ML20114E011
| ML20114E011 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/04/1992 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML20114E010 | List: |
| References | |
| NUDOCS 9209110002 | |
| Download: ML20114E011 (16) | |
Text
-
SOL I4ERN NUCLEAR OPERATING COMPANY aOSEPH M. FARLEY NUCLEAR PLANT STARTUP TEST REPORT UNIT 2 CYCLE 9 TABLE OF CONTENTS g
1.0 Introduction 1
2.0 Unit 2 cycle-9 core Refueling 1
3.0 control _ Rod Drop Time Measurement 7
4.0 Initial criticality 9
5.0 All-Rode-Out Isothermal Temperature 9
coefficient and Boron Endpoint 6.0 control and shutdown Bank Worth 10 Measurements
=
7.0 startup and. Power Ascension Proce+*re 11 8.0 Incore-Excore Detector Calibratio-12 9.0 Reactor coolant System Flow 14 Measurement' 4
APPROVED:.
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f General Manager - Nuclear Plant
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1.0 INTRODUCTION
The Joseph M.
Farley Unit 2 Cycle 9 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, Technical 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 commercial operations on July 30, 1981. The Sycle 9 core loading consists of 157 17 x 17 fuel assemblies, of whit, 100 are Westinghouse Low Parasitic (LOPAR) assemblies and the remaining 57 assemblies reprusent the first phase of transition to Westinghouse vantage 5 fuel.
All thimble plug and burnable poison inserts have been removed from the Cycle 9 core and two new double encapsulated secondary source inserts were added to be activated for use in Cycle 10. The new sources provide compatability with Vantage 5 fuel and the double encapsulation gives an additional margin of protection against source material leakage into the reactor coolant. The burnable absorber inserts have been supplanted by Westinghouse Integral Fuel Burnable Absorbers (IFBAs) incorporated in the Vantage 5 assemblies. The design burnup capability of the cycle 9 core is 17100 MWD /MTU.
Previous Cvele Completion Dates and Averace Burnupe Date St art-o f EOL EOL Burnup EOL Burnep Total Cycle Critical Cycle Rata (MWD /MTU)
(EFPD1 EEE1 1
05-08-81 05-27-81 10-22-82 15350 416.50-1.141 2
11-30-82 12-03-82 09-16-83 10371 281.68 1.913 3
10-22-83 10-24-83 01-05-85 14639 397.73 3.002 4
03-08-85 03-20-85 04-04-86 13183 359.48 3.987 5
05-11-86 05-13-86 10-03-87 16674 457.67 5.241 6
12-02-87 12-05-87 03-24-89 16138 444.09 6.458 7
05-18-89 05-21-89 10-13-90 17051 468.76 7.742 8
01-03-91 01-06-91 03-06-92 14757 405.69 8.853 2.0 UNIT 2 CYCLE 9 CORE REFUELIhG REFERENCES 1.
Westinghouse Refueling Procedure FP-APR-R8.
2.
Westinghouse WCAP 13201 (The Nuclear Design and Core Management of the Joseph M. Parley Unit 2 Power Plant Cycle 9)
Unloading of the cycle 8 core into the spent fuel pool commenced on 3-16-92 and was completed on 3-18-92 with no major delays.
During the of fload, each fuel assembly was inspected with binoculars for indications of damage or other problems.
Assembly U60 was found to have shiny rub marks on the grids.
No damage of any kind was found.
During insert changeout, thimble plug No. 275 fell out of the tool and could not be atraightened suf ficiently for storage in a spent fuel assem-bly.
Therefore, it was placed on top of the fuel assembly located in spent fuel pool position H02.
Following the insert changeout (which included removal of all thimble plug and burnable poison inserts) the Cycle 9 core loading commenced on 4-15-92 and was completed on 4-17-92.
The as-loaded Cycle 9 core is shown in Figures 2.1 through 2.5.
