ML19208A936

Startup Rept.

ML19208A936
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/31/1979
From:	ARKANSAS POWER & LIGHT CO.
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Shared Package
ML19208A934	List: ML19208A933 ML19208A936
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NUDOCS 7909180316
Download: ML19208A936 (183)

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ARKANSAS POWER & LIGHT COMPAhT ARKANSAS NUCLEAR ONE STEAM ELECTRIC STATION UNIT TWO STARTUP REPORT TO THE U.S. NUCLEAR REGULATORY C0!f1ISSION LICENSE NUMBER NFP-6 DOCKET NUMBER 50-368 FOR THE PERIOD ENDING July 31, 1979

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Foreword This Startup Report for Arkansas Nuclear One Unit 2 covers the period from the issuance of an operating license by the Nuclear Regulatory Commission, July 18, 1978, until completion of 20% power testing.

It is being submitted in accordance with Unit 2 Technical Specifi-cation 6.9.1.1 and Regulatory Guide 1.16, " Reporting of Operating Information - Appendix "A" Technical Specifications." The latter requires a startup report to be submitted within 90 days following completion of the startup test program or within 9 months following initial criticality, whichever is earliest.

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PREFACE Arkansas Nuclear One (ANO) - Unit 2 is located adjacent to Arkansas Nuclear 1 on a'-oenincula in the Dardanelle Reservoir on the Arkan-sas River in Pope County, Arkansas. The plant is about six miles West-Northwest. of Russellville, Arkansas, and about two miles South-east of the vi!14qe of Lcadon, Arkansas. The Nuclear Steam Supply System for both units is of the pressurized water reactor design.

Unit 1 is rated'pt 2568 Mwt and was supplied by Babcock and Wilcox.

Unit 2 is rated at 2815 Mwt and was supplied by Combustion Et;i-neering, Inc. Bechtel Corporation was the Engineer Constructor' for both units. Major design parameters are listed below:

7,esign Thermal Power 2815 Mwt Design Electrical Power 912 Mwe Average Temperature Operating (100% Power) 583 f Normal Operating Pressure (Primao) 2250 psia 6

Reactor Coolant Flow Rate 120.4 x 10 lb/hr Normal Operating Pressure (Secondary)(100%) 900 psia Steam low (Total Both Steam Generators 6 2.64 x 10 lb/hr s

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Page i TABLE OF CONTENTS SECTION PAGE

1.0 INTRODUCTION

AND

SUMMARY

l 2.0 INITIAL FUEL LOAD 3 3.0 POST CORE HOT FUNCTIONAL TESTS 14 3.1 NSSS TESTS 19 3.1.1 Intercomparison of Plant Protection System (PPS), 19 Core Protection System (CPC's), Main Control Board and Process Computer Input Tests i.

t 3.1.2 Reactor Coolaut System Flow Measurement Tests 20

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3.1.3 Control Element Drive Mechanism (CEDM)/ Control 27 Element Assembly (CEA) Tests 3.1.4 Primary and Secondary Water Chemistry Tests 34 3.1.5 Fixed Incore Instrumentation Tests 36 3.1.6 Movable Incore Instrumentation Operation Veri- f' fication Tests 3.1.7 Reactor Coolant System Leakage Measurement Tests 39 3.1.8 Reactor Coolant System Expansion Measurement 41 Tests 3.1.9 Reactor Coolant System Cold Leg Restraint Gap 43

\ Measurement (Shimming Verifications) Tests i

3.1.10 Core Protection Calculator / Reactor Trip Response 45 Time Tests 3.1.11 Pressurizer Spray Valve Control and Adjustment 53 Tests 3.1.12 Reactor Coolant System Heat Loss Tests 57 3.1.13 Chemical and Volume Control System Integrated 59 Tests 3.1.14 CEA Exercise Tests 61 3.I.15 Steady State Vibration Tests 66 3.1.16 Safety Injection System Check Valve Petest 67 3.2 PLANT TESTS 69

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Page 11 TABLE OF CONTENTS (Cont'd)

SECTION PAGE 3.2.1 Emergency Feedwater System / S/G Waterhammer Tests 69 3.2.2 Emergency Feedwater Flow Tests 72 3.2.3 Main Steam Safety Valve Piping Dynamic Transient 75 Tests 3.2.4 Emergency Feedwater Pump Turbine Dynamic Transient 76 Tests 3.2.5 Secondary Hydrostatic Tests 77 3.2.6 Pipe / Component Hot Deflection Preoperational Test 78 4.0 INITIAL APPROACH TO CRITICALITY 81 5.0 LOW POWER PHYSICS TESTS 87

5.1 INTRODUCTION

87 5.1.1 CEA Coupling Tests 90 5.1.2 CEA/Part Length Control Element Assembly (PLCEA) 91 Symmetry Tests 5.1.3 Isothermal Temperature Coefficient Tests 95 5.1.4 CEA Group Worth Tests 98 5.1.5 Differential Boron Worth Determination Tests 113 5.1.6 Critical Boron Concentration Tests 115 5.1.7 Pseudo Dropped and Ejec ' CEA Worth Tests 117 5.1.8 Stuck CEA Worth Measurement Tests 120 6.0 POWER ASCENSION TESTS 122 6.1 0% THRU 20% POWER PLATELs 126 6.1.1 RCS AT Power Determination Tests 127 6.1.2 Nuclear and Thermal Power ~ Calibration Tests 129

6. l'. 3 NSSS Calorimetric Tests 131 6.1.4 RCS Calorimetric Flow Measurement Tests 133

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Page iii TABLE OF CONTENTS (Cont'd)

SECTION PAGE 6 1.5 Linear Power Subchannel Calibration Tests 137 6.1.6 Process Variable Intercomparison Tests 138 6.1.7 Chemistry and Radiochemistry Tests 139 6.1.8 Core Performance Record Tests 141 6.1.9 CPC/COLSS Verification Tests 143 6.1.10 Variable T AVG ests 144 6.1.11 CEA Shadowing Factor Verification Tests 148 6.1.12 Unit Load Transient Tests 152 6.1.13 Shape Annealing Matrix and Boundary Condition 159 Measurement Tests 6.1.14 Reactor Trip with Shutdown Outside the Control 162 Room Tests 6.1.15 Incore Detector Signal Verification Tests 167 6.1.16 Movable Incore Detector Tests 168 6.1.17 Feedwater Control System Post-Trip Setting Tests 169 6.1.18 Condensate and Feedwater System Power Escalation 171 Tests 6.1.19 Ma.n Turbine Electro Hydraulic Control rests 172 6.1.20 Feedwater Heater Vents, Drains and Water Induction 173 Tests 6.1.21 Vibration and Loose Parts Monitor Tests 174 6.1.22 Heating, Ventilating and Air Conditioning Systems 175 Performance Tests 6.1.23 Biological Shield Survey Tests 176 6.1.24 Steady State Vibrations Tests 179 6.1.25 Dynamic Transient Tests 180 6.1.26 Turbine Generator Loading at Power Tests 181 s

7.0 CONCLUSION

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1. INTRODUCTION AND SUF%iRY

1.1 INTRODUCTION

The Startup Test Program was organized by Arkansas Power and Light (AP&L) personnel and administered by Combustion Engineering (CE) consulting Startup Engineers assisted by CE site Startup Engineers and home office personnel in Windsor, Connecticut. The Startup Test Program consisted of several phases. The test results from each phase were reviewed by the Test Working Group consisting of an AP&L Startup Supervisor, CE Lead Startup Engineer, CE Site Manager, Bechtel Project Startup Engineer, CE Chief Test Engineer, AP&L Nuclear Engineer, and others as required.

Test results falling outside of acceptance criteria re-ceived an additional review by the Plant Safety Committee and were resolved prior to beginning the next test phase.

The test phases are as follows:

A. Initial Fuel Load B. Post Core Hot Functional C. Initial Approach to Criticality D. Low Power Physics Tests E. 0% thru 20% Power Plateaus F. 20% thru 50% Power Plateau G. 50% thru 80% Power Plateau H. 80% thru 100% Power Plateau The maximum licensed reactor core power level (10(1) is 2815 MWth. The Startup Program began at 0102 7/?3/78 with the loading of the first fuel assembly into the reactor vessel and was completed (Later). -

1.2 SUMHARY 1.2.1 Initial Fuel Load Fuel loading was preceded by a response checkout of all neutron detectors that were utilized dur-ing the core loading. Two temporary incore de-tectors were used as well as the two permanently installed start-up range channels. Operability of the detectors was verified by positioning a neutron source near the detectors. Background count rate war also determined at this time.

. Fuel loading commenced on 7/2.'1/78 and the final fuel assembly was installed oi 7/28/78. During the evolution, minor delays were experienced due to equipment malfunctions and in one instance a fuel assembly was lowered on a previously loaded fuel assembly. A visual inspection of the involved assemblies revealed no damage.

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2 1.2.2 Post Core Hot Functional Tests Post Core Hot Functional testing commenced on 9/5/78 and was completed on 12/3/78. The pri-mary purpose of this testing was to verify that all required plant systems were operable prior to initial criticality. The Post Core Hot Functional Tests were conducted prior to bring-ing the reactor critical at selected pressures and temperatures ranging from ambient to zero-power, no-load conditions (545*F, 2250 psia).

A number of significant problems occurred during Post Core Hot Functionals that considerably lengthened the testing. These major problems included Control Element Assembly Control Sys-tem failures, Coil Stack replacement for CEDM

1. 39, failure of #2 Diesel Generator, failure of a Reactor Coolant pump flow DP sensing line root valve, and unit inverter problems. Each instance required a plant cooldown before re-pairs could be effected.

1.2.3 Initial Criticality The approach to initial criticality was ini-tiated on 12/4/78 at 1650 with the commence-ment of CEA withdrawal. Minor problems were encountered with the control element drive system that required adjustment of operating voltages and component replacement. A slow dilution of the Reactor Coolant System at ap-proximately one ppm / minute was established on 12/5/78 at 0900. The reactor was declared critical at 1455 cn 12/5/78. The measured boron concentration for criticality was in excellent agreement with the predicted value.

1.2.4 Low Power Physics Testing Low Power Physics testing commenced on 12/5/78 at 2330 and was completed on 12/16/78 at 1050.

All test results agreed favorably with predic-tions and were within acceptance criteria.

Numerous minor delays were caused by problems with the Process Computer System and the Con-trol Element Assembly Position System requiring additional grooming of the systems.

1.2.5 POWER ASCENSION TESTING The power ascension phase of testing commenced on 12/16/78 at 0800 and the 20% power test plateau was completed allowing escalation to 50% power on June 24, 1979. Testing is to be couducted at 3 additional major plateaus (50, 80 and 100%). The purpose of this test pro-gram is to verify as-built plant characteris-tics are acceptable and to verify the assum -

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3 2.0 INITIAL FUEL LC.tD Initial fuel loading of Arkansas Nuclear One - Unit II com-menced at 0102 aours on July 23, 1978, and was completed at 0620 hours0.00718 days <br />0.172 hours <br />0.00103 weeks <br />2.3591e-4 months <br /> on July 28, 1978. The fuel loading sequence is shown in Table 2.0.1 and Figure 2.0.1. Figure 2.0.2 shows fuel assembly location and CEA location by their respective serial numbers. As indicated by Figure 2.0.1, the core was loaded in a seven assembly wide slab from the west face to the east face. Once the slab was completed the south side of the core was loaded after which the north side was loaded.

Neutron multiplication during the fuel loading was monitored using four detector channels, the two permanently installed startup channels and two temporary channels. Inverse multi-plication plots were maintained for each detector channel throughout the loading to provide assurance that the reactor remained subcritical at all times.

Following completion of the fuel loading, a loading verification and alignment check was performed to provide the final position, identifi 2 tion, and orientation check of fuel assemblies, CEA's and sources. No major problems were encountered during the fuel loading.

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4 TABLE 2.0.1 FUEL ASSEMBLY LOADING SEQUENCE STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 1 AKC303 Neutron Source 1 A-8 2 AKC306 65 A-7 3 AKBD42 -

B-7 4 AKA014 51 B-8 5 AKC312 64 A-9 6 AKBT04 -

B-9 7 AKC414 -

A-10 8 AKC407 -

A-6 9 AKB007 -

C-8 10 AKA007 32 C-7 11 AKB043 -

C-6 12 AKC216 -

B-5 13 ACA021 44 C-5 14 AKA040 31 C-9 15 AKB039 -

C-10 16 AKC205 -

B-ll 17 AKA026 43 C-ll 18 AKB054 -

D-11 19 AKA022 23 D-10 20 AKB034 -

D-9

,21 AKA001 13 D-8 22 AKB033 -

D-7 23 AKA104 24 D-6 24 AKB041 -

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5 Table 2.0.1 (CONT'D)

STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 25 AKA047 F E-5 26 AKB011 -

E-6 -

27 AKA010 16 E-7 28 AKB020 -

E-8 29 AKA028 15 E-9 30 AKB051 -

E-10 31 AKA018 E E-11 32 .AKB052 -

F-ll 33 AKA108 10 F-10 34 AKB038 -

F-9 35 AKA009 7 F-8 36 AKBT01 -

F-7 37 AKA024 11 F-6 38 AKB009 -

F-5 39 AKA015 17 G-5 40 AKB023 -

G-6 41 AKA019 3 G-7 42 AKB026 -

G-8 43 AKA039 2 G-9 44 AKB028 -

G-10 45 AKA012 14 G-ll 46 AKB022 -

H-ll

'47 AKA033 6 H-10 48 AKB008 -

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STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE _ LOCATION 49 AKA005 1 H-8 50 AKB050 -

H-7 -

51 AKA107 8 H-6 52 AKBT03 -

H-5 53 AKA109 18 J-5 54 AKB049 -

J-6 55 AKA042 4 J-7 56 AKB027 -

J-8 57 AKA034 5 J-9 58 AKB024 -

J-10 59 AKA036 21 J-ll 60 AKB019 -

K-11 61 AKA016 13 K-10 62 AKB025 -

K-9 63 AKA037 9 K-8 64 AKB010 -

K-7 65 AKA003 12 K-6 66 AKB021 -

K-5 67 AKA004 G L-5 68 AKB035 -

L-6 69 AKA046 19 L-7 70 AKB048 -

L-8

'71 AKA008 20 L-9 72 AKB036 -

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STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 73 AKA017 H L-ll 74 AKBT02 -

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75 AKA020 28 M-10 76 AKB015 -

M-9 77 AKA038 D M-8 78 AKB037 -

M-7 79 AKA110 27 M-6 80 AKB029 -

M-5 81 AKA043 47 N-5 82 AKB002 -

N-6 83 /..*A0 23

_ 35 N-7 84 AKB045 -

N-8 85 AKA031 36 N-9 86 AKB012 -

N-10 87 AKA051 48 N-11 88 AKC210 -

P-ll 89 AKC105 60 P-10 90 AKB005 -

P-9 91 AKA011 58 P-8 92 AKB013 -

P-7 93 AKC203 -

P-5 94 AKC404 -

R-6

'95 AKC310 70 R-7 96 AKC305 Neutron Source 2 R-8

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8 Table 2.0.1 (CONT'D)

STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 97 AKC104 56 B-6 98 ACK304 71 R-9 -

99 AKC401 -

R-10 100 ACK402 -

P-12 101 AKC201 -

N-12 102 AKA027 41 M-12 103 AKB018 -

L-12 104 AKA106 29 K -12 105 AKB014 -

J-12 106 AKA035 A H-12 107 AKBT05 -

G-12 108 AKA013 22 F-12 109 AKB016 -

E-12 110 AKA002 38 D-12 111 AKC206 -

C-12 112 AKC416 -

B-12 113 AKC502 63 C-13 114 AKC207 -

D-13 115 AKA029 42 E-13 116 AKB046 -

F-13 117 AKA032 30 G-13 118 AKB040 -

H-13

'119 AKB044 -

J-13 120 AKA045 -

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9 Table 2.0.1 (CONT'D) 6TEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 121 AKA045 49 L-13 122 AKC208 -

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123 AKC504 72 N-13 124 AKC403 -

M-14 125 AKC212 -

L-14 126 AKC101 61 K-14 127 AKB047 -

J-14 128 AKA025 50 H-14 129 AKB004 -

G-14 130 AKC102 54 F-14 131 AKC209 -

E-14 132 AKC415 -

D-14 133 AKC405 -

F-15 134 AKC302 62 G-15 135 AKC309 -

H-15 136 AKC301 73 J-15 137 AKC406 -

K-15 138 AKC410 -

P-4 139 AKC204 -

N-4 140 AKA049 40 M-4 141 AKBT06 -

L-4 142 AKA030 26 K-4 143 AKB032 -

J-4 144 AKA103 C H04

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STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 145 AKB030 -

G-4 146 AKA048 25 F-4 .

147 AKB017 -

E-4 148 AKA006 39 D-4 149 AKC215 -

C-4 150 AKC408 -

B-4 151 AKC503 66 C-3 152 AKC213 -

D-3 153 AKA041 45 E-3 154 AKB031 -

F-3 155 AKA050 33 G-3 156 AKB001 -

H-3 157 AKA101 34 J-3 158 AKB053 -

K-3 159 AKA102 46 L-3 160 AKC214 -

M-3 161 AKC501 69 N-3 162 AKC412 -

M-2 163 AKC211 -

L-2 164 AKC103 58 K-2 165 AKB003 -

J-2 166 AKA044 52 H-2 l'67 AKB006 -

G-2 168 AKC106 57 F-2

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STEP NO. FUEL ASSEMBLY NO. CEA NO. CORE LOCATION 169 AKC202 -

E-2 170 AKC413 -

D-2 -

171 AKC409 -

F-1 172 AKC308 67 G-1 173 AKC311 -

H-1 174 AKC307 68 J-l 175 AKC411 -

K-1 176 AKC107 59 P-6 177 AKC108 55 B-10

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ARKANSAS POWER 6 LIGHT FUEL ASSEMBLY INDEXING CORE LOADING SEQUENCE A B C D E F G H J K L M N- P R .

1 171 172 173 174 175 170 169 168 167 166 165 164 163 162 2 151 152 153 154 155 156 157 158 159 160 161 3 159 149 148 147 146 145 144 143 142 141 140 139 138 4 5

12 13 24 25 38 39 52 53 66 67 80 81 93 8 97 11 23 26 37 40 51 54 65 68 79 82 176 94 6 2 3 10 22 27 36 41 50 55 64 69 78 83 92 95 1 4 9 21 28 35 42 49 56 63 70 77 84 91 96 8 9

5 6 14 20 29 34 43 48 57 62 71 76 85 90 98 7 177 15 19 30 33 44 47 58 61 72 75 86 89 99 10 16 17 18 31 32 45 46 59 60 73 74 87 88 11 112 12 111 110 109 108 107 106 105 104 103 102 101 100 113 114 115 '116 117 118 119 120 121 122 123 14 132 131 130 129 128 127 126 125 124 15 133 134 135 136 137 9

FIGURE 2.0-1 967 ~45

13-1 ARKANSAS NUCLEAR ONE UNIT 2 CORE INVENTORY MAP A B C D E F G H J K L M N- P R I

1 2

3 SECTIO 4A SEo' TION B 4 5

6 7

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10 S ECTICND SE :TIOB C 11 12 13 14 15

, FIGURE 2,0-2 (Page 1 of 5) 96/ 266 .

13-2 ARKANSAS NUCLEAR ONE UNIT 2 CORE INVENTORY MAP SECTION A A B C D E F G H AKC409 AKC308 AKC311 .

i 67 AKC413 AKC202 .AKC106 AKB006 AKA044 2

, 57 52 AKC503 AKC213 AKC041 AKB301 AKA050 AKB001 66 45 33 AKC408 AKC215 AKA006 AKB017 AKA0.48 AKB030 AKA103 4 39 25 C AKC216 AKA021 AKB041 AKA047 AKB009 AKA015 AKBT03 5 44 F 17 AKC407 AKC1J4 6 AKB048 AKA104 AKB011 AKA024 AKB023 AKA107 56 24 11 8 AKC306 AKB042  : AKAd07 AKB033 AKA010 AKBT01 AKA019 7 AKB050 52 16 3 FIGURE 2.0-2 (Page 2 of 5)

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13-3 ARKANSAS NUCLEAR ONE UNIT 2 CORE INVENTORY MAP .

SECTION B J K L \t N P R ACK307 ACK411 1 68 AKB003 AKC103 AKC211 AKC412 2

. 58 AKA101 AKB053 AKA102 AKC214 AKC501 3 34 46 69 AKB032 AKA030 AKBT06 AKA049 AKC204 AKC410 4 26 40 AKA109 AKB021 AKA004 AKB029 AKA043 AKC203 5 18 G 47 AKA049 AKA003 AKB035 AKA110 AKB002 AKC107 AKC404 12 27 59 6 AKA042 AKB010 AKA046 AKB037 AKA023 AKB013 AKC310 4 19 35 70 7.

FIGURE 2.0-2 (Page 3 of 5)

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13-4 ARKANSAS NUCLEAR ONE UNIT 2 CORE INVENTORY MAP SECTION C J K L M N P R I

AKB027 AKA037 AKB048 AKA038 AKB045 AKA011 AKC305 8

9 D 35 AKA034 AKB025 AKA008 AKB015 AKA031 AKB005 AKC304 9 5 20 36' 71 AKB024 AKA016 AKB036 AKA020 AKB012 AKC105 AKC401 10 13 28 60 4

AKA036 AKB019 AKA017 .KBT02 AKA051 AKC210 11 21 H 48 Q4*

AKB014 AKA106 AKB018 AKA027 AKC201 AKC402 12 29 41 .

I AKA105 AKB044 AKAl45 AKC208 AKC504 13 37 49 72 AKB047 AKC101 AKC212 AKC403 61 AKC301 AKC406 15 73 ,

.s FIGURE 2.0-2 (Page 4 of 5)

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13-5 ARKANSAS NUCLEAR ONE UNIT 2 CORE INVENTORY MAP SECTION D A B C D E F G H AKC303 AKA014 AKB007 AKA001 AKB020 AKA009 AKB026 AKA005 8 51 7 AKC312 AKBT04 AKA040 AKB034 AKA028 AKB038 AKA039 AKB008 9 64 31 15 2 AKC414 AKC108 AKB039 AKA022 AKB051 AKA108 AKB028 AKA033 55 23 10 6 10 s

AKC205 AKA026 AKB054 AKA018 AKB052 AKA012 AKA022 11 43 E 14 AKC416 AKC206 AKA002 AKB016- AKA013 AKBT05 AKA035 12 38 22 A AKC502 AKC207 AKA029 AKB046 AKA032 AKB040 13 63 42 30 14 AKC415 AKC209 AKC102 AKB004 AKA025 54 50 I

AKC405 AKC302 AKC309 62 FIGURE 2.0-2 (Page 5 of 5)

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14 3.0 POST CORE HOT FUNCTIONAL TEST INTRODUCTION The Post Core Hot Functional Test was composed of those tests required to be carried out prior to initial criticality which required the presence of the fuel and all reactor internals for their performance. The Post Core Hot Functional Test (HFT) phase was a period of hot operations and integrated testing of the Reactor Coolant System (RCS) and associated auxiliary sys-tems following the initial fuel load of the reactor.

The major objectives of the Post Core Hot Functional Test were to verify that all necessary plant systems were operable, that the operations personnel were familiarized with the operation of the integrated systems, and that the initial conditions for criticality were met. The test program essentially took the plant from the cold shutdown condition at the end of the core loading to hot standby at 545 F, 2250 psia, and then to 260 F, 460 psia for initial criticality. In addition to the major objectives, numerous specific objectives were satisfied during the Post Core Hot Functional Test program and are described in this section (see Table 3.0.1).

The successful completion of the Post Core Hot Functional Test program ensured that the plant was ready for initial criticality.

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POST CORE IIOT FUNCTIONAL TESTS 15 RCS Temperature / Pressure Plateau vs. Tests TABLE 3.0.1 120 F 160 F 200 F 260 F 29C T 320 F 360 F 400 F s50 50 50 460 460 460 460 460 TESTS psia psia psia psia psia psia psia psia Intercomparison of PPS, CPCs & Proc. Comp. Inputs X X Reactor Coolant System Flow Measurements CEDM/CEA Testing X X Primary and Secondary Water Chemistry Data X X Fixed Incore Instrumentation Verification X X X X X Movable Incore Instrumentation Operation Verification X Pressurizer Spray Valve and Control Adjustment Reactor Coolant System Leakage Measurement Reactor Coolant System Expansion Measurement X X X X X X X Emergency Feedwater System /Waterhammer

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POST CORE Il0T FUNCTIONAL TESTS 16 RCS Temperature / Pressure Plateau vs. Tests TABLE 3.0.1 (cont'd) 120 F 160 F 200 F [ 260 F 290*F 320 F 360 F 400 F '

~50 50 50 460 460 460 460 460 TESTS psia psia psia psia psia psia psia psia Reactor Coolaat System Cold Leg Restraint Gap Meas. X X X Pipe / Component llot Deflection Preoperational Checks X X X X Steady State Vibration Test X Safety Injection System Check Valve Retest Chemical and Volume Control System Integrated Test Core Protection Calculator / Reactor Trip Response Time X CEA Exercise Tests Main Steam Safety Valve Piping Dynamic Transient Emergency Feedwater Pump Turbine Dynamic Transient Secondary llydrostatic Test X*

sc ARCS Pressure 175 to 375 psia, Temperature >90 F c7s se k'

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POST CORE HOT FUNCTIONAL TESTS 17 RCS Temperature / Pressure Plateau vs. Tests TABLE 3.0.1 (cont'd)

, 450 F 470 F 500 F 530 F 540 F 545 F 260*F 1100 1550 1750 2000 2250 2250 460 TESTS psia psia psia psia psia psia psia Intercomparison of PPS, CPCs, & Proc. Comp. inputs X X X X X X Reactor Coolant System Flow Measurements X CEDM/CEA Testing X Primary and Secondary Water Chemistry Data X X Fixed Incore Instrumentation Verification X X X X X Movable Incore Instrumentation Operation Verification y Pressurizer Spray Valve and Control Adjustment X Reactor Coolant System Leakage Measurement X Reactor Coolant System Expansion Measurement X X X X X Emergency Feedwater System /Waterhammer X X Reactor Coolant System Heat Loss Measurement X s '.

