ML20073E526
| ML20073E526 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/22/1991 |
| From: | Woodard J ALABAMA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9104300258 | |
| Download: ML20073E526 (15) | |
Text
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~ Alabama Power Company
- --* '- 40 invemen Center Parkway :i Pov. Othee Box 1295 - '
. Wrrungham Alatwrm 3520t
' Teiephone 205 868 5086
<- .; w J, o. Woodard N ObOIIld b WCf' Vce Pteudent-Nuclear .
rarief ero,oct April 22. 1991- ""#**'*"#*
Docket No. 50-364 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Vashington, D. C. 20555 Gentlemen Joseph M.-Farley Nuclear Plant Unit 2 Cycle 8 < Startup Report 4
Enclosed is the Startup Report for Unit'2 Cycle /8 as referenced.-in our
- Cycle 8 Reload letter dated November 9, 1990.
I- If you have any questions, please advise.
i . Respectfully submitted,-
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.D dek
.- 'Voodard 4
JDV/HDRimaf2969 Enclosure--
1 cc: 'Mr..'S D. Ebneter Mr. S. T. Hoffman i Mr. G. F.: Maxwell i
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9104300258.910422 .
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. -ALABAMA POWER COMPANY ,
JOSEPH M. FARLEY NUCLEAR PLANT
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[
UNIT NUMBER 2 CYCLE 8
[
'I STARTUP TEST REPORT j (Test Activities From'10-24-90 to 01-22-91)
PREPARED BY THE PLANT REACTOR ENGINEERING GROUP !
.i I
TABLE 0F CONTENTS PAGE 1.0- Introduction . :1- -t 2.0 Unit'2 Cycle 8 Core. Refueling 1 l 3.0 Control-Rod Drop Time Measurement 6 4.0 Initial Criticality 8-5.0 All-Rods-Out Isothermal Temperature' 8-Coefficient 6.0 Baron. Endpoint; Contro1 Land-Shutdown 9 ,
Bank Worth Measurements 7,0 Startup and Power Ascension Procedure 10 <!
8.0 Incore-Excore Detector Calibration. 12 9.0 Reactor Coolant System Flow- "13 : '[
i Measurement-s
. 'l APPROVED:
1
[.OL h 4? #li Technical Manager .i ll
,y bM General; Manager - Nuclear Plant-f + >
PN/WSM/STARTRPT ' q n
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The Joseph M. Farley Unit 2 Cycle 8 Startup Test Report addresses the tests performed as required by plant procedures following core refueling. The report - provides a brief synopsis of each _ test and' gives a comparison of measured parameters with design predictions,-
1echnical Specifications, or values in the FSAR safety analysis.
Unit 2 of the Joseph M. Farley Nuclear Plant is a Three Loop Westinghouse pressurized water reactor rated at 2652 MWth. The Unit began comecreial operations on July 30, 1981. The Cycle 8 core loading consists of 15717 x 17 fuel assemblies and has a design burnup capability of 15,900 MWD /MTU. t Cycle' Completion Date and Average Burnup Cycle Number- Completion Date -Avg. Burnup (MWD /MTU) 1 October 22, 1982 15,350.5 2 September 17, 1983- 10.371.2 3 January 4. 1985 14,639.0 4 April 4, 1986- 13,183.8 5 October 2' 1987 16,674.0 6 March 24, 1989 _16,137.8 7 October 13, 1990 17,051.0 2.0 UNIT 2 CYCLE 8 CORE REFUELING ,
REFERENCES -
- 1. Westinghouse Refueling Procedure FP-APR-R7
- 2. Westinghouse WCAP 12704 (The_ Nuclear Design and Management of-_ '
the Joseph M. Farley Unit l2 Power' Plant-Cycle 8)
Unloading of the cycle 7 core- into ' the spent fue1~ pool commenced on 10-24-90 and was completed on 10-26-90 with=no' major problems.
