ML20125E582
| ML20125E582 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/10/1985 |
| From: | Mcdonald R, Morey O ALABAMA POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8506130167 | |
| Download: ML20125E582 (21) | |
Text
,'
ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER 2, CYCLE 4 STARTUP TEST REPORT PREPARED BY PLANT REACTOR ENGINEERING GROUP' APPROVED:
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Technical Superintendent (f r Plant Manager p;{6
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p)l:,ijd 8506130167 850610 DISK:
CYCLE 2 fDR ADOCK 050 364 QG Vi
TABLE OF CONTENTS PAGE 1.0 Introduction 1
2.0 Unit 2 Cycle 4 Core Refueling 2
3.0 Control Rod Drop Time Measurement 7
4.0 Initial Criticality 9
5.0 All-Rods-Out-Isothermal Temperature Coefficient, Boron Endpoint and Flux Distribution 10 6.0 Control and Shutdown Bank Worth Measurements 12 7.0 Power Ascension Procedure 14 8.0 Incore-Excore Detector Calibration 16 9.0 Reactor Coolant System Flow Measurement 18
1.0 INTRODUCTION
The Joseph M. Farley Unit 2 Cycle 4.Startup Test Report addresses the tests performed as re plant procedures following core refueling. quired by The report
-provides a brief synopsis of each test and gives a comaarison of measured parameters with design predictions, Tec1nical Specifications, or values assumed in the FSAR safety analysis.
Unit 2 of the Joseph M. Farley Nuclear Plant is a Westinghouse three loop pressurized water reactor rated at 2652 MWth.
The cycle 4 core loading consists of 157 17 x 17 fuel assemblies.
The Unit began commercial operations on July 30, 1981, completed Cycle 1 on October 22, 1982 with an average core burnup of 15350.5 MWD /MTU, completed Cycle 2 on September 17, 1983 with an average core burnup of 10371.2 MWD /MTU, and completed cycle 3 on January 4, 1985 with an average core burnup of 14,639.0 MWD /MTU.
1
2.0 UNIT 2 CYCLE 4 CORE REFUELING REFERENCES 1.
Westinghouse Refueling Procedure FP-APR-R3 2.
Westinghouse WCAP 10674 (The Nuclear Design and Core Management of the Joseph M. Farley Unit 2 Power Plant Cycle 4)
The refueling commenced on 1/12/85 and was completed in 10 days on 1/22/85.
The as-loaded Cycle 4 core is shown in Figures 2.1 through 2.4, which give the location of each fuel assembly and insert, includin locations and confi p rations.g the burnable poison insert The cycle 4 core has a nominal design lifetime of 15240 MWD /MTU and consists of 21 region 4 assemblies, 64 region 5 assemblies, and 72 region 6 assemblies.
Fuel assembly inserts include 48 full length control rod clusters, 40 burnable poison inserts, two secondary sources, and 67 thimble plug inserts.
i 2
Figura 2.1 APR Umt 2, Cycle 4 Referenco Loading Pattern 180' R
P N
M L
K J
H G
F E
D C
B A
l R-35 S-06 R-37 J-4 F
G-4 P-39 S-30 S-35 R-03 S-53 S-55 P-19 2
J-10 F
F L-3 F
F G-fo Rg40 P-48 S-O'3 S-67 R-60 R-22 S-13 S-46 P-28 3
J-15 F
F J-2 H-15 0-2 F
F G-15 A
R d2 S-44 R-S-29 R-N3 S-32 S-01 P-35 4
P-32 S-48 S-40 A-7 F
F K-83 F
L.