E[ Emergency Feedwater System Flow Settings X N

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POST CORE IIOT FUNCTIONAL TESTS 18 RCS Temperature / Pressure Plateau vs. Tests TABLE 3.0.1 (cont'd)

' 260 F 450 F 470 F 500 F 530*F 540*F 545*F 1100 1550 1750 2000 2250 2250 460 TESTS psia psia psia psia psia psia psia _

Reactor Coolant System Cold Leg Restraint Gap Meas. X X Pipe / Component llot Deflection Preoperational Checks X X X Steady State Vibration Test X Safety Injection System Check Valve Retest X Chemical and Volume Control System Integrated Test X Core Protection Calculator / Reactor Trip Response Time CEA Exercise Tests X Main Steam Safety Valve Piping Dynamic Transient X Emergency Feedwater Pump Turbine Dynamic Transient X Secondary llydrostatic Test

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19 3.1 NSSS TESTS 3.1.1 INTERCOMPARISON OF PPS, CPC'S, MAIN CONTROL BOARD AND PROCESS COMPUTER INPUTS TESTS 3.1.1.1 Purpose The PPS, CPC's, Main Control Board Process Instruments, Core Operating Limits Super-visory System (COLSS), and the Plant Com-puter have many parameters with common inputs. This procedure verifies the agree-ment between inputs as specified by the acceptance criteria.

3.1.1.2 Test Method The comparisons were performed at various temperature and pressure plateaus as the primary plant was brought from cold shut-down to a hot standby mode. (Refer to Table 3.1.1 for plateaus.)

3.1.1.3 Test Results The data taken at each plateau demonstrated that the inputs to the CPC's, PPS, Main Con-trol Board Process Instruments, COLSS, and the Computer were in agreement as required.

TABLE 3.1.1 Intercomparison Plateaus Test No. Temperature ( F) Pressure (psia) 1 260 460 2 360 460 3 450 1100 4 470 1550 5 500 1750 6 530 2000 7 540 2250 8 545 2250

- 3.1.1.4 Conclusion All acceptance criteria were met showing accurate and consistent comparisons of the inputs to the PPS, CPC's, Main Control Board Process Instrumer.ts, COLSS, and the computer.

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20 3.1.2 REACTOR COOLANT SYSTEM FLOW MEASUREMENTS TESTS 3.1.2.1 Purpose A. The Purpose in perf arming this test was to determine the characteristics of and collect data for the Reactor Coolant System (RCS) as follows: .

a. Determine the post-core RCS flow rates and pressure drops.

b. Determine the Reactor Coolant Pump (RCP) coastdown flow characteristics.

c. Collect data on the operation of the flow related po?.:tions of the Core Ope-rating Limits S2pervisor System (COLSS) and the Core Protection Calculators (CPC's) for steady state and transient flow conditions and adjust the appli-cable constants as necessary to reflect operating conditions.

d. Verify that the CPC's provide a reactor trip signal consistent with the trip times predicted by the CPC Fortran model.

B. In addition to the RCS flow-related objec-tives stated above, one portion of this test was designed to verify the response time as-sociated with a Low Departure from Nucleate Boiling Ratio (DNBR) CPC trip as described in the Technical Specifications (Section 3/4.3.1, Table 3.3-2, Page 3/4 3-6).

3.1.2.2 lest Method The test was performed at Hot Zero Power con-ditions of 545 F 10.5 F and 2250 psia, pres-surizer level was maintained at 34% 1 1%.

Data was obtained for both transient and steady state conditions for 26 various RCP configura-tions. These configurations are summarized in Table 3.1.2.1 and are given in more detail in Table 3.1.2.2.

For the purpose of verifying the conservative operation of the flow-related algorithms of the CPC's as well as to determine the DNBR trip times at various power levels, special Floppy test discs were loaded into the CPC's prior to performing the test runs. These discs simulated zero power ARO conditions on CPC Channel A; 50%

power, ARO on Channel B; 80% power, AR0, on Channel C; and 100% power, ARO, on Channel D.

s ll 6 / *2 ! I

21 CPC trip times were recorded by monitoring the RCP breaker status and CPC DNBR trip signals.

All continuous data collection was taken via BRUSH 260 strip chart recorders and a PDP 1104 mini-computer. In addition to continuous moni-toring, data was also obtained from the plant computer and Main Control Board periodically as required. .

The portion of the test which measured CPC trip response time from a low DNBR signal at simulated hot full power conditions was performed after run #26. Hot full power (HFP) conditions were simulated by loading the CPC channels with special F oppy test discs, different from those used pre-viously. The test was initiated by simultaneously tripping two diametrically opposed RCP's. The time delay was measured using a high speed Visi-corder. The above sequence was conducted 6 times, once for each CPC channel, each time alter-nating two RCP's.

The difference between the two portions of the test which measured low DNBR CPC trip response times is that the portion performed during the main bcav 'cf the test was necessary to verify Fortran mode.ing of the CPC's. The portion performed following Run #26 was necessary to demonstrate that the CPC's would generate a low DNBR trip signal within the tile specified by the technical Specifications.

3.1.2.3 Test Results The four-pump steady state volumetric flow rate was measured to be 361,468 GPM or 112% of de-sign. The corresponding 6m ss flow rate was de-termined to be 137 x 10 lbm/hr or 114% of de-sign.

Flow coastdown characteristics were recorded and determined from COLSS and CPC data acquisition.

Also, RCP AP's, speeds, and RCS temperatures were recorded on the BRUSH 260 chart recorders.

Paseline data for RCS pressure drops were re-corded on the BRUSH 260 recorders from reactor vessel AP signals.

An initial set of CPC flow constants FC1 and FC2 were supplied by CE Windsor in order to ensure the conservative operation of the CPC's through the 50% power testing. Further extensive analysis i of the RCP coastdowns yielded a final set of CPC flow constants to be incorporated prior to leaving the 50% plateau.

46l U

-l li O

22 Furthermore, the RCP coastdowns have been ex-amined and compared to those assumed in the safety analyses to assure conservatism.

CPC operation for both steady state and tran-sient conditions was verified via the data acquisition and subsequer.t reduction of such parameters as mass flow rate, DNBR trip times, and DNBR margin versus time. COLSS flow re-lated operations were verified via analysis of the COLSS snapshots collected throughout the test. COLSS addressable constants were ad-justed to reflect actual measured four pump steady-state and three-pump steady-state RCS flow, inclusive of the reverse flow algorithm.

The CPC's provided a trip signal on low DNBR within the acceptable times in all cases, (see Table 3.1.2.3).

The CPC Low DNBR Trip response times were ac-ceptable in all cases, (see Table 3.1.2.4).

t 3.1.2.4 Conclusions All objectives and acceptance criteria for this test have been met.

RCS flow for steady state four pump operation is 361,468 GPM cr 112% of design flow. The CPC's have demonstrated acceptable operation with re-gard to RCS flow considerations and have initiated low DNBR trip signals well below the time inter-vals given in the Technical Specifications.

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23 TABLE 3.1.2.1: RCP CONFIGURATION

SUMMARY

1. OF RUNS RCP DATA THIS CONFIGURATION CONFIGURATION COLLECTED 8 All 4 RCP's Operating Steady State 1 All 4 RCP's Stopped Steady State 1 4-lose-4 RCP Ceastdown Transient 2 4-lose-2 RCP Coastdown Transient 4 4-lose-1 RCP Coastdown Steady State 4 3 RCP's Operating, 1 Stopped Steady State 2 2 RCP's Operating, 2 Stopped Steady State 4 1 RCP Operating, 3 Stopped Steady State 26 h

24 TABLE 3.1.2.2: RCS FLOW & OCASTDOWN TEST SEQUENCE Test RCPs RCPs Static Noise Transient Run Running

• Tripped Data Data Data 1 A,B,C,D None X X 2 A,B,C,D X 3 None A,B,C,D X 4 A,B,C,D None X 5 A X 6 B,C,D A X X 7 A,B,C,D None X 8 B X 9 A, C,D B X 10 A,B,C,D None X 11 B,C X 12 A, D B,C X X 13 A B,C,D X 14 D A,B,C X

'5 A,B,C,D None X 16 C X 17 A,B, D C X 18 A,B,C,D None X 19 D X 20 A,B,C D X 21 A,B,C,D None X 22 A,D X 23 B,C A,D X X 24 B A, C,D X 25 C A,B, D X 26 A,B,C,D None X

• A corresponds to 2P32A, B to 2P32B, C to 2P32C, D to 2P32D

$s c,1 =

25 TABLE 3.1.2.3: CPC TRIP TIME 3 FOR FLOW COASTDOWNS I

Coastdown & Expected Range (sec.) Actual Trip Channel Power Min. Max. Time (sec.) Erro:- (sec. )

Run 2, 4/4 A 0 1.413 1.962 1.428 0 B 50 1.236 1.443 1.243 0 C 80 .297 .443 .370 .O D 100 .206 .355 .2T' 0 Run 11, 2/4 A 0 1.979 3.205 ' EC3 0 2

B 50 1.373 1.631 1.400 0 C 80 .638 .840 .655 0 D 100 .288 .445 .289 0 Run 22, 2/4 A 0 1.963 3.114 2.646 0 2

B 50 1.332 1.628 1.358 0 C 80 .616 .825 .646 0 3 .031 D 100 .296 .453 .265 Run 5,1/4 A 0 No Trip No Trip No Trip 2

B, 50 1.322 1.594 1.424 0 C' 80 1.32; 1.594 1.384 0 D 100 .394 .570 .476 0 Run 8, 1/4 A 0 No T-ip No Trip No Trip 2

B 50 1.386 1.660 1.450 0 2

C 80 1.386 1.660 1.475 0 D 100 .456 .660 .475 0 Run 16, 1/4 A 0 No Trip No Trip No Trip 2

B 50 1.393 1.667 1.422 0 C 80 1.393 1.667 1.455 0 D 100 .478 .666 .409 .069 Run 19, 1/4 A 0 No Trip No Trip No Trip 2

B 50 1.384 1.638 1.446 0 2

C 80 1.384 1.638 1.491 0 D 100 .497 .702 .510 0 1 Actual trip times plus 0.0833 seconds for breaker delay time.

2 90% RCP speed trip times.

3 Even though these are outside of the acceptance range, they are

-acceptable because they are conservative.

%7 282

26 TABLE 3.1.2.4: CPC RESPONSE TIME TEST RESULTS REQUIRED

• MEASURED PARAMETER RESPONSE TIME RESPONSE TIME Channel A, Low DNBR, RCPs A & D 10.80 sec. 0.58 sec.

Channel B, Low DNBR, RCPs B & C 10.80 sec. . 0.53 sec.

Channel C, Low DNBR, RCPs A & D 10.80 sec. 0.53 sec.

Channel D, Lev DNBR, RCPs B & C $0.80 sec. 0.53 sec.

• Technical Specification s

e

\

3.1.3 CEDM/CEA TESTS 3.1.3.1 Purpose The purpose of this test was to demonstrate that the Control Element Drive Mechanism Control Sys-tem (CEDMCS) and its associated peripherals, i.e.,

the Plant Computer, CPC/CEAC's, and the Operator Console, control the motion of the CEDM/CEA's and provide consistent position indication between devices. In particular the following objectives were defined:

A. Check the position indication systems and verify the proper functioning of the CEDM limits and alarms; B. Verify that all individual CEA's (81) have the proper drop time from a fully withdrawn position to the 90% insertion position; C. Verify the proper operation of the CEDM hald-ing bus for each subgroup; and D. Determine the CPC and PPS integrated reactor trip response time for both DNBR and LPD trip pa rame te r.s . These were limited to trip con-ditions caused by a dropped CEA.

3.1.3.2 Test Method CEDM/CEA testing was execut-1 at three stable plateaus of the HFT controlling procedure 2.650.01:

TEST RCS RCS RCS CONDITION FLOW PRESSURE TEMPERATURE 1 Shutdown Cooling N 50 psia 5 200 F 2 One pump each loop 460 psia 260 F 3 Four Pumps 2250 psia 545 F The CEDM Holding Busses were tested between the 5 200 F and 260 F RCS temperature plateaus, and the CPC response time testing was performed after the final rod drop sequence at the 545 F plateau.

b

28 All CEA movements during this test were executed in the Manual Individual mode. 4 Honeywell Visi-corder was used throughout the t'st to monitor the 5 motor coil (i.e., upper gr..pper, lower gripper, lift, load transfer, and pulldown) cur-rent traces. Startup nuclear instrumentation channels 1 and 2 were monitored throughout the test for any unexpected reactivity changes as-sociated with CEA motion. Whenever possible, all 81 CEA's were withdrawn to approximately 6 inches to verify the longevity of the CEDM power supply /

switch operation. All CEA disconnect breakers were de-energized except for the CEA under test.

A. CEA Testing at Cold Shutdown (0 RCP's Running):

At this plateau, a preliminary operational check was performed on each CEA before actual testing.

The CEA was energized and withdrawn to approxi-mately 10 inches, then inserted to its lower electrical limit. The rod was subsequently tripped and re-energized at the disconnect breaker. After each preliminary check, each CEA was individualy withdrawn to its upper electrical limit. Periodic stops were made at 12 in., 36 in., 60 in., 84 in., 108 in., and 132 in. during withdrawal to take coil traces and verify appropriate position indication.

The CEA was then inserted to its lower electri-cal limit with periodic holds at the aforemen-tioned points to complete the data aquistion.

After tripping and re-energizing the CEA, a continuous withdrawal to the upper electrical limit was performed to determine CEA withdrawal rate. After connecting the appropriate reed switch position transmitter (RSPT) signal to the visicorder, and initializing the computer rod drop program, the CEA was dropped to the bottom by opening its associated disconnect breaker. The 90% and 100% insertion times were then determined for the CEA by examining the visi-corder trace and the computer output. Upon com-pletion of the tests described above, for all 81 CEA;s, the two fastest and two slowest rods were retested a total of 3 times to determine the drop characteristics of the most limiting CEA's.

B. Hold Bus Test:

The Hold Bus Test was performed during the heat up to the 260 F plateau. Each CEA Subgroup (20) was placed on it appropriate holding bus. This was done by withdrawing each CEA in the subgroup to 6 inches and energizing the corresponding sub-group maintenance switch at the CEDMCS cabinets.

yhf b

29 It was then verified that rod motion for that subgroup was not possible from the operators console. The system was also tested to verify that only one subgroup could be assigned to a holding bus at a time.

C. CEDM Power Supply Burn In Test:

Whenever possible, all 81 CEA's were energized and withdrawn 6 inches to verify proper oper-ation of the CEDM power supplies.

D. CEA Drop Time Tests (260 l', 2 RCP's Operating):

At this plateau, only CEA drop time testing was performed. Each CEA was withdrawn to its upper electrical limit and dropped. The drop traces were analyzed for 90% and 100% insertion times using the same method. The two fastest and two slowest CEA's were retested.

E. CEDM/CEA Testing at 545 F, 2250 psia, (4 RCP's Running):

Testing at this plateau was identical to testing with the RCS <200 F with the following exceptions:

a. Fewer rod motion verification traces were taken, and

b. Since the various alarms checked during cold shutdown were not temperature dependent, they were not verified at this plateau.

F. CPC/PPS DNBR/HLPD Trip Response Time Testing:

Each CPC and CEAC channel were reprogrammed through interactive commands via a test disc and the teletype. This was to simulate full power operating conditions to the calculators, thus bypassing certain level inputs.

CPC A CPC B CPC C CPC D DNBR DNBR DNBR DNBR TEST CHANNEL HLPD HLPD HLPD HLPD

• CPC-A (DNBR) L T B B B T B B CPC-B (DNBR) T L B B T B B B CPC-C (DNBR) B T T B B T T B CPC-D (DNBR) B B T L B B T B l6/ 286

30 CPC A CPC B CPC C CPC D DNBR DNBR DNBR DNBR TEST CHANNEL HLPD HLPD HLPD HLPD CPC-A (HLPD) B T B B L T B B CPC-B (HLPD) T B B B T L B . B CPC-C (HLPD) B T B B B T L B CPC-D (HLPD) B B T B B B T L

• L = Live

• T = Tripped

• B - Bypass The above test matrix was employed to ensure that only the channel under test tripped, with the required two out of four PPS logic. In each case the test channel was placed in bypass long enough to withdraw a target CEA for that channel to its upper electrical limit.

The reed switch position transmitter upper electrical limit. The reed switch position transmitter (RSPT) for that CEA was then used as a live input to tb?

test channel. A visicorder was connected t. che fol-lowing signals:

a. The target CEA RSPT,

b. The target CEA Coil Monitor,

c. The CEA #1 Coil Monitor, and The Test Channel DNBR/HLPD trip output contact.

The test sequence was initiated by tripping the target CEA at its disconnect breaker, and monitoring the trans-ient with the visicorder. The total CPC/PPS response time was determined from the visicorder trace by mea-suring the time between the target CEA trip, and the interruption of current to CEA #1. All four CPC chan-nels were tested in the above manner for the DNBR cnd HLPD trip parameters.

3.1.3.3 Test Results General:

All 81 CEA's were tested per procedure at the three specified plateaus (<200 F, 260 F, and 545 F).

967 287

31 Few CEA's were tested successfully without some form of hardware trouble shooting required. Although problems experienced with rod motion were diverse, three general types of components were responsible for a majority of the failures. These were:

1) RCA CMOS (Complementary Metal Oxide Semiconductor) chips. ,

2) Timer board strip switches, and

3) Monsanto Opto Isolators.

Due to the repetitive nature of the aforementioned CEDMCS component failures, all CEDM Time Board strip switches were returned to the vendor for repair. In addition, all power switch opto isolators were replaced as well as ap-proximately 3500 other CMOS logic chips.

The above components are all part of the CEDMCS logic cicuitry i.e., upstream of the coil power switches. Only one CEA drive motor was found defective. This was CEDM

1. 39 which developed a low resistance condition on its upper gripper coil. The coil stack assembly was replaced and CEDM-39 was successfully tested.

It was intended that the computer drop time program be used at all three test plateaus to verify its calculational ac-curacy. However, a review of cold rod drop data showed that the computer program gave inaccurate and nonconser-vative results. This was due to a combination of software errors and program loading priority. As a result, use of the computer program was discontinued.

RCS < 200 F:

The following table summarizes the data taken at 160 F:

FASTEST SLOWEST PARAMETER DATA /CEA DATA /CEA RANGE UNITS 90% Drop Time 2.19/#2 2.48/#80 .29 sec.

100% Drop Time 2.45/#1 2.78/#80 .33 sec.

90% Drop Time (PL)* 2.17/#26 2.22/#25, .05 sec.

1. 27 100% Drop Time (PL)* 2.42/#26 2.51/#25 .09 sec.

CEA Speed 29.64/#18 30.5/#12 .86 in/ min CEA Speed (PL)* 19.85/#25 20.34/#26 .49 in/ min

• (PL) = Part length CEA RCS 260 F o [. j

32 The following table summarizes the data taken at 260 F:

FASTEST SLOWEST PARAMETER DATA /CEA DATA /CEA RANGE UNITS 90% Drop Time 2.30/#1 2.58/#71 .28 sec.

100% Drop Time 2.55/#1 2.89/#71 .34 sec.

90% Drop Time (PL)* 2.22/#26 2.29/#28 .07 sec.

100% Drop Time (PL)* 2.42/#24 2.55/#23 .13 sec.

CEA Speed ------- -------- ---

in/ min CEA Speed (PL)* ------- -------- ---

in/ min

• (PL) = Part Length CEA RCS 545 F:

The following table summarizes the data taken at 545 F:

FASTEST SLOWEST PARAMETER DATA /CEA DATA /CEA RANGE UNITS 90% Drop Time 2.42/#2 2.77/#71 0.35 sec.

100% Drop Time 2.69/#2 3.03/#71 0.34 sec.

90% Drop Time (PL)* 2.31/#26 2.42/#23, 0.11 sec.

25 100% Drop Time (PL)* 2.55/#26 2.81/#28 0.26 sec.

CEa sp.ed 30.4/#56 28.8/#31 1.60 in/ min CEA Speti(PL)* 20.11/#25 20.0/#28 0.11 in/ min

• (PL) = Part Length CEA CPC DNBR and HLPD Response Time Test:

All DNBR and HLPD test trips were conducted successfully per procedure. The results of these tests are as follows:

RESPONSE TIME (SEC)

TEST CHANNEL DNBR HLPD CPC-A .94 1.20 CPC-B .82 1.28 CPC-C .91 1.29 CPC-D .85 1.32 Two test runs were made for each CPC channel trip para-meter. The above table includes the most conservative response time.

3.l'.3.4 Conclusion The following applies to the CEDM/CEA's and their associated control systems:

n (t-..l

'lf

33

1) All CEA's have a characteristic 90% insertion drop time of less than three seconds;

2) All regulating and shutdown CEA's move at 30 +

2 in/ min;

3) All part length CEA's move at 20 + 2 in/ min.

4) all CEA position indication systems and alarms operate properly;

5) The CEDM holding busses operate properly for each CEA subgroup;.and

6) All CEA's exhibit a dashpot effect as verified by the drop time and recorder traces between 90% and 100% insertion.

Each CPC/PPS safety channel was tested for its response time to trip for the low DNBR and high LPD trip parameters.

In each case, the CPC/PPS DNBR and HLPD trip times were

<l.58 sec. respectively.

o (3 t

34 3.1.4 PRIMARY AND SECONDARY WATER CHEMISTRY TESTS 3.1.4.1 Purpose The purpose of this test was to verify proper water chemistry control for the primary and secondary cooling systems during post core hot functional testing. Further, to provide the necessary sampling frequency to comply with the Technical Specifications and Chemistry Manual (CENPD-28) and to provide assurance that chem-istry limits are not exceeded. Also, baseline chemistry data was to be established.

3.1.4.2 Test Method Both reactor coolant and secondary system sam-ples were obtained per the sampling frequency specified in CENPD-28. The results of these samplings were recorded on the appropriate plant data sheets.

A. RCS chemistry was maintained with the limits of CENPD-28 by chemical addition, feed and bleed, by the use of purification filters (2F3A and 2F3B), and by the use of ion ex-changers (2T36A, 2T36B, 2T70). These pro-cesses were carried out in accordance with plant operating procedures and instructions from the Radiochemistry Department.

B. Secondary chemistry was maintained with the use of chemical additions, in accor-dance with plant operating procedures and Chemistry Department instructions, and by performing Steam Generator Blowdowns.

3.1.4.3 Test Results A. Primary water chemistry results taken from the hot lab water report for the period of August 30th to November 9th, 1978, are listed below. The figures represent aver-age values for the time period, and are reported in ppm unless otherwise indicated.

I g n 6i -

35 PARAMETER AVERAGE VALUE PH 6.26 Cl- 0.07 F~ 0.04 .

Suspended Solids 0.01 Li~ 0.57 Total Gas 51.4 ml/ liter H 12.7%

2 N 82.2%

2 The values for Cl- and F~ were typically less than .01 ppm, except during plant transients. All the values above are well witdin those specified in CENPD-28.

B. The secondary plant was started up and shutdown numerous times during post core hot functional testing. These transients prevented the gathering of sufficient data to draw representative conclusions as to the adequacy of secondary chemistry control.

3.1.4.4 Conclusion A. This test showed that primary chemistry can be controlled within the specified limits and that the sampling frequency required to verify this is consistent with CENPD-28 specifications.

B. Secondary water chemistry data provided in-conclusive results due to varing plant con-ditions. However, further testing was per-formed during power ascension to verify pro-per secondary chemistry control.

'T h / 2Y'

36 3.1.5 FIXED INCORE INSTRUMENTATION TESTS 3.1.5.1 Purpose The purpose of this test was to verify that the leakage resistance of each fixed incore detector and its ass 9ciated cabling is equal to or greater than 1 x 10 ohms, and to provide a permanent rec-ord of fixed incore thermocouple data duying plant startup. A leakage resistance of 1 x 10 ohms is necessary to ensure proper detector response and to indicate proper electrical continuity between the fixed incore detector, its cabling, and amp-lifier system.

3.1.5.2 Test Method Incore detector leakage resistances were mea-sured at the 545 F/2250 psia plateau i.e. Hot Standby. During the test, incore detector sig-nal leads were disconnected at the incore amp-lifier input, one detector at a time, and leak-age resistance measured for each rhodium and background detector associated with the chosen instrument assembly. A megohmbridge using an applied voltage of 10.0 volts was used for all resistance measurements.

Core exit temperatures, as measured by the thermo-couple in each incore detector assembly at various temperature / pressure plateaus, were recorded via computer printout during the initial heatup to Hot Standby.