During the core unload, binocular _and TV inspection disclosed that fuel assembly U41, scheduled for reload in the Cycle-8, core, had a -
! gouge-like defect on one rod. As a result,-assembly U41 was rejected-for reload and the Cycle 8 core loading. pattern-was redesigned.
During Cycle-7 operation, radiochemistry'datalindicated that one or more fuel assemblies were leaking radioactivity into.the reactor coolant. Therefore, all fuel assemblies unloaded-from the, Cycle-7 core were subjected to ultrasonic leak testing. One-lenking fuel l assembly, S44, was. identified.1 However,;S44 was not: scheduled for reload, so the core design ~was not impacted. -Following insert! >
changeout, the Cycle 8 core reload began on:ll-13-90,and was completed on 11-16-90.
The as-loaded Cycle 8 core is shown in Figures 2.1- through 2.4, which give the - location ~ of each. fuel assembly and insert, - including wet annular burnable absorber insert locations and configurations.
1 F
e FIGURE 2.1: UNIT 2 CYCLE 8 REFERLUCE LOADING PATTERN A B C D E F G H J K L M N P R 205 283 291 , ,,,,,, g US$ U51 U32 2112 217 R107 250 244 8113 234 Y54 V16 YS5 W44 045 ,,, ]
_ 14
' V50 W35 740 233 16W97D t109 16WB7D t122 16W840 222 226 _ _ y3 U12 Y34 v41 Woe Y42 WS Y51 YD4 U29 274 R103 1 M 40 t127 232 $$04 225 R105 1M37 R111 242 -
-- 12 U24 W61 f35 W23 W57 W14 W3 WO6 Y31 W50 UO3 278 209 12W139 R116 1W910 236 16W940 254 16W990 R132 12W134 249 264 _ _ gy U39 Y29 Y15 v12 Y26 W27 Y12 W34 Y05 W52 Y32 YO7 U46 till 16 Weed R145 16W60 t144 i M 43 R104 1M42 R140 1W900 R106 I W 950 R123 - - 10 WO4 Y37 W47 Y27 U42 Y18 UO7 Y21 060 Y34 W3 Y45 W28 4134 1 M 30 a143 22 7 R102 1W133 221 234 R120 257 287 262 206 285 235 Y44 9
U23 Y43 W38 W64 W24 Y22 WOS W2 W25 Y28 WO7 W51 W19 U64 237 R121 IW790 212 1W100 R142 2 73 251 270 R144 16W50 230 1WiO1 R101 201 g U37 W10 Y47 WO9 Y30 U21 W53 W39 W60 ,U13 YOS W29 Y50 W1 047 1 M 44 R139 213 290 263 208 R141 228 til 1N135 R130 207 titt 261 216 f
- U63 Y49 W33 W54 W49 YO9 W21 W52 W17 Y23 W2 W59 WO Y40 U22 t124 t134 2144 1 M 45 d129 IW920 2117 16WB90 8137 R128 16WB10 16WB00 12W132 6 W13 YS4 WIS Y01 U52 Y13 U16 Y10 044 Y17 W20 Y34- W22 272 202 12W131 R133 16W90D 214 1W73D 210 1 WEN R110 12W141 219 239 062 Y16 Y20 W64 Y19 WO2 YO6 W57 Y33 W18 702 Y25 U40 73D R108 1 M 34 8131 254 S303 231 tilf 1 M 34 R125 254 4 U34 W55 Y24 W26 W54 W31 W54 W50 Y14 WS U31 296 220 1 W960 R126 1W93D R134 ledh 215 204 3 U25 Y11 YS3 W54 YS2 WO1 Yee YO3 UO6 223 R114 218 0135 224 R147 295 2 U54 W11 Y44 WOS T39 W5 044 25 M M i U20 USS R21 xxx = INSERT S/N xxx = FUEL ASSEMBLY S/N A untfu.
