L-05-026, Supplemental Information Supporting License Amendment Request Nos. 302 and 173, Appendix a of Attachment D
| ML051160431 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 02/23/2005 |
| From: | Pearce L FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-05-026 | |
| Download: ML051160431 (802) | |
Text
{{#Wiki_filter:Appendix A of Attachment D Beaver Valley Power Station Startup Reports
. Beaver Valley Power Station Unit No. 1 Startup Reports
*m.a .
*o03:5fa- 537 Duqu^isn L!,t * * (412) ItI 4 435 sith A.eg hltbwgr, VWflvfan llsl January 4, 1977 Beaver Valley Power Station, Unit Docket No. 50-334, License No. DI United States Nuclear P.eoulatorj Commissicn Director, Re'Zonal Office King of Prussia, Pennsylvania 19406 Ceatlc=zn:
In accordance %':th Rculatorry Guide 1.1', Repcrting s' Opzratir." Inforration - A.,Tceh.tcpl *^+/-44 e". ,' or:Ch~1 r report of plant startup and power escalation is being submitted 90 daYs folloiing declarat4i of :o~--eccnent of r- -1 rat^- for !ave-Valley Power StAtion Unit No. 1. Completion of the startup test program
'has not been accomplished and a supplemiental raport will be issued within three months of this report to further report the progress of the program.
Very truly yours, C. N. Dunn Vice President, Operations httachnent cc: Director, Office of Management Information and Progran Control (25) Director, Office of Inspection and Erforcement (2)
BEAVER VALLEY POWER STAT ION UNIT 1 STARTUP REPORT DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY
Table of Contents .. i Table of Contents
- - I :
iii List of Figures vi List of Tables - ,I,' ,. viii List of Abbreviations '- x Introduction ' 1.0 Initial Core Loading 2.0 Control Rod Testing 2.1 General 2.2 Control Rod Position Indication System Test 2.3 Control Rod Drive MechanismTiming Test 2.4 Control Rod Drop Time Measurement Test 2.5 Part Length Control Rod Mechanism Brake Test 2.6- Rod'Control System Test o . .Lod r i . . tIng. . I : -3.0 :Post Core toad, Precritical Testing,. . , 3.1 General 3.2 Incore Movable Detectfor System 3.3 Incore TC/RTD Cross Calibration - 3.4 Pressurizer Continuous SprAy Flow Verification 3.5 Pressurizer -leaters and Spray Capability 3.6 RTD Bypass Loop Flow'Verification 3.7 RCS Flow Measurement 3.8 Reactor Coolant System Flow Ccastdown 4.0 Hot Zero Power Testing., 4.1 General 4.2 Initial Approach to Criticility 4.3 Isothermal Temperature Coefficient 4.4 RCCA Worth Measurement', 4.5' Boron Endpoint and Boron Worth Measurement 4.6 Minimum Shutdown Verification 4.7 RCCA Pseudo'Ejeciion at Zero Power 5.0 Power Ascension Testing 5.1 Physics Tests 5.1.1 General 5.1.2 RCCA Pseudo Ejection and RCCA Out of Position 5.1.3 Power Distribution Measurements 5.1.4 Power Coefficient Measurement 5.1.5 Static RCCA Drop and RCCA Out of Position 5.1.6 Special Hot Channel Factor Test 5.1.7 Dynamic Rod Drop i
5.2 Transient and Trip Tests 5.2.1 General 5.2.2 Load Swing Test 5.2.3 Large Load Reductton 5.2.4 Net Load Trip Test 5.2.5 Plant Trip from 100% Power 5.2.6 Plant Load Follow Demonstration 5.2.7 Loss of All Off-Site Power at Load 5.3 Control Systems Dynauic Response Tests 5.3.1 General 5.3.2 Automatic Reactor Control 5.3.3 Automatic Steam Cenerator Level Control 5.3.4 Automatic Steam Dump Control 5.4 Instrumentation Calibration and Alignment 5.4.1 General 5.4.2 Operational Alignment of NIS 5.4.3 Startup Adjustments of Reactor Control System 5.4.4 Calibration of Steam and Feedwater Flow at Power 5.4.5 Operational Alignment of Process Temperature Instrumentation 5.4.6 rncore-Excore Detector Calibration 5.5 Secondary Plant Performance 5.5.1 General 5.5.2 urbine Overspeed Trip Test 5.5.3 Turbine Stretch Test 5.5.4 Steam Generator .Y4oistu're Carryover 5.5.5 Cooling Tower Performance 5.5.6 Auxiliary Feedwater Transient Test 5.6 Reactor Vibration Analysis 5.6.1 General 5.6.2 Vibration and Loose Parts Monitoring System 5.6.3 Nuclear Noise Measurement 5.6.4 Incore Detection of Fuel Assembly Vibration 5.7 Plant Radiation Survey and Shielding Effectiveness 5.8 Shutdown From Outside of the Control Room 5.9 Plant Sound Level Survey 5.10 Startup Testing Chronalog ii
LIST OF FIGURES FiRure T.tle.. Page 1.0-2 ICRR Plot, Temporary Source Range Detector A 1.0.3 1.0-2 ICRR Plot, Temporary Source Range Detector B 1.0.4 1.0-3 ICRR Plot, Temporary Source Range Detector C 1.0.5 1.0-4 ICRR Plot, Source Range Channel; 31 1.0.6 1.0-S ICRR Plot, Source Range Channel .N3i 1.0.7 1.0-6 Core I Assembly Xap - 1.0.8 2.1-1 Control Rod Locations 2.1.3 2.2-1 RPI Position Signal vs. Rod Position 2.2.3 2.3-1 Typical CRDH Withdrawal Cycle., 2.3.3 2.3-2 Typical CRDH Insert Cycle 2.3.4 2.*3-3 Control Rod Withdrawal Timing Trace. 2.3.5 2.3-4 Control Rod Insertion Timing Trace 2.3.6 IJ 2.4-1 Typical Rod Drop Trace 2.4.6 2-4-2 Rod Drop Times, Cold No Flow 2.4.7 2.4-3 'Rod Drop Times, Cold Full Flow.', 2.4.8 2.4-4 Rod Drop Times, Hot No Flow 2.4.*9 2.4-6 Rod Drop Times, Hot Full Flow 2.4.10 2.5-1 Part Length Control Rod Brake Mechanism Sound Level 2.5.2 end Power Supply Current vs. Time 3.2-1 Core Distribution of Flux Thimbles' 3.2.3 3.3-1 Distribution of Incore Thermocouples 3.3.4 3.5-1 Pressurizer Pressure vs. Time (Spray Valve Operation) 3.5.3 3.5 Pressurizer Pressure vs. Time'(Hep.Ler Operation) 3.5.4 3.8-1 RCS FlovCoastdown, One,Out of Three Loop Coastdown. 3.8.5 3.8-2 RCS Flow Coastdown, Three Out of Three Loop Coastdown 3.8.6 RSr
GIAure Title PaNe 3.8-3 RCS Flow Coastdovn, Two Out of Two Loop Coastdown. 3.8.7 No Loop Isolated. 3.8-4 RCS Flow Coastdown, Two Out of Two Loop Coastdown. 3.8.8 One Loop Isolated. 3.8-5 RCS Flow Coastdown, One Out of Two Loop Coastdown. 3.8.9 No Loop Isolated. 3.8-6 RCS Flow Coastdovn, One Out of Two Loop Coastdown. 3.8.10 One Loop Isolated. 4.2-1 RCS Makeup vs. ICRU-N31 4.2.4 4.2-2 RCS Makeup vs. ICRR-N32 4.2.5 4.2-3 RCS Boron vs. ICRR-N31 4.2.6 4.2-4 RCS Boron vs. ICCR-N32 4.2.7 4.2-5 ICRR vs. Rod Position-N31 4.2.8 4.2-6 ICCR vs. Rod Position-N32 4.2.9 4.3-1 Isothermal Temperature Coefficient Trace 4.3.4 4.4-1 Differential and Integral Reactivity Worth - 4.4.4 Control Bank A. 4.4-2 Differential and Integral Reactivity Worth - 4.4.5 Control Bank B. 4.4-3 Differential and Integral Reactivity Worth - 4.4.6 Control Bank C. 4.4-4 Differential and Integral Reactivity Worth - 4.4.7 Control Bank D. 4.4-S Differential and Integral Reactivity Worth - 4.4.8 Control Banks in Normal Overlap 4.5-1 Critical Boron Concentration vs. Integral Reactivity 4.5.4 Worth (Boron Worth) 4.6-1 Critical Boron Concentration vs. Integral Reactivity 4.6.4 Worth (Minimum Boron Concentration) 5.1.3-1 Flux Hap No. 2A, Power Distribution Heasurements 5.1.3.4 5.1.3-2 Flux Hap No. 3B, Power Distribution Measurements 5.1.3.5 5.1.3-3 Flux Hap No. 6A, Power Distribution Measurements 5.1.3.6 5.1.3-4 Flux Mpa No. 5A, Power Distribution Measurements 5.1.3.7 iv
Title Page Figure 5.1.3.8 5.1.3-5 Flux Map No. 7A, Power Distribution Measurements 5.1.3.9 5.1.3-6 Flux "ap No. 8A, Power Distribution Measurements Flux Map No. 9A, Power Distribution Measurements 5.1.3.10 5.1.3-7 5.1.3.11 5.1.3-8 Flux Map No. 9A, Power Distribution Measurements Flux Map No. lOA, Power Distribution Measurements ,. 5.1.3.12 5.1.3-9 5.1.3.13 5.1.3-10 Flux Map No. 12A, Power Distribution Measurements Flux Map No. 1SA, Power Distribution Measurements 5.1.3.14 5.1.3-11 Flux Map No. 21A, Power Distribution Measurements 5.1.3.15 5.1.3-12 5.1.5-1 Static Rod.Drop Thermocouple Map for.RCCA M-10 at 5.1.5.9 140 Steps Static Rod Drop Thermocouple Map for RCCA H-10 at 5.1.5.10 5.1.5-2 5 Steps 5.1.5-3 Static Rod Drop Thermocouple Map for RCCA F-12 3t 5.1.5.11 170 Steps 5.1.5-4 Static Rod Drop Thermocouple Map for RCCA F-12 at 5.1.5.12 5 Steps 5.1.5.13 5.1.5-5 Rod Position vs. Excore Detector Output for RCCA H-10 insertion 5.1.5.14 5.1.5-6 Rod Position vs. Excore Detector Output for RCCA F-12 Insertion
.NIS Detecror Locations 5.4.2.3 5.4.2-1.
5.4.6.2 5.4.6-1 N41 Detector Current vs. Incore Axial Offset Axial Offset
.ncore
- 5-4.6.3 .
5.4.6-2 N42 Detector Current vs. 5.4.6.4 5.4.6-3 N43 Detector Current vs. Incore Axial Offset 5.4.6.5 5.4.6-4 N44 Detector Current vs. Incore Axial Offset
. 5.5.3.2 5.5.3-1 Turbine Expansion Data Points -
5.6.2.3 5.6.2-1 VLPMS Vibration Signature I. . . v
LIST OF TABLES Table
'Title Page 2.4-1 Slowest, Fastest and Average P.od Crop Times 2.4.4 2.4-2 Range and Average of Additional Drops 2.4-jS 3.6-1 RTD Bypass Flows and Transport Times 3.6.2 3.7-1 Sample .'CS Flow Measurement Data 3.7.3 3.7-2 Sample RCS Flow Measurement Data 3.7.4 3.7-3 Measured RCS Flow Rates 3.7.5 3.8-1 RCS Flow Coastdown Time Response Summary 3.8.4 4.3-1 Isothermal Temperature Coefficients 4.3.3 4.4-1 RCCA Worth Measurements 4.4.3 4.5-1 Boron Endpoint and Boron Worth Data 4.5.3 4.6-1 Shutdown Margin Calculations 4.6.3 5.1.3-1 Flux Map Dara Summary 5.1.3.3 5.1.4-1 Power Coefficients Summary K) 5.1.4.3 II S.1.5-1 FAH and QT values from Thermocouple RCCA H-10 Data During I Static Drop 5.1.5.4 5.1.5-2 F6H and QT Values from Thermocouple RCCA F-12 Static Drop Data During 5.1.5.5 5.1.5-3 RCS Loop Temperatures Data During RCCA H-lO Static Drop 5.1.5.6 5.1.5-4 RCS Loop Temperatures Data During RCCA F-12 Static Drop 5.1.5.7 5.1.5-5 Static RCCA Drop Flux Map Data Summary 5.1.5.8 5.1.6-1 Data Su-mary, Special Hot Channel Factors Test 5.1.6.3 5.2.2-1 Loid Swing Test Parameters 5.2.2.2 5.2.7-1 Loss of All Offsite Power Recorder Data Summary 5.2.7.3 5.3.2-1 Pressurizer and Steam Generator Controls Response 5.3.2.3 5.4.3-1 Reactor Control System Adjustment Data 5.4.3.3 5.5.1-1 Seconary System Problems 5.5.1.2 vi I
I r 's , l, Table Title Nuclear Noise Measucement Test Sugary 5.6.3.2 5.6.3-1 Elev. 767 ft. 5.7.2 5.7-1 Radiation Survey of Reactor.Containment, Radiation Survey of Reactor.Containment, Elev. 735 ft. 5.7.3 5.7-2 Radiation Survey of Reactor Containment, Elev. 718 ft. 5.7.4 5.7-3 Radiation Survey of Reactor Containment, Elev. 692 ft. 5.7.5 5.7-4 Vii
LIST OF ABBREVIATIONS A.O. Axial Offset ARD All Rods Out
)
' BOL Beginning Of (core) Life BVPS Beaver Valley Power Station . CB Control rod Bank CRDH Control Rod Drive Xechanism DNB Departure from Nucleate Boiling EOL End Of (core) Life FCFH Full Core Flux Map F.P. Full Power FSAR Final Safety Analysis Report
- N Nuclear Enthalpy Rise Hot Channel Factor
*qN Nuclear Heat Flux Hot Channel Factor FqZ Axial Heat Flux Hot Channel Factor HFP Hot Full Power HZP Hot Zero Power ICRR Inverse Count Rate Ratio .61 Axial Flux. Difference 'hK Change in neutron population per generation M/D Moveable Detector HIN Minimum .TC Moderator Temperature Coefficient MTU Metric Ton Uranium MWD Megawatt Days NIS Nuclear Instrumentation System NSSS Nuclear Steam Supply System pcm Percent milli (10-5 Ap)
PI/L Part Length control rods ppm Parts per million PUR Pressurized Water Reactor PG Primary Grade (Water) QCFM Quarter Core Flux Map Qr Quadrant Tilt Change in reactivity RCCA Rod Cluscer Control Assembly RCP Reactor Coolant Pumps RCS Reactor Coolant Systea RPII Rod Position Indication RPS Reactor Protection System RTD Resistance Temperature Detector viii
SB Shutdown rod Bank Tavg RCS Hot and Cold.Leg Average Temperature - :- . TIC Thermocouple Tech Spec BVPS TechnicalSpecifications
- Tref Programmed Tavg based on turbine first stage pressure VLPHS ' Vibration and Loose Parts Monitoring System I
I. . . I ix
Introduction Beaver Valley Power Station, Unit 1. is the first of a two unit nuclear power generating scacion, with Unit 2 presently under construction, located in the borough of Shippingport, Pennsylvania. BVPS Is jointly owned by Duquesne Light Company, Ohio Edison Company, and Pennsylvania Power Company but operated solely by Duquesne Light Company. The NSSS Is a three loop, Westinghouse PWR with a rated capacity of 2652 NWT which incorporates the 17 x 17 fuel assembly array. The turbine generator is also Wescinghouse supplied and is rated at 923.4 MWe. On January 30, 1976, BVPS vas Issued an operating license to pernit fuel load and zero power physics testing to 1Z MP. Additional operacion to 52 FP vas permitted for some tests requiring flux maps. Initial fuel lcad was completed on February 7, 1976 and on iay 10, 1976 initial criticality was achieved. License to escalate power to 35Z FP was obtained on Xay 23, 1976 and full power operation was authorized on July :, 1976. The station was declared commercial for 502 power production on September 30, 1976. Thi3 report will present a synopsis of testing beginning with Initial core load and continuing to the 50Z FP plateau of the initial power ascension sequence where the station was declared partially commercial. Supplemental reports will cover those tests completed after partial commercial operatic., commenced until the completion of ths start-up test program. All start-up tests are listed in the table of contents. Those tests not started or completed by September 30, 1976 dll be identified in the text of this report, but comprehensive report on these tests will be withheld until they are completed.
1.0 Initial Core Loading assembly comprised The BVPS core consists of 157 fuel assemblies with each and one incore of ¶64 fuel rods, 24 rod cluster control guide thimbles, array. The guide instrumentation thimble all positioned in a 17 x 17 rod source rods,'or thimbles accommodate control rods, burnable'poison rods, by eight grids spaced thimble plugging devices. The assemblies are supported including grids at approximately equidistant along the length of the column the top and bottom assembly nozzles. the insertion of Core load commenced at 0308 hours February 4, 1976 with No significant'delays primary source assembly C-33 into core position C-l. at 0429 hours on vere encountered during the loading which was completed February 7, 1976 after approximately 73 hours. detectors (BO3 ion Just prior to initiating core load, three temporary neutron 1% introducing the first chambers) were installed in the reactor vessel. Before neutron count rate fuel assembly into the vessel, the average background the two permanently readings were taken on the temporary detectors and on tJ31 and N32. installed neutron flux detector source range detectors, was determined at Throughout core loading operations, RCS boron concentration between 2063 and regular intervals. RCS boron concentration was maintained requirement of 209L4 ppm, with an average 2070 ppm, exceeding the minimum 2000 ppm boron. Periodic reactor vessel boron samples were obtained within the vessel during fuel load. Prior to the start of core loading, samples that no obtained at preselected locations within the vessel verified stratification existed. 'I-- 1.0.1
Following the insertion of each fuel assembly into the vessel, the count rate was determined from the average of three 100 second counting intervals for each permanent source range channels and for each temporary detector. The inverse count rate ratio was calculated for each channel and a plot of TCRR vs. number of fuel assemblies loaded was maintained (Figure 1.0-1 thru 1.0-5). The response of all five source range channels was as expected throughout core loading except during loading step 92 and loading step 128. High count rates occurred when fuel assemblies were loaded into positions C-19 and K-2
.1 respectively. Locating the primary neutron source at the edge of the array apparently provided neutrons which were reflected back to the neutron detectors when a fuel assembly was loaded behind the source. High count rates or unexpected changes in the ICRR were not responsible for any loading operation delays.
The large discontinuities on the ICRR plots are the result of source or temporary detector relocation. New baseline data was established with e3ch source or detector movement. The final BVPS initial core load map is given on Figure 1.0-6. 1.0.2
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270'F Figure 1.0-6 Core I Assembly Map 1 2 3 4 6 7 8 9 10 11 12 13 14 15
<9 5
1 1 1 C26 I C41TIC:49 A I , IC51 C17-17IC_ C6 A34 C50 C3 C12 -4 B R 16P R 16P R C38 j1c B2 I A4 B8 I A3 I B28 C34 C24 C 16P IR s.12P. R 16? 1.2PSY C23 %51 B39 A38 B121 A49 B32 A29 B42 All Cll -I--' D R 16P R 16P RS 16P RL 16P R C48J C47 IB37 A42 B23 I A301 B3 IA37 IB4 IA71 B29 C1 C10 12PI16P R 16P SP 20P I16P R -16E 1 C7 B50 AS B18 A13 B47 A201B49 A16 B1I A17 B19 I_ - F R 16P R 116P IR- 16P R 116P R 16P1R 16P CS C52 A36 B33 A40 B31 A14 B14IA39 B34 A45 B26 A22 C33 C25 G 16P R 16P l16P1 R j20P R 16P SP 16P R 6PP C22 A18 B24 AIO B38 A21 843 Al B30 A47 525 A52 B44 A35 C31 _Oe9 R UP2S RS 20P R 20P RS 120P R 20P RS 2pS R B11 A33 B13 A28 C46 C2
~- J C3240 A27 8351 2 B9 A48 B35 A12 PPS R I16P SP 16? R 20PR 16?I 16P R - 16P _
C36 B17 A6 1B16 A44 B6 AB B21 A1S B10 1A41 B46 C8 K It 16P R
,- 4 1 16P R *t-I16P .4I R - 16P 4 i R I 4--16P IR -
16P I--I-RI - C13 IC1S IB15 32 IB221 A19 IB411A9- B52j A53_ B40 IC35 C44 L 12P 116P 16P 20P 16P R 16P _12PS C30 A26 JB7 A24 B5 h25 B36 A23 348 A31 C29 H R 16Pl 1 16P RS 16P R 16P I R C20 C9 B4.- ASO B27 A46 I B20 C16 C37 - L2PS? 16P IR 12P IR 16P 12P
-27 C45 C4
.IR _
16P I R A43 C26 C21 16P R T_- C42 _ P C43 C9IC14 R 9 I 7uel Assemblies 90. F Legend: A - Region 1 (2.10 w/o) R - Full length control cluster B - Rsgicz 2 (2.60 v/Q) RS - Part length control cluster C - Region 3 t3.10 v/o) P(S)S - Primary (secondary) neutron source (Crth ccrponent sArial,no.) XP - X rod burnable poison assembly X - number of rods Excore Neutron Detectors SP - Spare rod location Source Range N31, N32 Intermediate Range I35, 336 Pover Range N41, N42 N43, N44 i.0.8 ."I.
2.O CONTROL ROD TESTING 2.1 GeneraL Beaver Valley Unit No. I has 48 full length control rods and 5 part length control rods composed of a silver-indium-cadmium alloy contained in type I: 304 stainless steel tubes. The part length control rods contain the absorber material in the lower 25% of the cube length with the remainder of the space containing aluminum oxide for filler. The full length control,, rods are divided into 2 shutdown banks (SBA and SBB) and 4 control banks (CBA, CBB, CBC, CBD). Each of the 6 full length rod banks are further divided into 2 subgroups consisting of 4 rods per subgroup. The rod banks are all symmetrically located. The location of the various banks and individual RCCA's are presented in Figure 2.1-1. Full length control rod motion is obtained by the CRDK which are the magnetic jack type. Each CRD4 contains three induction coils which energize in a cyclic sequence to provide rod motion. Upon loss of power to the CRDX the rod drops to the fully inserted position. Although each shutdown and control bank may be moved individually, normal control bank motion is in a bank overlap sequence. CBA is withdrawn until it reaches a preses position near the top of the core where CBB starts moving out In sychronism with CBA. CRA stops when it reaches the top of the core vnd CHB continues to withdraw to a preset position where CBC begins to move. This overlapping sequence continues until only CBD is moving for control. Bank insertion is accomplished by a reversal of the withdrawal sequence. The subgroup of each bank alternate motion during withdrawal and insertion for a more uniform reactivity change.
- 2. A.1
The part length control rods employ reluctance motor drive and mechanical brakes which lock the part length rods in position upon loss of po'er to the drive mechanism. BVPS Technical Specification 3.1.3.6 states that all part length control rods will be fully withdrawn while the reactor is critical. This restriction ensures that adverse power shapes and rapid local power changes which may effect departure from nucleate boiling (DNB) considerations do not occur as a result of part length control rod Insertion during operation. 2.1.2 3I
Figure 2.'1-1 Control Rod
- Locations Called North A,., .
-_I
- ,, ' LOOP. '
_ . A_'.!..I R I . I I I A D P
- . i-SA Sa I -,.
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.- ,v C-- 8 PL IG M SP IP- .-
I I- SC IS _Is- - -, CB _s .II-L L A B8 _D C D K SA SB SB SP _ SA - J
.OOP H D PL PL C PL -
B I SA= SP _S SB . _SA G ) .. .. . .. . I ... A B _Q C D -B 4
.- .- . - j. L, SBI _
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- - , - . - - - 4 -
,. D AA B i.:
., -E AI 14-- 12I' 10 9
*.15 .14 13 12 11 10 9 - -8 '-7 6 '5 4 3 2 1V Function Number of Clusters ControkBsank D Control Bank C Control Bank B . 58 Control Bank A ,8 8
Shutdown Bank SB 8 Shutdown Bank SA 8 Part Length PL Spare Rod Locations SP S. . 2.1.3 S'..................
2.2 Control Rod Position Indication System Test The rod position indication systemu provides rod position indication for each full length and part length rod to the control room operator. This system also provides bank demand analog position for each control bank as generated by the rod control system by means of pulse-to-analog converters. Demand position for each subgroup of shutdown and control rods and the part-length rod bank are displayed on 3-digit step counters in the control room. The individual rod position signal is generated by the rod position detector which is a linear variable differential transformer (LVDT). The rod drive shaft serves az the core of the transformer. A signal conditioning module converts the A.C. signal from the detector to an adjustable linear D.C. analog position signal proportional to rod position. The D.C. analog signal is monitored by a rod bottom bistable which provides an alarm function when the signal drops below a preset voltage. Bank bypass bistables provide a blocking action on the rod bottom alarm when the rods of the associated bank are operated below the bypass bistable setpoint. The RPI system was calibrated at both cold and hot RCS conditions. At cold conditions, zero and span adjustments for each rod position indication LIVDT were performed at 0 and 228 steps. Zero and span adjustments for the pulse-to-analog converters digital readout were performed by manually pulsing the converters without rod motion. The rod bottom and bank bypass- bistables were also adjusted. At hot RCS conditions, zero and span adjustments for each rod position indication LVDT were performed to obtain the required position signals at 20.and 200 steps, respectively. A.C. test signal voltages required to generate equivalent position signal outputs were also determined for future recalibration purposes. 2.Z.l
Following the-hot calibration of the RPI System, a calibration check was performed in which the rod position signalloutput voltages were checked at twenty'step.withdrawal-increments for each rod.` The position sinarl voltages
- versus steps withdrawn for each rod were plotted to verify that all points were within the allowable tolerance band of,+/-.0.18 volts while maintaining the alignment of + 12 steps with the group step counter an required by Technical -
. Specification 3.1.3.2 for Position Indicator Channels. For the position signal voltages that exceeded the'allowable tolerance band, the attempt was made to
- --bing the position'signal voltage versus rod position points within the tolerance Uband by making zero and span adjustments. With the exception of control rods
- K-12, F-12. and M-lO between 100 and 160 steps, all position signal voltages were within or adjusted to be within the allowable tolerance band. Although attempts
' to adjust these position signal voltages to fall within the allowable tolerance
. band.were unsuccessful, these RPI's did indicate within the required + 12 steps
- of the associated group step.counter and were thus-considered acceptable.'
. Figure 2.2-1 shows a typical plot of RPI position signal output voltage versus rod position. There is a positive bowing of the plotted curve which was a
- characteristic exhibited by each of the RPI's. This bowing is normal for
- this-type of RPI system and is caused by the temperature effects of the hot control rod drive shaft entering the core of the'detector coils.
Mi rod 'bottom bistables and bank bypass bistable setpoints were checked to*-
--assure that they were set at 20 and 35 steps respectively.
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. 1, SiTlN A I BYnr JNTES I S.. ) I.E
2.3 Control Rod Drive Mechanism Timing Test Each full length control rod is withdrawn and inserted by.means of a Westinghouse, Model L-106A Magnetic Control Rod Drive Mechanism; Each CRDH contains an operating coil stack which consists-of a.stationary gripper coil, a movable gripper coil and lift coil. These coils are energized in the proper sequence by the rod control system slave cyclers to operate the movable gripper and stationary gripper latches which step the rod in and out. Proper coil polarities, coil currents and slave cycler operation were verified with the control rod drive shafts removed prior to core loading. After core loading all CRDH's were tested under cold shutdown conditions and again under hot conditions (547 F, 2235 psig) to verify proper timing of each rod control system slave cycler and proper operation of !ach CRDH with the control rods and drive shafts installed. Each full length rod was withdrawn and then inserted 50 steps while visicorder traces were recorded of the CRDM lift coil current, movable gripper coil current and stationary gripper coil current. The output ot a sound pickup attached to the top of the drive mechanism was also recorded on the traces obtained during cold shutdown for additional data on drive mechanism operation. The traces generated from each CRDM were evaluated for the proper sequencing of steps required for safe rod withdrawal and insertion and were found to closely agree with the expected results shown on figures 2.3-1 and 2.3-2. When withdrawing a control rod one step the stationary coil ia initially energized. The movahle coil is energized followed by the de-energizing of the stationary coil which transfers the load to the movable gripptr. The 2.3.1 I
lift coil Is energized lifting the rod one step. The stationary coil is then energized transferring the load to the stationary grippers. Finally the movable coil and then the lift coil are de-energized completing one cycle. Typical traces obtained for rod withdrawal and insertion shown on figures 2.3-3 and 2.3-4 demonstrate the proper coil sequencing. In addition to proper sequencing, the slave cyclers were able to control the current flow to the lift, movable and stationary coils such that the length of time each coil was energized and de-energized did not vary more than 0.01 seconds between each red. These times differed by no more than a few hundreds of a sec. from the typical times given in figures 2.3-1 and 2.3-2. I 2.3.2
.ln AMPERES.
I .MG AMPERES Figure 2.3-1 Typical CRDX Withdrawal Cycle SG AMPERES SOUND
- MODULATED I REDUCED CURRENT)
I II TYPICAL MAGNET POLE OPERATING
,., - 2200 PSI TIME 1 WITHDRAW CYCLE 540 OF MODEL L-106 CRDM WITHDRAW CYCLE TOTALTIME: MECHANICALTIME LIFT POLE CLOSE .185 SECONDS* - .085 SECONDS LIFT POLE OPEN .270 SECONDS .145 SECONDS MG POLE CLOSE .100 SECONDS .03l SECONDS MG POLE OPEN .100 SECONDS . . .050 SECONDS AFFECTED BY LIFT COll) - EO H SG POLE CLOSE .100 SECONDS i SO FOLE OPEN 055 SECONDS 112 SECONDS *.o8o SECOIDS TOTAL 867 SECONDS - -. 445 SECONDS
- 360 LB. LOAD - VALUE WILL VARY WMl LOAD
** AVERAGED VALUES -
_.3.3
LIFT MG Al SG AMI SOUND LIFT ON M OFF {_-- -- 7-- MG _-- MODULATED ON OFF (REDUCED CURRENT) i SG OFF ON M
-1i 1 0 SECONDS 5 - -- - *2 DEGREES
.8-.6-.
60 CYCLER CODE TtPIfrAl I~.
'a rgi.
. 'vmNLj POLE OPERATING TIME WNsERT CYCLE P PSI J.ODEL L-106 CRDM INSERT CYCLE 540°F LIFT POLE CLOSE _ h;ECHANICALTIA
- LIFT POLE OPEN .130 SECONDS
.165 SECONDS' .040 SECMDsS MG POLE CLOSE .070 SECONDS .100 SECONDS MG PO EOPEN .030 SECONDS
.07 SECODS SG POLE CLOSE .050 SECONDS SG POI Ec. N .130 SECONDS O110 SECONDS .060 SECONDS NO-A .110 SECONDS .080 SEOONOS JOT,.L .715 SECONDS .360 SEcoNDS
- 360 L.. LOAD - VALUE
** A'I'RAGED VALUES WILL VARY WIlTH LOAD
-2.3.4 2.3.4
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0 - '~I - 1_ 2.4 Control Rod Drop Time Measurement Trst The drop times for all full length control rods were measured to assure the maximum rod drop time allowed in the Technical Specifications was nct exceeded. Drop time measurements were performed for both.cold and hot RCS conditions with no flow and'full flow. The control rods were'dropped by first withdrawing a rod bank and then removing the movable gripper fuse and the stationary gripper fuse for each rod individually allowing the rod to drop into the core as in a normal trip. The
-rod insertion speed was slowed near the end of travel by the dashpot, a recked down section of the guide tube, and the entire assembly displayed a
- .slight bounce when bottoming.
The rod drop time measurement is obtained by recording the signals for the stationary gripper coil voltage, the RPI detector primary coil voltages, and a 60 Hz timing trace on a visicorder. Refer to Figure 2.4-1 for a typical
'rod drop trace. The first indication of voltage decay when the stationary I
gripper coil fuse is removed determines the initiation of the rod drop. A voltage is induced upon the RPI detector primary coil as the rod drops and is a function of the velocity zf the rod. The RPI coil-voltage increases with the velocity of-the rod until the rod reaches its terminal velocity. The
.rod is sloved by the dashpot resulting in a decrease in the voltage induced in the RPI coil. The RPI detector voltage is at a min.imum as the rod reaches the bottom of the dashpot then increases slightly as the rod bounces.
Each control rod was dropped a single time at each set of RCS condition. At cold conditions with no flow and at hot conditions with full flow, the slowest 2.4.1 I
and fastest rods were dropped ten additional times. All control rod drop times for each set of conditions were well below the HFF limit of 2.2 K seconds from the initiation of the rod drop to dashpot entry as required - by B'JPS Technical Specification 3.1.3.3. Drop times for B-6 and K-14 were slower than the remainder of the rods where PCS flow was Involved. This is 8 characteristic of 3-loop plants due to slightly different behavior at these core locations caused by hydraulic forces. J-7 also displayed slower drop times than the remainder of the rods. especially under influence of RCS flow. This caused no concern other than this rod should-be watched in future rod drop testing. The measured drop times for each control rod
-are recorded on Figure 2.4-2 for the cold no flow condition, Figure 2.4-3 for the cold full flow condition, Figure 2.4-4 for the hot no flow condition, and Figure 2.4-5 for the hot full flow condition. The slowest, fastest, and average drop time for.each set of conditions are presented In Table 2.4-1.
Table 2.4-1 presents both the drop times to dashpot entry and the drop times to bottom of dashpot. The average drop times to dashpot entry for each set of conditions ranged from 35.5% below the HFF limit of 2.2 seconds for the CFF condition to 49.1% below this limit for the IMF condition. The average drop time for the HFF condition, the condition at which the rods are most likely to be dropped to shutdown the reactor, was 41.42 below the 2.2 second limit. The average rod drop times were observed to be ighely faster than those of preceding 3 -loop plants. This fact may be due to the 17 x 17 fuel assembly design employed at BVPS. 2.4.2
The range and average of the ten additional drops for the slowest and fastest
- i rods for the CYF and 1FF conditions arc presented in Table 2.4-2. The range of the additional drops indicates that individual drop times may diffe: by 0.03 seconds but vill be within 0.02-seconds of the average for that rod.
I-i l 1% . . . i
. ..... . t i
I I, 2.4.3
Table 2.4-1 Slowest, Fastest and Average Rod Drop Times Rod Drop Times To Dashpot Entry RCS Slowest Rod Fastest Rod Average All Rods Condition (Tim;Rod) (Tile;R*4) (Time) CNF 1.22 sec; L-5 1.15 sec; G-7, G-9, J-9 1.18 sec. CFF 1.48 sec; K-14 1.38 sec; H-12 1.42 sec. IMF 1.14 sec; J-7, L-5, P-8 1.10 sec; B-10, E-1l. G-3 1.12 sec. G-7, J-9, f-10, IIFF 1.36 see; B-6, K-14 1.26 sec; P-3 1.29 sec. Total Rod Drop Time To Bottom Of Dashpot RCS l Slovest Rod Fastest Rod Average All Rods Condition l(Time;Rod) Mm~e;Rod) (Time) CIF 1.i2 sec; L-5 1.64 sec; F-8, GC7, G-9 1.68 sec. CFF 2.11 sec; K-14 1.97 sec; G-7 2.01 sec. HMF 1.57 sec; J-7, M-6 1.50 sec; C-3 1.54 sec. HFF 1.86 sec; B-6, K-14 1.73 sec; P-8 1.76 sec. 2.4.4
Table 2.4-2 Range .nd Average I . of Additional Drops
. 4 I I
'i ,. , , .
i I , . .
., ,. I I II I . ,
- RCS Condtiton RdD°
-CSCdtinRod Drop Time To Dasbpot Entry. I~d Drop Time To Bot'ton O Dahsot and Rod Range.0Afae nta Range Of Addition Drops Additional Drops Average CNP Slovest Rod (L-5) 1.22sec 1.21-1.22 sec 1.22sec i.72sec 1.71-1.73 sec 1.72sec CF Fastest Rod (G-9) 1.15sec 1.15-1.17 sec 1.16sec 1;64sec 1.64-1.67 sec 1.66sec HFF Slowest Rod (K-14 1.36sec 1.34-1.36 see 1.35sec 1.86s:c 1.84-1.86 Sec 1.85sec HFF Fastest Rod 04A) l.Z6sec' 1.26-1.28 sec L.276ec 1.73secl 1.73-1.75 sec 1.74se I
I 7. I I _. 2.4.5
Fasr Ca,, jty'p.pc I vo g*g 4*b*hI .1 II *.
** ,I I .) , 5 , ,l* i 2 ' *II § .,
e . I I . , In " *A * . I
' ' I
; * *, :'-,,,. . !.f ,
*I
.. 1
.' .1,. I
.~~~ :,; I~
rCc, 451 Cr s
'C4I KA Cce cc ..
*.A 4.*
4* C
.C
I Figure 2.4-2 Rod Drop Times Cold No Flow 6 R P N M L i. Q~-Op 3 H G F E D C c 2 1 Rod AArop Time to Dashpot Entry - in Seco:.IS Rod Drop Time to Rod Bottom - in LL-J I Seconds
I Figure 2.4-3 Rod Drop Time<J Cold Full Flow 4 900
.1 4 , 1.42
~-. A LOOP I .
R 1,42 1.42 1.43 2.01 2.00 1'.95
- I - S. -4. 4-4-4 2.00 1
- 9!
1.40 1. 4C
- I-S. 6. 44-4-4-4-4 2.00 1.95 N 1.38 1.42 1-
- 42 I-2.02 1.42 1.98 1.99 I2.0; 2.02
.b.2 11.4S .
. 200 2 207 2 _
K<L 1.48 2.11 1.40 1.99
- 1. 4C 2.OC 1.41 2.00 1.430 2.02-1.43 1.42 I k K I aI ---. - - - . - - - . 4- 2.02-2.02 1.41 1 1.42 1.481 1.40
- _ 1.99 2.0C . Z.09 1.99 LOOP O 2.01 1.40 1.4 1.96 1.39 1.42 2.00 1 J 1.41 _
1 .4 2.02 1800 _- 20C _ .99 1.97 _ 2. -,G
. , ^
1.42 1.40 1.40 1.41 1.44 -.41 2.00 1.9% 1.98 2.00 2.03 L.99
. 1.41 LI_ I- 1.401 1 F
--- . -- -I- -4 I -E....................
1.40
.'L 1.4: 1.44 1.40 2.00 2.0: 2.02 1.99 4I D
- 1. 4-4-1-4 1.42 1.39 2.00 1401 1.40 1.99 1."
.4-I -
. -I
.8C I 2.02 2 02 . 2.10 I I I I.. 4.el , I - - A IIIZI L.j LOOP 15 14 13 12 11 10 9 8 6 5 4 3 2 XX Rod Drop Time to Dashpct Entry - ir. Seconds Rod Drop Time tc Rol Bottoma - in Secondc 2.4.8
E Figure 2.4-4 Rod Drop Times Hoc No Flov 3P I-.90'
'i A R
oal2 1
. I L.LL' 1.*
L.11I I I ,,,
.1 ,
I I'I 4-4-4. 1.12 1.15 1.53 1.5 , 1.10 1.10 1.53 j 1.52 44I, 1.5f 0
. .52 11ll L. 54
.5_
1.13 L.57 a4 1.13 s1155 he
-M P
N 71.121
'W Vs., '11 1.53
;11
- 11
.11;53
- 1.53 1.12 ls55
, L _._.. J_
- J
- 1. 51 I
1.141 L.15 L.55 56 1.15 1.54 12 1 1.12 1.53 I1ek1l
-1
_- .1 .10 1.14 ___ .l 1.53 1.52 .S7 1.51
- .OOPj 11 130. 1
.53 - ;11.35 1 53 i80
-- 1.53 - l.;_10 _1 .53 5 11.50 _ !
-S
-G IL ..- 531.12
]
II131 1.11 1.'..
. 1.1) 1.12 11.1 1.51 1*55 1.52 1.54 1.51 1.54 1.5S 1.54
_1 _ 1.11 1.52 1.53 E
-- I I J _ . _ _ _1 1.11 1.12 1.11 1.15 1.54 1. 54 1.54 1.53 D
.11 1
- 1]
1,13 .11 1.-5 f.52 _Z K\N - -- 1.53 4 1.10 1 52 1.15 1.11 1.54
.52 J-1.11 1.11 1.54
- I -
-I PHe I 8 A
I a12701I - OO
.C Is 14 13 12 11 10 9 8 6 5 A 4 3 2 I;
.. i l l Rod t Drop Time to Dashpot Entry - in c conds I 1o IDrop. Tire to Rod Bottom - in Seconds 2.4.9
Figure 2.4-5 4 Rod Drop TIme; Hoe Full Flow A,3 9go lI LOC 3P A _ l I I . R 1.28 1. L.31 1.74 1.72 1.7~ p 11 9i 4-4-4-4-4 1.74 L.301.29 I.7
.q999-4 L.30 L.77 1.29 1.76 7-, N 1.2 1.28 L.77 1.76 -- 4-9 4-1.28 I.31 1.7 1.75 1.76 1.79 M II 1.21I2
_ 1....7.. 1 1 jI 1 1-,1.301 - _I i 1.36 1.86 1.27 1.14 L.29
,.77 1.28 1.76 1:301 1.77 j1.78 1.29 1.75 1
1.29
;1.77 t K
. .7 . 1.77 1.29 1.27 1.33 1.28 v- i 1.75 1.74 1.82 1.75 0- 1- .287 LOOP N I.74 281 _ - 1-2 180°
-H 1.77 11.74 7618 1.28 1.27 1.27 1.29
_~~ 1.75 1.75 1.761 _ 1.75 G~- 1.29 1.2Z 1.27 1.28 1.28 1.30 1.29 1.76 1.7 L. 74 1.74 1.76 1.77 1.76 I I I I I 1.25 F 71.4I 1.76 1.2f 1.28 1.29 L 21 1.7f 1.76 1.76 1.7' t-- 9 1.2'3 1.29 4-9-44 1.-6 1.76
- I.-- I
";Zz C 4 4-1 1 .301 1.30 1.36 a 1.75 § 0 l 1.79 1.86 I1-; 4X -- l l
- B
-~A I
Is 14 13 12 11 109 8 6 5 4 3 2 1 FZ] Rod Drop Time to Dashpot Entry - in Seconds LLJ Rod Drop Time to Rod Bottom - in Seconds 2.4.10 I ....
