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Text

Startup Rept
	ML20080K059
Person / Time
Site:	Millstone
Issue date:	07/22/1986
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20080K055	List: B12168, Forwards Summary Rept of Plant Startup & Power Escalation Testing Following Receipt of Plant,Unit 3 Operating License, Per Plant,Unit TS Section 6.9.1.1; ML20080K059;
References
	NUDOCS 9502280289
	Download: ML20080K059 (300)
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STARTUP.

REPORT DOCKET NO. 50-423

LICENSE NO. NPF-49 O

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MILLSTONE UNIT 3 O NORTHEAST NUCLEAR ENERGY COMPANY

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TABLE OF CONTENTS

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l SECTION PAGE NO.

l LIST OF TABLES v LIST OF FIGURES viii FORWARD 1

1.0 INTRODUCTION

2 2.0 SUM 4ARY CHRONOLOGY 3 3.0 PREOPERATIONAL TEST PROGRAM OVERVIEW 6 4.0 INITIAL FUEL LOAD 17 4.1 Summary Chronology 33 5.0 POST CORE HOT FUNCTIONAL TESTING 34 5.1

Shutdown Margin Verification 35 5.2 Incore Thermocouple /RTD Testing 36 5.3 Rod Control Slave Cycling /CRDM Timing 38 5.4 RCS Leak Detection 42 5.5 Pressurizer Heater and Spray Testing 45 5.6 Rod Orop Testing 49 5.7 RCS Flow Measurement 56 1 I

5.8 RTD Bypass Loop Verification 58 5.9 Movable Incore Detectors 60 5.10 Digital Rod Position Indication Testing 62 5.11 Loose Parts Monitoring System Testing 64 j 5.12 RCS Flow Coastdown 67 i 5.13 Rod Control Operational Testing 74 5.14 CVCS System Testing 77 5.15 RCS Loop Stop Valve / Coolant Pump Interlock Testing 80 6.0 INITIAL CRITICALITY 82 ,

6.1 Summary Chronology 88 O

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SECTION PAGE NO. )

Il 7.0 LOW POWER PHYSICS TESTING 89 7.1 Hot Zero Power Test Range Determination 91 7.2 Reactivity Computer Checkout 92 7.3 Boron End Point Determination 96 l 7.4 Isothermal Temperature Determination 98 j 7.5 Control Rod Bank Worth Measurement 101:

7.6 Zero Power Flux Maps 123 l 7.7 Pseudo Ejected Rod Test 130 j 7.8 Natural Circulation Test 133 l 8.0 POWER ASCENTION TESTING 141 l 8.1 NSSS Testing 8.1.1 RCS Flow Measurement 142 )

8.1.2 Power Coefficient Measurement 144 l I

8.1.3 RCS Boron Measurement 146 8.1.4 Core Power Distribution Measurements 147 8.2 Instrumentation Calibr.ation and Alignment 8.2.1 Operational Alignment of Nuclear Instrumentation 155 )

8.2.2 Operational Alignment of Process !

Temperature Instrumentation 157 8.2.3 Steam and Feedwater Flow l Instrumentation. Calibration 158 j 8.2.4 Axial Flux Difference Instrumentation Calibration 159 8.3 Control System Dynamic Testing 8.3.1 Reactor and Turbine Control 176 8.3.2 Atmospheric Steam Dump Control 179 l 8.3.3 Automatic Reactor Control 181 8.3.4 Automatic Steam Generator Water Level Control 187 8.3.5 Main Steam Line Isolation Valve j O' Closure Test 190 !

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TABLE OF CONTENTS (cont)

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8.4 Plant' Transient and Trip Testing 8.4.1 Turbine Overspeed 193 8.4.2 10 Percent Load Swing 194 8.4.3 Reactor Trip and Shutdown Outside

Control Room 202 8.4.4 Large Load Reduction 204

- 8.4.5 Loss of Power (20 Percent Power) 211 8.4.6 . Generator Trip (100 Percent Power) 217 8.5 General Plant Testing 8.5.1 Calorimetric .225 8.5.2 Secondary Plant Performance 227 8.5.3 Radiation Surveys and Process Radiation Monitoring Testing 235

8. 5. 4 ' Ventilation Systems Operability )

. Verification 243 ,

8.5.5 Chemistry and Radiochemistry Measurements ,

245 8.5.6 Neutron Shield Tank Cooling System Testing 249 8.5.7 Containment Hot Pipe Penetration-Cooling System Monitoring. 250 8.5.8 Turbine Plant Component Cool,ing Water ,

System Balancing 251' 8.5.9 Piping Fluid Transient Vibration Monitoring 252 8.5.10 Thermal Expansion and Restraint Monitoring 253 8.5.11 Loose Parts Monitoring System Testing 254 9.0 WARRANTY RUN 256 -l 9.1 Calorimetric 257 9.2 Secondary Plant Performance '258 1 e

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SECTION PAGE NO.

- APPENDICES A. FSAR Test Deviations 263 B. Startup Test Procedure Listing 266 C. Preoperational Tests ',ompleted During the Startup ,

Test Program 269 O. Summary of Problems Encountered During the Loss of Power Test (3-INT-8000, Appendix 8030) 273 9

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5.6-1 Rod Drop Times - Rod Drop Testing 52 5.7-1 RCS Flow Data - Precritical RCS Flow Measurement Test 57 5.8-1 RTD Bypass Loop Flow Data - Bypass Loop Flow Verification 59 7.2-1 BOL Delayed Neutron Parameters 94 7.2-2 Reactivity Computer Checkout Data 95 7.3-1 Summary of Boron Endpoint Test Results 97 7.4-1 Summary of ITC Test Results 99 7.5-1 Summary of Rod Worth Test Results 103 7.6-1 Core Power Distribution Measurement - HZP RIL 125 7.6-2 Core Power Distribution Measurement - HZP RIL D-12 ;

Ejected 126 7.6-3 Core Power Distribution Measurement - Control Bank D Inserted 127 7.6-4 Core Power Distribution Measurement - ARO HZP 128 7.6-5 Core Power Districution Measurement - Six Pass !

Symmetric Thimble Tilt Check 129 )

7.7-1 Pseudo Ejected Rod Test Results 132 l 8.1.1-1 RCS Flow Data - RCS Flow Measurement Test 143 )

8.1.2-1 Doppler Only Power Coefficient Verification 145 8.1.4-1 Core Power Distribution Measurement -

30 Percent Power 148 8.1.4-2 Core Power Distribution Measurement -

50 Percent Power 149 8.1.4-3 Core Power Distribution Measurement - ;

150 i 75 Percent Power - ARO '

8.1.4-4 Core Power Distribution Measurement -

75 Percent Power 151 8.1.4-5 Core Power Distribution Measurement -

90 Percent Power 152 1

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. i 8.1.4 Core Power Distribution Measurement -

100 Percent Power - Map 1 153 ,

.. 8.1.4-7 Core Power Distribution Measurement - ,

100 Percent Power - Map 2 154- !

8.2.4-1 Incore/Excore Cross-Calibration Preliminary Test -

50 Percent Power 163 )

8.2.4-2 Incore/Excore Cross-Calibration - Test 1 -

75 Percent Power 164 8.2.4-3 Incore/Excore Cross-Calibration - Test 2 - J 75 Percent Power 165

, 8.4.4-1 Plant Parameter Transient Extreme Values - I Large Load Reduction Test 206 8.4.6-1 ~ Plant Transient Data - Generator Trip From 100 Percent Power 220- I 8.5.1-1 Plant Calorian Mic Data . 226-8.5.2 Turbine-Generator. Performance Data - Secondary j Plant Performance Testing 232 8.5.2-2 Feedwater Heater Performance Data - Secondary-

-Plant Performance Testing 233 ]

8.5.2-3 Moisture Separator / Reheater Performance Data - !

Secondary. Plant Performance Testing 234-

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8.5.5-1 RCS Chemistry Limits 246 j 8.5.5-2 RCS Cheniistry Analysis Data 247 1

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@,_) LIST OF FIGURES SECTION PAGE NO.

4.0-1 Initial Core Loading Sequence - Steps 1 to 7B 20 4.0-2' Initial Core Loading Sequence - Steps 7C to 7D 21 4.0-3 Initial Core Loading Sequence - Steps 8 to 34B 22 I

4.0-4 Initial Core Loading Sequence - Steps 35 to 55C 23 4.0-5 Initial Core Loading. Sequence - Steps 550 to 56B 24 4.0-6 Initial Core Loading Sequence - Steps 57 to 86B '25 4.0-7 Initial Core Loading Sequence - Steps 87 to 1188 26 4.0-8 Initial Core Loading Sequence - Steps 119 to 158B 27 4.0-9 Initial Core Loading Sequence - Steps 159 to 193 28 4.0-10 Initial Core Loading Sequence - Figure Legend 29 4.0-11 Source Range Detector Response - Initial Core Loading 30 l 4.0-12 Temporary Detector Response - Initial Core Loading 31 5.3-1 Typical CRDM Oscillograph Trace - Rod Withdrawal 40 f

5.3.2 Typical CRDM Oscillograph Trace - Rod Insertior' . 41 5.6-1 Typical Rod Drop Computer Trace - Rod Drop Testing 55 5.11-1 Typical Spectrom Analyzer Plot - Loose Parts Monitoring System 66-

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5.12-1 Total Normalized Core Flow - One Loop Coasting Down 72 ,

5.12-2 Total Normalized Core Flow - Four Loops Coasting Down 73 6.0-1 ICRR VersusS' hutdown Bank Position - Preoperational Rod Withdrawal 84 6.0-2 ICRR Versus Control Bank Position - Precritical Rod Withdrawal 85 6.0-3 ICRR Versus Time - Dilution to Criticality 86 6.0-4 ICRR Versus RCS Boron Concentration - Dilution -

to Criticality 87

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LIST OF FIGURES (cont)

SECTION PAGE NO.

a.

7.0-1 Zero-Power Testing Connections 90 7.4-1 Rod Withdrawal Limits 100 7.5-1 Control Rod Worth Measurements - Typical Reactivity Trace 104 7.5-2 . Integral Control Rod Worth , Control Bank A 105 7.5-3 Differential Control Rod Worth - Control Bank A 106 7.5-4 Integral Control Rod Worth - Control Bank B 107 7.5-5 Differential Control Rod Worth - Control Bank B 108 7.5-6 Integral Control Rod Worth - Control Bank C 109 7.5-7 Differential Control Rod Worth - Control Bank C 110 7.5-8 Integral Control Rod Worth - Control Bank D 111 7.5-9 Differential Control Rod Worth - Control Bank D 112 7.5-10 Integral Control Rod Worth - Shutdown Bank A '113 7.5-11 Differential Control Rod Worth - Shutdown Bank A 114 7.5 Integral Control Rod Worth - Shutdown Bar.k B .115 7.5-13 Differential Control Rod Worth - Shutdown hnk B 116 7.5-14 Integral Control Rod Worth - Shutdown Bank C 117 7.5-15 Differential Control Rod Worth - Shutdown Bank C 118

. 7.5-16 Integral Control Rod Worth - Shutdown Bank D 119 7.5-17 Differential Control Rod Worth - Shutdown Bank D 120 7.5-18 Integral Control Rod Worth - Shutdown Bank E 121_ ,

7.5-19 Differential Control Rod Worth - Shutdown Bank E 122 $

7.8-1 Pretest Core Exit Thermocouple Map - Natural Circulation Test 136 7.8-2 Stable Core Exit Thermocouple Map - Natural l Circulation Test 137 !

7.8-3 Typical RCS T hot and T cold Plot - Natural Circulation Test 138 .

7.8-4 Pressurizer Level and Pressure Plot - Natural Circulation Test 139 7.8-5 Typical Ste'am Generator Level and Pressure Plot -

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' Natural Circulation Test 140 l l

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LIST OF FIGURES (cont)

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I 8.2.4-1 ' Axial Flux Difference Versus Time - Test 1 -

E' '75 Percent Power 166 6 8.2.4-2. Incore AQ Versus Channel 41 Current - Test 1 - t e 75 Percent Power. 167 !

8.2.4-3 Incore AQ Versus Channel 42 Current : Test 1 -

75 Percent Power 168 ,

8.2.4-4 Incore AQ Versus Channel 43 Current - Test 1 -

75 Percent Power 169 ,

, 8.2.4-5 Incore AQ Versus Channel 44 Current - Test 1 -

75 Percent Power 170 8.2.4-6 Axial Flux Difference Versus Time - Test 2 -

75 Percent Power 171 8.2.4-7 Incore AQ Versus Channel 41 Current - Test 2 -

75 Percent Power 172 8.2.4-8 Incore AQ Versus Channel 42 Current - Test 2 -

75 Percent Power 173 8.2.4-9 Incore AQ Versus Channel 43 Current - Test 2 - ,

75 Percent Power 174 8.2.4-10 _Incore AQ Versus Channel 44 Current - Test 2 - ,

75 Percent Power 175 .

L 8.3.1-1 RCS Temperature and Steam Generator Pressure as'a Function of Reactor Power 178 8.3.3-1 Typical Plant Transient Response Plot - Automatic Reactor Control Test 183 ,

8.3.5-1 Typical Plant Transient Response Plot - Main Steam Isolation Valve Closure Test 191 8.4.2-1 Typical Plant Transient Response Plot - 10 Percent

. . Load Swing Test ~197 ,

8.4.4-1 Typical Plant Transient Response Plot - Large Load Reduction Test 207 8.4.5-1 ' Typical Plant Transient Response Plot - Loss of

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Power Test 214 1

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SECTION PAGE-NO.

8.4.6-I- Typical Plant Transient Response Plot - Generator Trip From 100 Percent Power- 221 9.2-1 Specified' Heat Rate Warranty Curve - Secondary Plant Performance Testing- 261-9.2-2 Full ARC Specified Heat Rate Curve - Secondary -

Plant Performance Testi'ng 262 l

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FORWARD This ' report' addresses the conduct and results of the . startup test

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program for Millstone Unit 3 and spans the period from Initial Fuel Loading through Commercial Operation and Warranty Run. It is i

submitted in accordance with the requirements of USNRC Regulatory Guides 1.16, Revision 4, and 1.68, Revision 2, and Millstone Unit 3 Technical Specification 6.9.1.1. ,

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1.0 INTRODUCTION

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Millstone Unit 3 consists of a Westinghouse 4 loop pressurized water nuclear steam supply system rated at 3411 MWT and a General Electric turbine generator rated at 1204 MWE. The overall net electrical

, output of the unit is 1150 MWE. Millstone Unit 3 is located adjacent to Millstone Unit 1 (a 660 MWE General Electric BWR)- and j Millstone Unit '2 (a 870- MWE Combustion Engineering PWR) on ' an I i

approximately 500 acre site on the north shore of Long -Island Sound '

j in the town of Waterford, Connecticut. The unit utilizes a j subatmospheric containment design with a supplemental ' leak collection and release system (secondary containment) to further .

limit offsite releases in the event of a design basis accident.

The ownership of Millstone Unit 3 is divided among 15 joint owners.

F The majority owners are the Northeast Utilities subsidiaries.

Connecticut Light and Power Company and Western -Massachusetts >

Electric Company. The remaining portion is divided among -13

.New England public and private utilities.

4 The joint owners have designated' Northeast Nuclear Energy Company (NNEco), a wholly owned subsidiary of Northeast Utilities, to act as their agent and representative in matters relating to the design, ,

construction, testing, licensing, operation and maintenance of-Millstone Unit 3. NNEco presently performs a similar function for.

Millstone Units 1 and 2. The unit was designed and constructed by- l I

Stone & Webster Engineering Corporation.

The unit was constructed under Construction Permit CPPR-113 and t currently operates under Operating License NPF-49. Operating License NPF-44 was issued on November 11, 1985 to permit initial .

fuel load and low power operation (not to exceed 5 percent of rated ,

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thermal power). Operating License NPF-49 was subsequently issued on

- January 31, 1986 to permit full power operation.

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(,l.l 2.0 ~ PROJECT SUM 4ARY CHRONOLOOY The following is provided as an overview of the major milestones in the chronology of Millstone Unit 3.

DATE EVENT 08-09-74 Construction Permit CPPR-113 issued by the then Atomic Energy-Commission (AEC).

09-74 First structural concrete (turbine building) is placed.

04-75 Rebar placement for the containment mat begins.

09-78 First containment wall concrete is placed.

07-79 The turbine generator stator is set in place.

. 10-80 Reactor vessel and containment polar crane are set in place.

06-81 Steam generator erection is begun.

11-82 Emergency diesel generators are installed.

01-17-83 The system turnover process and preoperational test program are begun.

07-18-83 The reserve station service transformers (RSST) are energized.

12-09-83 Energization of 4160 volt switchgear is begun. .

12-03-84 to Perform steam generator secondary side A 12-22-84 hydrostatic testing.

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-DATE EVENT 1 04-16-85. ; Receive Special Nuclear Material (SNM) license.

SNM-1950.

04-19-85.to Perform RCS cold hydrostatic testing. ;

04-25-85 j i

l 04-24-85. The first shipment of reactor fuel is received. j

. i 0F-15-85 Unit 3 emergency drill is' conducted.

1 Perform turbine building hot functional testing.

06-10-85 to 10-19-85 s

07-10-85 to Perform .the ~ containment structural integrity l 07-24-85 test (SIT) and integrated leak rate test-(ILRT).

07-24-85 The last. shipment of reae. tor fuel is received. ~.

08-16-85 to Perform the engineered safeguard features (ESF)- j

' test 09-06-85 09-27-85 to Perform pkcore hot functional test. j 11-02-85 09-17-85 Perform initial turbine roll utilizing RCP heat as the heat source.

11-25-85 Receive Operating License NPF-44 permiting fuel l load and operation' up to 5% reactor power.

11-26-85 to Perform initial fuel loading. Startup test 12-03-85 program begins.

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' ' " EVENT-DATEL Perform post core hot functional testing.

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'01-23-86 Initial criticality achieved at 2200 hours0.0255 days 0.611 hours 0.00364 weeks 8.371e-4 months .-

01-24-86 to Perform low power physics testing (LPPT).

01-31-86. .

jJ 01-31-86 Receive Full Power . Operating License NPF-49.

02-01-86 to Perform the power ascension test program.

04-21-86 r

02-15-86 Achieve 30% power.

02-16-86 Initial synchronization to the grid. '

03-17-86 Achieve 50% power.

03-26-86 Achieve 75% power.

04-15-86 Achieve 90% power.

04-17-86 Achieve 100%-power.

04-23-86 Commercial operation is declared.

04-25-86 Perform the unit warranty run. Complete the .;

04-29-86 startup test program.

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- L3 .0 -PREOPERATIONAL TEST PROGRAM OVERVIEW

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The Preoperational Test Program officially began with the first-

. system turnover from Construction: to Startup, on 01-10-83,'of-9 ]

the' motor control centers to support the water treatment l system. This turnover process continued for both systems and I buildings ' through ' to completion- of system turnovers,. on 06-05-85, of the yard security system and the completion of building turnovers, on 11-04-85, of the yard area. This was the j last of 234 turnover packages.

4 The Preoperational Test Program included component' testing and i system flushing which, in most . cases, preceded the l preoperational testing of systems. System pressure testing (except steam generator and RCS hydros) was performed prior to ]

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-system turnover. Preoperational testing continued through 1983' and ' 1984, leading up to the transition to milestone' testing. !

Major milestones that were established are listed below along ]

with the start and completion dates for each milestone.

Milestone Date Start /Date Complete j Plant on Permanent Power 07-18-83 .

Steam Generator Hydro 12-04-84/12-20 i RCS Cold Hydro 04-14-85/04-24-85 .j Fuel Receipt 04-24-85/07-24-85' Emergency Drill 05-15-85 Containment ILRT 07-12-85/07-15-85 Engineered Safety Features Test 08-16-85/09-06-85 Turbine 8uilding Hot Functional Test 06-10-85/10-19-85 Precore Hot Functional Test 09-27-85/11-02-85 j i

i A summary description of each milestone follows.

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l V. Steam Generator Hydrostatic Test i l

ll The Steam ' Generator Hydrostatic Test involved the hydrostatic testing of the secondary side of the four steam generators and their associated piping. This milestone was subdivided into one test for each generator. The boundaries for each test included the attached piping. systems out to the nearest isolation points. For main steam piping, the main steam j isolation valves provided isolation and the ' main feedwater j piping was isolated at the steam generator feedwater stop : l l

valves. The remaining piping systems were isolated inside l containment- by installation of blank _ flanges or valve positioning.

The generators were filled for test with water from the condensate storage tank after being preheated to 180*F. A temporary transport system was utilized from the discharge side j q) t of the condensate system makeup pumps through the containment ]

equipment hatch to each generator. A recirculation skid was .

provided to assist in chemical addition and temperature :

maintenance prior to start of the test within the 120*F to 180*F test range. j The hydrostatic testing to 1570 psig began with the "A" generator, which completed its test on 12-04-84, and concluded .

l with the last generator test completed on 12-12-84. Tube to j tubesheet leaks were detected on generators A, B and C. i Subsequent to repair of the detected tube sheet leaks, retesting was performed. This activity incorporated six separate tests with a maximum test pressure of 840 psig. This testing commenced 12-12-84 and was completed on 12-20-84.

Following completion of the test, the Steam Generators were placed in a wet lay-up condition with a nitrogen overpressure.

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gf Reactor Coolant' System Cold Hy'drostatic Test i

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- The Reactor Coolant System (RCS)' Cold Hydrostatic Test involved i o

the pressure testing'.of the reactori ves'sel and associated piping / components to 3107~psig. In addition,-the test involved q

+ the initial fill and venting of the RCS as well.as the initial

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operation ofLthe reactor coolant' pumps (RCPs); Prior to assembling -the reactor vessel to close the RCS. pressure i boundary, the reactor vessel internals were installed. - Durin'g i l

p the test, the RCPs were utilized to heat the inventory of the i RCS . above the 4150*F. lower limit " based = on brittle fracture i

concerns.

The assembly sequence for the reactor vessel began on 04-03-85 when preparations were started for reactor vessel internals e

installation. On 04-04-85 the internals were installed - and

[_.. - preparations began . for installation of the vessel: head. The-head was. installed on 04-05-85. The RCS_ fill sequence began on-04-13-85 and. was complete on- 04-15-85. During this ' sequence, the tensioning of the reactor. vessel head was completed _ on 04-14-85. The RCPs were bumped on 04-19-85. The vibration testing runs of the RCPs were completed .on 04-20-85 and the heatup of the RCS was begun. During the period of 04-20-85 to 04-24-85, the pressure boundary was groomed and minor leakage. ]

paths repaired. Final pressurization to test pressure began on 04-24-85 and was completed that day.

Fuel' Receipt l' .The Fuel Receipt milestone was established to provide a framework to accomplish fuel receipt on site with subsequent

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fuel assembly transfer to a safe storage facility. Significant ,

prerequisites to this milestone included' testing of the j following systems: fuel pool cooling and purification,

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radiation monitoring, fuel building HVAC, . fuel building fire

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- },j; protectiori L and. detection, and = . fuel handling equipment. s

~ Additional prerequisites included. fuel building . turnover, establishment'of a physical security plan for_ the fuel building - !

and ' surrounding: areas, operator fuel handling training, and establishment of radiation.and fire protection programs for the fuel building, all of which would lead to receipt of a license ;

a from Lthe NRC to receive and store special nuclear material.

l>i dpon completion of all prerequisites, the NRC issued license

,' SNM-1950 on 04-16-85. Specific fuel shipment scheduling and 1 receipt concerns were re' solved with Westinghouse representatives over the next few days, and the initial receipt f

of 14 fuel assemblies occurred on 04-24-85. The final fuel shipment was received 07-24-85. .

4 I SIT /ILRT ,

The Structural Integrity Test / Integrated Leak Rate Test was.

performed to demonstrate the structural integrity of.

b containment at 1.15 times design pressure and to measure the leak rate from containment at peak accident pressure. Major ~ ,

test prerequisites included completion of Type B and C leakage ;

tests 'on containment isolation valves! and L penetrations (including equipment u and -personnel; hatches), installation of.

pressurization equipment, and containment turnover process. '

f During the performance of the prerequisite activities, some '

delays were caused by Type C test failures, rework and ;

subsequent retest of containment isolation valves.

Initial pressurization for the SIT commenced on- 07-10-85, but 'I this effort' was stopped when an, open containment leakage path was discovered. In this instance, misalignment of Leakage ;

Monitoring System lines penetrating containment resulted in an open-ended pipe. This deficiency was corrected by installation" of a jumper, and pressurization recommenced after an eight-hour delay. Peak pressure of 52 psig achieved within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 'j O.

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. the SIT : was . completed. the morning of 07-12-85 with no' t

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/ . deficiencies noted. Pressurization for the ILRT was' commenced '

nine: hours later; full pressure of 39.4 psig was ' achieved, and !

the test run commenced on- 07-13-85. After a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months hold, 4 leakage was determined _to-be 52.57 scfm (10% of the acceptance ,

> criteria). Depressurization was completed'07-15-85.

- Engineered Safety Features (ESA Test The ESF Test was started on *08-16-85, and ompleted on 09-06-85. The test - was divided into two separati sections:

ESF without loss of power and ESF with loss of power.

The ESF test without loss 'of off-site power was performed with

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the breakers of the major ESF-actuated equipment placed into the test position. This was done to verify safeguard logic before placing, the plant under. the dynamic transients of the i operating equipment. The ESF test with loss of off-site power was then performed to verify emergency diesel performance, ,

correct sequential loading of ESF equipment 'and proper train y separation.

l The performance of. the ESF ' test without loss of power revealed 1 some logic errors' with HVAC equipment and inadequately sized .q slave relays in the Main Steam Isolation Valve control logic. -l These concerns were subsequently corrected and satisfactorily I j

retested.

The ESF test with loss of off-site power revealed a deficiency in the diesel sequencer logic in that the diesel output breakers failed to close.due to incorrect time delay settings on certain control relays. Also, several electrical busses were not stripped during the LOP, Orange Train test. These problems were resolved and successfully retested. i 1 ...--

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Y Turbine Building Hot Functional Test if ,

The overall purpose of the Turbine Building Hot Functional Test.

(TBHFT) was to prepare, cleanup and test the secondary side of-

.the plant utilizing Auxiliary Steam to ensure system operability, .and to establish a- level of reliability for

' integrated system operation. This was all in preparation to support the- activities ; associated with Precore Hot Functional and subsequent Startup.'and Power Ascension Test'ing. The' test procedure (3-INT-2006) was utilized as a controlling document ,

which . integrated and. sequenced all the secondary plant activities, i.e.,- plant conditions, Phase II tests, condensate /feedwater train cleanup, operator training . and ;

validation 'of the plant's operating procedures. Major objectives for this' test included:

1. Demonstrate the ability to steam seal the main turbine and feed pump turbines utilizing the gland seal steam system.

Auxiliary boiler steam was utilized - for this process.

2. Demonstrate' the ability to draw vacuum and= maintain a

design pressure (1.5 in HgA) in the condenser. As required, condenser vacuum boundary leaks were to be located and corrected.

3. Demonstra.te' the ability to operate the condensate system )

in the.short and long recycle modes.

4. Demonstrate the ability to clean the hotwell, condensate ]

and feedwater systems prior to feeding forward through the j use of the condensate mixed bed deminerali7ers. In conjunction with this process, the proper operation of the condensate chemical feed system and portions of the l turbine plant sampling system was verified.

5. Perform the initial no-load uncoupled and coupled runs of the main turbine driven feedwater pumps utilizing .

auxiliary steam supplied from the auxiliary boilers.

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ye t

J ' Page 12-j' % ..

k During coupled runs, the feed pumps were operated in the . f recirculation mode only, due to limited steam supply from 1

(

"" -the auxiliary boilers. ]

6. Perform the initial. coupled run of the motor driven J l

feedwater pump. J

7. Perform the Phase ' II. tests for the following systems

1 l

gland seal steam

-condenser air removal secondary plant sampling ( artial)

- condensate system (partial) condensate chemical feed

- feedwater and recirculation The test was released for establishment of initial conditions and- performance of system lineup on 06-10-85. ;

Physical testing began on 06-14-85 when the main turbine was placed on turning gear. Testing. and secondary side system grooming continued until 11-06-85 when the test procedure was officially completed. The procedure was l kept open into the Precore Hot Functional. Test so it could ;

serve as a coordinating document for various balance' of-plant related Phase 3 tests. ;

Several major testing interruptions were experienced- :

during the performance of 3-INT-2006. No ispact 'on the precore hot functional testing or any other milestone l event was caused by these interruptions. l 07-08-85 to 08-11-85 A seawater leak into the hotwell was caused when the condenser air removal . piping in the B condenser, O {

waterbox separated from the' tubesheet face and allowed a seawater ingress into the hotwell. The ,

j I

separation was caused by corrosion of the bolts holding the penetration flange against the tubesheet face. During inspection of all waterboxes, corrosion Os 1

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l 1 $ c : L -

Page 13

)

QJ' s- 1 of - the inlet ! s'ide tubesheets .was ' observed.