The original w/o U-235 enrichments were:. No. of Fuel Assemblies Region 5 (R) assemblies .... 3.402% Region 5 ............ 1 Region BA (U) assemblies ... 3.598% Region 8A ........... 16 Region 8B (U) assemblics ... 3.994% Region 8B ........... 19 Region 9A (W) assemblies ... 3.792% Region 9A .......... 49 Region 9B (W) assemblies ... 4.202% Region 9B ........... 16 Region 10A (Y) assemblies .. 3.800% Region 10A ......... 36 Region 10B (Y) assemblies . . 4.200% Region-10B ......... 20 Total 157 2
.-. .- - _ - . . - ~ . .- .
FIGURE 2.2 CONTROL ROD LOCATIONS R P N M L K J H -0 F E D C B A 1
2 A D A 3 SA SA SP 4 C B SP B C 5 SP 58 SP $8 6 A B D C D B Al 7 SA $8 S8- SP SA ,,
8 po- D SP C SP C SP D 9 SA SP S8 SS SA 10 A B D C D B A 11 SS SP SS~ SP 12 C B SP 5 .C 13 ,,SP SA- SA 14 A D A 15 ' Abs Mgial:
o' ;
FUNCTION NUMBER DP CLUSTERS I Control Bank D 8 Control Bank C 8 Conwol Bank 8 8 Control Bank A 8 l l Shutdown Bank 38 8 Shutdown Bank SA 8 SP (Spare Rod Locadone) 13 3
T10URE 2.3 <
BUPNABLE ABSORBER AND SOURCE- ASSE!!BLY LOCATIONS R P N M L K J .H G F E D C B A i 1
2 3 16 16 16 4 12 SS 12 5 12 16 16 16- 12 6 16 16 12 ~2 16 -16 7 12 12 8 16 14 16 14 9 12 . 12-10 16 16 12- 12 16: 16 ~
~
11 12 16 16 16 -12 <
12 12 'SS 12 13 . 16 16 16 14 15 s
- - Number of WABAs . -Summary of Inserts i SS Secondary Source
! WABA Clusters .40 576 WABAs in Control Rods 48 40 Clusters mble Plugs R Sec. Sources- 2-
- To tal - 157
- 4.
1 l
FIGURE 2.4 BURNABLE ABSORBER AND SECONDARY SOURCE ROD CONFIGURATIONS O O S S
+
8 8 0 8 -8 8 0
- 8-O 8 0 8 0 0 8- 8 0 8 0 8- O 8 8 8 8.
8 0 .8 -
8 O8 .
I 12 BA CONFIGURATION 16 BA CONPIOURATION O O O 1 0 O O 0 O O -
O 8- -
O [E SECONDARY SOURCE LOCAT10N8'-
S
.= _ __ _ _ . . _ - . , - . . . m._ .. -
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1 8.0 CONTROL ROD DROP TIME MEASURDENT ~ (FNP-2-STP-ll2)- !
l t.
PURPOSE -j 1
l The purpose of this procedure was to measure the drop time of all 3 l full length control rods under hot-full flow-conditions-in the reactor C coolant system to insure compliance with Technical Specification Requirements.
l St> MARY OF RESULTS For the' hot full-flow condition (Tavg 1 541 deg.F and_all reactor coolant pumps operating) Technical Specification 3.1.3.4 requires that .I the drop time from the fully withdrawn-position shall be i 2.2 seconds !
from the beginning of stationary gripper _ coil-. voltage decay until dashpot entry. All full length rod drop. times were measured'to bei less than 2.2-seconds. The longest drop ~ time recorded was 1.48-seconds for rod B-6. The rod drop time results:for both dashpot' entry and dashpot botton,are presented in Figure 3.1. Mean drop times are '
sumanarized below:- ;
TEST MEAN TIME 10 MEAN TIME 'IO CONDITIONS DASHPOT ENTRY DASHPOT BOT 1VM Hot full-flow 1.374 see 1.829 sece j To confirm nonmal- rod mechanism ' operation prior. to conducting the - i rod drop test, the Verification of Rod Control' System,0perability.