2.5 Part Length Control Rod 'echanism Brake Test 11 -1 The brake mechanism of the part length control rods lock upon loss of power to prevent rod movement. The part length rod mechanism main transformer supply voltage was varied, using a powerstat, to engage and disengage the rod brake. The proper operation of this mechanism was verified by recording'the sound level at the rod drive mechanism housing and by aeasuring the power supply current to the Part Length Rod Control Cabinet. The recorded response clearly indicated a feedback voltage on the powerstat' immediately followed by an increased noise level a,- the rod mechanism housing, indicating proper operation of the brake mechanism. The attached curve (Figure 2.5-1) is characteristic of the noted response. BVPS Technical Specifications prohibit the use of the Part Length Control Rods during power operation. The Part Length Control Rods are always fully wihtdrawn and deenergized in modes 1 and 2 with Keff . 99. 2.5.1 I ..
FV-- * - -.. .-* - - . I I.- Sound Lavel AN.-oh**Wa J4n, A I 7.2 2.0 amps 2.2 2.0 amp Current disengage
- 1. amps II l1.I amF
' /i
- 1. @
Ong carc PI14 0,Ie i 1m XPr Ta
?0 0)a 042O Q C O ~rttj
(. 10 rtc 1.
I 2.6 Rod Control Syste'm Test Prior to initial criticality the proper operation of the rod direction, position, speed control and indication.functions of the full length rod control system In the manual mode was verified. Preoperational testing had previously been performed to verify the proper operation'of the control rod drive control system without actual operation of the rods. The control and logic circuitry were also systematically checked out prior to core loading. Included in these tests were checks of the motor-generator-sets,-reactor trip breakers, bypass breakers. power supply voltages, coil currents and polarities,
- DC hold cabinet voltage measurements, and rod speed and sequence'controls.
All controls used for normal system operations.are found on the main control
; board. The bank selector switch enables the.selection of either manual or automatic control modes utilizing a preset bank overlap and~also has the capacity to operate selected individual banks. The IN-OUT lamps indicate'thac rod motion has been requested by the rod control system. The IJ-11OLD-OUT lever is used for manual operation of the shutdown and control banks.
The position demand signal for each group of rods is displayed by a 3 digit step counter. Rod position is also observed on the individual rod position indicators (see Section 2.2). Bank stepping.rate is adjustable for operation in the manual mode. During rod motion the stepping rate is displayed on a control board indicator. Control bank overlap is preset by means of 6 digital thumbwheel switches. The full length rod banks were tested at reactor coolant system (RCS) conditions of 547F and 2235 psig, with shutdown margin assured by a RCS boron concentration 2.6.1
of > 2000 ppm. Each of the 6 RCCA full length banks were first tested for correct operation and indication of the I--OUT direction lights, bank selector switch, the IN-HOLD-OUr lever, the grcup step counters, the rod position Indicators, and the rod speed indicator. The individual rod position Indicators were compared to the group step counters to assure that they were in proper agreement. The control bank overlap was checked using a preset overlap which verified operation yet kept rod vithdrawal to a dinitma. The rod controls and indicators were all observed to operate as required. The individual rod position Indicators read within + 12 steps of their respective group step counters as required by BVPS Technical Specification. The control bank overlap for :wnual rod withdrawal and Insertion also performed satisfactorily. Testing did establish, however that the maximum manual rod speed was 66 steps per minute which was less than the expected 72 steps per minute. This stepping rate was within the acceptable tolerance, for the manual mode, of 72 +0, -10 steps/min. 2.6.2
3.0 POST CORE LOAD, PRECRITICAL TESTING 3.1 General -' II. This section describes the testing that was performed following core loading and prior to initial criticality in addition to control rod testing. With the station in a cold shutdown condition the incore movable detector system was operationally checked by fully exercising each drive assembly using a dummy detector drive cable. The Vibration and Loose Parts Monitoring System alarm setpoints and operational checks were performed and RCS and reactor core vibration vs. frequency spectrum traces for various RCP operating configurations were obtained for cold shutdown conditions (and later at hot standby conditions). During station heatup incore TC/RTD cross calibration data was obtained at various temperature plateaus. With the station being maintained in the hot standby condition the remainder of the precritical testing was completed. The pressurizer manually operated 2 continuous spray flow valves were adjusted to maintain the spray lines and nozzles above a minimum temperature to reduce thermal stress during spray valve operation. Pressurizer spray and heater capabilities were tested to verify the effectiveness of the pressurizer spray to reduce and pressurizer heaters to increase pressurizer pressure. Pressurizer alarm setpoints and heater capacities were also verified. Hot and cold leg RTD bypass flows were measured and the RTD loop transport times calculated to insure they were compatible with the assumptions made in the accident analysis. RCS flow with various combinations of operating RCPs and RCS loop stop valve arrangements was measured. The RCS flows were determined based on measured RCP input power and pump performance data supplied by the manufacturer. Finally RCS loop flow coastdown was measured for various RCP trip combinations and the RCS low flow reactor trip time delays were obtained. 3.1.1
During hot standby, baseline data for the alignment of temperature control l systems and steam and feedwater flow calibraticr were also obtained. Just
)
prior to initial criticality an operational check and alignment of the ItS was made (See Section 5.4 and 5.6). 3.1.2 i
3.2 Incore Movable Detector System The incore movable detector system contains five movable neutron flux detectors (fission chambers). The system is designed to scan the active length of 50 selected fuel assemblies and provide recordings of the relative three dimensional neutron flux distribution in the reactor core. The detectors are remotely positioned within the selected fuel assembly instrumentation guide tube by retractable drive cables. The position selection system for each of the Incore detectors contains i 5 and a 10 path rotary transfer device. During normal system operation'each'detector is aligned to its corresponding 10 path rotary transfer device. However, by mean; of the 5 path transfer device each detector can alsobe routed through several, other alternate insertion paths to various core positions. For calibration purposes, each detector can be routed one at a time into a common fuel assembly providing correlation of each of the five detector output signals. Figure 3.2-1 shows the location of the flux thimbles within the core. The incore movable detector system transmits the neutron detector output signals to the PRODAC-250 station computer for rapid flux Capping data acquisition for later reduction off site. The reduction program (Westinghouse "Incore 2.9" computer program) consists of calculations to provide parpmeters such as axial peaking factors, nuclear enthalpy hot channel factors, nuclear heat flux hot channel factor, relative assembly power, radial tilt and axial offset ratios. Proper operation of the system weas verified during cold shutdown. Each detector drive assembly was fully exercised in the insert-withdrawal mode using a dummy detector drive cable to check drive equipment, safety and withdrawal limit switches, five and ten path transfer devices, readout and control equipment along with establishing preliminary top and bottom of core setpoints. 3.2.1
Following fuel loading and station heatup to 547F the real detector drive cables we:e installed and proper power supply input checked. The Incore gas purge system and leak detector system was also checked for proper operation. All checkouts were completed satisfactorily. So problems were encountered with the exception of the E detector drive assembly. While testing this detector the dumy drive cable could not be fully inserted Into numerous positiona. A defective drive clutch resetting I plunger was subsequently found. A new plunger was installed and no further problems encountered. Final top of core setpoints were obtained during tow Power Physics Testing just prior to initial flux mapping data collection. The serpoints were determined by noting the digital readout of the detector position at the grid induced flux depressions and correlating these to the distances from the grid locations to the top of the fuel taken fron the nuclear design report. I 3.2.2
Core Dlstributlon of Flux Thimbles A ~~A
' I1 1I E A D C .
B
- I -
B I - t -9 9- I C 4 4-
- 1- - . - -9, B
- C C A . C _ B:. D
-I I I i -I - - .. A I. E
-I --- - - - I -
B C B D B C B F
- C
- H E A D B D A C
-- D - B ~-CP . -. D . BE -
- J C
.I K
C _ I AI I I B IE I B I A
- - L B
I- I 4I 444-4 D E A I -
- x
_ I _ _ I-I I. - I . I . _- p I ---1 'I C
-I I -.
-I
-1
-- I-- I I _4 R
_ _ . 9 ,1 1 2 3 4 5 6 7 8 9g ' ,,- 10 - - 11 - - 13 14 15 A - Flux Thimble Detector A - E - Flux Thimble Detector E B - Flux Thimble Detector B CP - CalibrationFlux Thimble C -'Flux Thimble Detector C (Co-On Path) D - Flux Thimble Detector D: I.,. .
- 3.2.3
3.3 Incore TC/RTD Cross Calibration Fifty-one chromel-alumel Incore thermocouples are provided (Figure 3.3-1) to measure the RCS temperature at the exit of selected fuel assemblies. Primary readout of the incore thermocouples is provided by the station computer, with backup readout by a precision indicator with manual point selection in the control room. Temperature distribution maps and relative power distribution maps may be obtained using the Incore thermocouples and the station computer. RCS temperature indication and AT/Tavg control and protection signals are provided from 24 RTD's, There are 3 RTD's on each of the cold and hot leg manifolds of the three loops. Two of these 3 RTD's on each manifold provide temperature Inputs to the AT/Tavg control and protection systems. The remaining RTD on each manifold is in installed spare. These 18 RTD's are narrow range units with temperature indication from 530F to 630F. An additional RTD is installed in each hot and cold leg to provide wide range temperature indication (0-700F) of the RCS from ambient condition through normal operating conditions. During normal operation of the RCS the AT circuitry must measure small differences in large temperature values. To accurately measure these differences the temp-erature sensors must first be normalized co each other at some point within the normal operating range. To accomplish the normalization. TC/RTD temperature data was recorded at 150F. 250F, 350F, 450F, 525F, and
' 7F during RCS heatup.
The "true temperature" of each temperature plateau was calculated as the average of the narrow range RTD temperatures. Isothermal correction factors were then calculated to each TC and RID. These correction factors were used for calibration of the low level amplifiers for each RTD channel and the computer input modules for each incore thermocouple. Correction factors could not be applied to the 3.3.1 I
control room multipoint indicator since one instrument is used to readout all
.f the thermocouples.'
The TC/RTD cross calibration was performed twice, once during hot functional testing and again after core loading.' During the first performance of this test, several of the wide range RTD temperature readings and various computer printout and panel'indicator readouts for the incore thermocouples did not meet the acceptance criteria of + 2F of the calculated true RCS temperature. The wide-range .RTD readings ranged from 3-4 degrees lower than the narrow range RTD's at all the temperature plateaus."' In one instance the computer power supplies for'calculating the thermocouple'temperatures were found to be drifting which resulted in erroneous thermocouple printouts. The incore thermocouple reference junction box temperatures wire monitored after hot functional testing because they were cycling more than specified by the manufacturer. One reference junction box was found to be missing a sheet of insulation. Two incore thermo-coupleswere not installed and one failed during the test. One RID appeared to have failed during the test but a faulty connection was later found to be the problem. A second TCIRTD cross calibration test w'a performed during station heatup after , core-loading. In this retest all narrow range RTD's were within the required
+ 2F of the RCS true temperature.
With the following exceptions the incore thermocouple data at each temperature plateau was within + 2F of the RCS true temperature:
- 1. The 150PFplateau data was considered unusable because of wide disagreement with the true temperature probably due to cold water infection from the residual heat removal system.
3.3.2
- 2. One TC computer readout O.llF out of specification low at the 250F plateau.
- 3. One TC panel readout 1.33F out of specification high ct the 547F plateau.
- 4. One thermocouple failed at the 35OF plateau but was later corrected.
The performance of the thermocouples was Considered acceptable because of the small magnitude of the deviatfons. As in the original performance of this test, the wide range RTD readings were again lowet than the narrow range RTD's. Although the wide range RTDT s generally were not within the required + 2F of RCS true temperature, their performance was considered adequate for the following reasons:
- 1. Their sole control functions are loop isolation stop valve interlocks and indications.
- 2. The maximum deviation between wide range RTD's is 2.67F (at the 350F plateau).
- 3. The first performance of the test also produced low readings giving some confidence in their repeatability.
- 4. The offsets can be compensated for by appropriate scalings applied to the low level amplifiers.
Correction factors for the low level amplifiers for each narrow and wide range RTD's were calculated and applied. Two uninstalled spare narrow range and two wide range RTD's were returned to the manufacturer for calibration. New resistance verses temperature tables were generated. Comparison of the new to old tables indicated that the narrow range RTD's were indicating approximately
.F higher than the actual RCS temperature. With this situation the margin between temperature dependant safety limits (DNB and F,
) as determined by the overtemperature and overpower AT setpoints was slightly increased. The error in RCS temperature has minimal effect on the core physics parameters during startup.
3.3.3 I , -
FIGURE 3.3-1 Dis:ribution of Incore Thermocouples II
. 2 3 4 9 10 'I1 12 96 13 14 15 A
B Li
~
TT TkdiV0 _ - r -- - I 7ITJ D a E - - -- J_, I I T--1T ., _ I,.. F r T -I T T- -T - TI
-TTjfT T- LOOP I K T T T T L __I_ I _- T - _IT ' T- T IV--
T T T I - t t t 4. 4. - 4. 41. IL T T T T
- I -l --
T I - I - *4.-- I 4. -4 P TI T R LOOP A . . . . _ ___j fIndicates Incore thermocouple location. 3.3.4 Al i . _ - - - -...
3.4 Pressurizer Continuous Spray Flow Verification The manual continuous spray flow valves are installed in parallel with each pressurizer power operated spray valve. Their function Is to supply the minimum amount of bypass spray flow necessary to warm the pressurizer spray lines and nozzles in order to limit thermal stresses during operation of the mal> spray valves. A preliminary setting of the continuous spray valves was performed during hot functional testing, however, due to the low differential pressure across the empty reactor vessel and lack of insulation on part of the spray lines, a sufficient bypass flow to clear the spray line low temperature alarms could not be obtained. Following ccre loading and the heatup for Initial criticality the *alves were again adjusted to provide sufficient flow to maintain the spray II lines above the low temperature alarm setpoint of 500 F and within 200 F of the pressurizer vapor space temperature. Final spray line temperatures were 510 F and 505 F for loops A and C respectively. When compared to the pressurizer vapor space temperature of 652.1 F the resulting temperature differentials were 142.1 F and 147.1 F for loops A and C respectively. 3.4.1
3.5 Pressurizer Spray and Heater Capability Pressure control of the RCS is accomplished by maintaining liquid and vapor in equilibrium inside the pressurizer. Automatic activation of a spray valve tnat is fed by the RCS cold legs and electric immersion heaters within the pressurizer control the saturation temperature thereby controlling the RCS pressure. The ability of the pressurizer spray valve and heaters to control pressure transients was determined prior to initial criticality. With the core Installed and the RCS at 2235 psig and 547F the pressurizer spray valves were fully opened to initiate a RCS pressure reduction. A chart recorder was used to monitor pressurizer pressure versus time. The pressurizer heaters were deenergized during this pressure transient so that only :he effect of the spray valves was recorded. The transient was terminated at approximately 2000 psig. The RCS was then returned to the original operating conditions. The heaters were then energized and the system pressurized to approximately 2300 psig. Again the chart recorder monitored pressure versus time while the spray valves remained shut in manual control. Upon completion of the transient the heaters were deenezgized and the RCS returned to normal operating pressure. During the pressurizer spray capability check it was found that the activation of the spray valves reduced pressure at the rate of 3.03 psi/sec. This compares favorably with the W nominal response of figure 3.5-1. The heater effectiveness was determined to be 0.2425 psi/sec. which fell within the lover limit of figure 3.5-2. The control systems for the pressurizer were tested during the 3.5.1
hot functional phase of the test program. Included in the hoc functional testing were the power operated relief valve3, the heater outputs, and the pressure and level controls, alarms, and the associated Interlocks. 3 .5.2
-C- ~Q... .. .
NOMINAL PRESSURM RESPONSE TO OPENING OF BOIT PRPSSUaZER SPRAY; VALVES (with allowable deviation) 2250 I a Si/sec.- 0.p 2200 2.8 pisi/aec. Cn W 2150 i i i I U, pi 2100 2050
.. --. Aj .
7.. -:
, 2000 4 10P 20 40 60 0#IP .
80 100 120 140 10 LI' 01.91 TIME (seconds) 0?W N'I m A I t__ '-S I
I NOXNAL PRESSURE RESPONSE TO ACTUATION OF ALL PRESSURIZER EWERS (with allowable deviation)
...... ._ . - '_, . _,..... -. 45 psi/sec.
2350 ....
- ._ _ _ . 4___:
35 psi/sec. 44 go
.... . . * /ec.
D. 2330 -- - *. * - ;.: .5; Wj
.-.-- ........ .. Na 2310 ** '-
U) .; '..j-. - ._ - -'C. ,. . . -- ---- p/Respons. 1-4
* , l - * ! * -:i . . .......... , . I......
. 2r Ws *- -*.--. t fA 2290 _~~ -.
t_7f~~~ _ ~
... ,)t -; :..:L;-! I 2270
-7}Y -- , _ , .. _ 0~bii.^
2250
*1 0 40 80 120 160 N.*
200 240
- ft TIHE (seconds) ft' 0b
- A (0' I.
3.6 RTD Bypass Loop Flow Verification RTD manifolds are required for monitoring RCS hot leg and cold leg temperature by the AT/Tavg circuitry of the RPS. Each RTD manifold recei-es a relatively small amount of reactor coolant flow from the RCP discharge (cold leg) or the reactor vessel outlet (hot leg). The flows are then combined and discharged into the RCP suction. RTD's in each bypass loop sense RCS hot leg and cold leg temperatures and provide voltage signals for the RPS. - The bypass loop flow is crucial to adequate AT and Tavg reactor protection. An abnormal delay in coolant flow through the bypass loop to the RTD's could reduce the safety margin associated with the RPS response time. Transport times were calculated for hot leg and cold leg bypass loop flows from "as built" piping measurements and flow tests conducted at hot standby conditions prior to intial criticality. Low flow alarm setpoints were established at 90% bypass flow at these conditions. In calculating the transport times for the hot leg and cold leg bypass flows a correction factor was employed since the sum of the individual isolated bypass flowa was greater than the total combined bypats flow. The measured bypass flow rates and associated transport times are reported in Table 3.6-1. All transport times were within the 1.0 second time limit required. The flow rates are higher and the transport times are lower for the cold leg In each case because the total head for the cold leg bypass flow is considerably higher than the head of the hot leg bypass flow due to the location of the bypass piping connection on the RCS cold leg. 3.6.1
I Table 3.6-1 i RTD :ypass Flows and Transport Times I I Flow Rate (GPKj Transport Time (Sec) Actual Min. Req'd Actual Max. Allowed I IA Loop Hot Leg 113 e5 .84 1.0 Cold Leg 175 49 .28 1.0 1B Loop Hot Leg 113 100 .88 1.0 Cold Leg 162 48 .30 1.0 lC Loop Hot Leg 110 96 .87 1.0 Cold Leg 165 49 .30 1.0 I i I ii i 3.6.2 II
3.7 Reactor Coolant System Flow Measurement RCS flow indication is obtained by the measurement of differential pressure across the elbow of the coolant system piping Ilocated between the steam generator and the RCP in each loop.' Tihere arei3 differential pressure transmitters per loop-each having
'an associated indication in the main control room in percent of full flow from 0 to 120. After core loading and prior to initial criticality, Cata was obtained to interrelate loop elbow differential pressure to actual RCS flow and to compare these flow rates to Technical Specification requirements.
Actual RCS flow was determined based on the RCPs performance data of pump input power vs. flow supplied by the manufacturer and measured RCP input power. Data was obtained for various RCP and RCS loor configurations. Using the measured pump input power and an iterative process, a flow rate was derived that best described the shape of the pump input power vs. flow curve of the manufacturer's performance data. In addition to pump input power, loop temperatures (to normalize water density to manufacturer's data), p..z;p speed and elbow tap differential pressure were also recorded. The elbow tap differential pressures recorded provide a means of correlating actual RCS flow rates to measured elbow tap differential pressures and corresponding RCS flow indication in the main control room. Only one transmitter output was recorded for each loop. Sample data is given on Table 3.7-1 and 3.7-2. Data for each measured operating configuration was recorded at one minute intervals for ten minutes. 3.7.1 ..K)
The Technical Specifications for RCS total flow rate require at least 265,500 gpm for 3 loop operation and at least 189,000 gpn and 187,600 gpm for 2 loop operation, loop stop valves open and stop valves closed respectively. The RCS flow rates obtained are sumarized on Table 3.7-3. All coabinations of loop configurations tested provided flow rates in excess of Technical Specification requirements. 3.7.2
_( (.. y "I IC Lov. Under Test Date f41A&3/76 Elbov Ta_ nriNot Loop Leg I..Cold Loop Lg RC; Elbow~' DVH - ,D/P rBidge I Temp Configurition ILit 1 Time
' D/P. (VOLTS) (PSI) WREf (OHMS) (F) 'RTD (OHHS) (F) Y-n t cell' 1 - -
0 -F TP-436-1 3.557 11.5384 TE-431D 426.8 564.41 TE-431C 424.8- - 559.56 ALLi, No0.E fpu4 l1L752 TP-436-I 3.546 11.448i TE-431D 426.9 564.77 TE-431C 424.9 559.80 'ALL NONE f2pw 1753 TP-436-1 3.567 11.5836 TE-431D 426.9 564.71 -T-431C-j, 426.9 565.51 AJLL jOE NE fpO 1754 . TP-436-1 3.581 11.6468 TE-6i1 426.8 564.46 jE-4iC 426.9 565.51- ALL NONE w 1755 TP-436-i 3.532 11.4257 Te-411D 426.9 564.71 TE- 4 3I 426.9 565.51 fpw - 1756 TP_436-1 3.566 11.5791 TE-431D 426.9 564.71. TE-43iC 426.8 565.27 ALL NONE fpw 11757 TP-436-1 3.541 11.466 TE-431D 427.0 564.95 T
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_ fpw 7 15 TP-436-1 3.548 11.4974 TE-431D 427.2 565.451 TE-431C 425.2 561.33 ALL NONE fip -' 1760
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TP-436-1 3.887 13.027f TE-431D 425.7 561.78 TE-431C 423.6 557.> 1A.AC -IB fpW 1821 TP-436-1 3.897 - 13.0727 TE-4310 425.6'. 561.54 TE-431C 423.5 557.15 - ,1C iB fpw 1822 TP-436-1 -3.877 i2,982 TE-431Dl 425.5 561.05 TE-431C 423.5 557.15 1A.1C 13 fpw 1823 5 cn P.- CIIm TP-436-1 3.917 13.163t LE_31)D 425.5 561.05 TE-431C 423.4 536.90 ~0 lA.1C 18 fpv 1824 Po o-TP-436-1 3.910 I3.1314 TE-43iD 426.8 564.46 TE-431C 424.8 r" m 562.80 1B.1BC -1A fpW 1859 TP-436-1 3.856 12.8871 TE-4311)1 I
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Ii k- J IC LOOP UWDER TEST Date 4/23/76 Loop Reactor Coolant PumP Input Pover Under Speed Line Volta e CT Phase Amps Power Test IRPH) Ainitial (KWeTie
- - A-B C-B A B ____ ()Inta rC 1_C 1200. 123.8 124.8 3.04 3.04 3.02 5006.4 JF 1755 IC 1200 123.8 124.8 3.05 3.05 3.02 4989.6 .JF 1756 IC 1200 123.9 124.8 3.05 3.05 3.02 5006.4 _ JF 1757
_c 1200 123.9 124.8 3.05 3.05 3.02 5006.4 JF 1758
.C 1200 123.8 124.7 3.05 3.05 3.02 5006.4 JF 1759 IC 1200 123.8 124.7 3.05 3.05 3.02 5006.4 J. 1759 IC 1200 123.8 124.5 3.05 3.05 3.02 5006.4 JF 1760 1C 1200 123.8 -124.5 j.05 3.05 3.02 5006.4 JF 1800 IC i2OO 123.8 124.5 3.05 3.05 3.02 5006.4 JF 1801 1200 123.8 124.6 3.04 3.04 3.02 5006.4 JY 1802
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Z 1200 124.2 125.0 2.96 2.96 2.94 4905.6 JF 1815 - IC 1200 124.1
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IC 1200 124.2 .125.0 2.97 2.97 2.94 4922.4 JF 1818 IC 1200 124.0 125.0 2.97 2.97 2.95 4922.4 JF 1819 iC 1200 124.2 125.0 2.97 2.97 2.95 4922.4 JF 1820 IC 1200 124.2 125.0 2.97 2.97 2.95 4922.4 JF 1821 Ca Wlu a, IC 1200 124.1 175.0 2.97 2.97 .2.94 4939.2 JF 1822 " . IC 1200 124.1 1^5.0 2.96 2.96 2.95 4922.4 JF 1823 X :s 1 n 0 N I_IC
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Table 3.7-3 Xeasured RCS Flow Rates THREE LOOP FLOW. i . . T . PUMP CONE IG;UKArIUN i.ND1VIDuAlL TOTA&"L CCOwLA#%%*T TEST TEST ISOLATED R UN PUWIP FLOW FLOW RATE SECT._ RUNNING SECURED L.OOP - LOOP NO. RATE (GPX)- (CPGP)
.t - i-
-_._......5 .I VII.A _ lA.lBIC ------- _LA ____-- I 91,100 277,800 l VII.B ' A1B1C__ __
----- - 7 92.400 277,800 VIT.C 1IA,B,1C _I__C _---- 13 94,300 277,800 I TWO LOOP FLOW VII.A 1A, 1B iC 1A 2 96,147 194.183 VII.A 1A. IC IB-A 4 96,530 196,233 VTI.B 1AX 1B IC .1B . 8t. '98,036 194,183 VII.B 1B. IC lA_ 1B 10 98,471 .198,203 VYI.C_ 1A. IC lB IC if1_ 99,703 196,233 VII.C 1B, 1C lA IC ____ 15 99,732 198,203
[,;,I.A VT1.A 1A IA, IC 1C _A TWO LOOP FLOW-ISOLATED LOOP I lA -B IC 3 5 95,783 VII.B IA, 1B 1C LB AC .9 97,066 192,420 lV.f lB lC lA 1B - 1l - 97,454 _
. SINGLELLOOP FLCW-3.7.5
3.8 RCS Flow Coastdown The RPS is designed to protect the reactor core in the event of a partial loss of coolant flow accident. A primary coolant low flow reactor trip is actuated to prevent a rapid increase in coolant temperature which could result in departure from nucleate boiling (DBN) and subsequent fuel damage upon partial loss of flow. The BVPS FSAR Partial Loss of Forced Reactor Coolant Flow accident analysis uses computer codes to calculate core flow, reactor trip time, the nuclear power transient following a 4? reactor trip, and the resultant heat flux transient. The DNB Ratio is determined from flow and heat flux information based on pump coastdown and nuclear heat transfer characteristics. A test was performed to obtain data necessary to verify the conservatism of the FSAR analysis. The reactor coolant flow transient following a RCP trip was monitored and recorded so that time delays associated with the loss of flow incident could be determined. The parameters recorded during RCP trips included the reactor coolant loop flow transmitter output voltage, position, the RCP breaker position, the reactor trip breaker and the RCCA detector coil voltage. With the station in hot standby, all control rods fully Inserted, and three reactor coolant pumps operating, the slowest RCCA (at hot full flow conditions, K-14) was fully withdrawn. The IA RCP was then crriped (Case 1). The tine delay from pump trip to the start of rod motion and the deenergizing of the 1A loop low flow bistables was determined. The res'elts are reported in Table 3;8-1. 3.8.1
MOMMEMEN Following the restart of the lA RCP, a sequence of RCP trips and starts provided conditions necessary for the collection of the balance of the flow coastdown data. Reactor coolant flow anoj RCP breaker response data was collected under the following conditions: Case 2) With three pumps operating/three pumps tripred and coasted down. Case 3) With two pumps operating, no loop isolated/one pump tripped and coasted down. Case 2) With two pumps operating. no loop isolated/two pumps tripped and coasted down. Case 5) With two pumps operating and one loop Isolated/one pump tripped and coasted down. Case 6) With two pumps operating and one loop isolated/two pumps tripped and coasted down. Independent analyses were repeated twice on the test data to prepdre normalized ~J average core flow vs. time curves for comparison with the curves used in FSAR sections 14.1.5, "Partial Loss of Forced Reactor Coolant Flew, "and 14.2.9, "Complete Lors of Forced Reactor Coolant FPlo". In each case the test curve fell below the FSAR curve. The six loss of flow cases were reanalyzed by 'Westinghouse using the same methods and computer codes that were used in the original BVPS FSAR analysis. Changes were made to the dynamlc pump characteristics used in the iHOENrX computer code to reflect BVPS test data. The resulting flow and DBE calculations are conservative relative to the BVPS test data. The minimum DBN Ratio was not. violated in the calculation: The BVS normalized average core flow vs. time curves meets or exceeds these asmended TSAR assumed core flow requirements In each case. (See Figures 3.8-1 to 3.8-6)
The Low Flow Time Delay. the RCP Under Voltage Trip Delay, and the RCP Under Frequency Trip Delay time responses were below the maximum acceptable values. (See Table 3.8-1) 3.8. 3
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4.0 HOT ZERO POWER TESTING 4.1 General During this phase of testing, various reactivity parameters were measured and core power distribution maps were obtained to verify the conservatism of the core design values used in the FSAR. The data obtained verified correct core loading, and also provided operations personnel with the information (such as, RCCA worths, boron worths, isothermal temperature coefficients, etc.) necessary to conduct reactivity manipulations. A Initial criticality was established by fully withdrawing all shutdown and control banks except for control bank D which was withdrawn to 160 steps. A RCS boron dilution was initiated until reactor criticality was achieved. Reactivity computer checkouts were then performed and the HZP physics test decade wa3 determined. Low power physics testing was performed at O-1% full power with limited excursions up to 3% full power during M/D flux mapping. The RCS was maintained at approx-imately 2235 psLg and approximately no load Tavg of 547F. A RCS boron dilution was initiated and the control banks and shutdown bank B were fully inserted into the core in sequence. Isothermal temperature coefficient, RCCA worth measurements, boron endpoint) boron worth measurements, and core power distribution (see Section 5.1.3) data were collected during this time period. Shutdown bank A was then inserted and the most reactive rod (RCCA H-14) withdrawn to maintain criticality. With ARI except RCCA H-14 which was fully withdrawn, minimam shutdown data was obtained. RCCA H-14 was dynamically dropped from this control rod configuration to obtain dropped rod worth data. As soon as the data associated with the dynamic rod drop was recorded, the reactor was manually tripped and borated to
approximately a 2% tK/K shutdown margin. Shutdown banks A and B were then fully withdrawn, reactor-criticality was established and additional RCCA worth data obtained by withdrawing control banks A, B, C, and D in nornul overlap. The control bank positions were then adjusted to the BVPS Tech. Specs. HZP Insertion i:: limits and the RCCA pseudo ejection of RCCA H-14 performed. Following this test, nuclear noise measurement data was obtained and a plant radiation survey at > 1% and c 3% full power was performed. The reactivity computer input was then changed to monitor all four NIS power range channels and a rescaling of the power range channels was performed. The initial power
- . , ascension testing program was then begun.
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4.2 Initial Approach to Criticality The inital approach to criticality began on May 7, 1976. After establishing the baseline for the Inverse Count Rate Ratio plots and with the RCS borated to 2149 ppm, all control rod banks and shutdown banks were fully withdrawn except control bank D which was withdrawn to 160 steps. RCS loop and pressurizer borcn sampling was then initiated to determine boron concentration at twenty and sixty minute intervals respectively, and boron dilution commenced. ICRR plots, boron concentration logs and dilution logs were maintained during rod motion and boron dilution (See Figure 4.2-1 to 4.2-4). ICRR plots were used to predict the initial criticality conditions during boron dilution. PG water supply problems caused a suspension in dilution for approximately 42 hours. Dilution resumed when the problem was resolved. Dilution was secured at 1830 hours on May 10, 1976 when the ICRR value reached approximately 0.05. Initial Criticality was achieved during RCS mixing at 1845 hours on May 10, 1976. Initial Criticality conditions were: RCS temperature - 547.5F RCS pressure - 2235 psig RCS boron concentration - 1332 ppm Control Bank D position - 160 steps These conditions were within the acceptance band of the BVPS design. Once criticality was established, the reactivity computer was calibrated in preparation for physics testing. Proper calibration of the reactivity computer was verified by comparing the reactivity calculated by the reactivity computer with the reactivity calculated from the neutron flux doubline time. 4.2.1 I
With the reactor critical, nuclear heating effects were observed at a power level corresponding to 3 x 10 7 amps as indicated on the reactivity computer and at approximately 1.5 x 106 amps on N35 and N36, the intermediate range NIS channels. The HZP Physics Testing Decade was chosen as 1 x 10 9 to 1 x 10 8 amps with an upper limit of 3 x 10 amps for HZP tasting. During the initial critical checkout of the Westinghouse model BVPS reactivity computer, the reactivity trace displayed a noticeable tendency to drift toward zero following a poaitive or negative step change in reactivity. Attempts to determine the cause of the drift were not successful and a Westlnghou3e PC-12 reactivity computer was brought on site to minimize time lost in the progress of the physics test program. The PC-12'reactivity computer also exhibited a reactivity drift toward zero following a step change in reactivity. After eliminating items such as improper delay neutron fraction values and/cr improper programing and setup of the reactivity computer, it was concluded that the problem source must have been the neutron detector to which the computer was connected. As BVPS is a three loop plant, the NIS power range channel N44 is not part of the overpower and overtemperature protection circuit. Channel N44 provided an input to the automatic rod control circuitry and was chosen for use with the reactivity computer to avoid placing one of the three overpower and over-temperature protection circuits in a tripped mode. When the reactivity computers were removed from channel N44 and connected to channel N43, the undesirable drift was virtually eliminated. The remainder of the zero power physics program was completed using the BVPS reactivity computer connected to channel N43. 4.2.2
The problem with channel N44 was eventually traced to a problem with the upper section of the dual section nuclear detector. Replacement of the N44 detector is planned for a future shutdown. 4.2.3
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4.3 Isothermal Temperature Coefficient, in order to confirm design objectives It was necessary to. determine the isothermal temperature coefficients of reactivity for various control rod configurations. These measurements were made by determining the magnitude . the reactivity change due to a change in RCS temperature (Tavg) while other RCS parameters were held constant. A RCS heatup rate of approximately lOF per hour was established and a plot of reactivity versus Tavg obtained using the x-y plotter of the reactivity computer for a RCS temperature change from 542F to 547F. This process was then repeated for a similar coo'ldown rate. The isothermal temperature coefficient was determined from the average value of the slope of both the heatup and cooldown plots. The isothermal temperature coefficient results are presented in Table 4.3-1. A typical trace is shown in Figure 4.3-1l All isothermal temperature coefficients fell within the tolerance of the design predictions. All va!~..s obtained were negative including the ARO value which was found to be -0.62 Pcm/F. The isothermal temperature coefficient is the sum of the.moderator temperature coefficient and the Doppler coefficient. The design value of -2.2 pcM/F for the Doppler coefficient at hot zero power, beginning of core life, was subtracted
. from the isothermal temperature coefficient determined in the above manner to obtain the 1MTC. Using this calculational technique it was determined the MC was a positive value for the ARO condition. BViPS Technical Specification 3.1.1.4, however, requires that the MTC be negative for all operating conditions except during low power physics test~s. This requirement ensures that the assumption used in the accident and transient analyses are valid. The MTC was maintained negative by administrative control over the alloajablemaiza.mum 4.3.1 i I- I - -I - -- ... --- I - . - - -- - - - - - --
coolant boron concentration during critical operation. This was accomplished by , limiting bank D withdrawal to not more than 70 steps withdrawn until a xenon concentration equivalent to 30% power equilibrium xenon was esrtblished during l the first 200 'IWD/)MI of core burnup. i.)
.,3 2
.1 Table 4.3-1
_3 *Isothermal Temperature Coefficients Isotherm Temperature Re~activityC-...fficient Rod Position .. -. -~ --- Acceptance (rods not-desixiatid are. fully withdra..m)- Actual (pcm ) (2) (pm) CWD 697 . -0.62 -1.1 +3' CBD 8O & BC' 210 -4.26 -4.2+ 3 CBC'&CBD.~6O,CBB 6205 -7.63 -7.9+/-+3 CBB, dCBCD& CBD 6 0, CI 8 200' - -12.77 -12.9 +/- 3 CBA, CB, ECBC&CED 60, SBB 6 192.5(1 -13.8$ -12.4 +/- 3
* (1) 553 position-is the averagi i of the-rod position during heatup'(SBB 6 185) and cooldoini (SBB3 200)
(2) Act ual value is1the averagee of-the -isothermal temperature coefficients obtained durli aig heatup, and cooldown.m reactivity I _1%1t. Ij I i " I i i . II - . - I I I f I I 9-- 4.3.3
W.E x9,01 9.6 Sam a 46 1320 Reactivity vs. RCS Tavg All. rods at 228 steps except CBC EsCBD 8 0, CUBB 205 546.8F to 543.8F
4.4 RCCA Worth Xeasurements Under hot zero power conditions the nuclear design predictions for rod cluster control assemblies group differential reactivity worth were validated by RCCA control bank reactivity measurements. The' RCCA control banks were stepped into or out of the core as required to compensate for reactivity changes due to RCS boron concentration changes. Reactivity worths were determined for the individual banks during boron, dilution, for the control banks in normal (100 step) overlap during boron addition and for RCCA H-14 during both addition and dilution. In each base, reactivity worth was determined by evaluating boron concentration sampling data, RCCA group position, and reactivity measurements obtained during rod motion-Two independent anaylses were performed on each reactivity trace obtained during boron :hanges and rod bank motion. Differential and integral reactivity worths were determined from the reactivity measurements in each analysis. The integral reactivity Worths for the two analyses were averaged in each case and the results were compared with design requirements (Table 4.4-1). Excellent reproducibility is apparent in the comparison between the worths of the banks in overlap and the sum of the individual banks. The average differential and integral reactivity worths are plotted on figures 4.4-1 thru 4.4-5. All control bank's integral reactivity worth& were within the range of the expected values except Control Bank A worth which exceeded the expected value. (Ref. Table 4.4-1). The expected values were obtained based on "state of the art" nuclear design calculations and previous start up measurements. Observed discrepancies between expected values and actual measured values do not in 4.4.1
themselves constitute a safety problem. Control Bank A is fully withdrawn during power operation and effectively functions as a shutdown bank. The fact tha: K excess negative reactivity is available outside the reactor core for trip insertion is not of safety significance. The integral worth of Shutdown Bank B was determined from reactivity computer traces obtained during the stepwise insertion of SBB into the core during RCS dilution. From two independent data analyses, results were averaged and then were compared with design requirements. SBB reactivity worth was within the range of the expectel value. The Integral worth of Shutdown Bank A was not measured due to BVPS Technical Specifications which prohibited the RCCA configuration necessary for the measurement. The part length banks' reactivity worth was not determined as the part length rods will not be.used. BVPS Technical Specifications require that the part length control bank be completely withdrawn from the core and the drive mechanism deenergized during power operation. 4.4.2
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Tab le ....mmalry Integral Hud Worth Measured Pred icated Difference Rod Bank (pcw) (pcm) (X) 1265 11110
- 113) + 11.9 CB A 19b2 [1870 +/- 1871 + 4.9 CB B 1209 [1140 +/- 1141 + 6.1 Cll C 1399 11370 + 1371 + 2.1 CB D 5835 5510 + 5.9 Tlotal Cuntro1 Wlinks 5847 5510 i 6.0 :1 In NnrmaL overlap . .
Shutidown 1020 ')80 + - 4. 1 O. wisk U 11-14 401 785* _ - - -
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4 .5 Boron Ettdpoint and Boron Worth Measurement Boron endpoint measurements were performed to determine the critical boron concentration at selected rod configurations.- Due to the uncertalnty associated uith determining the exact-just critical-bank configuration after.a RCS boration or dilution.-boron endpoints are determined uslng a measured reactivity increment as opposed to attempting to measure the actual boron concentration with the reactor just critical and the bank at its endpoint.-- With the reactor critical at HZP the crltical boron concentration was measured at a specified control bank position. The controlling bank positl.n was then changed from a position near its endpoint of travel tb its actual endpoint and the change in reactivity recorded via the reactivity computer. The boron endpoint value was calculated as follows: - F J ~~ ~ ~ ~ ~ ~
,(J- - ., ; '. ,, '+ 4 '. .- .. . .
Where: CB(E.P.) Endpoint boron concentration. CB(J.C.) - Just critical boron concentration with the controlling bank at the bank initial position. Aa Reactivity change caused by controlling bank withdrawal or insertion as measured on the reactivity cocputer.
- &CB n Inverse boron worth (ppmlpcm)
CB(J.C.) was corrected to the no load Tavg condition, if tie test was performed with f.value of Tavg other than 547 F as follnoas: CB(J.C.) CB(J-C)measured AT ACB 4.5.1
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- Isothermal temperature coefficient of reactivity
-T D - Differential boron worth ACB A summary of the measured boron endpoints along with the corresponding bank position and predicted boron concentrations are given on Table 4.5-1. The acceptance criteria for the ARO critical boron concentration is taken from design predictions with an Acceptable tolerance of + 4% of the design value. All other acceptance criteria for critical boron concentrations are then derived from a formula which adjusts the acceptable value based on the last measured endpoint critical boron concentration as the rods are inserted.