- Engineering analysis determined the cofrosion of both

" T 1

the bolt heads and tubesheets' was' the result of improper material compatibility which was. accelerated L.m. ' if by non optimal performance of . the waterbox cathodict protection system. Repairs undertaken included

- - changeout of all air removal line flange bolts with a more resistant alloy, epoxy coating of the inlet side,

~

. tubesheets and inlet waterboxes. Cathodic protection system setup, testing and operator training were

- performed to ensure optimum system . performance.

While the measures were being .taken to correct the cause and results of the corrosion, a full scale flushing program was performed on the condensate and feedwater system, up to feed stops, in - order to

- remove the chloride contaminationu^ caused by the seawater intrusion. The chloride levels in the G\ condensate .and connected systems were brought to acceptable levels and with the mechanical repairs ,

effected, testing was restarted on 08-13-85.

08-18-85 to 09-23 , ;

On- 08-15-85 a crack was discovered- in the upper.

crossover. piping between the A ' and B condensers.

Efforts to temporarily seal the crack using a mastic compound were unsuccessful and the secondary plant was shut. down, vacuum. broken and the hotwell pumped.

down to facilitate repairs. During the process of correcting the crack, additional internal condenser support damage was discovered. Engineering analysis indicated insufficient internal bracing had been ,

installed, and supplemental supports were specified.

After this additional material had been installed, a

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..Page 14 q

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s[ further delay was experienced while the ESF test with.

loss :of.. normal power (3-INT-2004) iwas performed.

3..

During _ this latter_ delay, the . outlet side of the waterboxes were epoxy coated as -a preventative measure.

This was the l'ast delay due to: an: equipment malfunction. TBHFT testing was- recommenced on 09/23/85. By this point the Precore Hot Functional

~ ~

Test was underway, plant heatup was in-progress, and the remaining TBHFT activities were performed in~

parallel with HFT.

In addition to the initial scoped testing for TBHFT on '10-17-85, the initial roll of the main turbine ,

took place. On 10-19-85, the main turbine was.

synchronized to the grid for the first time' and -

approximately 65 MW - generated. .The TBHFT was concluded at this point.

All objectives of the test were met with minor exceptions.

Due to testing and system grooming which took place during the TBHFT, . the secondary side was ' able to fully support PCHFT and the' subsequent startup' testing.

't Pre

Core Hot Functional Test 3

The Pre-core Hot Functional Test started on 09-27-85 and was completed on 11-02-85. In general, all systems required for -

plant operation were tested under normal operating conditions. ;

The major objectives of the test were to take the unfueled plant from a cold shutdown condition, through heatup, testing at normal operating temperature and pressure, and return to a !

cold condition. During this time the following design requirements and system functions were verified:

i. -

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Page 15 j-

%f - Freedom- of movement during < thermal , expansion for major components.

- The capacity of the Chemical and Volume Control System to maintain Reactor Coolant System (RCS) pressure during solid pressure control and to purify the letdown steam while the RCS was at operating' s

pressure.

The operation of the atmospheric steam dump valves and the condenser dump, valves during cooldown and at normal operating system conditions.

- The RCS heat loss to ambient at operating temperature and pressure.

The operability of both the primary and secondary sample systems and chemical addition systems.

- The operability of both the main and auxiliary feedwater pumps, f~ -

The starting up and paralleling of the main 5

N" turbine generator to the grid.

JThe RCS leakage calculation method.

The capability for remote shutdown and cooldown of the reactor plant.

- The initial vibration testing and monitoring of components during normal operation.

- The operability with a heat load of the plant's ventilation systems.

The initial check of the RCS thermocouple /RTO cross-calibration.

- The ability to isolate an RCS loop while maintaining primary pressure control within the isolated loop. l The operation of the ' pressurizer pressure and level control systems.

- The functionality of the Voice Page and Evacuation .

.l Alarm systems with normal plant background noise. j

- The ability of the plant to withstand a loss of l instrument air.

4

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I J Page 16 f~.

L '

This test was also used to passivate the RCS by operating at an elevated (>500*F): temperature for 28 days and to obtain a i 1

minimum of 10 days of RCP flow induced vibration cycles on the i 1

reactor internals.

I E All testing was. covered in the base procedure (3-INT-3000) and 34 associated appendices. All planned testing was . completed :

except for that on the boron thermal regeneration system which, ,

due to equipment problems, was delayed until a la'ter date. The l

deficiencies discovered during ' testing were addressed on a :

schedule consistent with plant and system operability requirements.

4 l

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,f" Page 17

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. 4.d INITIAL FUEL LOAD 3-INT-4000 I

OBJECTIVE L The Initial Fuel Load procedure provides a safa, organi' zed l' method for the initial core load.

DISCUSSION Initial fuel load was conducted over the period'of 11-26-85 to 12-04-85. The operation is summ' a rized in Section 4.1, Initial Fuel Load Chronology.

Prior to fuel load, proper alignment and calibration of the two 1 Source Range channels (SR 31, 32) and the three Temporary Detectors (TD A, B, C) were verified in accordance with' l i

3-INT-4000, Appendix 4003, Core Load Instruments and Neutron

.p. Source Requirements. Baseline background count rates were

( . taken. In addition, a neutron source was lowered near each detector to verify correct channel response. This latter check was required to be- performed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of beginning core load.

From dry storage in the Spent Fuel Pool (SFP), . each fuel assembly was transferred by the Spent Fuel Pool . Bridge and '

Hoist (SFP8H) to the Fuel Transfer System (FTS). After the FTS cart

moved- the fuel into containment, the- Single Integrated' Gripper Mast Assembly (SIGMA) refueling machine would engage the fuel assembly and load it in the proper core location.

Fuel movement in containment was under the direction of a fuel handling Senior Reactor Operator. Overall fuel load operations were directed by Reactor Engineering Personnel. The actual ]

^

loading sequence was controlled by 3-INT-4000, Appendix 4005, "

Initial Core Loading. In addition to delineating all movements ,

for each fuel assembly, this appendix also governed TD movement I

~and provided guidance for obtaining count rate data.

(

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'Page 18-y.

Neutron monitoring was provided by SR 31 and 32 and TD A, B' and f C. 'As each fuel assembly was lowered into the core,. count rates were monitored.~ During the loading sequence, count rate ,

data was collected and analyzed in accordance with 3-INT-4000, l '

Appendix 4004, Inverse Count Rate Ratio Monit.oring. After count rates. had stabilized, two counting trials of.100 seconds each were taken on all detectors. The counts ..were used to >!

calculate an Inverse Count Rate Ratio - (ICRR), which was then plotted versus the number of- fuel assemblies ~1oaded. The ICRR is used as an indicator of-the ap'proach to criticality and this plot ensured there was no unanticipated . approach to criticality. Appendix 4004 also provided for statistical verification of detector performance during' extended fuel load operation suspensions.

After the core was loaded, Appendix 4006,. Core- Map, was 3 performed to verify correct core _ loading. Reactor Engineering f

\ and QA performed a visual scan of all fuel-assemblies and inserts using~an underwater camera. Correct fuel assembly, and fuel assembly insert locations were verified. The core was 4 '

further verified to be free of debris. A permanent video record was also made.

1 RESULTS As stated previously, the initial fuel load began on 11-26-85~

at 1825 and was completed on 12-2-85 at 2310. The initial core loading sequence is shown in Figures 4.0-1 through 4.9-10. All five neutron monitoring channels responded as expected,- and there were no unexpected increases in subcritical multiplication.- Noise was intermittently observed.on'SR 31 and was determined to be from SIGMA machine movement and nearby welding activities. Inverse Count Ratio Response (ICRR) plots ~

for SR31 and 32 and TO A, B and C are.shown in Figures 4.0-11 and 4.0.12. .

n - - .-

Page 19

,m Due to a bow in an adjacent fuel assembly, assembly B49 could j not be loaded into core location E04 per the loading sequence. l

, The sequence was changed per the recommendation of Westinghouse l Fueling Services personnel to' leave ,E04 vacant and load around it. When E04 was " boxed in" by adjacent assemblies, B49 was successfully loaded into E04.

Throughout the entire loading operation, approximately 2 days were lo'st due to various probl, ems with the SFPBH and SIGMA machine. Problems with the SFPBH were mainly due to overload limit switches and spurious resetting of control setpoints.

Problems with the SIGMA were mainly: 1) The SIGMA machine did not realize when it was fully down; 2) The overload / underload trips were set too low /high, respectively; and 3) The east side motor and associated drive system were not functioning l properly. Corrective maintenance was performed in each case to ,

allow fueling operations to continue. No problems were y encountered with performing the Core Map.

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ui,,","$$stuon d 0-7 unit No. s STEPS 87 TO I188

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unit m. s STEPS 159 TO 193 1

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Assembly loaded in permanent position in previous step.

m-

-- Assembly loaded in temporary position in previous step.

-- l Q

1 g Assembly loaded into position during loading step number N.

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g Location of Tenporary Detector A (5 and C).

i

' M Assembly with primary source insert.

Not as yet loaded.

i Note: Arrows indicate detector or fuel movement.

3 i

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Figure u,,"%',"se sisuon INITI AL CORE LOADING SEQUENCE FIGURE LEGEND unit No. s

Source Range N-31 p.g. 30 1.1 w w wu

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t.

Page 33' o 4.1 INITIAL FUEL LOAD CHRONOLOGY ,,

DATE TIME EVENT 11-26-85 1600 All Initial Conditions for fuel load met - core loading instrument alignment checks performed.

1825 Primary source bearing assembly C04 loaded into core location L15.

Four fuel assemblies loaded.

2200 .

'11-27-85 0140 Operations personnel find bolt lying on control rod retainer plate in SIGMA mast. Fuel loading suspended.

0245 Bolt removed by ' Operations personnel SIGMA machin'e inspected - two empty bolt holes found on mounting plate above SIGMA mast.

0400 Visual scan of core and refueling cavity performed. No debris found.

0730 Fuel load recommenced.

g 1555 SIGMA machine inoperable. SFPBH inoperable.

2300 Begin count rate data acquisition to verify detector performance (anticipating delay in fuel loading of greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ).

11-28-85 0510 Recommenced fuel load.

1525 I&C personnel working on SIGMA.

11-29-85 0605 I&C personnel working on SIGMA.

1955 SIGMA no~w operable.

2128 Seventy-one fuel assemblies loaded.

11-29-85 2200 Assembly B49 could not be lowered into core location E04 -

adjacent ' assembly is bowed.

11-30 0240 Loading sequence modified to box in location E04 per Westinghouse recommendation.

0300 Fuel load recommenced. ~

12-01-85 0155 113 assemblies loaded. . ;

12-02-85 .0100 145 assemblies loaded.

'L 0729 157 assemblies loaded.

f _

2310 193 assemblies loaded -

fuel load complete.

12-04-85 2200 Core map complete.

1

- ._ . _ _ _ _ _ _ ~ . - _ . . _ . - _ ________________.____.4

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

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Page 34 O v.

5.0 POST CORE HOT FUNCTIONAL TESTE -

The major objectives ~ of this test were to' ensure all necessary p . plant systems were- operable, Operations personnel- were j familiarized with the .. integrated - operatio'n -of: the ' plant, the j RCS functioned properly with the core . installed and that .the initial. conditions for initial- criticality were met. The' test -

~

procedure .took tha plant. from a cold shutdown . condition to a' .

hot standby condition of 557*F and 2250 psia. Testing' was -

'l 1

conducted at'various predetermined temperature plateaus.

Major testing conducted during this milestone involved:

RCS loop RTD to incore thermocouple cross-calibration i Functional verification of the RCS leak detection computer program and surveillance procedure Proper operation of the rod control slave cycler and CROM _,

operation with rods attached was verified Rod drop times were measured under cold no-flow, cold

- full-flow, and hot full-flow conditions Proper pressurizer spray and heater operation was verified Proper operation of the flux mapping and rod position ,

indication systems was verified The RCS flow and RTD bypass flow were verified to be acceptable ,

RCS flow coastdown timing following a trip of a single RCP and the simultaneous trip of all four RCPs was measured and compared to the FSAR assumed values t

Extensive ope rational testing of " the CVCS system was

. conducted Proper operation of the RCS loop stop valve and RCP 7

interlocks wts verified L ,

Testing was conducted over the period from 12-13-85 to i~ 01-23-86.

VD

6.

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5.1 SHUTDOWN MARGIN

~

3-INT-5000, Appendix 5001

, j OBJECTIVE l The objective of-this test was to ensure that the core remains

~

subcritical and that the Technical ' Specification Shutdown Margin (SDM) requirements are met throughout Post Core Hot Functional (PCHF) testing. ;

DISCUSSION Based on information from the Westinghouse Nuclear Design Report, a RCS Boron concentration of > 1850 ppm was determined to maintain adequate SDM in Modes 3, 4, 5 regardless of rod position and 'RCS Tavg. The following data was' recorded at

24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months intervals during PCHF testing: RCS' boron concentration, pressurizer boron concentration, Tavg, reactor f- coolant pump status, residual heat removal system status and k[ control rod position.

RESULTS Adequate SDM was maintained throughout PCHF. RCS boron concentration was verified each day to be greater than 1850 ppa' '

(average = 2054 ppm). Pressurizer boron concentration was verified to be within i 50 ppe of the RCS while the RCS was in a cold condition. However, when the RCS heatup began, the ~

pressurizer boron samples became unreliable. Investigation revealed that the loop seal drain line for the pressurizer safety valves was connected to the pressurizer liquid sample line. With the RCS heated, condensate from the pressurizer vapor space . accumulated in the loop - seals and diluted the U pressurizer liquid samples. Plar.t deficiency DDR 996 covers this issue. ' While not - affec' ting the ability to operate the plant safely, this' situation represents an inconvenience. , {

Engineering is investigating possible solutions to the problem.

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Page 36 l

{N.g i j () 5.2 INCORE THERMOCOUPLE /RTD TESTING 3-INT-5000, Appendix 5002 -

l OBJECTIVE i The objectives of this test were to: l

1. Perform a functional check and obtain cross-calibration data for core exit thermocouples and reactor coolant RTDs.

2. Verify expected resistance versus temperature-characteristics of reactor coolant RTDs.

3. Verify expected millivolt versus temperature I characteristics for core exit thermocouples.

4. Verify temperature and pressure of the Inadequate Core Cooling System (ICCS) at each temperature plateau.

5. Obtain data for preparation of the RTD calibration-report.

DISCUSSION ,

The test was conducted on 01-15-86 and 01-16-86 during the

. heatup of the plant. Data was collected - from the incore thermocouples and RCS RTDs during four periods of constant RCS heatup instead of the traditional method where data is collected during four periods of isothermal RCS conditions.

The constant heatup rate method greatly increased testing flexibility and reduced the amount of time required for the test.

During each of the data collection periods, a constant rate of RCS heatup was achieved by first placing steam generator levels in the normal operating band with all generator levels approximately equal. Feedwater flow and blowdown were secured .

30 minutes prior to collecting the data. Data collection began when a constant heatup rate was achieved. Data was collected in the R'CS temperature bands of 355-365*F, 415-425*F, 480-490 F and 530-550 F.

1

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. Incore thermocouple temperature data was obtained by initiating

, a plant process computer printout at the beginning of the col.lection period. The incore. temperature data was from the

' Inadequate Core Cooling System (ICCS) . Data frum the RCS RTDs was' obtained, from the RTD inputs .to the Westinghouse 7300 -

process control system. Additional measurements of signal and compensating lead. resistances were made for the three-wire RCS wide range hot leg RTDs so. that the actual RTD resistance could be determined. After each RTD 'n the loop under test was measured, the procedure was repeated for the remaining loops.

Four sets of data from each loop were collected during each temperature band.

RCS wide range pressure was obtained from the ICCS computer via the plant process computer, and appeared on the printout of incore thermocouple temperatures. RCS narrow range pressure

. f. was obtained from the control room main control board indicators.

, RESULTS ,

i The incore thermocouple to RTD cross-calibration acceptance

criteria was achieved in that the incore thermocouple temperatures were within 2*F of each other, and within 2*F of the RTD cross-calibration resu*lts.. Tne acceptance criteria for RCS and ICCS pressure indication was also satisfactorily met in .i that the RCS wide range and narrow range pressures were within 40 psia of each other.

The RTD data was supplied to Westinghouse for evaluation and preparation of the RTD calibration report, a

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Page 38 A/

5. 3 : ROD CONTROL SLAVE CYCLER AND CRDM TIMING TEST .,

3-INT-5000, Appendix 5004 OBJECTIVE Under cold shutdown conditions, provide verification of proper slave cycler timing and Control Rod Drive Mechanism (CRDM) timing, and an operational check of each CRDM with a Rod Cluster Control Assembly (RCCA) attached.

DISCUSSION The test was performed from 12-15-85 to 12-27-85 under a Cold Shutdown (Mode 5) condition.

Proper slave cycler timing was verified by, in turn, selecting one rod from each rod control power cabinet and monitoring the CRDM lift coil, stationary coil, and moving coil currents, and hT the CRDM microphone output, while moving the rod from zero to 48 steps and then back to zero. All other rods in the group under test were prevented from moving by opening the appropriate lift coil _ disconnect switches. Proper slave cycler .

timing was verified by - comparing the CRDM coil current oscillograph traces with examples provided in the Westinghouse CRDM technical manual.

The operational check of each CRDM was accomplished by, in turn, withdrawing each shutdown and control bank to 48 steps, disabling all rods in the group except the one under test, and then alternately withdrawing and inserting the rod under test 10 steps while obtaining oscillograph traces of the lift, stationary, and moving coil ' currents. This process was repeated twice for each rod, and the resulting oscillograph traces were compared for timing to each other and to examples provided in Westinghouse CRDM Technical Manual.

v'O

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, jj;t It Figure 5.3 -1 shows .a typical oscillograph trace of lift,-

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' moving, and stationary coil currents during rod withdrawal f operation. Figure 5.3-2 shows the same during an insertion operation.

RESULTS Proper slave cycler timing and CRDM timing .were verified by- ,

comparing lift, moving, and stationary coil current oscillograph traces to examples provided in thel Westinghouse CROM Technical Manual. All comparisons indicated satisfactory equipment performance.

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A 5.4 - RCS LEAK RATE ~

3-INT-5000, Appendix 5006 OBJECTIVE This test performed two functions:

, -1) It reverified that the plant's computer. Leakage Calculation Program, SP 3J3, could detect a 1 gallon per ,

minute (GPM) UNIDENTIFIED LEAK from the Reactor Coolant ,

System (RCS) and connecte"d , portions of the Chemical and Volume Control System (CVCS).

2) In parallel, it validated the manual RCS Leak Test Surveillance Procedure, SP 3601F.6.

DISCUSSION This test was essentially a repeat of the Pre-Core Hot Functional Leak Rate Test, 3-INT-3000, Appendix 3030. ,

O This test was performed on 01-22-86 with the plant stable at normal operating temperature and pressure (557'F and I 2250 PSIA). The boundary of the test included the entire RCS, those i portions of the CVCS that delivered letdown to the Volume

Control Tank and returned it to the RCS, 'and to' the first isolation valve of all systems connected to the CVCS and the RCS. No changes were made to any of the valve lineups "

associated with the RCS or CVCS during the leak rate test. All

^

normal means of removing or adding water to the RCS and CVCS were secured and then a mass balance was performed using the change in pressurizer and primary tank levels. These volume changes were individually adjusted for any change in temperature over the test period; The test's initial conditions required extensive system lineup .

verifications. Once these were complete and the plant was

verified in a stable condition, a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months manual mass balance ;)

calculation was performed concurrently with both the computer j.

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[f'} program, SP 3J3, and -the surveillance procedure SP 3601F.6. ~

This' 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months test run was to obtain baseline information on the -

stable plant leak rate and to. document in Appendix 5006 that the plant met the Technical Specifications of no greater than-1.- -

e GPM . UNIDENTIFIED LEAKAGE ~ from ' the RCS (TS 3.4.6.2.b) and no !

. greater. than 10 GPM IDENTIFIED LEAKAGE- -from the RCS (TS 3.4.6.2.d).

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months mass balance portion of the test was successfully completed with the following data'being obtained:

1) IDENTIFIED LEAK RATE = 0.74 GPM

2) ' UNIDENTIFIED LEAK RATE = 0.73 GPM

. Upon completion of the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months test run, a 1 GPM known leak was I

induced off the low pressure section of the CVCS _ letdown line. -[

The failed fuel radiation monitor drainline was chosen for the

. source of the leak so as to allow the use of permanently installed flow detector -(3CHS-FI391) to monitor the induced leak.

After stabilizing the 1 GPM known leak (actual reading on -

3CHS-FI391 varied between 0.98 GPM and 1.17 GPM), a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months mass balance calculation was performed, again,- concurrently with both the Computer Program SP'3J3, and the Surveillance Procedure, SP 3601F.6. The data from-the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months test run and-

~

the change in relation to the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months test run was compared to 4 the following acceptance criteria for both the computer program

" and the surveillance procedure.

1) No greater than 10.GPM IDENTIFIED LEAKAGE from the RCS (TS 3.4.6.2.d). .

2) The change in the' UNIDENTIFIED LEAKAGE shall be one 1 GPM

[

~

19 percent (0.91 to 1.09 GPM).

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Page 44 N .

'U -The outputs of the ' leak rate tests were recorded as follows:

I l

SP 3J3 SP 3601F.6 <

IDENTIFIED 0.61 GPM 0.654 GPM LEAK RATE Change in UNIDENTIFIED 1.263 GPM 1.263 GPM LEAK RATE The leak rate change gave a conservative output since it actually indicated slightly more leakage than was present.

However, the change in UNIDENTIFIED LEAK RATE did not meet the acceptance criteria of 0.91 to 1.09 GPM. To document this, plant deficiency UNS 7495 was submitted, p' ' -

RESULTS Performance and evaluation of test results for the RCS Leakage Program, SP 3J3, showed genera 11y' satisfactory performance.

Although prograened-calculated leakage was higher than that for

the hand-calculation, identified in plant deficiency UNS 7495, this anomaly is explainable by a varying induced leakage flow (0.98 GPM to 1.17 GPM). The deficiency recommended to accept-as-is, in part, due to the conservative results_ of the test, i.e., indicating more leakage flow (1.263 GPM) than was actually present (acceptance criteria 0.91 to 1.09 GPM). The proposed disposition of UNS 7495 was approved by the Joint Test Group with the added requirement that it be sent to the Unit 3 Reactor Engineer for review. The subsequent review by the Reactor Engineer determined the installed leak detection program to be satisfactory.

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Page 45 j i

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b 5.5 PRESSURIZER HEATERS AND SPRAY TESTING ,

j j

3-INT-5000, Appendix 5007 )

OBJECTIVE The objectives of this test were to:

1. Establish optimum pressurizer spray valve bypass valve position in order to maintain the spray lines in a warmed j

condition (to minimize thermal shock on the lines when pressurizer spray is initiated) and at the same time l maintain bypass flow so that proportional heater output is kept at approximately 50 percent of rated capacity. Once the final position for the bypass valves have been set, the spray line low temperature alarms will be set. It should be noted that a preliminary setting of the bypass valves was completed during the precore hot functional test (3-INT-3000, Appendix 3011).

2. Verify pressurizer spray effectiveness is within design

(]

'" tolerances.

3. - Verify pressurizer . heater effectiveness is within design tolerances. q

4. Verify pressurizer heater capacity is within design ;

1 tolerances. j l

DISCUSSION l The test was conducted between 01-20-86 and 01-21-86 with the !

plant in a Hot Standby (Mode 3) condition.

The first objective was to be accomplished by recording pressurizer spray line temperatures while incrementally opening the spray valve bypass valve. This data would then be plotted

}

and the opt'imum position of the bypass valves selected. The ,

f optimum positions correspond to the point on the curves where spray line temperature flattens out. The spray valve bypass o valves would then be set to these optimum throttle positions

~ ' ~ ' ~ ~^ ~

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Page 46 ;

fT ,

M - and plant conditions ' maintained at steady _ state so that f equilibrium data could be'taken on the pressurizer spray lines.

The purpose of this data is to confirm that the spray line temperature is at > 540*F and the proportional heaters are at l approximately 50 percent of rated capacity. Adjustments to the J valve position would be made as required to achieve these l desired conditions. Once final bypass valve positions were '

established, spray line low temperature alarm s'etpoints' would

- be established and. reset if required. - These setpoints were .

required to' be > 530*F so as to conform to the. Westinghouse Precautions, Limitations and Setpoints (PLS) Document. ;

l The- second objective of the test was. accomplished by ,

establishing normal no-load operating teperature and pressure ,

l' in the RCS with the charging system flow controller in manual

, and all pressurizer backup and control heaters off. Once these

' conditions were established, both pressurizer spray valves were- 1

. fully opened and kept open until RCS pressure was reduced to approximately 2000 psia.

l The . third and fourth objectives were accomplished by l reestablishing normal no-load RCS . temperature, pressure and pressurizer level with both pressurizer Power Operated Relief '!

Valves (PORVs) in the closed position, the charging system flow l controller in manual and bot'h pressurizer spray valve

~

controllers in manual with the spray valves . closed. At that ;

point, all pressurizer - backup and control heaters were -

energized manually to full output and RCS pressure monitored until it reached approximately. 2300 psia. Once this pressure - )

was reached, all pressurizer heaters were returned to automatic >

~

as well as the charging system and pressurizer spray valve ,

I controllers. Concurrent with this transient, 3-phase voltages ;

and currents were taken on all pressurizer heater groups to l

i. verify that they were.within design specification.

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o Page 47 l

.'.. t

RESl;LTS I *

The setting of' the pressurizer spray valve ~ bypass valve

-positions could not be performed as initially proposed in the )

i

, ' test procedure due to excessive pressurizer spray valve seat leakage. A test change was written ' to first monitor proportional heater output and pressurizer spray line temperature with the spray valves shut (as indicated on the

main control board) and then secure instrument air to the '

' valves (the valves are fail-closed in design) td determine if

7. ,

the valves were being maintained partially opened due to improper control signals. Results of this test change indicated the valves were in fact fully closed. However, the- 1 g

seat leakage past these valves with the bypass valves open 1/16 ,

of a turn was such that the proportional heaters were operating at 100 percent 1of-rated capacity. As a next, step the bypass-e valves ~ were fully closed to ' determine if the leakage past the .3 4.

4 -

spray valves was sufficient to maintain the pressurizer spray '

. line temperature above the' low - temperature alarm setpoint of L,N 530*F. With the- bypass valves ~ shut, the flow . was not, i sufficient and the low temperature alarm was received. The I

bypass valves were then opened approximately 1/16 of a turn. ,

This resulted in spray'line temperatures of 539*F - for loop 1 ,

and 543*F for loop 2 with the proportional heaters operating at -1 approximately 80 percent capacity. A unit ~ deficiency, UNS 7485, was written to document the inability to generate the required spray line temperature versus - bypass valve position curves and the excessive proportional heater output. The deficiency was reviewed by Engineering and Westinghouse and dispositioned accept-as-is. The spray line temperature alarms setpoints were left at their initial settings of 530*F. This- :

was due to spray line equilibrium temperatures being approximately ~ .10*F higher than the setpoints and the '

requirement not to lower the setpoints below the 530*F >

Westinghouse Precautions, Limitations and Setpoints (PLS)

~

Document design value.

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' .The ' pressurizer spray effectiveness was successfully verified.

The verification was done with the ' plant at a no-load ,

temperature and - ' pressure with the charging system flow i

controller min ' manual . and all pressurizer heaters. turned off.

Initial. pressurizer level was ' 26 percent. - The . pressurizer j spray , valves ' were . then fully opened using the RCS master-

- .c pressura controller. The RCS pressure was : lowered from ' an ,

" initial value of 2240 psia.to the desired endpoint of 2000 psia in 114 seconds. While this time was slightly . slower than.the nominal response, it was well wit'hin design tolerances and test acceptance criteria.

The pressurizer heater effectiveness was successfully verified.

Two runs of the test were performed. These values were within the acceptance criteria. During both runs, the overall pressurizer heater capacity was below . design specification,.

being 1703.7 KW versus the design range of 1710-1890 KW. In O addition,- the group C proportional heater capacity . was -

393.99 KW versus ;the design range of 394.25-435.75 KW; the I

group D heater capacity was 324.3 KW versus the design range of 328.7-363.3 KW; and the group E heater capacity was 325.5 KW '!

versus the design range of 328.7-363.3 KW. Pressurizer heaters. .

groups A 'and .B (which are powered off vital-buses) had '

~

capacities of 329.9 KW and 330.0 KW, respectively. These values were within the 328.7-363.3 KW acceptance criteria. Therefore, i all Technical Specification requirements were met. >

Plant deficiencies UNS 7489 and UNS 7496 were initiated to document the discrepancies in heater capacities. Both !

deficiencies were reviewed by Engineering and Westinghouse' and dispositioned accept-as-is.

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-Page 49 j-]. .

5.6 ROD DROP TESTING

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3-INT-5000, Appendix 5008 OBJECTIVE The objectives of the test were to:

1. Determine the drop time of each control rod with the Reactor Coolant System in a cold condition. The drop times were measured . at no-flow and again, at full-flow.

2. Determine the drop time of each control rod with the Reactor Coolant System at ' normal operating temperature.

The drop times were measured at full-flow conditions.

Any rods having a drop time exceeding the acceptance criteria were required to be dropped 10 additional times. In addition, any rods having a drop time exceeding the average drop time for all rods by more than the two (two standard deviations) sigma limit were dropped three additional times.

