ML20071F713
| ML20071F713 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/31/1978 |
| From: | Passman N POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML100271751 | List: |
| References | |
| NUDOCS 7811070256 | |
| Download: ML20071F713 (144) | |
Text
{{#Wiki_filter:' l'UdEl' AUTil01(ITY OF Till: STATE OP IIDI YOIO II;DIAll POII;T 3, !JUCLEAlt POWER PLA!JT CYCLE 2, STARTUP PIIYSICS TEST REPORT by N. Passman i i I I i 7811C7ot5~A i t l' f
l 1 TABL1; Ol' CO!D'1;llTS l
. i Page flo. l l
1.0 I!!TRODUCTIO!! l-1 1.1 Plant Descriptaon 1-1 l t i 1.2 Test Objectives 1-1 1.3 Relevant Design Information 1-2 1.4 Sequence of Startup Events 1-2 1.5 Summary of Measured and Predicted Core Parameters 1-2 2.0 2-1 MEASUREMENT TECl!!:IQUES 2.1 General 2-1 2.2 Reactivity Measurements 2-1 2.2.1 Suberitical Measurements 2-1 2.2.2 Critical !!easurements 2-2 Power Distribution Measurements 2-3 2.3 Instrumentation Calibration Deta Collection 2-4 2.4 2-4 2.5 Thermal Power Measurements 3-1 3.0 TEST RESULTS - PIACTIVTTY MEASUREMENTS 3.1 Core Loading 3-1 Initial Criticality 3-1 3.2 Low Power Physics Tests 3-1 3.3 3.3.1 Preliminary Measurements 3-1 Doron Endpoints 3-2 3.3.2 Temperature Coefficient 3-2 3.3.3 RCC Bank Worths 3-3 3.3.4 3-3 3.4 At Power Tests Power coefficient Measurement 3-3 3.4.1 3.5 novable Detector I' lux Maps 3-4 4-1 4.0 IN::TntirinNT CALI11 RATIO!:S 4-1 4.1 Incore - Excore Calibr.ition Incore Titerniocouple and Wide li.mge IG'c Calibrat ion 4-1 4.2 i
APPEliDIX A Measurement of tlie Power Coefficient n_1 h'.'P E11D I X 13 Corc Loading Scquence B-1 APPE!! DIX C Generation of Withdrawal C-1 Limits to Insure a IJegative Moderator Coefficient e ii
LIST OF PIGuitl:0 FIGUlt1:0 TITII PACI; NO. 1.1 Core Layout 1-5 1.2 Control Rod Locations 1-6 3.1 ICRR During Core Loading SR-32 3-6 3.2 ICRR During Core Loading SR-31 3-7 i 3.3 ICRR vs Bank Position 3-8 3.4 ICRR vs Gal. Dilution Water Added 3-9 3.5 Control Bank D Differential and Integral Worth 3-10 3.6 Control Bank C Differential and Integral Worth 3-11 3.7 Centrol Bank B nifferential and Integral Worth 3-12 3.8 Control Bank A Differential and Integral Worth 3-13 3.9 Measured and Expected FD1111 - IIZP 3-14 3.10 Measured and Percent Diff. of FDH!! - HZP 3-15 3.11 Measured and Expected FDl!N - IIFP 3-16 3.12 Measured and Percent Diff. of FDH11 - IIFP 3-17 A-1 Data Analysis Flow Diagram A-5 C-1 Moderator Temperature Coefficient C-4 for Successive Rod Insertion C-2 Moderator Temperature Coefficient for Banks C-5 in Overlap C-3 Moderator Coefficient vs Boron Concentration C-6 at IIZP, 547 F C-4 Moderator Coefficient vs Boron Concentration C-7 at 251. Power, 553.7 r C-5 Moderator Coefficient vs Boron Concentration C-7 at 50% power, 560.4 F C-6 Boron Cocentration vs Rod Position for Various C-8 Power Levels to Insure a Negative MTC iii
4 LIST OF Tl W L1:S Table Title Page No. 1.1 Core Design Information Parameters 3-3 1.2 Summary of Measured and Predicted Parameters 1-4 3.1 Reactivity Computer Checkout Results 3-2 3.2 Plux Map Summary, Hot Zero Power - All Rods Out 3-4 3.3 riux Map Summary, llot Full Power - All Rods Out 3-5 4.1 Incore - Excore Calibration 4-2 e. 4.2 Incore Thermocouple Isothermal Correction Factors 4-3 A-1 Summary of Temperature Data from Power A-2 Coefficient Measurement A-2 Results of Data Analysis A-3 l l l L iv l' (