3.1.5.3 Test Results All incore detector leak ge resistance values 9

were greater than 1 x 10 ohms. Towever, two rhodium detector strings, specifically Rhodium

1. 3 of detector R-10 and Rhodium #4 of detector N-12, had measured leakage resistances three ordersofmagnitudegreaterthanalltggother detector strings (approximately 1 x 10 ohms).

These two detector strings were examined for continuity with a time domain reflec.ometer, but no indication of open connections was dis-covered. The incore thermocouple data was col-lected and the results tabulated and included with the test data.

3.1.5.4 Conclusion The acceptance criteria requiring fixed incore detector leakage rysistance to be greater than or equal to 1 x 10 ohms for all rhodium and background detectors was verified by the test results. The fixed incore thermocouple readings were collected for baseline data only and no specific acceptance criteria was required.

p6,7 )hJ

37 3.1.6 MOVABLE INCORE INSTRUMENTATION OPERATION VERIFICATION TESTS 3.1.6.1 Purpose The purpose of this test was to:

A. Obtain movable incore detector path length measurements for input to the plant computer.

B. Verify the proper operation of the moveable incore detector system (MICDS) when control-led by the plaat computer.

C. Verify that the detectors can be inserted into all assigned incore locations.

3.1.6.2 Test Method A. Path Length Measurements The movable incore detector path lengths were physically measured at ambient tem-peratures (N 120 F) using dummy detector cabling. The path lengths were remeasured using the encoder and portable control box to verify the physical measurement data at Hot Standby to verify that the effects of expansion were negligible.

B. Computer Control Computer control of the movable incore de-tector system (MICDS) was tested in the manual, semi-automatic, and automatic modes at Hot Standby.

3.1.6.3 Test Results During the testing, several system problems were identified:

A. Encoder and computer interface problems were found and identified.

B. The original rhodium detectors were over-sized and would not fit into the calibra-tion tubes at hot conditions.

C. Computer software problems were encountered.

D. Cable kinking problems prevented insertion of the detector into several drive paths.

E. Two drive machine motors failed.

~I U () I ]()Ik

38 The encoder and computer interface problems were corrected. New Rhodium detectors were ordered but were not present for testing. The MICDS was tested with cabling representative of the actual cable to be used with the incore detectors. The two drive machine motors were replaced and/or re-built and a cable straightening tool was obtained and used on both drive cables to correct the kink-ing problems-Manual, semi-automatic, and automatic computer control modes were verified for selected paths for both MICD A and MICD B. A software modi-fication was instituted to allow automatic mode to be restarted in any given path.

3.1.6.4 Conclusion The movable incore detector path lengths were measured using two different methods and the results were in agreement within the acceptance criteria. The performance of the manual, semi-automatic, and automatic MICDS computer control modes was demonstrated.

• aI i

39 3.1.7 REACTOR COOLANT SYSTEM LEAKAGE MEASUREMENT TESTS 3.1.7.1 Purpose Three methods were used to measure leakage from the RCS. Monitoring equipment in containment was installed to continuously monitor the en-vironmental conditions via remote observation in the Control Room. Instrumentation was also installed to monitor containment sump water level. The third method of leakage detection was a daily operator's calculation of RCS water inventory balancing losses against additions.

This procedure verified that the RCS leakage rate operating procedure gives accurate results for RCS water inventory.

3.1.7.2 Test Method With the RCS at 545 f, 2250 psia, the following parameters were monitored over a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period in order to determine RCS leakage:

A. Containment sump level, B. Safety injection tank level and pres-sure.

C. RCS pressurizer safeties and quenc'1 tank level and pressure, D. Holdup tanks and vacuum degasifier levels, E. Temperature in reactor vessel gasket telltale drain line, F. Flow to sample sink from RCS sample lines, G. RCS and CVCS test connections, drains and vents, H. Heat exchangers having an RCS or CVCS interface, I. Volume control tank level, and J. Charging flow rate.

Data was collected before and after the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold period. Data reduction was performed by two methods: (1) the graphical reduction used in the operating procedure was compared to, (2) the more precise reduction method as detailed in the test procedure.

U6l )U

40

• TECHNICAL SPECIFICATION TEST OPERATING ALLOWED PROCEDURE PROCEDURE LEAKAGE RATE LEAKAGE RATE LEAKAGE RATE  % DIFFERENCE

<l.0 gpm 0.346 gpm 0.342 gpm 1.2

• unidentified source .

3.1.7.3 Conclusion The operating procedure gives results that are consistent with the leak rate determined by the test procedure method and thereby meets the acceptance criteria. The total RCS leakage rate was less than the allowed leakage rate as stated in the Technical Specifications. (Section 3.4.6.2).

s6I -

d

41 3.1.8 REACTOR COOLANT SYSTEM EXPANSION MEASUREMENT TESTS 3.1.8.1 Purpose This test was a repetition of the one performed during Pre-Core HFT because of the many restraints, shims, and snubbers that have been added, or modi-fied, since the original performance of the-test.

The purpose was to demonstrate that the shims in-stalled as a result of the Pre-Core RCS Expansion Measurement Test were correctly sized, and to verify that RCS components (Steam Generators and Reactor Vessel) are free to expand and contract during plant heat up and cooldown.

3.1.8.2 Method A. During the initial heatup, inspections and/

or measurements were made at the following locations.

a. Reactor Vessel lower shear keys

b. Reactor Vessel upper lateral restraints

c. Steam Generator sliding bases

d. Steam Generator snubber assemblies During the initial heatup, visual inspections were made at 160 F, 200 F, 320 F, 400 F, 480 F and 520 F to ensure that the components were free to expand without binding. Measurements were taken at 120 F, 260 F, 360 F, 450 F and 545 F. Due to an unscheduled cooldown, mea-surements required at 260 F and 120 F, during cooldown, were not taken. The test was run a second time in order to obtain all required data.

B. The following acceptance criteria were de-veloped for the hot standby plateau.

a. The sum of the RV lower shear key di-mensions shall be 2 0.030 in, with any one dimension 2 0.030 in. and 5 0.230 in.

b. The sum of the RV upper lateral re-

, straint dimension shall be 2 030 in.

with any one dimension 1 0.030 in. and 5 0.040 in.

c. The steam generator sliding base di-mensions shall be > 0.200 in. and 5 0.230 in.

n}

f 99

42

d. The steam generator snubber pin to pin distance shall not be greater than 36.750 in.

3.1.8.3 Test Results A. During the initial RCS heatup, none of the components were binding nor did any of.the gaps being monitored constrict to the point of requiring shim adjustment. All required measurements were completed with the excep-tion of the final 545 F and 260 F stable readings.

B. RCS freedom to expand and contract was veri-fied during the first heatup. During the second heatup, all required measurements at stable conditions were performed with one ex-ception of the final 260 F and 120 F ambient readings. These measurements were omitted with the concurrence of Combustion Engine-ering. All measurements were within the required tolerance with the following ex-ceptions:

a. Steam Generator A sliding base

b. Plant 0 reactor vessel upper lateral restraint

c. Plant 120 upper lateral restraint 3.1.8.4 Conclusion No points of contact were observed throughout this test. The RCS expanded and contracted within the tolerances established in the test procedure, with the above exceptions. These were documented and transmitted to Combustion Engineering to expedite an engineering review of the out of tolerance gaps. C.E's response established the acceptability of the gaps as measured.

i (> I

43 3.1.9 RCS COLD I.EG RESTRAINT GAP MEASUREMENT (SHIM'!ING VERI-FICATIONS) TESTS 3.1.9.1 Purpose This test was repeated during PCHF Testing since many restraints, shims and snubbers had been added or modified since original -

performance of the test during precore hot functional testing. The purposes were to:

A. Measure RCS cold leg restraint - restraint lug hot gaps for shim installation veri-fication.

B. Measure RCS cold leg lateral restraint shim surface impact pad hot gap for shim installation verification.

C. Verify RCS cold leg angular restraint cold gaps are within tolerance as specified by the Test Procedure.

D. Verify RCS cold leg piping and insulation do not structurally interfere with re-straints prior to and during RCS heatup to hot standby no load conditions.

3.1.9.2 Test Method After the RCS was stabilized for six hours at 120 F 1 5 F, initial ambient measurements were taken at all gap locations described in A through C above. Final acceptance criteria for the cold leg angular restraint (C), re-quires ambient measurements only, therefore, these were not measured at any further temp-erature plateaus.

During heatup, visual inspections of all gap locations were made at 260 F, 360 F, 450 F and 545 F to assure that cold leg piping and insulation did not structurally interfere with restraints. Certain gaps suspect of contact were also measured at these plateaus to keep a close watch on their movement.

After the RCS was stabilized for six hours at 545 1 5 F, measurements were taken at all lo-cations, with the exception of the angular re-straints. In order to measure any long term thermal expansion of the RCS, these measure-ments were repeated just prior to cooldown fol-lowing hot functional testing.

Y67 300

Due to an unscheduled plant cooldown during PCHF, the measurements required just prior to cooldown, were obtained following the second heatup.

3.1.9.3 Test Results The initial heatup to 545 F yielded several gaps either in physical contact or exceeding clearance requirements. Deficiencies were written against all gaps not meeting acceptance criteria and work was done to correct them during the shutdown period.

Measurements taken following the second heatup showed that the contact points had been cleared but several of the gap clearances did not come within tolerance. Ten of the twenty-four RCS cold leg restraint - restraint lug gaps did not meet the original acceptance criteria of .0625"

+ .0625" - 0". Two of the thirty-two cold leg lateral restraint shim surface-impact pad hot gaps did not meet the acceptance criteria of 0.375" + 0.125" - 0.172".

3.1.9.4 Conclusions All measurement data was forwarded to Bechtel and CE project offices for analysis. Each con-tractor reviewed the data and following analysis, determined the data was acceptable and would meet the design criteria of both CE and Bechtel.

F 67 301

3.1.10 CORE PROTECTION CALCULATOR / REACTOR TRIP RESPONSE TIME TESTS 3.1.10.1 Purpose The objective of this test was to measure the response times associated with the following Core Protection Calculator Trips: ,

A. Departure from Nucleate Boiling Ratio

a. Th

b. Tc

c. Pzr Pressure

d. Neutron Flux Level B. High Local Power Density

a. Neutron Flux Level 3.1.10.2 Test Method The CPC response time tests measured the over-all CPC system response to a single input para-meter change. Two test discs were employed in order to load Response Time Test Software. Test Disc #1 was used to load Ch. A CPC, Ch. B CPC, and Channel B CEAC. Test Disc #2 was used to load Ch. C CPC, Ch. D CPC, and Channel C CEAC.

The test software was identical to that which resides in memory during normal system operations with the exception that the test software allowed one or more live inputs with the rest simulated by the software.

The functional unit tests for the CPC response time testing were divided into two main groups, High Local Power Density (LPD) trips, and De-parture from Nucleate Boiling Ratio (DNBR) trip tests. The test methods were as follows.

A. LPD Trips

a. Neutron Flux Power from Excore Neu-tron Detectors The neutron flux power test was a step change in all excore detector outputs to 116.5% of their initial values.

The test apparatus as shown in Figure 3.1.10.1 was utilized. Total response time measured was frem signal initia-tion, via the current generator, to the pre-amplifiers until the Trip Cir-cuit Breakers opened deenergizing the CEDM coils. The response time was mea-sured with a high speed visicorder.

.- 302

46 B. DNBR Trips

a. Neutron Flux Power from Excore Neu-tron Detectors The DNBR Trip neutron flux power test methodology was identical to the test producing the LPD trip. In this test case, the LPD trip was jumpered out so as to produce a DNBR trip only,

b. Cold Leg Temperature The cold leg temperature test was a step change in one cold leg tempera-ture input to the CPC from 553.5 F to 583.5 F. The step temperature change between the sensor and CPC's was simulated with the use of decade boxes, in conjunction wiht the test apparatus as shown in Figure 3.1.10.2.

For the total response times calcula-tion, the difference (DT) between the computed total response time and the step response time was added to the results of the procedure. This DT accounts for the calculated time dif-ference between a step change and a ramp input to the sensor. Since the test was performed using step changes, the DT was calculated and added to yield accurate time response data.

Measured response time was from sig-nal initiation to CEDM coil deenergi-zation. A high speed visicorder was used to measure the response time.

c. Hot Leg Temperature The hot leg temperature test was a step change in one hot leg tempera-ture input to the CPC from 612.5 F to 642.5 F. The hot leg tempera-ture response time methodology was the same as that used in the cold leg response time test.

Y67 30a

47 D. Reactor Coolant Pressure from the Pressurizer The PZR pressure response time test consisted of a step decrease in pres-sure input from 2175 psia to 2125 psia. The test rig shown in Figure 3.1.10.3 was employed for this test sequence. Test initiation was by means of a fast acting solenoid ope-rating a test valve that allowed ac-cumulator pressure to be applied to the pressure transmitter, thereby decreasing indicated pressure to be-low the DNBR trip setpoint. Response time measured was from the moment of solenoid actuation to CEDM coil -

deenergization.

3.1.10.3 Test Results The results of all response time tests are tabulated in Table 3.1.10.1. For each parameter tested, for all channels, the re-sponse times were within those specified in the ANO - 2, Technical Specifications, Table 3.3-2.

The major problem encountered during the conduct of this test sequence was with noise during the performance of High Neu-tron Flux Power initiated trips. The a-mount of electrical noise initially con-tributed to inaccurate response times of the NI Safety channels. These problems were methodically traced down and elimi-nated. Suspect response time tests were re performed and the results yielded ac-curate and consistent data.

3.1.10.4 Conclusion The intent and objectives of all response time tests conducted were met successfully.

The total response times were within the limits prescribed as shown by the tabu-lated results of Table 3.1.10.1.

967 304

48 TABLE 3.1.10.1 CPC Response Time Test Results Required Parameter Response Time MT(1) DT TT(2)

Ch. A Low DNBR, NI Safety Ch. A 10.39 sec. 0.240 0 0.240 Ch. B Low DNBR, NI Safety Ch. B. 10.39 sec. 0.355 0 0.355 Ch. C Low DNBR, NI Safety Ch. C 10.39 sec. 0.344 0 0.344 Ch. D Low DNBP, NI Safety Ch. D $0.39 sec. 0.229 0 0.299 Ch. A High LPD, NI Safety Ch. A 12.58 sec. 2.533 0 2.53 Ch. B High LPD, NI Safety Ch. B $2.58 sec. 2.475 0 2.48 Ch. C High LPD, NI Safety Ch. C $2.58 sec. 2.375 0 2.38 Ch. D High LPD, NI Safety Ch. D $2.58 sec. 1.527 0 1.53 Ch. A Low DNBR, TH1-4610-1 11.54 sec. 0.191 1.26 1.45 Ch. B Low DNBR, TH1-4610-2 11.54 sec. 0.148 1.26 1.41 Ch. C Low DNBR, TH1-4610-3 $1.54 sec. 0.149 1.26 1.41 Ch. D Low DNBR, TH1-4610-4 11.54 sec. 0.130 1.26 1.39 Ch. A Low DNBR, TH2-4710-1 $1.54 sec. 0.147 1.26 1.41 Ch. B Low DNBR, Th2-4710-2 $1.54 sec. 0.154 1.26 1.41 Ch. C Low DNBR, TH2-4710-3 51.54 sec. 0.148 1.26 1.41 Ch. D Low DNBR, TH2-4710-4 11.54 sec. 0.134 1.26 1.39 Ch. A Low DNBR, TCl-4611-1 $3.79 sec. 0.175 3.49 3.67 Ch. B Low DNBR, TCl-4611-2 13.79 sec. 0.190 3.49 3.68 Ch. C Low DNBR, TCl-4611-3 $3.79 sec. 0.188 3.49 3.68 Ch. O Low DNBR, TCl-4611-4 13.79 sec. 0.181 3.49 3.67 o67 305

49 6

TABLE 3.1.10.1 (cont'd)

- s Required Parameter Response Time MT(1) DT TT(2)

Ch. A Low DNBR, TCl-4711-1 0.171 '.49,3.66 13.79 sec. a Ch. B Low DNBR, TCl-4711-2 $3.79 sec. 0.186 3.49 3.68 Ch. C Low DNBR, TCl-4711-3 $3.79 sec. 0.179 3.49 3.67 Ch. D Low DNBR, TC1-4711-4 13.79 sec. 0.168 3.49 3.66 Ch. A Low PZR Press., PT-4601-1 13.19 sec. 0.257 0.011 0.268 Ch. B Low PZR Press., PT-4601-2 $3.19 sec. 0.204 0.011 0.215 Ch. C Low PZR Press., PT-4601 $3.19 sec. 0.356 0.012 0.368 Ch. D Low PZR Press., PT-4601-4 13.19 sec. 0.261 0.010 0.271 (1) Via a Visicorder (2) TT = DT + MT DT = Differential Time TT = Total Time MT = Measured Time NOTE: 1) All times are in secanis.

2) Differential time is the difference between the computed total response time and the step response time.

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53 3.1.11 PRESSURIZER SPRAY VALVE CONTROL AND ADJUSTMENT TESTS 3.1.11.1 Purpose The purpose of this test was to:

A. Establish the proper flow settings for the Pressurizer continuous spray valves (2RC-8A and 2RC-8B) to minimize thermal stresses across the pressurizer spray nozzle.

B. Measure the rate at which the pressurizer /

reactor coolant system pressure can be re-duced utilizing pressurizer spray flowing through the pressurizer spray valves (2CV-4651 and 2CV-4652) both individually and in parallel.

C. Verify that the pressurizer level control system is capable of automatically main-taining program level.

D. Verify proper nperation of the pressurizer level alarms and the Reactor Coolant Sys-tem high letdown flow alarm.

E. Verify the vibration of the pressurizer spray line is acceptable during pressuri-zer spray operation.

3.1.11.2 Test Method A. Continuous Spray Valve Settings A temporary thermocouple was placed on the spray line piping just before the horizontal run to the pressurizer. Reactor Coolant System temperature / pressure was established at 545 F/2250 psia per the HFT controlling procedure. The temperature of the spray line piping as indicated by the thermo-couple was compared to the average reactor coolant system cold leg temperature. Both continuous spray valves (2RC-8A and 2RC-8B) were then positioned to adjust the spray line temperature to 25 F to 30 F lower than the average reactor coolant system cold leg temperature.

B. Pressurizer Spray Effectiveness Test The Reactor Coolant System was stabilized at the 545 F/2250 psia plateau per the Post Core Hot Functional Controlling Procedure.

tl(ll b

54 Pressurizer Spray Valve 2CV-4651 and Pres-surizer Pressure controller 2PIC-4626A were selected for automatic operation.

The setpoint of 2PIC-4626A was then in-creased to 2300 psia and pressure was al-lowed to stabilize. Next, the setpoint of pressurizer pressure controller 2PIC-4626B was adjusted to 2000 psia. All pres-surizer heaters were deenergized and 2PIC-4626B was then selected for pressure con-trol. The resulting pressure transient was recorded on a strip chart recorder.

Following the pressure transient, control was returned to 2PIC-4626A and pressure was allowed to stabilize at 2300 psia.

Once again control was shifted to 2PIC-4626B and the resulting transient was re-corded on a strip chart recorder. After the second pressure transient control was shifted to 2PIC-4626A and pressure was allowed to stabilize at 2300 psia. The above described pressure transients were repeated, first with only 2CV-4652 in auto-matic and then with both 2CV-4651 and 2CV-4652 in automatic. Upon completion of the final pressure transient, pressure was re-turned to 2250 psia and the pressurizer pressure control system was returned to normal. During the pressure transients, the vibration of the spray line piping was monitored.

C. Pressurizer Level Control Test Reactor Coolant System was stabilized at the 545 F/2250 psia plateau per Post Core Hot Functional Controlling Procedure.

Pressurizer level control, charging pump interlock setpoints, pressurizer heaters level interlock setpoints and level de-viation alarm setpoints were verified utilizing an external test source to vary the pressurizer level setpoint above and below the actual pressurizer level. Pres-surizer level was then manually decreased to verify the low-low level alarm setpoints.

Following the level decrease both pressuri-zer level setpoint and pressurizer level were returned to 40% and the letdown con-troller placed in automatic. Letdown con-trol setpoint was decreased below the act-ual pressurizer level. After the maximum O67i 5

55 letdown flow for automatic operation was recorded, the letdown control valve con-troller was placed in manual and the let-down flow increased to the high letdown flow alarm setpoint, then to approximately 148 gpm where controller data was obtained. Pressurizer level and pressurizer level setpoint were matched, and with the let-down controller in automatic, the level setpoint was increased above the actual pressurizer level. After minimum let-dowm flow for automatic operation was re-corded, the letdown control valve control-1er was placed in manual and letdown flow decreased to approximately 16 gpm where controller data was obtained. Pressurizer level and pressurizer level setpoint were matched and returned to 40 percent and testing secured. 3.1.11.3 Te_st Results A. Pressurizer Continuous Spray Valve The pressurizer continuous spray valves (2RC-8A and 2RC-8B) were positioned at various valve settings to record piping temperatures. The temperatures yielded a temperature difference (delta-T) be-tween the spray line and the average cf . the A & B RCS cold legs. The delta-T and pressurizer heater requirements were recorded. From these, an optimum con-tinuous spray valve setting of one quar-ter (1/4) of one turn was selected. This valve setting resulted in a 38 F delta-T and required either one (1) bank of back-up heaters or two (2) banks of proportional heaters to maintian the RCS pressure at 2250 psia. B. Pressurizer Spray Effectiveness Test The results of this test are listed below: INITIAL FINAL TRANSIENT RATE OF SPRAY PRESSURE PRESSURE TIME PRESSURE (PSIA) (PSIA) (SECONDS) (PSI / MIN) 2CV-4651 Run 1 2300 2050 139.2 108 Run 2 2300 2050 145.4 103 2CV-4652 Run 1 2300 2050 148.8 100.8 Run 2 2300 2050 153.6 97.6 2CV-4651 Run 1 2300 2050 111.36 134 and Run 2 2300 2050 121.92 123.03 2CV-4652 l b\ .

All results of this test met design criteria specifications. C. Pressurizer Level Control Test The results of this test showed that the pressurizer level control system is capable of automatically maintaining the programmed level and that control and alarm setpoints operate properly. 3.1.11.4 Conclusion A. Continuous Spray Valve Settings The continuous spray valves were not set in accordance with the test because the required delta-T of 25 F - 30 F between the spray line nozzle and the average RCS cold leg temperature resulted in excessive use of the pressurizer heaters. Combustion Engineering reviewed the results of the continuous spray valve setting test and stated that a delta-T not exceeding 40 F is acceptable. Hence, the present setting yielding a 38 F delta-T is sufficient to ensure that the pressurizer spray nozzle will not undergo excessive thermal trans-ients. B. Pressurizer Spray Effectiveness Test The prersurizer spray system operated as designed to reduce pressurizer pressure at the rates required in the ecceptance criteria for all spray valve combinations. The pressurizer spray valve line piping vibration was verified acceptable by vis-ual observation during spray operations. C. Pressurizer Level Control Test It was proven that the pressurizer level control system is capable of automatically maintaining the programmed level. Also, proper operation of the control and alarm setpoints were verified. 01 56I -

57 3.1.12 RCS lEAT LOSS TESTS 3.1.12.1 Purpose The purpose of this test was to determine the RCS heat loss under hot, zero power conditions (545 F, 2250 psia). The results are used: A. As a measure of the effectiveness of the RCS insulation, and B. as input to the secondary plant calorimetric which will be used to determine reactor power. 3.1.12.2 Test Method A. The total RCS heat loss was determined using the steamdown method. The plant was ini-tially stabilized with steam generators filled to a specified level. Blowdown and feedwater were then secured and the gene-rators were steamed to a specified target level while RCS conditions were maintained. Energy inputs to the system were determined by monitoring reactor coolant pump and pressurizer heater parameters. Similarly, energy losses were determined by calculating energy transferred out of the steam gene-rators and by monitoring charging flow and temperature . A heat balance was then per-formed to determine the heat loss. B. Pressurizer heat loss was measured by main-taining the plant at essentially constant conditions. The energy input to the sys-tem by the pressurizer heaters was deter-mined by measuring the power input over a specified time interval. The pressurizer heat loss was then equal to the pressurizer heater input. This value was measured for conditions of:

a. no spray to the pressurizer, and

b. bypass spray only to the pressurizer.

3.1.12.3 Test Results A. The total RCS heat loss was detgrmined as described above to be 3.80 x 10 BTH/hr. This was neglecting the heat loss of the Chemical and Volume6Control System (CVCS) which was 3.09 x 10 BTU /hr. Including this term yielded a total heat loss of 6 6.89 x 10 BTU /hr.

                                                     "6/      314

5G B. Results of the pressurizer heat loss mea-sure.,ents were:

a. 5.96 x 105 BTU /hr. for the case of no pressurizgr spray, and

b. 1.04 x 10 BTU /hr. with continuous bypass spray.

3.1.12.5 Conclusions The measured value of the RCS heat loss satis-fled the acceptance criteria. The measured pressurizer heat ] css values, however, were not within the acceptance criteria. Recommendation of the vendor after subsequent evaluation was that the measured pressurizer heat loss values were acc2ptable. r J ' S}(>l i

59 1.13 CIEMICAL AND VOLUME CONTROL SYSTEM INTEGRATED TESR 3.1.13.1 Pu rpose The Chemical & Volume Control System Integrated Test was performed to accomplish the following objectives: A. Demonstrate the proper operation of the CVCS letdown system. This included the following:

a. Back pressure controller and back pressure control valves can maintain pressure downstream of level control valves automatically during pres-surizer level transients without lifting back pressure relief valve and stay within specified limits.

b. Letdown flowrates can be maintained within specified limits.

c. Letdown temperature controller can control letdown temperature in auto-matic within specified limits.