DISCUSSION ,

The test was performed between 12-19-85 and 1-20-86 during Cold Shutdown (Mode 5) and Hot Standby (Mode 3) conditions. During the test, the drop time of each control rod was measured under cold no-flow, cold fall-flow, and hot full-flow conditions. l The acceptable rod drop time in each case was less than 2.2 seconds from the beginning of the decay of the stationary )

gripper coil' to dashpot entry. Any rods which failed _the 2.2 l second acceptance criteria were required to be dropped ten additional times and any rods with drop times outside the two sigma limits were dropped three additional times.

I Rod drop times were determined by simultaneously dropping all !

rods in a group from a fully withdrawn position (228 steps). I Data from the group under test was collected using a computer ,

based data acquisition and analysis system developed by 1 Westinghouse exclusively for rod drop testing. Drop data for

, )

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Page 50 g .- -l i.)9"

. the group under test was- collected from the Digital Rod Position Indication -(DRPI) system. Testing progressed through s each group in sequence until all rods had been dropped.

4 Once all data had been collected, it was analyzed to determine ,

the drop and turnaround time of each rod, and the mean and two 1

'igma limits. Hardcopy drop traces for each rod were provided s ,

as well as summary tables listing individual rod. drop times and indicating those rods falling outside the two sigma limits. ,

Figure 5.6-1 provides a typical rod drop trace. Table 5.6-1 summarizes the rod drop times for cold no-flow, cold full-flow, and hot full-flow conditions.

i.

During the cold full-flow portion of the test, rods K14, J03, H06, and H10 exceeded the two sigma limits and were each -*

three additional times.

dropped As a result of these

. additional drops, K14 remained outside the two sigmas limit but y varied only 10 msec from the initial drop. J03. was within the ;

two signa limit on two of . the three additional drops; H06 '

remained outside the two sigma limits, but within 2 msec of the' initial drop time; and H10 was within the two sigma limits 'on two of the three additional drops. The additional drop data was reviewed and determined acceptable.

D'uring the cold, full-flow portion of the test, rods H06 and ~

F08 were determined to'be outside the two sigma limits and were each dropped three additional times. The supplemental drop.

times were within the two sigma limits. )

o During the hot, full-flow rod drop data, rods B04, M02, and LOS were determined to be outside the two sigma limits and were each dropped three additional times. The additional drops of -

M02 and LOS were within the two sigma limits so that only B04 remained outside the limit. This was reviewed and determined acceptable.

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b RESULTS' All rod drop times under cold no-flow, cold fu11-flow, and hot full-flow conditions were less than the 2.2 second acceptance

criteria.- The performance of the rods was demonstrated to be acc'eptable.

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Page 52 ROD DROP TIME TO DASHPOT ENTRY (msec)

ROD CORE COLD COLD HOT g POSITION NO FLOW FULL FLOW FULL FLOW SBA D02 1302 1500 1412 B12 1294 1492 1402

~

M14 1296 1506 1416 PO4 1294 1508 1422 H04 1288 1514 1404 B04 1298 1492 1274 D14 1308 1488 1398 P12 1298 1496 1394 M02 1298 1494 1418 H12 1304 1508 1408 SSS 003 1300 1486 1396 C09 1312 1498 1400 J13 1290 1480 1376 N07 1308 1500 1416 D08 1290 1496 1410 C07 1310 1492 1400 013 1290 1504 1398 N09 1300 1498 1402 003 1320 1494 1398 M08 1308 1494 1406 SBC E03 1298 1492 1398 C11 1294 1512 1396 L13 1294 1502 1388 NOS 1298 1512 1398 O

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Y ROD DROP TIME TO DASHPOT ENTRY (msec) l (Continued)

ROD CORE COLD COLD HOT DAME. POSITION NO FLOW FULL FLOW FULL FLOW SBD C05 1300 1476- 1394 E13 1302 1496 1402 N11 1302 - 1498 1402 LO3 1288 1494 1392 SBE A07 1296 1486 1398 016 1294 1494 1402 R09 1294 1498 1396 J01 1298 1480 1406 CBA H06 1320 1558 1406 F08 1304 1542 1386 H10 1284 1488 1364 K08 1300 1494 1408 EOS 1296 1510 1392 E11 1306 1508 1410 L11 1298 1498 1400 j LOS 1300 1498 1366 )

CBB F02 1304 1504 1420 B10 1302 1490 1418 )

K14 1326 1506 1422 P06 1308 1500 1410 B06 1294 1480 1400 F14 1290 1482 1398 P10 1296 1496 1408 K02 1288 1482 1386 0 .

wuciEIow".c st uon ROD DROP TIMES j*1a, unnwo.s ROD DROP TESTING 99 2

i Page 54 fl v

ROD DROP TIME TO DASHPOT ENTi1Y (msec) l (Continued)

R00 CORE COLD COLD HOT BANK POSITION NO FLOW FULL FLOW FULL FLOW CBC H02 1286 1488 1420 808 1296 1498 1400 )

H14 1302 1498 1398 I P08 1304 1519 1396 F06 1310 1490 1402 s F10 1302 1496 1404 K10 1294 1492 1392 K06 1298 1492 1402 CBD D04 1296 1480 1392

( M12 1292 1482 1406

( D12 1294 1488 1382 M04 1290 1498 1402

. H08 1296 1.508 1398 j l

I MEAN DROP TIME 1299 1497 1399 l

MEAN MINUS 2 Sl0MA 1283 1471 1361 MEAN PLUS 2 SIGMA 1315 1523 1437 RODS OUTSIDE LIMITS K14 H06 B04 J03 F08 H06 ;

H10 Acceptance Criterie Rod Drop Time < 2200 msec '

Nuclesr o or Stauon ROD DROP TlMES {j,

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PLRNT NAME: MILLSTONE III TEST OPERATOR: D SIPPLE REACTOR OPERATOR: P LRNG TEMPERATURE: 553 DEG F PRESSURE: 2250 PSIG FLOW RATE: 100 % FLOW DATE: 01/'.7/88 TI ME.: 01: 32 .

ROD #: H12 DROP #: 3

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5.7 PRECRITICAL REACTOR COOLANT SYSTEM FLOW MEASUREMENT

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3-INT-5000, Appendix 5009 .

OBJECTIVE The objective of this test was to obtain the data necessary to

, relate reactor coolant system (RCS) installed elbow tap differential pressure (0/P) to RCS flow and to determine RCS flow.

DISCUSSION The test was conducted on 01-18-86 with the reactor plant at steady-state . conditions, temperature at approximately 557*F, pressure at approximately 2250 psia and four reactor coolant pumps running. The test consisted of collecting voltage data from the RCS flow elbow tap transmitters. From this data, the RCS flow was numerically determined. Acceptance criteria g required that each loop flow be at least 90 percent of the FSAR Q design value of 94,600 gpm (85,140 gpm) and the total RCS flow to be at least 90 percent of the FSAR design value of 378,400 gpm (340,560 gpm).

i RESULTS :

All data was successfully obtained with the exception of ;

RCS-F436 and RCS-F446 on RCS loops 3 and 4, respectively.

These transmitters read abnormally low. Plant deficiency UNS 7466 was issued to document this problem. Upon evaluation, it was decided the data on the two good transmitters on each of loops 3 and 4 along with the data from loops 1 and 2 would be adequate for RCS flow determination. The RCS flow which was calculated met all acceptance criteria and is summarized on Table 5.7-1.

Subsequent to the test, corrective maintenance was performed on j the two transmitters which were the subject of UNS 7466. 1

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Subsequent performance of the units has been satisfactory.

Page 57 i A(J.

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Looo 1 RCS-F414 102,087 OPM RCS-F415 103,679 OPM RCS-F416 102,359 OPM Loop Average 102,708 OPM Looo2 .

RCS-F424 102,220 GPM RCS-F425 103,520 09M RCS-F426 101,560 GPM LoopAverage 102,433 OPM Looo 3 RCS-F434 102,806 OPM

> RCS-F O 5 104,918 OPM x RCS-F436 see text ;

Loop Aver' age 103,862 GPM Looo 4 RCS-F444 101,462 OPM

- RCS-F445 104,681 GPM RCS-F446 see text l

Loop Average 103,072 GPM Total Calculated Core Flow: 412,075 GPM Acceptance Criteria:

I Calculated Loop Flow 2. 85,140 GPM Calculated Core Flow >.340,560 GPM

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~'- 5.8 RTD BYPASS FLOW VERIFICATION !

a INT-5000, Appendix'5010 OBJECTIVE The objectives of this test were to:

1. Measure the . flow rate in each RTD bypass loop to verify acceptable bypass loop coolant transport time.

2. Establish the alarm . setpoints for the RTP bypass flow alarm in the control room.

DISCUSSION Prior to performing the test, the RTD bypass line as-built measurements. were obtained. Based on these measurements, the minimum flow rates to obtain a 1-second bypass loop transport time were calculated, l

The test was performed over the period 01-21-86 to 01-23-86.

With all four reactor coolant pumps in operation and the RCS at

'\

hot zero power, no-load condition, the RTD bypass loop flow measurements were taken. The measurements were obtained by '!

recording total RTD bypass flow in each loop with the manifold isolation valves open. After the total flows were obtained, the hot leg RTD bypass manifold isolation valves were closed and the cold leg flow was recorded. The process was then l

reversed in order to record hot leg flow. The individual hot leg and cold leg bypass loop flows were then compared to the minimum acceptable flow established based on bypass loop configuration. Then, using the total measured flow values for each loop, the RTD bypass loop lo flow alarm setpoints were established at 90 percent of the total flow in each RTD bypass flow manifold.

RESULTS .

All acceptance criteria were met. The results of the flow measurements are presented in Table 5.8-1.

Page 59 O

. sd CALC TOTAL- MINIMUM ~ MEASURED ALARM VOLUME- FLOWRATE FLOWRATE SETPOINT LOOP _ (FT3) (6PM) (GPM) (6PM)

Hot Leg 1 0.216 105.9 117 N/A 0.115 N/A Cold Leg 1 51.'6 155 Total Loop 1 N/A N/A 266 239.4 4

Hot Leg 2 0.242 108.6 118 N/A Cold Leg 2 0.111 49.8 160 N/A x' Total Loop 2 N/A N/A 265 238.5 Hot Leg 3 0.230 103.2 115 N/A Cold Leg 3 0.117 52.5 150 N/A Total Loop 3 N/A N/A 263 236.7 Hot Leg 4 0.235 105.5 118 N/A 1 Cold Leg 4 0.097 43.5 158 N/A

, Tota 1 Loop 4 N/A N/A 269 242.1 i

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5.9 ' MOVABLE INCORE DETECTORS i

INT-5000, Appendix 5011 OBJECTIVE The objective of this test was to demonstrate the operability 'I of the movable incore detector system (flux mapping) by: ,

1. Demonstrating proper system performance in manual and ,

automatic modes of operation.

2. Determining actual detector path lengths.

3. Verifying all detector thiables free of obstructions.

4. Installation of permanent system detectors.

DISCUSSION The test was performed on an intermittent basis over the period. ;

of 12-12-85 through 01-02-86. Proper system operation was ,

verified with dummy incore detectors-installed. All operations-were performed from the flux mapping console located in the

\ control room. In addition,, detector path lengths were measured I

\

in . order to provide alignment data for the automatic flux ,

mapping control system. Once these steps were performed, the actual detectors were installed and proper system operation, 'l including performance of an automatic full core flux map, was verified. .

Although the majority of the test was performed with the plant in a cold condition, a full core map was taken - under hot -;

standby conditions to ensure the detector paths were . free of obstructions and binding would not occur. During this ,

operation, the data link between the flux mapping system and the plant process computer was verified.

i RESULTS .

The test was performed satisfactorily with no deviation from test acceptance criteria. All thimbles were satisfactorily accessed with both the dummy and permanent detectors. No 9

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[~ evidence of binding was experienced. Some minor equipment

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problems were encountered, but these were readiiy resolved and operation of control circuitry and indicators was satisfactory.

A problem was encountered when the path lengths determined using the dummy detectors were utilized with the permanent core assemblies. Normal manufacturing tolerances associated with

-! the drive cables results in each cable being inserted a 1 slightly different length for . each revolution of the drive wheel. By performing a path length measurement for several paths using the permanent detectors, a correction factor was derived to allow using the original path length data withoat repeating every path length measurement following the installation or replacement of detector core assemblies.

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5.10 DIGITAL R00 POSITION INDICATION 3-INT-5000, Appendix 5015 -

OBJECTIVE To verify that the Digital Rod Position Indication (DRPI)

~

satisfactorily provides the required indication for each individual rod, under Hot Shutdown conditions (Mode 3).

DISCUSSION The test was performed over the period from 01-17-86 to 01-21-86. Each bank of shutdown and control rods was individually withdrawn in 24 step increments to 228 steps. At each 24 step increment, the DRPI on the main control board was l compared to the group step counter and plant computer. The DRPI display was required to be within 12 steps of the group step counter and the plant computer. In addition, the control group step counters were required to be within one step of the Q rod control pulse-to-analog converter output at every 24 step

\ increment.

l Each bank was then inserted to within 6 steps of the bottom and I jogged to the zero position. The rod bottom indicators were required to actuate at zero steps on the group step counters.

The DRPI main control board display and group step counters were continuously monitored during rod withdrawal and insertion for any indications of improper rod motion.

l Initially the plant computer was ' not providing rod positions due to a software problem in the program that processed the data from DRPI. This was corrected and the test was completed satisfactorily. .

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' Page 63 RESULTS The DRPI system provided indications of -rod : position that -

agreed - with the group step counters and ' plant computer. No indications' of improper rod motion were observed. Rod bottom ,,

indication occurred at zero steps. Control bank group step counters ' agreed within one step with the. rod control pulse-to-analog converter. -

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i g D 5.11 LOOSE PARTS MONITORING 3-INT-5000, Appendix 5016 -

OBJECTIVE The objectives of this test were to: -

1. Obtain baseline system signal data during the reactor plant heatup.

2. Obtain baseline system signal data with the plant at normal operating temperature and pressure.

3. Determine the approximate frequency of spurious alarms.

DISCUSSION The majority of testing was performed from 01-13-86 to 01-18-86 during the plant heatup at RCS temperatures of 250*F, 350*F, 420*F and 557*F. Testing was completed on 01-20-86.

Baseline signal data was obtained by using a spectrum analyzer which was connected to the auxiliary output jack on the Loose

(. Parts Monitoring system (LPM) cabinet. Hardcopy spectrum .

analysis data was obtained for all 8 monitoring channels during the various heatup temperature' plateaus. Additional data was taken at normal operating temperature and pressure by, in sequence, stopping a single reactor coolant pump and monitoring LPM response. The frequency of spurious alarms caused by noise of normal plent operation was also monitored.

The LPH was supplied by Rockwell and consists of a monitoring cabinet with audio output system and integral cassette recorder. There are eight accelerometers located on the primary system: 2 located on the reactor vessel head, 2 located on the lower reactor vessel and one on each steam generator in the channel head area. Prior to test performance, the system was modified by the addition of a 1500 hertz bandpass filter to enhance the capabilities to detect loose parts of a large mass (30 pounds).

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. Page 65 l 1

' :7 j RESULTS All baseline LPM signal data was obtained with no problems encountered. Refer to Figure 5.11-1 for an example of ' a typical spectrum analyzer output. However, during the test, an I excessive number of alarms were received from the' lower reactor vessel channels. The accelerometers for these channels are mounted on the incore detector guide tubes just- below the

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bottom of the reactor vessel. Further investigations ' indicated j L the alarms were being caused by the noise generated by the l incore detector thimbles which were rattling in the guide tubes due to RCS flow. Based on engirieering ' analysis, gain adjustments on the system's 1500 hertz filter were recommended on the affected channels. ,

Refer to Section 8.5.11 for a discussion of LPM testing conducted during the power ascension program.

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5.12 REACTOR COOLANT SYSTEM FLOW COASTDOWN 3-INT-5000, Appendix 5017 -

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OBJECTIVE-

. The objectives of this tes't were to:

1. . -Verify for a trip of one Reactor Coolant. Pump (RCP) with the other three pumps in operation that the low flow time delay is less than 2.5 seconds. ;

- 2. Verify for a ~ trip of one RCP . that . all points on the

~

faulted loop' flow coastdown curve are above the ,

, corresponding points on the predicted curve assumed'in the FSAR.

3. Verify for a trip of one RCP that all' points on the total core flow coastdown curve are above the corresponding points on the predicted curve as assumed in the FSAR.-

4. Verify the Reactor Coolant System (RCS) low flow reactor ,

trip response time is less than the value assumed in the FSAR for the case of four RCPs coasting down.

' \ 5. Verify that all points on the total core flow coastdown curve are above the corresponding points on the predicted curve in the FSAR.

DISCUSSION, j

1. One Loop Coasting Down Strip chart recorders were connected to the process rack cards containing the elbow tap d/p transmitter output for all four RCS loops, RCP breaker position, and reactor trip !

breaker position. A data logger. was connected to the .

process rack cards containing the signals for all three - i low flow bistables on the RCS loop (loop 1) to be tripped. ;

4 ,

Once the recorders were connected, the P-8 permissive was simulated (indicating reactor power above 37.5 percent) by .

jumpering a relay in the SSPS cabinets. With the P-8 permissive present, a reactor trip occurs by tripping one

-( RCP.

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The RCP in loop L1 was manually tripped from the control-room to initiate the test. The traces, dat:a logger output I

and plant process computer sequence of events output were then analyzed to measure the trip delay time and to createE the RCS flow coastdown curves for comparison to the FSAR curves.

2. Four Loops Coasting D.own During this portion of the test, the strip chart recorders were again connected to ali four RCS loop elbow tap d/p transmitter outputs and the contacts to monitor reactor trip breaker position and RCP breaker position. The data

< logger was connected to all twelve RCS low flow bistables.

As before, the P-8 permissive was simulated.

The test was initiated by simultaneously tripping'all four RCPs via a common RCP trip switch installed for the test. .;

The traces, data logger output' and plant process computer :

sequence of events data were again analyzed to determine the RCS-loop low flow reactor' trip response time and.'the.

~

total. RCS flow. coastdown rate for comparison to the FSAR Curves.

l RESULTS

1. One Loop Coasting Down I The low flow response time for the one loop coating down . .j case was 0.88 seconds which was less than the acceptance l

criteria of 1.00 second. A break down of the results is as follows:

Time from when the measured loop flow had decreased to the low flow trip setpoint until the last reactor trip breaker .

1 had changed position: ;

0.43 seconds (from sequence of events data) i t ..

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I Sensor. delay time

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0.40 seconds r

Gripper. delay time:

0.05 seconds

- Total:

0.88 seconds

' Acceptance Criteria: 1 1 second A secondary acceptance was that the time from the' reactor coolant pump breaker opening to the time that the rods o- were free to fall be' less than 2.5 seconds. Actual test ,

results are:

Time from the Reactor Coolant Pump Breaker opening to the Reactor Trip breaker opening:

1.8 seconds Gripper Response time:

.05 seconds f

Total: .

1.85 seconds Acceptance Criteria: 112.5 seconds.

In addition to the response time, the total core flow was l compared to the flow assumed in the FSAR following'a pump .

trip. As shown in' Figure ' 5.12-1, the total core flow remained above'the FSAR assumed value. ]

2. Four. Loops Coasting Down ;

The acceptance criteria for the four loops coasting down test was that the time from when the loop flow had decreased to the low flow trip setpoint until the control rods were free to fall shall be < 1.00 second when- .

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. considering the worst possible case. The results were: :

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Time from when,the measured loop flow has decreased to_the low flow trip setpoint until when the last Reactor Trip I Breaker has changed state:

Loop 1 Loop 2 Loop 3 Loop 4 0.327 0.317 0.287 0.232 seconds O.367 0.327 0.252 0.332 seconds 0.397 0.327 0.252 0.262 seconds Maximum T2 = 0.397 seconds Sensor delay times:

Loop 1 Loop 2 Loop 3 Loop 4 0.271 0.593 0.435 0.450 seconds ,

0.346 0.515 0.495 0.354 seconds 0.321 0.609 0.454 0.373 seconds Maximum Td = 0.609 seconds p

Gripper Release Time T = 0.05 seconds

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Low Flow Trip Time Delay (T1+Td+Th g TLF = 1.056 seconds-1 Acceptance Criteria TAC = 1.00 seconds

'As the test was originally written, the worst case value from each of the measurements was added to the worst case - a sensor time delay, and then to the gripper coil release time to determine the overall response time. This process yielded a result of 1.056 seconds which exceeded the test acceptance criteria of 1 second. After discussions with Westinghouse, a different analysis technique was used in determining the response times. This method involved .

i calculating the_ response times on a loop by loop / sensor' by sensor method rather than on a worst case basis. The new ld,v results are as follows:

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- Loop 1 Loop 2 Loop 3' Loop 4 I' O.648 - 0.960 0.772 0.732 seconds 0.772 0.892 0.797 0.736 seconds O.759 0.986 0.756 0.685 seconds All values were below the acceptance criteria of 1.00 seconds. j 1

The second acceptance criteria for the four loops coasting down test was that the total normalized core. flow for the ten seconds of the test was to be greater than the value Initial review of the test results

~

assumed in the FSAR.

indicated that the acceptance criteria was not met.

However, prior to performance of this test, Westinghouse 1

had performed a reanalysis of the RCS loss of flow ' l l

accidents. Based upon the new FSAR curves which had been generated by Westinghouse, the acceptance had been met. l

. These test results can be seen in Figure 5.12-2.

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5. 13 . ROD CONTROL OPERATIONAL TESTING

' 3-INT-5000, Appendix 5018 -

r OBJECTIVE ,

To demonstrate and document, prior to initial criticality, that l the a rod control system satisfactorily performs the required control and indication functions. ,

i DISCUSSION The test . was performed on 01-20-86 with the plant in . Hot .l' Shutdown (' Mode 3). . Prior to the start of the test, the rod speed control was adjusted to permit maximum rod speed, and the bank overlap setpoints were adjusted to permit the verification :j of proper operation with minimum rod motion.

.The test began by withdrawing each shutdown and control bank, !

in turn, to 48 steps while comparing Digital Rod Position -

Indication (DRPI), group step counters and rod motion lights to , J

k. verify that-. all rods in the bank under test were .being ;

withdrawn. Each bank was then inserted, again verifying proper ^

i rod motion on the DRPI, group step counters, and rod motion f lights. .l l

After verifying the rod control system could reliably control bank positions, the control bank overlap feature, control bank 0 full rod withdrawal limit (C-11 interlock) and rod bottom alarms and annunciators were verified. As a prerequisite, all shutdown banks were withdrawn to 30 steps to provide a large source of negative reactivity that could rapidly be inserted, i if required. Then control banks A, B, C and D were withdrawn in manual . control, while verifying that each bank. began motion and ceased motion in accordance with the bank overlap settings.

in the rod control logic cabinet. During this process, all control banks were stopped at 30 steps. Banks A, B, and C were stopped automatically by bank overlap settings, and 0 by manual i

Page'75 (s - operator. control. ' With all control banks now at. 30. steps, the rod control pulse-to-analog converter was. advanced to 220 steps using the test pushbutton in the logic cabinet. Manual control bank withdrawal of the D bank was then resumed, . and proper operation of the control bank 0 full ~ rod withdrawal limit (C-11 interlock) was verified by observing that bank' D withdrawal t halted at 223 steps on~ the pulse-to-analog converter and' that

. this action was properly annunciated on the main control board.

At that point Bank D _ was then returned to 30 steps and the pulse-to-analog converter was decremented using the t'est pushbutton in the logic cabinet, while verifying that the C-11 ;

interlock annunciator cleared.

Next, the "one rod bottom" and "two ' rod bottom" annunciators were tested by opening the control rod drive mechanism lift c coil disconnect switches for all but one rod in shutdown bank 1 E, and inserting the bank E rod in manual. When the single i

operable rod in shutdown bank E reached zero steps, the "one rod botton" annunciator was observed to energize. A second rod in shutdown bank E was then enabled by shutting its lift coil disconnect switch and ma.nually inserting this rod. When the second rod reached zero steps, the " two rods bottom" annunciator was observed to energize. At this point, the two shutdown bank E rods were returned to 30 steps and lift coil disconnect switches for all shutdown bank E rods were shut, restoring the rods to service. A With all shutdown and control rods at 30 steps, manual control' {

was again selected and control banks A, B, C and D were inserted while verifying proper bank overlap. The . shutdowr ;

banks were then restored to zero steps.

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' Page 76

- Restoration included returning the rod control logic cabinet

}

bank overlap settings, shutdown banks C, 0 an E tod speeds, and

[

l. process control system shutdown and control bank speeds to l- their normal settings.

l RESULTS Proper operation of control and shutdown banks, and proper control bank overlap was demonstrated. Operation of the-control bank D full rod withdrawal limit, and rod bottom alarms and annunciators were verified.

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Page 77 ]

, N . 5.14 CHEMICAL AND VOLUME CONTROL SYSTEM 3-INT-5000, Appendix 5031 -

1 OBJECTIVE

.The objectives of this test were to:

1. Verify the ability of the chemical and volume control system to perform boration and dilution of the reactor coolant system.

2. Verify the~ hot functional degasification capability of the letdown system using the degasification portion of the' radioactive gaseous waste (GWS) system. ,

1 f.

DISCUSSION The test was performed over the period of 01-18-86 to 01-22-86.

Testint; consist.ed of a series of operational verifications of

?

the Chemical Volume. and Control System (CVCS) to operate as intended and meet the limits of the acceptance criteria listed 1 below. All system operations were controlled from the control

~

-\~ room. Test data was obtained 'from permanent plant instrumentation, augmented as required with local te:st

[ '

instrumentation. .l The acceptance criteria for the test can be summarized as follows: !

1. The GWS degasifier operates within design limits for feed pressure inlet temperature, operating pressure, level and return flow temperature. !

2. The Charging System (CHS) is capable of increasing or !

decreasing RCS boron concentration by 100 1 10 ppm within l one hour. ,

3. The letdown system operates within design limits for flow -

rates,. temperature and filter differential pressure across various system filters. This also served to verify proper i

. sizing of letdown system flow restricting orifice.

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4. The hydrogen regulator is -capable of maintaining pressure

p T on the CVCS Volume Control Tank (VCT[ within design limits.

5. The boric acid and makeup flow controllers are capable of maintaining flow within design limits.

RESULTS Test acceptance criteria were met with .the following exceptions:

1. GWS degasifier feed pressure controller did not operate within design limits. Plant deficiencies, .UNS 7477 and !

UNS 7478, document this problem. Corrective maintenance was performed on the controllers with satisfactory retests.

2. Once testing began, the VCT high temperature alarm setpoint was determined to be too low. Plant deficiency DDR 815 documents this problem. The setpoint was revised

- and the alarm recalibrated satisfactorily.

s

3. Differential pressure across various letdown filters-exceeded acceptance criteria. Plan,t . deficiencies, UNS 7472 and UNS 7473, document this problem. Based on review of each specific situation, the filter (s) were either replaced or determined to be acceptable as installed.

4. The degasifier outlet conductivity cell provided readings which exceeded the actual conductivity of the outlet flow.

Plant deficiency UNS 7476 documents this problem. The conductivity cell was determined.to be defective. A replacement unit was installed and . satisfactorily retested.

5. The manual miikeup to the VCT could not be controlled 'in accordance with system design. Plant. deficiency UNS 7484 .

documents this problem. Corrective maintenance and recalibration of the controllers was performed. -The system was satisfactorily retested.

~

1 %

, . Page 79

.a -

4 .

6.- During letdown' flow orifice verification, the letdown flowrate through 3CHS*FCV121 exceeded th( nominal design limit by approximately 20- percent. Plant deficiency.

UNS 7488 documents this problem. The actual flowrate was reviewed by Engineering and determined acceptable. ,

7. .During testing, the design' VCT -hydrogen concentration could not be obtained. Plant: deficiency UNS 7491 documents this problem. Further purging of the VCT with

- hydrogen achieved an ! acceptable hydrogen ' concentration.

> Thel deficiency was closed ba' sed on this action.

8. The desired RCS. dilution ' rate of 100 pps/hr was '- not achieved during the test. Plant deficiency UNS 7490 ,

documents this. problem. Further investigation revealed a system lineup problem. This was corrected - and a satisfactory dilution rate verified by retest.

p In addition, as noted under Section 5.1, Shutdown Margin, it j was not possible to obtain accurate pressurizer boron samples once the plant was hot. This was because the loop seal drain

- line for the pressurizer safety valves is connected to the I pressurizer sample line. With the .RCS heated, condensate from d the pressurizer vapor space that had accumulated'in-the. loop seals diluted the pressurizer liquid samples. Plant deficiency DDR 996 covers' this issue, 'and is- currently under evaluation.

l 1

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Page 80 l I

7.s 5.15 REACTOR COOLANT SYSTEM LOOP STOP VALVE AND .. PUMP INTERLOCKS.