1.0 INTL;ODUCTION 1.1 P1Jd.'T DESCRIPTION Tht: Indian Point Unit 3 Nuclear Plant is a four loop closed cycle pressuri:cd light water moderated and cooled nuclear react 6r operated by the Power Authority of the State of New York. The reactor core is designed to produce 3025 megawatts thermal power resulting in a gross electrical generating capacity of 1000 megawatts of electrical energy. The Nuclear Steam Supply System was designed by Ucstinghouse Electric Corporation. Plant construction and design were performed by United Engineers. The plant is located on the cast side of the !!udson River, 30 miles north of New York City. 1.2 TEST OBJECTIVES This report documents the results of nuclear tests perforrned as part of the cycle 2 startop testing program: The objectives of the nuclear tests were: (1) to verify that the operating characteristics of the core are consistent with design predictions, (2) to demonstrate that measured core parameters are consistent with values used in t.hc Safety Analyses, (3) to demonstrate that the core can be operated at licensed thermal power safely and within the limits of the Technical Specifi cations,_ and (4) t o provida dat a for nuclear and temperature l in:it rumentat ion calibration. 1-1
1.3 RELINANT DESIGli INFORMATIOtt Tabic 1.1 presents selected design information of the Indian Point
;;uclear Plant, rigure 1.1 shows the core layout with control rods, sources, and enrichment Jocations. Figure 1.2 shows the core layout with individual control and shutdown bank locations.
1.4 SEQUE!;CE OF STARTUP EVE!!TS o Following core loading, June 30 - July 14, 1978, a series of pre-operational tests was performed both in the cold shutdown and hot shutdown conditions. Criticality was achieved on August 17, 1978 followed by a program of low power nuclear tests. The unit synchronized to the grid on August 25, 1978. Power escalation to maximum licensed power (100%) was accomplished by stopping at approx-imately lot power level increments, for purposes of data collection and testing. Maximum power (1001) was achieved on September 10, 1978. 1.5 SU?tilsRY OF MEASUPID A!!D PREDICTED CORE PARAMETERS Presented in Table 1.2 is a summary of selected results of zero power nuclear tests and at-power distribution measurements. The results reported in this document, indicate that all acceptance criteria specified in the cycle 2 reload submitta] were met. 1-2 I
TABLE 1.1 Core Design Information Parameters
!; umber of Fuel Assemblics 193 Region 1.._. 1 kugion 2 and 3 64 Region 4 64 Lattice Configuration 15X15 1; umber of fuel rods per assembly 204 Fuel loading, MTU 88.06 tiumber of Assemblics Containing RCC Full Length 53
!Jumber of Absorber kods per RCC Assembly 20 t; umber of Control Rod Assembly Guide Thimble per Assembly 20 1: umber of Instrumentation Thimbles per Assembly 1.
