ML20237J531
| ML20237J531 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/31/1987 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20237J489 | List: |
| References | |
| NUDOCS 8709030612 | |
| Download: ML20237J531 (329) | |
Text
{{#Wiki_filter:__ _ _ l ,, ~ - V GEORGIA POWER COMPANY PLANT V0GTLE UNIT 1 l INITIAL STARTUP REPORT l l l TO THE UNITED STATES NUCLEAR REGULATORY COMMISSION OPERATING LICENSE NPF-68 NRC DOCKET 50-424 FOR THE PERIOD JAhTARY 16, 1987 THROUGH JULY 31, 1987 l l l Og. 8709030612 870831 ; PDR ADOCK 05000424 ' p PDR I
4 ACKNOWLEDGEMENTS The Initial Startup Report for Plant Vogtle Unit I was edited. by W.C. Phoenix under the supervision of W.L. Burmeister. The Startup Test Team personnel are listed below: M.D. Barker S.N. Bennett* R.R. Bone M.E. Chance
- B.T. Cleveland
- M.W. Davis R.E. Elder L.W. Fly
- J.T. Gasser
- P.T. Green W.J. Guilford T.S. Hargis 0.D. Hayes
- l C.B. Holland
- l G.L. Hooper
- I W.F. Jocher l
S.P. Johnson
- N.C. Lee
- E.C. Longenecker*
O W.E. Moore
- C.L. Narmi
'C.A. Oakley*
W.C. Phoenix
- A.G. Rickman*
T.W. Ryan* R.D. Seibert* S.W. Semmes J.B. Sills
- B. Sloan*
D. Sowsawat D.E. Tamplin* J.J. Trawber* B. Turpin* J.A. Williams
- J.D. Williams D.T. Woodfin*
Denotes contributor to startup report. Thanks to Veronica Johnson for her typing, and to C.B. Holland and L.W. Fly for their help in assembling the report. I
TABLE OF CONTENTS (Test Numbers Are Listed in Parenthesis.) [FSAR Sections Are Listed In Brackets.] O SECTION PAGE NUMBER LIST OF TABLES VIII LIST OF FIGURES XI
1.0 INTRODUCTION
1-1 2.0 STARTUP TEST PROGRAM OVERVIEW 2-1 3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY 3-1 4.0 INITIAL FUEL LOAD 4-1 4.1 Overview and Summary Of Initial Fuel Loading 4-2 4.2 Initial Fuel Load Test Sequence (1-500-01) 4-6 [14.2.8.2.20] 4.3 Core Loading Instrumentation And (1-5SC-01) 4-17 Neutron Source Requirements [14.2.8.2.7] 4.4 Inverse Count Rate Ratio (1-500-02) 4-22 Monitoring For Core Loading [14.2.8.2.21] 4.5 Reactor Systems Sampling (1-5SJ-01) 4-28 For Core Loading [14.2.8.2.18] 5.0 PRECRITICAL TESTING 5-1 5.1 Precritical Test Sequence (1-500-04) 5-2 [14.2.8.2.23] 5.2 Auxiliary Feedwater Test (1-5AL-01) 5-7 [14.2.8.1.5 14.2.8.1.6] 5.3 RCS Final Leak Rate (1-5BB-01) 5-9 [14.2.8.2.1] 5.4 Pressurizer Heater and Spray (1-5BB-02) 5-10 Capability and Continuous [14.2.8.2.2] Spray Flow Verification 5.5 Reactor Coolant System (1-5BB-04) 5-16 Flow Measurement [14.2.8.2.3]
Table Of Contents (Continued) SECTION PAGE MUMBER 5.0 PRECRITICAL TESTING (Continued): 5.6 Resistance Temperature Detector (1-5BB-05) 5-21 Bypass Loop Flow Verification [14.2.8.2.4] 5.7 Reactor Coolant Flow Coastdown (1-5BB-06) 5-24 [14.2.8.2.5] 5.8 RCS Heat Loss Measurement (1-5BB-07) 5-28 5.9 Proteus Computer Operational (1-5RJ-01) 5-32 Test [14.2.8.1.109] 5.10 RVLIS Final Calibration And (1-5RP-01) 5-33 Operational Checkout [4.4.8.5] 5.11 Moveable Incore Detectors (1-5SE-01) 5-35 [14.2.8.2.9] 5.12 Operational Ali nment 8 (1-5SE-02) 5-37 Of Nuclear Instrument- [14.2.8.2.10] ation System Precritical 5.13 Rod Control System (1-5SF-02) 5-39 [14.2.8.2.12] 5.14 CRDM Operational Test And (1-5SF-03) 5-40 Rod Position Indication [14.2.8.2.13] Test [14.2.8.2.15] 5.15 Rod Drop Time (1-SSF-04) 5-42 [14.2.8.2.14] 5.16 RTD Mini-Cross Calibration (1-5SF-09) 5-45 [14.2.8.1.15] 5.17 Voltage Survey For (1-600-15) 5-47 Class 1E Buses [8.4.1] 6.0 INITIAL CRITICALITY AND LOW POWER PHYSICS TESTING 6-1 6.1 Initial Criticality and (1-600-04) 6-2 Low Power Test Sequence [14.2.8.2.41] 6.2 Initial Criticality (1-600-02) 6-8 [14.2.8.2.39] III
l 1 Table Of Contents (Continued) ! l SECTION PAGE NUMBER ] O i 6.0 INITIAL CRITICALITY AND LOW POWER PHYSICS TESTING (Continued): ) 1 6.3 Inverse. Count Rate Ratio (1-600-01) 6-14 i Monitorir.g For Approach To [14.2.8.2.38] Initial Criticality { 6.4 Determination of Core Power (1-500-03) 6-15 i Range For Physics Testing [14.2.8.2.22] I 6.5 Boron Endpoint Determination (1-6SF-09) o-17 [14.2.8.2.37] 6.6 Isothermal Temperature (1-600-03) 6-21 , Coefficient Measurement (14.2.8.2.40] 6.7 RCCA Bank Worth Measurement (1-6SF-06) 26 1 At Zero Power [14.2.8.2.35] l i 6.8 Pseudo Rod Ejection Test (1-6SF-04) 6-45 l [14.2.8.2.34] O 7.0 POWER ASCENSION TESTING 7-1 7.1 Power Ascension Test Sequence (1-600-13) 7-2 [14.2.8.2.50] 7.2 NSSS Testing 7-7 l 7.2.1 Power Coefficient (1-6SC-01) 7-8 Determination {14.2.8.2.26] 7.2.2 Incore Movable Detector (1-6SE-02) 7-12 And Thermocouple Mapping [14.2.8.2.30] 7.2.3 Thermal Power Measure- (1-SSC-02) 7-21 ment and Statepoint [14.2.8.2.8) Data Collection IV l L _ _ _ _ _ _ _ - _ _ _ - - _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ _ ____ ___ __
Table Of Contents (Coniinued) SECTION PAGE NUMBER 7.0 POWER ASCENSION TESTING (Continued): 7.3 Instrument Calibration and Alignment 7-23 7.3.1 Operational Alignment (1-6SE-03) 7-24 Of Nuclear Instrument- [14.2.8.2.10] ! ation System At Power 7.3.2 Operational Alignment (1-5SF-06) 7-29 , of Process Temperature [14.2.8.2.16] Instrumentation 7.3.3 Calibration Of Steam (1-6AE-02) 7-34 And Feedwater Flow j Instrumentation At Power 4 7.3.4 Axial Flux Difference (1-6SE-01) 7-37 Instrument Calibration (14.2.8.2.29] 7.3.5 Reactor Protection Test (1-5SB-01) 7-49 , [14.2.8.2.6] 7.3.6 At-Power Intercomparison (1-6RJ-01) 7-51 Of Reactor Protection [14.2.8.2.57] System Inputs And Plant Computer Outputs Test 7.3.7 Startup Adjustments of (1-5SF-07) 7-53 l Reactor Control System [14.2.8.2.17] l i 7.4 Control Systems Dynamic Testing 7-58 l 7.4.1 Large Load Reduction (1-700-01) 7-59 [14.2.8.2.52] 7.4.2 Load Swing Test (1-6SC-02) 7-60 [14.2.8.2.27] 7.4.3 Automatic Reactor (1-6SF-01) 7-69 l Control [14.2.8.2.31] l l l O v l l I I
Table Of Contents (Continued) SECTION PAGE NUMBER ! l l 7.0 POWER ASCENSION TESTING (Continued): 7.4.4 Auto Stram (1-6AE-01) 7-75 Generator Level [14.2.8.2.25] Control l 7.4.5 Dynamic Automatic (1-6AB-01) 7-81 Steam Dump Control [14.2.8.2.24] 7.5 Transient Tests 7-84 7.5.1 Plant Trip From (1-700-02) 7-85 100% Power [14.2.8.2.53] ' 7.5.2 Natural Circulation (1-600-10) 7-92 Demonstration [14.2.8.2.47] ! ( ! 7.5.3 Remote Shutdown (1-600-08) 7-120 Test (14.2.8.2.45] ' i 7.5.4 Loss of Offsite Power (1-600-09) 7-122 l At Greater.Than [14.2.8.2.46] 10-Percent Power [ l 7.6 Other Tests 7-126 1 t 7.6.1 Ultimate Heat Sink Heat (1-6EF-01) 7-127 Rejection Capability [14.2.8.2.60] l Test 7.6.2 Waste Evaporator (1-5HB-01) 7-130 Performance Test I 7.6.3 Boron Recycle (1-5HE-01) 7-131 ' Evaporator Performance 7.6.4 Metal Impact Monitoring (1-55Q-01) 7-132 ; System Test [14.2.8.2.19] ] VI l l i
Table Of Contents (Continued) SECTION PAGE NUMBER 7.0 POWER ASCENSION TESTING (Continued): 7.6.5 Gross Failed Fuel (1-6BG-01) 7-135 Detector [14.2.8.2.59] 7.6.6 Process And Effluent (1-6SD-01) 7-137 Radiation Monitoring [14.2.8.2.28] l l Sytem Test l I 7.6.7 Biological Shield (1-600-05) 7-139 Survey [14.2.8.2.42] l 7.6.8 Inplant Communications (1-6QF-01) 7-141 [9.5.2.2.6] I 7.6.9 Dynamic Response Test (1-600-06) 7-143 [14.2.8.2.43] 1 ! 7.6.10 Thermal Expansion Test (1-600-11) 7-149 l l [14.2.8.2.48] j 7.6.11 Primary and Secondary (1-600-12) 7-153 ! Chemistry [14.2.8.2.49] i ! 7.6.12 Ventilation Capability (1-600-14) 7-155 Test [14.2.8.2.58] 7.6.13 Steam Generator (1-700-03) 7-168 Moisture Carryover [14.2.8.2.54] 7.6.14 Plant Performance (1-800-01) 7-169 l [14.2.8.2.55] 7.6.15 Nuclear Steam Supply (1-800-02) 7-171 System Acceptance Test [14.2.8.2.56] I 8.0
SUMMARY
8-1 l i L b VII l - - - - - - - - - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _
LIST OF TABLES TABLE NUMBER _ TITLE PAGE l 4.1.1 Initial Fuel Loading Tests 4-3 i 5.1.1 Precritical Tests 5-4 5.5.1 Precritical RCS Elbow Tap Differential 5-18 Pressures And Flows l 5.5.2 RCS Flow Results 5-19 I 5.6.1 RTD Bypass Loops Required And As-Found 5-23 1 Flow Rates { 5.8.1 Parameters During RCS Heat Loss Test 5-30 5.15.1 Control Rod Drop Times 5-44 l 6.1.1 Initial Criticality And 6-5 l Low Power Physics Tests 6.1.2 Reactivity Computer Checkout 6-6 Results l 6.6.1 Predictions And Results Of 6-23 i ITC Test 1 6.7.1 Results Of RCCA And Bank Worth 6-28 i Measurements J 7.1.1 Power Ascension Tests 7-3 7.2.1.1 Power Coefficient Results 7-10 Versus Power Level 7.2.2.1 Zero Power Flux Map 7-13 (All Control Rods Fully Withdrawn (AR0)) 7.2.2.2 Zero Power Flux Map 7-14 (Control Bank D At Zero Steps Withdrawn, All Others Fully Withdrawn) 7.2.2.3 Zero Power Flux Map 7-15 (Control Rods At Insertion Limit) ( ( VIII
I List of Tables (Continued): I l TABLE NUMBER TITLE FAGE ; l l l l ' 7.2.2.4 Zero Power Flux Map 7-16 (Pseudo Ejected Rod Fully Misaligned) 7.2.2.5 30% Power Plateau Flux Map 7-17 7.2.2.6 50% Power Plateau Flux Map 7A18 7.2.2.7 75% Power Plateau Flux Map 7-19 1 1 7.2.2.8 100% Power Plateau Flux Map 7-20 7.3.2.1 Full Power Reactor Coolant System 7-31 Differential Temperatures 7.3.3.1 Steam And Feedwater Flow 7-36 Transmitter Ranges f j 7.3.4.1 Full Power AFD Calibration Values 7-41 1 l 7.3.6.1 Parameters And Acceptance Criteria 7-52 I W l
) 7.4.2.1 Feedwater Control System Settings 7-67 i
7.4.4.1 Control System Settings 7-79 l Feedwater Bypass Valve Controller 7.4.4.2 Main Feedwater Control System Settings 7-80 l 7.4.5.1 Steam Dump Controller Settings 7-83 7.5.2.1 Sequence Of Events For Vogtle Unit 1 7-96 Natural Circulation Test 1 7.5.2.2 Core Exit Temperatures From Selected 7-98 Thermocouple During Power Operation And Natural Circulation i 7.5.2.3 Discussion of Figures (For Natural 7-99 Circulation Test) l l l 7.6.4.1 1/2 Ft-Lb Sensitivity Test Results 7-134 l 7.6.9.1 Steady State Vibration Velocities 7-147 l l , 7.6.9.2 Main Feedwater Line Dynamic Response 7-148 Velocities And Support Forces IX l l
I l t i List of Tables (Continued): i TABLE NUMBER TITLE PAGE l 7.6.12.1 Computer / Recorder Data 7-157 I l ! 7.6.12.2 Area Temperatures 7-159 7.6.12.3 Temporary Thermocouple Temperature 7-164 i l Readings 7.6.12.4 Heat Exchanger Data 7-167 l .s I l I i l l l l
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LIST OF FIGURES D FIGURE NUMBER TITLE PAGE l f 4.1.1 RCS Boron Concentration and RHR Temperature 4-4 During Initial Fuel Loading i 4.1.2 Fuel Load Step Number During Initial Fuel Loading 4-5 i ! 4.2.1 Fuel Load Sequence 4-8 l l 4.3.1 High Voltage Response Curve 4-20 ; [ 4.3.2 Discriminator Response Curve 4-21 4.4.1 Inverse Count Rate Ratio: Permanent Detector NI-31 4-24 j 4.4.2 Inverse Count Rate Ratio: Permanent Detector NI-32 4-25 j l 4.4.3 Inverse Count Rate Ratio: Temporary Detector A 4-26 ( 4.4.4 Inverse Count Rate Ratio: Temporary Detector C 4-27 5.4.1 Pressurizer Minimum Flow Spray Valve Opening 5-13 l 5.4.2 Pressurizer Spray Effectiveness Test 5-14 5.4.3 Pressurizer Heater Effectiveness Test 5-15 1 j i 5.5.1 Total RCS Flow Versus Power Level 5-20 l 5.7.1 RCS Flow Coastdown 5-26 5.7.2 Inverse Flow Fraction 5-27
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i i 5.8.1 Steam Generator Water Level During Steamdown 5-31 l 6.1.1 Zero Power Physics Test Sequence 6-7 6.2.1 ICRR During Rod Withdrawal 6-10 , 1 6.2.2 ICRR During Dilution - Time 6-12 l 1 6.2.3 ICRR During Dilution - Boron Concentration 6-13 l 6.5.1 Acceptance Criteria And Test Measurement 6-19 l l Results For Boron Endpoint Test ! l ! i 6.5.2 Test Measurement Results For 6-20 I l Differential Boron Worth
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XI , I
LIST OF FIGURES (Continued): FIGURE NUMBER TITLE _ PAGE 6.6.1 ITC And MTC Results 6-24 6.6.2 Rod Withdrawal Limits 6-25 6.7.1 Individual Control Bank D Differential Worth 6-29 6.7.2 Individual Control Bank D Integral Worth 6-30 6.7.3 Individual Control Bank C Differential Worth 6-31 6.7.4 Individual Control Bank C Integral Worth 6-32 6.7.5 Individual Control Bank B Differential Worth 6-33 6.7.6 Individual Control Bank B Integral Worth 6-34 6.7.7 Individual Control Bank A Differential Worth 6-35 l 6.7.8 Individual Control Bank A Integral Worth 6-36 l 6.7.9 Individual Shutdown Bank E Differential Worth 6--37 j l 6.7.10 Individual Shutdown Bank E Integral Worth 6-38 6.7.11 Individual Shutdown Bank D Differential Worth 6-39 6.7.12 Individual Shutdown Bank D Integral Worth 6-40 6.7.13 Individual Shutdown Bank C Differential Worth 6-41 6.7.14 Individual Shutdown Bank C Integral Worth 6-42 6.7.15 Control Banks In Overlap Differential Worth 6-43 6.7.16 Control Banks In Overlap Inte8ral Worth 6-44 6.8.1 Pseudo Ejected Rod D-12 Differential Worth 6-47 6.8.2 Pseudo Ejected Rod D-12 Integral Worth 6-48 7.2.1.1 Power Coefficient Test At 50% Power 7-11 7.3.1.1 Source Range Detector Discriminator Setting 7-27 7.3.1.2 Source Range Detector High Voltage Setting 7-28 XII
LIST OF FIGURES (Continued): FIGURE NUMBER TITLE PAGE 7.3.2.1 Reactor Coolant System Temperatures 7-32 7.3.2.2 Reactor Coolant System Enthalpies 7-33 7.3.4.1 RCS Boron Concentration Versus Time 7-43 7.3.4.2 Axial Flux Difference Versus Time 7-44 l 7.3.4.3 Control Bank D Position Versue Time 7-45
/.3.4.4 Reactor Power Versus Time 7-46 7.3.4.5 Schematic Of Excore Nuclear Instruments 7-47 7.3.4.6 100 Percent Power Currents 7-48 7.3.7.1 Thot, Tcold, And Tavg Versus Power 7-55 1
7.3.7.2 Average Steam Generator Pressure Versus Power 7-56 7.3.7.3 Turbine First Stage Pressure Versus Power 7-57 7.4.2.1 10% Load Decrease- Steam Generator Levels 7-62 V 7.4.2.2 10% Load Decrease- Reactor Power And Control 7-63 Bank D Position 7.4.2.3 10% Load Increase- Steam Generator #1 Level 7-64 7.4.2.4 10% Load Increase- Reactor Power And Control 7-65 Bank D Position 7.4.2.5 10% Load Increase- Pressurizer Pressure And Level 7-66 7.4.3.1 +6 F Tref Mismatch - 7-71 Reference Temperature And Auccioneered T Average 7.4.3.2 +6 F Tref Misms tch - 7-72 Pressurizer Pressure And Control Bank D Position 7.4.3.3 -6*F Tref Mismatch - 7-73 Reference Temperature And Auctioneered T Average 7.4.3.4 -6 F Tref Mismatch - 7-74 Pressurizer Pressure And Control Bank D Position
'( XIII
f l LIST OF FIGURES (Continued): FIGURE NUMBER TITLE PAGE l 7.5.1.1 100% Trip - Neutron Flux 7-87 7.5.1.2 100% Trip - Pressurizer Pressure 7-88 7.5.1.3 100% Trip - Pressurizer Level 7-89 l 7.5.1.4 100% Trip - Nuclear Flux And Delta T Power Loop i 7-90 7.5.1.5 100% Trip - Loop 1 RCS Temperatures 7-91 1 7.5.2.1 RCS Forced Flow 7-105 { t I l 7.5.2.2 RCS Temperatures 7-106 7.5.2.3 Steam Gerierator Narrow Range Water Levels 7-107 ! 7.5.2.4 Auxiliary Feedwater Flow Rates 7-108 ) 7.5.2.5 Steam Generator Pressures 7-109 7.5.2.6 Pressurizer Level 7-110 j
) 7.5.2.7 Pressurizer Pressure 7-111 ;
4 l 7.5.2.8 Narrow Range Steam Geicrator Water Levels 7-113 f 7.5.2.9 Steam Generator Pressures 7-114 7.5.2.10 RCS Wide Range Loop Temperatures 7-115 7.5.2.11 Loop Wide And Narrow Range Delta-T Power 7-117 7.5.2.12 RCS Wide And Narrow Range Tcolds 7-118 7.5.2.13 Narrow Range Delta-T Behavior 7-119 7.5.4.1 Reactor Coolant System Wide Range Temperatures 7-124 7.5.4.2 Steam Generator #1 Water Level 7-125 7.6.5.1 Activity Levels Versus Reactor Power Levels 7-136 m b ~/ XIV
l 1
1.0 INTRODUCTION
s 1.1 The Startup Report l The startup report for the Vogtle Electric Generating Plant Unit 1 (Vogtle l Unit 1) discusses the results of testing from fuel loading through full power ? operation. This report is written in response to Technical Specification 6.8.1.1 which requires that a summary report be written to address each of the startup tests identified in the Final Safety Analysis Report (FSAR) and other license commitments. The report includes a description of the measured values of the operating conditions or characteristics obtained during the test i program and a comparison of these values with design predictions and i obtain specifications. Any corrective actions that were required to satisfactory operation are also described. I This report addresses the requirements by describing each of the tests and problems encountered during the tests. The Technical Specification requirement for submitting the startup report is within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. Item (2) is being satisfied since commercial operation was declared May 31, 1987. I Several startup tests remain to be completed and are identified in the report. Q Per Technical Specification 6.8.1.1, a supplement to this startup report will V be submitted every three months until all tests have been completed. 4 1 I I 1.2 The Facility Vogtle Unit 1 is one of two 3411 megawatt thermal (Mwt) four loop pressurized water reactors situated in Burke County, Georgia, approximately twenty-six miles south-southeast of Augusta, Georgia, on the Savannah River. It is jointly owned by Georgia Power Company, it:: operator, the Municipal Electric Authority of Georgia, Oglethorpe Power Corporation, and the City of Dalton. l The Nuclear Steam Supply System (NSSS) was supplied by the Westinghouse Electric Corporation and is similar to other recently-constructed four loop plants including the SNUPPS design. Vogtle's 1150 megawatt electric (Mwe) (nominal gross output) turbine-generator was supplied by General Electric and the architect-engineer was Bechtel Power Corporation. A more complete description of the facility can be found in the Final Safety Analysis Report. 1-1
2.0 STARTUP TEST PROGRAM OVERVIEW The startup test program was developed from commitments described in Chapter 14 of the Final Safety Analysis Report and other licensing commitments as discussed in the U.S. Nuclear Regulatory Commission's (NRC's) Safety Evaluation Report (SER), NUREG-1137 and supplements. Testing of the NSSS generally followed generic Westinghouse format except for natural circulation testing which occured using decay heat. The testing program was conducted in a deliberate stepwise program under the direction of pre-approved procedures. The plant was taken from fuel load to full power in a highly-controlled and documented manner that demonstrated acceptable conformance with design and permitted adjustments and minor modifications. The program started with the receipt of the fuel load license on January 16, 1987, and progressed with fuel loading, precritical testing, initial criticality and low power physics testing, and power ascension testing. Fuel loading directed the initial core loading in a safe manner. Precritical testing brought the plant to hot standby conditions, made measurements and demonstrated that the plant was ready for critical operation. Initial criticality on March 9, 1987 brought the reactor critical for the first time. Zero power physics testing made measurements on the critical reactor to demonstrate conformance with desi 8 n prior to power operation. Power ascension testing brought the plant to full power while making instrument adjustments and demonstrating the plant's ability to withstand selected transients. Fuel load, precritical testing, initial criticality and low power physics testing, and power ascension testing, are discussed in separate sections of the report. The report details the objectives, methodology, results, and problems of each startup test procoedure. The test procedures were numbered by the following convention: X - YDD - NN Where: X is the unit number; Y is evolution identifier: 5 - Precritical Test; 6 = Criticality or Later; 7 = Power Ascension Testing; and 8 = 100% Power Related Testing. DD is identifier for plant system, with 00 being entire plant or many systems; and NN is sequential number. The initial part of each section has a summary that lists the procedure by number with a one line description. 2-1
3.0 V0GTLE UNIT 1 STARTUP CFRONOLOGY D The startup chronology is a visual history of the startup test program. It starts just before fuel loading, continues through the end of July 1987, and l details the plant's status and testing. l l l l l \ l
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l l l l l 4 i ! l i ( l l l D 3-1 c_____-________ _ _ _ _ _ _
f 1 1 3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): J w MAJOR ACTIVITIES , 1/9- , 0 amb - Lineups in progress for RCS fill and vent l _j w_ 1/10 _, 0 amb -
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1/11 ' _ ' _~ 0 amb - Started RCS fill and vent
- 1/12 0 amb - Started Source Range Detector calibration Started temporary detector calibration l 1/13 Z Z Zf__._ I 0 amb - 1 Started RCS Boron Sampling procedure 0 amb - Filled Reactor Cavity i
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1/14 i r
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i 1/15._ 0 amb - Installed Primary and Secondary Neutron Sources C._ ' .~t ~
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jfig ~ 0 amb - Temporary Detector electronic noise
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1/17' '~~~ 0 amb 6 First bundle placed in core p _ _ _ _ _ _ _ Upender malfunction 1/18,-- .. 0 amb 6 1/19 0 amb 6 10 Bundles loaded
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1/20 0 amb 6 Core loading suspended due to Nuclear Instrument noise i .. . 1/21'
}' 0 amb 6
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1/22 0 amb 6 NI31 replaced
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1/23 0 amb 6 Core loading resumed, 25 bundles loaded
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1/24'- ]~ ,' 0 amb 6 76 bundles loaded m L' J 3-2
3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): w , MAJOR ACTIVITIES
'/25 -
- i 0 amb 6 86 bundles loaded
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1/26 0 amb 6 Sigma Refueling Machine required reprogramming Til$~2:7 133 bundles loaded
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1/27 j 0 amb 6 171 bundles loaded 1/28 ZL.-~ . l _i_ 0 amb 6 Core loading completed
..b 1/29 --
. ... 0 amb 6 Completed Core verification
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1/30 0 amb 6 ZT_ .. . Started installation of reactor vessel head package
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1/31 0 amb 6 En. .. p. _ 2/1 4
$[ ~[' 0 amb 5 Entered Mode 5
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Initiated Prteritical Test Sequence 2/2 0 amb 5
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2/3 0 90 5 Initiated Movable Incore Detector Test 1 2/4 -0 90 5 2/5
,' 0 90 5 Initiated DMIMS Test, Thermal Expansion, CRDM Operability Test 2/6 Cleaned Movable Incore Tubing
' 0 90 5 Initiated System RVLIS Test, Alignment of Nuclear Instrumentation j
2/7 0 90 5 Initiated Alignment of Process Temperature Instrumentation 2/8- 0 90 5 Initiated Dynamic Response Test 2/9 ._l_ 0 90 5 Ti - 3-3
3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): O n MAJOR ACTIVITIES h. 2/10 - ~ Iai 0 150 5 Initiated RTD Mini Cross-Calibration
- Z-~ ~ Discovered cable problem on CRDM's, initiated repair
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2/11 .~ _.-__ 0 150 5 Completed Bio-Shield Background Survey
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2/12---~~ ~ 0 150 5 2/13 _ 0 150 5
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Q ~~~ :: l 2/14 . . . . ._ 0 150 5 Resumed CRDM Operability Test
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1 2/15 l' ' O 150 5 Initiated Cold Rod Drop Test
!.~. ~ ~ } ^ZT Completed NSCW portion of Thermal E::pansion 2/16
.: 0 150 5 Performed RTD Mini Cross-Calibration
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0 150 5 ( ) 2/17 . . .
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2/18 , O 150 5 Completing Mode 4 entry requirements
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2/19 0 180 5 Performed RVLIS Test
. N' 2/20 0 250 4 Entered Mode 4
"! i Initiated Chemistry Test, Ventilation Capability Thermal Expansion 2/21 0 250 4 RTD Mini Cross-Calibration i
2/22 0 344 4 Initiated Thermal Expansion, RTD Mini Cross-Calibration
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for 350 F plateau 2/23 0 450 3 Entered Mode 3 Initiated Thermal Expansion, RTD Mini Cross-Calibration for 450 F plateau 2/24 0 450 3 i 3-4
1 l 3.0 V0GTLE UNIT 1 STARTUP CHR01.0 LOGY (Continued): , D I g MAJOR ACTIVITIES u., I j 2/25 - ,
!$I 0 557 3 Initiated Movable Incore Test, l
' } Z -- Auxiliary Feedwater Test 2/26 ((
J~T 1 0 340 4 Cooldown due to RCP #2 vibration Entered Mode 4/ Mode 5
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2/27 0 100 5 RCP #2 Aligned 2/28 R.%
, _ _ , 0 557 3 Returned to Hot Standby, Initiated CRDM Operability, j
gs - - -- -* Auxiliary Feedwater Test,RTD Mini Cross-Calibration, 3/1 S RCS Flow Measurement Test
, 0 557 3 Initiated Rod Drops, Rod Control, MSIV Stroke Test e--- -
Proteus Computer Test, Pressurizer Heater / Spray Test 3/2 _T~ ~ T ; M ~- 0 557 3 Performed Gross Failed Fuel Detector Test, l Alignment of Process Temperature Instrumentation l 3/3 ~f:. 0 557 3
".1~ 1:. Performed Dynamic Eff-scts on Pressurizer Spray Line, RVLIS Test D
l 3/4 3..... 0 557 3 Initiated Class IE Voltage Survey 3/5 {[ 0 557 3 Performed RCS Flow Coastdown Test, RCS Leakrate
'~.
3/6 , 0 557 3 Completed Class 1E Voltage Survey 3/7 0 557 3 Completed RCS Leakrate, MSIV Stroke Test 3/8 0 557 3 Completed Precritical Test Sequence D 3-5 l l
l 3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): MAJOR ACTIVITIES j "w O w i A ! E @
- 3/9 l
I~ TT 0 557 2 Entered Mode 2; Initiated Low Power Physics Test Sequence t Achiaved Initial Criticality 3/10 w 0 557 2 NI-35 replacement due to ground ec . T-.1T Performed RCS Heat Loss Test W.:1 _i_- _ 3/11 - 05572 l 2.._p... ___ _-I. _ i~~ 3/12 s _
<1 557 2 Initiated Biological Shield Survey, Performed Determination l
-. --_ of Power for Physics Tests, Isothermal Temperature '
3/13
-4 --
Coefficient
.g. <1557 2 Performed Flux Map, Bankworth Measurements, t 77, Boron Endpoint Measurement 3/14 <1 557 2 Continued Low Power Physics Tests l
' ._J.3a h 3/15 . li i .1] <1 557 2 3,..,__ Tripped reactor to prepare for N-1 Rod Measurement W -- Performed N-1 Bankworth Measurement 3/16 _ 0 557 3 Reactor shutdown due to dropped rod in Shutdown Bank B 3/17 -
- i. 0 150 5 Replaced rod J-3 coil stack 3/18- N <1 557 2 Resumed Low Power Physics Tests 3/19 g , l35572 Reactor trip on high rate i !
3/20 3 557 2 Performed Psuedo Rod Ejection Test Completed Low Power Physics Tests Reactor trip on low steam generator level l 3-6 l I t l
3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): . D "4 MAJOR ACTIVITIES b eE 3/21 . , - 0 557 3 Initiated Power Ascention Test Sequence, Feedwater Test
.- Reactor trip on low steam generator level 3/22-L 1 557 2 k~ ^,.'
k~ 3/23-ai- l~ 3 559 2 Automatic Steam Dump Test, Feedwater Test o 'J'T TIT Reactor trip on low steam generator level j Entered Mode 1 3/24 (; q ~ - 1 __f. , _ _ _. 7 560 1 Reactor trip on low steam generator level 3/25- 6 560 1 Tuning of Feedwater Bypass Valves in progress 3/26
,'~.}f~~ 12 563 1 Rolled Turbine Initiated Auto Steam Generator Level Control
-lT Reactor trip on low steam generator level f--
3/27 j 12 563 1 Auto Steam Generator Level Control Test ___ 6 Main generator synchronized to grid i 3/28 - ' 17 565 1 Performed Loss of Offsite Power Test I tF l 3/29 A 4 559 2 Performed Automatic Steam Dump Test N A~ 3/30 , 17 565 1 Tuning of Main Feedwater Regulating Valves i 3/31 29 571 1 Initiated Biological Shield Survey I 4/1 30 572 1
- Initiated Chemistry Test, Calorimetric, Startup Adjustments of Reactor Control, Alignment of Process Temperature Instrumentation 4/2 30 572 1 Gross Failed Fuel Detector, Metal Impact Monitoring Initiated Computer Intercomparison Test, Alignment of the Nuclear Instrumentation System, Steam /Feedwater Calibration, RCS Flow Measurement 4/3- - 30 572 1 Performed Flux Map, AFD Calibration 4/4 30 572 1 Performed Auto Steam Generator Level Control Power Coefficient Measurement 4/5 '
' 0 557 3 Turbine trip on generator fault 1
i Reactor trip on low steam generator level D 3-7
l 3.0 V0GTLE Ui :T 1 STARTUP CHRONOLOGY (Continued): . D "A MAJOR ACTIVITIES
$E W
C3 b 3 ! 4/6 0 557 3 ! :~.kf!1... i l l 4/7' ~.. d .J
.. 9 562 1 Performed Dynamic Effects on Fcedwater Pumps
~~
4/8 20 567 1 Performed Remote Shutdown Test ,
' I - 3 .._W
~~t ~ - '
I 4/9 2 557 2 Recovered to 30% power l h :.7h - 30 572 1 Performed Rod Control Test I j 4/10['T : J_~ l l 30 572 1 Performed Load Swing Test, Automatic Steam Generator 1.: A Level Control F -- _ : 0 557 3 Reactor trip on low steam generator level l 4/11 ' 5 560 1 Reactor trip on low steam generator level M.7 Main Feedwater Pump A discharge check valve broken l 4/12 s' ~~ 0 557 3 Outage on Main Feed Pumo A Rj _,
- ~
' ' ~~ f 5~ 0 557 3 D 4/13 . . . .
3_ l ! 4/14 ~F f '! O 557 3 Remote Shutdown Retest (Stability Portion)
! 5 558 1 l __ F ~.T 1
~
l 4/15 10 560 1 Power increase to 50%
' k'.
_ 148 570 1 4/16 8 570 1 Initiated Bio-Shield Survey, Ventilation Capability,
' Chemistry. Performed Calorimetric, Steam / Feed Calibration
, Align of Process Temperature, RCS Flow Measurement 4/17; 48 570 1 Performed Thermal Expanrion, Flux Map l
4/18; 48 570 1 Performed Automatic Steam Generator Level Control, l Computer Intercomparison 4/19- 49 570 1 Performed Power Coefficient Measurement Axial Flux Difference Calibration l 4/2b' 49 570 1 i 4/21 49 570 1 Started power increase to 75% 3-8 i I L
9 I 3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued):
- 1 "w MAJOR ACTIVITIES i
h I l$ 2#E 4/22 75 582 1 Performed Calorimetric, Bio-Shield Survey, DMIMS
" l ~.-
RCS Flow Measurement, Gross Failed Fuel Detector 4/23 biZ
^
._ V I~'~75 582 1 Performed Computer Intercomparison Reduced power'due to EHC problem .
.4/24 25 565 1 Completed repair of EHC
^ 3[~~ I Returned to power i 4/25 h^f1 T" 48 570 1 Main Feed Pump B pump seal leaking 4/26 [~_, ', 63 576 1 Repair of MFP B l
.. p _ ._
T.J~.q:. 4/27 j 2 l..,j,,_ 3 565 1 Reduced reactor power to repair EHC leak MFP B work continuing j
- ~~} : .--.
4/29 23 565 1
~ h h~~.33 580 1 Generator synchronized to grid
.,. _ . _ _. Returned to 73% power. Performed Plant Performance D4/29 . , , _
.i 73 580 1 MFP B returned to service 0 557 3 Roactor trip 1 4/30 5 559 1 Mode 1 entry 1 5/1 * ..~I .}f .~ 12 560 1 Return to 75% power
. . ~73 580 1 5/2 ., . 73 580 1 l
5/3 - l _ -74 530 1 Performed Power Coefficient Measurement i 5/4 75 580 1 Performed Axial Flux Difference Calibration I O 557 3 Reactor trip 5/5 -- 0 557 3 5/6 , 0 557 2 Returned to power 1 5/7 ' l .. 75 '580 1
~
'{',..
%9 i
M
3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): , O "w MAJOR ACTIVITIES 5/8 iRi 75 580 1 Performed Xenon oscillation for AFD Calibration !
. 7Z 1 Initiated 10% Load Swing Test
_. } _. ... .. 5/9 7 O 357 3
~~ ~~~ Reactor trip on low steam generator level during performance of Load Swing
~ ~.1 ZZ 5/10 _ ,, 0 557 2 Ret.urned to 66% power
---w. .__ 50 568 1 Reduced reactor power
~ 4:, ^:
5/11 _ , 55 576 1 Performed +10% Load Swing Retest 5/12 .- Q Y5 580 1
, Performed 50% Load Reduction
-.__._ f'- "75 58 0 1 Initiated increase to 90% power 5/13 - ~ 90 585 1 Performed Gross Failed Fuel Detector Test )
0 557 3 Reactor trip due to MSIV closure
-'i 5/14 0 350 4 Outage to inspect turbine bearings
. _.i 0 200 5 5/15 _
0 150 5 w } ::..- 5/16 4 ' -' 1 h [' _-', 0 150 5 5/17 _ , 3 0 150 5
~:} :.
5/18 , 0 150 5 4 5/19 ' 01955 1 0 200 4 Recovery from outage 5/20 \' O 367 3 5/21 , 0 557 3 5/22 O 557 3 Performed Process and Effluent Radiation Monitoring i Test t 5/23 l 0 557 3 O - 3-10 I
l l 3,1 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): 1 . i n , MAJOR ACTIVITIES 5/24 iRi 4 558 2 Performed Automatic Steam Generator Level Control
.f.(..{ Retest 5/25 ( 30 567 1 Returned to 90% power T 90 585 1 Performed Secondary Calorimetric
' y } '. ~ ! -
5/26 90 585 1 Completed Alignment of the NIS, Startup Adjust of Reactor
E, ~'
Centrol, RCS Flow Measurement, Chemistry, Thermal Expansion, Alignment of Process Temperature Instrumentation
- 1 5/27 ,_ _ ._ \ 95 586 1 Increased to 100% power 5100 588 1 Started Warranty Run 5/28 '~~~ "100 588 1 Performed Flux Map T:. .
5/29 _ p00 588 1 Performed Secondary Calorimetric, Bio-Shield Survey 7 Chemistrs
- 5/30 ~ ~~~100 588 1 Performed RCS Flow Meaurement, AFD Calibration, Thermal l . . _ . . ..
Expansion, SU Adjust of Reactor Control, Align.nent of Process Temperature Instrumentation, Steam / Feed Calibration D5/31 _._4__ t ___ ,iOO 588 1 Completed Warranty Rur, 3 _- _ . . , Plant declared commercial 6/1 3'00 588 1 Performed Power Coefficient Measurement,
' {; :_ -~~-" ] Computer Intercomparison, Alignment of the NIS 6/2 . .-. 100 588 1 Completed Thermal Expansion i: '~'
6/3 ' 100 58f 1
. 0 557 3 Reactor trip, performed Natural Circulation Test l 6/4 - 7 0 557 3 Performed Alignment of the NIS (Post trip portion) b e '
! 6/5 , 0 557 3 Completed Reactor Protection Test I '
' k i :
6/6 -J 0 557 3 Reactor trip on Source Range high level 5 558 1 Entered Mode 1 6/7 \
- 0 557 3 Reactor trip on low steam ger.erator level 0 557 3 Reactor trip on low steam generator level due to g jg MFP A check valve stuck open 6/8 , 0 557 3 MFP A outage i l
3-11 1
3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): ,
" ca. MAJOR ACTIVITIES e,
$@ b2 g 6/9 _. , 0 557 3
' ' .-~_: :~.! i 6/10- '9 ~ ~ - '
0 557 3 i ..j
-4 ~
6/11- ~ i a 0 557 3 T_.4 _q 6/12 g.: $rj 0 557 2
, Return to power
. . ._7 _ . .
~~~ -
6/13 1 557 2 6/14 "~ ~ ;~. _C'98 , 588 1 _ _ _4. _ _ ... 0 557 3 Reactor trip
~ ~
6/15 0 557 3
~-
6/16 0 557 3
. _. .1_ . .
6/17 0 557 3
'[ _
rix: ~
' ~
6/18 0 557 3 _.o . _. + . 6/19 , , 0 557 3 6/20-- 0 557 3
~[~~~
._ 1 i
6/21- 1 557 2 6/22 50 570 1 6/23 , K 93 585 1 Reactor trip from generator trip 6/24 . . . 0 557 3 a e.. ir @ 3-12
3.0 V0GTLE_ UNIT l_STi,RTUP C!lR_0NOLOGY (Continued): t )
'\ / j w MAJOR ACTIVITIES h
6/25 eI$a 1 557 2 6/26 $- 100 588 1 6/27 100 588 1 I 6/28 l 100 588 1 ' l 6/29 1 100 588 1 1 i a. 5 6/30 -- 100 588 1 7/1 , 100 588 1 7/2 , 100 588 1 7/3 100 588 1 1 7/4 100 588 1 l 7/5 100 588 1
, l 7/6 100 388 1 7/7 100 588 1 1
7/8 - 100 588 1 0 557 3 Reactor trip from generator trip l 7/9 0 557 3 i 7/10 ---- 0 557 3 ; 3-13 I 1
3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): , O b 1
" ra. I MAJOR ACTIVITIES b2 7/11 0 557 3 1
1 7/12- 1 557 2 q_, b_. ' 7/13 35 565 1 7/14
~j 100 588 1 g-y.-.. .
I l .; _ . ..
' ~ ~
7/15 - _- 100 588 1 I
-~12Z .
7/16 ; 100 588 1
~ ~5~ '
.t 7/17 _]'~~ 100 588 1
_ . 4 . _. _. . p "TQ ~.~ . ( 7/18 . , 4_ 100 588 1
. .T_K .' ~
7/19 'j
. , _ b.
100 588 1 l 7/20 _.. 100 588 1
. _. 4 7/21 , . _ 100 588 1 7/22 - 100 588 1 Reactor trip on generator trip 7/23 0 557 3 7/24 " 75 580 1 Returninu to power N
7/25 100 588 1 rea O s.m
3.0 V0GTLE UNIT 1 STARTUP CHRONOLOGY (Continued): , [J MAJOR ACTIVITIES "w iBi 7/26 100 588 1 j_ . __ \.
.y$ -...
Wj 7/28 $
. g. !
! 100 588 1 Reactor trip on turbine trip signal !'
7/29 ) _._ }._. 1 557 2 Return to power _ .L _ 7/30
.___'....y100 588 1 7/31 D 100 588 1
, ' ~ . . 'N .
..q:-
qU l ( i l 1 l l l l v 3-15
l 1 SECTION 4 INITIAL FUEL LOAD ( l 4-1
4.1 ,0VERVIEW AND
SUMMARY
OF INITIAL FUEL LOADING
~
The initial fuel loading at Vogtle Unit I was accomplished from January 17 to January 29, 1987. Fuel load was dry with water borated to refueling water concentrations in the reactor vessel and interconnecting systems but at least one foot below the level of the reactor vessel flange. Demineralized water filled the refueling cavity to at least one foot above the upper lip of the fuel transfer tube but at least one foot below the reactor vessel flange. ] Reactor coolant system (RCS) boron concentration and residual heat removal (RHR) temperature are shown in Figure 4.1.1. Boron concentration was periodically adjusted but at a slow enough rate that it did not disturb inverse count rate ratio monitoring. RHR temperature also varied somewhat but i likewise was slow. Some fuel bundles were moved more than once, especially those bearing sources, so fuel loading had more steps than the 193 bundles. The number of steps versus time is plotted in Figure 4.1.2. Delays are due to upender and nuclear 1 instrument problems described later. After the core was loaded a video taped verification of proper bundle ! placement was conducted. I Tests associated with initial fuel loading are listed in Table 4.1.1. I
/
( N I I I 1 O V 4-2
4.1 OVERVIEW AND
SUMMARY
OF INITIAL FUEL LOADING (Continued?- * , l l TABLE 4.1.1: INITIAL FUEL LOADING TESTS i l PROCEDURE NUMBER TITLE -DESCRIPTION 1-500-01. Initial Fuel Load Test Controlled initial fuel loading Sequence sequence. 1-500-02 Inverse Count Rate Ratio Specified the nuclear monAtoring Monitoring For Core Loading for reactivity monitoring during initial fuel loading. 1-5SC-01 Core Loading Instrumentation Verified alignment, calibration, and and Neutron Source neutron response of the temporary Requirements neutron monitoring instruments, and neutron response of the permanent plant instruments. 1-5SJ-01 Reactor Systems Sampling Sampled the reactor coolant system For Core Loading and connected auxiliary systems for i correct and uniform boron i concentration prior to initial fuel ) loading. 1 3 4-3 1 a i
l i 4.1 OVERVIES AND
SUMMARY
OF INITIAL FUEL LOADING (Continued)s ] 1 D FIGURE 4.1.1: RCS BORON CONCENTRATION AND RHR TEMPERATURE DURING INITIAL FUEL LOADING i i l FPM BORON CONCDfrRAfl0H DURING FUEL LOADING 2200 n0 2190. A RHR BORON CONCENTRATION, PPM 0 RHR TEMPERATURE, F y9 2180. %, , 2170. 2160. k i 2150.