, '3-INT-5000, Appendix 5033 OBJECTIVE The objectives of this test were to verify: ]

1. RCS loop stop valves and bypass valves are capable of l being operated only when the appropriate RCS temperature and valve position criteria are satisfied.- ;

2. Remote valve position in ,the control room corresponds satisfactorily to actual valve position.

3. Opening and closing stroke times for the RCS loop stop valves are 1 210 seconds.

4. Opening and closing stroke times for the RCS loop bypass valves are 1 40 seconds.

5. RCPs can be operated when the oil lift pump pressure criteria (< 600 psig) and loop stop valve position criteria (stop valves open) are met.

y 6. RCP breaker will trip if locked rotor signal is present or if- the associated loop stop/ bypass valves are in an unacceptable position.

DISCUSSION The test was performed over the period of 12-28-85 through 01-03-86 with the reactor in a Cold Shutdown (Mode 5) condition. All system manipulations were performed from the <

control room. Where possible, personnel were positioned to observe equipment operation.

RESULTS The acceptance criteria were met' with the following exceptions:

1. A pressure switch on the D RCP oil lift pump did not function properly. Plant dyiciency UNS 7420 was written to document the problem. korrective maintenance was performed and the component was satisfactorily retested.

{qf M

Page 81 9 2. The closed loop stop valve annunciators on the B and D loop did not function properly. Plant deficiency UNS 7381 was written to document the problem. Corrective maintenance was performed and the components retested satisfactorily.

3. Several loop stop and bypass valves exceed the stated stroke times. No valve exceeded the acceptance criteria by greater than 5 percent. Plant deficiency UNS 7417 was written to document the problem. The stroke times were evaluated by engineering and determined to be accept-as-is.

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o Page 82 y

h..[a 6.0 INITIAL CRITICALITY OBJECTIVE The objective of this testing was to-ensure that criticality was achieved in a safe and controlled manner and to verify that the critical boron concentration was within 1 percent AK/K of the Westinghouse' Nuclear Design Report predicted value.

DISCUSSION Testing was conducted on 01-23-86. Two procedures were used; the 3-INT-6000 base procedure covered the majority of testing and . Appendix _6001 to the base- procedure controlled the collection and analysis of Inverse Count Rate Ratio (ICRR) -

data. A summary chronology is provided in Section 6.1.

Prior to starting the approach to initial criticality, a verification of. all Mode 2 Technical Specification requirements was performed. In addition, the startup related surveillances were performed' on the Source Range (SR) and Intermediate Range (IR) nuclear instrumentation. Baseline count rates were determined and RCS samples were taken for determination of boron concentration. Initial RCS boron concentration 'was measured at 1870 ppe. The approach to criticality was begun at 1410 on 01-23-86. The shutdown- and control banks were withdrawn, observing proper sequence and overlap in 114 step increments, until control bank D.was at 160 steps. ICRR data

! was taken after each rod pull and plotted. When control bank D was at 160 steps, rod bank withdrawal was stopped and a new set -

of baseline data was taken. The reactor coolant system dilution was then . begun at a . rate of approximately 80 gpm.

i .During this procedure, boron samples were taken at 30 minute intervals and ICRR data was taken every 15 minutes. One hour ,

.and forty-five minutes after the dilution was started, the dilution rate was reduced to 30 gpm. Ten minutes later the ICRR indicated .2 and the dilution was stopped. The RCS and

_a ..i-m , . . . _ _ _. _ . _ . . - . . _ - _ _ _ _ _

Page 83 0 ,

CVCS were allowed to mix until criticality was achieved. The reactor was declared critical 20 minutes after the dilution was stopped at 2200 on 01-23-86. ICRR data for rod withdrawal and dilution to criticality is shown on Figures 6.0-1 through 6.0-4.

RESULTS The initial criticality test results are as follows:

Me'asured Predicted Control Bank D Position 160 steps 160 steps RCS Boron Concentration 1591 ppm 1559 ppm T 557*F 557*F ava The acceptance criteria of 1%aK/K was met although the RCS boron concentration was slightly above the predicted value.

This was due to boron mixing that was still occurring in the RCS and CVCS and due to increased Volume Control Tank (VCT) makeup. A more accurate measurement of the All Rods Out (AR0) critical boron concentration was made during low power physics testing.

1

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Nuclear Power Station unit n.. s PRECRITICAL ROD WITHDRAWAL 60-8

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Page 88

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6.1 , INITIAL CRITICALITY SulflARY CHRONOLOGY This section describes the major key events during the approach to initial criticality. All listed activities were performed on 01-23-86.

i Time Event 1400 All prerequisites and Initial Conditions are met.

~

1410 RCS boron concentration is measured as 1870 ppm.

Started taking baseline counts for 1/m plots during

. rod withdrawal.

1449 Started pulling shutdown bank A.

1602 All shutdown banks at 228 steps. RCS boron concentration measured as 1868 ppe. I 1649 Started withdrawing control bank A.

1751 Control bank D is at 160 steps.

p 1800 RCS boron concentration measured as- 1872 ppe.

V Started taking baseline counts for 1/m plots during dilution. -

1937 Started diluting the RCS at a rate of 80 gpa.

2130 Reduced dilution rate to 30 gps.

2140 Dilution stopped. 3 2145 RCS boron concentration is 1616 ppa.

2200 Reactor critical. RCS boron concentration is 1591.

2215 P-6 interlock is met. The source range trip is blocked.

2318 Reactor power is in the zero power' testing range and l low power physics tests are started.

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'Page 89 l

7.0 LOW POWER PHYSICS TESTING The objectives of. the low power physics testing (LPPT) program

-were. to obtain _ the physics', characteristics of the as-installed reactor core and to use this information to verify core design calculations. Demonstration of conformance with applicable Technical Specifications was also an objective. The LPPT was conducted with the RCS at normal operating temperature 'and pressure, 557'F and 2250 psia, respectively. Reactor power was maintained below 1 percent of full power. This power level

^

ensured a good signal-to-noise ratio- but was low enough to-avoid. nuclear heat effects. A reactivity computer system, diagrammed - in Figure 7.0-1, was used for reactivity measurements.

The LPPT 'is summarized in the following sections. In addition to the core physics related testing, a low power natural w/ circulation test was conducted under Appendix 7006 and is k described in Section 7.8. _j O

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TRIAX CABLE (DINECTOR 53175 (IN DRAWER) POWER RANGE DETECTOR CO-AX CHASSIS CONNECTOR 81- 1 R (ON PIC0 AMMETER)

N4X-A CABLE I PICOAltlETER SUPPLIED REACTIVITY 9 2-PIN COMPUTER MIKE N4X-B ~

HV SUPPLY O

't NOTES: The Nuclear instrumentation Detector cables are Triax cables terminated with Amphenol 43175 connectors. The Keithley picoammeter and power supply inputs are 83-1R Co-ax connectors.

Triax connector 53175 mates with Amphenol 52975 for cable-to-cable connection or Amphenol 34475 for cable-to-chassis termination. Chassis Co-ax connector 83-1R mates with cable connector 83-15P.

O Millstone Nuclear Power Stauon Unit No. 3 ZERO-POWER TESTING CONNECTIONS N

f Page 91

(3.

, Y 7.1 DETERMINATION OF THE HOT ZERO POWER TESTING RANGE

~

3-INT-7000, Appendix 7001 OBJECTIVE The objective of this test was to establish the hot zero power testing range to be used for Low Power Physics Testing (LPPT).

DISCUSSION The test was performed on 01-24-86. In order to' determine the point of adding heat, the core flux level was increased, at a rate of approximately 0.25 dpm, by manual withdrawal of control bank D. During the withdrawal, RCS temperature, intermediate range (IR) and power range (PR) nuclear instrumentation, and reactivity computer output were monitored. The core flux was increased until evidence of nuclear heat addition was detected by an increase in average RCS temperature and a decrease in n reactivity. The point of adding heat is the upper limit of the testing range. The lower limit ~ of the testing range was

. established 2 decades below the upper limit'.

RESULTS The addition of nuclear heat was observed at approximately

~

3x10 amps on both IR channels (N35 and N36) and at

.6 approximstely 1.6 x 10 amps on PR channel N44. Channel N44 was used to provide the power input signal to .the reactivity computer.

.8 .7 l The range of 1.6 x 10 to 1.6 x 10 amps on PR channel N44 was used as the hot zero power testing range fo'r LPPT.

i 9

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Pagn 92 j jq 1 7. 2 . REACTIVITY COMPUTER CHECK 0UT.

3-INT-7000, Appendix 7002 ..

OBJECTIVE The objective of this test was to verify proper operation of the analog reactivity computer as a prerequisite to performing LPPT.

' DISCUSSION -

This test was performed on 01-23-86 and 01-24-86. As a prerequisite to performing this test, the Beginning Of Life (BOL) delayed neutron parameters from the Westinghouse Nuclear Design Report were entered into the the reactivity computer.

These BOL delayed neutron parameters are listed in Table 7.2-1.

A dynamic check of the reactivity computer was then performed using the computer's internal exponential test circuit. .

( Following criticality, another dynamic check of the computer was performed by comparing the reactivity .value calculated by the computer to an inferred value based on stable reactor period. Results of this dynamic test are listed in Table 7.2-2. During LPPT, daily response checks of the computer were performed using the internal exponential test circuit.

RESULTS An internal exponential response check conducted on 01-24-86 indicated a malfunction with the reactivity computer. The unit j was immediately replaced with a second unit. After satisfactorily checking out the replacement unit, LPPT j i proceeded. Results of the checkout of the replacement computer are listed in Table 7.2-2. In order to validate the test data from the original reactivity computer, the problem with the '

unit was investigated. This indicated a problem with the _

exponential test circuit of the computer. The malfunction only l o

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~

i Page 93-

, affected the output of the computer while in the exponential test mode. Based on this, the - data collected during previous -

testing was determined to be valid. The replacement unit was used during the remainder of LPPT.

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2 0.001460 0.0308 3 0.001348 0.1153 )

l 4 0.002814 0.3113 Q 5 0.000955 1.2466 6 0.000319 3.3466 Where P - 18.92 usec -

. T = 0.970 1

1

'h Millstone gg, Nuclear Power Staua BOL DELAYED NEUTRON PARAMETERS 7.2-1 unit No. s

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Page 95 l

Original Reactivity Comouter j l

Indicated stable Interred Percent

Reactivity Reactor Reactivity Difference

Period #g (#,,,- #,g ) (100)

(pcm) (pcra) (pcm) p perted 106.2 50.51 105.46 0.70 63.88 100.27 63.71 0.27 36.0 200.89 35.8 0.56 O 19.4 400.76 19.3 0.52 i

l Reolacement Reactivity Computer l

l 105.35 50.59 105.36 -0.01 63.3 100.76 63.4 -0.16 35.7 / 200.94 35.8 -0.28 l 19.20 401.98 19.21 -0.05 -

Checkout Acceptance Criteria: Percent Difference 514.0% -

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""" Tme Nuclear Power Stada REACTIVITY COMPUTER CHECKOUT DATA Unit No. 3 7.2-2

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Page 96

- -7.3 BORON EN0 POINT 3-INT-7000, Appendix 7003 ~

.0BJECTIVE The objective of this -test was to ' determine the just-critical RCS boron concentration for the following. control. rod configurations:

1. All- Rods Out (ARO)

2. Control Bank D in

3. Control Banks C and 0 in

4. Control Banks A, B, C and D in

5. All Rods In (ARI) except rod F-021 DISCUSSION For each of the desired control rod configurations, critical conditions were established in the reactor (through borations or dilutions) with the rods as close as possible to the desired .,

configuration. The. RCS boron concentration was allowed to stabilize and samples were taken, then the appropriate rods

-l were withdrawn or inserted to achieve the ' desired 4

configuration. During thf s final adjustment, the reactivity )

-worth of the rods being moved was measured. The. measured reactivity was then converted to an equivalent boron concentration. The RCS boron concentration was then adjusted l using the equivalent value. The final adjusted number was the boron endpoint for the applicable control rod configuration.

RESULTS The boron endpoint.s determined by this test are given in Table 7.3-1. Also given are the predicted endpoints from the- )

Westinghouse Nuclear Design Report. All -test-determined endpoints compared favorably with .the design report . values. ,

I wa' i

1 Rod F-02 is the Most Reactive Rod Stuck Out

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Page 97 0 ..

Measured Predicted M-P

Bank _ Configuration (ppm) (ppm) (ppm)

ARO 1571 1566 +5 D in 1517 1499 +18 OV. D+CIn 1384 1357 +27 l

D+C+B+Ain 1116 1086 +30 1

ARI Less RCCA F-02 767 725 +42

Acceptance Criteria: Difference 1 100 ppm ui,[yo7.c sinuon

SUMMARY

OF BORON ENDPOINT T*ie UniL No. 3 TEST RESULTS 73-1

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4 ISOTHERMAL TEMPERATURE COEFFICIENT !

- i INT-7000, Appendix 7004 OBJECTIVE The objective of this . test was to determine the Isothermal

, Temperature Coefficient-(ITC). Using the measured ITC and the i l fuel, vendor . supplied -design fuel temperature coefficient data,_

the Moderator Temperature Coefficient (MTC) was determined.

DISCUSSION The test was performed from 01-23-86 to 1-25-86.

A heatup and cooldown of the RCS at a rate of between 10*;and 20*F per hour was initiated. During this operation, the change in reactivity versus the change in temperature was recorded on ,

an X-Y plotter. . The ITC was determin,ed by measuring the slope '

of the - X-Y plot. The value of the MTC was ~ determined by subtracting out the effect of the fuel temperature coefficient, supplied in the Nuclear Design Report from the ITC.

RESULTS The'~ test results of the ITC measurements are shown- on Table 7.4-1. All results are all within the design acceptance criteria as supplied by the fuel vendor. The all rods out value of the MTC was found to be - positive. Rod withdrawal limits, as- required by Technical Specification 3.1.1;3, were established to maintain the MTC negative at .all _ times during operation. _The rod withdrawal limits are shown on Figure 7.4-1.

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l Measured Predicted M-P

hqutg((gg locm/ 0F) focm/ 'F) locm/ 'F)_ f ARO -1.03 -1.69 +0.66

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D in -2.50 -3.24 + 0.74 D+Cin -6.07 -6.52 + 0.45 Acceptance Criteria: Difference 1

3 pcm/ 'F i tiillstone

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SUMMARY

OF ITC TEST RESULTS 'Ofi

Page 100 jm

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Unit No. 3

Page 101 7.5 CONTROL ROD WORTH MEASUREMENTS 3-INT-7000, Appendix 7005 OBJECTIVE The objective of this test was to determine the differential and integral worths of the control and shutdown rod banks, both individually and in overlap.

DISCUSSION The test was conducted from 01-24-86 to 01-28-86.

Starting from as close to the all rods out (ARO) critical condition as possible, control banks D, C, B, and A and shutdown banks E, D, and C were inserted individually. In each case, a dilution was started using primary grade water. As reactivity was added to the core from t?e dilution, control rods were inserted in increments to compensate for the reactivity addition. The reactivity inserted by each incremental rod insertion was measured using the reactivity computer. A typical rod worth trace during dilution is shown in Figure 7.5-1. At various points, the dilution was stopped

, to perform boron endpoint (Appendix 7003) and isothermal l

temperature coefficient (Appendix 7004) measurements. Prior to l

the insertion of shutdown bank E, a reactor trip was performed to meet the surveillance requirements of Technical l Specification 3.10.1. When shutdown bank C was fully inserted, the dilution was stopped and the F-02 control rodl was borated out of the core. The remaining two shutdown banks, A and B, were then diluted into the core to measure the N-1 boron l

endpoint 2 At the completion of the N-1 boron endpoint measurement, the reactor was tripped and then borated to the

" shutdown banks out/ control banks in" critical boron concentration. The reactor was then brought to a critical condition with all shutdown banks out and all control banks in.

O 1 Rod F-02 is the Most Reactive Rod Stuck Out 2This is a condition with all rods inserted except the Most Reactive Rod Stuck Out

a J.

Page 102 F f Following criticality, flux was increased to the zero power

-- (m)- .,

, testing range, and the control banks were , borated out in sequence and overlap. As boron was added to the . RCS, the.

control rods ' were withdrawn in incremental steps, and the reactivity added by each increment was measured on the reactivity computer in order to measure control rod worth 'in overlap.

RESULTS All acceptance criteria for the rod worths were met.

Table 7.5-1 summarizes the measured rod worths. Figures 7.5-2 through 7.5-19 show the measured integral and differential rod worth curves.

P O

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Measured Predicted (M-P)/P **

Bank IRGal IDCml jX)

D 619.5 593 + 4.46 C (D In) 1223.0 1254 -2.47 B (D+C In) 1239.5 1208 + 2.61 A (D+C+B In) 1216.3 1239 -1.83 i SDE (D+C+B+A In) 185.7 188 -1.22 f SDD (D+C+B+ A+SDE in) 547.8 526 + 4.14 :

b]~ j SDC (D+C+B+A+SDE+5 DD In) f,79.6 655 + 3.74 j l

ARI Less RCCA F-02 7925.7 7571 + 5.58 D-12 (HZP ins Limit) 386.9 491 *** N/A ;

Control Banks in Overlap 4365.6 4298.3* + 1.56

Sum of Individual predicted control bank values.

- Acceptance Criteria: Percent Difference .t 10%

- - Acceptance Criteria: Measured < 491 pcm O

nmston.

mi..e pow.c st.ua rei.

Unit No. 3

SUMMARY

OF ROD WORTH TEST RESULTS 7.3_ i '

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1 Page 123 7.6 ZERO POWER FLUX MAPS i

%].~

3-INT-7000 (Testing controlled by Base Procedure)

OBJECTIVE The objective of the zero power flux maps was to measure the core power distribution at hot zero power conditions and verify that core peaking factors were within the technical specification limits.

DISCUSSION The zero power flux maps were performed on 01-29-86 and 01-30-86. With . control banks at the desired rod position, reactor power was increased to between 1 and 2 percent power and a full core flux map was performed using the moveable incore detector system. During the flux map, data was collected on the plant process computer and later analyzed using the Westinghouse Incore 3.7 computer program. The b

L' J results of the analysis were compared to the core design and technical specification limits.

Flux maps were performed at the following conditions:

1. Zero Power Rod Insertion Limit (RIL): Control Bank A at 228 steps, Control Bank B at 164 steps, Control Bank C at 50 steps, and Control Bank D at 0 steps.

2. The Zero Power RIL with the control rod in core location D-12 withdrawn to 228 steps -(ejected rod measurement).

3. Control Bank D fully inserted with all other Control Banks i fully withdrawn.

4. All Rods Out (ARO)

All acceptance criteria were met for the zero power flux maps with the exception of the incore tilt measured in the "AR0" and "D in" flux maps. Both flux maps showed that the incore quadrant power tilt ratio design limit of 1.02 had been i 19

%./

exceeded. As the "D in" flux map and the "ARO flux" map had l

. -u-,- ,Aa - # ~

Page 124 3

been ~ performed -approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the "D-12 ejected f' '

rod" flux map, it was determined that localized xenon due to the simulated ejected rod configuration had caused the tilt. A fifth flux map using 21 symmetric thimbles was performed approximately 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after the "AR0" flux map to check the incore tilt at approximately 2 percent reactor power. The map l

e- showed the incore tilt to be less than the design limit of 'l 1.02. For specific test results see Tables 7.6-1 i through 7.6-5.

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Page 125

.O Test Date: 01-29-86 0315 - 0415 Map ID: HZP RIL Power Level: 1%

Boron Concentration: 1395 ppm Rod Position: CB A 228, CB B: 165, CB C: 53, CB D: 0 Maximum Measured Fq:

1.78 @ B7 Maximum FO : 2.76 @ F 15 L Maximum $H: 1.54 @ A6 Maximum F$g Error

-5.1 % @ C l 2 (from predicted):

Total Core

-36.5%

! Axial Offset:

Maximum Quadrant Design Limit: QPTR f 1.02 1.006 -

Power Tilt Ratio: -

Safety Limit: QPTR < 1.04 o

In locations unexcluded by Technical Specifications. F}y = 2.04 at 30% RTP l

L O

u,[y,Te sinuon CORE POWER DISTRIBUTION MEASUREMENT we UniL No. 3 HZP RIL 76-1

o ,

Pega 126 n

(

Test Date: 01-29-86 1100 - 1200 Map ID: HZP RIL D-12 Ejected Power Level: 1%

Boron Concentration: 1429 ppm Rod Position: CB A 228, CB B: 165, CB C: 52, CB D:0, D-12: 228

'n V FSAR Fg Limit: 11.5 Maximum Fo: 7.00 0 D12 Maximum Qg: 4.02 o C13 I

i l

See Section 7.7 for more information on the Pseudo Ejected Rod Testing I

,b ,

u,""f,'w7c sinuon CORE POWER DISTRIBUTION MEASUREMENT we unn No. s HZP RIL (D-12 EJECTED) 7.e-2

. .~

Page 127 i-M a

Test Date: 01-30-86 0600 - 0700 Map ID: CB D in; all other banks Out Power Level: 1%

Boron Concentration: 1511 ppm Rod Position: CB D: 0, all other banks >209 Maximum Measured Fy:

1.86 @ J-2 Maximum F0 : 2.81 0 G2 1

Maximum (H: 1.705 O J2 Maximum (H Error -7.9 9 D- 12 (from predicted):

Total Core g

Axial Offset:

Maximum Quadrant - ** 9 I' 1.023 "

. Power Tilt Ratio: Safety Limit: OPTR < 1.04

In locations unexcluded by Technical Specifications. Fh = 2.04 at 30% RTP.

" Design limit exceeded - see text.

o u,['E,',".c st uon CORE POWER DISTRIBUTION MEASUREMENT r ei.

UniL No. 3 CONTROL BANK D INSERTED 76-3

Page 128 Test Date: 01-30-86 1030 - 1130 Map ID: ARO HZP Power Level: 1%

Boron Concentration: 1566 ppm Rod Position: CB D: 228-Maximum Measured Fxy*: 1.578 @ J2 Maximum FQ : 2.36 @ J2 Maximum (H: 1.45 e J2 Maximum $H Error -4.7% e D12 (from predicted): ,

Total Core 2'37%

Axial Offset:

e Maximum Quadrant Design Limit: OPTR 31.02 1.023** '

Power Tilt. Ratio: Safety Limit: QPTR f 1.04 I

In locations unexcluded by Technical Specifications. Fh = 1.77 at 30% RTP.

- Design limit exceeded - see text.

1 O

%.)$7.*.st.uon CORE POWER DISTRIBUTION MEASUREMENT N' unit No. 3 ARO HZP 76-4

Paga 129

-' k 4

Test Date: 02-01-86 2146 - 2210 Map ID: Six Pass Symmetric Thimble Tilt Check Power Level: 2%

Boron Concentrationi NA

.f3 V

Rod Position: CB C: 107 Total Core: -38.82 Maximum Quadrant Des @n Umit: OPM i 1.02 1.003 Power Tilt Ratio: Safety Limit: OPTR f 1.04 O

ui,0$'l"*stmu. CORE POWER DISTRIBUTION MEASUREMENT we unn e. s SIX PASS SYMMETRIC THIMBLE TILT CHECK 7.6-5

, 2 Pags 130

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7, 7.

, 1,j "

PSEUD 0 EJECTED ROD TEST 3-INT-7000 (Testing controlled by Base Procedure)

OBJECTIVE.

The objectives of this test were to:

1. Measure the worth of the ' highest worth inserted rod to verify that the rod worth' used in the rod ejection l accident analysis was conservative.

2. Verify .that the core peaking factors measured by a1 flux .

map with the highest worth rod fully withdrawn from the core and the other control _ rods at ;the zero power rod insertion limit were less than the value assumed in the ,

accident analysis. I

.; DISCUSS {g .

The control rods were positioned at' the zero power rod' '

/~ .

insertion limit (RIL). Through control rod motion, reactor i power was- increased to approximately 1 percent and a flux map was performed. This provided a base line condition for the -

ejected rod. The power level was then reduced to the zero ,

power testing range and the rods were again repositioned 'at the zero power RIL. ?

5 The lift coils for all control bank D rods, except 0-12, were then deenergized. A boration was started, and, to compensate ,

for the negative reactivity addition, control rod D-12 was ;

withdrawn in discrete increments. The reactivity of each l withdrawal operation was measured on the reactivity computer.

Once rod 0-12 was fully withdrawn, core power was increased to i

approximately 1 percent and a' flux map was performed. The

~

power level during the performance of the first flux map in the ejected rod configuration was very unstable due to oscillations e in steam generator level. As a result, this flux map was not t analyzed and a second flux map was performed. This second map was used in the analysis.

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Paga 131 RESULTS

[- The worth' of' the' ejected rod and the peak FQ for the core were both less than.the safety analysis limits. The results of this' test are shown on. Table 7.7-1.

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e; Paga 132 O ..

Flux D-12 Measured Tech Safety Location Measured Map Position FQ Spec Analysis Maximum Limit Limit . Axlel Posit FjH Zero Power D-12 Aligned 2.99 4.64 4.64 F-15 1.54 in A-6 Rod with Control insertion Bank D et 0 Limit Steps 245

. O Zero Power D- 12 et 228 7.004 NA 11.5 D-12 4.02 in C- 13 Rod insertion Steps -

Limit D- 12 Ejected 372 Note: 0-12 Rod Worth = 365.9 pcm Predicted D-12 Worth = do! pcm l l

, O Philstone Nuci..r Power Staua PSEUDO EJECTED ROD TEST RESULTS LNL No. 3 '$_*3

l f .'f 3

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Pgge 133

!q p _

j _ 7.8 NATURAL CIRCULATION ,

.3-INT-7000,-Appendix 7006

y

. OBJECTIVE The objectives of this test were to:

1. Demonstrate plant performance capabilities and provide l operators with ' experience and training in core heat J removal by natural circulation with offsite power _

available. Satisfactory verification of natural l 1

circulation sIiall be confirmed by the establishment of 'l u

stable. reactor coolant loop temperatures subsequent to the~ ;

initiation of the transient. -

l

2. Verify the ability to bring the reactor to a hot zero-l power condition using natural ' circulation and the ,

atmospheric steam dump valves.

3. Determine. the length of time necessary to achieve and

- stabilize natural circulation. <

4. Determine reactor core flow distribution.

5. Verify - and monitor subcooling margin performance . under-natural circulation conditions. Through natural d circulation, the subcooled margin in the reactor shall be maintained > 30*F. Saturation conditions shall not exist -

in the RCS with'the exception of the pressurizer.

I DISCUSSION l The test. was performed on 01-30-86 with the reactor initially at slightly less than 5 percent power. The test transient was initiated by. tripping all reactor coolant pumps from the.

.]

control room. Monitoring of temperature indications- provided verification of the establishment of natural circulation flow.-

After steady state. conditions were verified, the reactor was-brought to . hot zero power conditions. Forced circulation ~was i then reestablished. .

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Page 134 ;]

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.. . l Data collection was accomplishediusing a process computer with .

4 ~ '

special programs, a computer trend block with data printer and

, the use of strip chart recorders.

' Verification .of satisfactory natural circulation flow 'was.

accomplished by monitoring plant parameters and the review of'

! collected data. !

RESULTS The reactor coolant pumps were tripped at 1910. Prior to ,

tripping the pumps,.a core exit thermoccuple map had been taken to document pre-transient conditir.is. Refer to Figure 7.8-1.

. . Natural circulation conditions we.e verified to exist at 1930.

This was based on stable core exit thermocouple readings as

, well as stable T hot and T cold readings. Natural circulation ~ ,

was maintained for approximately 30 minutes. Refer to Figure ;

7.8-2 for a typical core exit thermocouple map during natural' circulation. Plant cooldown was then initiated using the

% atmospheric dumps. This continued for approximately 40 minutes -

during which a cooldown rate of 30.7'F/hr was achieved. During i the cooldown, t'he lowest T,yg was 552.8*F which was above the .

test established lower limit of 551*F. .

l Once the cooldown was completed, the plant response to charging / letdown flow and ' pressurizer heater / spray valve d 1

operation was determined. At all times RCS subcooled margin '

(except in the pressurizer) was maintained above 30*F. When the plant response testing was completed, the reactor was shut q down and forced circulation established.

During the ' test, the lowest T,yg value observed was 552.3*F which was above the limit of 551*F. The lowest subcooled margin observed during the test was approximately 49*F which was above the 30 F limit. No L, expected responses were observed during the test.

l'

i 1

Page 135 l

A- typical ~. plant transie,nt. response plots covering the initial

, phase of the test where natural circulatio'n conditions were being established is provided as Figure 7.8-3 through 7.8-5. -i l

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U Nuclear Power Station STABLE CORE EXIT THERMOCOUPLE MAP rigur.

ut s. 3 NATURAL CIRCULATION TEST 7.s-2

l Paga 138 l y~g ,

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1 NATURAL CIRCULATION RCS C00GXM INITIATED l 600- ESNSE USIM EM IWS -6@@

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530 -

530 l

520 520'O 30 60 90 120 1% )

TIME (MIN) i PLOT 1 - RCS LOOP 1 WIDE R44GE T m r PLOT 2 - RCS LOOP 1 WIDE RN4GE Tc0LD

THE ARROWS ON W IS AND OTHER PLANT TRANSIENT RESPONSE PLOTS INDICATE THE VERTICAL AXIS ASSOCIATED WIE EACH PLOT. l O

,3jy,Ostuon TYPICAL RCS TwoT & TCRD Figure UniL No. 3 NATURAL CIRCULATION TEST 78-3

)

' ~ ' - -

Page 139

~

REACTOR C00LAffr RCS COOLDCWN INITIATED USitG STEN 4 DUTS C00LDOWh Pufs TRIPPED

" "A" -1,150 65-60 -

1 NATURAL CIRCULATION 2

% ESTABLISHED --->

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PLOT 1 - STEM GDERATOR 1 PPESSURE PLOT 2 - STEM GENEPATOR 1 WIDE RANGE LEVEL 1

O l wuci.No7.*r st uon PRESSURIZER LEVEL & PRESSURE PLOT rigure 7"

l UniL No. 3 NATURAL CIRCULATION TEST ,

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l Pags 140 i

(.:

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C00LIXW4 WUML CIRCULATION 50- ESTABLISHED RCS C00LDOWN INITIATED

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PORY CYCLES

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PLOT 1 - PPESSURIZER LEW.L PLOT 2 - PRESSURIZER PPESSURE O

'" TYPICAL S/G LEVEL & PRESSURE rigur.