F F-3 F
F R-7 P-45 S-52 S-27 P-07 S-15 R-45 S-61 R-24 S-59 P-52 S-50 S-72 P-37 5
F-7 F
F P-8 F
K-5 F
F-5 F
H-2 F
F K-7 S-49 S-45 R-38 S-57 R-29 R-59 R-fI R-28 R-30' S-14 R[d3 S-63 S-12 6
F F
C-6 F
N-5 K-14 H-7 F-14 E-3 F
N-5 F
F R-55 S-31 R-49 S-56 R-64 R-50 R-O'l S-38 R-26 R-Si R-42 S-60 R-13 S-62 R-28 7
W-7 F
P-7 F
L-4 S-6 M-5 F
E-4
.P-8 E-6 F
S-7 F
D-7 90*
S-71 R-58 R-63 R-34 S-07 R-52 S-70 P-I5 S-08 R-54 S-22 ReI4 R-33 R-53 S-Il 8
270*
F N-Il A-9 W-!!
F
'J+5 F
H-il F
4*G F
D-5 R-S C-5 F
f R-82 5-23 R-10 5-37 REUI RA R[17 S-36 R-0' 9' R-20 RSIS S-16 R-06 S-54 R-I9 g
M-9 F
P-9 F
L. O S4 L-12 F
D-18 P-TO E- 0 F
8-9 F
D-9 S-42 S-64 R-47 S-39 R-04 R-27 R-07 R-48 R-OS S-17 R-32 5-29 S-10 o
F F
C-lO F
L-13 K-2 H-9 F-2 C-Il F
N-10 F
F P-05 S-09 S-19 P-27 S-51 R "8 S-65 R-25 S-20 P-24 S-04 S-02 P-51 ii F-9 F
F H-I4 F
KI F
F-il F
B-8 F
F K-9 P-41 S-26 S-69 R-59 S-66 R-IS S-68 R-12 S-25 S-24 P-22 12 A-9 F
F N-3 F
E-12 F
F-3 F
F R-9 Rg05 R-d6 S-34 S-10 P-49 33 P-02 5-47 S-43 R-30 J-l F
F J-84 g.g G-14 F
F G-l P-38 S-05 S-58 R-SI S-41 S-33 P-33 g4 J-6 F
F E-13 F
F 0-6 I
R-44 S-21 R-57 65 J-12 F
G-12 O'
ID 20 ID COLOR CCOE LEGENO:
REGION 4
5 6
10 - ASSEMBLY ICENTIFICATION i
FRou CYCLE 3
3 5EED SS - SECCtJOARY SOURCE LOCATICt1 T-u-22S
'3.100 3.40Z 3.45 xx PREvICUS CYCLE
. c c T IerJ 3
FIGURE 2. 2 CONTROL ROD LOCATIONS R
P N
M L
K J
H G
F E
D C
B A
I A
D A
C 8
8 C
SP 4
I A
8 D'
! C D
8 A
6 S
S 5
S A
a 8
sp A
7 D
SP C
spj C
sp D
8 S
S S
S A
spi S
e g
9 A
B D
C D
9 A
10 5
0 9
SP 8
SP II C
B 8
C 12 3p Spl S
S A
A I3 A
D A
I4 Absorber 'Aaterial:
Ag-In-Cd 15 i
Function Number of Clusters Centrol Bank 0 Control Bank C 8
Control Bank B 8
Control Bank A 8
Shutdown Bank SB 8
Shutdown Bank SA 8
SP (Scare Rod Locations) 8 13 4
FIGURE 2-3 BURNABLE ABSORBER AND SOURCE ASSEMBLY LOCATIONS R
P N
M L
K J
H G
F E
D C
B A
I 2
8 SS 8
3 12 12 12 12 4
12 12 12 12 12 5
8 12 12 8
6 12 8
12 7
12 8
8 12 8
12 8
12 9
8 12 12 8
'~
12 12 12 12 12 II 12 12 12 12 12 8
ss 8
13 14
^
15 SS Secondary Source 432 Fresh Standard BA's
]
5
FIGURE 2-4 BURNABLE ABSORBER CONFIGURATIONS O
O O
g g
0 0
E O
O O-O E
O O
O E
O O
-O O
O 8 Fresh BA Configuration E
0 O
0 E
O E
O E
O E
E O
E O
E O
E E
O E
12 Fresh BA Configuration 6
i 3.0 CONTROL. ROD DROP TIME MEASUREMENT (FNP-2-STP-112)
PURPOSE The purpose of this test was to measure the drop time.of all full length control rods under hot-full
. flow conditions ~in 2e reactor coolant system.to ensure compliance with Technical Specification requirements.
