DCL-85-174, Suppl 1 to Startup Test Rept, for Nov 1984 - Jan 1985
| ML20116N091 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 04/29/1985 |
| From: | Norem M, Shiffer J, Womack L PACIFIC GAS & ELECTRIC CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| CON-#285-862 DCL-85-174, OL, NUDOCS 8505060526 | |
| Download: ML20116N091 (61) | |
Text
{{#Wiki_filter:_ _ _ _ - - _ - l 0 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT l UNIT 1 SUPPLEMENT 1 TO THE STARTUP REPORT . TO THE UNITED STATES NUCLEAR REGULATORY COMMISSION LICENSE NUMBER DPR-80 i FOR THE PERIOD NOVEMBER 1, 1984 THROUGH JANUARY 31, 1985 1 l
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PREPARED BY: C. E. BARBEHENN L. L. COSSETTE R. G. JOHANSEN T. S. 0HARA J. B. ROSENECK P. G. SARAFIAN R. A. SAVARD i l REVIEWED BY: W. E. COLEY M. N. NOREM C. G. RA0 L. F. WOMACK APPROVED FOR ISSUE: M. N. NOREM GENERAL CONSTRUCTION-LEAD STARTUP SUPERVISOR ,
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I L. F. WOMACK NUCLEAR PLANT OPERATIONS ENGINEERING MANAGER , (.
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( l DIABLO CANYON POWER PLANT j UNIT 1 STARTUP REPORT - SUPPLEMENT 1 ! Contents Page No. 1 ! Summary . . . . .................. ........ . 1.0 Thermal Power Measurement and Statopoint Data Collection . . . . . . . 2 2.0 Radiation Sunroys and Shielding Effectiveness. . . . . . . . . . . . . 8 3.0 operational Alignment of Nuclear Instruentation System ....... 9 4.0 Operational Alignment of RCS Temperature Instruentation . . . . . . . 18 20 5.0 Calibration of Steam and Feedwater Flow Instrumentation at Power . . . i 6.0 Main Turbine overspeed Trip Test . . . ................ 21 7.0 Incore Power Distribution . ..... ................ 23 8.0 Effluent and Effluent Monitoring . . . ................ 25 9.0 Chemical and Radiochemical Analysis ................. 26 p, 10.0 Control Systems Checkout 10.1 Automatic Reactor Control ......... ............. 27 10.2 Automatic Steam Generator Level Control ............... 29 10 3 Dynamic Automatic Steam Dump Control . . . . . . . . . . . . . . . . . 30 l 10.4 Startup Adjustments of Reactor Control System ............ 32
, 11.0 RCCA Pseudo Ejection and RCCA Above Bank Position Measurements . ... 33 12.0 Static Rod Drop and RCCA Below Bank Position Measurements ...... 36 13.0 Rod Group Drop and Plant Trip .... ................ 39 i
14.0 Plant Shutdown from Outside the Control Room . . . . . . . . . . . . . 41 i \ 15.0 L oa d Swing Test s . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 16.0 Doppler Power Reactivity Coefficient Measurement . .......... 45 48 17.0 Incore-Excore Detector Calibration . . . . . . . . . . . . . . . . . . 18.0 Not Load Trip from 505 Power ........... ......... 52 19.0 RCS Primary Coolant Plow Measurement . ................ 54 l l l
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DIABLO CANYON POWER PL ANT UNIT I STARTUP REPORT - SUPPLEMENT 1 List of Firures Page No. Diablo Canyon Power Plant Unit 1 Core Map . . . . . . . . . . . 4 Fig. 1 Core Average Radial Power Distribution-30% Test Plateau . . . . 6 Fig. 2 Fig. 3 Core Average Radial Power Distribution-505 Test Plateau . . . . 7 Souroe Range Detector N31 Characteristic Curve ...... .. 16 Fig. 4 Fig. 5 Source Range Detector N32 Characteristic Curve ........ 17 Fig. 6 Power Distribution for Pseudo Ejection (T.P. 42.2) . ..... 35 Fig. 7 Power Distribution for Rod Drop (T.P. 42.3) . . . . . . . . . . 38 Fig. 8 Load Cycling Pattern for Power Coefficient Yorification . . . . 47 Fig. 9 STP R-13 Saaple Plot - Incore vs. Excore Axial Offset . . ... 50 Fig. 10 STP R-13 Sample Plot - Incore Axial Offset vs. Normalized Detector Current .... .. . ...... .......... 51 11 l
l I DIABLO CANYON POWER PLANT UNIT 1 STARTUP REPORT - SUPPLI! MENT 1 j List of Tables , Page No. 5 Table 1 Pwer Distribution Results . . . . . . . . . . . . . . . . . . . , Table 2 Source Range Instrumentation Data Prior to Core Loading . . . . . 11 Nuclear Instrumentation Data Prior to Startup . . . . . . 12 Table 3 .... Table 4 Nuclear Instrimentation Overlap Data . . . . . . . . . . . . . . 13 Table 5 Pwer Range Detector Currents to give 100% Reading . . . . . . . 14 Table 6 Power Range High Level Trip SetPoints . . ... . ... . .. . 15 19 Table 7 AT and Tavg at Isothermal Conditions . . . . . . . . . . . . . Table 8 Turbine Overspeed Setpoints . . . . . . . . . . . . . . . . . . 22 Automatic Reactor Control System Response . . . . . . . . . . . 28 Table 9 Table 10 Pwer Distribution Results - Pre and Post Pseudo Rod Ejection . 34 Table 11 Pwer Distribution Results - Pre and Post Rod Drop . . . . . . 37 Table 12 Rod Group Drop and Plant Trip . . . ... .. .. .. . . ... . 43 Table 13 Doppler Coefficient Measurements . . . . . . . . . . . . . . . . 46 Table 14 Plant Response to Net Load Trip from 505 Pwer . . . . . . . . . 53 l l 1 I I l iii
SUMMARY
The Diablo Canyon Power Plant Unit 1 Startup Program activities included in this report cover the period from November 1,1984 to January 31, 1985. The full power operating license was received on November 2, 1984. Preparations for power ascension began on November 3, 1984 and mode 1 was entered for the first time on November 9, 1984. The generator was synchronized on November 12, 1984 and testing at 15% power was completed on November 22, 1984. The power level was increased to 30% on November 22.,1984 and testing at 30% power level was completed on December 19, 1984. The major delay at this test plateau was the un-satisfactory steam generator level control system performance which resulted in design changes. The power level was increased to 50% on December 20, 1984 and testing at this plateau was completed on January 5, 1985. After the rod group drop and plant trip from 50% power on January 5, 1985, capability to maintain hot stand-by conditions from outside control room was demonstrated and the plant was cooled down for a maintenance / inspection outage. As of January 31, 1985 the unit remained in cold shutdown conditions. I 1
1.0. Test Procedure No. 42.5 - Thermal Power Measurement and Statepoint Data Collection d TEST PROCEDURE The objective of this test was to collect statepoint data and verify core power level at various power ascension test plateaus. The data included
- temperatures, pressures, and steam generator water levels related to control and protection instrumentation as well as neutron flux distribution measurements.
Core power level was determined by secondary system heat balance calculations. 4 TEST DESCRIPTION ) This test was performed at nominal power levels of 15%, 30%, and 50% of j rated thermal power. Initial conditions at each plateau consisted of stable j plant parameters, equilibrium xenon, and control rods at or near fully with-1 drawn positions. Upon establishing these conditions, data were collected as concurrently as possible. Recorded information included: i - Full core flux maps through the use of the Incore Movable Detector System, (except at 15% power)
- Primary plant parametere such as reactor coolant system temperatures and i pressures,
- Secondary plant parameters such as steam and feedwater flows, steam generator
, pressure, steam generator levels, and various temperatures and pressures. 1 The collected data had a variety of applications. Full core neutron flux and power distributions were derived from the flux mapping data. Core power was determined by a secondary side heat balance on the steam gen-
' erators. Parameters.related to control and safety systems provided as input to other tests, which set forth the guidelines for necessary adjust-
! ments (if needed) to control systems. The collected information also served as a data base for steady state conditions at each of the power plateaus during the power ascension test program. This test will also be performed at the 75%, 90%, and 100% test plateaus. ] . 1 i l 1 ! 2
j RESILTS l Results specific to this test procedure included the calculated power levels l and the core power distributions. i Measured stea% state, equilibrim power distributions (i.e., relative assembly power, radial power shape, axial power shape, quadrant power tilt, peaking factors) were within Acceptance Criteria and very close to design predictions. Peaking factors were well below safety limits specified by the Technical Specifi-cations. Results of the flux amps are stuumarized in Table 1 and Figures 2 and 3 At each power plateau, Fg and FT obtained were compared to the limiting values at the next power plateau and found acoeptable. Collected data served as input to other testa discussed elsewhere in this report. i e h 3 l
O!ABLO SANYDO CCCES PLANT SINIT Wo. l CORE MAP o' l C29 CSB C37 C10 CS2 Cia C44 15 e C48 C57 C62 A19 C30 A34 C31 A09 C21 C23 C16 SG B2 5 A45 A57 B41 A08 B29 C59 C35 13 C46 C33 Bos A54 B44, B27 B21 A07 B60 A12 B30 A31 B51 B36 B54 C19 12 CIB B55 B42 A50 B04 A25 C08 C53 B40 A59 B52 A20 BSB A27 11 - C49 C64 A65 B56 A02 C09 10 - C39 A39 B05 A04 B26 A55 B01 All B23 A18 B35 A17 A47 833 A64 850 A58 C61 C20 C11 A4B B64 A56 B02 A44 B17 9- C24 818 A37 C22 A24 B57 A21 B32 A35 B53 A29 Bil A60 B38 A36 3- C13 A53 B12 A13 C03 C41 A61 B06 A63 B13 A03 807 A15 B15 7- C40 CSS A41 B62 A05 B37 A06 C38 A46 B63 A10 B31 A26 BIO 6- C34 A62 B14 i B20 A51 816 A49 C42 C06 A28 B22 A14 B09 A23 B61 A32 5- C45 C25 A43 B48 B24 B28 C47 B49 A01 B43 A52 B34 4 C05 B39 859 A40 B03 A30 B46 C51 C56 C17 CIS B47 A33 B19 A42 B45 3 C04 A22 Col A38 C54 C32 C36 2 C63 CO2 C28 A16 e C26 C50 C07 C12 C60 C43 C27 1 I l i I I I I F G H J K L M N P R A B C D E REMARKS Final Fuel Assembly Locations Core 1 ePrimary Source Assembly L
- Secondary Source Assembly rigun I
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l Table 1 Power Distribution Results 30% POWER 50% POWER ITEM TEST PLATEAU TEST PLATEAU , I CONDITIONS * - temperature ~556 deg. F '562 deg. F
- boron concentration 1080 ppa 1020 ppa
- burnup 57 MWD /MT 148 MWD /MT DATE 11-24-84 12-22-84 F - Measured value 1.427 1.391 M M12-IH
- location ** D04-IJ i
FT - Measured value 2.103 2.033 Q - location ** D04-IJ @77" M12-IH 074" Fg - Measured value 1.369 1.359 Quadrant Tilt - Measured value 1.008 1.006
- Common conditions include stable plant parameters, equilibrium xenon, control rods at or near fully withdrawn positions.