The measured boron endpoints were within the acceptance criteria with the exception of the critical boron concentration at the N-I rods condition. Since this small deviation (7.7 ppm above the upper acceptance criteria boundary) did not involve a violation of the BVPS Tech. Specs. and failure to meet the predicted value did not constitute a safety problem, the resulrs were considered acceptable. Boron worth is defined as the ratio of Integral reactivity change to the corresponding change in critical boron concentrations. Boron worth is calculated using data collected during rod worth measurements. Plots of integrated reactivity versus critical boron concentration were made and the average boron worth determined from the inverse of the slope of these plots (ex. Figure 4.5-1) and are summarized in Table 4.5-l. The avera-e boron worths determined were all within the predicted value of 10.3 + 1.03 pcmlppm.
Table 4.5-1 Boron Endpoint and i Boron Worth Data
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- Boron Endpoint Data Rod Position . - .
- Measured Critical . Acceptance
,..odsnot deilgnated are fully withdra`n) Boron Concentration (ppm) Crite:la (ppm)
,rod no.
All rods out 1344 1319 + 50 CBD @ O.- -1198 1210.3 + 19.4 CBC 5 CBD e o 1088 1086 + 16.4 900...... 903.6 + 23.5 CBB, CEC &-CBD O 0 CBA. CBB, CBC & CBD 0 791, - 793.1 + 14.9
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.S Boron Worth Data
.lntejral Reactivity ChanRe - Aveirge Boroti Worth (pctnIDpm)
Boron Dilution - ARO to . - 10.54 Control Banks Fully Inserted Boron Dilution - ARO to . 10.72 N-1 Condition (All rods in xcept U-14 at 228 steps) Boron Addition - Control ' 10.68 Banks Fully Inserted to ARO 4.5.3w-
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4.6 Hinimun Shutdown Verification As defined is rhe BVPS Technical Specifications the shutdown margin is the instantaneous amount of'reactivity by-which the reactor is or would be subcriti.-al from its present condition assuming:' a) 'no change in part length' rod position and b) all full length rod cluster assemblies (shutdown and control) are fu My inserted except for the single rod cluster assembly of highest reactivity worth vhIch is assumed to be full7 withdrawn. The required shutdown margin as explained in the basis of the BVPS Technical Specifications insures that 1) the reactor can be made subcritical from all operating conditions,
- 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
The shutdown margin wps evaluated by obtaining the critical boron concentration with all full length rods inserted and the nost reactive rod fully withdrawn (N-i condition). The minimum boron concentration required to maintain a shutdown margin of > 1.77% AKIK per BVPS Technical Specification 3.1.1.1 was calculated by converting the required shutdown margin to an equivalent boron concentration using the inverse boron worth (Figure 4.6-1) and adding this value to the measured N-1 configuration critical boron concentration of 603.6 ppm. Table 4.6-1 summarizes the shutdown margin calculations for BVPS at the beginning of core life at hot full power. Control rod integral reactivity worth to the N-1 Condition was obtained by multiplying the difference between the measured all rods out critical boron concentration and the N-I 4.6.1
condlotin critical boron concentration bv the boron worth of Figure 4.6-1.. This value was reduced by 10% for measurement uncertainty. Rod requirements value were taken from design predictions. The calculated shutdown margin at BOL verifies that adequate shutdown is available with an excess snutdown margin of 2851 pcm. K)L 4.6.2
. . . i ; Table 4.6-1 i Shutdown Margin
-- . Calcuiations Worth (POI)
. . . .. : t Control Rod Integral Reactivity I 7934 Worth to the N-i Condition I I , ,
10% Measurement Tolerance: I ! 793 Rod Worth Less Measurement ! ' 7 : 7141 I
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Doppler I 1260* Void and Variable.TMOD 260* ..Redistribution 500*
*500*
Rod Insertion Allowance, . Soo Rod Requirements, (2) - 2520-Calculated Shutdown Margin 4621- [(1) - (2)] Required Shutdown Margin 1770
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4.7 RCCA Psuedo Ejection at Zero Power Features of the PWR are intended to limit the possibility or consequences of a rod ejection accident. These include conservative mechanical design of the rod housingsia thorough quality control program-during assembly, and a,- nuclear design-which lessens the potential worth of an.ejected RCCA and minimi'.es the number of assemblies inserted at power. The analysis of RCCk ejection accidents and their consequences presented in the BVPS FSAR demonstrates the adequacy of-the nuclear design based on calculated values for ejected rod worths moderator and doppler reactivity coefficients,. hot channel factors and other important parameters. The RCCA ejtction analysis was performed for four different accident conditions:' 1) BOLD HZP, 2) BOL, HFP,
- 3) EOL,'HZP ard A)'EOL, HFP.
Data was obtained-during hot zero power testing't-o-verify the conservatism of the assumed worth of the worst case ejected-RCCA (R-14) and the conservatism of associated calculated maximum Nuclear Rot Channel Factor used in the FSAR accident analysis for UZP, BOL. With the reactor critical, Shutdown Banks A and B and Control Bank A fully withdrawn, Control Bank B near fully withdrawn, Control Bank C at the HZP Insertion limit and with Control Bank D fully inserted, H-14 was manually withdrawn during a slow boration. The integral worth of RCCA H-14 was calculated in two separate data reductions of reactivity changes m2asured on the reactivity computer during RCCA H-14 movement. 4.7.1
A boron endpoint and a moveable incore detector flux map was obtained for RCCA H-14 fully withdrawn. A slow dilution of the RCS was then initiated and a second rod worth measurement made as RCCA H-14 was inserted. The results of the analyses on each of two RCCA H-14 rod worth measurements (boration and dilution) were averaged to obtain an RCCA }1-14 rod worth of 401.15 pcm, the highest value was 410.8 pcm. An additional ten percent for measurement and other uncertainities results in an average of 441.3 pcm with a high value of 451.9 pcm. The FSAR assumed value for the worst case ejected rod worth was 0.785% AK, or 785 pcm, demonstrating the conservatism of the assumed value. The INCORE computer code reduction of the "H-14 ejected" moveable Incore detector flux map (see figure 5.1.3-6) yielded a maximum Nuclear Hot Channel Factor (FqN ) of 6.3422. Applying a ten percent engineering and uncertainty Zactor results in an Fq of 6.9764. The FSAR calculated value of Fq was 13.0 which demonstrates the conservatism of the accident analysis. 4.7.2 A-. - .. - t.. --- . . . -
5.0 POWER ASCENSION., 5.1 Physics Test 5.1.1 General The detailed program of react c physics measurements undertaker.'duringpower ascension physics tests verify that measured static and kinetic nuclear core characteristics are as expected and that the kinetic coefficients assumed in the safety analysis are conservative. Physics tests provide the control rod configurations necessary'for pover distribution measurements. These measurements are obtained'for normal and . abnormal rod configurations and provide information useful-in determining the
- static and dynamic effect of ejected, dropped and out of position control rods on the nuclear characteristics of the core. Additionally, power level maneu-
' vering provides data for determining the power coefficient and generated core axial power imbalances provide data for determining hot channel factors obtained near the BVPS Tech. Spec. limit for axial power imbalance.
The power ascension program establishes power level plateaus for the performance of tests. Portions of some physics test are performed at many platesus (i.e. special hot channel factors and power distribution and coefficient measurements). Other physics tests'are performed entirely at a specific plateau. (i.e. dynamic
'static rod drop, pseudo rod ejection and rot out of position tests.)
This section describes the power ascension physics testing completed up to the 50% F.P. plateau. 5.1.1.1 .
I- r 5.1.2 RCCA Psuedo Ejectior. and RCCA Out of Position At approximately 30% FP data was obtained to verify the conservatism of the HFP assumed worth of the worst case ejected RCCA (H-14) and associated calculated maximum nuclear heat flux hot channel factor used in the FSAR ejected RCCA accident analysis as discussed in Section 4.7 of this report. With the control rods in their full power rod configuration (control bank D at 190 steps and all other rods at 228 steps) and the reactor at 30% YP with equilibrium xenon, a reference (base case) full core fl/D flux map and Encore TIC map was obtained. Secondary plant calorimetric data, NIS detector currents and delta flux meter readings were also recorded during the performance o6f these maps to provide data for initial incore-excore detector calibration. RCCA H-14 was then withdrawn from 190 steps to its full out position of 228 steps. Reactivity changes were measured on the reactivity compacter during rod withdrawal and the rod worth calculated. With RCCA H-14 fully withdrawn a M/D flux map and T/C map was again obtained along with secondary plant calorimetric data, NIS detector currents and delta flux meter readings. RCCA R-14 was then inserted until it was realigned with its control bank while a second rod worth measurement was obtained. The FSAR assumed a ejected rod worth of 200 pem. Analysis of the two rod worth measurements resulted in an average RCCA H-14 rod worth of 11.42 pcm, verifying the conservatism of the assumed value. The measured values for Fq maximum, FaH maximum, QT and A.G. at the base case flux map and the flux map taken with RCCA H-14 ejected are given an Table 5.1.3-1 5.1.2.1
I 1I (Haps 07A8A). Relative fuel assembly powers obtained are shown on figure 5.1.3-5 S I and S.1.3-6. The measured-values for FqN-(increaaed.:by 3Z to account for I i manufacturing tolerances and further increased~by 5% for measurement uncertainty), N FAU (increased by 4% for measurement uncertainty) and QT for the basecase-were The all within the BVPS Tech. Spec. values of 4.44, 1.77 and 1.02 respectively. I difference between the measured and predicted relative fuel assembly powers were all acceptable. The maximum measured value for FqN with RCCA H-14 ejected (increased by 8% to account for manufacturing tolerances and measurement uncertainty) was 2.1459. This value was well within the FSAR assumed value,of .7.07.:
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5.1.3 Power Distribution Measurements A detailed power distribution measurement program is Included in the BVPS test program which provides the conditions necessary for many different power distribution measurements so that the effects of variations in rod config-urations and power levelson core performance can be analyzed. Power distribution measurements are accomplished via the incore instrumeitation system. This system consists of moveable minature fission chambers (moveable detectors) and fixed fuel assembly outlet thermocouples. These measurements are used in verifying compliance with design requirements and technical spe:ifications for hot c:tannel factors and axial flux distribution. They also are used for calibrating the excore detector flux difference indications, and for detecting core power asymmetrics, misaligned control rods, core loading errors, or local power distribution abnormalities. The INCORE computer coae processes core flux informat:on obtained by the incore instrumentation system, calculates peaking factors anc. compares the measured data to predicted values. Output from the INCORE code provides information on the core power distribution, maximum assembly power, and local rate of burnup. Table 5.1.3 provides a summary of pertinent core parlmeters from the power distribution measurement program. Excore detector current data taken during flux mapping is also shown where applicable. The flux map table lists data for the major flux maps taken during the startup test program, with both full core and quarter core maps numbered sequentially. Core positions for the maximum hot channel factors are identified by individual fuel rods. Quadrant power tilts are listed by quadrant. So adjustment has been made to the listed data to account for nuclear or engineering uicertainity. 5.1.3.1
The power distribution M.ps, Figures 5.1.3-1 to 5.1.3-2, presented in this section are provided for reference and compariscn for other "physics testing" secti'tns. Flux maps display relative assembly power Eorrspecific rod configurations. All relative assembly power,;distributions were with'in an acceptable range of the expected values. Thermocouple maps were also obtained during selected tests. These maps'provide additional data for comaparison with flux map information. Thermocouple data can be reduced by the on-site computer to provide hot channel factor values derived from temperature distribution as a rough, rapidon-site technique for determining hot channel factors. Thermocouple maps taken .!uring power distribution measurements provide a backup to moveable detector maps but are not sufficiently accurate for the calculations required by Technical Specifications. I) I a
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.Z- .0.760 0.637 1.329 1.427 1.002 0.448 0.597 0.704 0.598 0.442 0.965 1.406 1.309 0.612 0.710 5.3 0.8 -2.7 1.1 2.2 3.9 6.8 9.2 7.1 2.5 -1.6 -3.4 -4.2 -3.2 -1.7 J -
.678 1.132 1.458 1.403 1.132 0.707 0.689 0.607 .647 0.685 1.096 1.336 1.389 1.106 0.645 5.3 0.8 0.4 0.5 1.6 4.4 8.1 8.6 1.6 1.3 -1.6 -4.3 -4.3 -1.5 0.1 1.270 1.316 .357 1.044 0O52 10.70, 0.45410.698 0. 5j2 1.075 1.371 1.310 1.256
. 0.9 :2.1 -1.6 -1.3 i.2 ,4.0 5.3 3.1 3.2 1.6 -0.6 -2.5 -0.2 I 9 4 4. I I - I W .444-0.858 130 10.933 1.133 1.026 11.119 1.023 1.152 1.105 1.204 11.106 1.147 0.830 L -1.9 -5.1 -0.6 -3.0 0.5 4.3 3.4 4.5 5.6 5.6 -0.4 -0.4 2.9 I- I - I .
1 - 4. - M 0.618 0.475 1.008 1 .3431.396 1.483 1.413 1.389 1.081 0.517 0.625 3.3 x IV tj C C 3 Ic: 2.1 -2.8 -3.7 - .6 -Oo. 1.9 1.2 0.7 3.3 6.5 L X >I; j w - - i i j - -- i - -- i ;; , - i O I N 0.588 11.077 1.296 11.444 1.365 1.423- 1.Z88 11.10810.648
-2.8 -6.5 -3.6 -0.6 . -0.1 -4.1 1.41 7.0 _ KEY
_ . ID VI measured - 0.000 P 0.780 1.213 1.135 0.642 1.125 1206 0.845 ar tvZl
- 00.0
-6.5 . --3.6 I - 1.0 4. - 1.6 I10.2 -4. - 1 - 1.4 X deviation I -
from expected 0O.675 0.75610 675 I-. A:3 I- 0 4.8 4.8 4.8
- 9 - I-( C.)
01 (8 1 2 3 4 5 6 7 ' ii 'L 9 i1.0 11 12 13 14 I 15 I . . I.. I I..- 1r% A - I - I I I % *.306 .0.3S110.359,: I 6.3 _4.*2 ; 1.8 13 _ -I - - 4 1 , 0.435 0.608' 0.548' 0.J54 0. 710i .912'1 0.'684
' '5.9 11il.4I '6.5S '5.6 '5.2- ;'6.2 '0.'8 C - - - 0.384 0.599 0.666 0.734 0.813 0.954 1.04i 1.110- 0.804 4'.6 6.7 66.0 6.4- 7. -4.5 '-0.8' -1. '-6.7 _ _
0.342 0.426 0.544 0.698; 0.79 0.936 i.0oi5 i. 134 i.282 1.-258 1.230 D _ 5.6- 1 5.8 7.2 0.2 -3.8 1.2 '5.3 0.8', 3.0 -2.2 -1.8 0.308 0.501 0.534 0.598 0.591 0.735 0.8 9 0.970 1.087- 1.586 1.839 2.104 1.480 E -Q A _q- t 5.0 R.0 w J11n L z 1 0.411. 0 n5100.597 0.523 .324 0.633 0.760 0.836 0.711 1.526 .243 2.443 2.492
. '.10.1 15.3 '11.0 5.3 - 8.5 7.0 1.6 ; 1'.9', '-7.3 -5.1 '-9.1 4-.8 C
- 0.199 0.373 0.510 0.573 0.554 0.541 0.700Q 0.814 1.046 1.199 1.739 2.465 3.006 3.081 1.973 10.8 11.2 11.4' 10.'9 9.1" 9.9 -12.0', 9.7 12.0' ',1.1' -9.9 -8.9 ,-8.1 "-8.3 8.5
- 0.222 0.220 0.504 0.546 0.535 0.576 0.712 0.911 1.004 i.188 1.737 2.598 3.314 3.209 2.567 Z., . 14.0 15.6- 6 15.6S. . . 6.7 9.8 '14.3 14.3 I.3 '-0.'9' .5 -9.5' '-7.3' '-'.0 11.2 3II os 0.205 U.387 0'.522 0.563 0.535 3.544 0.704 0.842 0.908 1.146 1.778 2.409 2.965 3.008 1.879 13.I 15.3 1-4.1" -9'.0' >'5.4 '10.4' 12.:6:' '13.5 '-2.9 '-3.4:'-7.8' 11.O: -9.3' ;'10.'4' -12.8 K 0.448 0.506 0.564 O.Sl 0.337 0.619' 0.738 0.857 0.725 1.625 .253 2.505 2.561 14.3 9.2 8.9 8.5 9.6 6.0 3.9 4.0 3.9' -1.3 -4.7 -6.8 ' -6.2 0.336 0.497 0.509 0.612 0.592 0.709 0.798 1.046 1.194 1.594 1.922 2.147 1.599 L 2.0 10.0 9.1 5.7 -5.7' -0.9 -1.8 15.6 7.2 3.0 15.6 7.1 3.0
- . 0.362 0.444 0.594 0.728 0.809; 0.937 1.081 1.253 1.355 1.308 1.311 9.1 1.8 7.2 10.2 7.5' '7.4 '6.7' 6.3 7.0 9.4 7.4
.384 0.574 5.674 0.773 0.814 .958 1.0621 1.210 0.965 I
*n6 ' IVO . -, - .
4.5 1.6 7.9 12.1 7.8 '5.0 2.9 7.7 12.0 KEY
.. = ue. .~ Id 0.387 p.592 0.571 0.371 725 10.884 10.732 measured - o.oool p . .
3.1 R.5 10.9 10.5 7.4 13.0 { 7.9] Z deviation - 00.0 __ _ _ from expected R 0.321 10.388 0;396 rJ. I 11.5 11.9 12.1 C
1 3 4 5 6 7 8 9 10 11 12 13 14 15 2 I . . . 4 a . 0.634 O.A53 0.642 A 0.5 , - 1.8 1.8 7 0.601 0.921 I1.026 1. ,1241.050 10.918 0.601 B - I - 2.0 0.I -1.0 2.9 1.4 -0.3 2.1
- . -4* 4 4-9I9
~ 0.612 10.90910.979 1.146 L.207 1.175 10.998 0.924 0.636 C , - -
0.81 -1.5 -1. 5 -0.5 0.9 0.4 2.4 5.4 1.5 0.62 r0.836 10.956 1.100 1.082 1.1.2 1,.151 1.134 O.,97710.832 0.619 D 4-,-- 3~.0 11.6 1-0.11 -1.7 -5.4 -3.2 0.5 1.3 2.1 1.2 2'.6 0.59610.915 10.971 I1.12611.124 1.166 1.066 1.174 1.113 1.112 0.944 0.8781 0.561 1.3 1.4 1.4 2. -0.1 -1.8 -2.0 -1.1 -1.0 1.0 -1.3 -2.7 -4.7 0.922 0.996 1.123 1.139 1.207 L.162 1.182 1.107 1.163 1.071 1.080 0.948 0.883 F -
-4.1 0.21 0.2 0.4 1.3 0.9 1.4 1.2 -3.4 -2.8 -4.al -3.5 -4.61 0.642 1.030 1.160 1.145 1.199 1.160 1.174 1.051 1.155 L.-18 1.148 1.099 1.116 0.998 0.609 G -4.2 -3.6 -3.4 1.9 0.3 -0.4 0.0 0.9 1.2 1.4 0.2 -0.2 -2.4 -3.3 -3.9 a1 - .
1.161 1.062 0.830
!II 0.862 1.099 1.197 1.220 1.092 1.184 1.066 1.123 1.038 1.151 1.074 1.173 !4 3.0 0.6 -1.4 0.7 0,3 1.3 1.6 1.1 -1,1 _l. 5 -1_1 --k 1 -4 t _-?F _n 4)
Ji 0.649 1.044 1.174 1.142 1.170 1.151 1.179 1.068 1.122 1.113 1.156 1.101 1.115 1.022 0.634 3.0 n.8 0.8 -0.2 -1.5 -0.5 1.8 1.8 -3.1 -2.9 -2.7 *-3.8 -4.2 -1.3 0.7 O.93210.987 1.107 1.109 1.188 1.144 1. 168 1.131 1 .187 1.11611.103 0.997 0.949 K 1.4 -0.6 -0.2 -0.1 -L.3 -0.7 -0.8 -1.4 0.3 3.1 1.2 -0.7 -1. I .4 __ 4 . 0.624 0.913 0.935 11.10511.102 L.166 1.07811.182 11.138 1.131 0.983 10.9441 0.616 L 6.0 1.2 -2.34 -2.0 0.3 91019134 -1.81-1.01 -0.5 I 1.2 2.7 +7 -2.2.7 - 4.7 + - 4.7 4 -. 8 0.634 0.831 0.950 1.116 1.139 1.208 1.136 1.118 0.99710.894 0.655 to I0_ - H -I M0C C ..
-i 5.2
-1 1.0 -0.8 -0.3 -0.5
- t. - P_ - +- - i-
-0.2 I-
-0.8
-- 4.
-0.1 4.2 8.7 L--....-
8.6 Cg X xX: w 0.618 0.90311.000 1.175 1.213 1.153 0.977 0.9511 0.679 .¶ 2.4 0.1 0.6 0.9 -0.0 -0.9 -1.7 12.5 KEY O. 000 rv, C 4 - t_ __ >_ 0.589 0.926 1.056 1.11711.048 0.905 .0.620 measured - rl M p . _ _ . _ . . _ . 0.1 0.6 ' 1.9 2.3 _1.2 -1.7 5.4 Z deviation - 00.0 a, kA I
- 4. 4. - - 9 - I - from expected z7 >
0.665 0.883 0.665 3-. 0 5.5 5.5 5.5
- ' - I.
i C C. C.
.Ic. c' 1 2 3 4 S -6 7 8 10 "11 12 I .. I 14 T 1S A I-0.602 10.7951 0.618 I I -: -.1 t 1-4-I -
-3.3 -3.3 -0.8 4 -, - 4I.
0.635 I0.918 0.995 11.028 11.012 n.91610.607
-5.1. -0.1 -1;91 -1.8 1-0.1I -0.41 0.5
- . - I- P44 -,-I-4-4-.-.
-I -
0.639 f0.924 10.979 1.13111.170 1.143 1 0.97510.945 10.651 I 1.2 0.4 1-2.0 1--2.0 -1.8 1,-O.Y -2.4 0.4 0.878 ' 3.1 _ 0.652 0.64',
- 0. '.868 0.982 64,.,, 'i,868 0.'982 1.120 1.130 1.195 1.124 '1.102 0.984 O.'878 'O.652 D &---- - - - - -A-
. 2.8 1 J.3 1 0.3 --0.8 -1.2 -0.9 -m.2.4
- 0.6 2.6 3.3 E ~ 0.599 0.919 -0.99211.125.1 1.1911 179 11. 069 1.156 1.0911 1. 109 10. 976 0.923 0.576
-1.0 -0.1 .1.4. 1.0 -0.2. -0.4 -1.8 -2.4 -2.7 -0.3, 0.2 0.21 -4.8 0.878 0.982 1.1'164 l1.lSO' 1.189 1.153 1. 189 11.154 1.16211.107 1.102 0.96410.888
_4-4.5 -1.7 3.2 2.5 -2.5 :1.S 1.8 :1.S - 0.21 -1.3 -2.4 -3.5 -3.4' 0.621 1.000 1.126 1.124 1.194 i.166 1.184 1.074 1.181 1.135 1.157 1.114 1.123 0.987 0.608
-- 0.4 -1.4 -2.4 -1.7 0.9 2.6 _*2.4 2.0 2.2 -0.' -2.3 -2.6 -2.6 -2.6 . -2.5 J -~ 0.843 1.037 1.156 1.170 1.084 1.195 1.094 .170 1.073 1.173 1.062 1.190 1.173 1.033 0.809 3.2 -0.9 -3.1 -3.0 -0'.'.s ,,L3 --3.9 4.7 --1.9 :05 2 ,4 -1.4 -1.I 6 ' -1 3 :-1.n 0.644 1.006 1.119 1.109 1.177 1.161 1.209 1.098 1.106 1.101 1.155 1.128 1.140 1.011 0.627 3.4 -- 0.8 -3. - 0 - 2.1
-;6 -. 6 -4 -4.3 -3.1 -2.4- -1.4 -1.2 ,-0.3 O.6 x 10.894 0.970 -101 1.113- 1.169 1.132 .191 1.130 1.1, lh.113 1.138 0.98810.919
.-I-2.8 -2.9 -0.71 -0.8 0.8 -0.4 2.1 -0.6 1.0 -0.7 0.8 -1.1 -0.1 L 0.63. 10.927I 0.999 1.115 1.131 1.15111.098 1.200 1.163 1.149 11.01 O.9;3 0.599 4.3 0.7 - 2.1 0.2 0.8 --2.8 . 0.9 1.31 3.7 3.2 3.2 -0.9 -0.9 a.6 I .1 0.5 0.637 0.860 0.i64 1.127 50 1 . 2 3 6 1.153 M 1.i 1.140 1.007 0.911 .0.658 W O -. .
'1l.0 -5.'6 -1 .S -n0I 0.6I I .-[ n-A 1.0i 'g0 A .s L .It *.JUJ* b c .
_ 1 e ::t I- : A :- __, .- -_` A A - I -v . _w~ I ev.. c0 i 'n x .; N -1.3 W 1. 197 11.238 11. 158 0.95810.946 0.091 I-0.91 -5.6 -1.3 3.8 3.9 1 0.4 -4.0 2.7 . S.5 KEY L g n tn p 0.570 0.903 1.048 11.080 11.025 0. 831 0. 621 - measured - 0.000 D n? ft rn 0O.
-5.6
-1.3 I --
- 3.4 I
3.1 I-1.1 I-
*-4.01I - 2.7 %'deviation -00.0 from expecti'd Cr > ,
R 0.639 0.836 0.639 PI-k 2.5 2.3- - 2.5 0 a3
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A
- I I
S- I - t* - p- p-I - . 0.597 10.795 0.622 Il I
. , . -3.2 -1.9 __
0.5 . , B _i _ I _.i I 0-61-5 0.907 0.997 1.034 1.021 0.932-10.614 I 3.4 0.2 -0.5 -0.6 _. 1.3 0.8 _ 1.6 C I - . -i 0.639 0.915 0.98211.137 1.182 1.13410.967 0.923 0.654 3.0 1.1 . _. _. _
-0.1I 0.1 .
-0.g -3.0 -0.2 2.4 *f_
D I i O. .661 0.868 0.976 1.121 1.143 1.210 1.109 2.L93 0.987 0.892 0.663
- . - + A.R
--- 4 -.- - 3.5 4 -
1 .64 1.1
--- 4 --
1.3 4 1.5 I- 7-24 I -2.0 4 n.i 4 I. 2I -.- *
.2 I 0.595 0.91610.'991 1.113 1.119 1.182 1.063 1.127 1.086 1.117 0.99510.932 O.585
-.4 - 0.6 4 1.6 3.3 2.0 1.2 1.3 -1.5 -4.2 -13. 3 -0.5 -0.6 -2.1 -7.2
--- 4-4 - I- 4-1k 4-4 4 F
0.853 0.966 1.144 1.133 1.18011.154 '1.162 1.12011.153 1.142 1.13310.976 0.905
-5.3 -1.3 3.6 2.8 2.7 2.6 0.5 -1.1I -0.8 0.3 -2.2 -6.5 -6.6 G 0.603 0.982 1.116 1.126 1.187 1.163 1.182 1.054 1.145 1.123 1.181 1.142 1.149 1,028 0.634
-1.6 -1.4 -1.2 0.4 2.2 3.8 3.6 0.7 -0.6 -1.9 -2.0 -3.6 -5.3 -5.7 -1.6 U't 0.821 1.035 1.164 1.161 1.065 1.176 1.095 1.153 1.042 1.157 1.072 1.210 1.210 1.119 0.872 2.0 0.0 -0.5 -1.6 -0.3 -2.4 5.2 4.1 -1.0 -1.5 -3.7 -3.1 -4.4 --2.6 -2.2 1.01 0.672 1.004 1.127 1.103 1.151 1.136 1.202 1.098 1.094 1.102 1.160 1.147 1.168 1.081 0.686 2.4 0.8 -0.2 -1.6 -0.9 1.4 5.3 4.9 -5.0 -3.6 -3.7 -3.2 -3.7 --0.8 1.4 n nnn X ~ U.U 0.993 .1I0O 1.091 1.13911.129 1. .163 1.104 11.155 1.116 1.14811.02i 0.983 1.5 -0.4 -1.0 -0.8 0.4 0.6 -2.5 I -0.6 -1.9 -o.91 5 1.5 L '0.626 0.937 0.988 1.102 1.108 11.148 1.063 1.162 1.140 1.152 1.024 0.967 10.639 5.8 3.9j 3.0- 0.91 0.2 -1.7 -1.4 -1.2 1.5 2.5 2.3_ 1.5 1.2 J.85 _4 0.6 M '- ' 0.645 0.852 0.960 1.108 1.133 1.198 1.133 1,133 1.005 0.898 0.666
- 0 P-
-4.21 1.5 -0.1 -0.1 10.4 0.5 -0.3 I 0.6 1.9 3.2 2 . 7 0 saoC X a
- - - a _ - C M %
N 0.626 10.872 0.978 1. .162 1.205 1.15110.987 0.9940 0.666i r a 1.1 -3.6 -0.6 2.2 2.1 0.5 -1.0 1.7 4.3li KEY ___ _ _ _ _ C 0 0.573 0.89911.035 11.069 '1. 031 0.906 0.615 measured - 0.000 = n n n o-p ddeviation - 00.0
-3.6 -0.6 3.3 2.8 2.8 -1.0 1.7 from expected C 0>
R 0.653 0.856 0.656 5.9 5.7 6.0 0 9
)
- I :0 C 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15
- A _ ~ , 4 , ~ ,. _,
_ .42 10.84610.643 _ .1 . _Lir 2.0 -2.2 I
.. B,, _ 1.~ - . -
0.64910.993 1.036 1.086 1.039 0.937 0.609
'8.3 _ ~
8.3' 1.5 1.4 1,-8 2.3 - 1.7 _ _ _ 4 ~ -
'C p . - 4 0.679 0.98B1 24 I ClJ.. _, 1.20111.177 11 01810.92910.631 8.8 1 -'.31 3.0 . _._ _ '0.4 2.0'1 2.4 1 -1.81 '1. a _
0.682 0.87410.938 1.092 1.131 l.l9S 1.150 1;1521 0.98710.857 O.631 D'-- - - 3.2 -3.4 -2.81 --0.9 ';2.6 "'1.6 '1.2 1.l 9.3 -0.7 `.8 _- I - _4-4-1 -_-
. - - I I I I.
E - 0.63610.949 0.987 1.109 11.120 1.83 i.070 1.186 1.129 1.11810.969 0.91610.597 6.1 6. 4.0 .0 _I 1.6 16- -0.1 -0.4 -0.1 -1.5 0.2 0.4 0.7 ' -0.2 ' 0.4 1-0.4 I 1 F - 0.945 1. 006 1.103 1.126 1.192 1.153 .. 154 1.i31 1.173 l.i16 .112 O.959 0.881 3.1 1.2 -1.9 0.1 -0.7 1.0 -1.1 -0.9 -0.9 -1.61 -1.0 -3.S -3.9 0.640 12038 1.154 1.123 1.187 1.164 1.164 1.041 1.146 1.133 1.165 1.129 1.109 0.977 0.583 1.6
. 1.8 0.0 -1.5 0.3 2.0 0.6 -1.1 -1.0 -0.8 -1.6 -1.0 -3.9 -4.2 -7.4 I.
iot . . .- 0.830 1.060 1.170 1.201 1.093 1.118 1.059 1.132 0.973 1.079 l.ol01.132 1.117 1.046 0.-812 0.1 1.0 -2.2 -0.2 0.6 1.8 - 0.7 1.6 -7.5: -7.5 -6.7 -5.9 -6.6 -2.3 -2.0 Oj -- 0.631 1.011 1.129 1.140 1.195 1.164 1;166 1.068 1.070 1.056 '1.105 1.073 1.086 1.053 0.649
- 0.3 '-.0.9 -
-2.2-0.0 0.9 1.9 0.8 1.6 -7.5 -7.5 ' 4.7 -5.9 -
-5.9 3.1 '3.1 I -K--- -- -,.C893 0.975 1.118 i.126 1.197~ 1.151 1.'16811.108 1.174 11.101 1.125 10.935 10.946
--2.0 I
*:'.8I -O.5 _ I 0.1 I1,1 0.8 I 0.11 -2.9 -0.9 .
-2.11
_ .- _ I 0.1 '-S.9 13.2_. 0.590 0.898 0.951 1.089 1.095 1.184 1.080 1.176 1.154 1.183 1.035 0.972 10.619
-1.6
-1.6 -2.1 I I
-2.2 -2.6 -0.0 -0.6 -0.7 -I
-2.7 6.6 6.5 6.5 3.3 nf ce M _ I 0.614 0.822 0.941 1.093 1.136 1.19611.133 1.153 '1.025 0.894
-1.6' -3.0 -3.1 -2.7 -0.4 -0.61 -0.6 2.7 - 5.5 5.5 ,4.5 I I - l - I - - -- - - -. _ _ ,
-N _I 0.593 0.868 10.966 1.146 1.188 11.140 1.038 0.953 0.652 n 0 uX
-4.9 -4.9 1 -2.8 -0.7 "-0.7 '--1.2 4.5$ 4.51 4.5 KEY 0.570 10891 1.048 1.100 1.084 0.976 0.626 measured - 0.000
-4.9 I - -2.8~ 44 2.7 -
-2.7 I I 6.2' 4. 6.5 4.5 X devi tion -100. I M0 0.668 0.879 0.668 from expected -'o
'6.2 6.2 6.2
1 2 3 4 S 6 7 9 10 .11 12 13 14 15 A I. _ - 0.706 0.902 0.700 2.7 0.8 _ '9 I 3.6 I 0.713 0.98311.150 1.151 1.137 0.93910.619 B _-I-I-I - 5.3 .L -2.1 L 2.5 1 1.9 ; 4.0.1 a.-1.1 4 a. -0.8.4 -
- J. .L _______ _______ ____ ___ __
_ 0.730 1.020 1.059 1.232 1.292 1.224 1.018 0.922 0.632 D I 0.754 1.5
-0.5 -1.5 0.967 1.081
-0.2 -1.5
-4.0 1.3 1.173 1.214 1.217
-2.0 -0.4 1.8 0.0 4.5 -0.7 1.161 1.065 0.5 -1.5
-0.9 0.942
-1.2 0.980
-1.2 0.812 0.606
-0.8 -1.0 0.874 0.818 0.509 0.701 1.084 1.131 1.261 1.253 1.267 1.086 1.094 1.017 E 1.6 4.7 2.6 5.0 -1.B -1.7 -3.5 -1.8 -3.3 -4.1 -8.6
-0.3 1.9 1.032 1.174 1.315 1.361 1.331 1.1250 1.126 1.057 0.980 0.931 0.910 0.829 0.762 F 6.5 2.7 3.0 0.1 -2.9 -4.8 -7.5 -7.4
-2.0 2.1 6.3 9.2 7.8 0.729 1.175 1.299 1.311 1.362 1.303 1.203 1.006 0.932 0776 0.806 0.899 0.941 0.883 0.53:I G 2.9 3.9 -1.7 -6.9 -4.9 -4.0 -4.1 -6.2
-0.9 -1.7 0.7 1.8 7.4 7.2 6.1 I
0.963 1.217 1.378 1.328 1.240 1.243 1.105 1.034 0.790 0.457 0.684 0.902 0.999 0.918 0.71! Lig 0.1 0.0 0.6 0.6 2.7 3.3 4.2 6.3 4.8 -1.0 -6.8. -4.5 -3.1 -1.7 -3.90 W- 1.264 1.203 1.035 0.867 0.744 0.809 0.901 0.957 0.927 0.5S50 0.737 1.199 1.303 1.297 1.300 0.2 0.3 1.0 0.6 2.5 4.0 1.6 5.9 -3.3 -5.7 -6.5 -4.7 -2.4 0.6 -1.6 1.^93 1.161 1.268 1.248 1.257 1.162 1.112 1.011 0.981 0.932 0.928 0.838 0.795 I K 1.4 -1.4 0.2 -2.8 -3.0 -6.5 -3.5 3.8 0.9 2.4 0.1 1.8 -1.0 0.704 1.053 1.136 1.211 1.228 1.151 1.096 1.107 1.074 1.017 0.921 0.807 0.527 Ii 0.1 -1.0 2.0 0.5 0.6 -4.7 -0.8 -0.5 1.9 2.0 1.9 -5.4 -5.4 0.737 0.473 1.084 1.188 1.202 1.224 1.188 1.062 0.941 0.837 0.604 C& c c 0X
-1 -03 2 --0. -1.4 0.6 -1.6 -1.8 -1.8 2.3 -1.4 el n X c 0
0.714 0.981 1.075 1.236 11.293 1.141 0.937 0.903 656 N -2.5 1.7 I 1.9 -2.6 -8.6 -2.6 __measured - n n x
-2.6 -5.3 _-3.0 n I' .91.4 _ measured:- rO.OOO C a n r f n :1 0.979 .142 1.149 1.081 0.867 0.605 ___
P _1 W.U.UW
-5.2 -2.51. 1.8 1.7 -1.1 -8.6 -'3.0 X deviation - 1[.0 I (12.a 03 from expected
- 9. I - 7 -
0.701 0.911 0.689 1.% R 1.9 1.8 1.9 0
< ( C
.~~~~~~ . i 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A -~ § * , . l l I
- 2 I -3.7 4.
.TO.6C 16 0.80210.627 I I.
-3.3 4
-0.4
-_ _L .- L .
O..64 0.97171.011 1.061 1.033 0.914 0.611 7.5 .'0.'0 -1.0 . __-1.1 - __ _ 1.2 . -0.2 I2.0 C -~ , . . 0.639 0.9151 .97511.148 .l951.i170 0.990 0.925 0.643
'.4- 0.4
- VS.- 4--V.-Va_
.-1.4 -0.5 -0.1 I 1I.4 1.4 1.4 1
'3.1 0.64 0.866' O.96ft1.11911.166 1.214 11.157 1.098 0-.9761 0.86210.6044 I -,i D ~ I-.-§
, r%Id-I'd 2. k -.h.
2.3A I~7-I 2.6 1.4A -2.2 O. S ,1.8 .3.2
. v.. Id I *-, _' -I - &_b_ i., , ._ I ._ . _. ___' -
E - .0.611 0.937 0.996 1.125 '1.126 1.195 1.079 1.167' 1.093 I1.101 0.96010.910' 0.572 2: 1 281 2. - 0 .9 . -0.Q -1Li5 -2.8 -O.'S I .1 ILL -- 02 -4. - F- 0.897 1.001 l.179 .11.167 1.205 1.146 -1.189 11.176 1.192 1.098 1.08810.956 0.883
-2.1 0.8 5.'O 3.7 1.6 0.4' ,1.9 3.0 0.6 -2.4 -3.1 -3.7, -3.6 IG- 0.628 1.009 1.149 .142 1.204 1.146 1.166 1.074 1.202 1.140 1.132 .114 1.142 1.009 0.612
-0.3 -'1.2' -0.5 0.2 1.7
, 0.4 0.8 2.1 . 3.9 -0.1' -4.4 -2.3 -1' 0 -1.2 -2.8
- .11 0.841 1.068 1.185 1.190 1.087 1.175 1.067. 1.165 1.096 1.172 1.027 1.171 1.183 1.079 0.822 1 .
LI( 1.4 -0.3 . -1.0 . 1.1 e 0.1 0.7 1.4 4.5 4.2 0.5 -5.4 -2.6 -1.2' 0.7 . -0.9 0.639 1.019 -1.144 1.128 1.184 1.160 1.195 1.092 1.112 1.087 1.120 1.107 1.142 1.047 '0.635 3.4- -0.2 -
-0.9
-1.1 -0.0 - 1.6 3.3 -
3.9 -3.9 -4.8 -5.4 -2.9 -1.0 2.5 0.9 0.921 -0.99611.137 11.119 1.18811.124 L1i69 11.115. 1.168 11.090 1.09010.952 0.911 0.5 . .3.3 . 1. 2 2-0.511.0.21 -1.6 _ 0.2 -2.3.