' ^ 36 h
1 2140. g b
*f%
3 s
> Q1e a25
?Q i
II Q 4%.\ V' r M. l 23 0 \ 0 ' 21
,,,2 09 0.).
l s j2080. kj c Ig c~a====gm, a==3 ,0 , r2= h060..^, Z 2050. I l\ if
?$
1, c b8 % cd f g2040. 4 57N
$2030, g 16 0 to 4 020 P c5 14 l'5 l'6 1h l'8 19 20 2'1 22 2'3 24 25 2'6 2'? 2'8 21 DAY IN JANUARY 1987 4-4 I
4.1 OVERVIEW AND >UMMARY OF INITIAL FUEL LOADING (Continued): FIGURE 4.1.2: FUEL LOAD STEP NUMBER DURING INITIAL FUEL LOADING LOADING STEP HUMBER UERSUS TIME AND DATE Ill 299 A 90GTLE UNIT 1 CORE LOAD 199 189 170 169 159 l 140 tY 139 f 120 [ l l 110 I 99 D 89
,e 1 69 [
h 59 f f $40 Y 0 39 [ 29 1 _f ! C 10 - b9 -" f I? 18 l'9 N9 Ni 2'2 N3 24 2'$ N6 N7 N8 20 DAY OF JANUARY 198, 4-5
4.2 INITIAL FUEL LOAD TEST SEQUENCE (1-500-01) Objectives Objectives of the Initial Fuel Loading Test Sequence were to: (1) establish the conditions under which the initial fuel loading was to be accomplished; and (2) accomplish initial fuel loading in a safe and orderly manner. The abstract for this test is FSAR Section 14.2.8.2.20. Methodology The core was loaded by first loading bundles around bundles with sources, loading a slab of bundles from one side of the core to the other, completing the core on one side, and filling in the other. The core load was continuously monitored by temporary and permanent neutron detectors. A small core was first built near startup detector NI31, shown as SR31 on Figure 4.2.1 starting at the core shroud. Bundles with sources are shown with black dots and are shown in the figure for Steps CA to 7E as 1 and 2, the first two bundles in the core. The temporary detectors are shown as A, B, and C. The second source-bearing bundle was moved next to the one near the core baffle, which was then moved to a storage location as shown for Steps 8 to 9. Bundles were then loaded to start a " bridge" across the core from one permanent source range detector to the other as shown in Steps 10 to 37. The bridge was nearly completed, with shuffles for source-bearing bundles, in Steps 38 to 59B, and 60 to 62. The bridge was completed and another section of the core started toward the 270 portion of the core in Steps 63 to 93. This section was completed in Steps 94 to 126. With the core now more than half loaded, the remainder of the bridged section was filled out to the edge of the support barrel flat and a section toward 90' started in Steps 127 to 167. The core was completed in Steps 168 to 205. The temporary detectors were removed as required to be replaced by fuel bundles. In all, 193 fuel bundles were loaded. The core was then verified loaded as expected by a portable television monitor and video tape. 4-6
I I 1 l i 4.2 INITIAL FUEL LOAD TEST SEQUENCE (1-500-01) (Continued): I Results l The objectives of the core loading were met. Vogtle Unit 1 core was loaded and verified as shown in Figure 4.2.1 in a safe and orderly manner. There were no reactivity problems. Problems The fuel transfer car upender in the containment malfunctioned during the initial portion of fuel loading but an adjustment was made that prevented further problems. The source ran8e nuclear instruments suffered from noise problems due to grounds and fuel movement was periodically halted due to either high noise on the channel or the containment evacuation alarm. The grounds were isolated, and containment evacuations ceased. The Sigma refueling machine required reprogramming and was successfully reprogrammed. The Sigma's television camera could not be used ac planned to verify proper loading because the camera's electronics were not properly aligned. A portable h F camera was borrowed from Plant Hatch, a Georgia Power nuclear plant, and connected into the Sigma's television circuitry. The camera was lowered over the side of the refueling machine. The machine moved over the core to coarsely locate the camera and then finely by hand. Verification then proceeded. I i l 4-7
4.2 INITIAL FUEL LOAD TEST SEQUENCE (1-500-01) (Continued): FIGURE 4.2.1: FUEL LOAD SEQUENCE
- 1. co R P N M L K J M G fr g 0 C S &
l u . e. 4 .s .2 l i 2 . B 6 '6 l7o A j 3 .. 4 i 5- ' l 4 7- , SC* 37ce 9- l 10 -
'I - ' '
- 7ei 12 l
(3 - 14 l l 15 - ,- #] I 0' Core Loading Sequence Steps M to 3 1 4-8 1
l 4.2 INITIAL FUEL LOAD TF4T SEQUENCE (1-500-01) (Continued): FIGURE 4.2.1: FUEL LOAD SEQUENCE (Continued): 1 l l
. , 1 A P N M L K J H G F g g c g g a e i e e i i -
cl XX.lL 2 . a XX el A lx 3 ) 4 l 5- l
.h* l 7-SC* - 8 * - 27C'
! g-10 - ) l
"- I l-t-
'2 II I 13
\
i4 l2 9 l$ h I 0* . Core Lbeding Sequence Steps _!_ to _9, l 4-9
1 ! 4.2 INITIAL FUEL LOAD TEST SEQUENCE (1-500-O'.)_(Continued): FIGURE 4.2.1: FUEL LOAD SEQUENCE (Continued): i is o. l R P N M L x J H G F E O C 8 A i c XX io " ' 2 s XX.X a a 3 IG. 15 M /3 4 20 19 8 17 5~ 211 23 22 21 6- 26 27 f.C gg 7~ - 52 31 30 29 sc*- s - +37 36 55 3T 3% - 27e* ' 9-10 - . j i i! - 11 h l 13 j l 14 i
.t 15 i
C* l Core Leading Sequence Steps 10 to 37 t
4.2 INITIAL FUEL LOAD TEST SEQUENCE (1-500-01) (Continued): FIGURE 4.2.1: FUEL LOAD SEQUENCE (Continued):
@ IS C' fj .f { E O C 3 A XXXX 2
3-a XXXX A
.XXXX 5-
.XXXX-
-XXXX 5-XXXX 7-XXXX e XXXX
*~ - z2ce l}l W 39 gg
'*
- 6 Vf 43 y
~
IR 49 y) MS
- iz '
S5 52 5/ z l I3 ~ ' 57 !"6 55 gij I l [8 i fi9 l ,{ IS ._ t 892
~
ce
, Core Lbeding Sequence Steps M to g 4-11
1 I i 4.2 If TIAL FUEL LOAD TEST SEQUENCE (1-500-01) (Continued): l l FIGURE 4.2.1: FUEL LOAD SEQUENCE (Continued): ) i l
~@
15 0' i i , R P N M L 1 J H G F E O C S A i XXXX 2 . s XXXX. A ! XXXX 4 XXXX J 5- XXXX a- XXXX 7- XXXX y *-=- c XXXX -27o-C $- X.XXX l m- XXXX l ii - XXXX ! iz XXXX l is XXXX
- i. .
, , o2 .a , eX .:
l$ 0 0 0' Core Lbading Sequence Steps M to g i l
4.2 INITI,;, F_UEL LOAD TEST SEQUENCE (1-500-01) . Continued):
)
FIGURE 4.2.1: FUEL LOAD . SEQUENCE (Continued): A P N N L M J H G F E O C 5 A t g i XXXX e 93 a a XXXX e 4 3 XXXX n 4-
-XXXX = i s-XXXX 77 XXXX u 7- -
XXXX 7s ,2 er ,, e, O e- * - e XXXX 74 si sa se ' dz7e-xxxx 7172 .<z 42 .<<< eo ! io - XXXX
,i -
XXXX 7/ i iz XXXX 7o XXXX
=
is - a ! i. XXXX '8 ' ss 63 64 GS ,% G7 I, C* Core Lbeding Sequence Steps g to 9} ! 4-13 l L_-_____---_-__,_______-__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ .
[. ! l 4.2 INITIAL FUEL LOAD TEST SEQUENCE (1-500-01) (Continued): p FIGURE 4.2.1: FUEL LOAD. SEQUENCE (Continued):
}
i A P N M L K J G F E O C 8 A i iXXXXX 2 . e XXXXX as as s XXXXX m a2 '2+ 4- - X X XXX.ns "e af 5-
.XXXXX <> 2 </4 //7 '2o 7-XXXXX "o "> "i >/c g .e - * - XXXXXXXX X i
- <c XXXXXXXX - 27c- A l m- .XXXXXXXX X XXXXX- ev er e7 es
'i -
XXXXX * * *
- i2 a
~ IXXXXX e a d XXXXX o7e d a t i*
XXXXX m 1 is -
@ XXXXX i C*
Core Loading Sequence Steps 94 to 126 4-14
[ 4.2 INITIAL FUEL LOAD TEST SEQUENCE (1-500-01) (Continued): FIGURE 4.2.1: FUEL LOAD'. SEQUENCE (Continued): i 1 i . e. a * ~ y;ifirio .
- e i les m XXXXX 2
3-- -
- s 154 /*
/se XXXX'XXX m XXXXXXXX 167 4 iG2 138 mm
///YYYN/
s-XXXXXXXXX a ira m XXXXXXXXX
*- iso o4
.c \.r- la m ive w xXXXXXXXX - 27o-we is9 13 s D - -
is XXXXXXXX A
.-u a Si m s iss XXXXXXXXX io ive in XXXXXXXXX n-XXXXXXXXX
/ys ar ~
iz . m ix'XXXXXXXX is le m XXXXXXXX i. m as XXXXXXX is e m XXXXX l O' Core Loading Sequence Steps J12 to J)62 l
4.2 INITIAL FUEL LOAD TEST SEQUENCE (1-500-01) (Continued): FIGURE 4.2.1: FUEL LOAD SEQUENCE (Continued):
@.. c.
R P N M L K J 'M F I G E O C' S A 8' 1 4 i e I 2
*'m XXXXXXXXX XXXXXXX l 3 5 5 "S e2 e ms XXXXXXXXXX 5- e el e XXXXXXXXXX :
a:XXXXXXXXXXX
<- so e ss
- XXXXXXXXXXX a'oq* - XXXXXXXXXXXXXXX'
'a' aw XXXXXXXXXXXXX A:d O f.- xxxxXxxxxxxxxxxN, 10 - 174 171 f(n w XXXXXXXXXXX m l n-m os m i*
XXXXXXXXXXX l2 iM 17, 173 lllll lllll ) 13 - ff2 IW 177,XXXXX'XXXXX i* 5:XXXXXXXXX c XXXXXX
= / < @
Core Loading Sequence Steps 168 to 205 l .O 4-1e o
4.3 CORE LOADING INSTRUMENTS ION AND NEUTRON SOURCE l REQUIREMENTS (l-5SC-01) f Objectives Objectives of the Core Loading Instrumentation and Neutron Source Requirements Test were to: (1) verify alignment, calibration, and neutron response of the temporary core loading instrumentation prior to the start of. initial fuel loading; and (2) verify the neutron response of the nuclear instrumentation system (NIS) source range channels prior to the start of fuel loading. The abstract for this test is FSAR Section 14.2.8.2.7. Methodology Reactivity changes during initial core loading are monitored by temporary neutron detectors (also known as dunkers) lowered into the reactor vessel and the permanently installed source range nuclear instrument system (also known as source range instruments, or source ranges). The purpose of this test was to verify these instruments were operating properly before core loading started. Temporary Detectors Three temporary detectors were supplied by Westinghouse and consisted of BF-3 y detector tubes and their watertight housings, ; connectors, cables, ' preamplifiers, amplifiers, discriminatory, scaler-timers, rate meters, a strip chart recorder, and a high voltage power supply. After their arrival on site, they were connected and initially calibrated per Westinghouse procedures by Westinghouse personnel. Alignment by test procedure consisted of determining the high voltage plateau and discriminator settings for each detector. The equipment was connected and allowed to warm-up for at least thirty minutes. A temporary neutron source (borrowed from the boronometer) was placed in the dry refueling cavity floor next to the detectors. High voltage was increased in steps from 1500 to approximately 2300 volts and a detector response curve of counts per minute versus voltage was developed. Discriminator settings were then increased and a similar curve was obtained. Since there was only one high voltage power supply that supplied all detectors, a final high vcitage of 2100 volts common to all detectors was obtained from the best fit for all detectors (see Figure 4.3.1 for an example high voltage response curve). The setting for cach detector's discriminator was nominally 2.5 (see Figure 4.3.2 for an example discriminator response curve). f3 V 4-17 L
4.3 _ CORE LOADING INSTRUMENTATION AND NEUTRON SOUF9E REQUIREMENTS Il-SSC-01) (Continued): ..m j Background Count Rates A background count rate was determined for each channel. The source was removed from the vicinity of the detectors and 10 sixty second counting intervals taken. From this data an average background count rate and reliability factor for each detector was obtained. Response Checks The permanent source range instruments had been aligned by another procedure and ie:sponse checks of the temporary and permanent instruments was performed following alignment of the temporary instruments. l A response check, or verification that detectors were reliably responding to neutrons, was required following any delays of fuel movement of 8 hours or more. Three methods could be used. First was actually bringing a source to the detector and observing increases in count rates. Second was bringing a source-bearing fuel element close to the detectors. The third, used extensively, was a statistical reliability factor method that was especially useful because it allowed neutrons being generated by the partially loaded core to serve as the ! source. The third method's mathematics were programmed into a personal l computer that greatly facilitated the otherwise laborious calculations of determining average, standard deviation, and reliability factor for the five. M to ten count intervals required by the method. The only additional effort
.j required for the statistical method was to make at least five counts.
The first method was required for initial response verification. A 60 second sampling time was selected for all instruments and the neutron source was then placed as close to each detector as practical to verify a source count rate greater than background. After the core had been partially loaded the statistical method was used almost exclusively. Results All objectives were met. The permanent and temporary neutron detectors were aligned, background determined, and response checked prior to loading the initial fuel bundle. The temporary core loading instruments' initial calibration and checkout agreed with vendor data. Each channel responded l indicating a change in count rate as the neutron level increased and/or decreased near its associated detector. A statistical reliability factor method was used to verify operability. If an instrument produced a factor between 0.53 and 1.48 for five samples the instrument was declared operable and considered reliable, m 4-18
I i I i 4.3 CORE LOADING INSTRUMENTATION AND NEUTRON SOURCE REQUIREMENTS (1-5SC-01) (Continued): Problems The background counting interval was changed to sixty seconds because the original one hundred second interval provided too little variation in total counts to satisfy the statistical reliability (Chi-squared) program. Frequent failures plagued the temporary neutron detectors and one of the channels, initially termed C, did not respond most of the time. The Reliability Factor method was used to qualify detectors and was useful in
- determining when a detector was not reliable. The cause of channel C's ,
problems was never determined even after extensive troubleshooting 8 including multiple replacement of detectors, cables, and preamplifiers, and other work. , I i The FSAR required two operable detectors so C was not used and loading l proceeded without an operating spare temporary detector. The lack of a spare did not impede fuel loading. f No response to the source was observed when it was lowered into the reactor vessel as close to the permanent excore source range detectors as possible. Response checking the permanent detectors could only be accomplished by placing the temporary source next to the detectors. This was done by lowering the source in its temporary stainless steel holder down the NIS detector access hole on the refueling cavity floor. The D temporary detectors did not have a firm footing in the core because they were suspended by their cables and were somewhat free to move in reactor coolant flow driven by the RHR system. This movement sometimes caused the count rate to vary. Permanent actuated the source range detector N-31 suffered noise problems that occasionally containment evacuation alarm and stopped fuel movement. The detector response prior to restarting portion fuel of this procedure was used to verify operability movement. 1 i l I r i l 4-19
L i 4.3 CORE LOADING INSTRUMENTATION AND NEUTRON SOURCE REQUIREMENTS (1-5SL -01) (Continued): FIGURE 4.3.1: HIGH VOLTAGE RESPONSE CURVE l 1 TEMPORARY CORE LOAD DETECTOR HICH UOLTAGE RESPONSE 18
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4.4 INVERSE COUNT RATE RATIO MONITORING FOR CORE LOAD 7G (1-500-02) i Objective ' The objective for the Inverse Count Rate Ratio Monitoring for Core Loading Test was to describe the frequency, core conditions, and methods for obtaining nuclear monitoring data during initial core loading. The abstract for this test is FSAR Section 14.2.8.2.21. 1 Methodolgy There are two purposes for making inverse count rate ratios. The first is to i monitor for inadvertent criticality. The second is to monitor for dilution or temperature changes in the refueling water. The boron concentration of coolant )1 in the reactor vessel and connected systems was at least refueling boron concentration of 2000 ppm. Count rate data from the two permanent source range (NI-31 and NI-32, also shown as SR-31 and SR-32 for source range) and temporary nuclear detectors A, B, and C were used to calculate inverse count rate ratios. Data from five to ten counting intervals were tested for statistical accuracy using a Chi ' i Squared test and averaged. Background was subtracted to produce net counts per counting interval. This was then divided into a base ceunt rate established at D ( the beginning of fuel movement, or renormalized'when temporary detectors were moved or repaired. The Chi Squared calculations and averaging were performed using a BASIC program on a personal computer. The printouts were then signed and served as i data records. i j The temocrary detectors and their problems are described in the preceeding section. 1 l l Counts were taken for background prior to loading the first bundle, after each
- fuel bundle loading step (shown earlier in Figure 4.2.1) to monitor for criticality, and after the detectors had been repaired or replaced for normalization. Counts were also taken every hour while fuel movement pas suspended to monitor for possible dilution. t Normalization is using the new count rate as a baseline and dividing succeeding count rates into it for succeeding inverse count rate ratios. .
sJ Inverse count rate ratios were not meaningful, according *.o docedure, until ;!, the tenth movement but they do show some interesting geouletric effects prior - Jo the tenth bundle and when fuel bundles were loaded next to detectors. Geometric effects include when a fuel bundle is loaded between a neutron
, m 4-22 '
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4.4 INVERSE COUNT RATE RATIO MONTTORING FOR CORE LOADING (1-500-02) (Continued): source and a detectot, causing a much greater increase in count rate than could be attributed to reactivity. A comparison with Figure 4.2.1 and Figures 4.4.1 through 4.4.4 illustrate the change in count rate ratio due to geometry. i The procedure specified when a detector was to be considered " responding" for the purposes of reactivity monitoring. Generally a detector was considered 4 nonresponding if it was experiencing geometric effects, t Fuel movement would have been stopped if for a single bundle any channel showed greater than a factor of five change in count rate, or two channels
; increased by a factor of two, or if boron concentration decreased below 2,000 I
ppm, or if RHR temperature changed by more than 10*F. Any significant change in the trend of ICRR plots (indicating change in reactivity), or failure of a
, detector as signified by its count rate falling below 0.5 cps also would have j stopped loading.
i i _, Results The temporary and permanent nuclear detectors were used to successfully monitor the reactivity of the core as it was loaded.
. All objectives were met.
Problems There were no problems with the ICRR methodology or monitoring reactivity changes during fuel loading.
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t 4.4 INVERSE COUNT RATE RATIO MONITORING FOR CORE LOADING (1-500-02) (Continued): FIGURE 4.4.3: INVERSE COUNT RATE RATIO: TEMPORARY DETECTOR A Note: Sharp decreases in the ratio are due to geometry (see text). INUDSE COUNT RATE RATIO US CORE LOAD STD MIMBER 1.1 A A roe 0RARY DETECTOR A
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l 4.4 INVERSE COUNT RATE RATIO MONITORING FOR CORE LOADING (1-500-02) (Continued): FIGURE 4.4.4: INVERSE COUNT RATE RATIO: TEMPORARY DETECTOR C Note: Sharp decreases in the ratio are due to geometry (see text). INYDSE COUNT RATE RAfl0 US 1,0ADlHC STEP EMED 1.2 0 TDIPORARY DETECTOR C 1.1 1 n , n -M I.D .sm _ e k
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4.5 REACTOR SYSTEMS SAMPLING FOR CORE LOADING (1-5SJ-01) Objectives The objective of the Reactor Systems Sampling for Core Loading test was to sample the Reactor Coolant System (RCS) and directly connected auxiliary systems to verify correct and uniform boron concentration prior te irr irl core loading. The abstract for this test is FSAR Section 14.2.8.2.18. Methodology The RCS and directly connected auxiliary systems (chemical and volume control system (CVCS), RHR, and safety injection (SI)) were filled with 2000 to 2200 ppm borated water and vented in accordance with appropriate system operating procedures. Each auxiliary system was placed in recirculation and samples were obtained from various system locations. Baron concentration at each location was verified to be between 2000 and 2200 ppm. The reactor vessel was borated by ci.culating borated water via the RHR system. Samples were obtained from the operating RHR loop and in the reactor vessel at the following locations: water surface in the reactor vessel; approximately six feet beneath the surface; approximately fourteen feet beneath the surface; and at the lower core support plate. Circulation and sampling were continued until all locations were between 2000 and 2200 ppm and within 50 ppm of each other. This procedure did not apply to the transfer canal since one-time provision in the form of a footnote in the 5% power Technical Specification permitted the transfer canal to be filled with demineralized water. Results Boron concentration of the RCS and directly connected auxiliary systems was verified to be uniform and between 2000 and 2200 ppm prior to initiating core loading. Acceptance criteria and objectives were met. Problems There were no significant problems. 4-28
1 I D l SECTION 5 PRECRITICAL TESTING i i D 1 1 l 5-1
I l 5.1 PRECRITICAL TEST SEQUENCE (1-500-04) Objective w 1 ( The objective of the Precritical Test Sequence was to specify the sequence of ! events which constitute the precritical test sequence. I The abstract for ttis test is FSAR Section 14.2.8.2.23. I l Methodology The precritical test sequence specified tests that were performed to ensure 1 that the facility was in a final state of readiness to achieve initial criticality and perform low power tests. These tests commenced with the completion of core loading and ended with the plant at normal operating temperature and pressure ready to begin the approach to initial criticality. Testing began shortly after the reactor vessel head was fully tensioned and Mode 5 declared. The reactor was at ambient (<100*F) and about 350 psig. Initial checkouts were performed on the Control Rod Drive Mechanisms (CRDMs), drives for the Movable Incore Detector System (MIDS), Reactor Vessel Level Instrumentation System (RVLIS), and RCS temperature instrumentation. Data was gathered for the Thermal Expansion Baseline Measurement, Digital Metal Impact Monitoring System (DMIMS), and Gross Failed Fuel Detection System. Plant temperature was increase ( in a series of plateaus. Measurements based on plant conditions were made at 250*F, 345*F (just below Mode 3 entry temperature), 450*F, and hot standby temperature of 557*F. Measurements included thermal expansion, RCS temperature cross-calibration, RVLIS, and DMIMS. Normal operating procedures were used without difficulty during the heatup. The precritical tests are listed iri Table 5.1.1 and each test section has its own discussion. The RCS boron concentration remained at refueling concentration throughout precritical testing to ensure subcritf ality. Despite being at refueling concentrations, I reactivity monitoring continued on a limited basis during I initial rod withdrawls for CRDM and control rod position indication testing, control rod drop time testing, and rod control testing. The precritical test sequence also performed some required surveillance, including the required surveillance to allow control rod motion. Some tests including dynamic effects were part of an ongoing test series that is rept,rted in another section of this report.
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l 5.1 PRECRITICAL TEST SEQUENCE (1-500 04) (Continued): ) i l Results Precritical testing was successfully completed and all objectives were met. { Problems i Early in testing Reactor Coolant Pump #2 experienced hid h vibration (14-20 ' mi.'_ s ) . The plant was returned to Mode 5 and depressurized so the pump and motor could be realigned. Vibration was reduced to normal 1-3 mils after aligr'r en t . Recovery from the evolution was delayed because the absence of decay ' cat the RHR system cooled the reactor vessel and piping but the steam generators did not cool as rapidly. Time was required to cool the generators - to an acceptable temperature for reactor coolant pump restart. The Reactor Coolant System Heat Loss Test 1-5BB-07 was not performed during the precritical phase of testing. This test was delayed due to plant stability problems due to other testing but was performed at a convenient time shortly after initial criticality during the nuclear instrumentation systems outage. I l 5-3 L
r-5.1 PRECRITICAL TEST SEQUENCE (1-500-04) (Continued): TABLE 5.1.1: PRECRITICAL TESTS PROCEDURE _ NUMBER TITLE DESCRIPTION .-- 1-5AL-01 Auxiliary Feedwater Test Demonstrated proper operation of the Auxiliary Feedwater System. 1-5BB-01 RCS Final Leak Rate Measured RCS leak rate at hot standby. 1-5BB-02 Pressurizer Heater And Set the pressurizer continuous spray Spray Capability And valves; checked heater and spray valve Continuous Spray Flow effectiveness in raising and lowering Verification pressurizer pressure; collected data on i ' depressurization rates with auxiliary ! spray and simulated decay heat. l 1-5BB-04 Reactor Coolant System Measured RCS flow rate and calibrated Flow Measurement RCS flew transmitters. j 1-5BB-05 Resistance Temperature Determined f:,w rates in the RTD bypass Detector Bypass Loop lines and adjusted low flow alarm Flow Verification setpoints. l 1-5BB-06 Reactor Coolant Flow Measured decrease rate of RCS flow Coastdown and time delays associated with low flow reactor trip following a simultaneous trip of all four reactor coolant pumps. 1-5BB-07 RCS Heat Loss Measured insulation heat loss of the Measurement RCS and pressurizer with spray, and without spray. 1-5RJ-01 Proteus Computer Verified operation of the Offsite Operational Test Communications Link (OCL) and movable detector flux mapping sof tware. l 1-5RP-01 RVLIS Final Calibration Demonstrated that the system operates as And Operational designed, obtained plant heatup data, Checkout and confirmed predicted level indications. 1-5SE-01 Movable Incore Verified operation of the movable incore Detectors detectors and set limits for detector travel. D t 5-4
5.1 PRECRITICAL TEST SEQUENCE (1-500-04) (Continued): TABLE 5.1.1: PRECRITICAL TESTS (Continued): PROCEDURE __ NUMBER TITLE DESCRIPTION 1-SSE-02 Operational Alignment Calibrated the intermediate and power Of Nuclear Instrument- range detectors and set associated ation System Precritical alarms and trips. l-SSF-02 Rod Control System Verified proper operation of the rod control system. j 1-5SF-03 CRDM Operational Test Verified timing of the Control Rcd And Rod Position Drive Mechanisms (CRDMs) and accuracy of ' Indication Test the Digital Rod Position Indication ! (DRPI) during rod motion. 1-5SF-04 Rod Drop Time Measured time required for control rods i i to drop under hot full flow conditions from fully withdrawn and verified { deceleration. { ' 1-SSF-06* Operational Alignment Verified RCS temperature instrument of Process Temperature calibration in the 7300 process Instrumentation : D system, speciff cally RCS delta T. Isothermally ali ned 8 all RTDs. { I 1-SSF-09 RTD Mini-Cross Determined insulation and lead Calibration ; resistance of the RVLIS RTDs at ambient ' and hot standby conditions, and performed incore thermocouple temperature surveys. 1-SSQ-01* Metal Impact Monitoring Test Collected t'ata and verified calibration. 1-600-05* Biological Shield Survey Recorded baseline radiation measurements. 1-600-06* Dynamic Response Test Observed the auxiliary feedwater lines as auxiliary feedwater pumps were started and stopped. 1-600-11* Thermal Expansion Test Measured piping thermal growth during heatup. 1-600-12* Primary and Secondary Monitored the RCS and secondary chemistry Chemistry including feedwater. 5-5 l
5.1 PRECRIT10AL TEST SEQUENCE (1-500-04) (Continued): TABLE 5.1.1: PRECRITICAL TESTS (Continued): PROCEDURE 1 NUMBER TITLE DESCRIPTION 1-600-14* Ventilation Capability Verified heat transfer capability of the Test Emergency Safeguards System chillers. 1-600-15 Voltage Survey For Verified voltages on class 1-E 1 Class 1-E Busses (safety related) equipment during operation. These tests were partly performed during precritical testing i but continued into power ascension testing and are reported in l the power ascension testing portion of this report. I l l I 1 i V l l 1 i
- O 5-6 l
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5.2 AUXILIARY FEEDWATER TEST (1-5AL-011 Objective The objectives of the Auxiliary feedwater Test were to: (1) verify that the motor-driven auxiliary feedwater pumps (MDAFPs) could supply water to the steam generators within the design limits of the pumps; (2) demonstrate automatic operation of the MDAFPs and valves during simulated emergency Auxiliary Feedwater (AD) actuation; (3) perform dynamic tuning of the turbine's contrcl circuit; (4) verify that the turbine-driven auxilfary feedwater pump (TDAFP) could supply water to the steam generators within the pump's design limits; and (5) demonstrate the TDAFP's ability to cold-start ! i during a simulated AFW emergency initiation. This test completed preoperational tests described in FSAR Sections 14.2.8.1.5 (MDAFPs) and 14.2.8.1.6 (TDAFPs). l l Methodology Modifications to the auxiliary feedwater system following hot functional testing prompted reperformance of this test during precritical testing. The MDAFPs miniflow valves were tested to determine their capacity. These valves close as auxiliary feedwater flow increases. Performance tests were conducted ; on each pump's miniflow valve by opening the auxiliary feedwater valves for each pump and verifying sufficient flow wheu the miniflow valve closed. The auxiliary feedwater valves were then closed and the amount of flow at the B miniflow valve opening setpoint verified. The AFW actuation test was performed on each MDAFP and consisted of simulating an AIN actuation signal with the steam generators at no-load pressure and temperature, and the auxiliary feedwater valves closed. MDAFP pump discharge pressure and discharge valve positions were monitored by test recorders. A stop watch was used to time from AFW actuation until total flow was greater than 510 gallons per minute (gpm). Pump flow was then raised to 630 gpm and pump discharge pressure and flows were recorded to validate pump performance curves. Dynamic tuning of the TDAFP's turbine controls was performed for both local and remote controllers. The testing sequence was: (1) with the speed controllers in manual, the TDAFP was brought to half and then to full speed settings. At each speed, turbine and pump pressures and bearing temperatures were recorded. Speed was then adjusted to 4200 revolutions per minute (rpm) and vibration readings, speed, temperature and pressures were recorded for the pump and turbine. Speed was then adjusted te obtain 530 pounds per square inca differential (psid) between steam inlet z.nd pump discharge pressures and the turbine speed controller was then placed in automatic. Speed, pressures, and temperatures were monitored as automatic operation continued. l Proper (stable) operation of local and remote speed controllers was separately 5-7 l 1 l t _ _ _ _ _ _ - _ _ _ _ _ _ _ -
5.2 AUXILIARY FEEPWATER TEST (1-5AL-01) (Continued): and independently verified. For each controller, speed control was returned q to manual and decreased to obtain 330 psid, then increased to obtain 630 psid. j At each speed the controller was placed in automatic and proper operation verified. Turbine speed was then decreased to 1535 RPM and control isolation between local and remote stations was verified. Speed control was then transferred to the other station and the preceeding steps repeated. A cold quick start test of the TDAFP was performed. A cold quick start forces the TDAFP to automatically start from cold conditions. Recorders were attached to TDAFP's feedwater flow instruments and the AFW actuation signal. The steam generators were verified to be at no-load conditions to ensure maximum steam and feedwater backpressure, and the MDAFPs were secured. An AFW actuation signal was simulated and proper TDAFP starting and feeding of all four steam generators in less than sixty seconds was verified. TDAFP flow was then adjusted to 1175 gpm and total develcped pump head was calculated to verify d pump performance. 1 i Results ! { All acceptance criteria and objectives were met. j The MDAFP miniflow valves opened and closed properly to provide minimum flow l required by the pumps. The MDAFPs delivered 510 gpm to the steam generators l within the required sixty seconds following an AFW actuation. Time for train l I A's pump was 11.84 seconds and 12.78 seconds for train B's. Total developed ) head for train A's MDAFW pump was 3539.4 feet and 3540.7 feet for train B's. Both pumps met acceptance criterion of 3500 +105, -0 feet. l TDAFP dynamic tuning was satisfactorily performed. Proper operation of the speed controller and control circuits in manual and automatic modes was demonstrated. I Total developed head of the TDAFP was 3539.1 feet, satisfying acceptance l criterion of 3500 +105, -0 feet. I During the TDAFP cold quick start test the pump came to speed and delivered 510 gpm to the steam generators in fifty seconds, satisfying acceptance criterion of less than sixty seconds, 1 i i i Probleras j The transfer to local control portion of the TDAFP test was repeated due to recorder pen problems. Actual start time for the TDAFP quick cold start was visually observed to be much less than 50 seconds; however, due to problems using the Visicorder, the Emergency Response Facility (ERF) computer had to be used to provide hard data for demonstrating that acceptance criterion was met. After applying appropriate conservatism to the ERF data, a value of 50 seconds was obtained. 1% 5-8 l l l w_______-_____-_-__-_-__-_-__-__--___
j 5.3 RCS FINAL LEAK RATE (1-5BB-01) l Objectives The objective of this test was to determine the amount of identified and unidentified leakage from the reactor coolant system (RCS) and verify that the l l leak rate was within allowable limits. The abstract for this test is FSAR Section 14.2.8.2.1. I Methodology l i l The test was performed once during precritical testing on March 7, 1987. The normal surveillance procedure 14905-1, Revision 5, was performed to determine l RCS leak rate. The procedure stabilized RCS temperature and secured evolutions ! l that added or drained water from the RCS and associated systems. The associated systems included: reactor coolant drain tank; pressurizer relief tank; and volume control tank. Sampling was also stopped. Initial readings were taken for RCS temperature and pressure and the various associated system tank levels, l After two hours RCS temperature was restored within 0.5'F of initial l temperature and final readings of temperature, pressure, and tank levels were made. Corrections were made to adjust for temperature and pressure changes and the final leak rate was calculated. ~ Results l The objective wm met and the acceptance criteria were met. Total leakage was 4.91 gpm. This included identified leakage from an RCP seal injection filter isolation valve. Unidentified leakage was calculated to be 0.76 gpm. These leakage rates met the acceptance criteria of less than 10 gpm identified and less than 1.0 gpm unidentified leakage. Problems After this test was performed it was discovered that improvements to the surveillance procedure were needed. The modified procedure has successfully been used to determine RCS leak rates. No retest was required because the procedure was verified during normal plant operation. l 5-9
5.4 PRESSURIZER HEATER AND SPRAY JAPABILITY AND CONTINUOUS SPRAY FLOW VERIFICATION (1-5BB-02) _l l Objectives ! i Objectives of the pressurizer heater and spray' flow capability and continuous spray' flow verification test were to: establish the optimum continuous minimum spray flow rate; determine effectiveness of the pressurizer-heaters and normal i control spray; demonstrate depressurization rate by turning off pressurizer. l heaters and using auxiliary spray,. and opening the Power Operated Relief Valves (PORVs); and cellect data for benchmarking the training simulator. The abstract for this test is presented in FSAR Section 14.2.8.2.2 and discussed in FSAR Question 640.04. 3 i Methodology This precritical test had many parts, each discussed separately below. Optimum Minimum Spray Flow Optimum spray flow was established by adjusting each manual bypass flow spray valve separately. The valves were first closed, then opened in approximately one-eighth turn increments until spray line temperature as determined by installed spray line temperature instruments remained essentially constant. D The valve opening where temperature ceased rising rapidly and started to become constant (called the " knee" in the temperature curve) was marked as the J optimum opening and the valve was opened to that position. Spray line temperature was then monitored to ensure that maximum temperature difference of 125 F between the spray line and pressurizer was not exceeded. The spray line bypass valve opening versus spray line temperature adjustment is illustrated in Figure 5.4.1. 1 1 ! Pressurizer Normal Spray Effectiveness ; The pressurizer normal control spray effectiveness test established steady state conditions, turned off all pressurizer heaters, and then opened' both spray valves together. The resulting depressurization rate was then' compared, against upper and lower depressurization limits as shown in Figure 5.4.2 to determine satisfactory flow. I I D 5-10 _ _ _ _ _ _ _ _ _ . _ _ _ _ __________________-_- - _____-_-____-_-__ a
5.4 PRESSURIZER HEATER AND SPRAY CAPABILITY AND CONTINUOUS SPRAY FLOW VERIFICATION (Continued) (ll5BB-02) Pressurizer Heater Effectiveness Pressurizer heater effectiveness was determined by establishing steady state conditions, closing the spray valves, and adjusting automatic pressure control to 2300 psi. The rate at which pressurizer pressure increased from nominal 2235 psig to 2300 psig was then compared to upper and lower limits as shown in Figure 5.4.3 to determine heater effectiveness. )
)
Pressurizer PORV Blowdown Pressurizer PORV blowdown, response time, and leak inte8rity testing was , performed on each POR\ separately. This test involved connecting a chart i recorder to the open and closed signals to and from the valve. The PORV was ! then opened using the main control board handswitch and a blowdown ] (pressurizer pressure reduction) of approximatley 200 psi was conducted. The time for the valve to fully open when given an open si nal 8 was then determined from recorder trace. After the PORV was shut a leakage test was performed by monit; ring downstream pipe temperature for a decrease below the high temperature annunciator setting. 1 Depressurization Rate and Margin To Saturation The Depressurization Rate and Margin To Saturation portion of this test was l (
\
intended to provide training for reactor operators and information for I benchmarking the training simulator. A limited number of operators were i gathered in the control room prior to the start of the test. I 1 Depressurization rate and control of margin to saturation testing was performed in three steps. In all parts, plant conditions were first stabilized I and all but one reactor coolant pump was secured to simulate decay heat. In the first part, all pressurizer heaters were deenergized and reactor coolant I system pressure was allowed to decrease 100 psi. In the second part, plant conditions were stabilized with pressurizer heaters in automatic and without normal spray flow. Auxiliary spray flow was used to decrease pressure 100 psi. l l l For the third part, ylant conditions were stabilized and the steam generators were fed with the turbine driven auxiliary feedwater pump while reactor pressure was allowed to drop 100 psi. Results l The pressurizer minimum flow spray valves were successfully adjusted approximately one cuarter turn each to maintain spray line temperature of 546 F (for valve 1-1201-U4-090) and 547'F (for valve 1-1201-U4-091). 5-11 l
- 5.4 PRESSURIZER HEATER AND SPRAY CAPABILITY AND CONTINUOUS j)
SPRAY FLOW VERIFICATION (Continued) (1-SBB-02) The pressurizer normal spray effectiveness test decreased pressurizer pressure-200 psi in 102 seconds, within the acceptance criceria of 65 to 125 seconds. The pressurizer heater effectiveness test resulted in a pressure increase of j approximately 60 psi in 200 seconds, within acceptance criteria of 47 to 95 j psi in ~200 seconds. { 4 Overall opening times for PORVs were well within the two seconds or less - acceptance criterion. PV-0455A opening time was'O.25 seconds and- PV-0456A opening time was 0.'24 seconds. PORV PV-0456A was verified leaktight. 1 The depressurization rate and mar 18 n to saturation test was completed satisfactorily and operators observed the test. All objectives were met. 1 j Problems Pressurizer spray valve adjustments for optimum minimum spray valve was j initially completed with a lower spray line temperature than desired (with 1 the spray valve in a more closed position) to keep from using backup heaters. It was then determined that the auxiliary spray valve was leaking so the D auxiliary spray valve's manual isolation valve was closed and a retest' was performed to reset the valves to a better position and more acceptable temperature of approx 1mately 547"F. The short duration of the pressurizer heater effectiveness test prevented pressurizer heater powers from being measured so the latest surveillance test data was attached to the procedure, i Problems were encountered when PORV PV-0455A was tested., The energency' i response facility (ERF) computer failed to record blowdown data. The main control board (MCB) indication of valve position operated erratically and did not give proper indication. The MCB indication was reworked and functioned properly when PORV PV-0455A was retested. Blowdown data recorded on the ERF computer then indicated proper blowdown and control board indication functioned properly. Leak integrity of PORV PV-0455A could not be verified due to excessive packing leakoff from the upstream PORV block valve. This leakoff enters the piping downstream of the PORV and resulted in high downstream pipe temperatures. The block valve was shut and later repacked. There were delays in conducting the depressurization rate and margin to saturation test due to ERF computer problems that were resolve 6. 5-12
5.4 PRESSURIZER HEATEV AND SPRAY CAPABILITY AND CONT 1dUOUS-SPRAf FLOW VERIFICATION (Continued) (1-5BB-02) O FIGURE 5.4.1: PRESSURIZER MINIMUM FLOW SPRAY VALVE OPENING 1 PRESSURIZER MINIMUM SPRAY UALUE OPINING j 550 549 549 O_UALVE .1-1291-04-991 > 547- n. = ~~ 546 / ._.- 545 / 544 / ! 54.3 / /
$542 - N i 541 / /
W 546 / 539 / 538 [ j j~ c VALUE 1-1291-04-999 536 / h / 535--
!534 /
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> 532I e _
K 531 539 bl .b . '1 .l5 . '2 .d5 .3 .35
.4 .45 .d TURNS VALUE OPENIHC 5-13
! l 5.4 PRESSURIZER HEATER AND SPRAY CAPABILITY AND CONTINUOUS 'i SPRAY FLOW VERIFICATldN (Continued) (1-5BB-02) 1 j '~) FIGURE 5.4.2: PRESSURIZER SPRAY EFFECTIVENESS TEST l l I l l .; PRESSURIZER PRES $URE RESPONSE TO BOTH SPRAY YAl#ES I 24e '! (Q
, xg 0 TEST RESULTS l
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c , 5.4 PRESSURIZER HEATER AND SPRAY CAPABILITY _A1D CONTINUOUS SPRAY FLOW VERIFICATION (Continued) (1-5Bb-02)
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FIGURE 5.4.3: PRESSURIZER HEATER EFFECTIVENESS TEST i 1 PRESSURE RESPONSE TO ALL PRESSURIZER HEATERS 2348 0 TEST RESULTS UPPER DEVIAfl0N LIMIT 2339 - q 2329 2310 - \ 2,,, / A i a2290
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5.5 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (1-5BB-04) Objectives The objectives of the Reactor Coolant System Flow Measurement test were to: (1) demonstrate adequate full flow during precritical conditions; (2) demonstrate 75% power; adequate full flow during power range testing prior to exceeding and (3) recalibrates the RCS flow transmitters following the precritical measurement to read 100% prior to initial criticality. I The abstract for this test is FSAR Section 14.2.8.2.3. Methodology Flow was measured during precritical testing using elbow taps and by calorimetric during power range testing. I Precritical Measurement Method' ology Flow was measured during precritical testing using differential pressures developed by elbow taps on cold leg piping between the steam generctors and reactor coolant pumps. Since flow is proportional to the square root of differential pressure, a simple relationship was used to correlate flow and differential pressure. Differential pressure transmitters connected to the taps were initially calibrated for a single value of 550 inches water column at 68*F for full scale of 120% flow. This preliminary calibration corresponded
.- ) to an estimated flow of 112,000 gpm at indicated 100% flow.
As-found flow was determined from the average of ten consecutive Proteus computer readings of percent flow multiplied by 1,120 8pm/ percent. Main control board readings were also compared to the Proteus. The transmitters were recalibrates following formula: based on original scaling of 550 inches of water by the i 2 120% Flow d/p = 4.387E-08 x (As-Found Full Flow In GPM) Flow and differential pressure determined during precritical testing are
. listed in Table 5.5.1. Flow for each loop and total flow were determined.
1 Am
) 5-16
5.5 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (1-5BB-04) (Continued): At-Power Measurement Methodology The required flow determination was made at 50% power although flow was measured at every major power plateau (nominal power levels of 30%, 50%, 75%, 90%, and 100%). The at-power measurement used a variation on primary calorimetric power determination in conjunction with data gathered during secondary calorimetric determination of reactor power in the Thermal Power Determination and Statepoint Measurement test, procedure 1-5SC-02. This data included differential pressures developed by feedwater flow venturies, feedwater temperature near the venturies, feedwater and steam generator pressures, steam flow differential pressure, and reactor coolant system narrow range hot and cold leg temperatures. Differential pressures were measured by special high-accuracy differential pressure transmitters and temperatures were measured by permanently-installed narrow-range RTDs. Feedwater flow and temperature, steam generator pressure, a constant RCS net heat input from reactor coolant pumps, and heat losses were used to determine power in each reactor coolant loop. Enthalpy rise across each loop was determined from RCS temperatures at a constant pressure of 2250 psia. Loop power divided by loop enthalpy rise produced loop flow rates and loop flow rates were summed to produce total flow. Results RCS flow exceeded minimum requirement of 95,700 gpm per loop. Test results were also used to satisfy the surveillance requirement of Specification 4.2.5.2. Technical Precritical test results are shown in Table 5.5.1: All objectives and acceptance criteria were met. Problems . No significant problems were encountered. 5-17
)
5.5 REACTOR COOLANT SYSTEM 310W MEASUREMENT (1-5BB-04) (Continued'): TABLE 5.5.1: PRECRITICAL RCS ELBOW TAP DIFFERENTIAL PRESSURES AND FLOWS ; I l Transmitter Inches D/P Water Column j Loop Number. 100% Flow 120% Flow
- GPM Flow l
1 F0400 .313.7 451.7 101',472 F0401 340.4 490.4 728 i F0402 331.2 476.9 105,384 104, Loop 1 Average: 103,846 2 F0420 331.9 478.0 104,384 l F0421 313.7 451,7 101,472
- F0422 324.9 467.8 103,264 .
l Loop 2 Average: 103,006 ) 1 i 3 F0430 311.6 448.7 101,136 I F0431 327.0 470.9 103,600 F0432 307.5 442.8 100,464 j Loop 3 Average: 101,774 4 F0440 347.1 499.8 106,736 F0441 378.3 544.8 111,440 F0442 331.3 477.0 104,272 loop 4 Average: 107,890
)
i Total of Loop Averages = RCS Total Flow: 416,763 gpm Full range as-left instrument calibration (120% of full flow). 1 5-18
=l j
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5.5 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (1- 5BB-04) (Continued): TABLE 5.5.2: RCS FLOW RESULTS Nominal Actual GPM Flow Rates Plateau Power Loop Flows Power % Level % 1 2 3 4 Total _ 0* 0 103,846 103,006 101,740 107,464 416.057 30 34.4 104,017 104,668 99,391 104,240 412,316 50 45.3 103,881 101,103 103,830 104,171 412,985 75 72.7 101,870 100,380 102,410 106,800 411,460 90 90.4 101,653 99,979 101,737 105,998 409,367 - 100 99.4 102,132 100,347 101,387 106,352 410,217 l
- Flow measured during precritical test.
l l m 5-19
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1
5.5 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (1-5BB-04) (Continued): FIGURE 5.5.1: TOTAL RCS FLOW VERSUS POWER LEVEL U0GTLE I RCS TOTAL FIAW DURING INITIAL STARfuP 417 l TLOW Af 9:/. TROM EL90W TAPS 416!i $ FLOW Af POWER BY CALORINEfRIC l l 415 l l l 41t l
"3 0 l U2 i 3 3 e 4 til i E l &
U 6 41 9 a o Z g l499 3 0 H 498 11 $ l'8 1h 2'8 2h 3'8 3h 48 45 5'8 5h 6'8 6h Y9 7h 8'8 8'5 9'8 951911 PERCDff POWER Note: The zero power level point is from the precritical test. 5-20
5.6 _ RESISTANCE TEMPERATURE DETECTOR BYPASS LOOP FLOW VERIFICATION (1-5BB-05) Objectives Objectives of the Resistance Temperature Detector (RTD) bypass loop flow verification test were to: (1) determine the flow rate required to achieve a reactor coolant transport time of 1.0 seconds in each RTD bypass loop; (2) verify acceptable transport time for each loop; and (3) establish the setpoint and verify proper low flow alarm initiation and reset. The abstract for this test is FSAR Section 14.2.8.2.4. Methodolo8y The temperature of fluid in the reactor coolant system (RCS) hot and cold legs is measured by RTDs mounted in piping, called bypass loops, connected to RCS piping. These RTDs are used for reactor protection and control. Each loop has its own hot and cold leg bypass line and manifold. The hot leg line is attached to the hot leg at the hot leg scoop tube and leads to RTDs mounted in a portion of the pipe called a manifold. The cold leg line is attached to the cold leg downstream of the loop's reactor coolant pump and leads to a separate manifold. The two manifolds discharge to a common pipe that returns water to RCS cold leg piping bctween the loop's steam generator and reactor coolant pump. A flow sensing element (an orifice) is mounted in the common piping and is used to determine flow rates and for the low flow alarm. The required RTD bypass loop flow rate was determined from pipe length measurements recorded in Preoperational Test Procedure 1-3BB-01. Calculations established the minimum flow required to achieve a reactor coolant transport time of 1.0 second in each RTD bypass loop from the loop connection at the RCS piping to the last RTD in the manifold. These required flow rates are listed in Table 5.6.1. Acceptable transport time verification was performed by measurement of the actual flow rate in each RTD bypass loop to verify that the actual flows met or exceeded the flow necessary to achieve the required transport time. Each loop was measured separately and independently of the others. Total loop flow through the flow element was first recorded. The low flow alarm setpoint was then adjusted to actuate at 90% of total loop flow rate. Flow was then throttled using the hot leg RTD manifold isolation valve until the low flow alarm actuated. The actuation flow rate was recorded. The valve was then fully closed and the remaining cold leg RTD bypass flow recorded. The hot leg manifold isolation valve was then fully opened and the low flow alarm reset was recorded. The cold leg manifold isolation valve was closed and the remaining hot leg RTD bypass flow recorded. 5-21 i - _ _ - _ . k
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5.6 RESISTANCE TEMPERATURE DETFCTOR BYPASS LOOP FLOW VERIFI'l,hTION ' (1-5BB-05) (Continued): $ s a wN m f , L , 3 4
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Actual hot and cold leg RTD bypass flow rate 2!are listed in Table 5:6.'1 and'o were compared to calculated flows. **.