1
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FIGURE 2.1: fMTT 2 CYCLE 9 ??F1'PDJCF TI%fYfm ramN i
A B
C D
E F
G H
J K
L M
N P
R 4
ur0 ust v21
- 15 "TiF IT*T
- iUr WO W4 2L62 atte 2L34
.J W57
- 14 atts atis
~ 13 W28 2LO3 2L35 734 v39 YM 2L64 2LOS W11
~
liiff-
"iiir sia 117 ITF
......... 12 W22 f23 2L52 f24 2L26 YOf 2L23 714 2L57 T13 W5 "li]T*
a141-
..... 11 W58 2LO1 2L50 T40 T34 Y27 2L26 Y34 YS6 764 2L56 2 LOT W52
~iTF 7tr iiTF HT F liTF IT F ITF
..... 10 W1 2L30 T32 764 W5 2L14 a34 2L11 W55 T53 til 2L34 W3 a116 a164 4116 "ITF 9
WIT 2L36 Y29 2L25 Y26 2L1T YS2 Y12 767 2LO9 705 2L20 Y07 2L61 U23 TITC "iIF iT F
~IT F "Tir iT F 8
v37 2L60 764 v10 2L29 e64 Y30 2L16 70s R$4 2Lis vis v69 2L63 u67 a164 n164 4134 a119 7
v22 2L42 fie 2L28 fit 2L15 150 704 Y62 :
2L13 Y33 2L27 Y25 2L33 905 W
"iTF
~iTTi'
'iiF "iTF IT F atti 6
W54 2L49 702 Y11 W50 2L12 ast gLie M1 v61 v20 2L31 W10 "ITF R143 ut 2L06 2L55 T55 T65 701 2L22 Y17 T3T f63 2L54 2LOS W51 a165 aiu "iir TiF TiF 4
W35 YZ1 2L53 Y35 2L21 T28 2L19 Y31 2L51 Y22 900 a128
~iTF 3
W44 2LM 2L64 T11 Y54 -
T03 2t45 2LC2 W13 711T" 7TF "iT F 2
W54.
Wie 2L39 2L37 2L32 W59 W53 1
W2s usa u32
-xxx
= INSERT S/N xxx
= EUEL ASSEMBLY S/N
/
\\
NORfN
'Ihe original v/o U-235 enrichments were No. of Puel Assemblies Region.5 (R) assemblies.....
3.402%
Region 5........... 4 Region 8A (U) assemblies....
3.598%
Region 8A.......... 4 Region 8B (U) assemblies....
3.994%
Region 8E.......... 4 Region 9A (W) assemblies....
3.79 M Region 9A.........
16 Region 9B (1.) assemblies.... 4.202%
Region 9B.....'....
16 Region 10A (Y) assemblies...
3.806%
Region 10A........
36 Region 10B- (Y) assemblies... 4.185%
Region 10B........
20 Region 11A (2L) assemblies..
3.600%
Region 11A........
29 Region 11B (2L) asse:tblies.. 4.000%
Region 11B........
28 Total 157 l
2
- - - - - - = -. -. _. _ _
a FIGURE 2.2: Control and Shutdown Rod I.ocations R
P N
M L
K J
H G
F E
D C
8 A
1 2
A D
A 3
C B
P 3
c 5
A B
D C
D B
A 7
'S so-O SP C
SP.
C SP s
A B
D C
D R
'5 C
D A
15 Mgtat:
C' PUNCTION NUMSER OF CLUSTERS Control Bank D 8
Control Bank C 4
Centrol Bank 5 8
Control Bank A 8
Shutdown Bank SS S
Shutdown Bank SA 4
SP (Spare Red Loestions) 13, 3
i.
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-, - ~.... -. -. -. -.. - - - -. -
- -. - - - - ~ - ~ -.
. ~. -.. - - - - -
t FIGURE 2.3: Burnable' Absorber and Sourco Assembly Locations i
R P
N W
I, K
(
H F
ll D
C B
A 1
311 32.1 311 2
3I1 311 3
001
) MI 4884 NI 801 4
t 001 NI 1
MI 5
l 321 481 485 311 4
L l.