!Jumber of Grids in Active Core IIeight 7 Ileat Output, !@lth 3025 Percent licat Generated in ruel 97.4 Ilot Zero Power Coolant Temperature,0F -547.0 Operating Pressure, psia 2250
!!ot Channel Factors
!! cat Plux Pg T 1; 2.17 IJuclear Enthalpy Rise, FLil 1. '. 5 Average Clad Surface IIcat Flux, btu /hr-ft 2 ,
193,000 Average Linear Power Density, kw/ft Puel 6.24 Specific Power, kw/kg Uranium 33.2 Power Density, kw/ liter of Core 91.8 Enrichments, w/o Uranium 235 Region 1 2.28 Region 2 2.80 Region 3 3.29 Region 4 3.10 l i I e i i i 1-3 I f
TABL1; 1.2 Summary of ficasured and Predicted Parameters
- 1. fu:act.ivity Measurements 1.1 IIZP Critical Boron Conc. (ppm) l (Acceptance Criteria: P 50 ppm)
Rod Position Measured Predicted M-P ARO 1446 1443 +3 D IN 1362 13G4 -2 D+C IN 1255 1256 -1 1.2 Isothermal 'Iemperature coefficient (pcm/ r) (Acceptance Criteria: P 3 pcm/OF Inferred MTC* (pcm/OF) ARO -0.50 +0.4 -0.90 +1.10 D IN -2.14 -1.0 -1.14 -0.54 D+C IN -4.07 -3.14 -0.93 -2.47 1.3 RCC/RCCA Worth Measurements (pcm) (Accepance Criteria: P 15 %) D 721 750 -29 C(D IN) 1031 968 +63 B(D+C IN) 552 - 571 -19 A(D+C+B IN) 866 823 +43 Total 3170 3112 +58 1.4 Differential !!casured Power Defect -9.30 pcm/t Predicted Power Defect -8.67 pcm/t
- 2. Power Distribution Hot Zero Power (ARO) Technical Specificatien licasured Limit Value T*
F 1.85 1.64 g111 T r 4.09 3.06 Q Ilot Full Power (ARO) I' l.55 1.50 21 11 T r 2.08 1.83 0
- Assumes -1.6 pcm/ I' for doppler temperature coef ficient -
1-4 isi .is i.ii si i ,
i l'19 ure 1.1 90* . 15 14 13 12 11 10 9 8 7 6 5 4 1 3 2 1 POS P27 P30 P46 P4U P03 P59 230 236 195 207 241 246 209 P12 P39 P41 'B56 P35 B38 P37 U37 P26 P29 Pol 220 223 228 R64 12 R91 194 203 i<GG 243 253 .j P45 P24 CG1 bo4 h33 B50 bl3 B12 B15 B16 CO3 'P14 P53 235 1:82 191' k80 J41 R77 12.I.' R75 152 ECS 182 R108 249 PSI C31 E3G C35 C34 b54 C26 h28 C49 C56 D24 Col P43 221 13 R97 20 145 186 R73 254 164 ~18 RG9 170 237 ' M l PG4 P42 n51 C59 l B35 'C19 C53 l33 COG C47 bG2 C51 'B48 P38 P50 212 225 P79 177 l 10 155 14G I M. P67 160 9 17G IOG 229 226 P13 1:41 b2J C64 C16 C1J ' C.:V cgs C41 ' C27 C36 C37 n10 bv.s Pn 217 R39 C 169 165 LG1 156 R74 147 F84 158 171 134 - V, R9G 215 P11 P28 BJO B29 C44 C52 'CO2 bil C18 CGO C54 L46 E34 P49 'P33 222 208 Elli 172 R87 154 1G 193 179 139 150 174 .'.110 d 106 185 O. 211 R100 C24 n2O CG2 n17 A24 307 C32 n r,2 C23 m nG1 p54 "If R109 153 R95 135 P94 '192 RG5, 11 R71 33 N PG3 POG BGO LG8 C58
] R92 219 C3G ' C4 2 253 C21 C12 CLO' B4/ B43 PGe Pu/
216 234 R105 173 133 136 15 1G7 14 144 R98 149 RG3 202 251 -G P19 B25 .nol C43 C57 C28 C11 C25 C4 0 C29 C14 CG3 B31 L19 P10 206 P86 118 187 1G2 R78 143 207 148 R90 140 188 -I 137 RG8 223 P20 P31 B21 C48 L2C C4G CG4 542 'C45 COL b63 Cib u4> P.n Pib 200 248 R62 178 7 IGG E81 132 157 161 8 175 163 E 247 204 P47 C10 E2 'i C33 C38 B55 C22 B39 C09 CJ7 b59 C07 P3G 245 190 h101 19 168 183 E70 184 1G3 17 R104 189 244 . P50 P34 C39 B40 h02 B64 B57 h0G L13 n22 CSL P16 P23 199 R72 181 P113 151 R112 12 P3:- PlO2 15'; E83 100 R99 230 C P09 Poa ' P17 HOL P44 E32 P57 DJ4 Plt PO4 PO2 227 231 213 R7G 240 P.80 252 R103 197 242 250 0 P23 pGl P ',8 PhD P22 P21 P4( ~ 239 224 205 232 19c 201 216 A Dirichtnen ts : 270'
- Insert:s : i AXX - 2.28 w/o XXX
. . t.
