ML20148B453
| ML20148B453 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/04/1988 |
| From: | ALABAMA POWER CO. |
| To: | |
| References | |
| NT-88-0086, NT-88-86, NUDOCS 8803210521 | |
| Download: ML20148B453 (21) | |
Text
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4; AIABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER 2, CYCLE 6-STARIUP TEST REPORT PREPARED BY PIRE REACIOR ENGINEERING GROUP P
APPRCNTD:
[.O. Technical Manager J-4-2B hs ,W /Sp (f neral Manager - Nuclear Plant DISKt CYCLE 2/7 ,
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TABLE OF CGrrDrrS PAGE 1.0 Introduction 1
-2.0 Unit 2 Cycle 5 Core Refueling 2 3.0 Control Rod Drop Time Measurement 7 4.0. Initial criticality - 9 l
l 5.0 All-Rods-Out Isothermal Temperature coefficient and Boron Endpoint 10 6.0 control and Shutdown Bank Worth Measurements 12 7.0 Startup and Power Ascension Procedure 14 8.0 Incore-Excore Detector Calibration 16 9.0 Reactor Coolant System Flow Measurement 18 l
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1.0 INTRODUCTION
he Joseph M. Farley Unit 2 Cycle 6 Startup Test Report addresses the tests performed as required by plant procedures following core refueling. he report provides a brief synopsis of each test and gives a comparison of measured parameters with design predictions, Technical Specifications, or values assumed in the FSAR safety analysis.
Unit 2 of the Joseph M. Farley Nuclear Plant is a h ree Loop ,
Westinghouse pressurized water reactor rated at 2652 !sth. he cycle 6 core loading consists of 157 17 x 17 fuel assemblies.
'Ihe Unit began comercial operations on July 30, 1981, completed cycle 1 on October 22, 1982 with an average core burnup of 15350.5 MWD /MIU, completed Cycle 2 on September 17, 1983 with an average core burnup of 10371.2 MND/MIU, completed Cycle 3 on January 4,1985 with an average core burnup of 14,639.0 MND/MIU, completed Cycle 4 on April 4, 1986 with an average core burnup of 13,183.8 MWD /MIU, and completed Cycle 5 on October 2, 1987 with an average core burnup of 16,674 MWD /MIU.
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2.0 UNIT 2 CYCLE 6 CORE REFUELING REFERENCES
- 1. Westinghouse Refueling Procedure FP-APR-R5
- 2. Westinghouse WCAP 11542 (%e Nuclear Design and Core Management of the Joseph M. Parley Unit 2 Power Plant Cycle 6)
Unloading of the Cycle 5 core into the spent fuel pool comenced on 10/14/87 and was completed on 10/17/87. During a routine underwater
'IV inspection below the lower core plate of the reactor vessel (Procedure FNP-0-ETP-3640), a thermal sleeve from.the B-loop cold leg injection line was found lodged in the vessel lower internals. We lower internals were pulled and the thermal sleeve was removed prior to reloading the core.
Inspection of the Cycle-5 fuel assemblies unloaded into the spent fuel pool (Procedure FNP-0-ETP-3636) disclosed that assemblies S-02, S-57,
> S-06 and S-33 had suffered grid damage. W e damage to S-02 and S-57 did not impact the Cycle-6 core design because S-02 was scheduled for discharge and the problem with S-57 (a slight dent on grid No. 3) did not preclude re-use of the assembly. However, damaged assembly S-06 was replaced with assembly S-55, and damaged assembly S-33 was replaced with assembly S-05 in the Cycle 6 core. To obtain more favorable peaking factors, the substitute assemblies were not put into the core M ations planned for assemblies S-06 and S-33, but were placed else:where; and the two assemblies displaced by substitute assemblies S-55 and S-05 were put in the locations scheduled for S-06 and S-33. his resulted in four changes to the original core design even though only two substitutions were made.
Core reload comenced on 10/22/87 and was completed on 10/29/87. We as-loaded Cycle-6 core is shown in figures 2.1 through 2.4, which give the location of each fuel assembly and insert, including burnable poison insert locations and configerations. We Cycle-6 core has a nominal design lifetine of 16800 MND/mU and consists of 25 region 6 assemblies, 68 region 7 assemblies, 36 region 8A assemblies, and 28 region 8B assemblies. Fuel assembly inserts consist of 48 full length control rod clusters, 56 wet annular be nable absorber inserts (WABAs), two secondary sources, and 51 thimble plug inserts.