SUMMARY
-OF RESULTS For the Hot-full flow condition (T
> 541*F.and all reactor coolant pumps oaerating) TeeKHical Speci-fication 3.1.3.4 requires tiat the rod drop time from the fully withdrawn position shall be < 2.2 seconds from the beginning of stationary gripper coil voltage decay.until dashpot entry.
All full length rod dro?
times were measured to be less than 2.2 seconds.
Tae longest drop time recorded was 1.47 seconds for rod-B-6 and K-14. -The rod droa time results for both dashpot entry and' dashpot 3ottom 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 1.38 sec 1.89 sec To confirm normal rod mechanism operation prior to conducting the rod drops, a Control Rod Drive Test (FNP-0-IMP-230.3) was performed.
In the test, the stepping waveforms of d e stationary, lift and moveable gripper coils were examined, and the functioning of the Digital Rod position indicator and the bank overlap unit was checked.
Rod stepping speed measurements were con-ducted.
All results were satisfactory.
h 7
UNIT 2 CYCLE 4 900 R
\\
1.40
- 1. 3[
1.40
~
1.82 1.89 1.87 P
1.38 1.35 a
N 1.89-1.90
[
I
'1.37 1.38 1.38
'1.36 M
1.85 1.88 1.87 1.87,
[L 1.38 1.37 1.91 1.91 x
~. 3 8' 1.43 1.47 1.38 1.35 1.36 1.35 1
g 1.96 1.87 1.87 1.87 1.88 1.89 1.91 1.32 1.33 1.36 1.38 1.85 1.86 1.91 1.91 E
]
0 1.41 1.39 1.36 1.37 0
180
-N 1.94 1.91 1.88 1.89 1.38 1.34 1.37 1.42
_g 1.86 1.83 1.88 1.93-1.43 1.39 1.38 1.34 1.39 1.36 1.41 1.90 1.90 F
1.89 1.91 1.92 1.85 1.91 1.35 1.38 1.85 1.88 E
1.38 1.36 1.33 1.37 0
1.82 1.87 1.86 1.85 1.40 1.38 1.90 1.89 C
1.41 1.42 1.47 i
1.88 1.92 2.04 8
s g
270'
[
(
h 15 14 13 12 11 10 9
8 7
4 5
4 3
2 i
ORIVE LINE "0 ROP TIME" TA80LAT10N 5470F 2235 psig 100 TEMPERATURE -
PRESSURE -
g pg,
X.XX BREAKER "0PENING" TO DASHPOT ENTRY - IN SECONOS DATE -
3-7-85 X.XX BREAXER "0PENING" TO DASHPOT 80TTOM - IN SECONOS FIGURE 3.1 8
j 4.0 INITIAL CRITICALITY (FNP-2-ETP-3601)
PURPOSE The purpose of this procedure was to achieve initial reactor 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
0F RESULTS Initial Reactor Criticality for Cycle 4 was achieved during dilution mixing at 0735 hours0.00851 days <br />0.204 hours <br />0.00122 weeks <br />2.796675e-4 months <br /> on March 8 1985 The-reactor was allowed to stabilize at the following critical conditions:
RCS pressure-2235 psig, RCS temperature 548.0 F, intermediate range power 2 x 10-8 amp, RCS boron concentration 1900 ppm, and.Contr61 Bank D position-181.5 steps.
Following stabilization, the ooint of adding nuclear heat was determined and a checiout of the reactivity computer using both positive and negative flux periods was successfully accomplished.
In addition, source and intermediate range neutron channel overlap data were taken during the flux increase preceding and immediately following initial criticality to demonstrate that adequate overlap existed.