liXX - 2.00 w/o - Plu99 1n9 Device HXX - Control Rod CXX - 3.29 w/o SP!;XX PXX - 3.10 w/o - Secondary Source 12PXX. - Hurnal$le Poi:;on
*" I
2.0 MMASUluf!ENT TECIINIQUES 2.1 GE!!EkAL The methods for nuclear test data acquisition can be grouped into four distinct areas: (1) reactivity measurements, (2) measurements of core power distribution (3) collection of instrumentation calibration data, and (4) thermal pow r measurements. The purpose of this section is to describe the methods used in each of these areas. 2.2 REACTIVITY MEASURD4ENTS Measurements of core reactivity were performed both in suberitical and critical core conditions. In the subcritical mode, measurements were made during initial ccre loading and the approach to criticality. In the critical mode, measurements were made to determine core kinetics parameters. k 2.2.1 Suberitical Measurements During core loading, the core reactivity was monitored using the response of the two plant source range channels,. Monitoring was accomplished by observing the normalized inverse count rate ratio (ICRR) for each channel as the core was loaded. During the approach to criticality, ICRR plots usinq data from the two plant source range channels were used to predict c::pected criticality. ICRR data woro plotted an a f unction of rod punition durinq rod wi t hdrawal, and a:. a f unction of sacanut ed bot on concentration during dilut ion.
.'- ]
. Pigure 1.2
, . 15 14 13 12 1. I 10 9 8 7 6 5 4 3 2 1 i)
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o00 t 4 P G El G n G) v @ }@ v G Il i S A ' d\f fSA ll W F1 l\y
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- 180 a
@ l IE.] (N I r G $a@ ! Gl E
\'/ bl 7 3
!(y D A
/h\f lS.lt M c (D y () k(N N/ l(T) .
8 l@ IM e 4 1 2700 DA!!K SYMBOL NUMBER OF ROD CLUSTERS SA 8
! SB S
Q 8 4 c SD Q 4 8 l A 9 B 6 4 o. 1 . c a
- D Q _9 .i
~~
Control God Location: . l I 1-6
2.2.2 Critical Mea:,urements Small core reactivity changes were determined with the aid of an analog (reactivity) computer which provided an on-line solution to the point kinctics equations. Reactivity recor'ds were naintained on a continuous basis during a tout via a strip chart recorder which logged the output from the reactivity computcr. The input signal to the reactivity computer was provided by the Nuclear Instrumentation System (NIS) power range channels. During zero power measurements, one NIS power range channel was taken out of plant service and used as input to the reactivity computer. Signals from the top and bottom section were summed, and the common signal was converted to voltage by a picoammeter for input to the reactivity computer. At power, the voltage output cf the isolation amplifiers of all four NIS power range channels were summed and fed directly to the reactivity computer. , Differential and integral worth of individual rod control cluster assemblics (RCCA's) or rod control clusters (RCC) were obtained from the reactivity computer response to stepwise movement of the control element. During the measuremer.t, the reactor was maintained, on the average, in the critical state. Isothermal temperature coefficient data were obtained by measuring the reactivity computer response to small temperature changes, a few degrees below no load temperature. Junt critical boron concentration data were obtained from plant chemistry boron analyses of reactor coolant system samples (l'CS ) under equilibrium condi tionn. l'or boron concentration endpointn, corrections to the meacured concentration utilir.ed reactivity computer meanurements of the reactivity difference between .ictual and slenign core configurations. 2-2