3tI MI 481 48I MI 321
-7 i
908 323 4084 MI 481 MI 4M4 328
-8 l-333 MI 481
'48I MI 311
-9 3tt det 443 311 10 et MI 801 11 803 MI 43A MI OSI 12 7'
33I 33I 13 t.
l 333 3RI 331 14 l
15
.08 Typt TOTAL I.. (IRAMR F I FRA A0N)................. 2480
..(MAMR F EDGEMRT SOL 5M N#LETS)...
18 Note - 14eations M 4 and D.8 contain the new dully. compatible Seeendary sources for first ties irradiation in Cycle 9 4
- -. -. -.. =.. - -... -
FIGURE 2.4:
Secondary Source Rod Configurations
,O O 0, O
O O
O O
O O
O O
O O
O O
O O
O O
E Secondary Source Rod 5
Fig. 2.5: Durnable Absorber Configurations i i ii ii
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I 64 IFRA ASSDSLY 80 IrSA ASSOSLY IFBA- " Integral Fuel Burnable Absorber" 6
...2 l
3.0 CONTROL ROD DROP TIME MEASUREMENT (FNP-2-STP-112)
PUPPOSE The purpose of this procedure was to measure the drop time of all full length control rods under hot-full flow conditions in the reactor coolant system to insure compliance with Technical Specification Requirements.
SUMMARY
OF PESULTS For the hot full-flow condition (Tavg 2 541 'r and all reactor coolant pumps operating) Technical Specification 3.1.3.4 requires that the drop time f rom the fully withdrawn position shall be $ 2.7 seconds f rom the beginning of stationary gripper coil voltage decay until dashpot entry.
All full length rod drop times were measured to be less than 2.7 seconds.
The longest drop time recorded was 1.50 seconds for rod B-6.
The rod drop time results for both dashpot entry and dashpot bottom are presented in Figure 3.1.
Mean drop times are summarized below:
TEST MEAN TIME TO MEAN TIME TO CONDITIONS DASHPOT ENTRY DASHPOT BOTTOM Hot full-flow 1.354 sec.
1.827 sec.
To confirm normal rod mechanism operation prior to conducting the rod drop test, the Verification of Rod Control System Operability (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.
During the Rod Cuntrol System Operability test, the bank overlap unit switch settings and functions were verified to be correct. In addition, the actual fully-withdrawn position of each RCCA was measured using stationary gripper coil traces to provide data for planning RCCA repositioning f or future fuel cycles. The current Unit 2 fully withdrawn rod bank position is 231 steps.
During the rod operability test, an urgent failure alarm occurred while attempting to drive the rods powered from the 2AC power cabinet.
A defective circuit card was found and replaced in the cabinet to restore normal operation.
During the rod drop test, it was observed that rod G9 decelerated slightly prior to dashpot entry (although the drop time was a normal 1.30 seconds), and the trace for this rod was reversed.
Westinghouse evaluated the G9 rod drop trace and concluded that, since the trace did not indicate that the rod was binding or hanging up, the rod was accep-table for use.
To verify that rod G9 was falling freely, the drop time was measured a second time as 1.317 seconds, well within the 2.7 second Tech. Spec. limit and f aster than the average rod drop time of 1.354 sec.
The reversed trace was attributed to a reversal in the magnetic polarity of the drive shaft.
This same type of polarity reversal has been observed during previous startups, and has been proven to have no bearing on the drop time.
7 l
-4 C:IT 2 CYC.E 9 i
ACRTH 90 0
\\\\y 1.37 1.37 1.40 1.82, 1.85 1.85 P
\\
1.37 1.35
[
N 1.83 1.83 1.33 1.35 1.33 1.37
-N 1.80 1.83 1.78 1.85 yN 1.37
.1.33 M
L 1.80 1.80 N
1.45 1.33 1.35 1.32 1.33 1.33 1.38 1.93 1.80 1.82 1.85 1.80 1.78 1.87 K
1.35 1.33 1.32 1.33 1.83 1.75 1.77 1.80
-J g
N o
1.37 1.33 1.33 1.35 o
IN
-H 1.83 1.87 1.87 1.80 1.35 1.32 1.35 1.37 1.85 1.73 1.78 1.85
-G 1.35 1.37 1.33 1.33 1.35 1.37 1.35 1
1.87 1.83 1.80 1.87 1.82 1.82 1.82 F
1.35 1.35 1.82 1.82
-E 1.37 1.32 1.32 1.33 1.82 1.78 1.78 1.82 0
1.35 1.37 1.83 1.85 C
1.38 1.42 1.50 1.85 1.90 1.98 8
/
g f
0 270
(
15 14 13 12 11 10 9
8 7
6, 5
4 3
2 i
ORIVE LINE "0 ROP TIME" T ABULATION 35 W TEMPERATUAE -
PRESSURE -
% Flow.