B. Record Purification Filter D/P, Letdown Strainer D/P, and Ion Exchanger D/P at various letdown flow rates for baseline information. C. Verify proper flow rates to the Borono-meter and process radiation monitor. 3.1.13.2 Method The back pressure controller 2PIC-4812 in com-bif" tion with the letdown flow control valves ar.d the back pressure control valves were test-ed at 545 F and 2250 psia plateau by placing the controller in automatic with a specified pressure setpoint set and simulating a pres-surizer level error by adjusting the manual setpoint 5% below level setpoint. This re-sulted in maximum transient on the back pres-sure control valves.

 ,               Letdown and Purification System operations were performed at the 545 F and 2250 psia plateau. Letdown flowrates of 29, 40, 80, 120 and 138 GPM were established with various combinations of letdown flow control valves, back pressure control valves, purification filters, and purification ion exchangerp           f)k b During these tests, data was recorded foh'l}

baseline information.

60 3.13.3 Results The back pressure controller restored the back pressure to the initial value during the per-formance of system transients, and the back pressure relief valve did not lift. The let-down temperature and pressure was regulated by their respective controllers at normal flow-rates. 3.13.4 Conclusions It was verified that the letdown system func-tions properly. The Boronometer and process radiation monitor flowrates were satisfactorily set. e 967 3I7

61 3.1.14 CEA EXERCISE TESTS 3.1.14.1 Purpose The purpose of this test was to verify: A. Proper operation of the CEDMCS B. Proper operation of the CEA related com-puter alarms and interlocks C. CPC indicated group positions D. CPC calculated peaking and penalty factors E. CEAC calculated penalty factors F. CPC calcuiated rod shadowing factors G. Proper transmission of the CEAC penalty factors to the CPC's 3.1.14.2 Test Method The CEA exercise check test was performed dur-ing the Post Core Hot Functional Test sequence at the 545 F/2250 psia plateau i.e. Hot Stand-by. Prior to increasing the neutron population, a determination of base count rate (Co) was per-formed for each of the two startup channels. The scaler output of each startup channel drawer was connected to an individual counter-scaler, thus allowing simultaneous count rates to be obtained from the two separate channels. With the boron concentration at 2038 PPM, CEA Groups A, B, and P were withdrawn one group at a time in manual group mode. CEA group with-drawal was stopped at approximately equal re-activity insertion levels to monitor startup channels 1 and 2 from which count rates were obtained to construct inverse count rate ratio (1/M) plots. When group P reached the upper electrical limit, manual sequential mode was used to withdraw the regulation CEA groups. Again 1/M plots were maintained until all CEA groups were fully withdrawn to the upper elec-trical limit. At that point, the CEA groups and individual CEA's were moved into various configurations to test the proper operation of the CEDMCS and its associated computer alarms and inter-locks; and to verify the proper calculation of

                                                       ,b/       ))UO

62 CEA group positions, radial peaking factors,

 .            rod shadowing factors, and CEA deviation pen-alty factors by the CPC's and CEAC's.

3.1.14.3 Test Results A. Computer Alarms and Interlocks A total of 49 comparisons were made at various rod configurations to test com-puter setpoints, permissives, alarms, and motion inhibits. In all, 15 deficiencies were found. The Lower Sequential Permis-sive (LSP) and Upper Sequential Permis-sive (USP) actuation deviations were more than expected in 9 cases but all were con-servative from a rod overlap point of view and thus deemed acceptable. The other de-ficiencies were reviewed and the results were found to be acceptable for all de-ficiencies except for one group minor de-viation alarm. This deficiency was re-tested and the results found acceptable. B. CEA Group Positions Each CPC channel was examined to determine agreement between CPC CEA group positions and the corresponding lowest CPC subgroup position. Each subgroup indication was required to be within 1 75 inches of the indicated group position. In all, 31 CEA group configurations were tested for a total of 125 comparisons. The CPC indicated position of each sub-group was checked for agreement with the position of the target CEA as showa by the CEAC's. The agreement tolerance was 2 75 inches if both CEAC and CPC receive input from the same reed switch position transmitter (RSPT) and +3.75 inches if the CEAC and CPC receive input from dif-ferent RSPT's. A total of 69 CEA group configurations were tested. All compari-sons indicated that tolerances were satis-fied. C. CEA Related Factors

a. CPC Planar Radial Peaking Factors (PRPF)

Each CPC channel was checked for pro-per calculation of its CEA position dependent PRPF's. Two separate rom-parisons were made: 1) each CPC (167 319

63 channel PRPF was verified at all 20 axial core nodes for 6 CEA group con-- figurations, and 2) each CPC channel PRPF was verified at 3 nodes for 26 CEA group configurations. In all, 6 PRPF's were found dificient.

b. CPC Rod Shadowing Factors (RSF) .

Each CPC channel was checked for pro-per calculation or its CEA position dependent rod shadowing factors. Data was gathered for 10 separate CEA group configurations. Each RSF corresponding to its appropriate Excore Detector was verified for each CPC channel. In all, 120 comparisons were made. Three RSF's were found deficient, one each for CPC channels B, C, and D. Upon review of the data, it was discovered that the RSF deficiencies were due to a pro-

• cedural errot and retestir.g eliminated the deficiency.

c. CPC Penalty Factors All four CPC channels were checked for proper CEA position dependent penalty factor calculations. In all, 32 CEA group configurations were tested. Values of TPEN (Total Penalty Factor) and GRPPEN (Group Penalty Fac-tor) were recorded for above CEA posi-tions. Out of 128 comparisons, one value of TPEN was out of tolerance compare ( to the expected value.

d. CEAC Penalty Factors Both CEAC's were checked to verify proper calculation of CEA position related penalty factors. CEAC - 1 penalty factors (PF1 RAW) and CEAC -

2 penalty factors (PF2 RAW) were re-corded for 17 CEA group configurations. Out of 34 total comparisons of ex-pected vs. recorded values, 1 penalty factor was found out of tolerance, PF1 for CEAC - 1. o n>i T20

64

e. CEAC/CPC Penalty Factor Transmission In addition to the aforementioned CEAC penalty factor verifications, each penalty factor input to the CPC's, PF1 RAW and PF2 RAW, was compared to the actual penalty factor output from the CEAC's (PF OUT). .

Verification was made at 17 different CEA group configurations. The required deviation between the two varialles was 5 0.001. Out of 34 total comparisons, one out of tolerance condition was found for each CEAC.

f. Deficiency Review Discrepancies between the expected and recorded values of the CEA - related factors were attributable to small differences between the actual CEA positions used by the CPC's relative to the recorded positions from the CPC report obtained via the data link to the plant computer. Such differ-ences between the positions used by the CPC's and the recorded positions arose from minor analog to digital conversion differences. These de-ficiencies were reviewed by Combus-tion Engineering and were found to be insignificant.

3.1.14.4 Test Conclusion The deficiencies and Startup Field Reports is-sued against this procedure identified several CEDMCS inadequancies. Due to the frequency of problems associated with the CEDMCS, it was concluded that hardware changes were necessary in order to increase the CEA maneuvering re-liability. In summary, a whosesale replace-ment of the opto-isolators were performed. The components were replaced with new, state-of-the-art opto-isolators which were more reliable. Y67 321

65 In addition, large numbers of strip switches were replaced and/or repaired. These changes along with extensive troubleshooting solved the large majority of the CEDMCS problems. During the performance of this procedure when CEA motion was occuring in group mode, the CPC DNBR and LPD margin analog indicators cycled in an oscillatory type motion when the shutdown groups or part length CEA's crossed penalty fac-tor boundaries. Review of the situation re-vealed that this was due to failure to provide an adequate dead-band at the CEA exercise limit boundaries in CEA groups with more than one subgroup. Operating procedure precautions were added to avoid unnecessary protection system trips until a software change could be prepared and approved. This problem caused no non-conservatisms. e

                                         %7       322

66 3.1.15 STEADY STATE VIBRATION TEST 3.1.15.1 Pu rpose The purpose of th2. test is to monitor pipe vi-brations of the systems listed below during all significant plant operating modes that are likely to cause vibration in the subject piping system. A. Shutdown Cooling System. B. Steam Generator Blowdown System. C. Reactor Coolant System. D. Charging System. E. Letdown System. 3.1.15.2 Method A walkdown and visua: examination of each sys-tem was conducted at each specified test mode. Piping -sas observed for excessive or abnorn 31 vibration. In addition to the visual inspec-tion of the RCS piping, the reactor coolant pump vibration monitors were checked to veri-fy that no alarm condition was present. 3.1.15.3 Results No excessive or abnormal vibration was de-tected in any of the above listed piping sys-tems. All acceptance criteria were met. 3.15.4.1 Conclusions Vibrations of all piping systems are acceptable as determined by visual inspection.

                                                              - 6i -

67 3.1.16 SAFETY INJECTION SYSTEM CHECK VALVE RETEST 3.1.16.1 Purpose The Safety Injection System Check Valve tests were performed to accomplish the following objectives: A. Verify that Safety Injection Tank (SIT) 2T2D discharge check valve 2SIl6D will pass flow at normal operating pressure and temperature. B. Verify that the Safety Injection loop check valve 2SIl5B will pass flow to the Reactor Coolant System at normal operation pressure and temperature. C. Verify proper operation of the ECCS hot leg injection check valves (2SI-26A, 2SI-27A, 2SI-28A, 2SI-26B, 2SI-27B and 2SI-28B) with normal operating back pres-sure. 3.1.16.2 Test Method The following tests were performed under hot standby conditions (545 F/2250 psia): A. The test of valve 2SI-16D was performed by routing flow from Safety Injection Tank 2T2D through the check valve to drain valve 2CV-5061-2 then to the RWT via re-turn header valve 2CV-5082. The discharge rate from the SIT was controlled by throt-tling 2SI-17 (SIT drain to RWT). B. Proper operation of SIT check valve 2SI-15B was verified by charging in' o the RCS via the High Pressure Safety Injection Sy. tem. Normal charging flow was established into the flow then passes through 2SI-15B into the RCS. C. The HPSI ECCS Hot Leg Injection Check Valve test was performed as follows. Nor-mal charging flow was established with 2CV-ll5 (charging pump discharge to HPSI system) open. Flow through check valves 2SI-26A, 2SI-27A and 2SI-23.". was established by opening 2CV-5101-1 (HPSI header to Shut-down Cooling suction). To establish flow to check valves 2SI-26B, 2SI-27B, and 2SI-28B, the crossover valves (2SI-30 and 2SI-

31) from HPSI header #1 to HPSI were opened.

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68 3.1.16.3 Test Results A. Proper operation of check valve 2SI-16D was verified by noting a decrease in the SIT level and pressure. B. Flow through 2SI-15B was verified by not-ing a decrease in normal charging flow-(2FIS-4863) and an increase in HPSI header pressure (2PI-5020) to slightly above RCS pressure. C. Proper operation of the Hot Leg Injection check valves was verified by noting a de-crease in the charging flow as indicated on 2FIS-4863. 3.1.16.4 Conclusions All valves tested, performed satisfactorily and passed flow under normal operating conditions.

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                                                   ' 6i 8

69 3.2 PLANT TESTS 3.2.1 EMERGENCY FEEDWATER SYSTEM S/G WATERHAMMER TESTS 3.2.1.1 Purpose A. The objective of the test was to demon-strate that the emergency feedwater syr,- tem (EFWS) will automatically supply water to the steam generator following an emerg-ency feedwater actuation signal (EFAS). B. The second objective of this procedure was to show that the system design is adequate to prevent a damaging waterhammer. To ensure proper system performance, the fol-lowing acceptance criteria were established.

a. The indicated flow rate from 2P7A to SG2E24A, and from 2P7B to SG2E24B was 1 575 gpm.

b. The EFW valves actuated by an EFAS moved to their proper positions.

c. EFW pump 2P7A starts and reaches 3 rated speed within 5 20 seconds.

d. Following ar. EFAS, visual inspections, and evaluation of test data do not reveal any indications of damaging waterhammer.

3.2.1.2 Test Method A. The RCS is maintained at 450 F 1 5 F and 1100 1 15 psia. The EFW system is aligned for normal operoting conditions. Both S/G 1evels are greater than 50% and the pres-surizer level 2 45%. The levels in both Steam Generators were gradually reduced until S/G levels were at the trip point

• and the PPS initiated an EFAS signal.2P7A was actuated first followed by 2P7B after an approximate 90 second time delay. After attaining a 50% level in the steam genera-tors, 2P7A was stopped and flow from 2P7B

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was regulated to maintain levels in the operating band. The test was also per-formed at the 545 F test plateau.

• Temporarily reduced to 10% S/G level for the purpose of this test.

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70 B. In order to check for indications of waterhammer occurence, the test was per-formed at various flow rates. These flow rate combinations were:

a. 2P7B to B S/G at 200 gpm flow.

b. 2P7B to B S/G at 400 gpm flow. .

c. 2P7B to B S/G at full flow (2 575 gpm).

d. 2P7A to A S/G at 250 gpm flow.

 ' ~~
      ~_                e. 2P7A to A S/G at full flow (2 575 gpm).

f. 2P7A and 2P7B to A and B S/G's at full flow (2 575 gpm to each S/G).

Following each test of the EFW system, a vis-ual inspection and evaluation of test data was performed to ensure that no indication of dam-aging waterhammer was incurred. 3.2.1.3 Test Results Several problems were encountered during initial attempts to perform this test. These included: (1) 2P7A tripped on overspeed. Maintenance was performed upon the governor and throt-tle adjustments were made. (2) During EFW actuation, the emergency suc-tion valves from the service water system opened due to low suction pressure tran-sients. The following maintenance ac-tivities were performed:

a. The pump suction startup strainers removed, cleaned and reinstalled.

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71

b. Air pockets existed in the high points of the pump suction piping and venting the pump sucidon piping was required.

(3) Additionally, in order to achieve the re-quirement that 2P7A deliver a total flow to both steam generators of 650 gpm, the turbine controller was adjusted. . (4) 2CV-1039 indication in the control room was inoperable. This was corrected by adjustment of limit switch settings. Following completion of maintenance activities, the test was repeated successfully. 3.2.1.4 Conclusion After initial problems were resolved through maintenance and adjustments, the system was re-tested to verify proper function. The objectives of the test were met. The Emer-gency Feedwater System (EFWS) will automatically supply water to the steam generators following an Emergency Feedwater Actuation Signal (EFAS). Also, the system design was demonstrated ade-quate to prevent damaging waterhammer. 32S 96I -

72 3.2.2 EMERGENCY 'r4EDWATER FLOW SETTINGS 3.2.2.1 Purpose The purpose of this test was to flow balance the Emergency Feedrater System (EFWS). To as-sure design basis system performance, the fol-lowing acceptance criteria were established > A. EFW pumps 2P-7A and 2P-7B shall have a minimum recirculation flow of 50 gpm. B. Given that steam generator secondary pres-sure is 1015 psia + 5 psia; EFW pumps 2P-7A and 2P-7B shall develop 575 + 5 - 10 gpm flou through their corresponding dis-charge trains while operating independently. 3.2.2.2 Test Method t EFW Pump 2P-7B: EFW pump 2P-7B was started to verify proper recirculation flow. The flow path to steam generator 2E24B was then lined up and EFW flow was established to 2E24B. The stroke on the control valve was then limited to pro-duce the required flow rate with maximum con-troller demand. To verify train B system per-formance, an EFAS was simulated. The flow path to steam generator 2E24A was lined up and EFW flow was established to 2E24A. The stroke on the control valve was then limited to produce the required flow rate with maximum controller demand. To verify train A perfor-mance, an EFAS was simulated. To determine system performance with trains A and B operating simultaneously, both flow paths were opened and the associated flow rates r..- corded. Trains A and B from 2P-7B were then isolated and 2P-7B secured. 9

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73 EFW Pump 2P-7A: EFW pump 2P-7A was started and proper recircu-lation flow verified. Trains A and B were lined up to steam gener-ators 2E24A and 2E24B. EFW pump 2P-7A speed was then increased until a combined reference flow rate of approximately 650 gpm was esta-blished. The demand on speed controller was recorded and 2P-7A secured. System performance was verified by simulating an EFAS to 2P7A. This was accomplished by energizing steam admission valve and verifying that 2P-7A accelerated to the reference speed and flow rate. Train A performance was deter-mined by isolating train B and recording the flow rate to 2E24A. Train B performance was determined by isolating train A and recording the flow rate 2E24B. 3.2.2.3 Test Results The mass flow rate developed by the EFW System to the steam generators depends upon secondary pressure, pump speed, recirculation flow rate, system leakage, and control valve position as well as inherent system head losses. Initial testing of the EFWS revealed several component related deficiencies that prevented system performance from meeting the acceptance cri-teria. The following summarizes the deficien-cies and their solutions:

a. EFW pump recirculation flow control valves would not hold their position while throt-tied. This caused fluctuations in pump discharge pressure and thus flow to the steam generator secondary. Flow Orifices were replaced with " Rodman" type variable restriction orifices. By increasing the head loss factor for these orifices, it was possible to open the recirc. valves to their backseats thus providing stable pump recirculation flow rates.

74

b. Some discharge flow stop valves would not isolate their respective trains thus rob-bing flow from the other train under test.

The valve seats and seals were inspected for these valves and reworked as necessary.

c. Speed control for the 2P-7A driver was erratic. The speed controller was re -

worked by the vendor and the turbine governor controller was adjusted.

d. A flow element was installed backwards.

This increased the line resistance in train B from EFW pump 2P-7B. As a re-sult, flow data recorded with the flow ele-ment in this condition was inaccurate. The flow element was reinstalled with the pro-per orientation and its corresponding flow transmitter was calibrated. After perform 2ng the above maintenance, the fol-lowing results were obtained:

1. Recirculation flow rates for EFW pumps 2P-7A and 2D-7B were successfully set to 50 gpm.

2. The maximum flow rate developed by 2P-7B to Steam Generator 2E24B was 570 gpm a-gainst a secondary pressure of 1020 psia.

3. The maximum flow rate developed by 2P-7B to Steam Generator 2E24A was 567 gpm a-gainst a secondary pressure of 1018 psia.

4. The maximum flow rate developed by 2P-7A to Steam Generator 2E24A was 575 gpm at a turbine speed of 3600 rpm. Secondary pres-sure was 1018 psia.

5. The maximum flow rate developed by 2P-7A to Steam Generator 2E24B was 582 gpm at a turbine speed of 3596 rpm. Secondary pres-sure was 1015 psia.

3.2.2.4 Conclusion The Emergency Feedwater System performs ade-quately after an EFAS to mitigate the conse-quences brought on by reactor decay heat fol-lowing a design basis accident.

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75 3.2.3 MAIN STEAM SAFETY VALVE PIPING DYNAMIC TRANSIENT TESTS 3.2.3.1 Purpose The purpose of this test was to verify the adequacy of the piping restraints for the main steam line during blowdown of one main steam safety valve. . 3.2.3.2 Mut i Test instrumentation to measure the dynamic response of the system piping was installed on the main steam line and on the safety valve. Testing required increasing the system pressure until safety lift occured. 3.2.3.3 Results The main steam header pressure was increased from normal operating pressure (s 980 psia) until the safety valve lifted @ 1080 psia. Blowdown continuted until system pressure was

                 @ 676 psia. All monitored parameters were ac-ceptable except the measured strain @ SG1 of 625p in/in vs. the max. expected value of 300p in/in.

Bechtel (S.F.) evaluated the measured strain of 625p in/in and determined that it was well with-in the design limits of the system piping. 3.2.3.4 Conclusions For the purpose of this test, the safety valve operation and piping restraint performance was satisfactory. 9 O L J' a O, )

76 3.2.4 EMERGENCY FEEDWATER PUMP TURBINE TRIP DYNAMIC TRANSIENT TEST 3.2.4.1 Purpose The purpose of this test was to verify the ade-quacy of the piping restraints for the steam supply line to the emergency feedwater pump, turbine during fast closure of the turbine trip and throttle valve. 3.2.4.2 Method Test instrumentation was installed to monitor the dynamic response of the system piping dur-ing fast closure of the trip and throttle valve which included load cells, pressure transducers and displacement transducers. During the test, the emergency feedwater pump was brought up to speed to deliver 240 + 5 gpm. After stable con-ditions were obtained, the turbine was manually tripped. 3.2.4.3 Results All dynamic responses were within the expected value= axcept valve closure time, which was

                 .065 oc.. (measured) vs. .050 sec. (calcu-lated).

Bechtel, S.F., evaluated the measured valve closure time and determined that all affected transients were acceptable. 3.2.4.4 Conclusion The dynamic response of the piping system was acceptable.

                                                                  /
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77 3.2.5 SECONDARY HYDROSTATIC TESE 3.2.5.1 Purpose The purpose of this test was to perform a hydrostatic test on,ita non-isolable feed lines and other lines that had been reworked. The rework was required herause sections of. the feedwater piping failed to meet the material brittle fracture specifications. 3.2.5.2 Test Method The main steam lines were filled by overflowing the Steam Generators. After the system was filled and vented, the Emergency Feedwater Pump was used to raise system pressure to approxi-mately 500 psig. At 500 psig, the Hydro Pump was started and system pressure slowly raised. The Hydro Pump could not raise pressure above approximately 1100 psig. It was found that several Main Steam Safety valves were leaking as well as the "B" MSIV packing. System pres-sure was reduced to tighten the MSIV packing and safety valve gags. System pressure was then increased and held at 1360 + 25/-0 psig for 10 minutes followed by a reduction in pressure to 1100 + 25 psig for final inspec-tion. The system was depressurized and re-stored to its normal condition. 3.2.5.3 Test Results The secondary system was inspected by the Bech-tel Hydro Engineer and the ASME Boiler Code Inspector. No leakage was observed on the welds and piping that was required to be in-spected. 3.2.5.4 Conclusions The Main Steam and Feedwater systems met all acceptance criteria for this hydrostatic test and one hydro cycle was used. O e4 yo]-

78 3.2.6 PIPE / COMPONENT HOT DEFLECTION PREOPERATIONAL TESTS 3.2.6.1 Purpose The purpose of this test is to verify that the piping systems listed below respond to thermal expansion in accordance with the design intent.

                   >A. Main steam.

B. Main steam bypass to condenser. C. Main steam to emergency feedwater pump turbine driver. D. Steam Supply to main feedwater pump tur-bine drivers. E) Charging system. F. Shutdown cooling. G. Pressurizer relief valve discharge piping. . H. Pressurizer surge piping. I. Pressurizer spray system. J. Steam generator to blowdown tank. K. Letdown line. Note: The design intent is that:

a. The piping expands freely with constraints only at rigid restraints and anchors: i.e.,

the expansion is not contrained at spring hangers, snubbers, pipe whip restraints or any other obstructions.

b. The pipe returns to its approximate origi-nal position in the cold condition.

3.2.6.2 Method A. Temperature Measurements: Type 'K' mag-netic thermocouples were installed at all specified locations in the containment. Temperature was read by and printed by a Data Logger, usually @ 4 hr. intervals. Temperatures outside the containment were taken by magnetic dial indicators and by a hand held thermocouple read by a Fluke Digital Thermometer. B. Pipe Displacement Measurement: Pipe dis-placement at selected points was measured by either of the below methods: 1qCs JJ

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79

a. Measured displacement of the spring in spring hangers.

b. Measured displacement of mechanical snubbers.

c. V.;asured displacement of pipe with respect to a fixed reference point.

e.g., stanchion, wall, restraint or scribe mark.

d. Measured displacement of pipe as mea-sured by scribe trace or a scratch gage or pencil trace on cardboard tar-get.