,,,"y,Esuum 78-5 UnlL No. 3 NATURAL CIRCULATION TEST

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

y, f Page 141

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y+

hd '8.0 ' POWER ASCENSION TESTING

SUMMARY

3-INT-8000 ..

The base procedure controlled the sequence of events during initial' power operation.- Most of the testing occurred at power level plateaus. of 30, 50, 75, 90, and 100 percent. At each of j these power levels, both the primary and secondary systems-(plus auxiliaries) were' observed for operation within design

. specifications.. Plant and-test instruments were'used to verify. l proper operation, not'only at steady-state conditions, but also for selected transients. Prior- to proceeding from one plateau to another, the test data was reviewed to assure operation .at a ;

higher -- power level was- permissible. This test established j plant conditions necessary for specific tests, called for l

'~

individual power ascension tests to be performed, provided direction when in transitory periods between individual tests, i 4

and provided restoration requirements as needed. Major testing accomplished included the following: *

~

Instrumentation and controls ~ systems calibration and grooming

~ Plant performance verification (steady-state) ]

10 percent-load swing j

Reactor trip and shutdown outside the control room :

2 -

Large load reduction Loss of power trip , ;)

Generator trip from 100 percent MSIV closure The power ascension test sequence was accomplished over the period from 01-31-86 to 04-21-86.

, O :

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

4 Paga 142

,3 Nj 8.1.1, REACTOR COOLANT SYSTEM FLOW MEASUREMENT I

3-INT-8000, Appendix 8015 OBJECTIVE The objectives of the Reactor Coolant System Flow Measurement were: i

1. Determine the Reactor Coolant System (RCS) flow utilizing .

a precision heat balance.

2. Calculate correction factors for the RCS flow elbow taps j l

h in order to correlate their indications of flow with the. j precision heat balance flow.

3. Ensure that adequate Reactor System flow is present as ]

l required by Technical Specifications. f

\

1 DISCUSSION With the reactor plant operating at a 50 percent power level, a

, precision heat balance was performed to determine exact reactor thermal power. Reactor power was measured taking high accuracy readings from the protection cabinets and analyzed in accordance with a flow uncertainty analysis performed for this test. An overall uncertainty of 2.1 percent for reactor-coolant flow was achieved with this method. Based on this 50 percent power level, the elbow tap instrumentation was normalized. This test was repeated at' 90 percent power. The- i 50 percent preadjustment data and the post-adjustment flow data !

l taken at 90 percent power are presented in Table 8.1.1-1. j

. 3 RESULTS All acceptar.ce criteria were met. RCS flow was verified to be above the Technical Specification required level of 387,500 gpm (T.S. 3.2.3.1.a). Based on the RCS flow data taken at 90 percent power level, no adjustment to the RCS flow

. instrun:entation was required.

O l

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1 l

l Paga 143 l

~

i

.Yp]- l 50% POWER LEVEL 90% POWER LEVEL LOOP MEASURED INDICATED MEASURED INDICATED I 107.26% F414 : 101.95 110.1% F414 : 106.9 F415 : 103.78 F415 : 107.28 F416 : 102.78 F416 : 107.25 2 111.0% F424 : 102.7 109.7% F424 : 102.7 F425 : 103.9 F425 : 108.7 F426 : 102.08 F426 : 110.0 3 108.5% F434 : 103.4 108.13 F434 : 103.4 F435 : 105.78 F435 : 108.35 F436 : 103.18 F436 : 108.5 l

4 104.5% F444 : 102.58 104.8% F444 : 104.25 j F445 : 105.38 F445 : 103.68 i I

F446 : 94.0 F446 : 103.8

' RCS FLOW DATA u )y,w7c st uon Tm, unit m. :s RCS FLOW MEASUREMENT TEST e .1.1 -1

- 1

A Pagi 144 f~g L.L )' POWER COEFFICIENT '

3-INT-8000, Appendix 8020 ,,

OBJECTIVE

. The objective of this test was to verify - the Westinghouse RENCE Nuclear Design Report prediction of the doppler only power ,

coefficient.

O.

DISCUSSION At the 30, 50, 75, 90 and 100 percent power plateaus, the reactor was allowed to attain equilibrium xenon. Once steady state conditions were achieved, thermal power was measured and rod control was placed in Manual. Then, using the turbine controller, a series of step load decreases / increases of approximately 40 MWE each were made. During these transients, reactor power, AT, and Tavg were recorded. This data was used to calculate, at each power level, a doppler only power .

M coefficient verification factor (C ) which was compared to the Westinghouse Nuclear Design Report predicted doppler only power coefficient verification factor (C ),P RESULTS The results of the test are listed in Table 8.1.2-1. The i acceptance criteria requiring that the absolute difference between C M and CP be less than 0.5 F/% power was met.

Power l .2- 1

r l

Page 145 l l

O ..

l l

i ABSOLUTE DIFFERENCE l POWER LEVEL g CM CP vS Cp i 1.3.1. ("F/E POWER) (*F/E POWER) (*F/E POWER) 30 -2.75 2.66 0.09 )

50 -1.63 1.66 0.03 O

75 -1.05 1.13 0.08 l

l 90 -0.91 0.96 0.05 100 -0.90 0.90 0 Acceptance Criteria: Absolute difference between CM and CP is < 0.5 "F/% Power 4

O wi.))wer station DOPPLER ONLY Tm, Unit No. 3 POWER COEFFICIENT VERIFICATION 5.1.2-1 l

l

.__-_____--________n

f5 ~ , ,

Page 146

.1. '3 'RCS BORON MEASUREMENT X .3-INT-8000,. Appendix 8031 ,

OBJECTIVE-q The. objective of this test was to perform a core reactivity

. balance in' order to support comparison of the actual full power

. equilibrium RCS boron concentration to the Westinghouse Nuclear Design Report predicted value. i

,1 DISCUSSION ;

i .: The ' test was performed on 04-19-86. With the plant operating in a steady state condition at a 100_ percent power level with j control bank 0 at 210 steps and equilibrium xenon, three RCS l

boron samples were taken. In addition, primary side data l necessary to support calculation of a core reactivity balance ' l were also taken. A plant calorimetric was then performed to ,

4 accurately determine thermal power output. Using this l

. information, a. core reactivity balance was performed and used

~~

to correct - the measured RCS boron concentration for actual Tref, xenon, samarium and rod position. The corrected value was then comrarad to the predicted value of 1058 ppa.

RESULTS The corrected RCS boron concentration was required to be within i 1% AK/K of the predicted concentration. The ' corrected concentation was determined to be 1071 ppm which was within 0.124% AK/K of predicted. The acceptance criteria was met.

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-Page 147 )

I s i

^8.1. 4 ' CORE POWER DISTRIBUTION MEASUREMENT' 3-INT-8000 (Testing controlled by Base Procedure) ,

.l OBJECTIVE

.The. objective of .this' test was to measure the core power distribution at various core power levels in order to verify

^

l the measured peaking factors were within the limits specified j in Technical Specifications and the Westinghouse Nuclear Design l Report predictions. .

DISCUSSION Testing- was conducted over the period of 02-17-86 to'04-28-86.

A total of seven full core maps were taken and analyzed - one at 30, 50, and 90 percent power and two at 75 and 100 percent power. All flux maps were analyzed using the Westinghouse Incore 3.7 computer program.

O y RESULTS

[ The. results of the testing is provided in Tables 8.1.4-1 through 8.1.4-7. All the test acceptance criteria were met

^

with the exception of the 30 percent power level measured F xy value of 1.56 which exceeded the stated Technical Specification RTP F

x limit of 1.55. Review by Reactor Engineering: indicated that the measured F xy value did not exceed the Technical Specification F xy limit of 1.768. Considering this and since +

'an additional full core flux map was to be taken prior to

, increasing. power an additional 20 percent as required by Technical Specifications, the F xy. was considered acceptable. L All subsequent measured F values were within the Technical

' RTP Specification F x limits. ,

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Page 148

-7'4 Test Date: 02-17-86 Map ID: 30% Power Flux ~ Map Power Level: 1013 MWT Boron Concentration: 1303 ppm Rod Position: CB D: 184 .

Maximum Measured xy F *: 1.56 @ B7 Maximum Fn : 2.1IB @ B7 Maximum (g: 1.41037 Maximum dH Error 3.3% e Gi 1 (from predicted):

), Total Core

-3.924 Axial Offset:

Quadrant Power Top Half Bottom Half Tilt Ratios: of Core of Core Quadrant 1 0.9984 0.9984 Quadrant 2 0.9900 1.0018 e Design Limit: 11.02 Safety Limit: 31.04 i Quadrant 3 1.0077 1.0054 s ;

Quadrant 4 0.9950 0.9941 l

In locations unexcluded by Technical Specifications. l l

NOTE: The FyP limit of 1.55 was exceeded; however the F lxy limit for !

I 30% RTP of 1.768 was not exceeded. Fhwas less than the Technical

( .

Specification limit of 1.49 at RTP.

u,""$t" suum CORE POWER DISTRIBUTION MEASUREMENT we Unit No. 3 30 PERCENT POWER e.1.+ 1

Pega 149 A Test Date: 03-18-86 L/.

Map ID: 50% Power ARO ~

Power Level: 1700 MWT Boron Concentration: 1217 ppm Rod Position: CB D: 216 .

Maximum Measured Fxy*: 1.51 O B7 Maximum Fo: 2.014 @ 59 Maximum dg: 1.386 087 l Maximum dg Error 4% eE8 (from predicted):

Total Core

-2.616 Axial Offset:

Quadrant Power Top Half Bottom Half I Tilt Ratlos: of Core of Core -l Quadrant 1 0.9985 0.9987 Quadrant 2 0.9979 0.9999 Design Limit: 31.02 Quadrant 3 1.0106 1.0083 Quadrant 4 0.9930 0.9931

In locations unexcluded by Technical Specifications.

NOTE: Rxy F 9tmit of f 1.55 was met. F[g was less than the Technical Specification limit of 1.49 at RTP.

u,N".c st uon CORE POWER DISTRIBUTION MEASUREMENT we Unit No. 3 50 PERCENT POWER e.1 +2

Pcgn 150 I

Test Date: 03-27-86

&rm Map 10: 75% Power ARO -

Power Level: 2589.0 MWT Boron Concentration: 1125 ppm Rod Position: CB D: 222

)

Maximum Measured Fxy*: 1.48 o B7 !

Maximum Fg : 2.008 o B7 Maximum Faf ;g: 1.368 o 87 ;

1 Maximum dH Error 2.4% o G7 (from predicted):

O Tot i core Axial Offset:

-4.733 i Quadrant Power Top Half Bottom Half )

Tilt Ratlos: of Core of Core i Quadrant 1 0.9988 0.9989 Quadrant 2 1.0024 1.0018 e l Design Limit QPTR I 1.02 '

Ouadrant 3 1.0049 1.0048 % Limit WTR i 1.04 Quadrant 4 0.9937 0.9945

In locations unexcluded by Technical Specifications R

NOTE: RCS Flow = 104%. F fP limit of < l.55 was met. Fh was less than the Technical Specification limit,0f 1.49 at RTP.

u,"ToDst.uon CORE POWER DISTRIBUTION MEASUREMENT Tale unit No. 3 75 PERCENT POWER - ARO 8 1 A-3

Page 151

/'l Test Date: 04-14-86 V

Map ID: 75% Power incors/Excore Cross Calibration Power Level: 2566.0 MWT Boron Concentration: 1125 ppm Rod Position: CB D: 210 Maximum Measured Fxy*: 1.48 0 B7 Maximum F0 : I 900 # 07 N

Maximum F H: 1.364 O B7 N

Maximum FAH Error 3.8% 0 G7 (from predicted):

O retai Core Axial Offset:

3.05 Quadrant Power Top Half Bottom Half Tilt Ratlos: of Core of Core Quadrant 1 0.9967 0.9947 Quadrant 2 0.9984 0.9994 "

Design Limit: QPTR I 1.02

Quadrant 3 1.0053 1.0055 Quadrant 4 0.9996 1.0005

In locations unexcluded by Technical Specifications.

NOTE: RCS Flow = 104%. F%P11mit of 11.55 was met. Ffg was less thsn the Technical Specification limit of 1.49 at RTP.

U gg,j,'w7e st uon CORE POWER DISTRIBUTION MEASUREMENT we Unit No. 3 75 PERCENT POWER 01 Ad

Paga 152 Test Date: 04-17-86 0r"' .

Map 10: 90% Power ARO ~

Power Level: 3050.0 MWT Rod Position: CB D: 202 Maximum Measured Fy:

1.49 @ B7 .

Maximum Fg: 2.04 @ B7 Maximum $H: 1.36 e B7 Maximum da Error _ gg ,

(from predicted):

Total Core

-8.89 Axlal Offset:

Quadrant Power Top Half Bottom Half Tilt Ratios: of Core of Core Quadrant 1 0.9978 0.9975

, Quadrant 2 0.9970 0.9995 "

Design Limit: QPTR S 1.02 Quadrant 3 1.0074 1.0078 yeWm QMR 41.04 Quadrant 4 0.9978 0.9953

In locations unexcluded by technical specifications R

NOTE: Burnup = 670 MWD /MTU. RCS Flow = 107%. FdP limit of f 1.55 was

, met. Fh was less than the Technical Specification limit of 1.49 at RTP.

O ui,[y,7.c st.uon CORE POWER DISTRIBUTION MEASUREMENT rei.

Unit No. 3 90 PERCENT POWER 81'*3 1

Paga 153 Test Date: 04-19-86 w/

Map ID: 100% Power ARCr l

Power Level: 3411.0 MWT l Boron Concentration: 1078 ppm I Rod Position: CB D: 213 l

Maximum Measured Fxy*: 1.47 o 87 Maximum Fo : 1.99 0 B7 i

Maximum Fyg: 1.35 o B7 !

l Maximum F H Error % o R11 (from predicted):

O Toteicore Axial Offset:

7.2e Quadrant Power Top Half Bottom Half Tilt Rattos: of Corg of Core Quadrant 1 0.9965 0.9973 Quadrant 2 0.9973 0.9993

Design Limit: QPTR ( 1.02 Safety Limit QPTR 11.04 Quadrant 3 1.0068 1.0080 s Quadrant 4 0.9995 0.9955

In locations unexcluded by Technical Specifications.

NOTE: Surnup = 760 MWD /MTU. FRP limit of 11.55 was met. F[H was less than the Technical Specification limit of 1.49 at RTP.

O ui,["'j'st uon CORE POWER DISTRIBUTION MEASUREMENT Table unit No. 3 100 PERCENT POWER - MAP 1 81 A-6 4

> Pcgg 154

'( '

s Test Date: 04-28-86 Map ID: 100% Power ARO-Power Level: 3410.0 MWT I

Boron Concentration: 1090 ppm Rod Posttion: CB D: 212 Maximum Measured Fxy*: 1.47 0 B7 Maximum F0 : 1.98e87 '

Maximum (g: 1.35087 Maximum dH Error 4.5% 9 R 11 (from predicted): a Total Core

-6.88 Axial Offset:

Quadrant Power Top Half Bottom Half Tilt Ratlos: grCore of Core Quadrant 1 0.9979 0.9974 Quadrant 2 ~ 0.9985 0.9978 "

Design Limit: QPTR 11.02 Safety Limit: 09TR 11.04 Quadrant 3 1.0060 1.0070 %

Quadrant 4 0.9971 0.9970

In locations unexcluded by Technical Specifications.

NOTES: Burnup = 977 MWD /MTU. RCS Flow = 107%. F%P limit of 11.55 was l

met. FM was less than the Technical Specification limit of 1.49 at RTP. l g'%)

%,,['$,7 cst.uon CORE POWER DISTRIBUTION MEASUREMENT w. j uniLNo.3 100 PERCENT POWER - MAP 2 8 1 A-7 l

1

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l

7 -

Page 155 8.2.1' OPERATIONAL ALIGNMENT VERIFICATION OF NUCLEAR INSTRUMENTATION 3-INT-8000, Appendix 8002 08JECTIVE The' objectives of this test were to:

1. Calibrate the excore power range instrumentation utilizing the power level calculation from the plant process computer calorimetric.

" 2. Determine overlap indication between the Source Range (SR), Intermediate Range (IR) and Power Range (PR) channels.

3. Verify that PR currents versus reactor power exhibit linear response.

DISCUSSION The test was conducted on 02-15-86, 3-15-86, 3-17-86, 3-26-86, 4-16-86 and 4-18-86 with the plant at 30, 40, 50,'75, 90 and

_p 100 percent power levels, respectively. At each plateau, plant-calorimetrics were performed in order to obtain data for PR adjustments. In addition, at 30 percent power, the flux deviation alignment was verified by manually mahipulating the output of a single channel and observing the flux level at which the deviation alarm occurred.

Between the 75 and 90 percent test plateaus, PR detectors N42 and N44 were replaced when water was discovered in their wells in the neutron shield tank. ' When the water was found in the

- wells, an inservice leak test was performed on the Neutron Shield Tank (NST). No leaks were found and it 'was therefore postulated that the water entered the wells during NST fill or testing operationt. The original N42 and N44 detectors had-exhibited higher detector current than those of N41 and N43, due to the additional moderation from the water in the N42 and N44 wells. The original' detectors exhibited normal response to power level changes and trips and good overlap with the

"~ ~ ~ ^ ~ ^ ~ ~ " ' ' ~ ~ ~ ~ ~ ~

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Page 156

[ . intermediate range channels. After replacing N42 and N44, the PR checks- were -again performed at 30, 40, 50 and 75 percent-power levels. _ The initial tests at 90 and 100 percent power -

levels were then performed.

Throughout the test IR and PR output data was recorded and evaluated to ensure proper detecter overlap. SR'and IR overlap

. data taken during initial criticality was reviewed in order to ensure at least bne decade of overlap existed.

I RESULTS

)

The required overlap of at least one decade between SR to IR and IR to PR was successfully verified. After adjustments, all PR channels consistently . agreed within 2 percent 'of the secondary calorimetric reactor. power level. All PR channels exhibited a linear response in the power range.

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.Page 157 7% 8.2.2 OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE INSTRUMENTATION 3-INT-8000, Appendix 8004

~

1 ;

OBJECTIVE The objective of this test was to acquire data to align the AT and- T ,yg process instrumentation such that individual instrumentation channels are consistent with each other and consistent with core thermal power.

DISCUSSION The test was performed on 02-15-86, 03-17-86, 03-26-86, 04-16-86 and 04-18-60 with the plant at power levels of 30, 50, 75, 90 and 100 percent, respectively. Process control system T

hot and T cold data was collected during thermal equilibrium at listed power levels. Using this data, full load T,yg and AT values were extrapolated and used to align the process control system T,yg and AT loops at each power level.

. C'

\ '

. RESULTS The AT and T,yg process loops were successfully aligned. At '

100 percent each channel's average AT was within the acceptance criteria of 55*F to 60*F. The AT values were 55.00*F, 55.02*F, 56.03*F, and 55.65*F for loops 1, 2, 3 and 4, respectively. In addition, each channel's. T ,yg was below the high limit of )

SS7.1*F. The values were 585.77*F, 584.53*F, 585.40 F, and l 585.30*F for loops 1, 2,'3 and 4, respectively. I All acceptan'ce criteria were based on the Westinghouse Precautions, Limitations and Setpoints (PLS) document.

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" 7 7 , 8.2.3L CALIBRATION OF STEAM FLOW AND FEEDWATER' FLOW '

3-INT-8000,1 Appendix 8003-u Y

.y OBJECTIVE

'To determine recalibration data for Steam Flow Transmitters to conform ~to actual plant conditions as determined" by 'the calorimetric program.

I 1

DISCUSSION The test was performed on 02-15-86, 03-17-86,' 03-26-86 and i 04-18-86 with the plant at 30, 50, 75, and~100 percent power l levels, respectively. During the test, process control- system i parameters for feedwater flow, steam flow, and steam pressure were recorded. Using this data, the process control loops were

)

then adjusted so that steam flow matched feedwater flow during l steady state conditions.

As - a first step, based on test data, corrected steam flow j L\ transmitter ranges were calculated and used to recalibrate the j steam flow transmitters. Then the process control system was adjusted to its original settings so that its alignment matched the new transmitter calibration. This process was' repeated at each of the power plateaus. Since this procedure was strictly a data collection and. adjustment . evolution, there were no.

acceptance criteria.

RESULTS Steam flow, feedwater flow and steam- pressure data was collected and used to adjust the steam flow' instrumentation at each of the power plateaus. Based on data obtained from the test, the steam flow transmitters were recalibrated following. 1 1

the completion of the Power Ascension testing program. All

, activities were successfully completed.

H l

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Paga 159 -

'[ ,

ME 8.2.4 INCORE/EXCORE NUCLEAR INSTRUMENTATION CROSS-CALIBRATION ;

b/ 3-INT-8000,. Appendix 8028 OBJECTIVE The objective of this test was to determine .the relationship between the axial offset determined by an incore flux map and the axial. offset as indicated by the excore power range nuclear

~

~ '

instrumentation. Using the. measured incore to excore relationship, calibration factors were ' determined for the excore power range neutron detectors and the T'ilting Factors computer program. ,

DISCUSSION The test was performed during the period on 03-28-86 and ,

04-14-86 at a power level of 75 percent. This test consisted of taking a series of incore flux maps over several different axial flux conditions. The measured incore axial offset was '

q then compared to the axial offset determined from the upper and A_/ lower excore detector currents which had been measured at the time of the flux maps.

The first. calibration: was performed at a 50 percent power level. This was to determine' the preliminary calibration factors for the excore detectors prior to exceeding 50 percent power and to provide initial calibration of excore detectors.

During this time, two full core flux maps and two quarter core flux maps were performed over a 15 percent change in axial offset. The results of the preliminary calibration are shown in Table 8.2.4-1. This data indicated that the excore power range channels were capable of being calibrated. However, the result's for channels N42 and N44 were of concern in that they did not produce the expected test results as seen in channels N41 and N43. As the excore detectors sit inside dry wells in a water-filled, natural circulation cooled neutron shield tank,

~

it was felt that the unexpected test results could have been

c E Ptge 160 L.'. h' --

' Y\ due to temperature variations within the tank. Based on this proposed explanation, the decision was made to increase power to 75 percent and to perform the test at 75 p&rcent power or above as required by technical specifications.

At' the 75 percent power plateau, three full core flux maps and five quarter core flux maps were performed over a 23 percent swing in axial offset. The plot of axial offset versus time is shown in Figure 8.2.4-1. The results of the test are shown in Table 8.2.4-2 and in . Figures 8.2.4-2 through 8.2.4-5. These results once again showed that the detectors were capable of being calibrated but the data for detectors N42 and N44 did not produce the expected results in that the current for detector N42 Bottom was approximately twice the current of N42 Top and the current for detectors N44 Top and Botton were approximately 10 times higher than the current found on channels N41, N42, and N43.

O V

Based on this anomalous data, a decision was made to check the excore detectors in containment. This was performed during a .

cold shutdown for steam generator water chemistry cleanup prior to increasing power above 75 percent. A series of electronic checks had already been made on the excore detector channels from the instrument racks. No problems had been noted. During the cold shutdown, the detectors were checked for loose connections and general detector condition inside the detector wells. Inspection of the detector wells indicated . that the well for channel N42 contained approximately 1.5 feet of water ;

and the N44 well contained approximately 3.5 feet of water. In addition, the aluminum can that houses the detectors for channel N44 was full of water. The other six excore detector wells were examined and found to be dry.

}

l

.Page 161

[]

V After this discovery, the detectors for channels N42 and- N44 were removed from the detector wells and a leak test was performed on the neutron shield tank. The leak' test applied a pressure of 15 psig to the tank and was held for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The test results showed no leakage of water into the detector wells and it was subsequently decided that, during the initial fill of the neutron shield tank, water had spilled out of the tank manways on the top of the tank and into the detector wells.

Although the detector wells were inspected after the initial fill, the water was evidently not noticed. The detector wells were pumped out, dried and two new power range detectors were installed. As channel N44 was used as the input channel to the reactivity computer during Low Power Physics Tests (LPPT), an evaluation was done on the acceptability of the physics test results. Sirice testing of channel N44 indicated no damage had been done to the detector, and since previous incore/excore cross-calibration test results showed the detector to be O capable of being calibrated, it was determined that LPPT

~

results were still valid.

The third incore/excore cross-calib' ration was performed during the power ascension following the outage. Prior to startup, the two new detectors which had been installed were adjusted using the calibration factors determined in the previous incore/excore cross-calibration using symmetrically opposite detectors. Channel N42 was adjusted using channel N41's calibration factors and channel N44 was adjusted using N43's calibration factors. At 50 percent power a check of Quadrant Power Tilt Ratio (QPTR) and excore axial flux difference was performed. The indicated QPTR was less than the technical specification limit of 1.02 and greatest difference between the highest and lowest indicated axial flux difference channel was less than 2 percent. Power was then increased to 75 percent and the third set of incore/excore cross-calibration Q measurements were taken.

V

I Page 162

-[L The third calibration consisted of two full core flux maps and 1 b two quarter core flux maps over an 18 percent change in axial flux offset. The plot of axial offset versus time is shown in J Figure 8.2.4-6. Additional' quarter core flux maps and one full core flux map had been planned; however, it became necessary to reduce power after the second quarter core flux map due to an l oil leak in the turbine generator electro-hydraulic control system. The data from the four flux maps was analyzed. The results for the two detectors which had not been replaced was

~

consistent with the results of the previous calibration and the-results for the two new detectors was consistent with the expected results. The results of the third calibration are shown on Table 8.2.4-3 and Figures 8.2.4-7 through 8.2.4-10.

RESULTS The objectives of the test were met. As discussed above, problems with power range detectors N42 and N44 were corrected.

h., The performance of the excore detector system- has been V satisfactory with the original N41 and N43 detectors and the replacement N42 and N44 units. ,

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b Paga 163

'f%

V Detector 41 Calibration Curves:

Incorego =.934(Excore Ao)I 6.35 Upper CURR =.575( Ag) + 113.65 LowercuRR = .785( Ag) + 123.95 Detector 42 Calibration Curves:

1 incorego = 1.318(ExcoreAo) + 43.5 l Upper cuRR =.910( Ag) + 105.04 LowercuRR "-l 74( Af) + 209.15 l

' - Detector 43 Calibration Curves: l l

Incore3o =1.357(ExcoreAo) + 4.6 Upper cuRR " 979( Af )

I I7'I3 j

LowercuRR = .746( Ag) + 125.64 )

i l

l Detector 44 Calibration Curves:

IncoreAo =1.357(Excore Ao) + 0.416 1

UppercuRR =9.023(Ag) + 1157.9 Lower cuRR --7.868( Ag) + 1165.8 i

Notes: Number of data points 4 Axial Flux Difference swing 8.2%

Duration 3-18-86 to 3-24-86 e 50% RTP h %,,$,'w7c sinuon INCORE/EXCORE CROSS-CAllBRATION Tabi.

unn m.s PRELIMINARY TEST - 50 PERCENT POWER om

Ptgs 1616 l

)' Detector 41 Calibration Curves:

~

Incore3o =1.355(ExcoreAo) + 6.1 Upperm =.833( Ag) + 111.5 Lowerm - .812( Ag) + 122.4 Detector 42 Calibration Curves:

1 IncoreAo =1.420(ExcoreAo) + 43.1 )

Upperm =.800( Ag) + 108.77 Lowerm --1.79( Ag) + 203.45 Detector 43 Calibration Curves: l O J I IncoreAo = 1.380(ExcoreAo ) + 3.78 I

Upperm =.894( Ag) + 118.47 ;

1 Lowerm - .852( Ag) + 125.30 Detector 44 Calibration Curves: -

IncoreAo = 1.520(ExcoreAo ) + 2.50 Upperm =7.15( Ag) + 1126.76 Lowerm --7.94( Ag) + 1165.04 Notes: Number of data points 9 1 Axial Flux Difference swing 255 Duration 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months "i"*t'a' INCORE/EXCORE CROSS-CAllBRATION Tel.