100 5-7-92 X.XX 3REAKER "0PEMING* TO DASHPOT ENTRY - IN SECONOS DATE -
X.XX BREAKER "0PENING" TO DASHP0T BOTTOM - IN SECOMOS FIGURE 3.1 O
4.0 INITIAL CRITICALITY
( FNP- 0-ETP-3 601 )
PURPOSE The purpose of this procedure was to achieve initial criticality under carefully controlled conditions, establish the upper flux limit for the
{
conduct of zero power physics tests, and operationally verify the calibration of the reactivity computer.
SUMMARY
OF RESULTS Initial reactor criticality for Cycle 9 was achieved during dilution mixing at 2212 hours0.0256 days <br />0.614 hours <br />0.00366 weeks <br />8.41666e-4 months <br /> on May 8,
1992.
The reactor was allowed to stabilize at the following conditions:
RCS Pressure 2236.9 psig RCS Temperature 548.9 *F Intermediate Range Power 1 x 104 Amp RCS Boron Concentration 2013.8 ppm Bank D Position 206 steps once criticality was achieved, the point of adding nuclear heat was determined in order to define the flux range for physics testing, and the reactivity computer calibration was verified by making positive and negative reactivity changes and comparing the reactivity indicated by the reactivity computer with values determined using the Inhour Equation.
5.0 ALL-RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT (FNP-0-ETP-3601)
PURPOSE l'
l The objectives of these measurements were to determine the hot, zero power isothermal and moderator temperature coefficients for the all-rods-out (ARO) configuration and to measue the ARO boron endpoint concentration.
SUMMARY
OF RESULTS The ARO, hot zero power temperature coefficients and the ARO boron endpoint concentration are tabulated belows
@O, HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT Boron Measured ITC Design Acc. Calculated Rod Conc.
ITC Criterion MTC Confiouration
.. ppm.
eem/*F fgfff pcm/*F All Rods Out 2078.9*
+3.81
+3.23 1 2
+5.57 i
9 1.
i
.m w
w
+
where:
ITC = Isothermal Temperature coefficient, includes -1.65 pcm/'F Doppler coefficient HTC = Hoderator Temperature Coefficient, corrected to the ARO condition
Rod Conficuration Measured C.
focm1 Desion-oredicted C. form) l All Rods Out 2078.9 2103 +50, -70
]
Since the measured HTC
(+5.57 pcm/*F) was within the Technical Specification limit of
+7.0 pcm/'F, no rod withdrawal limits were 4
required.
The design review criterion for the ARO boron concentration was also satisfied.
1 6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS (FNP-0-ETP-3601)
PURPOSE The objective of the bank worth measurements was to determine the integral reactivity worth of each ecscrol and shutdown bank for 1
comparison with the values predicted by design.
SUMMARY
OF RESULTS The rod worth measurements were performed using the bank interchange method in which: (1) the worth of the bank having the highest design worth (designated as the " Reference Bank") is carefully measured using the standard dilution method; then (2) the worths of the remaining control and shutdown banks are derived from the change in the reference bank reactivity needed to offset full insertion of the bank being measured. The measured worths satisfied the review criteria both for the banks measured individually and for the combined worth of all the banka.