-(FNP-0-ETP-3643) was performed. .In this test, the stepping waveforms--
of the stationary, lift and movable gripper coils were examined,'-rod' speed was measured, and the functioning of the Digital: Rod position?
Indicator (DRPI) and' bank overlap unit were checked.
r As a .part of the RCCA wear: reduction program . the Cycle 18 control rod fully-withdrawn position was changed,from 228 steps?to 2311 steps: +
by_ increasing bank. overlap from:100. steps'to:103 steps. 'During the: R Rod Control System Operability testh the bank overlap unit; switch l settings and functions were verified correct for?103 steps of. bank i overlap.. In addition, the actualofully-withdrawn position =of each-RCCA was measured using stationary gripper coll: traces: to : provide, data for planning RCCA repositioning-for. future fuel cycles.1 ~
Initially, during.the_ Rod Control' System Operability test.;none.
of the Group-2 rods _.would nove. cSince half- of the Group-2 rods' are i driven by -the LAC power cabinet, and the other half by.the 2BD power-cabinet this' suggested that a problem existed;in both cabinets _. Uponi investigation, one of -the power fuses 'on top of the 2AC powerLeabinet '
-was discovered to be blown'and was replaced. 1Although the-specific problemi with the 2BD power cabinet was'never identified,,the;cabineti became operational following the removal 4 and reinsertioniofc several~
circuit cards during the course:of troubleshooting.?=Thus,=the probleme j was attributed to a-loose circuit card connection.. -
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, UNIT 2 CYCLE 8 MORTH 900 1.38 1.81 1.37 1.83 1.41 1.86
\V ~
P 1.38 1.37 , ;
1.83 1.8 y M i 1.37 1.37 1.35 1.39 1.8 1.83 ,1.8 1.83 1.38
'l.83
'1.35 1.79 [L
'N 1.47 1.35 1.33 1.32 1.34 ~1.36 ,
1.40 1.91 1.8 1.85 1.85 1,83 1.78 1.85 K
'1.37 1.34 1.33 1.38 1.79 1.77 1.78 1.81 J s o 1.41 1.35 1.35 1.38
, 0 o 1.86 -
1.87 1.87 1.81 180 -H 1.37 1.32 1.37 1,79 1.77 1.78 '0 _g 3,g 1.41 1.40 1.36 1.33 1.37 1.38 1.37 1.83 1.85 1.85 1.81 1.87 1.81 1.83 I 1.35 1.38 1.8 'l.87 E a 1,38 1.36 1.36 1.37 1.8 1.81 1,79 1.83 0 1.40 1.38 1.86 1.83 0 1.40 .l.43 1.48 l1.88 1.88 I 1.97 1 g 270 0
g i
15 14 13 12 11 10 9 -8 7 6 5- 4 3 2 i ORIVE LINE "0 ROP TIME" TABULATION TEMPERATURE . 548.27 PRESSURE - 2291.021- y p(gy _ 100 X.XX BREAKER "0PEMING' TO DASHPoi ENTRY - IN SECOMOS I-l-91 DATE -
X.XX BREAKER "0PENING"'TO DASHPOT BOTTOM - IM SECOMDS
.FIC. 3.1 I
-7
}
During the Rod Drop Tine Me u urement, the came'reveraal was'noted in the rod D12 rod drop trace that had been observed previously during
. the Cycle-7 startup . Although no anomalies were observed in the step-ping traces for rod D12 during the Rod. Operability Test, the traces' 1 were re-examined and were confirmed to be normal. It is believed that the magnetic polarity of the rod D12 drive shaft was reversed, causing its rod drop trace to be reversed with respect to the remaining rod drop traces. The drop time of rod D12 was normal (1.38 seconds to -!