N7s*"" " TEST 1 - 75 PERCENT POWER s.2+2

Page 165 1

if Detector 41 Calibration Curves:

IncoreAo =1.340(ExcoreAoi+ 6.35 UppercORR " 002( Af) + 107.9 LowercuRR = .768( Af) + 118.95

~

Detector 42 Calibration Curves:

incorego = 1.350(ExcoreAo) + 8.72 Upper cURR " 03d( Af) + 106.6 LowercuRR - .838( Ag) + 121.5 l% y _ _ _ - _ _

Dete.ctor.43 Calibration Curves:

Incore Ao = 1.340(Excore Ao) + 3.85 Upper CURR " 034( Af) + I I4 04 LowercuRR = .795( Ag) + 121.76 Detector 44 Calibration Curves:

IncoreAo_ =1.340(ExcoreAo) + 19.8 Upper CURR =.902( Ag) + 112.94 LowercuRR =-1.13( Ag) + 152.24 Notes: Number of data points 4 Axial Flux Difference swing 14:22 Duration 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months O " " ""

INCORE/EXCORE CROSS-CAllBRATION re.

$$73 TEST 2 - 75 PERCENT POWER 82A-3

Page 166 r:

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O v us,)' Der st8uon INCORE AO YERSUS CHANNEL 41 CURRENT rigure unit no. s TEST 1 - 75 PERCENT POWER e.2.+2

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g Page 176 f4 8.3.1 REACTOR AND TURBINE CONTROL b[ 3-INT-8000,-Appendix 8005 OBJECTIVE-The objectives of this test were:

1. To - determine the T ,yg program resulting in the highest possible steam pressure and optimum plant efficiency-without exceeding pressure limitations for the turbine, or the maximum allowable T,yg.

2. To obtain primary system temperatures, steam pressures and reactor thermal power data at steady-state conditions for zero, 30, 50, 75, 30 and 100 percent power levels.

DISCUSSION The test was performed on 02-01-86, 02-15-86, 03-17-86, 04-13-86, 04-15-86, and 04-18-86 with the plant at power levels of zero, 30, 50, 75, 90' and 100 percent, respectively. Plant j performance data, including loop Thot, Tcold, T,yg, feedwater

\, flow, feedwater temperature, steam pressure, turbine inlet pressure, turbine impulse chamber pressure and plant gross electrical output, was collected at each power plateau. .This data was then analyzed and compared .to the design T,yg and steam pressure. Based on this comparison, adjustments to the T,yg control program were to be made to achieve the design steam pressure for each power level while still maintaining parameters within design limitations.

At the zero, 50, and 75 percent power level plateaus, data was taken twice -

once with steam supplied to the moisture separator / reheaters (MSR) and once with steam isolated. The tests with steam supplied to the MSRs were intended to closely ;

approximate actual plant performance conditions. Steam was continuously supplied to the MSRs during the 90 and 100 percent l power level data collection periods. '

Page 177

.f~y , During the 100 percent power testing, required plant conditions

'- included full load steam generator pressures between. 980 and 1000 psia, and T,yg less than the upper design limit of 587.1*F. This was - to verify that the T,yg control program was properly adjusted.

RESULTS The T,yg control program was verified to function properly in that T,yg and full load steam pressures were within design ;

limits. No adjustments to the control program were required.

r Figure 8.3.1-1 provides the T,yg and average steam generator pressure as a function of power level, determined during the .

test.

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Note: The above graphs are averaged representations Of numerous data points taken during testing and should not be considered Official test results.

O w,[$,",e st,t,,, RCS TEMPERATURE AND STEAM GENERATOR Figure Unit No. 3 PRESSURE AS A FUNCTION OF REACTOR POWER 8.3.1 - 1

Pags 179 j fN 8.3.2 DYNAMIC AUTOMATIC STEAM DUMP CONTROL TEST k./ ' 3-INT-8000, Appendix 8013 OBJECTIVE The objective of this test was to verify the proper closed loop response of the steam dump control system in the T,yg and steam pressure modes of operation. The'T,yg mode was tested in both the plant trip and load reject submodes.

DISCUSSION The test was performed on 02-11-86.

The plant trip submode was tested by increasing T,yg to 567 F with power maintained at 15 percent by manual rod control. A reactor trip was then ' simulated to the steam dump system so as ,.

to control T,yg on the plant trip controller. The steam dump was then placed in T,yg mode and data collected for 10 minutes fq. to ensure the plant trip controller achieved and maintained a V stable T,yg. The acceptance criteria was for T,yg to be maintained within l'F of the program value of 562*F with no divergent oscillations in temperature.

The load reject submode was tested by maintaining power at ,

15 percent and T avg at approximately program level (562*F) in manual rod control with a high rate of load rejection and zero impulse pressure simulated (load reject to 0 percent.). The steam dump was placed in T avg mode and data collected for 10 minutes to ensure that the load reject controller achieved and maintained a stable T,yg. The acceptance criteria was the T,yg to be maintained 1.5 to 4*F above the 557'F no load value with ;

no divergent temperature oscillation. ,

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The steam pressure mode was tested by setting the steam header pressure controller to 1078 psig at 15 percent power, placing .j the dump valve controller in automatic, and monitoring plant

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pressure response for 10 minutes following a= slight increase in

,fs M' reactor . power, - The acceptance ' criteria was that the steam generator pressure- controller ~ response could miintain a stable !

1078 psig pressure. , I RESULTS All test acceptance criteria were met. In the plant trip ,

submode, T,yg was ' maintained at 561*F which was within the acceptance criteria o f. 562*F 11*F. For the load' reject

, submode, T,yg was maintained' at 561*F. which was - within the acceptance criteria band of 558.5*F to 561*F. In the steam pressure mode, steam header. pressure was maintained at' 1078 psig which was as required by the acceptance criteria. No divergent oscillations were observed during any of' the

, transient testing of the steam dump system.

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'3-INT-8000,? Appendix 8017 1 f: y - -.

s OBJECTIVE-

. The' objective of this test was to verify the performance of the y _ . automatic ~ reactor control system in maintaining reactor coolant-average temperature, T,yg, within~ acceptable steady-state. -

limits.

DISCUSSION The test was performed on 02-18-86 with the reactor and turbine D

generator at a steady-state power " level of 30 ' percent. The pressurizer level and . pressure control, steam generator water level control, and ' turbine . driven feed pump speed ; control ,

systems were all in automatic. The steam dump system was in~

automatic i n ' the ' T,yg mode. The rod control system . was -in ,

manual. The following plant parameters were monitored:

auctioneered nuclear flux, power mismatch, compensated power mismatch, auct.ioneered hi T,yg, compensated T,yg , . T error' E

9 compensated T ref, rod speed demand, steam ' header pressure, ,

turbine impulse pressure, and pressurizer pressure.

The test consisted of switching the rod ' control system; to automatic and monitoring the plant response. Rods were then

. shifted to manual and withdrawn to create a 6*F mismatch between T,yg and T ref. The rods were shifted to automatic to allow T,yg to return to T ref. This' step was then repeated with rods driven in to create the 6'F mismatch. ,

The acceptance criteria was that no manual -intervention was required and that T,yg returned to within 1.5'F of Tref' '

g RESULTS The plant responded as expected. The rod control system l controlled T,yg in a stable manner. No adjustments were t i

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..'l required to' fine tune the instrumentation. Following rod-withdrawal, T ref was at 566*F and T,yg was' at 572*F. Once

, automatic control was established,~ T,yg ret'urned to 566 F within 398 seconds. Following rod insertion, T ref was at 566*F 'l and T,yg was at 560*F. Once automatic control was established, ;

T,yg ' returned to 566*F within 259 seconds. At no time ' was manual intervention required.

The transient response of Thot, Tcold, pressurizer level and pressure during this test is illustrated in Figure 8.3.3-1.

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, ['; 8.3.4 AUTOMATIC STEAM GENERATOR WATER LEVEL CONTROL 3-INT-8000, Appendix 8018 OBJECTIVE The objectives of'this test were to:

1. Demonstrate the level control stability 'of the steam generator feedwater bypass valves in automatic control at low power.

2. Demonstrate the stability of the steam generator water level control system when transferring control from the feedwater bypass valves to the main feedwater. valves.

3. Demonstrate proper response of the automatic steam generator level control system during plant transients at power levels of 50, 75, and 100 percent with adjustments being made as required. to optimize system performance.

4. Demonstrate proper operation of the turbine driven feedwater pump speed control during power escalation.

5. Verify proper automatic programming of the steam generator

' level during power escalation.

DISCUSSION The test was performed over the periods of 2-10-86 to 02-15-86, 02-16-86 to 03-23-86, 03-18-86 to 03-23-86, 03-28-86 to 03-30-86, and 04-20-86 to 04-21-86 at power levels of <5, 30, 50, 75 and 100 percent, respectively.

With the unit operating at less than 5 percent power and on the feedwater control bypass valves, a set of +5 percent. and

-5 percent narrow range steam generator level deviations were imposed on the plant. The system response was recorded as ,

steam generator water level control was switched from manual to automatic. This verified the bypass valve control system l before proceeding to higher power levels. i (2

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Page 188 l e

i Testing the transfer of steam generator water- level control f]

~v from the feedwater control bypass valves to the main feedwater control valves was' performed at 20 percent powe'r. During this operation, the main feedwater control valves were slowly opened l in manual while observing the feedwater control- bypass valves closing in automatic.

At.'30 percent power, the steam flow and feedwater flow

' .i nstrument calibration was conducted in accordance with' -

Appendix 8004. Level deviations of +5 percent and -5 percent were then used to observe the steam generator water level control system's transient response. At the 50, 75, and 100 percent power levels, tests consisted of repeating the-steam flow and feedwater flow transmitter calibrations, followed by recording the system response to the 10 percent ,

load swing test (Appendix 8022). The 75 percent power level test included system performance throughout a 50 percent load reduction -(Appendix 8026). The plant parameters monitored u during the tests included:

Steam Generator Programmed Level Setpoint Narrow Range Steam Generator Water Level Level Controller Output Nuclear Instrumentation Power Level Feedwater Flow Steam Flow Flow Error Flow Valve Controller Output i

Data was collected on strip chart ' recorders during the tests ;

below 30 percent power. A computer was used as a data-logger 9-for the 30, 50, 75 and 100 percent power tests.

r During each test the process control loops for feedwater control valves and feed pump speed control were adjusted as

Page 189 f~l required to achieve optimum performance. In addition, data on j V the control loop settings and the actual feedwater control valve differential pressure was recorded so that'the scaling of the ' control valves could be adjusted to match plant peiformance. Though separate from this test, steam generator j water level oscillations were observed at 58 percent power, and !

additional adjustment was performed to optimize system response before increasing power level. The feedwater control- valve l position was increased and feedwater pump speed decreased to j stabilize the levels, and then further testing and control system adjustment resumed. .

l RESULTS Automatic steam generator water level control demonstrated the J ability to meet the est.ablished acceptance criteria: l

1. Level overshoot /undershoot was less than 14.0 percent-

/~3 following a level increase / decrease.

2. Level returned to within 2 percent of reference level, within 10 minutes following a transfer of level control, or within 20 minutes following a change in level or level setpoint.

Automatic feedwater pump speed control was demonstrated to meet .!

the established acceptance criteria: l

1. Feedwater pump discharge pressure oscillations were less J than 13 percent following a steam flow change.

2. Main feedwater control valve stem position was:

Steam Flow (%) Valve Position (%) '

30 10-30 50 20-40 'I 75 40-60 100 60-85 i

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11 ); 8.3.5 MAIN' STEAM ISOLATION VALVE CLOSURE TEST .;

3-INT-8000, Appendix 8037 , ,

7

'3 OBJECTIVE ,

The objectives of this test were to:

1) Verify, under dynamic steam flow conditions, the ability >

of the valves' to close in less than'5 seconds.

2[ Verify the ability of the primary plant, secondary plant, and plant _ automatic control systems to sustain the simultaneous closure of all MSIVs and bring the plant to stable hot st'andby conditions without initiating safety-injection or lifting primary / secondary safety valves.

y DISCUSSION The test was performed on 03-31-86 with plant power being ,

maintained at 20 percent. The test was . initiated by the.

.. simultaneous manual closure of all four main steam isolation valves. The plant was brought to hot standby conditions by use of the atmospheric steam dumps. Final' steam generator pressure was 1092 psig. Plant conditions were monitored using installed instrumentation, the plant computer, and a - high' speed d'ata

. logger.

RESULTS ,

All MSIVs closed in less than 5. seconds with A, B, C, and 0 ;

closing in 3.11, 2.76, 3.05, _ and 3.20 seconds ' respectively.- ;

During the test,.neither the pressurizer safety valves nor main- .;

steam safety valves lifted, nor did safety injection initiate. ;

, All acceptance criteria were met. Plant performance following~ ;

closure was as expected. .The transient response of various plant parameters during this test is -illustrated in Figure ;

8.3.5-1. ,

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Page.191 sO MAIN STEAN l'ALATION VALVES CLOSED 40- -40 35 -

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8. 4.1 ' .

TURBINE OVERSPEED TEST ,

3-INT-8000, Appendix 8016 OBJECTIVE The objective of this test was to demonstrate the capability of

~

the turbine generator to consistently trip at acceptable speeds e during an overspeed condition.

DISCUSSION The test was performed on 02-15-86 with the plant, initially, at a 15 percent power level. Prior to' performing the actual overspeed. tests, the electrohydraulic control- (EHC) system was put through a series of electrical and mechanical tests. After these were successfully performed, the unit's backup, overspeed trip feature was tested by running the turbine generator up to 105 percent of rated speed and observing the trip. This was 4 performed three times. .The mechanical overspeed mechanism was then functionally checked at a reduced speed.

With the backup overspeed system and elements of the mechanical overspeed system tested, the turbine generator was then set to overspeed in order to perform a functional . check of the mechanical overspeed trip and verify that the unit tripped at an acceptable level. This was also performed three times. ,

RESULTS All checks and trips were successfully performed. During the 105 percent trip of the backup o'verspeed trip feature, the unit tripped consistently at 1894 RPM during each of the three runs. ;

This was well within the acceptance criteria range of 1845 to !

1935 RPM. During the mechanical overspeed trip portion, the unit tripped at 1962, 1963, and 1963 RPM. This compared well-to the acceptance criteria of < 1998 RPM.

l

. . _ _ _ _ . _ _ _ _ _ . - _ . . _. . _ _ . . _ _ - _ . _ _ _ _ :. _ 2 _ ._ _l

i p i Page 194

[ 8.4.2 10 PERCENT LOAD SWING TESTS 3-INT-8000, Appendix 8022 ..

'08JECTIVE The objective of this test was to verify proper plant transient response, including automatic control system performance, when 120 MWE step load changes were introduced at the turbine 'l generator.

. l DISCUSSION l The test was performed on 2-18-86, 3-23-86, 3-29-86 and 4-21-86 at reactor power levels of 30, 50, 75 and 100 percent, respectively. The test consisted of rapidly lowering the generator load by approximately 120_ MWE by adjusting the EHC load limiter to a predetermined target value. When the plant had stabilized at the new power level, the generator load was rapidly increased to its original level using the EHC standby Ox load set potentiometer.

During and after each transient, the following plant parameters were monitored:

Auctioneered nuclear flux Loop 1 T hot narrow range Loop 1 T cold narrow range Loop 1 T,yg Loop 1 AT T

ref SG 1 feed flow Steam flows Steam generator levels Steam header pressure Feed pump discharge pressure Pressurizer pressure ;

Pressurizer level 4

Pege 195-

'(u,[ Auctioneered T avg Loop l' overpower AT trip setpoint ,

Loop 1 overtemperature AT trip setpoint=

Generator output (MWE)

Feedwater temperature Acceptance criteria for this test were:

1. Reactor trip does not occur

2. Turbine trip does not occur

3. Steam generator atmospheric dump valves do not lift

j. 4. Steam generator code safety valves do not lift j 5. Pressurizer power operated relief valves do -not lift

6. Pressurizer code safety valves do not lift

7. Unexpected manual operator intervention is not required

8. Plant parameters do not incur sustained or divergent oscillations

9. Nuclear power overshoot or undershoot is <3 percent RESULTS The test was successfully performed with the following exceptions:

1) On the 10 percent decrease from 75 percent, the atmospheric dump valve for steam generator A lifted. The setpoint selected on the main board hand-indicating controller for that valve was set too low. The setpoint was readjusted by Operations personnel.

2) On the 10 percent decrease from 100 percent power, feedwater flow started oscillating. Manual intervention was required to stop the oscillation. I&C personnel investigated and determined that the steam generator water level controller characteristics had been changed by a recent repacking of feedwater regulating valves. The valves had been made less responsive due to tighter

,p packing. The steam generator level control system was

-l adjusted to compensate for the tighter packing.

b.

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Page 196 l

. : (A, r, The above discrepancies were corrected..as noted or evaluated to

.% /

be acceptable. During each induced transient,

. t , undershoot/ overshoot was within the 3*F acceptace criteria.

The maximum value observed was ' approximately 2*F undershot during the increase to 100 percent power. Figure 8.4.2-1 provides a representation of typical plant response to a load )

change. The information was taken during the testing at !

100 percent power. ,

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,- Page 202

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W M 8.4.3 - REACTOR TRIP AND SHUTDOWN OUTSIDE CONTROL ROOM 1

.3-INT-8000,l Appendix 8023 ,,

OBJECTIVES The' objectives of this test were:

1. To demonstrate plant trip and.- shutdown '. capability from-outside; the control- room, resulting in hot standby >

condition, utilizing the Technical Specification minimum ,

shift crew. -

~ 2. To , demonstrate that the plant can be maintained 'in hot ,

standby condition from outside the. control room.

3. To demonstrate that plant' control can be transferred back

- to the control room from the remote control location.

As. an initial condition of the test, reactor power level was required.to be greater than 10 percent.

,-- DISCUSSION With the reactor operating ~ at a power level of approximately 1 15 percent, the test commenced at 1630. on 02-18-86 by initiating- a remote reactor trip from the' reactor trip breakers .

located on the 43'6" level of the Auxiliary Building. Turbine

^

trip occurred. automatically following .the reactor trip. Plant control was then transferred to the Auxiliary Shutdown Panel located on the 4'6" level of the : Control Building. A Hot Standby condition -(Mode 3) was achieved at 1635. ' After Mode 3'

' had been maintained for more than thirty minutes, control was transferred back to. the Control Room. Reactor startup commenced'at 1730 hours0.02 days 0.481 hours 0.00286 weeks 6.58265e-4 months .

l No abnormal conditions - occurred during the test. System, equipment, and instrument response was as expected for a normal ;

plant trip.

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Pagm 203-(3. . RESULTS The acceptance criteria for the reactor trip and shutdown outside the-control room test were:

1. The plant can be remotely tripped with' transfer to the Auxiliary Shutdown Panel. Hot standby condition (Mode 3) can be achieved from outside the control room per plant Emergency Operating Procedures.

2~. Plant Hot standby condition (Mode 3) can be maintained for-at least 30 minutes from outside the contro.1 room.

3. With stable plant conditions, control can be transferred y

back to the control room from the remote control location.

All acceptance criteria for the test were demonstrated satisfactorily.

In addition to the above test,.the ability to take the plant to Hot Shutdown (Mode 4) from .outside the Control Room was

(

successfully demonstrated during the precore hot functional-test.

fC l

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l Page 204~

s 3 -8;4.4L LARGE LOAD REDUCTION 3-INT-8000, Appendix 8026 _ i

, 1 ..

OBJECTIVE c The. objectives of the test were to:

1) Verify the' ability of the primary plant, secondary plant and the automatic reactor control system to sustain a 50 percent step load reduction from . a 75 percent power a level. -

2) To obtain transient response data for the evaluation of the interaction of plant systems.

2

3) To obtain transient response data for determination if control system setpoint changes were. required to improve transient response based on actual plant operation. ,

DISCUSSION , ,

The test was performed on 03-30-86. Prior to the start of the

. test, _ the plant was operating in steady-state conditions at' 75 percent power. Additionally, the reactor rod control system, the turbine bypass system, steam generator water level

' control system, pressurizer pressure and level control systems -

and the feedwater' pump speed control system were in automatic' control.

, 1 4

The reduction in power was: accomplished by a rapid lowering of ,

the setpoint of the turbine control load limiter to a .

previously determined target value. ;

.j Acceptance criteria for the test was that the plant could 1 sustain the transient without a reactor or turbine trip, safety injection, . lifting of steam generator or pressurizer safety '

valves or unexpected manual intervention. In addition to these-acceptance criteria, there also were predicted values for the-extreme values of several plant parameters during the ,

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Page 205 )

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n. 1 transient. These included T avg, generator and

- Q steam

. pressurizer levels, pressurizer pressure and time duration of j

j l

maximum rod speed and steam dump actuation. 1

- I RESULTS 1 The plant responded as expected. The transient was successfully performed and all acceptance criteria were met.

The plant electrical load was reduced from 861 WE to 214 WE, a drop of 56.3 percent. Of this reduction, 550 WE were shed in the first 25 seconds of the transient. Figure 8.4.4-1 indicates the reduction in generator output during the ,

performance of the test.

The only operator involvement in the establishment of stable conditions aft.er the transient was to place the feed pump speed-controller to manual. This was to minimize the interaction between the two pumps at low power levels.

b) w During the duration of the test, the predicted transient extremes of several parameters were exceeded. This was due to the load reduction being larger than 50 percent and were not deviations from the acceptance criteria. The predicted extreme and actual extreme values are shown in Table 8.4.4-1. The primary system pressure transient was controlled by pressurizer spray and a 4.5 second opening of the PORV.

After completing the test, the plant was returned to a 75 percent power level to permit the continuation of the testing program.

The transient response of various plant parameters during this test is illustrated in Figure 8.4.4-2.

i

l Paga 206 I

PARAMETER EXPEC1ED EXTREME ACTUAL EXTREME Tave peak <5'F above initial 7'F steady state value Tave undershoot (5'F below final 2*F steady state value Tave oscillation <5'F during steam dump 0*F

+80 psi +75 psi Primary pressure f

-100 psi -125 psi

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S/G D -25% + 18%

Maximum Rod Speed s 30 seconds 1 minute 16 seconds Steam Dump ( 8 minutes 8 minutes 30 seconds Actuation Note: The above values are expected results and do not represent acceptance criteria.

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V 8.4.51 LOSS'0F: POWER TEST (20 PERCENT POWER)

INT-8000, Appendix 8030' .,

08JECTIVE The objectives of this test were to:

?!" 1. . Demonstrate that the plant responds as designed following a plant. trip with no offsite power.

2' . Demonstrate that. the . turbine driven auxiliary feedwater~

pump (TDAFP) will' maintain adequate steam generator levels

, for a minimum of. two' hours with the motor driven auxiliary feedwater pumps- (MOAFP) and the auxiliary feedwater pump cubicle ventilation system out-of-service.

3. Demonstrate the capability of the batteries to provide vital power without any AC support (battery chargers and AC power to the inverters out-of-service) for a minimum of two hours.

DISCUSSION

'The test wa's performed on 03-31-86. Just prior to initiating a loss-of power, the MDAFP and the auxiliary feedwater pumps' cubicle ventilation system were removed from service by placing applicable switch controls in pull-to-lock. This ensured that 2' only the TDAFP would:be available to provide feedwater to the steam generators and that it would run without any ventilation.

Also, AC pow'er breakers to the battery chargers and . inverters

were opened.

The test was initiated with - the' plant at 16 . percent power level. The - turbine was off-line and steam was being dumped. to the condenser through the condenser' dump / turbine bypass valves.

The tran.sient was begun by manually ' tripping- the reactor and then opening all off-site feeder breakers for the 4.16KV and

.. 6.9KV buses. The emergency diesel generators started and' sequenced on vital loads. Plant response was. monitored with- '

- - . ~ , . . . - - . , - . . - - - .. , , - - - - - - . ~ -r.--. -. - ,m - a

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Pag) 212 l

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( the cpeputer and control board indications. Natural ~

.v-J-i

_ circulation was established in the primary sys, tem. The TDAFP and . atmospheric dump valves were used to remove heat for a period of two hours.

Following the- test, a plant startup was performed to support further testing.

RESULTS All acceptance criteria for this test were met, with exceptions noted, as follows:

1. The diesel generators started and sequenced on loads as required except that that the auxiliary building filter fan 3HVR*FN68 and cold shutdown air compressor 3IAS-C2B failed to ' start. In addition, control building chiller 3HVK*CHL1B started as designed, but tripped shortly

.i thereafter. See Appendix 0 for a discussion of problems

( . encountered during LOP and their resolution.

2. The TDAFP operated well within established design limits throughout the two-hour run as indicated below.

ITEM MAXIMUM READING LIMIT Bearing 134*F 1200 F Temperature Bearing Supply 94 F 1150 F Oil Temperature Bearing Return 106*F 1180 F Oil Temperature Turbine Rotor .6 mils 11.5 mils Vibration (peak-to peak)

7, - . .. _ - _ _ . _ .. - __

l Page 213

- . j T

Q . ITEM MAXIMUM READING LIMIT Pump Shaft Vibration- .90 mils '11.2 mils 4400 RPM (peak-to peak)

Pump Shaft Vibration .95 mils $1.5 mils 3400 RPM (peak-to peak) -l The maximum readings for Bearing Temperature (134*F) and !

Turbine Rotor Vibration (.6 mils) were recorded ,

immediately after. startup. Within 15 minutes, 'both ,

readings were down to 120*F and .45 mils respectively. As the pump operated, the vibration continued to decrease {

with all bearing vibrations stabilizing between .18 and .

.24 mils. ,

3. The TDAFP cubicle temperature steadied out at a maximum of ,

'97'F, well. within the 50 to 120*F normal temperature range. The. EEQ Design Basis maximum abnormal excursion, the transient-considered for the TDAFP cubicle on-a loss of all AC power, .is a 58'F increase from 104 to 162'F. :

Relative' Humidity- (RH) reached a maximum of 58.6 percent ,

approximately 80 minutes into' the. run, and then decreased to 53 percent at the end of the two-hour run. The design range is from 10 percent to 75' percent RH. ;

1 The transient response of various plant parameters during this test is illustrated in Figure 8.4.5-1.

E LOP PROBLEM SlM4ARY ;

Refer to Appendix 0 for a summary of problems . encountered during the LOP test.

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8.4.6 : GENERATOR TRIP FROM 100 PERCENT POWER

~3-INT-8000, Appendix 8032

{ , .

OBJECTIVE ,

.The objectives of this test were to: ,

1. . Verify the ability of the primary and secondary plants to sustain a trip:from 100 percent load..

2. Verify' the ability of control systems to bring the plant .

, to a stable Hot Zero Power (HZP) condition.

DISCUSSION ,

The test was performed on 04-21-86. Plant load was established 1 at approximately 100 percent. Prior to initiating the ~ trip rod control, steam generator water level control, pressurizer -;

pressure / level control, and steam generator feed pump speed j control were all placed in AUTO. In addition, RCS ..Tavg, AT, .i

=

' team s generator levels, .and pressurizer pressure and level were - ,

? verified to - be within the normal full power operating bands. 5 Test personnel . were stationed to observe the Main Control

Boards, pressur,izer safety valves, and steam generator- safety valves. A high speed data. acquisit. ion system' was - set up to !

record key plant parameters. . With the plant operating at t 100 percent power . the . test transient was initiated when the ;

2 generator output breaker was opened by jumpering- contacts on the Reverse Power Relay. The generator output breaker ~ opened ,

at 0513 on 04-21-86. Recovery from the resulting turbine trip - :

and reactor trip was in accordance with ~ plant procedures. ;

The following acceptance' criteria applied to the test.

1. All' rods fully inserted and nuclear power decreased to ,

less than 15 percent in two seconds. l

2. Safety injection did not occur. I

3. Pressurizer safety valves did not lift.

l

4. Steam generator safety valves did not lift. l

5. RCS T,yg remained above the P12 setpoint of 551 F.

t

' 2; .

h. .,

'Page'218

. 6. Pressurizer pressure remained above 1925 psia.

7. Pressurizer level remained above 17 perc'ent.~

8. A reactor trip resulted from the turbine trip.

L 9; Turbine speed remained less than 1980- rpm.

10. The overall RCS T hot resporse time was less than 6.0 seconds.

RESULTS All test acceptance criteria were met: -

1. Nuclear power was observed to decrease to less than 15 percent in two seconds.

2. Safety injection did not occur.

3. Pressurizer safety valves did not lift.

4. Steam generator safety valves did not lift.

l 5.

The lowest observed *RCS T,yg was 552.9'F which was above the acceptance criteria of 551*F.

6. The lowest observed pressurizer pressure was 2003 psia which was above the acceptance criteria of 1925 psia.

7. The lowest observed pressurizer level was 24.6 percent-which was' above the acceptance criteria of 17.0 percent.

8. A reactor trip resulted from the turbine trip.

9. Peak turbine speed was 1868 rpm which was' less than the acceptance criteria of 1980 rpm.

10. The acceptance criteria for the overall RCS hot leg i response time was 6.0 seconds.

This ' response time was calculated by measuring the time interval' between the point where neutron flux had decreased to 50 percent of its original value to the point where T hot started to decrease.

This method of calculating the loop response times yielded a 4.0 second response time for loops 1 and 2. Loops 1 and 2 were the two RCS loops where hot leg response time was measured during this test.

s

'[ _l

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, - v Yi Page 219 p.