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e e p.we D FIGURE 2.1: UNil 2 CYCLE 6 REFERENCE LOADING PATERN H J K L M N P R A B C D E F G 235 293 250 - ------ ------ ------ ------ -- --- -------15
$66 147 S68 201 Ril4 229 R146 287 R139 239
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$55 140 U40 161 U62 154 S57 217 4W1 12W39 R125 16W14 R101 12W28'4W7 202 - ------ ------
13 S05 U42 059 139 051 157 U54 V64 S46 233 R 09 12W33 R132 12W19 SS04 12W36 R120 '12W17 R116 218 - ------
12 i8 U34 117 004 131 035 164 031 101 S42
$10 257 4W4 12W34 R137 12W26 207 254 242 12W50 R145 12W43 4WS 249
11 U29 T28 U15 T20 144 156 U26 146 U12 060 $11
$51 U55 ...... ...... ...... ...... ......
R142 12W49 R104 12W41 R121 16W24 R144 16W23 R130 12W51 R134 12W35 Ril3 -
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...... .. !!.. .!!!.. .!!!.. .!!!.. .!!!.. .!!!.. .!!!.. .!![.. .!!!.. ,,,,,,
-- 9 R122 258 12W52 R107 291 216 232 211 All8 12W45 730 16W16 R131 12W18 16W17 S60 U44 135 U18 115 U21 142 U36 127 D22 T02 U28 118 1 056 $56 210 261 16E13 R102 275 -- 8 205 R!li 16W20 226 273 R112 12W27 220 12W31 R105
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.!!!.. .!!b.. .!.... .!$... .!!!.. .!!I.. .!.g,, ,,7,, ,ggg,, ,{p,,, ,(({,, ,}},,, 3 7 3 7 R103 12W40'223 16W22 R143 12W48 Riel 16W18 221 12W32 R148 274 256 -- 7 270 237 S16 045 105 U09 155 U23 il9 001 T25 U13 114 U10 103 U50 S37 R108 12W46 R106 12W25 R127 16W15 R135 16W21 R124 12W44 R126 12W37 R147 --------- 6 132 U16 163 U20 T48 U19 T22 U53 134 I41 U49 160 U17 244 4W2 12W22 R138 12W20 278 264 214 12W23 Ril5 12W29 4W8 203 ---- ---- 5 006 123 U33 138 109 107 002 133 U25 U41 559
$71 U52 ...... ...... ...... ...... ......
R136 R123 12W38 S$03 12W42 R117 12W21 R140 260 ------- -------- 4 283 12W24 S12 124 U24 153 U14 166 Ull T37 U03 112 S49 272 4W3 12W30 R110 16W19 R133 12W47, 4W6 251 ----------------------- 3
$52 U63 U38 145 U58 104 057 U48 S30 285 R129 263 R128 262 Rl;9 269 ----------------------------- 2
$17 T08 046 1 50 U39 T21 521 296 290 740 ------------------------------ ---- -------- 1 S44 168 S29
(--- FUEL ASSEMBLY INSERT SERIAL NUMBER North (--- FUEL ASSEMBLY SERIAL NUMBER g Theoriginalw/oU-235enrichsentswere: Reg.on6(S)assesblies . 3.443 Reg .on7(ilassemblies 3.603: Reg,on8A(U)assentlies . 3.506: Region 65 (U) assemblies 3.9941 3
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FIGURE 2.2 CONTROL ROD LOCATONS R P N M L K J H G F E D C B A 1 2 A D A 3 SA SA SP 4 C B SP B C 5 SB SP SB 6 A B D C D B A 7 SA SB SB SP SA 8 D SP C SP C SP O I 9 SA SP SB SB SA 10 A B D C D B A 11 SB SP SB 12 C B SP B C 13 SP SA SA 14 A D A 15 Absorber Mgtgrial: FUNCTION NUMBER OF CLUSTERS Control Bank D 8 Control Bank C 8 Control Bank 8 8 Control Bank A 8 Shutdown Bank SB 8 Shutdown Bank SA 8 SP (Spare Rod Locations) 13 LOCATIONS N5 & C11 = CORE WATER LEVEL THERMOCOUPLE PROBES 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 1 2 4 12 16 12 4 3 , 12 12 SS 12 12
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4 12 12 12 12 4 5 12 16 16 12 12 6 12 12 16 12 16 12 7 16 12 12 16 B 12 16 12 16 12 9 l 12 16 16 12 12 10 12 4 12 12 12 4 11 12 12 12 SS 12 12 12 4 12 16 12 4 13 14 15
