

Maximum reading obtained was 2.45 ppm following first Reactor Coolant Pump run during initial heatup after fuel reconstitution.

Secondary Systems 1.

The moisture carryover test has not been performed due to small chloride problems and lack of sufficient equilibrium power levels greater than 99%.

Page 78 6.20 CHEMISTRY AND RADIOCHEMISTRY TESTS AT POWER (cont)

Secondary Systems (cont) 2.

Small condenser tube leaks caused the unit to vary power.

ximum concentration of'chlorides in steam generators was 1 ppm.

Through blowdown and isolation and repair of water boxes chlorides were quickly reduced and maintained below operating limits and generally <.05 ppm.

3.

The Hotwell Cation Conductivity cells have proven to be very responsive to small tube leaks in the condenser.

They have been very reliable for detecting chloride leaks and enabled us to isolate and repair the indicated water box normally before exceeding the chloride specification.

4.

General chemistry on the secondary side was operated with-in specified parameters with few exceptions.

These ex-ceptions were generally predictable as in power transients and pump starts and were quickly reduced and returned to within specifications.

5.

pH pH additions again were not necessary to control feed or steam generator pH while maintaining a hydrazine resi-dual of approximately 20-30 ppb.

6.

S.S.

Suspended solids in the Feed system and steam gen-erators were controlled by generator blowdown and were generally <1.0 ppm.

Exceptions to this were during startup of pumps and when vacuum was broken on condenser causing increased oxygen.

These periods were short in time and reduced quickly by going to full blowdown on generators.

7.

Cl Chlorides were generally

<.05 ppm except on a few oc-casions when sea water leaks were detected in the conden-ser.

Chlorides were quickly reduced by increased blowdown and isolation of faulty water box.

The Hotwell Cation Con-ductivity cells were the mainstay of our leak detection pro-gram and functioned better than expected in keeping chloride from accumulating in the steam generators.

Maximum Cl,found in generators was

~l ppm, for a very short period.

I Conclusions Chemistry controls in primary and secondary systems as specified in the St. Lucie Unit 1 Chemistry Manual were implemented and found to provide adequate protection for plant system.

Surveil-lances and test procedures gave sufficient indications of adverse trends to allow ample time to restore conditions normally prior to exceeding operational limits.

TABLE 6.20 RCS 'CHEMISTRY AT STEADY STATE POMER Page 79 ANALYSIS pH Conductivity Cl D.O.

S.S.

Boron Lithium Diss.

H2 Gas Act.

Gross Act.

Crud Act.

Tritium I Ratio DE/ I-131 Spectrum LIMITS 4.5-10.2 Varies

<.15 ppm

.1 ppm

.1 ppm 5 ppm Varies

.2-2.0 ppm 10-50 cc/kg

< 1 uCi/gm 1/17/77 80%

6.43 12.0

<.05

<.005

.01 624 1.10 34.4 3.51E"01 3.0E 1

4.315E 4

7.47E 2

.517 2.03E 3

Performed 3/3/77 3/4/77 100%

6.40 10.67

<.05

<.005

<.Ol 591 1.34 16.38 4.216E 1

8.34E 1

1.977E 3

1.03E 1 1.085 5.81E 3

Performed

0

Page 80 6.21 TOTAL RADIAL PEAKING FACTOR This measurement was taken at the 50% power level.

Total radial peaking factor (FTr) is defined as the product of the unrodded planer peaking factor /Fr) ynd the quantity one plus the azmuthal tilt (Tq).

F~ ~ F (1 + T ).

This necessitates a calculation of azmutkal power tilt. loth Tq and F

are calculated by the GINCA computer incore analysis program /see section under DDPS Snapshot heading).

Fr is also a required monthly surveillance item that is performed, by the Reactor Engineering Department.

See Table below for selected measurements taken during the power ascension testing program.

Table 6.21-1 DATE

% POWER T **

1130 0630 0230 0600 1512 1048 2200 1018 12/10/76 12/11/76 12/17/76 12/21/76 01/Ol/77

,01/10/77 01/15/77 01/29/77 20 30 42 51 61'0 81

0. 0157

0. 013

0. 010 0.008

0. 006

0. 007

0. 004 0.007

1. 328 1.287 1.301 1.306

1. 300

1. 307 1.310 1.310 1412 0754 0800 02/10/77 02 24 77 03/22/77 89 100*

0. 004 0.004 FTr limit is 1.36 at 100% power

- Tq limit is 0.02 per Technical Specifications

0

Page 81 6 ~ 22 TURBINE OVERSPEED TRIP TEST Purpose The purpose of this test was to demonstrate that the Turbine overspeed trip mechanism would trip the turbine at a speed of 1998+0 -10 rpm and to verify that the overspeed trip weight will operate when the trip weight body is subjected to oil pressure and record this pressure.

Test Results Holding the test handle in "Test" position, oil pressure was applied to the trip mechanism to verify it would actuate.

The pressure was recorded for future use during periodic testing.

Holding the test handle prevents the turbine from actually tripping while allowing an operability check of the trip mechanism.