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** Assembly locations (i.e., D12) as shown in Figure 1. Pin location within assembly (i.e., IH) based on 17x17 matrix ranging from AA to QQ.
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FIGURE 2 CORE AVERAGE RADIAL PthfER DISTRIBUTION - 301 TEST PLATEAU Assembly Average Powers From Unrodded Flux Map i RELATIVE AS$DBLY PthfER(PI) N:ASURED Pi - UPCTED Pi DPECTED Pi .604 .692 .820 .741 .817 .679 .577 3.7 .9 1.2 .3 .8 .9 .9
.503 .909 1.057 1.023 1.064 1.045 1.058 1.002 1.010 .5b9 .520 2 .8 3.7 3.7 1.2 1.5 .6 .9 .9 .9 .9 2.0
.512 1.063 .592 1.148 1.138 1.184 1.139 1.177 1.088 1.097 .985 1.081 .520 3 1.0 .9 8 3.7 3.7 2.5 1.9 1.9 .9 .9 2.6 2.6 2.6
.bb6 .967 1.276 1.057 1.212 1.157 1.16b 1.13b 1.1J/ 1.UWb 1.J44 . Web .901 4
1.0 .8 .8 .8 3.7 1.2 .5 .7 -2.7 2.8 2.8 2.7 2.7
.587 1.017 1.106 1.050 1.171 1.133 1.169 1.115 1.155 1.096 1.196 l.095 1.138 1.047 .599 5
.8 .2 .1 -1.4 .6 .6 .8 .-3 .-5 -2.7 2.8 ?.8 2.8 ?.8 2.8
.680 .996 1.071 1.140 1.114 1.133 1.016 1.077 .994 1.094 1.087 l.181 1.109 1.022 .704 6
.8 -1.5 -2.4 2.5 -1.1 .3 .4 .6 -1.8 -3.1 -3.5 l.1 1.1 1.1 2.8
.807 1.038 1.127 1.115 1.146 1.009 1.041 .952 1.027 .977 1.145 l.146 1.169 1.063 .839 7
.5 -1.0 -2.5 -2.5 -1.2 .3 .1 .3 -1.5 -3.5 -1.3 .2 1.2 1.4 3.5
- .738 1.038 1.086 1.158 1.086 1.075 .947 1.008 .941 1.067 1.105 1.196 1.133 1.050 .750 g
.1 .1 -2.9 -2.9 -2.9 .8 .9 .4 -1.5 -1.6 -1.1 .3 1.3 1.1 1.6
.810 1.047 1.116 1.104 1.120 .984 1.017 .941 1.047 1.017 1.170 1.153 1.176 1.066 .822 9
.1 .1 -3.5 -3.5 -3.5 -2.8 -2.5 -1.5 .4 .4 .8 .8 1.7 1.6 1.4 ;
.711 1.050 1.060 1.128 1.087 1.091 .982 1.064 1.002 1.134 1.133 1.176 1.104 1.023 .693 l 10 3.8 3.8 -3.5 -3.5 -3.5 -3.4 -3.0 -1.9 -1.0 .4 .6 .6 .6 1.2 1.2 l
.605 1.0$6 1.150 1.029 1.144 1.101 1.131 1.084 1.127 1.069 1.146 1.050 1.091 1.045 .591 II 3.8 3. 8 3.8 -3.5 -1.7 -2.2 -2.0 - 3. 0 -2.9 -3.2 -1.5 -1. 5 1.5 P.6 ?.6
.9 02 .978 .309 1.075 1.178 1.131 1.15E 1.110 1.031 1.028 1.241 .958 .904 12 j
!.9 1.9 .8 .9 .8 -1.2 -2.9 -2.9 -3.2 -3.6 -3.6 .1 3.1
.516 1.074 .977 1.118 1.093 1.137 1.082 1.150 1.094 1.111 .953 1.062 .524 13 1.9 1.9 1.8 1.0 .5 -1.6 -3.3 .5 .3 .4 .7 .8 3.5
. bib .891 1.J34 1.009 1.044 1.038 1.078 1.035 1.043 .897 .521 14 1.9 1.6 1.5 .2 .4 .1 2.8 2.4 2.3 2. 2 2.8
.590 .679 .810 .740 .833 .702 .596 15 i.3 . 9 .1 .3 ?.8 2.4 2.3 l R P N N L K J H G F E D C B A DIABLO CANYOK POWER PLANT - UNIT I 1
F18URE 3 CORE AVERAGE RADIAL POWER DISTRIBUTION - 5 01 TEST PLATEAU Assembly Average Powers Free Unrodded Flux Nap RELATIVE ASSEMBLY PERER (PI) , IEASURED Pi - EIPECTED Pi j EXPECTED Pi
.586 .686 .812 .741 .812 .683 .581 1
.9 .5 .8 .8 .7 .0 .0 1
.517 .892 1.020 1.013 1.03t 1.026 1.026 .933 .997 .859 .503 2 2.6 2.6 .9 .5 .5 -1.0 -1.5 -1.5 -1.4 -1.2 .1
.511 1.056 .%2 1.114 1.105 1.154 1.110 1.143 1.00J 1.051 .940 1.040 .509 3 1.5 1.5 1.1 .9 .9 .2 .6 -1.1 -3.0 -2.1 -12 .1 1.1
.879 960 1.271 1.059 1.167 1.133 1.195 1.136 1.164 1.047 1.2L9 .952 .875 4
1.1 .9 .4 .5 .5 -1.3 .2 .9 .7 -1.7 .6 .1 .7
.587 1.014 1.108 1.059 1.173 1.138 1.19E 1.154 1.198 1.136 1.158 1.061 1.103 1.010 .578 5 1.0 .3 .4 .5 .4 .5 2.5 2.5 2.5 .4 .9 .3 .1 .1 . 6
.681 1.003 1.077 1.156 1.125 1.149 1.04d 1.122 1.038 1.138 1.120 1.167 1.091 1.004 .679 6
.2 .5 -1.7 -1.4 .5 .8 2.1 2.2 1.4 .2 -1.1 .5 .4 .4 .6
.798 1.035 1.142 1.133 1.161 1.031 1.071 .987 1.066 1.016 1.172 1.139 1.155 1.038 .816
-1.0 .7 -1.2 -1.2 .7 .8 1.8 2.0 .8 .7 .2 .7 .0 .4 1.2 f
.721 1.022 1.099 1.183 1.107 1.10 3 .979 1.037 .%5 1.094 1.12] 1.191 1.118 1.034 .746 8 -1.9 -1.4 -1.6 -1.2 -1.6 .5 1.1 .8 .3 .3 .2 .5 .0 .1 1.5
.790 1.027 1.142 1.136 1.153 1.015 1.051 .968 1.059 1.023 1.17! 1.148 1.165 1.042 .817 9 -1.9 -1.4 -1.2 -1.0' -1.4 .7 .1 .0 .1 .1 .8 .1 .8 .0 1.4 l
.683 1.007 1.071 1.160 1.In 1.127 1.01! 1.092 1.023 1.146 1.14: 1.184 1.114 1.022 .692 1 10_
.1 .0 -2.2 -1.0 -1.1 -1.1 .8 .4 .0 .5 1.0 1.0 1.7 1.4 1.4 j .594 1.0 34 1.129 1.054 1.15E 1.120 1.160 1.113 1.156 1.128 l.179 1.074 1.112 1.039 .597 11 2.3 2.3 2.3 -1.0 .9 -1.0 .8 -1.1 ' 1.1
.3 .9 .9 .7 2.8 2.8
.880 .954 1.290 1.071 1.179 1.131 1.176 1.126 1.167 1.070 1.282 .967 .898 12 1.2 .2 1.0 .7 .6 -1.4 -1.8 -1.8 .4 .5 .4 1.6 3.2 1 1
.507 1.047 .965 1.117 1.094 1.130 1.088 1.137 1.088 1.112 .%0 1.060 .522 13
.6 .6 1.4 1.2 .1 -2.2 -2.6 -1.6 .7 .8 .9 1.8 3.7
.507 .877 1.02] 1.007 1.022 1.017 1.034 1.01J 1.UZ4 .551 .bl5 14
.7 .9 1.2 .1 -1.9 -1.9 .8 .6 1.3 2.0 2.8 15 .578 .673 .7% .727 .809 .68E .588 l V .4 -1.4 -1.2 -1.0 , .4 .6 1.3 I hon 1H R P N N L J K H G F E D C B A DIABLO CANYON POWER PLANT - UNIT I 7
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- I 2.0 Test Procedure No. 1.15 - Radiation Surveys and Shielding Effectiveness TEST OBJECTIVE I The objective of this procedure was to verify the adequacy of the radiation surveys and shielding ef fectiveness program as prescribed by Nuclear Power Operations (NPO)
Procedure TC 8401. The main objective of the test program was to measure radiation levels in accessible areas of Unit 1 at various reactor power levels and identify any location where shielding may be deficient. [. TEST DESCRIPTION < Radiation measurement locations were selected such that locations with the highest l uncertainty of proper shielding were measured. Common measurement locations were entrances to labyrinths, shield wall penetrations, and primary shield walls. Each j j measurement location was identified with a Radiation Base Point (RBP) number or pen-1 etration number. Neutron and gamma radiation measurements were performed at each REP for each reactor {, power level plateau. The radiation dose rates at each RBP were expected to increase linearly as a function of increasing reactor power. A linear regression by a least l squares fit of the measurements was performed for each RBP. The resulting linear j equation was extrapolated to 100% reactor power and the extrapolated dose rate compared to FSAR zone requirements. j! A correlation coef ficient was calculated for each RBP and used to measure the degree ! of linear relationship between reactor power and the measured dose rates. A larger ' correlation coefficient indicates a greater degree of linear relationship, with a correlation coefficient of 1.00 being an exact linear relationship. i TEST RESULTS I The maximum reactor power level achieved at the time of this report was 50%. Further 1 radiation measurements will be performed as higher reactor power levels are achieved, the results of which will be included in a final report. l A preliminary review indicated that all surveyed areas, when extrapolated to their l maximum value at 100% power, will meet the FSAR radiation zone requirements. Most i RBPs exhibit a positive linear relationship between increasing reactor power and ' dose rate measurements. Special survey procedures have been initiated in those areas identified as possibly exceeding FSAR zone limits in the " Shielding Design Review". Close monitoring and review of survey data will continue throughout j the remainder of the Bio-Shield Survey of all RBPs at the remaining test power plateaus. i l 4 8 4