-1.51
-3.1 2.9 -4.2 -0.5 l l L 0.6341 0.939 1.008 1.112 1.115 11.131' L069 1.166' 1.123 1.118 0.978 0.893 0.586 5.9 I 3.014. 3.8 4. a.1 44.j4 .- 08 -4. -1. 6 1-1.5, 4-1-4-1-4--0.2 - 0.7 - 0. 1 -2'.1~ -2.1 M4 -
0.644 0.866 0.963 1.105 1.124 1.204 11.130 1.099 0.982 0.888 0. 644 0 c- Pa _ 3.3 12.3 -0.8 P-1.61 -1.5. 0.01 -0.9 -2.1 1.2 4.8 3.3
-_I-I. - + - -
C t N -I 0.632 0.884 0.96611.168 1.219 1.152, 0.936 10.925 0.678 D1 t2 if-1.3 -3.1 -2.7 1.3 1.9 1 -0.1 -5.8 1.4 8.-7 KEY St I- 'a %A _ _ _ _ n V. 0.581 0.89111.046 .100 1.036 0.863 0.608 measured - 0.0001 p X deviation -100.0
-3.1 -2.7 2.5 2. 1i.4 -5.81 1.5 F1 >
_. _ _l _4 _ _ __ from expected _ e0 R .-. 0.658R
--- 0.867 10.658 4.5 1 4.5 .4.5. 0a
, I
1 2 3 4 5 6 7 8 9 1H 11 12 1.3 14 15 A - I4 q. - - I. 4 I O..73 0.85110. 3.1 _ . _-~9 1.31 t 651 1., , ,? 4 .- , ., . 0.674 0. 35 1.093 1.079 1.0045 0.837 0.531 2.21 -- 3.5-i 2-.0 1.7 3.3Y -? -1.8 _S
-1.7 __
0.716 0.98811.027 11.179 1.218 1.128 0;914 0.780 C1.495 C $ I ,
-0.8 -0.41 -1.41_
-2.4 -4.0 1.3I - 1.7 - 4.5T 0.4 V
- - 4444 0.757 0.959 1.066 1.155 1.174 1.154 1.084 0.960 0.76710.589 0.406 0 2.4 0.1 -1..3 -1.3 -0.7 -0.5 , 1.1 1.0 -0.11 -0.7 -1.4 Il -I - I I 4 +- t -I *~
E - l 0.71011.089 1.127 1.254 1.253 1 1.2621 1.078 1.062 10.9251 0.75410.5391 0.5101 0.339' 0.9 2.4 1.6j- 4.8 3.3 5.8 -0.7 0.2 -1.11 -0.21 -2.2 , -3.1 -6.0 F l 1.053 1.183 1.306 1.365 1.350 1.280 1L.153 1.077 0.945 0.71110.3651 0.496 0.525
-0.4 2.4 5.3 8.9 8.3 7.1 2.51 2.5 0.7 -1.3 -4.5 -7.4 -7.3 G - .0.743 1.186 1.307 1.320 1.384 1.345 1.265 1.099 1.087 0.961: 0.798 0.679 0.694 0.682 0.418 0.3 -1.4 0.8 1.5 7.5 7.6 6.3 2.6 3.1 -0.8 -4.4 -3.5 -4.5 -4.5 -?3.5
.0.980 1.216 1.382 1.353 1.291 1.304 1.188 1.175 1.055 1.014 0.851 0.863 0,862 0.776 0.603 wJ - 0.9 -1.0 -0.2 1.2 4.3 4.1 4.0 5.3 2.2 -0.9 -6.3 -4.4 -4.7 -3.0 -5.1 0.751 1.202 1.309 1.330 1.362 1.339 1.290 1.181 1.078 1.036 0.976 0.940 0.906 0.846 0.500
. 1.0 -0.8 0.1 1.2 4.2 4.8 5.3 5.4 -4.8 -5.3 -6.3 -4.6 -3.5 -0.4 -1.9 1.089 1.169 1. 290 -1.291 1.319 1.254 1.23611.167 1.143 1.04710.98l 0.842 0.773 1.61 -0.3 1.8 0. 1 1.7 -0.6 1.5 -1.9 -0.8 -3.61 -3.31 -6.7 -3.9
- -, . _ . _ . - _ I l L 0.71911.078 I1.152 1.23411.265 11.235 11.193 1.223 1.1871 1.11310.981 10.8361 0.532 0.3 -11 0.6
.,* 1.0 -1.2 -3.8. -.-.0.51 -0.3 - -.1.4 . 2.0 2.0 1 -3.21 . -
0.756 10.990 1.11211.228 1.266 .297 1.222 1.149 1 01010.879 0.625 M -0.9 -0.8 -2.3 -1.8 -1.6 .0.1 -1.8 -1.7 -1.7 1.0 -2.6 CD 0 W P-
- r- - I- I 4- 4I I- .a *1 0.739 1.01811.123 11.283 1.35011.205 1.014 0.96010.682 ba
-2.?7 -5.5 -z.7 0.6 0.6 -3.2 . -7.4 -3.7 0.1 KEY 0 n ,X .u v I~_
0.668 1.023 1.191 1 1.201 1.141 0.934 0.640 _ measured - l0.0001 P
-5.5 -2.7 1.0 I 0.9 -1.5 -7.4 -3.7~ X deviation - 100.0,° V , I
- I- II - I- I- -
from expected . D. R 0.7361 0.95910.727 C 0 1.8 1.8
-1.
t.,, IC I ( C.
(IIC. . i i
; 8 9 1 I 1' 2 3 4 5 6 7 1.1 1 lo 11 13 . - 14 15
.- .1 A . I -1
~O:5_----- 6.774 0. .90 590 A
0.600 0.916 0.976
-5.6 -5.5 1.008
-506 0.976 0.916 0.600 L
B :_______ 1.2 3. -5.9 -3.5 1 2 1.2 C , 0.699 0.913 0.998 1.108 1.088 1.108 0.998 0.913 0.699 12.8 12.8-1.2 1.2 --3.6 - -8.8 -3.6
. 1.2 1.2 12.8 _
_____0.670 0.905 0.987 1.141 1.142- 1.218 1.142 1.141' 0.987 0.905 0.670 . D
- 8.0 7.1 2.0 1.6 0.1 1.0 0.1 1.6 '2.0 7.1 8.0 _
E 0.669 0.969 1.033 1.108 1.111 1.150 1.078 1.150 1.111 1'108 1.033 0.969 0.669 12.8 7.4 6.7 -0.6 -1.8 ~-3.6 "-1.4 -3.6 -1.8 -0.6 6.7 7.4 12.8 P 0.846 0.968 1.151 1.120 1.187 1.124 1.178 1.124 1.187 1.i20 1.151 0.968 0.846
- -6.S -I.8 2.4 -1.0 -1.4 -2.5 -0.1 -2.5 .4 -1.0 2.4 -1.8 -6.5 _
G 0.585 0.946 1.143 1.167W L.195 1.139 1.159 1.060 1.159 1.139 1.195 1.167 1.143 0.946 0.585
-6.5 -6.5 -0.6 2.3 -0.2 -1.2 -1.1 -0.4 -1.1 -1.2 0.2 2.3 -0.6 -6.5 -6.5 0.770 1.002 1.189 1.239. 1.135 1.236 1.113 1.162 1.113 1.236 1.135 1.239 1.189 1.002 0.770 w -6.5 -6.5 .- 0.3 2.8 ;3.9 48 ' 4.6- -2.9 .4.6 -4.8 3.9 -:2.8 -0.3 -6.5 -6.5 K-0.58 0.946 :1.143 1.167 1.195 1.139 1.159 1.060 1.159 1.139 1V.195 1.167 1.14 3 0.946 0.585
-6. -6;5 -0.6 2.3 --0.2 -1>.2 -1.1- --0.4 -1.1 -1.2 ' 0.22-2.3 -0.6 -6.5 -6.5 0.846 0.968 1.151 V.120 1.387 1.124 1.178 1.124 1.187 1.120 -1.151 0.968 0.846 K
-6.5 -1.8 2.4 -1.0 -1.4 -2.5 '-0.1 -2.5 -1.4 -1.0 2.4 -1.8 -6.5 L r 0.669 0.969 i.033 1.108 1.111 1.150 1.078 1.1!u 1.111 1.108 1.033 0.969 0.669 12.8 7.4 6.7 -0.6 -1.8 -3.6 -1.4 -3.6 -1.8 0.6 6.7 7.4 12.8 M
- 0.670 0.905 0.987 1.141 1.142 1.218 1.142 1.141 0.987 0.905 0.670 0, 0 ,-V o C C ri 8.0 7.1 2.0 1.6 0.1 1.0 0.1 1.6 2.0 7.1 8.0 0 D X:=
N _10.699 0.913 0.998 1.108 1.088 1.108 0.998 0.913 0.699 ~ -. -'.i~ 12.8 1.2 1.2 -3.6 -8.8 -3.6 -1.2 1.2 12.8 KEY 3 P- a- ,I C,0 0.600 1.008 0.976 0.916 0.600 measured p _ 1.2 0.916 0.976 1.2 -3.5; -5.9
-1.590 0.7579
-3.5
-. 90 1.2 1.2 X from expected
-0.oo deviation -100.0 3 rl &
C-tn w
--5.6 1-5.5 -5.61 0 I, - .- .-
I 2 3 4 6 7 9 10 1 11 12 13 1 14 is A - I 4. I4 -4. 4 0.66910.881 0.669 7.0 6.9 7.0 - B ~4I4I _ I I 0.592 0.903 1.060 1.149 1.060 0.90310.592
-0.2L -0.2I 4.7 I17.24. 4.7 4. -0.2 -0.2
_______ I I C ~~-4I I - . O.b35 10.900 0.984 1. 128 l. 299 1. 192 0.98410.900 10.635 I-. 1 .. -An I . en
- R8. A£11 -0.2 -0.2 I 2.5 I. _._II4 .- _
D - 4 - 0.618 0.858 0.974 1.146 1.148 1.233 1.148 1.146 0.974 0.858 0.618
-n. I I I. i6 4. no;f4. 2.8 I 0.7 I 2._ 0.74 2.014 _0.6 II 1.6 4. -0.34 4
E ,-4-. 0.6107 0.897. 0.981 1.106 1.108 1.144 1.039 1 144 1.108 1.106 0.981 0.897 01607 _ IL-v, ,, - _n f. 5L.L I LAl _n I.I
-7 1 -A I _4 -A-I ---' 4 --
-4A1 4 -2.I 4 _no A 4
1 -1.t -0.6 72.5
-.. . -. -. 4 F I 0,962 1 1201.114 1.164 1.092 1.108 1.092 1.164 1,14 j1.120 0.96210.903
-0..2 -2.4 -0.11 -1,6 -3.3 -5.3, -6.0 -5. 3 -3.3 -1.6 -0.3 -2.41-0.2
-I 4~ - -- - - - - -2. -0.2 -
G - 0.624 1.010 1.151 1.141 1.218 1.117 1.129 0.958 1.129 1.117 1.214 1.141 1.15111.101 0.624
-0.2 -0.2 0.1 -. 1.8 -3.2 -3.7 -9.9 -3.7 -3.2 1.8 0.0 0.1 -0.2 -0.2 0.822 1.070 1.196 1.212 1.14/ 1.230 1.098 1.045 1.098 l.230 1.147 1.212 1.196 1.070
-0.2 -0,2 0.882 w 02 O, S.O 4,3 32 7.5 3.2 - r.1 5-n f.l n2 . Q OJ 0.624 1.010 1.151 1.141 1.214 3.117 1.129 0.958 1.129 1.117 1.2'4 1.141 1.151 1.010
-0.2 -0.2 0.6:4 0.1 0.01 1.8 -3.2 -3.7 -9.9 -3.7 -3.2 1.8 0.0 0.1 -0.2 -0.2 X 0.903 1O.9h2 1.12011.114 1.164 I.09211 108 1.O'2 1.16411.114 1.120 0.96210.903
-0.2 1-2.4 --0.31 -1.6 -3.3 I -5.35 -6.0 -5.3 -3.3 -1.6 -0.3 -2.4 -0.2
- I- 4-4 4-4-4 ---- 4-4 -
0.607 10.897 0.981 1.106 1.108 1.144 1.039 1.144 I-I.- A L 1.108 1.106 0.981 0,897 0.607
-4 2.5 -0.6 44 1.41 -0.8 -2.1 -4.1 -4.9 -4.1 -2.1 4. -0.8 1.4 -0.6 2.5
- .1 4 4 I. 4 1. 1 & -
M 0.618 0.858 0.974 1.146 1.148 1.233 1.148 1.146 0.974 0.858 0.618
._. -0. 3 1.61 0.6 2.0 0.7 2.3 0.7 2.0 0.6 1.6 -0.3 4 0 .-_
. _.__ . __ . -- i . - - -- - ob C C Ja 0.635 0.900 0.984 1.19811.299 1.198 0.98410.900 0.635 C 4 *i 2.5 -0.2 -0.2 4.2 8.9 1-4-I-U 4.2 -0.2 -0.2 - 2.5 .1 4-4. -4.- KEY P. .. 0.592 0.903 1.06011.149 1.060 0.903 0.592 measured - 0.000
-0.2 -0.2 4.7 7.2 4.7 -0.21 -0.2 2 deviatLon - 00.0 n
- -- t - p-v - l X
V - x M0 Rn 0.669 0.881 0.669 LroW expected 3- - 7.0 6.9 7.0 1' I C (
5.1.4 Power Coefficient Measurement The power coefficient or power defect is the reactivity change associated with a correspoding core thermal power level change. It is the combined effect of the Doppler defect and a corresponding Moderator temperature defect. Data necessary to calculate the power coefficient of reactivity has been obtained during the power escalation phase of the start up test program through 50X F.P. vith the balance of data to be obtained as higher power levels are reached. Power defect measurements are made for a number of unit load changes from several initial power levels. Secondary plant calorimetric data arc obtained before and after each load change to determine reactor thermal power output. Reactivity measurements are made on the reactivity computer during load changes and accompanying rod motion. Reactivity effects due to xenon and boron con-centration changes are also considered when determining the power coefficient. Xenon effects are approximated by calculations based on reactor power history, and boron concentration changes are determined via RCS samples obtained period-ically during the test. The data obtained at approximately the 30% and 50% F.P. plateaus is tabulated and reduced on Table 5.1.4-1. The table lists the parameters necessary for determining power coefficients including reactor thermal power, reactivity changes and boron concentration changes. At %302 F.P., power was decreased to %20% F.P. and then increased to 'I30%F.P. The power coefficients wer within the range of the expected values for these power levels. At 1%45% F.P., power was reduced to -.35% F.P., reduced again to "t25% P.P., increased to
-.35% F.P., and finally returned to %452 F.P. The initial thermal power measurement (c.lorimetric) obtained at 1.451 F.P. was in error and was ne.: used 5.1.4.1
so a different conservative value was assumed, based on the NIS reading for the same time. The power coefficients were within the range of the expected values for these power levels also. 5.1.4.2
Pover Ccefficient
-Sumary
, , I
'Station Total Boron Power Coef.
Power X 'Reactivity Conc. (PPX) (PCMIZ Pover) Change. Tested Accept CALOR NIS(ave) -(PCH) RCS - PZR _ Band 33.89 29.10 1038 1049
¢ -10.25 -9.45 20.44 18.75 -137.85 (-11.83) -17.55 20.44 18.75 1043 1050
-14.85- -9.52 32.50 30.25 179.10 (-13.08) -17.68 (45.99) 41.90 45.25 1025 _
-18.14 -9.28 36.10 35.33 -105.21 1016 1048 (-12.79) -17.23 36.10 35.13 ___ _ _ _ _
* -10.31 -9.35 25.20 24.75 -112.40 1043 1028 (-13.71) -17.36 25.20 24.75 I. 36.70 35.35 L53.05 1046 1059
_-13.67 12 .73)
-9.35
-17.36 36.30 36.00 1000 1023
-.- _ _ -10.36 --9.24 46.20 45.30 _ 102.55 999 1002 (-11.97) -17.16
- Calorimetric pover value used at conclusion of one step and at the start of the next atep.
** Perforted 1000 gallon dilution. - -Power Coefficients are calculated both with and without xenon and boron concentration effects.-
Power coefficient values in parenthesis are adjusted for xenon and boron effects. . . . 5.1.4.3
5.1.5 Static RCCA Drop and RCCA Out of Position The purpose of this test is as follows: 1) to determine the response of the incore and excore Instrumentation to a power maldistribution caused by a RCCA being below its bank position, 2) to demonstrate that the core radial power peaking factor limit (Fxy) is not exceeded at '.30% F.P. with a single completely misaligned rod and 3) to determine the integral reactivity worth of the misaligned rod. The test wan performed using the highest worth RCCA (H-10) and then repeated u3ing the RCCA (F-12) predicted to cause the highest hot channel factors. The test was started with the reactor at '302 F.P. and all control rods fully with-drawn except CBD. whith was at 213 steps. Just prior to the start of the static rod drop of RCCA H-10. secondary plant calorimetric data, RCS boron concentration data, a full core flux map and an incore T/C map were obtained. RCCA 1-10 was then Inserted into the core from 228 steps to 0 steps by diluting the RCS boron concentration. Rod worth measurement data was recorded on the reactivity computer _, during the Insertion. At approximately 30 step increments.of rod insertion, incore TIC maps and partial flux maps were obtained. With RCCA 1-10 fully inserted In the core, a full core flux map, secondary plant calorimetric and incore TIC map were obtained. RCCA H-10 was then withdrawn to its full out position While boron concentration was increased. During the withdrawl, data was again obtained as during the RCCA insertion. After allowing perturbations caused by the insertion of RCCA H-10 to return to normal, the test was repeated using RCCA F-12. The response of the incore thermocouple, the RCS loop temperature RTD's, the NIS excore detectors and the incore moveable detectors to the static rod drop was analyzed. Each instrumentation system was able to detect an abnormal rod configuration. 5.1.5.1
Incore thermocouple maps taken during the RCCA insertion vere analyzed for the relative temperature change compared to the base case map vith the'respective RCCA fully withdrawn. The results are shown on Fiit'res 5.1.5-1 through 4. The presence of a perturbation due to the insertion rod is clear in each case. Tables 5.1.5-1 and 2 provide values of nuclear enthalpy rise hot channel factors N and quadrant tilts tained from incore thermocouple data. TheF H values are the deviations from the base case values. Absolute values were-not used since the normalization factors for < using incore thermocouple maps based on'incore flux maps had not yet been determined. Comparison of the F A-,Hdeviations for selected assemblies near the dropped rod vilth their respective symmetric assemblies, indicate a rod misalignment. The quadrant tilts provided from the'incore thermo-couples also properly detected the dropped rod location.' RCS loop temperatures also sensed the change in core power distribution.- Tables 5.1.5-3 and 4 provides a suviary of RCS loop temperature deviations,'from base - case data, for RCCA H-lO and RCCA F-12 insertion. A relative increase in loop A and loop C T VS and aT during rod insertion, as opposied to the relative decrease in loop B Tavg and aT, indicates that the rod insertion within the core is close to loop B in both cases. Figures 5.1.5-5 and 6 are plots of excore detector currents as a function of RCCA position. RCCA H-10 and F-12 insertion plots indicate-a dropped rod in the quadrant of the core nearest channel N42 detector. A sumary of the incore flux nap data is given on Table 5.1.5-5. Measured values for _ and F were increased by 8% to account for manufacturing tolerances and measurement uncertainty. Meauured values'for F" were increased by 4% to account for measurement uncertain'ty. The acceptance criteria for F vith a RCCA conpletely
- .e- :--,
5.1.5.2
misaligned is taken from design predictions. All other acceptance criteria is based on the BVPS Tech. Specs. Although all hot channel factors for the dropped rod were within the boundry for RCCA misalignment analyzed in the FSAR. the predicted value for FNY was exceeded slightly. Investigation into the predicted value for F revealed that the value provided from design predictions was in fact the predicted value for FN and had incorrectly been labeled F . The value provided in the FSAR was also incorrectly labeled as FXY. Since the measured value for for either RCCA inserted was within this predicted value the results were determined to be acceptable. The relative fuel assembly ?owers for the base case vere also all within the acceptable tolerance of the design predictions (see Figures 5.1.3-7 through 5.1.3-10). The increased power maldistribution as the RCCA's were inserted were reflected by an increasing quadrant tilt ratio. Quadrant tilt ratios of 1.1607 for RCCA H-10 at 20 steps and 1.1614 for RCCA F-12 at S steps were obtained. The measured integral reactivity worth of RCCA H-10 was 132.1 pcm during insertion and 130.5 pcm during withdrawal. Although this value met the acceptance criteria of 1.1 x measured value <250 pca, the value was lower than expected. The test was performed at a lower power level (030%F.P.) than originally intended (,.50Z F.P.). This would result in lower measured rods worths. Also, the small reactivity change associated with a single step of the rod increased the uncertainty of the measured value as determined from the reactivity trace. 'using the boron worth determined during the rod drop of RCCA F-12 and the reported RCS boron concentration for RCCA H-10 at 228 steps and 5 steps, an integral reactivity worth for H-10 of 184.1 pcm was obtained and was considered a more conservative value to be used. The measured integral reactivity worth of RCCA F-12 was 145.4 pcm during insertion and 143.8 pcm during withdrawal. This also met the acceptance criteria of 1.1 x measured value <250 pcm. 5.1.5.3
Table 5.1. 5-1 FAH a'd Qr Values From Therm&Kouple Data During RCCA H-10 Static Drop Assembly Near - Acmy FAu - -Symlmet.ic Assemblv FMi RCCA H-10 @ 110 Steps *O5 Steps -Assemblv _ 140 Steps @ Steps H-11 -.090 -. 297 H-5 + .024 +.097 J-10 -. 095 -. 298 C-6 +.010 +.049 K-li -. 047 -. 146 F-5 +.019 +.094 F-11 -. 023 -. 142 K-5 +.025 +.104 E-10 -.023 .L-6 +.08 +.079
. Quadrant Tilts Q
RCCA H-10 Position (ste;'s) -_ .. I 223 140 5 A1 15 A 1 8 15 A 1 8 15 I 0.993 0.987 1.023, 0.959 . 1.119 0.861 H X UH U 1.027 0.992 0.999 1.020 1.107 0.913 R R R 15 1 8 . 8 .15 1 . 8 15 0.990 0.995 1.005 H 0.99 X 1.014 UH 1.02 0.972 H 1.115 X0.838 R R R
. I i..
r I. - NOTE: X is position of P.CCA being inserted into core. I .. .. i I .. . 5.1.5.4
Table 5.1.5-2 F4H and QT Valueu From Thermocouple Data During RCCA F-12 Static Drop Assembly Near Assembly FAH Symuetric Assmbly FAH RCCA F-12 .170 Steps 5 Steps Assembly I 170 Steps ?5 Steps F-1 -.021 -.304 F-5 +.014 +.102 K-5 +.014 +.028 L-6 +.013 +.136 F-13 -.041 -.492 F-3 +.013 +.137 N-10 +.010 +.106 U-6 +.016 +.162 K-3 +.012 +.158 E-12 -.030 -.386 E-4 +.006 +.107 __ ___ _ _ _ _ __ _ _._ _ D-5 +.008 +.084. Quadrant Tilt 5.1.5.5
C "'- ..
.... -C'.
I
.LOOP TAVW DEVIATION FROM BASE CASC RCCA 11-10 PosLion 200 170 .140 l 110 80 s0 50 5 (s3tips). __ - - .
.aRft~ o 1 - -. -
+1.85 +0.3 +0.4 +0.55 +O.9 +0.5 +3.8 Ta +2.35
+2.25 +1.80 -0.5 -I -0.7-- -0.9 -1.05 -1.85 -2.15 TOVR Dev. LooP 2 0.0 l 0.0 O.0 +0.05 '-.55 t1.2 Tavg Diev. Loc-p +2.30 +.75 Base Case Tavg: Loop 1 552.50F-
.oop 2 553.20P Loop 3- 553.65F I.-
LOOP AT DEVIATION FROM BASE CASE RCCA 11-10 -P--iti-on 200. 170 140 110 80 50 20 5 1lI'os
+0.1 -0.05 +0.2 - +0.55. +0.7 +1.55 +1.9S +0.4 AT Dev. Loop 1 0.0 -0.3 -1.45 -1.85 -2.55 -2.55 -3.40 -2.4 AT Dcv. L.cop 2 0.0 -.- 0.35 1-2.5 -0.1- -0.4 +0.1 0.0 -0.75 kT 'Dev. Loo 3 .A lo WV rt11 WV*
Base Case AT:. Loop1 35.90F- I'-ri ri. t-.m
- -1 Loop 2 35.12F Oa 0 I.
.I .
r t0
- I.oop 3 *32.60F :
I .-.I 013 O U1 I. m 0 w
I LOOP TAVG DEVIATION FROH BASE CASE
,CCA F-12 PosLt)on 200 170 140 110 80 50 20 5 (steps)I Tavg )ev. Loop 1 +1.45 +1.45 +1.5 +0.8 -0.35 +0.3 +1.5 +0.7 Tavg, Dcv. Loop 2 +1. 1.2 +0.4_ .3_ -2.1' - 2.15 -2.55 -2.3 Tav& Dav. Loop 3 +1.5 +1.35 +0.5 0.0 -1.65 -1.5 -1.8 -1.5 Base Case Tavg: Loop 1 554.0?
Loop 2 555.2F Loop 3 554.61 IA LOOP AT DEVIATION FROMt BASE CASEt 200 170 140 110 80 so 20 5 RCCA F-12 Positlon (stIeps) va tr ;a H rA n41 tJ AT Dev. Loop 1 +1.3 +2.45 +2.7 +3.6 +4.25 +4.3 +4.25 +5.7 ntb P n nr AT Dev. Loop 2 +0.85 +1.85 +1.2 +0.95 +0.5 -0.75 -1.15 -0.7 ,1~ t-.m 0: V i . AT Dcv. LoR 3 +1.0 +2.1 +1.65 +1.6S +1.55 +0.5 +0.5 +1.1 V 09 m0 VI g~ ID Base Case AT: Loop 1 38. 1? Loop 2 37.2? ~I,rl I.Dop 3 36.0F rt. ( (11 \1 I
Table 5.1.5-5 Statir RCCA Drop Fltx a3p Data Sunmary Base Case Flux Hap Prior to Insertion of cceptance RC~l-lM)eas. Value* -Crierianc 14 2.0735 c 4.4154 Nuclear Heat Flux Hot&Channel Factor: (Z) Nuclear Enthalpy Rise Hot Channel Factor I7 1.4557 < 1.7620 Mjax. Quadrant Tilt'Ratio ' 1.0206 NiA
, : I ' -
Flux Map taken with RCCA H-10 at 5 steps Heas. Valuea ACcetanc ___ ____ ___ ___Criteria Nuclear Reat Flux Hot 'hannel Factor F (Z) 2.2438 4.4177
.. I Nuclear Enthalpy Rise Hot Cbcnnel Factor F 1.5807 1.7592 Core Radial PowetrPeaki'- iactor Limiit (Fxy) 1.6624 1.1 tc:1.64 Hax. Quadrant Tilc Ratio ; 1.1454 l *N/A I Base Case Flux Hap Prior to Insertic.n of , Acceptance
*RCCA F-12-U - ~~M s . . - --u__e_ Criteria I
1 Nuclear Heat Flux Hot Channel Factor F (Z) 2.0869 < 4.4377 Nuclear EnthalpyRise Hot Channel Factor tr 1.4522 < 1.7570 Hax. Quadrant Tilt Rates .. 1.0121 N(A I flux Hap taken vith RCCA F-12 at 5 steps Meas.- Value* Acceptance i Criteria _u Nuclear Heat Flux Hoc Channel Factor F' Nuclear Entbalpy 1U~se Hot Channei Factor (Z) 9_ 2.2946 1.5923 ______ 4.4177 1.7539 i Core Radial Po`erj Peaking Factor LimitW(Fxy) 1.6696 .1.1 to,1.64
!!ax. Quadrant Tilt Ratio - 1.1614 NtA
- F(Z) and F v3lues include-an 8% increa!e to accou'nt 'or manufacturing th1erances aed measurenent uncertainty. rvalues include a 4% increase to account for measurement uncertainty.
5.1 .5.8
Figure 5.1.5-1 Static Rod Drop Thermo-Couple Map for RCCA H-10 at 140 Steps. 1 2 3 4 5 6 7 8 9 10 11 12 15 IM43 I I Loool IA I I I L ' A B C D iKs) I I 4 Temperature Difference (F) RCCA H-10 140 steps T/C Map. Between Base Case T/C Map and 5.1.5.9
Figure 5.1.5-2 Static Rod Drop Thermo-couple Map for RCCA H-10 at 5 Steps. 1 2 10 -, 11 t A B C E
-F G
K L P
.. R
-.Ij I a- . . . . ; e v Temperature Difference ('F) Between Poise Case T/C t!ap and RCCA N-10 @ 5 steps T/C 'ap. 5.1.5.10 . . A,
Figure 5.1.5-3 Static Rod Drop Thermocczple Map for RCCA F-12 at 170 Steps I) A B C D E F C L a H 0 P I K L H N P R IZ - Temperature Difference (-F) Between RCCAF-12 e 170 step; T/C .43p. PJse Case TIC 'bp and 5.1.5.11
Figure 5.1.5-4 Static Rod Drop
. I I . * . Thermocouple' Kap For RCCA F-12 at 5 Steps 1 -. 2 .. ,3 . 4- 11 12 13 14 15
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5.1.6 Special Hot Channel FP.;tors Test As part of an augmented startup and cycle 1 physics program, measurements of K the core hot channel factors (FH and FQZ) are obtained during transient maneuvers with the indicated axial flux difference outside the target band but within the limits specified by zhe BVPS Technical Specifications. These measurements are to investigate and confirm that such transient maneuvers do not result in unacceptable hot channel factors and to provide part of the data base required to allow Westinghouse design code Verification. With the reactor at a stable power level of '47Z F.P., CBD e 184 steps, equilibrium Xenon, Tavg approximately l.5F greater than Tref, and Al within the target band, a QCFH (Map 114) was obtained. CBD was then inserted to 114 steps in an attempt to get Al within 2Z of the AlL in the negative direction. The RCS was eiluted to compensate for reactivity changes. SZ decreased to only -20.74% and P second QCFH (Map #13) was obtained. CBD was then withdrawn to 180 steps and the RCS was borated to compensate for reactivity changes, until Al was within the target'i band. When the power level stabilized a third QCFMI (M!ap 416) was obtained. A fourth (Map #17) and fifth (Map 418) flux map were obtained at two hour intervals after the third map. Throughout the test, sixteen plant parameters were trended to permit a computer simulation of core conditions. Incore T/C maps were taken with each QCFM. Each flux map was reduced using the INCORE computer code to determine the hot channel factors. Table 5.1.6-1 provides a summary of the results of the analysis. All HCF's except F for the most negative case were within the technical specification limits. When F exceeds its limit, the F Z limit is reduced at xy Q least 1% for each 1% Fxy exceeds its 'Limit. AccountimS for this zequirement, 5.1.6.1
the applicable F 7 was still within its limit. At all timcs, 'I remained well within th^ negative region (-4.56 to -20.74%). With the reactor at a stable power level of %482 power, the test was repeated i-lnthe same sequence for positive 'I, Ta was decreased to ,.F below T ref
-avg~
#20)
With CBD @ 164 steps and Al within the target band so that the base'QCFM (Xap couldbe obtained. CBD was then withdrawn and the RCS was borated to maintain T 1n an attempt to achieve a dl of '+28Z. The maximum .I achieved with ARO; was +2.91% and a QCFH (Map 021)-was obtained. CED ,as inserted' to '194 steps' Three and the RCS was borated until At was again within the target band. T/C maps additional QCFX's (Map 022, 23, 24) vere t'nken at two hour intervals. were taken with each QCFM and plant para'eters were trended to provide a computer simulation of core conditions. Table 5.1.6-1 also shows'the results of the positive .I flux map data reduced I-via the INCORE :.ompurer cod,.- All measured HCF's wwere within Tech. Spec. limits.' a At appro.imately 50% F. P., with it at its-most negative value, F' was at I. I
' minimum and FQZ was also at a maximum. When AIWas Vost positive, pAH was a maximum and FQ7 was at a aminimum.
5.1.6.2
Table 5.1.6-1 Special Hot Channel Factors Data Susar, Special Hot Channel F3ctors - Negative Al - 47? Full Pover K) Base (J1) Most Neg tt2) Just in Band (03) 2 Hrs. in Band (04) 4 Hrs. in Band (#5 FgH 'lax. Heas. 1.4104 1.3450 1.3478 1.3938 1.3810 FAH Limit 1.5800 1.5800 1.5800 1.5800 1.5800 FrJ Max. Meas. 1.4740 2.3104
- 1.4015 1.4304 1.4148 Fxy Limit 1.5873 1.7512 1.5873 1.5873 1.5873 FOZ Max. Heas. 2.1226 2.6832 2.1370 2.0759 2.0241 FqZ Limit 4.111 4.111 4.111 4.111 4.1012
- This Fxy is for source 158 R 8HF, the measured FqZ for this source is FQZ - 1.9494 the limit for FQZ reduced to account for Fxy being 32% over it's limit is Fq2 - 2.7955.
Special Hot Channel 'Factors - Positive AI - 48% Full Power Base (#I) Most Neg (f2) Just in Band (03) 2 Mrs. in Band (04)4 Mrs. in Band (if F'H Max. Mleas. 1.3639 1.47Aq 1 3RAb I - An 1 1MSO
--- -Z
-'~ - - -~- - - _^ - >- ----------- . ;,
F_ Limit 1.5768 1.5768 1.5768 1.5767 1I-767 \, ) _ _. 4_ = _ 1.4 -- -w Fxv Max. Heas. 1.4999 1.4'84 _ 1.4195 _ 1.4440 1 1.4163 Fxv Limit 1.5844 1.5844 I 1.5884 1.7480 1.7480 FQZ Max. Mens. 2.1646 1.9106 1 1.9710 2.0080 2.0041 FQZ Limit 4.1111 4.0223 I 4.0815 _ 4.0914 4.0914 5.1.6.3
'J
._I
5.1.7 Oynamic&Rot Drop Performance of this test had not been started by September 30, 1976, and hence the results are not:available for inclusion in this initial report. Subsequent to the performance of this test, a revision to the Start-up Report will be issued in accordance with VRC Regulatory Guide 1.16,discussing the performance and results of.this test. - . - I. .
. I I ;
'.1 7 .1
I 5.2 Transient and Trip Tests 5.2.1 General of the Transient operation of the reactor control and protection system and escalation . emergence. power system is tested during station start-up and power to verify conformance with design requirements.
- Each test establishes a transient situation and monitors system responses before, during and after the transients. Data analysis verifies predicted for station behavior during load changes or provides information necessary control set point adjustment. Transient performance is optimized under the most adverse but acceptable operating conditions.
of the
- The loss of all off-site power at load test was completed and portions
! load swing test were performed through 50% FP. The balance of the transient and trip tests will be completed as higher power levels are reached. I I II iI II I Z 5.2.1.1 II
5.2.2 Load Swings The NSSS transient response, including the automatic reactor control system was tested to verify satisfactory performance when + 1O0 step load changes were made via the turbine generator at a 2001mmin. rate. This test was performed at 302 FP and will be performed at the 75% FP and 10OZ FP levels as they are reached. The first transient was a102 step load decrease from 30Z power. Various primary and secondary plant'parameters were monitored using multi-channel recorders, installed plant instrumentation and computer trend. '(See Table 5.2.2-1). The second transient was a step load increase to 30X power with important'parameters again beige monitored. When-this test is repeated for higher power levels, power
- distribution measurements'will'be obtained in'addition to the data taken at the -l lover power level. -- -
Data acquired at the 30% FP level den^nstrated satisfactory transient response of the primary and secondary plant. The reactor and turbine generator did not
- trip, safety Injection was not initiated and neither the steam generator or
- pressurizer relief or safety valves lifted. Nuclear power overshoot/undershoct was less than 3% for the load swings and since no sustained or diverging oscillations of plant variables were incurred, manual intervention was not required to reach steady state conditions. 5.2.2.1
Table 5.2.2-1 Load Swing Test Parameters K) Table of Monitored Purameters Load Swing Test II I TI Power Range Channel I Q Xain Generator Gross MW II Power Range Channel 2 Q Reactor Total Thermal Q Power Range Channel 3 Q Boric Acid 1 Hour Flow Symmetric Offset Q1 Scm. Gen. A Feed Wtr. In (I) I Symmetric Offset Q2 Srm. Gen. B Feed Wtr. In (T) Symmetric Offset Q3 Stm. Gen. C Feed Wtr. In (T) Symmetric Offset Q4 Stt. Gen. A Stm. Out (P) Control Rod Bank D Step Count Stm. Gen. B Sem. Out (P) Auctioneered Tavg Stm. Gen. C Stm. Out (P) Reactor Coolant Tref Stm. Gen. A Stm. Out (F) Reactor Coolant Average DT St=. Gen. B St%. Out (F) Reactor Coolant Loop Temp. Sem. Gen. C Stm. Out (F) AT Setpoints, O.P. and O.T. Scm. Gen. A Feed Wtr. In (F) Stm. Hdr. Press (P) Scm. Gen. B Feed Wtr. In (F) (P) - Prcssure Stm. Gen. C Feed Wtr. In (F) (T) Temperature Stc. Gen. A N.R. (L) (F) (L) Flow Level Sc.. Gen. B N.R. (L) Scm. Gen. C N.R. (L) J 5.2._.2
5.2.3 'Large Load Reduction Performance of this test had not been started bvy Septemrbvr 30, 1976, and hence the results are not available for incluson in thi's initial report. Subsequent to the performance of this test, a revision to the Start-up Report will be issued in accordance with NRC Regulatory Guide'l.l6.'discussing the perforrince and results of this test. 5.2.3.1
5.2.4 Net Load Tr-ip Test Performance of this test had not been started by September 30, 1976, and hence the results are not available for inclusion in this initial report. Subsequent to the performance of this test, a revision to rlhe Start-up Report will be issued in accordance with NRC Regulatory Guide 1.16.discussing the performance . and results of this test. 5.2.4.1
5.2.5 Plant Trip From 10OZ Power
- Performance of this test had not been started by September 30, 1976, and he~nce
- the results arc not availab!.e for inclusion in this initial report. Subse~uenL
, f to the performance of this test, a revision to the Start-up Report will be
' ' issued in accordance uith NRC Regulatory GuIde l1.6,discusslng the performance t t and results of this test.
i
. 1 5 t
- f. i ti
- S I ,
.I I
- 3
5.2.6 Plant Load Follow Demonstration Perform.2nce oa this test had not been started by Septeirber 30, 1976 and hence the results are not available for inclusion in this initial report. Subsequent to the performance of this test, a revision to the Start-up Report will be issued in .accordance with NRC Regulatory Guide 1.16,discussing the performance and results of this test. 5.2.6.1 I
l 5.2.7 Loss of'All Off-Site Power at Load I A test was performed to verify the nbility of the station to sustain a unit trip from a power level'of >10Z when coobined with a loss of all off-site powe: sources and to verify that the station can be brought to a hot shutdown condition in a safe manner using emergency'diesel power only. With the'station operating at alsteady power level of 30% and with all auxiliary loads being supplied by the turbine generator, ail station service was prevented
; from automatic transfer to':4fi-site sources by opening breakers for the system - station service transformers. '
X 9. - . ..;- ,*. A unit trip was initiated by manually tripping the turbine generaror requiring the emergency diesel generators to pick up and supply all vital loads. This condition was maintained until no load Tavg (547 F) was reached and stable station conditions werei established. A variet.7 of parameters were recorded
' throughout the'test to monitor their response through the tran.sient. Table 5.2.7-1 'contains a listing of those parameters m6nitored during 'the test and contains a summary of the recorded responie during the transient 2nd station
' recovery.
A review of incore T/C data revealed no abnormal temperatures within the reactor following loss of power and reactor trip. All associated reactor
-coolant tenperature response was as expected during the transient. Pressurizer pressure and level behavior was as anticipated, and the pressurizer safety
- valves did not lift.
?Most of the problems that were experienced following the unit trip resulted fron less o! .ation instru-ent air pressure to several valves since the station 5.2.7.1
air compressors are not supplied by a vital bus. These problems include the following:
- 1) Loss of component cooling water to the sample coolers caused overheating of the sample system ion exchangers.
- 2) Loss of normal RCS letdown flow when a valve in the letdown path drifted closed caused a letdown relief valve to open. Letdown was suspended.
- 3) Loss of RCP Stator and Thermal Barrier Supply Valve, with failure of installed valve lock-up devices, made RCP seal injection flow (charging) mandatory to prevent RCP seal damage. Charging was maintained.
Other problems resulted from loss of electrical power including the following:
- 1) Excess letdown heat exchanger operation could not be initiated because electric power was not available to operate valves required for that evolution.
- 2) Loss of power resulted in the loss of both the A. C. motor driven Air Side Seal Oil Pump and the Hydrogen Side Seal Oil Pump of the main generator. Seal oil pressure was maintained by the D. C. motor driven Air Side Seal Oil Pump. When A. C. power was restored, the D. C. pump was not immediately shutdown apparently resulting in exce..s seal oil pressure and an injection of oil into the main generator frame. The main generator incurred no damage as a result of this incident.
Station shutdown from 30%. power concurrent with a loss of all off-site A. C. power was accomplished in a safe and orderly fashion with all transient response requirements satisfied. 5.2.7.2
f- TAP' I Requilred Recorder . Paarameter dentificntion _ Response (time zero Is trippoint)
- -e initially 555F immediate drop to 547F, decreasc to:532F at TAV__ _ntoneeed_' TR-RC-408 15 min.-minimum, < 530F, occurred after 60 min.,
Initially 555F, Immediate drop to 549.5F, decrease to 547F ___ Rt'F-..TR-RC-40R . @ 25 min. erritic decrease from t 574F to " 560F in "l lotop ff1 TR-RC-413 1v 20 min. * - -: C Loop ll1 TR-RC-410 immediate drop from 532F to 515F.* erratiedecreave from 5S74F. to % 560F in Tit Loop t3 TR-RC-413 ' 20 min.* Loop 73 TR-RC-410 errntic decrease from' 535 to 515F in 20 min.*
.-Thermocouples C-8 Analog trend erratic decrease from 570F to 551F in " 20 min '
(core E-12 1 -recorder (A) erratic decrease from 568F to 550F in ' 20 min location) G-l Analog trend errativ decrease from 572F to 551P in 15 min ,i. C-8 recorder (B) -erratic decrease *rom 558F to 548F in u 15 mini
; immediate drop from 2250 psig to 2170 psig, eratic increase to
- 4
- wtI 2295 psig . 45 min. decrease to 2120 psig @ 70 min, increasing Pressurizer pressure 1PC-RC-444 until end of recovery.
.underdamped transLent response to a step decrease In progranmed I'ressurizer leveli LR-RC-450 - ' zr.level from 29% to 20% - 18.5%, muinimumlIevel reached.