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Flow rates required to produce a one second or less ihancfds rise ~ar.d as-found flow rates are listed in Table 5.6.1. The as-found rateb nessded required c, d s i s s , t Each loop low ficw alarm setpoint actuated within l'.N% or' the setpoint,
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'g, TABLE 5.6.1: RTD'sYPASS. Eft)FS REQUIRED AND AS-FOUND FLOW RATES s
3- ' COLD LEG HOT LEG
. LO")P # h!NLMUM RE_ QUIRED /.5-FOU'ID
- - MINIMUM REQUIRED AS-FOUND
{ 1 82.1 gpm 92 gpm 117.6 gpm 15'1 gpm
. 2 82.5 gym 94 gpm .
l'!O.7 gpm 182 gpm
; 3- 33.9 gpm ~94 gpm ,
120.3 gpm 173 gpm 4 821.6 gpm 92 gpm 120'3 gpm 179 gpm A f s 4 4 4 4 i e b g
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5.7 REACTOR COOLANT FLOW COASTDOWN (1-5BB-06) Objectives The objectives of the Reactor Coolant System (RCS) Flow Coastdown test were to: (1) measure the rate at which reactor coolant flow changes, subsequent to tripping all reactor coolant pumps; and (2) measure various delay times associated with the loss of flow accident. The abstract for this test is FSAR Section 14.2.8.2.5. Methodology The test verified that the rate of decrease of reactor coolant system's flow, also known as coastdown, following simultaneous deenergization of all four reactor coolant pumps was slower (more conservative), and that time delays in the flow-sensing circuitry that trips the reactor were shorter (more conservative) than assumed in the analysis. The test consisted of a single coastdown initiated by simultaneously tripping all four reactor coolant pumps from stable hot stcndby conditions of 557 F Tavg and 2235 psig pressurizer pressure. RCS flow differential pressure transmitters and trip circuitry bistables and reactor trip breakers were monitored by strip chart recorders. The Emergency Response Facility (ERF) computer's Sequence Of Events (SOE) also monitored reactor coolant pump trip and reactor trip breaker opening times. Voltages from the strip chart recorders were used to determine flow coastdown time constants, sensor delay times, and breaker opening times. Special equipment was required for the test. A temporary test switch was installed and wired parallel to the "stop" contacts of each reactor coolant pump's main control board control switch to simultaneously trip all four reactor coolant pumps. More than thirty inputs were connected to six high speed strip chart recorders. Most of the inputs were signals from the reactor protection circuitry including three analog flow signals from each of the four reactor coolant flow transmitters, a total of twelve, and digital trip signals from all twelve low flow trip bistable signals were added for information. The other signals were reactor coolant pump (RCP) breaker positions and reactor trip breaker (RTB) positions. Reactor coolant pump breaker position showed when the reactor coolant pumps had been tripped and the RTBs showed when the plant tripped. These trip signals and the analog flow signals were required for the test. In order to provide breaker positions to the recorders and normal still retain position indication as required by the reactor operators, the "open" indicating lights were replaced by a light emitting diode (LED) and resistor. The LED provided visual indication and the resistor provided a voltage signal for the recorders. At-power operation was simulated so the protective circuitry would trip on low flow. This signal was supplied by removing the 118 vac supply to the P-8 bistable to enable the single loop low flow reactor trip. 5-24
I b.7 REACTOR COOLANT FLOW COASTDOWN (1-5BB-06) (Continued): A computer trend was initiated for selected points on the ERF. Pre-trip RCS parameters were recorded, the recorders were started on high l speed, and after a 10 second countdown the four reactor coolant pumps were { simultaneously tripped by the trip switch. { l Data analysis and reduction was performed using the strip chart recordings and l a printout of the SOE. A method recently used at Northeast Utilities Millstone j III Nuclear Power Plant, the " sensor by sensor, loop by loop" method was used ' to more accurately reflect response of the plant to actual process conditions. This method adds the T1 (the time from when meas'ared loop flow has decreased to the low flow trip setpoint until the last reactor trip breaker has changed state) and Td (sensor time delay for each loop) for each loop sensor. The i maximum value is added to Tg (the time required for the control rod dri ve motor grippers to release) to produce T1f (the total low flow delay time). l Flow coastdown was compared to a curve specifically calculated for the faster coastdown due to operating at hot standby conditions. The FSAR Chapter 15 i (accident analysis) curve was not used because at-power coastdown is not representative at hot standby test conditions, j Overall flow coastdown is plotted in Figure 5.7.1. The inverse flow coastdown l curve is plotted in Figure 5.7.2. D Results All objectives and acceptance criteria were met. Reactor flow from 3 to 10 seconds after tripping the reactor coolant pumps had an inverse time constant of 13.687 seconds. The acceptance criteria was >11.58 seconds and flow coastdown criteria was met. l The sum of flow coastdown, sensor time delay, and gripper release time l comprise the low flow trip time delay. The maximum of T1 and Tg was 0.903 seconds for loop number four's FT-444. Gripper time delay, Tg, was 0.033 i seconds for a total Tif of 0.936 seconds. This total u.s less than the 1.0 second acceptance criteria. Problems A total of fcur coastdowns were performed due to scaling problems with the recorders and loose connections on the reactor trip circuit breaker open light temporary fixture. The sequence of events (SOE) record and data from the fourth trip were sufficient. Manual data analysis was laborious and resulted in several minor arithmetic errors that did not affect passing the acceptance criteria. (A personal computer program has been developed for possible use on I the Vogtle Unit 2 startup.) D 5-25
5.7 REACTOR COOLANT FLOW COASTDOWN (1-5BB-06) (Continued): FIGURE 5.7.1: RCS FLOW COASTDOWN U0CILE 1 TOUR RCP TLOW C0A$fD0fM Af H0T STA@M 1" A PREDICTION FOR HOT STADM
.9s n U0cTLE 1 DATA Af HOT STANDM
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4 4 4 4 ,,, SECONDS flME FROM RCP IRIP 5-26
5.7 REACTOR COOLANT FLOW COASTDOWN (1-5BB-06) (Continued): FIGURE 5.7.2: INVERSE FLOW FRACTION INVERSE FLOW FRACTION, 4 RCP FLOW C0ASIFRt
.75
.7 6 SLOPE : 9.07314/SEC : 1/ TIME CONSTANT
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5.8 RCS HEAT LOSS MEASUREMENT (1-5DB-07) Objectives The objectives of the Reactor Coolant System (RCS) Heat Loss Measurement test were to determine: (1) heat loss through the insulation of the RCS and pressurizer; (2) heat loss due to combined pressurizer minimum spray and insulation; and (3) heat losses and gains to process systems connected to the RCS. Measured RCS heat loss was used in the Natural Circulation Demonstration test (1-600-10) to determine core decay heat and natural circulation flow rate. Values determined during the measurement are used in calorimetric determination of reactor power. There was no FSAR abstract for this test. I Methodology The heat loss tests quantified heat sources and sinks which, in a steady-state system, must be equal. The RCS heat loss measurement was conducted prior to power operation to avoid difficulties in quantifying decay heat. RCS heat loss was measured separately from the pressurizer. Each ir discussed separately. I RCS Heat Loss During the RCS test, heat inputs included power entering the RCS from the s reactor coolant pump (RCP) shafts, charging and seal injection flow, and pressurizer heater input minus insulation heat losses. Heat losses included ; letdown flow, seal thermal barrier cooling, and steam generator heat " rejec. ion. All sampling was stopped during this test to minimize unquantifiable heat losses. RCS heat loss inputs and results are listed below in Table 5.8.1. All power entering the RCS from the RCP shafts immediately or eventually degrades to heat. Shaft power was determined by measuring motor voltage, current, and power factor and using the motor manufacturer's value for efficiency (92%) under the existing load. RCP shaft power was: 5 x Voltage x Amperage x Power Factor x Efficiency All four RCP shaft powers were summed for total RCP heat input, by far the largest single heat input. Fluids entering the RCS are considered to bring heat even though they are cooler than the RCS. These included charging flow and seal injection flow, whose flow rates and temperatures were measured at the beginning and end of the RCS heat loss portion of the test. Fluids leaving the RCS are heat losses. These include letdown flow and seal bleedoff, whose temperatures and flow rates were also measured at the beginning and end of the test. The envelope for charging and letdown was the regenerative heat exchanger. O 5-28
l 5.8 RCS HEAT LOSS MEASUREMENT (1-5BB-07) (Continued): l Another heat loss was RCP seal thermal barriel cooling. Flow rates and ; temperature rises across all four RCPs were measured. A "steamdown" method was used to quantify the largest single heat loss, heat ; rejected through the steam generators. This method measured the rate of l decrease in steam generator water level after the generators had been allowed l to become saturated and the steamdown rate was linear, and calculates heat required to change the water lost to steam. This heat loss is the enthalpy change from saturated water to saturated steam of the mass of water lost during steamdown. The generators were first filled to approximately 60% narrow range water level and all feedwater, blowdown, and sampling were isolated. Levels increased slightly as the water in the generators came to saturation. The atmospheric relief valves were used to establish constant RCS temperatures and steam j generator pressures. Water levels then decreased over approximately forty ) i minutes in response to their respective heat loads. Figure 5.8.1 is a plot of steam generator water level during the steamdown. At the end of the test auxiliary feedwter was slowly reestablished and blowdown was restored. I Pressurizer Heat Loss Pressurizer heat loss was measured by manually cycling a known heat input, one bank of backup heaters, to maintain pressure within a small (approximately 15 l psi) band about the normal operating pressure of 2235 psig while maintainin8 l pressurizer level constant. Integrated heater electrical energy input divided D by test time was heat loss. i l The test consisted of two parts. The first, pressurizer heat loss with spray, ! measured heat required to warm incoming normal minimum spray flow and balance insulation heat losses and required approximately three hours. The second, pressurizer heat loss without spray, required spray flow to be isolated by closing the minimum flow spray valves and measured only insulation losses and required approximately two hours. Results All objectives were met (there was no acceptance criteria). RCS insulation heat loss was measured to be 9.688E6 BTU /hr (2.84 Mwt). Net heat input to the RCS other than the reactor core (heat sources, mainly RCPs, j minus beat sinks excluding the steam generators) was measured to be 5.529E7 l BTU /hr (16.205 Mw). Pressurizer heat loss without spray was 3.53E5 BTU /hr (103 Kw); with spray it was 1.658E6 BTU /hr (486 Kw). Problems No major problems were encountered. 5-29 l
5.8 RCS HEAT LOSS MEASUREMENT (1-5PB-07) (Continued): TABLE 5.8.1: O PARAMETERS DURING RCS HEAT LOSS TEST RCP MOTOR POWER: RCP #1: 5.376 Mw RCP #2: 5.712 Ifw RCP #3: 5.376 Mw RCP #4: 5.952 Mw HEAT INPUTS: RCP Energy At Shaft: 7.039E7 BTU /hr (20.63 Mw) Pressurizer Heaters: 9.82 E5 BTU /hr ( 0.272 Mw) Chargin8 Flow: 8.23 E6 BTU /hr ( 2.412 Mw) RCP Seal Injection: 1.382E6 BTU /hr ( 0.405 Mw) TOTAL HEAT INPUT: 8.09 E7 BTU /hr (23.719 Mw) l l HEAT LOSSES: Steam Generator:* 5.529E7 BTU /hr (16.205 Mw) l RCP Seal Leakoff: 3.79 E5 BTU /hr ( 0.111 Mv) Letdown Flow: 1.451E7 BTU /hr ( 4.253 Mw) RCP Thermal Barrier: 7.83 E5 BTU /hr ( 0.229 Mw) Pressurizer Loss: 3.53 E5 BTU /hr ( 0.103 Mw) RCS Insulation Loss:** 9.69 E6 BTU /hr ( 2.84 Mw) O
- Steam generator heat loss was measured during the steamdown.
RCS insulation loss was the net of all sources minus all other losses. l 5-30 l l I _________________a
B 5.8 RCS HEAT LOSS MEASUREMENT (1-5BB-07) (Continued): FIGURE 5.8.1: STEAM GENERATOR WATER LEVEL DURING STEAMDOWN i i RCS HEAT LOSS TEST, V0CTLE I, 3-19-87 79 A STD M CINERATOR 51 LEVEL 6 65I . 1 ST D M CENERATOR 82 LEVEL 0 STDM CINERAf0R 83 LEVEL l STDM CDERATOR 54 IEEL X 55 AAA- - - 4 W "A l 50 A- 4 L 4 ,* a ' h 2
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v 11 $ l'8 15 2'O 2'5 39 35 IS A5 5's 5'5 6'8 6'5 76l
!!ME IN MlHilfES FROM 1924 CST, 3-10-07 Note:
The faster steamdown rate of Steam Generator #4 is believed due to its greater flow rate (as reflected by its RCP power). l 5-31
5.9 PROTEUS COMPUTER OPERATIONAL _ TEST (1-5RJ-01) ~ ; n Objectives-b Objectives of the Plant Computer Test were to: (1) test the Offsite i Communications Link (OCL) software and hardware to ensure that the plant computer could both send and receive incore data to and from the Atlanta Data , Center; and (2) test the :uoveable detector flux mapping software- to ensure l that the plant computer could properly recognize and process data. received from the flux mapping panel.
]
i The abstract for this. test is FSAR Section 14.2.8.1.109.A.1 & 3. I Methodology Offsite Communications Link Testing of the OCL verified that the plant computer could properly format, transmit, print, or punch the incore data and then receive the results. Test case data was entered into the computer and was formatted and output on the desired device. This output was compared to the desired format to verify that the computer was picking up the correct variable and placing it in the correct location in the formatted output stream. The communication link was tested by transmitti:g this data echoed back to the desired data format. All options of the OCL package were exercised to ensure that the Vogtle Unit 1~ site could transmit data, initiate incore calculations at the Atlanta Data Center, and 1 receive the results. O Flux Mapping Software Testing the Flux Mapping Software involved manipulating the selector switches on the deenergized Flux Mapping Panel while simulating flux maps by simulating ' detector voltages with a power supply at the plant computer terminations. All thimble identifications, path selections, and error checking capabilities.were tested. The flux mapping software calculations were also tested by injecting known voltages and comparing the outputs to known results. Results ! The objectives and acceptance criteria were met. The OCL and Flux Mapping software were satisfactorily tested. Problems Minor software problems were discovered but were easily corrected. 5-32
5.10 RVLIS FINAL CALIBRATION AND OPERATIONAL CHECKOUT (1-5RP-01) Objectives The objectives of the Reactor Vessel Level .strumentation System (RVLIS) Final Calibration and Operational Checkout Test were to: (1) demonstrate that the system operates as designed by simulating various combinations of pressure and temperature inputs and observing that the output of the system, as displayed on * .e control room Primary Safety Monitoring System (PSMS) displays as predicted; (2) obtain plant heatup data so that plant specific curves representing wide range differential pressure versus reactor coolant temperature and primary system pressure are created; and (3) confirm predicted level indications with different combinations of reactor coolant pumps operating. j This test demonstrated that the system is installed and functions as described in FSAR Appendix 4A, Section 4A.1.C. Methodology = Thirteen tests were performed on the two similar but independent RVLIS systems. Input parameters were simulated over their range and the PSMS display's digital readouts for reactor vessel were observed and verified to be as expected. These inputs consisted of three differential pressure measurements pressures, taken across the reactor vessel at various levels, two RCS one RCS temperature, seven RVLIS sensing line compensating RTDs, RCP " running" signals, and three line isolator switch contacts. During plant heatup the following data was taken at least every 50 F: two RCS pressures; one RCS temperature; seven RVLIS sensing line temperatures; and three reactor vessel differential pressures. Data was recorded on a data logger and transmitted to Westinghouse for final scaling of the dynamic level portion of the system. Prior to heatup, and again with the RCS at operating tempera *.ure and pressure, the dynamic range reactor vessel levels were recorded with various combinations of operating reactor coolant pumps. 5-33
5.10 RVLIS FINAL CALIBRATION AND OPERATIONAL CHECKOUT (1-5RP-01) l (Continued): 1 Results
]
This test showed that the RVLIS subsystem of the PSMS system operated properly under plant conditions and field input constants had been entered correctly. Data taking during heatup was adequate to determine the plant-specific constants that were then entered into the RVLIS. Close correlation was shown for dynamic range reactor vessel levels read prior to heatup and again at plant operating conditions for various i combinations of reactor coolant pumps. ! I
.All objectives and acceptance criteria were met. I I
Problems Readings for upper and full range reactor vessel level were taken at RCS ! temperatures of 200 F and 557 F with no RCPs operating. Acceptance criteria required that level readings be within 3% of each other, but readings were 104% and 108%, respectively, a difference of 4%. The problem was investigated and it was determined that the instruments could be used as-is and the criteria is valid when cross-comparing on a train-to-train basis under the same temperature conditions. The problem occured because under cold conditions the higher density of the reactor coolant water overranges the differential D pressure tran mitter because it is spanned from the bottom to the top of the reactor vessel. t 1 5-34
5.11 MOVABLE INCORE DETECTORS (1-5SE-01) Objectives Objectives of the Movable Incore Detector Test were to: (1) set up and demonstrate the operation of the incore instrumentation system; and (2) verify adequacy of the incore instrumentation system for flux mapping. The abstract for this test is FSAR Section 14.2.8.2.9. Methodology l l The incore system consists of six drive units, six ten-path devices, six five-l path devices, incore instrument tubing, all located in the containment, and a i control panel located in the control room. This procedure prepared the movable incore detector system for normal operation by setting path length limits for the top of core, bottom of core, and storage path. An initial test was performed on each drive unit using a dummy cable. The manual mode of operation was used to traverse each path associated with each drive unit with the dummy cable. These paths included the Emergency, Storage, Common, and Calibrate paths. After successful tests with the dummy cable each l drive unit received its own operating incore detector. The operating cables , I were inserted and withdrawn using the Automatic mode of operation, simulating ( N a full core flux map, and including the Emergency, Common, paths. and Calibrate , l l l Data taken during the Manual and Automatic traverses determined the lengths of insertion and withdrawal for each path (Top of Core and Bottom of Core limits). The insertion and withdrawal limits determined from the traverses were then fixed using Veri-Selector tabs. q Results All objectives and acceptance criteria were met. The movable incore detector drive system was prepared for operation and demonstrated to be capable of operation in minual and automatic modes. Limits for the Emergency, Storage, Common, Calibrate, and Core Top and Core Bcttom were determined and fixed. Circuitry between the detector cables, control panel, and Proteus plant computer was tested. 5-35 l
l 5.11 MOVABLE INCORE DETECTORS (1-5SE-01) (Continued): Problems During the dummy cable traverses, several paths for each drive unit exhibited , blockage. For some of these blocked paths, repeated cycles of manual ! withdrawal / insertion cleared the blockage. For the remainder this method was ! not effective. Because of the number of blocked tubes, it was decided to clean and lubricate with Neolube all incore tubes, taking approximately three days. The Drive A detector cable intermittently jammed but operated satisfactorily ; after it was replaced. Inspection of the jamming cable revealed a burr on the cable that had prevented free movement. ; I l 1 3-36
l l 5.12 OPERATIONAL ALIGNMENT OF THE NUCLEAR INSTRUMENTATION SYSTEM PRECRITICAL {1-5SE-02) D Objective l The objective of the Operational Alignment of Nuclear Instrumentation System test was to establish and determine voltage, trip, operational, and alarm settings, of source range, intermediate range, and power range instrumentation prior to initial criticality. In addition to establishing initial trip i setpoints for all source range, intermediate range, and power range channels; the initial gain settings for the flux deviation averaging amplifiers were also set. The High Flux at Shutdown alarm, a part of the source range instruments, was adjusted to maintain protection as count rate increased during initial core loading. This abstract for this test is included in the portion of FSAR Section 14.2.8.2.10 dealing with initial adjustments. l Methodology Prior to the start of initial fuel loading, the high voltage and discriminator settings of Source Range Nuclear Instrumentation Channels (Source Ranges), N31 l and N32, were adjusted using a temporary Americium-Beryllium neutron source to ensure that the Source Range neutron flux monitoring instrumentation was operational. l The High Flux at Shutdown alarm bistable was set to the square root of ten i (3.162) times the measured natural background after the temporary neutron ! source had been removed. The bistable was readjusted to square root of ten l times the measured neutron flux count rate after initial loading of two, ten, , and sixty fuel bundles. l l After fuel loading was completed, but prior to the starting dilution of the Reactor Coolant System for initial criticality, the Intermediate and Power Range Nuclear Instrumentation Channels were aligned using test circuits integral to the appropriate channels. Relay bistable trip and reset setpoints were verified and recorded for reference. The High Range Flux Trip setpoint was conservatively set to 25% to correspond to the requirement for the Initial Criticality Test Sequence prerequisites. 5-37 i
5.12 OPERATIONAL ALIGNMENT OF THE NUCLEAR INSTRUMENTATION SYSTEM PRECRITICAL (1-5SE-02) (Continued):
]
l Results ! All objectives and acceptance criteria were met. 1 All the following operating settings and trip setpoints were . determined, J verified, and recorded in accordance with this test, the Technical' q Specifications, and the FSAR: _! Source Range Instruments:
- 1. Initial trip setpoints,
- 2. Pulse amplifier attenuator and discriminator voltages, !
- 3. High volta 8e power supply plateaus and operating voltages, j
- 4. High Flux at Shutdown Alarm setting, and
- 5. Initial operating settings.
Intermediate Range Instruments:
- 1. Initial trip setpoints,
- 2. Operational checks, and
- 3. Initial operating settings.
Power Range Instruments:
- 1. Initial trip setpoints,
- 2. Operational checks of each power range drawer,
- 3. Positive and Negative Rate Trip Circuit Settin8s,
- 4. Flux deviation averaging amplifier gain settings, and j
- 5. Initial operating settings. l l
The Nuclear Instrumentation Source Range Channels responded satisfactorily to monitor neutron flux during initial fuel load. l Problems Noise spikes attributed to grounds.were encountered during the early. portions ; of the test. The grounds were caused by frayed cable insulation, excessive i moisture, and broken insulators. The detectors and interconnecting cables.were rewerked to eliminate all grounds. The equipment'then performed as~. designed and satisfied all objectives. 1 5-38
5.13 R0D CONTROL SYSTEM (1-5SF-02) t Objectives \,' I The objectives of the Rod Control System test were to demonstrate and document that the Rod Control System performed control and indication functions properly just prior to initial criticality. This abstract for this test is FSAR Section 14.2.8.2.12. Methodolo87 With the plant at normal operating temperature and pressure, the Rod Bank Overlap Unit was temporarily reset to limit total rod withdrawal while still demonstrating proper bank overlap. Overlap started at 18 steps withdrawn. Withdrawal was limited by temporary settings to a maximum withdrawal of 30 steps. Rod control was placed in manual so rod withdrawal was controlled by the In/ Hold /Out switch and the Rod Bank Overlap Unit. Rods were withdrawn j until all control banks were at 30 steps. The point at which temporary 1 settings in the Rod Bank Overlap Unit caused each bank to start and stop l moving were recorded from the group step counters. All rods were inserted to i 0 steps and rod positions from the Proteus, Pulse to Analog Converter, and I DRPI recorded. Overlap unit was returned to original setpoints. I Results (% ( ,j Rod Control System operated correctly. Rods overlapped at the proper points. All position indications were in agreement. All acceptance criteria and objectives were met. Problems There were no s18 nificant problems. s 5-39
l- .5.14 CRDM OPERATIONAL TEST AND ROD POSITION. INDICATION TEST (1-5SF-03) L l Objectives l l D The objective of the Control Rod Drive Mechanism (CRDM) portion of this test i was to verify proper operation of CRDMs under both. cold and hot plant conditions including demonstrating proper slave cycler timing. The objective of the Rod Position portion of this test was to' verify that the rod position ; indication system satisfactorily' performs required indication and alarm ; functions for. each individual rod and that each rod operates satisfactorily over its entire range of travel. The two tests were combined into one { procedure for convenience. 1 1 The abstracts for this test are FSAR Sections 14.2.8.2.13 and 14.2.8.2.15, I respectively. Methodology CRDM operation and Rod Position were demonstrated at both hot standby (557 F) ] and cold shutdown plant conditions by the same method. Rods were pulled, one i bank at a time to 48 steps so all banks were at 48 steps. All rods but one were disabled by disconnecting their lift coils at the lift coil disconnect' switch box located in the control room. The one rod still connected was then withdrawn 10 steps, inserted 10 steps, ) withdrawn again 4 steps and finally reinserted a steps. A Visicorder (high i
^
speed oscillograph) was used to record voltage traces of each CRDM coil (Movable, Lift, and Stationary) for each rod motion. The rod's lift coil was then disconnected and another's connected, and the process repeated for remaining rods in the bank, and then for the other banks. l Each trace was examined to ensure proper slave cycler timing. The bank was then withdrawn in 48 step movements to the fully withdrawn position of 228 steps verifying proper rod position by comparing indication from the Group : Step Counters, Digital Rod Position Indication (DRPI), Proteus, and Control Bank Pulse to Analog Converters. The bank was then inserted and another bank was tested until all Control and Shutdown Banks were tested. l y During test of Control Bank D the FULL ROD WITHDRAWAL alarm was verified to be operating correctly. After completion of all CRDM operational testing for all banks the "TWO OR MORE RODS ON BOTTOM" alarm was verified by withdrawing all rods to six steps and then lowering two rods to bottom and then verifying annunciator alarm. 1 5-40
5.14 CRDM OPERATIONAL TEST AND ROD POSITION INDICATION TEST (1-5SF-03) (Cont.) () r Results I All objectives and acceptance criteria were met. CRDM coil traces and traces from the slave cyclers were all acceptable at both hot and cold plant conditions. The rod control system withdrawals and insertions of each rod bank showrd proper indication on DRPI, Proteus, and Group Step Counters. The FULL Rou WITHDRAWAL and TWO OR MORE RODS AT BOTTOM alarms operated correctly. Problems l The IN/ HOLD /0UT switch on Reactor Operator Board on the Main Control was not j adjusted correctly. Adjusting switch components corrected problem. Several times while attempting to pull one bank of rods, either a rod in f another bank would be pulled or one rod in the bank would fail to withdraw. Two problems were discovered; connectors were swapped on the reactor vessel head, and there was a Bechtel/ Westinghouse interface drawing error. Both were resolved by swapping wire connections. A bad card was also found in the Pulse to Analog converter but was repaired. A wiring problem was discovered between the Proteus and DRPI when the Proteus g did not show correct rod positions. This was corrected and a timing problem between the data transfer from DRPI to Proteus was subsequently discovered. A time delay was incorporated into the Proteus software and the Proteus displayed correct readings for all rods. i i
5.1 ROD DROP TIME (1-5SF-04) Objectives The objectives of the Rod Drop Time test were to: (1) determine the rod drop time of each Rod Cluster Control Assembly (RCCA) under full-flow conditions, i with the reactor at normal operating' temperature and pressure; 'and (2) verify ! the operability of the control rod deceleration device. The abstract for this test is FSAR.Section 14.2.8.2.14 Methodology I 1 The rod drop time measurement test determined the' control rod drop time of each RCCA under full flow conditions, with the reactor at normal operating 1 temperature and pressure, and verified operability , of the control rod l deceleration device, also called the dashpot. Control rod drop time is the I time required for control rods to drop from fully withdrawn to the point when they entered the dashpot section of the fuel bundle's guide tubes. This time was less than 2.2 seconds listed in Technical Specification Section 3.1.3.4.
)
Initial core loading had been completed, the excore source range neutron detectors were operating, and all control rods were initially fully inserted. A baseline count rate was measured for each neutron detector and a limit of five times the baseline was established to stop rod withdrawal during the test. (This limit was never approached.) I Control rods were then withdrawn in groups of four rods and dropped by- I deenergizing the individual drive mechanisms. A computer capable of automatically measuring drop time and deceleration was connected to the Digital Rod Position Indication (DRPI) position sensing coils and provided individual dashpot entry times and total time to rod bottom (called. turnaround times). The computer consisted of three parts, a central processor with printer j located at the control rod drive control cabinets in the basement of the ' control building, and two remote units located at each Digital Rod Position Indicator (DRPI) cabinet on the upper floor of the containment. The DRPI 4 cabinets had been previously modified by additional cables and connectors to ! accommodate the computer. i Timing measurements were first made to demonstrate operability of the computer under cold (135F) conditions with only one reactor coolant pump operating. Hot measurements at hot standby conditions of 557 F Taverage-and.all four reactor coolant pumps operating were then made. Rod drop timing was completed in less-than a shift after computer and , power supply problems were resolved. 5-42 s
5.15 ROD DROP TIME (1-5SF-04) (Continued):
/'T (s_) Results All objectives and acceptance criteria were met.
Dashpot entry time for all control rods but two were within a two-sigma limit of 40 milliseconds about 1445 milliseconds average drop time. These rods, H08 and H02, were three milliseconds outside the limit and were each dropped an additional six times. Drop times are listed in Table 5.15.1. Each rod's dashpot was shown to decelerate the rods. Average turnaround time was 1933 milliseconds; avera8e turnaround time was 488 milliseconds longer than dashpot entry time. Problems Power supplies in the rod drop computer's remote units required modification before the computer could be used. Identification and solution of this problem did not present a critical-path time delay. There were no other significant problems. 1 h V l t l l l 7-~x (s) 5-43
5.15 R0D DROP T)ME (1-SSF-04) (Continued): I TABLE 5.15.1: CONTROL ROD DROP TIMES l (HOT STANDBY (557 F, 2235 PSIG) FULL FLOW CONDITIONS) DASHPOT ENTRY
- TURNAROUND DASHPOT ENTRY
- TURNAROUND ROD TIME (MSEC) TIME (MSEC) ROD TIME (MSEC) TIME (MSEC) i I
I D02 1462 1962 F08 1430 1930 l B12 1464 1974 K08 1446 1926 l M14 1478 2008 H10 1444 1924 i B04 1454 1964 F02 1428 1868 , D14 1462 1942 B10 1444 1934 P12 1474 2004
)
K14 1482 1992 l M02 1480 1920 P06 1420 1870 i G03 1444 1924 B06 1476 1936 C09 1438 1938 F14 1478 1968 J13 1438 1988 PIO 1414 1904 j N07 1466 1926 K02 1428 1848 I C07 1446 1946 H02 1402 1902 I G13 1452 1942 B08 1454 1914 I N09 1414 1964 H14 1442 1972 1 J03 1424 1894 P08 1424- 1914 I E03 1458 1918 F06 1432 1902 i C11 1464 1944 F10 1434 1904 l D L13 NOS 1456 1448 1996 1938 K10 K06 1414 1448 1934 I 1868 I C05 1444 1934 D04 1428 1918 l E13 1448 1958 M12 1432 1892 N11 1432 1952 LO3 1476 1936 ! H04 1434 1914 D12 1428 1898 i D08 1438 1918 M04 1428 1918 H12 1428 1948 H08 1488 2028 M08 1446 1916 H06 1410 1910 Dashpot entry time is used to satisfy Technical Specifications. The term " msec" is millisecond (0.001 second).
SUMMARY
OF TABLE VALUES (DASHPOT ENTRY TIMES ONLY) Mean Dashpot Entry Time: 1445 msec. Two Sigma limits (Dashpot Entry Times): 1405 msec and 1485 Mmsec. Rods outside the two sigma dashpot entry time limits were: H02 and H08. 5-44
t 1 5.16 RTD MINI-CROSS CALIBRATION (1-5SF-09) i Objective O The RTD mini cross-calibration test: (1) determined insulation and lead I resistances. at ambient conditions for the Reactor Vessel. Level Indication. System (RVLIS) instruments and compared the RVLIS RTD'(RTD TE-0433C) to the. I Loop 3. Hot leg RTDs at four temperature. plateaus (250*F, 345*F, 450*F, and. l 557*F); (2) at the 557*F plateau, measured for insulation and lead resistances of all reactor coolant system (RCS) RTDs except TE-0413B; .and (3) performed' incore thermocouple temperature surveys at the four temperature plateaus and. compared selected thermocouple to adjacent thermocouple for congruence. The abstract for this test is FSAR Section 14.2.8.1.15. Methodology Prior to heatup for initial criticality,. baseline temperature data for the RVLIS RTD was taken using a standard ~ digital multimeter (DMM). During heatup, , temperature data was taken at 250'F, 345'F, 450'F, and 557*F for the RVLIS RTD i and loop 3 hot leg RTDs. Test equipment consisted of a digital voltmeter (DVM) ) connected through a special RTD switching box to the Process Protection Set ) cabinets. The switching box provided almost simultaneous readings of RTD resistances to minimize drift error. j Incore thermocouple maps .were also taken at each of the four temperature plateaus. In order to check accuracy, ten thermocouple were selected and O compared with other thermocouple in close proximity. Thermocouple maps were obtained from Proteus or ERF computer printouts. l
)
i At the 557'F plateau, individual RTD insulation and lead resistance values ) were measured using a standard DMM for all RCS loop RTDs except TE-0413B. J (TE-0413B was found acceptable during precore hot functional testing and therefore was not included in this test.) Combined lead resistance data was reduced to obtain actual RTD lead resistances a..d errors through the use .of standard formulas. l i (3 5-45 (/
5.16 RTD MINI-CROSS CALIBRATION (1-SSF-09) (Continued): m Results j l All objectives and acceptance criteria were met. I l l Deviation between RVLIS RTD'and Loop 3 Hot Leg RTDs was. well within the required 0.5*F criteria at all temperature plateaus. Maximum deviation was
-0.36*F at 345*F, while at 557*F the deviation was -0.28*F.
l Incore thermocouple surveys demonstrated the overall operability of the i computer processed thermocouple mapping system. Technical Specification requirement of two thermocouple operable per quadrant per train was met at ; all plateaus. l L RCS loop RTD lead resistance errors varied from a minimum of 0.0 ohms to a maximum of 0.17 ohms. Most resistance errors were between 0.0 and 0.055 ohms. Problems . After testing had started, the RVLIS RTD was replaced due to grounded conductors. Routine retesting was performed in accordance with administrative guidelines to demonstrate its operability. l Thirty nine of one hundred sixteen incore thermocouple comparisons fell j s outside the accuracy requirements specified by the procedure. However, accuracy requirements were obtained based only on RTD accuracy requirements, thus the readout error included by the computer processing system decreased the accuracy. Where actual instrument problems were suspected, corrective l action documents were generated to resolve the discrepancies. No retesting was I required because the Technical Specification requirement of at least two operable thermocouple per quadrant per train was met and no other test
- accceptance criteria exists for incore thermocouple. ,
i 1 1 l q
- d 5-46
5.17 VOLTAGE SURVEY FOR CLASS 1E BUSES _(1-600-15) Objective To obtain bus voltage, -current, active and reactive power on IE buses when loaded to 30% or more and supplied by normal supply transformers. Additionally, to obtain voltage and current of the' lowest voltage but, string during transient condition created by non-concurrent starting of largest class 1E and non-1E 4.16kv motors and to supply this information to Engineering for voltage analysis. This test satisfied Vogtle Safety Evaluation Report (SER) Section 8.4.1, which calhd for data to be gathered in accordance with NRC Branch Technical Position PSB-1 for plant voltage analysis, i Methodolo8Y
]
The 'A' and 'B' Trains were tested independently using the following format- I Watt and var transducers with recorders were connected to the 4.16kv i switchgear. Recorders for measuring voltage were connected to the 4.16kv I switchgear and each 480v switchgear. I Steady state testing was performed by measuring power and reactive power on the 4.16kv switchgear and the switchyard reserve auxiliary transformer feeder. Bus voltage and current were also recorded on the 4.16kv switchgear, 480v motor control center (MCC) switchgear and associated 240/120vac distribution panels. The transient test was performed by starting the largest 1E motor (auxiliary feedwater pump) followed by the largest non IE motor (TPCW pump for Train- 'A or River Water Makeup Pump for Train 'B'). During the transient, voltage, power and reactive power were recorded on the 4.16ky switchgear, and 480 volt , switchgear voltage were recorded. The data was then evaluated. Results The acceptance criteria and objectives were met. I 1 Testing of both trains was completed satisfactorily. Analytical voltage analysis indicated that test results were satisfactory. I 5-47
5.17 VOLTAGE SURVEY FOR CLASS 1E BUSES (1-600-15) (Co,tinued): Problems No major problems were encountered. The 'A' train steady state and transient tests were performed over a period of approximately eight hours. It was subsequently determined that plant evolutions during the time frame varied the loading below minimum required load. A retest of train 'A' was performed over an approximately two hour time frame and plant conditions were more closely controlled. The second test yielded more consistent data. 4 a
v ---w O SECTION 6 INITIAL CRITICALITY AND LOW POWER PHYSICS TESTING
+
t O i l eP
6.1 INITIAL CRET1'ALITY AND LOW POWER TEST SEQUENCE (1-600-911 Objective - The objective of the Initial Criticality and Low Power Test Sequence was to define the sequence of tests and operations, beginning with initial criticality, which constitutes the Jow-power testing program. , The abstract for this test is FSAR Section 14.2.8.2.41. Methodology The Low Power Physics Tests were performed to verify tue Jesign characteristics of the nuclear core. They commenced with initial criticality and ended with a pseudo ejected rod test and measured core power distritntions and the reactivity worth of the control rods, critical boron concentrations, and isothermal temperature coefficients. Table 6.1.1 lists tests performed exclusively as zero power physics tests; others that carried into power. range testing such as flux mapping and operational alignment of the nuclear instrumentation system are not listed in the table. } Initial criticality was the first zero power physics test. The plant was at normal operating temperature and pressure with reactor coolant system berun concentration greater than 2000 ppm. Initial criticality was achieved by withdrawing the control rods in their normal sequence and then diluting the reactor to criticality. The reactivity computer's operation was then verified prior to making any measurements with it, including the point of adding nuclear heat. The reactivity computer was supplied, initially calibrated, and temporarily installed by Westinghouse and this test provided a dynamic checkout of the computer, and its interconnecting cables and power range detectors. Control bank D was withdrawn and inserted to provide approximately +25 percent mille (pcm), then -25 pcm, +50 pcm, -50 pcm, +75 pcm, and finally -75 pcm of reactivity as measured by the reactivity computer. The time required for the flux to double (called doubling time) or decrease by half (called halving) was measured with a calibrated stopwatch which yielded reactivity inserted. The cot :ptter 's calculation and measured reactivity mur.t agree for positive incurtions. A comparison of the checkout values is listed in Table 6.1.2. A daily checkout using an internally oenerated test signal continued throughout the the physics test program. The second test established the power range for the physics tests by determining the point where reactivity disturbances due to feedback from nuclear heating of the nuclear fuel occurs. Physics tests are performed below the point of adding nuclear heat, about 0.1% reactor power. Determination of Core Power for Low Power Physics Tests set the upper power level at one decade , below and the lower limit two decades below the point of adding nuclear heat. 6-2
mg , n
.3 y ~
4 1 6.1 N INITIAL CRITICAL 1'lY AND LOW POWER TEST SEQUENCE _(1-600-04)_ (Continued): ~ , ,
. . . s
, 4, ,
's 1 The reactivity worth of individual e Atdol Eonks wai measur D from unrodded (all rods fullr withdrawn, or all rods out (ARO)) to all control rods inserted l
with the" single most worthy "s%ch" rod fully with'drawin in the " N-1" , configuration. This measurement was'made by trading boron worth-(dilution) for y negative control rod worth (control, rod insertion) to maintain approximately a
'~
just crit 1'en1 reactor. . I l4 Figure 6.1.1 illustrates the test scrpehce. '
- w '
'1 Isothermal Teagerature (Coeffici6th,LEoron Endpoint Measurements, and Flux Map L s s Measurements twere made' after the control rod banks had been individually and 1 O sequentially siiserced via boro'n dilution. When each control bank was fully L
incerted, dilution was stopped and the designat(d tests performed. [ J' y s ( During the dilutions, thecontrol.bankreactivit[worthwasmeasuredas the t
- bar.k was inserted.
s The reactor was tripped when ,;11 control bankb ' were fully inserted to demonstrate the ability of the remaining rods to trip as required by Technical Specification Special Test Exceptica. After the N-1 test the redctor was tripped an6 borated. Criticality was achieved with all shutdown banks withdrawr. rand control banks inserted. The worth of control banks in overlap was there measured. Shortly after the Hot Zero Power Rod Insertion Limit (HZPRIL) had. been reached a control rod ' dropped - i
,, and its motor coil stack was replaced as discussed in the problems section ..I below; Rod worth measurmonp continued with dilution from essentially all rods out to the HZPRIL where ~ the ' pseudo ejected red worth measurement was made. A flux map was taken, the highest worth rod was fully.. withdrawn and a second
, flux map was taken. Thii Completed the Pseudo Ejected Fod test and the Low Power Physics Test Seque7ce. \
s Throughout the physics test program, the Technical Specifications Special Test Ccceptions were used. Special Test Exception 3.10.1~ and 3.10.3 allowed [ ' operation with a_ positive moderator temperature coefficient, _ the control rods below the rod insertioria limits, and less'than 1.3% shutdown margin. These conditions were required due to the off-normal conditions needed for physics tests . ' L , t Results The Low Po ur Physics Tests all met acceptance criteria except for positive moderator temperature coefficient (MTC) and tilt _be ed on all incore detectors. 111 objectives were met. 1 \ L Ol - e-s J
6.1 INITIAL CRITICALITY AND LOW POWER TEST SEQUENCE (1-600-04) (Continued): Problems ! As expected based on preceeding similar plants, the. moderator temperature , coefficient (MTC) at unrodded configurations was slightly positive. Operation l with a positive MTC was not a problem during zero power. physics.because the test sequence started by invoking special test exceptions. Vogtle Reactor ! Engineering issued the required guidelines on rod withdrawal. limits prior to I finishing the tests and revokin8 Special Test Exception that ' allowed exceptions to the limits of Technical Specification 3.1.1.3. Intermediate Range detector NI-35 failed to respond to the initial increase in flux levels following initial criticality. A grounding problem in the detector j{ was discovered and repaired. j Early in the testing the letdown system experienced pressure spikes that lifted the letdown pressure relief valve and occasionally delayed the program to allow the RCS to stabilize. These delays were eliminated after the valve { was repaired.- The initial calibration of the power ran8e: channels was conservatively adjusted, so 100 microamps detector current equaled 100% power. Actual 100% power currents are in the 300-400 microamp range so the. power range channels i were more sensitive by a factor of 3-4 times. . This meant the_ power range O channels indicated power of 3-4% when the' reactor was actually at 'about 1% power, somewhat below the useful range for flux mapping of 2% to 4% actual i I power. It was recognized that the power range channels were used for reactor protection and a delta-t power signal was used to determine actual reactor l power. A strip chart recorder with delta-t power was located near the reactor i operator's console to assist him in establishing and maintaining the required J flux mapping power level. " Shutdown Bank B's rod J-3 dropped due to a burned out connection on the top of the CRDM coil stack. The connector was integral with the coil stack so the entire coil stack required replacement. The plant was cooled to Mode 5 so the containment equipment hatch could be opened and another stack brought into the I containment. The missile shield was removed, the motor stack replaced, missile a shield reinstalled, and plant returned to zero power physics testing within i three days of the drop. ! A reactor trip occured due to power range high rate during the flux map at the zero power insertion limit after control bank C was moved one step. Bank C was ' in a high worth region so one step produced a sudden small step increase' in power. The rate trip setting was found too sensitive and was recalibrates. I i O 6-4 i i
6.1 INITIAL CRITICALITY AND LOW POLfER TEST SEQUENCE (1-600-04) (Continued): TABLE 6.1.1: INITIAL CRITICALITY AND LOW POWER PHYSICS TESTS
- PROCEDURE NUMBER TITLE DESCRIPI' ION 1-500-03 Determination Of Core Determined high and low power operating Power For Physics level for zero power physics testing, Testing except for flux maps.