## Number of WABAs 656 WABAs in SS Secondary Source 56 Clusters 5
e FIGURE 2.4 BURNABLE ABSORBER AND SECONDARY SOURCE ROD CONFIGURATIONS
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O O O O O O O E O E O E O O O O O E E O O O O O O E O E O i E I O n O O O E O E~ 4 BA Configuration 12 BA Configuration { 1 l
,E O E, , 0 0 0, E O E O E O O O O O O E E O O O O O E O E O E O O O O O E O E O O O 16 BA Configuration Secondary Source Rods 6
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3.0 CCNTROL ROD DROP TIME MEASURDENT (FNP-2-STP-ll2) PURPOSE 2e purpose of this test 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 RESULTS For the Hot-full flow condition (T > 541'F and all reactor coolant pumps operating) Technical SpecUlcation 3.1.3.4 requires that 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 drop times were measured to be less than 2.2 seconds. W e longest drop time recorded was 1.48 seconds for rod B-6. h e rod drop time results for both dashpot entry and dashpot bottom are presented in Figure 3.1. Mean drop times are sununarized below: TEST MEAN TIME 'IO MEAN TIME 'IO CONDITIONS DASHPCTF ENTRY DASHPOT BOTIOM Hot-full Flow 1.385 sec 1.86 see To confirm normal rod mechanism operation prior to conducting the rod drops, a Control Rod Drive Test (FNP-0-ETP-3643) was performed. In the test, the stepping waveforms of the 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 also conducted. All results were satisfactory. 7
.' UNIT 2 CYCLE 6 Y .
NORTH 900 1.37 1.40 1.39 &' L 1.82 -
- 1. 83 1.89 g \ 1.37 1.36 ,
1.85 ~1 .83 [ 1.37 1.36 . 13( 1.40 H 1.86 .l.87 1.82 1.86 1.h0 1.38 M 1.85 . 1.83 1.h6 1.37 1.35 1.35 1.30 1.38 ' 1.39
' N 1.96 1.37 1.8h 1.83 1.36 1.83 1.35
- 1. 8t 1.87 1.36 1.87 {g 1.8h 1.83 1.80 1.8h ~J
) Qo 1.43 1.36 1.36 1.ho I80* -g 1.92 1.83 1.8h 1.89.
1.37 1.35 1.38 1.39 1.82 1.80 1.82 1.89. ~0 1.39 1.h1 1.38 1.3h 1.40 1.hh 1.38 F 1.86 1.86 1.85 . 1.82 1.88 1.86 1.87 1.39 1.h2 1.83 1.89 1.37 1.37 1.37 1.36 1.85 1.86 1.86 1.82 1.h0 1.39 0 1.90 1.87 1.ho 1.45 1.h8 8 A.88 1.91 2.00 s 27o g r' A 15 14 13 12 11 10 9 8 7 6 5 4 3 2 i ORIVE LINE "0 ROP TIME" TABULATION TEMPERATURE - sh7.9 PRESSURE - 2260 usic % Flow - 100 X.XX BREAKER "0PENING" TO DASHP0T ENTRY - IN SECOMOS DATE - 11-30-87 X.XX BREAKER "0PEMING" TO DASHP0T 80TTOM - IN SECOMOS Figure 3.1 8
4.0 INITIAL CRITICALITY (FNP-2-ETP-3601) PURPOSE he purpose of this procedure was to achieve initial reactor criticality under carefully controlled conditions, citablish 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 6 was achieved during dilution mixing at 1555 hours on December 1, 1987. %e reactor was allowed to stabilize at the following critical conditions: RCS pressurp 2230 psig, RCS temperature 549.0'r, intermediate range power 1 x 10- amp, RCS boron concentration 1943.5 ppm, and Control Bank D position 179 steps. Following stabilization, the point of adding nuclear heat was determined and a checkout 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 initial criticality to demonstrc.te that adequate overlap existed. 9