a Power coefficient data were obtained by measuring the change in core reactivity resultant from a change in core power level. Under conditions of constant RCS boron concentration, the core power 1cvel was decreased and increased incrementally approximately 101 by controlling RCC bank motion. Changes in core reactivity resultant from RCC bank motion were measured using the reactivity computer. Changes in core power level were determined from secondary side calorimetric measurements made prior to and follcwing the power level decreases and increases. This method of measuring the power coefficient yicided poor results due to a ' very small signal to noise ratio on the signal coming from the power range instrumentation. The measurement method was revised to clininate the use of the reactivity computer by using the change in Tavg as an indirect measurement of the power coefficient, Appendix A provides details of this measurement. 2.3 POWER DISTRIBUTIONS The Moveable Detector (M/D) Flux Mappping System was used to collect power distribution data. The power distribution measurements were performed throughout the startup program with standard control bank positions to verify correct fuel loading patterns and design calculations to cross-calibrate the incore and thermocouple system; and to provide calibration data for the correlation of excore detcetor response. Data from the M/D system was input to the IUCORE computer code to generate detailed three dimensional core power profiles. The INCORE code combines measured flux distributions with design calculated powcr flux distributjons, to ycid specific fuel rod powers, local burnup, core power til ts , core average axial offset, etc. 2-3
},
2.4 11:CThUMUlTTATIO!! CAbil:ISTIO!! DATA COLLECTIO!3 At each stable power level (statepoint) during the power escalation program (approxinintely each 10'.) measurements were made of RCS loop temperatures (Tavg and 3 T), Steam Generator pressure and HIS power range detector currents. 11IS currents were read from the power range current meters. Temperature and pressure da- were obtained from the meters on the control board and from the P-250 process computer. Correlations between incore axial power distribution and excore power range detector response were made through simultaneous measurement of core power IcVel, excore actector currents, and core power distributior.s (flux Maps). Calibration data for the incore thermocouple system were obtained during isothemal measurements prior to criticality and during simultaneous flux and thermocouple (T/C) maps taken at power levels of 100t. 2.5 THEP2AL POWER MEASUREME!ITS Core thornal power was determined by performing a heat balance across cach of the steam generators. This measurement required the accurato determination of steam generator pressure, feedwater inlet temperature, and fecdwater flow. For each steam generator steam pressure was taken as the average of the three channel P-250 readings; feedwater temperature was taken from the resistance temperature detector (RTD), located in the feedwater header utili::ing the PRODAC nystem; and feedwater f]cw was determined from Harton prencure gaugcc inntalled on a venturj tap of each leedwater lobe. l 2-4
- 3. 0 TL'ST jd;OULTS - ;!; ACTIVITY 111;A!;UPJ:M1;llTS 3.1 COII LOALItiG Core loading was accomplished by adding fuel assemblics to the vessel following the prescribed sequence shown in api >cndix B, ICRR data obtained from the IIIS source range channels are presented in rigures 3.1 and 3.2. There were reo unexpected changes in core reactivity during the loading of the fuel assemblics.