After, review of the test results with Westinghouse, it was determined that the method used for def.ermination of the ;

. overall hot leg response time should have .been the time .

interval between the point -where neutron flux had decreased to 50 percent of its original value to the point where the . hot leg . temperature had decreased by 331/3 percent of the initial delta T.

Using this new method to calculate overall hot leg response time resulted in the following:

New Acceptance Criteria Loop 1 (w/o pressurizer) 6.7 seconds 5 6.8 seconds l Loop 2 (w/ pressurizer) 8.7 seconds 1 8.4 seconds r

. g Westinghouse ' reviewed the failure of the loop 2 hot leg transit time and based on a sensitivity study concluded that the additional 0.3 seconds did not, impact the conclusions in the FSAR. However, a reanalysis of five accidents in the FSAR which rely on the overpower and overtemperature delta T reactor trips was, determined to be required.

The following five accidents being reanalyzed are:

1. Loss of Load

2. Rod Withdrawal at Power

3. RCS Depressurization

4. Steam Line Break at Power .

5. Steam Generator Tube Rupture It is anticipated this reanalysis will be complete on or about 09-01-86.

Figure 8.4.6-1 illustrates the response of various plant

~

parameters to the transient. Table 8.4.6-1 details the responseofYariousplantparametersduringthetest. -

1

Pega 220 1

p 1 pramintar_funiini Initial Minimme Maximum Einal.

Nuclear Power, Channel 41 (N) 99.9 0 99.9 0 Tevg . Loop 1(*F) 587.0 552.9 587.0 558 Tror (OF) 587.5 557.6 587.5 557.7 A T Loop 1 (N) 100.6 1A6 100.6 1A6 OP A T, Loop 1 (N) 109.7 108.6 109.9 109A OTA T. Loop 1 (N) . 112.1 109 2 149.9 145.8 Pressurizer Pressure (psie) 2261.3 2003.0 2261.3 2206.3 Pressurizer Lowl (N) 61.6 24.6 61.9 27.6 Steam Generetor NR Level (N)

Loop 1 51.1 1.8 51.1 5A Loop 2 47.7 0 47.7 2.1 Loop 3 50.0 2A 50.0 3.0 Loop 4 50.3 2.9 50.3 2.9

/ Steam Flow (199H)

- Loop 1 3684 0 3691 0 Loop 2 3684.1 0 3686.5 0 Loop 3 3754.9 0 3754.9 0 Loop 4 3671.9 0 36 % .3 0 Steam Generetor Pressure (psig) j Loop 1 9782 9782 1082.3 1082.3 Loop 2 976.3 975.6 1079.7 1079.7 Loop 3 970.0 970.0 1075 1075 Loop 4 971.0 971.0 1076 1076 Main Feedwater Flow (MPPH)

Loop 1 3833 0 3847 0 Loop 2 3725.6 0 4045A 0 Loop 3 3898.9 0 4826.7 0 Loop 4 3781.7 0 3781.7 0 Note: The above data was taken from a combination of direct indicator observation, data trer.ds, and the temporarily installed high speed data acquisition system.

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'*5i$ E"s*'" GENERATOR TRIP FROM 100% POWER */83 ,

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NOTE: m ER iN m 2. 0uTSuRsE CAUSED BY TE LOSS OF LOAD #O TURBINE BYPASS VALVE M:TUATION, PRESSURIZER 2,300 -

LIM!L MO PRESSURE ARE ESSENTIALLY couTRou.ED BY MAT RDOV4. THROUGH o TE STEAM GENERATORS.

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, , 8. 5.1, CALORIMETRIC-7 3-INT-8000, Appendix 8001

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OBJECTIVE The objective of this test was.to determine, at selected power !

levels,. plant thermal power by means of 'a manually calculated - ,

calorimetric. These calculated values were used as . input' to

'the readjustment of -the power range (PR) instrumentation. In

' addition, .the manually calculated values were compared against the values from the plant procest computer calorimetric program '

(3P3) as~a validation process.

DISCUSSION The test was conducted at the 25, 30,-40, 50, 75, 90, 98, and 100 percent power plateaus. Once stable plant conditions were' established, data was collected on selected plant. parameters.

In each case, data was taken for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period at 5 minute intervals. This data was then reduced and the plant power

.- level calculated. ;

4 RESULTS The results of this test are summarized in Table 8.5.1-1. In each case the process computer (3P3) calculated power levels compared favorably with those from the manual calculations.

All objectives of the test were met.

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.4 Manually Calculated Computer Calculated >

Nominal Power Power Level Power Level (X) (X) (X) 25 23.48 24.37 30 30.40 30.44 40 41.00 40.75 i O. 50 50.69 50.47 75 74.70 7488

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90 89.69 89.58 !

l 98 97.26 97.50 100 99.99 99.91 1 Millstone gg, Nucisar Power Statia twt wo. s PLANT CALORIMETRIC DATA 3333

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[(M). L8.5.2 SECONDARY PLANT. PERFORMANCE

,' 3-INT-8000' Appendix 8006 OBJECTIVES The objectives of this. test were to:

1. Obtain baseline plant operating data at 10, 40, 50, 75, 90 and 100 ~ percent power plateaus for use 'in the Secondary

- Plant Performance Monitoring Program.

2. Determine the turbine generator and se'condary plant ,

o performance. as' an initial condition to conducting.

] performance testing during Warranty Run (3-INT-9000, Appendix 9002). ,

3. Acquire specific system and component data- to permit

[ proper comparison of initial. performance test results to

- turbine generator manufacturer's guarantee values,

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DISCUSSION

. The secondary plant performance test made maximum use of-permanently installed plant instrumentation and plant Process Computer for data acquisition. In addition precision test instruments were installed to= monitor low pressure (LP) turbine exhaust pressures,.. main t'urbine control valve positions and l makeup flow to the hotwell(s). Local gauges.were used for low pressure extractions steam pressures. The ' test precedure was prepared using the ANSI /ASME PTC-6 Steam Turbine Performance ,

Test Code for guidance. The ' plant process computer data .

acquisition software was designed to.' allow data to be recorded 3 on both hardcopy and magnetic tape. !

l The test was performed, based on plant status' over the period of 02-16 through 04-19-86. During the 30, 40, and 50 percent power plateaus, a single data run was performed. Two-data ; runs, were performed at 75, 90 and 100 percent power x

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levels. Each test run consisted of four distinct steps; . cycle

' isolation, steady state verification, data collection and data reduction / correction.

.y The cycle isolation step required a systematic check of drain -

valves, traps, turbine bypass ' valves, feedwater heater and MSR emergency drain volves, steam seal system valves and pump minimum flow valves. During this step test personnel used portable infrared imaging equipment, digital heat. probes and an ultrasonic leak detector to determine the condition of each isolation point. Plant Trouble Reports , were submitted for malfunctioning equipment. The overall purpose of this step was to ensure optimum plant component / system performance existed prior to performance data collection.

Steady state verification consisted of acquiring two hours of '

computerized and local performance data. Variations in

'Q selected parameters were compared to a predetermined steady state projected value. Once test personnel determined steady j state conditions existed, the data run portion of the test began.

The test run required two additional hours of steady state data ,

1 collection. At 75, 90 and 100 percent test points, steam j generator blowdown was isolated and auxiliary steam

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requirements' were supplied by the auxiliary boiler. This minimized calculational uncertair ty in steam flow to the main turbine.

The final section of the test involved averaging and correcting specific parameters to reference cycle conditions. These corrected test values were compared to target or predicted values at each power level. The predicted values and corrections were developed from performing a computer heat balance simulation for each test power level. These computer e

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. based heat balances' were based : on - vendor des'ign ..' data modified to reflect both plant "As ' Built" configurat. ion and: actual 1 system alignment.-

After all appropriate ' corrections were made, corrected net

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i turbine heat rate and generator load ' were - calculated. In between the two test data runs conducted at the 75, .90 and 100 percent power plateaus', turbine control- valve positions were modified and then returned to their initial pdsitions. Once steady state conditions were reestablished, the second data run commenced. This process ensured independence of data runs.

The corrected heat rates from duplicate test runs were required to agree within 0.25 percent.

RESULTS During the . test,. a total of ten test runs were performed as !

power was escalated from -30 to 100 percent NSSS ' rated power.

Table 8.5.2-1 summarizes the corrected net turbine heat rate q and generator. load -calculated for each test. hold point and l compares them to the heat balance predictions at each power.

level. As _ indicated, overall turbine generator performance exceeded ' predicted across' the various load ranges. In addition, below is a summary of other major component testing.

1.- Condenser During - this test, no attempt was made to evaluate main condenser thermal performance. This- was because the original design information was made obsolete as a result of tuoe change, during construction, out of the original 70-30 Cu-N tubes with titanium alloy tubes. ;

2. Feedwater Heaters Overall feedwater heater . performance .was- close to l

l predicted at: rated ponr. The final feedvater temperature 1

of 439'F was at or slightly above pred1,-ted. The only significant performance . deviations were noted at three l'

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b . specific points within the three feedwater heater strings.

These are the drain cooler approaches (DCA) on 1A,18 and 1C heaters, subcooler approaches (SC) on 4A, 48 ' and 4C heaters, and terminal temperature differences (TTD) on 6A, 6B and 6C heaters. Suspected causes and remedial actions are as follow:

Heater No.

Problem Suspected Cause Recommended Action '

1A, 18, 1C Steam / Vapor bypass Raise level to High DCA into drain cooler break the vapor Temperature inlet bypass; reestablish at proper operating level 4A, 4B, 4C e

(,; Low SC Higher than normal None; the high approach operating level- level's needed to temperature maintain drain .

pump NPSH 6A, 68, 6C Low TTD Drain level low in Establish and heaters maintain loop seals Further in-service testing is planned to establish proper DCA values on 1A, 18 and 1C feedwater heaters. Trouble '

Plant maintenance requests have have been issued to ensure loop seals on 6A, 6B and 6C heaters are filled. Table 8.5.2-2 provides a comparison of test to predicted performance data for all three feedwater heater strings.

3 Paga 231

3. Moisture Separator / Reheater Performance Moisture Separator / Reheater (MSR) performance, was reasonably close to predicted performance at rated power.

Test data for the two MSRs showed remarkable similarity.

This . indicates an approximately even flow and duty split between the HSRs. Table 8.5.2-3 gives a comparison . of tests to predicted performance values. As noted on this table, most test values are lower than predicted. It is suspected that the reason is more likely a result of difficulty in heat balance modeling than any performance deficiency.

The key MSR performance index is the thermal temperature difference (TTD). A lower TTD indicates more efficient heat transfer.

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CORRECTED NET CORRECTED

]' ' TEST TEST LOAD TURBINE HEAT RATE 2 GENERATOR LOAD '

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DATE X RATED POWER TEST PREDICTED TEST PREDICTED l 2/16/86 30 11646 1240311240 308.8 290129 3/14-15/86 40 11283 1146211146 424.2 418142 l'

3/17/86 50 10700 109I411091 548.7 538154 3/26/86 75 10G'. 2 101081202 868.1 861 18 )

l 3/29-30/86 75 10012 101081202 869.5 ,861118 4/12/86 75 3 9982 100421201 879.8 875118 4/15/86 90 9776 9867i197 1077.7 1067121 4/16/86 90 9805 98671197 1074.0 1067121 l

4/18/86 100 9722 9790198 1202.0 1194112 4/19/86 100 9741 9790i98 1199.9 1194112

1. First test run with Motsture Separator Reheaters in service

2. NTHR = Gross E ec ric o er (MWE O

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una m.s SECONDARY PLANT PERFORMANCE TESTING o52-1

Page 233 g~

T HEATER TTD ('F) DCA ( *F) . TR ( *F)

NUMBER TEST PREDICTED TEST PREDICTED TEST PREDICTED 1A 3.0 3.1 55.4 9.6 74.8 75.3 2.7 68.2 9.6 77.6 75.3 18 3.1 1C 3.0 3.1 43.8 9.6 78.4 75.3 2A 4.4 7.3 5.3 10.2 39.4 41.9 2B 6.8 7.3 4.9 10.2 41.2 41.9 2C 6.9 7.3 4.9 10.2 37.0 41.9 3A 2.5 3.7 5.7 10.6 40.4 38.1 3B 3.5 3.7 7.I 10.6 41.0. 38.1 3C 3.2 3.7 4.5 10.6 39.9 38.1 4A 6.3 5.5 37.4 56.3 64.2 66.2 48 56.3 62.5 66.2

. 8.6 5.5 37.8 4C 8.1 5.5 39.2 56.3 65.3 66.2 5A 0.4 3.0 10.1 12.8 64.3 62.8 58 4.8 3.0 6.1 12.8 65.0 62.8 1 SC 5.5 3.0 7.2 12.8 63.6 62.8 6A -0.6 3.1 N/A N/A -

69.8 61.5 !

6B - 1.0 3.I N/A N/A 67.9 61.5 6C -0.3 3.1 N/A N/A 68.7 61.5 ,

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Data from Test Hold Point 100.2.1 at 100% Rated Power O

u,,Dcstauen FEEDWATER HEATER PERFORMANCE DATA Tele unit m. s SECONDARY PLANT PERFORMANCE TESTING e.s.2-2

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REHEATER A REHEATERB 4

STEAM FLOW test: 607.18 606.52

-(Klbm/hr) predicted: 660.02 660.02 l

I lNLET TEMP test: 365.9 366.3

(*F) predicted: 369.5 369.5

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OUTLET TEMP test: 504.6 504.2

(*F) predicted: 513.18 513.18 TEMP RISE test:- 138.7 137.9

('F) predicted: 143.7 143.7 SUPERHEAT test: 139.5 138.9 (oF) predicted: 145.5 145.5 DRAIN TEMP test: 525.8 523.6

('F) predicted: 537.7 537.7 TTD test: 21.2 19.4 (OF) predicted: 23.9 23.9

Data from Test Point 100.2.1 at 100% Rated Power

,,j'yl"w st.um MOISTURE SEPARATOR / REHEATER DATA Tabi. ;

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.8.5.'3 : RADIATION MONITORING SYSTEM .

3-INT-8000, Appendix'8007 -

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OBJECTIVE j The objectives of this test were to:

1. Measure and document the gamma and neutron radiation j levels in selected areas of Millstone Unit 3 during power ascension testing.

2. Determine ~ locations where permanent ' shielding or engineered barriers (i.e., high radiation area doors, labyrinth entrances, etc.), are deficient or not in' conformance with the Millstone Unit 3 FSAR. .

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3. Compare permanently installed area radiation monitor readings to portable radiation instrumentation results.

Compare selected permanently iristalled process monitor readings with grab sample results.

4. Identify high radiation areas and verify access is V controlled as required.

5. Determine ' neutron spectrum factors for various areas inside the containment building.

6. Log the permanently installed area radiation monitor alarms at the 100 percent reactor power test plateau, the reason for the alarms, and their. final' disposition.

7 DISCUSSION ,

A total of 378 Radiation Base Points (RBPs) were selected to be surveyed at each power level (zero, 30, 50, 75, 90 and 100 percent) during the power ascension testing program.

Survey points were chosen at each installed radiation monitor location,- along all shield walls, at gate or labyrinth entrances to cubicles projected to become High Radiation Areas, and along boundaries where the prescribed FSAR dose rate !

changes. The RBP survey locations were labeled. with sequentially numbered 11" X 14" signs to aid survey personnel and ensure sampling reproducibility.

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-4 / 'A training program was developed and administered to all: survey. .

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. personnel. - This' training program outlined survey. methodology, -!

documentationi requirements, ALARA considerations,. expected -

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survey instrument' responses to the containment subatmospheric

-environment and- nitrogen-16 gamma fields, . containment subatmospheric entrance / egress. procedures, . and biological shield survey. experiences at other nuclear power plants.

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A. mock survey was conducted in containment prior to initial criticality and the - drawing of a containment vacuum. This survey was performed in 'BioPak-60 units to simulate realistic #

survey conditions. Special attention was given to the movable incore detector regions of containment and the overexposure hazards associated with this " Extra High. Radiation Area." - This i sock survey was used to develop a survey man-hour estimate which would be used - to develop a man-rem - estimate for the surveys done at power. In addition,; the- water jugs used-.for the neutron spectrum- factor ' determination were 'placed in

~ containment prior to initial criticality and .the establishment L of containment vacuum. This was done in a further attempt to - , ;

maintain personne1' exposure ALARA L and. to lower the number of personnel entries required into the containment subatmospheric environment.

l 4 In addition to the general surveys conducted at the 378 RBPs, '

an extensive radiation, monitor /TLD/ survey meter comparison survey was conducted on two. containment radiation monitors.

One survey was conducted on 3RMS-RE32 at 90 percent reactor power, and the other on 3RMS-RE35 at 100 percent, reactor power.

The survey consisted of comparing extrapolated gamma TLD results and various survey: meter readings with the plant I radiation monitoring system computer readout information.

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In addition, a surveyJ eeter/ installed radiation area monitor -

ecomparison survey was 'also conducted on: 11 area monitors located in the Auxiliary, Waste Disposal,' and Fuel Buildings; The surv3y consisted of- simultaneously exposing the installed -

area monitor and selected survey instruments to'a Cs-137 source '

and comparing the various readouts.

Experiments were performed at the University of Lowell in order ,

to determine station survey instrument, TLD, and pocket ionization chamber response to the highly energetic nitrogen-16 gamma radiation. Experiments were' also performed at the station to study survey instrument response to subatmospheric conditions. Since both of these conditions exist -in containment, it was ' desirable to' determine which instruments responded in the' most accurate and reliable manner. Neutron

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survey meters were sent to the University of ' Michigan . for analysis and_ calibration using a heavy-water moderated Cf-252 j O.

1 source.

. 3 The following installed process monitor readings were compared to grab' sample results. This was done to determine the accuracy of the installed process monitors.

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CHS69 RCS Gross Activity / Specific Nuclide Monitor !

HVQ4,9 ESF Building Ventilation Monitor HVR108 Ventilation Stack Monitor LWS70 Radioactive Liquid Waste Monitor ARC 21 Steam Jet Air Ejector Monitor CMS 22 Containment Atmosphere Monitor ' !

DAS50 Turbine Building Sump Monitor HVC16 Control Building Ventilation Inlet Monitor i i

These process monitors do not represent all process monitors-but represent monitors that sample important plant processes, and/or are required by Plant Technical Specifications. !

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Page 238 p.

I RESULTS

'1. Shield Surveys -

A. Zero Percent Power The inside containment portions were con' ducted on 12-13-85. The outside containment portions were conducted on 12-23-85 and 12-30-85. All surveys were conducted prior to initial criticality and were intended to' verify no sources of radiation were-present that would affect subsequent surveys. There were no abnormal findings.

B. 30 Percent Power This portion was conducted on 02-15-86. This survey indicated steam generator loop general area radiation levels of up to 2.6 R/hr (gamma). Contact readings on the RCS loop crossover lines (coolant line connecting reactor coolant pump to steam generator cold leg) read. between 7.9 to 93 R/hr (gamma). No kJ appreciable neutron dose rates in these areas were observed. In the loop areas on the 24'6" elevation of the containment, readings were ,700 to 800_ mR/hr.

(gamma). These rates were consistent between loop areas on this elevation. Surveys of the -11'3" elevation of the containment produced readings of 1800 mrem /hr (neutron).

A neutron radiation area was discovered outside the containment equipment hatch on top of the Hydrogen Recombiner Building. This area was posted and levels never exceeded the 15 mrem /hr neutron limits of the FSAR. Also, an additional radiation area was discovered on the 43'6" elevation of the auxiliary building. This was determined to have been caused by radiation streaming through a penetration in the volume control tank shield wall. Other than these O

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V two items,~th'e results of the 30 percent survey were as expected.- -

C. 50 Percent Power This portion was conducted on 03-17-86. This survey indicated steam generator loop general area readings of . up to 5.0 R/hr (gamma). Contact readings on the RCS loop crossover lines read between 14.0 to 18.0 R/hr (gamma). No appreciable neutron dose rates in these areas were observed. In the' loop areas on the 24'6" elevation of the containment, readings were approximately 2 R/hr. Again, the readings between loops were very consistent. Surveys of the -11'3" elevation of the containment produced readings of 500-mrem /hr (neutron). This .same survey location at

- 30 percent reactor power indicated 1800 mrem /hr neutron. It appears that the neutron reading taken

,- at 50 percent power was not in the exact location as

-Q the survey point taken at 30 percent power. It should be noted that at 100 percent power the surveyor, - while approaching this survey location, detected neutron levels exceeding 1000 mrem /hr.

All survey readings were within the levels discussed in the Millstone Unit 3 FSAR.

D. 75 Percent Power i This portion was conducted on 03-26-86. At the time of the survey, the containment personnel air lock inner door was inoperable making the containment inaccessible. Only the points outside the containment were surveyed. All survey points were .

I within specification except for point number 109 l which is located adjacent to 3CHS-RE69 (failed fuel monitor) on the 4'6" elevation of the auxiliary building. Upon evaluation, the larger than expected l

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(/) ~ dose rate, was the result of the - letdown piping on 3CHS-RE69 and not due to a deficiency in adjacent j

zhield walls. .l 1

E. 90 Percent Power This portion was conducted on 04-15-80. Due to ALARA !

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concerns, the containment survey points were eliminated from this power level. The 90 percent I radiation values were considered redundant to the values scheduled to be taken at 100 percent power. No new problems were encountered during the out of containment portion of the survey.

F. 100 Percent Power This portion was conducted on 04-18-86. For ALARA considerations and because previous readings between loops had been similar, only one loop area on the 24'6" elevation of containment was surveyed. General g area readings of between 7 to 10 R/hr (gamma) were C/ '

measured. Two' loops on the 3'8" elevation were surveyed from 10' outside the loop area using a teletector and readings of 30 R/hr (gamma) were observed. From this 10' approach distance to the loops at elevation 3'8", no appreciable neutron dose rates were observed. Neutron radiation levels on

-11'3" elevation were measured in excess of 1000 mrem /hr. No further neutron rad level quantification was attempted at the -11'3" evaluation in order to minimize exposure to the survey personnel. Outside containment, five survey points were determined to be in excess of the FSAR established limits. In each case, these discrepancies were the result of adjacent component piping and not deficiencies in shielding.

At the 100 percent plateau, two monitors were alarming because the actual normal exceeded the expected normal and setpoints for these monitors were revised.

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- 2. Installed Area Radiation Monitor Evaluation The permanently installed area radiation monitoring system was. evaluated at the 90 and 100 percent power plateaus.

. This evaluation was conducted during -the period -from 04-17-86 to 04-30-86. This evaluation was done to verify ~

the response of the area radiation monitors at other than very low levels of radiation. This evaluation, plus comparison of - containment area monitors at 100 percent power, indicated a good correlation between radiation monitor readings and survey meter readings.

3. Installed Process Radiation Monitor Evaluation The comparison of process radiation monitor readings to survey results indicates that process monitors show accurate radiation trends, but are not all accurate in determining the absolute value of radiation in the '

process. Monitors that require accuracy do provide :

- q. accurate readings. r X/, - 4. Neutron Spectrum Factor Determination TLDs used to determine the f.eutron spectrum factors in containment have been removed and data reduction is in progress. The results of this analysis will be utilized to enhance the Unit' 3 neutron dosimetry program through

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the determination of accurate quality factors.

5. Conclusion This test verified that radiation , levels in the plant are as stated in the FSAR with the exception of a radiation area caused by _the letdown piping to the failed fuel monitor, 3CHS-RE69. A Plant Modification Request has been-submitted to provide permanent shielding of' the letdown

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piping, and additional shielding is being installed in various identified areas to keep. exposure ALARA. }

Comparison of area radiation monitor readings to survey :

s results shows' that the' area monitors provide a good l indication of radiation levels. Some process radiation fa

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Page 242

.x 4 L N4 monitor results were not as accurate. They do, however, -

provide a good indication of trends in the monitored

- process. Significant is the fact that the liquid waste discharge monitor, 3LWS-RE70, the failed fuel monitor, 3CHS-RE69, and the containment atmosphere monitor, 3 CMS-RE22 do provide accurate radiation levels.

Investigation is continuing on other process monitors to provide more accurate source term calibration correlatable to field results.

Approximately 1.3 man-rem and 260 man-hours were expended in performing the Reactor Power Shield Survey. An ALARA review of the job estimated that 3.795 man-rem would be expended for the entire survey. Because observed dose rates were lower than expected, and survey points were deleted at various power plateaus, less exposure was-received than originally predicted.

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. m j 8.5.4 VENTILATION. SYSTEM OPERABILITY 3-INT-8000, Appendix 8008 - '

OBJECTIVE The objectives of the test were to: ,

1. Veri fy that the' containment air ventilation systems (containment air recirculation ~ system and CRDM cooling systems) are capable of maintaining the containment air' temperature less than the EEQ equipment design limit of 90'F.

2. Verify that the Main Steam Valve Building (MSVB) ventilation system can maintain the MSV8 within the EEQ equipment design range of 50*F to 104'F.

The acceptance criteria for the test was to verify that the containment air ventilation systems maintain containment air

temperature within the Technical Specification range of 80'F to O

120*F. $

DISCUSSION Temperature data for the containment was monitored using 41 ;

permanent RTDs. located throughout the containment structure.

In addition, the reactor plant chilled water (CDS) temperature to the containment air coolers were monitored as well as containment pressure, outside ambient air temperature, and reactor power level. Data was taken at 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months intervals during power ascension testing. !

Temperature data for the MSVB was monitored using 5 permanent RTDs located at various levels -in the structure. In addition, ,

outside ambient air temperature and reactor power level were g also monitored. Data was taken at 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months intervals throughout< ;

power ascension testing.

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. ,3 RESULTS .

At .the 100 percent power level, all upper elevation areas in the containment exceeded the EEQ equipment design temperature of 90*F by .an average of 15*F. However, the Technical Specification upper temperature limit of 120*F (TS 3.6.1.5) was l satisfactorily met at all power levels. l In the MSVB, the area between the main steam isolation valves -

exceeded the upper EEQ equipment design temperat6re of 104*F by an average of 3*F. All other building areas were maintained within the required limits.

Temperature excursions similar to the above were noted during .l precore hot functional testing. At that time, plant deficiency UNS 6300 was written to cover the containment excursion and - I UNS 6452 was written to cover the MSVB. These - prior -

deficiencies were considered enveloping for the power ascension temperature deviations and no new deficiencies were generated.

These deficiencies, while not affecting equipment operability, .,

are being reviewed by Engineering to assess the impact on EEQ '

qualified life of various equipment in the noted areas. In addition, per the requirements of the Facility Operation License, Section 2.C.3, Millstone 3 must, prior to startup following the first refueling, recalculate the qualified service lives of all applicable components located in the containment. These calculations are to be based on actual temperature readings over the first fuel cycle.

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Page 246 I

1 RCS CHEMISTRY ATTRIBUTE SPECIFICATION LIMIT I pH 4.2-10.2 l Conductivity - N/A-Expected range:

1.0-40.0 uMhos/cm Dissolved Oxygen s100 ppb

Chlorine s150 ppb Fluoride s150 ppb Dissolved Hydrogen 25-50 cc/kg water Lithium 0.2-2.2 ppm as Li

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Boron 0-4000 ppm Sillca s1000 ppb Aluminum ISO ppb Calcium & Magnesium 150 ppb Magnesium (25 ppb Specific Activity (D.E.1-131) s1.0 uC1/gm Gross Activity As required by procedure O

Millstone

""j;jst ua RCS CHEMISTRY LIMITS eIs*j'_,

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POWER LEVEL (%) .3.0. jliQ. l 1023 1706

' POWER LEVEL (MWT) l SAMPLE DATE 02-16-86 03-18-86

. SAMPLE TIME 1648 0850 1 l

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1 ANALYSIS RESULTS UNIIS l I

pH/ temperature 5.95/26.1 6.26/24.0 pH/t conductivity / temperature 25.8/25.5 21.5/24.0 uMhos/cm/t !

Dissolved oxygen <5.0 < 5.0 ppb

, Chloride <10 (I ppb Fluoride <20 <1 ppb Dissolved Hydrogen 40 36 cc/kg Lithiom 1.6 1.73 ppm Baron 1297 1201 ppm S111ce 423 450 ppb Aluminum 14.4 21.0 ppb Calcium + Magnesium 1.5 <1 ppb Magnesium <1 <1 ppb D.E. I- 131 1.94E-04 2.66E-04 uC1/gm Gross Activity 3.14E-02 5.63E-03 uCi/gm

o klear Power Stalla Unit No. 3 RCS CHEMISTRY ANALYSIS DATA m.

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POWER LEVEL (2) 21 .l.0.0.

POWER LEVEL (MWT) 2558 3411 SAMPLE DATE 03-27-86 04-19-86 SAMPLE TIME 0840 0900 ANALYSIS RESULTS UNITS pH/ temperature 6.53/26.0 6.48/26.9 pH/t conductivity / temperature 23.1/26.0 22.2/25.0 uMhos/cm/t Dissolved oxygen <5 <5 ppb Chloride <10 (10 ppb j Fluoride <20 (20 ppb l Dissolved Hydrogen 35.5 43.5 cc/kg l Lithium 2.02 1.98 ppm j l Boron 1133 1076 ppm

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Silice 388 335 ppb j Aluminum 25.0 8.0 ppb i Calcium + Magnesium 1.52 10.9 ppb Magnesium 0.47 2.8 ppb D.E. I- 131 6.79E-04 8.38E-04 uC1/gm j Gross Activity 1.25E-01 1.579E-01 uC1/gm h '.