dashpot entry). '
4.0 INITIAL CRITICALITY (FNP-2-ETP-3601)
PURPOSE
]
The purpose of this procedure was to achieve initial criticality I under carefully controlled conditions, establish the upper flux limit -- 1 for the conduct of zero power physics tests, and operationally verify the calibration of the reactivity computer. !,
StM1ARY OF RESULTS Initial Reactor Criticality _for Cycle 8 was achieved during dilution mixing at 1816 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.90988e-4 months <br /> on January 3, 1991.- The-reactor was allowed to stabilize -at- the following conditions:
RCS Pressure- 2230 psig RCS Temperature 547.00F Intermediate Range Power 1.5 x 10-8 Amp RCS Boron Concentration 1893.5-ppm Bank D Position 186.5-steps-Following stabilization, the point of. adding __ nuclear heat was ,
determinud and a checkout:of the reactivity computer using positive __ !
and negative flux periods was performed. In_ addition, NIS source _an'd- 1 intermediate range overlap data.were taken_during the flux! increase .;
prior to criticality to demonstrate.that adequate overlap existed. ; 1 Initial criticalitt for Cycle 8'was achieved with.the flux'in the source range as a result of withdrawing -the source-range (SR) detec-tors to the maintenance position to-reduce the SR count _ rate. _During _
previous cycles, the high SR count rate resulted in criticality being-achieved in the intermediate range. . Testing performed prior'to criticality (FNP-0-ETP-3652)-demonstrated that withdrawal of the SR ~.
detectars reduced the count. rate to one-third of the -value with the detectors in their normal, inserted position.-
5.0 ALL-RCDS-0UT ISOTHERMAL TEMPERATURE COEFFICIENT AND. BORON ENDPOINT :
(FT-2-ETP-3601)'
-PURPOSE The ob.jectives of these measurements were to determine the -hot. ;
zero power-isothermal and moderator temperature coefficients' for :the? ;
all-rods-out (ARO) configuration and to measure the AR0; boron endpoint-concentration. '
8 '
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4 , .
. . St> NARY OF- RESULTS The ARO, _ hot zero power temperature coefficients and - the ARO -
boren endpoint concentration are tabulated below.- As described-in Par. 6.0, these data have been_ adjusted to correct for an error in the delayed neutron data provided for reactivity computer calibration.
ARO. HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT l Boron Measured 'ITC Design Acc.- Calculated Rod Conc. ITC Criterion MTC Configuration ppm - pem/0 F _pcm/0 F pcm/0 F All Rods Out' 1913.7 -1.13 -0.38 1 2 +.694 ITC = Isothemal temperature coefficient, includes -1.72 pcm/o r doppler coefficient -!
MTC = Moderator only temperature coefficient,-normalized to_the-ARO '_ j Condition i ARO. HZP BORON ENDPOINT CONCENTRATION Rod Configuration -Measured ce (ppm)- ' Design-predicted Co (ppm) i All Rods Out 1913.7 1955 1 50
=;
Since the measured MTC (0.694 pcm/a r)'was well within thei Technicali- I Specification limit of +5.0 pcm/oF, no rod withdrawal . limits- were -
required. The. design neceptance criterion for.the*ARO boron concen --
tration also was satisfied, a
6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS (FNP-2-ETP-3601)-
q PURPOSE ,
The objective of the bankiworth measurements was-+i determine the !
integral reactivity worth of each control and shutdown _ bank- for . -l" comparison with the valuescpredicted by design.
Sl> MARY OF RESULTS The rod worth measurements were . performed using the bank' inter - l change. method in which:x(l) the worth-cof.the bank'hav'ingithe highest, J design-worth _(the'" Reference Bank")>is carefully measured us.ng the _
i standard dilution method then (2) the worths of the remaining control and shutdown banks are^ derived from the change 11n reference bank; reactivity needed to offset: full insertion'of the bank being measured.- !