The measured displacement of the pipe during heatup from ambient to various temperature plateaus was compared to an expected displace-ment range. Duringheatup)pipingsystemswere walked down and visually checked for any in-terference or binding. 3.2.6.3 rest Results During heatup, it was observed that the main steam lines were restrained by a whip restraint just downstream of the steam generators. Con-sultation with Bechtel Stress Group revealed that no excessive stress problem existed. Dur-ing a subsequent cooldown, shims were removed from the restraints and main steam lines could expand freely. The major deficiency in pipe movement occured on the main steam lines @ Hangers 2 EBB-1A-H7 and 2 EBB-2A-H7 where the pipes deflected to the West instead of to the East as expected. The North-South movement of the main steam headers in the turbine build-ing was only a fraction of what was expected due to the piping leads to turbine being only at ambient temperature. Both main steam headers were up against their stops just downstream of the MSIVS. Several hanger locations of the shutdown cooling system were inaccessible when pipes were hot. Several of the installed scratch gages failed due to the flexible arms not being able to ab-sorb large pipe displacement. Also, several pencil gages had failed due to location prob-lems. During subsequent heatups, failed scratch and pencil gages were replaced and new measurements were taken. b'D'(O 9oI

80 Shims in the main steam header restraints just downstream of the MSIVS were machined down and during a subsequent hot measurement, an ade-quate gap was measured. Bechtel evaluated all questionable pipe movement, including main steamline movement at Hangers 2 EBB-1A-H7 and 2 EBB-2A-H7 and determined that no stress problem exists. 3.2.6.4 Conclusion All piping systems in the scope of this test were observed for free expansion during system heatup and no interference was observed. Evalu-ation of the test results by the Bechtel Stress Group revealed that no stress problems exist in the subject piping system and that the intent of the test has been met. W e

81 r 4.0 INITIAL APPROACH TO CRITICALITY 4.1 Pu rpose To provide a safe organized method for attaining the ini-tial criticality of the Arkansas Nuclear One - Unit 2 Re-actor. 4.2 Test Method The approach to criticality commenced on 12/4/78 with the Reactor Coolant System at 260 F and 460 psia, all CEA groups fully inserted, and two reactor coolant pumps run-ning. The Reactor Coolant System was at a boron concen-tration of 1996 ppm. Prior to increasing the neutron population, a determination of base count (C was performed for each of the two start-up channels. Th,)e scaler output of each startup channel drawer was connected to an individual counter-scaler, thus allowing simultaneous count rates to be obtained from the two separate channels. A similar connection was made to two logarithmic power channels to provide instrumentation overlap and continuity in the event criticality was not Prior achieveg% 1 x 10 power.tothestartupchanneldeactivationat The neutron population was increased initially by with-drawing the CEA Groups in incremental steps resulting in an essentially unrodded core (CEA Group 6 was returned to N 75" withdrawn for later maneuverability when criti-cal). The CEA withdrawal sequence, and intervals are shown in Table 4.1, and correspond to approximately equal reactivity insertion levels. Following each withdrawal increment, the count rate (C ) on each startup channel was 4 recorded and a ratio, C,/C =H, calculated. From this, a plot of inverse count riteg(1/M) versus CEA withdrawal was generated. Figure 4.1 shows this plot for each start-up channel. Several CEDMCS related problems occurred during the CEA withdrawal sequence. Troubleshooting corrected these problems so that the withdrawal sequence could be continued. After CEA withdrawal had been completed, the dilution of the RCS boron concentration was commenced at an initial rate of approximately 2 ppm / minute (88 gpm DMW). During the RCS dilution from 1996 to 1196 ppm, the RCS was sam-pled for boron concentration at 30 minute intervals and a plot of 1/M versus dilution time was maintained for each startup channel. After a 30 minute stabilization period at 1196 ppm, the RCS dilution was recommenced at a rate of approximately 1 ppm / minute (44 gpm DMW). The 9

82 RCS was sampled for boron concentration at 15 minute intervals and plots of 1/M versus dilution time were continued. The plot was frequently extrapolated to provide estimates of the time remaining to initial criticality. Figure 4.2 shows 1/M versus dilution time curves for each startup channel, and Figure 4.3 shows a curve of RCS boron concentration versus dilu-tion time. The RCS boron concentration was decreased-until criticaity was,9ghieved at a boron concentration of 980 ppm at 1 x 10 power. CEAGroup5gwasthen used to increase reactor power to 1 x 10 at which point it was stabilized. CEA Group 6 was at 82 inches withdrawn at criticality. 4.3 Test Results_. Initial criticality of the Arkansas Nuclear One - Unit 2 reactor was achieved at 1455 on December 5, 1978. The start-up was performed by first withdrawing CEAs and by dilut-ing the reactor coolant system. Criticality was achieved with CEA Group 6 at 82 inches withdrawn and a boron con-centration of 980 ppm. This was in good agreement with the predicted value of 996 ppm with CEA Group 6 at 75 inches withdrawn. No abnormal conditions were observed throughout the approach to criticality other than the CEDMCS problems previously mentioned. 4.4 Conclusion Arkansas Nuclear One - Unit 2 was brought to criticality in a safe and organized manner. The critical CEA posi-tion and boron concentration agreed well with the pre-dicted values. F

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83 TABLE 4.1 CEA Withdrawal Sequence Procedure Step # CEA Group Inches Withdrawn 7.9 All CEA's Full in - 7.11.1 Shutdown A 24

 ,,      7.11.2                Shutdown A             34 7.11.3                Shutdown A             56 7.11.4                Shutdown A             Full out 7.11.5                Shutdown B             30 7.11.6                Shutdown B             52 7.11.7                Shutdown B             Full out 7.11.8                Part Length P          Full out 7.12.1                Regulating 1

• 78 7.12.2 Regulating 2
• 78 7.12.3 Regulating 3
• 99 7.12.4 Regulating 4
• 138 7.12.5 Regulating 6 v 146 7.12.13 All CEA's Full out
• Withdrawal of regulating groups performed in manual sequential mode.

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85 FIGURE 4.2 Arkansas Nuclear One Unit 2 INITIAL APPROACH TO CRITICALITY BOL, 1st Cycle, 2600F, 460 psia 9 l . :: . 1 ___- . i .-

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86 FIGURE 4.3 ARKANSAS NUCLEAR ONE UNIT 2 INITIAL APPROACH TO CRITICALITY BOL, 1st CYCLE, 2600F, 460 psia ,

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                                                                                                                                                                                                                                   %l

87 5.0 LOW POWER PHYSICS TESTS

5.1 INTRODUCTION

The objectives of the Low Power Physics Test program were to measure the physics characteristics of the as-built core and to demonstrate conformance with applicable Technical Specifications. The measurements were conducted at260F,460psiaang545F,_3g50psiawiththereactor operating between 10 and 10 of full power - high enough to provide a good signal to noise ratio but low enough to avoid sensible heat effects. A reactivity com-puter system was used for reactivity measurements with two dual ex-core uncompensated ion chambers being used

 ,           as input. Measurements of CEA group worths, isothermal temperature coefficients, dropped, ejected, and stuck CEA worths, and of inverse boron worths were made. A summary of the results appears in Table 5.1. Refer to the appli-cable sections of this chapter for more detaileo informa-tion. The Low Power Physics Test measurement results were in close agreement with the predicted values.

9 I

88 TABLE 5.1 LOW POWER PHYSICS TEST RESULTS I. 260 F, 460 PSIA MEASUREMENT PREDICTED MEASURED ARO Critical Boron Conc. 1006 ppm 999 ppm-4

                                                -4 EARO ITC                               .076x10 Ap/ F    +.08x10 Ap/ F CEA Group 6 Worth                    0.376%Ap           0.404%Ap CEA Group 5 Worth                    0.662%Ap           0.624%Ap CEA Group 4 Worth                    0.331%Ap           0.373%Ap CEA Gps. 6, 5, 4 in Differential Boron Worth                                64.89 ppg /%Ap CEA Gps. 6, 5, 4 ITC                 65  ppm /%Ag
                                       -0.22x10   Ap / F  0.08x10 Ap/ F Sequential Worth of CEA Gps. 6,5,4                           1.38%Sp            1.39%Ap All Rods Out Differential Boron Worth             65 ppm /%Ap        64.97 ppm /%Ap II. Heatup to 360 F, 460 PSIA ITC ARO                                .035x10Ap/ F     .096x10Ap/ F III. Pressurization to 1100 PSIA
                                               -6                   -6 Pressure Coefficient -               .41x10 A / psia    .462x10 A / psia IV. Heatup From 360 F to 450 F ITC ARO                               .018x10Ap/ F   .103x10" Ap/ F V. Pressurization From 1100 PSIA to 2250 PSIA
                                                                   -6 Pressure Coefficient    '-
                                         .38x10' Ap/ psia   .58x10      / psia VI. Heatup From 450 F to 545 F ITC ARO 450 F to 500 F

• N/A .069x10[Ap/F ITC ARO 500 F to 545 F
• N/A .016x10 Ap/ F
• For information only n -

89 TABLE 5.1 (Cont'd.) LOW POWER PHYSICS TEST RESULTS VII. 545 , 2250 PSIA MEASUREMENT PREDICTED MEASURED l Worst Dropped PLCEA(P-24) .0272p .0342Mp Worst Dropped PLCEA Sub-Group (P-1) . ll6Mp .133Mp Worst Dropped CEA (6-1) .132Mp .127%p Next Worst Dropped CEA (6-47) .0717%p .085Mp ARO Critical Boron Conc. 1001 ppm 1012 ppm CEA Group 6 Worth 0.557Mp 0.568Mp CEA Group 5 Worth 0.514Mp 0.524Mp CEA Group 4 Worth 0.766Mp 0.743Mp CEA Group 3 Worth 0.778% 0.792% Gps 6-3 @ LEL, Gp 2 @97.5 WD Differential Boron Worth 76.5 ppg /M p 72.56 ppg /Mp ITC, ZPDIL .5x10 ,p/ F .48x10 Ap/ F

                                                                   -4 MTC, ZPDIL                      .34x10 Ap/ F         .32x10 Ap/ F Worst Ejected CEA at ZPDIL (4-11)               . 376Mp              . 351Mp
     ! Next Worst Ejected CEA at ZPDIL (6-1)                .348Mp               . 322Mp CEA Group 2 Worth             .942Mp               . 916Mp CEA Group 1 Worth             1.258M p             1. 275Mp CEA Groups 6-1 in Differential Boron Worth      76.5 ppmfMp          72.30 ppmfMp ITC, CEA Gps 6-1 in          0.81x10 Ap/ F         -1.07x10 -4Ap/ F
                                             -4 MTC, CEA Gps 6-1 in             .65x10 Ap/ F         .908x10 Ap/ F PLCEA Group P Worth           . 400Mp              . 418Mp ,

CEA Group B Worth 3.112Mp 3. 402Mp CEA Group A Worth minus Stuck CEA A-52 1.99 Mp 2. 032Mp Total Inserted Worth of CEA Gps 6-A (minus the Stuck CEA) and w/Gp. P 10.736Mp 10.665%Ap

      , Sequential Worth of CEA Groups 1-6                4.85Mp               4. 746Mp_4 ITC, ARO                      0.0Ap/,*F            0.323x10 Ap/ F
                                                                    -4 MTC, ARO                      .16x10 Ap/ F         .1923x10 Ap/ F CEA 6-01                      .135Mp                .1207%p PLCEA Group P Integral Worth                .226Mp                .25Mp
                                                              }

ohl

90 5.1.1 CEA COUPLING TEST 5.1.1.1 Purpose To confirm that the CEAs are coupled to their respective CEA extension shafts. 5.1.1.2 Test Method . The test was conducted at 260 F and 460 psia. Each CEA was inserted one at a time until a reactivity decrease of >.5C had been achieved. The CEA was then returned to its original posi-tion. 5.1.1.3 Test Results With each movement of a CEA, a resulting change in reactivity was seen. All CEAs exhibited this, thus indicating all were properly coupled. 5.1.1.4 Conclusion Since all CEAs, when moved, caused an appro-priate decrease or increase in reactivity, it was concluded that the CEAs were properly cou-pled. y 6'l

91 5.1.2 CEA/PLCEA SYMMETRY TBSTS 5.1.2.1 Pu rpose CEA/PLCEA symmetry tests were performed to verify that the CEAs and fuel were loaded in the core as designed. 5.1.2.2 Test Method The test was performed with the NSSS at 545*F, 2250 psia, and CEA Groups A, B, 1, 2, 3, 4, 5 and I at the upper electrical limit (UEL). CEA Group 6 was initially at

• 75" withdrawn.

The test program began by inserting the center CEA (CEA 6-1) to its lower electrical limit, (LEL). Reactivity and Power swings, as moni-tored on the reactivity computer, were compen-sated for by CEA Group 6 movement. Subsequent symmetric CEA subgroups were checked for sym-metry as follows: the first CEA of a subgroup (the reference CEA) was inserted to its LEL while reactivity and power swings were compen-sated for with CEA Group 6 movement. (After conditions stabilized, Group 6 was not moved again until the first CEA of the next sym-metric subgroup was inserted.) The second CEA was then inserted to its LEL while the reference CEA was withdrawn to its UEL. Any reactivity difference, as indicated on the re-activity computer, was noted. The third CEA was then inserted to its LEL while the second was withdrawn to its UEL. This trading pro-cess was continued until all the CEAs of a particular symmetric subgroup were traded. The last CEA was then withdrawn as the refer-ence CEA was reinserted. This provided a means of evaluating for reactivity drift which may have occurred since the reference CEA was ini-tially checked. The above described procedure was carried out for each cf the symmetric Groups of Table 5.1.2.1. 5.1.2.3 Test Results Each CEA of Table 5.1.2.1 was checked within its symmetric CEA subgroup as described above. Data was recorded and analyzed per procedure and all symmetric CEAs were found to agree (within a subgroup) to within the acceptance criteria of

                  + 1.5C of the symmetric CEA subgroup average.

Figure 5.1.2.1 summarizes the results of this test. 1b3

5.1.2.4 Conclusions Since the acceptance criteria for this test (as described above) were satisfactorily met, it can be concluded that the Fuel and CEAs were correctly loaded. 9 M

                                                   "f H (3 l

93 C) if7 r

                                                                                                                 'I3 TABLE 5.1.2.1 T

SYMMETRIC CEA/PLCEA GROUPS Symmetric CEA/PLCEA Groups are shown below. The R designation indicates the CEA/PLCEA that the group is referenced to for purposes of this test: Group 1 Group 2 Group 3, Group 4 Group 5 Group 6 Group 7 R-6-46 R-5-58 R-4-10 R-3-62 R-2-71 R-2-6 R-1-38 6-47 5-59 4-11 3-63 2-74 2-7 1-39 6-48 5-60 4-12 3-64 2-77 2-8 1-40 6-49 5-61 4-13 3-65 2-80 2-9 l-41 3-66 1-42 3-67 1-43 3-68 1-44 3-69 1-45 Group 8 Group 9 Group 10 Group 11 Group 12 Group 13 Group 14 R-A-50 R-A-70 R-B-14 R-B-30 R-B-2 R-P-22 R-P-26 A-51 A-72 B-15 B-31 B-3 P-23 P-27 A-52 A-73 B-16 B-32 B-4 P-24 P-28 A-53 A-75 B-17 B-33 B-5 P-25 P-29 A-54 A-76 B-18 B-34 - A-55 A-78 B-19 B-35 A-56 A-79 B-20 B-36 A-57 A-81 B-21 B-37

94 FIGURE 5.1.2.1 CEA/PLCEA SYMMETRY TEST RESULTS N t . A B C D E F G H J V. L M tl ~ R P 9 9 1

                           -_                    -0.38         -0.30 4            2               4
                                          -0.70         -0.74           -0.57 5                  8              7             7                8               5 l
               -0.10              -0.46         -1.22         -1.46             -0.24           -0.05                     3 1                 11            13              11              1
                        - 0.31            -0.49        - 0.19           -0.20          -o.06 8                 14              to           to                14              3 5
               -0.24             -0.11          -0.48           0.13           -0.07           -0.09 4                 if                3             6              3               il                4
       -0.24             o.20             -0.52        -0.37                                                              6
                                                                       -0.49           -0.05             - 0.15 9               7                  to     6_     12            12               to              7                9 7

+ 0.13 - 0.3 S -0.18 -0.14 -0.12 -1.20

                                 -0.21'-                                                                        + 0.15 2                 13-               Y                            6               13                2
      + 0.2 8           -0.03            -0.18                         -0.17           -0.03            -0.48 9              7                  to             12            12               to              7                9

+ 0.13 + 0.05 - 0.12 k + 0. 0 4 - 0.01 +0 22 + 0.13 4 M 3 6 3 U 4

      - 0.11            4 0.0f              N             b            -0.32           + 0.11           -0.02 8                 14              10            10               14              8
              + 0.05               N              E           - 0.01           48.005          + 0.0 6                    11 1                11            13              11              1 A                 A            A             + 0.0 2         + 0.0 b 5                  t              7             7                g               r N              N           + 0.03           + 0.07          0. oo 4             2               4 74 N            M             -0.14 i             T M           -0.02 G                                                                                                           -

x Symmetric Group # y (See Table 5.1.2.1) Deviation from reference CEA (R) in c. (See Table 5.1.2.1) Acceptance Criteria = + 1.5c

                                                                                             = Uncompensated Ion Chamber at 10'8" Radius              ,

N ()l '3

95 5.1.3 ISOTHERMAL TEMPF2ATURE COEFFICIENT TESTS 5.1.3.1 Purpose Isothermal temperature coefficient (ITC) tests were performed to determine the as-built values of ITC as a function of soluble boron concen-tration and CEA configuration. A ' measured'- moderator temperature coefficient (MTC) was then derived to verify compliance with Techni-cal Specifications and to ensure consistency with certain assumptions in the Safety Analysis. 5.1.3.2 Test Method The ITC for a given temperature and core config-uration was measured by changing the primary system temperature and recording the change in reactivity. Each particular measurement in-volved a series of temperature changes wherein the primary system temperature was changed and returned to the base temrerature two to three times while reactivity was being calculated and recorded on the reactivity computer. From each temperature change, an individual ITC was calculated (change in reactivity divided by change in temperature), and the final ITC value was taken as the average of the individual ITCs. For ITC measurements performed at 260 F, the initial temperature change was an increase of 10 F. For ITC measurements at 545 F, the ini-tial temperature change was a decrease of 10 F. 5.1.3.3 Test Results Seven ITC measurements were performed during the Low Power Physics Testing for various base temperatures, CEA configurations, and soluble boron concentrations. These measurements and their predictions, acceptance criterion, and results are delineated in Table 5.1.3.1. The corresponding moderator temperature coef-ficients (MTCs) for the ITCs of Table 5.1.3.1 at 545 F, are given in Table 5.1.3.2 along with their predictions, and acceptance criteria. 5.1.3.4 Conclusion All measured Isothermal Temperature Coefficients as well as Moderator Temperature Coefficients

,             are within the applicable acceptance criteria, and are therefore acceptable.

_-o

                                                              ~

b b 56'i

96 TABLE 5.1.3.1: ISOTHERMAL TEMPERATURE MEASUREMENT RESULTS PREDICTED MEASURED 4

                                        ~             -

ERROR MEASUREMENT CONDITIONS (X10{ALUEAk/k/F)(10{ALUEAk/k/ F) (X1 260 F, EAR 10 , 460 psia, N996 ppm -0.076(1006 ppm)+0.080 -0.156 260 F, CEA Groups 6, 5 & 4 at LEL, 460 psia, s907 ppm -0.220(917 pp.n) -0.079 -t.141 Heatup (260 F to 360 F), Gp. 6 at 115" WD, 460 psia -Os035 +0.096 -0.131 Heatup (360 F to 450 F), Gp. 6 at 115" WD, 1100 psia -0.018 +0.103 -0.121 Heatup (450 F to 500 F), Gp. 6 at 2 115" WD, 2250 psia NA +0.069 N/A Heatup (500 F to 545 F), Gp. 6 at 2 115" WD, 2250 psia NA +0.016 N/A 545 F, ZPDIL , 2250 psia, s809 ppm -0.500(N/A) -0.479 -0.021 545 F, CEA Gps. 6 through 1 at LEL, 2250 psia, N657 ppm -0.810(638 ppm) -1.068 +0.258 545 F, EAR 0 , 2250 psia. N1016 ppm +0.000(1001 ppm)+0.032 -0.032 NOTES:

1. EAR 0: Essentially All Rods Out, i.e., all CEAs at their UELs ex-cept Group 6 which is > 130" withdrawn.

2. No accepta ice criteria or predictions applicable.

3. ZPDIL: Zero Power Dependent Insertion Limit, i.e., CEA Groups 6 through 3 are at their LELs and CEA Group 2 is at 97.5" withdrawn.

4. ERROR = Predicted Value - Measured Value.

The error is compared to the acceptance criteria given below.

            "The measured ITC shall be within + 0.5 x 10' Ak/k/ F of the predicted ITC value for the test condition.

o h)1) f.'

97 TABLE 5.1.3.2: ISOTHERMAL TEMPERATURE MEASUREMENT RESULTS PREDICTED MEASURED 4 _ ALUE _gRROR MEASUREMENT CONDITIONS (X10{Ak/k/F)(10{ALUE _ Ak/k/ F) (X10 Ak/k/*F) 1 545*F, ZPDIL , 2250 psia, s809 ppm -0.340(N/A) -0.320 -0.020 545 F, CEA Gps. 6 through 1 at LEL, 2250 psia, s657 ppm -0.650(638 ppm)-0.908 +0.258 545 F, EARO , 2250 psia, N1016 ppm +0.160(1001 ppm)+0.192 -0.032 NOTES:

1. ZPDIL: Zero Power Dependent Insertion Limit, i.e.; CEA Groups 6 through 1 at their LELs and CEA Group 2 is at 97.5" with-d rawn.

2. EAR 0: Essentially All Rods Out; i.e., all CEAs at their UELS ex-cept for Group 6 which is > 130" withdrawn.

3. ERROR: Predicted value - Measured value The error is compared to the acceptance criteria given below:

                 "The measured MTC shall be within + 0.5 x 10 'Ak/k/ F of the predicted MTC value for the test condition. The measured MTC for the test condition at 545 F or greater
                                                        ~

shall be less positive than 0.5 x 10 *Ak/k/ F, (Tech-nical Specification 3.1.1.4 - BOL condition)." ( n > O o

98 5.1.4 CEA GROUP WORTH TESTS 5.1.4.1 Purpose This series of tests provided for the measure-ment of the reactivity worths of the CEA's, to verify the accuracy of models used for design and safety calculations by demonstrating that they accurately predict the measured CEA worths. The group worths which were measured are as follows: A. RCS at 260 F/460 psia

1. CEA Groups 6, 5, 4 (no overlap)

2. CEA Groups 6, 5, 4 (with overlap)

B. RCS at 545 F/2250

1. CEA Groups 6 thru 1 (no overlap)

2. CEA Groups P, B, A

3. CEA Groups 6 thru 1 (with overlap)

4. CEA Group P (group 6 @ 120")

5.1.4.2 Test Method All CEA group reactivity worths were measured by introducing a continuous change in the RCS boron concentration (boration/ dilution) and maintaining reactor power by inserting or with-drawing CEA groups in increments. The result-ing reactivity traces were then reduced to ob-tain integral reactivity worths. During dilution of the CEA's into the core, the group sequence was 6-5-4-3-2-1-P-B-A. Hence, the core was in a different rodded state for each group worth measurement. Also, since safety considerations preclude the reactor being critical with all CEAs inserted, mea-rurement of the worth of shutdown Group A was slightly modified. This was done by using a combination of boration/ dilution, group trips, and extrapolating to find the total integral worth of Group A. 5.1.4.3 Test Results Tables 5.1.4.1 and 5.1.4.2 present the results of the CEA group worth measurements in compari-son to predicted worths. Also, the integral

                                                                  <t y))

O,

99 group reactivity worths measured at 545 F are displayed in Figures 5.1.4.1 through 5.1.4.10. The worth of Group P was measured twice, once during the sequential insertion of all of the CEA's and again with ARO except Group 6 which was at 120" withdrawn. Figure 5.1.4.11 pre-sents the overlapped Regulating CEA Group in-tegral worth curve. - 5.1.4.4 Conclusion All of the measured CEA integral reactivity worths are in good agreement with predicted values and satisfy all acceptance criteria. I.. \

• 100 TABLE 5.1.4.1 CEA GROUP REACTIVITY WORTHS 260 F/460 psia PREDICTED MEASURED ACCEPTANCE CEA NUMBER WORTH WORTH LIMITS GROUP OF CEA's FUNCTION (% Ak/k) (% Ak/k) (% Ak/k) 6 5 Regulating 0.376 0.404 .276 .476 5 4 Regulating 0.662 0.624 .562 .762 4 4 Regulating 0.331 0.373 .231 .431 at 6

l)1l

101 TABLE 5.1.4.2 CEA GROUP REACTIVITY WORTHS 545 F/2250 psia PREDICTED MEASURED ACCEPTANCE CEA NUMBER WORTH WORTH LIMITS GROUP OF CEA's FUNCTION (% Ak/k) (% Ak/k) (% Ak/k) 6 5 Regulating 0.557 0.568 0.457-0.657 5 4 Regulating 0.514 0.524 0.414 .614 4 4 Regulating 0.766 0.743 0.651- .881 3 8 Regulating 0.778 0.792 0.661 .895 2 8 Regulating 0.942 0.916 0.801-1.083 1 8 Regulating 1.258 1.275 1.069-1.447 P 8 Shaping 0.400 0.418 0.300-0.500 B 20 Safety 3.112 3.402 2.645-3.579 A 16 Safety 3.567 3.55 3.032-4.102

*A        16         Safety         1.99          2.032    1.692-2.288

• • P 8 Shaping 0.226 0.25 0.15-0.35
• Worth of Group A without stuck CEA.
  • Group 6 at 120" withdrawn.

l 9 () I

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FIGURE 5.1.4 9 7 001 9 6b'

111 INTEGRAL CEA GROUP WORTH BOL, FIRST CYCLE CEA GROUP P ' (Group 6 0 125")

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112 INTEGRAL CEA GROUP WORTH BOL, FIRST CYCLE Regulating Groups Overlap Worth

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                                                                                                                                                                                                                                                                                          %bT

113 5.1.5 DIFFERENTIAL BORON WORTH DETERMINATION 5.1.5.1 Purpose The purpose in determining differential boron worth is to provide a means for predicting the resultant change in boron brought about by a change in reactivity, (or vice versa), for a varying set of core conditions. 5.1.5.2 Test Method Differential boron worths were calculated by solving the following equation: Differential Boron Worth (%Ak/k/100 ppm Boron) = Inserted CEA Worth (Initial) - Inserted CEA Worth (Final) CBC (Initial) - CBC (Final) Where: CBC = Critical Boron Concentration co-- responding to the particular (initial or final) core configuration. The inverse of differential boron worth, IBW, which is often more convenient to use, was calculated as follows: Inverse Boron Worth (IBW, ppm /%Ak/k) =

                          . 100 Differential Boron Worth it should be noted that predictions of boron worths were provided in the IBW form as was the acceptance criteria.