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3.0 Test Procedure No. 42.9 - Operational Alignment of Nuclear Instrumentation TEST OBJECTIVE l The objective of this test is to align and monitor the Nuclear Instrumentation System (NIS) prior to and during core loading, and through power ascension. t i TEST DESCRIPTION Prior to core loading, the pulse amplifier attenuator and discriminator voltage settings, the high voltage power supply plateau, and the operating voltage settings for the source range channels were determined. Prior to Startup, the initial trip setpoint for all the nuclear instrumentation channels was determined. During Startup, the overlap between source range and intermediate range and between the intermediate range and power range channels were determined. During power ascension, the power range detector currents vs. ' core power were determined and the flux deviation alarm settings were monitored. At the 50% power test plateau, the intermediate and power range operating detector voltages were checked. t Af ter shutdown from power operations at the 50% power test plateau, the source range operating voltage was checked, the intermediate range detectors' compensation voltages were set and the current for each power range channel which gives 100% , _ power level indication were obtained. 4 TEST RESULTS Required adjustments, calibrations, and setpoint determinations were accomplished without significant problems using standard I&C procedures. The source range instru-mentation data prior to core loading is listed in Table 2. The Nuclear Instrumen-tation data prior to Startup is shown in Table 3. Results of the nuclear instrumen-tation overlap data taken prior to criticality and at various power levels are shown i in Table 4.
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Intermediate range and power range detector characteristics were determined at the 50% Detector power plateau prior to the power range Incore-Excore detector calibration. ] 4 plateaus were also determined at this power '1evel. The following were performed shortly af ter shutdown with a core burnup of at 1 least 1200 MWD using I&C procedures. (1) Source range detector high voltage settings were determined to be 2200 Vdc for N31 and N32 using the reactor as neutron source. Figures 4 and 5 show the source range detector characteristics. + + 9 I l I _ - - __. - -- - -. ,_ .. _ _ , _ _ . _ _ ~ __ _ - _ - _ . - -
l (2) Power range detector currents equivalent to 100% indicated power were determined. Results are shown in Table 5. (3) With an overlap of about three decades existing with the source range, each intermediate channel was adjusted to provide proper compensation (6.5 Vdc for N35 and 25.8 Vdc for N36). Power range detector high level trip setpoints were reset prior to power increase to the next power plateau as listed in Table 6. This test will continue through the rest of the power ascension program. 'l i 5 i 10
Table 2 Source Range Instrumentation Data Prior to Core Loading Detector Parameter Units N31 N32 Attenuator Setting db. 10 10 Discriminator Voltage Vdo -1.02 -1.013 Detector Voltage Vdo 2350 2350 Detector Voltage bistable trip Vdo 2240 2175 High nux alars ops 18.7 18.5 High nux trip ops 8.578 x to 8.33 x 10 l I I l l 11
Table 3 Nuclear Instrumentation Data Prior to Startup Intermediate Banas Qiannels Detector Parameter Units N35 N36
- 1. High Voltage Setting Vdo 800.76 799
- 2. Omnpensating Voltage Vdo -40.287 -40.093
- 3. Cepensating Voltage Bistable Trip Vdo -19 999 -20.0
- 4. Loss of Detector Trip Vdo 703 700.2
- 5. P-6 Bistable Trip amp 1 35 x 10-1 1 33 x 10
- 6. Rod Stop Bistable Trip amp 7.4 x 10-5 7 1 x 10
- 7. Reactor Trip Bistable amp 9.8 x 10-5 9.8 x 10 '
Power Banae Q1annels Detector Parameter Units N41 N42 N43 N44
- 1. High Voltage Setting Vdc 800 800 800 800
- 2. High Voltage Bistable Trip Vdo 698.7 698.3 699.1 698.8
- 3. P10 Bistable Trip $ 10.3 10.26 10.21 10 .27
- 4. P8 Bistable Trip 5 34.61 34.51 34.54 34.52
- 5. Overpower Rod Stop Bistable Trip 5 103 09 103.07 103 04 103 02
- 6. High Neutron Flux Rate Trip 5 4.704 4.752 4 728 4.776
- 7. Flux Rate Time Constant soo 2.14 2.145 2.15 2.145 12
I Table 4 Ikselear Instrumentation Overlap Data i PRECRITICAL DETECTOR READINGS 01 POWER '3% POWER ~15% POWER. 30% POWER ~50% F0WER SOURCE RANGE (cps) N31 - Control Board 45 3.8x10' Blocked Blocked Blocked Blocked 3.1x104 N31 - NI Drawer 45 Blocked Blocked Plocked Blocked N32 - Control Board 55 4.8x10' Blocked Blocked Blocked Blocked N32 - NI Drawer 55 4.8x104 Blocked Blocked Blocked Blocked INTEIDEDIATE BANGE (amps) N35 - Control Board 1.0x10-II 2x10-10 1.2x10-5 8x10-5 1.0x10-4 2.5x10-4 N35 - NI Drawer 1.0x10-II 2x10-10 9,9x10-6 6x10-5 1.0x10-4 1.4x10-4 N36 - Control Board 1 1.0x10-II 2x10-10 1.3x10-5 8x10-5 1.0x10-4 2.5x10-4 N36 - MI Drawer .1.0x10 -II 2x10-10 1.0x10-5 6x10-5 1.0x10-' l.4x10-4 POWER RANGE (%) N41 - Control Board 0 0 3.0 15.0 32.0 51 N41 - NI Drawer 0 0 2.9 14.5 31.0 50 N42 - Control Board 0 0 3.0 14.0 31.0 46 , N42 - NI Drawer 0 0 3.0 14.0 31.5 46.5 l N43 - Control Board' O O 1.0 13.0 30.0 50 N43 - NI Drawer 0 0 2.9 14.0 31.0 50.5 N64 - Control Board 0 0 2.1 14.0 31.0 47 ! M44 - NI Drawer 0 0 2.9 14.5 32.5 47.5
Table 5 Power Range Detector Currents to Give 1005 Reading Upper Detector Lower Detector Current Current Detector
%a) %)
21 432.7 432.2 l N42 432 3 467 4 434.4 450.5 23 m4 441.1 439.6 l f I i 4 1 l I l I i 14 l l l
Table 6 Power Range High Level Trip Set Points _ Desired Actual Set Point (5) Power Plateaus Setpoint ($) N41 N42 N43 N44 ($ RTP) 24.6 24.8 24.5 24.1 0 to 5 25 + 0.5
- 1.0 1
25 + 0 25.0 25.0 25.0 24.9 15
- 1.0 40 + 0.5 40.1 39.8 40.1 40.0 30
- 1.0 l
60 + 0 59 9 59 7 60.0 60.0 ! 50
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a 4 i 15
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4.0 Test Procedure 42.8 - Operational Alignment of Reactor Coolant System Temperature Instrumentation TEST OBJECTIVE The purpose of this test procedure is to align the AT and Tavg instrumentation channels during power ascension. TEST DESCRIPTION AT and Tavg data collected in Test Procedure 42.5 - Thermal Power Measurment and State Point Data collection, was transcribed to this test at each power level. At isothermal conditions, AT and Tavg values were also determined from Thot and Teoid values. Linear regression analysis will be performed at the 75% and 100% plateaus to determine the extrapolated AT and Tavg at 100% power and the channels calibrated, if required. TEST RESULTS i At isothermal conditions AT and Tavg values agreed with the values calculated from Thot and Teold values within the specified tolerance as shown in Table 7. Further analysis and calibration, if any, will be performed at the 75% and 100% j power plateaus. f f i l 18
. l l
j Table 7 AT and Tavat at Isothem=1 Condi tions Loop 1 Loop 2 Loop 3 Loop 4 Parameter (des. F) ' 546.17 546.14 546.2 546.08 Thot 5 % .54 Toold 54.42 5% .45 5M.36
-0.25 -0.31 -0.16 -0.46 AT (calculated) -0.14 -0 .37 AT (measured) -0.19 -0.22T 546*.3 546'.3 546.28 SM .31 Tavs (calculated) 54.45 5W.6 Tavs (measured) 54.45 54.43 Accentance criteria AT (measured) = AT (calculated) 01 3 deg. F Tavs (measured) Tavs (calculated) A1 deg. F i
1 19
l
. l 5.0 Test Procedure No. 4.1 - Calibration of Steam and Feedwater now Instrumentation at Power TEST QILIECTIYE5 he objective of the test was to onlibrate the steam flow instrumentation to feed-water flow and to perfom a cross-check verification of all signals indicating feedwater and steam flow with the reference feedwater flow determined by high accuracy differential pressure gauges.