3.2 IIIITIAL CRITICALITY The approach to criticality began on Augist 17, 1978 at 23:15 hrs. with the incremental withdrawal of shutdown and control banks. Primary System boron concentration during rod withdrawal was approximately 1760 ppm. Inverse count rate ratio data from the two cource range channels during rod withdrawal are shown in rigure 3.3. Criticality was achieved with the addition of reactor makeup water at a rate of 60 gpm. Inverse count ratios during boron dilution are shown in rigure 3.4. Throughout the critical approach, count rates from the two source range channels were both adequate and consistent for good monitoring of core reactivity. Count rates at the beginning of rod withdrawal were 82.6 and 55.9 cps for 1:-31 and !!-32 respectively. 3.3 LOW POWER PilYSICS TrSTS 3.3.1 Preliminary Measurements lmmediately followin<j criticality, the uj>per limit of flux leve] for zero pwcr testing was established an about one decade below which detectabic nuclear heat wan added to the coolant. Ao increasing Tavy was observed at a flux level 3-1
, -G of 1.3 x 10- urups on power rango 11-41 (connected to the reactivity computer)
-6 and about 6 x 30 ampa on both intermediate range channelu. All reactivity
-7 measurements were performed below 3 x 10 amps on channel fi-41. This tenting
-10 was also about three decades above the gamma background IcVel of 4 x 10 anps on 11-41.
i Next a check of the reactivity computer performance was made by measuring three values of reactivity and comparing the value with that inferred from the resultant reactor period from parameters given in the core design report. The results of this test, given in Table 3.1, indicate proper operation of the reactivity computer. TABLE 3.1 PIACTIVITY COMPUTER CIIECT,0UT PISULTS Doubling Time Period Inferred Measured Difference Reactivity Reactivity (Meas. - Inferred) (sec~l ) (sec-1) (pcm) ' (pcm) (pcm) 143.8 207 28.8 29.0 +0.2 3.5 94.5 55.2 55.0 -0.2 39.5 57.0 80.3 80.5 +0.2 3.3.2 Boron Endpoints The just critical boron concentration was measured for three rod configurations. The test results were sum:narized in Table 1.2 along with design predications. ; l 3.3.3 Isothernal temperature coefficient measurements were performed at three core conditions au summarized in Table 1.2. An seen in Table 1.2 the all-rodu-out moderator-cidy temperature coef ficient (MTC) J: ponitive i 3-2
and therefore rod withdrawal limits had to be imponed to innure a negative HTC au required by technical cpecifications. This calculation in presented in Appendix C. 3.3.4 RCC Bank Worths Dank worth measurements were performed over the four individual control banks (i.e.: non-overlap mode). Measured and predicted integral worths of these four banks arc summarized in Table 1.2. Also the differential and integral worth of each bank is shown as a function of bank position in Figures 3.5, 3.6, 3.7 and 3.8. 3.4 AT POWER TESTS 3.4.1 Power coefficient Measurement See Appendix A i e i e 1-3
3 . ',a Movable Intector l']ux Mapu Orte movable detector flux map was taken at low power (w4*.) in the all-rods-cut cor. figuration. The results of this map arc shown in l'igure 3.9 and 3 10 with the measured Fall compared to predicted l'A11 for cach assembly. In general Figures 3.9 and 3.10 shows that the measurements are consistently larger than the predictions around the core periphery while the opposite is true in the core interior. Ilowever all values are within the acceptance criteria of i 15t. Additional flux map results are summarized in Tib'.e 3.2. TABLE 3.2 FLUX MAP
SUMMARY
HOT ZERO POWER - ALL RODS OUT
.996 .986 .971 /
Tilts f 1.018 1.000 1.025 1.008 N
.996 Core Avg. Lxial Offset: +46.46%
N Peak F - 1.642 at location J-14 DJ All n Most limiting F : 3.0614 at location G-ll El, axial point 8 3-4
A u cond f ull core flux sul, was taken at f ull g>0wer (v991) in the all-rodu-out configuratiore. The result.s of thic map are shown in riguren ~).11 and 3.12. All valuun are within the acceptance criteria of 115*. Additional flux map results are cummarized in Table 3.3. TABLE 3.3 FLUX 14AP SUMMAP.Y If0T FULL PCMER - IsLL RODS OUT Tilta 1.0085 1.0003 .9997 1.0023 .9890 1.0073 .9'359 r1
/ .9972 Core Avg. Axial Offset: + 4.932 4 11 Peak r : 1.503 at location 11-13 JL t II !!