The initial results of the reference bank--measurement were :i approximately 10% high.- During :the review and evaluation' process, san error was discovered -in the delayed neutron data provided in -Ref. : 2 ' l for calibration of the reactivity computer. -InLorder to correct this- j 9
____:__=_--_=__== -- - - _ - - _ ._.
- error. Westinghouse indicated that all reactivity computer reasure-ments (including the boron endpoint and_ isothermal temperature ,
coefficient measurements outlined in Par. 5.0) should be reduced by a factor of 0.9055. The' adjusted control and shutdown bank worths (tabulated below) satisfied the review criteria both for the banks measured individually and for the combined worth of all banks.
SlMtARY OF CONTROL AND SHUTDOWN BANK WORTH MEASURPMENTS -
Control or Predicted Bank Shutdown Worth & Review Measured' Bank' Percent Bank Criteria (pem) Worth (Dem) Difference A 335.1 100 326.84 -2.44-B (Ref.) 1168 1 117 1169.7* +0.15 0 848 1 127 849.65 +0.'19 D 940 1 141 923.06 ' -1. 80 ~
SDA 908 + 13G 870.16 -4.17- i SDB 1070 1 100 1085.17 +1.42 Combir.ed 5269 1 526.9 5224.58 -0.84-
- Measured by dilution method 7.0 STARTUP AND POWER ASCENSION PROCEDURE:(FNP-2-ETP-3605)
PURPOSE The purpose of this procedure was to provide controlling-instructions for'
). NIS intermediate and power range setpoint changes,-as required prior.to startup and during power as'c'ensjon,
- 2. Ramp rate limitation and controlfrod' movement = recommendations.
- 3. Conduct-of startup and power ascension: testing, toxinclude:
~
a._ ' Hot zero power.(HZP)' physics tests'(FNP-2-ETP-3601).:
- b. Incore movable detector system alignment 1(FNP-2-ETP-3606).
- c. -Incore-excore AFD channel recalibration (FNF-2-STP-121).-
- d. Core hot channel factor surveillance:(FNP-2-STP-ll0)'.
- c. Reactor coolant system flow measurement-(FNP-2-STP-115.1)'.
SIMIARY-OF RESULTS-In ordier to . satisfy Technical: Specification requirements for invoking special -HZP physics test exceptions,? preliminary t rip . set-points-of=less.than or equal to 25%~ power were used for the NIslinter-
~
mediate _and-power range channels. :In addition, the intermediate = range. -
10 j
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a + . M channel setpoint currents were reduced-to eddreas the effects of SR/IR detector repositioning (described-_in Section 4.0) and projected changes in-core-neutron leakage from-the previous core _ cycle. When physics tests were completed,the power range setpoint was increased to -
80V to allow power escalation above -25% for calorimetric. recalibration I of the. power' range channels..-(The 80V setpoint was administratively imposed to address the possibility that the uncalibrated power range channels could be indicating non-conservatively.) At approximately 34% power, the power range channels were recalibrated and setpoint currents were determined for.the_ intermediate range channels. ,
Following recalibration of the power range channels, the NIS PR l high-range trip setpoint was restored to 109%. ]
The Westinghouse fuel warranty limits the power ramp rate to 3%_
of full power per hour between 20% and 100% power'until full power has-been sustained for'72 cumulative hours out_'of any seven-day' operating:
period. This' ramp rate was observed during the' ascension-to full power.
The determination of incore movable detector ' system core _ limitc _ '
settings (FNP-2-ETP-3606):was accomplished during the ascension:to 34% '
power. This was followed by the incore-excore recalibration testr (FNP-2-STP-121) at 34% power _ and the reactor coolant flow- measurement -r (FNP-2-STP-115.1) atL100% power, which are described in Sections;8.0' and 9.0 of this report. _As summarized in Table'7.1, core hot channel a factor surveillance was performed on the incore-excore full-core base
- fi case flux map _taken under non-equilibrium conditions'at.34* power,can'd _;
on the full-core flux maps performed at. equilibrium xenon at;33.6F and 100% power. jr TABLE 7.1:
SibHARY OF POWER ASCENSION FLUX MAP' DATA i
Parameter Map 196 .