TEST DESCRIPTION De foemater and steam flow instruentation output signals were checked against the reference feedwater flow at steady state power levels of 305 and 505 Rated j Rernal Power. Test data collected as part of Startup T.P. 42.5, Thermal Power Measurement and Statopoint Data Collection, were analyzed by this procedure to determine deviation of steam and feedwater flow ooepared to the reference feedwater flow. l l I TEST RESULTS On November 24, 1984, data were analyzed for the 30% test plateau. Five Steam now Transmitters and two Feedwater now Transmitters were outside the tolerance of $25 and 11.55 respectively. As allowed by the procedure at 305 plateau only, it was j decided to recalibrate all Steam now TranJ51tters but Done of the Feedwater now
- Transmitters.
On December 27, 1984, data was analyzed for the 50% test plateau. No Steam now Transmitters and one Feedwater now Transmitter were outside of tolerance. Rose will be calibrated during the outage and reverified to be within tolerances at 505 power prior to escalation to 755 power. l The test will be repeated at the 755, 90% and 1005 power plateaus. I J 20
6.0 Test Procedure No. 22.9 - Main Turbine overspeed Trip Test TEST OBJECTIVE The objective of this test was to test the Main Turbine Overspeed Protection System. TEST DESCRIPTION The turbine was run at 80MW for a ten hour " Soak" period and then unloaded. The overspeed setpoints (103%,111% and 111.5% of normal speed) were verified by Nuclear Plant Operations Surveillance Test Procedure (STP) M-218. l TEST RESULTS M-218 was performed satisfactorily and the results are shown in Table 8. 21
Table 8 Turbine Overspeed Setpoints Trip setting Actual Acceptance Criteria DEN 1035 1855 rps 1854 1 5 rpm Neobanical 1115 1966 rps 1998 + 2, -36 rps 111.55 1928 rpm 1926 5 rps DEHe
# NOTE: The setpoints are automatically reduced by 4.5.5 while testing the 11.55 trip setting from the disitial electrohydraulio (DEH) Unit.
l 1 l i 22
j 7.0 Surveillance Test Procedure R-3A - Incore Power Distribution TEST OBJECTIVE The purpose of this procedure was to obtain flux maps using the Moveable Incore Detector System (MIDS). The detector outputs were used to determine such core parameters as axial flux distributions, peaking factors, and core tilts for several startup tests during the 30% and 50% power plateaus. The flux maps were also used to fulfill the routine surveillance requirements. TEST DESCRIPTION Various full core flux maps and quarter core flux maps were performed during power ascension testing. Full core maps nominally involved 12 passes through the core by the six incore detectors. Quarter core maps involved three passes of the detectors in selected locations and were used only for determining axial flux distribution. Digitized detector output from the flux maps served as input to the INCORE computer code which calculated relative assembly powers, peaking factors, and quadrant power tilts for the full core cases.
! Below is a chronological summary of the flux maps taken during this portion of the power ascension test program to 50% power:
) -30% power, all rods out (ARO), equilibrium xenon; provided base line data for T.P. 42.5 - Thermal Power Measuremert and Statepoint Data Collection.
-30% power, Control Bank D at ~177 steps (100% RTP Rod Insertion Limit),
equilibrium xenon; provided reference data for T.P. 42.2 - RCCA Pseudo Ejection and RCCA above bank position measurements. J j -30% power, Control Bank D (except RCCA B-6) at ~177 steps, RCCA B-6 fully withdrawn to simulate a rod ejection; provided post-ejection data for T.P.
- 42.2 - RCCA Pseudo Ejection and RCCA above bank position measurements.
-50% power, ARO, equilibrium xenon; provided baseline data for T.P. 42.5 -
Thermal Power Measurment and State Point Data Collection and provided reference data for T.P. 42.3 - Static Rod Drops and RCCA below bank position . measurements. 23
-50% power, ARO (except RCCA H-4), RCCA H-4 fully inserted to simulate a rod ,
drop; provided data for T.P. 42.3 - Static Rod Drop and RCCA below bank pos- l ition measurements.
-50% power, ARO equilibrium xenon; provide reference data for STP R Nuclear Power Range Incore-Excore Detector Caliuration.
-7 quarter-core maps at 50% power provided data for STP R-13 Nuclear Power Range Incore-Excore Detector Calibration.
TEST RESULTS Each of the flux maps listed above was analyzed and determined to be satisfactory. Results are discussed in more detail in other sections of this report. The incore power distribution will also be determined at 75%, 90% and 100% power levels. t 24
8.0 Test Procedure No. 1.16 - Effluents and Effl'uent Monitoring TEST OBJECTIVE The main objective of this procedure was to document the existence of an adequate program to verify the icvel of radwaste releases. Specifically, this test verifies the calibration of the effluent monitors by comparing with laboratory sample analy-sis results. TEST DESCRIPTION , Effluent monitoring is an ongoing program by Nuclear Power Operations (NPO) which involves following the procedures in the Plant Manual. The test collects data to verify the effluent monitoring program and from these data verifies the calibration of the effluent monitors. The intent was to perform this verification at the 30, 50, 75, and 100% power test plateaus. TEST RESULTS A minimum activity level is required to adequately judge the calibration of each monitor. However, through the 50% power test plateau, the activity levels at each monitor had not been large enough>to verify the calibration of the monitors. The test will continue through the remainder of the Power Ascension Program. 4 i l l l l l l 25 i
)
-V.
9.0 Test' Procedure No, 1.17 - Chemical and Radio' chemical Analysis TEST OBJECTIVE The objectives of this procedure were to document the ability to perform reactor plant chemical and radiochemical analysis and to document the ability to control water chemistry.
't TEST DESCRIPTION Plant systems sampling, analysis, and chemistry control are part of an ongoing program by N' u clear Plant Operations (NPO) involving the use of approved procedures in the Plant Manual. One specific area of interest was to evaluate the performance +
of the Boron Concentration Measurement System (BCMS). i j TEST RESULTS Reactor plant chemistry has been maintained within Technical Specifications Limits up to and including the 50% power test plateau. Main Control Room indication of j RCS boron concentration has agreed with chemical samples to within 10 ppa when
- the plant is operating under steady state conditions.
9 J 4 4 d e 26 l
10.0 Control Systems Checkout 10.1 Test Procedure No. 38.1 - Automatic Reactor Control TEST OBJECTIVES The objective of this test was to verify the performance of the Automatic Reactor Control System in maintaining reactor coolant average temperature within acceptable steady state limits. TEST DESCRIPTION The Rod Control System was switched from manual to automatic control with the reactor at equilibrium conditions at 30% power and system response monitored. With the Rod Control System in manual, Tavg was increased approximately 6 deg. F above Tref by withdrawing Control Bank D. The Rod Control System was then trans-ferred from manual to automatic and plant response recorded. Af ter the plant stabilized, Tavg was decreased approximately 6 deg. F below Tref by insertion of Control Bank D with the Rod Control System in manual. The system was transferred from manual to automatic and plant response was recorded. i TEST RESULTS The Automatic Reactor Control System responded properly to a +6 deg. F temperature
~
, mismatch between Tavg and Tref. The Test Acceptance Criteria were met and the plant stabilized properly, within acceptable limits, af ter the Rod Control System automati-cally compensated for the temperature mismatch and brought Tavg back to Tref. Table No. 9 compares the actual data obtained with acceptable data limits. l 1 . i e 27
Table 9 Automatic Reactor Control System Response Initial Acceptable Actual Condi- Description Data Limits Data tion Tref - Tavg (Initial Condition) 11.5 deg. F 0.5 deg. F l Tavg Maximum - Initial Pressurizer Pressure 165 psig 15 psig
> Initial'- Minimum Pressurizer Pressure 165 psig 60 psig Tref Peak-to-Peak Amplitude of Tavg Oscillation 15 deg. F 3 5 deg.F Minimum Period of Tavg Oscillation 160 sec. 99.6 sec.
Tref - Tavg (After Transient) A1.5 deg. F 4 .4 deg.F Maximum - Initial Pressurizer Pressure 165 psig 5 7 psig Tavg Initial - Minimum Pressurizer Pressure 165 psig 12 psig Peak-to-Peak Amplitude of Tavg Oscillation 15 deg. F 2.6 deg. F Tref Mininta Period of Tavg Oscillation 160 see 138 see Tref - Tavg ( After Transient) A1.5 deg. F 4 .63 deg.F 28
10.0 Control Systems Checkout 10.2 Test Procedure No. 38.2 - Automatic Steam Generator Level Control TEST OBJECTIVE The objective of this test was to verify proper operation and stability of the Automatic Steam Generator Level Control System and Automatic Feedwater Pump Speed Controller. , l TEST DESCRIPTION f This test was performed at a nominal reactor power of 30%. The programmed level setpoint signal was disconnected from the level controller and a constant test signal of equal magnitude substituted. The test signal was then raised and lowered with the controller in AUTOMATIC while system response .<as recorded.