9
~
5.0 ALL-RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT, BORON ENDPOINT AND FLUX DISTRIBUTION ~(FNP-2-ETP-3601)
-PURPOSE The objectives of these measurements were to:
.(1)-determine the hot, zero power isothermal and moder-ator temperature coefficients for the all-rods-out,(ARO) configuration; (2)) measure the ARO boron endpoint con-centration;bution(3. determine the hot, zero power ARO and flux distri in the reactor core. (optional)
SUMMARY
OF RESULTS The measured ARO, hot zero power temperature co-efficients and the ARO boron endpoint concentration are shown in Table 5.1.. The isothermal temperature coeffi-cient was measured to be +1.22 pcm/*F which meets the design acceptance criteria.
This gives a calculated moderator temperature coefficient of +3.92 pcm/*F which is within the Technical Saecification= limit of +5.0 pcm/*F.
Thus, no rod witidrawal limits are needed to~
ensure the +5.0 pcm/*F limit is met.,The design acceptance criterion for the ARO critical boron con-
-centration was also satisfactorily met.
Flux distribution was determined by the performance of a flux map at 34% power (see section 7.0 for results).
4 10
s TABLE 5.1 ARO, HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT-Rod Configuration Boron Measured Calculated a Design Acceptance T
Concentration aT
" mod Criterion ppe pcm/*F pcm/*F pcm/'F All Rods Out 1909.5
+1.22
+3.92
+0.6 3
T
- Isothermal temperature coefficient, includes -2.7 pcm/'F doppler coefficient a
a
- M derater only temperature coefficient nod ARO,'HZP BORON ENDPOINT CONCENTRATION Rod Configuration Measured CB (ppa)
Design predicted CB (ppa)
All Rods Out 1900 1911 i 191.1 9
11
6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS (FNP-2-ETP-3601)
PURPOSE The objective of the bank worth measurements was to determine the integral reactivity worth of each control and shutdown bank for 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; and (2) the worths of the remaining control and shutdown banks are derived,from the change in reference bank reactivity needed to offset full insertion of the bank being measured.
The control and shutdown bank worth measurement results are given in Table 6.1.
The measured worths satisfied the review criteria both for the banks measured individually and for the combined worth of all banks.
12
~
~
./
a,;
r i
,. 7, TABII,,6.1
SUMMARY
OF CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS Predicted Bank Measured Worth & Review Bank Percent Bank Criteria (pcm)
Worth (pcm)
Difference
'~
Control A 622 i 100 598~.8
-3.7 Control B (Ref.)
1308 131 1223.d*,
-6.4 Control C S 1 1 140 885.4
-5.1 Control D 983 147 955.6
-2.3 Shutdown A 1113 1 167 1051.6
-5.5 Shutdown B 900 135 833.3
-7.4 All Banks Combined 5859 585 5548.5
-5.3
- Measured by dilution method l
l 13
'[m 42 7.0 POWER ASCENSION PROCEDURE (FNP-2-ETP-3605)
PURPOSE The purpose of this procedure was to provide controlling instructions for:
5 1.
Ramp rate and control rod movement limitations 2.
Incore movable detector system final alignment 3.
Flux map at less than 50% power 4.
Adhering to the delta flux band during ascension to 75% power 5.
Incore/Excore calibration at 75% power.
J
SUMMARY
OF RESULTS Westinghouse fuel warranty provisions recommends 3
that the power ramp rate be limited to 3% of full power per hour between 20% and 100% power until full power is
-2 achieved for 72 cumulative hours out of any seven-day operation period.
This ramp rate was followed through-4s out the ascension to 100% power except for 3 occasions when the indicated NIS power went up between 3 and 4%
}
in one hour.
Alignment of the incore movable detector system
)
normal, calibrate and emergency paths was accomplished during power ascension (at power levels above 5%)
=
prior to performing the flux map at 34% power.