Tiurbine speed SnsR-Tn-/o; smooth constdown increa.se to;,' 1000 psig from . 980 psig, fluctuate e 11000+/-25 psig Steam header press. : PR-MS-474 for 60 min,'remained with +/-50 psig for the duration of recovery. reedwater flow 1A FR-MS-478 initially 0.99 rnpplh, dropped off scale e trip. - Sta. immediately dropped to 0.88 mpph from 152 mmph, drifted down St cam flow lA FR-4S-478 slig1utly through remainder of recovery. Steum generator . Immediately dropped to 6.5% from 43%, increased to 26Z In 1,20 rmin, 1evel 1A FR-HS-478 returned slowly, with little fluctuation, to 40%'durung recovery. A Temperature indication is not necessarIly representntive of RCS loop conditiono, since tC reed tCis ri.w utcps when the tint trips. wfl. o r r1 73 f3
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5.3 CONTROL SYSTEMS DYNAMIC RESPONSE TESTS 5.3.1 General KIj During initial power escalation the proper dynamic response of the Reactor Control System, the Steam Generator Lev.l Control System and the Steam Dump Control System was verified. The Reactor Control System was tested at the 30% FP plateau to verify its ability to maintain RCS T39V within its programmed value. After verifying the proper operation of the system when switching the rod control Into the automatic mode with stable conditions, the transient response of the system to 6F deviations generated between the RCS Tavg and T ref was observed. The proper operation of the Steam Generator Level Control System was also tested at the 30% FP plateau. With steady state conditions established, each steam generator was subjected to transients using simulated signals to optimlza _ the level and flow controller settings. The Steam Dump Control System was tested at approximately 10% FP. This test observed the response of the Turbine Trip Controller to deviations between RCS Tavg and no load Tavg, the response of the Load Rejection Controller to deviations between RCS Tavg and simulated programmed Tavg, and the ability of the Steam Header Pressure Controller to maintain a stable Steam Header Pressure. In addition to these tests, further information on the characteristics of various automatic control systems are obtained during plant trans.ent testing. 5.3.1.1
5.3.2 Automatic Reactor Control. The reactor rod control system enables the reactor power level to automatically follow turbine load changes as sensed by the control system. The reactor avg rod control system Is designed to handle step load changes of up co 10%Y or a ramp change of 57. per minute from 15 to 1007. fu~llpover, without a reactur trip occuring, steam dump valve actuation or pressurizcr -relief valve actuation.. operation of the reactor rod control system Is as follows:, Actual RCS T-y is avavg
- signal Is generated. This error signal is then added to the error signal-obtained by comparing the indicated rate of change of reactor power and the turbne lad.The et erorgenerated determines the speed and direcino full length rod travel. The system is designed such that T avg is maintained.
within a control deadband of + 1.5? during automatic rod-operation. The reactor rod control system was tested with the reactor at approximaztely 30Z full power. While maintaining T avg w ithin + iF of T
- refad adwt~h ihh rsuie rsuie pressure and level control. steam generator level control and steam dump control systems in the automatic mode, the rod control system was switched from manual to automatic and various station parameters were monitored.. The-rod control.-
system was then returned to manual and T was adjusted to 6F higher than TC eve The system was again switched to automatic and the .response of-the rod control
- system dnd the station (in particular T avg pressurizer pressure and level,-and steam generator level),were~monitored during the~transiont. After reaching a..
stable T the rod control system was returned tozmanual and T was lowered avg..* ... avg 6F below T The rod control system was again switched to autonatic and the ref' response of the rod control system and the station monitored. 5.3.2.1
When switching rod control to nutomatic with Tavg within + 1.5F of T f no noticible variation of pressurizer pressure or level and no varLF:ion of T avg was observed. During the transients the rod control system acted to i=mediately return T to within its deadband with no appreciable overshoot or instability observed. avg A summary of the pressurizer and steam generator controls response during tne transients is provided on Table 5.3.2-1. During the performance of the test, the transients were considered ended shortly after recovering T avg to its programed value, hence the final values on Table 5.3.2-1 are not Indicative of steady state values for all parameters listed. Pressurizer pressure decreased noticablity during the transients (expected variatior during normal operation is + 10 psi) however, the lowest value was still approximately 245 psig above the low pressure reactor trip point of 1945 psig. Pressurizer level response was satisfactory. The programed level setpoint for the pressurizer is a function of Tavg and thus variations with changes in Tavg were observed. Also, the initial value of pressurizer level for the two transients
- were taken at different values of Tavg and thus are notably different.
. All three steam generator level responses were satisfactory. The maximum level deviations recorded during the ramp up transient were attributed to the swelling of the water in the steam generators during the initial 6F change in T V.. The e maximum devia:ions recorded during the ramp down were not considered significant. 5.3.2.2
Table 5.3.2-1
. ,. I Pressurizcr and Steam Generator Controls Response I' I . ,
.n ." 1 * -
I Iw . - - - , , I i@, . I . . ;7 - r . -;_ . 11
. I . . . .I . - .. I I t, r - !' I '
- .-.. 5.3.2.3
5.3.3 Automatic Steam Generator Level Control The automatic steam generator level control system is designed to maintain each steam generator level in agreement with a set point programmed according to turbine load. A flow balancing controller determines ant differe:ce between .steam and feedwater flow and regulates a feedwater contl'i valve to correct an% flow imbalance. The level controller determineF the difference between the measured level and programmed setpoint level and provides a second input to the feedwater flow controller. While the reactor was at 30% power with the generator connected to the grid and steady state conditions established, each steam generator was subjected to level and steam flow transients using simulated test signals to optimize the level and flow controller settings. Level and flow controller output signals, steam and feedvater flow rates, flow error signal, and stea-m generator level were monitored on a recorder to evaluate controllers response and stability. <2 The level reference signal to each level controller was replaced by a D. C. test signal. With the test signal adjusted to match the programmed level signal to the actual level signal, each level controller was switched from manual to automatic mode and the ability to maintain stable feedwater flow rate and steam generator level was observed. The test signal was then varied for 5% increase and decrease of level span and the response for each change was observed. Each level controller was adjusted as necessary to V=m1L t:.C overshoot to less than 5% of the level span. Overshoots of 4X, 3%, and 2% were observed for the 1A, 1B, and IC steam generators, respectively. A similar test was performed for each steam generator flow controller. A test signal was used to vary the steam flow signal by 3% to observe the ability 5.3.3.1
flow imbalance and of the controller to correct for steam flow/feedvacer Each flow controller achieve steam generator level and.feedviater flow stability. was adjusted for stable operation. system was to manually The final test of each steam generator level control the programmed-reference increase the actual steam generator level to 5Z above the response.of the systev.- level, then switch to the automatic mode .and observe steam generator level Each level control system was required to return its iI times taken to recover to the programmed level within 5 minutes. Theactual I for the 1A, 1B, and IC from the.5X deviation were %.53.-30, and ^A60 minutes were greater than the required I steam generators, respec:ively. Although the tiures i returning II 5 minutas, the controllers demonstrated stable operation in smoothly level with little or no i the steam generator level to the programmed setpoint I Operating Supervisor I overshoot. Instrumentation perso.nel ani the Station I judged the slow but smooth response acceptable. i u.. performed twice. The i Testing of the steam generator lev,1 control system i steam secured to the feedvater I original p-trformance was conducted with extraction II to the steam generators. heaters, ,resultLng in unusually low feedwater temperatures i and "swell" in the steam I I The cold feedwater,2ccented theeeffects of "shrink'! although stable and able
,generator levels. It 4as believed that the controls, I the allotted oscillation/.
to return the levels to normal setpoints vit'in because of the low gain overshoot/time limit, were still not optimized and flow transients. settings might not respond rapidly enough for more severe previously were obtained. The test was performed n.gain and the results discussed i 5.3.3.2 \I- j
.I ..
5 3.4 Atitnatic Stenn ;thrn Canrro1 The condenser steam dump system is comprised of 13 dump valves divided into four groups. The system is capable of dumping 85% of full power steam flow to the main condenser. This design feature allows a loss of all generator electrical load from full power except station sarvice (about 5% load) without requiring a reactor trip. The rod control system Is designed to compensate for the remaining 10% pover mismatch. The system has two automatic modes of operation, steam pressure mode and Tavg mode. In the steam pressure mode the first two banks of steam dump valves are operated by the steam pressure controller to maintain a constant steam header pressure. In the Tavg mode, two steam dump controllers are available. Thetturbine trip controller operates the first two banks of steam dump valves to decrease Tavg to Its no load value (547F) following a turbine and reactor trip. The load rejection controller operates the first two banks or all four banks for small and large load rejections respectively to restore Tavg to its program value which is dependent on power level. An initial checkout was performed on the condenser steam dump system during Hot Functional Testing. Due to the limited steam supply only selected dump valves were operated with simulated test signals. Checks were performed in the steam pressure and Tavg mode. The trip open time was also measured for all 18 dump valves. After the completion of low power physics testing, power was ascala:ed to approximately IOZ full power and the dynamic response of all three condenser steam dump controllers (turbinc trip, load rejection. and steam pressure controller) tested. 5.3.4.1
The turbine trip controller response was tested by creating a temperature error between RCS Tavg and no load Tavg while in the steam header pressure mode and then switching control to the turbine trip controller. No
- oscillations or Instability were observed in the system response. Since the
- reactor was not tripped when the transient was initiatedTavg stabilized at an equilibrium value of 549.5F instead of the no load 547F TavE value.
The response of the load rejection controller was tested by creating a temperature error between'RCS Tavg and a simulated program Tavg of 547F with the turbine
- latched. The system responded satisfactcrily stabilizing Tavg at 553.2F.
This temperature was slightly above the 5F deadband of 'the controller (547F to 552F) since'some temperature error was necessary to maintain condenser steam ' dump providing a'load for the reactor. The steam header pressure'controller was tested in both manual and auto with no operational problems encountered. The controller maintained header pressure
- at its corresponding setpoint with no instability or oscillations.
Several dump valves in the first bank failed to operate when testing the turbine trip controller. The instrument air pressure to these valves was increased and the valves exercised. The valves operated properly throughout the remainder of the test. 5.3.4.2
5.4 Instrumentation Calibration and Alignment K) 5.4.1 General Several instrumentation systems require data to be obtained during initial power escalation to provide information necessary for their proper calibration and alignment. The NIS power range channel gains are conservatively set prior to initial startup and are adjusted during power escalation based on accurate secondary plant calorimetrics. Process temperature instrumentation and the Reactor Control System are initially calibrated based on design RCS Tava and 4T across the RCS hot and cold leg loops at 100% FP. The projected 100% FP RCS tavg and AT are determined during power escalation based on secondary plant calorimetrics and adjustments to instrumentation systems made as required. The proper calibration of steam and feedwaer flow instrumentation is checked at each power level plateau, and the response of the NIS excore power range detectors to axial power distributions within the core is determined. This report discusses the calibration and alignment of these items only up to the 50% FP plateau. Further analysis will be available when the power escalation program is completed. 5.4.1.
5.4.2 Operational Alignment of NIS The NIS consists of eight independent channels: two source range channels, two Intermediate range channels and four powec ringe' channels. Each of the three ranges (source, intermediate and power) prcvides some form of overpower reactor trip protection. The three ranges overlap, providing continuous protection from the source range (lowest neutron flux) through the power range. In-addition there are four auxiliary monitoring'channels: the source range visual - audio-count rate channel, the source/intermediate range startup - rate channel, the power range comparator channel and the power range flux deviation channel. Each power range'detector is separated into two'ion'chamber sections which transmit two independent current signals.' These two signals are proportional to upper and lower core neutron fluxes. The source, intermediate and power range channel detectors relative positions with respect to the core are shown on Figure 5.4.2-1 Just prior to initial criticality proper metering indication and trip setpolnts for the source, intermediate and power range channels were verified and the power range high I.evel trip was set at 20% FP. The power range channel gains were conservatively calibrated such that 120 mieroamps from both the top and bottom detector sections was equivalent to 120% FP. During initial startup and power escalation, overlap data between the source and intermediate range channels and between the intermediate and power range channels was obtained with acceptable overlap (> 1.5 decades) observed between each range. During power ascension the pcwer range channel gains were adjusted based on secondary plant calorimetrics. Due to the conservative gains, attempts to perform a turbine roll after the completion of low power physics testing
without exceeding the licensed reactor power limit of 5% FP was impossible. The gains were adjusted from 120 microampL to 180 microamps equivalent to 120% F? until further adjustments could be made at higher power levels, based on secondary plant calorinetrics. Detector current vs. percent full power data is recorded during power ascension to verify detector linearity. Intermediate and power range channel detector current vs. high voltage saturation curves will also be obtained. Intermediate range detector compensating voltage effectiveness and source range detector high voltage plateaus will be checked and the intermediate range rod stop and high level trips w): t be accurately set after shutdown from 100% power. Throughout the initial startup and power escalation, power range channel N44 has had excessive noise and Intermittent spikes. The source of the noise has been determined to be the N44 detector upper Ion chamber. Plans have been made to have the entire detector assembly replaced. Further evaluation; and discussion of the response and operation of the NIS during Initial startup and power ascension will be provided after the completion of the NIS alignment testing. 5.4.2.2 I
Figure 5.4.2-1 NIS Detector Locatlions I. IS 1ACKr II - NORTH
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5.4.3 Startup Adjustments of the Reactor Control System The reactor Automatic Rod Control System is initially calibrated for design Tavg value (576.3) at 100% power. This will not necessarily be the operating value of the system. At hot zero power and during power operations data is obtained to determine adjustments to be made to the Tavg program. This will allow full power operation with the maximum allowable main steam pressure (optimum plant efficiency) with-out exceeding the design Tavg or :urbine design steam inlet pressure. erimary system temperatures, steam pressures, and thermal power data are obtained and evaluated to provide a basis for making adjustments to the Reactor Control System. Table 5.4.3-1 provides the data obtained up to the 50% FP plateau. Since the relationship between Steam Generator Outlet Pressure and Power Level is nearly linear in the region between 50% and 100% power, the 100% value of Steam Generator Outlet Pressure is estimated by extrapoliting data obtained during power escalation. The results of this extrapolation based on data up to the 50Z FP plateau were a projected 100Z FP steam pressure of 817 psig. To insure a design main steam pressure of 765.7 to 775.7 psig at 100% FP a Tavg adjustment from 576.3 to 570.36 was calculated. After determining that the moderator temperature coefficient would remain negative at this new lower Tavg the Tavg programgeneration modules (7*-RC-408F
& W) and the overpower AT Trip setpoint reference Tavg modules (TH-RC-412M. 422H and 423H) were resealed and recalibrated to control Tavg in accordance with these results. Additional data will be obtained at 75, 90 and 100% FP and adjustments made as necessary to optimize main steam pressure.
5.4.3.1 i
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- l. I i % Rated Therma1 Pover (Calorimetric)
I o Lfoi:. 0Z (Nom.) 30%(Nom.) O0%(Nom.) 75%(Nom.) 90% (Nom..) _ (Ac ) _ (Act) I..: : . . 2 L.2 (Act) :312 (Act) 5LZ(Act) __.(Act) ___ _ 7_
.r-A KA l. Header A (PSIG) Pr-AV 8 = __8____ _ __-_l_.
STEAfI Ieader B (PSIG) Posr-Adi 11.0 06 7_ 8_.8 .- PRSUEPre- A - 0fj, 906 Ba _33 C (P1 PRESSURE prNQ-A(l. olender NA) A N/ _ .- -- -n - (ACTUAL) Average (rsMC) y i 1 1 1F -, 9 4__ 870.0 ,_l 0W'ER STFAX PRESSURE (PSIC) 775.7 775.7 817 ; _.A EX_1 POd TED 100%X EXTRAPOLnTi 100% POWtERI SATUHATION:T.hIP. (PF) 516.76 516.76 522.7 - _ l YT (SATLRTA10N TEMP. - 516.76 (OF)-' 0 . 5.94 AClIAL (OF) 576.3 576.)j 576.3 _ _ _ : E.XTA;AroATED "Ij TAVG (AUCT) DESTRED 0F) 576.3 576.3 570.36: ; _ _ tAIN _.7 T +.S tV0MTS)
. 57- _3_
n .328 1.076 0.328 1.076
. .70.3, 0.261 1.607
; V 1.0 L.TMIT (VOL.Ts~** 1.6S 1.680 1.680 1.680 1.680 1.680 SETTTIfCS 11I1 ITMIT (VtI.TS)* 2.852 2.852 2.614 l - ,
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5.4.4 Calibration of Steam and Feedwater Flow at Power Proper calibration of the steam and feedwater flow instrumentation is verified at power levels of 30, 50, 75 and 100% FP. Transmitter output voltages and multiplier divider output voltages (voltage proportional to actual flow) are recorded for each steam and feedwater flow instrument at each power level. Concurrent with this data collection, secondary plant calorimetrics are performed using accurate test feedwater flow measuring instruments. Measured multiplier/ divider voltage outputs are compared with expected calculated voltage outputs based on feedwater flows obtained during calorimetric measurements. If deviations exist between measured steam and feedwater flow voltages such that a steam flow/feedwater flow mismatch bicvable trip may occur (which coincident with a low steam generator level In the same loop results in a reactor trip) prior to reaching the next power plateau, adjustments are made to the instrument channel. Fin.l adjustmenra are made at 100.. F? to Insure that the measured multiplier/ divider output voltages are within tolerance of the expected output voltage. Some adjustment of the steam and feedwater flow instrumentation was performed at the 30% FP plateau. Data obtained at the 50% FP plateau indicated presant calibration acceptable for additional power escalation. Further evaluation and discussion of steam and feedwater flow instrumentation calibration will be provided after data collection at the higher power levels Is completed.
5.4.5 Operational-Alignment of Process Temperature Instrume~ntatfin The &TIT avg..control andprotection instrumentation are aligned prior to initial criticality and during initial power escalation. Prior to inicial criticality the zero of the AT and Tavg signal susmators in the AT/T .*. - .avg protection and control channels were checked and aligned. This was accomplished by simulating values measured at the 547F data collection period of the Incore T/C and RTOD
-Cross Calibration Test and making any necessary biasadjustments at the AT and T signal su-Atcrs.T ,
avg The AT across the hot and cold leg loops of the RCS is used as a measure of reactor power in both the AT/T control and protection channels. Therefore,
- ~~avg .-
the AT signal summators must be aligned at power such that with the plant at nominal 100% full power as determined by secondary system calorimetrics, all AT channels will read 100%. For initial power escalation.the AT will be 55F. Temperature data has been obtained during power escalation concurrent with secondary side calorimetric power measurements at powerlevels of.30 and 50% full power. At 75% pouer temperature and calorimetric data will arsin be obtained. The datawill then be plotted and extrapolated to 100% power to determine actual lOO1 AT values and gain and bias adjustments made, as required, at the signal summators. At 100% power additional data will be obtained and the signal summators checked and realigned as necessary.. 5 .4 5.4.5.1 -. . . . - . . - .. ..- , - _I -.. .. ...-.. . - . .... - - I. .. - .. .. .. ..
5.4.6 Incore - Excore Detector Calibration During st;rtup testing data is obtained to determine the response of the NIS power range excore Ietectors to axial power distributions within the core. This information is then used to calibrate the delta flux meters and the delta flux penalty input to the overtemperarure AT setpoint to reflect actual incore axial power distributions. Formal data collection for Incore - excore detector calibration is not scheduled until the 75% FP plateau. A preliminary calibration on the delta flux meter indication was performed at the 30% FP plateau. The delta flux meters were calibrated based on the graphs of detector currents vs. incore axial offsets d provided in Figures 5.4.6-1 thru 4. These graphs were generated from data obtained during flux map no. 7A, 9A and 11A (see Table 5.1.3-1). Power range detector currents obtained during these maps were normalized to 100% F? based on secondary plant calorimetric performed during the performance of these flux maps. Due to limitations in the circuitry, the large deviations between the top and bottom detector slopes on Figures 5.4.6-1 thru 4 made the calibration of the delta flux meters difficult and only a rough calibration was possible. As more data points are available, a more accurate relationship between the detector currents and the incore axial offset will be generated. Data obtained during map no. 6A was also intended to be used for these calibrations but was not used due to the fact that control bank D was inserted to 155 steps. -his deep bank insertion resulted in excore detector currents that were not consistent with other data obtained at this time. This bank configuration tends to shape the flux distribution towards the outside of the core, causinglarger excore detector currents than expected for the respective power level. 5.4.6.1
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5.5 Secondary Plant 5.5.1 General The performance of some secondary side systems is checked during startup and power escalation by the startup test program to verify proper operation of those systems. -The measurements taken'during testing provide information necessary for comparison with design specification, provide baseline data for comparison with future operating data,,and provide data useful for transient analysis. Problems In a number of secondary side systems have caused delays in the startup and pover escalation schedule. Tab1e 5.5.1-1 provides a brief review of some of the major secondary systems problems.
- - ,5.,.1
Table 5.5.1-1 Secondary Systens Problems PROBLEM RESOLUTION Feedwater System: Excessive vibration of Redesign recirc. flow path, the main feed pumps (lA realign pumps and IB) at low flow (recirc. flow) Loss of IA main feed pump Replace 1A main feed pump due to loss of TIC well at implelIcr bearing and subsequent loss of oil Loss of lB main feed pump Replace 1B main feed pump due to loss of thrust bearing pump, impeller and outboard pump seal: Loss of 3A auxiliary feed Replace 3A auxiliary feed pump due to over heating pump, recirc. path under on recirc. engineering evaluation "High conductivity" in Revise Tech. Spec.A secondaiy the main steam generator water chemistry transient limits during power transients Condensate System: "High conductivity" in the Revise Tech. Spec.* secondary condenser hotwell during water chemistry transient limits power transients Chronic clogging of Repeated cleaning of strainers condensate pump strainers Heater drain tank emergency Plug damaged condenser tubes, dump condenser baffle failure repair baffle and add baffle (twice) and resultant condenser supports. tube damage. Circulating Water System: Cooling tower pumps fail Under engineering evaluation to meet flow requirements by approximately 10% F.F. Technical Specifications were modified slightly to adjust secondary water chemistry specifications to more tolerable limits during heat up transients. The change provides for 14 days in the transient limits of 7 m!.cromhos. The previous specification provided for 7 days in the transient limit of 10 micronhos. 5.5.1.2
5.5.2 Turbine Overspeed Trip Test U The main turbine is provided 'ith various protection' trip devices including the turbine overspeed trip >mchanism. This device consists of a weight mounted ' in the turbine rotor extension shaft. The weight is mounted offset from the center of the shaft so that centrifugal force tendsito move it outward. It is held in position by the compression of a spring such that when the turbine exceeds a specified speed (1974 to 1998 rpm) the compression of the spring is
- - overcome by the centrifugal force of the weight on the rotating shaft. The weight moves outward and operates a device that opens the auto stop oil'drain valve. The release of this oil, in turn, causes the electro hydraulic emergency'
- trip fluid to drain which closes all steam valves to tb. turbine'. Provisions'are available to operate the'cvcrapeed trip mechanism for test purposes,'vithout actually overspeeding or tripping the turbine. While holding a test lever in the test position to prevent actual draining of the auto stop oil, high pressure U oil is directed under the overspeed trip weight and the oil pressure under the weight increased until the auto'stop oil drain'valve is operated. A local' pressure gauge is provided to check the oil pressure at the trip point.
After the completion of low pover physics testing and prior to initial turbine generator sychronization following core loading, the overspeed trip mechanism oil pressure trip setpoint was tested and found to trip at 48 psig. This value was used to correlate the oil pressure trip to actual trip speed when the unit was oversped. After turbine e2nerator sychronization and operation at approxriately lOX load for about 13 hours to insure stable turbine conditions, the generator was taken off the line and the turbine speed was increased to the overspeed trip point. The actual overspced trip point was checked a total of three tines to insure good repeatability. The overspeed trip mechanism performed satisfactorily for each tarbine trip with overspeed trip points of 1998, 1996 and 199i rpm ' Cobtained with a frequency counter durirn, the three checks. 5.5.2.1
5.5.3 Turbine Stretch Data has been taken during station start-up and will- continue through power escalation, to determine the turbine axial and lateral expansion at each reactor power level plateau scheduled in the start-up sequence. Throughout the start-up program, data is also obtained on the turbine rotor position, vibration and the casing and differential expansions monitored by the turbine supervisory instrumentation. Baseline data was recorded with the turbine on the turning gear. During start-up and power escalation, data is collected for four hours, or until expansion stabilizes, at each power plateau. Figure 5.5.3-1 illustrates the turbine expansion data collection points. Turbine rotor end positions are also measured when requirtd. A complete analysis of the turbine expansion will be made at the completion of the start-up program. Data obtained through 50 FP indicates no unusual turbine expansion. 5.5.3.i
Figure 5.5.3-1 Turbine Expansion Data Points
- A6 1 - - A5 L5 L4 LP TURBINE #2 41 L? TURBINE I1 L8 -4.- Ll A9 A2 HP TURB1nE Alo Al P1 Pt
,.5.3.2
i 3.5.4 Steam Cenerator Moisture Carryover and hence Performance of this test had nor been started by September 30, 1976, Subsequent the results are not available for inclusion in this initial report. b6 to the performance of this test, a revision to the Start-up Report will issued in accordance with NRC Regulatory Guide 1.16,discussing the performance and results of this-test. 5.5.4.1
5.5.5 CoolLng Tower. Perforrince During start-up and power escalation, cooling tower operating data is obtained to calculate psychrometric values (te. range and approach) and to calculate the cooling tower capacity for comparison with design values. At station loads of 0% (hot standby), 30?, 75% and 100% power, atmospheric data such as barometric pressure, wet aad dry bulb temperature. wind speed and wind direction arerecorded. The change in temperature of the circulating water through the cooling tower, the change in pressure through the cooling
- tower pumps and the turbine exhaust hood temperature are also recorded. The data obtained at 100% F.P. will be compared to the cooling tower manufacturers t
- design curves to determine the properoperatiott of the cooling tower.
! iA review of the test data obtained at 30% F.P. indicates that the cooling tower is performing as expected at ambient conditions of % 76'F, a dew point of " 56*F and a relative humidity of '. 25%. A final analysis will be performed when all testing is complete at the higher power levels and all available data can be reviewed. C<4...... 5. 5.51
5.5.6 Auxiliary Feedwater Transients Performance of this test had not been started by September 3t, 1976 and hence the results are not available for inclusion in this initial report. Subsequent to the performance ot this test, a revision to the Start-up Report will be issued in accordance with 4RC Regulatory Guide 1.16sdiscussing the performance and results of this test. 3.5.6.1
5.6 Reactor Vibration Analysis 5.6.1 General Comprehensive vibration measurements made during hot functional testing prior to core loating at Rochester Gas and Electric, R. E. Cinna Unit 1, Carolina Power and Light, H. B. Robinson Unit 2, and Consolidated Edison, Indian Point Unit 2 were documented and submitted to the AEC Director of Reactor Licensing. That data Is the basis for acceptance of later plants of similar design including BVPS Unit 1. During hot functional testing at B'JPS Unit 1, the unit was operated at full flow for at least 240 hours and achieved approximately 20 million cycles on the internal components. Following hot functional tenting and prior to core load the core vessel and Internals were removed and inspected for vibration effects and for measurement verification with no vibration problems detected. Additionally, as part of a continuing Westinghouse vibration data ccllection And evaluation program, nuclear flux data was obtained via th~e incore instrumentation system and the excore detectors of the NIS. Perturbations in the nuclear flux signatures obtained with the incore and excore detectors provide information that may be useful in determining the presence of vibration of the nuclear core, reactor vessel internals and core barrel. During normal operation, the VLPMS provides vibration detection and alarm functions. This system was checked out and adjusted during start-up testing. 5.6.1.1 1
5.6.2. Vibration and Loose Darts Monitor System Input co the VLPMS electronic equipment comes from five detection elements (acceleremeters) located at three points on the RCS and two points on the reactor vessel. Each detector is positioned to pick up vibration and accoustic impacts from loose parts at its specific location. Additionally, reactor power signals are provided from the NIS. Table 5.6.2-1 lists VLPMS Inputs. Tests were performed to verify RCS vibration, loose parts alarms, plant nuclear vibration alarms, and to obtain baseline vibration amplitude vs. frequency spectrums (Vibration Signature Analysis) under stable RCS conditions of cold shutdown, hot standby and various power levels. Loose parts monitor alarm serpoints were checked using a pendulum impulse calibrator. RCS vibration alarm setpoints were checked by inputing a square wave signal into an input amplifier card in the VLP41 cabinet to simulate a vibration input 2nd checking for a comparator trip to verify alarm actuation. Nuclear vibration alarm setting were checked in a similar manner using square wave input to the input amplifier cards. With the alarm setpoints verified, an operational check of the VLPHS was performed. The RCS piping In the vicinity of the sensors was tapped and the VLPFIS response was observed. After verifying proper operation of the VLPMS, baseline vibration traces were obtained at cold shutdown, hot standby, 0%, 307 and 30% full power. Data will also be obtained at the 75% and 100% power levels when they are reached. 5.6.2.1
Figure 5.6.2-1 is a sample of vibration'data obtained at zero pow.er. The K.) vibration data obtained during startup testing can be used for comparison with vibration signatures obtained later: in ,core... life.- I . : .... I I i I I i I . . . i - ,, .
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- Localtlon 1 -,Lower Vessel (attached to incore conduicm.imediatcly below vessel)
Channel 2 -Upper.-Vessel (attached to one vessel head closure stud) I .- I Channel 3 - Stean Generator A (attached to one hot leg primary manway closure stud) Channel 4 - Steam Generator B (attached to one hot leg primary manway closure stud) Channel 5 - Steam Generator C (attached to one hot leg primary manway closure I: stud) Channel 6 - Spare 9i Channel 7 -Nuclear Reactor Power (NIS Chane. N43) I Channel 7 - Nuclear Reactor Power (NIS Channel N44) . : S .......... . . . . .
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5.6., Nuclear Noise Measurement
\Il Vibrations of the reactor core, or reactor vessel internals can cause characteristic nuclear flux perturbations corresponding to the location and degree of the vibration. An indepth analysis of nuclear flux signals from different core locations is required to accurately determine vibration characteristics.
As part of a continuing Westinghouse program, a test was performed at BOPS to obtain simultaneously the top and bottom detector outputs of the four Power Range channels from which nuclear signatures could be evaluated. Data was collected at three different control rod configurations, (see Table 5.6.3-1), two rod configurations at 3% power and one rod configuration at the thirty percent power level. In addition to nuclear flux data, control rod bank reactivity worth data was collected at the 31 power configuration and two incore M/D traces were obtained at the 30z power configuration. Power range channel N44, both top and bottom detectors, had excessive noise at all three rod configurations and had intermittent signal spikes at 30% power. Investigation revealed that noise was being generated In the 'A' (top) detector. The entire detector assembly will be replaced to eliminate the noise pzoblem. A report discussing the test is being prepared for Duquesne Light Company by Wesci:ghouie. 5:6.3.1
Table 5.6.3-1 Nuclear Noise Measurement Test
' Summary
' . TEST SUMOMRY TABLE
.Rod, Case P'WR Coniguration* Data Acquisition I 3% Bank D Q 208 steps satisfactory**
' II 3% Bank D @ 67/68 steps satisfactory**
Bank C.@ 67, steps . 1 ; 30% Bank D @ 186 steps _ satisfactory**
- All rod banks are completely withdrawn unless noted otherwise.
* **Power Range Channels N44 had excessive noise in all cases.
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5.4. ! Irneoro- i:eteercios o-f ::uel A-tsscably Vibration Fuel asse-sbly vibra ions within the core a~ssrbly can possibly be detected at power by evaluatint: nuclear flux data obtained with Incore moveab'.- detectors. Fuel assewbly vibr.:tions cause nuclear flux. perturl..2tions corresponding to the location and degree of the vibration. As part of a continuing Westinghouse Investigation in this area. a test was performed at BYPS to obtain nuclear flux signatures at the 30% power level using the incore noveable flux detcc:ors. The detectors were first inserted fully into the core at preselected core positions. Thcy were then individualuy withdrawn and positioned at each successive grid strap by observing the local flux depression caused by the grid. When all five detectors were in position, flux data was collected on nagnetIc tape. Additionally, eacore bottom deetecor currents, Intormedijrt' ran;Lv current rendinP. and power rxane percent pcver rendings wtxr.* also obtained. No rot notion was pwrsltted during daza recording. A fina: report discussir.g the test wf.U be provlded by ';eti:-hc'use fao Duquesne Light Conpany.
- 5. . ..
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i Plant radiation. :urvLeys ~.s:cre' prfornt'd at po~wer !e.!.1;"of 3'. 307 a:Id ,O0.~ ii S. sch'i4cIuCS! 01bt 1OO' S (partial survey) full rowlr.' A radl.ni.'n survey !.A :3 I I C. imma .md neutron' radiatioin Wtert, -ceisured durhar. eachd -A;irvey to~ F? plateau. 4 i determinc.the effectiv£?ress of the biologicail shielding. Tables 5.7-1 thru A prov.ides a',summary of the g~rnma and neutron radiation redingi taken wtthin the II reactor containment 5 Several points were ruchccktd at the 50* Vrp Icvel due i i readings were cor-p.2red to the previous~ S'ur ~vs The i t to Inconsistencies %:e waich I inconsiatccles weri. dase m.iirly to, slight ii;rtaiions In the locaticns at I II the readings veret~akor'. at dIifferont ppier levels.' I i Caw-a and/or neutron, radia~t tn levels o bs or ve161 ntrstricted arva.i (auxillary. j fuel and decontamination buildizags) for each surivey we-re all wl n I 0t! a z.: cis II ii levels (c2.5 :nrecrnhr-wtth the exception of r-ee-'l-it. tken-insidee;-h!%u1dtd equipment co partments). .. I II II arvas n r-:F,'of urbint-z:i~ i Cam=a radi.atin r..*ad I n . in un e trl:atc! I buildin-'anid outside groun~ds) for ca.-h surv.ey shated' o e~i. 1 raJ La tLor. iI i levels 2 0ove bilkground. i I II ani b !,1o I -a! shielding e~fctivc:xes Further evaluation' of plant. ra-1a: inn 'le:voib i I will be 'ossible a f C r 't h. resuit.; of tiII ra Iat Ion s9urvye at 10OO'FP art, II II I avail t ab le aI i II t I I II Ii I II II i I i .
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5.3 Shutdown From Outside of the Control Room An emergency shutdown contril panel (SDP) located outside of the main control room provides the capability for shutdown and maintenance at hot standby of the uWit in the event that the main control room beLoT.;:s uninhabitable. During hot functional testing proper operation of controls on the SDP and the ability to safely maintain the unit in a hot standby condition from the SDP were demonstrated. With the unit .at approximately 101 generator output, during power ascension testing, the reactor was manually tripped and a shutdown from outside the control room conducted. The station was maintained in a hot standbv condition from the SUP for two hours. Approximrately 25 minutes uas :equired to obtain stable hot standby conditicnA. while Maintaining the unit in this condition one of the motor driven auxillar: feedwater puzps was damaged due to inadequate flow through its recirculatlin line. A teporary replacement pump has been installed
- while the original auxili.ary fvedwater pump is repaired. An engineering evaluatinn of the ade"Quacy of the pumps recirculation line flow has been initiated. A few minor changes were gener
- ated to the BVPS Emergency Operating Procedure for control roon inaccessibility as a result of this test. No other oper3tion problems were enco'niterwd during the te-.;
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- -' : ' 5.9 Plant .Sund Level Survey the areas in che station where A sound level survey uas perforned tc idertify and areas where P3ge-Party high noise levels exist during normal operation comnwanications-are not audible.
With the station at approximately .307FPa general "walk through" sound survey I. of areas within the major structures of the station was performed. A snand at various locations. The level meter was used to measure sound level (bA) the survey. audibility of the Pige-Party was also'clecked during FP, high noise areas were Based on the results of the general survey at 30% areas to determine the general identified and detailed surveys were made of thcse these areas will be made ag-iin noise source and magrtitude. 'Detailcd surveys of I,, with'the'station operating at 100% FP.' II in those areas where hsigh General and detailed sound surveys were not made those areas whore noise levels
- radiation levels preclude general'occupancy or
, preclude meaningful results.
a:iere sound levels exceeded Tl'e 30% EP survey' identified a large'number of areas i as cossible inmthose areas. I 90 dbA. 'Noise sources were identified as well of the data collected at 100% F.P. i Recommendations will be, made following a revicw required. i II to provide for procection of station personnel as iIf -~~ .. * .. . i I i i I I , ! . . II 5.9.1 I I I I -
5.10 STARTUP TESTI;G CHRO:.ALOC DA.TE TIME (},rs)**~ EVENT* 21L17rk 0308 Commenced core loading. 217.'7r, 0429 Completed core loading. 2/14/76 0930 Started initial Incore Instrumentation System Operational Check. 2/19/76 0700 Completed initial Incore Instrumentation System Operational Check. 2119176 1900 Started VLPHS operational checkout and alarm setpoints check. 2/21/76 1600 Obtained initial turbine stretch readings with turbine on turning gear. 2/22/76 1300 Started CRDYK timing measurements for cold shutdown conditions. 2125/76 1500 CRDM timing measurements at cold shutdown completed. 2/26/76 2300 Started cold shutdown channel calibration on the RPI system and Pulse-to-Analog Conver:er calibration. 2/28176 0.00 Cold shutdown calibration of RP' system completed. 2/28/76 2300 Completed operational checkout of VLPXS. 2/29/76 1600 Started cold shutdown rod drop tine measurements. 3/1;76 2300 RCS depressurized for maintenance on 1A RP - red drop tine measurements suspended. 3/i/76 1800 'aintenance work completed - rod drop time ceasure-ments .oitinued. 3:6/76 0400 Rod drop time measuresen s ccnpleted. 3/6/76 0700 Started YLP'IS spectrum analysis for cold shutdcwn conditions. 3/7/76 1630 Completed VLPMS spoctru= analysis for cold shutdown. 3/8176 1500 Cbtalned RTD/TC rross Calibration data on 15CF plateau. 3/9/76 to Various retests perform.ed. 3/28/76 3/29/76 1800 Obtained RTD/RC Cross Calibration data on 250F plateau.
- Does not inxclude vari ous retests that were perfor-.ed in tne initial scage..
of the startup.
** 24 Hour Cleck 5.10.1
DATE TIMIE (hrs) EVENT 3130/76 1100 Obtained fVD/TC Cross Calibration data at 350F plateau. U 4/2/76 0200 Obtained RTD/TC Cross Calibration data at 45DF plateau. 4/2/76 1800 Obtained RTDITC Cross Calibration data at 525F plateau. 4/3/76 0700 Obtained VLPMS speccrum analysis traces for hot standby conditions. 4/3/76 ; -000 Obtained RTD/TC Cross Calibration data at 547F plateau 4/4/76 0700 Plant cooldown to 200F for maintenance on Main Steamlines. 4/12/76 0445 RCS temperature at 547F.
-Performed RCSPre3sure Test e 4/12/76 1500,,
41l3/76 0700 Started Pressurizer Heater and Spray Capabilities Test 4/13/76 1500 , Completed Heater and Spray Capabilities Test, . 4/15/76 1300 Stacted RCS Flow Measurement Test 4/15/76 1500 Started installation of real detectors for Incore Instrumentation System 4/16/76 1300 I Installation of real detectors-for Incore ; Instrumentation System Completed 4/18/76 0100 Plant cooldown in progress - unable to start IC RCP and safeguards, logic probles.' 4/21/76 0400 RCS at 547F, continued with RCS Flow '!easurement Test. 4j21/76 0900 Plant cooldown In progress - Main Steam Leaks. 4/23i76 0800 RCS at 547F, continued with RCS Flor. Measurement Test. 4/23/76 1930 RCS Flow Measure-ent Test Completed. 4/23176 - 2000 Started RID Bypass Loop Flow Measurements. 6/24/76 1900 - Started.CRDM timing zeasurements for RCS at 547F. 4/25/76 1200 RTD Bypass Lcop Flow Measurements Cmpcletea. 5.10.2
DATE TIME (hrs) EVENT 4/25176 1900 CRD:{ Timing Xeasurements Completed. 4/27/76 2300 Started Rod Drop Time Measurements for RCS at hot standby. 4/28/76 1700 Rod Drop Time Measurements Completed. 4/28/76 2300 Starred RPI System Calibration Check at 547F. 4/30/76 1500 Started Part Length Control Rod Mechanism Brake Test. 5/1176 'l200 Completed Part Length Control Rod Mechanism Brake Test. 5/3/76 1000 Started Reactor Plant Protection System Setpoints Verification. 514/76 0700 - Completed RPI System Calibration
- Started Rod Control System Checkout
- Started Initial Operational Alignment of Process Temperature Instrumentation prior to InitiaL Criticality.
5/4176 1300 Rod Control System Checkout Completed. 515/76 0400 Completed alignment of Process Temperature Instruments. 5/6/76 0000 Started RCS F.'.ow Coascdoun Test. 5/6/76 0530 RCS Flow Coasteown Completed. 5/6/76 1200 Realigning Process Temperature Instrucentation. 516176 2000 Performing Operational Check of NIS prior to Initial
- riticality.