1-600-01 Inverse Count Rate Obtained count ra*es and computed inverse 2 Ratio Monitoring count rate ratios for reactivity l For Approach To monitoring during approach to initial Initial Criticality criticality. q 1-600-02 Initial Criticality Directed the initial criticality. 1-600-03 Isothermal Temperature Determined the isothermal temperature Coefficient reactivity.of the critical reactor at j l Measurement certain endpoint configurations and q derived the Moderator Temperature l Coefficient (MTC) of reactivity. l 1-600-04 Initial Criticality Controlling document for. initial And Low Power Physics criticality and low power physics Test Sequence tests. Included checkout of the reactivity computer. l 1-6SF-04 Pseudo Rod Ejection Determined reactivity worth of the Test worst pseudo-ejected control rod and measured core power distributions in ! ' the aligned and fully misa11gned i configurations. i i 1-6SF-06 KCCA and Bank Worth Measured individual bank reactivity Measurement At worths, and control banks in overlap - Zero Power reactivity worths with a reactivity computer. 1-6SF-09 Boron Endpoint Determined boron concentration in the Determination critical reactor at certain endpoint configurations. This and the RCCA worth measurement results were combined for inverse boron reactivity worths. These exclusively physics tests are listed in numerical order whereas they are discussed more or'less in order of performance. Other tests, including Flux Mapping (1-6SE-02) and Operational Alignment Of Nuclear Instrument Systems (1-5SE-02) are discussed in-the Power Ascension section of this report because they continued into power range testing. 6-5 u______________.____________________________________.____________
6.1 INITIAL CRITICALITY AND LOW POWER TEST SEQUENCE (1-600-04) (Continued): TABLE 6.1.2: REACTIVITY COMPUTER CHECKOUT RESULTS POSITIVE REACTIVITY INSERTIONS PCM SECONDS PCM REACTIVITY PERCENT
- l TARGET DOUBLING TIME MEASURED CALCULATED DIFFERENCE
+25** 231.7 23.5 22.77 3.21%
+50 97.5 48.2 48.05 0.62% .
i
+75 53.23 78.0 78.19 0.24%
Acceptance criteria is <4.0%. I l NEGATIVE REACTIVITY INSERTIONS PCM SECONDS PCM REACTIVITY PERCENT
- t TARGET _ HALVING TIME MEASURED CALCULATED DlFFERENCE
-25** 219.59 -29.5 -30.04 1.78%
-50 133.02 -51.0 -55.18 7.58%
l
-75 85.16 -85 -108.12 21.38 I
1 ) No acceptance criteria for negative reactivity insertions.
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- 6. 2 - INITIAL CRITICALITY (1-600-02)
Objective The objective of initial criticality was to achieve initial criticality in 'a safe controlled manner. The abstract for this test is FSAR Section 14.2.8.2.39. Methodology In addition to achieving initial criticality in a safe controlled manner, the procedure also provided instructions for increasing reactor power (flux) level so the Source Range and Intermediate Range Nuclear Instrument System overlap data could be obtained during Startup Procedure 1-6SE-03. The procedure also left the reactor in a configuration to support continuation of low: power physics testing. Initial core loading had been completed, the reactor coolant system was' at normal operating pressure (2235 psig) and no-load average temperature(Tavg of 557*F), RCS boron was at refueling concentration (2000 ppm), the excore i nuclear instrumentation system was operable, and all control; rods were fully l inserted. A baseline count rate for each source range channel was determined and recorded as initial count rate per startup procedure 1-600-01 (Inverse Count Rate Ratio Monitoring For Approach To Initial Criticality). Estimated critical condition (ECC) for Control Bank D at 160 steps withdrawn was 1302 ppm RCS boron concentration with an upper limit of 1357 ppm and lower limit of O 1247 ppm. Shutdown banks A through E were then withdrawn sequentially in '114 step increments (approximately 0.5% delta K/K) until the shutdown banks were fully withdrawn. Control banks A through D were then withdrawn in sequential overlap in 50 step increments until bank D was at 160 steps. Inverse count rate ratio (ICRR) calculations and plots were made to monitor reactivity changes during control rod withdrawal and are shown in Figure 6.2.1. A 1/M plot decreases as count rate increases; the expected decrease between sections thirteen (13) and sixteen (16) is due to geometric effects as control banks C and D are withdrawn past the neutron sources. The sources are at 25% core height. In preparation for dilution to criticality, the ICRR plots were renormalized to 1.0 for rods withdrawn. Pressurizer heaters were energized for -boron equalization between the pressurizer and reactor coolant system. Demineralized water was added to the reactor at approximately 120 gpm (2000 pcm/hr reactivity change rate, 2 ppm / minute dilution rate) while the source range instruments were continuously monitored. ICRR data and plo*s were made for changes in time and boron concentration. These are shown in Figures 6.2.2 and 6.2.3, respectively. The plots were used to predict critical boron-concentration and time. RCS loop and pressurizer boron samples were obtained and analyzed at thirty minute intervals. The RCS boronometer was monitored continuously during the dilution.
'O 6-8
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6.2 INITIAL CRITICALITY (1-600-02) (Continued): The reactor.was diluted from refueling concentration of approximately 2000: ppm-D to approximately 1450 ppm. At 1450 ppm the dilution was stopped to allow the reactor coolant system to mix. The dilution was resumed at a slower nominal rate of 1 ppm / minute but was stopped to allow the reactor to become critical-as boron mixed between the volume control tank (VCT) and RCS. The' reactor operator declared the reactor critical at 0837 Central' Standard Time on March 9, 1987, thirteen hours after the start of rod motion, and inserted control' rods from 160 .to 135 steps to maintain low power level and to counter the effects of dilution already in the piping between-the charging pumps and reactor coolant system after dilution had been stopped. Control bank D was then withdrawn to increase power to IE-8 amps as indicated by the Intermediate Range channels to obtain NIS overlap data. One of the Intermediate Range detectors, N35,'did not respond so the reactor was reduced to make repairs. Data was obtained three days later, following repairs. Results The objectives and acceptance criteria were met. Initial criticality was achieved in a safe controlled manner at a RCS boron of 1330 ppm, bank D height of 160 steps, and RCS temperature of 557*F. Critical boron concentration was within predicted and design limits of 50 ppm and all acceptance criteria were met. Source and intermediate range nuclear instrument system overlap data- was obtained following repair of NI35. Problems During the initial dilution pressurizer concentration decreased more rapidly than the RCS loops. The auxiliary spray valve was determined to be leaking past its seat and was manually isolated to stop leakage. Dilution flow rate of 120 gpm was too rapid to permit boron concentration to equalize between the pressurizer and RCS. This problem was solved by alternating dilution flow rates between 60 gpm and 120 gpm during the initial dilution to 1450 ppm. Intermediate range channel NI35's detector was found 8 rounded to its housing and the detector was replaced. This repair delayed the test program 3 days. No further problems were encountered. 6-9
6.2 INITIAL CRITICALITY '(1-600-02) (Continued): FIGURE 6.2.1: ICRR DURING ROD WITHDRAWAL O, , i INVERSE COUNT RATE RAfl0 - APPROACH 70 CRiflCALITV 1.95 i A s0URCE RAEE CHANNEL - N131 1, _p -8 a O SOURCE RAEE CHANNEL - H132 w
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l 6.2 INITIAL CRITICALITY (1-600-02) (Continued): FIGURE 6.2.1: ICRR DURING R0D WITHDRAWAL (Continued): O l CONTROL ROD POSITIONS DURING WITHDRAWAL STEP NUMBER CONTROL ROD CONFIGURATION O All control rods fully inserted (baseline count rate). 1 Shutdown Bank A at 114 steps withdrawn. 2 Shutdown Bank A fully withdrawn at 228 steps. f l 3 Shutdown Bank A fully withdrawn, Shutdown Bank B at 114 steps. I 4 Shutdown Banks A and B fully withdrawn. ) 5 Shutdown C at 114 steps. 6 Shutdown Banks A, B, and C fully withdrawn. 7 Shutdown Bank D at 114 steps. ! 8 Shutdown Banks A, B, C, and D fully withdrawn. " 9 Shutdown Bank E at 114 steps. 10 Shutdown Banks A through E fully withdrawn. 11 Control Bank A at 50 steps withdrawn (withdrawing in overlap). l 12 Control Bank A at 100 steps withdrawn. 13 Control Bank A at 150 steps, Control Bank B at 35 steps. , 14 Control Bank A at 200, Control Bank B at 85 steps. 15 Control Bank A at 228, B at 135, C at 20 steps. 16 Control Bank B at 185, C at 70 steps. i 17 Control Bank B at 228, C at 120, D at 5 steps. l 18 Control Bank C at 170, D at 55 steps. 19 Control Bank C at 220, D at 105 steps. l 20 Control Bank C at 228, D at 160 steps (final configuration). O V 6-11 l l
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6.3 INVERSE COUNT RATE RATIO MONITORING FOR APPROACH TO INITIAL CRITICA ITY (1-600-01) Objective l The objective of the Inverse Count Rate Ratio Monitoring for Approach to Initial Criticality was to describe methods for obtaining and evaluating ~ data used as an indication of core reactivity-during the approach to initial criticality. The abstract for this test is FSAR Section 14.2.8.2.38. Methodology The approach to criticality was monitored by the inverse count rate ratio l method and used the same computer program describ'ed in Section 4.4, " Inverse l Count Rate Ratio For Initial Fuel Loading". The major difference was shorter count intervals because the count rate was higher and the approach to criticality was being performed by boron dilution so count rate- increased i during a long counting interval.-
-]
i The installed source range instruments and their' scaler-timers were used to monitor -count rate. This information was relayed by headset to a personal. ' computer located just outside the control room's control. board area. From this position the computer's results were relayed to personnel in the control raom l- for graphing. 1 D Results The objective and acceptance criteria were met. Inverse count rate ratios, the same method of monitoring reactivity used' during initial core load, were used to monitor the approach to initial criticality. Problems There were no significant problems, l 4
l 6.4 DETERMINATION OF CORE POWER RANGE FOR PHYSICS' TESTING (1-500-03) Objectives O The objectives of the determination of core power. range for physics testing were to: (1) determine the reactor power (flux) level where the effects of t l fuel heating are first detected; and (2) establish the range of neutron flux ; in which zero power physics measurements were to be performed. The abstract for this test is FSAR Section 14.2.8.2.22. i Methodology This was the second test following initial criticality. The first was. checkout of the reactivity computer since the reactivity computer was to be used to determine reactivity changes due to fuel heating. The reactor was critical at approximately 2E-08 amperes (Intermediate Range Channel current reading), c.ontrol bank D was at 160 steps withdrawn, and plant conditions were stable. Control bank D was then withdrawn for a positive reactivity addition of 20 to 60 pcm as indicated by the reactivity computer. A positive startup . rate is required to determine reactivity changes associated with heating and approximately a 200 second doubling time is acceptable. The reactivity computer's picoammeter was reranged as necessary to follow increasing power. \ Nuclear heat was detected by a steady decrease in reactivity, not by increasing Tavg. When the test supervisor was satisfied that nuclear heating had been reached, intermediate range and reactivity picoammeter readings were taken. Control bank D was inserted somewhat to lower flux approximately one half decade and the reactor stabilized. Subsequent determinations of nuclear heat were made in similar fashion. The intermediate range channel and reactivity computer picoammeter readings j for each determination were averaged. The upper limit for testing was l established as one decade below the average, and the lower limit two decades below. The upper and lower flux levels were recorded and control bank D was then positioned to establish the neutron flux level within this zero power. physics testing range. This range coincided with the flux levels used for the j earlier reactivity computer checkout. { 6-15
6.4 DETERMINATION OF CORE POWER RANGE FOR PHYSICS TESTING (1-500-03) (Continued): Results The objectives and acceptance criteria were met. Nuclear heating was detected by reactivity at 1.8E-06 amperes on the reactivity computer's picoammeter and 1.75E-06 amperes on the intermediate range channels. The zero power physics testing range flux level was determined to be 1.8E-08 to 1.8E-07 amperes on the reactivity computer picoammeter and 1.75E-08 to 1.75E-07 amperes on each intermediate range channel. ) Problems The only problem was during the second approach to nuclear heating power level increased beyond the picoammeter's range so another approach was required. The additional approach was within the picoammeter's range and problem-free. 6-16
L q
- 6.5 , BORON ENDPOINT DETERMINATION _(1-6SF-09) i Objective.
The objective of the Boron Endpoint Determination measurement was to determine l the critical reactor coolant system boron concentration appropriate to .an ! l endpoint configuration (RCCA configuration). The e$stract for the test is FSAR Section 14.2.8.2.37. 1 Methodology I l The reactor was critical and stable at zero power, no-load temperature and' reactor pressure, with the control rods at or near the' desired endpoint position. The RCS boron concentration was stabilized and determined by chemical analysis. -Adjustments to existing boron concentration were'made if. the reactor was noc exactly at no-load temperature and pressure, or if the rods were not fully withdrawn or inserted (as required for the endpoint). 1 If the rods were exactly at the desired position (banks either fully inserted or withdrawn), the critical. boron concentration was the existing boron-concentration. Otherwise, the rods were moved as necessary to the~ desired: position and the reactivity change resulting from their _ movement measured, converted to an equivalent boron concentration, and added.to the existing RCS boron concentration. Changes from as-found to endpoint rod position. were repeated at least twice for each measurement. The resulting boron ; concentrations were then compared to Westinghouse predictions, illustrated in-Figure 6.5.1. There were two types of acceptance criteria, as shown in Figure 6.5.1. The all-rods-out (all RCCA's fully withdrawn) value was an absolute value; all others were the difference between successive measurements. Boron endpoints were measured at the following rod configurations:
- 1. All-rods-out (all RCCAs fully withdrawn);
- 2. Control bank CD fully inserted;
- 3. Control banks CD and CC fully inserted;
- 4. Control banks CD, CC, and CB fully inserted;
- 5. All control banks fully inserted; and
- 6. At the N-1 configuration, q During the N-1 measurement, the controlling shutdown bank was stopped short of being fully inserted. The endpoint was then obtained by fully inserting the partially withdrawn controllin8 shutdown bank. Inserting the bank from- just-critical insured that the reactor would not be critical with the stuck rod even partly inserted. Being critical with the stuck rod' inserted would violate the special technical specification test exception for the N-1 test.
6-17
6.5 BORON ENDPOINT DETERMINATION (1-6SF-09) (Continued): Differential boron worth, the change in reactivity for a change in boron l concentration, was determined from boron endpoint measurements of the control banks. Reactivity changes resulting from control rod motions between endpoints were divided by corresponding changes in boron concentration between endpoints to produce data points. These points were then linear least square fitted (see Figure 6.5.2) to produce an overall differential boron worth. Boron endpoint measurements from this procedure were used with rod worth measurements (test number 1-6SF-06, Section 6.7 of this report) to determine i differential boron worth.
, Results All objectives were met. All acceptance criteria were met, as illustrated in Figure 6.5.2.
The all-rods-out measurement was made with the rods fully withdrawn so no corrections due to rod position were required. Measured boron concentration was 1337 ppm, within the acceptance criteria of 1307 1 50 ppm. All rodded configurations required corrections and met their acceptance criteria as summarized in Figure 6.5.1. 3 Measured differential boron worth was -10.42 pcm/ ppm, satisfying acceptance criteria of -10.09 1 1.0 pcm/ ppm. Problems No significant problems were encountered. The test procedure as-written did not permit a measurement if the rods were exactly at an endpoint configuration so for the all-rods-out measurement a minor procedure change was made to permit the measurement. 6-18
6.5 BORON ENDPOINT DETERMINATION (1-6SF-09) (Continued): FIGURE 6.5.1: ACCEPTANCE CRITERIA AND TEST MEASUREMENT RESULTS FOR BORON ENDPOINT TEST Acceptance Measurement Criteria, ppm Data, ppm Condition Relative Absolute Relative Absolute
- 1. All Rods Fully Withdrawn --- --
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- 2. CB D Fully Insertee --i[--- --"
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N-1 Configuration is all control rods fully inserted e .ept the stuck-out rod. 6-19
6.5 BORON ENDPOINT DETERMINATION (1-6SF-09) (Continued): FIGURE 6.5.2: TEST MEASUREMENT RESULTS FOR DIFFERENTIAL BORON WORTH l' l DIITDtDff! AL BORON WORTHS .PRDICTIONS & DATA
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6.6 ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT (1-600-03) Objectives The objectives of the Isothermal Temperature Coefficient (ITC) test were to determine the isothermal temperature coefficient and derive the moderator temperature coefficient (MTC) from the ITC data. The abstract for this test is FSAR Section 14.2.8.2.40. Method The I'lC test varies reactor temperature so slowly that the fuel and moderator are essentially at the same (isothermal) temperature. The reactivity computer measures changes in reactivity due to combined changes in moderator and fuel temperature, and the amount of reactivity change divided by the moderator temperature change that produced the reactivity change is the ITC. Predicted fuel temperature coefficient (FTC, due almost exclusively to Doppler broadening of the U-238 resonance peaks) is subtracted from the ITC to produce MTC. The test was performed by lowering and raising RCS temperature. Temperature was first decreased 3 to 4 degrees F at a rate of approximately 10 F/ hour by opening the atmospheric relief valves. The cooldown was terminated and a heatup to original temperature at approximately the same rate was initiated. Control rod motion was avcided during the cooldowns and heatups. The reactivity computer's x-y recorder was configured to record Tav8 and reactivity. The slope of the line was the ITC. Separate reactivity computer traces were made of the cooldown and heatup. All ITCs were averaged to produce a single value for a given condition. The FTC was then subtracted from the ITC to produce MTC. ITC changes with changes in boron concentration and the test was performed at three configurations:
- 1. All rods out (AR0);
- 2. Control Bank D (only) inserted (CD in); and
- 3. Control Banks C & D (only) inserted (CD and CC in).
The ITC measurement requires stability in the reactor's boron concentration, so it was performed after boron concentration in the RCS and connecting systems (especially the volume control tank (VCT) and pressurizer) had equilibt ated . The boron endpoint test preceeded thi; test and provided statepoint data for the ITC. Vogtle Unit 1 Technical Specifications require a negative MTC except while special test exceptions are invoked during physics testing. 6-21
6.6 ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT (1-600-03) (Continued): Results All acceptance criteria except one (concerning positive MTC, see below) were met. The measured ITC for each rod configuration was within + 3 pcm/'F of prediction as shown in Table 6.6.1. Differences resulted almost exclusively from the difference between as-found and predicted boron concentrations for , the test configurations. The results are also graphed as a function of boron j concentration in Figure 6.6.1. l l All objectives were met. I l l Problems j i The all rods out MTC was calculated to be.+0.89 pcm/ F, greater than Technical l l Specification 3.1.1.3's requirement for a negative MTC. Per the Technical l l Specification's action statement for non-negative MTC, control rod withdrawal l limits shown in Figure 6.6.2 were imposed. These effectively reduced boron I concentration to values that resulted in negative MTCs. The non-negative MTC was never a problem during zero power testing because the l applicable special test exception was invoked prior to initial criticality in anticipation (based on previous plant experience) that MTC would be positive. There were no other problems. , j 4 l t l ( i 6-22
6.6' ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT (1-600-03) (Continued): I O TABLE 6.6.1: PREDICTIONS AND RESULTS OF ITC TEST I PCM/ F ITC PCM/ F MTC Configuration PPM Boron Predicted , Measured Calculated ARO 1340.7 -2.12 1 3 -1.04 +0.89 CD In 1290.4 -3.43 1 3 -2.29 -0.36 CD & CC In 1148.16 -6.70 1 3 -5.947 -4.017 4 l 4 O l l O e-23
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6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) Objectives Objectives of the RCCA and Bank Worth Measurement at Zero Power test were to: (1) determine the differential and integral reactivity worth of the Rod Cluster Control Assembly (RCCA) banks; and (2) determine that the highest worth Report. rod is less than or equal to the value used in the Nuclear Fuel Design The abstract for this test is FSAR Section 14.2.8.2.35. Methodology Control rod worth was measured by the standard technique of trading boron for rod reactivity and measuring reactivity with an analog reactivity computer. Individual control banks were diluted into the core starting with all rods out and finishing with all rods inserted except the most worthy " stuck out" rod, which was fully withdrawn. The control banks were then borated to nearly fully withdrawn to determine overlap worth and compare total worth measured by overlap with the total control bank worth by individual rod bank measurement. All zero power rod worth measurements were made with the reactor at zero power level and within a narrow reactivity band (nominally <i30pcm) of critical. Reactivity was measured with the reactivity computer. A boron trade is either a dilution or boration. A dilution results in rods entering the core. Demineralized water is supplied to the CVCS and enters the reactor coolant system, diluting its baron concentration. The reactor responds to the dilution's positive reactivity by increasing its power level. The excore nuclear detectors sense this increasing power level and the reactivity computer, attached to one set of power range detectors, calculates reactivity based on signal changes. The reactor operator periodically inserts control rods The (or banks of rods) to insert negative reactivity to lower power level. result is nearly constant reactor power level while RCS boron concentration and rod height decreases. The reactivity computer's trace shows a sawtooth-shaped curve with dilution producing a slow nearly linear positive reactivity increase and a few steps of rod motion producing a sharp sudden periodic negative reactivity decrease. Conversely, a boration inserts negative reactivity so RCS boron concentration and rod height increase. The reactivity computer's trace is also sawtooth with a slow linear negative reactivity trend due to boration balanced by periodic sharp positive insertions due to rod withdrawal. The reactivity computer's output and average temperature and core power are recorded on a strip chart recorder and reactivity worths are calculated based solely on these recordings. 6-26
.. r i 6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued):
1 I ' I Results All objectives and acceptance criteria were met. Differential and integral rod worths were measured for individual control . I banks and shutdown banks E, D, and C. The worth of all rods except the stuck rod vere measured. Differential and integral worths of the control banks in \lli overlap were measured and the integral worth was compared to the sum of individual control bank worths. (The worth of the pseudo ejected rod is l discussed in Section 6.8, startup test 1-6SF-04). l Predictions Figures 6.7.1 and as-found data are tabulated in Table 6.7.1 and graphed in ' through 6.7.16. All measurements were within acceptance - } criteria. The graphs show close agreement with predictions; some of them ' "- essentially overlap. The differential rod worth figures are sufficiently detailed to show depressions in reactivity due to greater neutron absorption at grid strap locations of 36, 69,102,135, and 168 steps withdrawn. The grid at 200 steps is generally not shown on the figures. 3 e 3 Problems There were no problems aside from interruptions discussed elsewhere in this report. The actual measurements were made without significant problems. 6-27 e
6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POEIER (1-6SF-06)'(Continued): (O j TABLE 6.7.1: RESULTS OF RCCA AND BANK WORTH MEASUREMENTS COMPARISON OF INDIVIDUAL INTEGRAL WORTHS FOR CONTROL AND SHUTDOWN BANKS i INTEGRAL WORTH I ROD BANK PREDICTED MEASURED Control Bank D 630163- pcm 674.3 pcm Control Bank C 12401124 pcm 1221.53 pcm Control Bank B 960196 pcm 1027.3 pcm Control Bank A 690169 pcm 693.4 pcm Shutdown Bank E 880188 pcm 857.6 pcm Shutdown Bank D 710171 pcm 756.1 pcm i Shutdown Bank C 930193 pcm 995.6 pcm i O
- /
{ l t COMPARISON OF TOTAL WORTHS LESS STUCK ROD Total Worth Of All 63791638 pcm 6487.4 pcm Control & Shutdown (>5739 pcm) Banks Less Stuck l Rod (F-10) l COMPARISON OF TOTAL INTEGRAL WORTHS FOR INDIVIDUAL AND OVERLAP CONTROL Ratio Of Sum Of 1.010.04 1.0026 Control Bank Individual Worths To Control Bank In Overlap 6-28 t
. . . . . . , , . . , . . . , , , - - - - - - - - - - - - - - , - - - - - - - - ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
- j j .i 1 6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): '3-i t
h FJGURE 6.7.1: INDIVIDUAL CONTROL BANK D DIFFERENTIAL WORTH
-l 4
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I VOCfLE I LOW POWER PHYSICS ?EST RESULTS I 7 g 6 CONTROL BAHX D DATA 0 WESTINGHOUSE PREDICTION A 6 It *
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6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): , ; FIGURE 6.7.2: INDIVIDUAL CONTROL BANK D INTEGRAL WORTH I l I l Y0CTLE UHlf I CYCLE I LOW POWER PHYSICS RESULTS 7N - -- A CONIROL BANK D MTA IB % 0 WESTINGHOUSE PREDICfiON 4 599 '2 _ \
\
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,6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued):
FIGURE 6.7.3: INDIVIDUAL CONTROL BANK C DIFFERENTIAL WORTH l
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.(
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Y0CILE UNIT 1 CYCLE 1 LOW POWER PHYSICS RESULTS 13 ( A C0terROL MMX C NTA 0 WES11NGHouSE PRDICfl0N
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6.7 RCCA BANK HORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): FIGURE 6.7.4: INDIVIDUAL CONTROL BANK C INTEGRAL WORTH l l l l l t U0CILE UNIT I CYCLE I LOW POWER PHYSICS RESULTS i 1390 Q, 8 CONTROL BANX C MTA l 0 wEsflNGHOUSE PRDICTION 1100 19 _ 900 800 $ 700 D . 1
$590 g?_
la 488 a EY e 1 8 300 W. 200 , M 100 %. 9 -
- A 11 2'8 48 68 8'8 1h8 1h8 ide Ide Ide 2h6 2d8 2 441 STEPS WifHtRAWM 6-32
6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): ) FIGURE 6.7.5: INDIVIDUAL CONTROL BANK B DIFFERENTIAL WORTH U0CTLE I CYCLE 1 LOW POWER PHYSICS TEST RESULTS 11 A CONTROL BAMX B DATA 18 D J' STINCHOUSE PRDICTION 9 4 k
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l 6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): FIGURE 6.7.6: INDIVIDUAL CONTROL BANK B INTEGRAL WORTH i l YoGTLE UMit I CYCLE I IM poler PHYSICS RESULTS 1199 _ A CONTROL BAMX B DAlA 1999 AA G L"STINGH00SE PRDICTION 999 $> 899 Y 799 > l l 699
\
2 599 i .99 % j399 h, a 1 W 299 'A l k s s b 199 . 9 hn il 2'8 48 6'8 89 Ad8 139 1k9 id8 id9 2h8 239 2 411 STEPS WITHDRAWH l 6-34
6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): FIGURE 6.7.7: INDIVIDUAL CONTROL BANK A DIFFERENTIAL WORTH U0GTLE UNif I CVCLE I LOW POWER PHYSICS RESULTS 7 3 b CONTROL M MX A M TA 0 WESTINGHOUSE PRDICTION 5.5 5 ^ 4.5 i . A T\ i,.5 /T \M O !, /t %
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6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POMER (1-6SF-06) (Continued): ( l FIGURE 6.7.8: INDIVIDUAL CONTROL BANK A INTEGRAL WORTH l U0GTLE UHlf I CYCLE I LOW POWER PHYSICS RESULTS 750 7,, A CONTROL BAMX A DATA D WESTINGHOUSE PREDICTION 600 k j 550 500 % ' 450-. l . % D % ! 2 300 2 250 A a 200 1 150 . 100 x 50 4
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11 80 2d8 2461 STEPS WITHDRAlef 6-36
J 6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POMER (1-6Sr-06) (Continued): _ p r1cuaE e.7.e, 1so1v1ouit Saurooms eisx E e1rrExEsr14t woara U0GTLE IINIT I CYCLE I 14W POWER PHYSICS RESilLIS 19 A SHUID0684 BAMX E DATA 9 n mt!NcHOUSE PREDICTION 8 2 1 \ \ E
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j 6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06') (Continued): FIGURE 6.7.10: INDIVIDUAL SHUTDOWN BANK E INTEGRAL WORTH U0CTLE UNif I CYCLE I LOW POWER PHYSICS RESULIS 1999 A SHUfD0let MMX E DATA 9H n WESilNCH00SE PREDICTION
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6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): k FIGURE 6.7.11: INDIVIDUAL SHUTDOWN BANK D DIFFERENTIAL WORTH U0GTLE UNIf I CYCLE I LOW POWER PHYSICS RESULTS 8 8 SHUIDOWN BAMX D DATA D WESTINGHOUSE PREDICfl0H 7 , 6
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6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): FIGURE 6.7.12: INDIVIDUAL SHUTDOWN BANK D INTEGRAL WORTH
=
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6.7 RCCA BANK WORTH MEASUREMENT AT ZER0__ POWER ___.(1-6SF-06) (Continued): _ 1 FIGURE 6.7.13: INDIVIDUAL SHUTDOWN BANK C DIFFERENTIAL WORTH YOCTLE UNIT I CYCLE I LOW POWER PHYSICS RISULTS 11 8 SHUTDOWN BANX C MTA 10 0 :: WESTINGHOUSE PRDICTICtl 9 0 00 .A 8 7
- 6 A I 5 I -
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6.7 RCCA BANK WORTH MEASUREMENT _AT__ZER0_ POWER _(1-6SF-06) (Continued): _ FIGURE 6.7.14: INDIVIDUAL SHUTDOWN BANK C INTEGRAL WORTH s. UOGTLE UNIT I CYCLE I LOW POWER PHYSICS RESULTS ltse 4 SHUTDOWN BAMX C DATA 1000 t C %ESTINGH00SE PREDICTION 900 >- 800
,. \s.
600 ^ 4 500 \ -> 400 2
\1 J 300 k 4.
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r i l l 6.7 RCCA BANK WORTH MEASUREMENT AT ZERO POWER (1-6SF-06) (Continued): FIGURE 6.7.15: CONTROL BANKS IN OVERLAP DIFFERENTIAL WORTH i U0CTLE I LOW POWER PHYSICS TEST RESULTS 17 16 6 CONTROL BANXS IN OVERIAP DATA 15 0 MESTMM MIGIM A ! y n -1 l13 i 0 12 14 l 11 h , a is 46 h i !
, 1%R 1 i o i ! 41 %
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6.7 RCCA BANK WORTH MEASUREMENT AT ZER0_ POWER (1-6SF-06)_(Continued): FIGURE 6.7.16: CONTROL BANKS IN OVERLAP INTEGRAL WORTH i i VOCTLE I 1,0W POWER PHYSICS TEST RESULTS
=
4000
, A c0NTROL BAMXS IN OUERIAP DATA 3600 0 WESTINGHOUSE PRDICTION .*
3200... _._ __ _ __ g . ., 2800 .4 2400 2000 g1600 2
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l 6.8 PSEUD 0 ROD EJECTION TEST (1-6SF-04) Objective The objective of the Pseudo Rod Ejection Test was verify that the power distribution resulting from a simulated ejected Reactivity Control Cluster Assembly (RCCA) is within acceptable limits. The abstract for this test is FSAR Section 14.2.8.2.34. Methodolo87 l The test was performed at the end of the low power physics test sequence after { control bank overlap worth had been measured. Rod control was in manual so overlap existed and control bank C was diluted. to Zero Power Dependent 'i Insertion Limit (ZPDIL) of 228 steps (control bank A), 161 steps (control' bank ! B), 46 steps (control bank C) and less.than 5 steps (control bank D). At the ZPDIL control bank D was fully inserted and RCCA D-12 had the highest' ejected j worth. This test provided core power distribution at ZPDIL with the control ) be. 6 uQ..ca , lin UW pseuuo-ejecte6 rud v- d iully ullsaligned, and the l worth of the pseudo ejected rod.at zero power. The flux map with banks aligned and positioned at the ZPDIL was- taken after , reactor power was raised by dilution to the flux mapping power level of approximately 3%. This base case full core flux map determined F(N-deltaH) prior to rod withdrawal. Reactor power was then returned to the zero power band and control bank D was m) raised six steps. The lift coils on three of the four RCCAs in the bank were disconnected to allow the remaining D-12 to continue to be withdrawn.. l D-12 was withdrawn in short steps of approximately -10 pcm to +10. pcm, alternating with batch borations. The excore power range nuclear detector closest to the rod, N42, had been used throughout low power physics to permit the N-1 measurement to be conducted with the lowest possible power level, and ! spatial effects from D-12 were substantial. The combination of short [ withdrawal steps and batch borations enabled the reactor to stabilize sufficiently for accurate measurements. Boration was terminated at 185 steps on D-12. The worth of D-12 from 6 to 185 steps was obtained by summing the worth in pcm of all rod height and reactivity increments from the reactivity computer's strip chart. The differential and integral worth curves for the misaligned D-' 12 are shown in Figures 6.8.1 and 6.8.2, respectively. ' l The remaining worth of D-12 was obtained by continuously withdrawing D-12 to ! its fully withdrawn position and measuring the reactivity change on the strip l chart recorder. This also added positive reactivity to start to increase ! reactor power for the fully misaligned flux map. Total worth of D-12 is the sum of the small steps to 185 steps withdrawn plus the withdrawal to fully i ! withdrawn. I t l 6-45 3 l l l
6.8 PSEUDO ROD EJECTION TEST (1-6SF-04) (Continued): l l N A full core flux map was teken to determine F q. Reactor power was then D l returned to the zero power level by inserting and aligning D-12 to its bank. The lift coils were reconnected and the bank realigned to its overlap position l for further testing. l Results The objective and acceptance criteria were met. 1 The worth of D-12 was measured to be 492 pcm, whereas prediction was 490 pcm. The acceptance criteria required inflating the measured value by ten percent { to 541.2 pcm for comparison with the safety limit of <860 pcm, and it passed this acceptance criteria. T The incore flux map with D-12 fully withdrawn verified that F was 6.18 by: ) T N q F = F x 1.03 x 1.05, l q q l which is less than the limit of 13.0. l I l l Problems
)/ There were no significant problems. d l
The large spatial effects due to a highly tilted core as D-12 was withdrawn made determining worth somewhat more difficult than normal, but the measured worth was very close to prediction. l l 1 l i k l 1 l
) 6-46 ~/ ,
1 l l l l
6.8 PSEUD 0 ROD EJECTION TEST -(1-6SF-04) (Continued):- FIGURE 6.8.1: PSEUD 0 EJECTED ROD D-12 DIFFERENTIAL WORTH PSulDO-EJECfD R0D D-12 LOW POWER PHYSICS TEST
]
6 6 D-12 DI)TERDGIAL WORTH hifA l 5.5 A 5 k 4.5 l 1 4 E 2 H3 A.
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6.8 PSEUDO ROD EJECTION TFST (1-6SF-04') (Continued):' FIGURE 6.8.2: PSEUD 0 EJECTED ROD D-12 INTEGRAL WORTH PSalD0-EJECTED R0D D-12 IM PCMEs PHYSICS TEST 559 -
^
A D-12 INTECML WORTH MTA 599 450 l 400 359 399 l 2 259 9 E
$299 d
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O 6-48
t l l
'SECTION 7 l l
1 POWER ASCENSION TESTING t 1 l l l l D 7-1
7.1 POWER ASCENSION TEST SEQUENCE (1-600-13) Objective The objective of the power ascension sequence was to define the sequence of operations, beginning at approximately 5% power, which constitute the power ascension testing program. The abstract for this test is FSAR Section 14.2.8.2.50. Methodology The plant was brought from less than five percent power to full power in a stepwise manner that permitted control systems and plant operation to be observed and adjusted as necessary. There were five major testing plateaus, at 30%, 50%, 75%, 90%, and 100% of full power with minor plateaus every five percent for turbine-generator monitoring. Initial power increases were conducted slowly according to Westinghouse fuel preconditioning guidelines. The Vogtle Unit 1 Startup Chronology section details the time and power levels for each test. Ic aeneral, the plan for testing at each major plateau was similar. The plant .s brought to power and stabilized for approximately a day before meaningful power measurements or adjustments could be made. A calorimetric determination of reactor power was performed and nuclear instruments adjusted. Control systems were tested and adjusted. Tests requiring xenon equilibrium were then conducted, followed by a transient or escalation to the next plateau. The listedconduct in Tableof7.1.1. each test is described. Tests included in this section are Results The results and acceptance criteria for each power range test is included in the test's separate discussion. Supplements will be issued as required by Technical Specification 6.8.1.1 until all testing has been completed. Problems Each test summary contains a section that discusses problems encountered during the tests and their solutions. 7-2
[ l 7.1 POWER ASCENSION TEST SEQUENCE (1-600-13) (Continued): TABLE 7.1.1: POWER ASCENSION TESTS (Tests are listed in alpha-numeric order.) l PROCEDURE i NUMBER TITLE DESCRIPTION l i 1-SHB-01 Waste Evaporator Demonstrates operation of the liquid I Performance Test waste evaporator. I 1-SHE-01 Boron Recycle Demonstrates operation of the boron Evaporator Performance recycle evaporator. 1 1-SSB-01 Reactor Protection Verified and documented reactor Test protection system trip setpoints. 1-5SC-02 Thermal Power Measurement Gathered data for determining core and Statepoint Data thermal power and other statepoint Collection measurements including RCS flow and loop differential temperature. 1-5SE-02 Operational Alignment Calibrated the Nuclear Instrumentation Of Nuclear Instrumentation System (NIS) and intially set the System Precritical reactor trip setpoints prior to critical operation. 1-SSF-06 Operational Alignment Of Verified full power temperature rise Process Temperature across the reactor vessel during Instrumentation precritical and power ascension. 1-5SF-07 Startup Adjustments Of Calibrated the reactor temperature Reactor Control System control program for optimum plant efficiency. 1-5SQ-01 Metal Impact Monitoring Collected baseline data for Digital System Test Metal Impact Monitor System setpoints. 1-600-05 Biological Shield Surveyed the plant for radiation Survey levels during the test pro 8 ram. 1-600-06 Dynamic Response Monitors response of selecteil Test piping to transients. 1-600-08 Remote Shutdown Test Demonstrated the capacity of the plant to be shut down a1d stabilized in hot standby from outside the control room. conditions. 7-3
7.1 POWER ASCENSION TEST SEQUENCE (1-600-13) (Continued): TABLE 7.1.1: POWER ' ASCENSION TESTS (Continued): PROCEDURE NUMBER TITLE DESCRIPTION 1-600-09 Loss Of Offsite Power Demonstrated the capacity of the At Greater Than plant to withstand a total loss of 10 Percent Power offsite power from normal low power 1-600-10 Natural Circulation Demonstrated natural circulation j Demonstration with decay heat, gathered data for ! benchmarking the training' simulator, J and provided-operator training. 1-600-11 Thermal Expansion Monitored thermal growth of selected Test piping'during heatups and power operation. 1-600-12 Primary and Secondary Monitored primary and: secondary Chemistry chemistry. I 1-600-14 Ventilation Capability Demonstrated the capability-of.the Test plant ventilation system to cool rooms containing safety-related ! equipment, j 1-6AB-01 Dynamic Automatic Steam Operated the steam dump (turbine. Dump Control bypass) contre . valves and made adjustments. l-l 1-6AE-01 Auto Steam Adjusts the steam generator level l Generator Level control system at various' power l Control levels and during transients. 1 l 1-6AE-02 Calibration of Steam and Calibrated steam flow transmitters i Feedwater Flow Instrumen- and determined if feedwater flow tation At Power transmitters required calibration. !. 1-6BG-01 Gross Failed Fuel Performed adjustments and gathered l Detector operating data for'setpoint adjustment. , 1-6EF-01 Ultimate Heat Sink Demonstrated the ability of the , Heat Rejection ultimate heat sink to reject i Capability Test heat. 1-6QF-01 Inplant Communications Verified operability'of the inplant communications system. l-7-4 l I
I I
- 7.1 POE!ER ASCENSION TEST SEQUENCE (1-600-13) (Continued)
TABLE 7.1.1: POWER ASCENSION TESTS (Continued): PROCEDURE _ NUMBER TITLE DESCRIPTION l 1-6RJ-01 At-Power Intercomparison Compared protection system control Of Reactor Protection board instrument readings with the System Inputs And Plant Proteus computer's readings at various Computer Outputs Test power levels. 1-6SC-01 Power Coefficient Measured the ratio of power and Determination temperature change for comparison with predictions. 1-6SC-02 Load Swing Test Varies load to obtain data for adjusting control systems. 1-6SD-01 Process And Effluent Verified operation and gathered Radiation Monitoring operating data. System Test 1-6SE-01 Axial Flux Difference Calibrated the power range nuclear i Instrument instrumentation for correct response
- Calibration to axial core power changes.
1-6SE-02 Incore Movable Detector Determined core power distribution And Thermocouple by flux maps using the incore movable Mapping detector system. 1-6SE-03 Operational Alignment Calibrated the nuclear instruments at Of Nuclear power based on calorimetric power Instrumentation determination and adjusted trip System At Power setpoints. 1-6SF-01 Automatic Reactor Verified proper response of the reactor Control in automatic rod control. l 1-700-01 Large Load Reduction Reduces load 50% to obtain data for adjusting control systems. 1-700-02 Plant Trip From 100% Obsersed plant response to a generator ! Power trip from full power. l 1-700-03 Steam Generator Determines amount of moisture in j Moisture Carryover steam leaving the steam generators. l l 1-800-01 Plant Performance Monitors plant performance and , makes adjustments for maximum j plant efficiency. 7-5 l
-7.1 POWER ASCENSION TEST SEQUENCE (1-600-13) (Continued): TABLE 7.1.1: POWER ASCENSION TESTS (Continued): PROCEDURE NUMBER-TITLE DESCRIPTION 1-800-02 Nuclear Steam Supply. Demonstrated ability of the nuclear-System Acceptance steam supply system to operate at Test full power for one hundred hours. I 7-6
y 7 D 1 l SECTION 7.2 l l 1 ! l l NUCLEAR STEAM SUPPLY SYSTEM ' l l (NSSS) TESTING i i 1
i l 7.2.1 POWER COEFFICIENT DETERMINATION (1-6SC-01)' ;
.)
Objective The objective of the Power Coefficient Measurement was to. verify nuclear design predictions of the Doppler-only power coefficient. q The abstract for this test is~FSAR Section 14.2.8.2.26. 1 1 Methodology The test acquired data of reactor power changes resulting from changes in 1 reactor average temperature (Tavg) and compared change to design' predictions at the 30%, 50%, 75%, and 100% power plateaus. This simple method relies only , on changes in Tavg and reactor power level, without control rod motion. The reactor was operated at a constant power level long enough to establish j xenon equilibrium. A two pen strip chart recorder was connected to Tavg and ! delta-T and calibrated for expanded ranges. A range of approximatley ten degrees was used for Tavg and two to three degrees was used for delta-T. A strip . chart recorder .was used because it allowed easy visualization of l stability, could easily be marked with a pen, and was a convenient permanent i record. A Proteus computer trend of the same parameters and others was also ) established in case the recorder failed or was miscalibrated. I The rod control system was placed in manual control to prevent rod motion _and { pressurizer level and pressure, and steam generator level ' controls were ! verified to be in automatic. A sampling program was started to verify that RCS born had not changed significantly over the course of the test by sampling the RCS and pressurizer approximately every half. hour. The test was started by reducing turbine load approximately 40 Mwe, sufficient to increase Tavg two to three degrees F. The plant was allowed to stabilize- ') l and turbine load was returned to approximately its original value. 'This cycle J was repeated at least three times. The 50% power test is illustrated in Figure 7.2.1.1. Data consisted of the change in Tavg and reactor differential temperatures at ! the stability periods. The data was analyzed according to vendor specified- 1 procedure and compared to predictions. Results are listed in Table'7.2.1.1. The 100% power determination was originally scheduled for 90% power, but so little ' time was scheduled for the 90% power plateau that three extra days waiting for requisite Xenon stability was not warranted and it was decided to perform the test at full power. Load swings were always initiated by decreasing power and care was taken not to overshoot full power when returning to the original power level. 7-8
7.2.1 POWER COEFFICIENT DETERMINATION (1-6SC-01) (Continued): I
- )
Results W s The acceptance criteria was met at all plateaus.on all- measurements. Table 7.2.1.1 lists actual results. The objectives were met. l Problems 'l 4 q The recorder's delta-T indication was misscaled for the 50% power test .and. I based on the misscaled results the test initially failed its ~ acceptance criteria. The delta-T pen actually was calibrated to 2*F change per inch pen travel, not 3 F as was marked on the chart itself. The Proteus printout'taken1 ) simultaneously was plotted (and is included as Figure 7.2.1.1) and showed the i scaling problem. Recalculations using the revised data produced acceptable results. . It was observed that the test should be conducted in a short time with no more than approximately five to fifteen minutes between major power' changes and ; only one smooth continuous turbine power change should be made. Prolonged I operation at a lower power level than original increases xenon concentration and appears to affect results. O ; 1 f Ii g ,
. . _ . . _ _ _ _ . _ . . _ _ _ . _ _ . _ _ _ . _ _ _ _ _ = _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ . _ _ _ . _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ . - _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _
7.2.1 PORER COEFFICIENT DETERMINATION (1-6SC-01) (Continued): . {\ TABLE 7.2.1.1: POWER COEFFICIENT RESULTS VERSUS POWER LEVEL k[
=-
PERCENT POWER PREDICTED MEASURED RATIO PLATEAU 1TC* PWR COEF** _ RATIO *** _ RATIO # jfA DIFFERENCE (# 30 -5.38 pcm/ F -13.3 pcm/% 2.4721 -2.1964 0.2757 50 -7.27 pcm/'F -12.7 pcm/% 1.7590 -1.6817 0.0773 75 -9.33 pcm/ F -11.6 pcm/% 1.2433 -1.1412 0.1021 100 -12.13 pcm/ F -10.53pcm/% 0.8681 -1.0155 0.1474 ITC (Isothermal Temperature Coefficient) contains a bias due to the zero power physics testing difference between predicted and as-found boron concentrations of +1.08 pcm/ F. This bias was added to the raw predicted value to produce biased ITC. .n PWR COEFF (Power Coefficient, pcm/% power) is due to the reactor ~ power level. RATIO is PWR COEFF/(biased ITC) and is *F/ percent power.
# MEASURED RATIO is the ratio of *F change per percent power change with percent power change being calculated by changes in reactor vessel differential temperature. i
## RATIO DIFFERENCE is the arithmetic sum of MEASURED RATIO and PREDICTED RATIO. The acceptance criteria is 10.5 and was always met.
,t
) 7-10 i - b
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7.2.2 'INCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING'(1-6SE-02)- i l l Objectives The objectives of the Incore Movable Detector and Thermocouple Mapping test were to: (1) obtain'and analyze core power distribution for various control. ) rod configurations and' power levels; (2) verify that the actual. core. power distribution is maintained within- the power distribution limits of the i Technical . Specifications at applicable power levels; .and (3)- obtain -{ thermocouple map. data as required by the power ascension test sequence. The abstract for this test is FSAR Section 14.2.8.2.30. ) 1 i Methodology . Core power distribution maps were performed.throughout-the' power' ascension program using the movable incore detector system to obtain data.for input into the INCORE code. Measurements were made at low power (less than 5% power for Control Bank D at 0 steps, all rods'out, pseudo ejected rod and rod insertion limits) and during the 30%, 50%, 75% and 100% power plateaus (all rods out)'to verify design calculations and show compliance with Technical Specifications.- i i For each flux map the INCORE code generated values for the FQ(z) and FXY hot channel factors at 61 axial points along the active fuel length. The top 20 FDH values were also generated. The INCORE code also determined the quadrant l power tilt ratios for the top and bottom of the core and the core average axial offset. $ Results j All objectives were met. Results for all maps are seen in Tables 7.2.2.1 through 7.2.2.8. Results for core parameters at applicable power levels [ Axial Flux Difference (Mode 1 >15% power), Hot Channel Factors (Mode 1 > 5%. power), Quadrant Power Tilt Ratio (Mode the 1 > 50% power)] acceptance criteria. were well within Technical Specification-limits, meeting criteria. All but one of the low power maps met all acceptance Thermocouple maps were obtained as required. Problem i In only one case was there a problem. With rods at the zero power insertion limit and the core at approximately three percent of full power, the measured incore tilt derived from Fdelta-h values from all assemblies exceeded its Westinghouse acceptance criterion of 1.02, but was within the Westinghouse safety criterion of 1.04. An evaluation concluded that flux maps at low power levelo 'have large uncertainties, and since no safety criteria had been i violated, maps would be taken at higher power levels. Testing at 35% and 50% power indicated no excessive tilt. D 7-12 l
i 7.2.2 INCORE MOVABLE DETECTOR AND THERM 0COUDLE MAPPINGL(1-6SE-02) (Continued): O TABLE 7.2.2.1: ZERO POWER FLUX MAP 1 (ALL CONTROL RODS FULLY WITHDRAWN (ARO)) Plant __ Conditions
.]