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8.5.61 NEUTRON SHIELD TANK COOLING TEST-3-INT-8000,~ Appendix 8010 l l

OBJECTIVE The objective, of this test was to verify that .the Neutron Shield: Tank . Cooling System performs within design limitations

. . at 100 percent' power. The shield tank consists of. an annular tank surrounding the reactor' vessel. Its purpose is to serve -

as neutron shielding to-adjacent areas of the containment structure. Cooling water in the tank circulates under natural convection from the ' tank 'to the neutron shield tank cooler where it is cooled with ' water L from the reactor plant chilled water system. In addition to the shielding function, the tank serves as the support structure-for the reactor vessel.

DISCUSSION The test was performed on .02-16-86, 03-17-86, 03-26-86, 04-15-86, and 04-18-86 at plant power levels of 30, 50, 75, 90, and 100 percent, respectively. The temperature of the neutron shield tank was monitored and recorded at each power plateau during the power ascension. The shield tank outlet temperature (inlet to the neutron shield tank - cooler) and the neutron ,

shield tank return water temperature (outlet -from the neutron shield tank cooler) was recorded at each power level and compared against the acceptance criteria.

RESULTS All data obtained met the acceptance criteria which required .3 that the tank temperature be maintained less than 135*F at all power -levels. The highest neutron shield tank temperature recorded during the test was 123*F.

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Page 250 8.5.7 . CONTAINMENT, PENETRATION TEMPERATURE-MONITORING 3-INT-8000, Appendix 8011 -

08JECTIVE The purpose of this test was to verify that.the hot _ containment ;

5 piping penetrations were within design temperature during power )

~ '

4 ascension and .at full reactor power. The penetration coolers !

consist of liquid cooled annular structures surrounding selected hot containment piping penetrations. ' They form an integral _ portion of the piping penetrations and run the entire *

.i depth of the containment structure. The coolers are supplied y cooling water from the reactor plant component cooling water system. Liquid cooled penetrations are used on the main steam, 3 feedwater, RCS letdown, steam generator blowdown and steam supply lines to the turbine driven auxiliary feedwater system.-

DISCUSSION-Vf The test was performed on- 02-17-86, 03-17-86, 03-27-86, j 04-15-86 and 04-18-86 at' plant power levels of 30, 50,-75, 90, and- 100 percent, respectively. With the reactor- plant j component cooling flow at a minimum to the penetration coolers, the containment concrete temperature adjacent to the ,

penetration was measured. Data ' was obtained at four points (90" apart) on each' penetration.

4 RESULTS All data met the acceptance criteria which required- all temperatures to. be less than 150 F. Actual temperatures were ,

. between 58*F and 140 F. l l

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Paga 251 f~y V '8.5.8 TURBINE PLANT COMPONENT COOLING WATER SYSTEM BALANCING 3-INT-8000, Appendix 8019 -

~

OBJECTIVE The ol'jective of this test was to verify adequate flow balancing of the turbine plant component cooling water system (CCS) at 100 percent power.

DISCUSSION The test was conducted as plant conditions permitted over the period from 02-08-86 to 05-06-86. The CCS flow rates to system .

i heat exchangers were initially adjusted as part of the 1 preoperational test program. These flows were then modified in response to increased turbine plant heat loads, at 30 and 100 percent power. Final flow modifications were completed at 100 l percent power and the final throttle valve positions were i q recorded in the test appendix for future reference. Flows were V, verified to be adequate by monitoring temperatures and flows at various system locations using permanently installed and temporary instrumentation.

RESULTS The objective of this test was satisfied. Adequate cooling water flow was verified to all CCS heat loads.

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.M, jj 8.5.9- ~ PIPINGFLUIDTRANSIENT.VIBRNTIONHONITORING e

3-INT-8000,-Appendix 8029 d

'0BJECTIVE The objective of this test ~was to verify, by visual inspection and instrumented measurement, the vibrational response of plant' p.iping systems during selected fluid transient events that are credible within plant operating modes. ,

DISCUSSION The test was conducted over the period of _04-21-86 to 04-24-86. ,

The transients selected for this test were:

!- 1. Main turbine trip

2. Closure of the feedwater isolation valves o During each transient event, qualified test personnel observed

- the.- response of piping and associated supports. In addition, '!

temporary test instrumentation was installed at selected pipe supports. ;

RESULTS

, All test acceptance criteria for the main turbine trip' and

, feedwater isolation valve closure transients were met. No' permanent deformation or damage was observed.

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li Page 253

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, '8.5.10 THERMAL EXPANSION AND RESTRAINT MONITORING i

3-INT-8000, Appendix 8034 -

OBJECTIVE The objective of this test wasito verify, by visual inspection and instrumented measurement, that the feedwater and main steam piping systems are free tu thermally expand as designed.

DISCUSSION This test was conducted over the period of 02-03-86 through 04-21-86. The inspections were performed at plateaus of zero,.

30, 50, 75, and 100 percent power levels. Test data which was collected by visual inspection, system walkdowns and instrumented measurement, was compared to design ranges.

Discrepancies (piping interferences or snubber indication out of design range) were evaluated and resolved by Engineering.

, RESULTS All potential contact of piping with structures, components and conduit was evaluated by Engineering. This evaluation noted no potential interference which could restrict piping or components from expanding. Furthermore, all data points outside of the predetermined acceptance criteria were evaluated and found to be acceptable by Engineering.

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'8.5.11 : LOOSE PARTS MONITORING 3-INT-8000,' Appendix 8035 l

. -0BJECTIVE ;

The objectives of'this test were to:

1. Obtain baseline system signal., data during the power ascension test phase.

~

2. Obtain baseline system signal data with the. plant at full power.

3. Determine the ~ approximate frequency of spuriousi alarms.

DISCUSSION performed on 02-16-86, 03-17-86, 03-26-86, The test was 04-16-86, and 04-18-86 with the plant power at levels of 30, 50 75, 90, and 100. percent, respectively.

Baseline. signal data was obtained by using a spectrum analyzer which:was connected to the auxiliary output jack on the Loose Parts Monitoring ' system (LPM) cabinet. Hardcopy spectrum analysis data 'was obiained for all eight monitoring channels during the testing plateaus. The frequency of _ spurious alarms caused by the noise of' normal plant operation was also ,

i. monitored.

The LPM was -supplied by Rockwell and consists of a monitoring cabinet with audio output system and integral cassette recorder.

There are eight ' accelerometers located on the primary system: two located on the reactor vessel head, two located on the lower reactor vessel and one on each steam generator in _ the channel head area. The. system has. been ,

modified by the addition of a 1500 hertz bandpass filter to enhance the capabilities to detect loose parts of a large mass-(30 pounds).

Page 255 JA

< $v), . ' RESULTS All baseline LPM signal data was obtained with no problems encountered. The frequency of spurious alarms was approximately three per day. In accordance with Engineering direction provided .following the phase five testing -(see Section 5.11), the gains of the 1500 hertz filter were' adjusted for the upper and lower reactor vessel LPM channels to reduce the number of spurious alarms. No adjustments were required on the remaining channels. The alert-levels for power ascension and initial commercial operation were determined to be between-0.1 to 0.38 ft-lbs for a 30 pound object impacting 3 feet from the tra.1sducer and between 0.01 and 0.08 ft-lbs for a 0.25-pound abject impacting 3 feet from a transducer. Additional-testing indicated that the alert levels may -need to be increased further to obtain a false. alarm rate'of approximately one per day. It is -anticipated that any further adjustments

. will result in alert levels no greater than 0.5 ft-lb kinetic energy, 3 feet away from a transducer.

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.Page 256 9.0 WARRANTY RUN TEST SUMARY 3-INT-9000 -

This test proved the reliability of the NSSS system. The plant was maintained at rated power for 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months . Appropriate data was recorded to allow plant performance to be analyzed. The warranty run was conducted from 04-25-86-to 04-29-86.

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- 3-I'NT-9000, Appendix 9001 :

s OBJECTIVE

- The. objective of this test was to determine plant thermal power-by .means of- the' plant. process computer- calorimetric c -calculation, plant process computer data collection with manual calculation, and manual data collection with ' manual calculation. These cal'culated values were used.hs--input to the readjustment of the power range (PR) instrumentation.

DISCUSSION The test was conducted at 100 percent power. Once stable plant conditions ' were established, l data was collected : on selected plant parameters. In each case, data was taken for 15 minutes

- at 5 minute intervals. This data was then reduced and --the '

. plant power level calculated.

RESULTS The results of this test are as follows:

- Plant Process Computer Calorimetric' Calculation 100.5%

Plant Process Computer Data Collection with manual data reduction 100.1%

Manual Data Collection with Manual Calculation 100.2%

In each case the calculated power levels compared favorably with the power range' instrumentation. All objectives of tilis test were met. There was no formal acceptance criteria for this test.

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9.2. SECONDARY PLANT PERFORMANCE E 3-INT-9000, Appendix 9002 - l OBJECTIVES The objectives of the test were to:

1. 'Obtain performance data needed to properly compare actual performance to General Electric Company -(GE) warranty values for the turbine generator.

2. Acquire baseline operating' data at rated power for routine '

monitoring and reporting requirements. *

, 3. Estimate the loss of efficiency associated with operating the turbine in the full arc ' steam admission mode.

DISCUSSION

.This test was performed over the period 04-19-86 to 04-29-86 with the unit operation at a 100 percent' power level. The test j procedure was prepared using the ' ANSI /ASME PTC-6 Steam Turbine

/, Performance Test Code for guidance. Prior to performing the i test, an uncertainty analysis on all heat rate inputs was

(: .

performed. Heat rate uncertainty was determined to -be approximately 0.7 . percent. Overall test uncertainty was I '

calculated at less than 1.0 percent.

1 Test prerequisites required calibration checks . of. -selected plant instrumentation within 30- days of testing. During testing, steam generator blowdown was isolated and ainxiliary"

! steam was supplied by the auxiliary boiler. The test procedure i

required inventory losses of less than 0.25 percent of valve:

wide open.(WO) main turbine throttle flow. In addition, cycle component alignment was . verified and a systematic isolation =

check was completed within two hours of testing. ,

. . 1 Each test point required four hours .of data acquisition. ,The )

first two hours were taken to verify steady-state operation. . j The plant process. computer provided most data acquisition needs 1

.:. l , - . . . . . -- - .. - - .. .. A

x Pag 2 259 x

with very. limited local data taking required. _ Duplicate test runs were conducted with turbine control valve positions upset between tests.

Corrected test heat rates from duplicate tests were compared according to ASME PTC-6 which requires agreement of parallel runs within 0.25 percent.

RESULTS Turbine generator net turbine heat rate -l (NTHR) exceeds the warranty value by approximately 0.1 percent (12 Btu /kWh) at the warranty point. Refer to Figure 9.2-1, Specified Heat Rate Warranty Curve, for comparison. Overall test uncertainty is approximately 0.75 percent. Per the ANSI /ASME PTC-6 Steam Turbine Performance Test Code, verificatiori of NTHR also verifies that warranteed electrical load has been achieved.

The mass flow warranty value was verified from valve wide open test results.

Corrected test values obtained during the Initial Performance Test at 100 percent rated power (3411 MWTH) with steam generator blowdown isolated and auxiliary steam load supplied by the auxiliary boiler were:

Gross Generator Load 1203.9 MW

, , E Station Service Load 47.7 MW E

Net Turbine Heat Rate 9707 Btu /kWh Valve Wide Open Volumetric Flow 2 igg; Ft3/5 l

Note: During normal plant operation, gross generator load )

will be lower and NTHR higher by approximately 0.5 )

percent since steam generator blowdown will be in operation with auxiliary steam load supplied from the

. main steam system.

Meam Generator Power 1 Net Turbine Heat Rate = Gross Generator Load 2 Indicates turbine is passing approximately 4.0 percent excess flow

(

Page 260 I

The .' test NTHR and. gross generator load exceeded predicted full are admission target values by 0.75 to 1.0- percent. The full -

I arc target values for NTHR and gross generator load at 100 h percent rated power are 9785 Btu /kWh and 1194.3 MWE '

respectively. Refer to Figure 9.1-2, Full Arc Specified Heat Rate Curve.

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Nucteer Power Station 9.2-1 unn w..s SECONDARY PLANT PERFORMANCE TESTING

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Page 263 APPENDIX A J FINAL SAFETY ANALYSIS REPORT TESTING DEVIATIONS

. l

. Introduction FSAR Chapter 14 details testing and operational commitments from Initial Inspection and Component Testing through Warranty Run. During the l 4 Startup Program, certain aspects of test procedures and performance i

deviated from FSAR Chapter 14 as stated. These deviations were 1

documented and approved by the use of Quality Assura'nce forms and l procedures relating to FSAR Changes. As such, the changes were reviewed l by the site Plant Operations Review Committee for unreviewed safety ,

question significance.

Preoperational/ Acceptance Test Deviations

1. Boron Thermal Regeneration System (BTRS) Testina - This FSAR Change allowed BTRS. testing to be performed after completion of HFT as plant conditions permit due to lack of system availability as well i as the fact that BTRS is not covered by Technical Specifications nor {

is it a safety-related system. l

2. Spent Fuel Pool Coolina (SFC) System Testina -

This FSAR Change l

allowed SFC testing to be performed after completion of HFT as plant conditions permit due to lack of system availability as well as the fact that SFC is not covered by Technical Specifications nor were were the untested portions of the system safety-related.

3. Control Rod Drive Mechanism (CROM) Testina -

This FSAR Change deleted CRDM testing at hot standby conditions because j equivalent /more limiting testing was performed during cold shutdown conditions.

4. Rod Drop Testina This FSAR Change deleted hot, no-flow rod drop time testing because equivalent /more limiting testing was performed during cold full-flow conditions.

S' Rod Drop Testina - This FSAR Change administrative 1y took exception i

to the RG 1.68 requirement to perform hot no-flow rod drop testing '

deleted in (4) above.

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6. Main Feedwater Testing - This FSAR Change allowed certain transient Feedwater system ' testing to be performed post-HF.T during Power !

Ascension when plant conditions were better able to support testing.

Startup Test Deviations

1. Natural Circulation Testing -

This FSAR Change eliminated some specifjc natural circulation testing requirements which were incorrectly identified for performance during Post-Core Hot ~

~

Functional Test.

2. Shutdown From Outside The Control Room Test -

This FSAR Change allowed credit to be taken for the required Cold Shutdown demonstration as part of the Shutdown from Outside the Control Room Test because of equivalent testing performed previously.

3. Loss of Power Test - This FSAR Change deleted a prerequisite for the Station Blackout test which required all plant loads to be supplied from the turbine generator because it allowed greater test flexibility and the fact that equivalent turbine generator testing would be performed during the 100 percent Power Trip.

Pseudo Ejected Rod Test

4. -

This FSAR Change deleted the Pseudo

. Ejected Rod Test at 30 percent power because of the excessive flux tilt it would have caused, credit taken for like testing at other I '

similar design plants, and previous similar testing performed at zero percent power.

5. Pseudo Ejected Rod Test -

This FSAR Change administrative 1y took exception to the RG 1.68 requirement to perform a Pseudo Ejected Rod i Test at greater than 10 percent power which was deleted in (4) above.

6. 50 Percent Reactor Trip - This FSAR Change deleted the requirement to perform a 50 percent Power Reactor Trip and substituted a 10 percent Load Swing for the following reasons: ,

i. a. There was no regulatory - requirement to perform a 50 percent trip.

b. The NSSS supplier deleted the requirement to perform a rod drop / negative rate' trip test at 50 percent power.

.. . ._ , _ .m.. ., ... . . _.__.,

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6 b' r e' y..- _ . ,.

, . APPENDIX'A- - Page 265 ,

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c. The NRC requested performance .of -a 10 percent Load Swing at' i

50 percent power.

j

- d. The -plant challenge involved was significantly less.

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APPENDIX B Page 266 STARTUP TEST PROCEDURE LISTING STARTUP-PROCEDURE NUMBER TITLE REPORT SECTION 3-INT-4000 Initial Fuel Load -

4.0 Appendix 4003( ) Core Load Instruments and Neutron Source Requirements 4.0 Appendix 4004 Inverse Count Rate Ration Monitoring 4.0 Appendix 4005- Initial Core Loading 4.0 Appendix 4006 Core Map 4.0 3-INT-5000 Postcore Hot Functional Test 5.0 Appendix 5001 Shutdown Margin 5.1 Appendix 5002 TC/RTD Testing (Incore TCs-RCS RTDs) 5.2 Appendix 5004. Rod Control Slave Cycler /CRDM Timing . 5. 3 Appendix 5006 RCS Leak Detection 5.4 Appendix 5007 Pressurizer Heaters and Spray 5.5 Appendix 5008 Rod Drop Testing 5.6 Appendix 5009 RCS Flow Measurement 5.7 Appendix 5010 RTD Bypass Loop Verification 5.8 Appendix 5011 Movable Incore Detectors 5.9 Appendix 5015 Digital Rod Position Indication 5.10 Appendix 5016' Loose Parts Monitorin~g 5.11 Appendix 5017 RCS Flow Coastdown 5.12 Appendix 5018 Rod Control 5.13 Appendix 5031 Chemical and Volume Control System 5.14 Appendix 5033 RCS Loop Stop Valve / Pump Interlocks 5.15 3-INT-6000 Initial Criticality 6.0 Appendix 6001 Inverse Count Rate 6.0 3-INT-7000 Low Power Physics Testing 7.0 Appendix 7001 HZP Testing Range Determination 7.1 Appendix 7002 Reactivity Computer Checkout 7.2 Appendix 7003 Boron Endpnint 7.3 Appendix 7004 Isothermal Temperature Coefficient 7.4 n Appendix 7005 RCCA or Bank Worth Measurement 7.5

~

Appendix 7006 Natural Circulation (Low Power) 7.8' !

(1) Some appendices were deleted prior to performance and remaining appendices were not renumbered. Therefore, some numbers were not listed.

-D \

l APPENDIX B Paga 267

. w STARTUP

~

PROCEDURE NUMBER TITLE REPORT SECTION

3-INT-8000 Power Ascension Testing -

8.0 Appendix 8001 Calorimetric 8.5.1

. Appendix 8002 Operational Alignment of Nuclear Instrumentation 8.2.1 Appendix 8003 Calibration of Steam and Feedwater Flow 8.2.3

Appendix 8004 Operational Alignment of Process Temperature Instrumentation 8.2.2 .

~

Appendix 8005 Reactor and Turbine Control 8.3.1 Appendix 8006 Secondary Plant Performance. 8.5.2 Appendix 8007 Radiation Survey and Process Radiation 8.5.3 Appendix 8008 Ventilation System Operability 8.5.4 Appendix 8009 Chemistry and Radio Chemistry 8.5.5' Appendix 8010 Neutron Shield Tank Cooling 8.5.6 Appendix 8011 Containment Penetration Temperature Monitoring 8.5.7 Appendix 8013 Steam Dump Control 8.3.2 O+ Appendix 8015 RCS Flow Measurement 8.1.1 1 Appendix 8016 Turbine Overspeed 8.4.1 -

Appendix 8017 Automatic Reactor Control 8.3.3 Appendix 8018 Automatic Steam Generator Level Control 8.3.4 Appendix 8019 Turbine Plant Component Cooling System

, Balancing 8.5.8 L Appendix 8020 Power Coefficient 8.1.2 Appendix 8022 10 Percent Load Swing 8.4.2 Appendix 8023 Reactor Trip and Shutdown From Outside the Control Building 8.4.3 Appendix 8026 Large Load Reduction 8.4.4 Appendix 8028 Axial Flux Difference Instrumentation Calibration 8.2.4 Appendix 8029 Pipe Fluid Transient Vibration, Testing 8.5.9 Appendix 8030 Loss of Power (20 Percent) 8.4.5 Appendix 8031 Reactor Coola'nt System Boron Measurement 8.1.3 Appendix 8032 Generator Trip (100 Percent) 8.4.6 4 l

7, , .

c, APPENDIX B Pagn 268 4

STARTUP TITLE

~

PROCEDURE NUMBER REPORT SECTION Appendix 8034 . Thermal Expansion and Restraint - 8.5.10

. Appendix 8035 . Loose Parts Monitoring 8.5.11 Appendix 8037 . Main Steam Line Isolation Valve Closure 8.3.5 3-INT-9000 Warranty Run. 9.0 Appendix 9001 Calorimetric 9.1 Appendix 9002 Secondary Plant Performance ,

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Pag 2 269

,m APPENDIX C PREOPERATIONAL TESTS COMPLETED DURING THE STARTUP TEST PROGRAM )

The following preoperational tests were completed during the start 9p test program. The individual tests were completed consistent with Technical Specification system operability requirements. )

Test Number Title Date Completed 3307AP001 Low Pressure Safety Injection 12-07-85 3308-P002 High Pressure Safety Injection 12-06-85 3309-P001 Quench Spray 12-30-85 3311CP Post Accident Sampling 01-29-86 3312CP Containment Atmospheric Monitoring 01-12-85 33130P Containment Filtration 03-05-86 3313FP (Rev 1) Containment Vacuum 12-31-85 3314BP Fuel and Waste Disposal Building HVAC 03-03-86 33140P ESF Building HVAC .

12-06-85 3314FP Control Building HVAC 12-19-85 3314IP Supplemental Leak Collection and Release 12-31-85 3315BA (Rev 1) Main Steam Valve Building HVAC (Retest) 01-29-86 3317-A Moisture Separator Reheater 02-03-86 3319CP001 Condensate Polishing 03-24-86 3320-P Feedwater Heater Drains and Vents 01-11-86 3322-P Auxiliary Feedwater 12-16-85 33240A Stator Cooling 01-30-86

. 33250A Condenser Tube Cleaning 04-05-86 3330AP Reactor Plant Component Cooling Water 01-03-86

~ Q[

APPENDIX C Page 270 ,

4

' Test Number Title Date Completed n .

3330CP Reactor Plant Chilled Water ~

11-26-85 33310A Hot Water Heating / Preheating 11-25-85 3335BP Radioactive Liquid Waste 02-23-86 3335CP Boron Recovery 01-27-86 ,

3337-P' Radioactive Gaseous Waste 01-27-86

~

3341BP'(Rev 1) Fire Protection-Halon (Retest) 11-25-85 3341CP Fire Protection-C0 2 12-30-85 3345CP006 Battery Duty Cycle-Testing 01-10-86 3 3404-P Digital Radiation Monitoring 11-18-85 341Wf Reactor Vessel Level 12-31-85 3720BP (Rev 1) Station Emergency Lighting (Retest) 12-20-85 3999-P Pipe / Pipe Support Steady-State Vibration 02-04-86

, 3-INT-2001 Appendix P5 l

(Rev 1) Secondary Plant Performance 12-21-85 3-INT-2001-Appendix R10 Incore (Power Distribution) 01-12-86 i

3-INT-2001 Appendix R11 Estimated Critical Position 04-21-86 3-INT-2001 Appendix R12 Shutdown Margin 01-06-86 3-INT-2007 ISI Valve Stroke Time Testing 01-09-86 O.

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APPENDIX C Page 271 h

.Q '

The following_ preoperational"tes'ts were completed after the startup test program was completed..

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- Test Number Title Date Completed 3721-A001'- Electrical Distribution - Security 05-22-86

. 3721-A002 Integrated System Test - Security 05-30-86 3-INT-2001 Computer Programs Test 05-23-86 3-INT _2008 Efficiency Testing of Air Filtration Units 07-18-86 The following preoperational -tests are yet to be completed. Provided is a summary of test status and plan for test comp 1etion.

Test Number Title 33040P_ Boron Thermal Regeneration N The preoperational test has not yet-been begun due to equipment problems. .

The system is currently isolated and not required for plant operation.

. Testing will be completed in accordance with plant requirements' but no {

1ater than startup fn110 wing the first. refueling outage. As this test is '

referenced in Chapter 14 of the Millstone 3 FSAR, a proposed revision to the FSAR has been submitted to permit performance'of the test as dictated by plant requirements.

3305-P Spent Fuel Pool Cooling and Purification The safety-related portion of the system was satisfactorily tested as a

- prerequisite to receiving nuclear fuel. The remaining (non-safety) portions will be tested once the spent fuel pool is filled to support refueling and subsequent fuel storage activities. It is therefore anticipated the remaining testing will be completed prior to the first refueling outage. As this test is referenced in Chapter 14 of the

.- Millstone 3 FSAR, a proposed revision to the FSAR has been submitted to permit completion of the test as dictated by plant requirements.

'--e e -r-. , , w g e

APPENDIX C Page 272 i Test Number Title .

3311EA EEQ Area Temperature Monitoring System Physical testing is complete but the test procedure is being kept open while a revision to various EEQ area temperature alarm setpoints are made. The procedure will then be utilized to cover the system retest with the revised setpoints.

3319CP002 Condensate Liquid Waste l l

The test is partially complete.

Currently the system is not required to support plant operations. Plans are to complete the test in a manner l

consistent with plant operations requirements.

3328-A Chlorine During the startup of Millstone 3, the medium used for biological growth control in the service water system was switched from chlorine gas injection to sodium hypochlorite injection. The sodium hypochlorite system is presently in service and performing its intended function. The testing of the system will be completed consistent with plant requirements.

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Y A Page 273 j,

(f APPENDIX D SUMARY OF PROBLEMS ENCOUNTERED DURING -

THE LOSS OF POWER TEST (3-INT-8000, APPENDIX 8030)

PROBLEM COMMENTS / RESOLUTION T

1. CCP*PIB did not go Test logic was incorrect in that ,

from 0FF to ON during PIB was in pull-to-lock at the Loss of Power (LOP). time of LOP. PIC'was aligned to train B and was observed to function properly. A test change was . issued to correct this problem with the test procedure.

2. CHS*P3B did not go from Test logic was incorrect. P3A 0FF to ON during LOP. was running initially, tripped-on LOP and subsequently automatically restarted. A test change was issued to correct this problem with the test procedure.

3. FWA*A0V26 did not go from Plant deficiency UNS 7572 was OPEN to CLOSE during LOP. issued to document' this problem. l Plant maintenance personnel investigated and found a limit switch problem. Limit switch was-adjusted and retested satisfactorily.

4. HVK*CHLIB did not go from Plant deficiency UNS 7573 was OFF to ON during LOP. issued to document this problem.

Contrary to the problem I description, review of the 1

Sequence of Events (SOE) digital

[ printout indicate:

' APPENDIX D Page 274 -

1

..n(9 ~ PROBLEM COMENTS/ RESOLUTION HVK*CHL1B did not go from . 1. HVK*CHL1A which was running 0FF to ON during LOP. at the time of LOP, tripped (4. continued) on LOP.

2. Approximately 80 seconds after restoration of power, HVK*CHL1B automatically started. This is as per design.

3. Approximately 148 seconds after starting HVK*CHL1B tripped. The postulated cause is low Freon level.

4. Approximately 15 minutes after tripping on LOP, HVK*CHL1A, responding to operator action, started.

(ps' An automatic timer feature prevents the restart of a chiller for 15 minutes after a chiller is stopped.

Therefore, with the exception of the B chiller tripping, both chillers operated per design.

Regarding the B chiller trip, based on past operating history of these chillers, it is postulated the B chiller. tripped

, because of low Freon level.

Plant Maintenance personnel

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j APPENDIX D Pags 275' ;

l 1

\,

): PROBLEM COMMENTS / RESOLUTION i

recharged the Freon in the B '

chiller. The unit has performed satisfactorily since then. j

5. HVR*FN6B did not go from Per a change to the system 0FF to ON during LOP. operating procedure (OP 3314A),

the variable inlet vanes (VIV) on the fan must be placed in MANUAL at a 20% open position for the l fan to start automatically.

During LOP, VIV were in AUTO.

This was an improper system alignment. Plant Operations personnel .ere w advised of this O ~ and action was taken to ensure I proper system alignment in the .

future.

6. IAS-C2B did not go from Plant deficiency UNS 7574 was OFF to ON dur,ing LOP. issued to document this problem.

Plant Electrical Maintenance personnel investigated and.

determined the problem was caused by a. fault in an overload heater circuit which caused an inoperable control circuit.

After repair, retest under a simulated LOP condition was satisfactorily.

p. -

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1

n APPENDIX 0 Pag) 276 N PROBLEM COM4ENTS/ RESOLUTION l

. ,, . . .c ;.

7. 'SWP*MOV130B'did not go Error in test procedure. #

from CLOSED to'OPEN HVR*ACU1B was in pull-to-lock .so duririg LOP. no open signal.was sent .to. valve 2 ,

A test change.,was issued 'Ao correct this problem with the test procedure.

.u

8. SWP*P1A was not running Error in test , procedure < The beforesbafterLOP. -

procedure assumed the alternate <

e ~

Thid is c6n..,.,i t'rar to the pump on each SWPMrai.n.would be f.sstprocddu'r$. running. A test.. : cha,nge >was issued .,,tp,,.,corgect this administrative problem.

.n s s.s t ' or. . J& '

,_., -5e .

9. SWP*Plc was running,.,. See discussion under number 8.

f s. before a'nd after LOP.'

kJ Thisiscontrarytot5[

test proced6Ee.

1

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