5/7176 0700 Process Temperature Instrumentation Alignment Cowpleted. 51/176 1300 Couipleted Reactor Plant Protection System Setpoint Verification. Complete Operational Check of NIS prior to Initial Criticality. 5/8016 0100 Dilution to initial criticality commenced. 5/8/76 2100 Dilution delayed due to need for H7 blanket on VCT and problems with primary grade water 02 content. 5/10/76 1600 Dilution resrarted. 5/10/76 1845 Achieved Initial Criticality. 5.10.3
DATE TIME Mhrs) EVENTI ' 11-j 5/10/76 2200 Reactor Trip due to loss of Vital Bus Inverter No.3. 5111/176 0530 Reactor Critical, determining low power physics testing decade. 5111/76 ,0700 Insestegaring problem with Reactivity Computer 5/12/76 1400 'Reactivity Computer functioning properly - low power physics testing decade determined. II 5112/76 2000 Borated Bank D to 204 Steps. 5/11276 2030 'Obtained boron end point and isothermal temperature coefficient measurements - SBA, S8B, CBA, CBB and CBC @ 228 seeps, CBD 0 200+/-0 s-eps. 5/13/76 0600 Diluted Control Bank D.to 168 Steps 5113176 0900 Reactor Trip due to adjustment of a pot setting in an ovcrtemperature'AT circuit. i 2 5/13176 2300 Reactor critical, obtaining nuclear noise measurement I.I data with ARO. 5114/76 0230 Reactor trip due XIS channel N&4 noise spike with channel N43 OOS for connection to Reactivity i Computer. Li i 5114176 0800 Reactor Critical, obtaining incore flux map - SBA, SBB, CBA, CBB, CBC 0 228 steps, CBD Q 175 steps. i 5/14176 2300 Reactor trip due to failure to reset rate I trip on one.NIS power range channel prior to I placing another channel in test. II T I 5/15/76 0300 Reactor Critical. , i i 5115/76 0730 Diluting Control Bank P to 0 steps. I 5/15/76 1500 Obtained boron endpoint and isothermal temperature coefficient measuremeults - SBA, SBB, CBA and CBB
@ 228 steps, CDC @ 20018O steps, CBD Q 0 steps.
5/16/76 0130 Obtained incore iluxsmap - SBA, SBB, CBA, CRE,
' 228'steps,,CBC 0 175 steps, CBD @ 0 steps.
5/16176 0430 Diluting ControlBank C to 0 steps. I 5.1C.'.
DATE Tl.ME (hrs) EVENT 5/16176 1000 Obtained boron endpoint and isothermal temperature coefficient measurcemnts - SBA, SBB and CBA 0 228 steps, CBB l 200 t ' steps, CBC and CBD I 0 steps. 5/16/76 1500 Obtained incore flux map - SEA, SBB and CB, @ 228 steps, CBB e 184 steps, CBC and CBD Q 0 steps. 5116/76 2200 Obtained baseline data for Startup Adjustments of the Reactor Control System Test. 5/16/76 2300 Diluting Control Bank B to 0 steps. 5/17/76 0800 Obtained boron endpoint and isothermal temperature coefficient measurements - SBA and SBB e 228 steps, C3A @ 200 tb° steps, CBB, C8C and CBr. @ 0 steps. 3/17/76 1700 Diluting Control Bank A to 0 steps. 5/18/76 0030 Obtained boron endpoint and isothermal temperature
'.oeff46 ent measurements - SBA @ 228 steps, S3B
@ 200-0 steps, CBA, CBS, CBC and CBD e 0 steps.
5/18/76 0230 Reactor manually tripped for surveillance require-ments prior to exceeding shutdown margin for Minimum Shutdown Verification Test. 5/18176 0330 Reactor Critical. 5/18/76 0800 Diluting Shutdown Bank B to 0 steps. 5/18/76 1200 Obtained boron en.point measurements - SBA 5 190 +1G steps, S8B, CBA, CBS, CBC and CB9 @0 steps. 5/18/76 1500 M3nually tripped reactor due to control rod inoperabilicy while inserting SBA and withdrawing RCCA H-14 5/19/76 0230 Reactor Critical. 5/19/76 1030 Elanually tripped rvactor again due to control rod inoperability while inserting S&A and withdrawing RCCA H-14. 5/20/76 0300 Reactor Critical. 5/20/76 0830 Completed stuck RCCA Worth and <5.S worth ueasurements Obtained boron endpoint nmasure=dnt - SBA > 20 stops, SBB, CBA, CBB, CB0. and CBD I 0 steps, RCCA H-14 s 228 steps. 5.'0.5
DATE TlHE (hrs) - EVENTJ - - ' 5/20176 1400 Performed dynamic drop RCCA 11-14 worth measureanzt. S/21176 0230, Started boration for control bank worth measurement using normal overlap. 5/21/76 0830 - Reactor subcritical'following an inadvertent boration dueto,RPI problems. 5/22/76 1830 ReactorCritical,continuing boration for control bank worth measurements. 5/23/76 0430 Boration completed, performing RCCA configura:ion adjustment'to HZP insertion limits. 5/23/76 0830 Reactor shutdown due to RYI's being out of spec. 5/24/76 0430 Reactor Critical. 5/24/76 - 1100 'Started RCCA Pseudo Ejcction at Zero Power Test. 5/24/76 I. 1500 Obtained boron endpol-it measurement and incore flux map with RCCk 11-14 'ejected. 5/25/76 0030 RCCA Pseudo Ejection Test Completed. 5/25176 0230 Reactor at approximately 3X F.P.,'obtaining nuclear noise measurement' data with control banks C and D at position near max.'differential worth. 5/25176 0500 Nuclear noise measurements completed, starting
- station radiation survey.
. . I 5/26/76 0300 . Radiation survve completed.
5127/76 0430 Attempted turbine roll - unable to rmaintain without increasing reactor power above indicated 5X F.P. due to conservatism in NIS gain settings. 5/28/76 0300 Adjusted gains on .NIS. 5/28/76 0500 Performed turbine roll.
'5/Z8/76 '0600 Reactor trip due to throttle valve close limit siitches still made with turbine increasing in speed and RPS sensing greater than .lO ReactorPower.
5/28/76 1400 Received a-35% F.P. License. 5/28/76 2042 Reactor Critical. 5.10.6 ,
DATE TIME (hrs) EVENT 5/28/76 2300 Reactor power at approxinatt1y 7% F.P., started Dynamic Automatic Steam Dunp Control Test. 5/29/76 1530 Reactor tripped due to low-)',w level in iC Steam Generator caused by overfeeding the steaM generator and level shrink resulting. 5/29/76 1940 Reactor Critical. 5/30/76 0100 Performed turbine roll for checkout by Westinghouse. 5130176 0800 Reactor tripped due to turbine trip signal when test signal simulated turbine Ipulse pressure >10% load. 5/30/76 1131 Reactor Critical. 5130/76 2000 Completed Dynamic Automatic Steam Dump Contro. Test. 5/31/76 0000 Reactor tripped due to turbine trip. Antimcnitoring protection of the turbine while attempting t:o synchronize generator to system. Wiring error in EHC corrected. 5/31/76 0150 Reactor Critical. 5/31/76 0500 Reactor tripped due to turbine trip caused by a spike on first stage pressure while transferring from throttle valve to governor valve control. 5/31176 0633 Reactor Critical. 5/31/76 1400 Synchronized and loaded Turbine - generator to It 45 MWE. Obtained secondary plant calorimetric data for NIS gain adjustment. 5131/76 1500 Reactor tripped due to 214 higb flux rate on NIS. 6/1/76 0101 Reactor Critical. 6/2/76 0900 Reactor tripped when NIS source range was unblocked due to error in Procedure. 6/2176 1130 Started RCS cooldown because steam generator chemistry out of spec. 6/7/76 1200 Depressurized RCS and draining pressurizer to repair leaking manway gasket. 6/13/76 0000 Reactor Critical. 6/13/76 0400 Obtained incore l:.c; map witlh A\RC. 6/13/76 1230 Reactor trip due to ma:tual turbine trip when Tavg dropped too lo'0whten Increa;'n. load due to insufficient re3ctor power Intcermediate R.tner Red Sto ). 5.10.7
TIME (hrs) EVE4T 6/13/76 1456 Reactor Critical. 6/13/76 2000 Turbine - generator at 90 XWE. 6/13/76 2150 Reactor trip due to'negative rate trip on NIS power
-range channels.
6/13/76 2245 Reactor Critical. 6114/76 1300 - Turbine - generator at 90 M.E, obtaining secondary plant calorimetric-data. 6/14176 1500 Performed Turbine Overspeed Trip Test. 6114/76 2000 Reactor trip dueto negative rate trip on NIS. 6115176 1010 Reactor Critical. 6115/76 1700 Reactor at ' 23% F.P., obtained secondary plant calorimetric data... 6/15/76:' 1930 Reactor trip due to low-low steam generator level caused by clogged condensate pump strainer. 6/16/76 0330 Started power escalation to 30% F.P. 6116/76, G900- Reactor trip, NIS 2/4 negative rate, one channel negative rate not reset prior to placing second channel in test and tripping. 6/16176 2140 -Reactor Critical. 6/16/76 2230 Reactor trip.due co loss of feedwater flow caused by clogged condensate pump strainers. 6/17/76 0236 Reactor Critical. 6117176 1000 ';,Reactor trip due to shorting 120 VAC Vital Bus causing overpower trip on intermediate range channel. I II 6/17/76 1151 Reactor Critical. 6/17/76 1700 Obtained Cooling Tower performance data at 30% F.P. I 6117/76 2000 Started station shutdown due to RCS -leak on pressure transmitter line. I 6126176 0430 Reactor Critical. I 6/26176 1900 Reactor trip due to high-high steam generator level i while putting feed regulatirg valves in auto.' i 6/26/76 2027 Reactor Critical. -- I II 5.10.8 I Li
DATE TIE Cars) EVEN;T 6/27/76 1200 Reactor at 30% F.P. Performed secondary plant calorimetric and adjusted gains on NIS. Obtained 30Z power plateau data for the fo:!owing tesrs:
- Sta:tup Adjustments of Reactor ConLL3I System.
- Operational Alignment of Process Temp. Inscr.
- VLPMS spectrum analysis.
- Calibration of Steam and FU Ylow Instr. at Power 6/27/76 1500 Started Automatic Steam Generator Level Control Test.
6/28/76 0530 Reactor trip due to overpower trip while attempting to place rod control system in auto. 6/28/76 0912 Reactor Critical. 6/28/76 1700 Reactor trip due to low-low steam generator level while transferring from main feed regulating valves to bypass valves. 6/28/76 1830 Reactor Critical. 6/29/76 0530 Started station radiation survey at 30% F.P. 7/1/76 0400 Radiation survey conpleted. 7/1/76 1030 Reactor trip due to high-high level in IC steam generator during Autonatic Steam Generator Level Control Test. Commenced cooldown due to out-of-spec. steam generator chemistry. 7/5/76 1218 Reactor Critical. 7/6/76 0100 Reactor trip due to high-high steam generator level caused by turbine trip while testing turbine over-speed trip device. 7/6/76 0205 Reactor Critical. 7/6/76 0600 Completed Automatic Steam Generators Level Control Test. 7/6/76 2300 Performed Automatic Reactor Control Test. 7/7/76 1800 Commenced station shutdown and steam geacrators to correct water chemistry and to check for condenser tube leaks. 7/24/76 2030 Reactor Critical. 7/25/75 0200 Reactor trip due to low-low stean generator level. 7/25/76 0400 Reactor Critical. 5.10.9
DATE TIME (hrs) ' EVENT '
.7125176 ., 2330 ' Rechecking calibration of steam and F.W. flow instrumentation.
7127/76 1300 Reactor trip due to turbine trip. 7/28176 1805 Reactor CriLical. 7/30176 1500 Recheck of steam andF.W. flow instrumentation ca'ibration completed.
, 7/30176 1930 Reactor trip due t%> low-low steam generator level
;I . resulting from IA main feed pump trip. Rotating elements of pump damaged beyond repair.
i 8/1/76 2028 Reactor'Critical.' I 8/2/76 1500 Reactor trip due to turbine trip when 1A Steam II Cenerator Feed Valve failed wide open. 8/2/76 1830 1B Hain Feed Pump outboard scal fatled, pump I manually tripped..,,, i 8/18/76 1230 Reactor Critical. II 8/19/76 0600 Reactor trip due to low-low level in steam generators causedby'clo ged condensate pump strainers.
.. I
8/19/76 0928 Reactor Critical.
U., II 8/20/76: -,.1900 Started Station Sound level survey at 30Z F.P. I 8/21/76 0130 Obtained preliminary incore flux map, plant calorimetric'and' IS data for incore-excore i detector calibration. I I 8/22/76 1330 Reactor trip due to indication of loss of 1A RCP
,above'30%'F.P. when'vital bus was shorted.
8/22/76 2227 Reactor Critjcal. - I 8/22j75 2252 Reactor trip due to low-low stea gen.erator l.vel when =ain feed pump discharge valve was left-closed.' 8/22/76 2350 Reactor Critical. 8/24/76 0930 Station sound level survey at 302 F.P. completed. i/24/76' 1800. -Reactor crip due tobhtgh flux rate on NIS while
- performing maintenance on NIS.
8/24/76 2300 Reactor Critical.
.; .. '5.10.10
DATE TITE (hrs) EVENT 8/24/76 2310 Rea:tor trip due to low-lnw stean generator levels. 8/24/76 2335 Reactor Critical. 8/25/76 0200 Started Shutdo'.n from Outside the Control Roon Tesc. 8/25/76 0215 Reactor Manual Trip. 8/25I76 0600 Shutdown from Outside the Control Rodm lest completed.
*8/26/76 1230 Started station cooldown because 8 7 1auxiliary of inoperabillty of feedwater pump which was damaged during Shutdown from Outside the Control Room test.
9/5/76 1000 Reactor Critical. 9/5/76 1130 Reactor shutdown due to high vibrations feed pump. on 1B main 9/7/76 0300 Reactor Critical. 9/7/76 0530 Reactor trip due co low-low steam generator level. 9/7/76 0740 Reactor Critical. 9/10/76 2330 Obtained equilibrium xenon at 30% F.P., pulling CDB to 187 steps to start RCCA Pseudo Ejection and RCCA Out of Posicion Test. 9/11/76 0700 RCCA Pseudo Ejection and RCCA Out of Position Test Compleced. 9/11/76 1730 Obtained nuclear noise measuret.ent data at 30Z F.P. 9/11/76 1900 Performed Fuel Assembly Vibration Test. 9/12/76 1300 Started Static RCCA Drop and RCCA Out of Position Test. 9/14/76 0500 Static RCCA Drop and RCCA Out of Position Test completed. 9/14/76 0900 Reactor trip due to RCS loop low flow signa1 caused by shorting vital bus. 9/14/76 1725 Reactor Critical. 9r14/76 1900 Reactor trip due to turbine trip caused by a s?iLke in turbine impulse pressure giving permissive 1. 13 with throttle valves closed. 9/14/76 1950 Reac:tor Critical. 9/16/76 2300 ObLainad power coefficient and ir.tearal power dofect
.nesuremn.irsat 30. F.P. plateau.
5.10.11 I_ I
DATE TIME (hrsl EVENT 9/17/76 0400 Started retest of Automatic Steam Generators Level Control Test. 9/17176 2000 Started Loss of all Offsite Pocr at Load (30% F.P.) Test. 9117176 2300 Loss of all Offsite Power at Load Test Complete. 9119/76 2153 Reactor Critical. 9/20/76 0600 Automatic Steam Generator Level Controls Retest completed. 9121176 0130 Obtained Load Swing Test data for 30% F.P. plateau. 9/21/76 0500 Reactor trip due to channel N41 of NIS in test and fuse Llew in channel N44. 9121/76 0838 Reactor Critical. 9/21/76 1500 Reactor power escalated to 50% F.P. plateau. 9/22/76 0700 Performed secondary plant calorimetric and adjusted gains on NIS. Obtained 50% power plnteau data for the following tests:
- - Startup Adjustments of Reactor Control System.
- Operational Alignment of Process Temp. Instr.
- VLP.MS Spectrum Analysis.
- Calibration of Steam and NW Flow instr. at Power.
9/22/76 2100 Obtained power coefficient and Integral power defect measurements at 50% F.P. plateau. 9124/76 0900 Performed Special Hot Channel Factors Test (Negative al).
- 9/28176 2000 Pertormed Special Hot Channel Factors Test (Positive AI).
9129/76 1700 Reactor trip due co turbine trip. P-13 permissive actuated after the turbine was manually tripped due to excessive noise fron miain Zenerator.
- 9/30/76 2315 Reactor Critical.
9/30/76 2400 BVPS Unit 1 declared available for coimercial operation up to 50Z of rated reactor power. 3.10.12
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*.. Deaver Valley Pomr Statlon, Unit So. I .
-Docket 'o. 50-334, Lizenas No. DINK-J6
-1 Startuo Roiort - Su-o1ement I
. I I -
16nlotd States '.Nucelar P 1:.erula.tiry Co=_1svlo M~ractor. "*&Lonstl Of" tca King, of Prw eia, Par...sylv.nla ;' . In accordance with Regulatory Guile 1.16, Reportfnj of Onesra:tr.g Information - Appendix A, Technical Specl [catLons. the attachod supplenont..-
-submitted,90 da93 after the Lnitiil report was issued. The stJrtup test pro-ran has not biee. cerpleted ad a.. a"itlor.3a; suppc Ientary report wtdl b.t isiuc! WLthin : ~.O o, thes swi'ric.en.: tC further rcort the progress o.tA
- of the prcgrar..
Var:y truly yours, C. ":. Dunn Vice Prou'xdcr.$, rpN..:rt!e.s cc: Director, OffIcc of :ML-.ai.c~cr: 1rfoor-;4tLon an' i'ro.;r.i: Ccntr.-1 (' 3Irector, ). c cb etr -.- r.-* Vr.i- rc:.r. ts '
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- .- *'s* .?y
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1.... S .'-& e..,Jy gL Revtse the Startup Report as follos: irtQ S..f,
. Insert Pages
. I
- - - - xli ' a xli 5.1.3.3 ;t 5.1.3.3 5.1.3.6 5.1. 3.6
- 5. t.6. 1-3 5.1.6.1-3 5.1.7.1 5.1...1-3 5.5.1.2 5.SA.2 5.10. 13-17
;one ccisLi ron d p are detteoC.' by r4. iI lines..
'tu p ib 't ly ov -Sthtt6 ';5czdrtup4RprcO'U .1;
- .
- b'r 3. .197.6 -iihon~.tle. stationwa 4va prdc-na 5Zratr.pgwer 'and
;.rl:.this timpe' rlod the Statlon esaclated to the 752 FP test plateau after. 1/4 completlon of all testing scheduled for the 50t VP and 65Z FP plateaus uxctipt ior shutdy= and transiient tesits, BUF 1.1 -92.,Auxiliary Feedwaier Trans11enL ai..ni BVT 1.1I 9.4.3, Net Load Trip Tait, respectively. rf,4 eo tliese teAts w4;; deforred urntil later in the startup pr.ogramn for schaduler e'tam dump bff~lo supporte. in tht% condens.or. .ll'~ df ~,~14aavv supmplemenntal report. relatively 1Lttl1., progreo.1 was -it-."the coWmI~etion rof
~'thOe Statup-test' program ani only one test, BUT 1.1I 9.4.5. Dynamic r.*'J Drop, wa:I totally cowmptlerLd In this time +/-rnterva.2i A major Station outage for -ciaintenanc.e and pcrerfo.-inc of reactor corntalnmint 1-tolit i.rn valves tvjpq C 1'.k tet oocirr,.'J during N.Iv%..bvr And ecarly Do;cvtn*obr. In addition to force~d imts.tges and ~-0ldr.n' boCCa'zr. tif failure~of Ow. reco plant rlv..r wa.ter purnp,. condecnseor tt:bv. lenag.t%' Un.! attendnt alteati g.er..tor vat.er ciho?.L.try pruob1lm=. two InciJents. of Avero, fe-diviter. llst vtbratto'n oc.-tirred. The (irs.t irnc!dent occturred %n Nov 1emr , 19-4 -and1roprwci:,A t'b
.Zt %:[.Iiti i..'i .1 .4.i-leatlq I :wA ntvz;.m:eo. oot- ,,~* *tevv~r i: . *.*. 'arIv. Afte*r returnir.: to powe~r .imcl codll'ietinm a t.--it :. !wnv IAtfg-t t!'%- :t1brut ion nh-e?.~cmer.
X 11 .i.:1r N..'* I
.zr:at l.sr Ilest*l *l red obn ;ove:t-ril '., @he:.
Aq~..- *)i the *:..xr tIe. -. .. j'. st.ttionl wa-. toitt.!ovnq .nd mak!n;:. pr.epalr~la.:!; tor re:ttrnin tog p...er opet-rarle.ll the basLe causc of the stevcre feedwvatr line vibr.)LiLrns li:J beein r5utativ41P:- fIad'aea-ulfia the dc htchLn 9gn paramker.t -or the tcetdvf 1
*tl u lod:'.api&rtof tho ;S5S ,'nckage, I IO od -the rembLner ofW ysitm bIance of plont iujpied. Interim n4tloits to reduce the pro bality .,
Of, farther severe vibration was being Implemented until long term correctLve
*actlon could be Identified and complectd. A hold wvs placed on continuatfon of the startup test program because of tbe fuedwater line vibrationlncideati.
Special testing performed as a reiul: of the' feedwator llnc vibrationA. are not.:
!:nltlded In thl.i report, but =ay be wo:nd in the Annual Operating Report cubmttted ' :he h Station to the Nuclear Regiflltury Coi=islon.
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}.( ._I 2W 1 -1.a. s.eh 58 7 .-! IL.' -lII 5.3, I0.8l
!. I .
). 8 11.5 lI.SJ:
- - - t-1I.591 t t).59i 1 ;I.735I O.31.2 t-II.M970 II.t-87 I I.'.'b t- ..
I.as9 I 2.104 1.ASOj _s,-A s S-,)~I,Lw .(J to X7e s - x _R _4U . (.5111 1.5.7 18.5S'" *). J ' 0.)633 U. Its)I 0jil.s. 0-.1 5.0 1).lI 5. J Is52 I _ 5J- _.1
- I 4.44 85 7. 1 L.6 03.711 I.9 I'.
-7.3 6
NA. 243.1 2.443
'1 .1; 1.91:). }31 (.51(110 .571 (1.554 :.541 0.700 o. 4 1.046 373J4 !.465 3.otw O~11. I I-1..8 44I 13 I I 1.4 1 It.'
-4* -
- 9. I I).'J I - 444 12.z0 9.7 '.1) *I. L 1.
-9.9 -
-8.9* ,- _8A I rp- I~ 0 Iui.-~ OU so, In.s U.535 ).576 (.712 0. 9 1. U( 1.188 1. 17 1.598 3.314.
- 4 1 V.U~ 15.6 15.hl S.U s.7 9. 8 14.3 14.31 7.1 -0.9 _9.5s -g.5 .99 -7J.3:
-4. I--I-E-4 4 I 99 S2.
_- it. I.:ns ft 87 n34 s o. 9.0 t). 5 15 ji. 544 U.84 ' .963 1.146 1.778 409 ! 2.965.
, 1.0o 15. 3 1 4 . 1 9.X( 5.4 113.4 12-.6 11S .5 2.,i -. 1.4 -7.8 tl). ! 9-3
- Y- I- I- O . - - . ---- -. ---- - --- ;w-l -- I --- . . --- 11.
,t.448 0 .506 0 . 5h4 U.511 P.337 (1.619 0.7 s: .857 U3.725 1.625 1.253. 12.S505 9.2 9.9
- 14. 3 I 14 8.5 9.h, 4-
=
4 3.9 I 1
.z1 L
1.9 I
-1.3 -4.7 -6.8
(}. JI .1.497 50.9 1).n1 143.592 0.709 U..798 'i . 1-46 1. AM 1,'9 1m t.922.i Is.6 17.'! 30.0 15.6 I.l 3.6- _O.I '1A.n - 9.A V 5.7 I ,7V i-014.9
-- _- _ M I ... _ - -- I - 2- l l - -Ss- . 7- - w I., . 0.162 P1.444 .3.728 , 809 0X.937 1.0$1 t1.253 1.355 1.308 f.311 :* - .
'1. C. I:.
- 9. I 1.5 1(1 12 7. 4 it. I 6.1 7.0 19.4"I i.4*
E . . l _ _ _ I
't^S
'184 0.574. ).674 U. 7371 I.h14 1.9s8 I1.06 1t1.210 I.- - s:
4.5 . 1.6 7.9 12.1.1 7. 15.0 1 2.9 ' 7.7 I 17.'
- 9 I. 1 - - ~
. I - - I - 9 - -; ;'.f_ r. -..
0,J87 1.592 0.57f' . J71 1.125 0.84 0.732 eensured = re CL -
- r. " 0 ..
zj2 I~~,.I 2.1 8.S 10.9 10.5 7.4 3.0 7.9 fron ex;pecL4 n.*i2( 0.3" I J. J96 I1 1511X.9 I-S..
.V1 r..
L
.t . -.. . *:'t. .f .... , ,
.pr'e hot ;thanneo 1. ati-Lors .(F' and F Z)are obtai'ied durlng transient maneuvers V.-
deslpcda.v .- . .::. .;
~
utltc t~hagrlf ecots-i than'rqy Pat.1f dthinte tatrget bar d to complon cc tinrhocacc alp co~ngs e Mvrdfeaft~e uwiY-0.7%anm Nino brTe tppo oiwnatcl 1fr greacter::thangc~ andtil with ~in the target ,and aQ1ap414 CFM wai ot~itdbilrDwas thenInsrtedC toa $1'.)stesi abnatte. A togt' ihn2.o II tcit n'egativedirction. T/e C wapsdituredtk
~
to~ ~ ~~~~~he1~OF hnesrpatvt compvnsttefr 11ors de onvto 207'!ermand -theebt ob talled (1C~.. C-LI?0015 ovHDdwsthnwfh ar~a-.- cto theArsutep- on the .ina::ls wa- boatetocomenste or eacurtivalty canges, util !.I wasurwithn thun tarntv to ;%:rmt jcompter imuatin ocr Tcclni.:al i:nc Icat-ornTI mapls. wereatak~en ere within ti-e Technical Specification limits with OttILL.?.ietlon o f F vC' fo te et nlCp3tiC .eas. Ulien F~. ex-Cteei tts. l=Itt, the Ft::, 11'?tI.4 'I1 5. .., . . :.-'1 . '!I :.: -'"e. I
M
- .-:aht ;n' " I,~t
.' F!; :i'; wX: hX:wII . Ii~':3 1 Aa: tI:W. .. -- i:;. w.! WI LI:II:i Om.' giL.sr L. r.*g I-n 5-1.*te -2').0:.
',;t ! rreactor at 4 Stabte ppowtr l.v.e' o: dZ power, the tesc was repeated in tiw sa" Sequence for poaitive 41. T av VaW decreased to IF.beloTrref.-with :: C3D 2 16; st.eps and It within the target bznd so that the b.is QCFMt (.'lsp 020) eulld be obtnined. CBD vas-thtn withdrawn and the RCS wag borated to 6alntaln T in an attempt to achieve a4.: of INg . TS+ mzimum .'I achievod with AR. thc w.as +2.917 .nd a qCFRI (:!.p 021) was obtained. CBD was lnbarted to 194 steps and thse R(:%S vs borittcd untilIAl wais wair. within the target balnd. Three addl.ional a:I'!': t I.:;i *'. 21, 221 were taken in two hour Interv419. T/C maps were trkon a ea.. : *!t:F:I *n p::nt p ar4-etmers wtre trended co provide a eomputer -il.ulatlon c.oiro :.n.- I.-us.
;.ab I.* 1
. .- .1 l' :;hov; tO!v r*.;ilkta of tlim no-ft Iv ! ! lux -it;st 1I.it. redi'1,-*! t
;i.i Ose z::hx': *.':Jctr c.J-t. AI az&.aiur).. gvr. w*Ithia T.-lanl..ea ::li. $..slI.:
.': a;t,-:alt.: S'Y F.P.. wit!h .Al at It zori tg.i:l* v.:ltv, F!,, W34.a t .
. :;inW.a .: J 'i l.-. a' a m1t;lun. Vln [.1 was -.t:t posltLve. F. W .1
,.ixI::.:, .:n.' F Z W.ii :a anfntan.
i I
I ( r - I I M_ __ t i I I :1 ~. .A M I . - .. A:.::*..u
- Wl~.Itit- rAI-'.14 L~\!' - M#-:' *':
-: 4! 111 IN4}IX
.. ~ .. : . .
I.t I i'- tt .. n ...... 1 ., - - . .-- . . ,I I
~~~~~~~4. . *. 4............. .__.____......................_...
'** t ; 2.t. .. ................. -' l ,
I' ;...... **tt
.I. . . _...*..... t.%t_ ,__ ._ i...... ..-. . ____.
,_ . __ '.,__ , _. __,_ * "^.a I1
;,,.!.:.%w.! n 1..
.- ' - .' ,. - st. -. !. 9. ;
. .st' t- . -is -I., , .,t--.
- 1' -~ . ' 1s..
*.i % in , Xn -e1-s*s
'PICIM. "Ot tt'ASN:L?. tA. tlIWS -I. :::Vt't - )1-!
Il' *'2 F-i' tt r'x rflte IdP II
-i~tI !-2ii" ' .
4mLA . JI.. r v t*ts F' I for sourst* 1. HS lit t lit- titt.,I F Z fnr. thi soturce Ini _ Z !- 2.O .
)LV- .. . 7"i tbn- llkili (mr F11Z rL.lI.wI' *ni ntcotant tor %,v h*I'tn: J',. over tts I int IS .3 1192. ....
.. ~~~~~~. . : ,-- - -* . :
1 ~ ~~ ~ - ~ 3 *I S
- C.'
- 11 lv ,
- S -T.
i.
- 5. l . *7 r..l-:^ nrov
.a.*~ *.. !t!.- ioiar 1!S poer r.Inge Jn*.:.J:nt. li ent thi:4 -.. te
- rtui r-uly
- . blc f 0 .1-1parlng tle actual detector pove ainal witls a pr-estib ielhyed .-wor
,iclnal received through a lag netvork. The di(freni*;e Is anplifled ind the' differential signal £s.iolivered to a.blstable which trips whea the ditferenue..
exceeds a preset amount (_ 5' F.P., time constcnt - 2.sec.). When two or more. of the foir rate trip biattblas are in the tripped stace a.reactor trip signal is generated. The Dynamic Rod Drop Tcst deconstrated the ability o' the negative ra:e trip circuitry to properly respond to power level changes within the core. rhe rwe.r level change was Initited by dropping the two :nntrol rods mct Jifticult to detect b-. :the excore detectors diso to low worth andfor core location. in addition this test provile., a revie of thle -tation response and control
;s.t. _%be;vior due to the nueative rat.e reactor trip prior to the performance of th.- planc trip from 10)'Y F.P.
- he tes: was perfor~ed at approxia.ately 50). F.P. RCCA H-6 Ind F-A wcre J!.n.anm anlt dr.,pped into the cere. %IS flux leveU, rate clre.:'t slgnals.
r.og.ative rate histable trips and con:rol rod bo:tom light In.1Ljcailon were
-... nitor..*.I on high speed recorders alonzi with 'Iat.t r -Ltcd to the p1.1nt rctpunse
- . t!iv rva-1-tor trip.
acc:itance critcria for the tent, w.-s: aTe 1) the ractor tripped miS . retlvit of
- h.- -. zti:c r.ate trip blstibIe for a: least three of the !o-r NIS pnw r r.ar..
.. h.an:el'; 2) -&II futll length RCC:A.4 dropp.d .ind arv on the o.tton of :he Coure;
- 3) tne preo.iurizer and stea-%gener.tor vateey Jdi n:et lift; .a:d :) -a.tfety 0c a: not initlare .
or -t-v4 It aeti
- z trlfn% to. thv tr
- -a'lIpg :z re r: rI u-f'@
Ot cha:! tl Mev -all.rm tvpt.-;ritvr of thse satiot~n ;zmputer v.eri:let thisi ;ze-penc. ; also veriftcd th-.t ZS diannel ':.1 tripped pritr to c!te opening of the reactor trip breakers while channel N44, which has its detectr positioned furthest from the tvo drorped rods, tripped after the reactor trip breakors hid opene.d. It can be concluded huwever 'that ttae trip of channel N;; wait the result of the two dropped r (11-6 anS F-B) hecause the response time tiehau:tel S!3 from trip actu.tion until the stationary grippers release va. iound to be 0.304 and 0.297 second '.* for Tr~iin A anj B reipectivoly by the Reactor Protection System time reeponwe te;t Ing. The test data shoved a 0.1I. AeconJ Internal between the trips of ch.tintel ';.1 (trip activation) and N!.!. estibli-thing th!at ti.e control r.)J b.nk
.ell not ha..' been dr.wp'ng tint mI after Ihnnnel ':* .1! Lr 1;'ip.J.
- *l RC;:r:A' a
- er nu.se ; f:ull: .1ro;pe.! te t'he bot tonr . ti e ,ur.* vithunt diffl e1 ty.
.he =ax!-urn reading of pro :Iurizer preosure diurini; the pust trIp tra.iu.lvor w.11
- '05 paig. The setpo1nt fnr the prvsturf.i.r s.afec:- v.1Ive4-1s 2ISS psi . The
- c-:..Ate~ar ; ... erat~o: o ti.t-t pr .:..........uresi were d94;'.9. P4*1.1X............. '>*ps^
resp.!ctlm'l .jr the 1A., B. and !C 4te-i= ;?lCer.3tr-. Mne. 1rWvs:t =.Iln 4fety v.alve setpolet Is 1)75 o:1 g. The j.atiazl pre.sare vat :e.: wer. nll well beow t!;at re^,uirod for -4: 'e: : .tulen ac! -i nere wui ne S'fe:, InJectien it.nal I 1S) !n!:13t.d. s! *:'rf bFth !I H igh
.t.az Line-Flsw and V') Le-I.Vo eccurr d durln4 the Po'4t trlp trangs n.iv ..,It buit were not 'oInci len: .i. r :.vr :!; "S. :!t :!I . .'r 1!:: rlr: *'*.*.
Ir etl.ee:tel after Cie rip .in! qa.n1e: re.r!. .!1: -1..na ! , x!tis.r onll M__ M_ __ M
.~~~~~ ~ ,, .. ~v*^ . WiD
*4.t.anJ.- according. to the-sitquonce f isA~-rdcrd~.(
; c:urr aIpproxima~tely 2 minutes af~ek t~to. tcl.;. There wa
<° . of - thisi
.~3; hecause the SLA ueaory bank nprti43'64 W4 ;
*,4t t4'rip. ~thfs'j in' vn erond~printoUt 'sh5-.:fijiP~Yr r~N _ IM approxcu.It-6 eh. l Ytnid p ro a3 Zkto f fol~owed by Auto Start~o h UxIlsa4ri zB1~
level; D~ita obtained during trip 'ust' Indicated ilii~"'
;tcc,%s H-6 and F-A iUluuinat.d Approxiontely 2.3 seconds.,aft4it%
!;;3 negati:e rate bistable. Frcm the 3eactor Protection Sy'esea e.re¶Pon 6:.*
cestlng. tho uigative rate reactor trip rexponre tircw'was found.to 4econds including; an assus.cd value o' 0.15 seconds for the atattonory'grippS r t r.lva.e. tlurinf the perfor~ance of Control Rod Drop Tine Me3sirementR, th. hot .
*ulI R;S .ow drop tize4 of these rods were found to be 1.74 seconds. Thus the:, .:4 drup tlse appe.1red to be excesqlve. Upon further £nv..tlgacion itvas".detorxfn64 "
that, duc to Lproper terninal =arkin&9 (not aceord iv ngw .C le.d.a werv actually connected to the rod bottom indcaSting iightsiofj rod: K-8,- and R-l'). The test data thu* represents the rod drop :tie of rods dropppd:de
- o the reaccor trip and not those dropped rods which caused the.reactor tiip'.J :
.ho termta-' 'b!OC W*j l4ter changed to the pr.per libing..
., . '. -a._ j bs,;
. -.,2I
. .1 I-.
e.? I. i.. I - P~l~g ~! '-:.I. !
. . .: i. ;P.e. ,
§6,1.4. I...
Table 5.5.1-1 Secondary sygteris Prob lem
"_...v. .. _..:
DINM Irrimm UW'f1f IITTflJ
,. J 4
.V1 v.-. A
.,. .. -. : At.
r'- V '~T Ml 0 . .AReplec. IA tal t--eed pump 11--h
.t . -
. t ' - 3 1 iruvI
. *1"'-1-nt Lrfied-pup Replace 3A auxiliary feid pm rhbosting on -recirc.' -Path under engineering evaluation ei "fligh1con'ductivit-y". in -the rain Ravine Tech. Spei.* secondary water
.:uteaq gene rator during power chemistry transient limits I;
- a. "-tv o Incid ents o f s e e e fedt r line vibration .4nin reed rOKUlattng valvL. contr.ler Circtift rnodtfirattons I . . ..
Condensate System: . -. > .
.'Hish con~ductivity" inu the condenser Revive Tech. Spc.* voc vnddrry, C~hronic clogging' of condeinsate pumip Repeat cl alnjn of strain.!rs
-strainers Heater droin tank emergency dump PI--;darnaced conden.4or Otiabtiem r&.pair condenser baffle'-failure (ctwice),
- baffle and *idu b.i~ffj .41'app.rtH and res'u'lant conden~oor tube djra4%Q.
CUrculating water system: Cooling toiwer puznx fa~il to meet Under en %CIn-.'r Ing ev-z1'S..-It.II': flcw retui're:ents by 40proxiti4ttiy
'i
*Technic-al Specifications we rt modifie~d slightly to) idju,4t seconesurv w't' chemistry Spocificatlons to mare tokriahi 11=it:i Iurin., !W.L up train;.tvnti.
;~.>:yTha chng. providesorfr 16 days in -the transient' limits of 7 .microzihou.
*.KprevFlous specifftictio~n pr'ovided -for 7 days In the transient 1 imit a( 19) Thai mtcr6=nho%.'
- 3. S. 1. . . I
- v ~.-. . .- .5'R;t.-
- ,;1*-0 DATE TIMIE (hrs. EVENT~
1)11/76 1843 Reactor trip fro= 40:0 power wheistart nt 6 recorder connected to negatlvu rate birtab1e;. *-:-.. in preparation for Dynamic. Rod Drop Test.. ;-.i
.:- ReitOr CrIticat11 ... :-10/2/76 . -.....
0640 ...
- 5c g
- StatrotDynao Rotl >
I 10/2/76 .. 0755 *cospl.tid, Dynxialc ......... Ro$>4.......; j10/3/76 to Ststion to Not Standby-. Reacoto rivi 10/5/76. vater PultP repaim*- ;W .^X;X
*eco
*rtcl ** . ,
-. iO/5/76 1533 RReactor Critical.
1015/76 2151 Reactor tripped manually due to major leak -n' :' . . main food pump outboard 0a41 cooling line wlth other food pump out of service. 0161/76 0040 Reactor Critical. 10/6/76 06310 Obtained overlap Jat.a for Operatlonal Alignment of NIS while ti.N4alatlng to 50t power. 1)/7/ ;6 0830 Started Station Radlation Survey. Completed Station Radiation Suarve::. Turbino-Gonerator at 05 MWZ. Obtained sotriqa reCrics and NIS data for various bhnk D positions. - l*)/.i;6 . 1813 NIS adjusted to conform with calortmutriwe. Reactor trip due to momentary vital bus ground caused by short circuit In radi.ition monitoring rack when technicLan's scrkewdriver slipped. Adjusted ;-termedtate range Rcnictor Protection System Netpoints. 1i/10/76 1522 Ro.¶ctor Critical. 10/10/76 1610 Reactor trip due to Intcrmediate r.ing trip.:- setpoint improerly vet low. Rcactnr Crtt kal. 0 10/15/7'. Z025 Reactor shutdown fron !5% power cempleted.ddee.- to loss of condenser vacuum when opening condnesar sec:icn for Ins.ec:ion c.used b-.- s7earil condonser
- -.be '.:Il;rosv.
iI I
DATE . TIME Curs.) . LENT.-
- 109176 2015 Reactor Critical.
10./2P, , . ,td~1atton su<v .chocko preyo.Di80.4, iW
~~:~: - SQ~power. . '
10/22/76 ..:; : o85t .oi trip and eafety.injiction due to f IC# U
* > ri s. . 5 i4~
4:*.~ ; fital bus .. 0123/.76. . , 2210 Ractor Critical. ... a .10/24/76 -: -0319, *:.,.Perforicec .econeary plant calorimetric *'.' ko. * .W .. :
'0 . 'Mjus 141S ci;amiei
*t*OL correspond.to calkr
. -metric.
10/26176 1645 Started Load SvIng Test at 50% power. I 10/27/76 0530 Corpleted Load Swing Test at 50: power. 101281;6 0330 Adjusted overpower trip and .od stop'retpoints. 10/28/76 0700 .. Started turbine stretch test for 652 pouer. Started Control Room Air Conditioning roteot. 10/28/76
- 1509 Renctor trip cau'ed by turbine triptrey. Iu auto wtop oil pregstiare. Stespunde turbine
. , stretch test. -
10)/281;6 1615 Completed Control Room Air Conditioning rettst. 101'9176 0415 start.d turbinestr.tch test. - 10:29/76 0805 Completett tu:rbiti stretch tist for 65% power. 10 9/76 2016 ReActor trip due t. turbine trip from false low lou EH reservoir lovel electro-hydraulic control pump trip and lockout. 10/29/176 2140 Reactor Critical. 10/31176 , 1230 Performed secondary plant calorimetric. ll/1/76 ; 0305 Obtalned Incore flux map. 1l/l/76 1000 Started Special tiot Channel F.actor Test at 65" power. I1l//7b 1730 Completed pec.ial1 Hot Cat.nnel 1'.s.:or Te.'t: at 63. power. S. 1m). ! *. , N' IKXF'. ; Nr I. '.
i E F.?':T I/ I tIP, s I(s Heacter trtp caiuaed by steam generator low level:. coincident with low feed flow when steam flow transmitter was 1dotaCed for-reseeaing Y'. I.1 W.. Reactor Critie4l. Obtained basoline.turbine : -x stretch data. Aff
^X : 76 ' 10421' Reactor trip due to steasm &fer l coincident with steamfeedvauter tfovu isrutch:
due to Instrument rescaling. ch 4;
*11/12/76 0700 Reactor CritIcil., 4 1/21/76 0800 Obtalned baseline turbine stretch data.