Test Date: 3/18/87 Power Level: 2% Boron Concentration: 1353 ppm Rod Position: CD = 228 steps. I Core Burnup = <10 MWD /MTU Summary of Core Parameters
- i FQ(z) - All measured values, including uncertainties, were within the limit.
The most limiting FQ(z) value was 2.325 at core location D-4, axial point 29. This was 49% below the limit 4.577.- I l FDH - Technical Specification limit is 1.989. l All measured FDH values, including uncertainties, were within the limit. The limiting FDH value was 1.494 at core location D-4. This was 25% below the limit. FXY - All measured FXY values, including uncertainties, were
/ within the limit.
The most limiting FXY value was 1.5856 at core location D-4, axial point 48 where the limit was 1.831. This was 13.4%' below the limit. ; l Quadrant Power Tilt Ratios Top Half of Core Bottom Half of Core Quadrant 1: 1.0236 1.0263 i Quadrant 2: 0.9979 0.9931 Quadrant 3: 0.9777 .) 0.9730 Quadrant 4: 1.0008 1.0075 ! i Technical Specification limit for Mode 1 > 50% power-is 1 1.02. ; Westinghouse safety review criterion is 1.04. i Core Average Axial Offset: +1.963
*Although limits for hot channel factors and quadrant power tilt ratios are not applicable until Mode 1 above 5% power and 50% power respectively, the INCORE calculated values were reviewed for any significant deviations form.
the limits. 7-13
l i i 7.2.2 TNCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING (1-6SE-02) __(Con tir ued ): TABLE 7.2.2.2: ZERO POWER FLUX MAP l 1 (CONTROL BANK D AT ZERO STEPS WITHDRAWN, ALL OTHERS FULLY WITHDRAWN) '
]
I Plant Conditions l ! Test Date: 3/13/87 Power Level: 3% l Boron Concentration: 1293 ppm Rod Position: CC = 200 steps, l Core Burnup = <10 MWD /MTU CD = 0 steps. l Summarv of Core Parameters
- FQ(z) - All measured values including uncertainties were within the limit.
The most limiting FQ(a) was 2.661 at core location J-2, axial point 29. This was 42% below the limit of 4.577. FDH - Technical Specification limit is 1.988. 1 All measured FDH values, including uncertainties were within the limit. The most limiting FDH value was 1.681 at core location J-2. This was 15.4% below the limit. FXY D - All measured FXY values, including uncertainties, were i within the limit. The most limiting FXY value was 1.8000 at core location J-2, 1 axial point 48, where the limit was 2.033. This was 11.5% below the limit. j l l Quadrant Power Tilt Ratios
)
Top Half of Core ( Bottom Half of Core j Quadrant 1: 1.0252 1.0290 i Quadrant 2: 0.9529 0.9888 Quadrant 3: 0.9770 { 0.9734 Quadrant 4: 1.0009 1.0088 Technical Specification limit for Mode 1 > 50% power is 11.02. Westinghouse safety review criterion is 1.04 Core Average Axial Offset: -1.728 ' 1
*Although limits for hot channel factors and quadrant power tilt ratios are not applicable until Mode 1 above 5% power and 50% power respectively, the )
INCORE calculated values were reviewed for an significant deviations from i the limits. D 7-14 s _______.________.___-_.i
7.2.2 INCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING (1-6SE-02) (Continued): ) TABLE 7.2.2.3: ZERO POWER FLUX MAP (CONTROL RODS AT INSERTION LIMITS) Plant Conditions Test Date: 3/19/87 Power Level: 2% Boron Concentration: 1260 ppm Rod Position: CB = 161 steps, Core Burnup = <10 MWD /MTU CC = 46 steps, CD = 0. steps. Summary of Core Parameters
- FQ(z) - All measured values, including uncertainties, were within the limit.
The most limiting FQ(z) value was 2.738 at core location L-2, axial point 39. This was 40% below the limit of 4.6 FDH - Technical Specification limit is 2.008. All measured FDH values, including uncertainties, were within the limit. The most limiting FDH value was 1.647 at core location D-8. This was 18% below the limit. FXY - The maximum measured FXY value, including uncertainties, exceeded the limit.. ) The , aximum value of 2.1469 occurred at axial point 11 where the ilmit was 2.047 This was 4.9% above the limit. Quadrant Power Tilt Ratios Top Half of Core Bottom Half of Core Quadrant 1: 1.0256 1.0352 Quadrant 2: 1.0225 0.9964 Quadrant 3: 0.9794 0.9719 Quadrant 4: 0.9926 0.9965 Technical Specification limit for Mode 1 > 50% power is i 1.02. Westinghouse safety review criterion is 1.04. Core Average Axial Offset: -31.292.
*Although limits for hot channel factors and quadrant power tilt ratios are not applicable until Mode 1 above 5% power and 50% power respectively, the INCORE calculated values were reviewed for any significant deviations from the limits. The FXY limit being exceeded by 5% (2.1469) can be attributed to the rod configuration and the uncertainty in power level below 5% at which the flux map was taken. The 5% power limit on F is 2.2.
XY Measured incore tilt derived from Fdelta-h values from all assemblies exceeded its Westinghouse acceptance criterion of 1.02, but was within Westinghouse safety criterion of 1.04 See the " Problem" portion of this section. 7-15
I l 7.2.2 INCORE MOVABLE DETECTOR AND TIIERM0 COUPLE MAPPING _(1-6SE-02) i l (Con tinued )- i i TABLE 7.2.2.4: ZERO POWER FLUX MAP i w (PSEUD 0 EJECTED ROD FULLY MISALIGNED) ! ) Plant Conditions ; i i Test Date: 3/20/87 Power Level: 2% ! ! Boron Concentration: 1231 ppm Rod Position: CB = 158 steps, i
- Core Burnup = <10 MWD /MTU CC = 43 steps, {
l CD = 0 steps, l Rod D-12 = 228 steps. For core parameters see Section 6.8 of this report (test 1-6SF-04). s ! l l ! s Le .sd l J j l l l l 7-16 i l L _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _
7.2.2 INCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING (1-6SE-02) (Continued): TABLE 7.2.2.5: 30% POWER PLATEAU FLUX MAP flant Conditions Test Date: 4/03/87 Power Level: 35% 4 Boron Concentration: 1064 ppm Rod Position: CD = 215 steps Core Burnup = 46 MWD /MTU ; Summary of_ Core Parameters
- FQ(z) - All measured value.s. including uncertainties, were within j the limit.
( The most limiting FQ(z) value was 2.024 at core location D-4, ' axial point 29. This was 56% below the limit of 4.577. i FDH - Technical Specification limit is 1.85. All measured FDH values, .l including uncertainties, were j within the limit.
)
The most limiting FDH value was 1.438 at core location D-4. ~ This was 22% below the limit. l l FXY - All measured values, included uncertainties, were within the limit. The most limiting FXY value was 1.52 at core location D-4, axial point 47, where the limit was 1.739. This was 13% below the limit. I Quadrant Power Tilt Ratios Top Half of Core Bottom Half of Core Quadrant 1: 1.0086 1.0127 Quadrant 2: 1.0002 0.9965 Quadrant 3: 0.9892 0.9860 Quadrant 4: 1.0020 1.0049 Technical Specification limit for Mode 1 > 30% power is f_ 1.02. Core Average Axial Offset: -1.286. l 7-17
(
~
! 7.2.2 TNCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING (1-6SE-0.2) (Continueo): TABLE 7.2.2.6: 50% POWER PLATEAU FLUX MAP Plant Conditions Test Date: 4/16/87 Power Level:~ 4P", Boron Concentration: 1010 ppm ' Rod Position: CD = 215 steps t Core Burnup = 116 MWD /MTU Summary of Core Parameters
- FQ(z) - All measured values, including uncertainties, were within i t
the limit. I The most limiting FQ(z) value was'2.004 at core location D-4, axial point 30. this was 56% belcw the limit of 4.589. FDH - Technical Specification limit is 1.794. All measured FDH values, including uncertainties, were within the limit. The most limiting FDH value was 1.433 at core location D-4. This was 20% below the limit. FXY
- All measured values, including uncertainties were within the limit.
The most limiting FXY value was 1.5127 at core location D-4, axial point 49, where the limit was 1.702. D below the limit. This was 11% Quadrant Power Tilt Ratios Top Half of Core Bottom Half of Core Quadrant 1: 1.0062 1.0128 Quadrant 2: 1.0023 1.0000 Quadrant 3: 0.9910 0.9856 l Quadrant 4: 1.0005 1.0016 l Technical Specification limit for Mode 1 > 50% power is 1 1.02. l-Core Average Axial Offset: -1.305. i 7-18 i i l i
7.2.2 INCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING (1-6SE-02) (Continued): TABLE 7.2.2.7: 75% POWER PLATEAU FLUX MAP Plant Conditions Test Date: 05/03/87 Power Level: 74% Boron Concentration: 940 ppm Core Burnup = 442 MWD /MTU Rod Position: CD = 210 steps Summary of Core Parameters
- FQ(z) - All measured values, including uncertainties, the limit. were with.1n The most limiting FQ(z) value was 1.965 at core location h-5, axial point 36. This was 37% below the limit of 3.110 FDH - Technical Specification limit is 1.67.
All measured FDH values, including uncertainties, were within the limit. The most limiting FDH value was 1.395 at core location H-5. This was 16% below the limit. FXY - All measured values, including uncertainties, were within the limit. The most limiting FXY value was 1.4728 at core location M-12, O axial point 47, where the limit was 1.620. below the limit. This was 9% Quadrant Power Tilt Ratios Top Half of Core Quadttnt 1: Bottom Half of Core 1.0024 1.0118 Quadrant 2: 1.0075 0.9990 Quadrant 3: 0.9895 Quadrant 4: 0.9895 1.0006 0.9997 Technical Specification limit for Mode 1 > 50% power is < 1.02. Core Average Axial Offset: -3.713. I l l
7.2.2 INCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING f_1-6SE-02) (Continued): TABLE 7.2.2.8: 100% POWER PLATEAU FLUX MAP Plant Conditions Test Date: 5/31/87 Power Level: 100% Boron Concentration: 880 ppm Rod Position: CD = 228 steps Core Burnup = 840 MWD /MTU ! Summn:ir_of Core Parameters
- l FQ(z) - All measured values, ircludic.c uncertainties, were within the limit. l I
The most limitin8 FQ(z) value was 2.018 e core location H-5, axial point 37. This was 13% below the lin t of 2.31. 1 FDH - Technical Specification limit is 1.552. I Ala. measured FDH values, including uncertainties, were' wikhin the limit. { 1 The most limiting FDH value was 1.419 ut core location D-9. ' l This was 9% below the limit. FXY - All measured FXY values, including uncertainties, were within the limit. l The most limiting FXY value was 1.4789 at core location H-5, axial point 35, where the limit was 1.542. . This was 4% i r% below the limit. O Quadrant _._ Power Tilt Ratios Top Half of Core Bottom Half of Core Quadrant 1: 1.0100 1.0015 ' Quadrant 2: 1.0060 1.0044 Quadrant 3: 0.9886 0.9915 Quadrant 4: 0.9954 1.0027 Technical Specification limit for Mode 1 > 50% power is 1 1.02. , Core Average Axial Offset: -6.783. 1 I l 1 i 1 7-20 l
r i i 7.2.3 THERMAL POWER MEASUREMENT AND STATEPOINT DATA COLLECTION (1-SSC-02) J Objectives l Objectives of the Thermal Power Measurement and Statepoint Data Collection l Test were to: (1) determine reactor power by performing a heat balance at 30, 50, 75, 90, and 100 percent power; and (2) identify instrumentation for statepoint data collection. l The abstract for this test is FSAR Section 14.2.8.2.8. Methodology The statepoint data collected by this procedure provided required data for the following startup tests: Reactor Coolant System Flow Measurement (1-5BB-04); l Calibration Of Steam And Feedwater Flow Instrumentation At Power (1-6AE-02); Operational Alignment Of Process Temperature Instrumentation (1-5SF-06); i Startup Adjustments Of The Reactor Control System (1-5SF-07); Operational ' j Alignment Of Nuclear Instrumentation System (1-6SE-03); and Nuclear Steam Supply Acceptance Test (1-800-02). l 1 Q Special instrumentation permitted accurate feedwater flow, temperature, and m) pressure, steam pressure, and reactor hot and cold leg temperatures. High j { precision (f% accuracy) temporary feedwater differential pressure gauges were ) installed parallel to one of the permanent 1% feedwater flow transmitters on ! each steam generator flow measuring venturi. High precision pressure gauges 1 l were also added to monitor feedwater pressure and steam pressure to each steam ; l generator. Temporary instruments were installed to monitor the hot and cold l ' 1eg resistance temperature detectors (RTDs) for each reactor coolant loop. The l temporary instruments were typically calibrated within seven days prior to their use at each power plateau. l Data taking at each major plateau consisted of recording data at five minute intervals for an hour. Data from the temporary and control room instruments was recorded by hand. Data was also collected on the Proteus computer. l This data was then used to determine feedwater mass flow rate and enthalpy, I steam enthalpy, core thermal power, reactor loop differential temperatures and power. l The results from these calculations and additional data determined whether or j not adjustments were required prior to proceeding to the next major plateau. The 30, 50, 75, and 90 percent power plateau calculations were manually l performed. Prior to the 100% power plateau a personal computer program was !] v developed and validated using data from the 75% tnd 90% power plateaus. All 100% power plateau calculations, including the NSSS acceptance test, were completed using the computer. 7-21 l
7.2.3 THERMAL POWER MEASUREMENT AND STATEPOINT DATA COLLECTION (Continued) (1-5SC-02)
/~'% i k -) m At all power plateaus the secondary chemistry requirements were within !
specification so steam 8enerator blowdown could be isolated. All statepoint j measurements were obtained without steam generator blowdown. ' I Results ! i All objectives and acceptance criteria were met. Data was successfully collected. Instruments required for precision calorimetric and statepoint were ; identified, calibrated, and installed. ' i i Problems i Durin8 the initial data collection for the 50% and 75% power plateaus the i requirement for stable plant parameters could not be maintained within the 11/2% of full power tolerance band. At the 50% plateau a grid instability l caused by the trip of Florida Power and Light's St. Lucie Unit 2 Nuclear Power j Plant interrupted date taking. At the 75% plateau the load dispatcher adjusted i ('"3 system load and caused a slight load reduction. In each case data collection i ,_/ s was terminated, data voided, and when plant stability was achieved data collection was restarted and successfully retaken. The temporary instruments were in six different locations so a large number of personnel and communication and coordination was required. The plant paging system and sound powered phones were successfully used for communication. l N 7-22 1 _.__-_-__ . - __ __--__ -___________-____-___-__ ._. a
r l l l l 1 i l I 1 SECTION 7.3
- l l
l t l 1 INSTRIMENT CALIBRATION AND ALIGNMENT l D l l l l l 7-23 I
7.3.1 OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION SYSTEM AT POWER (1-6SE-03) Objectives The objective of the Operational Alignment of Nuclear Instrumentation System at Power Test was to establish and determine voltage' settings, trip settings, operational settings, alarm settings, and overlap of channels on source ran8e. intermediate range, and power range instrumentation from prior to initial-criticality to at or near full reactor power. l The abstract for this test is FSAR Section 14.2.8.2.10. Methodology '1 This test consisted of recording data, verifying instrument linearity, and adjusting setpcints. Each evolution will-be discussed separately below. ' Source Range / Intermediate Range Overlap 'l During the approach to initial criticality the count rate readings from each source range channel and current readings from each intermediate range channel-were recorded. starting when the first intermediate range channel started to respond and increase from E-11 amperes. Readings were taken until the souree range channels were blocked to prevent a reactor trip from high' flux. i Intermediate Range / Power Range Overlap During power ascension from low power physics testing (<5% powet) to the 30% power plateau the intermediate range channels current readings and the percent power readings from each power range channel were recorded and at least 1.5 decades of overlap was verified. l Resetting The High Flux Trip Setpoint Before escalation to the next higher plateau power level the power range -high flux trip was reset to'20% above the new power plateau's using the appropriate i Instrumentation and Control procedure. Power Range Instrument Gain Adjustment Soon after the plant had reached a new power plateau a precision calorimetric was perfo.rmed to determine actual reactor power and the gain on each power range channel was adjusted so power indicated by the power range , nuclear instruments read within 2% of actual. 7-24 1
'l 7.3 1 OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION SYSTEM AT i POWER (1-6SE-03) (Continued):
Detector Linearity Currents from each power range detector were recorded and plotted at each new power plateau to verify linearity. Flux Deviation Amplifiers At each power plateau from 50% to 100% the flux deviation averaging amplifier' ; power deviation alarm was checked and reset, -if . need ed , - using calibrated voltage sources to produce an alarm if any detector sensed a power . deviation 8reater than 2% from the average of the four power range channels. Intermediate Range Detector Voltage Plateau Upon reaching stability at the 100% power plateau the intermediate range detectors voltages were adjusted in 50 volts direct current (vdc) steps from 300 to 1500 vde and corresponding detector currents were recorded and plotted. 1 The initial 800 vde setting was verified to be on the flat part of the curve (in the proportional range); the response from 300 to 1500 vdc was constant. l 1 Power Range Detector Voltage Plateau Steps taken to determine the high voltage plateau curve for intermediate range detectors were repeated for the power range detectors. Detectors voltages were adjusted in 50 volts direct current (vdc) steps from 300 to 1500 vde -and corresponding detector currents were recorded and plotted. The initial 800 vde setting was verified to be on the flat part of the curve; the response ~from 300 to 1500 vde was constant. Source Range Discriminator And High Voltage Setting When the reactor had been shut down after operating at 100% power, the discriminator bias voltage setpoint and the high voltage power supply setpoint was reset for each source range channel. This was accomplished using the irradiated reactor core as a neutron and gamma source in contrast to the initial calibration which used a 3.0 Ci AmBe neutron source. The discriminator's bias voltage was decreased from -0.05 vde to -4.0 yde in 0.025 volt steps for each detector and the corresponding count rate was recorded and plotted (see Figure 7.3.1.1). The bias voltage which produced the smallest slope was selected as the new operating point to discriminate against background gamma radiation not present during initial calibration. The high voltage power supply output was then adjusted-from 1900 to 2300 vdc in 50 volt steps for each detector and the corresponding count. rate recorded and plotted (see Figure 7.3.1.2). The high voltage power supply was adjusted to a setting that gave e count rate on the flat portion of the curve at least D 7-25
7.3.1 OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION SYSTEM AT POWER (1-6SE-0 ) (Continued): thirty percent greater than the " knee" voltage. Intermediate Range Channel Compensating Voltage Adjustment Each intermediate range channel's ion chamber compensating voltage was adjusted following the trip from full power to produce a detector response of ; IE-11 amperes. The voltage was changed from its original setting of -40 1 l vdc to -14 vde for N35 and -22 vdc for N36. Power Range Detector Currents The power range channels full power detector currents were determined by placing each power range channel in the "DET A" position and adjusting the A detector test signal until the full power meter read 50%, then placing the channel in "DET B" and adjusting the B detector test signal until the full power meter read 50% again. The channel was then placed in "DET A+B" and the full power meter was verified to read 100%. The individual detector currents i for each channel were then recorded. This process was repeated for each power ' range channel in succession. i Results All objectives and acceptance criteria were met. The operational alignment of the NIS was performed at power. Overlap between the source range and intermediate range channels, and between j the intermediate and power range channels was verified to be greater than 1} ' decades. The power range detector currents versus core power were determined and verified to be linear. Prior to each power escalation the power range high flux trip setpoint was readjusted to 20% of full power above the next plateau. The flux deviation alarm from average power was checked at each power plateau from 50% to 100% of full power. At 100% power the operating voltages for the intermediate and power range channels were verified to be on the flat portion of the response curve. After shu t <' awn from 100% power the operating voltages of the source range channels were verified to be on the flat portion of the response curve. Intermediate range detector compensating voltages were set to read 1E-11 amperes and the power range test currents for 100% full power were determined. l Problems Power range detector N43's response to high voltage became erratic above 1150 l vde, but since its operating range was 800 vdc no immediate corrective action t was taken. An investigative work order was written for the next extended outage. 7-26 1 1
7.3.1 OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION SYSTEM AT POWER (1-6SE-03) (Continued): FIGURE 7.3.1.1 SOURCE RANGE DETECTOR DISCRIMINATOR SETTING SOURCE RANCE DISCRIMIMf0R UCLTAGE SEff!NG 399 [ A S0llRCE RANCE H31 (SET 9 9.7Y) p D SOURCE RANCE M32 (SET 9 9.8U)
! 250 200 ;
159 R 0 199 'N , 59 3 l9. s , is ,
\h Rs DISCRIMIMf0R UOLTACE SEfflNC
, ,s o 7-27
7.3.1 OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION SYSTEM AT POWER (1-6SE-03) (Continued): FIGURE 7.3.1.2 SOURCE RANGE DETECTOR HIGH VOLTAGE SETTING SOURCE PAME HIGl UOLTACE SEff!NG 300 A source RANCE M31, SET P 2200U 0 S0uRCE RAEE N32, SET P 2158V 250 l 200 159 ._ g a higg / / 50
/
/
I h 0 en . .* 1.7 II8 If9 d 2)1 2.'2 22T XV YOLTAGE SEfflNC 7-28
I l l 7.3.2 OPERATIONAL ALIGNMENT OF PRCCESS TEMPERATURE INSTRUMENTATION (i-5SF-06) Objective The objective of the Operational Alignment of Process Temperature Instrumentation Test was to align the delta temperature and average temperature process instrumentation under isothermal conditions prior to i criticality, and at power. The abstract for this test is FSAR Section 14.2.8.2.16. Methodology This procedure assured proper alignment of the Delta T and Tavg process instruments prior to initial criticality and during power L temperature I ascension. At full power (90% to 100% reactor power) the alignment was checked j and final adjustments performed. ( Initial alignment of the RCS process temperature instruments was verified by substituting RTD simulators for the actual RTDs and simulating temperatures corresponding to an assumed full power differential temperature (Delta T) of 51 F. Correct response was verified and the circuits were restored. With the plant stable at no-load hot standby condition of RCS average temperature (Tavg) of 557t}*F the outputs of the process temperature circuits were measured and recorded. Delta T and Tavg were then calculated and compared to D the indicated values to verify proper calibration at assumed isot.hermal : j conditions. l During power ascension the output of selected process instruments in the I protection racks corresponding to hot leg temperature (Thot), cold leg temperature (Tcold), Delta T and Tavg were measured and recorded. The resistances of RTDs installed in the RTD bypass manifolds but not connected to instrumentation (called " spare") were also measured and recorded. This process was repeated at all power plateaus in conjunction with precision secondary calorimetric reactor power level determination, startup procedure 1-5SC-02. At the 75% power plateau the Thot and Tcold temperatures measured at each plateau up to that point were used to determine hot and cold leg fluid i enthalpies. The enthalpy values were correlated versus power and extrapolated ! to values expected at 90% and 100% power. These extrapolated enthalpys were ' then converted to corresponding Thot and Tcold temperatures and expected Delta T and Tavg calculated for each coolant loop. The expected full power Delta T l values were averaged and the Delta T summing amplifiers recalibrates to reflect these values. 1 i Final values for differential temperature at full power for each loop are shown in Table 7.3.2.1. I i 7-29
7.3.2 OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE INSTRUMENTATION (1-5SF-06) (Continued): O V Delta T is displayed to the reactor operators in terms of percent of full power, or Delta T Power. The values of Delta T Power obtained at the 90% and 100% power pleteaus were compared for agreement with the power levels calculated by 1-5SC-02. The output of the Tavg circuits was compared with calculations of expected Tavg from the output of Thot and Tcold circuits and outputs of the Thot and Tcold circuits were compared with That and Tcold values calculated from resistances measured at the spare RTDs. At the 100% power plateau fluid enthalpies at all power plateaus levels were determined, plotted, and inspected for linearity. The full power enthalpies were then used to calculate new values for full power Thot and Tcold temperatures. These temperatures were used to calculate full power Tavg values for each loop. Any loop whose Delta T value was not within 11% of 100% reactor i power or the Tavg vah e was not within 11% of the average Tavg value was l l rescaled to bring it within this tolerance. Temperatures are shown in Figure 7.3.2.1 and enthalpies in Figure 7.3.2.2 for data taken during the power ascension. Results 1 All objectives and acceptance criteria were met. Proper alignment of the Delta T and Tavg process temperatures was verified
%) prior to criticality.
During subsequent power ascension testing, at each power plateau, the response ! of the temperature instruments was verified and recorded. Before escalation I above the 75% power plateau, the Delta T power was rescaled from its conservative setting of 51*F to mere adequately reflect actual plant response. At the 100% power plateau the calibrations were checked against actual plant response and the Delta T summing amplifiers rescaled to reflect 100% power. l l Problems Two of the four reactor coolant loops (#2 and #4) produced greater power due to higher flow rates so the same scaling factors could not be used for all four loops. Separate factors were used as necessary. ' Reactor temperatures also tended to be depressed somewhat at lower power levels due to an excessive Tavg-Tref mismatch that was corrected prior to the 90% power level and final alignrent of the process temperature instruments at full power. The full power values were used and at the full power plateau temperatures were verified correct within acceptable tolerances. 7-30
7.3.2 OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE INSTRUMENTATION (1-5SF-06) (Continued): A Q TABLE 7.3.2.1: FULL POWER REACTOR COOLANT SYSTEM DIFFERENTIAL TEMPERATURES l l LOOP _ NUMBER F DIFFERENTIAL- TEMPERATURE l 1 55.46*F 1
- l. 2 56.20*F l l
3 55.70*F k 1 4 57.05 F l l 1 1 4 O 7-31
7.3.2 OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE INSTRUMENTATION -- (1-5SF-06) (Continued): FIGURE 7.3.2.1: REACTOR COOLANT SYSTEM TEMPERATURES j 1 I l f RDCTOR C00! ANT SYSIDI IDIPERATURES US RDCTOR PWR { [ 629 ' A LOOP 1 THof A LOOP 1700D A LOOP 2 iH0T A LOOP 2 fC0 0 . #/ 619 0 LMP 3 TH0f 0 LMP 3 TC00 I l LOOP 4 TH0f l LOOP 4 TCOD f-
,/
699 l 599 580 / l 579 h e 569 j ;(__ - i g W A 559 il 19 2'8 39 49 59 6'8 Y9 88 98 1 96) PERCENT RDCf0R POWER i 7-32
7.3.2 OPERATIONAL ALICNMENT OF PROCESS TEMPERATURE INSTRUMENTATION (1-5SF-06) (Continued): FIGURE 7.3.2.2: REACTOR COOLANT SYSTEM ENTHALPIES RDCTOR C00! ANT SYSTDI DffHALPIES VS RDCTOR POWER t 649 6 LOOP 1 TH0f A LOOP 1 TCOLD ' 639 A LOOP 2 IHOT A LOOP 2 iCOLD ,/ 0 LOOP 3 TH0T C LOOP 3 TC0LD l LOOP 4 7H0f l LOOP 4 fCOLD l 629 .
'Y i
t 619
,7
> 699 f _ l /
/
599 l'
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a 589 6' 579 j 569 ___ _
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__A 9 w l R 559 I il l'8 2'8 3'8 49 5'8 69 Y9 8'8 M9 190 PERCDG REACTOR POWER l 7-33 I a
L L ! 7.3.3 CALIBRATION OF STEAM AND FEED 9ATER FLOS INSTRUMENTATION AT POWER (1-6AE 02) Objective l- The objective of the Calibration of Steam and Feedwater Flow Instrumentation l At Power test was to assure proper calibration of the steam and feedwater flow' transmitters. The main feedwater flow transmitters are the main contributors to secondary calorimetric determination of reactor power, which has a variety of uses including calibrating the reactor protection circuitry. 'The main steam flow transmitters must be calibrated to as-found differer,tial pressures developed mostly by flow restrictors in the steam generator'steem nozzles. t This test was not discussed in the FSAR, but was performed as part _of the standard Westinghouse startup test methodology. { Methodology
\
Calibrations of the feedwater flow and steam flow differential pressure .j transmitters were correlated to and . corrected by temporary instruments. j installed for precision calorimetric determination of reactor power conducted 'l during the thermal power measurement and statepoint data collection test. .A i precision secondary calorimetric was performed at each power plateau. High i precision (1/4% accuracy) differential pressure gauges were temporarily installed parallel to permanent' plant feedwater flow . differentia 1' pressure ' instruments (1/2% accuracy) to measure pressure developed by feedwater flow D venturies. (The venturies ad been calibrated to 1/4% accuracy. by an independent laboratory, and since the only copper in the feedwater system is at the condenser tube sheets, decalibration of the venturies by copper fouling is considered unlikely.) The temporary instruments were generally calibrated-within seven days of their use, and so are considered to provide the most accurate differential pressure signal and a standard against which other flow l l instruments were calibrated. I l At each major power plateau, the output of the steam and feedwater flow l transmitters were measured, recorded, and converted to differential pressure sensed by the transmitter based on its previous calibration. This . measured differential pressure was then extrapolated to full scale and compared to the actual transmitter calibration. Any steam flow transmitter found significantly I out of tolerance was recalibrates prior to escalation .to the next power plateau. Steam flow and feedwater flow rates were compared' to verify mismatches were within allowable tolerances. Results
.{
All objectives and acceptance criteria were met. 1 All eight steam flow transmitter calibration ranges were obtained, as listed in Table 7.3.3.1. Operation of-the permanent feedwater flow transmitters was verified. D 7-34 i 4
L i 7.3.3 CALIBRATION OF STEAM AND FEEDUATER FLOW INSTRUMENTATION AT POWER (1-6AE-02) (Continued): Problems At the'30% power plateau the four steam flow transmitters on steam generators
#2 and #3 were found to have a zero offset of approximately ten inches water column and one feedwater flow transmitter was suspected of not being properly i calibrated. Steam flow is sensed by differential pressure between condensing pots on the upper level taps of the steam generators to pressure taps at approximately the same elevation on the main steam lines inside containment.
Differential pressure is developed by steam flow through the steam generator internals, the flow restricting nozzle, two right angle bends in the main steam line, and three short runs of main steam piping. The ten inch offset was due to the low pressure tap on the main steam lines I being ten inches lower than the condensing pots. The offset resulted in a l constant flow rate of about one million pounds per hour flow; it was large j because flow is proportional to the square root of differential pressure and a i small differential pressure offset at zero flow appears as substantial flow. l Revised calibration values incorporating the offset provided' correct steam ' flows. The suspect feedwater flow transmitter was recalibrates. At the 75% power plateau, four steam flow transmitters were recalibrates. One feedwater flow transmitter required recalibration. Three of the feedwater flow transmitters were found to have malfunctioned and were replaced amd D successfully retested. l At the 100% plateau the four steam flow transmitters not recalibrates at the i 75% plateau required recalibrations. A cross-check between feedwater and steam flows at each power level was performed and demonstrated that all transmitters were operating satisfactorily. l \ l l l 1 l 7-35 _.--__.--__._--___-_--J
7.3.3 CALIBRATION OF STEAM AND FEEDWATER FLOW INSTRUMENTATION AT POWER (1-6AE-02) (Continued): I O
\,) TABLE 7.3.3.1: STEAM AND FEEDWATER FLOW TRANSMITTER RANGES STEAM STEAM FLOW TRANSMITTERS FEEDWATER FLOW TRANSMITTERS GENERATOR INSTRUMENT NO. W.C.* RANGE INSTRUMENT NO., W.C.* RANGE.
I 1-FT-512 0 - 457.3 1-FT-510 0 - 1847.09 i 1 1-FT-513 0 - 441.4 1-FT-511 0 .1842.60 2 1-FT-522 10 - 471.0 1-FT-520 0.- 1835.39 l~ 2 1-FT-523 10 484.5 1-FT-521 0 - 1838.36 i 3 1-FT-532 10 - 455.4 1-FT-530 0 - 1840.86 , 3 1-FT-533 10 - 466.7
~
1-FT-531- 0 - 1844.96 j 4 1-FT-542 0 - 461.0 1-FT-540 0.- 1845.54 i 4 l 4 1-FT-543 0 - 457.3 1-FT-541 0 .1847.40 j W.C. means inches water column at ambient. temperature. Range is calibration range for the span of the instrument. This span is greater than the full power' flow. 4 i 7-36 i
7.3.4 AX1AL FLUX DIFFERENCE INSTRUMENT CALIBRATION (1-6SE-01) m Objective The objective of the Axial Flux Difference (AFD) Instrumentation Calibration b' Test was to derive calibration factors for overpower and overtemperature differential temperature (delta-T) setpoints, based on incore flux data,
- calorimetric data, and excore nuclear instrumentation detector currents.
d l The abstract for this test is FSAR Section 14.2.8.2.29. Methodology l l This procedure calibrated reactor protection circuitry that protects the core from axial core power distributions that at high power levels could lead to i departure from nucleat boiling (DNB). Gain settin8s in the power range nuclear instruments are adjusted so core-average axial power shape is inferred from excore power range detector currents. Setpoints in the overpower and overtemperature circuits are also adjusted. l Calibration factors for overpower and overtemperature delta-t setpoints were ! derived based on incore flux map data, calorimetric data, and excore nuclear l instrumentation detector currents. The response of the excore power range l detectors was demonstrated to be linear with respect to the incore axial power distribution. l The procedure had three main parts. First was a preliminary adjustruent at 50% j i power based on full core flux maps taken at intermediate low power plateaus ; and the 50% plateau. Second, a substantial change in axial core power shape 1 was forced by inserting control bank D and this change provided the greatest ; range of calibration. Finally, calibration was checked at full power. l At the 30% and 50% power plateaus, a full core flux map was performed and the corresponding excore detector currents were recorded. Reactor core thermal ! power was determined by calorimetric data taken during the flux maps. The j axial offset was then determined at each power level, first using the incora data and then the excore currents. Incore versus excore linearity was also verified. Calibration factors were derived to allow use of excore detect &c 4 current signals to represent actual incore flux input to the various proces's ' , indications and protection circuits. The process indications and protection circuits were then recalibrates to reflect these new factors. This initial calibration was performed prior to escalating above the 50% power plateau. The response of the process indicators and protection circuits was verified by injecting test signals into each excore power range channel and recording the outputs and/or indications of P various components downstream (ie, test points, indicators, . recorders, and computer points). . 7-37 Y
.T ,
__ . _ _ . 1. _________i
7.3.4 AXIAL FLUX DIFFERENCE ldSTRUMENT CALIBRATION _(1-6SE-01) (Continued):
- i During testing at the 75% plateau, a xenon oscillation was induced by ' '
inserting control rods while diluting, and then withdrawing rods while / , borating to produce incore core axial offsets from -21.1% to +11.6%. quarter-core flux maps were taken at various incore axial effsets and Fulland{'/ excore l, ' detector currents were recorded during each flux map.
*!s f Importantparametersfortheaxialxenonoscillationat75%powerareshownin$;j\
Figure 7.3.4.1 (RCS Boron Concentration Versus Time), Flux Difference Versus Time), Figure 7.3.4.3 (Control Bank D Posir ior.. (Arial - t> Figure 7.3f..? Tame), Vd suG and Figure 7.3.4.4 (Reactor Power Versus Time). Each grapd shc,ws tne approximate test time, starting at approximately 2200 hours t May A 1987 qad continuing for the next twelve hours, and the twenty two houns preceeding tde test. The preceeding time is shown for the stabilit.y period prior to initiating the oscillation. yl At test time secondary plant problems threatened to cut short the test so it was started as soon as boron fad ' stabilized, indicating a reasonably stable bulk xenon concentration. .f s t As shown in Figure 7.3.4.1, once bulk xenon had been stabilinc', little l further , distribution. change was observed despite considerable change in ,the po;;er d
, y+ <
> 1 As shown in Figures 7.3.4./ and 7.3.4.3, the oscillation insertin vas. e ducec(3 by
-9.5% , g control bank D to drive and maintain AFD to an intermodsdry step i of i for a quarter core flux map. Another quarter core nap wasdtahen at -
12 %, and a full core map at -153%. Rods were then withdrawn 'in ile:ervals to f/ ) f,\ obtain and maintain AFD in approximately 2}% steps for quarter ' core fitV maps and finally were fully withdrawn. AFD increased in this unrodded condition to a peak of approximately +9.5 and a full core flux map was taken. The oscillation was then damped by rods for further testing. Damping the 1 oscillation with rods satisfied Regulatory Requirement 1.68, ApMndix A.5.d. s , A precision calorimetric was performed prior to starting the xenon osciUbb, / , ' to determine actual core thermal power. ' Excore detector currents were normalized to 100% power by linear scaling using data from the precision calorimetric. The normalized excore detector currents were then plotted versus the incore delta-q to detendne the
+
slope (proportionality constant) of these two factors. (See Figure N 3.4.6.) i The intercept of the tcp and bottom excore detector equations was used to \ calibrate of a conversion factor for each detector. The sum of the absolute value the product of proportionality constant and the conversion factor for each power ran8e channel (top and bottom detector) yielded the "masu red" relationship between the incore and excore axial flux signals. . This was then divided delta into the calibration span factor (percent span divided by percent flux) of the process indication and protection r!rcuits to produce the necessary gain to be used for the overtemperature delta temperat'2re (OTDT) delta flux summing amplifier and for the process computer. 1 ) 7-38 9 E 1 L
7.3.4 A%IAL_ FLUX DIFFERENCE INSTRUMENT CALIBRATION, (1-6SE-01) (Continued): i s'3 . U At the 100% power plateau another full core flux-map was performed and the computer's delta flux and process indicators delta flux readings were tocorded. Ang the INCORE computer program the flux mp data was analyzed to determin- actual core power distribution. Tvis was cocipared to the process indicetors and computer readings for agreement. The 100% power AFD calibration values are shown in Figure 7.3.4.5 g jid listed in Table 7.3.4.1. 'The three data points and their extrapolations are shown in the Figure and Ifsted in the Table. (A scheme.tir. 6f the excore nuclear instruments affected by this procedure is shown in : Figure 7.3.4.5.) These final values were used for calibration. 1 Results l i All objectives and receptance criteria were met.
\ Too ax!al flux difference cal:.bration constants were determined and the necessar'/ calibrations and verifications were performed on the proceas computer, process indicators,and recorders, and the protection circuits. Tte resnonae cf the excore power range '
' detectors was verified to 'ae linear to core !
powcr flux distributions.
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e' .; i r l The Instrument and Ccattol sur+e111ance procedures used to calibente the power range excore chaar.els and i deltN temperature protection circuits were based on biase ;grteinperature griginal dqsign values from the~ earlier Precat?.iorts . Limitaticas,:aad Jetpoints (PLS) had were revised to reflect the latest values. i f 1. An attempt yas mafe<, to eluninate a potential calibration deficiency related to possible ground loops in the input to the delta flux summing smplifier by 1 changing the input amplif.1ers.,-ain g from 0.8 to 1.0. The result, while , , eliminating the ground loom calibration, was an, incorrect scaling in the /' f, overtemperature delta tempeaiture protection circuits. The incorrect scaling was discoverei during the pb dr. ascension from 50V power plateau to 75% power I'
, plateau end vaa corre.: fed e:, V.e 75% power pkteard n< t f , i ('
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7.3.4 AXIAL-FLUX DIFFERENCE INSTRUMENT CALIBRATION (1-6SE-01) (Continued): 1 The techniques used to davelop the calibration constants for' incore power distribution versus -excore power range channel response allowed for the _ difference in the relative flux densities sensed by the excore power range j detectors but assumed their relationship with' average core power was linear. This is apparent by the use of a linear relationship for the AFD target curve. After the calibration of the.. excore power range channels and the. temperature protection circuits at the ,75%~' power overtemperature c' - ~ . plateau, the actual nonlinear relationship was proven.by verification of AFD l- at the 100% power level. At'the 100% power plateau, while allowing xenon _ to reach equilibrium levels,_ three full core flux maps were' performed and the - excore detector currents were recorded. New calibration constants were derived. using the data from these flux maps and the excore power range channels .and i overtemperature delta temperature protection circuits were recalibrates. Another full core flux map was then performed, axial power. distribution determined, and agreement .obtained from the excore power range' channel instruments. E y I i l 4 1 - l i > 7-40
- __ ___ =____.__ _ __ _ ____1__o
7.3.4 AXIAL FLUX DIFFERENCE INSTRUMENT CALIBRATION (1-6L~.-01) (Continued): 1 TABLE 7.3.4.1: FULL POWER AFD CALIBRATION VALUES 2 (Refer to Figure 7.3.5.5 for a sketch of the excore nuclear instrument system i showing where the following values were used.) 1 The following is a synopsis of data from the three' flux maps taken at full power: uA I (Normalized-top) uA I (Normalized-bottom) I(Delta-0)# N41 N42 N43 N44 N41 N42 N43 N44
-5.085 314.42 297.17 327.25 307.67 349.92 341.17 370.33 352.25 .j -6.783 310.80 292.80 322.80 303.40 353.30 344.30 373.80 355.30 -8.220 307.00 290.00 319.60 399.90 355.80 346.30 375.40 356.80 }
l i Notes: # I(Delta-Q) is celta-q from the INCORE program and from incore detectors. DEFECTOR l EXCORE NORMALIZED uA CURRENT ## SLOPE CONVERSION FACTORS l CHANNEL I* top I* bottom btop bbottom Rtop Rbottom N41 326.5 340.4 2.353 -1.886 0.0255 0.0245 ! N42 308.8 333.0 2.305 -1.633 0.0270 0.0250 1 1 N43 339.7 362.2 2.462 -1.640 0.0245 'O.0230 l N44 320.3 345.1 2.489 -1.445 0.0260 0.0241 Notes: ## I* top and I* bottom are detector currents (normalized to full power) at delta-Q = 0.0. l 1 n I = I* + b x I(Delta-Q) ! R = 8.333/I* 7-41
7.3.4 AXIAL FLUX DIFFERENCE INSTRUMENT CALIBRATION (1-6SE-01) (Continued): TABLE 7.3.4.1: FULL POWER AFD CALIBRATION VALUES (Continued): SUMMING AMPLIFIER GAINS I 1 EXCORE CHANNEL SUMMING AMPLIFIER GAIN # ! N41 1.569 N42 1.617 I N43 1.700
)
i N44 1.674 1 G = Rtop x btop - Rbot x bbottom i Note: # G is overtemperature delta-t summing amplifier gain. These values are also the excore/incore conversion factors used in the Proteus software. l l 1 l l l 7-42 l _
7.3.4 AXIAL FLUX DIFFERENCE INSTRUMENT CALIBRATION _(1-6SE-01) (Continued): FIGURE 7.3.4.1: RCS BORON CONCENTRATION VERSUS TIME l
)
l i 75x POWER XDt0N OSCILLATION j lose ! l l 8 PPN RCS BORON CONCDURATION
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11 d k d d l'8 l2 l'4 l'6 l'8 2'9 2'2 2'4 2'6 28 3'8 3'2 34 36 38 40 42 4( HOURS flNE STARf!NC TRON HIDNICHT 5-7-87 O 7-43
l 7.3.4 AXXAL FLUX DIFFERENCE INSTRUMENT CALIBRATION. , (1-6SE-01)-(Continued): l FIGURE 7.3.4.2: AXIAL FLUX DIFFERENCE VERSUS TIME l 1 i 75% POWER XENON OSCl!JATION is
, A AFD FROM N141 Q 6
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t t -29 il d k d d [9 [2 [4 [6 [8 I6 2'2 N4 26 IB [9 I2 [4 [6 [8 48 4'2 40 HOURS flME STARflNG FROM MIDNIGHT 5-7-87 l $ 7-44
7.3.4 AXIAL FLUX DIFFERENCE INSTRUMENT CALIBRATION (1-6SE ,1) (Continued): FIGURE 7.3.4.3: CONTROL BANK D POSITION VERSUS TIME 75% POWER XENOH OSCILLATION 230 A STEPS WifHDRAWN 229 210 A M ,g I 199 url 1 3 e f 189 _
/
2 [170 ' o 169 14 Z
--APPROXIMATE TEST DUPAfl0N--
150 d 5 k d d [0 [2 [4 [6 [8 29 N2 2'4 26 [9 N9 N2 N4 N6 N8 49 42 44 HOURS flME STAFTING FROM MIDNICHI 5-7-87 7-45
7.3.4 AXIAL FLUX DIFFERENCE INSTRUMENT ___ CALIBRATION (1-6SE-01) (Continued): FIGURE 7.3.4.4: REACTOR POWER VERSUS TIME l l I q 75% POWER XENON 0$CILIAfl0N ! 89 A PERCENT REACTOR POWER 79 A I 18 I l ! l 77
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7.3.4 AXIAL FLUX DIFFERENCE INSTRUMENT CALIBRATION D -6SE-0.8) (Continued): FIGURE 7.3.4.6: 100 PERCENT POWER CURRENTS 1 E i l i f 199x POWER ATD DATA l368 ] O 355 h350
' 345 -
0 B0TTON DETECTOR - N41 N l Z i 335
/
I LINEAR DGRAPotATION OF DATA 330 325 - l A TCP DETICTOR - H41 315 / 319-3h
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7-48
7.3.5 REACTOR PROTECT'ON TEST (1-5SB-01) Objectives d Objectives of the Reactor Protection Test were: (1) to verify setpoints for the reactor protection system were verified to be at their initial values prior to intitial startup; (2) during power ascension, to document at successive power plateaus any adjustments and intermediate values of trip ) setpoints; and (3) for the 100% power plateau, to adjust and record final reactor trip setpoints as specified by test program data. The abstract for this test is FSAR Section 14.2.6.2.6. Methodology The procedure consisted mainly of verifying trip setpoints in the reactor protection system to insure they were in agreement with the Westinghouse . Precautions, limitations, and Setpoints (PLS) document except when superceded i by changes to Technical Specifications Limiting Safety System Settings. i Prior to initial criticality, trip setpoint data was obtained from review of completed surveillance procedures. During the remainder of the startup program, setpoint data was obtained through review of maintenance work orders I generated by st trtup test program sequencing procedures. p l x. Results All objectives and acceptance criteria were met. I All reactor protection system setpoints recorded in this procedure agreed with ! the Westinghouse PLS except in three cases where the PLS was superceded by Technical Specifications. Technical Specifications were used in: pressurizer low pressure safety injection setpoint; containment high pressure; and low-low steam generator level. Problems For the pressurizer low pressure safety injection (SI) setpoint, 1885 psig was used instead of 1870 psig until resolution of uncertainties associated with
- Veritrak pressure transmitters are resolved. For containment high pressure, a setpoint of 3.8 psig instead of 3.5 psig was used to conform with Technical Specifications. Finally, for low-low steam generator water level, 22.5% narrow range instead of 18.5% narrow range was used until resolution of uncertainties associated with Veritrak transmitters have been resolved.