~
The steam flow signal input to the flow balancing controller was next substitu-ted with a test signal of equal value. This test signal was then increased and decreased 5% with the controller in AUTOMATIC while system response was recorded. The controllers were restored to their operational configuration and integrated system response was checked by manually increasing Steam Generator level 5%, switching the controller to AUTOMATIC and monitoring system response. The entire procedure was completed on one Steam Generator Level Control System before proceed-ing to the next. The Feedwater Pump Speed Controllers were tested by varying the master controller +5% of feedpump operating speed with the master / final control stations in AUTOMATIC. Feedwater pump speed response was monitored and the procedure was repeated for the second feedwater pump. TEST RESULTS The test was started with initial settings suggested by Westinghouse. This resul-ted in unacceptable oscillations in level. The problem was traced to too fast a response of the level control valves. These valves had volume boosters to meet closing time requirements. Changing the control system gains and resets did not resolve the problem. After discussions with Westinghouse, a design change was imple-mented to delete the volume boosters, modify the pneumatic tubing and install proper solenoid valves to meet the required closing times. After the change was made, the test was successfully completed with the settings initially suggested by Westinghouse. The Feedwater Pump Speed Controllers worked satisfactorily without any adjustment to controller settings. l l 29
10.0 Control Systems Checkout 10.3 Test Procedure 41.8 - Dynamic Automatic Steam Dump Control TEST OBJECTIVE , l The objective of this test was to verify proper operation of the Turbine Trip, 1 Load Rejection, and Steam Pressure controllers in the Steam Dump System and to l adjust controller setpoints, if needed, to obtain satisfactory response. 1 TEST DESCRIPTION With the Main Turbine tripped, reactor power at 1%, and steam dump being control-led in the steam pressure mode, the turbine trip controller was tested by raising Tavg to 550 deg. F and then transferring into the Tavg mode. The Steam Dump System and Tavg were monitored for proper response. Reactor power was then increased to 6% at a fast rate while monitoring the Steam Dump System and Tavg for proper response. Testing the load rejection controller required the Main Turbine latched, reactor power at 3% and the Steam Dump System in the steam pressure mode. Two additional special requirements were to have: (1) The sudden load loss interlock actuated to place the load rejection control-1er in the Tavg control circuit and to unblock the Steam Dump Valves, and (2) A simulated Tref signal of 543 deg. F into the load rejection controller to create a temperature mismatch. The Steam Dump System was then transferred to the Tavg mode while monitoring the Steam Dump System and Tavg response. Testing the Steam Header Pressure Controller required the reactor to be at 1% power and the Steam Dump System in the steam pressure mode. With the steam pressure control-1er in automatic mode, reactor power was increased to 5% while monitoring the Steam Dump System and steam pressure response. TEST RESULTS During the conduct of the Turbine Trip Controller Test, a controller module (TM500) in the Tavg control circuit was discovered to require re-calibration. The out-of-tolerance controller module created too much of a temperature mismatch which caused the Reactor Coolant System to cooldown below the expected Tavg value of 548 deg. F. Af ter it was calibrated, this portion of the test was repeated satis-factorily. On the power increase transient from 1% to 6%, the Steam Dump System responded satisfactorily and Tavg stabilized at 550.1 deg. F (within the acceptance criteria of 549.4 deg. F to 554.6 deg. F). The testing of the Load Rejection Controller was performed without any problems. During the transient, the Steam Dusp System responded satisfactorily and Tavg stabil-ized at 547.2 deg. F (within the acceptance criteria of 546.4 deg. F to 551.6 deg. F). l 30 i
i
.
- I I
The Steam Dump System responded satisfactorily during the Steam Pressure Controller Test. The transient steam pressure stabilized at 987 psig (within the acceptance criteria of 986.2 to 1023.8 psig). Since the Steam Dump System responded satisfactorily, there was no need to adjust any of the controller setpoints. d 1
,5
= I
~31
10.0 Control Systems Checkout 10.4 Test Procedure No. 38.6 - Startup Adjustments of Reactor Control System TEST QIMECTHE The objective of this test procedure was to determine the reactor coolant average temperature program required to maintain the design full load Turbine Impulse Chamber pressure. TEST DESCRIPTION Reactor Coolant Tavg, Steam Generator pressure and Turbine Impulse Chamber pres-sure were recorded at 05, 305 and 505 Rated thennal Power (RTP). Each of these parameters was extrapolated to 1005 RTP. A temperature program correction was then computed from the difference between the saturation temperature of the extrapolated Steam Generator pressure and the design full power Tavg. This correction was applied to the design temperature program generated by the Reactor Control System, Steam Dump Control System and plant computer. With Tavg controlled at the new Tref, Turbine Impulse Chamber pressure is compared to the 505. load design value and agreement verified. This entire process is repeated at 755 RTP to obtain a further refinement in the temperature program. Upon reaching 1005 RTP, the temp-erature program is adjusted (if necessary) to obtain the design value of Turbine Impulse Chamber pressure. Throughout this procedure, changes in the temperature program are constrained to design limitations on Tavg and Turbine Inlet pressure. TEST RESULTS Test data were taken at 05, 305 and 505 RTP and the results were plotted and extra-polated to 1005 RTP. A Tref correction of -3.46 deg. F was subtracted from the max design Trof of 576.6 to yield the new projected 1005 RTP Tref of 573.14 deg. F. This value correlates closely with the plotted and extrapolated 1005 Tave of 573.4 deg. F. Tref as a linear function of percent load was used to determine the desired voltage as a linear function of power for recalibration of the Turbine Impulse Controllers TC-505 and TC-505A. The calibration will be done during the outage and the readings referified at 505 power. A calculation of the change in Moderator Temperature Coefficient (MTC) due to the reduced Tref at 1005 RTP resulted in the equivalent of approximately 1 ppm boron. This change in Tref will have no appreciable effect on rod withdrawal limits and will not result in a positive Moderator Temperature Coefficient for the current rod withdrawal limits. l This test will be performed at 755 and 1005 power plateaus also. , i t 32 l - .- . .. _ - .___ _
~ - - . _ _ . . . _ -
11.0 Test Procedure No. 42.2 - RCCA Pseudo Ejection and RCCA Above Bank Position Measurements TEST OBJECTIVE The objective of this test was to determine the power distribution and rod worth associated with an ejected RCCA. TEST DESCRIPTICN The test was performed at the 301 power plateau with the plant stable and Control Bank D at the hot full power rod insertion limit of 177 steps. 4 Af ter verifying these conditions, the movable incore detectors were used to 3 perform a flux map to determine the " pre-ejection" power distribution. l While maintaining constant turbine power and boron concentration, RCCA B-6 was withdrawn from 177 to 200 steps. At 200 steps a partial flux map (i.e., data from 6 of the 58 flux thimble locations) was taken, af ter which the rod was fully withdrawn. With RCCA.B-6 withdrawn, a full core flux map was taken for the post-ejection power distribution. Finally, RCCA B-6 was returned to its initial posi-tion. ] TEST RESULTS j Power distribution results are summarized in Table 10 and core average radial j power distributions are shown in Figure 6. All Acceptance Criteria were met and no significant problems were encountered during the performance of this test. The post-ejection.value of F (i.e., heat flux hot channel factor) was 2.294, 1 l well below the Acceptable Cr teria (FSAR) limit of 7.07. Ejected rod worth was 14.3 pen, well below the Acceptance Criteria lindt of 200 pcm. 33
l Table 10 Power nistribution Results - Pre ar.d Post Pseudo Rod Election POST-EJ ECTED l PRE-EJ ECTED ITEM ' FLUI MAP FLUI HAP 1 Conditions - power 305 305 l
- temperature (RCS) ~553 deg. F ~556 deg. F ,
1060 ppa 1060 ppa
- boron concentration
- burnup 200 aud/st 200 sud/at Date 12-11-84 12-11-84 Rod Configuration Bank D RCCA B-6 0 177 steps i 228 steps F - measured value 1 393 1.490 AH
- location
- M12-IH D04-lU T 2.218 2.294 F - measured value Q
- location
- LO2-QQ P11-QM e 58 in f 58 in 1.447 1,49 F
g
- measwed wh 1.006 1.044 QUADRANT TILT - measured value eAssembly location (i.e., M12) as shown in Figure 1 .