1112/76 1315 Obtained turbine stretch data at 652 power. 11/2/76 1500 Started turbine stretch test at 750 poier. 11/2/76 1830 Completad turbine strctch test at 75' power. 11/2.Nh 1900 Porformed uecondary plant calorimetric. 111/216 2131) Obcained C~tltbratlon of Steami and Feadwater Flow InarrismentatIon ,t Power tat.t at 75: power. 2'30 StarteJ O)peratLon;l Alignment of NIS ga In 1112/76 adjustnentv. IIJ/i/7 " 2300 Started Opertional Allgnmnt of Procaso
.... . b.^
..f...........
Temper;Lure Instrumentation.at452 paverwr.t 1I.76 2310 Started Star! p .Adjustmetnta to Reactor Control System tast. 11/3/;6 0110 Completed Operastiontl Alignment of XuS gain adjustments. 11/3/76 0225 Performv.! scI)fndJrzt plant cilor-metric. 11/31 76 0335 ita-ied R~cttor Care Vlbrtati!n Monitoring Test. Ll/3/76 0545 Started CoolIng Tcr..r rPrforuance Teut. 11/3/76 1400 Conpleted C(oollnc Tower rirforsance Test 4t 75Z power. 11/3/76 2330 Completed Startup Adji.'thirLts of Reactor Control IS:4tem at 75' pcwer. 110/476 1905 Completed Operatrlonil .AlIgncent of Procan Temperatur:e tnq-rtjninnracaon at 75; p.-.r.
;.V). I; j * ;r ';. I
ti
.n r~: TIME (bra.) . C~~VENT,..
11/5/76 1600 Started Load Swing Test at 75.. power. I. 1l/5/76.- oaeartor.:.trip due. to-purlousjist'-trlpit ft one piwer range channel out' of xcrvice'ind' .Y current readings beingtaken on. another channel. 11/5/76 , 1727i'^ Ra3ctor..Critical. -
- -..* - '. I. %..,V^
11/51.6 Suspended Reactor Corc Vibration Test., 7 11/5/76 1853 Reactor.tripped due to an impulse pressure spike when Turbine.Control Valve& overspeod protectln *. *; control teit was performed. 11/5/76 1931 Reactor Critical. . 11/5/76 2200 Unit was rsaped down and was tripped on low steam generator level after heavy rumbling was heard and felt and feedline leak was discovered. 1115/ 7b Con~enced cooldovn for repair of feedline leak. 1116176 to Station In cold shutdcwn for Investiganion of 12/1.1 ! 6 feedwater line vibration Incident atnd performance of contalnemnt ilo1.Ation valve type C letk tcuts.-
- .O S ...,.. -
k_U S 121;17h
. .. W 1 /.6. ..
2; 3 1500 Started Auxiliir7 River W.atr Ss t.et Tt. t.
-a . .
"I11,: 1500, Completed Auxiliry River Water SVOYtez Temt.
; ScarcedStatlon hetp4tt.<
.. ..-- '>.':oeuwWl
..121!.1l7.6.. '
Reactor Critical.
*-. 12I !A176 - Reactor trip on low-ioY stean gencrator level due to large load Increasae.
U/ I1/176 Reactor startup restratned by two of thr~e cha'rging pumps out of service. 2I216/76 1649 Reactor Critical. I '117/76 0625 Started NIS Channel -444 C..in AdJu.stent aftcr new detector Installation LteAt. 1211 717i6 1730 Completed NIS Channel NtU Gain AJJusttznt after new detector lnstahl.ltion. 12/20/76 1343 Obtained Incore flu:x ma.p to determine new top of core liI-its.
- 12/21/76 2235 Adjtstted new .11n top .a, 'so.nt i:. of core I !=.Its.
5*. I. I.f *;I;';:.'!::;I NOI . !
I
. ms oy*
s
.* *.; fwrfmfl'?14 ^e¢-e.mJry plZ.sn: iogatag"rii..*
i:::.x :. Ferformd alcc a ptant otserkic. I)wo0 0103 O tai isatnted V1IVMZ*asp. Compl*te4 Fetedwiate ipitng vibration Test. 2013 Obtgai ePt~04iW V1on eWorimsvUl adM . .X d *s fl pqgr.g ' R..ctof ttrp on taw get.rterito 1 .. r, C.Il"st witb'sto vmd ft IlW 12 22. 13049 follom s5 f*ebmagor IIw iabeg.titty. Htner : LA"mg tgo f46tefr yustei. Rvedeot Crttic4L. tOM iasetot trip due :.0 turbine trtp with turbine imtpu. meber presswr tbove seototnt. I 1.19
*wvtor CritiIE41.
2o v4s.ctor trip *'g to turbine' trtp wIth urhine tyipua *:h.Ambtr pro-44uer labvv motpiant.
- ' I -: ;f X4Kccur Crtticsa.
I-. I.- 00.0 Pertormed ,.couderY.pnnt, calarluiric. 1 .".' J.'1 : 40A Xeactor trip Jue : turbine trip on low stso4top aLI preutu fro% tuknum :Auxe.
- '.'i .'
XaC.*tor itartuP dtLay.-d Ii:L! ctrusit -ie.-ctlun.
- o lin feloJ tgulAteinz, vAlvr controll.re cJn be compZate4.
F 6 aD33Y--SY3. s^
.<.p I DuqesneWii Jn29 1977 . * . 1- -, .- - .
- -.- ;# W 2'tq'^ June 29, 1977*
Beaver Vailley Power Station, Unit No. 1
. .;..: Docset, No. 50-334, .License No. DPR-66 t'&atup Report- SuHRlemant 2 t * *. . ' **. . . .4N
.,S.
United States Nuclear Regulatory Commission Director, Regional Office King of Prussia, .Pennsylvanii 19406 Gentlemen: in accordance with Regulatory Guide 1.16, Reporting of Oporating Information - Appendix A, Tcchnical Specification3, the attached supplerient to tho summary report of plant startup and power escalation is being submitted 90 days after the first supplement was issued. The startup test program has not been completed and an additional supplementary, report will be issued within three months of this supplement to further report the progress of the program. Very truly ycurs. C. N. D'znn Vice President, Ooerations Attachment cc: Director, Office of Management Information and Program Control (2 Director, Office of Inspection And Enfor emcnt (25) lb
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£ S IV Lv v-vt VI vl vii viL vLLi x
Ix x xi xL xLi xiL xiv 5.1.1.1 5.1.3.2. 4-15 5.1.1.2. 1-2t 5.1.6.2-3 5.1.6.2-6 5.2.3.1 5.2.3.1-1 S.4.1.1 5.4.1.1 5.4.S.1 5.4.6.1-S 5.4.6.1-12 S.5.1.2 5.5.1.2 5.6.2.I 5.10.1. S.10.17-211 Cha.n,;,o m..la (vn re!':v:ade{ p14w; r.lre Jon..ted !-v mr-.Irm C.lS .!%.":
I *:1 . _ _iA R.. A .. _ b..A. _pd . .. a:-. - -i , I' - , "
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,,O It.tile of Cqntentg LiL List of Flgures . .
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v'I LUit of Tables
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ix List of AbbroViatleios - .5 . I f. xi Introdzuctlon Le,.. 1.0 Initial Care Loading - : 2.0 Control Rod Testing I. It 2.1 GenerAl 2.2 Control Rod Position Indication System Test , '. 2.3 Control Rod Drive .MechanLsm Tiamint: rest I 2.4 Control Rod Drop Time !)eauurement Test 2.5 Part Length Control Rod Mechanlsm Brake Test 2.6 Rod C:ontrol System Test 3.0 Post Core LoatI, Precritical Testling 3.1 General 3)- Incore Yif..bhl& D)etector Systsnma I -.) 0 - 3.) 3.4 Incore TC/RTD Cross Calibration - Pressurizer Continuous Spray Flow Vterficatcion 3.5 Presstr'izer Heaters andi Spray Capability 3.6 RTD Bypass Loop Flow VerificUation-3.7 RCS flow Measuremont 3.8 Reactor Coolant System Flo%'C%-.Utdown 4.0 [lot Zero Power Testint; 4.1 General 4.2 Initial Approach to Criticality 4.' Tiothermal Termperature Coeffielent 4.4 RCCA Worth' Measurement ; 64.5 Boron Endpoint and Boron Worth Measurement 4.6 Minimum Shutdown Verification 4.7 RCCA Pseudo Ejecetion at'Zoro Power
.5.0 Power Asecniton teTtin; -
5.1 Physics Tests 5.1.1 General S.1.2. RCCu PiCudo EJection ~and RCCA Out of 1'osItion, 5.1.3 Power Distribution Meul:4trements
. 'O'
.. i 5.1.5 St.atic: R::'.. Dro; .an.1 Ri:C.% I)Oit t.oPiittlo 5.1.7 Dvn.inmle ;r*'
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IXS
Ftl. are Rest Flow tCo.1:4t"ou8. 0'1} ltt %II Two t."lltil l".-.:;t.l-'W:. No Loop Llaited. RM. Flow Coastdown, Two Out of Two Loop Co.1stati. One Loop Isolated.. 3.8-5 RCS Flow Coastdown, One Out of Two Loop Constdown. 3.8.9 4o.Loop Isolated.. 3.S.9IO
.;RCS Flow Coautdown, One Out of Tuo Loop Coastdown.
One:.Loop Isolated. RCS Htakeop vs. ICRR-31 4. 2.'-i' 4.2- RCS Mbakeup vs... ICRR-N32 4.2.5 t , ;* . :.,*-.*.*-- 4.2-3 RCS Boron vs. ICRR-N31 4.2.6
- 4.2-3 RCS Boron vs. ICCR-132 4.2-5 ICRR vs. Rod Postrion-N31 4.2.8 **
... 2-6 ICCR vs. Rod Position-S3 4.2.9
- 4. 3-1 tsothernal Temperature Coefficient Trace 4.3.4
!lIfferentlal and Integral Reactiv!ty W.rth - 4.4.4 Control Bank A.
Dtfferential and Integral Reactivity Worth - 4.4.5 Control Bank B.
- 4. 4e 4.4-3 D(fferentlal and tntegral Reactivity Worth - 4.4.6 Control Bank C.
Differential and Inc-ral Reactivity Worth - 4.4.7 Control Bank D. Differential and Integral Reactivity Worth - Control Banks fn% NorniL Overlap
- 4. 5-I Critical Boren Concentration vs. Integral Reactivity 4.5.4 Worth (Boron Worth)
Critical Boron Concentrition vq. Integral Reetivityv 4e.6. *. Worth (NinimumBoron Concentration) S. 1.3- Flux Hap lo. 2.A. Paw:r Di.tributtIon Xet'sreents 5.1.3.5 5.1. 3- 2 Flux .Up o':. 33, Power Dlbtribut ion M!ea:iurmennts4 5.1.3.6 3.1.1.7 1 1. I. 3- *'lutcx Mi'p ::1 1'aswor Di.4crM1'at i *....&..rlhstc 5.1. 3.I.
;. I. 3- *; .-:Flxe .sL; ':. 5.A. t'.)wer 1)1.4itr tolgttions..-zr*.l
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5.. 3-8 J.u%Mip No.. , l^X ) -)I;itribut Ionl t 5.1.3-8 Flux ~tU-Xo FlIux .".p .o. 1OA, Power Plstrlbutionlea stirev-:~tu; 5.1.3.JI
*. 5.1.3-9 Flux .. .p o. Power DLsitrlbution ,s.j,4ure.-4csttdt I A
- 5. 1. -10 Flux K'pvio. PowerDhlt*ribuLion .X..amuremnts 5.1..7
- 5. l.3I5 5.1. 3-11 ISA Power Disatribution Measurements 5.1. 3-12 Flux Hap No. 21A Pover DistrLbut lon Hvasureeent3 5.1.3.191 5.1.3-13 Flux %Up %bpNO.No. 31A Power Distriblot ton MeXasurements 5.1.3.1 9
5.1.3-14 Flux Map No. 3. A Powvr Distr~bu Lot Hamrerm.ots'I ltt 5.1.3.19i 5i.1.3-15 ;;A Po)wer b1strlbut Lozn Hvleisur-:=vntL S. 1.3. 19 5.1. 3 -Ih Flux map ;o. 4 SA Powve r Distribut ion Measurvnents 5.1.3.20! FLux Xlap No. 48A1, Pow.-r Distribiat I.':I1 .IeIjrenvut;t it 5.1.1.19 S;twtie: Rotl Drop *lI.up
. MlherO'.r RCCA, HtI10 at, 5.1.S.9 1.) Stieps
. I 1- St3th! ROd Prop Th.rrocousplet Hap 'or RC:CA I{-lO .t 5. 1.5.10 5 Ste..pt
,. 1. 5-3 StZLt. Rod Drap Tht..r-ocot:plv 'Hap tor RCCA F-12 lat 5.1.5.11 18O Stteps Stat ic Rod Drop herm:ocouzple Nlip for XCCA F-12 it 5.1.5.12 5 Steps p.od Posution vA. Exc-.) re!),tector t)iiList for RC:CA 5. 1.5. 13 H-IC 1:sertio:s 5.1.5-6 ROd Pos 1LIon vs . Excore Detector Output for RCCA 5.1.5.14 F-'1 InDo rtior Cn 5.'. 2-1 NtIS Detv^:ozr WILtion z~s 5.4.2.3
- 3. 4. 6-1 %':1IDe-tuotor Coarr nt. v.;. hicore . 5.4.6.4 I
I I
.
- 6-_ 3 ':'D.eteCt.%Ir Curre. 1t, vs.- hIn r. . %Xg;.11 ItI: ;- ' I I
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- i. e ,n- ct-cr t I;tc Lt:r rat vs. 1tncore A.' il Ofi(et 5.6
- i. *."- 2;:* ot-.-tor C'urrenst vs. Ilssiore .kxlal Offseet-*~.f1 5-4..- 9 Incore vs. Excorc M.kLxl Oufset7.vcP.-Irw 5.5.3-1 turbio- Exp.ansLon Duta Polnts i **.'
VLPHS V~bration SIgnaturs .
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3.,-1 Sam.ple RCS Flow Measurement D&L
- -' . Sample RC's Flow eiurcn'ent Data 3 . 4 3.'6-1
- Measurd RCS Flow Rates 3.-1 - RCS FLow Coastdown Time.Respowe .Summary 3.8...
4.3-1Isothermal Tvmpterature Coefficienta 4.3.3 3.-1 I~*.4-1 RShu RCCA Flu WorthKoatown zInr 1.-laisIe.SuIar Measurements 4..;. 3. 4.5-1Baron Enidpolnt and Boron Worth Data 4.5.3 4.6. I
)... 1-1 Flux ".p Data -urimtry
- .3- P.ouer Coeff~icie.utsumm-try 5.1.5-1 FMo.fl and QT Values fron Thermocouple Data During 5.1.5.:;-
RCCA R-10 Static Drop 1 S. 36 e.
.l.5-2 F.AH Ind QT VJlucs fron Thernecouple* Data Duhring 5.1.5.5 RCCA F-12 Static Drop
,.1.5-3 RCS Loop Tumperaturei Data During RCCA 11-10 Static Drop 5.1.S.6 RCS Loop Temperaturn Da.t.a Durlng RCCA, F-12 St.atic Drop 5.1.3.; !
5.1.5-; Static RCC.t Drop Flux Xip Data Sutzmary 5.1.5.8 5.1.6-1 Data Sumz.4ry, Sp.d.tl Hot Chiannel Faxctors Test 5.1.6.5 5.1.6-2 P Dti Suznxary. Sptcial Hot Channel F'aezor% T.st 5 .,I 4
;. '1.'-1 Load Sving rFe-t 0;rameters 3.2.1.4.
- 5. _ . 2,.
. '3Large Lond Rvductileo Par.ant.tr.4 Ob ;-_.;- i~~~ll4rb.. I. ad R4 -ho-' -in :'.a-r v:-.: . ^r-rU.- : w"..r
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- . 5*l-3l. Seconda.y S.ystv Y*roble, 5.6.3-1 Nucldar NoLse Measurtment Tesct tm.ry 3.t,^ . *. Radlation Survay of Reactor Contalnment, EleY. 767 ft...
5.7-2 't RdIaton urvuy of Reactor Contaltimnt. EIev. 735 ft.
, -,;.gr.1 3.7-3 Rad laion Survey of Reactor Contsinmtent, Eley. 718 ft. 56 7.4 ,-.f-if .4,
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5.7-4 tadiation Survey of Reactor Containment, Elev. 692 ft. 5.76 5 -f -.. : V; ':
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A.0t. AEX1.11 OIs. t 7ARO * :. , - - AII '.oAd;: OutL - : ' BOL Beginning Of )'aLife ;
.. : BVPS . I. .Beaver V'alley Vower Station *;4>s ' .
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CB a .. ,..' 4, Control rodx-k. Control Rod:Drtv.Pechnniu :
. *... -CRD X Departure from Nutleateo:o1Ling
'$End Of co0j' Full C.e Flux map
- DF.P. Full Power ESAR Final Safety Analysis Report Nuclear Enthalpy Rise Hot Channel Factor F{1 Nuclear Heat Flux Iot Channel Factor Axial Heat Flux Hot Channel Factor
- NFP HFP Hot Full Pouwr Hot Zero Power Inverse Count Rate Ratio AII Axial Flux Difference Change In neutron population per generation Moveable Detector Minimum MTC MIoderator Tenperature Coefficient M~TU Metric Ton Uranium MUD Megawatt Days Nuclear Instrernentatien System
,4Iss NSSS Nuclear Stea-a Supply System pcn Percent mtilli (10-5 .p)
P/L !'.lrt Lengrh control rods ppm ;'arts per million PU'R Pressurized Water Reactor PC, Primary Grade (Water) QCFX i.asrter Core Flux Map QT Quadrant Tllt Chan.iie In reactivit-y RCC.A Rod Cluster Control Assem1biv RCP Reactor Coolant I'i:.ips KI'I I.R1 iod i'n itttmn Ind1 icat logn
*~~~~~~~~~~i I................... ; Peactor Proteetlion Sv:ten Resitl tince t:; lir Detector
SW {
- 2.vt ,. ;ti . .; .rf ;. . 5. * - 4 .* .
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X.. -H *w -. ofl Shutdown rod 3Bink T.tvg RCS Hot and CoA Lieg Average Tesmptvractart TIC . Termocout~1e-,, . :. :.,?.a
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Tech Spec. . tVPS.Tech iLSpecificactons re f 5** Prorad vgble o tuirbinef Lirst stzige pfre::.urgl I
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btrouitl of 'Ssxi;1,Wtirt. leIiii~1vNnt. DVPs hi jointly '-:wed by l';zwe.'i^ ' Co1iny Oho Ediison Cmp.any, and Pennsylvania Po.vr ComLpanyt .in but operated aiolely by Duquessne Light CompIn:ve The NSSS Is a three loop, W atlnghvuae P WR with s rated capacity of 2652 r.'AT which incorp.rawes the -' 17 fuel a Iccblyaray. The turbine genrator Is":.
.lso Vedtinghouse -supp lilJ ;and it rated .t 923.4 X.e.
O-n Januar 30, 1976., BvPr was .u..d .noperating licen'e tO per:It .ue1 load. a'nJ zero powo r rhys cs tes;ing to 1%FP. -~AddItlon.l oper.ation to 5; FP Wa38 pvrrsItte.d 44. ... tesi ruqtiLrIng flux
'otc C1r4@7 Init i.al i1ll l w.1.s -c.l:;plet.d J* '-u- bm.,ry ., :1976and on May 10,- 1976 lnttli eritlc alit%.was .aLhiuv..l.
-t - - tOt a "Cjlt.lt. .power to 35. Fl' wY-S obtalned' i!.iv. ., 19;t n ¢:Il ptnrr operntion was authorized on .Isily 2, 197h. The st.ation waS declare4 ea'-m-rcial for 56.. power product ion on September 30, 197h.
Thhis report will pre~ent a Sytioppsi. of.testing b'o[nning with Initi.al core bad
-.Ind %.niLnuIit' to the 50; FP plateau oftthe Init t.l pover 1eascnien Wquence
- where.the statton Uaf declaired 'p.rtial ly comr.crcia1. Supplemenint.al reportF wil crvr. thoeee trt e mcCletdt iter part [l omarrmo.r l 'per.at Lon c'mien:iedl SintiI
- te copletcii of the .saXt-'Ip test pr.r.1b.
.,-S r - ..- . . , . . .. .
.All &t.art-t:p*te:;ts ,re 1i-ated In the tablbe oi c.ontfnt: . .Thsa)e test;; no: :2t.rteid
;: h.'-;'r
- t i:.:*; rep.'r. --:a thvse -*;Ill I!rt. *Lis- : ;6..... It
r 8;< ; *. .t. +
-- 's i.I:i ir: .suP,,!e=_,nt to thle 3etaver 'V:.l!*y ?.;uer S:.-:ten Startup :te-vr: caw~
the atirt:'p test ao:'vltis betucer. Sep:ernbr 30, 1976 whn the s:aaicn was . declared uavilab!e i.or co=ercLal power production ac 50.: reactor pover ar.
!A~cimber31. 197. Ai.
DuringSthis ti= pariod the Station escala:ed to the 75' FP test plat t a complticn of all c3itP i; scheduled for the 50 FP and 65v FP piateau except er shu-:trn and transitnt tests, BVT 1.1 - 9.24.1, AuxiliaryFevt -a-T Te3t, and BVT 1.1 - 9.4.3, .'e: Lcad Trip Test, respectively. Perforance of Z these tests was deferred until later in the startup program for scheduler c:nvenlen~e and becausa oC marginal perfor:ance of the =ain turbine bypass Sean du-p baffle supporzs in the condenser. Althoue. additional Jiata was
- abtained for nu-erous st4rtu- tests during the tile Interval covered by this supplezvntal report, relatIvely little progres -was =de on the co:*letion of the s:t::up test program and onl7 one tei:, 3';T 1.1 - 9.a.5, Dynamic Rod Drop, ?
was totail: co:pleted In th' s tioe interval. A =ajor Station out3ae for mntenance and performance of reactor contair.ment 4 Isolatlor. val'ves t:pe C leak tests occurred during Navecber and earl:y Decerber.., 1 addiftion to forced oucages and cooldowns because of failure of the reactor. plan: river water purps, condenser tube leaka;e and attendznc steam generator wtte: che=Lszry p:cb'ens; rwo incidents ol severe feedwarer line vibration occurred. The first incident occurred on '.;ova-ber 3, 1976 and promptad the W S:aticn to be;'n a schedule =aIntena..mnz u:age sov ral weeks earl:y. Aftar re:sr'r.- to po-er a-d conduzt:!n4 a :ast to in.!:e+/-:'ate the vibrat'on phezone'r.
'^htzh *41..4 .o ':p'.* th-! e:. nt or reveo: .e _3u.4e, i second .coa:er '"ne x Ii .....
-Ar~it~ittn Incident 'occurredt on Deceiber 27. 1976.'A f th o~~n eart e The basic cause of the sceverie fO',, awar line vibrations had been tentat~ively' idtit f edas-a mismatch in the desL~n porameters-for the faedwater flov.reglatidni dh odae~-
- &~~-.vves,~uuppli" ;as Part of tha NSSS packago.e sndr.o~h k§lo ,upl Zi'q.nttrlzmact$QI t~ d i41 thru'seere-yibration a e`4r l'nt 'n' "CCUbun LedtL1¶'l& impldmnttf nt
*~~~~t~i~ :coud~b~d~nti Ledand complested. 'AholA "s1adl niU tif 9 f (1- -program hecadge of.,th teedwai n Sci l'tst SfSPdr!ormed as a result: of .the feda ter itic-v~b r onf~
lt~ Included In this ieport, but may be-found in the.Annual Operating Raport submitted by the Station to the Nuclear Regulatory Co~ission.
~m
E - 9.,:. **-.w*t .c 1ITRODUCTKO. :,<<!i**2 ' 4:..- . SUPPL FlFTr so. ' This second supplement of the Beaver Valley Pow.erESation Startup Report civers the startup acttv'itiei between December 3.i, 1976 and March 33, 197t. All power ascension testing during this tiie period was parformed ac th* 75?P twit plateau. In addition to the *cheduled testing, several special a I tesmt engineering' toots wvoruperforf ed. Addtltonal dasa was obtalned for numerous startup tests and one cost, 8VT 1.1-8.2.3, was completed in the tle:. Interval covered by this supplemental report. A third Evedwater line vibration incident similar to the two incidents that occurred late lit 1976 occurred on Jan.uary S. 1977. Thle main feedwat.er rv'guLattng valve delign was determined to be the eause of the vibra.tion ineWIJents. A forced maintenance outage to repl.ce the f(vdwa.ter regulatIng valves trim with 4 ncw erla to match theddasign parameters for the valves beg.-n at this tieae. During this outage, which lasted most of the time Interv.l covered by this supplemental report, repairs and modifications to supports and equipment dasaged by the vibration incidents were effected. Several special test en.OnetLring tests were performed to check the operation and stability of the feedwater flow regulating valves both before and lmuediately following the return to power. After satiefactory completion of thesu special tosts, the starttip test program was continued. 1NV ::l. so l. ! h _ ..
5 P*1Vlt T st 5.1.1 *e*iv.r.iL The detailed program of reactor physics meawurements undertaken during power .ascen ilcah cbsr.t masured static and kinatic nuclear core at nee an cofiiksasmdino charactr tIics 4r-a's expocted and that the s asumed in th safttF 'analysis are :conservative.
* '. - .e
- Physics tests provide the control rod configurations necessary for power distribution measurements. These measuremcnts are obtained for rlrmal'and abnormal rod configurations And provide informatiLon useful In determining the static and dynamic effect of ejected, dropped and out of position control rods on the nuclear characteristics of thl core. Addititinally, power level m:MnIL-
- vorint provideit data for determining ihe power coefficient and genuratvd core axial power imbalances provide data for determining hot channel factors obtained near the BVPS Tech. Spec. limit for axial power imbalance.
The power ascension progrnm establishes power level plateaus for the performance oi tests. Portions of some physics test are performed at many plateaus (i.e. special hot channel factors and power distributlon .rnd coefficlent me.surements). Other physIcs tests are performed entirely at a specific plateau. (i.e. dynamic static rod drop, pseudo rod ejection and rod out of position test$.) This section describes the power ascension physics teiting compicted up to the 75: F.P. plateau. I 5.1.1.1 SUP'PLE'I.:imr ':1o. 2
thu p* dtdtrlbu~ 5.1. 3-1 *o 5.1.3-l;, pyresenLud ln ii:; 5rlgurs uttor. art' provLdJ'd foeorr eftdrncatipriLon for V'ter ';,::lc:; te.st tlU" vt&4't Ios. \, Flux maps display relative assembly powvr for speclric rod conflzur.:t t :. All reladve assembly povwr distributions were within an acceptable range of the expocted yaluad. . Thermocouple maps were also obtained during Belected teats. These maps provi1&d, additional data for comparison with flux map Information. Thermocouple data can be reduced by the on-site computer to provide hot channel factor values derive from temperature distribution as a rough, rapid on-site technique for determining hot channel factors. Thermocouple maps taken during power distribution measuremer provide a backup to moveable detector maps but are not sufficiently accurate for the calculations required by Technical Specifications.
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0.71') 1.1)7N 1.152' I. 1.2'5 1.215 1.19'3 1.223 1.187 1.11310.9f31 0.8361 0.53 U.-3 -1.1 t0.6 1.U -1.2 I-3.X -0.5 -0.3 1 4 2 .0 2.0 -5.2 -5.3 I. .5 0_* 2 "f. 7 5h c,.99(1 1.1121 1.2.2X 1.2:>6 1. 21' 1."I' 1.149 1.(10 0.879 -. 625
-O'. 'i -t. XI -2.1 -l . b O.1 -1.8 -1.7 -1.7
- 1.0 -f .6 _1 . _ .
. _ I i v M v I
_ C Ol.N 14t.0l811.1 1 1.- .3 1.350 1 .215 0.96010.682
- %.7 I--5; 5 7-w! 11.6
- 1.6 -3._ -7.I.4. -3.17 0.1I - KEY A. .*-
- .T- I' - I' - t 1 0.93!.
I' 9I- _4 - -- o.tdsI (.t(, II.I.1123_) I 1. 191 I .201 1.141 7.934 1O. (64 r easured -l O. lU ': - :o _7:,4
-5. 51I -
- I & 4-1.11 (l. 9 -1.5 I - - I -
i 7 .4 -S.7 z dvvlat Ion - I . , . fron expected (1. 7 3h 0.95910.727 I
. 1.a 7 .
*1
~: .. **...
1411
A. 3 4 5 6 7 a 9C 10 11 12 13 14 15 I I I i. 2 1 1 1 1
- .Y-9 919 -
().59u (. 7J7' 0.59U I
-5.6 -5.5 _
-5 6 . _
i I . . _ U.boo 0.916 U.976 1. 0(J U.976 0.916 0.600 II l . l _1.2)9 - I.,.,9 - -3.5 B1 -5.9 -3.5 I.2 1r2
- - - ,- p. -. *. - 9 - r - - * -
0.913 0.699 C 1 u.699 1. 0.913310.998 1.108 1.2 1.2 -3.6 l.u8h
-8.8 1.108 0.998
-3.6 1.2 1.2 1 2.8 n0.670 0.905 0.907 1.141 1.142 1.218 1.142 1.141 0.987 0.905 0.670 EJ
_ 8.0 7.1 2.4) 1.6 _G.E 3.0 0.1 1.6 2.0 7.1 8;0 . (1.h69 0.969 1.i433 1.108 1.111 1.150) 1.078? 1.150 1.111 1.108 1.033 0.969 0.669 1-2.8 7. 6h.7 -U.6 -Ia.N -3.6 -1.t -3.6 -1.8 -0.6 6.7 7.4 12.8 0.a46 0.968 1.151 1.120 1.167 1.124 1.178 1.124 1.187 1.120 1.151 0.968 0.846 I -6.5 -1.8 2.4 -1.0 -1.4 -2.5 ..0.1 -2.5 -1.4 -1.0 2.4 -1.8 -6.5 - G U.585 0.946 1.143 1.1b7 1.195 1. 1'19 1.159 1.060 1.159 1.139 1.195 1.167 1.143 0.946 0.585
-6.5 -6.5 -43.6 2.3 0.2 -1.2 -1.1 -0.4 -1.1 -1.2 0.2 2.3 -0.6 -6.5 -6.5 1-.770 1.002 1.189 1.239 1.135 1.236 1.113 1.162 1.113 1.236 1.135 1.239 1.189 1.002 0.770
-6.5 -h.5 -0.3 2.8 3.9 4.8 4.6 2.9 4.6 4.8 3.9 2.8 -0.3 -6.5 -6.5
'I,
- %S . 9.46 1.143 1.167 1. 195 1.139 1.19 1.06U 1.159 1.139 1.195 1.167 1.14 3 0.946 0.585
-6.v -,.5 -0.6 2;3 - 0.2 -1.2 -1.I -0.4 -1.1 -1.2 - 0.2 2.3 -0.6 -6.5 -6.5
_0. 544 0.968 1.151 1.1 2( 1.187 1.124 1.178 1.124 1.187 1.120 1.151 0.968 0.846
-u,.5 -1.8 2.4 -1.0 -1.4 -2.5 -O.1 -2.51 -1.4 -1.1) 2.4 -1 .8 -6.5 L __ (3.6h9 0.969 1.033 1.108 1.111 1.150 1.0.8 1.150 1.111 1.108 1.033 0.969 0.669 12.8 ;-7.4 b.7 -0.6 -1.8 -3.6 -1.4 -3.6 -1.8 0).6 6.7 7.4 12.8 Z
1.142 1.141 0.987 0.905 0.670 - I 0.670 0.905 0.987 1.141 1.142 1.21S
$.0 7.1 2.0 1.6 0.1. 1.0 0.1 1.6 2.0 7.1 8.0 Z s
_ _ _ _ 0.913 _. 699) 0.998 1.168 1.aSS 1.1418 0.9998 0.9131 0. 699 *: ;: 1.8 1.2 1.2 -3.6 -8.8 -3.6 1.2 1.2 12.8 KEY
- I. - - T- 1-I-
- n. WU 0.916 0.976 1 .(00't 0.4i6 0.916 0.6(H) _tcasured - F -55°°l _ r.
p - Z deviation - 2.0 I . - 1.2 - 1.2 -3.5 6
-5.9 h
-3.514-a 1.2 1.27 from expected 3.590 0.779 0.590 V-R -5.h t-
. -5.5
-5.h .4 j
- r.w --
_- C ( ...
w 3 4 S 0 11 12 13 14 15 A B. ..-.-.-.- I 94a44I 0.592
-0.2 T
0.903 L1 2(. DJ.6b' I.*hI)
. I l.
1.149 4.7 . 7.2 6.9_ I, O. 1.060 69
- 7.0_
- 4.7 0.903 0. 592
-0.'2 -0.2
- - - I 444-I - I-
-0.2' 0.900 -0.2 I
' C. .7 . - 9 0.635 0.900 0.984 1.198 1.299 1.198 0.984 0.635
'.5 .- 0.2 * -0.2 -, 4.2 I.9 . 4.2 -0.2 -n-2 1.2.5
--- 4-I I -4 -4 - 4-, -- I- I 9 - 'I . .
- I 0.6'18
- 0. h0l7 U).89 4
0.858
-0.37 1'). 9.S1 I .6 I0.974 I-0.6 I.
.146 1.148
- 1. U I 1 . l08 1 i. 144 l-.,(39 0.74 1.233 1.148 4- - 0.7
;1.144
_IA4.1 4 I.i46 0.974 1 .108 2.U0
- 4
- 0.6 1.106 I I'.0i .- I 0.858 1-4 I- ."'61 0.981 0.618
-U.3 4-.-
0.607 0.897 0.6 07 C. _ _ _ _ _ _ _ .- ft. f. .4 _n-. X' -:. 1 -4 . I -C . ) -_ 6 -*
-2.1 *
-r .
l _ -
- 1. 4
_ -x_
-n-'i v 7 S' 0.9b2 I .120 1.114 1.164 1.092 1'.108 1.0 1.164 1.114 1..120 - 0.962 0.903 I . -'1.2 * . I# -0. 3 -1.6 ..-33 -5. 3 -6.0 -5.3 I33-. -0.3 -2>.4 -0.2 U.O.24 3.010 .151 1.1..4 1.218 1.117 1.129 0.958 1.129 1117 1.214 1.141 -. 151 . I01 0. 624 I
-0.2, -o.. 1.1 *O. u 1.8 -3.2 -3.7 - ,9.3.7 -3.2 . 1.8 0.0 0.1 -0.2 -0.21
. I U.822 1.070 1.196 i.2 1.47 .230 1.09 .045 1.098 1.230 1.147 1.212 .1196 1.070 0.882
-'1. -0^.! '.2{. 0.6 s.n 4.3§ * .... 1 7.S Ji.. : 4.3.A s.n.i .O.h.L . ...L2 @-. !2 u.' Q4 1.1310, 1. 151 I1.1' I 1.214 1.117 1.129 03.958' 1. 129 1.117 1.214, 1.141 1.151 1.010 0.624 I . . ... .... ....
-u. 0.1 0.0 1.8 -3.2 -3.7 -9.9 I -3.7 -3.2 '1.8 0.0 '0.1 -0.2
.9b- -0.3'21.1I14, 1.164 1.092 1.108 1.-)92' 1.164 1.114 1.120 (3.96210.903
-2 .. --1.6 - -3. i . -5.3 -h.0 -5. 3 '-3.3 -1.61 -0.3 -2.4 -0.2 (Y
1?.b0J 1(3.8I . I - "8-OS 0.98! -ub t-4-us b 11.1 - I -. II.l1.1441039 1.144 1.108 1.106 0.981 *. 0.89710.607 -1.0
- n.61.4 I.5 -0.8 -2.1 -4.1 -4.9. - -4.1 -2.1 -0.8 1.4 -n.6 2.5 I I I UI
_ _ 0.lbls O.35 s03t.974 1.146 I148 1.233 1.148 1.i46 0.974 0.858 0.618
-0.3i iS 1.61-..
Ii
. O.e
-I
. 2.0 0.7 '2.3i I 0.7
_ _ 1 . a
. 2.0
¶ 0.6 lI 1.6 I
-0.3 I 1% 1.
.. I
. ., 0.635 0.900 0.984 1.198 1. 299 1. .198 0.984 0.900 0.635 O1*' 2 KEY _ 1.
. 2.5 -0.2 a4a4a
-0.2 4- 4.2 414 8.9 4.2 a
-0.2 .1 -0.2' .1.. 2.5, L.? I=Leasured - 0.oiir _! .
= 0.592 0.903 1.060 1.149 1.060 0.903 0.592 *: i.
-0.2 -0.2 4.7 4 - 7.2 *a 4.7 "O0.22 . -0.2 L deviation -0° 0
-I - * - 4 - 9 - = 0
£KUU VA _L 0.669 0.881 0.669 7.0 6.9 .7.0 .
I
.I 0
5 6 .7 8 9 10 11 12 13 14 15 1 2 3 4 0.583 0.769 O. 583 II , . 7-I-I-I-I-.
-6.8 -6.7 -6.8
- - - a......-. - I 0.963 0.899 0.589
,b _
0.589 0,589 0.899 0.963 0.997 0.963 0.899 0.589
-0.6 -0.6 -4.8 -7.0 -4.8 - -0.6.V - -0.6 I- I - I - I - V - I - -4.8 * -0.6. -0.6V 0.674- - -
-0.6 0.674 0.897 0.980 0.674 1.098 1.091 1.098 0.980 0.897 0.674 I h ,
-0.6 -0.6 -4.5 -8.6 -4.5 -0.6 + - -0.6I -48.7 I ,
i S.7 8.7 -
- - * -t---* 0-4--
.0.6530.653 8.7 0.885 0.978 1.137 1.145 1.236 1.145 1.137 0.978 0.885 0.653
.0.653 5.4 4.7 1.0 1.2 0.4 2.6 0.4 1.2 1.0 4.7 5.4
-:-I. 0.644 0.947 1.011 1.113 1.129 1.180 I1.10311.180 11.1291 1.11311.0l1 4.
0.947 0.644 5.0 8.7 8.7 5 .0 4.5 -0.1 -0.2 -1.1 1.0 -1.1 -0.2 -0.1 0.973 1.146 1.129 1.210 1.154 1.207 1.154! 1.210 1.12911.146 0.973 0.853 0.653
-1.3 2.0 -0.2 0.5 0.1 2.4 0.1 0.5 -0.2 2.0 -1.3 -5.7
-5.7 1.143 1.163 1.199 1.162 1.183 1.090 1.183 11.162 1.199 1.163 1.143 0.954 0.590 0.590 0.954 -0.6 -5.7 -5.7
-0.6 2.0 0.6 0.8 1.0 2.5 1.0 I 0.8 0.6 2.0
-i.7 -5.7 1.188 1.232 1.124 1.236 1.116 1;191 1.116 11.236 1.12411.232 1.188 1.01).10.777 0.777 1.011 7 5-7 i
_-04 , 2.27 2.8 4.8 4.9 5.5 4.9 4.8 2.8 2.2 -0.4 1 -5.7-
-5.7 -5.7 1.143 0.954 0.590 I 0.5j4 0.954 1.1431 .163 1.199 11.162 1. 183 1.090 1.183 1.162 1.199 1.163 2.0 -0.6 5-.7 -5.7 1-5.; -5.7 -0.6 2 .0 0.6 0.8 1.0 2.5 1.0 0.8 0.6 I-4-4 I- I- I.- J-lt - t 14 0.853 0.973 1.'46 1.129 1.210 1.154 1.207 1.154 1.210 1.12911.146 10.9-73 0.853
-5.7 -1.3 2.0 -0.2 0.5 0.1 2. 0.1 0.5 -0.2 2.0 -1.3 4 a -5.7 1 I a[p a 4a I at T F 7 1 0.6:4 0.947 1.011 1.113 1.129 1.180 1.103 1. 180 1.129 1.113 1.011 0.947 0.644
-0.1 2 -1.1 ' 1.0 -1.1 -0.2 -0.1 4 S5 5.0 .7 8.7 5.0 4.5 -0.
-t _- -
I 0.653 0.885 0.978 1.137 1.145 1.236 1.145 1.137 0.97810.885 0.653 5.4 4.7 1.0 1.2 - 0.4 . 2.6 0.4 11.2 1.0 4,7 5.4 r. . X- _.. C -
-- . j : - _I I 0.674 0.897 0.980 1.098 1.091 1.09$ 0.980 0.89710.674 s'_c:
8.7 1-0.6 V --0.6 V -4.5 44 -8.6 -4.5 -0 0.6 4 -0.6 8.7 j KEY tri - -4 measured - 0.000 0.589 D.899 0.963 0.997 0.963 0.899 0.589 tr.. .1_ 4' I ,
-0.6 -0.6 -4.8 -7 ,0o -4.8 _ -0.6 -0.6 2 deviation -00.0 I -
._ _ A - A*-
P 0.769 10.583 . fro; expected I -
-.583 _
I -6..8 -6., -6.8 a_ - __ C c C-.I
I- -- (., C *' 2 3 5 6 7 *9 10 11. 12 13 14- 15 _ _ l- l.05 10.795ii0.605 - : A _ 1 1 3.1 -3.;. -3.3' . _ _ 0.581 0.587 0.984 1.039 0.984 0.887 0.581
'3 _ _. - -
.-.. 0 . _ .* 0 -2.8 -3.0 -2.8 -2.0 -2.0
- - - I - Vf9-9 '1 94 0.884 0.653I 0.653 0.88!.0.96611.135
~0.653 0. 88;-P 0.96611.135 1.187 1.135 .0.966 0.884 0.653-
_ -_ 9
-5.3 . ,-I2 .0 .-2.0 -1.3 -0.51-1-.3 ---2-.0 --- 2.,I
'-2.01-1.3
_
- 5.3 0
- 636 5.3 0.962 0.636 0.864 .1119 1.127 1.211 1.127 1.119 0.962 0.864 0.636 D _- _
'. . I
- 2.3 -0.6 _ -0. -1.2'. _ 0.5 -1. 2 -0.4 -0.6 2.3 P . 2.6
_ 2 6 24 0.62 0.922 0.988 1.112 1. 131 1.187 1.098 1.187 1.131 1.112 0.988 0.922 0.6i 1 5.3 .X2. 2.