Then the turbine speed was raised until the turbine actually tripped.

The speed at trip was low so adjustments were made and the trip repeated.

Final overspeed trip value was 1971 rpm.

This is lower than the specified value but was considered acceptable as it is conservative, will have no effects on reliable turbine operation, and the next adjustment would have been close to the upper rpm limit of 111% of design speed.

Conclusions This procedure verified proper setting of'he turbine overspeed'rip, This was performed during Cycle 1 but was inadvertently omitted from the original startup report.

Page 82 6'3 XENON FOLLOW Purpose This was not part of the scheduled power ascension

testing, but rather a verification that the shape annealing factors (SAF) that were calcu-lated and input as a gain adjustment in the linear power range sub-channels during testing in Cycle 1 were still valid for Cycle 1A.

Results The reactor was at equilibrium xenon with CEA group 7 at approzimately 68 inches.

CEA group 7 was borated out and the resultant axial shape index (ASI) oscillation was monitored for each power range channel.

By correlating the GINCA ASI to the ezcore ASI, the SAF for each channel can be generated.

Two methods were used to reduce the data that was generated.

The first technique plotted GINCA ASI vs ezcore ASI and visually deriving the slope of the line (SAF) that was generated for each channel.

See figure 6.23-1 for a typical geometric display.

The second technique used a mathematical analysis known as least squares fit.

The two techniques are compared with the reference values for each chan-nel, in table 6.23-1.

This verification was done at the 50% power plateau and again at the 80% power plateau.

The graphs of calorimetric power, CEA group 7 posi-tion, and average RPS ASI, vs time for the 50% power level and the 80%

power level may be found on figures 6.23-2 and 6.23-3, respectively.

Conclusions The reference (from fuel Cycle 1)

SAF values were still valid and the gain adjustments for the linear power range subchannels were left as they were.

Combustion Engineering confirmed this decision.

The oscillation for Cycle 1 began with equilibrium, all rods out (ARO).

CEA group 7 was diluted to mid-core and as the swing reached a maximum

value, CEA group 7 was borated to ARO.

The oscillation for Cycle 1A verification began with equilibrium conditions and CEA group 7 at mid-core.

GINCA ASI solution assumes oscillation beginning at ARO, hence there are some discrepancies.

Page 83 TABLE 6.23-1 SHAPE ANNEALING FACTORS 50%

POWER CHANNEL A"

REFERENCE 3.46 3.77 L.S.F. d

3. 282 3.597 GEOMETRIC 3.385 3.678 D

3.92 3.07 3.21 3.588 2.918

'.186 3.676 2.946 10 3.68 3.448 80%

POWER A

3.46 3 ~ 25 ~>

3.31 3.77 3.53

3. 62 3.92 3.48 3.56

3. 07 2.81 2.95 3.21 3.09

3. 14 10 3.68 3.32 3.54 a

b c

SAF ~ (Ginca Axial Shape Index)

(Excore Axial Shape Index)

Reactor Protective Channel or Nuclear Instrument Control Channel Cycle 1 Forced Xenon Oscillation (Ref. Dec.

'76 Startup Report, Sec.

6.11)

Least Squares Fit

Page 84 FIGURE 6.23-1 GEOMETRIC FIT FOR SAF (CHANNEL A~

50%

POWER)

GINCA ASI 0.25 0.20 0.15 SLOPE

-" 3.385 0.10

-0.04

-0.02 0.05

-0. 05

0. 02 0.04 0.06 EXCORE ASI

-0. 10

-0.15

-0.20

-0.25

Page 85 FTGURE 6.23-2 50%

X ViOiV FOLLOW OH

~

4%4 gJ CC g

CQ O ~

~ O O

54 52 50 48 O ~

H C O A H

O M O

Q O C 'H 4+0 140 120 ~

100 T 80 60 40 20 0

OCtg H

+2.00

+1. 00 0.00

-1.00

-2.00

-3.00 DEC 30 O

O O

O COO DEC 31 O

O O

JAVi 1

O O

JAiV 2 O

O

'C O

JAiV 3

Page 86 FIGURE 6.23-3 80%

XENON FOLLOW U

CC oM 0 O A Da O

88 84 80 76 72 O

Ch O

O CC O

5Ã 140I.

120 j-100I'0l 60(

40l 20 x

+2 ~ 00 tg

+1.00 ch

0. 00 (a

-1. 00

+1

-2.00

-3.00 OO CQO O

O O

OO OO O

O O

O CO uD O

28 JAN 29 JAN 30 JAN 31 JAiV

Page 87 6.24 EFFLUENT MONITOR CORRELATION Purpose The purpose of this evolution was to compare the indicated count rate on the various monitors to actual effluent activity.

This provided a check of monitor linearity and provided data for ac-tivity versus count rate curves for in-plant use.

Results The following monitors were included in the correlation:

Liquid Radioactive Waste Gaseous Radioactive Waste Plant Vent Iodine, Particulate, Gaseous Air Ejector Vent Steam Generator 1A and 1B Blowdown All monitors were checked at least. twiceone of these checks was during nominal 100% power operation.