14ost limiting r : 1.826 at location E-07 LK, 9 axial point 10 4-:i
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- 4.0 I!CTRUMUltr CALIDI<ATIONS 4.1 Incore - Excore Detector Calibration One f ull-core map and 8 quarter core maps were taken at 90% power to om. in calibration data for the excore instrumentation. Thecc maps covered a range in axial offset from -4.7% to 11.9% generated by insertion of control bank D and the resultant axial xen,on oscillation produced upon withdrawl.
INCORE analyses provided a measu. red value of the core average axial offset which was used as the basis for excore calibration. The results of the calibration are listed in Table 4.1. 4.2 Incore Thermocouple and Wide Range RTD calibration The primary purpose of this test was to determine isothermal correction factors as a function of temperature for individual thermocuples. During the heatup of the reactor subsequent to the refueling, reactor temperature was held constant at plateaus of approximately 29C#F, 44L'F and 544*P, where individual thermocouples readings were compared to the average of the narrow range RTD reading. Table 4.2 lists the results of these measurements including the isothermal correction factors for each operab]c thermocouplc. New calibration curves based on those correction factors where entered into the plant p-250 computer. 1-1
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TA1:1,1; 4. 2 Incoru ThurtnucoupJe Inot.herinal correction l'actorn Tavy (from narrow ranqe RTD) = 295. 72 ^!' Correctior, factor = Tavy - T.C Cor:cction factors for each thermocouple in ascending order (all values in v r) :
-0.50 -1.03 -0.83 -0.18 -0.93 -1.28
-0.73 -
0.02 -
-0.83 i -0.23 -1.03 -0.68 -0.23 -0.88 -0.78 -0.08 -0.43 -0.73 -0.38 -0.23 -0.83 -0.03 -0.93 -0.88 -0.93 -0.78 -1.13 -
1 -0.68 -0.63 -1.08 -0.08 .l.43 -
-0.63 -1.08 -0.98 -0.88 -0.68 -1.48 0.12 -0.93 -0.93 -0.73 -068 - -0.28 -0.33 -0.98 -0.43 -0.43 -1.18 -0.18 -0.73 -1.03 -0.28 -0.83 -0.68 -0.48 -0.98 -0.63 -0.52 -0.83 Tavg = 449.76 P Correction Factors:
3.16 2.61 2.91 3.61 2.76 1.96 - 3.21
- 4.11 -
2.86 3.87 2.36 3.81 3.36 2.41 3.11 4.11 2.86 3.11 3.01 3.71 2.61 3.8G 2.76 2.56 2.71 3.17 2.36 3.41 2.96 2.41 4.16 1.96 - 2.9G 2.56 2.86 3.21
- 2.01 4.'46 2.76 2.61 3.26 3.01 -
3.66 2.81 2.41 3.66 3.41 2.21 4.16 2.51 2.91 1 1- 3
Tavy. = !,43.90"I' Correction l' actors: 1.90 1.35 1.85 2.80 1.65 1.10 - 2.35 3.30 - 1.60 3.05 1.15 7.90 2.55 1.50 2. 4 5. _. 2.40 1.35 2.30 1.70 3.00 1.55 2.40 1.55 0.80 1.70 2.20 1.10 2.50 1.90 1.40 3.65 0.60 - 1.00 1.35 1.75 2.30 0.85 3.65 1.90 1.60 2.20 1.95 - 2.65 1.90 1.40 2.70 2.75 1.15 3.65 1.40
~'
1.20 2.40 1.40 - 2.80 1.40 2.25 3.45 2.15 o 4-4
a APPE!IDIX A licasurement of the Power Coef ficient 1.0 IITRODUCTICII A measuremont of the power coefficient was performed at the beginning of cycle 2 of Indian Point Unit 3 on September 19, 1978. The method utilized, involved exchanging power for nCS temperature while maintaining constant rod position and boron concentration. Power level was monitored via the core 6T sinco no other power signal is either accurate endugh or unaffected by the changing RCS temperature. Specifically, calorimetrics are not accurate enough because of the small power changes involved ( 2t) and the
!!IS are also inaccurate due to the RCS temperature deviating from the programmed temperature. The purpose of this report is to present the measurement results and the data analysis.