Pin location within assembly (i.e., IH) based on 17x17 attrix ranging from AA to QQ. 34
FI8URE 6 POWER D15falOUT10N5 FOR P5EU00 EJECTION (T.P.42.2) ( NOTE: a) Box entries are relative essembly PRE EJECTION average powers, ASSE mLY POWER b) Ejected location: 8-6 POST-EJECTION ASSEMELY POWER ,
.603 .671 . 807 .730 . 799 . 649 .595 1 .571 .650 .785 721 .826 .725 .624
.972 .856 .515
.521 . 894 1.062 .974 1.035 1.022 1.012 .925
.967 1.032 .962 1.034 .869 .555 2 .556 .940 1.021 .934 1.016 1.130 1.150 1.045 1.072 .944 1.073 .531
.521 1.064 .970 1.177, 1.157 1.180 3 1.100 1.104 1.074 1.085 1.002 1.135 .599 TV 1.091 1.039 1.111 1.130 1.113 1.200 1.140 1.151 1.052 1.293 .976 .902
.89E .963 1.288 1.071 1.174 1.159 1.o?1 1.086 1.126 1.151 1.145 1.141 1.102 1.301 1.067 .965
_ ne i nip 1 746 1.124 1.167 1.080 1.123 1.013 .573 1.000 1.113 1.066 1.181 1.145 1.209 1.154 1.195
.576 1.146 1.152 1.122 1.136 1.067 .596 5 .597 1.031 1.106 1.076 1.012 1.109 1.151 1.150 1.165 1.125 1.121 1.184 1.096 .960 .667
.645 .935 1.072 1.170 1.143 1.136 1.027 1.091 1.021 6 1.117 1.152 1.181 1.146 1.013 .702
.701 .94E 1.072 1.124 1.121 1.086 1.028 1.066 1.039 1.054 1.016 1.181 1.156 1.163 1.033 .813 1.016 1.140 1.142 1.17 0 1.012 1.045 .946
.785 1.174 1.098 7 1.125 1.008 1.004 .952 1.012 1.025 1.151 1.166
.845 1.056 1.095 1.135
.951 1.103 1.142 1.210 1.130 1.034 .7
.721 1.020 1.108 1.192 1.119 1.075 .931 .951
.951 1.056 1.129 1.166 1.158 1.041 .794
.753 1.003 1.101 1.141 1.116 1.033 .934 .928
.948 1.065 1.042 1.1% 1.161 1.165 1.031 .805
.783 1.016 1.138 1.146 1.17 0 1.018 1.036 9 .945 1.014 1.021 1.133 1.143 1.144 1.070 .835
.809 1.03E 1.085 1.126 1.130 1.016 .992 1.087 1.024 1.161 1.153 1.187 1.093 .961 .666
.676 .972 1.066 1.165 1.133 1.133 1.004 gg 1.105 1.008 1.053 1.134 1.104 1.129 1.127 1.075 .961 .681
.700 .988 1.083 1.130 1.141 1.129 1.185 1.077 1.103 1.024 .551 1.06f 1.184 1.060 1.17 8 1.13S 1.169 1.107 1.147
.60! 1.044 1.057 .592 1.124 1.122 1.109 1.109 1.121 1.129 1.134 1.067 11 .647 1.15E 1.262 1.065 1.167 1.117 1.160 1.073 1.294 .972 .89E
.912 .975 1.294 '1.076 1.1E2 1.139 12 1.130 1.106 1.126 1.115 1.117 1.056 1.222 1.023 .942
.93( 1.011 1.279 1.067 1.119 1.081 1.126 1.076 1.139 1.093 1.121 .965 1.071 .522
.516 1.075 .976 13 1.061 1.074 1.075 1.081 1.102 1.097 1.111 .990 1.0E9 .579
.535 1.068 1.022
.945 1.018 1.025 1.052 1.001 1.023 .881 .518
'.516 .888 1.012 i
g4 .567
.537 .888 1.019 .932 1.051 1.019 1.077 .967 1.058 .919
.578 .652 .803 .742 .839 .692 .581 15 .633 .779 .859 .EED .590
.572 .654 ,
} R P N N L K J H G r E D C B A nyt, 9 DIABLO CANYON POWER PLANT - UNIT 1 35
. I 12.0 Test Procedure No. 42.3 - Static Rod Drop and RCCA Below Bank Position Measurements TEST OBJECTIVE The objective of this test was to determine the power distribution associated with a dropped rod configuration.
TEST DESCRIPTION Initial conditions were established as follows: The test was performed at the 50% power plateau with the plant stable and all RCCAs at or nearly at full withdrawn position. Af ter verifying these conditions, the movable incore detectors were used to per-form a full-core flux map to determine the reference, or " pre-drop", power distri-bution. RCCA H-4 was inserted in steps (H-4 was chosen because this RCCA was predicted to " cause the most severe change in power distribution) maintaining reactor power and temperature nearly constant through boron dilution. Rod motion was suspended for RCCA H-4 at positions 190, 150, 100, and 50 steps while partial flux maps (i.e. , data from 6 of 58 incore flux thimble locations) were taken. Rod insertion was completed in approximately 45 minutes, leaving a configuration with RCCA H-4 fully inserted and all other rods withdrawn. A full-core post-drop flux map was immediately performed, after which RCCA H-4 was returned to the top of the core. During the rod withdrawal, reactivity changes were balanced with boron addition. The entire RCCA H-4 insertion / withdrawal process was completed in less than four hours. TEST RESULTS All Acceptance Criteria were met and no significant problems were encountered during the performance of this test. , N The pre- and post-drop nuclear enthalpy rise hot channel factora (Fg) occurred ! in the same assembly (M-12). The value increased from 1.391 to 1.652, below the j Acceptance Criteria of 1.660. The results are summarized in Table 11 and the j core-average radial power distributions are shown in Figure 7. The negative l reactivity associated with the rod created a flux tilt across the core, as shown , by the quadrant tilt map in Table 11. Due to the symmetric location of RCCA H-4 l (see Figure 1), the upper two quadrants had almost identical decreases in rela-tive flux and power levels; the lower quadrants indicated corresponding increases in flux and power.
" Dropped" rod worth was 97 PCM, a result based on reactivity computer data during the rod insertion. However, this value was a rough estimate because strong dop-pler feedback at 50% power led to poor resolution of the reactivity computer data.
I I 36
l l Tabie 11 Power Distribution Results - Pre and Post Rod Droo ITEM PRE-DROP POST-DROP FLUI MAP FLUI MAP Conditions - power 50 5 505
- temperature (RCS) ~562 deg. F ~562 deg. F
- boron concentration 1020 ppm 997 ppm
- burnup 220 MWD /MT 220 MWD /KT Date 12-22-84 12-23-84 Rod Configuration ARO RCCA H-4 0 0 steps N
F - measured value 1 391 1.652 AH
- location
- M12-IH M12-JH T
F - measured value 2.033 2.409 Q
- location
- M12-IH M12-JH e 74 in e 74 in F - measured value 1.359 1.356 g
QUADRANT TILT - measured value 1 .0 06 1 .137
# Assembly location (i.e., M12) as shown in Figure 1 .
Pin location within assembly (i.e., IH) based on 17x17 matrix ranging from AA to QQ. 37
PI2URE 7 POWER D75TRIBUTIONS POR ROD DROP (7.P. 42.3) l PRE-DROP ASSEMBLY POWER NOTE: a) Box entries are relative assemb3y POST-DROP powers. ASSEMBLY POWER b) Drop IOcation: E-4
)
.586 686 .812 741 .812 .683 .561 1
.464 .511 .579 .519 .616 .567 .502
.517 .692 1.020 1.013 1.036 1.026 1.026 .993 .997 .859 .503
.493 .910 .934 .740 .719 .653 .727 .754 .B24 .711 .453
.511 1.056 .962 1.114 1.105 1.154 1.110 1.143 1.063 1.081 .940 1.04: .509 3 .623 .706 .775 .85a ,826 .977 .535
.493 .970 .E97 .908 .850 .725
.879 .960 1.271 1.059 1.167 1.132 1.195 1.136 1.164 1.047 l.269 .952 .8?b 4
.n16 .e?6 1.10F, .873 .842 .669 .408 .674 .838 .879 l.101 .935 .870
.589 1.014 1.108 1.059 1.173 1.138 1.198 1.154 1.198 1.136 1.158 L.061 1.103 1.010 .576 5 .621 1.040 1.068 1.000 .985 .911 .819 .716 .810 .898 .970 ,984 1.032 .990 .55E
.681 1.003 .077 1.156 1.125 1.149 1.044 1.122 1.036 1.135 1.120 i.Itz 1. 0H 1.004 679 6 .671
.765 1.045 1.107 1.110 i.057 .981 .866 .862 .852 .963 1.034 l.098 1.067 .971
.795 1.035 1.142 1.133 1.161 1.031 1.076 .987 1.066 1.016 1.172 . 139 1.155 1.035 .815 7
.934 1.152 1.163 1.171 1.126 .991 .961 .902 .951 .982 1.118 1.156 1.151 1.074 .E?E g .721 1.022 1.099 1.183 1.107 1.103 .979 1.037 .965 1.094 1.127 1.191 1.118 1.034 .746
.846 1.119 1.200 1.219 1.169 1.084 .986 1.013 .983 1.081 1.151 1.204 1.176 1.065 .795 l
.790 1.027 1.142 1.136 1.153 1.015 1.057 .968 1.059 1.023 1.179 1.148 1.165 1.042 .817 c1n 1 1R6 1_712 1_??5 1 21n 1 neF 1_002 1_ndt 1 not 1 non 1. 210 1 212 1 212 1 11F Rc1
.683 1.007 1.071 1.160 1.119 1.127 1.015 1.092 1.023 1.146 1.143 1.184 1.114 1.022 .692
.7P9 1.140 1.247 1.263 1.259 1.229 1.136 1.178 1.141 1.223 1.253 1_2Lo 1.21F i 124 7RS
.594 1.034 1.129 1.054 1.158 1.120 1.160 1.113 1.156 1.126 1.179 1.074 1.112 1.039 .597
.69f 1.232 1.322 1.199 1.290 1.2E5 1.290 1.265 1.277 1.257 1.292 1.212 1.199 1.21C tc:
.860 .954 1.290 1.071 1.179 1.131 1.176 1.126 1.167 1.070 ;.162 .9f' .895 1.027 1.159 1.445 1.235 .1321 i.315 1.318 .298 1.302 1.225 1.404 1.153 1.057
.5c7 1.047 .965 1.117 1.094 L 130 1.088 L137 1.088 1.112 .960 1.060 .522 13 1.286 1.279 1.116 1.214 .64!