SUMMARY
OF CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS Control or Predicted Bank Shutdown Worth & Review Heasured Bank Percent l
Juh_
Criteria foem)
Worth foemi Difference A
244 1 100 259.5
+6.4 B (Ref.)*
1280 1 128 1309.1
+2.3 C
716 1 107 632.1
-11.7 D
987 1 148-958.0
-2.9 SD - A 975 1 146 1017.8
+4.4 SD - B 205 t 165 1033.7
-5.9 All Banks J04 1 530.1 5210.2
-1.7
- Measured by the dilution method 10
4 7.0 STARTUP AND POWER ASCENSION PROCEDURE (FNP-2-ETP-3605)
PURPOSE The, purpose of this procedure was to provide controlling instructions fors 1.
NIS intermediate and power range setpoint changes, as required prior to startup and during power ascension.
2.
Romp rate limitation and control rod movement recommendations.
3.
conduct of startup and power ascension testing, to include a.
HZP physics tests (FNP-0-ETP-3601).
b.
Incore movable detector system alignment (FNP-2-ETP-3606).
c.
Incore-excore AFD channel recalibration (FNP-2-STP-121).
d.
Core hot channel factor survaillance (FNP-2-STP-110).
e.
Reactor coolant system flo; eanurement (FNP-2-STP-115.1).
SUKMARY OF RESULTS In order to satisfy technical Specification 3.10.3 requirements for invoking speck HZP physics test exceptions, preliminary NIS trip setpoints of less - than or equal to 25% power were used for the inter-mediate and power range channels.
In addition, the intermediate range channel setpoint currents were reduced to address the projected reduction in core neutron leakage from the previous cycle, and to account for re-placement of the M35 detector.
When physics tests were completed, the NIS power range high range-hi h 7
flux trip setpoint was increased to 80% to allow power escalation above 25%.(The 80% setpoint was administrative 1y imposed to address the pos-sibility that the uncalibrated power range channels could be indicating non-conservatively. ) At approximately 33% power, the power range channels were recalibrated and setpoint currents were determined for the inter-mediate range channels.
Following power range channel recalibration, the power range high flux trips were increased from 80% to 109%.
Subsequently, however, as power was increased, it was found that the loop ATs were indicating about 3%
power higher than calorimetric power.
(The control board percent AT indicators were still calibrated to the cycle-8100% loop ATs; the cycle-c 9 loop ATs were not measured until 78% power.) Therefore, for conserva-tism, the power range NIS high flux trips were reduced from 109% to 100%
until the AT channels could be rescaled.
Subsequently, the trips were further reduced to below 55% due to the apparent QPTR problems described in Section 8.0.
Once it was resolved that the QPTRs were satisfactory, the power range high flux trips were 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 sus-tained for 72 cumulative hours out of any seven-day operatino period.
This ramp rate was observed during the ascension to full power.
11 1
I 1
The determination of movable detector core limit settings (FNP-2-ETP-3606) was performed during the ascension to 33% power. The incore-excore recalibration (FNP-2-STP-121) was performed at 33% power and completed at 100% (see Section 8.0).
The RCS flow measurement (FNP-2-STP-115.1) was performed at 78% power and was repeated at 100% to refine the flow measurement (See section 9.0).
This procedure was also used at 78%
power, and was repeated at 100% power, to determine the RCS loop 100% ATs for scaling the OTAT and OPAT protection channels.
As summarized in Table 7.1, core hot channel factor surveillance was initially performed under non-equilibrium conditions using the incore-excore base case full core flux map taken at 33% power, and then under equilibrium conditions using full-core flux maps performed at 33%, 48%, and 100% power.
TABLE 7.1
SUMMARY
OF POWER ASCENSION FULL CORE FLUX MAP DATA Parameter Fuel Tvoe Mao 219 Mao 226 Mao 227 Mao 228 Avg. t I)wer N/A 33%
33%
48%
100%
Max FDH Lopar 1.5166 1.4924 1.4940 1.4517 Vantage 5 1.6364 1.6256 1.6293 1.5773 Max power tilt
- N/A 1.0071 1.0052 1.0031 1.0032 Avg. core t A.O.