4 5.0 ALL-RODS-OUT ISCTIMERMAL TEMPERA'IURE COEFFICIENT AND BORON ENDPO.'NT (IWP-2-ETP-3601) 1 PURPOSE he objectives of these measurements were to determine the hat, zero power isothermal and moderator temperature coefficients for the all-rods-out (ARO) configuration an:t to measure the ARO boron enopoint concentration. I
SUMMARY
OF RESULTS he measured ARO, hot zero power temperature coefficients and the ARO boron endpoint concentration are shown in Table 5.1. %e isothermal temperature coefficient was measured to be +0.26 pcW'F which meets the design acceptance criteria his gives a calculated moderator temperature coefficient of +2.6 pcW'F which is within the Technical Specification limit of +5.0 pcy 'F. h us, no rod withdrawal limits were needed to ensure the +5.0 pcW'F limit was met. We design acceptance criterion for the ARO critical boron concentration was satisfactorily met. 10 i
TABLE 5.1 ARD, HZP ISOIhERMAL AIO MODERATTR TEMPERATURE COEFFICIENT
..., fi Nrat. ion Boron Measured g Design Acceptance Calculated Concentration g Criterion a, , ,
ppn pc W'F pcig/'F pc W'F All Rods Out 1962.5 +0.26 -0.63 +3 +2.6 g - Isothermal temperature coefficient, includes -2.225 pcW'F doppler coefficient a,,, - MraW dy tyrahe emcied, m.uMM to & E MRm ARO, HZP BORON ENDPOINT CONCENTRATICN Bod Configuration Measured C, (ppa) Design - predicted C, (ppm) All Rods Out 1970.7 1976
6.0 OCNTROL AND SHLHDOW BANK WORH1 MEASUREMDirS (FNP-2-ETP-3601) PURPOSE ne' objective of the bank worth measurements was to determine the integral reactivity worth of each control and shutdown baak for comparison with the values predicted by design. S'JMMARY OF RESULTS We 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 reference bank reactivity needed to offset full insertion of the bank being measured.
%e control and shutdown bank worth measurement results are given in Table 6.1. W e measured worthr, satisfied the review criteria both for the banks measured individually and for the combined worth of all banks.
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TABLE 6.1 SIMERY OF CamO* APO SHIFIDOW BANK WOR'IH MEASURE 2ENTS Predicted Bank Measured North & Review Bank Percent Bank Criteria (pcm)- Worth (pcm) Difference Control A 347 100 322.9 -6.9 Control B (Ref.) 1242 124 1190.2 -4.2 Control C 860 129 733.6 -8.9 Control D 1071 160 975.0 -9.0 0 Shutdown A 971 145 916.0 -5.7 Shutdown B 1063 159 992.9 -6.6 All Banks Combined 5554 555.4 5180.6 -6.7
- Measured by dilution mettod N
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7.0 STAR'IUP AND POWER ASCENSION PROCEDURE (fNP-2-ETP-3605) PURPOSE he purpose of this procedure was to provide centrolling instructions I fort l
- 1. NIS intstmediate and power range setpoint changes, as required prior to startup and during power ascension.
- 2. Ramp rate limitation and control rod movement recomendations.
- 3. Conduct of startup and power ascension testing, to include: )
- a. H2P reactor physics. tests (fNP-2-ETP-3601).
- b. incore movable detector system alignment (INP-2-ETP-3636).
- c. incore-excore AFC channel recalibration (INP-2-STP-121).
- d. core hot channel factor rutveillance (ENP-2-STP-110).
- e. reactor coolant system flow measurement (FNP-2-STP-115.1).