O 7-49 1 \ 1
7.3.5 REACTOR PROTECTION TEST (1-SSB-01) (Continued): f, ')s Nuclear Instrumentation overlap data (from test 1-6SE-03) turned out to be less accurate and less conservative than calorimetric data (from test 1-5SC-
- 02) for determination of final setpoints for C-1 rod interlock and IR high level trip. Calorimetric data (available at higher power levels) was used to determine final setpoints.
Dttring testing the overtemperature delta T trip parameter K2 was discovered not to reflect the latest version of the Westinghouse PLS. The calibration procedure was changed and the value was verified by retest. No other problems were encountered during performance of the test. 1 I l l l l l { b
%) 7-50
7.3.6- AT-POWER INTERCOMPARISON OF REACTOR PROTECTION SYSTEM
~
INPUTS AND PLANT COMPUTER OUTPUTS TEST (1-6RJ-01) Objective The objective of the At-Power Intercomparison of Reactor Protection System 4 Inputs and Plant Computer Outputs Test was to demonstrate satisfactory. agreement between inputs to the Reactor Protection System (RPS) and plant computer (Proteus) outputs. l j The abstract for this test is FSAR Section 14.2.8.2.57. i 1 Methodology This test was performed at the 30%, 50%,. 75%, and 100% power plateaus. A comparison between Proteus and permanent plant instrumentation was -performed at each specified plateau to verify readings were within the required acceptance criteria band. All readings for comparison with Proteus were taken from the main control board and remote shutdown panels except RCS narrow range cold leg temperatures. These temperatures were determined by measuring the voltage signal in the process cabinets and converting the voltage to temperature by the following formula: Volts x 12*F/ volt + 510*F = F. Table 7.3.6.1 lists parameters and acceptance criteria. Results-i The ir.tercomparisons met acceptance criteria. All objectives were met. 1 Problems i Various instruments were found out of calibration. After recalibration these instruments were verified to meet the required acceptance criteria, where applicable. I j i O 7-51
k 7.3.6 AT-POWER INTERCOMPARISON OF REACTOR PROTECTION _ SYSTEM INPUTS AND PLANT- COMPUTER OUTPUTS TEST (1-6RJ-01) (Continued): TABLE 7.3.6.1: PARAMETERS AND ACCEPTANCE CRITERIA PARAMETER ACCEPTANCE CRITERIA
- RCS Narrow Range Cold Leg Temperature 0.5'F Overpower /0vertemperature Delta T Trip 1.5%
Setpoint RCS Differential Temperature (Delta T,) 21.5% RCS Taverage 1.0'F RCS Wide Range Thot and Tcold 7.0 F RCS Flow 11.2% f j Pressurizer Pressure 18.0 psi l Pressurizer Level 1.0% Steam Generator Wide Range Leve.1 11.5% Steam Generator Narrow Range level 1.0% Steam Generator Pressure 13.0 psi Containment Pressure 0.5 psi Main Turbine First Stage Pressure 8.2 psi Comparison of parameters had to be within the acceptance ] criteria. { l ( l 1
)
l I l 7-52 l 1 I
7.3.7 STARTUP ADJUSTMENTS OF THE REACTOR CONTROL SYSTEM (1-5SF-07) Objective The objective of the Startup Adjustments of the Reactor Control System Test I was to obtain optimum plant efficiency. The abstract for this test is FSAR Section 14.2.8.2.17. j Methodology l Thermal efficiency of the cecondary plant depends on inlet steam temperature, so this test maximized efficiency by obtaining the highest possible secondary steam pressure without exceeding main turbine pressure limitations or the maximum allowable Tavg of 588.5F. The test was di,vided into two sections, one prior to initial criticality, and the other at the major testing plateaus of 30%, 50%, 75%, and 100% power. Each test section began with verifying plant stability and recording overall plant equipme.nt status. The plant computer (Proteus) was used to initiate trends of selected parameters that were then averaged over a ten minute interval. These parameters included: (1) Reactor thermal power; (2) Main Turbine First Stage Pressure; (3) Main Generator Gross Electrical Output (Mwe); A (4) Reactor Dynamic Total Thermal Heat Output; V (5) Steam Generator Total Thermal Heat Output; and (6) Pressurizer Pressure. The following were also n.onitored for each of the four reactor coolant loops: (7) Hot and Cold Leg Temperatures (Thot and Teold); (8) Reactor Average Temperature (Tavg); (9) Steam Generator Pressure; (10) Main Feedwater Flow; and (11) Main Feedwater Temperature. l Graphs of selected parameters were prepared for comparison with design curves and extrapolation to full power values after the 50% power data collection. These parameters included: Average loop Thot, Tcold, Tavg, Steam Generator Pressure, and Main Turbine First Stage Pressure. The extrapolated steam generator pressure was compared with design steam generator pressure of 1000 psia and a saturated steam temperature difference was computed. This temperature difference was used for any necessary rescaling of the cutput of the Treference (Tref) program for the Rod Control System. The extrapolated Turbine First Stage Pressure was also compared to design values of approximately 665 psia for rescaling the input to the Tref program. This 7-53
l 7.3'.7 STARTUP ADJUSTMENTS OF THE REACTOR CONTROL SYSTEM (1-5SF-07) [' .(Continued): process was continued at the 75%, 90%, and 100% power'platea'us until a final turbine first stage -pressure value and corresponding Tref were achieved. f J , Figure 7.3.7.1 is RCS average temperatures, Figure 7.3.7.2 is steam generator, I
- pressure and 7.3.7.3 is turbine first stage pressure, versus reactor power.. j l
! Results i The final Tref- program was determined to be 0 to 651.3 psig for 0 to 100% j power and 557 to 588.5 F. ' l The objective was met. l Problems Plant equipment availability (including in-service or out-of-service moisture separator reheaters and feedwater heaters) impacted achieving design turbine l
.first stage pressure at low power levels so the curves of RCS temperature, !
steam generator pressure, .and turbine first stage pressure are low in the 30% ! l to 40% reactor power range. This ceased to be a problem at high power levels ' ! when the plant was brought closer to design conditions. Final values of the ! Tref program were made at the 100% power level at maximum plant efficiency. i i I 7-54
l 7.3.7 STARTUP ADJUSTMENTS OF THE REACTOR CONTROL SYSTEM (1-5SF-07) ~ (Continued): l FIGURE 7.3.7.1: THOT, TCOLD, AND TAVG VERSUS POWER l I RIACTOR COOLANT SYSTEM TEMPERATURES US RIACTOR PWR 629 A H07 LEC fMPERATURE j 615 0 AVERACE fBPERATURE (IAUG) 619 $ COLD LEG f B PERATURI 695 699 3, 595 599 /
/
585 589 _ 575 . . l 579 / m. 565 - 569 . A _a [
~
W E 555 Nb c W A 559 il 19 29 39 49 5'8 69 78 88 9G 1 961 PERCENT REACTOR POWER 7-55
7.3.7 STARTUP ADJUSTMENTS OF THE REACTOR CONTROL SYSTEM (1-5SF-07) (Continued):
.1 FIGURE 7.3.7.2: AVERAGE STEAM GENERATOR PRESSURE VERSUS POWER i
i
- STEAM CDERATOR PRESSURES VERSUS POWER 1199 a 6 AYERAGE PRES $11RE, PSIA 1999 \
1989 s f im \ ; W h869 I a l b959 I = 0 l k1949 l l \ I g839 N l g1929 d y NN )
> N l'
1919 m . g w i $1999 , Il 19 2'S 39 4'8 5'8 6'8 f0 89 N9 1 911 l PERCENT REACTOR POWER ; l 7-56
! 7.3.7 STARTUP ADJUSTMENTS OF THE REACTOR CONTROL SYSTEM (1-5SF-07) (Continued): l FIGURE 7.3.7.3: TURBINE FIRST STAGE PRESSURE VERSUS POWER j l l l l TURBINE TIRST STAGE PRESSURE UERSUS POWER 799 ) Tm1 STM PRM 659 , 699 / l i 559 599 V i 459 [ l l l 499
/
/_
350
/ \
399 l 256 i l l l W l K 299 e 159 l W X l A.199 0 59 ./ 9 9 [9 N9 N9 4'O 59 6'8 N9 88 N9 1 941 PDCmt RDCTOR POWER 7-57
O SECTION 7.4
.ONTROL SYSTEMS DYNAMIC TESTING C
O l 1 a l l O 7-58 l 1 4
7.4.1 LARGE LOAD REDUCTION (1-700-01) L Objective D The objective of the Large Load Reduction Test is. to demonstrate satisfactory 1 plant. transient response to a 50% power load reduction without tripping or -) actuating safety equipment. The abstract for this test is FSAR Section 14.2.8.2.52.- l 1 1 Methodology The test is performed at 75% power and 100% power. The load reduction of 50% l power is initiated from the turbine control panel via the 133%/ min' load- ' reduction circuit by holding in the ramp button for 23 seconds. The automatic ! control systems are allowed to respond and stabilize the plant. Initially the automatic rod control ~ system will reduce reactor power by 10% and the steam , dumps to the condenser will absorb the other 40%' power. As rods automatically insert to match Tavg and Tref, reactor power and turbine power will be matched at the new power level of 25% (for the load rejection from 75%) or 50% (for the load rejection from full power). 1
.{
the Steam Generator Level Control System; the l The systems Feedwater Pump monitored Controllers;are:. the Reactor Control System; the Pressurizer Level and Pressure Control System; and the Steam Dump Control System. Results All results have not been obtained in time to be included in this report but will be discussed in later s'applements. t Problems 1 At 75% power the control rods did not move due to a faulty relay card. j Operators immediately responded by manually inserting rods as turbine power ' was decreased. Turbine power was reduced to 26% and reactor power was reduced to 56.5%. Approximately three minutes into the transient, steam generator 2 main 1 feedwater regulating valve began to oscillate causing. fluctuations in flow. The controller was placed in manual to stabilize the steam generator level. ] The plant stabilized without further manual control by the operators. The. l other steam generators did not experience the same oscillations experienced by steam generator 2. Investigation into the problem led to an adjustment in the booster relay associated with the control valve. l 7-59 j
7.4.2 LOAD SWING TEST _(1-6SC-02) Objectives The objective of the Load Swing Test is to verify nuclear plant transient response, including automatic control system performance, when step load changes are introduced to the turbine-generator at 30, 75, and 100% power-levels. The abstract for this test is FSAR Section 14.2.8.2.27. Methodology With plant conditions stable at the desired power levels and all reactor ) automatic control loops in automatic, turbine load is decreased by about 10% ! of rated power.'The plant is allowed to respond and stabilize while important-plant parameters are monitored for any necessary operator intervention. If plant parameters remain within allowable tolerances, turbine load is t increased by about ten percent and the same plant parameters are again ] monitored. Trend data for detailed analysis is obtained using the Proteus J computer and Emergency Response Facility (ERF) computer. Data is recorded from main control board indicators before and after the transient. Step change load decreases are obtained from the turbine-generator by' setting the Standby Load Set potentiometer to a setting representing a load 10% less than the existing load and switching turbine controls to Standby. Step load increases are obtained by returning the turbine control to Normal. See O Figures 7.4.2.1 through 7.4.2.5 for results of the 75% power tests. Results Testing is incomplete for the 100% power plateau and will be reported in j a supplement to this report. Problems i During the first load swings at 30% power some data from the computer was not available so control room strip charts were used to obtain the needed data. At the 75% power level, the Automatic Rod Control circuit did not function properly. The circuits were investigated but no fault could be found. The problem was believed to be caused by a sticking relay on one of the control j cards. A functional test demonstrated that the rods worked properly. The { i card was replaced shortly thereafter anyway. The -10% load swing was repeated and the results were satisfactory. The +10% load swing was then attempted but the plant tripped on low-low steam generator . water level as reactor power increased to approximately 80% and generator load reached 949 Mwe. The power excursion was due to improper setting of the turbine load limiter. 7-60 1 1
)
7.4.2 LOAD SWING TEST (1-6SC-02) (Continued): ) The method of increasing load was then changed to a manually-controlled ramp to preclude reactor trip. The +10% load swing was again attempted but was - aborted due to excessive svings in steam generator water levels from approximately 28% to approximately 71% narrow range level. The results of the 75% power test revealed that the gains associated with the steam generator level control needed to be increased. Modest adjustments were made for an . improved transient response. See Table 7.4.2.1 for a listing of the controller , L settings. c
=
I i {
$ i 6
o 4 -
,' >(
) 7-61 e4ss s
.4 4 .
7.4.2 LOAD SWING TEST (1-6SC-02) (Continued):
FIGURE 7.4.2.1: 107. LOAD DECREASE - STEAM GENERATOR LEVELS LOAD SWING AI 75x POWER - 10% LOAD DECREASE 78 i d A STEAM GENERATOR 81 LEVEL A STEAM GENERATOR 82 LEVEL A 0 STEAM GENERATOR 53 LEVEL Z
W ( $ STEAM GENERATOR 84 LEVEL U L i 60 N O%f o ,
! I i
kk I 55 a M ij Q Ns j a rk 4 8 5fMA R' sq' k K Z 40 il $ 10 lb 20 2t' ; l TIME FROM START OF TRANSIENT (MINUTES) l l ^ 7-62
7.4.2 LOAD SWING TEST (1-6SC-02) (Continued): FIGURE 7.4.2.2: 10% LO D DECREASE - REACTOR POWER AND CONTROL 3ANK D POSITION 1 I l LOAD SWING AT 75x POWER - ISx LOAD DECREASE 89 299
]
6 REACf0R POWER X A COMTROL BAMX D IN STEPS l l 199 RLA k I
- 75. l 189 l 3
l l .!?9 79. 169 , a 159 kg E X 149 h g 2
- L l h 4 h39 L 69' Q m Al,4A--.--n p 3. 3 ? ? .1l,4 ' 29 l
2 c w a j E 55 _ , 1119 8 I II h 19 l'5 29 2$ flME FROM START OF TRANSIENT (MINJfES) l l 7-63
7.4.2 LOAD SWING TEST (1-6SC-02) (Continued): FIGURE 7.4.2.3: 10% LOAD INCPEASE - STEAM GENERATOR #1 LEVEL
> LOAD SWING Af ?$X POWER - 10 PERCEHI LOAD INCREASE I h 56 0
1 A STEAM GENERATOR 1 LEVEL h 54
$ o jAk.
i
, So 52
,i R A 4
t[ l48 \s A A W i p
$46 \ A
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# 28
~1 '
h d h d $ l'8 l'1 12 l3 l'4 15 TIME of TRANSIDif (MINUTES)
~
7-64
7.4.2 LOAD SWING TEST (1-6SC-02) (Continued): A FIGURE 7.4.2.4: 10% LOAD INCREASE - } REACTOR POWER AND CONTROL BANK D POSITION l LOAD SWlHG Af 75x POWER - 10x POWER INCRFASE 80 205 4 RFACTOR POWER 78.
, , ,_A_a A_A_A3_r#_ L, p u ffy ?p * *D ". *I.N..'$!! '
76. 6 g0r.s[>COff/CO@!r/AT;00ZOOCOT/0000007;C000c0e s
- 74. 0 185
! O CONTROL BANX D (SIEPS) 189 72.
175 ) 70. f 170
^
X E v 68. 165 C 2 $
$ 169 E O 66 7 ,,
g g m E , 155 o u 0 64. u me;l 150 a s a K 62 145 5
-1 hI $ $ k$ d $d d 18 11 l2 13 14 15 TIMI 0F TRANSIENT (MINUTES) 7--65 i
r
>1 +j i 7.4.2 LOAD SWING TEST (1-6SC-02) (Continued):
c l f3 FIGURE 7.4.2.5: 10% LOAD INCREASE - PRESSURIZER PRESSURE AND LEVEL ) k, i l l l LOAD SWlHG Af 75x POWER - 10 PERCINT LOAD INCREASE 58 2255 f 2254 $ l FPESSURIEER FRESSURE (PSIC) 2253 I k 56. l 2251
- 55. I IH 22'Ja 54 3 1 48 k 53.
I i 2246 bl 0 52. l[ , 2245 igC" '2244 f 2243 i j 50- @ 0 PRES!URIZER LEVEL (x) 4
$ 49. 2248 5 d
2239h
! 48; p 2238$
[ h237$ I
$ 'G J236 y
# 0 2235 E 2?.34 h b 45 2233 E
-1 hI d h k$ dhd h I'0 l'1 l'2 13 14 15 flME OF TRANS!ENT (MIN) l l
l 7-66 !q l L.) l l l l
/
7.4.2 IDAD SWING TEST (1-6SC-02) (Continued): TABLE 7.4.2.1: FEIDilATER CONTROL SYSTEM SETTINGS JIP i j Feedwater Pump Speed Controller Settings A. I Settings At i Plateso Power 30% 75% l Slave Controller 509B/509C* 509B/509C* s Delta P Setpoint 1.08/1.1 1. .,08/1.1
; Ge.in
! Lag Time Constant 0/0 0/0 (heconds)
Slave controllers; 509B is for feedwater pump turbine A, 509C is for feedwater pump turbine B, i hkh' Delta P Controller 2 Settings At Plateau Power _,
; 30% 75%
Gain 0.4 0.4 1 Reset Time 100 100 Constant (Seconds)
^- ..
7-67
7.4.2 LOAD SWING TEST (1-6SC-02) (Continued): I TABLE 7.4.2.1: FEEDWATER CONTROL SYSTT)1 SETTINGS (Continued) Main Feedwater Valve Controller < (Settings Were The Same For All Steam Generators) i Settings At Flateau Power 30% 50% 75%* Level Gain 2.0 2.0 2.2 Level Reset 830 830 330 Time Constant (Seconds) Flow Mismatch 0.32 0.32 0.5' Gain
~
Flow Mismatch 500 500 500 Reset Time Constant (Seconds) These settings were made after the 75% test 4
- was completed.
l i l l 1 t
) 7-68
7.4.3 AUTOMATIC REACTOR CONTROL __(1-6SF-01) Objective The objective of the Automatic Reactor Control Test was to demonstrate the l capability of the reactor control system to respond to input signals. l The abstract for this test is FSAR Section 14.2.8.2.31. l i l Methodology This test demonstrated the capability of the Reactor Control System to respond and maintain reactor coolant average temperature (Tavg) within steady state l limits of 11.5*F of the programmed reference signal (Tref). Tref is computed from main turbine first stage pressure. The Reactor Control System positions ] control rods to maintain temperature. Both sections of the test started with the plant stable at 30% power and control bank D nominally between 114 and 150 steps. Tavg was adjusted both higher and then lower than Tref. After each adjustment the Reactor Control ; (rod motion control) was chan8ed from Manual to Automatic. Plant parameters l were recorded as the automatic control responded and restored Tavg to within 11.0 F of Tref and was able to maintain a band of 1.5*F. In the first part of the test, rod control was placed in Manual and Bank D was D withdrawn from 138 steps to 155 steps. Tavg was increased from 565*F to 572 F. When control was returned to Automatic, Tavg returned to withir. 11.0'F of Tref i l and was maintained within 1.5 F. See Figures 7.4.3.1 and 7.4.3.2 for results. In the second part, rod control was again placed in Manual and Bank D was inserted to 143 steps. Tavg was lowered to 557 F. When control was returned to Automatic, Tavg returned to within 1.0*F of Tref and was maintained within 1.5*F. See Figures 7.4.3.3 and 7.4.3.4 for results. Pressurizer pressure was monitored during the transients. As . expected, pressure followed Tavg and only increased above the initial value by 20 psi during the transient. This was well below the criteria of 50 psi. Results This test met the acceptance criteria and objective. No control settings were changed based on the performance of this test. 7-69
7.4.3 AUTOMATIC REACTOR CONTROL (1-6SF-01) (Continued): Problems During the initial performance of this test the system did not respond : properly. Tavg continued to decrease when control was placed in automatic , after the +6*F mismatch. The rod speed demand did not decrease to zero as ! Tavg approached Tref. This resulted in the undershoot. Investigations revealed a bad relay card and an amplifier gain out of adjustment in the 7300 control system cabinets. When these were corrected the test was completed and the system responded satisfactorily. j D , l i 1 I l i 7-70 1
- 1
_.__________j
7.4.3 AUTOMATIC REACTOR CONTROL (1-6SF-01) (Continued): FIGURE 7.4.3.1: +6*F TREF MISMATCH REFERENCE TEMPERATURE AND AUCTIONEERED T AVERAGE AUTOMIIC RFACf0R CONTROL 1-6SF-91 TOR +6F
$74 6 AUCTIONEERED HIGl T AYERACE 573 - a 1 I REFERENCE (fRIF) 3 572 ,A 571 A
< 3 570 \
,, \
5., 1
, T 562 A u 1 E \
f566 e k; E l56541IIAii1441111111_.LA1111111111111111" j
!,56
\44&aa4 g &
0 E 563
-1 h I $ $ k $ d y g:
flME IN MINUTES 7-71
7.4.3 AlTIT)MATIC REACTOR CONTROL (1-6SF-01) (Continued): FIGURE 7.4.3.2: +6*F TREF MISMATCH PRESSURIZER PRESSURE AND CONTROL BANK D POSITION Auf0MA11C REACTOR CONTROL 1-6SF-91 FOR +6F 2269 160 A PRESSURIZER PRESSURE PSIC 2249. [ A
,AA' 155 2239. A
& 350 h229. [
fa k ge p219. g/ l
- & 1 45 ,
299. l 31,0. g
\g "::::::: c' 1 49 "
E
, s o E G g
$189. 5 nnnnnnnn a e
! w " " " " " " " " " " O c0NTROL BAMX D STEPS POSIfl0H li! K Q179 . . 135$
-1 h I h h k h d h 1; flME IN MlHUTES 7-72
( 7.4.3 AUTOMATIC REACTtR CONTROL (1-6SF-01)-(Continued): FIGURE 7.4.3.3: -6'F TREF MISMATCH REFERENCE TEMPERATURE AND AUCTIONEERED T AVERAGE : Auf0MATIC REACTOR CONIROL 1-6SF-91 FOR -6 F l 568 j 3 4 RCS AUCTIONEERS HIGH f AUG 567 i T REF q 566 *^^ >-
~ 'i. g
,,, axa u u u i u . u a u . . . . . u u u s u u 564 563 L
~
O ~ j 4 e 561 5 /
;56. 4 1 p5, /
w 8
/
*x 558A - -
b E 557
-1 h } $ $ k $ d h I flME IN MINUIES MIRING TRANSIENT t
7-73 l
l 1 i 7.4.3 AUTOMATIC REACTOR CONTROL (1-6SF-01) (Continued): I FIGURE 7.4.3.4: -6*F TREF MISMATCH I l PRESSURIZER PRESSURE AND CONTROL BANK D POSITION i Auf0MAfl0 REACTOR CONTROL 1-6ST-01 FOR -6 F 2270 t80 1 0 PRESSURIZER PRESSURE ) i 2265. j 175 2260. 3 2255. < N 170 6 'n s 2250.
/ s i 3
OCCC0 f245 3 .....
;2240.
A 160 W [g*"""% \ D l2235. ; ; = = H-H " " " " " " " " : : " " " ll l lh2230. 155 $ 1 I CONTROL BAHX D STEPS POSIfl0H % K 8 g2225. 150 $ W t; N 2220. w i g i
! n... 145 $ ]
$2159==== a i W E i a e )
12210 140 8 1 l
-1 h I d d k $ d h d TIME IN MlHUIES i
7-74
'7.4.4 _ AUTO STEAM GENERATOR LEVEL CONTROL __(1-6AE-01)
Objectives The objectives of the Automatic Steam Generator Level Control Test are to: (1) verify the stability.of the automatic steam generator-level control system following simulated transients at low-power and high power conditions; and verify operation of the variable speed. feature of the feedwater pumps. The abstract for this test is FSAR Section 14.2.8.2.25. Methodology The test is performed in separate low and high power sections whose purposes are to: (1) verify the stability of the steam generator level control system following simulated transients at low power levels-(1-15%), and at the high power levels of 30%, 50%, 75% and 100%; (2) verify stable operation of the variable speed control system of the main feedwater pumps; ~ and (3) verify the stability of the steam generator level control system when transferring control from the bypass feedwater regulating valves to the main feedwater regulating valves. Control system adjustments were made dur1ng this test. Low Power Testing Between 1 and 15% reactor power, steam generator level undershoot and overshoot were verified to be less than four percent narrow range level while l on the Bypass Feed Regulating Valves (BFRVs). Level stability was observed when transferring from the BFRVs to the Main Feed Regulating Valves (MFRVs). Initially steam generator level was raised 5% above the normal level _setpoint of 50% narrow range level. One at a time the BFRV's were placed in automatic control. The response of the BFRV control system was monitored to verify that steam generator level undershoot was less than four percent beyond the i setpoint as the system attempted to decrease level. Steam generator level ( returned to and remained within 12% of setpoint within 30 minutes (three time l constants of 600 seconds reset time). A similar transient was performed to ensure steam generator level overshoot was less than four percent. Level was first lowered, then the controller was placed in automatic. The initial test of the BFRV's failed. The valves were dynamically tuned by lowering the gain and raising the reset time constant. A satisfactory retest was performed. Control system settings are listed in Table 7.4.4.1. Reactor power was raised to 15% to prepare for transfer from-the BFRV's to the MFRV's. This was accomplished by placing the MFRV's in automatic and slowly closing each BFRV in manual control while ensuring the MFRV's opened. Steam generator level response was verified to ensure the level returned to and remained at 50 2% within three reset time constants of 600 seconds. s 7-75
l 7.4.4'- ' AUTO STEAM GENERATOR LEVEL CONTROL'(1-6AE-01)__(Continued): i During the transfer of control from the BFRV's to the MFRV's, the methodology D i was eventually revised based on operating experience. originally with the BFRV's in manual and the MFRV's in automatic. The' transfer was It was found that better control could be achieved with the BFRV's in aut'omatic end the MFRV's in manual during the transfer. The transfer is performed by slowly opening the MFRV's (one at a time) and allowing the BFRV's to slowly close by reducing their demand. Fewer oscillations occur when using this method. 1 (The most recent methodology, developed after the test program with help from Westinghouse, is to place both valves in manual and make the transfer while maintaining constant feedwater flow.) I 30% Power Testing. This portion of the test verified automatic operation and stability of the ] Main Feedwater Pump Turbines (MFPTs) and the overshoot and undershoot of the steam generator level control system. Observation of both control systems was performed while step load changes of +10% and - 10% were made in the . Turbine g Control System. j L After placin8 the MFFT main controller in automatic, each individual I controller was placed in automatic, and main feedwater pump (MFP) discharge ) < pressure and MFRV stability observed for five minutes. _A one minute average l of pump discharge pressure and steam header pressure was obtained from the plant computers and actual differential pressure-(delta-p) computed. This-was compared to the delta-p setpoint computation and verified to be within 13% by l the following formula: Delta-p Setpoint (psid) =
-5
! psid/lbm/hr + 45 psid
=
Total Steam Flow (lbm/hr) x 10-Observation of steam generator level overshoot and undershoot was performed similarly to 15% power testing except the MFRV's were used. Step load changes of +10% and -10% were imposed on the MFPT and steam l generator level control systems while concurrently performing 6SC-02, " Load l Swing Test". This was to verify steam generator level variations due to the i load swing were within 110%, that changes in steam generator level, steam l ' flow, or feedwater control valve position or flow did not induce a diverging level oscillation, and any induced oscillations were within 12%, MFP discharge pressure within 139 psi within 3 delta-p controller reset time constants and, that the MFRV positions stabilize within the maximum and minimum values of the valve lift curve. 1 7-76 1 l'
l i 7.4.4 l AUTO STEAM GENERATOR LEVEL CONTROL (1-6 E-01)_(Continued): ' 50% Power Testing , D- The 50% Power Testing for this test verified stability of the Stt w Generator Level and MFPT Control Systems upon reaching the plateau by comparing the plant computer data.with installed instrumentation to ensure comparability within 11%. Steam generator level was verified to be maintained at 50 12%,. MFP discharge pressure oscillations ( 39 psi and MFRV positions within the maximum and minimum limits'of the valve lift curve. 75% Power Testing i Control system stability was verified similar to the'50% plateau'by comparing l parameters displayed by the plant computers with installed instrumentation. j Step load changes similar to the.30% plateau of +10% and -10% were also imposed on the control systems. Finally,- a 50% load reduction was performed l t concurrently with 1-700-01, "Large. Load Reduction". During the load reduction l steam generator level was verified not to vary more than 225% from initial j level, steam flow or feedwater control valve position or flow changes did not ; induce a diverging level oscillation, and any induced oscillations were within i 2%, MFP discharge pressure within 39 psi within 3 delta-p controller reset i time constants and that the MFRV positions stabilize within the maximum and-minimum values of the valve lift curve. i 100% Power Testing ) D Full power testing full power has not been completed. However, the plant has operated successfully and stably operated at full power. The results of testing will be reported in a later supplement. l Results The main feedwater control system has been demonstrated stable and capable of controlling steam generator water levels up to full power. 9 m l l
7.4.4 AUTO STEAM GENERATOR LEVEL CONTROL (1-6AE-01) (Continued): Problems Initial tuning of the BFRV's and the MFRV's was required to improve level ; control and response of the system. Table 7.4.4.2 lists the controller i settings. After the 10% Load Swing at 75% power, the MFRV controller settings were changed for a quicker valve response by increasing level and flow compensation gain settings. : The MFRV position appeared to be a problem when it was determined that the valves were opening too much for the given steam flow conditions. Subsequent investigation revealed that the strip chart recorder used to monitor the l lanyard type transducers attached to the MFRVs for position indication had an j l improper gain setting. Once this was corrected, valve positions were shown to ! be within their maximum and minimum limits based on steam flow. i 1 E l l l t (o) l l ! l l 7-78 i
7.4.4 AUTO STEAM GENERATOR LEVEL CONTROL (1-6AE-01) (Continued): I TABLE 7.4.4.1: CONTROL SYSTEM SETTINGS ! FEEDWATER BYPASS VALVE CONTROLLER i l
)
FEEDWATER BYPASS VALVE CONTROLLER PERCENT POWER l STEAM GAIN (%/%) RESET TIME CONSTANT (SEC.) i GENERATOR 1-15% 15-20% 1-15% 15-20% i 1 1.3 1.3 620 620 , 2 1.3 1.3 620 620 1 3 1.2 1.2 620 620 4 1.3 1.3 620 620 1 1 l l l N l l O 7-79 '
7.4.4 f,yTO STEAM GENERATOR LEVEL CONTROL (1-6AE-01) (Continued): O TABLE 7.4.4.2: MAIN FEEDWATEF CONTROL SYSTEM SETTINGS l V FEEDWATER PUMP SPEED CONTROLLER Percent Power Parameter 30% 75% 100% Delta-P Setpoint 1 1 (Later) l Controller (Slave) Gain (%/%) Lag Time Constant 200 200 (Later) (Sec) Delta-P Master 0.382 0.382 (Later) Controller (Master) Gain (%/%) Reset Time 100 100 (Later) Constant (Sec) i O MAIN FEEDWATER VALVE CONTROLLER (One Value For All Steam Generators.) Percent Power I Parameter 15% 30% 75%* 100%
~
k Level Compensation Gain (%/%) 2 2 2.3 (Later) 1 i Level Compensation Reset 830 830 830 (Later) j Time Constant (Sec) l Flow Compensation Gain (%/%) 0.32 0.32 0.5 (Later) , Flow Compensation Reset 500 500 500 (Later) Time Constant (Sec)
- 1 These setpoints were made after the 75% test had been performed. l t
1 N 7-80
7.4.5 DYNAMIC AUTOMATIC STEAM DUMP CONTROL (1-6AB-01)
,_ Objective
(
\
The objectives of this test were to: (1) verify automatic operation of the Taverage (Tavg) steam dump control system; (2) demonstrate controller setpoint adequacy; and (3) obtain final settings for steam pressure control of the condenser dump valves. The abstract for this test is FSAR Section 14.2.8.2.24. Methodology This is one of the first tests as reactor power is initially raised to low levels following low power physics testing. The steam dump system consists of turbine bypass control valves and controllers. Both modes of control, Tavg and steam pressure, were tested. In the Tavg mode, an RCS average temperature signal controls the turbine bypass valves. In steam pressure control, pressure in the main steam header is controlling. The Plant Trip Controller was first verified to respond properly by simulating a reactor trip to the steam dump control system with the steam dump controller in steam pressure mode and the plant at hot standby of 557 F Tavg. Reactor power was then increased to approximately cne percent, causing Tavg to increase slightly. When Tavg reached 560 F, steam dump control was switched to
'~ the Tavg mode. resulting in the steam dumps opening to control Tavg at 558 +
1 F. Reactor power was further increased to approximatley four percent and ~ Tavg verified to increase to and stabilize at 562 + 2 F. Reactor power was then decreased to approximately two percent and tee steam dump control was placed in the steam pressure mode and the simulated reactor trip signal was removed. A sudden loss of load was simulated and a Tref (Treference, from turbine first stage pressure) signal of 553*F was injected into the steam dump control system. The steam dump control mode selector switch was placed to Reset, then to the Tavg mode, and plant response was monitored for proper control stability at setpoint. The sudden loss of load and Tref signal of 553 F were i removed, steam dump control was changed to steam pressure mode, and reactor power was increased to approximately four percent. Plant response was : monitored to verify stable response and proper setpoint control. The final steam dump controller settings are shown in Table 7.4.5.1.
) !g 7-81
7.4.5 DYNAMIC AUTOMATIC STEAM DUMP CONTROL (1-6AB,-01) (Continued): Results D All acceptance criteria and objectives were met. The automatic steam dump control system responded to a simulated plant trip sudden loss of load, and reactor power chan8es to control steam pressure and Tavg as designed. Plant stability and proper setpoint control were verified to be within acceptable tolerances. Final gain settings were recorded. 1 Problems l There were no major problems. Due to the need to verify proper steam dump control prior to increasing reactor power above five percent, the procedure was changed to use auxiliary feedwater instead of main feedwater. At the time, the main feedwater system was unstable at this low power level and resulted in a reactor trip on low I steam generator water hvel. Another procedure problem was encountered during performance of the loss of load controller response. The procedure specified Tref signal with reverse polarity and caused the steam dump valves to briefly fail open. Immediate operator response restored the plant. The procedure was corrected and a retest successfully conducted. i n [ 7-82
7.4.5 DYNAMIC AUTOMATIC STEMi DUMP CONTROL (1-6AB-01) (Continued): TABLE 7.4,5.1: STEAM DUMP CONTROLLER SETTINGS PARAMETER VALUE-Tavg Amplifier Gain (TC-500D) 3.21 Load Rejection Amplifier Gain (TC-500A) 7.8 Steam Header Controller Gain (PC-507) 5 Steam Header Controller Time Constant (PC-507) 17 Seconds I i i l D l I L 7-83 u 1
i D l l l 1 SECTION 7,5 l i TRANSIENT TESTS 1 J l 1 i l l l
7.5.1 PLANT TRIP FROM 100% g R (1-700-02) Objective. Objectives of the Plant Trip From 100 Percent Power Test were to:- (1) verify ' the ability of the plant automatic control systems to sustain a trip from 100-percent power and to bring the plant to stable . conditions following the transient; (2) determine the overall response time of the hot leg RTD; and (3) optimize the control systems setpoints, if necessary. The abstract for this test is FSAR 14.2.8.2.53. Methodology The reactor trip is initiated by opening the main generator breakers. Data was recorded on the primary system and the main steam system during the transient and again when steady state conditions had been established. The acceptance criteria for this test are:
- 1. Pressurizer Safety Valves did'not lift. !
- 2. Main Steam Safety Valves did not lift.
- 3. Safety Injection was not initiated.
- 4. All contro1~and shutdown rods released an'd dropped.
- 5. The plant is stabilized in Hot Standby conditions. 4
- 6. Nuclear Flux drops to less than 15% within 2 seconds. '
- 7. The RTD response time in the RCS is less than 6.8 seconds. !
Data collected during two unplanned trips were used'to satisfy acceptance I criteria. Post trip logs from both the Proteus computer and the Emergency j Response Facility (ERF) computer were used to collect the data required to satisfy the acceptance criteria. The first trip occured on June 3, 1987 with the plant at 100% power. A lightning strike resulted in a trip of the main generator which produced a reactor trip. A generator trip which was originally required by the test. This trip satisfied acceptance criteria except for RTD response time and a portion of the dynamic effects test. The post trip logs were collected and used to complete the pretrip and posttrip data requirements. The ERF computer has a rolling 2 hour history and i automatically saves it in the event of a trip. Review of the data j
- demonstrated that all acceptance criteria were met except for the RTD response i l
time. This criteria required information from the narrow. range Thot and Tcold, which is not normally recorded. Three important parameters are shown in Figures 7.5.1.1 through 7.5.1.3. Flux response ~ time was extrapolated from data taken at 2 second intervals. 7-85
7.5.1 PLANT TRIP FROM 100% POWER (1-700-02) (Continued): p The ERP computer has the capability to monitor additional parameters through a Q special patchboard (called t.be ERF jack panel). Narrow range RCS temperatures and nuclear flux were connected through temporary cables to the Jack Panel. These points were monitored along with the rolling history so they would be saved in the event of another unplanned trip. l The second trip occured from 92% power on June 23, 1987. Again the trip was due to a main generator trip. The ERF computer collected the RTD response time ,! data. Since the plant was only at 92% power instead of 100%, the data was sent to Westinghouse for analysis to ensure that the acceptance criterion on RTD response time was met. Figures 7.5.1.4 and 7.5.1.5 show flux decay and temperature response, respectively, that satisfied the acceptance criterion. I Results The test met all acceptance criteria. The RTD response time was measured to be 6.3 seconds. See Figures 7.5.1.1 through 7.5.1.5 for plant parameters during the transients. I Problems O v The plant experiences cooldowns if manual operator action is not taken shortly i j after the trip to shut several two inch main steam line drains that open'with i a turbine trip. A solution is under engineering evaluation. The cooldown also briefly causes low pressurizer level and pressure although they recover after the cooldown is terminated (see Figures 7.5.1.2 and 7.5.1.3). l The dynamic response of the plant piping systems has not been measured because the monitoring equipment was not prepared for the unanticipated trips. The ' l 1 dynamic response will be measured and reported in a supplement to this report. 7-86
i 7.5.1 _ PLANT TRIP FROM 100% PO'#ER (1-700-02) (Continued): l t h FIGURE 7.5.1.1: 100% TRIP - NEUTRON FLUX l l l
'~
MDITRON FLilX POWER RANCE (CH 1) 6/3/87 tes = A PERCENT FMX 9e., _ 8e l 7e ; l l ! l se ! D
- e 58 l r
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^ ^ A A 4 A A A A 4 A A A 3 l
-be d s's ide! IIe 2de 2$8 3h 3de 4de 4Is 5de 5$8 6esi TIME IN SECONDS FROM TRIP q
) 7-87 c-_-------------------_-----_--
7.5.1 PLANT TRIP FROM 100% POWER (1-700-02) (Continued): FIGURE 7.5.1.2: 100% TRIP - PRESSURIZER PRES 5tTRE 4 U0CfLE 1 100% POWER TRIP 6/3/87 223e M A PRES $URIZER PRESSURE 2200 E aL 2150 ^ n / t2100 t [ { 050
/ ~
h A AA 4 000 .1A [ 41950
-10 h 5'8 1h8 1$9 2h8 2$9 3h0 350 4h8 4$9 She 5$0 600 ,
flME IN SECONDS FROM IRIP 7-88
7.5.1 PLANT TRIP FROM 100% PO!IER (1-700-02) (Continued): l FIGURE 7.5.1.3: 100% TRIP - PRESSURIZER LEVEL I l l , i U0CTLE 1 100% TRIP 6/3/87 70 A PRESSURIZER LIUEL 65 l i I 60 n 4 55 50 d i y45 ' W u 3f O V W L l 35 'r 4 W D 30 25 Ea -
^
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= 20 ^^^**^AAaaa
. . . . a l z_ - = ;. a - ..
3 l I
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-00 h 5'8 1ho 150 2de 250 3h8 3$9 4h 450 Eb6 550 6061 I l
flME IM SECONDS TROM TRIP 1 4 7-89
7.5.1 PLANT TRIP FROM 100% F0WER (1-700-02) (Continued)J FIGURE 7.5.1.4: 100% TRIP - NUCLEAR FLUX AND DELTA T POWER, LOOP 1 i U0GTLE I TRIP 0F 6/23/f 7 FROM 92X (lND.) POWER I 199 un u unu in uni .i,, 8 MUCLEAR FLUX 99M N A. BELTA T POWER LOOP 1 89. 79 4 .__ 1 i ! 69 H 59 D Z W l 0 g g49 f 39
' i i, W __,,,'">,,,
g 29 1,.'> . s 0 9 19 l { MY^*^V">@iWWitMt#&AVA4AVA%MA4Ay#f#Av.yAVAyg j
-1 h f d 5 k $ d h d h l'8l'1l'2l'3l'4lb1'61h1'8l'92ll flME IN SECONDS 7-90
J ! 7.5.1 PLANT TRIP FROM 100% FOWER (1-700-02) (Continued): i 3 FIGURE 7.5.1.5: 100% TRIP - LOOP 1 RCS TEMPERATURES l l I i i
]
i l U0CTLE I fRIP 0F 6/23/87 TROM 92x (IND.) POWER 628 1 O I HOT LOOP 1
$ f COLD LOOP 1 610 -
i 690 l l 590 . h ! A,, l 5'a i a l 9 E 550
-1 h d d d d d d d d d I'0l'1l'2l'3l'4lb1'61hl'8l'9261 flME IN SECONDS l
1 h 7-91 l
7.5.2 NATURAL CIRCULATION DEMONSTRATION 0-600-10) t Objectives Objectives of the Natural' Circulation Demonstration Test were: (1) demonstrating the capability to remove decay heat by natural circulation; and (2) demonstrating that primary parameters can be controlled within acceptable limits. The test also provided information for the Vogtle training simulator and experience for reactor operators. The abstract for this test is FSAR Section 14.2.8.2.47. The test is also addressed in FSAR Questions 440.45, 440.98, and 640.04 Methodology l The test was conducted June 3, 1987, following a reactor trip from full power. The test was intended to be performed after a planned or unplanned trip. It occured enti:aly under hot standby conditions with normal offsite and onsite power available and consisted of deenergizing the reactor coolant pumps from hot standby conditions, observing establishment of natural circulation, measuring natural circulation flow and flow distribution, and restarting the reactor coolant pumps. Natural circulation flow was measured by stopping all feedvater and measuring the rate of steam generator level decrease by a "steamdown" method. The test's sequence of events is listed in Table 7.5.2.1. , t The plant had tripped but had operated at full power long enough to provide sufficient decay heat. Test preparations were made as soon as the cause of the trip had been identified. Data-gathering computers were started and operators were dispatched to the #4 atmospheric relief valve (ARV), which had not been responding to commands from the control room to open or close (stroke). The l test required all four ARVs to be operating for a uniform rate of decrease in i steam generator water levels. l Local operation of the #4 ARV was successfully demonstrated and additional licensed operators who were to observe the test were gathered in the control room. The reactor coolant pumps were deenergized to initiate the transition from forced to natural circulation. Natural circulation was declared stable after sixteen minutes, and a half hour stability demonstration period began. Steam generator blowdown and sampling had earlier been isolated and auxiliary feedwater was stopped. Steam generator water levels decreased at a rate of l approximately 34% (narrow range level) per hour during the "steamdown" portion. The rate of steam generator water level decrease in a saturated isothermal steam generator measures heat being removed by steaming. All other heat sinks have been measured. RCS insulation heat loss had been measured during an earlier test and charging and letdown, reactor coolant pump seal cooling, etc., was measured during the test. The sum of all heat sinks was balanced by core decay heat. 7-92
1 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): _ Reactor coolant system flow was determined by a variation of a simple primary o coolant system calorimetric. Hot and cold leg fluid enthalpies were determined I I from wide range RCS temperatures and pressures. Core heat divided by the l enthalpy difference produced total reactor flow, i Recovery involved restoring blowdown and auxiliary feedwater to the steam generators, measuring blowdown temperature (by procedure) to verify that no large temperature difference existed between steam generators and the reactor i coolant system, and restarting reactor coolant pumps. i Results All objectives and acceptance criteria were met. i Core cooling with natural circulation was shown to be easily controlled using auxiliary feedwater and atmospheric relief valves in either remote automatic or local manual control.