The plateaus and core configurations for which boron worth calculations were carried out are summarized in Table 5.1.5.1. 5.1.5.3 Test Rasults The results of boron worth calculations and cor-responding data for initial and final cases are given in Table 5.1.5.1. 5.1.5.4 Conclusion

  .            All inverse boron worths were compared to their appropriate predicted values and were found to be within the acceptance criteria of + 15 ppm /
                %Ak/k of the predicted values.

O OkO i6: u

114 TABLE 5.1.5.1: DIFFERENTIAL BORON WORTH DETERMINATION INSERTED BORON MEASURED

• MEASURED PREDICTED REACTIVITY CONCENTRATION BORON WORTli IBW IEW MEASUREMENT CONDITION (%Ak/k) (ppm) (%Ak/k/100 ppm) (ppm /%Ak/k) (ppm /%Ak/k) ERROR 260 F, 360 psia:

1. EARO, CBC (INITIAL) 0 999 CEA Gp 6,5,&4 at LEL with CEA 1.541 64.89 65.0 0.11 Gp 3 at 146.3" WD (FINAL) -1.4485 905

2. CEA Cp 6,5,&4 at LEL (INITIAL) -1.4485 905 1.539 64.97 65.0 0.03 CEA Gp 6 at

• 115" WD (FINAL) -0.0783 994 545 F, 2250 psia:

3. EARO, CBC (INITIAL) 0 1012 CEA Cp 6.5,4&3 at LEL with 1.378 72.56 76.5 3.94 CEA Gp 2 at 97.5" WD (FINAL) -2.7719 811

4. CEA Cp 6,5,4&3 at LEL with CEA Gp 2 at 97.5" WD (INITIAL) -2.7719 811 CEA Gp 6 through I at LEL with 1.387 72.30 76.5 4.20 CEA Gp B at 120.8" WD (FINAL) -4.902 657 sc

&.

• Measured IBW: Measured Inverse Boron Worth. This value must agree with the PREDICTED IBW within +15 ppm /%Ak/k.

CC

    **EARO:     Essentially All Rods Out. This refers to all CEAs at their UELs except for Group 6 which is > 130 inches withdrawn.

1:i .

115 5.1.6 CRITICAL BORON CONCENTRATION 5.1.6.1 Purpose The purpose of this section was to determine the RCS boron concentration for a particular all rods out critical condition. 5.1.6.2 Test Method A. At the 260 F plateau, boric acid was in-jected through the charging pumps with the reactor critical until all rods were fully withdrawn. Chemistry samples were then taken to determine the ARO critical boron concentration. B. At the 545 F plateau, boric acid was in-jected through the charging pumps with the reactor critical until all rods were fully withdrawn except group six which stabilized between 130" and 140" withdrawn. Chemistry samples were then taken to determine criti-cal boron concentration with the above con-trol rod configuration. CEA group six was witudrawn to the fully withdrawn position (the reactivity computer was used to de-termine tha residual worth of group six) and reactivity was allowed to stabilize. Group six was then returned to between 130" and 140" withdrawn to return power to its base level. The above manipulation of Group six was done two more times so that a more defined value for Group 6 residual worth could be obtained. 5.1.6.3 Test Results A. At the 260 F plateau, the ARO critical boron concentration was measured and de-termined to be 999 ppm. B. At the 545 F plateau, the ARO critical boron concentration was calculated to be 1012 ppm. 5.1.6.4 Conclusion The acceptance criteria of within + 100 ppm of predicted values was met at both temperature plateaus as summarized in Table 5.1.6.1. o I

116 TABLE 5.1.6.1 ARO CRITICAL BORON CONCENTRATION MEASUREMENT RESULTS TEMPERATURE I TEST RESULTS ACCEPTANCE DPM PLATEAU ( F) (PPM) CRITERIA (PPM) DIFFERENCE 260 999 1006 + 100 7 PPM 545 1012 1001 + 100 11 PPM t

                                                            ))7 5.1.7 PSEUDO DROPPED AFD EJECTED CEA WORTH TESTS 5.1.7.1   Purpos; The purpose of these tests was to determine the reactivity worths of the worst case drop-ped and e32cted CEA's. In addition, the worth of Part Length CEA subgroup P-1 was determined.

5.1.7.2 Tast hathod "Dum,; ped" and " Ejected" CEA worth measurements were performed at the 545 F, 2250 psia plateau. The initial CEA configurations for each case are given below. To ensure the predicted worst case Regulating Group CEA was selected for each condition, the predicted second most reactive CEA was traded with the predicted most reactive CEA and their worths compared. Reactivity measurements were made with a Reactivity Compu*. . A. Dropped CEA: For the full-length CEA's, CEA 6-1 was predicted as the most reactive dropped CEA whereas, CEA 6-47 was predicted to be the second most reactive. For the Part length CEA's, CEA P-24 was predicted to be the most reactive. Part Length CEA subgroup P-1 was predicted to be the rost reactive Part Length subgroup. This was verified as follows. With Regulating Group 6 between 130" and 150" withdrawn, Part Length CEA P-24 was diluted from its Upper CEA Limit (UCL) to its most negative worth insertion point. The CEA was then inserted to its Lower CEA Limit (LCL) to determine the re: idual (posi-tive) worth of the rod. P-24 was subse-ouently borated to its UCL. The worth of Part Length Subgroup P-1 was determined via the same method as CEA P-24 while using Regulating Group 6 to compen-sate for reactivity swings. Supgroup P-1 was also retrieved to its UCL after test-ing. O((

118 B. Ejected CEA: In the case of the ejected CEA, CEA 4-11 was predicted to be the worst ejected CEA while CEA 6-1 was predicted as the second worst. This was verified as follows. With the reactor stable and CEA Regulating Groups 3, 4, 5, and 6 at their LEL's and Group 2 at 97.5 inches withdrawn, CEA 6-1 was borated to its UEL to determine the ejected rod worth. CEA 6-1 was then traded with CEA 4-11 and Regulating Group 2 was adjusted such that CEA 6-1 was at its LEL and CEA 4-11 was at its UEL. CEA 4-11 was taen realigned with group 4 using Group 2 to compensate for reactivity swings. 5.1.7.3 Test Results Table 5.1.7.1 summarizes the " Dropped" and

           " Ejected" CEA worths as determined by this test. As can be seen in the Table, CEA 4-11 was determined to be more reactiva than CEA 6-1 for the ejected CEA case, which was as predicted.

5.1.7.4 Conclusions Each of the measured " Dropped" and " Ejected" CEA and subgroup CEA worths agrees with its predicted counterpart within the range of ac-ceptance criteria as shown in Table 5.1.7.1. e

119 TABLE 5.1.7.1

             " DROPPED" AND " EJECTED" CEA WGaTH MEASUREMENT RESULTS MEASURED        PREDICTED     MEETS ACCEPTANCE CEA               WORTH (%Ak/k)   WORTH (%Ak/k) CRITERIA ?(1)(2)(3)

DROPPED RESULTS: PLCEA P-24 -0.034 -0.027 Yes/Yes Subgroup P-1 -0.133 -0.116 Yes/Yes CEA 6-1 -0.127 -0.132 Yes/Yes CEA 6-47 -0.085 -0.0717 Yes/Yes EJECTED RESULTS: CEA 6-1 0.322 0.348 Yes/Yes l jCEA 4-11 0.351 0.376 Yes/Yes

1. The acceptance criteria for the " Dropped" cases is: "The measured dropped CEA/PLCEA worth shall be within 1 25% of the predicted dropped worth of the CEA/PLCEA, or within 0.1%

Ak/k of the predicted worth, whichever is larger."

2. The acceptance cr*teria for the " Ejected" cases is: "The measured ejected CEA worth shall be within + 25% of the pre-dicted ejected CEA worth, or witnin 0.1% Ak.'k of predicted worth, whiche7er is larger.

3. The first entry here corresponds to the :omparison for 1 25%

of the predicted value. The second entr:. here corresponds to the comparison for 0.1% Ak/k of the pred;cted value. 4

120 5.1.8 STUCK CEA WORTH MEASUfEMENT 5.1.8.1 Purpose Technical Specifications require that available shutdown margin shall not be less than 5.0% Ak/k, with the highest worth CEA stuck out, whenever tho reactor is critical. In order to verify the validity of the predictions, the shutdowa worth with the predictes most reactive CEA with-drawn was measured. 5.1.8.2 Test Method CEA A-52 was predicted to be the worst case stuck CEA. The test was performed at 545 F and 2250 psia, with CEA groups 6 through B (including the part length CEAs) tripped; Group A was initially at s 156" withdrawn. CEA A-52 was withdrawn to its UEL and secured. T' remainder of Group A was ins .ted (diluted) to s 10 inches withdrawn

                  + 5 inches. Conditions were allowed to stabi-lize and then the remainder of Group A was in-serted to its LEL to determine its resiuaal worth. 12 residual worth determination was repeated two more times so that a more accurate value could be obtained.

5.1.8.3 Test Results The worth of all CEAs (full and part length) inserted except for the most reactive CEA, which was stuck out, was determined to be 10.67% Ak/k versus 10.74% Ak/k as predicted. This in turn, corresponds to a stuck (CEA A-52) worth of 1.52% Ak/k versus a predicted worth of 1.58% Ak/k. The worth of Group A minus the stuck CEA was determined to be 2.03% Ak/k versus 1.99% Ak/k as predicted. Stuck CEA results and evaluations are summarized in Table 5.1.8.1. 5.1.8.4 Conclusions Acceptance criteria related to stuck CEA cases were applied to the following determinations:

1) Worth of Group A minus the most re-active stuck CEA (CEA A-52).

2) Worth of all CEAs inserted (including part length CEAs) minus the most re-active stuck CEA (CEA A-52).

These acceptance criteria were met satisfactorily (see Tablo 5.1.8.1). OhR

121 TABLE 5.1.8.1: STUCK CEA RESULTS MEASURED PREDICTED ' MEETS ACCEPTANCE CASE WORTH (%Sk/k) WORTH (%Ak/k) ERROR CR7TERIA  ! Stuck CEA A-52 *l.52 1.58 3.8% N/A - Group A Minus Stuck CEA A-52 2.032 1.99 2.07% YES(7)

                                               /.042
                                               %Ak/k All CEAs Inserted- 10.666         10.736        .65%l YES(2)

(with PLRS) Minus Stuck CEA A-52 l l

• Based on a Group A measured worth of 3.55% Ak/k.

(1) The acceptance criteria for this case is: "the measured worth for this case shall be within i 15% of the pre-dicted worth or within 0.1% Ak/k of the predicted worth, whichever is larger. (2) The acceptance criteria for this case is: "the measured worth for this case shall be within 1 10% of the pre-dicted worth".

                                                                         .O O~ \ u o ( ?s)

122 6.0 POWER ASCENSION TESTS 1he Power Ascension tests were conducted to determine the as built plant characteristics during steady state and transien* operation from Hot Zero power to 100% power and to demonstrate that the plant is capable of withstanding the accidents and transient; analyzed in the FSAR. The majority of the tests requiring steady state conditions occurs at the major plateaus of 20%, 50%, 80% and 100% power. Minor testing was also accom-plished at other power plateaus. Plant performance data was recorded during testing at the major plateaus (see Table 6.0.1). All test data for a given plateau was evaluated prior to increasing power to the next major pla-teau. The Power Ascension test report is divided into four sections, each one corresponding to a major power plateau. Each section contains details of the tests performed during power ascension to a major plateau and all tests performed at the plateau. N l

3  % 2 0 1 0 X X X X X X X X X X X 1 5 X X X 9 0 X X 9 0 X X 8 5 X 6 0 X X 5

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S t  % T s 0 X X S e 4 E T T s 1  % N v . 0 X X O 0 3 I u S a 6  % X X X X X N e 0 ' I X X X X X X X X X E t E 2 C a L S l B  % A P A 5 X X X X X T 1 R r E e '% W O o w 0 1 X X X P P

                 %                                                                    X 5   X    X  X 0                               _         _            X n

n o o i i n t t o s t a a t i t n n c r n t n s o e i b e a o e i m f i m r s T t e i l n e b i a r r a o r i r y c u e C i u l a r i s V n t s a p t f a o r a a C m s i e n i e l e o i r t  !

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a e e r i u l R c f F s c a a D h y t r S b n i n n e t t T t e t a d a r G g a i D a a r i m e r i n m e V n r l D D e d c i m e r a r V A i T a e w n a r i w a o T w e r t t o a p o r o V y f S o d n u b n n P a o P r r S e d a n t a e r C l a l s t e L l a o A a C {% G l P i s T A a e S C C a r a e s e i s P e O C b a h S L e p r e p

Cnc T n l C S c m / i t S S a S c B S S n o e r C r A i a m E S r C u D S C i r h o P a E n h e N T R N S N R P C C V C U S T T_ L_ C_

POWER ASCENSION TESTS 124 Power Plateau vs Test . TABLE 6.0.1 (cont'd) TEST '0% 5% '10% 15% '20% 30% 40% ISO % 60% 70%'80% 20%'50% 65% 80% '90% 95% 100% Radial Peaking Factor Verification fX 20% Rx Trip with S/D outside Control Room X 80% Total LOF/ Natural Cire. Verification X Loss of Offsite Power _ X Ejected CEA X Dropped CEA X Load Rejection from 100% X Fast Trip Recovery /XE .ollow X PLCEA Xenon Control X Incore Detector Signal Verification X X X X Movable Incore Detector Checks X X X X FWCS Post Trip Setting X 100% Turbine Trip _ X Field Adjustment of Inct re Dynamic Comp. X 50% Turbine Trip X _C T , C

POWER ASCENSION TESTS 125 Power Plateau vs Test TABLE 6.0.1 (cont'd) TEST 0% 5% 10% 15% 20% 30% 40% 50% 60% 70% 80% 20% 50% 65% 80*s 90% 95% 100% Turbine / Generator Loading X X X X X X X X X X X X X Main and Reheat Steam Test X* X* X X X X X X Condensate and Feedwater System Test X X X X Main Turbine Electro liydraulic Control X X X X FW Vents, Drains and Water Induction X X X X Vibration and Loose Parts Monitoring X X X X X llVAC Performar.cc X X X X X Pipe / Comp Ilot Deflection X Piping Dynamic Transient X X X X Steady State Vibration X X X X Biological >hield Survey X X X X X

• Performed during Power Ascension following 20% Rx Trip

~C T-CC C , N N

126 6.1 0% thru 20" POWER PLATEAU INTRODUCTION Power Ascension testing commenced on December 16, 1978 with the plant at Hot Zero power and continued thru the 20% power plateau. A total of 27 individual tests were performed at the 20% power plateau. Testing was also accomplished at the 0%, 5%, 10% and 15% power plateaus. All testing requiring power levels between Hot Zero power and 20% power has been performed and those major objectives of the Power Ascension Test program have been satisfied. Sections 6.1.1 thru 6.1.27 provide a detailed description of the 0% thru 20% power tests.

                                                                     *1
                                                           %2>

127 6.1.1 RCS AT POWER DETERMINATION TESTS 6.1.1.1 Purpose The RCS AT Power Determination was performed in order to measure the reactor thermal output at low power plateaus (5%, 10%, 15% and 20%). 6.1.1.2 Test Method A primary system calorimetric was completed after steady state conditions were achieved at each plateau. *emperature, pressure, and flow data were obtained from the CPC's. The core thermal output was then determined by ' wiltiplying the primary mass flow rate by the change in enthalpy across the reactor. 6.1.1.3 Test Results RCS AT Power data was collected and a primary system calorimetric was calculated. The cal-culated thermal output is compared with the average values of calibrated neutron flux powe , static thermal power, and core average power in Table 6.1.1.1. 6.1.1.4 Conclusion The RCS AT Power Determination was completed successfully at the 5%, 10%, 15%, and 20% power plateaur.

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128 TABLE 6.1.1.1 POWER CALCULATED RCS AVG. NEUTRON STATIC TliERMAL CORE PLATEAU AT POVER _ FLUX POWER POWER AVG. POWER 5% Run 1 5.025

• 1.55 5.3 5.66 Run 2 4.76 5.31 5.06 6.07 10% 9.93% *3 9.91 10 52 15% 17.32% 14.51 14.99 15.9 20% 19.54% 19.58% 19.35% 19.56%
• Prior to adjustments on excore linear calibrate potentiometer.

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                                                         %B

6.1.2 NUCLEAR AND THERMAL POWER CALIBRATION TESTS 6.1.2.1 Eurpose To adjust the Excore Linear Power Calibrate potentiometers and the CPC addressabl con-stants (KCAL and IPC) relating to the core power level to agree with the RCS AT power.. 6.1.2.2 Test Method The Nuclear and Thermri Power Calibration Test was performed at the . , 15%, and 20% power plateaus as part of the power ascension test sequence. For each safety channel, the input to the High Linear Power Bistable and the CPC values, PHICAL (calibrated neutron flux power) and BDT (static thermal power) were recorded and compared to the RCS AT power. Adjustment of the Excore Linear Power Calibrate potentio-meters, or the addressable CPC constants KCAL or TPC was necessary if the High Linear Power, PHICAL or BDT readings varied from the RCS AT power by more than 10.2% of Rated Thermal Power. For each safety channel (one at a time) the following adjustments were performed as neces-sary. A. The Exco;e Linear Power Lalibrate potentio-meter was adjusted so that the input to the High Linear Power Bistable, as monitored by an external DVM at the PPS cabinet, equaled the following value: DVM Reading = % Power x 5 Volts 1 0.0tiV 100 B. The CPC addressable constants KCAL and T2C were adjusted as follows: KCAL (NEW) = % Power x KCaL (OLD) PHICAL (OLD) TPC (NEW) = % Power x TPC (OLD) BDT (OLD)

                                                     %b

130 Af ter the initial adjustments were performed, readings from all 4 channels for High Linear Power, PHICAL, and BDT were taken and compared to the RCS AT power. If any of the readings varied from the RCS AT power by more than 1 0.2% of hated Thermal Power, the adjustments were repeated until the 1 0.2% criteria could be met. , 6.1.2.3 Iest Results Ir.itial attempts at 5 and 10% power to cali-brate the Excore Linear Power Signals because t'aere was not enough adjustment range in the calibration potentiometers. The linear power subchannel gains were adjusted by changing taps on fixed resistors in the amplifier circuit and all four channels were adjusted using the pre-viously described method and the acceptance criteria of agreement within 1 0.2% of Rated Thermal Power was met. Appendix E was successful]v completed at the 15% and 20% power plateaus thout incident. 6.1.2.4 Conclusion At the 5%, 15% and 20% power plateaus, the Ex-core Linear Power Calibrate potentiometers and the CPC addressable constants KCAL and TPC were adjusted using information obtained from the RCS AT power determination. The High Linear Power, PHICAL, and BDT readings for all safety channels agreed with the RCS AT power to .ithin 1 0.2% of Rated Thermal Power. 96b

131 6.1.3 NSSS CALORIMETRIC TESTS 6.1.3.1 Purpose The purpose of this test was to: A. Determine core thermal power by means of a secondary plant heat balance: - B. Verify the COLSS core thermal power calcula-tions; C. Verify that OP 2103.16 (Heat Balance Calcula-tion) will provide a satisfactory indication of core power. 6.1.3.2 Test Method Plant parameters were maintained essentially constant while steam generator data and reactor power information was collected over a 3 hour period. This data along with the energy input and loss terms measured during the RCS Heat Loss Test was used to calculate the net heat output. The calculated core thermal power was compared to the COLSS secondary calorimetric power (BSCAL) to verify the accuracy of the algorithm and to the COLSS primary calorimetric power (BDELT). Adjustments were made as necessary to the AT Power Gain factor (in the BDELT algorithm) to provide agreement between BDELT and BSCAL. OP 2103.16 (Heat Balance Calculation) was completed concurrently and compared to the calculated cora thermal power to verify its accuracy. 6.1.3.3 Test Results Due to various difficulties which were encoun-tered, this test was performed a total of three times. During the first test run,.one of the feedwater flow transmitters was found to be out of calibration, causing minor errors in the secondary calorimetric results. 7esults of the second test run were satisfactory; however, OP 2109.16 was not performed. All results of the third test run were satisfactory. Table 6.3.1 reflects the results of each of the test runs. 6.1.3.4 Conclusions The plant computer secondary calorimetric was found to be within the acceptable limits. Also OP 2103.16 (Heat Balance Calculation) was found to provide acceptable results.

                                                      ;3 g     02B

N O D e a' TABLE 6.1.3.l_ 132 RESULTS OF NSSS CALORIMETRIC DATE CALCULATED BSCAL BDELT RESU LTS CALCULATED BSCAL BDELT PERFORMED CORE (BEFORE ADJUSTMENT) 0F VALUE FOR (AFTER ADJUSTMENT) THERMAL POWER OP 2103.16 AT POWER GAIN RUN #1 12/30/78 16.86% 16.91% 18.57% (1) (2) (2) RUN #2 12/31/78 20.76% 20.69% 18.33% (1) 1.132 19.32% 19.50%(3) RUN #3 1/14/79 18.72% 18.99% 17.40% 18.53% 1.0916 19.05% 19.03% NOTES: (1) OP 2103.16 not performed. Deficiency written. (2) Test aborted prior to this step due to miscalibrated flow transmitter. (3) Actual power was adjusted slightly prior to this step.

133 6.1.4 RCS CALORIMETRIC FLOW MEASUREMENT TESTS 6.1.4.1 Purpose The purpose of this test was to determine the reactor coolant flow rate based upon the com-puter secondary plant calorimetric and the measured primary pressure and temperatures , (T and T ) and to provide guidance for ad-h justment of the CPC and COLSS flow algorithm constants if necessary. While this method yields more accurate results at higher power levels, it was performed at lower power levels as well to provide additional info rmation. No adjustments are made below 80% of rated power. 6.1.4.2 Test Method Calculation of the reactor coolant mass flow rate was based upon secondary plant calorimetric power and primary pressure and temperatures. Over a specified period, during which time plant con-ditions were maintained essentially constant, RCS data was recorded from both the CPC's and the plant computer. Following this collection period, the average enthalpy rise of the reactor coolant was determined and used with secondary calori-metric power to calculate the mass flow rate of the reactor coolant. The calculated coolant mass flow rate was com-pared to CPC's and COLSS values for RCS flow. New values were calculated for the constants in the CPC and COLSS algorithms to provide the desired agreement. 6.1.4.3 Test Results Average core thermal power during this test was 19.80% (COLSS secondary plant calorimetric power). The average enthalpy rise of the reactor coolant across the core as determined from CPC data was 13.330 Btu /lbm. Hence, the reactor coolant mass 8 flow rate was calculated to be 1.4266 x 10 lbm/hr. This translates tg 118.49% of the base mass flow rate (120.4 x 10 lbm/hr). By com-parison, all four CPC channels indicated ap-proximately 111% of base flow and COLSS indi-cated 111.68% of base flow. More detailed re-sults are shown in Table 6.1.4.1. DAO U' 96S

134 New values were calculated for the COLSS flow bias constants and for the CPC flow constants and for the CPC thermal power (BDT) scaling constants (TPC). These values are shown in Table 6.1.4.2 and are the values required to make the CPC and COLSS flow rates agree with the measured coolant flow rate and to offset the change to CPC AT power caused by changing CPC calculated flow. Since this test was per-formed for information only at this power level, none of the new constants were entered. 6.1.4.4 Conclusions The calculated RCS flow was slightly higher than expected but was within acceptable limits. g

135 TABLE 6.1.4.1 REACTOR COOLANT MASS FLOh' VALUES - INDICATION VALUE (1) Calculated (2) 118.49% CPC A 111.07% CPC B 111.16% CPC C 111.04% CPC D 111.06% COLSS 111.68% (1) All values are given as percent of base flow (120.4 x 106 lbm/hr. ) . (2) As calculated using COLSS secondary calorimetric power and coolant enthalpy rise across the core. e 9h 96B

TABLE 6.1.4.2 136 COLSS AND CPC FLOW ADJUSTMENT FACTORS CPC VALUES COLSS VALUES Flow Constants Thermal rawer Constants FC-1 FC-2 TPC D15 (1) D15(2) D15(3) D15(4) Previous Values: CPC A 1.1054 .022094 .99929 CPC B 1.1049 .022094 .95633 CPC C 1.1044 .022094 1.1503 CPC D 1.1044 .022094 1.1106 COLSS 0 0 0 0 Calculated

• Values:

CPC A 1.1792 .023569 .92839 CPC B 1.1778 .023552 .89118 CPC C 1.178 .023576 1.0670 CPC D 1.1783 .023573 1.0373 COLSS -5434.85 -5484.85 -5484.85 -5484.85

• Values calculated at 20% power for informatien only (not input into CPCs).