N/A 11.175 11.430 11.220 5.000 Limiting FQ(2)"
Lopar 2.1070 2.0711 1.9906 1.8395 Vantage 5 2.2671 2.2324 2.1627 1.9928 FQ Limit Lopar 4.5467 4.5467 4.5053 2.2993 vantage 5 4.8125 4.8015 4.7577 2.4226 Flux map N/A Non-equi-Equi-Equi-Equi-conditions librium librium librium librium
- Calculated power tilts based on assembly FDHN from all assemblies.
" Based on percent to FQ limit.
Fuel types referenced above are low parasitic (Lopar) fuel (100 assemblies), and vantage 5 fuel (57 assemblies).
8.0
.INCORE-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 of f set 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.
l~
L 12 1
I
e
SUMMARY
OF RESULTS At an indicated power of approximately 33%, a full core base-case flux map was performed but, before the test could be continued, the reactor tripped due to control power fuses blowing on two of the NIS power range channel drawers.
However, it was still possible to utilize the (non-equilibrium) full-core flux map to perform a tentative evaluation of the core hot channel f actors, which were satisf actory (See Table 7.1). When power was restored to 33%, the base case map was repeated as a quarter core flux map.(since the hot channel factors had been evaluated).
Five additional quarter-core flux maps were performed at various positive and negative axial of fsets to develop equations relating detector current to incore axial offset.
The power range channels were adjusted to incorporate the revised calibration data.
In order to improve the delta flux calibration accuracy of the NIS chan-nels, additional data was collected and used. In addition to reading the NIS detector currents from the NIS drawer panel meters during the incore-excore test, more accurate detector current measurements were also obtained using a digital voltmeter (DVM).
The DVM data were used for calculation of channel calibration curreats and a DVM (rather than the panel meters) was used to read the test currents input for NIS drawer calibration. The DVMs were connected across the precision test resistors that are wired in series with the detector current input meters.
The initial test setup consisted of eight DVMs connected with unshielded
)
test leads to the upper and lower detector current inputs of the four NIS power range drawers. Subsequent investigation of the reactor trip refer-enced above demonstrated that this tent setup was the cause of the trip:
Thermal power was so close to the 35% P8 and P9 satpoints that the 60 Ha pickup from the unshielded test leads entered the drawers and caused the P8 and P9 bistables to rapidly vibrate on and off at the 60 Hz rate.
This rapid switching action, reflected to the control power through transformer coupling, overheated the drawer control power fuses, even-tually causing an N41 control power fuse to blow. Shortly afterward, in drawer N43, a second control power fuse blew, rseulting in a reactor trip.
As a result of this event, subsequent DVM measurements were performed in only one NIS drawer at a time using shielded test leads to reduce 60 Hz pickup.
The detector currents read visually from the panel meters were used to derive data for use in manual QPTR calculations.
The detector currents read by-DVH were used to develop data used for NIS channel calibration and surveillance testing.
When power was escalated to about 51%, QPTR calculations derived from panel meter readings gave a result of 1.116, so the power. range NIS setpoints were reduced to comply with the Tech.
Spec. Action for a QPTR exceeding 1.09.
However, it was found that the 1.116 QPTR was the result of a spurious indication error caused by a high contact resistance in the meter range switch.
Exercising the range switch lowered this reading, but calculations from panel meter data continued to border or si' Jhtly exceed the 1.02 Tech Spec limit for QPTR.
QPTR calculations made from DVM data never exceeded 1.003.
An incoro l
flux map was run at 48% pcwer ( Table 7.1 ) which proved that the core l
hot channel factors had remained well within limits and that the incore tilt values were consistent with the QPTR calculations f rom the DVM data.
It was concluded that the high QPTRs calculated from panel meter data were a result of (1) the accuracy level of the meter ( = 2% ),
(2) the sensitivity of the QPTR calculation to small changes in current
- data, 13
l l
l (3) lower detector currents from the extrems_f 1.ow leakage core pattern for Cycle 9, and (4) range switch contact resistance.
QPTR was monitored as power was increased.
As excore detector currents increased with increasing power, the QPTR calculated from panel meter data started to consistently fall within the Tech Spec limit.
Survo111ance procedures were changed to allow the use of DVM dt.ta to calculate QPTRs as an alternative to the data taken from the panel meters.