.624 1.259 1.171 1.243 1.279 . 304 1.283 1.305
.507 .877 1.023 1.007 1.022 1.017 l.034 1.013 1.024 .887 .518 34 1.203 1.272 1.184 1.226 1.026 .625
.630 1.030 1.184 1.145 1.253
.578 .673 .796 .727 .809 .686 .588 15
) .E87 .501 9F3 .910 1.003 .E20 .695 i n0F.1H R P N M L K J N G F E D C 8 A DIABLO CANYON POWER PLANT - UNIT I 38
l J 1 l 13.0 Test Procedure No. 43.5 - Rod Group Drop and Plant Trip TEST OBJECTIVE The main objective of this test was to demonstrate the ability of the Excore Detector System Negative Rate Circuitry to detect a two rod drop and to review plant response and control system behavior to the resulting plant trip. TEST DESCRIPTION With the plant stable at a nominal power level of 50% and on automatic control, two rods (N-13 and C-3) were simultaneously dropped. The rod motion caused an Excore Detector System negative flux-rate trip which tripped the Reactor and the Turbine. The plant response was monitored during the transient. TEST RESULTS The two rods dropping simultaneously caused a negative rate trip which caused a Reactor trip and a Turbine trip. The acceptance criteria for the test was met as the transient did not cause (i) a safety injection (ii) reactor coolant pump tripping (iii) steam line safety valve lifting or (iv) pressurizer safety valve lifting. The plant responded as expected to the plant trip with the exception of pressuri-zer level and Tavg both of which went below expected values due to the Auxiliary Steam demand. A summary of selected parameter response to the transient is shown in Table 12. 39
Table 12-Rod Groun Dron and Plant TriD TRANSIENT Parameter Units Initial Max. Min. Final 49 - 0 Reactor Power 5 - Electrical Output (Gross) W 4 91 - - 0 Tref deg. F 560 - - 549 Tavg deg. F 559 - - 535 Pressurizer Pressure psig 2250 2231 2047. 2080 Pressurizer Level 5 39 35 15.2 18 Steam Header Pressure psig 831 906 831 856 (Loop 1) Steam Generator Level 5 45 44.5 1.3 35 Core Exit Thermocouple (F5) deg. F 580 _ _ 528 Pressurizer level and Tavs did not meet expected values of }_205 and 547 deg. F respectively. The reason for this was the continued use of auxiliary steam. Pressurizer level dropped with Tavg and went below 205 about 80 sec. after trip. Tavg dropped below 547 deg. F about 17 seca, after trip and settled around 535 deg. F and MSIVs had to be closed to raise Tavg. Feedwater isolation occurred about 8 secs. after trip. . l l l 40
4
\
14.0 Test Procedure No. 41.1 - Plant Shutdown from Outside the Control Room i TEST OBJECTIVE The purpose of this test was to demonstrate that normal Hot Standby conditions can-be established and maintained from outside the Main Control Room. ! TEST DESCRIPTION , Following a reactor trip, essential primary and secondary system conditions such as RCS Temperature and Pressure, RCS Boron Concentration and Steam Generator levels and pressures was controlled from outside the Main Control Room - primarily f rom the Hot Shutdown Panel (HSDP) - as required to establish and maintain stable shutdown conditions. TEST RESULTS The test was performed on January 5,1985, with the unit operating at approxi-mately 50% power. Following a reactor trip from outside the Main Control Room (ref. T.P. 43.5), a minimum Operations Test crew consisting of six members evacu-
! ated the Main Control Room, assumed control of the plant from the HSDP and manned other stations to monitor and control plant parameters in accordance with Operat-ing Procedure OP AP-8 " Control Room Inaccessibility."
The on-shif t Shif t Foreman and his crew remained on watch during this test to monitor the plant and note any problems encountered. Control was maintained from the HSDP for approximately three hours. During this period, additional actions were required from various locations within the plant by the test crew. Once the test crew had established Hot Standby conditions (RCS temperature 1547
- deg. F, Pressurizer pressure ; 2235 psig and Pressurizer level ; 22%), these conditions were maintained for approximately 30 minutes. The test was then terminated by transferring control back to the Control Room.
] Some Control Room Operator actions were performed during this test. These actions and their justifications are listed below: (1) MSlVs were closed to isolate the steam losses on the secondary side thereby, preventing any further decreases in Tavg. The plant staff will update Operating Procedure AP-8 to instruct the operating crew to close the MSIVs prior to leaving the Control Room. (2) The test crew operator could not initiate letdown by manually actuat-ing a relay because the instrument rack for the relay was locked. The plant staff will update AP-8 to instruct the operating' crew to take the keys with them prior to leaving the Control Room. (3) Containment High Pressure Alarm actuated. Control Room Operators placed the Containment Fan Cooler (CFCU) in high speed. Due to a thermal over-load problem, the CFCUs will not automatically restart in high speed after an. auto transfer. A Design Change has been initiated to resolve i this problem. 41 l
(4) A safety valve lif ted on Main Steam Lead 1-2. The Control Room Opera-tors lowered the pot settings on the 10% steam dtamps to reseat the safety valve. Manipulation of the 105 steam dumps from the Control Room was perfomed so that Operations could detemine the correct pot setting to prevent the safety fra lif ting. Operations will update AP-8 to instruct the operating crew to set the correct pot settings on the 105 Steam Dump Controllers prior to leaving the Control Room. Because the Control Rom actions could be resolved administrative 1y and all acceptance criteria were met, this test demonstrates satisfactorily the capability to remotely maintain the plant in Hot Standby conditions. 4 I 1 1 i 42 1 l l l l L__---_________ _ ._ . _ _ . ___. ._ , .
l i 1
- 15.0 Test Procedure No. 43.1 - Load Swing Tests l
i TEST OBJECTIVE The objective of this test was to verify nuclear plant dynamic response, including automatic control system performance, to 10% step load changes introduced at the Turbine Generator at each of the power test plateaus. TEST DESCRIPTION The plant was stabilized and on automatic control at the specified power. The plant output was reduced by 10% at the maximum rate (2200 MWe/ min.). During the load decrease and until plant conditions stabilize, the following parameters were recorded on Chart Recorders: Nuclear power Controlling Bank position Reactor Coolant temperatures Steam flow and pressure Feedwater flow Steam Generator level Pressurizer pressure and level The plant output was then increased by 10% at the maximum rate (2200 MWe/ min.) to the previous power, again recording the above parameters. The plant parameters are evaluated for acceptable dynamic response. TEST RESULTS The test was performed at the 30% power plateau on December 19, 1984 and at the 50% power plateau on December 31, 1984. On both occasions, all Acceptance Criteria were met satisfactorily as follows: Reactor and Turbine did not trip Safety Injection did not initiate Main Steam Safety Valves did not lift Pressurizer Relief Valves or Safety Valves did not lift Manual intervention was not necessary to bring the plant to steady state conditions Nuclear power under and overshoot was less than 3% For the load swing from 40 to 50% power it was necessary to run both feedwater pumps in order to meet the "no manual intervention" criteria. In each run, Steam Generator level variation was within the expected value of +10%, however the expected Pressurizer pressure swings of less than 50 psi was not met for the foll-owing load swings: 30% - 20% (56.4 psi) 20% - 30% (54.0 psi) 50% - 40% (58.5 psi) 43-
The expected steam pressure overshoot or undershoot of +25 psi from the final value was not met for the following load swings: 30% - 20% _(32.5 psi: overshoot) 50% - 40% (32.5 psi: overshoot) 40% - 50% (36.2 psi: overshoot) These deviations from expected values were evaluated by Westinghouse both at 30% and 50% plateaus. Westinghouse identified Control System setting changes for fine tuning the systems to be implemented prior to load swing tests at 75% power. The test will also be performed at 75% and 100% (only -10% swing) power levels. s t 44
, . , . .- - ,- -,,,e,
4 16.0 Test Procedure 42.1 - Doppler Power Reactivity Coefficient Heasurement GLTECTIVE The objective of this test was to verify nuclear design predictions of the Doppler only power coefficient. TEST DESCRIPTION At each test plateau, af ter establishing stable plant conditions with equilibritan xenon, and axial flux difference at or near its target value, .the turbine load was decreased approximately 22 We (25) at a rate of 2200 We/ min (2005 per min). This ! action caused about 25 drop in reactor power level. Subsequent 44 We load swings were performed in order to vary reactor power by about 45 in each case ana a final load swing of 22 We returned power to its initial value. This sequence is abovn in Figure 8. The period between each load swing was long enough to allow stabilization of Tavg and AT. Boron concer.tration and control rod position were maintained constant throughout the test. For each increase in load,.4T increases and Tavg decreases. The increased AT osuses a negative reactivity effect due to the fuel's doppler coefficient, which is offset by the positive reactivity due to isothermal temperature coefficient (ITC) effect on decreased Tavg. In a similar manner, load decreases involve a decrease in AT and an associated increase in Tavg. The load swings done in this test directly measured the change in oore average cool-ant tape.rature required to offset a change in AT. By relating the changes in AT to changes in reactor power level, ratios of doppler coefficient to ITC vere calculated by dividing the change in Tavg by the change in power for each load swing. The acceptance criterion was that this ratio aust be within o.5 deg. F/5 of design value. Doppler coefficient was then inferred by multiplying the average ratio by an ITC based on design calculations. l TEST RESULTS ' At the 305 and 505 power test plateaus, the ration of doppler ooerficient to ITC were within acceptance criteria. The inferred doppler coefficients wore -11.84 pom/5 power at 305 power and -10.94 pen /5 power at 505 power. This trend of decreasing doppler coefficient with power was expected because higher fuel temper-atures cause a reduced degree of broadening of U-238 resonance absorption peaks. l I This test will be repeated at the 755 and 905 test plateaus. 45
)
l Table l'3 Doppler Coefficient Measurements DESIGN RATIO
- INFERRED DOPPLER DESIGN DOPPLER TEST PLATEAU. MEASURED RATIO *
(% RTP) (deg.F/% power) (deg.F/% power) COEFFICIENT COEFFICIENT (NOMINAL) (pca/% power) (pcm/% power) 30 2.83 3.22 -11.84 -13.5 50 2.03 2.40 -10.94 -12.8 i 8 RATIO = DOPPLER COEFFICIENT ISOTHERMAL TEMPERATURE COEFFICIENT Acceptance criterion: measured ratio = design ratio +0.5 deg.F/% power i 9 J 46 4 m -. . , - - , - . -,. . ~ - - . .,- -- ., - -- ,- y, , ,,3 --
---- -m-, - - - - , < ~ . A -~
m e a O O e FIGURE a T.P.42.1 LOAD CYCLING PATTERN FOR POWER COEFFICIENT VERIFICATION c s c M7 oi+2 Ob 3 4 e s" oi +#, oo ,o 4 g g C f r i 1 1
.j oi-2 o, i
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- 17.0 Surveillance Test Procedure No. R Incore-Excore Detector Calibration TEST OBJECTIVE The objective of this test was to determine the relationship between the axial power distribution in the core (as established by use of movable incore flux detectors) and power range excore upper / lower detector signals. The scaling factors that were calculated were used to calibrate the excore nuclear instru-ments. The test was performed initially at the 50% power plateau and will be repeated at 75% power.