h A-3
2.0 !41:A.",Ulti;!4r!!T P.!$ULTS i 1:CS temperature (Avg Tavg) and core average BT were hooked up to a two pen stripcharticcorder and scaled to 1(T/ inch. After observing the averagea T signal for about an hour, it was clear that the signal was too noisy to allow an accurate aT reading. It was decided to look at an individual loop dT rather than the core average oT. The nT signals from loops 1 and 2 were found to be stable whereas the signals from loops 3 and 4 were considerably noisy. The measurement was therefore performed with loop 1 lit input to the stripchart recorder and then repeated using loop 2 AT. The conditions of the plant at the time of the measurement were as follows: Power = 92% Bercn Conc. = 953 ppm
- Tavg = 565. 35'F D position = 220 steps Temperature results from both measurements are summarized in Table 1.
TABLE 1 Summary of Temperature Data from Powcr Coafficient I'easurement Indian Point Unit 3 Cycle 2 i f Tavg Tavg dT t)T Q Q i f (From calorimetrics) i til 565.35 566.50 46.10 45.55 92.42 90.51 566.50 565.06 45.55 46.22 90.51 92.18 (!2 565. 566.32 47.50 46.08 92.18 ----- 566.32 565.22 46.88 47.40 ----- ----- l'or measurement F1, Inop 3 aT was used. l'or measurement l' 2 , I nop .' A T wa s u sed . f.u te - i o initial i !inal O j . owe r
3.' O DATA AnAj,ycis The analynic of the power coef ficient data is somewhat involved arid in presented in flow chart form in Pjgure 1. Note that even though the RCS temperature changed during the experiment, the final power icvol (Pg ) is evaluated using the initial RCS temperature. This is donc in order to insure that the final answer is in the form of the ratio of the isothermal temperature coefficient to the doppler-only power coefficient. That is, the doppler-only power coefficient is defined as a change in power at constant RCS temperature, therefore in the analysis no RCS temperature change should be used. Results of these calculations are shown in Table 2. TABLE 2 Results of Data Analysis Tavg (OF) P P P (t) Power Coef. Afg Tavg Avg P i f i f (Tavg - Tavg) (t) (%) (p -P) U.Tavg/. P) ( F) (t) i f 61 -1.15 92.18 91.04 . 1.14 -1.009 565.93 91.61 1.44 91.16 92.55 -1.39 -1.036 565.78 91.85
#2 -1.32 92.42 91.17 1.25 -1.055 565.66 91.79 1.10 91.30 92.35 -1.05 -1.046 565.77 91.83 Averages -1.036 5G5.70 91.77 c
A-3
4 '. 0 Gut *:1hlG > Thre power coefficient calculated in Tabic 2 has the units of "r per percent power that is: -1.036 Or/t. To arrive at a doppler-only power coefficient in terms of the more familiar pcm per percent power a calculated isothermal temperature coefficient must be used. This value can be obtained from UCAP-9244, rigure 5.1 using the measured conditions of 565.7800 Tavg and 953 ppm boron concentration. Using linear interpolation in rigure 5.1 (BOL) the expected isothermal temperature coefficient is -8.48 pcm/Or. During the zero power physics testing it was found that the predicted isothermal temperature coefficients in rigure 1 are biased more positive that the measured values by 0.50 pcm F. That is, the measured isothermal temperature coefficent is 0.50 pcm/ r more negative than predicted. Therefore, the proper isothermal temperature coefficient to be used in this analysis should be the predicted value biased 0.50 pcm/ r nore negative: 4 Bo
~~Pr Aso = 8.48 - 0.50 = -8.98 pcm/ r Therefore the measured doppler-only power coefficient is:
-(8.98 pcm/Or x 1.036 Or/*.) = -9.30 pcm/t This compares very favorably with the design value of -0.67 pcm/t found in rigure 5.5 of WCAP-9244.
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