Once equilibrium 100% power conditions were achieved, a full core flux map was performed to verify that the core hot channel f actors were satis-factory and to refine the delta-flux calibration of the power range NIS Channels.
Prior to readjusting the NIS channels, an evaluation of the existing calibration (performed at 33% power) using data measured by DVM yielded a maximum calculated QPTR of 1.0056 (consistent with the change in incore tilt), and an agreement within 0.6% between core axial offsets calculated from detector currents and the flux map incore AO.
The detector current vs Ao equations derived f rom 100% power DVM data (in which both the slopes and zero-offset currents were revised to account for core leakage changes between 33% and full power) are tabulated below:
TABLE 8.1 DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHANNEL N411 0.8130 *AO
+
144.23 uA I-Top
=
-0.8327 *AO
+
142.37 uA I-Sottom
=
,Q]ihBNEL N42:
I-Top 0.8342 *AO
+
147.00 uA
=
-0.8602 *AO
+
141.55 uA I-Bottom
=
CHANNEL N43:
I-Top 0.8201 *AO
+
149.81 uA
=
-0.8662 *AO
+
144.88 uA I-Bottom
=
CHANNEL N44:
0.9228 *AO
+
162.24 uA I-Top
=
-0.9960 *AO
+
160.56 uA I-Bottom
=
9.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (FNP-2-STP-115.1)
PURPOSE The purpose of this procedure was to measure the flow rate in each reactor coolant loop in order to confirm that the total core flow met the minimum flow requirement given in the Technical Specifications.
14
t A-t s
-t L
SUMMARY
OF RESULTS The Unit 2 RCS flow measurement was initially performed at 78% power, primarily for the purpose of measuring the RCS loop ATa for rescaling the OPAT and otAT protection channels.
Previously, the main control board percent AT indicators had been calibrated to the Cycle-8 100% loop ATs and, for conservatism, the OPAT and oTAT protection setpoints had been adjusted to 90% of the Cycle-8 loop ATs.
Twelve sets of calorimetric data and RCS spare T-hot and T-cold RTD readings were obtained, averaged and extrapolated to 100% power to yield the following 100% loop ATs:
Loop 1 (TE-412): 100% AT = 66.546 'F Loop 2 (TE-422): 100% AT = 69.363 'F Loop 3 (TE-432): 100% AT = 65.307 'F Tha average of these ATs is 2.47
'F (about 3.8% AT) higher than the Cycle-8 AT average, which accounts for the disagreement between percent AT and calorimetric power noted du ing power ascension to 78%. The OPAT-and OTAT protection loops were recalibrated to these ATs prior to escal-ating power to 100%.
'The RCS flows calculated from the 78% power flow test data were Loop 1 92,502 gpm Loop 2:
88,624 gpm Loop 3:
92,897 gpm The total. core flow was 274,023 gpm, which satisfies the 267,880 gpm minimum requirement given in the Unit 2 Technical Specifications.
-When equilibrium, 2004 power conditions were achieved, the RCS flow measurement.was repeated simultaneously with the 100% power flux map-described in Section 8.0.
Twelve sets of calorimetric data and RCS spare
-T-hot and T-cold RTD. readings were obtained, averaged and normalized to j
100.0% power to yield the following RCS flows:
Loop is-93,256 gpm Loop 2
-89,150 gpm -
Loop 3:
93,280 gpm The total core flow was 275,687 gpm, which satisfies the 267,880 gpm minimum requirement given in the Unit 2. Technical Specifications.
In addition.to nmasuring RCS flow, the~1oop ATs for 100.0% power were also determined and the final calibration of the CPAT and OTAT protection Joops was performed. The revised loop 100% ATs were:
. Loop.1 (TE-412)i 10Cn AT = 64.627 'F Loop 2 (TE-422): 100% AT = 67.424 'F Loop 3 (TE-432):--100% AT = 63.349 'T These values-are close to the original Cycle 8 100% loop ATs (63.894
'F,-
67.011 'F, Land 62.901='F for loops 1,-2,-and 3, respectively).-
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