t TEST DESCRIPTION Each of the four (4) power range nuclear instrumentation channels consists of a pair of uncompensated ion chambers stacked vertically. Each detector in a given channel is located symmetrically above and below the core axial midplane. l The calibration test was performed to provide the data necessary to calibrate A the pairs of detectors and provide upper / lower signals that are proportional to the power. split between the upper / lower halves of the core over a wide range l of axial power distributions. ! To provide such data, the control rod position and soluble boron content of the I RCS were varied initially in such a way to provide power distributions axially j- skewed toward the bottom of the core. Flux maps (initially full core, quarter-l core thereaf ter) were recorded using the movable incore detector system to deter-
- mine the amount of asymmetry in the axial power distribution (expressed as axial i flux difference, AFD*, or axial offset. AO**). The excore detector signals also
- were measured during the flux maps so that a direct comparison could be made of incore detector vs. excore detector AO. Control rods then were returned to their initial position and the asymmetric buildup of xenon in the core was allowed to j produce a xenon-induced axial power oscillation, shif ting power toward the top of the core.. Periodically, flux maps and excore signals were recorded. At a prescribed point in the xenon / power oscillation (end of the test), a control rod maneuver was performed to dampen out the axial oscillation and return core conditions to normal.
3 i The data from the test were used to plot incore A0 vs. excore A0 for each power l range excore channel and also incore A0 vs.' normalized (full power) detector ; j currents for each power range excore upper and lower detector. These plots provided l i best-fit straight lines from which excore detector gains (slopes) and offsets l (intercepts) were obtained. A subsequent I&C surveillance test procedure STP I-2D used these gains and offsets to calibrate the power range excore nuclear instruments. I TEST RESULTS '(50% Power Plateau) l I Control Bank D rods were inserted in two (2) increments of approximately twenty (20) steps each. Incore A0 shifted from about -3% initially (Bank D at 208
' steps to about -9% (188 steps) and then to about -17% (168 steps). Upon hold-ing the latter configuration for approximately two hours, rods were returned to their original position. During the subsequent axial xenon oscillation, five more quarter-core flux maps were produced at incore AO's ranging from about -7.8% to +2%.
Thus the total span of minimum to maximum A0 was from about -17% at about +2% i 1 48
. : l The entire test required approximately 27 hours to complete. )
Slopes (gains) of the incore vs. excore A0 plots were approximately 1.6 for all channels. Offsets ranged from about -0 3% for channel N44 to +6.4% for channel N 42. The upper detector in channel N42 had a somewhat lower current (sensitivity) than the other detectors, accounting for the slightly different calibration oon-stants for this channel. A sample of each type of plot is enclosed for channel N41 (Figures 9 and 10). Completion of the I&C calibrations por STP I-2D took approximately four (4) days. Rather than perfoming only the applicable gain adjustments a full calibration was performed on all power range channels. This test will be performed again at the 755 power level, eAFD (5) = AO (5) x 5 core power / 100 HA0 (5) = (Upper detector current or oore power - Lower detector current or core power) x 100/ (Upper + Lower) l 1 1 1 49
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18.0 Test-Procedure No. 43.7 - Net Load Trip From 50% Power TEST OBJECTIVE l l The objective of this test was to demonstrate the ability of the unit tc sustain l a net load rejection f rom nominal 50% power. TEST DESCRIPTION The plant was stable at nominal 50% RTP and on automatic control. The load rejection was initiated by opening the main transformer high side breakers. The plant was then stabilized using Nuclear Plant Operations (NPO) Emergency Operating Procedure OP-AP-2
- Full Load Rejection. During the transient, various plant parameters were monitored for analysis of the plant response to the transient.
TEST RESULTS The first attempt of the test resulted in a reactor trip caused by low RCS flow. This trip was caused by a solenoid valve in the Turbine Control System that stuck in the open position not allowing the Turbine Governor Valves to reopen. This resulted in a drop in turbine speed, generator frequency and hence in RCS flow. The solenoid valve was replaced and tested successfully before the test was repeated. The second attempt was successful with the unit sustaining the transient. The rod control was taken on manual at about 20% so that the Steam Generator levels could be controlled by the Main Feedwater Regulating Valves. (The bypass valves do not have automatic controls.) Response of selected parameters is tabulated in Table 14. The acceptance criteria were met and the plant response was reviewed by engineering and Westinghouse. No additional setpoint changes were recommended. (Changes were recommended by Westinghouse af ter reviewing Load Swing Tests at 30% and 50% power and the Rod Group Drop and Plant Trip Test at 50% power. These changes will be incorpor ated prior to performing the Load Swing Test at 75% power). i 52
i l f Table 14 Plant Resoonse to Net Load Trin From 501 Power INITIAL DURING TRANSINT FINAL Unita Minimum Maximum , Parameter 47.8 23 50;s 27 Reactor Power - 5 I Electrical Output MWe 47 8 - - 47 Trer deg. F 56 1 - - 550 Tayg deg. F 559 552 562.5 5 56 deg. F 30 14 30 14 ZL T 2250 2184 2268 2250 Pressurizer Pressure psig s Pressurizer Level 5 38 33.5 40 30 Steam Header Pressure paig 839 2 837.5 931.2 932 S. G. Lev el 5 43 27 .5 53 31 FW Heater 1-1 A Outlet Temperature deg. F 363.6 233.6 363.6 237 Control Bank D Position steps 1 82 - - 120 l f f 4 1 1
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s 53 I 2-_-__-_-___ ._
% l 19.0 Surveillance Test Procedure R RCS Primary Coolant Flow Measurement l 1
TEST OBJECTIVE l The objective' of this test was to verify the calibration of RCS flow instruments
< at 30% and 50% of full power and to confirm that the total flow of all four loops is greater than the Technical Specification requirement of 363,000 gpm.
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1 i TEST DESCRIPTION 4 Prior to obtaining data, the plant load was stabilized at a constant value and plant parameters were checked or adjusted to be within normal operating limits.
' Data was then obtained during a nominal 30 minute period with plant conditions stabilized and plant load constant.
Reactor coolant flow was determined by performing a heat balance on the RCS. This was done by using the gross steam generator thermal output calculated in the high accuracy heat balance test (STP R-2A) and narrow range hot-leg and cold-leg temperature measurements. The heat balance across the secondary side of the steam generators (STP R-2A) produced an accurate determination of primary system heat rate. The heat race results were then refined by compensating for RCS peripheral and convective heat loads to determine actual core heat generation. Actual RCS flow was then calculated. TEST RESULTS 30% Full Power Test Total RCS flow was measured as 374,500 gpm which is approximately 3% more than required by Technical Specifications. No recalibrations of loop flow meters were required; however loop flow constants (in STP I-1A Surveillance Test Procedure done each shift by the operators) were updated. 50% Full'Pcwer Test Total RCS flow was measured as 376,100 gpm which is 3 1/2% more than required by Technical Specifications. One loop flow meter indicator required recalibration. Since the flow change was in the conservative direction, no changes'were required for the Surveillance Test Procedure flow constants mentioned above. s This test will be repeated at 75, 90 and 100% power levels. 1 0 54 P
N PAC I F'IC OAS AND ELECTR,IC C O M PANY bbWb { 77 BEALE STREET . SAN FRANCISCO, CALIFORNIA 94106 * (415) 781-4211
- TWX 910 372 6587 JAuss o. SHIPPER l oaE7o"221o o., April 29,1985 PGandE Letter No.: DCL-85-174 l Mr. George W. ' Knighton, Chief Licensing Branch No. 3 Division of Licensing
.0ffice of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington , D.C. 20555 Re: Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Supplement 1 to Startup Report
Dear Mr. Knighton:
As required by the Operating License for Unit 1 (Section 6.9 of the Technical Specifications), Supplement 1 to the Startup Report for the period from November 1,1984, to January 31, 1985, is transmitted herewith. Kindly acknowledge receipt of this material on the enclosed copy of this letter and return it in the enclosed addressed envelope. Sincerely, J. . iffer Enclosure
-cc: J. B. Martin H. E. Schierling Service List 4
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i PGandE Letter No.: DCL-85-174 ENCLOSURE 1 ( 0306S/0032K l}}