Map 202- Map'203- - '
Avg. % power 34.0% - 33.6% 100.2%
Max FDH 1.5577 1.5873 1.4783-Max power tilt * .l.0007- !
1.0007' l.0005 !
Core avg.
- A.O. 10.468 J10.885' :1.742 y Limi t ing' FQ(Z)' ' 2.1390-- ' 2.2195- 1.8508L FQ Limit- 14.5240' .4.52'40 L2.3084L Xenon conditions Non < : Equilibrium - Equilibrium'- -
Equilibrium;
- Calculated power 'ilts. t based on assembly ~FDHN fromLall assemblies 1:
- The most l limiting FQiZ), ' based on margin to the 'FQ climit.
4 11 1 3i
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- p ,. ,-w- , ,y e ~,-p. p. e.,e--my-.. g- e v. ey s e, s- a y y + w , y , g p. +
8,0 fNCORE-EXCORE DETECTOR ~ CALIBRATION (FNP-2-STP-121)
PURPOSE The objective of this procedure was to determine the relationship between power range upper and lower excore detector currents and axial offset for the purpose of: calibrating the control board and the plant :
computer axial flux ~ difference ( AFD) channels, and for calibrating the. ,
delta flux penalty to the overtemperature delta-T protection system.
SUMMARY
OF RESULTS At an indicated power of approximately 34%, a full-core base case-flux map-and five' quarter-core flux maps were performed at various positive. and negative axial offsets to develop equations relating - ,
detector current to core axial offset. To reduce error, an effort was made to perform all flux maps at the same reactor. coolant system tem -
perature. The power range NIS channels were adjusted to incorporate the revised calibration data.
Following the recalibration, escalation to full power _ proceeded without incident. At 1004 power, under equilibrium xenon' conditions, a full-core flux map was performed- to _ correct the incore-excore calibration for the, effects of .the rise. in ~ RCS average temperature: as power is. increased. Table 8.1-gives the final detector. current vs-- -
axial offset equations obtained1following the calibration correction -
at 100& power.
TABLE 8.1 DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS
-0BTAIN FROM INCORE-EXCORE CALIBRATION TEST-CHANNEL'N41:
4 I-Top. = 0.8136*AO -+ 156.91.uA I-Bottom = -0.9486*AO +- 152.87 uA-CHANNEL N42:
,I-Top = 0.8266*A0 -+ 157.63-uA I-Bottom = -1.000l*AO, +. 152.59 tui- ;
CHANNEL N43:
I-Top = 0.8258*AO- + 160.81 uA I-Bottom = -1.0066*AO -+ 155.46 uAL 1
CHANNEL N44: i I-Top = 0.9208*AO ' +: 173. 93 ,tV(
I-Bottom = -1.1521*A0 + -176.06 uA' 12 I' - - - . - . . . . ,- .. . . . , . - - . - , , - , - - , - ., ,- +
9.0 REACTOR COOLANT SYSTEM FIDW MEASUREMENT (FNP-2-STP-115.1).
PURPOSE The purpose of this procedure was to measure tho flow rate in each reactor coolant loop in order to confirm that the total core flow met the minimum flow requirement given in the Unit 2 Technical Specifications.
St> NARY OF RESULTS To comply with the Unit 2 Technical Specification, the total reactor coolant system flow rste measured at normal operating temperature and pressure must equal or exceed 261,600 gpm for three loop operation. From the average of twelve calorimetric heat balance measurements, the total core flow was determined to be 279,780 gpm..
which meets above the criterion.
The RCS flow test data (which include direct RTD measurements of-RCS T-hot T-cold and percent thermal power) were also used to deter-mine delta-T for each RCS loop in ordn to evaluate the need to-rescale the delta-T protection and control systems. Based on the ,
results of this measurement, Loops 2 and 3 were rescaled.
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