)
i The transition from forced to natural circulation was smooth, as shown by the absence of large temperature or pressure excursions. RCS flow appeared well mixed in the reactor vessel and there were no flow-starved areas as shown by nearly uniform temperatures at core exit ) thermocouple above the core (see Table 7.5.2.2). J
/3 As expected, the narrow range temperature indicators were shown to be j Q unreliable indicators of natural circulation flow or changes in flow, '
verifying training that wide range indicators must be used under natural circulation conditions. Manual operation of the atmospheric relief valves was shown to be an effective and managable method of controlling steam generator pressure and RCS temperature during natural circulation. Capability To Remove Decay Heat The capability to remove decay heat was demonstrated by the plant's ability to attain and sustain natural circulation. The plant was shown to be very stable during natural circulation flow. Figures discussed in Table 7.5.2.3 show plant parameters as natural circulation was established and maintained, and illustrate the plant's stability. Core decay power of 23.2 Mwt (0.68 percent of full power) was measured and produced a reactor coolant system flow of 3.67E6 pounds per hour (2.41% of RCS flow as measured at full power) for a power-to-flow ratio of 0.282. A temperature rise of 16.85 F was measured by thc. wide range hot and cold leg RTDs; this correponds to 28.2 percent of full-power core differential temperature. l s 7-93
M 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued; , Uniform Core Flow Core exit thermocouple showed a essentially uniform temperature, indicating no flow-starved areas in the core. This uniform distribution persists from the time reactor coolant pumps were deenergized, as shown in Table 7.5.2.2. Wide range hot leg tv.perature sensors also indicated that flow in the reactor b vessel was well-mixed. Loop number four's cold leg temperature varied considerably due to manual operation of one atmospheric relief valve, and a much smaller but uniform variation in temperature was detected by all four wide range hot leg temperature sensors. Uniform response indicated considerable mixing between the cold and hot leg temperature sensors. Since the core exit thermocouple showed uniform temperature, mixing had to occur in the core itself, the lower plenum, or the downcomer. Narrow Range Temperature Instruments Not Designed For Natural Circulation Flow Conditions The narrow range temperature instruments, in their bypass manifolds, were verified not to be accurate indicators of temperature, magnitude of core } cooling, and changes in core cooling. This result was expected considering a they are not designed for natural circulation flow conditions. Table 7.5.2.3 J and its graphs illustrates their behavior. Information For Benchmarking The Simulator The test was also intended to gather information for benchmarking the Vogtle training simulator. Information gathered included: steam generator pressures, levels, and auxiliary feedwater flow rates; reactor coolant system hot and cold leg wide and narrow range temperatures; pressurizer pressure and level; and chemical and volume control (CVCS) and reactor coolant pump seal cooling flows and temperatures. F Reactor Operator Training The test provided training for reactor operators. A total of twenty one operators observed or participated in the test. The test resembled simulated loss of offsite power conditions conducted on earlier plants where loss of offsite power was sometices simulatid by closing the main steam isolation valves, using only the atmospheric relief valves and auxiliary feedwater, and deenergizing the reactor coolant pumps. Test plans provided the option to use turbine bypass valves, but at test time the main feedwater isolation, main steam isolation and bypass valves had been closed. 7-94
.4
l 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): Loop Transit Time Loop transit time for forced flow is about 10.2 seconds. Calculated loop transit time for the test's power level and corresponding natural circulation flow was 10.2 seconds (full flow loop transit time)/0.0241 (flow fraction ! under natural circulation at the test's power level), or 477 seconds (7.9 minutes). i l The longer natural circulation times are clearly shown by the sharp decreases l in steam generator #4's pressure, are reflected some time later in Tcold, and still later in hot leg temperature. Problems There were three minor problems. First, the #4 ARV required manual operation. It was recognized prior to deenergizing the reactor coolant pumps that the test required all four atmospheric relief valves to be opccating if it was to be conducted as pl.3nned at hot standby temperatures; otherwise during natural circulation, stearn pressure in the #4 steam generator pressure would rise to safety valve setpoints and uniform steamdown would be impossible. A team of operators was dispatched to the number four atmospheric relief valve and ; communications established via sound powered phones between operators in the { control ro'om and at the-valve. Manual operation using the syotem operating j procedure fer local valve o;cration was verified prior to deenergizing the l
. reactor coolant pumps. Throughout the test, the control room operator directed pperators at the valve.
l l Manual operation worked well. The valve is hydraulically operated and approximately eighty strokes of the valve's manual hand pump were required to initially move the valve but thereafter a few strokes would move it in either , direction. Graphs of steam generator number four's pressure and level show that cor. trol was not as fine as automatic, but it was acceptable for this tent, which required fairly stable control. Manual control can certainly be expefted to be adequate in less demanding situations. [The second mitor problem was determining steam guerator blowdown temperature, which delayed restarting the reactor coolant pumps. By operating procedure, steam generator blowdown temperature must sufficiently match reactor coolant
)/ ,,, temperatures to prevent overpressurization when a reactor coolant pump is restarted. The preferred instrumentation was not operating so alternates were j used. ,
The third minor problem das that steam generator __ levels were established higher than the normal control band to enrare r.9fficient water inventory for the steamdown portion of the test. The test wocedure allowed narrow range water levels to be' as high as 70%. However, at 70%, operation of an atmospheric relief valve . created sufficient. swell to actuate automatic auxiliary feedwater '$9 . atior~ at 75% level. Levels were reestablished at approxirnptely 60% and'no j'prther high level probi.ews occured. 7-95 ! l ( 0 f' y Jt ._-- ~
t 7.5.2~ _ NATURAL CIRCULATION DEMONSTR ATION (1-600-10) (Continued) TABLE 7.5.2.1 SEQUENCE OF. EVENTS FOR V0GTLE UNIT 1 NATURAL CIRCULATION TEST CDT Test
- Clock _ Time Elapsed Time Event Discussion _
16:05 N/A Plant Tripped Plant tripped.from full power. . Main steam isolation valves I were shut shortly afterward. 18:08:42 N/A Blowdown Isolated Steam generator blowdown was isolated, i 19:16:42 00:00 Reactor Coolant All four RCPs were tripped d (19.28)*** Pumps Manually within 0.21 seconds of each Tripped other. , 19:25:35 08:53 Auxiliary Feed . Feedwater'to Steam Generator l (19.43) to S/G #3 Stopped #3 wt stopped. 19:32 16:00** Steady State Steady state natural circulation t (19.53) Natural was declared by shift ] Circulation supervisor. 19:36:53 20:11 Auxiliary Spray Spray was initiated to reduce i (19.62) Initiated pressurizer pressure. 19:41:34 24:52 Auxiliary Spray Spray valve fully closed. (19.69) Completely Stopped I 19:49:53 33:11 Auxiliary Feed Auxiliary feedwater to steam (19.83) to S/G #2 stopped generator #2 was stopped. 19:57:13 40:31 Auxiliary Spray Auxiliary spray valve was (19.95) valve opened opened. 20:02:00 45:20 Auxiliary Feed Auxiliary feedwater to steam (20.03) to S/G #1 stopped generator #1 was stopped. 20:02:05 45:25 Auxiliary Feed Auxiliary feedwater to steam I (20.03) to S/G #4 stopped generator #4 was stopped. Steam generator steamdown and natural circulation RCS flow rate and core power measure- j ment started. ' 20:36:59 80:17 Auxiliary Spray Auxiliary spray completely ; (20.62) valve was closed stopped. ! 7-96 l w_-______-_____-__-___-____-____-___________
I i . I ! 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): ! l h TABLE 7.5.2.1 SEQUENCE OF EVENTS FOR V0GTLE UNIT 1 NATURAL CIRCULATION TEST l i (Continued) i l CDT Test I Clock Time Elapsed Time Event Discussion I 20:4G:04 91:22 Auxiliary Feed Auxiliary feedwater to steam (20.80) to S/G #4 started generator #4 was started. The natural circulation measure-ment stopped when feedwater was reinitiated. 20:48:17 91:35 Auxiliary Feed Auxiliary feedwater to r?.eam (20.81) to S/G #3 started generator #3 was started. 20:48:30 91:48 Auxiliary Feed Auxiliary feedwater to steam (20.81) to S/G #2 started generator #2 was started. l l 20:48:46 92:04 Auxiliary Feed Auxiliary feedwater to steam l (20.81) to S/G #1 started generator #1 was started. 21:04 D 108 Blowdown Restored Blowdown and sampling restored to all steam generators. Steam generator temperature, measured by steam generator blowdown temperature was prerequisite to restarting the reactor coolant j pumps. 21:56:23 159:41 RCP #3 Restarted The first reactor coolant pump was restarted. 22:00:08 163:26 RCP #4 Restarted Second reactor coolant pump was restarted; test and special l test exception was terminated.
- Elapsed time is minutes: seconds.
Control room clock time was used, so this is approximate time.
*** Numbers in parentheses ( ) beneath clock time are fractions of hours to facilitate cro.m correlation with the Figures.
Each 0.1 hour is 6 minutes. l l l 7-97
._ - a
4 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): TABLE 7.5.2.2: CORE EXIT TEMPERATURES FROM SELECTED THERMOCOUPLE DURING POWER OPERATION AND NATURAL CIRCULATION
.O F TEMPERATURE CORE PRE- 2 7 12 17 LOCATION TRIP MINUTES
- MINUTES
- MINUTES * -MINUTES
- A60 602.2 571.00 572.57 573.00 572.32 C10 602.6 571.20 572.60 573.37 572.60 C04 617.5 1570.50 572.00 572.57 571.98 C06 625.5- 573.68 574.90 575.75 574.95 C08 622.2 568.04 569.54 570.27 569.72 C10 623.6 571.52 573.20 573.57 572.98 C12 615.6 570.76 572.19 572.55 572.12 C14 592.0 571.41 573.40 573.16 572.13 E02 620.8 573.22 574.65 575.36 574.69 E04 622.4 568.21- 569.90 570.38 569.63 E10 628.0 573.89 575.08 575.89 575.44 i E12 619.5 570.09 571.40 572.12 571.58 ;
E14 619.8 574.86- 576.40 576.85 576.07 G02 622.0 570.73 572.80 573.73- 572.73 G06 625.3 572.49 574.50 -574.73 574.49 G08 620.5 572.50 .573.86 574.34 573.98 G10 622.5 573.37 575.00 575.51 575.28 G12 628.1 574.04 575.35 575.96 575.38 G14 619.5 571.18 572.50 573.25 572.40 H01 605.8 571.52 572.91 573.48 572.80
'O- i HIS J02 604.7 621.2 569.36 570.82 571.11 .570.30
.571.28 572.28 573.25 572.70-J04 628.3 573.80 575.61 575.91 575.30 J06 625.8 573.16 574.70 575.13 574.70 i J10 622.6 571.44 573.27 573.69 573.25 J12 628.4 575.77 577.49 578.04 577.40 LO2 615.9 567.88 569.61 569.87 569.40 LO4 621.4 570.84 572.52 573.17 572.61 LO8 629.1 573.60 575.19 575.83 575.25 L10 628.1 573.31 574.83 575.32 575.70 L12 622.0 570.81 572.30 572.60~ 571.95 L14 618.5 571.52 572.95 573.26 572.80 N06 624.9 573.20 574.56 575.40 574.68 N14 589.4 568.37 569.92 570.32 569.60 R06 602.5 571.83 573.43 574.11 573.40 R08 606.3 571.72 573.68 574.58 573.93-R10 602.8 571.10 572.75 573.34 572.90 Range: Maximum 629.1 575.77 577.49 578.04 577.40-Minimum 589.4 567.88 569.54 570.27 569.40 Average 617.49 571.71 573.26 573.80 573.22 Std. Dev. 10.49 1.89 1.88 1.91 1.97 Minutes following trip of all four reactor coolant pumps.
7-98 L
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): i l l TABLE 7.5.2.3: DISCUSSION OF FIGURES _) Part I - Transition From Forced To Natural Circulation And Thirty Minute Stability Period - 19.25 to 19.92 Hours Figure Title , Discussion 7.5.2.1. RCS Forced Flow The reactor coolant loop differential pressure flow transmitters showed a rapid decrease in forced flow after the reactor coolant pumps were deenergized. Natural circulation flow was too low to register on these instruments. Flow from all three sensors on each loop was averaged to produce a single value. 7.5.2.2. RCS Temperatures Wide range resistance temperature device (RTD) readings are shown without correction for non-isothermal offsets. Temperatures diverged after forced flow , stopped. Hot leg temperatures behaved ' similarly and were approximately equal if the initial offsets are considered. (This indicates mixing in 1 m the reactor vessel.) Cold leg } temperatures varied considerably due to 1 cold auxfliary feedwater and changes in ] steam generator pressure from { atmospheric relief valve operation.
]
7.5.2.3. Steam Generator Narrow Levels decreased for steam generators 1 i j Range Water Levels and 2 and then recovered due to feeding. ' l Feeding raised levels but also cooled j l cold leg temperatures. 7.5.2.4 Auxiliary Feedwater Steam generat.or water levels responded Flow Rates to intermittent feeding. Feeding was stopped by the end of this time period. i 7.5.2.5. Steam Generator RCS heat source decreased when the l Pressures reactor coolant pumps were deenergized. The atmospheric relief valves operated smoothly in automatic for generators
#1 and 2 and maintained almost constant pressure. #3's valve was sticky and resulted in pressure changes as it modulated. #4 was being manually l
operated and remained in its pre-RCP trip position until it was closed to recover pressure at 19.36 hours.
-m
) 7-99 g,
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): TABLE 7.5.2.3: DISCUSSION '0F FIGURES (Continued) Part I - Transition From Forced To Natural Circulation And Thirty Minute Stability Period - 19.25 to 19.92 Hours - (Continued) Figure Title Discussion 7.5.2.6. Pressurizer Level Pressurizer level was in automatic control but level decreased and then increased '.n response to RCS temperature changes. 7.5.2.7. Pressurizer Pressure Normal spray was lost when the RCPs were tripped so auxiliary spray was used to. control pressure from 19.62 to 19.69 hours. Prior to deenergizing the pumps the normal _ spray valves were closed so auxiliary spray.would be forced into the l pressurizer rather than backflowing into the cold legs via the normal spray lines. The backup heaters were off but control bank C was on throughout the test. ( 7-100
/
r a 7.5.2 NATURAL CIRCULATION DEMONSTRATION () 600-10) (Continued): TABLE 7.5.2.3: DISCUSSION OF FIGURES (Continued) Part II - Natural Circulation Steamdown - 19.92 to 20.85 Hours ! Figure Title I Discussion - 7.5.2.8. Narrow Range Steam All feedwater had been stopped and the l Generator kater Levels object was to maintain steam generator pressures (and thus RCS temperatures) as constant as possible. Levels decressed smoothly for the generators with smooth automatic pressure control but in steps for the manually-operated #4 n The rate of level decrease provided , measurement of the heat being removed , by the steam generators. This heat , i loss was added to other known losses to produce a total heat loss that had r . to be balanced by core decay heat. . s The effect of sudden steam generator pressure changes are clearly shown in s the #4 steam 8enerator level. Prior to ) decreasing, level in the saturated steam generator increased when the valve was opened to reduce pressure. Level decreased as the valve was shut. Sudden level changes can be ' correlated to pressure changes in Figure 7.5.2.9. 7.5.2.9. Steam Generator Pressures reflected the degree of Pressures control. Automatic control produced essentially constant pressure. 7.5.2.10. Wide Range Loop Wide range temperature sensors are Temperatures mounted directly on their respective piping and sense fluid temperature in the pipe itself. Tcold reflected pressure control in the steam generators. #4 Tcold changed with generator #4 pressure, but delayed due to transport time between the steam generator tubes and RTD down- . stream of the reactor coolant pumps. , Hot leg temperatures showed only minor variations, delayed due to transport , time through the reactor vessel from the cold legs. . 7-101 p- -- es
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued) TABLE 7.5.2.3: DISCUSSION OF FIGURES (Continued)
} Part III - Illustration Of Behavior Of Narrow Range Temperatures As Indicator Of Natural Circulation Note: During reactor operator training the behavior of the narrow range instruments under natural circulation is emphasized; the purpose of this section is to discuss test findings supporting this training. ; _ Figure Title Discussion 7.5.2.11. Loop Wide And Narrow This illustration compares delta-T power Range Delta-T Power indications. Absolute temperature i
indications are shown in the next Figure. Delta-T power is power level corres-ponding to a teniperature difference. At full flow with the plant is operating at full power it will read 100%. Delta-T power is useful during natural circulation because it provides an indication of how adequately the core is being cooled and if cooling efforts are effective. Loop 4 wide range temperatures showed a delta-T power of approximately 25% while narrow range produced only about 10%. As expected, the narrow ran8e instruments did not provide an accurate assessment of the amount of temperature rise across the core or the amount of decay heat being generated. Loop 4 wide range also showed a greater delta-T when steam flow increased due to opening the atmospheric relief valve. The narrow range instruments gave the opposite indication. 7-102
'l
.i
.]
I l 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continue O : d,G i TABLE 7.5.2.3: DISCUSSION OF FIGURES (Continued) Part III - Illustration Of Behavior Jf Narrow Range Temperatures As Indicator Of Natural Circulation (Continued): Figure Title Discussion 7.5.2.12. RCS Wide and Narrow The four wide range cold leg temperatures i Range Tcolds and single Loop 4 narrow range cold leg I temperature are plotted together. Other I narrow range cold leg temperatures were I not plotted because they did not change. The steamdown time period is covered in this plot for comparison with other parameters. The narrow range cold leg RTD did sense temperature changes, but delayed somewhat from the wide range's.
Narrow range temperature was lower than wide range by about twelve degrees.
7.5.2.13. Narrow Range Delta T This plot covers the initiation of l Behavior natural circulation and shows how narr6w ! range delta-T power peaks and then l decreases sharply as natural circulation develops. Although this is the normally expected behavior, the time frames and magnitudes do not match wide range values, as shown by comparing Figure 7.5,2.2 to this graph. For instance, at 19.4 hours,, Loop 3 has an increasing differential temperature even though its narrow range delta-T power is decreasing, i o a U 7-103 l
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): 1 i f I l PART I: TRANSITION FROM FORCED TO NATURAL CIRCULATION 1 AND THIRTY MINUTE STABILITY PERIOD 19.25 TO 19.92 HOURS 1 l 1 l l l l l l l 7-104
[ l 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): l FIGURE 7.5.2.1: RCS FORCED FLOW UOGILE I RfACTOR COOLANT SYSTEM FORCED FLOW l 1 199 6 LOOP 1 AYERACE RCS FLOW 99 A f40P 2 AUERACE RCS FLOW l 0 LOOP 3 AVIRAGE RCS FLOW 89 E LOOP 4 AVERACE RCS FLOW 79 l I 69 59 2 0 4 49 'l b I J a l l j 39 ' k o 20 k Z l W 19 l 0 K W <
& 9 19.2 1h.3 1h.4 1h.5 1h.6 1h.7 1h,8 1h.9 25 TIME IN HOURS 6/3/87 Note: Flow is shown on the same time scale as the rest of the graphs in this section for comparison. Forced flow was zero within minutes after the reactor coolant pumps stopped.
7-105 L.
7.5.2 NATURAL CIRCIJLATION DEMONSTRATION (1-600-10) (Continued): FIGURE 7.5.2.2: RCS TEMPERATURES U0CfLE I F.lDRAL CIRCUIAfl0N RCS TEMPERATURES 585 A LOOP 1 WIDE RANCE TH0T A LOOP 2 WIDE RANGE TH0T 589 0 IMP 3 WIDE RANCE TH0T ! T40P 4 WIDE RANCE TH07-575 a 570. 569
.I
) &,,i, r---a ,u. 550 >F- 8 kdA JA h 545 'tt 1
,549 W A f40P 1 WIDE RANCE TCOLD _..A_ LOOP 2 il W RANCE TCOLD D LOOP 3 WIDE RANCE TCOLD l LOOP 4 WIDE RaNCE TCOLD A 535 19.2 1/.3 19'.4 1M.5 1M.6 19'.7 IM.8 1/.9 26l TIME IN HOURS 6/3/87
) 7-106
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): 1 FIGURE 7.5.2.3: STEAM GENERATOR NARROW RANGE WATER LEVELS i i UOGILE I HATURAL CIRC. STEAM CIM HARROW RHG LEVELS 85 A STEAM CINIRATOR 1 LEVEL A STFAM CENERATOR 2 LEVEL 0 STEAM CENERATOR 3 LEVEL l STEAM CEMERATOR 4 LEVEL 80 A 75 b ~ n 78
,a
- a. i
_L . O i '. j 1 65 ' 1 A-N _ a 69 4 z @ . l l l l A 55 19.2 19.3 19'.4 19'.5 19'.6 19.7 19.8 19'.9 2 61
!!ME IN HOURS 6/3/87 l
7-107 e
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): FIGURE 7.5.2.4: AUXILIARY FEEDWATER FLOW RATES (~'} NJ V0CTLE I HATURAL CIRCU!Afl0H STDM CEH. AUX. FIED j 148 6 STDM GDERATOR 51 A SIDM CENERATOR N2 0 STDM CDERATOR 33 l STDM CDERATOR 54 120 ll l l Igg WA%Lt ~ l A kA ; E l i U i
%,.l 60 l $l i
w I 49 b lMME I e 29 iJ mag , J l { 8-@ 19.3 19.2 JM 19.4 19.5 M 19.6 19.7 19.8 19.9 2 11 TIKE IN H0tlR$ 6/3/87 7-108
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): (Vl FIGURE 7.5.2.5: STEAM GENERATOR PRESSURES U0CTLE I HRIURAL CIRCULATION STDM CDi. PRESSURES 1159 4 STDM CDfERATOR 1 PRESSURE A SIBM CDEFAIOR 2 PRESSURE 1140 ;fEAM CDiERATOR 3 PRESSURE G STDM CDERAI0i14 PRESSURE l ' 1139 Un m m % 599BM W m\ 1129 r 0 1119 M T d W~ r k i l 1199 g [ 1999 1- 1 4 2,. m 1. L .u _A. iA .a .A A u A . id u . . i 1989 4- o (1979 m i i
/ T/N Eg _
# 96 9 N z
- b f
-N f959 M
e b1049 19.2 19'.3 19'.4 19'5 19'.6 19'.7 19'.8 19'.9 2 61 TIME IN HOURS 6/3/87 I e 7-109 m_____________ _ _ _ _ - - - -
l 7.5.2 NATURAL CIRCULATION DEMONSTRATION'(1-600-10) (Continued):. ; I D FIGURE 7.5.2.6: PRESSURIZER LEVEL i l YOGfLE I NATURAL CIRCULATION PRESSUR12ER LEVEL 27 A PRESSURIZER LEYEL 26.5 R 26 . 1 e
~
{ , sb ! 25.5 a 25 A - d i. < p [ Z l W U d4
$ d / !
44.5 j 19.2 19.3 19'.4 19'.5 19'.6 19.7 - 19'.8 19.9 2 11 flME IN HOURS 6/3/87 l l 7-110
l I l 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): ) l l O FIGURE 7.5.2.7: PRESSURIZER PRESSURE l YOCTI.E I HATURAL CIRCU!ATION PRESSURIZER PRESSURE 227d A PRESSURIZER PRESSURE 2272 2279 - 22.. I J\ i z 1 2264 q j 2262 4 A O ,,,, hL
'5 4
/ \e I 2258 k> >N 256 254 > 100k
[ *B 5 5 5 gi g2252 J e M258 19.2 1h.3 1h.4 19.5 19.6 19' 7 19'.8 19'.9 2 61 flME IN HOURS 6/3/87 l 7-111
7.5.2 NATURAL CIRCULATI0ii DEMONSTRATION (1-600-10) (Continued): I l
)
PART II: NATURA', CIRCULATION STEAMDOWN i 19.92 TO 20.85 HOURS l l i l i l l i i , 1 I I 1 D 7-112 1 l l
l I I 7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued); FIGURE 7.5.2.8: NARROW RANGE STEAM GENERATOR WATER LEVELS UOGYLE I SIDM GENER4 TOR HARROW ~RANCE 144IER LEVELS 75_ _ 6 STDM GD'JJtAIOR 81 - Lf517 7, ( A SIDM CDERATOR 82 - Lf527 0 STDM CDERATOR B3 - Lf537 65 bh l STDM CDERATOR 54 - Lf547 55 > 50
% m s%
45 % g 49 L-35 I U 3 ! i& 30 19.9 28 20'.1 2d.2 2d.3 2d.4 2N.5 2d.6 2d.7 2d.8 29.!'
!!ME IN HOURS 6/3/87 1
7-113
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): FIGURE 7.5.2.9: STEAM GENERATOR PRESSURES U0CTLE 1 HATURAL CIRC. STEAM GDERATOR PRESSURES 1130 6 STEAM GENERATOR 81 - Pf514 A STEAM CDEPATOR 82 - Pf524 l 1120 0 CIEAM GENERATOR 54 - Pf534 E STEAM CDERATOR 84 - PI544 1110 d 1100 AA - --. I n- n
+
1990 ui.a u L g' a" '#m a n yJ L 1 A u lli. AAA n1A b 10a0 1 1970 1NO /
\,i V' l1950 .
(4 P4a a
! h' 030 I i i I i I i I l 19.9 20 20.1 20.2 20.3 20.4 20.5 20.6 20.7 20.8 20.1 j flME IN HOURS 6/3/87 D 7-' '
l t i
- t. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): O FIGURE 7.5.2.10: RCS WIDE RANGE LOOP TEMPERATURES U0GTLE I RCS WIDE RANGE LOOP TEMPERATURES
$80 6 LOOP 1 IHOT - fE413A A LOOP 2 iHOT - TE423A j 0 LOOP 3 TH0T - TE433A l LOOP 4 TH0T - TE443A 575 )
l 579 h "Mh i;,;; ;;gn m u- <uinni e g; 565 6 LOOP 1 TCOLD - TE413B A LOOP 2 TCOLD - TE423B D LOOP 3 ICOLD - TE433B l LOOP 4 ICOLD - IE443B 569 i b 555
.2. u .sw m m .Ld . 11 ht "Il W
E C W A 559 19.9 29 29'.1 29.2 29'.3 29'.4 29.5 29.6 29.7 29' 8 20.0 flME IN HOURS 6/3/87 l l 7-115
7.5.2 MATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued)3 PART III: ILLUSTRATIONS OF CHARACTERISTICS OF NARROW RANGE TEMPERATURES AS INDICATORS OF NATURAL CIRCULATION l 1 l l l l I 7-116 i L----------_--
7.5.2 NATURAL CIRCULATION DEMONSTRATION (1-600-10) (Continued): FIGURE 7.5.2.11: LOOP WIDE AND NARROW RANGE DELTA-T POWER YOCTLE I LOOP I4 WIDE & NARROW RANCE DELTA-f POWER 48 A LOOP 4 WIDE RANCE DELTA T PWR 35 M 30 dbA -
.3 J
25 Rafa a w. d.:1
'emagt.
h 20 0 LOOP 4 MR DELTA-f POWIR TD441 7 C 15 la ar. ) Nd e y ' 3, f 5 19.9 2'8 20'.1 20'.2 20'.3 20'.4 20' 5 2N 6 20'.7 20'.8 20.1 TIME IN HOURS 6/3/87 7-117
7.5.2 NATURAL CIRCULATI,0N DEMONSTRATION (1-600-10) (Continued): FIGURE 7.5.2.12: RCS WIDE AND NARROW RANGE TCOLDS VOCfLE I RCS WIDE AND MRROW RANGE TCOLDS 564 6 LOOP 1 WIDE RANCE TE413B A LOOP 2 WIDE RANCE TE423B u LOOP 3 WIDE RANGE TE433B g LOOP 4 WIDE RANGE TE443B 560
.m 558 356 Wh [ - I
,,, %ypeMpagpeere="gtma!
552 M " 550
.c v 548 0 LOOP 4 NARROW RANCE TE441B -
546 . ' The test was conducted March 28, 1987. The plant was operated at 17% nuclear, power with the main electrical generator supplying power to the grid. 'The j primary and secondary plant were in a normal configuration, Coraputer trends 1 were established to monitor a wide range of plant parameters. . i The loss of offsite power was initiated by tripping all breakers' supplying' i power to the plant (to the 4160 and 13.8 Ky busses) and opening the geierator l ! output breakers. This caused a complete loss of all ac power to tha plant l including equipment normally in service such as instrument air. The . safety- J l related (IE) busses load shed and the emergency diesel: generators started and l loads' sequenced normally. The plant was stabilized in Mode 3 (hot standby)-for O thirty minutes. Systems were then restored and recovery began. ,., j 3 Results l The acceptance criterM and objective were met. . Stable conditions were declared approximately forty-five minutes following tce~ l trip. The plant was maintained in Hot Standby for an additionah thirtpu two minurm. The plant performed satisfactorily and decay heat was removed by ' natural circulation. i
/ ,
Figures 7.5.4.1 and 7.5.4.2 show imnortant parameters: wide range temperatures and steam generator narrow range ater level for loop 1. The slight cooldown >; j observed wasauxiliary balance cold expectedfeedwater due to theaslack of decay steam heat so generator in l,Iwas earlyleve water ant raised.life to The difference between hot and cold leg temperatures i's due to natural ', e circulation since the reactor coolant penps stopped when offsitd power was lost.
,('
1/ 7-122 m
4 1 e 7.5.4 'LOSSOFOFFSITEPOWER_ATEREATERTHAN10%POER(1-600j9] .j
., i j
,,, i
,t Yuf' " j
~
h, l, t Problems l The Proteus computer [s dei r was lost des to high computen coom temperature as I
,, a result ot*the loss'of power to its air conditioning system. However, the i Emergency Respon<;e '
Facility (ERF) computer's data was adeguatn for analyzing l test resulta. One of the valves for volume control tank (VCT? isolat',ca, Valve LV-1120, did not fully c"ose when t;ansfering charging pump suction from the VCT to
- refueling wa',er storage tank (RWST) tot this had no effect on the test. The 3
valve was re[ sited.
' / ..
.') ! ii
. 1 s
< ,, ,e t
4 c 4 i
)
r~ f 1
-)
l
\
i 'm _.) 7-123 l
7.5.4 LOSS OF OFFSITE POWER AT GREATER THAN 10% POWER (1-600-09) ( FIGURE 7.5.4.1: REACTOR COOLANT SYSTEM WIDE RANGE TEMPERATURES LOSS OF OTFS!fE POWER 1-699 RCS TFNFERAfilRES 373 A T HOT LOOP 1 572 565 h 560 ew? 555
- O i %
550 o l $ 545 LA >
'N i
A T COLD LOOP 1 g 548 A- l i l
- 535 44ka n
E 530 1 -1 h } d $ d $ d h d $ l'811l'2 l'31'4 l'5 l'61h I'819 2'8 2'12'2 2'3 2'42h 2'6 2'? 2'8 2'9E TIME DURING TRANSID(T IN MlHUTES 7-124 1 _ _ -__ _ -__ -_-_ a
7.5.4 LOSS OF OFFSITE PO' DER AT GREATER THAN 10% POWER (1-600-09) l-FIGURE 7.5.4.2: STEAM GENERATOR #1 WATER LEVEL l l I p LOSS OF OITSITE POWER 1-600 STEAM GEN. LEVEL i O ss E
$ 0 STEAM GEHmTOR 1 LEVEL
. 50 1 g
W W 45 W U ! I J l c 1 1 8 i g 1 3 0 K I 35 K OAY , 1 0
> 30 ^
^
25 l E ! c W I H i # 28
-1 h } d $ k $ d h d h [0[1[2 N3 N4[5[6I7[8[9N02122 23 N4252'62h N82910 flME DURING TRANSIENT IN MINUTES 7-125
\
SECTION 7.6 r .
. -se OTHER TESTS s .
7-126
I '7.601 ULTTMATE HEAT SINK HEAT REJECTION CAPABILTTY TEST (1-6EF-01) Objective i The objective of the Ultimate Heat Sink Heat Rejection Capability Test was to demonstrate the capability of the ultimate heat sink system, specifically the Nuclear Service Cooling Water (NSCW) cooling towers, .to reject heat and ; verifies that the Vogtle Unit 1 and 2 NSCW cooling towers can function as the ultimate heat sink during a design basis accident. The abstract for this test is FSAR Section 14.2.8.2.60. Methodology j The Ultimate Heat Sink Heat Rejection Capability Test was performed April 21, 1987. The test was performed in accordance with the Cooling Tower Institute (CTI) Standard ATC-105. This standard uses test methodology that compares the ' performance of a cooling tower at existing test. conditions to the 100% design capability and expresses the results as a percentage of the capability. For the NSCW cooling tower, the 100% design capability is that required to satisfy Design Base Accident conditions detailed in FSAR Section 9.2.5. The test was performed by. placing the NSCW system Train B in service with maximum available thermal heat load and monitoring various operating D parameters. The objective was to achieve a heat load of approximately 70E6-BTU /hr to create a measurable water' temperature differential of approximately ] 8'F across the tower. Heat load was maximized by operating the B diesel generator near 100% load, placing the ESF. Chiller and related HVAC in service, operating the Recycle Evaporator and transferring as many loads as possible from Train A to Train B. i For instance, the Auxiliary Component Cooling Water (ACCW) heat exchanger #1 l was isolated and bypassed and Train A containment coolers were placed in i standby. i The tower's four induced draft fans were placed in operation for the duration of the test with the NSCW water return in the Spray mode. NSCW flow was throttled slightly to obtain minimum flow and maximum tower differential temperature. The required NSCW heat load was achieved by mid morning of April 21, 1987, with an average differential temperature of 6.8'F. The test proceeded until 9 pm EST. l 7-127
7.6.1 ULTIMATE HEAT SINK HEAT REJECTION CAPABILITY TEST (1-6EF-01) (Cont.) q Parameters measured during the test and used to calculate tower performance g were:
,(1) NSCW Hot Water Temperature (Tower Inlet Water Temperature);
(2) NSCW Cold Water Temperature (Tower Outlet Water Temperature); (3) Tower Makeup Water Temperature; (4) Air Intake Wet Bulb Temperature; (5) NSCW Water Flow;
)
i (6) Tower Makeup Flow; (7) Fan Power; and (8) NSCW Pump Discharge Pressure. Basin surface water temperature and atmospheric conditions (wind speed) were also continually monitored. Instrumentation used to measure (1) through (8) and basin surface water /N were platinum resistance temperature devices (RTDs) accurate to d tem
- 0. $erature "F. Each RTD was individually calibrated.
minute intervals using a multi-channel data logger. Readings were taken at or> e Eight mechanically aspirated wet bulb psychrometer were located around the periphery of the tower to measure intake air wet bulb temperatures using platinum RTDs accurate to 0.1 F. The RTDs were also measured at one minute j intervals using a multi-channel data lo8ger. Annubars measured NSCW flow. They were verified to 1% accuracy prior to the test. Differential pressure across each annubar was monitored using a Heise j differential pressure gauge. Flow readings were taken at 20 minute intervals. l Tower performance was monitored over a ten (10) hour period. Two, one hour j intervals of maximum stability were selected from this data to calculate tower - performance. The extended monitoring period permitted determination of the thermal time lag associated with the tower basin's large water volume. The calculation was a quasi heat balance that used theoretical tower performance curves in conjunction with test data to determine theoretical flow necessary to produce tower differential temperature measured during the test with prevailing ambient conditions and design fan power. Tower capability was then determined as a percentage of design by comparing the calculated theoretical flow to actual test flow. 7-128
7.6.1 . ULTIMATE HEAT SINK HEAT R"JECTION' CAPABILITY TEST (1-6EF-01)-(Cont.) O~ System configuration and operational limitations caused the test method to V vary from CTI standard. The vendor concurred with all variations and care was f taken to minimize the effect of variations on final test results. The .CTI standard requires establishing a heat load on the tower equivalent to design heat load, in this case the design basis accident. It was not possible to simulate this load so the tower was loaded to the maximum availble load as previously discussed. This heat load was approximately 30% of design and was adequate to = provide sufficient differential temperature to calculate tower capability. To satisfy the accident analysis for evaporation, the NSCW system has a very large basin which causes a considerable thermal lag between the cold water at-the basin water surface and pump discharge where cold water temperature is monitored. Thermal lag was measured during the test and used to displace the hot water temperature time interval, in accordance with the CTI standard. The intent of the test was'to prove the tower has .at least 100% design capablity based on Design Base Accident conditions. This requires burdening the tower with hot humid weather, in addition to design heat load. The FSAR analysis used the hottest and most humid conditions recorded in this area the , last 33 years, from'1947-1981. Because ambient test conditions were different, i the results had to be corrected. Calculations in the procedure accounted for heat loss due to makeup water to the tower but not through the concrete walls to the ground. A conservative O calculation r,howed only an insignificant 0.043 F decrease ~ in basin temperature. Results The acceptance criteria and objectives were met. Calculated NSCW capacity was 128% of design. Test conditions on the April 21 1987 test were 5-10 mph winds, average wet bulb temperature of 64.8 F, and a, dry bulb temperature constantly in the high 80's F. Problems There were no significant problems. J 4 O 7-12e
1 7.6.2 WASTE EVAPORATOR PERFORMANCE TEST (1-5HB-01) i The Waste Evaporator Performance Test has not been performed and will be discussed ir a supplement to this report. l O I 7-130
7.6.3 BORONRECYCLEEVAPORATORPERFORMANCE(1-5HE-0,1] The Boron Recycle Evaporator Test has not been completed and will be reported l in a supplement to this report. 4 E l I 1 7-131
i 7.6.4 METAL IMPACT MONITORING SYSTEM TEST __(1-5SQ-01) Objectives Objectives of the Digital Metal Impact Monitoring System (DMIMS) test were: I (1) final channel calibration _ and_ setup of DMIMS equipment (signal conditioners, power supplies, failure and test modes); (2) establish settings .) for Alarm Signal Levels; (3) demonstrate DMIMS sensitivity; and (4) L demonstrate DMIMS operation under various plant conditions during Plant Startup Testing. The abstract for this test is FSAR Section. 14.2.8.2.19. - i Methodology. l l DMIMS testing began' prior to initial criticality. First, .the twelve signal i conditioners for the twelve DMIMS channels were removed and calibrated. Power supply voltages were checked and proper responses to power supply failures I were verified. DMIMS self-test software was verified. Signal conditioners were reinstalled and connected and channel integrity tests were conducted. To complete preparatory testing, threshold and alarm setpoints were entered for all channels via the DMIMS panel keyboard. Immediately prior to RCS heatup for initial criticality, a test rig was set up and operated successively at each accelerometer location. To demonstrate and quantify DMIMS sensitivity, the test rig applied a-known kinetic energy input near the accelerometer for comparison to DMIMS output for that accelerometer. O, Test method consisted of measuring and recording- back8 round noise. DMIMS response was then recorded for 1/2 ft-lb impacts caused by manually swinging test weights of 0.25 lb, 2.75 lb, and 30 lb. This' methodology was repeated for each of the twelve channels. During RCS dynamic venting, heatup and subsequent startu DMIMS was operated in accordance with system operating procedures,pand testing final testing took place. During control rod movements after initial criticality, the DMIMS Control Rod Drive Mechanism INHIBIT function was demonstrated. Background data was documented at 25%, 50%, 75%, and 100% power plateaus and i alarm / threshold setpoints were verified acceptable after the 100%_ plateau was reached. Periodically throughout the startup program, DMIMS Single Event and Multiple Event reports were generated and evaluated as required and recorded by surveillance procedure 14000-1. l i 1 7-132
7.6.4 METAL IMPACT MONITORING SYSTEM TEST (1-SSQ-01)_(Continued): Results All objectives and acceptance criteria were met. l Preparatory testing was completed with no discrepancies. All hardware associated with the DMIMS panel was adjusted to meet calibration tolerances i and software was verified acceptable. l ( DMIMS sensitivity testing was pnrformed with the reactor coolant system filled and the minimum amount of associated equipment in operation (one or two I reactor coolant pumps running). Average background rates under these l conditions varied from 0.0g to 0.lg. As expected, DMIMS response to test impacts was greatest (as much as 10-15g) for the high frequency inputs from the 0.25 lb weight. Low frequency inputs from the 30 lb weight evoked the I least response (as little as 0.2-0.3g). The data is listed in Table 7.6.4.1. . I 1 i l Background noise trends followed expected results throughout the power j ascension test program. Background noise was greatest at 0% power hot standby I with four RCPs in operation (0.2g for all channels). Background noise dropped I , to essentially zero at 30% and all higher power levels. CRDM cycling was ' l demonstrated to ba indistinguishable from background noise. Problems aj Prior to DMIMS panel testing, corrected. some minor hardware deficiencies had to be ,
' Channel XE-755 experienced a charge preamp failure and the signal j conditioner for channel XE-750 also failed. No retesting was required since !
j the deficiencies were corrected prior to initial calibration. Prior to DMIMS sensitivity demonstration, additional hardware deficiencies (associated with accelerometers and not DMIMS panel) were discovered and corrected. Softline cables were cut or damaged for accelerometers XT-751 and XT-756. The hardline cable for accelerometer XT-758 was severed, necessitating replacement of the entire transducer. No retesting was required since repairs were accomplished prior to initial testing. During DMIMS sensitivity demonstration, channel XE-754 did not register secondary impacts under initial test conditions. Secondary impacts, since they arise from impacts remote from the given accelerometer, are by their nature difficult to detect. Retesting at a later date under slightly more favorable conditions demonstrated the capability of channel XE-754 Eo register secondary impacts. No other significant problems were encountered. I m _) 7- 33 k . _ _ _ _ . _ _ _ . _ _ _ - _ ms
I 7.6.4 METAL IMPACT MONITORING SYSTEM TEST (1-5SQ-01) (Continued): TABLE 7.6.4.1: 1/2 FT-LB SENSITIVITY TEST RESULTS I AVERAGE RESPONSE (G)** CHANNEL NUMBER AVG BACKGROUND 0.25 LB* 2.75 LB* 30.0 LB* XE-750 0.1 4.8 2.3 0.5 l XE-751 0.0 8.5 1.8 0.2 1 XE-752 0.1 6.6 3.8 1.6 XE-753 0.03 4.5 3.2 1.8 XE-754 0.03 9.8 1.2 0.4 XE-755 0.0 14.9 2.5 0.8 XE-756 0.067 4.5 1.6 0.3 XE-757 0.0 8.8 1.9 0.6 XE-758 0.0 2.3 1.6 0.3 { f XE-759 0.067 10.8 2.2 0.6 XE-760 0.0 10.3 2.4 0.3 l XE-761 0.0 9.6 2.6 0.6 l i
* \
Test weight in pounds (LB).
** Data taken prior to heatup.
l l l l v 7-134 l l l _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ ~
7.6.5 GROSS FA1 LED FUEL DETECTOR __(1-6BG-01) Objectives Objectives of the Gross Failed Fuel Detector test were to: (1) calibrate the gross failed fuel detector; and (2) establish baseline activity levels. The abstract for this test is FSAR Section 14.2.8.2.59. l Methodology - 1 ( -The test first calibrated the gross failed fuel detector and then ' established l baseline activity levels at the various power ascension testing power levels. i Calibration of the electronics was according to the manufacturer's calibration procedures. An AmBe neutron source was then placed adjacent-to the. detector and input sensitivity and detector high voltage were set. The system was l aligned response time. for operation with the necessary flow rate to maintain required 3 j l During power ascension, at each power plateau (30%, 50%, 75%, 90% and 100%) five samples of count rates were recorded and averaged to provide a baseline l of activity levels versus power level. A Sraph of these count rates is l included as Figure 7.6.5.1. l Resu'ts l All-objectives and acceptance criteria were met. l The gross failed fuel detector was calibrated and responded as specified to l indicate gross activity levels of the RCS.
}
Problems No significant problems were encountered during this test. f 7-135
7.6.5 GROSS FAILED FUEL DETECTOR (1-6BG-01) (Continued): FIGURE 7.6.5.1 ACTIVITY LEVELS VERSUS REACTOR POWER LEVELS CROSS FAILD HIEL DETECTOR ACTIVITY US POWER LEVEL 200 1 190 $ UOGTLE I STARTUP DATA 180 f
; 170 &
~ 160 ! 4 150 [ g ue /
$ 130 !
120 [ s ua /
;1Bo /
; 90 /
E Ba I
/
y 70 1 _ h 60 50 i l40 g 30 8 20 _
! le 89/
h le 20 3'O 4'O 50 6'O 7'O 8'O 9'O 100 PERCENT REACTOR POWER 7-136
7.6.6 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM TEST (1-6SD-01, Objectives Objectives of the Process And Effluent Radiation Monitoring System (PERMS) Test were to: verify calibration of the PERMS system against an acceptable standard; monitor any activity buildup in the reactor coolant during low power and power ascension; establish baseline activity levels; and demonstrate that PERM 3 is responding correctly by independent laboratory or other analysis. The abstract for this test is FSAR Section 14.2.8.2.28. Methodolo87 Tests of the PERMS monitors were limited to those that are included in Technical Specifications, classified as or support Engineered Safety Features, are assumed to function in accident analysis, or are used to process, store, control, or limit the release of radioactive materials. Calibration and correct response of PERMS was verified by comparison of PERMS cathode ray tube (CRT) readout with sample analyses or direct test instrument readings. Activity buildup was first monitored prior to low power testing when I background health activity level for each detector were measured by chemistry or physics personnel. At 100% power, background activity levels were again measured. A sample or measurement was taken at each detector and the PERMS CRT reading was recorded. Comparison of sample analyses and CRT reading provided the basis for verification of PERMS. Comparison was considered acceptable if either of the following criteria were met: (1) For sample activities above detector sensitivity, acceptance criteria was met (CRT readout within specified tolerance of sample activity); or (2) For sample activities below detector sensitivity, CRT readout was also below detector sensitivity or within tolerance of detector sensitivity. Results All objectives and acceptance criteria were met. For the low power data, background levels at monitoring channels were taken on March 9, 1987, the day of initial criticality. data Comparison with'100% power showed significant differences in only two cases: (1) Containment area radiation power; levels increased from 0 mR/hr at 0% power to 10-15 mR/hr at 100% and (2) Backgound levels for radiation' monitor channels 1-RE-0018, RE-0021 and 1-RE-0848 increased by approximately a factor of ten from 0% 1-to 100% power due to contamination from a mid-May valve misalignment in the backflushable filter system. 7-137
7.6.6 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM TEST' (1-6SD-01) (Continued): Comparison of- CRT readings to sample analyses at 100% power yielded final results which were within the guidelines specified above for: each . detector j tested. As of July 31, 1987, four' monitors remain to be tested, 1-RE-0014, 1-RE-2656A, ARE-2533A, and'ARE-2533B. The ARE-2533A and ARE-2533B are awaiting parts.