2: CT' CO C tr( LJ'

137 6.1.5 LINEAR POWER SUBCHANNEL CALIBRATICN TESTS 6.1.5.1 Purpose The purppse of this test was to adjust the Linear Power Subchannel gains, the Excore Linear Power Potentiometers, the 200% Linear Calibrate potentiometer, and the CPC addres-' sable constants (KCAL and TPC) relating to the core power level. 6.1.5.2 Test Method The reactor was stabilized at 20% power. An NSSS Calorimetric Measurement was performed. Following the completion of the NSSS calori-metric, baseline power data was obtained from all four Core Protection Calculator cha..nels and the Plant Protection System (PPS) channels. The selected PPS channel (A, B, C or D) High Linear Power and Low DNBR trips were bypassed. The Excore Linear Subchannel amplifier for each of the three (3) detectors for the selected channel was adjusted to the calori-metric power. At the completion of the Linear Subchannel amplifier adjustment, the 200% Linear Calibrate potentiometer was ad-adjusted and a neutron power signal was simu-lated to each of the three Linear Subchannel amplifiers to verify proper amplifier operation. The Excore Linear Power calibration potentio-meter was adjusted such that the indicated Linear Power was in agreement with the computer secondary calorimetric power. Upon completion of adjustment to the Excore Linear channel, KCAL and TPC were adjusted as necessary, thereby ad-justing CPC Calibrated Nuclear Power (PHICAI.) and CPC Delta Temperature Power (BDT) respec-tively. The above process was performed on the remaining three (3) protection channels and as left power data was obtained from all four CPC and PPS channels. 6.1.5.3 Test Results All Linear Power Subchannel amplifiers were adjusted to the NSSS calorimetric value. The 200% Linear Calibrate potentiometer and the Excore Linear Power potentiemeters were suc-cessfully adjusted for all four channels. KCAL and TPC adjustments were performed as described in the body of the test. 6.1.5.4 Conclusions All necessary adjustments were made to the Linear Subchannel gains, the Excore Linear Power Potentiometers, the 200% Linear Cali-brate Potentiometers, and the CPC addressable constants relating to core power level (KCAL and TPC). 1 ,

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138 6.1.6 PROCESS VARIABLE INTERCOMPARISON TESTS 6.1.6.1 Pu rpose The purpose of this test was to compare Pro-cess Instrumentation readings obtained from the Plant Computer, Plant Protection System, Core Protection Calculators, and various con-sole meters to verify proper agreement between systems. 6.1.6.2 Test Method This test was performed at tne 20% power pla-teau. After establishing steady state RCS con-ditions (not necessarily equilibrium Xenon), data was recorded fo. tae following variables:

1. RCS cold leg temperature,

2. RCS hot leg temperature,

3. RCP differential pressure,

4. RCP speeds,

5. RCS pressure,

6. Pressurizer level,

7. Steam Generator levels, and P Steam Generator pressures.

Common process variable readings for each sys-tem were then intercompared against preset criteria to assure the accuracy of proc :s loop calibrations and system signal processing. 6.1.6.3 Test Results All process variable intercomparisons were with-(. x in specified tolerance at the 20% power plateau with the exception of one RCS cold leg tempera-ture indicator, three pressurizer pressure in-dicators, one pressurizer level indicator, and five steam generator level indicators. Following the 20% poste plateau trip and sub-sequent return to power, a hold was included in the power ascension at 20% power and re-testing was completed which resolved all oat-standing deficiencies. 6.1.6.4 Conclusion All procesc variable intercomparisons were within the specified tolerances at the 20% power plateau.

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139 6.1.7 CHEMISTRY AND RADI0 CHEMISTRY TESTS 6.1.7.1 Pu rpos e The purpose of this test was to conduct chemistry tests with the intent of establishing baseline corrosion data and activity buildup with power level. As a result of this, procedures for-sam-ple collection analysis were verified. Also, this test was used to verify the calibretion of the process radiation monitor. 6.1.7.2 Test Method A. Primary System Sample and analysis procedures were per-formed using the CE Chemistry Manual (CENPD-

28) as a guide. Three sets of RCS Chemistry Analyses were performed at the 20% plateau.

The RCS Chemistry Analyses included the fol-lowing tests:

a. pH

b. Conductivity C. C1

d. F-

e. Dissolved Oxygen

f. Suspended Solids

g. Boron

h. Lithium

i. Dissolved Hydrogen

j. Gamma Spec. Analysis (gas)

k. Degased Gross Beta

1. Crud Activity

m. Tritium

n. Iodine Ratio o, iodine Dose Equivalent

p. Gamma Spec. Analysis (liquid)

q. Total Gas (primary coolant)

S y /

140 B. Secondary System Sampling and analysis procedures were per-formed using CENPD-28 as a guide. Five sets of secondary chemistry analyses were performed at the 20% plateau. Each set of chemistry analyses included the following tests: ,

a. pH

b. Conductivity

c. Cation Conductivity

d. Dissolved Oxygen

e. Hydrazine

f. Ammonia

g. Silica

h. Sodium

i. Iron

j. Copper C. Process Radiation Monitor A s'mple was to be taken downstream of the Pro ess Radiation Monitor. Laboratory re-sults of the Gross Gamma Coolant Analysis were to be compared to the Process Radia-tion Monitor analysis for verification of proper Process Radiation Monitor function.

Agreement within + 20% was necessary to verify proper calibration of the Process Radiation Monitor. 6.1.7.3 Test Results The required radiochemistry and secondary sam-ples were obtained and analyzed. The require] process adiation monitor readings were not obtair.ao due to inoperability of this system. Baseline activities for the 20% plateau were established. 6.1.7.4 Conclusion It was proven that primary and secondary sam- . pling and analysis can be performed in ac-cordance with Technical Specifications and CENPD-28. Baseline activities for the RCS were recorded. The Process Radiation Monitor calibration was not verified at the 20% power plateau. ubb

141 6.1.8 CORE PERFORMANCE RECORD TESTS 6.1.8.1 Purpose The purpose of this test was to record core performance data from incore detectors, and o specify the acceptance criteria for com-

               } -ison of the measured results with predicted co_e operating parameters.

6.1.8.2 Test Met h -d A. While the reactor was being maintained at 20% steady state power, with equilibrium Xenon, in' ore detector data was collected for analysis B. The measured results were then compared to predicted values in the following man-ner:

a. The comparison of the measured power distribution with the predicted power distribution is a root mean squared statistical comparison of the relative power density distribution for each of the 177 fuel assemblies.

b. The comparison of the measured axial power distribution with the predicted axial power distribution is a root mean squared statistical comparison of the relative axial power distri-bution for each of the 100 axial nodes.

c. The me'sured values of total planar radial peaking factor (Fxy), total integrated radial peaking factor (Fr), core aver-age axial peak (Fz), ar.d cote 3-D power peak (Fq) were compared to predicted value .

6.1.8.3 Test Results A. Results of the statistical comparisons and peaking factors are summarized in Tables 6.1.8.1 and 6.1.8.2.

                                                                .- 4 3

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142 TABLE 6.1.8.1 Measured Results Acceptance Criteria (RMS) (RMS) Dower Density Distribution 1.903 $5 Axial Power Distribution 2.120 $5 TABLE 6.1.8.2 Acceptance Measured Predicted  % Difference Criteria Fxy 1.4145 1.3607 3.95 5 10% Fr 1.3786 1.3607 1.31 5 10% Fz 1.30111 1.30 0.08 5 10% Fq 1.79745 1.769 1.61 5 10% 6.1.8.4 Conclusions All acceptance criteria have been met for the comparisons between predicted values and measured results. As shown by the above results, the computer model predictions were adequate for determining core operating parameters. 9 O'

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143 6.1.9 CPC/COLSS VERIFICATION TESTS 6.1.9.1 Purpose The CPC/COLSS Verification Tests were performed to accomplish the following objectives: A. Verify that the CPC/COLSS DNBR and LPD-calculations are correct. B. Evaluate the effect of process input noise on the CPC/COLSS system. 6.1.9.2 Test Method A. At hot zero power, the process input noise on one CPC channel was recorded on FM tape, as a zero power data base. The CPC addres-sable constants and slowly varying input parameters were input into the CEDIPS* com-puter code. The CPC output parameters were compared to the CEDIPS* output in order .. to evaluate the effect of noise. Detailed verification of CPC/COLSS DNBR and LPD calculations were not perfor::.ed at this plateau. B. At 20% power with ARO and Xenon equilibrium, the process input noise was measured. Plant computer reports containing information on the CEAC, CPC's, and COLSS were obtained for use in the verification of the CPC/COLSS DNBR and LPD calculations. The CPC/COLSS data was compared to the results of the CEDIPS* computer code and the incore de-tector analysis results. 6.1.9.3 Test Results A. The process noise data from hot zero power was recorded and sent to CE-Windsor. B. The process noise data from the 20% plateau was recorded. The data required for veri-fication of CPC/COLSS DNBR and LPD calcula-tions was collected and compared to the re-sults of the CEDIPS* computer code. All data was transmitted to CE-Windsor for re-view. 6.1.9.4 Conclusions The CPC output parameters were compared to the CEDIPS* code and were found to be acceptable. The COLSS DNBR and LPD related calculations were reviewed by CE-Windsor and found to be adequate. j. p 1, , g in: q4J

  *CEDIPS is a FORTRAN program for statistical analysis of effects of process inputs upon the CPC system.

144 6.1.10 VARIABLE T AVG 6.1.10.1 Purpose The objective of this test was to detecmine the Isothermal Temperature Coefcient (ITC) and Power Coefficient. , 6.1 10.2 Test Method Two methods were used to determine the Isother-mal Temperature and Power Coefficients; one method was performed with no CEA movement and the other was performed with center CEA move-ment. These two approaches are described in more detail below: A. No CEA Movement With the reactor at steady state and equi-librium or near equilibrium xenon and CEA group 6 at 120 inches withdrawn, a small step change in the turbine control valve position is made and then adjusted to establish a new coolant inlet temperature. This change produces a small turbine load-reactor power mismatch. The temperature change results in a reactivity feedback and a resultant power change. The power change produces an opposite reactivity feedback and the reactor settles out at a new power and temperature condition. The cycle is then reversed by making a small step change in the turbine control valve position in the opposite direction. The ITC is calculated iteratively using tne resultant power and temperature changes along with an assumed power coefficient. The Moderator Temperature Coefficient (MTC) is then calculated a Stracting the pre-dicted Fuel Temperat.._ Coefficient (FTC) from the measured Isothermal Temperature Coefficient. B. With Center CEA Movement

a. Isothermal Temperature Coefficient With the reactor at steady state and equilibrium xenon and CEA group 6 at 120 inches withdrawn, a small step change in the turbine control valve position is made and then adjusted to establish a new coolant inlet temperature. This change produces a small turbine load-reactor power O (,9

145 mismatch. The temperature change re-sults in a reactivity feedback. This reactivity is matched with equal and opposite reactivity by movement of the center CEA (holding reactor power constant). The ITC is calculated iteratively knowing the power and temperature changes along with the center CEA integral worth curve and by using the test predictions as initial guesses for the Isothermal Temperature and Power Coefficients. The MTC is calculated as described previously. _

b. Power Coefficient A reactivity insertion is made using the center CEA, resulting in a change in reactor power. Average coolant temperature is held constant by changing turbine load to match re-actor power. The reactor settles out at a new power when the reactivity feedback due to change in power is equal and opposite to the CEA reactivity insertion. The Power Coefficient is calculated iteratively in a manner similar to the ITC calculation.

6.1.10.3 Test Results The Variable T AVG Test was performed at the 20% power plateau as part of the power ascension test program. During the ITC measurement with no CEA movement, T was swung approxi-mately.+5Fabout'2bNprogrammedT a 0% Id power or 546.5 F. However, during E8e ITC measurement with center CEA movement, T had eold to be lowered from the programmed T in order 97 toaccommodateacomplete10Ftemperafureswing and still maintain the center CEA position in an area of relatively high worth. The Power Coefficient measurement with center CEA movement was performed by withdrawing CEA 6-1 to 135" withdrawn and noting the change in reactor power in moving CEA 6-1 from 120" withdrawn (the group 6 average position) to 135" withdrawn. The reactor power was then decreased approximately twice the amount de-termined above by inserting CEA 6-1. This cycle was perfe~'ed three times. e, o@ l

146 The final ITC and power coefficient valec3 were the average value of the runs conducted. The measured value, test predictions, and acceptance criteria for the 20% power plateau are shown in Table 6.1.10.1. The physics test prediction for the ITC was corrected for the measured RCS boron concentration. 6.1.10.4 Conclusion The measured values for the Isothermal Tempera-ture Coefficient and Power Coefficient compared well with the predicted values. Agreement be-tween measurement and prediction was well with-in the uncertainties associated with each para-meter.

14/ TABLE 6.1.10.1 Nominal Reactor Power 20% Boron Concentration (RCS) 828 rpm Isothermal Temperature Coefficient

                                                      -0 Measured                        -0.200 x 10     Ap/ F Predicted                       -0.17  x 10 Ap/ F Acceptance Criteria             +0.5   x 10'    Ap/ F Power Coefficient
                                                      -4 Measured                        -1.189 x 10     A /% Power Predicted                       -1.17  x 10~ Ap/% Power Acceptance Criteria             +0.2   x 10' Ap/% Power 9

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148 6.1.11 CEA SHADOWING FACTOR VERIFICATION TESTS 6.1.11.1 Purpose The purpose of this test was to verify that the CEA shadowing factors used in the CPCs are valid. 6.1.11.2 Tast Method The test was performed at 20% power with initial conditions of eqtiilibrium Xenon, All Rods Out (AR0) and T at 545.5 F + 0.5 F. Data was eold taken for numerous CEA configurations (see Table 6.1.11.1) and reduced via procedure. Reatt;vity was monitored on a Reactivity com-puter. Power was monitored on the CPCs, COLSS, and Turbine-Generator MWe. The CEA shadowing factors were calculated as follows: For configurations without Part Length CEAs inserted: 3 F = 1 D* (with CEA's inserted) Pwer (ARO) 2-i=1 i

• 3 1 DI (ARO) Power (with CEA's inserted) i=1 For configurations with Part Length CEAs inserted:

F = D*2(with CEA's inserted) , Power (ARO) D* (ARO) Power (with CEA's inserted) Where: F = CEA shadowing factor for CPC Channel x (x = A, B, C, D) D = upper excore detector, Channel x D* = reiddle excore detector, Channel x D}=lowerexcoredetector,Channelx Power = reactor power (COLSS calorimetric value) f ai } U[ v f

149 After the CEA shadowing factors are s'alculated, they are compared to appropriate acceptance criteria. If any shadowing factor does not agree with its acceptance criteria, new cor-rection factors (ASM2, ASM3, ASM4) are calcu-lated. Revised acceptance criteria are then generated. If any shadowing factor does not agree with the revised acceptance criteria,. new power uncertainty factors (BERR1 and BERR3) are calculated to compensate for the aoditional error difference. Any revised correction and pcVer uncertainty factors are then loaded into the CPC's. 6.1.11.3 Test Results The channel-wise CEA shadowing factors were calculated for the CEA configurations listed in Table 6.1.11.1 and are presented in Table 6.1.11.2. The results of calculations indi-cated that trese measured factors were not in agreement with the appropriate acceptance criteria. The channel-wise CEA snadowing factors were averaged and cling the average value, new correction and uncertainty factors were calculated. These values were then loaded into the CPC's via addressable constants. 6.1.11.4 Conclusions Upon adjustment of the correction and uncer-tainty factors (CPC constants ASM2, ASM3, ASM4, and BERR1, BERR3) the CPC's now accurately com-pensate for the effects of CEA shadowing. b

                                 , u%

150 TABLE 6.1.11.1 CEA CONFIGURATIONS USED IN VERIFYING CEA SHADOWING FACTORS / CORRECTION FACTORS CORRECTION CEA C0KFIGURATION FACTOR-Group 6 @ LEL Group 6 and 5 @ LEL ASM2 Groups 6, 5 and 4 @ LEL Groups 6, 5, 4 @ LEL, P @ 37.5" WD ASM4 Groups 6, 5 @ LEL, P @ 37.5" WD Group 6 @ LEL, P @ 37.5" WD ASM3 Croup P @ 37.5" WD I (, . e-

151 TABLE 6.1.11.2 MEASURED CEA SHADOWING FACTORS CEA/ GROUP / POSITION CHANNEL A CHANNEL BI CHANhTL C i Cit-B3TL D 6/LEL 1.0531 1.0643 1.0605 1.0513 6/LEL, 5/LEL 0.8517 0.8534 0.8454 0.8485 6/LEL, 5/LEL, 4/LEL 1.0386 1.0497 1.04co 1.0438 6/LEL, 5/LEL, 4/LEL, P/37.5" 1.0906 1.1015 1.0861 1.0956 i 6/LEL, 5/LEL, P/37.5" 0.8808 0.8838  ! 0.8844 0.8856 l 6/LEL, P/37.5" 1.0848 1.0957 i 1.1096 1.1037

                  '37. 5"  1.0545      1.0604      1.0600    1.0638      ,

O r; C_

152 6.1.12 UNIT LOAD TRANSIENT TEST 6.1.12.1 Purpose The purpose of this test was to: A. Demonstrate that the following systems operate satisfactorily in the automatic. mode to maintain plant parameters within acceptable limits during steady state power operations, 5% per minute power down ramps, 1% per minute up ramps and 10% down step change in plant power:

a. Reactor Regulati g System (RRS)

b. Feedwater Control System (FWCS)

c. Steam Dump and Bypass Control System (SDBCS)

d. Megawatt Demand Setter (MDS)

e. Pressurizer Level Control System (PLCS)

f. Pressurizer Pressure Control System (PPCS)

B. Monitor the response of the RRS, FWCS, PLCS, PPCS and SDBCS to plant trips. 6.1.12.2 Test Method This test was performed at the 3% and 20% power plateaus. The tests which occured at each pla-teau are listed below: A. 3% Plateau The reactor was etabilized at 3% power and the control systems verified operational with PLCS and PPCS in automatic operation and steam pressure being maintained by the SDBCS in automatic controlling the 5% dump valve (2CV-0303). 2CV-0303 was placed in manual and slowly closed. As 2CV-0303 closed, 2CV-0302 automatically opened. When stable conditions were achieved (with 2CV-0303 fully closed), 2CV-0303 was re-turned to automatic mode (without balancing automatic and manual control signals). Plant conditions were allowed to stabilize with 2CV-0303 controlling steam pressure automatically. Varicus parameters were recorded on strip chart recorders and com-puter trends during the performance of the test. The test data was reduced and ana-lyzed to verify proper operation of the SDBCS.

                                                             / }/t b '/ /

153 6.1.12.2 (cont'd) B. 20% Power Plateau

a. Automatic Steady State Operation The reactor was stabilized at 20%

power and the control systems veri-fied to be in the automatic mode of operation. Strip chart recorders and computer crends were established as required and a 30 minute steady state run was performed. Following the 30 minute run, the test data was col-lected, reduced and analyzed to de-termine the acceptability of the control systems operations. Control system setpoint adjustments were per-formed as necessary based on the re-sults of the test data analysis. The above described process was performed until no further setpoint changes were required.

b. SDBCS Test The reactor was stabilized at 20%

power and the control systems veri-fied to be in the automatic mode of oper +. ion. The turbine load was de-creas s at approximately 1/2% per minute. the turbine load was de-creased, the SDBCS automatically opened 2CV-0303 to control steam pressure. When 2CV-0303 was 50% open, the down power transient was secured and turbine load was increased at approximately 1/2% per minute until 2CV-0303 was fully closed and the steam pressure was stable at its nor-mal level. Strip chart recorders and computer trends were used to monitor the transient, the data was analyzed to determine the acceptability of the control systems operations. Control system setpoint adjustments were per-formed as necessary based on the re-sults of the test data analysis. The test was rerun as necessary until no further adjustments of the SDBCS were necessary.

c. FWCS Tests The reactor was stabilized at 20%

power and the control systems veri-fied to be in the automatic mode of

                 \h '     operation. Steam Generator level Qg          transients were initiated by changing 1 -             the setpoint at the master controller.

154 6.1.12.2 (cont'd) Master Controller No. 1 controlled level in Steam Generator A and Master Controller No. 2 controlled level in Steam Generator B. After each of the transients listed below, strip chart recorder traces and computer trends were analyzed and the FWCS setpoints adjusted as required. The transient was repeated until no further adjust-ments were required. The following transients were completed first on FWCS No. I then on FWCS No. 2: INITIAL FINAL ' STEAM GENERATOR STEAM GENERATOR RATE OF CHANGE LEVEL LEVEL

1) 70% 60% 10% per min.

60% 70% 10% per min.

2) 70% 60% 1% per sec.

60% 70% 10% per min.

3) 70% 80% 10% per min.

80% 70% 10% per min.

4) 70% 80% 1% per sec.

80% 70% 10% per min.

d. RRS Tests The reactor was stabilized at 20%

power, CT: Group 6 between 113" and 135" withdrawn, the CEDMCS in manual, sequential. all other control systems in automatic. 'Ising RRS #1 (#2) for temperature control T was de-AVG creased 4.5 F less than T

• RE '

CEDMCSwasplacedinAutoaakicSe-quential and the resultant transient recorded on strip chart recorders and computer trend groups. The CEDMCS was returned to ,the Manual Sequential mode, the results analyzed and RRS setpoints adjusted as required. Either or both transients were repeated as necessary until no further adjust-ments were necessary.

155 6.1.12.2 (cont'd) e. MDS Test The reactor was stabilized at 20% power, CEA GR 6 between 113" and 135" withdrawn, the CEDMCS in Manual Sequential, the MDS in the Ready Mode and all other control systems in auto-matic. Turbine load was decreased by 20 MWe from the turbine contre! panel, control was transferred to the MDS. The MDS was placed in the Operator Set mode and the turbine load was increased 20 MWe at 1% per minute. The MDS was placed in the Ready Mode and turbine control was returned to the turbine control panel where load was increased by 20 MWe. The MDS was placed in the Operator Set mode and the turbine load was decreased 20 MWe at 5% per minute. Both transients were recorded using strip chart recorders and computer trends. The test data was analyzed and the MDS setpoints adjusted as necessary. The transients were re-peated until no further setpoint ad-justments were necessary.

f. Reactor Trip Test The reactor was stabilized at 20%

power. Strip chart recorders were set up to monitor the parameters specified in the test procedure. Re-actor was tripped in accordance with the controlling procedure and data taken to record the transient. 6.1.12.3 Test Results A. 3% Power Plateau The test was performed as described, however, the SDBCS Master Controller was in local set-point rather than remote setpoint. When in local setpoint, the setpoint is manually set using the thumbswitch mounted on the face of the controller. The 13% steam dump valve 2CV-0302 was automatically opened to main-tain steam pressure at the setpoint as the 5% steam dump valve (2CV-0303) was closed. When 2CV-0303 was returned to automatic operation, 2CV-0302 closed as 2CV-0303 opened to maiatain steam pressure at the required setpoint. 3 O3-9 (Jo

156 6.1.12.2 (cont'd) B. 20% Power Plateau

a. Automatic Steady State Operation The results of this test showed the following:

1. No spurious CEA motion -

2. Steam generator levels were main-tained at 70.5 1 5%

3. Pressurizer level maintained at the programmed level 1 5% -

4. Pressurizer pressure maintained at 2250 psia 1 15 psi

5. Reactor power was maintained at 20 1 5%

6. T was maintained at 1 1 F of thyge desired value

7. Steam generator pressure was main-tained at i 15 psi of desired value.

b. SDBCS Test The test was performed as required except the main feed pump turbine speed control was not in automatic as required. As the turbine load was decreased, the SDBCS controlled pres-sure around setpoint. The maximum value of steam pressure was 1035 psia (by plant computer). This value was reduced to and maintained at the set-point approximately five minutes after this maximum was reached. Computer trends and brush pen recordings showed no unusual transients.

c. FWCS Test The brush pen recorder data of steam generator level and flow show that proper feed water control was main-tained. The graphs indicate that the level demanded by the FWCS #1 (#2) would be achieved in steam generator A (B) while the level in the remaining steam generator was relatively unaf-fected. Dr .ng the transient, an overshoot o: the demanded set-point was seen with the level finally set-tling down to + 2% in a fairly short period of time!

4$

                                                                })

157 6.1.12.2 (cont'd) d. RRS Test The RRS tests were performed for RRS

                             #1 and RRS #2. Analysis of the data for this test revealed proper CEA motion for each transient. The artificially created power defect was damped quickly with very little overshoot.

e. MDS Test The 1% per minute up power test was performed as described using the MDS in the operator set mode. However, due to the initial CEA position, (114.9 inches withdrawn) power could not be increased 20 MWe prior to reaching an all CSA's out configuration as required by the test procedure. Therefore, power was increased until the group 6 CEA's were 146 inches withdrawn and then the 5% per minute down transient was performed. Data was collected as required.

f. 20% Reactor Trip The required data was monitored on the brush recorders, thus meeting the requirements of this test. After the trip, the feed flow was excessive, the Reactor Trip Override bias po-tentiometers in FWCS #1 and #2 were adjusted to provide proper feedwater flow following a reactor trip.

6.1.12.4 Conclusion A. 3% Power Plateau The SDBCS operated satisfactorily to main-tain steam pressure at the desired set-point during the test. B. 20% Power Plateau

a. Steady State Operation The ability of the various control systems when in automatic mode, to maintain steady state operations was demonstrated.

Ohj

158 6.1.12.2 (cont'd) b. SDBCS Test The SDBCS operated satisfactorily to control and return steam pressure to the prescribed setpoint.

c. FWCS The FWCS has been shown to operate as expected in the Automatic Control mode.

The ability of FWCS #1 and #2 to achieve demanded setpoints at various rates has been demonstrated.

d. RRS Test Both RRS #1 and #2 operated satis-factorily to maintain T within AV the T controlbandasbesigned.