Full flux maps were taken at 34%, and 78% power.
As summarized in Table 7.1, all results were within i
Technical Specification Limits.
j An incore/excore calibration check was performed 9
at 34% power.
Channel N44 currents were found to be out of calibration-and channels N42 and N44 showed QPTR's >1.02, therefore a correction to the incore-6 excore calibration data was performed and revised J
currents were issued for calculating AFD (STP-7.1) and QPTR (STP-7.0).
At approximately 75% power, a complete incore-excore recalibration was performed to comply with Technical Specification requirements.
2 The incore-excore recalibration is described in Section
=
8.0.
5 1
2 s
-i 4
1 14 5
e Y
TABLE 7.1
SUMMARY
OF POWER ASCENSION FLUX MAP DATA Parameters Map 100 Map 101 Date 3/22/85 3/25/85 Time 10:30 11:00
' Avg.'% Power 34%
78%
Max FAH 1.4888 1.4114 Max.- Power Tilt
- 1.0093 1.0055 Avg. Core % A. O.
+9.579
+7.008
{
l
- Calculated power tilts based on ass mbly FAHN from all assemblies.
e s
15 ig
. 7
'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 incore axial offset for the purpose of calibrating the delta flux penalty to the overtemperature AT protection system, and for calibrating the control board and paant computer axial flux difference (AFD) channels.
SUMMARY
OF RESULTS A preliminary verification of excore AFD channel calibration was performed at 34% power to insure that an AFD target band could be defined for ascension to 78% power.
Flux' maps for incore-excore recalibration were run at approximately 78% power at average percent core axial offsets of +7.008, -12.182, -17.478 and
+14.324 as determined from the INCORE code printouts.
The measured detector currents were normalized to 100% power, and a least squares fit was performed to-obtain the linear equation for each top and bottom detector current versus core axial offset.
Using these eguations, detector current data was generated and utilized to recalibrate the AFD channels and the delta flux penalty to the overtemperature AT setpoint.
(See Table 8.1) p 16
m TABLE 8.1 DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHANNEL N41:
I-Top 0.8465*AO +
198.7294 pa
=
I-Bottom
-1;3662*AO +
200.7462 pa
=
CHANNEL 42:
I-Top 0.8655*AO +
194.4327 pa
=
I-Bottom
=
-1.4081*AO +
199.3029 pa CHANNEL N43:
I-Top 0.8296*AO +
203.0991 pa
=
I-Bottom
=
-1.5133*AO +
201.7735 pa CHANNEL N44:
I-Top 0.9614*AO +
219.2990 pa
=
I-Bottom
=
-1.5816*AO +
225.0805 pa 17
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 Unit 2 Technical Specifications.
SUMMARY
OF RESULTS-To comply with the Unit 2 Technical Specifications, the total reactor coolant system flow rate measured at normal operating temperature and pressure must equal or exceed 265,500 gpm for three loop operation.
From the average of six calorimetric heat balance measure-ments, the total core flow was determined to be 281,443.2 gpm, which meets the above criterion. -
18
.m:
.W:
{ Memng Address :-
Alabama Power Company 3 -
600 Norm II:n Street -
- Post Office Box 2641
. Birmingham, Alabama 35291
- Telephone 205 7834090
- R. P Mcdonald.
n2'n'80fo*"'
AlabamaPower d
y tresu me+ac nstem June 10,1985
' Docket No. 50-364 Director, Nuclear Reactor Regulation
'U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention: Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Unit 2 Cycle 4 Startup Report Gentlemen:
Enclosed is the Startup Report for Unit 2 Cycle 4 as required by the December 18, 1984 letter from Mr. R. P. Mcdonald to Mr. S. A. Varga.
If you have any questions, please advise.
Yours very t uly,
/
l./
R. P. McDona d RPM /MDR:cl,D-5 Enclosure cc: fir. L. B. Long Dr. J. N. Grace Mr. E.-A. Reeves Mr. W. H. Bradford i
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