]
l Problems Main steam line radiation monitor channel 1-RE-13120 failed its acceptance criteria because the CRT readout was slightly above minimum detector l sensitivity while sample analysis indicated no activity. The CRT readout was only 8% above the detector's minimum sensitivity, so the failure was' a statistical anomaly and no instrument adjustments were warranted. Retesting yielded acceptable results. .j No other problems were encountered.- 1 l j D I 7-138
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7.6.7 BIOLOGICAL SHIELD SURVEY (1-600-05) Objective The objectives of the Biological Shield Survey Test were to: (1) document the radiation levels in accessible locations of the plant outside of the biological shic1d while the reactor is at power; (2) obtain baseline radiation levels for comparison with design radiation zone data and future measurements l of level buildup with operation; (3) evaluate the integrity of the biological and the secondary shield of.the reactor building; and (4) ensure that plant personnel will not be subjected to unnecessary radiation exposure as a result of inadequate. shielding. The abstract for this test is FSAR Section 14.2.8.2.42. Methodology - Precritical and Critical Six separate surveys of the biological shield were conducted - background (pre-criticality), 0% power, 30%. power, 50% power, 75% power and 100% power. 1 Gamma and neutron dose rate measurements were taken during each survey. Gamma measurements were taken at all survey. points within Containment, Control Building, Auxiliary Building and Fuel Handling Building. while neutron i measurements were performed in and immediately surrounding Containment. Initially, 390 Radiation Base Points were surveyed. A total of 179 survey p points were deleted following the 30% power and 50% power surveys either for l Q ALARA purposes for the survey teams or because adequate shielding was j demonstrated. For the 100% power survey, 156 survey points of labyrinths and i horizontal pipe chases were added in the Auxiliary, Fuel Handling and Control ' Buildings per NRC request. Surveys were performed by the following technique: personnel- stood approximately 1 meter from the Radiation Base Point, then while holding the survey instrument approximately 1 meter above the floor they rotated 360 degrees and raised the survey instrument to head level and then down to knee level. The highest observed radiation dose rate was recorded. As the surveyors traversed to the next survey. point, they scanned the wall from approximately 1 meter to 2 meters above the floor elevation, per figure 3 of ANSI Standard 6.3.1, to detect-streaming from voids in shielding. Walls with visible penetrations that wo'11d be accessible to personnel were scanned at a slower rate to ensure no radiation streaming existed. Horizontal pipe chases. were scanned'their entire lengths in 1 meter grids per Figure 2a of ANSI Standard 6.3.1. Survey instruments were held approximately 1 meter from the floor for pipe chases located under the floor and at head level for pipe chases located above a ceiling. ; 7-139
7 6.7 BIOLOGTCAL SHIELD SURVEY (1-600-05) (Continued): Survey. teams consisted of two to four. personnel and were usually accompanied by the Test Engineer. Survey team personnel were given ALARA briefings prior to each survey at 30% power level and above. General- safety, areas with'- potentially high dose rates, survey methods,- instrument response time and , instrument source check frequency were discussed. Instrument source checks ' were. performed prior to starting a battery of surveys, .at 8 hour intervals,.
~
! and at the completion of a battery of surveys. 1 Results i The objectives and acceptance criteria were met. l
.i No areas were found where the dose rate exceeded the radiation zone designations of Table 12.3.1-1 of the FSAR.
Eleven locations were identified where additional shielding'should be provided l inside Containment; however, they were not significant enough to delay l increases to higher power levels. In addition, a radiological protection j program enhancement, a means to allow radiation levels to be permanently I posted, was identified. Problems ! No significant problems were encountered. ! i 1 l D , 7-140 i t-
7.6.8 INPLANT COMMUNICATIONS (1-6QF-01) Objectives Objectives of the Inplant Communications Test were to: (1) demonstrate that the Plant Paging System is audible in areas where ambient' noise levels are high; and (2) demonstrate that the Tone Alarm System is audible in areas where ambient noise levels are high. Tne test was performed at plant power levels of 30% or groter. The abstract for this test is FSAR Section 9.5.2.2.6. Methodology Telephone / Paging System Testing of the Telephone /Page system verified that announcements from speakers in the plant, and on the paging system's five telecommunication lines could be clearly heard above back8 round noise. Test personnel were assigned to the Unit 1 Main Control Room Zone Merging Unit, and designated locations in the plant. At each location, a test message was broadcast over the plant paging system and personnel verified that the message from speakers in the area was clear and intelligible over background noise levels. Communication was established between the control room and each plant location for all five lines of the the paging system. At each location these lines were verified to be clear and intelligible above ambient noise levels. This process was repeated until all areas of a particular building had been tested. Tone Alarm System The Tone Alarm System is used to notify plant personnel of security and the various levels of radiological alerts. Personnel were assigned to the Unit I control room and various locations in the plant. The control room personnel activated a specific tone alarm and personnel assigned in the plant verified that the tone alarm was audible above plant background noise in their area. The areas process was repeated until all tone alarms were tested for all designated of the plant. Results The acceptance criteria and objectives were met. The telephone /page and tone alarm systems performed satisfactorily under ambient noise levels and met the requirements of FSAR Section 9.5.2.2.6. 7-141
7.6.8 INPLANT COMMUNICATIONS (1-6QF-011 (Continued): l
)
Problems l No significant problems were encountered. Problems that were discovered included finding a headset not installed in the control building elevator, the l inability to communicate over the containment elevator headset (repaired by ! maintenance work order), and the headset in the auxiliary building elevator li l not working (also repaired), i l l I i I l i i i
)
l t ! I i I I
~' 7-142 i i l
7.6.9 DYNAMIC RESPONSE TEST (1-600-06) l 1 Objective
'b Objectives of the Dynamic Response Test were to: (1) verify during power range testing that stress analysis of essential nuclear steam supply system (NSSS) and balance of plant (BOP) components under transient conditions is in accordance with design; (2) resolve discrepancies from hot functional testing; (3) test modifications made since hot functional testing and; (4) test systems not tested during hot functional testing.
The abstract for this test is FSAR Section 14.2.8.2.43. Methodology Systems and conditions tested were: (1) The Main Steam System, for flow induced vibration at a power level of approximately 100% reactor power. l (2) The Main Feedwater System, for response to starting and stopping of the feedwater pumps. (3) The Main Feedwater System, for flow induced vibration at a power level of approximately 100% reactor power. (4) The Main Feedwater System, for response to starting a feedwater pump while the other feedwater pump is operating, at 45% to 55% reactor power. (5) The Alternate Charging Line, for flow-induced vibration l with the RCS at full temperature and pressure. i (6) The Steam Generator Blowdown Sampling Lines, for flow-induced vibration with the RCS at full temperature and pressure. This test included only those lines whose design had changed since the Hot Functional Test. (7) The Reactor Coolant Loop Flow Transmitter Lines, for steady-state vibration with the RCS at full temperature l and pressure. (8) The Pressurizer Spray Lines, for steady-state vibration and for response to opening and closing of the spray control valves, with the RCS at full temperature and pressure. l ' (9) The NSCW Bypass Lines, for response to normal train startup, to auto start of the standby pump when an operating pump trips, and to restart of the pumps after a loss of offsite power.
, 7-143 i
_ _ _ _ . _ _ _ _ ________ a
7.6.9 _ DYNAMIC RESPONSE TEST (1-600-06) (Continued): D (10) The Spent Fuel Pool Cooling System, for steady state vibration and for response to starting and stopping the Spent Fuel Fool Cooling System pumps. Systems and cenditions which remain to be tested are as follows: l (1) The Main Steam System, for response to rapid closure of ! the turbine stop valves from approximately 100% reactor power. ! Test methodology varied among the several tests performed. For steady state ! vibration tests on accessible lines, visual inspection of the piping was ' performed, supplemented by readings from a portable vibrometer when necessary. ' Velocity transducers were temporarily mounted for steady state vibration measurements of the main steam and main feedwater lines inside containment because these lines were not accessible due to radiation. l Most dynamic transient tests were performed by visual inspection. Observers ; l were and stationed at various points on the system to look for excessive motion, performed post-test walkdowns to look for damage. Temporary D instrumentation was used for the inside containment portion of the main feedwater lines (support force and velocity), for the main steam lines 1 (support force and pressure), and for the later transient tests of the NSCW . l Bypass Lines after initial visual monitoring indicated problems (displacement, ' l support force, and pressure). Results Dynamic response of the following lines was found acceptable based upon visual inspections: l (1) The outside containment portion of the Main Steam System, for steady state vibration. l (2) The outside containment portion of the Main Feedwater System, j i for starting and stopping of the feedwater pumps. (3) The Alternate Charging Line, for steady state vibration. l (4) The Steam Generator Blowdown Sampling Lines, for steady i state vibration. (5) The Reactor Coolant Loop Flow Transmitter Lines, for steady state vibration. 7-144
7.6.9 DYNAMIC RESPONSE TEST (1-600-06) (Continued): ; (6) The Pressurizer Spray Lines, for steady. state vibration and i for opening and closing of the spray control valves. 1 (7) The Spent Fuel Pool Cooling System, for steady state vibration and for response to starting and stopping of the Spent Fuel Pool Cooling Sy a em pumps. Dynamic response of the following lines was found acceptable based upon instrumented measurements:
'(1) The inside containment position of the Main Steam System, j for steady state vibration.
(2) The inside contaiment portion of the Main Feedwater System, for steady state vibration and for starting and stopping of the feedwater. pumps. Peak values of the measurements for these cases are given in Tables '7.6.9.1. l Problems Test problems occurred on two tests, the steady state vibration of'the Main Feedwater Lines outside containment and the dynamic response to starting of the pumps of the NSCW Bypass Lines. l The outside containment portion of the Main Feedwater System was visually- ";i L inspected for steady state vibration with the plant operating at full power. The only portion of the system which was found unacceptable by _ visual-inspection was on the 1-1305-019-24" line at elevation 260' 9" in the turbine building. This line runs.from the 3rd point feedwater heaters to the 4th point feedwater heaters. Measurements were made at support 1-1305-019-H013 using a portable'vibrometer, , and because of the very low frequency, a scale (ruler).. The results were l evaluated by the design organization which concluded that the' vibration is acceptable for the piping but fatigue failure was likely in support 1-1305- j 019-H013. This support is presently being redesigned. l l 1 The NSCW Bypass Lines were visually observed for dynamic- response during j i normal train startup, autostart of the standby pump when an operating pump is tripped, and during restart of the pumps following a losa of offsite power. When the normal train startup and autostart of the standby pump transients were performed for Train A, the observers stationed along the . piping saw excessive dynamic response. In order to obtain quantitative data on the 3 j [ response, three' displacement transducers were added to the Train B . bypass. j i lines prior to . performing the normal train startup and autostart of the j l standby pumps for Train B. The data was recorded on a light beam j oscillograph. Both the data recorded on the oscillograph and the observers 7-145 i 1 l l
7.6.9 DYNAMIC RESPONSE TEST (1-600-06) (Continued): indicated excessive response. The data recorded at this time was not
- w. . sufficient to allow 'a full analysis of the event, so -additional instrumentation' consisting of a total of eight displacement . transducers, two force transducers, and one pressure transducer was installed on Train B. Also, three rigid supports were shimmed to bring the -piping into a closer correspondence with the analytical model of these lines. The full complement-of instrumentation was recorded durin8 the loss of offsite power test, .with l the' values recorded on Train B to be used in' qualifying Train A'due' to the l similiarity of the two trains.
l The recorded data was plotted and evaluated. The results of the evaluation were that the piping.was being overstressed during the transients and that the pipe supports for these lines needed to be redesigned. A design change package was written to address the problem and perform the required supports redesign. The required modifications are now in progress. l l l l 7-146
i 7.6.9 DYNAMIC RESPONSE TEST (1-600-06)_(Continued): i'
- TABLE 7.6.9.1: STEADY STATE VIBRATION VELOCITIES i
MAIN STEAM MAIN FEEDWATER INSTRUMENT ALLOWABLE MEASURED INSTRUMENT ALLOWABLE MEASURED NUMBER VELOCITY VELOCITY NUMBER VELOCITY VELOCITY i
)
VLAB-1 1.8"/sec 0.50"/sec VLAE-7 1.8"/sec 0.003"/sec i VLAB-2 1.8"/sec 0.45"/sec VLAE-8 1.8"/sec 0.24"/sec f I VLAB-3 1.8"/sec 1.02"/sec VLAE-9 1.8"/sec 0.35"/sec ) VLAB-4 1.8"/sec 0.40"/-sec VLAE-10 1.8"/sec 0.41"/sec l VLAB-5 1.8"/sec 0.80"/sec VLAE-11 1.8"/sec 0.012"/sec VLAB-6 1.8"/sec 0.21"/sec 1 VELOCITY _ TRANSDUCER LOCATIONS l [w INSTRUMENT NUMBER LINE NUMBER NODE POINT DIRECTION VLAB-1 1-1301-107-26" 40 North-South VLAB-2 1-1301-107-26" 40 East-West VLAB-3 1-1301-107-26" 70 Vertical VLAB-4 1-1301-107-26" 70 East-West VLAB-5 1-1301-104-26" 440 North-South VLAB-6 1-1301-104-26" 440 East-West. VLAE-7 1-1305-060-16" 20 Vertical VLAE-8 1-1305-060-16" 20 East-West I VLAE-9 1-1305-058-16" 420 East-West VLAE-10 1-1305-058-16" 420 North-South VLAE-11 1-1305-058-16" 435 East-West
- ,/^'
I 7-147 l l
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i I 7.6.9 DYNAMIC RESPONSE TEST (1-600-06) (Continued): TABLE 7.6.9.2: 1 , O* MAIN FEEDWATER LINE DYNAMIC RESPONSE VELOCITIES AND SUPPORT FORCES i i
. INSTRUMENT ALLOWABLE MEASURED VALUES FOR TEST CASES j NUMBER VALUE A B C D E l VLAE-7 1.8"/sec <0.02"/s <0.02"/s <0.02"/s <0.02"/s <0.02"/s l VLAE-8 1.8"/sec <0.06"/s <0.04"/s <0.04"/s <0.04"/s <0.04"/s VLAE-9 1.8"/sec <0.04"/s <0.05"/s <0.07"/s <0.04"/s <0.04"/s 1
VLAE-10 1.8"/sec <0.02"/s <0.02"/s <0.02"/s <0.02"/s <0.02"/s ! l VLAE-11 1.8"/sec <0.03"/s <0.03"/s <0.30"/s <0.03"/s <0.03"/s ) l CAE-1 120,000# <60# <400# <200# <60# <60# CAE-2 120,000# <100# <200# <200# <100# <100# CAE-3 120,000#- <60# <70# <200# <80# <80# CAE-4 120,000# <300# <200# <300# <300# <300# Note: The dynamic responses of the feedwater line were so small (in most I i ' cases below instrumentation noise levels) they are listed as less than- I some value. LOAD CELL LOCATIONS INSTRUMENT NUMBER SUPPORT NUMBER CAE-1 1-1305-060-H005 CAE-2 1-1305-058-H003 CAE-3 1-1305-062-H006 CAE-4 1-1305-064-H006 TEST CASE DESCRIPTIONS TEST CASE _ DESCRIPTION A Startup of Main Feedwater Pump A. B Shutdown of Main Feedwater Pump A. C Startup of Main Feedwater Pump B. D Shutdown of Main Feedwater Pump B. E Startup of Main Feedwater Pump A while i Main Feedwater Pump B is running and the l plant is at 45% to 55% power. 7-148
7.6.10 THERMAL EXPANSION TEST (1-600-11) Objectives Objectives of the Thermal Expansion Test were to: (1) demonstrate that essential Nuclear Steam Supply System (NSSS) and Balance of Plant (BOP) components can expand without obstruction and that the expansion is in , accordance with design; (?) demonstrate that during cooldown the components return to their approximate baseline cold position; (3) resolve by testing discrepancies from hot functicnal testing and to test modifications made since hot functional testing was con pleted; and (4) test systems not tested during hot functional testing (e.g., main feedwater). The abstract for this test is FSAR Section 14.2.8.2.48. Methodology The specific components tested were: (1) The RCS primary equipment supports (which had been shimmed based on measurements made during hot functional testing) were visually inspected durin8 heatup to ensure proper performance in the shimmed condition. (2) The pipe whip restraints (which had been shimmed based on measurements made during hot functional testing) were visually inspected during heatup to ensure proper performance in the shimmed condition. (3) The Pressurizer Spray Lines, from the Loop 1 and Loop 4 Cold Legs to the Pressurizer. (4) The Auxiliary Pressurizer Spray Line, from the Charging Line to the Pressurizer Spray Line. (5) The Alternate Charging Line, from the Normal Charging Line to the Loop 4 Cold Leg. (6) The portion of the RHR System containing snubbers V1-1205-007-H602 and V1-1205-007-H604. (7) The instrumentation tubing from RTD Bypass Line 1-1201-117-3" to instrument 1-FISL-0427. (8) The Main Steam Line outside containment, up to the main turbine stop valves. This was inspected during secondary side heatup to look for humping or bowing. 7-149
7.6.10 THERMAL EXPANSION TEST (1-600-11) (Continued): (9) All piping in the scope of Preoperational Test 1-300-08, Thermal Expansion Test, was walked down at full temperature conditions to j s, ) look for interferences, except for inaccessible portions of'the Auxiliary Feedwater Pump Turbine Steam Supply Lines. (10) The Main Feedwater System, from the 3rd point feedwater heaters to the. steam generators. This included the main feedwater flow ; path through the auxiliary feedwater nozzles. l Testing was conducted by walking down the test piping and supports at RCS l temperature plateaus during heatup and at reactor power plateaus during power ascension. During the walkdowns, the piping was visually inspected for l interferences, for gaps at pipe whip restraints, for spring hanger settings ! between design hot and cold settings, and for snubber swing clearances. l The position of each snubber was recorded during the walkdowns using a handheld scale. Additional data was collected at each plateau from 28 position transducers and 24 Resistance Temperature Detectors (RTDs) tanporarily mounted on the piping under test. Problems identified during the l l walkdowns were recorded on Thermal Expansion Test Trouble Sheets. A summary of these problems is listed later. All snubber position measurements, position transducer and RTD readings, and Trouble Sheets were evaluated and i disposited. ' l Testing has been performed at the following plateaus: Cold pre-test, 250 F, I s' 350 F, 450 F, Full Temperature, 30% power, 50% power, 75% power, 90% power, and 100% power. Cold post-test readings will be performed when the plant is brought to cold conditions. Results The test objectives and acceptance criteria have been met. Piping displacements from the snubber measurements and the transducer readings at each plateau were evaluated and found acceptable before proceeding to the next plateau. The 100% po w displacements have also been evaluated and found j acceptable. I 1 7-150 i L_____-__________
7.6.10 THERMAL EXPANSION TEST (1-600-11) (Continued): Problems A total of 184 Thermal Expansion Test Trouble Sheets were generated during the testing. After evaluation, 105 of these were dispositioned "Use as is, no action required". Of the 79 Trouble Sheets which required work to be performed, the work required was as follows: Work Required Number of Trouble Sheets Trim insulation to clear interference. 27 Reset spring hangers which are not 12 within design hot and cold settings. Install missing insulation or repair 12 damaged insulation. Reshim pipe whip restraints to proper 8 gap. Retighten loose nuts on pipe supports 5 Remove travel stops from spring hangers. 4 3 Remove temporary pipe supports. 2 Rework spring hanger which was assembled 2 (Two trouble with an incorrect component. sheets on the same problem.) Redesign Main Steam Line supports damaged 2 (Two trouble by humping. sheets on the same problem.) Redesign personnel barrier to clear 1 interference. Cut floor grating to clear interference. 1 Trim pipe support steel to clear interference. 1 Remove plywood s;acer left in steam generator 1 column clevis. Redesign smoke detector mounting plate to 1 clear interference 7-151
7.6.10 THERMAL EXPANSION TEST (1-600-11) (Continued): IT The main steam line humping problem was the most significant; it was the only
\- / prob]em which resulted in the failure under load of a load bearing compcaent.
The following is a discussion of this problem. During the Hot Functional Test, the main steam line was heated too rapidly. The main steam lines at Vogtle are quite long. Condensed water accumulated in I the line quicker than the condensate drains could remove it. Thermal expansion l resulting from temperature distribution due to accumulated condensation (hot steam on top, cooled condensate on the bottom) caused the lines to bow upward, or " hump". The lines lifted off some deadweight supports and caused a pair of vertical rigid strut supports to buckle. After the Hot Functional Test, the buckled rigid struts were replaced with spring hangers and the procedure for heating the main steam lines downstream of the Main Steam Isolation Valves (MSIV's) was revised. Walkdowns of the main steam lines during heatup and cooldown were added to the Power Ascension Thermsl Expansion Test to verify the adequacy of these changes. Heatup of the main steam lines during Power Ascension Testing was completed with only minor humping of the lines which was judged acceptable. On June 11, 1987, though, with the plant at hot zero power conditions, rigid strut 1-1301-l 007-H035 was found buckled and main steam line 1-1301-007-36" was humped so that it was lifted off deadweight supports 1-1301-007-H024 and 1-1301-007- l' H025. Minor damage was also noted at support 1-1301-007-H027 which is located between 1-1301-007-H035 and 1-1301-007-H025. [ ') N-The cause of the buckled strut is believed to have been operation of the line in the hot standby mode with the condensate drains out of service for repair. The humping was observed to come and go over the following several days as the i main steam line temperature changed, subsiding when the line had flow through j it and returning when the line had no flow and was allowed to cool. The i solution, a modification of main steam line hangers, is in the final review ! cycle. i l 1 f 7-152 1
4 7.6.11 PRIMARY AND SECONDARY,, CHEMISTRY (1-600-12) ) Objectives The objectives of the Primary and Secondary Chemistry test were to: (1) demonstrate the ability to maintain reactor coolant and secondary system water - quality within specifications durina the startup testing program; and (2) to compile baseline data for reactor ecolant system and secondary systems.
- The abstract for this test is FSAR Sectiva 14.2.8.2.49.
?
Methodology During the pre-critical test sequence and at each major power plateau, samples ' were taken from the nuclear sampling system and turbine plant sampling system and analyzed by chemistry lab technicians using standard operating procedures. / The lab quality assurance program was used to assure the " test stands" were calibrated. For some parameters, such as cation conductivity, in-line process - analyzers were used to obtain readings. Results Reactor Coolant System All objectives were met. At the pre-critical test and at all power plateaus, - ) the tachnical specification parameters; all kept within specification. chloride, fluoride, and oxygen, were Other diagnostic parameters were also s measured, and kept within specification, ' with the exception of hydrogen, as noted in the " Problems" section below. Secondary Systems All acceptance criteria and objectives were met.
- At the pre-critical and <5% power plateaus, the steam generators are fed directly from the condensate storage tanks by the auxiliary feedwater system.
Both feedwater specifications and steam generator specifications were met during chese test plateaus. Au the 30% plateau and higher plateaus, the main feed and condensate systems are used to feed the steam generators. Also, steam generator chemistry specifications are tighter at these power levels than at the power levels below 5%. Still, all specifications were met at each plateau. Baseline water caemistry data were collected and the plant systems needed to control water chemistry were proven satisfactory by the test. ) 7-153 4
7.6.11 URIMARY AND SECONDARY CHEMISTRY (1-600-l?) (Continued): Problems Reactor Coolant System
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At the <5% and 30% power plateaus, it was noticed that hydrogen concentrations I in the reactor coolant systems were often below the specified minimum of 25 cc/kg. The problem wa,s investigated and traced to an influx of nitrogen leaking into the volume control tank via the backflushable filter system. A design change has been initiated to install an isolation valve in the nitrogen l line. Operating procedures for the nitrogen purge were revised to empty the j nitrogen tank after each backflush and valves around the reactor coolant 1 filters are now manipulated to avoid nitrogen leakage. There were no indications of fuel cladding leaks or primary to secondary tube leaks. Secondary Systems I Though some contamination manifested itself during the startup testing j program, the cleanup mechanisms available (condensate polishers and steam ! generator blowdown) were effective in removal of these impurities. ' There were no indications of condenser tube leaks. l l l v) 7-154 e___-____-_-______
I l l l 7.6.12 VENTILATION CAPABILITY TEST _(1-600-14) L i 'N Objective ! The objective of the Ventilation Capability Test was to verify that various l heating, ventilating, and air conditioning (HVAC) systems for the containment l and areas housing Engineered Safety Features (ESP) equipment continue to t maintain design temperatures. The abstract for this test was FSAR Section 14.2.8.2.58. The test also was I discussed in FSAR Question 640.12. l l l Methodology Each test section will be discussed separately. Room Temperature Survey l With normal ventilation lineups in the auxiliary and control buildings, room l temperature measurements were taken at 0%, 50%, and 100% power. j i " Containment Temperatures With the designed minimum number of containment HVAC systems operating, measurements of containment air temperatures and concrete surface temperatures adjacent to high temperature piping penetrations and at selected locations on l the concrete were taken at 0%, 50%, and 100% power. Data was taken by 4 computers and recorders and is listed in Tables 7.6.12.1 and 7.6.12.2. ESF Room Coolers For each ESF room cooler, room cooler temperatures were recorded. Heat i i transfer capability was computed. Actual tube side heat exchanger heat l transfer capability was calculated from inlet and outlet water temperatu:es l ' and cooling water flow rates. Inlet air wet bulb and dry bulb temperatures and air flow rates to the coolers were also measured. Determining Coil Acceptance Criteria l The inlet air data, inlet water data, and coil data (including coil size, tube size, fin spacing) were input into the American Air Filter's COOLNUC computer program. The computer predicted outlet conditions and theoretical heat transfer capability of the coil. Actual and theoretical heat transfer values were then compared for acceptance at actual operating conditions. Using the actual values for the coil, data was extrapolated to design accident conditions to verify the unit had the required capacity at accident conditions. The heat removal capacities and loads for the coolers are listed j in Table 7.6.12.3. 1 I q s_ ) 7-155 i L__--------_--------
7.6.12 VENTILATION CAPABILITY TEST (1-600-14) (Continued): 1 (-s Results 4 All objectives and acceptance criteria were met. All auxiliary, control, and containment building air temperatures measured at 0, 50%, and 100% power were found to be acceptable. All concrete surface temperatures taken at 0%, 50%, and 100% were found acceptable. Each ESF room cooler has the required capacity to maintain design temperature in the area it serves in an accident environment. 1 Problems Several of the ESF room coolers did not perform as efficiently as the computer i predicted. The coils on one of the ESF room coolers had to be cleaned befori acceptable results could be obtained. Additional calculations were performed on the remaining ESF room coolers that were not performing efficiently. These units were still found to be capable of handling the heat load in an accident condition. To prevent dirt from reducing the performance of the units, the task of cleaning the cooling coils is being added to the preventive maintenance program for each ESF room cooler.
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1 l y en l 7-156
i l 7.6.12 VENTILATION CAPABILITY TEST (1-600-14) (Continued): I TABLE 7.6.12.1: COMPUTER / RECORDER DATA l POWER LEVEL (TEMPERATURE) l 0% 50% 100% ACCEPTANCE INSTRUMENT LOCATION (*F) Q1 ( F) CRITERIA __ TE-2563 Containment 101 91 111 -<120*F Top Center i TE-2612 Containment 83 89 92 1120 F Lower Lvl TE-2613 Containment 81 86 91 1120 F Lower Lvl TE-12258 RX Cavity 89 97.4 101 1200*F ! Upper Concrete TE-12260 RX Cavity 96 103.9 107 1200*F Upper Concrete TE-12259 RX Cavity 91 97.8 103 1135*F Lower Concrete
. TE-12261 RX Cavity 95 101.5 108 1135*F L Lower Concrete I
l TE-2664 RX Cavity 97 95.9 103 1200*F Upper (North) TE-2667 RX Cavity 102 96.5 107 1200*F Upper (East) TE-2665 RX Cavity 99 87.3 108 1200*F Upper (South) l TE-2666 RX Cavity 100 104 108 1200 F Upper (West) TE-2668 RX Cavity 80 77.9 81 1135'F Lower TE-2669 RX Cavity 79 75.2 80 1135'F Lower 7-157
7.6.12 VENTILATION CAPABILITY TEST (1-600-14) (Co-r nued): TABLE 7.6.12.1: COMPUTER / RECORDER DATA (Continued): POWER LEVEL (TEMPERATURE) OA 50% 100% ACCEPTANCE. INSTRUMENT LOCATION (*F) (*F) ( F) CRITERIA TE-12262 Neutron Det 99 102 111 1150*F. Air Passage TE-12263 Neutron Det 94 98 102 1150*F Air Passage TE-12264 Neutron Det 111 113 119 1150*F Air Passage TE-12265 Neutron Det 95 104 105 1150 F Air Passage TE-12266 Neutron Det 102 102 109 1150 F Air Passage TE-12267 Neutron Det 94 103 110 1150"F Air Passage n v TE-12268 Neutron Det Air Passage 101 105 106 1150*F TE-12269 Neutron Det 94 96 98 1150*F Air Passage TE-12468 FHB Corridor 72 70 82 1135 F TE-2688 RX Cavity 86 89 89 -<135*F Concrete j 1 TE-12971 Main Steam 60 79 90 1150 F l Line Control l Bldg. TE-12969 Main Steam 104 86 100 -<150*F Line Aux. Bldg. TE-12270 Containment 102 114 122 <200*F Concrete TE-12271 Containment 114 130 133 1200*F Concrete 7-158 _ _ . _ _ __.__._____._______i
~
7.6.12 VENTILATION CAPABILITY TEST (1-600-14) (Continued): TABLE 7.6.12.2: AREA TEMPERATURES F0WER LEVEL (TEMPERATURE) 0% 50% 100% ACCEPTANCE LOCATION (*F) (*F) ( F) CRITERIA Control Bld8 71 76 72 >70 F, <80*F Level 4 Battery Room R-409
; Control Bldg 70.3 77 71 170 F, 180 F Level 1 Battery Room R-183 4160V SWGR 66 70 70 165'F, 1100*F 1AA02 Room R-A48 'h 4160V SWGR 70 73 73 165*F, 1100 F IBA03 Room R-A50 ; Main Steam and 68 76 85 >17 F, <126 F Feedwater Valve Area SG2 and SG4 Room R-A56 RX Trip 68 78 70 SWGR 165'F, 1100 F Room R-B71 CCW HX 66 73 i
Room R-20; 78 140 F, 1100 F Train A CCW HX 65 71 78 >40 F, ~<100 F Room R-203 - Train B SWGR 1AB16 77 76 80 >40 F, ~<100 F Room R-207 ~ ACCW HX 69 73 77 >40*F, <100 F Room R-104 - ~ Train A 7-159
1 7.6.12 VENTII* TION CAPABILITY TEST (1-600-14) (Continued): TABLE 7.6.12.2: AREA TEMPERATURES (CONTINUED): POWER LEVEL (TEMPERATURE) ) 04 50% 100% ACCEPTANCE ! LOCATION (*F) ( F) ( F) CRITERIA j l ACCW HX 69 70 76 140*F,1100*F Room R-105 Train B Main Steam 68 80 88 J,17 F, 1126*F and Feedwater Valve Area SG1 and SG3 Room R-108 MCC 1BBB 80 77 79 ?_40*F, 1100 F Room R-118 l MCC 1ABB 77 78 78 140 F, 1100*F Room R-116 l i CCW Pump 72 67 77 ~
>40*F, ~<100 F Room R-A03 Train A ;
CCW Pump 72 72 75 )_40*F, 1100 F Room R-A05 Train B i Spent Fuel 75 77 75 ?_40*F, 1100*F Pool HX Room R-A53 Train A ACCW Pump 71 70 73 )_40 F, 1100*F Room R-B23 Train A ACCW Pump 74 75 74 140*F, 1100 F Room R-B24 Train B SI Pump 71 74 75 p_40*F, 1100"F Room R-B15 Train A O 7-160 t L-____-_-_---- - _ - _ _ - - - - - - _ - - - - - - - - _ - . - - - - - - -
7.6.17 VENTILATION CAPABILITY TEST (1-600-14) (Continued): I ! TABLE 7.6.12.2: AREA TEMPERATURES (CONTINUED): POWER LEVEL -(TEMPERATURE) 0% 50% 100% ACCEPTANCE LOCATION (*F) ( F) ('F) CRITERIA l SI Pump 71 79 74 140*F,1100*F , Room R-B19 Train B l Piping 90 95 92 140*F, 1100 F-Penetration Room R-B11 Train B 1 1 Chg Pump 73 72 79 140 F, 1100*F I Room R-C115 Train A Chg Pump 68 67 75 140'F, 1100*F Room R-C118 Train B Chg Pump 78.2 75 75 140*F, 1100"F Room R-C111 PDP Pump ' RHR HX 70 70 74 140*F, 1100*F -i Room R-C90 Train A j RHR HX 75 75 75 2,0'F, 4 1100*F Foom R-C91 Train B Piping 83 86 ~81 2,40*F, 1100*F Penetration Train A Room R-C105 Cont Spray 77 81 75 140*F, 1100*F Pump Room R-D77 Train A Cont Spray 71 74 76 140'F, 1100*F l Pump Room R-D76 Train B O- 7-161
I l 7.6.12 VENTILATION CAPABILITY TFRT (1-600-14) (Continued): TABLE 7.6.12.2: AREA TEMPERATURES (CONTINUED): POWER LEVEL (TEMPERATURE) UA 50% 100% ACCEPTANCE l LOCATION (*F) ( F) (*F) CRITERIA RHR Pump 70 69 75 >40 F, <100 F Room R-D48 Train A RHR Pump 70 72 75 140 F, 1100 F l Room R-D122 ' Train B Spent Fuel 73 69 78 140"F, <104 F l Pool Pump j Room R-A07 ' l l NSCW 74 75 ! CNMT Coolers 75 140 F,1100 F Room R-B11 Aux FW D Pump Room R-101 Train A 73 79 83 140 F, 1120*F Aux FW 70 80 Pump Room R-102 84 140 F,1120 F Train B Aux FW 78 79 89 140 F, <120 F Pump Room R-106 ~ Train C Diesel Generator 84 85 82 150*F, ~<120 F Train A Room 103 Diesel Generator 84 68 85 150 F, 1120 F Train B Room 101 l Steam Tunnel 66 66 82 SG1 and SG3 117 F, 1126*F Steam Lines Steam Tunnel 66 76 86 SG2 and SG4 117*F,~<126*F m Steam Lines 7-162 l
i 7.6.12 _ VENTILATION CAPABILITY TEST (1-600-14) (Continued):
,-~
k-s TABLE 7.6.12.2: AREA TEMPERATURES (CONTINUED): l POWER LEVEL (TEMPERATURE) 0% 50% 100% LOCATION (*F) ( F) ( F) i MET TOWER 61 64 81 Ambient Temp T6172 l MET TOWER 61 64 83 4 Ambient Temp T6438* MET TCVER 4 mph 7.3 mph 5.5 mph Wind Speed l S6170* (MPH) l
* ]
l This is the ERF Computer Point. These values may be read ' off the ERF Computer display. I h
\
I i l f (,)/ 1 l 7-163 1
7.6.12 VENTILATION CAPABILITY TEST (1-600-14) (Continued): 1 TABLE 7.6.12.3: TEMPORARY THERMOCOUPLE TEMPERATURE READINGS _ POWER LEVEL (TEMPERATURE)
- 07. 50% 100% ACCEPTANCE PENETRATION LOCATION ( F) (*F) (*F) CRITERIA 1 Top of penetration 97 113 109 -
<200 F l directly on concrete where it meets edge of anchor ring cover plate SG1 Main Steam.
1 Bottom of penetration 76 88 94 1200 F directly on concrete where it meets edge of anchor ring cover plate i SG1 Main Steam. ! 1 Approx 90 on 84 101 98 1200 F penetration directly
]
on concrete where it ! meets edge of anchor ! ring cover place SG1 Main Steam.
'Ok.)
2 Top of penetration 85 92 102 ~
<200 F directly on concrete where it meets edge of anchor ring cover plate 3
SG2 Main Steam. l 2 Bottom of penetration 73 88 92 <200 F ! directly on concrete l where it meets edge of l anchcr ring cover plate l SG2 Main Steam. 2 Approx 90* on 75 90 94 1200 F penetration directly on concrete where it meets edge of anchor ring cover place SG2 Main Steam. 3 Top of penetration 89 84 100 1200 F directly on concret) where it meets edge of anchor ring cover plate SG3 Main Steam. f%
\
7-164
7.6.12 VENTILATION CAPABILITY TEST (1-600-14) (Continued): TABLE 7.6.12.3: TEMPORARY THERMOCOUPLE TEMPERATURE READINGS (Continued): l POWER LEVEL (TEMPERATURE) 0% 50% 100% ACCEPTANCE PENETRATION LOCATION ( F) ( F) ( F) CRITERIA 3 Bottom of penetration 72 87 95 -<200 F : directly on concrete where it meets edge of ancher ring cover plate SG3 Main Steam. 3 Approx 90 on 80 85 96 1200 F l penetration directly i on concrete where it I meets edge of anchor l ring cover place i SG3 Main Steam. 1 4 Top of penetration 101 117 122 1200 F l directly on concrete where it meets edge of anchor ring caver plate SG4 Main Steam. 4 Bottom of penetration D directly on concrete 80 85 98 1200 F ; l where it meets edge of anchor ring cover plate SG4 Main Steam. I 4 Approx 90* on 83 98 118 1200 F penetration directly l on concrete where it l meets edge of anchor ring cover plate SG4 Main Steam. I 18 Top of penetration NA 86 98 1200 F directly on concrete where it meets edge of anchor ring cover plate SG1 Main Feed. 18 Bottom of penetration NA 83 93 1200 F directly on concrete where it meets edge of anchor ring cover plate SG1 Main Feed. 7-165
7.6.12 VENTILATION CAPABILITY TEST /1-600-14) (Continued): TABLE 7.6.12.3: TEMPORARY THERMOCOUPLE TEMPERATURE READINGS (Continued): POWER LEVEL (TEMPERATURE) 0% 50% 100% ACCEPTANCE _ PENETRATION LOCATION ( F) ( F) ( F) CRITERIA 19 Top of penetration NA 87 97 <200*F directly on concrete where it meets ed8 e of anchor ring cover plate SG2 Main Feed. 19 Bottom of penetration NA 83 93 1200*F directly on concrete where it meets edge of anchor ring cover plate SG2 Main Feed. 20 Top of penetration NA 102 109 1200 F directly on concrete where it meets edge of anchor ring cover plate SG3 Main Feed. 20 Bottom of penetration NA 95 104 -<200 F directly on concrete where it meets edge of anchor ring cover plate SG3 Main Feed. ~ 21 Top of penetration NA 105 108 <200 F directly on concrete ~ where it meets edge of anchor ring cover plate SG4 Main Feed. 21 Bottom of penetration NA 97 106 <200 F directly on concrete ~ where it meets edge of
, anchor ring cover plate SG4 Main Feed.
7-166
7.6.12 VENTILATIONCAPABILITYTEST(1-600-14)-(ContinueQ; TABLE 7.6.12.4: HEAT EXCHANGER DATA MINIMUM DESIGN -PREDICTED HEAT ACTUAL HEAT REQUIRED HEAT REMOVAL USING REMOVAL FFOM REMOVAL FROM THE C00LNUC' MASTER PARTS LIST FIELD CALCULATION PROGRAM, DESIGN EQUIPMENT TAG MEASUREMENTS X4C1592-V02 CONDITIONS *** i NUMBER BTU /HR BTU /HR BTU /HR I 1 1-1555-A7-001 90340 76800 113917 1-1555-A7-002** 76270 69900 137630 1-1555-A7-003** 20628 23300 40083* 1-1555-A7-004 29308 15900 31494 1-1555-A7-005 30511 32700 42346 1-1555-A7-006 28376 29325 40768 1-1555-A7-007 43954 43650 46883 { 1-1555-A7-008 42986 43800 67730 1-1555-A7-009** 24393 24000 46480
)
1-1555-A7-010 30691 24010 43141 1-1555-A7-011 54979 76100 86481 1-1555-A7-012 45190 76100 87932 1-1555-A7-013 39679 43710 70304 1-1555-A7-014 42314 43710 70304 _O 1-1555-A7-015 32986 31550 61885 1-1555-A7-016** 38240 31550 68635* 1-1555-A7-017 55069 64000 119520 1-1555-A7-018** 85960 64000 119520 1-1539-A7-001** 19837 17934 51834* 1-1539-A7-002 15514 11544 48988 1-1539-A7-005** 48456 55393 76017* 1-1539-A7-006** 29851 41567 47392* 1-1532-A7-001** 255810 581045 735673* l 1-1532-A7-002 344172 649756 796757 I These values were calculated based on the field data (field UA) applied to field conditions because the cooling coils were not as efficiently performing as predicted by the C00LNUC program. Evaluated and accepted. The predicted heat removal values were calculated using the actual heat removal field measurements and were compared to_the minimum design required heat removal requirements. 7-167
7.6.13 JTEAM GENERATOR MOISTURE CARRYOVER (1-700-03) () Objective The objective of the Steam Generator Moisture Carryover test is to. determine the moisture carryover performance of the steam 8enerators. The abstract for this test is FSAR Section 14.2.8.2.54. Methodology The moisture carryover test is performed using approximately one Curie of radioactive sodium (Na-24) solution introduced into the feedwater system by a I chemical addition pump. Samples are drawn in steam, feedwater, and steam generator blowdown for determination of radioactivity content due to the i l sodium. Moisture content is determined from the activity levels. l l The plant is operated in steady state and at er near full power for the test. l l Results The moisture carryover test has not been completed. It will be addressed in a supplement to this report. 1 J l C~jD 7-168 i l i
7.6.14 PLANT PERFORMANCE (1 v00-01) Objectives The objectives of the-Plant Performance test are to: (1) monitor secondary plant systems under loaded conditions during power ascension testing; and (2) make adjustments to p1r a secondary systems as necessary to improve plant efficiency. The abstract for this test is FSAR Section 14.2.8.2.55. Methodology Secondary plant performance monitoring started by establishing a Proteus computer trend that listed many of the secondary equipment's temperatures, pressures, and flow rates so performance of pumps and heat exchangers could be assessed. Certain primary plant parameters were also included because the primary plant's temperatures and power level impact the secondary. The monitoring program was aimed at bringing the plant as closely as possible to the heat rate drawings and making adjustments from that baseline. In order to do this, instrumentation was placed in service, feedwater heater levels stabilized at an efficient operating point, and the secondary inspected for equipment problems as the plant was brought to power. The heat rate drawings were in terms of turbine power but listed reactor power as an input. The flows, temperatures, and pressures were replotted as a function of reactor power so it was possible to compare as-found values with predictions based on the heat rate. Results The plant was monitored throughout its power ascension phase and this monitoring discovered problems (discussed below) that have been subsequently corrected. Gross generator electrical output is approximately at design. Testing has not been completed and results will be reported in a later supplement. Problems The main feedwater final temperature instruments to the Proteus used by operations for calorimetric determination of reactor power were found to be reading approximately 6*F lower than true at full power. The original Proteus calibration used the American standard curve for 100 ohm platinum RTDs whereas the plant calibrations were made to the European standard (DIN 43760). There is approximately a 6 F difference between the two curves at 440'F. 7-169
7.6.14 PLANT PERFORMANCE (1-800-01) (Continued) l The problem , discovered at full power while trying to perform an energy balance on the final feedwater heaters as part of the Plant Performance procedure. A difference between the Proteus inputs and RTDs at the steam generator nozzles had been previously noted and the latter RTDs had been used ( for the startup program's statepoint and data collection procedure (1-5SC-02) l that in turn calibrated steam flow, determined full power reactor coolant loop differential temperatures, and defined the state of the plant at full power so no retesting of already-completed startup procedures was required. Level controls for the final feedwater. heaters (termed #6 heaters at Vogtle) , were initially tc low for proper operation. These vertical heaters are 1 equipped with intebral drains coolers and the low water level at the entrance to the drains cooler caused steam to enter the drains cooler and be carried with slugs of water into piping cascading to the next stage heaters (called #5 l heaters). S3ug flow caused the cascade lines to vibrate excessively and they ' were temporarily restrained by wire. Excessive steam flow into the drains cooler can lead to tube erosion so the problem was quickly addressed and the level controls were raised. The low level problem was first noticed by excessive drains cooler approach temperature. Drain cooler approach (DCA) is the difference between subcooled condensate leaving the drains cooler and incoming feedwater temperatures. DCA should normally be 10 to 12*F on heaters with drains coolers and there is a narrow (one or t5o inch level) operating band below which DCA increases until it approaches the temperature rise across the feedwater heater as steam
, blows through the drains cooler. Heaters with integral drains coolers should be operated above the inflection point and raising the level controllers I allowed enough range for proper control.
The fourth point heaters do not have integral drains coolers and are not intended to operate with a liquid level. They are equipped with level sensing switches to stop incoming extraction steam from the turbine and cascade drains from upstream feedwater heaters in case of high level. The upper level taps on l the heaters are perturbed by steam flow inside the heater and produce a false high level indication when extraction steam is admitted. .This high level in I turn causes extraction steam to be isolated. The solution was to raise the level switches. The switches will still provide equipment protection but at a slightly higher actual water level. The option of raising the switches was chosen over tapping the thick feedwater heater shell.
/m b 7-170
I 7.6.15 NUCLEAR STEAM SUPPLY SYSTEM ACCEPTANCE TEST (1-800-02) l p V Objectives ! The objectives of the Nuclear Steam Supply System (NSSS) Acceptance Test were to: (1) demonstrate the reliability of the NSSS by maintaining the plant at rated output (+0, -5 percent) for one hundred hours without a load reduction resulting from an NSSS malfunction; and (2) measure and compare the NSSS output to its warranted rating of 3425 Mwt. The abstract for this test is FSAR Section 14.2.8.2.56. i Methodology 5 The plant was initially operating at 99 1% of rated thermal power with steady ' state conditions established and control systems in automatic mode. Operat. ions surveillance procedure 14030, Power Range Calorimetric Channel , Calibration, was performed; thermal power was adjusted to 99.5 0.5% rated ! thermal power and steady state conditions were established. I I 1 The Proteus computer was sat up to collect steady state trend data once per fifteen (15) minutes per procedure 1-800-01, Plant Performance. An hourly log was established to record gross generator output, reactor power, and acceptable accumulated time. The 100 hour reliability test was then started. ! After 55 hours of continuous steady state operation, the NSSS performance measurement was started. The following measurements were performed: 2 (1) Proteus steady state data collection was established at five minute intervals, I (2) Two half-hour precision calorimetric (usin8 test instruments) t were performed in accordance with procedure 1-5SC-02, Thermal l Power Measurement, to verify plant stability at rated NSSS thermal { output of 3425 +0, -1.0% MWT, (3) Four one-hour precision calorimetric were then performed in accordance with procedure 1-5SC-02, Thermal Power Measurement, and l (4) The results of the four one-hour calorimetric were then averaged l to determine the average thermal power output during the NSSS l Performance Measurement Test. After the NSSS performance measurement was completed, the Proteus steady state data collection was reestet.tished at once per fifteen minutes for completion of the 100 hour warranty run. O 7-171 l
7.6.15 NUCLEAR STEAM SUPPLY SYSTEM ACCEPTANCE TEST (1-800-02) (Continued): Results All acceptance criteria and objectives were met. The 100 hour reliability test was completed uninterrupted at an average reactor power level of 99.57% with a minimum power level of 98.4% and a maximum power level of 99.9% during the 100 hour interval. Average thermal power output during the four hour NSSS performance measurement ; was 3420 Mwt or 99.97, of warranted thermal power output. l i l Problems l The Proteus computer failed due to a buffer problem for a 52 minute period during the 100 hour reliability test. Three fifteen minute steady state data sets were missed during this failure. However, control room strip chart recorders and the ERF computer trends were utilized to record data during the Proteus' failure. No other problems were encountered during the test. , l 4 1 I I [ t., . 7-172
8.0
SUMMARY
(%I The Vogtle Unit 1 startup test program, although not entirely completed, has V shown that the plant operates as designed and can withstand transients that l can reasonably be expected during its lifetime. ' The remainder of the startup test program will be addressed in a- later supplement to this report. I d 1 I l l l [ 8-1}}