The first three listed above were done using effluent streams from which samples were taken for detailed analysis.

Since we have no steam generator tube leaks there was no activity in the effluent streams for the last two.

For these, representative samples were made up, traceable to the National Bureau of Stand-ards.

This allowed generation of curves comparing monitors count rate to actual activity from representative effluent streams.

In all cases, samples and standards used were counted on a gamma an-alyzer calibrated to NBS standards using appropriate geometries.

Conclusions All monitor readings were acceptably linear throughout the range of the instruments.

Meaningful curves comparing monitor count rate to actual activity were drawn for each monitor for in-plant use.

Page 88

6. 25 DDPS CALORIMETRIC AND DDPS SNAPSHOT At the beginning and end of every major power level, a digital data processor system (DDPS) calorimetric and snapshot was performed.

The reactor power, in million British thermal units (MBTU) was the main item of interest on the calorimetric printout.

The reactor power in MBTU can be divided by 87.3728 to get reactor power in per cent of rated thermal power.

The snapshot is a paper tape punch out containing time, date, CEA positions, signal readings from each of the 45 vanadium incore neutron detectors and signal readings from each of the four levels of the rhodium incore neutron detectors.

The snap-shot, normally, was read via an on-site terminet to Combustion Engineering's computer in Connecticut, where a core performance

analysis, GINCA, would be performed.

The GINCA in Miami, be put on gineering

more, the ed out at result can be transmitted to FP&L Power Resources in a "long form" format.

The GINCA result can also micro-fiche and the film mailed to PSL Reactor En-Department, in the "long form" format.

Purther-GINCA result, in a "short form" format can be print-the on-site terminet.

The table on the next page identifies the make-up of the GINCA short form that is returned to the terminet.

Page 89, TABLE 6. 25-,1 INFORMATION ON GINCA SHORT FORM Reactor Power RCS Inlet and Outlet Temperatures Inoperable Incore Neutron Detectors Core Axial Shape Index Core Average Axial Shape Index Core Average Axial Peak Location Core 3-D Power Peak Core Average Burnup Batch Burnup Azmuthal Tilt Amplitude:

Excores Azmuthal Tilt Angle:

Excores Azmuthal Tilt Amplitude:

Incores Azmuthal Tilt Angle:

Incores Alarm Setpoint (level vs value)

Total Radial Peaking Factor Integrated Radial Peaking Factor Core Maps with Max. and Min. Values Relative Power Density Linear Heat Rate F

Exposure PfWD/KTJ)

H f

Page 90 7.0 COMMENTS ON ORIGINAL STARTUP REPORT During our preparation of this followup Startup Report, we noted a few items in the initial report that were not completely discussed or were inadvertently omitted.

Below is a disccusion of these items.

Section 3.1.2, Test Results, CEDM/CEA Performance Tests has a

list of the tests performed.

That list should include:

(7) Manual scram tests.

This verified that the CEDM/CEA system responded properly to a manual scram and it was completed satisfactorily.

These manual scram tests were also performed as part of the initial Approach to Criticality procedure but were not reported in that section either.

Section 3.10 RCS Heat Loss.

This section'id not discuss the predicted value for heat loss and did not compare actual data to the predicted.

The actual total value obtained was 2.84 Mw.

The expected value was on the order of 3 Mw.

Section 6.9, Reactivity Coefficient Measurements compared the actual results to the Technical Specification or vendor numbers, but FSAR values were not given.

For MTC, the FSAR Accident Analysis assumes values from

+.5 to -1.4 x 104 AK/K/ F, The measured value is.1x 10 A K/K/ F which is within the band.

For Power Coefficient the FSAR gives

-1.6 x 10 3

AK/K/Kw/ftwhich corresponds to -.96xl0 " AK/K/%.

The measured value was -1.07 x 10 4

AK/K/% and the vendor design value is (-1.0 +.1) x 10 4 AK/K/%.

Section 6.13, Core Power Distributions, stated that the radial and axial peaks measured (1.3414 and 1.336 respectively at 50% power) were acceptable.

The accident analysis assumes values of 1.7 and 1.44 respectively.

Our measured values are conservative with respect to the FSAR values

and, as originally stated in 6.13, are satisfactory.

The Technical Specification limit for radial peaking is

<1.36 which is above the me'asured value.

Page 91 7 ~ 1 RCS FLOP COASTDOVil In the original Startup Report, the flow coastdown results were given.

In that report, they were not specifically compared to the FSAR flow coastdown

curve, Figure 15.2.5-1.

The attached figure has a reproduction of the FSAR curve (1) for total loss of flow and a plot of the actual test results (3).

In addition, curve (2) is CE's revised coastdown curve, using as-built data, which was used in their final safety and setpoint analysis.

It should be noted that the test results are con-servative with respect to both analytical curves and the acceptance criteria was met.

Page 92 FLOW COASTDOfPii 0

0 G

1 FSAR Fig. 15.2.5-1 2

CZ Fig. 15.2.5-1 received by FP&L 2/22/76 O

3 Test Results 0

Time Seconds

1 I

r

,S J

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