-1 -0.2 - O.0[-0.5 0.5 -0.5 -0.0 -0.2 -2.11 -2.2 ~.3 -
r _ - I- - I 9-9-4-9-4-4 0.861 0.955 1.,115 1.120 1.218 [1.165 1. 210 1.165 1.218 1.120 1.X15 0.955 0.8
.-".. r-3.1 -0.8 -- 1.0 'I 1.1- 1.0 2.6 *1.0 1.1- .0--0.8 '-3.1 .-4i.8 0.'.93 0.963 1.130 1.137 1.215 1.199 1.218 1.103 1.218- 1.199 1.215 1.137 1.130 0.963 0.593
-4.8 --4.8 -1.7 -0.3 -:1.9 -:3;9 - 4.0 , 3.9 4.0 3.9 1.9 -0.3 -1.7 -4.8 -4.8 I! -
O.#'?6 41.020 1.0175 1.209 A. 125 '1.22 1..147' 1.Z08 1.147: 1.272 1.125 -1.209 1.175 -1.020 0.784
.. 8 -4.8- -1.5 . 0.3 3.0 7.9 -. 7.8 ,'7.0 -7a'-- 7.9 --3.0 -0.3 -1.5 -4.8 -4.8.
..593 0.963 .1.130 '1.137 1.215 :1.199 1.218 1.105 1.21881.215 1.199 1. 1.11.130 0.963 0.593,
-. 8 -4.8 '-1.7 -0.3 . 1;9 3.9 40- 3.9 ;-74.0 3.9 - 1.9 --0.3 -1.7 -4.8 -4.8
-... . . - -- . - 4 K 0.561 0.955 I1.115 1.120 1. 218 1.165 1.210 1.165 I.218 1.120 1.115 0.955 0.861
-3.1 -(v.8 .-1.0 1.1 1.0 '.6 1.0 1.1 -1 .0 -O. 8 -3.1 -4.8
- t - ti- - - ti- - i - -i - i - i - *---4--
La
- 0. 624 0.922 0.956 1.112 1.131 1.187 1. 098 1.1187 1.131 1.112 0.988 0.922 0.624 5.3 2.2 2.11 -0.2 0.0 .- 0.5s -0.5 -0.5 0.0 -0.2 I 2.1 2.2 5.3 m ,_ . -- _______.
I -i 0.636 0.86.' 0.962 1.119 1.127 1.*>11 1.*127 . 1.*119 0.962 0.864 0.636 2.6 2.3 -0.6 -0.4 -- 1.2 O.S1 -1.21 -0.4 -0.6 - 2.3 2.6 :' A:
.- .. :. .- . .. - :J 0.651 0.984 0.966 1i135 1 18711.1 35 O.966 0.884 0.65:I
- 0, -1* 3 -0.51 -1.3 -s2On -2.0 ' KEY 5.3 999941 -2.0 5.:
0.5811 0.88710.984 1.039 0.95; 0.887 0.581 earured - 0.000 r. -.
-2.*0 -Z.0 -2.8 -3.0 -2.8 -2.0 -2 . 0 ;2 deviatiton - 00°° -
- - - 9 -
0.605
.. - 9-
- 0. 795 10. 605 frou expected - S R
-3*3 -3.4 -3.3
i 1 2 3 4 5 6 7 9 10 11 12 13 14 15 U.i75 U. 7'4i 0. 'd19 4 d 4-I 44 -
-U. *. 0. 1.9
-1.5'J11.648 0. 94b 0.979 0.96h4 0.885 0.603 0 . l l l l 4.1 I
-1.9 I -:!.U -0.6 0. U 1.0 '. 3
_I . 0.6.1 0Udd.f o,9b8 1. 1211 1. 11. I. I*3U 0. 9:$ 0.916 0.640 C l 2
-1.4 -(3.7 -0.7 0.2 - .1 2.2 3.3 2.9 -1. I
- - - 4-4-1-I - * - I"- I t? -
0.662 0.674 0.911", 1.1J2 1. 0SU 1. 15 1. 136 1. 002 0.872 0.639 D 0. rn
-o.4 -0.4 -0.1 1.5 2.1 3.2 h.Y 2. 3 -0. 1 -1.0 I.oOS U0.921) 1.OnIo. 11rs II. I) I...223 I1 3 1 i. I 1.1371 1.126 0.978 0.90010.572 2.5 .6 . 1- I -U.. I-.!, 0.1 -0. -0.1 1 -1.71 -O.f -0.9 0.41 -1.3 F ,I . ).2_.UU1 1.9:h5 1. 13; 1. 15J 1.198 ).207 1.194 1.171 1. 129 1.114 0.96010.845 6- - I l).:. -0_ 3 -Il.- I 41-4 I). 7 -DI. I -O. I -1.8 _ .4 -2.0 _s, v -2.3
. If.59. 9.6J 1.1 '7 1. IS 1.4 8 1. I'I 1.244 1.1 15 1. 223 1.174 1.1901 1.138 1.104 0.934 0.558
_. _I.9 -0.1) -0. 7 0. 5 1.1 1. 5 -0. I -U. -1.:9 _.6 -2'.0 -2.1 -2.2 _-1. 0.it) I l .477 I.lh) l.2'5 1. 156 1.24 ';' IvbU I._UI 1.129 ,1.22 1.I20 01.192 1. 116 __96 O.t;)
- 2. I -0. - -0.4 -0.2 1. I I1.$ 2. 1 0. -0.6 -0.8 -1 .8 -2.1 -2.7 -1.1 -I 1. 4 to. i91 (1.955 1.1 4 [.159 1.37 1.214 1.250 1.15( 1. 21 1.185 1.199 1.136 1.099 0.960 0.591
. . 0. 1.I I . . .1 -1 -1.9 -2.2 -2.6 0.5 2. 5 K- ~
'.3,86 ).976 1. I Jim 1.151 I . 192J I.1'95 : 1., 1.192 1.187 £.1 1.11I U. 0,hb U. wjd 0.5 3.7
-0.I -o. 6 -0. 0 -IJ. j -U.lJ -0.2 1.3 -0.5 -0.5 -1.I -1.6 4Ib .& 4 44 - .4 I 0. & -6
'). 574 0 . 883 0.987 I . 126 1. 14; 1.911 1. 140 1.221 I..155 1.138 0.991 0.937 U. 1U5
-U. 9 -0.9 0.0 -43.6 . -0.7 -0.9 -0 *. -0.1 -0.1 0.5 0.4 4; 4.5 I11 1 .0 0 1.88 0.651
=
U.614 .851 0.978 1.152 1.2:' 1. 154 1.131
-0.9 -0.4 -0.8 -o. 7 -01.9 -0.5 -0.7 -0.5 .1 3.2 5.1 1-4-4-1-4-I ---- I I I. -
-II
'I t(.610 0.:38 0.974 1.118 1.157 1.113 0.970 0.9 1 0.655
-1. 51I - -I1.4 '-(. a -0.8 -U. "J -1l .3 -1.1 2. 5.b
- - I. - - I*- I- I- I-7.P A1.571 0.A62 0. 93:E 0.959 U. h55 i.592 measured - 0.0fo°
_j .
-I . ' .5_______ LI4II 0..7 0.h 0.5 -1.1 2.1 2 deviatlon -
r a. j0 jJ 6A A
. grow CXPVCLca 0.61111 _ (7).400 3..
-I a-
- 3. 1.9
.~. . . ' -
( C.
w C I
--- I 2 3' - 4 - '- 6 - -7 a8 . 9 1.0 11 12 13 14 15
-I- . 'I r IIi I3.S)3 tl .7 72 (3.1.01 , .
p -t 4-4-4 13. 2'.(* .i. O _ . .~--~ _ -- _ _ I_ S t1..'71 V: 91' 0.8H7 0.591 i 4-s -f-4.- I.. Il ,, .1 '3.3 1.11 .! 2.0 Z.H 2.9
.I --- '1-*5-I* 1.1 0.-I-*. 2.8 - 9m
... ,., ,0.b32 0.977 1. 126 1.171 I.IJJ* -0.94 U.'9 I0.631 II. I -41. I (.9. ". I.9 3.0 'I I 1.) (il.1) -0.1
- - 199 49 9*
. -...... . ..... - - .. 0.i'68 0.9'82 1.129 1.145 -1.1I -. I S2 'I.12 0.99010'.859 1.. ', 1},6 0.631 72.9
':1.. ':
I .~ _.6 . . 0. j -. I.2 .1 -4.5S -0.7 9? 4-9 44 99 J - 0.892 .5-tJ.
- 0. '370 0.H9,) U.569 569 i- -
I U. fif l) ,1. 915 0.996 .1.111 I . 15'. .. 26 1.1)1 1.1 )I 1.22.0 1.140 _1.121 0.')70
... , 5 - 2.9 1.7 - .I _O'. I - i.1.3 - -U..1 .'-1I. 2 -1. 7 _(}." '-0.9 .. .11 ' -t ;O'.
- - - 1- - 1 -
a 1.133. I. I b J I .!!I) I (.098 I. ,20H 1.2125 1.1958 1.195 II S!I.IIU U.95.. 2.839
)O.hn!
, . -0.3 -;'.U -2.e4 _1.'9 -2.5 -2 *7
--s -. 2 11. t. S 'II.(U 0.4 -0. .
, *, '. 9.9a.7 1. 1:9 .15 1 1.22: I-.n
. I. 23 1.3 1.2';3 1.178 1.190 1.136 1.101 0.938 0. 558 1.M.t. . 1.2 -0. 0.:5 U.S -. I 1.5 -0.4 -0.3 -2.0 -- 2.6 -1.9 2.3 -2.4 -4.4 Q..7t.; 1.1)00 1 17 1.213 1.14sa 1.232 I.139 1.196 1.12i 1.213 1.112 1.186 1.136 0.996 0.75S 1.3 -IJ. -0.! -0.1 -ufe O.3 '0.4 '-0.1 -O.1 .'-1.3 -22.3 2.7 -3.0 -. 7 -0.1 I I .. . .. ~ *;; *.", 2 i.'.! 1.1t 7 I.157 I.29 1.210 1.240 1.14 7 l.'09 1.183 1.19 1.130 I .u98 0.975 0.607
. 0. O .. 1.1 .... 0.W3 (0.6 '1. I'I l -1.4' -1.6 -2. -2.4 -2.5 A.4 4.10 I
- I.., 1.2 41M t.1'14 - 1.27Uh I. l-J7 I;;1Uj 1). g9tz 0.'901 I JU,
. tl. h .'-p) . -0.5S .- I. .* -O1.* -0.7 -1.1 -2.0 .-_.3 - 0. 6 4.5
- - - -. I I -
ll) tI 907 2). _.I D,"R 1.1 10 1.1 1. 1.201 1.12_ 1.211 1.147 1.l12 0.972 0.*932 0.602 9.4
- 5. J ,-1 . -i.! -1.7 -(1.3 '-0.9 ' -I.1 -11.8 '-0.8
- 4.9 4.9 1~ 3- 1 t I.- , - I - i - ..- _
11.6-1 0.S o. 0.164 1.117 1. 146 1.207 1.1(6 I1 11S 0. 984 0.858 3.63h i= :2 =
!. 11 o.5 -I . '-l . -1.0 -O.6 4.()
1 -1.2 0.5
- 1.4 , 4.2
.s ---- I-n g7- ..... , ._ - 4.
)-. - I --_l.ZinII 1.9 14 I
- I 4 1.11,7 1.113 0.965s (1.899 0.634 _ .V *1 "
U
, .0 '-I .() --0.3. -0.3 f-.4 -. Ia -1.2 1.I ; 3.5
.1)13 -
- I 9-4-I 9-A4 f(I.1 I 5w 0st..461 ('.)7't !*"tS I.. I, n1.81sJ measured - Z ..
1.0 -0}. 3 1.6 ,I .:I 1-. I *_I. L} 2 deviation- 0. a -- 9- I- 1.4 I- 9--1.2 .- fiom expected (i.615 b.797 0.615 5.4 ').I 1 - S.:-
- I --.
w 1 2 3 4 5 6 7 8 9 10 11 12 13 14 1S 4-U 4 ,- ., - 5.0 I I
- 0. ul 7 0.60 0.61:
5.1 5.0 0.572 0.h62 0. 998 1. 069 0. 998 0.862 0.572 1-I-,-,- 0.2 U.2 3.3 4.7 3.3 0.2 0.2
- - 9-9-9. - I 9. _,--
1 I L 1 0.651
-7.0 0.885 0.2 0.975 1.L41 0.2 1.3 1.203 1.141 2.4 1.3 0.975 0.2 0.885 0.2 0.65 7.1
_ 0.632 0.8o8 0.976 1.125 1.14;2 1.201 1.142 1.125 0.976 0.868 0.63 .
. 4.0 3.3 0.1 -0.2 -1.l -0.8 -1.1 -0.2 0.1 3.3 4.0 0.611 0.916 1.005 1.133 1.158 1.209 1.122 1.209 1.158 1.133 1.005 0.916 0.611
_ 7.0 3.6 3.1 0.4
-1.0
-0.
-1.3
-1.0
-0.3 -
0.4 - 3.1 . 3.6 7.0
. 831 0.947 1.119 1.155 1.235 1.195 1.21's 1.195 1.235 1.155 1.119 0.947 0.831
-3.4 -2.7 -0.8 -0.6 -0.3 -1.0 -1.1 -1.0 -0.3 -0.6 -0.8 -2.7 -?.4 _
0-.5o 0.933 1.097 1.128 1.212 1.203 1.223 1.119 1.223 1.203 1.212 1.128 1.097 0.933 0.568
- i.., -3.4 -2'.6 -2.4 -0.8 -0.3 -0.3 -1.2 -0.3 -0.3 -0.8 -2.4 -2.6 -3.4 -3.4 o .'Q. 0. 9b7 1.145 1.188 1.121 1.22' 1.128 1.185 1.128 1.224 1.121 1.188 1.145 0.987 0.740
-3.4 -3.4 -2.5 -1.9 -1.4 -0.4 -0.4 -0.8 -0.4 -0.4 -1.4 -1.9 -2.5 -3.4 -3.4
- .~,ix 0.933 1.097 1.128 1.212 1.203 1.223 1.119 1.223 1-.203 1.212 1.128 1.097 0.933 0.568
-S.. -3.4 -'-.6 -2.4 -O.N -0.3 -0.3 -1.2 -0.3 -0.3 -0.8 -2.4 -2.6 -3.4 -3.4
... _ .t__ JI 0.947 1.119 1.155 1.235 1.195 1.21; 1.195 1.235 1.155 1.119 0.947 0.831
-3.4 -2.7 -0.8.' -0.6 -0.3 -1.0 -1.1 -1.0 -0.3 -0.6 -0.8 -2.7 -3.4 0.tl 0.91o 1.005
_ 1.133 .1.158 1.209. 1.122 1.209 1.158 1.133 1.005 0.916 0.611
,7.0 3.6 3.1 0.4 -0.3 -1.0 -1.3 -1.0 -0.3 0.4 3.1 3.6 7.0
_ 0.b32 0.866 0.976 1.125 1.142 1.201 1.1'2 1.125 0.976 0.868 0.632 . la "I
".0 3.3 0.1 -0.2 -1.1 -0.8 -1.1 -0.2 0.1 3.3 4.0
- a. n 0.651 0. 885 0.975 1.411 1.203 1.141 0.975 0.885 0.651I 8'.- .,
7.0 0.2 0.2 - 1.3 2.4 1.3 n.2 0.2 7.0.5 KEY I"
,. . =
9I 0.5721 0.86' 0.998
- 9. t 9E 1.069 0.998
-9.
0.862 0.572 ineOsured - r00001
. 2 deviatlon -. ° ]
O.- ' 0.' 3.3 ;.7 3.3 0.2 0.2 z - - - - - _- 4 ---. . - - _ i-- --- I ro cXpcCLea -W nAI 7 0.805 0.617 ry 5.0 5.1 5.0 4 __ (C c (Q.
rvlu I rvtt . th%, *zjp'pl C bl. Vy; .w.a st il I lt hiza It.s I1 I . At ,a 11 1 .- :. rnt. ILed wel1 ithiin t!e tL-9 .1LIvi rgi I.on (-4. 5i, to
-I- . - I . -.
. 74
?- ;.. . ..
aw 19 , ' ', . ' , ',,'...................................................... ; . i , _ With the reactor at a stable power levelof ^48; power, tht test was .repe.atvd In-the same- sequence- -for .. posItive U'.1. T vg was..decrea:sed
.7 to.-IF... below Tref with
*C8D -. 64 'step n it iwlithin the target band so that.the base QCFXM-(.Yap t'O)'t couli be obtained.. CED was then withdrawn and the RCS was borated to man'Italn TAvg in in attcmpt to ach Leve a .2 of. *28:. The maximum a I achieved with ARO';
was +2.912 and a qCFv4 (Xap #01) was obtained. CBD was inserted to 194 steps and the RCS was borated until Z1 was again vithin the target band. Three addltional QCF.'s (Map p22, 23, 24) were taken at tva hour Intervals. T/C maps were taken with each QCFM and plant parameters were trended to provide a computer slmulation of core conditions. . Table 5.l.b-l also shous the resIUll:; of the, poglrlve .tl flux mip datta redtwsed I 4 vlit theL INCORE eotputer code. All me.asured IICF's were wuthin Technitcal Sv¢cI Ic~at Lon 1.Im I v. At approxima tvLy 510 F.P., with -I at Its most n-goative value, wns ;t a Y mInLmum and F Z wag also at a maximum. When AJ wa:4 most positive, Fl was a
.Tx11- alld F ZWa3i at a Minimmt:.
thl* test wasx; re.r.teej fair n.;ltt i.,: *.1 wi:! tht. rtactor its a stable poIwer leve.l tif 64; F. P. '.Ith T -lF aboveT1 n th.* tar-eL hand, a base ca4.* QCFI (X p 037) wint obt. netd. C3D was thn Ilnm-r~ted us;:1n4 . [I'sn to cop.penalte for re.activltz c:lh.angv 'itil th. :t w.ei wlthItn ' of thi ne;zttJve
..1 . 'i CNI) .-t 151 .atn;d2.! .- o:md1 4e.C)nd C., I I Ms! .111 " 3.M. W.I.;
. ::.:) W.I.; Onvii l r L ' .4t -;:; ii-.. bhr.i, -sa. :-I ntX .l;.
UI
ceoIlstaKti power l:vel. Wtli :,.i just witbiu th? tuar.et b.1n4J. Ird tleitCF Ib tOt.mp d39) wa.s obtainvd. Two hours later .afourcih tQCFI (Mip '140) was obtained. The fifth QCF4 (Map #41) was delayed .1/2 hour boyond the two hours requested ; because it was necessary to realign COD and wait for stabilization. ThroughoutK:i
- the test, sIxteei plant parawters were trended to permit a computer Imulat of. core conditions and T/C maps ware taken with each QCF.I. Tablu 5.1.6-2 provides a summary~ of the results of the nqgatLve at special hot channel fceto analysis at 65% FP., in which all FSl and FQZ factors were acceptable. In ';
case 2, although the Fxy for source 158 H1 8HF exceeded Its limic, the 1. applicable FQZ wAs still acceptable when its limit was reduced as specified in, BVPS Technical Speciflca:ton 4.2.2.3g.1. in case 3, although the Fxy for source 163 HIO HI exceeded its limit, the applicable P Z was still acceptable when Itmi limit was reduced as specifLed in BVPS Technical Specif ication .4.2.2.2. All hoc channel factors quickly returned to normal after operatlon outside th..' target btnd. The Al remained welL within (-2.92 to -1i.95.b) the negative regton-s- 1 at all times. At a .table power level of -,74'. F.P. with CBD at 165 steps and .1 within the tar~et banJ, T va was decreased to .IF below Tre( 44id base cuse cCF.' ('lap 447) was obtained. Upon completion of the first I1CF., CBI) was fully withdrawn ueigtg boration to maintain a conrtant power level. A ..X within !2. of the positive
*.1 could not be reached. With the Maximum Al (+4.09%) that could be obtained, a second QCFN (Hap 448) was obtained. CBD was Inserted until .I was then obtained and two hours later QCFM (Map 450) was obtalned. A reactor trip made It impossible to obtai.n the 11sL QaCF.. As all Nit channel .1c:Ors w'.ere .keetOptablte I-.*elately .liter cthe return to tit.- t.ar.. t b.a:.!. 1: ua.; aa td.ided 0h1.t
- io .. ! ?'r .uthe.:r QUM't or !: rvp*-at it' '.t':c . T/C riapji v*r*
).1. ). I % %~~,.,
.ili%:t @,
.tell t1,107.t .&'.ILI :4IXLvni p1.m t p*arairntt **r:i wtere Lrs.*ded thrauVi1O'1L thb,. L..4t.
.x A Summary of thte mcaiureJ hor' channn1 fhctors and their..rspectlve,4 I g ai. 75%9 F.P. is shown Ln Table 5.1.:6-20' All F Z antd FNH factorsi were acceptable. In cease 2 althoug h Fxy for source 3190 H 9AA excLt-ede.,d Its limit, tie applicable FZwas StillI acceptabl.e when Its limitt wa re~duce~d as specifie~d in BVPS Tachnicak'Sp micifLeat~on 4~.2,20'2. g I
I* i. 1.6.4 1;11;1111.FMENT I'll. 11
0 SIICIAL HOT CMANELL FM.TuAS - ILj.4tIVE 4A - 47Z FULL h1(k Cosa Cb-D 0,1 F-- FiZ
.... I LIE h. u a_ .I tbI LICn -I -. 14 Ll sI t AA hI- IrlI I--ILtal t I I --- A6l,,Its. &.I .I .Inl t,
_ _ I - - I - - I -- - I. - - - Ibs I- .41114 I I.4 hbh 1.47'(0 I.5919 7.12724 j*1 $I II _I *_ _I _ _ 4 __. Ip.-t M. . I. II& IL.U% I - asO- - ______46 p 1.Nt1 2.4u12 4.44/Ra . I . -_t _ *-
- _- __
I - -- _ ;_ _ t.401'i I .71 l 2 1170 I1. I(11,11 I -at to llallj 11 _sa ._,___ 1.4 11# _
- 1. 14 _ 1. _ A I *Al 3A I ~ 4 I-_
- 2. 74:1.
i . .. _ . - I . 1417 _1._s. 14 1.1045 1.1343 7,.0 159 4.4431 1 _ 1th __
- .litb-. In 1 n.1.P- 1.114) -2. J1D-ij 4.:A'li) lDO, I. IdO I.4 Ia.? I .e.46 '2 __._34 1.414m 3 *h 2.1141 SUCIAL 1W ClAML VAfM - hr).TIVE A1 - 4bt FULL VIUE
-- . 'V - - ..
C4-D *-H F'rr r,, C i.; .Ahl IOaIttua . . . _ . I .
.__L.S_
iW.asurra A.1 iuallod Lieut &-asred A.Iaurtgd* Lisle- I FWASUVrd AJisme.
- a 4 - 44--4----4 4 -- --
r L".I 164 I-&IrS 1A.l6q I - L.Un, -.- I .1,S4 1.7112 I
- 2. 17 - I 4 .44(h' I ...... .. _ -- . _ .
U. 6 1.47fl 1.47T4 3_Ss1j7 1.7132 9--. 2.06.14 _ _§_ - ;__ - - .9-.- 1 .9E -- 1.* ... .. ._. ...
.a I ._ -__a 1. 1i46 I.bhIq- _ I_
1.713 3 2.12b _- I.401"1 *-nii a': VI.. lit NuJ I. . .._ _ n . a.a. 14 I ! Id; ___ 3. 151i2 i.4.14:A I L4LAW I.SaS 1.68S _ I* - - - - _ __
- _
2:1 biLH I .40til
- 1. Nv&. In boes;I ;; !151 3. Vail - g . 44)" I I.e. r'. I: I 41b I _.57.4' 1.5t1 _ I i.01141 2. 14.4 A tIz.marzd vueisc; ior ( 7. and xy Iner.twivi by UZ to account for manualacturlng tolerantvs
,aid wcet;saremsunt uncerLatnLy. ?tkesurel valueis for F Incr.uascd by 4S to acctiwt for "I CA uwaltsurenLtut wwtol rtaincy. (I. rn
.A iIls Fxy iS for "wurcu 15 118 IIF, tbqe aJlsnted F Z. for Oils source is FQZ a 2.040, .4 n w:
.3 itu limit Is 3.0192. thae li.Ic for IF Z. rudured Q to acutuiit for Fxy lbeing, 32Z civer 0 =
-~. ?'7
,-1 pi
- _n P
- Ca
. 1 . $1 '
.' ,: 11 - -M' -
; h ,.
-C C
.a
.;NW . . . ( / , ,
. 4-1.,
. I. .
snaCLM wrIcmunt,1 vi=oxS - YEzCATIvz LI - 65S nULL rt=~ .. V. 4 A- 11~ Iurd ' A_____I______ 1 Lxt N-X..,%,r- I AJI'.srt-d (A) Limit Fi-uvcJvrt is1it? to(-V U* !
- 2U. 1. Rit I.__1.5lu 1.a'h1 21.2eis9 I2."t1) 1b 1.11261.~8 2.ib -
CS-4
- 1 1.'5.tf.
.s.~. . :.'l I qi&11.%¶I 1 i.7u99 1.siii 2y$ 1.P'Sb___ .992 ). 20..
1.i~h d - ___ _ ___ 9I __9__ __
- -.- __ I.-
-I --- - ~ u ~ ns 17S -n -.- 6 __ _ _ _
1.0'i. AdjusttJ valwe lta V z Hvas. a I.0111f Z !n
~ o.* 1~-rwas. : i.fl:
.r. J-, t v., we rotrFzy
- Me~%. a IMU K
.4 i-.
teduced to atcctut for- U I3 F~AC lb for %jur~t 1SS N SHF at.Paittit"~ 1t. Chit AdJI$gtcdIT q for ithtiumrIce t3 1.95131;. the limit for? F h.1" ;3.aritv: tat Is :.zziz vith a YZ).Of 1:.t, b( .9510 wItI~drJ. 1%td ~A9~IV I.r I.,tscre 161 1110 Hi at Position Vt. the aJIusted V1Z tor thO ,aurcv'ls 2.11T0; tho lImIst for, T2 reduted to sccc'unt for !: C .9 ' ?1 it,~ .".oveIt II it t I ). 11 I
- 1. vItli K (Z) WI 0.I559 I;lu.J. . I -:1;-
fl: 6. k
- h A s £sfir SaMar';v W It ,U aPoelt n &S. ther adjultvd-F 6 Z (vt thtu source Is 1.92 itse limit-for T 2 rrducts to Accounit fnr
* .. 'L~'j10uOver Its likat Is 23144u witta 3 (Z) of 1 fnciudj.,
4?. '
,_,1IzL_
I Inordk'r to evAalmatv the interact ion of~ -.'veraI aiutomatitc catl'rol Ovstems Lai sustaining plant stability in large loaid reductions, a SOZ power decri~as:-- uc~~ -vate.ubn~ ~
*was inatfcidiat ArAte of '2O%/min. iateubiegeneratr Ithi-perform.'d,~t .75Z F.P.?and-will be performed at 100. It La rhan During theqO.1: Ad.:.,duceiodn.frq7Zpower.s vaau cL~
plant r P.A"O 7 L.we nianparmofev vre onitored using multi-hnelotdr' c plant ina~A:%ten't'a~tio and C66m utar ,trend (see Tablr$23-) changes sre needed to improve transient response. *, The requiremernts of the test were generatly satisfied with the.xc pcton dhita: the station operators assumed manual control of the-rod control system (se-e TAblN 5.2.3-2). Detailed evaluation of the test data later Intlcated that S -nnnual intervention of the rod control system was not necussary. The reactor
.and turbine did not trip, safety injection was not initlited, and pressurizer and steam generator saifety valves did not lift. Tavg-undershoot wail-1 th6 i; JF below the fInal ste3dy state valt and there were no Tavg oscillations kravg increased then decreased to final value). Several predicted parmeter.
maxlmum-t .an.d mInlaznts were exceeded. The -.avg peak was 14F above the initial sce.ady s*tte value while 5F above was expec:td. Primary pressure swing of 40 pol. above and 100 psi below the initial pressure were expected for the load reduction. ?ressure suings of 56 psi above and 200 psi below the Initial pressure were experienced. The lowest primary pressure was still S9 psi above
- he low pressure trip point. The steam generator levels varied by as uch as 1' (* 15. of luvrl spa.n from the Initial valuie was expected) but the steam
.:-e:0r's:.1r t -I.A.; recivers*.! II I sa fah:sh it':i.
0
;. '. !I,1 . . . .::; :O. :
~trI tt~Lh lt- ad -r,.d let isJni wa(
Wit %itid CLad.L1 4 ~zt~~ * ' to ope.ratv. TheL o~ther Nsy.At.m repnsp .i;ppe.iret :~narina whvi.t thi iavIjt '4.1$1 w.k,.?I Into aiccount. A blow.a fust in the arming4 cl~rculct Vis~ ldIntiii ed as Ltit.* cat for. thu. s~t.4m dump failing to-ooihte~." Opration Of tiw aiitomIat Ic oteam dump and automatic rod control set'msi"Wl1 be 'chocked durIng the 10O"I to 30% powe'r load reduction. G. . 1. - . . I
. I .:. ", '.
I- I I. !. 1._ :;;IIIIlJ:_III:NT NO. I le - ,
LARt;E L.')AD RMEDUCTItON
' .X .!..E ,
*- I Sr
" - ' -tI; *i'.:. . !t .. r Tablel of Xlonitord.' I'Paralme!ter4 S Largo Load Ra*diset Ion
- K I
-:v
- .. ,>Nddiuai PO'er . ,,, 9.
- ' Con Rro CrOsCi D
. CntrO1 Rod- BHnks- Cg. D.?osit . . . ..
.:t ...':- "'O
'v.
'-h- , " - - -,
.: . . O .
- 1. . . 4 . .. ..
J.1 Cen. A Cen. A Levelo S m. (:tfl. B Le v el Gun.
,f Stm. C Level St.. Gen. A Pressura 'V Stm. Cen. B Pressurd StM. Gemn. C Pressura I. ;' .
Stm. Gdn. A Stm. Flow S tm. Gen. 3 Stm. Flow S. Gen. C Smtm. Flow Stm. GCn. A Feed Wator Flow St -. (:en. B Feud Water Flow S ta. GCn. C F-evd WI4t-r Flow S ti. Cen. A Temp. Stm. Gen. B Feed Water Temp. Scm. Gen. C Fs--d SWator Temp.
..2.1.3 st'vvu.n-.,,;'r ,&). -,
II'
. . I 1 ':' 4 , ..
TABLE 5.2.l -2 ** z LAR(;E LOAD nEDUcCrl :
.. x.. ... . . *.. PARAMETRRStP ?AS
'0, INITIAL IXC I tMUM 41INIt!1 FRIAL Preuuuriter 2204 pNIM 2260 2004 2212
.:Peasura .. ,. *2250 peiC *2290 .2060 -. . '*2260 ,z. .
J)M K. jre3surtier - 4* 62 2$ . . 29 U*38Z , . *59 *25. :29 .
*:4A t *ea. *42. - 54 17. 42
-.C a 7': '35.
0er'toi.Loi *21 *55
-.C.--,t v.:,
1 ,.7 35 . *23 6ss. IC SRoam Candrator livo@l- *39 '49 i 1 '-49
*563.5r '577 *552 552 Auotionoroed 76t 76S 30J.3: 30.3S
- 6. *822 '822 632% *34% V.
Control ba-nk - 122 19 Dotto positian *195 .' 117.5 Trr..56'- w 2 - t = - 0563.5* *SS2 IT Ovet- *132 '*150 '10) *'1S
.Tmpor1turv:
SetpoLnt N~uclear Flux *76.3 *'6.3 '31.5 *31.5 01-43) tnuc1l'.ar Flux
'475.1 *75.77 '33 *33.0 Steam 798 1024.8 796 842.9 Hetader '800 10o:0 800 880 Prossure IA Fcan 2.63 x 10X 2.6S x 106 .69 x 106 .83 x 106 Cenerator p.h 1 feed tlow '2.75 a I . x 106 111 Stcrn Ccncrktor *2.75 x 100 _1.0S x 106
'eod Flow pph IC Ste=I *2.- X lo - -- -. 8 x lao Cenarrcor pph tfMFtov
*YJ1uv taken *rlva In-stalled8 pi-l-it le"trie.svistationl.
i w
*5, In:atrr' ntat l..n C.ai ihration .antl All£nmLilt 4 5.~54. I Ienernl .
S&uvral inatrumentation systems require dato tob obtainedaungnit aLll power escalatton- to provide information nece s tary-h(oithri eAb j and alignment. The SIS power range channel gaisn"are coisesrva'esotpril to initial startup and are adjusted during power escalation sh 4onaccurate. secondary plant calorimatrics. Process temperature instrum.entlon and.th . Reactor Control System are initially calibrated based on design RCdT*Avg and. . aT across the RCS hot and cold leg loops at 1002 FP. The projected 100%FP RCS T3V; and 4T are determined during power adcalation based on secondary plant; cnlortretriCs and adjustments to instrumentation systems made as required. Then:~ proper c.ltibr.atLon of stearn and feedwater flow In.trumentation 1s chocked at onels pover level plateau. and the response of the NIS excore power rante detectors to axld power dL.tributlons within the core Is determined. This report dlscuise J the c:tlibration and alignment of these items only up to the 15Z FP plato.au. ..l Further analysia will be availabla when the power escalation program is coopletod;
- A
; *S
,...I. I I!I"I';
-E-EN ^;t. -'
,TlI.TlT contro an protectton i Instraitluolr aligned prior
-ritIcallty nnd durtng initial power escalation. Prior to initial criticality the zero of the AT and T 0V signal sumnators in-theILT/T avg avg. protection and eoitrolchannels wer che'ckid land alignad. *This as accomplished by simu'laint values measured at ta l IncoroT coF.the T.li d-Cross Cmiabratlon ToitFand iakin" an' necesisry bias adjustments at tho A / .*
and signal ut R*ators. , . *.. The AT across the hot and cold leg loopi of the RCS is used as a measure of reactor powor in both the AT/Tavg control and protection channels. Therefore, ,'; the' AT signal mummators must be aligned at power such that with tha plant at nominal 100% full power as determined by secondary system calorinetrics, all
.Tchannels will read 100%. For initial power escalation the AT will be S5F.
Temperattare data has been obtained during power escalation concurrent with secondary sid. calorimetric power measurements at power levels of 30, 50 and 75Z full pover. This data was plotted and extrapolated to 100Z power to determine the 100% AT values and gain and bias adjustments made, as required, at the signal suvmatorR. These extrapolated 100% .3T values for each channel wv*re reqjuirvil to fall within O.SF of their average and their average was required LO .t'all witlili 2F of the expected value of 68F. Temperature data at 75' power w.as obtained at two different times. The original data yielded 100% 'T valueci ; that did not satisfy the above criterla. The temperature data obtalned the second time yielded sloilar results, with an avera.e 100t., AT of 60.35F and Individual chat'vnl AT'es of 61F, 0.16F, 59.3F, 61F, 60.9F and 59.4F for ch:nnaels 411.
;2I. 431. 412, .22 and 432 rvsp.ctively. The full lo.id .. T of 68F is bwaed on a 4a.,f.an tlow .of i,500 (WM. The be*:t e.Ytim.- I flVm 1i 91,1s)sP CNN w10eCI .otil!
.NO.
aAs weli as the 100 power Jat to be obtiniwd to cliv l rezi ,i V. tAnV It sue ators as necessary.
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. * .. a a Z sumai:%t.;lr .m1. ,
3j ... ,t) ~~r. Excoroiu Ik-Le~CIor t:a IlbratLoom
- ingItt~aiirtup tvstLn.g 41.~ti tbi~~nedtoieterm'Ine dth ru.sp c u NIS powa~r range' exeore dete.cLors to ~imLil puowr distributibils wlthdn the.
S. * *'* . Aw C es Thim Lniss.rm-ition Is thtan used to catibrart this dkolt flux ueti~rs' 4gV., .. an.tdltfu~e Alynput~to the oiertemnperature t&T setpointCo . rfot cUa tuncore axial power distributions. . 1-~
*prLminarycn*Lbr'tion
_ on the dolt' tlx eter-indfcation was.performod . at tho 30% FP plateau. The delta flux lutters weire calibrated based onl
* -the gr.wl'hs of dte cr currents ve Incore axial offsetat providedi Figures -5.4.6-I" hru' . These graphs' were geneirated fro'm'data 'obtained drnlu p o.7A, 9A lid 11A (saTAble 5.1.3-1). Power range dutector curronts obtalnetiduring thtli~'semvip-i owtrte'noirmial~zii to'10O2 Rio ba~iei.d an Seci'ntary plant calorintetr~ce pe~rforme'd d'aring the pe'rforat.inie Uf
- tht~s-flux ma-is Dici to Aicitat'io s 'in' th rut',twar, ~va~n
- bvtowen th,, top -.and bs'ttinm sOnis&orlFigures 5.4..b-1 tbru -a m-We the c.tihr~attIon of the dtelta flux-metors difficult and only A r~ugh calibration was eJ P0:.i14Lblo. - t).sta ubt~iintd during mmp no. 6A wast tlvi~o Intitnd"I'd t06 be used for thuse calibrations but was not used due to the fiset that. control bank D ow-%4.
Inserted to M5 steps. iiiLts deep brink inses.rtloon remulted In ex.orv Jvtvctor currents that were not cnsistent withl uthor dacta at thIuI t1mi.t. Th~is hank configuration tends to shape the. fl':x distribution towrds the outdido of th.e core, catisl4no larger ex'sore dette.~tor currents trhin expi-cted fior thi' respotcttve- power lev.:l.
'F":r 63al I t : ma .
!ux ':15.4 'e.1, . n..4i. 4) w.ort *'b1tdn.a1111I t eon:t rooo baink 4 *9*: IB 1), h .n. I r .- ;;. t :. : ais n !mr.'
hot
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.idd ItII .Ittd i LiI IIt 1 I' t IiiI :.lI lt. .MI'rIlroximil
!nln t 71; Iower. ad In 4ddition.
tILvrmLuOtIp! :11.8, pIt.lant %I.s 'rtin-ttr ies .and vlarfous plIant par~ameter5 were. recorded durlng the f Lux cup-t . *i*
- 'stmiary'of the me~aiured hot channel are shown on Table 5.4.6-1. ALL of the peaking f6ctors'and t ativ fitel.assembly power errors were satisfactory in each case. Tb'e,, F-Fxv nand" F peaking factor3 were evaluated In ccordance with che apolcak ,:Technct Specifications in order to correct for power level and to Include the extra ,"
conservatism of the Teeh. Specs. The ndjusted peaking factors are the e.asured values incre.asud to account for enalneering and measureoenc uncert.aintIW. 5 vas increased by a factor of 1.04. Fxy and F were each Increased by a faltor of 1.0.I15 ns required by the Technical SpecificatIon 4.2.4 The dat.l obt.alne.ed dxrLng the tvst was us.%d to cilibrate the recorder excore in~struenntation to the corresponding Incore nxial offsets. The calibration , dlal.t 1i4 di.Yplayt.J In FIgures 5.;.6-3 thru 9. The tines drawn through the d.tta ;;p'intr were fitted by L!se methed os least squares. P. *. I. ;.'i.- ,l:0. _;',
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POS 1T1:1 A
. XXIAL POWER
-OFFSET' (Z) 42 ' 190 _ 10.481 74.47 1516, 31
; 1516.31 4.3 f183 * -15.209 - I74.16 44 178 -18.329 74.39 1516.31 45 202 - . + 5.257 75. 57 1543.10
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X EIMASU r FAll LIMlT '(AXIMXV Fxy Fny LIXIT MEASUREP - MEASURED ADJUSTED ADJUSTED 42 1.33J. 1.3912' 1.5540 1.3J$1L 1.4940 1.6291 43 I.3S02 1. ; ;.' 1.5550 1.4223 1.5382 1.6301 44 - 1.34'70 l.2i^)O -1.5543 - 1.4360 D --1.5530 1.7975 1.4038 Q 1.5236 1.6293 45 1. 3.,2 I .3476 1.5506 1.3583 1.4690 1.6257 0od,Jd inro!Jdetd
;QUADR.ANT r I L.T MAP ) - nx1 [ imr F.I F LIMXIT MXA1MLI LIIT
.XFA U i;ED ' ;¶AJUSTED - MEASURED -
42 1.9069 2.0623 2.856 -. 1.0084 1.02 43 2.0344 2* 200 2.993 1.0104 4.'4 1.9995 7. 1625 - -2.838 - l.0059 1.02.
. I.. 45 1.8612 2.0129 2.793 i.010) 1.02
- ELNT1YVEF'EL ASSEM5LY ?t):;ER ERROR dT RELP IEVlEt[. POC.X
- 0.9 REl.,%rtIvE PO'*t-R .0.9 M\!INX=1t tLIMIT .474
.wqp-&orr-.-r6R CREN,-5SJ, INCOREAMOF'ST5 I I I t .
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