SUMMARY
OF RESULTS In order to satisfy Technical Specification requirements for invoking special core physics test exceptions, preliminary trip l setpoints of less than or equal to 25% power were used for the NIS intermediate and power range channels. When physics tests were completed, the power range setpoint was increased to 80% to allow power escalation above 25% for calorimetric recalibration of the power range channels. ( h e 80% setpoint was used instead of 109% in case the uncalibrated power range channels were indicating non-l conservatively.) At approximately 35% power, the power range channels l were recalibrated, the high-range trip setpoint was restored to 109%, and setpoint currents were determined for the intermediate range channels. We Westinghouse fuel warranty limits the poer ramp rate to 3% of full power per hour between 20% and 100% power until full pop.r hns been sustained for 72 cumulative hours out of any seven-d:.y operating l period. h is ramp rate was observed during the ascension to 100% power. Determination of incore movable cetector system core limit settings (FNP-2-ETP-3606) was accomplished during the ascension to 35% power. h is was followed by the incore-execre recalibration tenot (FNP-2-STP-121) at 35% power :ed the reactor coolant fic- measurement (FNP-2-STP-115.1) at 100% powet, which are described in 5 actions 8.0 and 9.0 of this repott. As summarized in Table 7.1, Core hot channel factor surveillance was performed on the incore-excore full-core base case flux map taken under non-equilibrium conditions at 35% power, and on the full-core flux maps performed at 35% and 100% power under equilibrium conditions. 14 w
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5 TABLE 7.1 SIPIMARY OF POWER ASCENSION FLUX MAP DATA Parameter Map 1_46 Map 152 Map 153 Avg. % Power 32.43 32.15 99.85 Max FAH 1.5096 1.4949 1.4684
. Max Power Tilt
- 1.0087 1.0073 1.0064 Avg. Core % A.O. 9.811 ,
15.529 3.689 Maximum FQ(Z) 2.0969 2.1533 1.8215 FQ Limit 4.5240 4.5124 2.3026-Xenon Conditions Non- Equilibrium Equilibrium Equilibrium
- Calculated pow r tilts based on assemblf FAHN from all assemblies.
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M 8.0 INCORE-EXCORE DE'IECIOR CALIBRATION (FNP-2-STP-121) PURPOSE he objective of this procedure was to determine the relationship between power range upper and lower excore detector currents PM incore axial offset for the purpose of calibrating the control board and plant computer axial flux difference (AFD) channels, and for calibrating the delta flux penalty to the overteroperature delta-T protection system.
SUMMARY
OF RESULTS We incore-excore recalibration was perfmwd in two phases: (a) At approximately 35% po q , a full-core base case flux map and five quarter core flux maps were run at various positive and negative axial offsets to develop equations relating detector current to core axial offset. To reduce error, all flux maps were performed at the same RCS temperature. %e power range NIS channels were adjusted'to incorporate the revised calibration data. (b) When power is increased to 100%, the increased Neutron leakage resulting from the rise in RCS temperature introduces error into the 35% power incore-excore equations. Werefore, the 35% power zero percent axici offset currents must be re-normalized to adjust the incore-excore calibration for the higher operating temperature at full power. h is was accouplshed by performing a full core 1 flux map under stable, 100% power conditions to verify core hot l channel factors were satisfactory and to develop the normalization corrections. Using the finalized incore-excore equations given in Table 8.1, the NIS power range channels were recalibrated. l l l \ l 1 l 16
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TABLE 8.1 DETECIOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHA?NEL N41: I-Top = 0.7004*A0 + 159.84 pa I-Bottom = -0.9360*A0 + 156.61 a - CHANNEL N42: I-Top = 0.7337*A0 + 162.18 pm I-Bottom = -1.082A*A0 + 158.31 ps CHANNEL N43: I-Top = 0.7577*A0 + 162.38 a 1-Bottom = -0.9180*A0 + 155.37 pa CHANNEL N44: I-Top = 0.7962*A0 + 173.09 pa I-Bottom = -1.1697*A0 + 171.32 pa 17 , im +
v .,' 9.0 REACTOR COOLANT SYSTEM FIDW MEASURDENT (FNP-2-STP-ll5.1) PURPOSE The purpose of this procedure was to measure the flov' 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 systert flow rate measured at normal operating temperature and pressure must equal or exceed 265,500 gpn for three loop operation. From the average of six calorimetric heat balance measurements, +.he total core flow was determined to be 278,731.9 gpn, which meets the above criterion, k r i K M E.
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NT-88-0086 E- -'T;. Alabama Power Comp:ny
.i. 600 North 18th Street Post Offee Box 2641 . .
- Birmingham, Alaeama 35291-0400 Telephone 205 250-1835-
- t032";Len, AlabamaPower the southem electric system March 7, 1988
. Docket No. 50-364 U. ~ S. Nuclear Regulatory Cotanission Attention: Document Control Desk Washington, D. C. 20555 ,
Joseph M. Farley Nuclear Plant - Unit 2 Cycle 6 - Startup Report Gentlemen: Enclosed is the Startup Report for Unit 2 Cycle 6 as outlined in the October 12, 1987 letter from Mr. R. P. Mcdonald. If you have any questions, please advise.
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Yours very trply, - (
. G, i v y R. P. Mcdonald RPM /MDR:emb Enclosure cc: Mr. L. B. Long Dr. J. N. Grace Mr. E. A. Reeves Mr. W. H. Bradford t\
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