REF No adjustments to the RRS setpoints were required.

e. MDS Test The MDS operated as designed during increasing and decreasing power tran-sients. Oscillation of + 3 MWe around the setpoint were observed during steady state operation between tran-sients. Adjustments were made to de-crease the ma3nitude of the oscilla-tion and improved steady state per-formance wac achieved.

f. Reactor Trip Test All the required data was taken during the reactor trip test at the 20% power plateau. Adjustment of the reactor trip override bias potentiometers pro-vided proper flow signals which provides proper feedwater flow following reactor trip. All control systems functioned properly during the reactor trip to maintain system parameters within their specified bands.

159 6.lsl3 SHAPE ANNEALING MATRIX AND BOUNDARY CONDITION MEASUREMENT TESTS 6.1.13.1 Pu rpose The objective of this test was to measure the Shape Annealing Matrix (SAM) and to verify the Boundary Point Power Correlation (BPPC) con-stants for the CPC's. The primary purpose for performance of this test at the 20% power pla-teau was to evaluate the test method and to pro-vide additional data which could be used to verify the adequacy of the SAM and BPPC con-stants to be measured at the 50% plateau. 6.1.13.2 Test Method The SAM coefficients and BPPCs are determined from a least squares of the measured excore detectar readings and corresponding axial power distribution determined from the incore detectors signals. Since these values must be representative for rodded and unrodded cores th ughout life, it is desirable to use as wide a range of core axial shapes as are avail-able to establish their values. This is done by initiating an axial xenon oscillation. Data is periodically gathered during the oscillations so that tha data will be representative of as wide a rang of axial shapes as possible. In-coce, excore and related data is recorded. This data is input to the CECOR incore code which relates the incore detector signals to power distribution and summarizes the neces-sary power distribution and excore detector data in a form and format which can be easily input to programs used to perform the least squares fitting. The output from CECOR in-cludes:

1) the excore detector fractional responses for each CPC,

2) the core peripheral power fractior. for the upper, middle, and lower third of the Core.

3) the core average power fractions for the upper, middle, and lower third of the co- ,

and

4) the upper and lower core boundary average power.

t.]. ()b

I.60 The above CECOR output is then used to determine a "best set" of SAM coefficients and BPPC con-stants by using least squares analysis. The results of these calculations are then used to adjust the power uncertainty factors (BERR1, BERR3) used by the CPC's in the LPD and DNBR calculations. 6.1.13.3 Test Results While it was intended to monitor the xenon oscillation for at least 60 hours, sufficient self-dampening had occurred within 50 hours to terminate the data collection. One hun-dred CECOR cases were run, of which ap-proximately 20% reflected the core in a rod-ded configuration. The core was unrodded for the remainder of the cases. The measured SAM is presented in Zable 6.1.13.1. Contrary to the initial intent, " measured" 20% SAM values were input to the CPC's prior to escalating to 50% power. This was done because the original predicted values were judged to be inadequate to support the 50% testing. Due to the fact that the BPPC acceptance criteria were developed for the 50% meacurement, no evalua-tion of these results were performed. 6.1.13.4 Conclusions Performance of this test verified the computa-tional methods for determining the SAM and BPPC coefficients and provided a satisfactory Shape Annealing Matrix for the 20% plateau. Two areas were identified for possibly improving the data to be collected during the 50% plateau measure-ment. First, the xenon oscillation during this test was initiated from an equilibrium xenon condition with CEA Group 6 at 120" withdrawn. By starting from an ARO equilibrium Xe con-dition, a " larger" oscillation would be obtained. Secondly, the withdrawal of CEA Group 6 to an ARO configuration during this test was performed very slowly. This caused some dampening of the oscil-lation. A more rapid withdrawal of Group 6 dur-ing the 50% plateau test should also result in a

          " larger" oscillation.

Y u(Vo

161 TABLE 6.1.13.1 SHAPE ANNEALING MATRIX COEFFICIENTS (20% Power) CPC CHANNEL SAM COEFFICIENT A B C D S(1,1) 5.1452 6.8268 6.6279 7.0577 S(1,2) 1.7594 -1.0973 .7470 -1.8424 S(1,3) -4.1878 -2.6112 -2.7996 -2.0400 S(2,1) -1.0207 -1.1478 .6828 .4026 S(2,2) 4.7274 4.9342 4.3819 3.8429 S(2,3) .3914 .4685 .3377 .0100 S(3,1) -1.1244 -2.6790 -2.9451 -3.6550

• S(3,2) -3.4868 .8368 .6349 .9995 S(3,3) 7.5792 6.0796 6.1374 5.0300

                                                %b

162 6.1.14 REACTOR TRIP WITH SHUTDOWN OUTSIDE THE CONTROL ROOM TESTS 6.1.14.1 Purpose This test was performed to demonstrate that the plant could be taken from 20% power to hot stand-by from outside the control room following a re-actor trip. - 6.1.14.2 Test Method The test was initiated from 20% power by trip-ping the reactor manually from the remote shut-down station. One operating crew performed the shutdown from the remote stations while a second crew stood by in the control room to take action if needed. The remote crew performed all im-mediate and follow-up actions of the remote shut-down operating procedure. To monitor the trip, 47 parameters were recorded on brush recorders and 12 parameters were placed on a 1 second com-puter trend. Watchstanders were stationed to verify proper EHC indication of the turbine trip. 6.1.14.3 Test Results The main events were recorded as follows: Re-actor trip breakers 2, 3, 6 and 7 were manually opened. The CEDM main power bus undervoltage relays 1 and 2 deenergized, tripping the main turbine in 1.143 seconds. Start-up transformer

                #3 auto transferred to supply vital busses 2H-1, 2A-1, and 2A-2. The generator output breaker (500 KV) tripped within 1.572 seconds.

Table 6.1.14.1 and Figures 6.1.14.1 thru 6.1.14.3 show the results of the computer trends. 6.1.14.4 Conclusion As indicated in Table 6.1.14.1, all parameters were established for hot standby conditions by 240 seconds following the reactor trip. All control actions were conducted from outside the control room. h

163

                     ,                       TABLE 6.1.14.1 POST TRIP REVIEW OF COMPUTER TREND CROUP DATA 20% Reactor Trip with Remote Shutdown TIME AFTER NEUTRON TRIP      POWER         LEVEL                           PRESSURE                  TEMP.

S/GA S/GB PZR S/GA S/GB RCS TH g TH TC TC 2 y 0 sec 18.6 71.3 72.8 36.0 961.7 964.4 2261.4 556.0 556.4 546.7 546.4 10 sec 0.4 55.0 54.6 33.0 997.9 1000.9 2234.0 551.7 552.4 547.1 546.5 20 sec 0.2 60.0 61.6 33.0 1006.3 1009.4 2223.0 549.0 549.3 547.5 546.8 40 sec 0.1 62.2 62.8 32.6 1011.8 1015.0 2231.0 548.2 548.4 547.6 547.5 60 sec 0.05 62.9 63.0 32.2 1012.2 1015.3 2232.7 547.9 548.4 547.7 547.7 90 sec 0.01 64.4 63.1 31.3 1009.4 1012.4 2232.7 547.2 547.6 547.3 547.3 240 sec 0.01 68.2 68.4 30.9 996.0 998.9 2262.0 545.5 545.9 545.2 545.2 . .C C ^. CC C

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167 6.1.15 INCORE DETECTOR SIGNAL VERIFICATION TESTS 6.1.15.1 Purpose - To verify the proper conversion of the signal from the incore detectors to voltage as read by the plant computer. This comparison of the signal generated by the incore detector to the voltage seen by the plant computer also veri-fied the proper operation of the incore ampli-fier. 6.1.15.2 Test Method The plant was maintained at a power level of approximately 20% 1 0.2%. The RCS pressure was at 2250 psia 1 15 psia. Following de-termination of the connector number for the core location desired, the amplifier associ-ated with the connector was determined. The input connector was disconnected and a special test cable connected between the connector and the amplifier assembly. Using a pico ammeter, the current was measured and recorded for the level 1 detector while simultaneously recording the raw incore signal. This was repeated for the remaining detector levels for the incore string under test. Following completion of the string, the special test cable was discon-nected and the input connector to the amplifier bin reconnected. This procedure was repeated for the remaining incore detectors. 6.1.15.3 Test Results Problems with unstable signals caused poor agreement between detector signal and computer raw signals. Voltages as read on the computer were consistently low. Following an extensive investigation, the problems were determined to lie with the process computer input multipliers and not with incore detectors or the amplifiers. The incore detector system was adjusted and a retest performed at which time it was found that 37 detectors out of 220 were cut of the acceptance criteria. Out of spec incores were checked, readjusted, and retested. Eventually all except 2 detectors were brought into agree-ment. These were determined to be failed in the detector assemblies. 6.1.15.4 Conclusion The proper conversion of the signal from the incore detectors to voltage as read by the plant computer has been verified. Also, this comparison verified the proper cperations of the incore amplifiers.

                                                                ~

9 h)b

168 6.1.16 MOVEABLE INCORE DETECTOR TESTS 6.1.16.1 Purpose This procedure was performed to provide base-line data on the moveable incore detector sys-tem (MICD). At the 20% power plateau, the MICD Manual Pause Interval Constant was to be de-termined and set into the computer. 6.1.16.2 Test Method The reactor was operating at 20% power with temperature, pressure and level maintained constant. A brush recorder was set up to monitor the output of the moveable incore detector amplifie . A computer trend group was established on a one second trend moni-toring the detector position and detector output. With the pause interval set for 15 minutes, the manual mode of operation of the moveable incore drive system was selected. Using the computer console, the #2 moveable in-core drive machine was selected and the de-tector was driven to Level 3 of path #13. After 15 minutes, the computer pulled the de-tector out of active core to the home position. The brush recorder tracing and the computer trends were analyzed for pause interval time. The detector output was considered stable when the detector build-up of current reached steady state. 6.1.16.3 Test Results The procedure was commenced and aborted after several attempts. Maintenance activities were initiated to cor-rect the deficiencies and the procedure was recommenced. All steps were performed satis-factorily. The pause interval was determined to be 291 seconds. 6.1.16.4 Conclusion The data collected, on the computer trend groups, correlated well with the data recorded on the brush recorders. MICDB machine operates as required and the new pause interval value has been entered into the computer. QY) U I[<$

169 6.1.17 FEEDWATER CONTROL SYSTEM POST - TRIP SETTING TESTS 6.1.17.1 Purpose The purpose of this test was to adjust the Feedwater Control System (FWCS) Reactor Trip Override (RTO) bias voltage which corresponds to feed flow at 5% reactor power. The 5% , override voltage automatically adjusts feed flow to prevent excessive cooldown of the RCS following a reactor trip. 6.1.17.2 Test Method The FWCS valve lineup was set to duplicate that which wculd exist immediately following a re-actor trip. While maintaining reactor power at 5% steady state, the flow demand voltages from FbCS's 1 and 2 were monitored to determine required flow for decay heat. After maintaining the required flow for fifteen minutes, the aver-age flow was determined and the RTO bias po-tentiometer was adjusted to a value corre-sponding to the average flow demand signal. 6.1.17.3 Test Results Several problems were incurred during the per-formance of this test: A. At this low power level, steam generator levels are very sensitive to any feed-water control and the operators were un-able to maintain levels within the band of 70.5% + 10%. B. There were problems with the feedwater pump recirculation breaker tripping and closing the valve when the controller was in automatic. In order to prevent trip-ping the plant, the controller was placed in manual. C. The third problem was the FWCS #2 Master Controller. The procedure required both Master Controllers to be in manual, how-ever, operators were having troub'e con-trolling FWCS #2 in manual so it was put 9 in automatic. During performance of the fifteen minute flow test, FWCS #1 was in manual. The voltage out-put had large discrete changes rather than smooth transients which made it difficult to determine an average voltage. The average

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17G voltage from FWCS #2 as determined from the

            '> rush recorder charts was 1.5V. This value was then input to the Reactor Trip Override bias on both FWCS #1 and #2.

6.1.17.4 Conclusions A. Although the FWCS was not operating satis-factorily during this test sequence, the voltages obtained were input as prelimi-nary settings until a later time at which better values could be obtained. During the trip from 20% power, the FWCS operation and RCS cooldown rate were monitored to determine the validity of the results. The settings were slightly high and were appropriately adjusted at that time. B. The test was not performed exactly as originally written. However, the intent, which was to set the RTO bias potentio-meter, was accomplished satisfactorily. 3 {\DI

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171 6.1.18 CONDENSATE AND FEEDWATER SYSTEM POWER ESCALATION TESTS 6.1.18.1 Purpos e The purpose of this test was to: A. Obtain base operating data while demon-strating the ability of the Main Feed . water System to supply the steam gener-ators at the required pressures, tem-peratures, and flows under all antici-pated steady state conditions. B. Verify the proper operation of the FWP recire, valves. 6.1.18.2 Test Method With the reactor at approximately 20% power, the Feedwater Control' system is placed in Mode 1 (full auto) and flows are allowed to stabi-lize. Following flow stabilization, main feed pump data and flow valve position data is re-corded from computer points and trend groups. The 20% baseline 65ca was also obtained from local readings and computer data points as noted in the procedure. 6.1.18.3 Test Results The results confirmed that the main feedwater pump minimum flow valves maintain the required pump suction flow. Also, the base data ob-tained a the 20% power level was in agreement with guidelines per the GE heat balance diagrams. 6.1.18.4 Conclusion At the 20% plateau test sequence, the base operating data was obtained and proper opera-tion of the FWP recire. valves was verified. Also, the ability of the Main Feedwater Sys-tem to supply the steam generators at the re-quired pressures, temperatures, and flow rates was demonstrated.

172 6.1.19 MAIN TURBINE ELECTRO HYDRAULIC CONTROL TESTS 6.1.19.1 Purpose This test was performed to verify that the Electrohydraulic Control System (EHC) func-tioned properly during the 20% trip. 6.1.19.2 Test Method Baseline data was collected prior to the tur-bine trip. Data was obtained on each hydraulic fluid pump while the other pump was in standby. During the 20% trip, a watchstander observed the Main Stop Valves, Main Turbine Control Valves, and Main Turbine combined Intercept Valves. 6.1.19.3 Results and Conclusion All the EHC controlled valve indications were verified as closed. The EHC system functioned properly during the 20% trip. 9 o (3 a 1

173 6.1.20 FEEDWATER HEATER VENTS, DRAINS AND WATER INDUCTION TESTS 6.1.20.1 Purpose The purpose of this test was two fold: A. To demonstrate the satisfactory operation of the Feedwater Heaters during steady-state conditions, and B. To demonstrate the satisfactory operation of the Feedwater Heater and Heater Drain Tank dump and dump bypass valves to per-form their function in the event of high heater shell and drain tank levels. 6.1.20.2 Test Method Each individual Feedwater Heater shell and drain was instrumented with appropriate pres-sure gauger, to allow test personnel to monitor the performance of the heaters. Baseline data including process computer per-formance calculations to determine Feedwater Heater Terminal Temperature Difference and Drain Cooler Approach Temperatures. 6.1.20.3 Test Results The required baseline feedwater heater data was obtained for this plateau. It was determined that the pressure transducers and transmitters installed on the #7, #6, and #5 feedwater heater shells were not ranged properly, i.e., they will not read sufficiently low enough to indicate the near vacuum conditions present in these heaters at low powers. 6.1.20.4 Conclusions Since design data does not exist for the secon-dary plant at 20% power, the data gathered serves an information purpose only. New pressure sensing equipment is presently on order for the #7, #6, and #5 heaters. c(js

174 6.1.21 VIBRATION AND LOOSE PARTS MONITOR (V& LPM) TESTS 6.1.21.1 Purpose The purpose of this test was to provide base-line data for core vibration and loose parts monitoring at 0% and 20% of reactor power. 6.1.21.2 Test Method This test consisted of basically two parts: Individual Reactor Coolant Pump (RCP) Baseline Data, and RCS V & LPM Baseline Operating Data. These are discussed below in more detail. A. Individual RCP Baseline Data (< 0% Power) For each individual RCP, a tape recording was made on the V & LPM capturing that RCP's start up as well as steady state operation. The recording was then aurally and spectroscopically (by frequencies) evaluated in order to identify any ab-normal indications. This sequence was performed for each RCP with the other RCP's secured. B. RCS V & LPM Baseline Operating Data At reactor power levels of 0%, and 20% power, baseline V & LPM data was obtained for steady state operating conditions. For each area of the RCS which is moni-tored by the V & LPM, data was acquired via tape recordings and frequency / power spectrum plots. In addition, during these data acquisition runs, various parameters were trended for N 5 minutes on the plant computer. 6.1.21.3 Test Results The data as described above was obtained at the required power levels. 6.1.21.4 Conclusions All data was obtained per procedure and ac-ceptance criteria were satisfactorily met.

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175 6.1.22 HEATING, VENTILATING AND AIR CONDITIONING SYSTEMS PERFORMANCE TESTS 6.1.22.1 Purpose The purpose in performing this test procedure was to:

1) Demonstrate the satisfactory performance of plant heating, ventilating and air conditioning (HVAC) systems under ac-tual operating heat load.

2) Demonstrate that the HVAC system will satisfy the design criteria at cold shut-down conditions.

3) Provide baseline temperature and/or pres-sure data at selected points of the plant for future reference.

6.1.22.2 Test Method This test was performed at cold shutdown and 20% power af ter plant conditions had stabilized for 24 hours. Cold shutdewn data was taken dur-ing the winter months to ensure the coldest ambient conditions. The HVAC system status was verified to be in the correct operating mode. Data was taken at selected points in the plant. Temperatures outside of containment were taken using a hand held thermocouple and containment temperatures were read remotely, using installed RTD's. 6.1.22.3 Tect Results Temperatures were taken throughout the plant in accordance with the procedure. 6.1.22.4 Conclusion The HVAC systems operate properly to maintain all plant temperatures within design tolerances. 968 072

176 6.1.23 BIOLOGICAL SHIELD RURVEV TE.STS 6.1.23.1 Purpose The test was conducted to accomplish the fol-lowing objectives: A. Determine background radiation levels - prior to initial criticality. B. Evaluate the adequacy of plant radiation shielding. C. Determine radiation levels throughout the plant at various power levels. 6.1.23.2 Test Method A comprehensive series of gamma and neutron dose rate level surveyr. were performed prior to initial criticality and at two low power level conditions. Low power shield tests were conducted at a steady state power level between 0% and 5% power (low power shteld test #1) and between 15% and 20% power (low power shield test #2). Dose rate surveys were taken at numerous loca-tions which included but were not limited to the following areas: A. Locations inside the Reactor Building. B. Areas adjacent to the Reactor Building wall. C. Around penetrations through the Reactor Building wall. D. Selected points in the Turbine and Auxil-iary Building. Radiation dose rate levels at each measurement point were compared at different power levels to verify that a linear relationship existed. This was done to ensure that an extrapolation of dose rates to 100% power could be considered valid thus allowing for identification of poten-tial problem areas. 6.1.23.3 Test Results There were several areas where the design ra-diation levels were exceeded or were expected to be exce:eded. These areas are risted in Table 6.t.23.1 along with suggested corrective actions

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177 6.1.23.4 Conclusion Three major acceptance criteria were established to judge radiation dose rate levels. A. Radiation levels should meet the radiation zoning criteria established by the FSAR.

a. This criteria was satisfied for the Low Power Shield Test #1 (0% - 5%

power) but not for the Low Power Shield Test #2 (15% - 20% power). See Table 6.t.23.1 for exceptions. B. Radiation levels in unenclosed areas out-side the Reactor Building should not be greater than 0.8 mrem /hr.

a. This criteria was satisfied for Low Power Shield Tests #1 and #2 and is expected to be met at full power based on extrapolation.

C. Radiation resulting from streaming through penetrations, shielding defects, etc., will not cause a significant hazard to personnel.

a. This criteria was satisfied for Low Power Shield Test #1 but not for Test

                     #2. See Table 6.1.23.1 for exceptions.
                                                         #'T .

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178 TABLE 6.1.23.1 Areas of Higher than Expected Dose Levels AREA PROBLEM SUGGESTED CORRECTIVE ACTIONS

1. Reactor Building Dose Rate is ex- These areas are not expected Elevation 424 East pected to exceed to require frequent or pro-and West of Canal. 100 mrem /hr at longed personnel access dur-100% power. ing power operation and therefore posting of these

2. Reactor Building Dose rate exceeds areas should be sufficient Elevation 405 100 mrem /br at to ensure personnel pro-East and West of 20% power tection.

Canal

3. Reactor Building Gamma streaming These penetrations are Elevation 357 exceeds 100 mrem / greater than 6' above the Penetrations 1, hr at 20% power floor and are thus noramlly 5, and 7 or is projected considered inaccessible to to exceed 100 personnel. However, special mrem /hr at 100% maintenance could require power access to these areas. If measurements at higher power

4. Reactor Building levels confirm that a dose Elevation 335 rate in excess of 100 mrem /

Penetration 4 hr is anticipated at 100% power, then an engineering evaluation should be made to determine if a simple fix could be found to shield these penetrations. If a simple fix is not available, the permanent posting of these penetrations should be adequate.

5. Auxiliary Buidling The shield wall An engineering analysis of Elevation 335 Sec- around the puri- the area concluded that no tion A fication deminer- radiation dose rate problem alizers does not existed at the present time.

go all the way up Further analysis may be re-to the ceiling. quired based on testing at When personnel higher power levels. stand above filter housings for 2F4A or B, work on over-head equipment in the area, or climb over the wall to the valve gallery for the ion ex-changers, they are in a direct line with the ion exchangers. Also, a general area ra-diation problem is anticipated-in the hallway due to scat- -r ter over the wall O{@ Ol3

179 6.1.24 STEADY STATE VIBRATIONS TESTS 6.1.24.1 Purpose The purpose of this test was to monitor pipe vibrations of the systems listed below during all significant plant operating modes that were likely to cause vibration in the subject pip-ing system. A. Gaseous Waste System (Surge Tank 2T17 to CV-2428). B. Penetration Room Ventilation System. 6.1.24.2 Test Method A walkdown and visual examination of each sys-tem was conducted at each specified test mode. Piping was observed for excessive or abnormal vibration. In addition to the visual inspec-tion of the RCS piping, the reactor coolant pump vibration monitors were checked to verify that no alarm condition was present. 6.1.24.3 Test Results No excessive or abnormal vibration was detected in any of the above listed piping systems. 6.1.24.4 Conclusions Vibrations of all piping systems are acceptable as determined by visual inspection. (\u \ O

180 6.1.25 DYNAMIC TRANSIENT TESTS 6.1.25.1 Purpose The purpose of this test was to verify the adequacy of the piping restraints for lines under transient loads as follows: A. Gaseous Waste System (Surge Tank to CV-2428). B. Penetration Room Vent System. 6.1.25.2 Test Method. A walkdown and visual examination of the piping and piping supports was conducted while the systems were repeatedly turned on and off to induce the transients. 6.1.25.3 Test Results No excessive abnormal pipe or hanger movement was noted during the transients. 6.1.25.4 Conclusions Dynamic response of the Gaseous Waste System and Penetration Room Vent System is acceptable as determined by visual inspection.

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181 6.1.26 TURBINE GENERATOR LOADING AT POWER TESTS 6.1.26.1 Purpose The purpose of this test was to perform the initial generator loading, overspeed testing, exciter adjustments, and turbine balance as required. Turbine and generator baseline data w- also collected for evaluation and future reference. 6.1.26.2 Test Method The turbine generator was accellerated to 1800 RPM in accordance with the Turbine Startup Operating Procedure (OP 2106.09) at which time

                 ! the following tests were performed:

A. Generator Excitation B. Generator Synchronization C. Loading to 10% Power D. Overspeed Testing 6.1.26.3 Test Results A. During the performance of the initial roll and subsequent generator loading, several unplanned turbine generator trips occurred. The trips were attributed to high vibra-tions of #2 Low Pressure Turbine. A maxi-mum vibration of 21 mils was recorded on bearing T-2. However, the vibration pro-blems were corrected with the addition of five balance shots. B. Exciter checks were made in accordance with GEK 14870, "Off-Line Tests Generator Run-ning," " Exciter Unloaded," during which time two turbine generator trips occurred, one from a high volts /Hz reading and one from an anti-motoring signal. Both were resolved and testing continued. C. All generator synchronization tests were conducted with the exception of the Auto Synchronization which will be run during a subsequent start up. The transfer of auxiliaries to the startup transformer was performed during an actual reactor trip rather than simulated as initially planned. gS3 07 B

182 D. During each of the three overspeed trip tests, the turbine tripped at 1959 RPM. Although the overspeed trips were expected to occur between 1980 and 1998 RPM, the results were determined to be satisfactory. 6.1.26.4 Conclusions All required tests were conducted, with the exception of the Auto Synchronization, as noted previously. Testing to this point was satisfactory and there are no problems related to this test procedure that would prohibit testing at higher power levels. e

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183

7.0 CONCLUSION

February 1, 1979, the reactor was tripped completing the majority of the 20% test plateau. An extended maintenance outage was commenced to modify a design defect in the reactor coolant pump motors, repair and retest the main steam safety valves, implement a design change in the main condenser hot wells, install a cooling tower by pass line, replace a hydro-gen seal on the main generator and repair the number 2 diesel generator engine. On June 6, 1979, the reactor was taken critical. Deficiencies from the 20% plateau were retested and all remaining testing required for escalation to 50% power was completed. 50% power was achieved June 24, 1979. The 50% power plateau testing commenced and is in progress at this time. Scheduled testing has been interrupted intermittently while investigating an anomaly with the RCS hot leg tempera-ture. A supplementary report will be issued describing the post 20% power test as required by the Unit 2 Technical Speci-fication and Regulatory Guide 1.16. f4 W e,@

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